ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 70

DOE Review Release 06/20/2013

Document Title:

Fission and Activation Product Assignment 

for Internal Dose-Related Gross Beta and 

Gross Gamma Analyses

Document Number: ORAUT-OTIB-0054

Revision: 01

Effective Date: 06/13/2013

Type of Document: OTIB

Supersedes: Revision 00 PC-1

Subject Expert(s): Robert E. Burns, Jr., Donald E. Bihl

Approval: Signature on File Approval Date: 06/07/2013

Robert E. Burns, Jr., Document Owner

Concurrence: Signature on File Concurrence Date: 06/07/2013

John M. Byrne, Objective 1 Manager

Concurrence: Signature on File Concurrence Date: 06/10/2013

Edward F. Maher, Objective 3 Manager

Concurrence: Vickie S. Short Signature on File for Concurrence Date: 06/10/2013

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 06/13/2013

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. 01 Effective Date: 06/13/2013 Page 2 of 70

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.

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.

Signature on File 10/10/2007

Donald E. Bihl, Document Owner

Signature on File 10/03/2007

John M. Byrne, Task 3 Manager

Signature on File 10/03/2007

Edward F. Maher, Task 5 Manager

Signature on File 11/13/2007

Kate Kimpan, Project Director

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

James W. Neton, Associate Director for Science

06/13/2013 01 Total rewrite of the document to implement the following: a limiting 

reactor approach in place of averaging, an expanded range of 

reactor modeling parameters, more rigorous treatment of 

radionuclide ingrowth and decay and parent/progeny behavior, 

updated beta and gamma yield values, removal of the counting 

adjustment factors, and explicit consideration of changes in the beta 

and gamma activity of urine samples as a function of time between 

sample collection and counting. Additional changes were made in 

response to comments received from the Procedures Subcommittee 

and to provide clarifications with respect to methods and application 

of the results. Incorporates formal internal and NIOSH review 

comments. Constitutes a total rewrite of the document. Training 

required: as determined by the Objective Manager. Initiated by 

Robert E. Burns, Jr. and Donald E. Bihl. 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 3 of 70

TABLE OF CONTENTS

SECTION TITLE PAGE

Acronyms and Abbreviations..................................................................................................................5

1.0 Introduction .................................................................................................................................7

2.0 Purpose.......................................................................................................................................7

3.0 Scope..........................................................................................................................................7

4.0 Background.................................................................................................................................8

5.0 Reactor Modeling and Determination of Fission and Activation Product Activities.....................8

5.1 ORIGEN2 Modeling for the Initial Set of Representative Reactors.................................8

5.2 Evaluation of the ORIGEN2 Results for the Initial Set of Representative 

Reactors..........................................................................................................................9

5.3 Fission and Activation Product Inventory Calculations for the 

Representative Reactor Cases .......................................................................................9

5.4 Decay Times .................................................................................................................10

6.0 Determination of Intake Fractions and the Dosimetrically Significant Radionuclides ...............11

6.1 Intake Fractions.............................................................................................................11

6.2 Development of Dosimetrically Significant Reactor Source Term.................................12

7.0 Determination of Intakes of Mixed Fission and Activation Products from 

Nonspecific Radionuclide Analyses ..........................................................................................13

7.1 Activity Fractions for Urinalyses ....................................................................................13

7.1.1 Radionuclide Mixture in the Sample at the Time of Collection ..........................14

7.1.2 Decay and Ingrowth After Collection and Impact on Sample Activity................14

7.1.3 Activity Fractions for Different Types of Urinalyses...........................................15

7.2 Activity Fractions for Air Samples or Surface Contamination Measurements...............16

8.0 Mixed Fission and Activation Product Dose Assignment Summary .........................................22

8.1 Interactive RadioEpidemiological Program (IREP) Distribution Assignment.................24

9.0 Attributions and Annotations .....................................................................................................25

References ...........................................................................................................................................27

ATTACHMENT A, RADIONUCLIDES FROM REACTOR SOURCE TERMS NOT 

INCLUDED IN DOSE ANALYSES ........................................................................ 28

ATTACHMENT B, RELEASE FRACTIONS ........................................................................................ 29

ATTACHMENT C, RADIONUCLIDES CONSIDERED IN COMMITTED ORGAN 

DOSE CALCULATIONS AND THE ABSORPTION TYPES 

ASSIGNED TO EACH SOLUBILITY CATEGORY................................................ 30

ATTACHMENT D, NORMALIZED INTAKE FRACTIONS FOR THE 

DOSIMETRICALLY IMPORTANT RADIONUCLIDES ......................................... 37

ATTACHMENT E, NORMALIZED INTAKE FRACTIONS FOR THE REDUCED LIST 

OF DOSIMETRICALLY IMPORTANT RADIONUCLIDES..................................... 41
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ATTACHMENT F, SUMMARY OF INTAKE RETENTION FRACTIONS............................................. 44

ATTACHMENT G, RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY.............................. 47

ATTACHMENT H, EXAMPLE INTAKES ESTIMATED FROM BIOASSAY AND 

FROM AIR SAMPLE DATA................................................................................... 67

LIST OF TABLES

TABLE TITLE PAGE

5-1 Reactors used to represent categories .......................................................................................8

5-2 ORIGEN-S irradiation parameters for the nine representative reactor cases...........................10

5-3 Fission and activation product decay times for general steps in the fuel cycle.........................11

7-1a Gross beta activity fractions in urine for minimally processed samples....................................17

7-1b Gross gamma activity fractions in urine for minimally processed samples...............................17

7-2 Urine activity fractions for samples following major chemical processing.................................18

7-3a Associated radionuclide activity fractions for assigning intakes: ATR 1 ..................................19

7-3b Associated radionuclide activity fractions for assigning intakes: ATR 2 ..................................19

7-3c Associated radionuclide activity fractions for assigning intakes: ATR 3 ..................................19

7-3d Associated radionuclide activity fractions for assigning intakes: FFTF 1.................................20

7-3e Associated radionuclide activity fractions for assigning intakes: FFTF 2.................................20

7-3f Associated radionuclide activity fractions for assigning intakes: N Reactor 1 .........................21

7-3g Associated radionuclide activity fractions for assigning intakes: N Reactor 2 .........................21

7-3h Associated radionuclide activity fractions for assigning intakes: TRIGA 1...............................21

7-3i Associated radionuclide activity fractions for assigning intakes: TRIGA 2...............................22

7-4a Activity ratios for use if gross beta results are reported for air samples and workplace 

measurements ..........................................................................................................................23

7-4b Activity ratios for use if gross gamma results are reported for air samples and 

workplace measurements .........................................................................................................23

A-1 Radionuclides from reactor source terms not included in analyses..........................................28

B-1 Elemental exposure fractions....................................................................................................29

C-1 Radionuclides considered in committed organ dose calculations and the absorption 

types assigned to each solubility category................................................................................30

D-1 NIFs for radionuclides that contribute ≥1% dose to committed organ dose..............................37

E-1 Normalized intake fractions for the reduced list of dosimetrically important 

radionuclides.............................................................................................................................41

F-1 Summary of IRFs used to compute urinary excretion...............................................................44

F-2 Half-lives used for decay corrections ........................................................................................45

F-3 Beta and gamma radiation yield values ....................................................................................46

G-1 Contributions to gross beta activity for an unprocessed urinalysis ...........................................47

G-2 Contributions to gross beta activity for urinalysis following major chemical processing ...........51

G-3 Contributions to gross gamma activity for an unprocessed urinalysis ......................................54

G-4 Contributions to gross beta activity for an air or swipe sample.................................................58

G-5 Contributions to gross gamma activity for an air or swipe sample............................................63

H-1 Gross beta urinalysis results and ATR-1 90Sr results................................................................67

H-2 ATR-1 radionuclide intake rates derived from 90Sr intake rate of 76.0 pCi/d ............................68

H-3 Gross beta urinalysis results and ATR-1 90Sr results................................................................69

H-4 ATR-1 radionuclide intake rates derived from 90Sr intake rate of 3.51 pCi/d ............................69

H-5 ATR-1 radionuclide air concentrations and intake rates derived from a gamma air 

concentration of 1 × 10-10 μCi/mL .............................................................................................70
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ACRONYMS AND ABBREVIATIONS

ATR Advanced Test Reactor

Bq becquerel

d day

DCF dose conversion factor

DOE U.S. Department of Energy

FFTF Fast Flux Test Facility

g gram 

hr hour

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis

IREP Interactive RadioEpidemiological Program

IRF intake retention fraction

keV kiloelectron-volt, 1 thousand electron-volts

m meter

MAP mixed activation product

MDA minimum detectable activity

MeV megaelectron-volt, 1 million electron-volts

MFP mixed fission product

mL milliliter

MT metric ton

MTHM metric tons of heavy metal

MTU metric tons of uranium

MW megawatt

MWd megawatt-day

NIF Normalized Intake Fraction

NIOSH National Institute for Occupational Safety and Health

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

pCi picocurie

PWR pressurized-water reactor

RaLa radioactive lanthanum

s second

SRDB Ref ID Site Research Database Reference Identification (number)

SRS Savannah River Site

SS stainless steel

t ton (short)

TIB technical information bulletin 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 6 of 70

TRIGA Training, Research, Isotopes, General Atomics

U.S.C. United States Code

yr year

μCi microcurie

§ section or sections
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1.0 INTRODUCTION

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 

historical background information and guidance to assist in the preparation of dose reconstructions at 

particular sites or categories of sites. They will be revised in the event additional relevant information 

is obtained about the affected site(s). TIBs may be used to assist NIOSH staff in the completion of 

individual dose reconstructions.

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 PURPOSE

The purpose of this document is to provide a basis for assigning 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 for dose reconstruction 

involving reactor-related source terms if site-specific guidance is not available.

3.0 SCOPE

The guidance in this document applies, if 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. The activity and intake fraction values were 

derived to address gross beta or gross gamma results specifically from air samples or urinalyses, but 

they 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. 

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., the Hanford 

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

reactors]. It does not apply to operations involving decay times shorter than 10 days (e.g., early 

radioactive lanthanum (RaLa) processing). It also does not apply to determination of intakes if 

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.

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.
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 8 of 70

4.0 BACKGROUND

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

surveys, air sampling, and 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. 

5.0 REACTOR MODELING AND DETERMINATION OF FISSION AND ACTIVATION 

PRODUCT ACTIVITIES

This section describes selection of the representative reactor cases and calculation of fission and 

activation product activity at decay intervals of interest following reactor shutdown. 

5.1 ORIGEN2 MODELING FOR THE INITIAL SET OF REPRESENTATIVE REACTORS

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:

• 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 selected for fuel irradiation calculations using 

the ORIGEN2 isotope generation and depletion code, version 2.1 (Croff 1980). More than one reactor 

or one type of fuel with cladding was selected for some of the categories. The reactors were selected 

to cover a wide range of fuel (cladding) types, enrichments, and burnups. 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 Sodium-cooled fast reactors

ATR High-flux reactors

TRIGA reactor (Al-clad fuel) Research reactors

TRIGA reactor (SS-clad fuel) Research reactors

PWR Generic reactor

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

production reactors and the TRIGA reactors were run with more than one ORIGEN2 cross-section 

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

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

libraries to evaluate the choice of library on the inventory results. Reactor-specific ORIGEN2 libraries 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 9 of 70

were available for the FFTF, the ATR, and the pressurized-water reactor (PWR). 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 to be representative of the compositions of typical source terms 

encountered by workers. Decay times are discussed further in Section 5.4. 

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, FFTF, and 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 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. Nominal values for burnup 

and specific power were selected as appropriate given the composition basis. The fuel plus 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. 

5.2 EVALUATION OF THE ORIGEN2 RESULTS FOR THE INITIAL SET OF 

REPRESENTATIVE REACTORS

The results from the 11 ORIGEN2 runs were compared on the basis of relative activity (to 137Cs) 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 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 reactors were selected as having isotopic inventories and ratios that would be 

representative of the variety of reactors and fuel combinations under consideration:

• The ATR

• The FFTF

• The N Reactor

• The TRIGA reactor

5.3 FISSION AND ACTIVATION PRODUCT INVENTORY CALCULATIONS FOR THE

REPRESENTATIVE REACTOR CASES

To facilitate the fission and activation product inventory calculations needed for this TIB, case-specific 

cross-section libraries were created for the ORIGEN-S code using the TRITON analysis sequence of 

the Oak Ridge National Laboratory (ORNL) SCALE code system. TRITON automates the process of 

resonance self-shielding correction and sequential fuel depletion calculations to create binary libraries 

that can be used by ORIGEN-S. Custom libraries derived from modern (ENDF/B-VII) evaluated 

cross-section data were developed for the ATR, FFTF, N Reactor, and TRIGA reactor. The ORIGENS libraries generated by the TRITON models were used in radionuclide inventory calculations to 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 10 of 70

obtain the needed fission and activation product inventories as a function of decay time. ORIGEN-S, 

in conjunction with the ORIGEN-ARP preprocessing utility, is the current version of the ORIGEN code. 

An ORIGEN-S input specifies the cross-section data to be used, the core composition of interest (fuel, 

cladding, structural materials, etc.), the specific power level (MW per unit mass of heavy metal), the 

irradiation duration, and the decay intervals of interest. To ensure broad representation of the fission 

and activation product inventory data for the four representative reactors, at least two (and in one 

instance three) sets of irradiation parameters were defined for each. This resulted in nine 

representative cases for which fission and activation product inventory data were computed using the 

custom-built cross-section libraries developed for each reactor. The parameters selected for the nine 

representative cases are listed in Table 5-2. Three cases were defined for the ATR and two were 

defined for the other three reactors.

Table 5-2. ORIGEN-S irradiation parameters for the nine representative reactor cases.

Case Parameters Basis

ATR 1 Specific power = 2,379.1 MW/MTU

Irradiation time = 132.27 days

Burnup = 314,684 MWd/MTU

Maximum burnup at nominal 

power.

ATR 2 Specific power = 8,651.2 MW/MTU

Irradiation time = 36.4 days

Burnup =314,904 MWd/MTU

Maximum burnup at maximum 

assembly power.

ATR 3 Specific power = 2,379.1 MW/MTU

Irradiation time = 56 days

Burnup = 133,230 MWd/MTU

Nominal burnup at nominal power.

FFTF 1 Specific power = 163.8 MW/MTHM

Irradiation time = 929.4 days

Burnup = 152,230 MWd/MTHM

Maximum burnup at nominal 

power.

FFTF 2 Specific power = 163.8 MW/MTHM

Irradiation time = 488.3 days

Burnup = 79,984 MWd/MTHM

Nominal burnup at nominal power.

N Reactor 1 Specific power = 10.4 MW/MTU

Irradiation time = 114.2 days

Burnup = 1,188 MWd/MTU

Production of weapons-grade

plutonium (nominal 6% Pu-240

content) at nominal power.

N Reactor 2 Specific power = 10.4 MW/MTU

Irradiation time = 285.6 days

Burnup = 2,970 MWd/MTU

Production of fuel-grade plutonium 

(nominal 12% Pu-240 content) at 

nominal power.

TRIGA 1 Specific power = 15.57 MWd/MTU

Irradiation time = 730.1 days

Burnup = 11,368 MWd/MTU

Maximum burnup at nominal 

power.

TRIGA 2 Specific power = 15.57 MW/MTU

Irradiation time = 115.2 days

Burnup = 1994 MWd/MTU

Nominal burnup at nominal power.

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 of 10 days, 40 days, 

180 days, and 1 year 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. 

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 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 11 of 70

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

any particular dose reconstruction if sufficient information is available. 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, a 

decay time of 180 days or 1 year should be chosen. 

The work activities described for each of the four decay intervals in Table 5-3 can be used for general 

guidance. As an alternative, if an appropriate decay time cannot be readily established, intakes and 

organ doses for all four decay times should be determined and the set that is most favorable to the 

claimant should be selected for dose reconstruction.

Table 5-3. Fission and activation product decay times for general steps in the fuel cycle.

Activity Decay time

Reactor operations, spent fuel storage, fuel examination: activity from fuel leakage/failure or 

segmentation.

10 d

Dissolution of fuels from 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 DETERMINATION OF INTAKE FRACTIONS AND THE DOSIMETRICALLY 

SIGNIFICANT RADIONUCLIDES

This section describes the process of determining the dosimetrically significant constituents of the 

radionuclide inventory data from the ORIGEN-S calculations through the use of element-specific 

exposure fractions and radionuclide- and absorption type-specific dose conversion factors (DCFs). 

6.1 INTAKE FRACTIONS

The fission and activation product content of irradiated reactor fuel and 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 parents were 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 only for 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 129I). In later years (approximately mid-1950s onward), it was rare to process 

fuel aged for less than 100 days.

Conversion of the radionuclide inventory data from the nine ORIGEN-S cases to intake fractions 

begins by extracting the fission and activation product activity values from the output files for the 

desired decay intervals and writing them to spreadsheets. The fission and activation product activity 

data are then summed as appropriate (ORIGEN-S reports them separately) and the radionuclides 

showing an activity value of zero at ≥ 10 days of decay are removed. (“Zero” in this content refers to 

the double-precision threshold for the hardware/software environment, typically less than 10-30. 

Values below this threshold are set to zero by the code.) Noble gases and radionuclides that do not 

have specifically defined inhalation DCFs (for committed organ dose) in the RadToolbox database are 

also removed. Radionuclides that do not have DCFs include those that are inexorably associated 

with a given parent (e.g., 137mBa) and those with very short or very long half-lives. A complete list of 

radionuclides removed from the fission and activation product inventory data (because they were 

noble gases or did have inhalation DCFs) is provided in Table A-1 in Attachment A. Although the 

noble gases showing nonzero activity at ≥ 10 days decay are removed, their progeny are not.
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 12 of 70

Each of the source terms for the nine representative reactor cases was processed as described 

above. Release fractions from DOE Standard 1027 (DOE 1997) were then used to convert from 

activity at each of the four decay intervals to intake fraction. For each radionuclide and decay time in 

a given source term, the product of the activity and release fraction was summed and the values were 

restated on a normalized basis. These values are referred to as Normalized Intake Fractions (NIFs) 

because the product of the radionuclide activity and the release fraction was defined as an intake 

fraction. This means the activity units are arbitrary as long as they are the same for each case.

The release fractions used to compute the NIF values were assigned as follows: gases (e.g., tritium) 

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 particulates and solids, the 

DOE standard 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. Hence, 

the smallest release fraction used for any radionuclide was 0.01, an order of magnitude greater than 

the minimum value in DOE Standard 1027 (DOE 1997).

The large majority of the radionuclides in the source terms considered in this document were 

nonvolatile particulates. Therefore, the choice of the release fractions has a limited effect on the 

results. The release fractions affect the ratio of the intakes of volatile and semivolatile elements 

relative to the nonvolatile particulates. The only volatile or semivolatile species of dosimetric 

importance are the radioiodines. The choice of the release fractions has no effect on the activity 

fractions presented in Sections 7.1 and 7.2. 

The release fractions used to convert from activity to intake for each radionuclide of interest are listed 

in Table B-1 of Attachment B. The values are assigned on an elemental (rather than isotopic) basis.

6.2 DEVELOPMENT OF DOSIMETRICALLY SIGNIFICANT REACTOR SOURCE TERM

The calculations of the NIF values for the nine representative reactor cases resulted in source terms 

consisting of approximately 300 radionuclides. Committed organ doses were computed using the NIF 

values for each reactor case and decay interval to determine the dosimetrically important constituents 

across the nine representative cases. 

Committed organ doses for each reactor case and decay interval were computed using inhalation 

DCF values from the RadToolbox software, version 2.0.0, distributed by the Oak Ridge National 

Laboratory (ORNL 2003). The DCFs from RadToolbox are consistent with those from International 

Commission on Radiological Protection (ICRP) Publication 68 (ICRP 1995). The inhalation DCFs for 

the collective set of radionuclides of interest across the nine reactor cases were partitioned into three 

solubility categories based on their ICRP-defined absorption types. The most soluble type for each 

nuclide was assigned to the soluble category and the least soluble types were assigned to the 

insoluble category. A moderately soluble category was also defined. Type F materials were always 

assigned to the soluble category, type M to moderately soluble, and type S to insoluble. If a 

radionuclide had only one absorption type defined, it was used in all three categories. Radionuclides 

with two absorption types were assigned as follows:

• Types F and M: Type M was assigned to the insoluble category.

• Types F and S: Type F was assigned to the moderately soluble category.

• Types M and S: Type M was assigned to the soluble category.

In some cases, absorption types were limited based on appropriate physical or chemical constraints 

for discharged reactor fuels. Examples include radiostrontium isotopes, for which only type F was 

assigned, and radioiodines, for which only the vapor form was considered. 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 13 of 70

Three tables of inhalation DCF values were compiled containing entries for each radionuclide of 

interest. Table C-1 in Attachment C presents the complete list of radionuclides considered in the 

committed organ dose calculations and the absorption types assigned for each of the three solubility 

categories. Three sets of committed organ doses (one for each solubility category) were computed 

for each reactor case and decay interval for the 27 ICRP Publication 68 (ICRP 1995) organs 

considered in RadToolbox (including effective and remainder). The collective set of organ dose 

results (across all nine cases) was then evaluated to identify all radionuclides that contributed ≥ 1% of 

the dose to any organ for any reactor case, decay interval, or solubility category. This resulted in a 

total of 36 dosimetrically important radionuclides being defined for the nine representative reactor 

cases. Next, the NIF values for each reactor case and decay interval were renormalized using just 

these dosimetrically significant radionuclides. The resulting NIF values are listed in Table D-1 in 

Attachment D. Renormalizing the NIF values in this manner is favorable to the claimant because it 

serves to increase the intake fraction values for the most important radionuclides over what the 

fraction would be if all radionuclides were included.

To reduce the workload for dose reconstructors, a further reduction in the number of dosimetrically 

significant radionuclides was performed. The Table D-1 NIF values were used to calculate effective 

dose for each reactor case, decay interval, and solubility category. Radionuclides that contributed 

≥ 1% to effective dose for any case, decay interval, or solubility category were then identified and a 

subsequent renormalization of the corresponding NIF values was performed. This reduced list of 17 

dosimetrically important radionuclides and the corresponding NIF values are presented in Table E-1 in 

Attachment E. As with the first reduction in the number of radionuclides of interest, this subsequent 

reduction is favorable to claimants with respect to assigned intakes.

The reduced list of dosimetrically important radionuclides (i.e., Table E-1) is used only to establish the 

relative intake fractions for the other components of an intake once the intake for the primary 

(indicator) radionuclide has been established. Intake assignments for the indicator radionuclides 

reflect the more comprehensive 36 radionuclide (Table D-1) list. This process is detailed in the 

following section.

7.0 DETERMINATION OF INTAKES OF MIXED FISSION AND ACTIVATION PRODUCTS 

FROM NONSPECIFIC RADIONUCLIDE ANALYSES

This section describes the calculation of activity and intake fraction values to be used for assigning 

intakes from gross beta or gross gamma assay data. The general process is to determine the activity 

fraction of a principal (indicator) radionuclide that should always be present in a given type of sample. 

The activity of the indicator radionuclide in relation to the total sample activity is then used to 

determine the corresponding intake given the specifics of the claim being evaluated. Once the 

indicator radionuclide intake has been established, the intake for the remaining radionuclides in the 

mix is assigned using the ratios established from Table E-1. Absorption types for each radionuclide 

are then selected to yield the most favorable dose assignment for the organ of interest. This can be 

done for a specific decay time or across all decay times, as appropriate for each case. 

Two indicator radionuclides are considered in this document: 90Sr for gross beta analyses and 137Cs

for gross gamma analyses.

7.1 ACTIVITY FRACTIONS FOR URINALYSES

This section describes the calculation of indicator radionuclide activity fractions relative to the total 

beta or gamma activity for a 24-hour urine sample following a 2-year (730-day) chronic intake.

The gross beta or gross gamma activity for a given sample depends on the radionuclide mixture at 

intake; the intake period (only chronic intake is considered), biokinetics (urinary excretion);
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 14 of 70

radionuclide decay and ingrowth that occur in the sample in the time between collection and counting;

separations chemistry (for sample preparation), if any; and the beta or gamma radiation yields for the 

radioisotopes present at the time of counting. All of these factors were considered in deriving the 

indicator radionuclide activity fractions for urinalyses.

7.1.1 Radionuclide Mixture in the Sample at the Time of Collection

For each radionuclide present in the intake, the contribution to the urinary excretion is given by the 

product of the NIF (Table D-1 value) and the intake retention fraction (IRF) for a 24-hour urine sample. 

Progeny radionuclides that grow in from parent radionuclides in the intake also contribute to the 

excreted activity. Some of these progeny radionuclides are also present in the intake mixture, 

meaning there are both supported and unsupported components of their total activity in urine. 

IRF values (see Table F-1 in Attachment F) were obtained from the Integrated Modules for Bioassay 

Analysis (IMBA) software for 24-hour urinalyses following a 2-year chronic intake period. IRFs were 

determined for each radionuclide in the Table D-1 mix. Absorption type F was used for all 

radiostrontium isotopes. Otherwise, only the least soluble absorption type was used for radionuclides 

for which multiple selections were possible. Limiting the activity fraction calculations to the least 

soluble absorption types maximizes the indicator radionuclide activity fractions because both of them 

can only be type F. However, this does not mean dose reconstructors should limit the absorption 

types used for determining assigned dose to the least soluble forms.

Following ICRP conventions, parent and progeny radionuclides present in the intake were treated 

independently (i.e., specific IRF values were used for each radionuclide). In addition, progeny 

radionuclides that would accumulate in the body from the decay of the corresponding parent 

radionuclides over the course of a 730-day chronic intake were also accounted for. The Dose and 

Risk Calculation (DCAL) program was used to determine the progeny activity in urine relative to the 

parent activity following a 730-day chronic intake. The specific progeny radionuclides considered and 

their activity fractions in urine were:

• 90Y (from 90Sr): 0.685

• 95Nb (from 95Zr): 0.401

• 95mNb (from 95Zr): 0.0056

• 125mTe (from 125Sb): 0.0927

• 140La (from 140Ba): 0.517

The progeny fractions above were used in conjunction with the parent radionuclide activity values 

(given by the product of the intake fraction and the corresponding IRF) to compute the supported 

component of the progeny activity in urine. The unsupported component, likewise the product of the 

intake fraction and IRF, was then added to get the total. 

Radioiodines were not included in the excreted activity calculations under the assumption they would 

be lost in the sample preparation process and thus not present at the time of counting. This is 

favorable to claimants because any radioiodine activity that was present would increase the measured 

beta or gamma activity and therefore the assigned indicator radionuclide intake.

7.1.2 Decay and Ingrowth After Collection and Impact on Sample Activity

Radioactive decay and ingrowth that occur in a urine sample during the time between collection and 

counting affect the sample’s total beta and gamma activity and thus the indicator radionuclide activity 

fractions. The magnitude of this effect depends on the delay time between collection and counting 

and the contribution to the total beta or gamma activity from short-lived radionuclides in the mixture at 

excretion. 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 15 of 70

The radionuclide mixture and corresponding beta and gamma activity were determined for each 

reactor case and decay time at 1-day intervals from 1 to 30 days. Decay and ingrowth were 

computed through a combination of standard radioactive decay calculations (for radionuclides without 

transient or secular equilibrium products) and using the ORIGEN-S code (for parent-progeny 

relationships). The half-lives used for the standalone decay corrections are given in Table F-2 in 

Attachment F. 

There are limited urinalysis data available that include information on both when a sample was 

collected and when it was counted. However, such data are available for gross beta and gross 

gamma assays by Hanford, Idaho National Laboratory, and ORNL. Review of count data from these 

sites showed a 30-day maximum to be appropriate for delays between sample collection and 

counting. Limiting the maximum delay time is necessary because in most cases the indicator 

radionuclide contribution to the sample activity increases with increasing delay. 

7.1.3 Activity Fractions for Different Types of Urinalyses

Activity fractions for the indicator radionuclides in gross beta or gross gamma urinalyses were 

established for the following three types of analyses:

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

beta emitters naturally in urine, such as 40K,

• 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.

The latter analysis was used at the Hanford Site from the late 1940s through the early 1960s. It was 

assumed that the chemical recoveries for all the elements captured by separations chemistry were 

essentially equivalent. Therefore, explicit consideration of chemical recovery values was 

unnecessary: the activity fractions are independent of the chemical recoveries for processed samples 

unless there are significant differences in recovery for some of the elements. Because the gross beta 

urinalysis procedure at the Savannah River Site (SRS) is known to have differed from that assumed 

here, the guidance that is provided here for chemically processed gross beta samples should not be 

applied to cases from the SRS. Rather, guidance on the interpretation of gross beta urinalyses 

performed at SRS will be provided in the SRS Site Profile. The SRS-specific guidance should be 

used in lieu of this document. 

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. However, if 

such information is not available it is favorable to the claimant to assume all of the activity is 

attributable to occupational exposure.

The sample activity data and beta and gamma radiation yield data (see Table F-3 in Attachment F) 

were used to compute the total beta and gamma activity for each of the three urinalyses types 

described above. Beta and gamma activity were computed as a function of delay time (i.e., the time 

between sample collection and counting) for each of the 36 radionuclide intake mixtures (four fuel 

decay times for each of the nine reactor cases). For processed samples, delay time between 

chemical separation and counting was assumed to be negligible. 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 16 of 70

The beta and gamma yield values used to convert from radionuclide activity to beta or gamma activity 

reflect energy thresholds that were applied to omit beta or gamma activity that likely would not have 

been detected. A threshold of 50 keV (average energy) was used for beta emitters, and a 100-keV 

threshold was applied for gamma lines. Beta and gamma yield values were combined for cases 

where parent/progeny pairs establish equilibrium within 24 hours. These include 89Sr/89mY, 

99Mo/99mTc, 129mTe/129Te, 136Cs/136mBa, 137Cs/137mBa, and 144Ce/Pr. 

The indicator radionuclide activity fractions are simply the fraction of the total beta or gamma activity 

that is contributed by the indicator radionuclide for the reactor case, decay time, and analysis type of 

interest. Delay between sample collection and counting is accounted for by selecting the maximum 

indicator radionuclide activity fraction that occurs between 1 and 30 days delay. This ensures 

favorability to claimants. The analysis type determines if the beta or gamma activity is used and if 

some radioelements are removed to reflect preparation chemistry. The beta or gamma activity data 

used to establish the indicator radionuclide activity fractions for each reactor case, decay time, and 

analysis type are given in Tables G-1 through G-3 in Attachment G. 

The applicable indicator radionuclide activity fractions for gross beta and gross gamma urinalyses are 

given in Tables 7-1a (for gross beta urinalysis without chemical separations), 7-1b (for gross gamma 

urinalysis without chemical separations), and 7-2 (for gross beta urinalysis following separations 

chemistry). The values in Table 7-1a and 7-1b should be applied to cases where minor chemistry 

might have been performed to remove natural emitters such as 40K. The Table 7-2 values should be 

used for cases where significant chemistry was performed before counting.

The activity fractions in Tables 7-1a, 7-1b and 7-2 are used, as appropriate, to determine the indicator 

radionuclide activity associated with a given gross beta or gross gamma urinalysis result. The 

resulting indicator radionuclide activity values are then used to determine the corresponding intake. 

Once the indicator radionuclide intake has been established, intakes for the other radionuclides in the 

intake are assigned using Tables 7-3(a) through 7-3(i) below. These tables list the intake fractions for 

the Table E-1 radionuclide mix for each representative reactor case stated relative to the two indicator 

radionuclide intakes.

The intake ratios for 131I in Tables 7-3(a) through 7-3(i) should not be used for sites where iodine was 

well controlled or if individual iodine monitoring data are available. The 131I intake ratios would 

primarily apply to the earliest programs, before the use of ventilation and collection systems. For later 

years, it is reasonable to assume there were no significant intakes from iodine at most sites.

7.2 ACTIVITY FRACTIONS FOR AIR SAMPLES OR SURFACE CONTAMINATION 

MEASUREMENTS

Activity fractions for gross beta or gross gamma assays for air samples or surface contamination 

measurements were computed in essentially the same manner as described in the previous section 

for urinalyses. The key differences are that IRF values are not needed and delay time between 

sample collection and counting was considered negligible. The calculations are identical otherwise. 

As was the case for the urinalysis evaluations, radioiodines were not included in the activity fraction 

calculations for the air sample/surface contamination measurements. This is an overestimating 

assumption for cases in which the measured results did include radioiodines.

Tables G-4 and G-5 in Attachment G list the contributions from the Table D-1 radionuclide mixtures to 

gross beta and gross gamma activity for air concentration or surface contamination. The indicator 

radionuclide activity fractions for gross beta and gross gamma air concentration or surface 

contamination assays are listed in Tables 7-4a and 7-4b, respectively. Assay results are multiplied by 

the appropriate Table 7-4a or 7-4b value to compute the indicator radionuclide activity or activity 

concentration on the sample. Indicator radionuclide intake is established based on the corresponding 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 17 of 70

Table 7-1a. Gross beta activity fractions in urine for minimally processed samples.

Radionuclide Activity fractions for gross beta urinalyses following minimal chemical processinga

Sr-90

(gross beta)

Fraction of beta activity in urine: ATR 1 Fraction of beta activity in urine: ATR 2 Fraction of beta activity in urine: ATR 3

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

2.14E-02 3.21E-02 1.18E-01 2.12E-01 1.24E-02 1.99E-02 9.19E-02 2.07E-01 1.43E-02 2.25E-02 1.02E-01 2.26E-01

Fraction of beta activity in urine: FFTF 1 Fraction of beta activity in urine: FFTF 2

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

4.13E-02 4.97E-02 7.06E-02 8.27E-02 3.21E-02 4.19E-02 7.27E-02 8.99E-02

Fraction of beta activity in urine: N Rx 1 Fraction of beta activity in urine: N Rx 2

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

1.96E-02 3.00E-02 1.19E-01 2.27E-01 3.68E-02 5.30E-02 1.47E-01 2.17E-01

Fraction of beta activity in urine: TRIGA 1 Fraction of beta activity in urine: TRIGA 2

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

1.40E-01 1.68E-01 2.47E-01 2.77E-01 2.01E-02 3.07E-02 1.26E-01 2.50E-01

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.

Table 7-1b. Gross gamma activity fractions in urine for minimally processed samples.

Radionuclide Activity fractions for gross gamma urinalyses following minimal chemical processinga

Cs-137

(gross gamma

Fraction of gamma activity in urine: ATR 1 Fraction of gamma activity in urine: ATR 2 Fraction of gamma activity in urine: ATR 3

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

1.49E-01 2.51E-01 3.90E-01 4.50E-01 7.66E-02 1.90E-01 4.39E-01 5.25E-01 1.03E-01 2.48E-01 5.82E-01 6.92E-01

Fraction of gamma activity in urine: FFTF 1 Fraction of gamma activity in urine: FFTF 2

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

2.14E-01 2.47E-01 2.97E-01 3.38E-01 2.37E-01 3.05E-01 4.07E-01 4.66E-01

Fraction of gamma activity in urine: N Rx 1 Fraction of gamma activity in urine: N Rx 2

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

1.67E-01 3.43E-01 7.06E-01 8.18E-01 2.67E-01 4.12E-01 6.08E-01 6.78E-01

Fraction of gamma activity in urine: TRIGA 1 Fraction of gamma activity in urine: TRIGA 2

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

6.23E-01 7.05E-01 7.75E-01 8.06E-01 1.75E-01 3.72E-01 7.86E-01 9.08E-01

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.
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 18 of 70

Table 7-2. Urine activity fractions for samples following major chemical processing.

Radionuclide Activity fractions for gross beta urinalyses following major chemical processing

Sr-90

(gross beta)

Fraction of beta activity in urine: ATR 1 Fraction of beta activity in urine: ATR 2 Fraction of beta activity in urine: ATR 3

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

2.37E-02 3.68E-02 1.77E-01 4.71E-01 1.34E-02 2.18E-02 1.22E-01 4.12E-01 1.54E-02 2.47E-02 1.33E-01 4.27E-01

Fraction of beta activity in urine: FFTF 1 Fraction of beta activity in urine: FFTF 2

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

1.03E-01 1.50E-01 3.75E-01 5.63E-01 6.00E-02 9.25E-02 3.07E-01 5.42E-01

Fraction of beta activity in urine: N Rx 1 Fraction of beta activity in urine: N Rx 2

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

2.17E-02 3.40E-02 1.68E-01 4.63E-01 4.41E-02 6.67E-02 2.58E-01 5.23E-01

Fraction of beta activity in urine: TRIGA 1 Fraction of beta activity in urine: TRIGA 2

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

1.98E-01 2.55E-01 4.76E-01 5.81E-01 2.17E-02 3.39E-02 1.67E-01 4.63E-01
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 19 of 70

Table 7-3a. Associated radionuclide activity fractions for assigning intakes: ATR 1.

Table 7-3 values: ATR 1

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Co-60 2.05E-04 2.03E-04 1.95E-04 1.85E-04 Co-60 2.03E-04 2.01E-04 1.93E-04 1.83E-04

Sr-89 6.72E+01 4.46E+01 6.61E+00 5.29E-01 Sr-89 6.66E+01 4.42E+01 6.54E+00 5.24E-01

Sr-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Sr-90 9.90E-01 9.90E-01 9.89E-01 9.89E-01

Y-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Y-90 9.93E-01 9.90E-01 9.90E-01 9.89E-01

Y-91 8.04E+01 5.64E+01 1.09E+01 1.23E+00 Y-91 7.96E+01 5.59E+01 1.07E+01 1.21E+00

Zr-95 8.67E+01 6.28E+01 1.39E+01 1.90E+00 Zr-95 8.59E+01 6.22E+01 1.38E+01 1.88E+00

Nb-95 7.49E+01 7.41E+01 2.66E+01 4.08E+00 Nb-95 7.42E+01 7.34E+01 2.63E+01 4.04E+00

Ru-103 4.50E+01 2.65E+01 2.26E+00 8.73E-02 Ru-103 4.45E+01 2.63E+01 2.24E+00 8.64E-02

Ru-106 1.74E+00 1.65E+00 1.28E+00 9.17E-01 Ru-106 1.72E+00 1.63E+00 1.27E+00 9.07E-01

I-131 1.21E+03 9.13E+01 5.29E-04 6.33E-11 I-131 1.20E+03 9.05E+01 5.23E-04 6.26E-11

Cs-134 7.48E-01 7.29E-01 6.47E-01 5.52E-01 Cs-134 7.41E-01 7.22E-01 6.40E-01 5.46E-01

Cs-137 1.01E+00 1.01E+00 1.01E+00 1.01E+00 Cs-137 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Ce-141 8.62E+01 4.55E+01 2.32E+00 4.54E-02 Ce-141 8.54E+01 4.51E+01 2.30E+00 4.49E-02

Ce-144 2.89E+01 2.69E+01 1.93E+01 1.25E+01 Ce-144 2.86E+01 2.66E+01 1.91E+01 1.23E+01

Pr-143 7.60E+01 1.65E+01 1.30E-02 1.04E-06 Pr-143 7.52E+01 1.63E+01 1.29E-02 1.03E-06

Pm-147 2.81E+00 2.96E+00 2.74E+00 2.42E+00 Pm-147 2.78E+00 2.94E+00 2.71E+00 2.40E+00

Eu-154 2.36E-02 2.35E-02 2.30E-02 2.23E-02 Eu-154 2.34E-02 2.32E-02 2.27E-02 2.21E-02

Table 7-3b. Associated radionuclide activity fractions for assigning intakes: ATR 2.

Table 7-3 values: ATR 2

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Co-60 2.07E-04 2.05E-04 1.97E-04 1.86E-04 Co-60 2.05E-04 2.03E-04 1.95E-04 1.85E-04

Sr-89 1.14E+02 7.61E+01 1.13E+01 9.02E-01 Sr-89 1.13E+02 7.53E+01 1.11E+01 8.92E-01

Sr-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Sr-90 9.91E-01 9.91E-01 9.90E-01 9.89E-01

Y-90 9.98E-01 1.00E+00 1.00E+00 1.00E+00 Y-90 9.90E-01 9.91E-01 9.90E-01 9.89E-01

Y-91 1.30E+02 9.16E+01 1.76E+01 1.99E+00 Y-91 1.29E+02 9.07E+01 1.74E+01 1.97E+00

Zr-95 1.34E+02 9.74E+01 2.16E+01 2.95E+00 Zr-95 1.33E+02 9.65E+01 2.14E+01 2.92E+00

Nb-95 6.28E+01 8.55E+01 3.94E+01 6.29E+00 Nb-95 6.22E+01 8.47E+01 3.91E+01 6.22E+00

Ru-103 8.53E+01 5.04E+01 4.29E+00 1.66E-01 Ru-103 8.45E+01 4.99E+01 4.25E+00 1.64E-01

Ru-106 1.88E+00 1.78E+00 1.38E+00 9.93E-01 Ru-106 1.86E+00 1.77E+00 1.37E+00 9.83E-01

I-131 4.24E+03 3.20E+02 1.85E-03 2.22E-10 I-131 4.21E+03 3.17E+02 1.83E-03 2.20E-10

Cs-134 5.40E-01 5.27E-01 4.68E-01 3.99E-01 Cs-134 5.36E-01 5.22E-01 4.63E-01 3.95E-01

Cs-137 1.01E+00 1.01E+00 1.01E+00 1.01E+00 Cs-137 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Ce-141 1.79E+02 9.46E+01 4.82E+00 9.44E-02 Ce-141 1.77E+02 9.37E+01 4.77E+00 9.33E-02

Ce-144 3.22E+01 3.00E+01 2.15E+01 1.39E+01 Ce-144 3.19E+01 2.97E+01 2.13E+01 1.38E+01

Pr-143 2.29E+02 4.97E+01 3.93E-02 3.13E-06 Pr-143 2.27E+02 4.92E+01 3.89E-02 3.10E-06

Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 2.02E-02 2.01E-02 1.97E-02 1.91E-02 Eu-154 2.00E-02 1.99E-02 1.95E-02 1.89E-02

Table 7-3c. Associated radionuclide activity fractions for assigning intakes: ATR 3.

Table 7-3 values: ATR 3

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Co-60 1.97E-04 1.95E-04 1.87E-04 1.77E-04 Co-60 1.95E-04 1.94E-04 1.86E-04 1.76E-04

Sr-89 1.02E+02 6.75E+01 9.99E+00 8.00E-01 Sr-89 1.01E+02 6.69E+01 9.90E+00 7.92E-01

Sr-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Sr-90 9.92E-01 9.92E-01 9.91E-01 9.90E-01

Y-90 9.97E-01 1.00E+00 1.00E+00 1.00E+00 Y-90 9.89E-01 9.92E-01 9.91E-01 9.91E-01

Y-91 1.16E+02 8.16E+01 1.57E+01 1.77E+00 Y-91 1.15E+02 8.09E+01 1.55E+01 1.76E+00

Zr-95 1.22E+02 8.84E+01 1.96E+01 2.68E+00 Zr-95 1.21E+02 8.77E+01 1.94E+01 2.65E+00

Nb-95 7.09E+01 8.52E+01 3.63E+01 5.72E+00 Nb-95 7.03E+01 8.46E+01 3.60E+01 5.66E+00

Ru-103 7.35E+01 4.34E+01 3.70E+00 1.43E-01 Ru-103 7.29E+01 4.30E+01 3.66E+00 1.41E-01

Ru-106 1.82E+00 1.72E+00 1.34E+00 9.61E-01 Ru-106 1.81E+00 1.71E+00 1.33E+00 9.52E-01

I-131 2.86E+03 2.16E+02 1.25E-03 1.50E-10 I-131 2.84E+03 2.14E+02 1.24E-03 1.48E-10

Cs-134 2.52E-01 2.46E-01 2.18E-01 1.86E-01 Cs-134 2.50E-01 2.44E-01 2.16E-01 1.85E-01
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 20 of 70

Table 7-3 values: ATR 3

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Cs-137 1.01E+00 1.01E+00 1.01E+00 1.01E+00 Cs-137 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Ce-141 1.51E+02 7.99E+01 4.07E+00 7.98E-02 Ce-141 1.50E+02 7.93E+01 4.04E+00 7.90E-02

Ce-144 3.15E+01 2.93E+01 2.10E+01 1.36E+01 Ce-144 3.12E+01 2.91E+01 2.09E+01 1.35E+01

Pr-143 1.71E+02 3.70E+01 2.92E-02 2.33E-06 Pr-143 1.69E+02 3.67E+01 2.90E-02 2.31E-06

Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00 Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 8.34E-03 8.30E-03 8.13E-03 7.90E-03 Eu-154 8.28E-03 8.24E-03 8.06E-03 7.82E-03

Table 7-3d. Associated radionuclide activity fractions for assigning intakes: FFTF 1.

Table 7-3 values: FFTF 1

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Co-60 5.79E-02 5.74E-02 5.51E-02 5.22E-02 Co-60 2.01E-02 1.99E-02 1.91E-02 1.81E-02

Sr-89 1.21E+01 8.04E+00 1.19E+00 9.52E-02 Sr-89 4.20E+00 2.78E+00 4.12E-01 3.29E-02

Sr-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Sr-90 3.46E-01 3.46E-01 3.46E-01 3.46E-01

Y-90 1.01E+00 1.00E+00 1.00E+00 1.00E+00 Y-90 3.48E-01 3.46E-01 3.46E-01 3.46E-01

Y-91 1.81E+01 1.27E+01 2.44E+00 2.76E-01 Y-91 6.25E+00 4.39E+00 8.44E-01 9.54E-02

Zr-95 3.26E+01 2.36E+01 5.24E+00 7.16E-01 Zr-95 1.13E+01 8.18E+00 1.81E+00 2.48E-01

Nb-95 3.60E+01 3.22E+01 1.03E+01 1.55E+00 Nb-95 1.25E+01 1.12E+01 3.56E+00 5.35E-01

Ru-103 4.51E+01 2.66E+01 2.27E+00 8.76E-02 Ru-103 1.56E+01 9.21E+00 7.85E-01 3.03E-02

Ru-106 2.64E+01 2.50E+01 1.95E+01 1.40E+01 Ru-106 9.14E+00 8.66E+00 6.74E+00 4.84E+00

I-131 6.25E+02 4.68E+01 2.63E-04 3.04E-11 I-131 2.16E+02 1.62E+01 9.11E-05 1.05E-11

Cs-134 3.31E+00 3.23E+00 2.87E+00 2.45E+00 Cs-134 1.15E+00 1.12E+00 9.92E-01 8.47E-01

Cs-137 2.89E+00 2.89E+00 2.89E+00 2.89E+00 Cs-137 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Ce-141 3.22E+01 1.70E+01 8.67E-01 1.70E-02 Ce-141 1.11E+01 5.89E+00 3.00E-01 5.88E-03

Ce-144 2.57E+01 2.39E+01 1.72E+01 1.11E+01 Ce-144 8.89E+00 8.28E+00 5.94E+00 3.83E+00

Pr-143 2.22E+01 4.81E+00 3.81E-03 3.03E-07 Pr-143 7.69E+00 1.67E+00 1.32E-03 1.05E-07

Pm-147 6.25E+00 6.21E+00 5.68E+00 5.03E+00 Pm-147 2.16E+00 2.15E+00 1.96E+00 1.74E+00

Eu-154 2.09E-01 2.08E-01 2.04E-01 1.98E-01 Eu-154 7.25E-02 7.21E-02 7.06E-02 6.85E-02

Table 7-3e. Associated radionuclide activity fractions for assigning intakes: FFTF 2.

Table 7-3 values: FFTF 2

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Co-60 5.17E-02 5.12E-02 4.92E-02 4.66E-02 Co-60 1.80E-02 1.78E-02 1.71E-02 1.62E-02

Sr-89 2.30E+01 1.52E+01 2.25E+00 1.80E-01 Sr-89 7.98E+00 5.30E+00 7.84E-01 6.26E-02

Sr-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Sr-90 3.48E-01 3.48E-01 3.48E-01 3.47E-01

Y-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Y-90 3.49E-01 3.48E-01 3.48E-01 3.47E-01

Y-91 3.41E+01 2.39E+01 4.60E+00 5.20E-01 Y-91 1.18E+01 8.32E+00 1.60E+00 1.81E-01

Zr-95 6.10E+01 4.41E+01 9.79E+00 1.34E+00 Zr-95 2.12E+01 1.53E+01 3.40E+00 4.64E-01

Nb-95 6.68E+01 5.99E+01 1.92E+01 2.89E+00 Nb-95 2.32E+01 2.08E+01 6.68E+00 1.00E+00

Ru-103 8.25E+01 4.87E+01 4.15E+00 1.60E-01 Ru-103 2.87E+01 1.69E+01 1.44E+00 5.56E-02

Ru-106 3.57E+01 3.38E+01 2.63E+01 1.89E+01 Ru-106 1.24E+01 1.18E+01 9.15E+00 6.56E+00

I-131 1.18E+03 8.87E+01 4.99E-04 5.75E-11 I-131 4.11E+02 3.08E+01 1.73E-04 2.00E-11

Cs-134 1.75E+00 1.71E+00 1.52E+00 1.30E+00 Cs-134 6.10E-01 5.95E-01 5.28E-01 4.50E-01

Cs-137 2.88E+00 2.88E+00 2.88E+00 2.88E+00 Cs-137 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Ce-141 6.05E+01 3.20E+01 1.63E+00 3.20E-02 Ce-141 2.10E+01 1.11E+01 5.67E-01 1.11E-02

Ce-144 3.72E+01 3.47E+01 2.49E+01 1.61E+01 Ce-144 1.29E+01 1.20E+01 8.65E+00 5.58E+00

Pr-143 4.13E+01 8.95E+00 7.08E-03 5.64E-07 Pr-143 1.44E+01 3.11E+00 2.46E-03 1.96E-07

Pm-147 7.82E+00 7.82E+00 7.15E+00 6.33E+00 Pm-147 2.72E+00 2.72E+00 2.49E+00 2.20E+00

Eu-154 1.37E-01 1.36E-01 1.33E-01 1.29E-01 Eu-154 4.75E-02 4.72E-02 4.62E-02 4.49E-02
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 21 of 70

Table 7-3f. Associated radionuclide activity fractions for assigning intakes: N Reactor 1.

Table 7-3 values: N Reactor 1

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Co-60 3.01E-03 2.98E-03 2.86E-03 2.71E-03 Co-60 2.73E-03 2.70E-03 2.59E-03 2.46E-03

Sr-89 7.29E+01 4.84E+01 7.17E+00 5.74E-01 Sr-89 6.60E+01 4.39E+01 6.49E+00 5.20E-01

Sr-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Sr-90 9.07E-01 9.07E-01 9.06E-01 9.05E-01

Y-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Y-90 9.06E-01 9.07E-01 9.06E-01 9.05E-01

Y-91 8.78E+01 6.17E+01 1.19E+01 1.34E+00 Y-91 7.96E+01 5.59E+01 1.07E+01 1.21E+00

Zr-95 9.90E+01 7.17E+01 1.59E+01 2.17E+00 Zr-95 8.97E+01 6.50E+01 1.44E+01 1.97E+00

Nb-95 8.10E+01 8.21E+01 3.02E+01 4.66E+00 Nb-95 7.34E+01 7.45E+01 2.74E+01 4.22E+00

Ru-103 6.58E+01 3.88E+01 3.31E+00 1.28E-01 Ru-103 5.96E+01 3.52E+01 3.00E+00 1.16E-01

Ru-106 4.07E+00 3.86E+00 3.00E+00 2.15E+00 Ru-106 3.69E+00 3.50E+00 2.72E+00 1.95E+00

I-131 1.62E+03 1.22E+02 7.07E-04 8.48E-11 I-131 1.47E+03 1.11E+02 6.41E-04 7.67E-11

Cs-134 1.15E-01 1.12E-01 9.95E-02 8.50E-02 Cs-134 1.04E-01 1.02E-01 9.02E-02 7.69E-02

Cs-137 1.10E+00 1.10E+00 1.10E+00 1.10E+00 Cs-137 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Ce-141 1.04E+02 5.51E+01 2.81E+00 5.49E-02 Ce-141 9.44E+01 4.99E+01 2.54E+00 4.97E-02

Ce-144 3.09E+01 2.87E+01 2.06E+01 1.33E+01 Ce-144 2.80E+01 2.60E+01 1.87E+01 1.21E+01

Pr-143 8.95E+01 1.94E+01 1.54E-02 1.22E-06 Pr-143 8.12E+01 1.76E+01 1.39E-02 1.11E-06

Pm-147 3.90E+00 4.10E+00 3.79E+00 3.36E+00 Pm-147 3.54E+00 3.72E+00 3.43E+00 3.04E+00

Eu-154 5.48E-03 5.45E-03 5.34E-03 5.19E-03 Eu-154 4.97E-03 4.94E-03 4.83E-03 4.69E-03

Table 7-3g. Associated radionuclide activity fractions for assigning intakes: N Reactor 2.

Table 7-3 values: N Reactor 2

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Co-60 3.10E-03 3.07E-03 2.95E-03 2.79E-03 Co-60 2.61E-03 2.59E-03 2.48E-03 2.35E-03

Sr-89 3.46E+01 2.30E+01 3.41E+00 2.73E-01 Sr-89 2.92E+01 1.94E+01 2.87E+00 2.29E-01

Sr-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Sr-90 8.42E-01 8.42E-01 8.42E-01 8.41E-01

Y-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Y-90 8.45E-01 8.42E-01 8.42E-01 8.41E-01

Y-91 4.47E+01 3.14E+01 6.03E+00 6.82E-01 Y-91 3.76E+01 2.64E+01 5.08E+00 5.74E-01

Zr-95 5.49E+01 3.98E+01 8.82E+00 1.21E+00 Zr-95 4.63E+01 3.35E+01 7.42E+00 1.01E+00

Nb-95 5.85E+01 5.31E+01 1.73E+01 2.60E+00 Nb-95 4.93E+01 4.47E+01 1.45E+01 2.18E+00

Ru-103 3.78E+01 2.23E+01 1.90E+00 7.34E-02 Ru-103 3.18E+01 1.88E+01 1.60E+00 6.17E-02

Ru-106 5.53E+00 5.24E+00 4.07E+00 2.92E+00 Ru-106 4.66E+00 4.41E+00 3.43E+00 2.46E+00

I-131 7.31E+02 5.52E+01 3.19E-04 3.82E-11 I-131 6.16E+02 4.64E+01 2.69E-04 3.22E-11

Cs-134 3.12E-01 3.04E-01 2.70E-01 2.31E-01 Cs-134 2.63E-01 2.56E-01 2.27E-01 1.94E-01

Cs-137 1.19E+00 1.19E+00 1.19E+00 1.19E+00 Cs-137 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Ce-141 4.80E+01 2.54E+01 1.29E+00 2.53E-02 Ce-141 4.04E+01 2.14E+01 1.09E+00 2.13E-02

Ce-144 2.64E+01 2.46E+01 1.77E+01 1.14E+01 Ce-144 2.22E+01 2.07E+01 1.49E+01 9.59E+00

Pr-143 3.68E+01 7.97E+00 6.31E-03 5.03E-07 Pr-143 3.10E+01 6.71E+00 5.31E-03 4.23E-07

Pm-147 3.87E+00 3.91E+00 3.59E+00 3.18E+00 Pm-147 3.26E+00 3.30E+00 3.02E+00 2.67E+00

Eu-154 1.51E-02 1.50E-02 1.47E-02 1.43E-02 Eu-154 1.27E-02 1.27E-02 1.24E-02 1.20E-02

Table 7-3h. Associated radionuclide activity fractions for assigning intakes: TRIGA 1.

Table 7-3 values: TRIGA 1

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Co-60 3.21E-01 3.18E-01 3.06E-01 2.89E-01 Co-60 3.13E-01 3.11E-01 2.98E-01 2.82E-01

Sr-89 5.15E+00 3.42E+00 5.06E-01 4.05E-02 Sr-89 5.02E+00 3.33E+00 4.93E-01 3.95E-02

Sr-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Sr-90 9.76E-01 9.76E-01 9.75E-01 9.75E-01

Y-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Y-90 9.77E-01 9.76E-01 9.76E-01 9.75E-01

Y-91 6.52E+00 4.58E+00 8.80E-01 9.95E-02 Y-91 6.36E+00 4.47E+00 8.59E-01 9.70E-02

Zr-95 7.89E+00 5.71E+00 1.27E+00 1.73E-01 Zr-95 7.70E+00 5.57E+00 1.23E+00 1.69E-01

Nb-95 8.70E+00 7.79E+00 2.49E+00 3.73E-01 Nb-95 8.49E+00 7.60E+00 2.43E+00 3.64E-01

Ru-103 3.38E+00 1.99E+00 1.70E-01 6.56E-03 Ru-103 3.30E+00 1.95E+00 1.66E-01 6.40E-03

Ru-106 6.18E-01 5.86E-01 4.56E-01 3.26E-01 Ru-106 6.03E-01 5.71E-01 4.44E-01 3.18E-01

I-131 8.05E+01 6.07E+00 3.51E-05 4.21E-12 I-131 7.85E+01 5.92E+00 3.43E-05 4.10E-12

Cs-134 1.25E-01 1.22E-01 1.08E-01 9.22E-02 Cs-134 1.22E-01 1.19E-01 1.05E-01 8.99E-02
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 22 of 70

Table 7-3 values: TRIGA 1

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Cs-137 1.03E+00 1.03E+00 1.03E+00 1.03E+00 Cs-137 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Ce-141 6.01E+00 3.18E+00 1.62E-01 3.17E-03 Ce-141 5.86E+00 3.10E+00 1.58E-01 3.09E-03

Ce-144 6.76E+00 6.29E+00 4.52E+00 2.92E+00 Ce-144 6.59E+00 6.14E+00 4.41E+00 2.84E+00

Pr-143 4.90E+00 1.06E+00 8.41E-04 6.70E-08 Pr-143 4.78E+00 1.04E+00 8.20E-04 6.53E-08

Pm-147 2.15E+00 2.13E+00 1.94E+00 1.72E+00 Pm-147 2.10E+00 2.08E+00 1.90E+00 1.68E+00

Eu-154 7.10E-03 7.06E-03 6.92E-03 6.72E-03 Eu-154 6.92E-03 6.89E-03 6.75E-03 6.55E-03

Table 7-3i. Associated radionuclide activity fractions for assigning intakes: TRIGA 2.

Table 7-3 values: TRIGA 2

Nuclide

Intake relative to Sr-90

Nuclide

Intake relative to Cs-137

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

Co-60 3.88E-01 3.84E-01 3.69E-01 3.49E-01 Co-60 3.84E-01 3.81E-01 3.65E-01 3.46E-01

Sr-89 7.33E+01 4.87E+01 7.21E+00 5.78E-01 Sr-89 7.27E+01 4.83E+01 7.14E+00 5.72E-01

Sr-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Sr-90 9.91E-01 9.91E-01 9.90E-01 9.90E-01

Y-90 1.00E+00 1.00E+00 1.00E+00 1.00E+00 Y-90 9.94E-01 9.91E-01 9.91E-01 9.90E-01

Y-91 8.69E+01 6.11E+01 1.17E+01 1.33E+00 Y-91 8.61E+01 6.05E+01 1.16E+01 1.31E+00

Zr-95 9.89E+01 7.16E+01 1.59E+01 2.17E+00 Zr-95 9.80E+01 7.09E+01 1.57E+01 2.15E+00

Nb-95 8.07E+01 8.19E+01 3.02E+01 4.65E+00 Nb-95 8.00E+01 8.12E+01 2.99E+01 4.60E+00

Ru-103 4.98E+01 2.94E+01 2.50E+00 9.68E-02 Ru-103 4.94E+01 2.91E+01 2.48E+00 9.57E-02

Ru-106 1.75E+00 1.66E+00 1.29E+00 9.23E-01 Ru-106 1.73E+00 1.64E+00 1.27E+00 9.13E-01

I-131 1.41E+03 1.06E+02 6.15E-04 7.37E-11 I-131 1.40E+03 1.05E+02 6.09E-04 7.29E-11

Cs-134 8.51E-03 8.30E-03 7.37E-03 6.29E-03 Cs-134 8.44E-03 8.22E-03 7.29E-03 6.22E-03

Cs-137 1.01E+00 1.01E+00 1.01E+00 1.01E+00 Cs-137 1.00E+00 1.00E+00 1.00E+00 1.00E+00

Ce-141 9.71E+01 5.13E+01 2.61E+00 5.12E-02 Ce-141 9.62E+01 5.08E+01 2.59E+00 5.07E-02

Ce-144 2.94E+01 2.74E+01 1.97E+01 1.27E+01 Ce-144 2.92E+01 2.72E+01 1.95E+01 1.26E+01

Pr-143 8.88E+01 1.92E+01 1.52E-02 1.21E-06 Pr-143 8.80E+01 1.91E+01 1.51E-02 1.20E-06

Pm-147 3.72E+00 3.91E+00 3.61E+00 3.20E+00 Pm-147 3.69E+00 3.87E+00 3.57E+00 3.16E+00

Eu-154 3.72E-04 3.70E-04 3.62E-04 3.52E-04 Eu-154 3.69E-04 3.67E-04 3.59E-04 3.49E-04

air concentration and assumed breathing rate. Table 7-3 is used to assign intakes for the other 

radionuclides in the mix once the indicator radionuclide intake is established. 

8.0 MIXED FISSION AND ACTIVATION PRODUCT DOSE ASSIGNMENT SUMMARY

This document provides activity and intake fraction values for use in interpreting gross beta or gross 

gamma activity measurements related to reactor source terms. The values were derived using 

representative parameters and models; site- and claim-specific information, if available, should take 

precedence over the guidance in this document. This includes site-specific models and guidance in 

Site Profiles.

Individual monitoring data should be used to determine doses from specific radionuclides if available. 

Site Profile and claim-specific information should be reviewed to determine if chronic iodine intakes 

were feasible. 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. 

The methods used in this document to compute intake ratios resulted in much larger values for 131I at 

short decay times than for particulates. However, particulates were more significant sources of 

internal exposure at most sites. 

To apply this document, the activity fraction data from Table 7-1a/b, 7-2, or 7-4a/b is applied as 

appropriate to determine the indicator radionuclide activity values associated with a given gross beta 

or gross gamma assay result. Selection of decay times should be based on the worker categories in 

Table 5-3 and consideration of site and claim information. If a specific decay time or times cannot be 

established, the dose evaluation will need to consider all four possibilities. Similarly, in most 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 23 of 70

Table 7-4a. Activity ratios for use if gross beta results are reported for air samples and workplace measurements.

Radionuclide Activity fractions for gross beta air or swipe samplesa

Sr-90

(gross beta)

Fraction of beta activity on sample: ATR 1 Fraction of beta activity on sample: ATR 2 Fraction of beta activity on sample: ATR 3

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

1.37E-03 2.66E-03 9.74E-03 2.05E-02 6.29E-04 1.51E-03 7.51E-03 1.80E-02 7.82E-04 1.75E-03 7.99E-03 1.85E-02

Fraction of beta activity on sample: FFTF 1 Fraction of beta activity on sample: FFTF 2

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

2.62E-03 4.12E-03 8.58E-03 1.33E-02 1.58E-03 2.61E-03 6.11E-03 9.88E-03

Fraction of beta activity on sample: N Rx 1 Fraction of beta activity on sample: N Rx 2

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

1.15E-03 2.28E-03 8.64E-03 1.82E-02 2.19E-03 3.86E-03 1.12E-02 2.05E-02

Fraction of beta activity on sample: TRIGA 1 Fraction of beta activity on sample: TRIGA 2

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

1.35E-02 2.09E-02 4.37E-02 6.73E-02 1.22E-03 2.41E-03 9.16E-03 1.97E-02

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.

Table 7-4b. Activity ratios for use if gross gamma results are reported for air samples and workplace measurements.

Radionuclide Activity fractions for gross gamma air or swipe samplesa

Cs-137

(gross gamma

Fraction of gamma activity on sample: ATR 1 Fraction of gamma activity on sample: ATR 2 Fraction of gamma activity on sample: ATR 3

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

1.89E-03 3.91E-03 1.83E-02 8.58E-02 8.35E-04 2.22E-03 1.24E-02 6.57E-02 1.06E-03 2.58E-03 1.36E-02 7.31E-02

Fraction of gamma activity on sample: FFTF 1 Fraction of gamma activity on sample: FFTF 2

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

9.97E-03 1.82E-02 6.31E-02 1.37E-01 5.82E-03 1.09E-02 4.39E-02 1.22E-01

Fraction of gamma activity on sample: N Rx 1 Fraction of gamma activity on sample: N Rx 2

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

1.71E-03 3.62E-03 1.77E-02 8.97E-02 3.57E-03 6.90E-03 3.10E-02 1.28E-01

Fraction of gamma activity on sample: TRIGA 1 Fraction of gamma activity on sample: TRIGA 2

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

2.24E-02 4.01E-02 1.35E-01 3.15E-01 1.67E-03 3.47E-03 1.62E-02 7.99E-02

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.
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 24 of 70

instances all nine of the representative reactor cases will need to be considered. If the dose 

reconstructor does not know if significant separations chemistry was performed prior to a gross beta 

urinalysis, it is favorable to the claimant to assume significant chemistry was performed. The activity 

fraction data from Table 7-2 should therefore be used for those cases. If both gross beta and gross 

gamma results are available for a given assay then the more sensitive measure [i.e., that with the 

lowest minimum detectable activity (MDA)] should be used. If the dose reconstructor does not know if 

a given assay result was a gross beta or gross gamma count (and both are realistic choices for the 

site under consideration) then both should be evaluated to determine which yields the larger assigned 

dose.

Corresponding radionuclide intakes are determined (using IMBA and activity fraction ratios) once the 

appropriate number of indicator radionuclide activity values are established. Table 7-3 is then used to 

establish the corresponding intakes for the other radionuclides in the mix. The goal is to find the 

combination of reactor, decay time, and absorption type assumptions that result in the most favorable 

dose being assigned. Recall that 89Sr and 90Sr are always treated as type F and 131I is always type V 

when applying this document.

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

unprocessed gross beta bioassay data, and air-sampling data.

In general, the information in this document is intended for application to gross beta or gross gamma 

assays. In some cases, it could be applied to a radionuclide-specific assay such as a 90Sr bioassay or 

an in vivo count as long as the assumptions and bases employed in developing the associated 

radionuclide activity fractions in Tables 7-3a through 7-3i are considered and deemed appropriate. 

Such considerations would be site or case specific and are outside the scope of the guidance in this 

document for gross beta and gross gamma assays.

If a site used a specific chemical recovery fraction to state gross beta counting results in terms of 

activity for a specific radionuclide, the fraction should be multiplied by the reported result to convert it 

back to gross beta activity. For example, gross beta results at Hanford were divided by a chemical 

recovery for 90Sr and reported as 90Sr-equivalent activity. The 90Sr-equivalent result should be 

multiplied by the recovery fraction to get back to gross beta activity before applying the methods in 

this document. However, if the dose reconstructor does not know the recovery fraction, then it is 

favorable to the claimant to treat the result as gross beta activity. 

8.1 INTERACTIVE RADIOEPIDEMIOLOGICAL PROGRAM (IREP) DISTRIBUTION 

ASSIGNMENT

The indicator radionuclide doses calculated as above are entered into IREP with the distribution 

appropriate for the manner in which the intake was determined. For example, if there were positive 

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

a geometric standard deviation equal to 3 would be assigned (ORAUT 2007). For missed dose, a 

triangular distribution is assigned (ORAUT 2007).

Given the conservatism in how they are assigned, the doses from the associated radionuclides should 

be entered as constants in IREP. Measures in this document to ensure favorability to claimants in 

relation to assigned doses include the following:

• Not accounting for reactor power history in the calculation of radionuclide inventories 

maximizes the short-lived radionuclide content in relation to the long-lived content. 

• Increasing the release fractions for particulates by an order of magnitude higher than those in 

DOE Standard 1027 (DOE 1997) increases the intake fractions for those radionuclides.
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 25 of 70

• Renormalizing the Table D-1 and E-1 mixtures to reduce the number of radionuclides 

considered further increases the intake fractions for the dosimetrically important radionuclides.

• Computing indicator radionuclide activity fractions using only the most insoluble form of the 

other radionuclides in the mix maximizes those values and, thus, the assigned intakes.

• The dose assigned for a given claim is the maximum obtained by considering all possible 

absorption types for all nine representative reactor cases and in some instances across all four 

decay times (if there is not sufficient information available to select a specific decay time or 

times).

In addition to ensuring favorability to the claimants, the ranges of irradiation parameters (specific 

power, exposure time, etc.) to establish the fission and activation product inventory data for the 

representative reactor cases in conjunction with the conservatisms and the maximizing approach 

discussed above were used to encompass uncertainties associated with dose assignments per this 

document. Such uncertainties 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.

Doses assigned per this document are likely to represent upper bounds. Therefore, it would be 

reasonable to assign them as such (i.e., constant distributions) and to ignore any further uncertainty 

from metabolic models, etc.

When intakes are determined using urinalysis data and other bioassay data are available (e.g., wholebody counts), the reasonableness of the intakes can be checked. Predicted body burdens of gammaemitting radionuclides based on the urine-based intakes for those that were also measured in whole 

body counts or would have been reported if they had been detected, and had well-defined detection 

levels, can be compared to levels that were measured in the whole-body counts or to MDAs, as 

appropriate. If predicted versus measured body burdens do not agree well, users should discuss the 

proper course of action with their Group Manager or the Principal Internal Dosimetrist. In general, 

these instances would be an issue for best-estimate cases or those in which the assigned dose from 

application of this document could have a significant effect on the probability of causation.

9.0 ATTRIBUTIONS AND ANNOTATIONS

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 (SRDB).
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 26 of 70

[1] Burns Jr., Robert E., CHP. ORAU Team. Senior Health Physicist. March 2007.

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

were evaluated relative to 137Cs as a convenience. This is a common practice because 137Cs 

is a prominent fission product and is easily quantified. 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 27 of 70

REFERENCES

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. [SRDB Ref ID: 33208]

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

Radionuclides of Workers, Publication 68, Oxford, England.

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.

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

Tennessee, December 31. 
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 28 of 70

The following radionuclides are those identified in the collective set of fission and activation product 

inventory data for the nine representative reactor cases for which dose conversion factors were not 

readily available

Table A-1. Radionuclides from reactor 

source terms not included in analyses.

Al-28 Cd-111m

Ca-48 Cd-116

Sc-45m In-114

V-50 Sb-118

Cu-66 Te-118

Ge-71m Te-119

Ge-73m Te-128

Ge-76 Ba-136m

As-75m Ba-137m

Se-72 La-133

Se-77m Ce-139m

Se-82 Pr-140

Y-87m Pr-144m

Y-89m Nd-140

Nb-91 Nd-144

Nb-91m Nd-150

Nb-92 Sm-148

Nb-92m Gd-153m

Nb-97m Ho-163

Mo-100 Er-167m

Rh-105m Tm-165

Rh-106 Tm-168

Pd-111 Hf-174

Pd-111m W-180

Pd-112 W-183m

Ag-107m Os-186

Ag-108 Ir-191m

Ag-109m Pt-190

Ag-110 Tl-206

Ag-111m Bi-208

Ag-113

ATTACHMENT A

RADIONUCLIDES FROM REACTOR SOURCE TERMS NOT INCLUDED IN DOSE ANALYSES
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 29 of 70

ATTACHMENT B

RELEASE FRACTIONS

These fractions, adapted from DOE Standard 1027 (DOE 1997), were used to calculate intake 

fractions from the fission and activation product inventory data for the nine representative reactor 

cases.

Table B-1. Elemental exposure fractions.

Element

Release 

fraction Element

Release 

fraction

Ag 0.01 Na 0.05

As 0.05 Nb 0.01

Au 0.01 Nd 0.01

Ba 0.01 Ni 0.01

Be 0.01 Os 0.01

Bi 0.01 P 0.05

Br 0.05 Pb 0.01

C 0.01 Pd 0.01

Ca 0.01 Pm 0.01

Cd 0.01 Po 0.01

Ce 0.01 Pr 0.01

Cl 1 Pt 0.01

Co 0.01 Rb 0.01

Cr 0.01 Re 0.01

Cs 0.01 Rh 0.01

Cu 0.01 Ru 0.01

Dy 0.01 S 0.05

Er 0.01 Sb 0.01

Eu 0.01 Sc 0.01

Fe 0.01 Se 0.01

Ga 0.01 Si 0.01

Gd 0.01 Sm 0.01

Ge 0.01 Sn 0.01

H 1 Sr 0.01

Hf 0.01 Ta 0.01

Hg 0.01 Tb 0.01

Ho 0.01 Tc 0.01

I 0.05 Te 0.01

In 0.01 Tl 0.01

Ir 0.01 Tm 0.01

K 0.05 V 0.01

La 0.01 W 0.01

Lu 0.01 Y 0.01

Mg 0.01 Yb 0.01

Mn 0.01 Zn 0.01

Mo 0.01 Zr 0.01
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 30 of 70

Table C-1. Radionuclides considered in committed 

organ dose calculations and the absorption types 

assigned to each solubility category.

Radionuclide

Absorption type assigned

Soluble Mod. soluble Insoluble

H-3 V (HTO) V (HTO) V (HTO)

Be-7 M M S

Be-10 M M S

C-14 V (CO2) V (CO2) V (CO2)

Na-22 F F F

Na-24 F F F

Mg-28 F M M

Si-32 F M S

P-32 F M M

P-33 F M M

S-35 F M M

Cl-36 F M M

K-40 F F F

K-42 F F F

K-43 F F F

Ca-41 M M M

Ca-45 M M M

Ca-47 M M M

Sc-46 S S S

Sc-47 S S S

Sc-48 S S S

V-49 F M M

Cr-51 F M S

Mn-54 F M M

Fe-55 F M M

Fe-59 F M M

Co-58 M M S

Co-58m M M S

Co-60 M M S

Ni-59 F M M

Ni-63 F M M

Ni-66 F M M

Cu-64 F M S

Cu-67 F M S

Zn-65 S S S

Zn-69 S S S

Zn-69m S S S

Zn-71m S S S

Zn-72 S S S

Ga-66 F M M

Ga-67 F M M

Ga-68 F M M

Ga-72 F M M

ATTACHMENT C

RADIONUCLIDES CONSIDERED IN COMMITTED ORGAN DOSE CALCULATIONS AND THE 

ABSORPTION TYPES ASSIGNED TO EACH SOLUBILITY CATEGORY

Page 1 of 7
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 31 of 70

Radionuclide

Absorption type assigned

Soluble Mod. soluble Insoluble

Ga-73 F M M

Ge-68 F M M

Ge-69 F M M

Ge-71 F M M

Ge-77 F M M

As-71 M M M

As-72 M M M

As-73 M M M

As-74 M M M

As-76 M M M

As-77 M M M

Se-73 F M M

Se-75 F M M

Se-79 F M M

Br-77 F M M

Br-80 F M M

Br-80m F M M

Br-82 F M M

Rb-81 F F F

Rb-83 F F F

Rb-84 F F F

Rb-86 F F F

Rb-87 F F F

Sr-83 F F F

Sr-85 F F F

Sr-89 F F F

Sr-90 F F F

Sr-91 F F F

Y-87 M M S

Y-88 M M S

Y-90 M M S

Y-91 M M S

Y-91m M M S

Y-92 M M S

Y-93 M M S

Zr-88 F M S

Zr-89 F M S

Zr-93 F M S

Zr-95 F M S

Zr-97 F M S

Nb-90 M M S

Nb-93m M M S

Nb-94 M M S

Nb-95 M M S

Nb-95m M M S

Nb-96 M M S

ATTACHMENT C

RADIONUCLIDES CONSIDERED IN COMMITTED ORGAN DOSE CALCULATIONS AND THE 

ABSORPTION TYPES ASSIGNED TO EACH SOLUBILITY CATEGORY

Page 2 of 7
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 32 of 70

Radionuclide

Absorption type assigned

Soluble Mod. soluble Insoluble

Nb-97 M M S

Mo-93 F F S

Mo-93m F F S

Mo-99 F F S

Tc-95 F M M

Tc-95m F M M

Tc-96 F M M

Tc-97 F M M

Tc-97m F M M

Tc-98 F M M

Tc-99 F M M

Tc-99m F M M

Ru-97 F M S

Ru-103 F M S

Ru-105 F M S

Ru-106 F M S

Rh-99 F M S

Rh-99m F M S

Rh-101 F M S

Rh-101m F M S

Rh-102 F M S

Rh-102m F M S

Rh-103m F M S

Rh-105 F M S

Pd-101 F M S

Pd-103 F M S

Pd-107 F M S

Pd-109 F M S

Ag-105 F M S

Ag-106m F M S

Ag-108m F M S

Ag-110m F M S

Ag-111 F M S

Ag-112 F M S

Cd-107 F M S

Cd-109 F M S

Cd-113 F M S

Cd-113m F M S

Cd-115 F M S

Cd-115m F M S

Cd-117m F M S

In-109 F M M

In-111 F M M

In-113m F M M

In-114m F M M

In-115 F M M

ATTACHMENT C

RADIONUCLIDES CONSIDERED IN COMMITTED ORGAN DOSE CALCULATIONS AND THE 

ABSORPTION TYPES ASSIGNED TO EACH SOLUBILITY CATEGORY

Page 3 of 7
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 33 of 70

Radionuclide

Absorption type assigned

Soluble Mod. soluble Insoluble

In-115m F M M

In-117 F M M

In-117m F M M

Sn-113 F M M

Sn-117m F M M

Sn-119m F M M

Sn-121 F M M

Sn-121m F M M

Sn-123 F M M

Sn-125 F M M

Sn-126 F M M

Sb-118m F M M

Sb-119 F M M

Sb-120m F M M

Sb-122 F M M

Sb-124 F M M

Sb-125 F M M

Sb-126 F M M

Sb-126m F M M

Sb-127 F M M

Sb-128 F M M

Sb-129 F M M

Te-121 F M M

Te-121m F M M

Te-123 F M M

Te-123m F M M

Te-125m F M M

Te-127 F M M

Te-127m F M M

Te-129 F M M

Te-129m F M M

Te-131 F M M

Te-131m F M M

Te-132 F M M

I-123 V (I2) V (I2) V (I2)

I-124 V (I2) V (I2) V (I2)

I-125 V (I2) V (I2) V (I2)

I-126 V (I2) V (I2) V (I2)

I-129 V (I2) V (I2) V (I2)

I-130 V (I2) V (I2) V (I2)

I-131 V (I2) V (I2) V (I2)

I-132 V (I2) V (I2) V (I2)

I-133 V (I2) V (I2) V (I2)

I-135 V (I2) V (I2) V (I2)

Cs-127 F F F

Cs-129 F F F

ATTACHMENT C

RADIONUCLIDES CONSIDERED IN COMMITTED ORGAN DOSE CALCULATIONS AND THE 

ABSORPTION TYPES ASSIGNED TO EACH SOLUBILITY CATEGORY

Page 4 of 7
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 34 of 70

Radionuclide

Absorption type assigned

Soluble Mod. soluble Insoluble

Cs-131 F F F

Cs-132 F F F

Cs-134 F F F

Cs-135 F F F

Cs-136 F F F

Cs-137 F F F

Ba-131 F F F

Ba-133 F F F

Ba-133m F F F

Ba-135m F F F

Ba-140 F F F

La-135 F M M

La-137 F M M

La-138 F M M

La-140 F M M

La-141 F M M

Ce-135 M M S

Ce-137 M M S

Ce-137m M M S

Ce-139 M M S

Ce-141 M M S

Ce-143 M M S

Ce-144 M M S

Pr-139 M M S

Pr-142 M M S

Pr-143 M M S

Pr-144 M M S

Pr-145 M M S

Nd-147 M M S

Pm-143 M M S

Pm-144 M M S

Pm-145 M M S

Pm-146 M M S

Pm-147 M M S

Pm-148 M M S

Pm-148m M M S

Pm-149 M M S

Pm-151 M M S

Sm-145 M M M

Sm-146 M M M

Sm-147 M M M

Sm-151 M M M

Sm-153 M M M

Sm-156 M M M

Eu-147 M M M

Eu-149 M M M

ATTACHMENT C

RADIONUCLIDES CONSIDERED IN COMMITTED ORGAN DOSE CALCULATIONS AND THE 

ABSORPTION TYPES ASSIGNED TO EACH SOLUBILITY CATEGORY

Page 5 of 7
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 35 of 70

Radionuclide

Absorption type assigned

Soluble Mod. soluble Insoluble

Eu-152 M M M

Eu-152m M M M

Eu-154 M M M

Eu-155 M M M

Eu-156 M M M

Eu-157 M M M

Gd-147 F M M

Gd-149 F M M

Gd-151 F M M

Gd-152 F M M

Gd-153 F M M

Gd-159 F M M

Tb-153 M M M

Tb-155 M M M

Tb-156 M M M

Tb-156m M M M

Tb-157 M M M

Tb-158 M M M

Tb-160 M M M

Tb-161 M M M

Dy-155 M M M

Dy-157 M M M

Dy-159 M M M

Dy-166 M M M

Ho-166 M M M

Ho-166m M M M

Er-165 M M M

Er-169 M M M

Er-171 M M M

Er-172 M M M

Tm-166 M M M

Tm-167 M M M

Tm-170 M M M

Tm-171 M M M

Tm-172 M M M

Tm-173 M M M

Yb-166 M M S

Yb-169 M M S

Yb-175 M M S

Lu-169 M M S

Lu-171 M M S

Lu-172 M M S

Lu-176 M M S

Lu-176m M M S

Lu-177 M M S

Lu-177m M M S

ATTACHMENT C

RADIONUCLIDES CONSIDERED IN COMMITTED ORGAN DOSE CALCULATIONS AND THE 

ABSORPTION TYPES ASSIGNED TO EACH SOLUBILITY CATEGORY

Page 6 of 7
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Radionuclide

Absorption type assigned

Soluble Mod. soluble Insoluble

Hf-175 F M M

Hf-180m F M M

Hf-181 F M M

Hf-182 F M M

Ta-180 M M M

Ta-182 M M M

Ta-183 M M M

W-181 F F F

W-185 F F F

W-187 F F F

W-188 F F F

Re-186 F M M

Re-187 F M M

Re-188 F M M

Re-189 F M M

Os-185 F M S

Os-189m F M S

Os-191 F M S

Os-191m F M S

Os-193 F M S

Os-194 F M S

Ir-192 F M S

Ir-192m F M S

Ir-193m F M S

Ir-194 F M S

Ir-194m F M S

Pt-193 F F F

Pt-193m F F F

Pt-195m F F F

Pt-197 F F F

Au-198 F M S

Au-199 F M S

Hg-197 F M M

Hg-197m F M M

Hg-203 F M M

Tl-204 F F F

Pb-205 F F F

Bi-210 F M M

Bi-210m F M M

Po-210 F M M

ATTACHMENT C

RADIONUCLIDES CONSIDERED IN COMMITTED ORGAN DOSE CALCULATIONS AND THE 

ABSORPTION TYPES ASSIGNED TO EACH SOLUBILITY CATEGORY

Page 7 of 7
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Table D-1. NIFs for radionuclides that contribute ≥1% dose to committed organ dose. 

Radionuclide

ATR 1 ATR 2 ATR 3

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

Mn-54 1.91E-07 8.69E-07 3.04E-06 5.59E-06 6.77E-08 4.70E-07 2.41E-06 5.15E-06 9.30E-08 5.56E-07 2.49E-06 5.12E-06

Fe-55 3.33E-06 1.58E-05 6.85E-05 1.67E-04 1.11E-06 8.05E-06 5.12E-05 1.45E-04 1.49E-06 9.34E-06 5.18E-05 1.41E-04

Co-58 2.12E-07 7.66E-07 9.29E-07 4.23E-07 8.03E-08 4.45E-07 7.93E-07 4.19E-07 2.10E-07 1.00E-06 1.56E-06 7.93E-07

Co-60 8.21E-08 3.94E-07 1.79E-06 4.64E-06 2.68E-08 1.97E-07 1.31E-06 3.95E-06 3.65E-08 2.31E-07 1.34E-06 3.89E-06

Sr-89 2.69E-02 8.67E-02 6.06E-02 1.33E-02 1.48E-02 7.30E-02 7.50E-02 1.91E-02 1.88E-02 7.99E-02 7.16E-02 1.76E-02

Sr-90 4.01E-04 1.94E-03 9.17E-03 2.51E-02 1.30E-04 9.60E-04 6.66E-03 2.12E-02 1.85E-04 1.18E-03 7.16E-03 2.20E-02

Y-90 4.02E-04 1.94E-03 9.18E-03 2.51E-02 1.29E-04 9.61E-04 6.66E-03 2.12E-02 1.85E-04 1.18E-03 7.17E-03 2.20E-02

Y-91 3.22E-02 1.10E-01 9.95E-02 3.08E-02 1.69E-02 8.79E-02 1.17E-01 4.22E-02 2.15E-02 9.66E-02 1.12E-01 3.89E-02

Zr-95 3.48E-02 1.22E-01 1.28E-01 4.78E-02 1.74E-02 9.35E-02 1.44E-01 6.25E-02 2.26E-02 1.05E-01 1.40E-01 5.88E-02

Nb-95 3.00E-02 1.44E-01 2.44E-01 1.03E-01 8.13E-03 8.21E-02 2.63E-01 1.33E-01 1.31E-02 1.01E-01 2.60E-01 1.26E-01

Mo-99 3.78E-03 9.46E-06 2.05E-20 3.03E-40 4.47E-03 1.71E-05 5.46E-20 9.36E-40 4.18E-03 1.38E-05 3.83E-20 6.33E-40

Ru-103 1.80E-02 5.15E-02 2.07E-02 2.19E-03 1.10E-02 4.83E-02 2.86E-02 3.51E-03 1.36E-02 5.14E-02 2.65E-02 3.13E-03

Ru-106 6.96E-04 3.19E-03 1.17E-02 2.30E-02 2.44E-04 1.71E-03 9.23E-03 2.10E-02 3.38E-04 2.04E-03 9.61E-03 2.11E-02

Cd-113m 7.23E-10 3.49E-09 1.63E-08 4.42E-08 2.88E-10 2.13E-09 1.46E-08 4.59E-08 1.04E-09 6.62E-09 3.97E-08 1.20E-07

Cd-115m 2.79E-06 8.49E-06 4.59E-06 7.19E-07 1.60E-06 7.47E-06 5.94E-06 1.08E-06 2.02E-06 8.09E-06 5.61E-06 9.81E-07

Sb-125 2.32E-05 1.12E-04 4.84E-04 1.18E-03 7.43E-06 5.66E-05 3.62E-04 1.02E-03 1.07E-05 6.92E-05 3.85E-04 1.05E-03

Te-129m 5.89E-04 1.54E-03 4.08E-04 2.49E-05 3.89E-04 1.56E-03 6.07E-04 4.30E-05 4.69E-04 1.61E-03 5.49E-04 3.75E-05

Te-132 3.95E-03 3.24E-05 1.79E-17 4.01E-34 4.67E-03 5.87E-05 4.76E-17 1.24E-33 4.36E-03 4.72E-05 3.34E-17 8.35E-34

I-131 4.85E-01 1.77E-01 4.85E-06 1.59E-12 5.50E-01 3.08E-01 1.23E-05 4.70E-12 5.31E-01 2.56E-01 8.95E-06 3.29E-12

I-132 2.04E-01 1.67E-03 9.23E-16 2.06E-32 2.41E-01 3.02E-03 2.45E-15 6.36E-32 2.25E-01 2.43E-03 1.72E-15 4.30E-32

Cs-134 3.00E-04 1.42E-03 5.93E-03 1.39E-02 7.00E-05 5.06E-04 3.12E-03 8.46E-03 4.68E-05 2.91E-04 1.56E-03 4.09E-03

Cs-136 8.78E-05 8.77E-05 2.62E-07 4.27E-11 4.01E-05 6.13E-05 2.69E-07 5.08E-11 4.70E-05 6.20E-05 2.38E-07 4.32E-11

Cs-137 4.05E-04 1.96E-03 9.27E-03 2.54E-02 1.31E-04 9.69E-04 6.73E-03 2.14E-02 1.87E-04 1.19E-03 7.23E-03 2.22E-02

Ba-140 2.81E-02 2.67E-02 6.30E-05 7.49E-09 2.85E-02 4.14E-02 1.44E-04 1.98E-08 2.93E-02 3.67E-02 1.11E-04 1.48E-08

La-140 3.23E-02 3.08E-02 7.25E-05 8.63E-09 3.27E-02 4.77E-02 1.65E-04 2.28E-08 3.36E-02 4.23E-02 1.28E-04 1.70E-08

Ce-141 3.45E-02 8.84E-02 2.13E-02 1.14E-03 2.32E-02 9.08E-02 3.21E-02 2.00E-03 2.81E-02 9.47E-02 2.92E-02 1.75E-03

Ce-144 1.16E-02 5.22E-02 1.77E-01 3.13E-01 4.17E-03 2.88E-02 1.44E-01 2.94E-01 5.84E-03 3.47E-02 1.51E-01 2.98E-01

Pr-143 3.05E-02 3.20E-02 1.19E-04 2.61E-08 2.97E-02 4.77E-02 2.62E-04 6.63E-08 3.16E-02 4.38E-02 2.10E-04 5.12E-08

Pr-144 1.16E-02 5.22E-02 1.77E-01 3.13E-01 4.17E-03 2.88E-02 1.44E-01 2.94E-01 5.84E-03 3.47E-02 1.51E-01 2.98E-01

Nd-147 8.74E-03 6.39E-03 4.42E-06 1.04E-10 8.40E-03 9.40E-03 9.54E-06 2.60E-10 9.48E-03 9.13E-03 8.09E-06 2.13E-10

Pm-147 1.13E-03 5.75E-03 2.51E-02 6.09E-02 3.17E-04 2.90E-03 1.91E-02 5.36E-02 5.81E-04 4.22E-03 2.39E-02 6.48E-02

Pm-148m 1.78E-04 5.24E-04 2.38E-04 2.96E-05 4.13E-05 1.85E-04 1.24E-04 1.78E-05 8.72E-05 3.37E-04 1.96E-04 2.73E-05

Sm-151 1.23E-06 5.95E-06 2.83E-05 7.82E-05 4.84E-07 3.60E-06 2.51E-05 8.05E-05 1.41E-06 8.99E-06 5.47E-05 1.69E-04

Eu-154 9.45E-06 4.56E-05 2.11E-04 5.61E-04 2.61E-06 1.93E-05 1.31E-04 4.05E-04 1.55E-06 9.83E-06 5.82E-05 1.73E-04

Eu-155 5.47E-06 2.62E-05 1.18E-04 3.04E-04 1.58E-06 1.16E-05 7.67E-05 2.29E-04 2.49E-06 1.57E-05 9.07E-05 2.61E-04

Ta-182 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

ATTACHMENT D

NORMALIZED INTAKE FRACTIONS FOR THE DOSIMETRICALLY IMPORTANT RADIONUCLIDES

Page 1 of 4
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Radionuclide

FFTF 1 FFTF 2

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

Mn-54 1.17E-03 4.77E-03 1.05E-02 1.22E-02 8.15E-04 3.61E-03 9.23E-03 1.18E-02

Fe-55 1.57E-03 6.69E-03 1.82E-02 2.82E-02 8.60E-04 3.99E-03 1.26E-02 2.14E-02

Co-58 7.42E-03 2.40E-02 1.83E-02 5.28E-03 6.40E-03 2.26E-02 2.01E-02 6.33E-03

Co-60 4.69E-05 2.02E-04 5.76E-04 9.47E-04 2.28E-05 1.07E-04 3.54E-04 6.39E-04

Sr-89 9.82E-03 2.83E-02 1.24E-02 1.73E-03 1.01E-02 3.18E-02 1.63E-02 2.48E-03

Sr-90 8.10E-04 3.51E-03 1.05E-02 1.82E-02 4.41E-04 2.09E-03 7.21E-03 1.37E-02

Y-90 8.15E-04 3.51E-03 1.05E-02 1.82E-02 4.43E-04 2.09E-03 7.21E-03 1.37E-02

Y-91 1.46E-02 4.46E-02 2.55E-02 5.01E-03 1.50E-02 4.99E-02 3.31E-02 7.14E-03

Zr-95 2.64E-02 8.30E-02 5.48E-02 1.30E-02 2.69E-02 9.21E-02 7.06E-02 1.84E-02

Nb-95 2.92E-02 1.13E-01 1.08E-01 2.81E-02 2.95E-02 1.25E-01 1.39E-01 3.96E-02

Mo-99 3.31E-03 7.42E-06 1.01E-20 9.49E-41 3.26E-03 7.96E-06 1.26E-20 1.30E-40

Ru-103 3.65E-02 9.35E-02 2.37E-02 1.59E-03 3.64E-02 1.02E-01 2.99E-02 2.20E-03

Ru-106 2.14E-02 8.79E-02 2.04E-01 2.54E-01 1.57E-02 7.05E-02 1.90E-01 2.59E-01

Cd-113m 2.11E-06 9.15E-06 2.70E-05 4.62E-05 8.57E-07 4.05E-06 1.38E-05 2.60E-05

Cd-115m 2.63E-05 7.18E-05 2.44E-05 2.41E-06 2.53E-05 7.52E-05 2.97E-05 3.22E-06

Sb-125 4.66E-04 2.00E-03 5.45E-03 8.43E-03 2.91E-04 1.37E-03 4.33E-03 7.35E-03

Te-129m 5.82E-06 1.36E-05 2.28E-06 8.81E-08 3.57E-06 9.11E-06 1.77E-06 7.50E-08

Te-132 3.62E-03 2.39E-05 5.03E-18 3.67E-35 3.73E-03 2.68E-05 6.57E-18 5.26E-35

I-131 5.06E-01 1.65E-01 2.75E-06 5.51E-13 5.22E-01 1.85E-01 3.59E-06 7.90E-13

I-132 1.86E-01 1.23E-03 2.59E-16 1.89E-33 1.92E-01 1.38E-03 3.38E-16 2.71E-33

Cs-134 2.69E-03 1.13E-02 3.00E-02 4.44E-02 7.74E-04 3.57E-03 1.09E-02 1.78E-02

Cs-136 3.18E-03 2.84E-03 5.36E-06 5.52E-10 1.48E-03 1.44E-03 3.16E-06 3.57E-10

Cs-137 2.34E-03 1.01E-02 3.02E-02 5.25E-02 1.27E-03 6.00E-03 2.07E-02 3.95E-02

Ba-140 1.93E-02 1.65E-02 2.45E-05 1.84E-09 1.99E-02 1.84E-02 3.18E-05 2.63E-09

La-140 2.22E-02 1.89E-02 2.82E-05 2.12E-09 2.28E-02 2.12E-02 3.66E-05 3.03E-09

Ce-141 2.61E-02 5.98E-02 9.07E-03 3.09E-04 2.67E-02 6.67E-02 1.18E-02 4.39E-04

Ce-144 2.08E-02 8.40E-02 1.80E-01 2.01E-01 1.64E-02 7.23E-02 1.79E-01 2.20E-01

Pr-143 1.80E-02 1.69E-02 3.98E-05 5.50E-09 1.82E-02 1.87E-02 5.10E-05 7.74E-09

Pr-144 2.08E-02 8.41E-02 1.80E-01 2.01E-01 1.64E-02 7.23E-02 1.79E-01 2.20E-01

Nd-147 7.13E-03 4.67E-03 2.03E-06 3.03E-11 7.05E-03 5.03E-03 2.55E-06 4.16E-11

Pm-147 5.06E-03 2.18E-02 5.94E-02 9.12E-02 3.45E-03 1.63E-02 5.16E-02 8.70E-02

Pm-148m 1.71E-03 4.48E-03 1.28E-03 1.01E-04 8.72E-04 2.50E-03 8.30E-04 7.17E-05

Sm-151 5.54E-05 2.40E-04 7.19E-04 1.26E-03 3.91E-05 1.85E-04 6.43E-04 1.24E-03

Eu-154 1.70E-04 7.32E-04 2.13E-03 3.60E-03 6.02E-05 2.83E-04 9.58E-04 1.77E-03

Eu-155 4.57E-04 1.96E-03 5.58E-03 9.11E-03 2.32E-04 1.08E-03 3.58E-03 6.41E-03

Ta-182 1.55E-04 5.63E-04 7.24E-04 4.15E-04 1.52E-04 5.99E-04 8.95E-04 5.62E-04

ATTACHMENT D

NORMALIZED INTAKE FRACTIONS FOR THE DOSIMETRICALLY IMPORTANT RADIONUCLIDES

Page 2 of 4
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 39 of 70

Radionuclide

N Rx 1 N Rx 2

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

Mn-54 1.72E-07 8.46E-07 3.12E-06 5.79E-06 3.40E-07 1.42E-06 4.18E-06 6.42E-06

Fe-55 4.79E-06 2.47E-05 1.13E-04 2.77E-04 1.03E-05 4.50E-05 1.63E-04 3.33E-04

Co-58 1.29E-06 5.08E-06 6.50E-06 2.99E-06 1.32E-06 4.40E-06 4.48E-06 1.70E-06

Co-60 9.37E-07 4.88E-06 2.33E-05 6.11E-05 2.05E-06 9.08E-06 3.45E-05 7.48E-05

Sr-89 2.27E-02 7.91E-02 5.84E-02 1.29E-02 2.29E-02 6.79E-02 3.99E-02 7.31E-03

Sr-90 3.11E-04 1.63E-03 8.14E-03 2.25E-02 6.61E-04 2.95E-03 1.17E-02 2.68E-02

Y-90 3.11E-04 1.64E-03 8.15E-03 2.25E-02 6.63E-04 2.95E-03 1.17E-02 2.68E-02

Y-91 2.73E-02 1.01E-01 9.66E-02 3.02E-02 2.95E-02 9.26E-02 7.06E-02 1.83E-02

Zr-95 3.08E-02 1.17E-01 1.29E-01 4.89E-02 3.63E-02 1.18E-01 1.03E-01 3.23E-02

Nb-95 2.52E-02 1.34E-01 2.46E-01 1.05E-01 3.87E-02 1.57E-01 2.02E-01 6.96E-02

Mo-99 3.77E-03 1.02E-05 2.34E-20 3.49E-40 3.44E-03 7.94E-06 1.45E-20 1.78E-40

Ru-103 2.05E-02 6.34E-02 2.69E-02 2.88E-03 2.50E-02 6.58E-02 2.23E-02 1.97E-03

Ru-106 1.27E-03 6.31E-03 2.44E-02 4.84E-02 3.65E-03 1.55E-02 4.77E-02 7.82E-02

Cd-113m 3.35E-09 1.76E-08 8.66E-08 2.37E-07 4.60E-09 2.05E-08 8.06E-08 1.82E-07

Cd-115m 3.72E-06 1.23E-05 7.01E-06 1.11E-06 5.20E-06 1.46E-05 6.63E-06 8.66E-07

Sb-125 2.50E-05 1.31E-04 6.00E-04 1.48E-03 6.34E-05 2.81E-04 1.02E-03 2.08E-03

Te-129m 6.79E-04 1.92E-03 5.39E-04 3.32E-05 8.13E-04 1.96E-03 4.37E-04 2.23E-05

Te-132 4.08E-03 3.63E-05 2.12E-17 4.78E-34 3.82E-03 2.89E-05 1.34E-17 2.50E-34

I-131 5.05E-01 2.00E-01 5.76E-06 1.91E-12 4.83E-01 1.63E-01 3.74E-06 1.02E-12

I-132 2.10E-01 1.87E-03 1.09E-15 2.46E-32 1.97E-01 1.49E-03 6.91E-16 1.29E-32

Cs-134 3.58E-05 1.83E-04 8.11E-04 1.91E-03 2.06E-04 8.99E-04 3.16E-03 6.18E-03

Cs-136 9.71E-05 1.05E-04 3.32E-07 5.45E-11 1.84E-04 1.69E-04 4.25E-07 5.77E-11

Cs-137 3.43E-04 1.80E-03 8.99E-03 2.49E-02 7.85E-04 3.51E-03 1.39E-02 3.19E-02

Ba-140 2.70E-02 2.78E-02 6.92E-05 8.30E-09 2.40E-02 2.11E-02 4.18E-05 4.15E-09

La-140 3.10E-02 3.20E-02 7.97E-05 9.56E-09 2.76E-02 2.43E-02 4.81E-05 4.77E-09

Ce-141 3.24E-02 9.00E-02 2.29E-02 1.24E-03 3.17E-02 7.49E-02 1.51E-02 6.78E-04

Ce-144 9.61E-03 4.70E-02 1.68E-01 3.00E-01 1.75E-02 7.27E-02 2.07E-01 3.06E-01

Pr-143 2.79E-02 3.17E-02 1.25E-04 2.76E-08 2.43E-02 2.35E-02 7.39E-05 1.35E-08

Pr-144 9.61E-03 4.70E-02 1.68E-01 3.00E-01 1.75E-02 7.27E-02 2.07E-01 3.06E-01

Nd-147 9.00E-03 7.12E-03 5.20E-06 1.23E-10 8.07E-03 5.44E-03 3.16E-06 6.20E-11

Pm-147 1.22E-03 6.71E-03 3.09E-02 7.56E-02 2.56E-03 1.16E-02 4.21E-02 8.52E-02

Pm-148m 6.28E-05 2.00E-04 9.58E-05 1.20E-05 1.46E-04 3.95E-04 1.51E-04 1.56E-05

Sm-151 3.50E-06 1.84E-05 9.23E-05 2.57E-04 4.24E-06 1.90E-05 7.56E-05 1.74E-04

Eu-154 1.71E-06 8.91E-06 4.35E-05 1.17E-04 9.99E-06 4.44E-05 1.72E-04 3.83E-04

Eu-155 1.03E-05 5.33E-05 2.53E-04 6.58E-04 1.95E-05 8.63E-05 3.27E-04 7.02E-04

Ta-182 1.72E-08 7.56E-08 1.64E-07 1.50E-07 6.42E-08 2.40E-07 4.13E-07 3.14E-07

ATTACHMENT D

NORMALIZED INTAKE FRACTIONS FOR THE DOSIMETRICALLY IMPORTANT RADIONUCLIDES

Page 3 of 4
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 40 of 70

Radionuclide

TRIGA 1 TRIGA 2

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

Mn-54 8.69E-05 2.79E-04 5.49E-04 6.20E-04 2.17E-05 1.01E-04 3.62E-04 6.81E-04

Fe-55 2.06E-03 6.93E-03 1.69E-02 2.53E-02 2.12E-04 1.04E-03 4.59E-03 1.14E-02

Co-58 1.62E-04 4.13E-04 2.82E-04 7.89E-05 1.22E-04 4.56E-04 5.65E-04 2.62E-04

Co-60 1.69E-03 5.73E-03 1.46E-02 2.33E-02 1.35E-04 6.67E-04 3.09E-03 8.18E-03

Sr-89 2.71E-02 6.15E-02 2.42E-02 3.27E-03 2.55E-02 8.45E-02 6.03E-02 1.35E-02

Sr-90 5.26E-03 1.80E-02 4.79E-02 8.07E-02 3.48E-04 1.74E-03 8.37E-03 2.34E-02

Y-90 5.27E-03 1.80E-02 4.80E-02 8.07E-02 3.49E-04 1.74E-03 8.37E-03 2.34E-02

Y-91 3.43E-02 8.25E-02 4.22E-02 8.03E-03 3.03E-02 1.06E-01 9.82E-02 3.11E-02

Zr-95 4.15E-02 1.03E-01 6.07E-02 1.39E-02 3.44E-02 1.24E-01 1.33E-01 5.08E-02

Nb-95 4.58E-02 1.40E-01 1.19E-01 3.01E-02 2.81E-02 1.42E-01 2.53E-01 1.09E-01

Mo-99 3.29E-03 5.81E-06 7.10E-21 6.45E-41 3.84E-03 9.89E-06 2.19E-20 3.31E-40

Ru-103 1.78E-02 3.59E-02 8.15E-03 5.29E-04 1.73E-02 5.10E-02 2.10E-02 2.27E-03

Ru-106 3.25E-03 1.05E-02 2.18E-02 2.63E-02 6.09E-04 2.87E-03 1.08E-02 2.16E-02

Cd-113m 2.75E-08 9.39E-08 2.47E-07 4.11E-07 1.93E-08 9.58E-08 4.58E-07 1.27E-06

Cd-115m 2.98E-06 6.41E-06 1.95E-06 1.88E-07 2.67E-06 8.36E-06 4.62E-06 7.39E-07

Sb-125 2.85E-04 9.59E-04 2.34E-03 3.50E-03 3.09E-05 1.53E-04 6.76E-04 1.68E-03

Te-129m 5.62E-04 1.04E-03 1.55E-04 5.82E-06 5.69E-04 1.53E-03 4.15E-04 2.59E-05

Te-132 3.45E-03 2.00E-05 6.22E-18 8.54E-35 4.00E-03 3.38E-05 1.91E-17 4.36E-34

I-131 4.23E-01 1.09E-01 1.68E-06 3.40E-13 4.91E-01 1.84E-01 5.15E-06 1.73E-12

I-132 1.77E-01 1.03E-03 3.20E-16 4.40E-33 2.06E-01 1.74E-03 9.83E-16 2.25E-32

Cs-134 6.57E-04 2.19E-03 5.18E-03 7.44E-03 2.96E-06 1.44E-05 6.17E-05 1.47E-04

Cs-136 1.44E-04 1.02E-04 1.71E-07 1.71E-11 3.10E-05 3.19E-05 9.74E-08 1.62E-11

Cs-137 5.39E-03 1.85E-02 4.91E-02 8.27E-02 3.51E-04 1.75E-03 8.45E-03 2.37E-02

Ba-140 2.41E-02 1.62E-02 2.15E-05 1.57E-09 2.81E-02 2.75E-02 6.63E-05 8.05E-09

La-140 2.76E-02 1.86E-02 2.47E-05 1.81E-09 3.23E-02 3.17E-02 7.64E-05 9.28E-09

Ce-141 3.16E-02 5.72E-02 7.76E-03 2.56E-04 3.38E-02 8.90E-02 2.19E-02 1.20E-03

Ce-144 3.56E-02 1.13E-01 2.17E-01 2.35E-01 1.02E-02 4.76E-02 1.65E-01 2.98E-01

Pr-143 2.58E-02 1.91E-02 4.03E-05 5.40E-09 3.09E-02 3.34E-02 1.28E-04 2.84E-08

Pr-144 3.56E-02 1.13E-01 2.17E-01 2.35E-01 1.02E-02 4.76E-02 1.65E-01 2.98E-01

Nd-147 7.98E-03 4.12E-03 1.60E-06 2.32E-11 9.32E-03 7.01E-03 4.95E-06 1.19E-10

Pm-147 1.13E-02 3.83E-02 9.31E-02 1.39E-01 1.30E-03 6.79E-03 3.02E-02 7.49E-02

Pm-148m 3.04E-04 6.30E-04 1.61E-04 1.23E-05 1.83E-05 5.53E-05 2.57E-05 3.26E-06

Sm-151 3.00E-05 1.03E-04 2.75E-04 4.67E-04 7.24E-06 3.61E-05 1.75E-04 4.95E-04

Eu-154 3.74E-05 1.27E-04 3.31E-04 5.42E-04 1.29E-07 6.42E-07 3.03E-06 8.25E-06

Eu-155 7.61E-05 2.58E-04 6.54E-04 1.03E-03 1.11E-05 5.49E-05 2.53E-04 6.64E-04

Ta-182 9.34E-04 2.67E-03 3.09E-03 1.72E-03 7.56E-04 3.15E-03 6.59E-03 6.13E-03

ATTACHMENT D

NORMALIZED INTAKE FRACTIONS FOR THE DOSIMETRICALLY IMPORTANT RADIONUCLIDES

Page 4 of 4
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 41 of 70

Table E-1. Normalized intake fractions for the reduced list of dosimetrically important radionuclides.

Radionuclide

ATR 1 ATR 2 ATR 3

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

Co-60 1.16E-07 4.48E-07 2.18E-06 6.77E-06 3.96E-08 2.27E-07 1.53E-06 5.61E-06 5.31E-08 2.65E-07 1.58E-06 5.56E-06

Sr-89 3.81E-02 9.85E-02 7.38E-02 1.94E-02 2.19E-02 8.41E-02 8.77E-02 2.71E-02 2.74E-02 9.15E-02 8.44E-02 2.51E-02

Sr-90 5.67E-04 2.21E-03 1.12E-02 3.67E-02 1.92E-04 1.11E-03 7.79E-03 3.01E-02 2.70E-04 1.36E-03 8.45E-03 3.14E-02

Y-90 5.68E-04 2.21E-03 1.12E-02 3.67E-02 1.91E-04 1.11E-03 7.79E-03 3.01E-02 2.69E-04 1.36E-03 8.45E-03 3.14E-02

Y-91 4.56E-02 1.25E-01 1.21E-01 4.50E-02 2.50E-02 1.01E-01 1.37E-01 5.99E-02 3.13E-02 1.11E-01 1.33E-01 5.56E-02

Zr-95 4.92E-02 1.39E-01 1.55E-01 6.97E-02 2.58E-02 1.08E-01 1.68E-01 8.87E-02 3.29E-02 1.20E-01 1.66E-01 8.39E-02

Nb-95 4.25E-02 1.64E-01 2.97E-01 1.50E-01 1.20E-02 9.46E-02 3.07E-01 1.89E-01 1.91E-02 1.16E-01 3.07E-01 1.79E-01

Ru-103 2.55E-02 5.85E-02 2.52E-02 3.20E-03 1.63E-02 5.57E-02 3.34E-02 4.98E-03 1.98E-02 5.89E-02 3.12E-02 4.48E-03

Ru-106 9.85E-04 3.63E-03 1.43E-02 3.36E-02 3.60E-04 1.97E-03 1.08E-02 2.99E-02 4.91E-04 2.34E-03 1.13E-02 3.01E-02

I-131 6.86E-01 2.02E-01 5.90E-06 2.32E-12 8.13E-01 3.54E-01 1.44E-05 6.68E-12 7.72E-01 2.93E-01 1.06E-05 4.70E-12

Cs-134 4.24E-04 1.61E-03 7.22E-03 2.03E-02 1.04E-04 5.83E-04 3.64E-03 1.20E-02 6.80E-05 3.34E-04 1.85E-03 5.84E-03

Cs-137 5.72E-04 2.23E-03 1.13E-02 3.71E-02 1.93E-04 1.12E-03 7.87E-03 3.04E-02 2.72E-04 1.37E-03 8.53E-03 3.17E-02

Ce-141 4.89E-02 1.00E-01 2.59E-02 1.67E-03 3.43E-02 1.05E-01 3.76E-02 2.84E-03 4.08E-02 1.08E-01 3.44E-02 2.50E-03

Ce-144 1.64E-02 5.93E-02 2.15E-01 4.57E-01 6.17E-03 3.32E-02 1.68E-01 4.18E-01 8.48E-03 3.98E-02 1.78E-01 4.26E-01

Pr-143 4.31E-02 3.63E-02 1.45E-04 3.81E-08 4.39E-02 5.50E-02 3.06E-04 9.42E-08 4.60E-02 5.02E-02 2.47E-04 7.31E-08

Pm-147 1.59E-03 6.54E-03 3.06E-02 8.89E-02 4.68E-04 3.34E-03 2.23E-02 7.62E-02 8.45E-04 4.83E-03 2.82E-02 9.25E-02

Eu-154 1.34E-05 5.18E-05 2.56E-04 8.19E-04 3.87E-06 2.22E-05 1.53E-04 5.75E-04 2.25E-06 1.13E-05 6.87E-05 2.48E-04

ATTACHMENT E

NORMALIZED INTAKE FRACTIONS FOR THE REDUCED LIST OF DOSIMETRICALLY IMPORTANT RADIONUCLIDES

Page 1 of 3
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Radionuclide

FFTF 1 FFTF 2 N Rx 1

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

Co-60 6.51E-05 2.44E-04 7.59E-04 1.29E-03 3.15E-05 1.27E-04 4.61E-04 8.83E-04 1.33E-06 5.53E-06 2.81E-05 8.76E-05

Sr-89 1.36E-02 3.42E-02 1.64E-02 2.36E-03 1.40E-02 3.77E-02 2.12E-02 3.42E-03 3.22E-02 8.97E-02 7.03E-02 1.85E-02

Sr-90 1.12E-03 4.25E-03 1.38E-02 2.47E-02 6.10E-04 2.47E-03 9.38E-03 1.90E-02 4.42E-04 1.85E-03 9.81E-03 3.23E-02

Y-90 1.13E-03 4.25E-03 1.38E-02 2.48E-02 6.13E-04 2.47E-03 9.38E-03 1.90E-02 4.42E-04 1.85E-03 9.81E-03 3.23E-02

Y-91 2.03E-02 5.39E-02 3.36E-02 6.82E-03 2.08E-02 5.91E-02 4.31E-02 9.86E-03 3.88E-02 1.14E-01 1.16E-01 4.33E-02

Zr-95 3.67E-02 1.00E-01 7.21E-02 1.77E-02 3.72E-02 1.09E-01 9.18E-02 2.53E-02 4.37E-02 1.33E-01 1.56E-01 7.01E-02

Nb-95 4.04E-02 1.37E-01 1.42E-01 3.82E-02 4.07E-02 1.48E-01 1.80E-01 5.47E-02 3.58E-02 1.52E-01 2.97E-01 1.50E-01

Ru-103 5.07E-02 1.13E-01 3.12E-02 2.17E-03 5.03E-02 1.20E-01 3.89E-02 3.04E-03 2.91E-02 7.19E-02 3.25E-02 4.13E-03

Ru-106 2.97E-02 1.06E-01 2.68E-01 3.46E-01 2.18E-02 8.36E-02 2.47E-01 3.58E-01 1.80E-03 7.15E-03 2.94E-02 6.95E-02

I-131 7.02E-01 1.99E-01 3.63E-06 7.51E-13 7.22E-01 2.19E-01 4.68E-06 1.09E-12 7.16E-01 2.27E-01 6.94E-06 2.74E-12

Cs-134 3.72E-03 1.37E-02 3.95E-02 6.06E-02 1.07E-03 4.23E-03 1.42E-02 2.46E-02 5.08E-05 2.08E-04 9.77E-04 2.75E-03

Cs-137 3.25E-03 1.23E-02 3.98E-02 7.15E-02 1.75E-03 7.11E-03 2.70E-02 5.46E-02 4.87E-04 2.05E-03 1.08E-02 3.57E-02

Ce-141 3.62E-02 7.23E-02 1.19E-02 4.21E-04 3.69E-02 7.90E-02 1.53E-02 6.06E-04 4.60E-02 1.02E-01 2.75E-02 1.78E-03

Ce-144 2.89E-02 1.02E-01 2.36E-01 2.74E-01 2.27E-02 8.57E-02 2.33E-01 3.04E-01 1.36E-02 5.33E-02 2.02E-01 4.30E-01

Pr-143 2.49E-02 2.04E-02 5.24E-05 7.50E-09 2.52E-02 2.21E-02 6.64E-05 1.07E-08 3.96E-02 3.60E-02 1.51E-04 3.95E-08

Pm-147 7.02E-03 2.64E-02 7.81E-02 1.24E-01 4.77E-03 1.93E-02 6.71E-02 1.20E-01 1.72E-03 7.61E-03 3.72E-02 1.08E-01

Eu-154 2.35E-04 8.85E-04 2.81E-03 4.90E-03 8.32E-05 3.36E-04 1.25E-03 2.45E-03 2.42E-06 1.01E-05 5.24E-05 1.68E-04

ATTACHMENT E

NORMALIZED INTAKE FRACTIONS FOR THE REDUCED LIST OF DOSIMETRICALLY IMPORTANT RADIONUCLIDES

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Radionuclide

N Rx 2 TRIGA 1 TRIGA 2

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

Co-60 2.85E-06 1.04E-05 4.37E-05 1.08E-04 2.36E-03 6.88E-03 1.93E-02 3.19E-02 1.92E-04 7.60E-04 3.76E-03 1.20E-02

Sr-89 3.19E-02 7.79E-02 5.04E-02 1.06E-02 3.78E-02 7.38E-02 3.19E-02 4.46E-03 3.62E-02 9.63E-02 7.35E-02 1.99E-02

Sr-90 9.21E-04 3.39E-03 1.48E-02 3.88E-02 7.35E-03 2.16E-02 6.31E-02 1.10E-01 4.94E-04 1.98E-03 1.02E-02 3.44E-02

Y-90 9.24E-04 3.39E-03 1.48E-02 3.88E-02 7.36E-03 2.16E-02 6.32E-02 1.10E-01 4.96E-04 1.98E-03 1.02E-02 3.44E-02

Y-91 4.11E-02 1.06E-01 8.93E-02 2.64E-02 4.79E-02 9.90E-02 5.56E-02 1.10E-02 4.30E-02 1.21E-01 1.20E-01 4.57E-02

Zr-95 5.06E-02 1.35E-01 1.31E-01 4.67E-02 5.80E-02 1.23E-01 7.99E-02 1.91E-02 4.89E-02 1.42E-01 1.62E-01 7.46E-02

Nb-95 5.39E-02 1.80E-01 2.56E-01 1.01E-01 6.40E-02 1.68E-01 1.57E-01 4.11E-02 3.99E-02 1.62E-01 3.08E-01 1.60E-01

Ru-103 3.48E-02 7.55E-02 2.81E-02 2.84E-03 2.49E-02 4.31E-02 1.07E-02 7.23E-04 2.46E-02 5.81E-02 2.55E-02 3.33E-03

Ru-106 5.09E-03 1.77E-02 6.03E-02 1.13E-01 4.55E-03 1.27E-02 2.88E-02 3.60E-02 8.64E-04 3.27E-03 1.31E-02 3.17E-02

I-131 6.73E-01 1.87E-01 4.72E-06 1.48E-12 5.92E-01 1.31E-01 2.22E-06 4.64E-13 6.97E-01 2.10E-01 6.27E-06 2.54E-12

Cs-134 2.88E-04 1.03E-03 4.00E-03 8.94E-03 9.18E-04 2.63E-03 6.82E-03 1.02E-02 4.21E-06 1.64E-05 7.51E-05 2.16E-04

Cs-137 1.09E-03 4.02E-03 1.76E-02 4.61E-02 7.54E-03 2.22E-02 6.47E-02 1.13E-01 4.99E-04 1.99E-03 1.03E-02 3.48E-02

Ce-141 4.42E-02 8.59E-02 1.91E-02 9.81E-04 4.42E-02 6.86E-02 1.02E-02 3.49E-04 4.80E-02 1.01E-01 2.66E-02 1.76E-03

Ce-144 2.43E-02 8.34E-02 2.62E-01 4.42E-01 4.97E-02 1.36E-01 2.85E-01 3.21E-01 1.45E-02 5.42E-02 2.01E-01 4.37E-01

Pr-143 3.39E-02 2.70E-02 9.34E-05 1.95E-08 3.61E-02 2.30E-02 5.31E-05 7.38E-09 4.39E-02 3.80E-02 1.55E-04 4.18E-08

Pm-147 3.57E-03 1.33E-02 5.32E-02 1.23E-01 1.58E-02 4.60E-02 1.23E-01 1.90E-01 1.84E-03 7.73E-03 3.68E-02 1.10E-01

Eu-154 1.39E-05 5.09E-05 2.18E-04 5.54E-04 5.22E-05 1.53E-04 4.37E-04 7.40E-04 1.84E-07 7.32E-07 3.69E-06 1.21E-05

ATTACHMENT E

NORMALIZED INTAKE FRACTIONS FOR THE REDUCED LIST OF DOSIMETRICALLY IMPORTANT RADIONUCLIDES

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Table F-1. Summary of IRFs used to 

compute urinary excretion.

Nuclide and 

absorption type

IRF for chronic intake

(24-hr urine, 2 years)

Cd-113m S 6.15E-04

Cd-115m S 5.33E-05

Ce-141 S 6.47E-07

Ce-144 S 9.10E-06

Co-58 S 1.60E-02

Co-60 S 2.01E-02

Cs-134 F 3.39E-01

Cs-136 F 6.65E-02

Cs-137 F 3.78E-01

Eu-154 M 8.73E-03

Eu-155 M 8.49E-03

Fe-55 M 9.65E-04

I-131 V 5.83E-01

I-132 V 1.20E-02

La-140 M 4.78E-06

Mn-54 M 4.33E-02

Mo-99 S 1.54E-03

Nb-95 S 1.98E-03

Nd-147 S 2.81E-05

Pm-147 S 1.88E-04

Pm-148m S 4.22E-05

Pr-143 S 3.77E-05

Pr-144 S 1.43E-09

Ru-103 S 1.06E-02

Ru-106 S 1.56E-02

Sb-125 M 4.73E-02

Sm-151 M 6.35E-03

Sr-89 F 1.91E-01

Sr-90 F 2.31E-01

Ta-182 S 4.59E-04

Te-127m M 9.57E-02

Te-129m M 8.06E-02

Te-132 M 4.39E-02

Y-90 S 3.39E-05

Y-91 S 8.58E-05

Zr-95 S 5.81E-04

ATTACHMENT 

Page 1 of 3

ATTACHMENT 

Page 1 of 3

ATTACHMENT F

SUMMARY OF INTAKE RETENTION FRACTIONS

Page 1 of 3
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Table F-2. Half-lives used for 

decay corrections.

Radionuclide Half-life

Mn-54 312.7 d

Fe-55 2.7 y

Co-58 70.8 d

Co-60 5.271 y

Sr-89 50.55 d

Sr-90 28.6 y

Y-90 64.1 h

Y-91 58.51 d

Zr-95 64.02 d

Nb-95 35.06 d

Mo-99 66.02 h

Ru-103 39.35 d

Ru-106 368.2 d

Cd-113m 13.7 y

Cd-115m 44.6 d

Sb-125 2.77 y

Te-129m 33.6 d

Te-132 78.2 h

Cs-134 2.062 y

Cs-136 13.16 y

Cs-137 30.17 y

Ba-140 12.789 d

La-140 40.22 h

Ce-141 32.5 d

Ce-144 284.3 d

Pr-143 13.56 d

Pr-144 17.28 m

Nd-147 10.98 d

Pm-147 2.6234 y

Pm-148m 41.3 d

Sm-151 90 y

Eu-154 8.8 y

Eu-155 4.96 y

Ta-182 114.74 d

ATTACHMENT F

SUMMARY OF INTAKE RETENTION FRACTIONS

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Table F-3. Beta and gamma radiation yield values.

Radionuclide Beta yield Gamma yield

Mn-54 0 1.0

Fe-55 0 0.0

Co-58 0 1.306

Co-60 1 2.0

Sr-89/Y-89m 1 9.3E-5

Sr-90 1 0

Y-90 1 0

Y-91 1 0.003

Zr-95 1 0.996

Nb-95 0 0.998

Nb-95m 0 0.2587

Mo-99/Tc-99m 1 1.0743

Ru-103 1 0.9306

Ru-106/Rh-106 2 0.342

Cd-113m 1 0

Cd-115m 1 0.0222

Sb-125 1 0.8562

Te-125m 0 0.00274

Te-129m/Te-129 1.02 0.11567

Te-132 1 0.9199

Cs-134 1 2.22876

Cs-136/Ba-136m 1 2.8793

Cs-137/Ba-137m 1 0.8998

Ba-140 1 0.40381

La-140 1 2.1542

Ce-141 1 0.48

Ce-144/Pr-144 1.98 0.133586

Pr-143 1 0

Pr-144 1 0.0257

Nd-147 1 0.1997

Pm-147 1 0

Pm-148m 0.955 3.2178

Sm-151 0 0

Eu-154 1 1.633

Eu-155 0 0.20653

Ta-182 0.99 1.4468

ATTACHMENT F

SUMMARY OF INTAKE RETENTION FRACTIONS

Page 3 of 3
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The values in the following tables show radionuclide contributions to the total beta or gamma activity in a sample for the delay time that 

maximizes the indicator radionuclide contribution for the given reactor case and fuel decay time Delay time refers to the time elapsed 

between sample collection and counting. 

Table G-1. Contributions to gross beta activity for an unprocessed urinalysis. 

Nuclide

ATR 1

(decay time/delay time)

ATR 2

(decay time/delay time)

ATR 3

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/ 1d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Mn-54 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-60 1.63E-09 7.82E-09 3.55E-08 9.30E-08 5.32E-10 3.91E-09 2.60E-08 7.92E-08 7.24E-10 4.58E-09 2.66E-08 7.81E-08

Sr-89/Y-89m 3.42E-03 1.10E-02 7.69E-03 2.51E-03 1.88E-03 9.26E-03 9.52E-03 3.61E-03 2.39E-03 1.01E-02 9.08E-03 3.32E-03

Sr-90 9.24E-05 4.48E-04 2.11E-03 5.81E-03 2.99E-05 2.21E-04 1.54E-03 4.89E-03 4.28E-05 2.73E-04 1.65E-03 5.07E-03

Y-90 9.24E-05 4.48E-04 2.11E-03 3.98E-03 2.99E-05 2.21E-04 1.54E-03 3.35E-03 4.28E-05 2.73E-04 1.65E-03 3.47E-03

Y-91 1.94E-06 6.59E-06 5.98E-06 2.61E-06 1.02E-06 5.29E-06 7.05E-06 3.57E-06 1.30E-06 5.81E-06 6.76E-06 3.30E-06

Zr-95 1.46E-05 5.12E-05 5.35E-05 2.74E-05 7.31E-06 3.92E-05 6.03E-05 3.59E-05 9.50E-06 4.39E-05 5.89E-05 3.37E-05

Nb-95m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nb-95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Mo-99/Tc-99m 3.03E-09 7.58E-12 1.64E-26 3.63E-43 3.59E-09 1.37E-11 4.38E-26 1.12E-42 3.35E-09 1.11E-11 3.07E-26 7.57E-43

Ru-103 1.12E-04 3.21E-04 1.29E-04 2.28E-05 6.88E-05 3.01E-04 1.78E-04 3.64E-05 8.49E-05 3.20E-04 1.65E-04 3.25E-05

Ru-106/Rh-106 2.05E-05 9.41E-05 3.46E-04 7.17E-04 7.18E-06 5.04E-05 2.72E-04 6.54E-04 9.95E-06 6.02E-05 2.83E-04 6.57E-04

Cd-113m 4.43E-13 2.14E-12 1.00E-11 2.72E-11 1.76E-13 1.30E-12 8.96E-12 2.82E-11 6.36E-13 4.06E-12 2.43E-11 7.38E-11

Cd-115m 9.32E-11 2.84E-10 1.54E-10 3.77E-11 5.36E-11 2.50E-10 1.98E-10 5.65E-11 6.73E-11 2.70E-10 1.87E-10 5.15E-11

Sb-125 1.07E-06 5.17E-06 2.24E-05 5.58E-05 3.44E-07 2.62E-06 1.68E-05 4.84E-05 4.96E-07 3.21E-06 1.78E-05 4.96E-05

Te-125m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Te-129m/Te-129 2.61E-05 6.82E-05 1.81E-05 2.01E-06 1.72E-05 6.89E-05 2.69E-05 3.46E-06 2.08E-05 7.15E-05 2.43E-05 3.02E-06

Te-132 2.93E-07 2.41E-09 1.33E-21 1.42E-35 3.47E-07 4.36E-09 3.54E-21 4.38E-35 3.24E-07 3.50E-09 2.48E-21 2.96E-35

Cs-134 9.88E-05 4.66E-04 1.95E-03 4.70E-03 2.31E-05 1.67E-04 1.03E-03 2.86E-03 1.54E-05 9.59E-05 5.15E-04 1.38E-03

Cs-136/Ba-136m 1.20E-06 1.20E-06 3.59E-09 2.69E-12 5.49E-07 8.40E-07 3.69E-09 3.21E-12 6.44E-07 8.49E-07 3.25E-09 2.73E-12

Cs-137/Ba-137m 1.53E-04 7.40E-04 3.50E-03 9.61E-03 4.93E-05 3.66E-04 2.54E-03 8.09E-03 7.06E-05 4.50E-04 2.73E-03 8.38E-03

Ba-140 1.34E-04 1.27E-04 3.00E-07 1.72E-10 1.36E-04 1.97E-04 6.84E-07 4.54E-10 1.40E-04 1.75E-04 5.29E-07 3.39E-10

La-140 1.54E-04 1.47E-04 3.46E-07 8.89E-11 1.56E-04 2.27E-04 7.88E-07 2.35E-10 1.61E-04 2.01E-04 6.09E-07 1.75E-10

Ce-141 1.18E-08 3.02E-08 7.26E-09 7.24E-10 7.91E-09 3.10E-08 1.10E-08 1.27E-09 9.58E-09 3.23E-08 9.96E-09 1.11E-09

Ce-144/Pr-144 1.94E-07 8.74E-07 2.96E-06 5.63E-06 6.99E-08 4.83E-07 2.40E-06 5.29E-06 9.78E-08 5.81E-07 2.53E-06 5.36E-06

Pr-143 2.48E-07 2.60E-07 9.72E-10 9.34E-13 2.41E-07 3.88E-07 2.13E-09 2.38E-12 2.57E-07 3.56E-07 1.70E-09 1.83E-12

Pr-144 0.00E+00 0.00E+00 0.00E+00 3.69E-35 0.00E+00 0.00E+00 0.00E+00 3.47E-35 0.00E+00 0.00E+00 0.00E+00 3.51E-35

Nd-147 3.70E-08 2.70E-08 1.87E-11 2.74E-15 3.55E-08 3.97E-08 4.03E-11 6.86E-15 4.01E-08 3.86E-08 3.42E-11 5.61E-15

Pm-147 2.07E-07 1.06E-06 4.63E-06 1.15E-05 5.84E-08 5.35E-07 3.51E-06 1.01E-05 1.07E-07 7.77E-07 4.40E-06 1.22E-05

Pm-148m 4.34E-09 1.27E-08 5.79E-09 1.17E-09 1.00E-09 4.51E-09 3.01E-09 7.06E-10 2.12E-09 8.21E-09 4.78E-09 1.08E-09

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Nuclide

ATR 1

(decay time/delay time)

ATR 2

(decay time/delay time)

ATR 3

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 8.20E-08 3.95E-07 1.83E-06 4.90E-06 2.27E-08 1.67E-07 1.14E-06 3.53E-06 1.34E-08 8.53E-08 5.05E-07 1.51E-06

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Ta-182 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nuclide

FFTF 1

(decay time/delay time)

FFTF 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Mn-54 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-60 9.31E-07 4.00E-06 1.14E-05 1.90E-05 4.52E-07 2.12E-06 7.03E-06 1.28E-05

Sr-89/Y-89m 1.25E-03 3.59E-03 1.58E-03 3.26E-04 1.29E-03 4.04E-03 2.06E-03 4.68E-04

Sr-90 1.87E-04 8.10E-04 2.41E-03 4.19E-03 1.02E-04 4.81E-04 1.66E-03 3.17E-03

Y-90 1.87E-04 8.10E-04 2.41E-03 2.87E-03 1.02E-04 4.81E-04 1.66E-03 2.17E-03

Y-91 8.80E-07 2.68E-06 1.53E-06 4.24E-07 9.04E-07 3.00E-06 1.99E-06 6.05E-07

Zr-95 1.11E-05 3.48E-05 2.30E-05 7.46E-06 1.13E-05 3.86E-05 2.96E-05 1.05E-05

Nb-95m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nb-95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Mo-99/Tc-99m 2.65E-09 5.95E-12 8.12E-27 1.13E-43 2.61E-09 6.38E-12 1.01E-26 1.55E-43

Ru-103 2.27E-04 5.82E-04 1.48E-04 1.65E-05 2.27E-04 6.32E-04 1.86E-04 2.28E-05

Ru-106/Rh-106 6.30E-04 2.59E-03 6.00E-03 7.90E-03 4.64E-04 2.08E-03 5.59E-03 8.07E-03

Cd-113m 1.29E-09 5.60E-09 1.65E-08 2.84E-08 5.25E-10 2.48E-09 8.48E-09 1.60E-08

Cd-115m 8.80E-10 2.40E-09 8.16E-10 1.27E-10 8.46E-10 2.51E-09 9.93E-10 1.69E-10

Sb-125 2.16E-05 9.24E-05 2.52E-04 3.98E-04 1.35E-05 6.33E-05 2.01E-04 3.47E-04

Te-125m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Te-129m/Te-129 2.58E-07 6.03E-07 1.01E-07 7.10E-09 1.58E-07 4.03E-07 7.83E-08 6.04E-09

Te-132 2.69E-07 1.77E-09 3.74E-22 1.30E-36 2.77E-07 1.99E-09 4.88E-22 1.86E-36

Cs-134 8.85E-04 3.74E-03 9.88E-03 1.50E-02 2.55E-04 1.18E-03 3.61E-03 6.02E-03

Cs-136/Ba-136m 4.35E-05 3.89E-05 7.34E-08 3.48E-11 2.03E-05 1.98E-05 4.33E-08 2.25E-11

Cs-137/Ba-137m 8.83E-04 3.83E-03 1.14E-02 1.98E-02 4.79E-04 2.26E-03 7.83E-03 1.49E-02

Ba-140 9.22E-05 7.84E-05 1.17E-07 4.23E-11 9.47E-05 8.78E-05 1.52E-07 6.03E-11

La-140 1.06E-04 9.03E-05 1.34E-07 2.19E-11 1.09E-04 1.01E-04 1.74E-07 3.12E-11

ATTACHMENT G

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ATTACHMENT G

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Nuclide

FFTF 1

(decay time/delay time)

FFTF 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Ce-141 8.90E-09 2.04E-08 3.10E-09 1.96E-10 9.11E-09 2.28E-08 4.01E-09 2.78E-10

Ce-144/Pr-144 3.49E-07 1.41E-06 3.01E-06 3.62E-06 2.75E-07 1.21E-06 3.00E-06 3.96E-06

Pr-143 1.46E-07 1.37E-07 3.24E-10 1.97E-13 1.48E-07 1.52E-07 4.15E-10 2.77E-13

Pr-144 0.00E+00 0.00E+00 0.00E+00 2.37E-35 0.00E+00 0.00E+00 0.00E+00 2.60E-35

Nd-147 3.02E-08 1.97E-08 8.60E-12 7.98E-16 2.98E-08 2.13E-08 1.08E-11 1.10E-15

Pm-147 9.33E-07 4.02E-06 1.09E-05 1.72E-05 6.36E-07 3.01E-06 9.51E-06 1.64E-05

Pm-148m 4.15E-08 1.09E-07 3.12E-08 4.00E-09 2.12E-08 6.08E-08 2.02E-08 2.84E-09

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 1.47E-06 6.35E-06 1.85E-05 3.14E-05 5.22E-07 2.46E-06 8.31E-06 1.55E-05

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Ta-182 5.90E-08 2.14E-07 2.75E-07 1.88E-07 5.76E-08 2.27E-07 3.40E-07 2.54E-07

Nuclide

N Rx 1

(decay time/delay time)

N Rx 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Mn-54 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-60 1.86E-08 9.67E-08 4.63E-07 1.22E-06 4.06E-08 1.80E-07 6.85E-07 1.50E-06

Sr-89/Y-89m 2.88E-03 1.00E-02 7.40E-03 2.44E-03 2.90E-03 8.62E-03 5.06E-03 1.38E-03

Sr-90 7.18E-05 3.77E-04 1.88E-03 5.20E-03 1.52E-04 6.81E-04 2.70E-03 6.19E-03

Y-90 7.18E-05 3.77E-04 1.88E-03 3.56E-03 1.52E-04 6.81E-04 2.70E-03 4.24E-03

Y-91 1.64E-06 6.06E-06 5.81E-06 2.56E-06 1.78E-06 5.57E-06 4.25E-06 1.55E-06

Zr-95 1.29E-05 4.92E-05 5.43E-05 2.81E-05 1.52E-05 4.93E-05 4.33E-05 1.85E-05

Nb-95m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nb-95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Mo-99/Tc-99m 3.02E-09 8.20E-12 1.88E-26 4.17E-43 2.76E-09 6.37E-12 1.16E-26 2.13E-43

Ru-103 1.27E-04 3.95E-04 1.68E-04 2.99E-05 1.55E-04 4.10E-04 1.39E-04 2.04E-05

Ru-106/Rh-106 3.74E-05 1.86E-04 7.20E-04 1.51E-03 1.08E-04 4.56E-04 1.41E-03 2.43E-03

Cd-113m 2.05E-12 1.08E-11 5.31E-11 1.46E-10 2.82E-12 1.26E-11 4.94E-11 1.12E-10

Cd-115m 1.24E-10 4.10E-10 2.34E-10 5.80E-11 1.74E-10 4.88E-10 2.22E-10 4.54E-11

Sb-125 1.16E-06 6.08E-06 2.78E-05 6.98E-05 2.94E-06 1.30E-05 4.73E-05 9.83E-05

Te-125m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

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RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

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Nuclide

N Rx 1

(decay time/delay time)

N Rx 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Te-129m/Te-129 3.01E-05 8.52E-05 2.39E-05 2.67E-06 3.60E-05 8.68E-05 1.93E-05 1.79E-06

Te-132 3.03E-07 2.70E-09 1.57E-21 1.70E-35 2.84E-07 2.15E-09 9.96E-22 8.89E-36

Cs-134 1.18E-05 6.04E-05 2.67E-04 6.48E-04 6.80E-05 2.96E-04 1.04E-03 2.09E-03

Cs-136/Ba-136m 1.33E-06 1.44E-06 4.55E-09 3.44E-12 2.51E-06 2.32E-06 5.82E-09 3.64E-12

Cs-137/Ba-137m 1.30E-04 6.81E-04 3.39E-03 9.41E-03 2.96E-04 1.32E-03 5.25E-03 1.20E-02

Ba-140 1.29E-04 1.32E-04 3.30E-07 1.91E-10 1.15E-04 1.00E-04 1.99E-07 9.51E-11

La-140 1.48E-04 1.53E-04 3.80E-07 9.85E-11 1.32E-04 1.16E-04 2.29E-07 4.92E-11

Ce-141 1.11E-08 3.07E-08 7.80E-09 7.84E-10 1.08E-08 2.56E-08 5.16E-09 4.29E-10

Ce-144/Pr-144 1.61E-07 7.87E-07 2.82E-06 5.40E-06 2.92E-07 1.22E-06 3.47E-06 5.49E-06

Pr-143 2.27E-07 2.58E-07 1.02E-09 9.87E-13 1.98E-07 1.92E-07 6.01E-10 4.82E-13

Pr-144 0.00E+00 0.00E+00 0.00E+00 3.53E-35 0.00E+00 0.00E+00 0.00E+00 3.60E-35

Nd-147 3.80E-08 3.01E-08 2.20E-11 3.25E-15 3.41E-08 2.30E-08 1.34E-11 1.63E-15

Pm-147 2.24E-07 1.24E-06 5.69E-06 1.42E-05 4.72E-07 2.13E-06 7.75E-06 1.60E-05

Pm-148m 1.53E-09 4.86E-09 2.33E-09 4.76E-10 3.55E-09 9.61E-09 3.67E-09 6.20E-10

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 1.48E

-08 7.73E-08 3.77E-07 1.02E-06 8.66E-08 3.85E-07 1.50E-06 3.34E-06

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Ta-182 6.53E

-12 2.87E-11 6.20E-11 6.79E-11 2.44E-11 9.11E-11 1.57E-10 1.42E-10

Nuclide

TRIGA 1

(decay time/delay time)

TRIGA 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/1 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Mn-54 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-60 3.35E-05 1.14E-04 2.94E-04 4.68E-04 2.68E-06 1.32E-05 6.13E-05 1.64E-04

Sr-89/Y-89m 3.44E-03 7.81E-03 4.58E-03 6.17E-04 3.24E-03 1.07E-02 7.66E-03 2.55E-03

Sr-90 1.21E-03 4.15E-03 1.11E-02 1.86E-02 8.02E-05 4.00E-04 1.93E-03 5.41E-03

Y-90 1.21E-03 4.15E-03 7.58E-03 1.28E-02 8.03E-05 4.00E-04 1.93E-03 3.71E-03

Y-91 2.06E-06 4.96E-06 3.58E-06 6.81E-07 1.82E-06 6.37E-06 5.91E-06 2.64E-06

Zr-95 1.74E-05 4.31E-05 3.48E-05 8.01E-06 1.44E-05 5.21E-05 5.57E-05 2.92E-05

Nb-95m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nb-95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Mo-99/Tc-99m 2.63E-09 4.66E-12 8.49E-24 7.71E-44 3.08E-09 7.93E-12 1.76E-26 3.96E-43

ATTACHMENT G

RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

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Nuclide

TRIGA 1

(decay time/delay time)

TRIGA 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/1 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Ru-103 1.11E-04 2.24E-04 8.45E-05 5.49E-06 1.08E-04 3.17E-04 1.31E-04 2.35E-05

Ru-106/Rh-106 9.59E-05 3.11E-04 6.79E-04 8.19E-04 1.79E-05 8.47E-05 3.18E-04 6.73E-04

Cd-113m 1.68E-11 5.75E-11 1.52E-10 2.53E-10 1.18E-11 5.87E-11 2.80E-10 7.78E-10

Cd-115m 9.96E-11 2.14E-10 1.02E-10 9.85E-12 8.92E-11 2.80E-10 1.55E-10 3.88E-11

Sb-125 1.32E-05 4.44E-05 1.10E-04 1.65E-04 1.43E-06 7.07E-06 3.13E-05 7.95E-05

Te-125m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Te-129m/Te-129 2.49E-05 4.60E-05 1.25E-05 4.69E-07 2.52E-05 6.78E-05 1.84E-05 2.08E-06

Te-132 2.56E-07 1.48E-09 2.21E-19 3.03E-36 2.97E-07 2.51E-09 1.42E-21 1.55E-35

Cs-134 2.17E-04 7.22E-04 1.75E-03 2.52E-03 9.76E-07 4.74E-06 2.03E-05 4.99E-05

Cs-136/Ba-136m 1.97E-06 1.39E-06 1.08E-08 1.08E-12 4.25E-07 4.37E-07 1.33E-09 1.02E-12

Cs-137/Ba-137m 2.04E-03 6.97E-03 1.86E-02 3.13E-02 1.33E-04 6.61E-04 3.19E-03 8.95E-03

Ba-140 1.15E-04 7.70E-05 4.93E-07 3.60E-11 1.34E-04 1.31E-04 3.16E-07 1.85E-10

La-140 1.32E-04 8.86E-05 2.55E-07 1.86E-11 1.54E-04 1.51E-04 3.64E-07 9.56E-11

Ce-141 1.08E-08 1.95E-08 4.92E-09 1.62E-10 1.15E-08 3.04E-08 7.47E-09 7.60E-10

Ce-144/Pr-144 5.96E-07 1.90E-06 3.89E-06 4.23E-06 1.72E-07 7.97E-07 2.76E-06 5.35E-06

Pr-143 2.10E-07 1.56E-07 1.44E-09 1.94E-13 2.52E-07 2.72E-07 1.04E-09 1.02E-12

Pr-144 0.00E+00 0.00E+00 2.55E-35 2.77E-35 0.00E+00 0.00E+00 0.00E+00 3.51E-35

Nd-147 3.37E-08 1.74E-08 4.23E-11 6.11E-16 3.94E-08 2.96E-08 2.09E-11 3.14E-15

Pm-147 2.09E-06 7.06E-06 1.75E-05 2.61E-05 2.39E-07 1.25E-06 5.57E-06 1.41E-05

Pm-148m 7.40E-09 1.53E-08 6.39E-09 4.88E-10 4.46E-10 1.35E-09 6.25E-10 1.29E-10

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 3.24E-07 1.10E-06 2.89E-06 4.73E-06 1.12E-09 5.57E-09 2.63E-08 7.20E-08

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Ta-182 3.54E-07 1.01E-06 1.40E-06 7.80E-07 2.87E-07 1.19E-06 2.50E-06 2.77E-06

Table G-2. Contributions to gross beta activity for urinalysis following major chemical processing.

Nuclide

ATR 1

(decay time/delay time)

ATR 2

(decay time/delay time)

ATR 3

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Sr-89/Y-89m 3.42E-03 1.10E-02 7.69E-03 2.51E-03 1.88E-03 9.26E-03 9.52E-03 3.61E-03 2.39E-03 1.01E-02 9.08E-03 3.32E-03

Sr-90 9.24E-05 4.48E-04 2.11E-03 5.81E-03 2.99E-05 2.21E-04 1.54E-03 4.89E-03 4.28E-05 2.73E-04 1.65E-03 5.07E-03

Y-90 9.24E-05 4.48E-04 2.11E-03 3.98E-03 2.99E-05 2.21E-04 1.54E-03 3.35E-03 4.28E-05 2.73E-04 1.65E-03 3.47E-03

Y-91 1.94E-06 6.59E-06 5.98E-06 2.61E-06 1.02E-06 5.29E-06 7.05E-06 3.57E-06 1.30E-06 5.81E-06 6.76E-06 3.30E-06

ATTACHMENT G

RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

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Nuclide

ATR 1

(decay time/delay time)

ATR 2

(decay time/delay time)

ATR 3

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Ba-140 1.34E-04 1.27E-04 3.00E-07 1.72E-10 1.36E-04 1.97E-04 6.84E-07 4.54E-10 1.40E-04 1.75E-04 5.29E-07 3.39E-10

La-140 1.54E-04 1.47E-04 3.46E-07 8.89E-11 1.56E-04 2.27E-04 7.88E-07 2.35E-10 1.61E-04 2.01E-04 6.09E-07 1.75E-10

Ce-141 1.18E-08 3.02E-08 7.26E-09 7.24E-10 7.91E-09 3.10E-08 1.10E-08 1.27E-09 9.58E-09 3.23E-08 9.96E-09 1.11E-09

Ce-144/Pr-144 1.94E-07 8.74E-07 2.96E-06 5.63E-06 6.99E-08 4.83E-07 2.40E-06 5.29E-06 9.78E-08 5.81E-07 2.53E-06 5.36E-06

Pr-143 2.48E-07 2.60E-07 9.72E-10 9.34E-13 2.41E-07 3.88E-07 2.13E-09 2.38E-12 2.57E-07 3.56E-07 1.70E-09 1.83E-12

Pr-144 0.00E+00 0.00E+00 0.00E+00 3.69E-35 0.00E+00 0.00E+00 0.00E+00 3.47E-35 0.00E+00 0.00E+00 0.00E+00 3.51E-35

Nd-147 3.70E-08 2.70E-08 1.87E-11 2.74E-15 3.55E-08 3.97E-08 4.03E-11 6.86E-15 4.01E-08 3.86E-08 3.42E-11 5.61E-15

Pm-147 2.07E-07 1.06E-06 4.63E-06 1.15E-05 5.84E-08 5.35E-07 3.51E-06 1.01E-05 1.07E-07 7.77E-07 4.40E-06 1.22E-05

Pm-148m 4.34E-09 1.27E-08 5.79E-09 1.17E-09 1.00E-09 4.51E-09 3.01E-09 7.06E-10 2.12E-09 8.21E-09 4.78E-09 1.08E-09

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 8.20E-08 3.95E-07 1.83E-06 4.90E-06 2.27E-08 1.67E-07 1.14E-06 3.53E-06 1.34E-08 8.53E-08 5.05E-07 1.51E-06

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nuclide

FFTF 1

(decay time/delay time)

FFTF 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Sr-89/Y-89m 1.25E-03 3.59E-03 1.58E-03 3.26E-04 1.29E-03 4.04E-03 2.06E-03 4.68E-04

Sr-90 1.87E-04 8.10E-04 2.41E-03 4.19E-03 1.02E-04 4.81E-04 1.66E-03 3.17E-03

Y-90 1.87E-04 8.10E-04 2.41E-03 2.87E-03 1.02E-04 4.81E-04 1.66E-03 2.17E-03

Y-91 8.80E-07 2.68E-06 1.53E-06 4.24E-07 9.04E-07 3.00E-06 1.99E-06 6.05E-07

Ba-140 9.22E-05 7.84E-05 1.17E-07 4.23E-11 9.47E-05 8.78E-05 1.52E-07 6.03E-11

La-140 1.06E-04 9.03E-05 1.34E-07 2.19E-11 1.09E-04 1.01E-04 1.74E-07 3.12E-11

Ce-141 8.90E-09 2.04E-08 3.10E-09 1.96E-10 9.11E-09 2.28E-08 4.01E-09 2.78E-10

Ce-144/Pr-144 3.49E-07 1.41E-06 3.01E-06 3.62E-06 2.75E-07 1.21E-06 3.00E-06 3.96E-06

Pr-143 1.46E-07 1.37E-07 3.24E-10 1.97E-13 1.48E-07 1.52E-07 4.15E-10 2.77E-13

Pr-144 0.00E+00 0.00E+00 0.00E+00 2.37E-35 0.00E+00 0.00E+00 0.00E+00 2.60E-35

Nd-147 3.02E-08 1.97E-08 8.60E-12 7.98E-16 2.98E-08 2.13E-08 1.08E-11 1.10E-15

Pm-147 9.33E-07 4.02E-06 1.09E-05 1.72E-05 6.36E-07 3.01E-06 9.51E-06 1.64E-05

Pm-148m 4.15E-08 1.09E-07 3.12E-08 4.00E-09 2.12E-08 6.08E-08 2.02E-08 2.84E-09

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 1.47E-06 6.35E-06 1.85E-05 3.14E-05 5.22E-07 2.46E-06 8.31E-06 1.55E-05

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

ATTACHMENT G

RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

Page 6 of 20
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 53 of 70

Nuclide

N Rx 1

(decay time/delay time)

N Rx 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Sr-89/Y-89m 2.88E-03 1.00E-02 7.40E-03 2.44E-03 2.90E-03 8.62E-03 5.06E-03 1.38E-03

Sr-90 7.18E-05 3.77E-04 1.88E-03 5.20E-03 1.52E-04 6.81E-04 2.70E-03 6.19E-03

Y-90 7.18E-05 3.77E-04 1.88E-03 3.56E-03 1.52E-04 6.81E-04 2.70E-03 4.24E-03

Y-91 1.64E-06 6.06E-06 5.81E-06 2.56E-06 1.78E-06 5.57E-06 4.25E-06 1.55E-06

Ba-140 1.29E-04 1.32E-04 3.30E-07 1.91E-10 1.15E-04 1.00E-04 1.99E-07 9.51E-11

La-140 1.48E-04 1.53E-04 3.80E-07 9.85E-11 1.32E-04 1.16E-04 2.29E-07 4.92E-11

Ce-141 1.11E-08 3.07E-08 7.80E-09 7.84E-10 1.08E-08 2.56E-08 5.16E-09 4.29E-10

Ce-144/Pr-144 1.61E-07 7.87E-07 2.82E-06 5.40E-06 2.92E-07 1.22E-06 3.47E-06 5.49E-06

Pr-143 2.27E-07 2.58E-07 1.02E-09 9.87E-13 1.98E-07 1.92E-07 6.01E-10 4.82E-13

Pr-144 0.00E+00 0.00E+00 0.00E+00 3.53E-35 0.00E+00 0.00E+00 0.00E+00 3.60E-35

Nd-147 3.80E-08 3.01E-08 2.20E-11 3.25E-15 3.41E-08 2.30E-08 1.34E-11 1.63E-15

Pm-147 2.24E-07 1.24E-06 5.69E-06 1.42E-05 4.72E-07 2.13E-06 7.75E-06 1.60E-05

Pm-148m 1.53E-09 4.86E-09 2.33E-09 4.76E-10 3.55E-09 9.61E-09 3.67E-09 6.20E-10

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 1.48E-08 7.73E-08 3.77E-07 1.02E-06 8.66E-08 3.85E-07 1.50E-06 3.34E-06

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nuclide

TRIGA 1

(decay time/delay time)

TRIGA 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/1 d 1 y/1 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/1 d

Sr-89/Y-89m 3.44E-03 7.81E-03 4.58E-03 6.17E-04 3.24E-03 1.07E-02 7.66E-03 2.55E-03

Sr-90 1.21E-03 4.15E-03 1.11E-02 1.86E-02 8.02E-05 4.00E-04 1.93E-03 5.41E-03

Y-90 1.21E-03 4.15E-03 7.58E-03 1.28E-02 8.03E-05 4.00E-04 1.93E-03 3.71E-03

Y-91 2.06E-06 4.96E-06 3.58E-06 6.81E-07 1.82E-06 6.37E-06 5.91E-06 2.64E-06

Ba-140 1.15E-04 7.70E-05 4.93E-07 3.60E-11 1.34E-04 1.31E-04 3.16E-07 1.85E-10

La-140 1.32E-04 8.86E-05 2.55E-07 1.86E-11 1.54E-04 1.51E-04 3.64E-07 9.56E-11

Ce-141 1.08E-08 1.95E-08 4.92E-09 1.62E-10 1.15E-08 3.04E-08 7.47E-09 7.60E-10

Ce-144/Pr-144 5.96E-07 1.90E-06 3.89E-06 4.23E-06 1.72E-07 7.97E-07 2.76E-06 5.35E-06

Pr-143 2.10E-07 1.56E-07 1.44E-09 1.94E-13 2.52E-07 2.72E-07 1.04E-09 1.02E-12

Pr-144 0.00E+00 0.00E+00 2.55E-35 2.77E-35 0.00E+00 0.00E+00 0.00E+00 3.51E-35

Nd-147 3.37E-08 1.74E-08 4.23E-11 6.11E-16 3.94E-08 2.96E-08 2.09E-11 3.14E-15

Pm-147 2.09E-06 7.06E-06 1.75E-05 2.61E-05 2.39E-07 1.25E-06 5.57E-06 1.41E-05

Pm-148m 7.40E-09 1.53E-08 6.39E-09 4.88E-10 4.46E-10 1.35E-09 6.25E-10 1.29E-10

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 3.24E-07 1.10E-06 2.89E-06 4.73E-06 1.12E-09 5.57E-09 2.63E-08 7.20E-08

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

ATTACHMENT G

RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

Page 7 of 20
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 54 of 70

Table G-3. Contributions to gross gamma activity for an unprocessed urinalysis.

Nuclide

ATR 1

(decay time/delay time)

ATR 2

(decay time/delay time)

ATR 3

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d

Mn-54 7.75E-09 3.52E-08 1.23E-07 2.26E-07 2.74E-09 1.91E-08 9.78E-08 2.09E-07 3.77E-09 2.25E-08 1.01E-07 2.07E-07

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 3.30E-09 1.20E-08 1.45E-08 6.59E-09 1.25E-09 6.94E-09 1.24E-08 6.53E-09 3.28E-09 1.57E-08 2.43E-08 1.24E-08

Co-60 3.26E-09 1.56E-08 7.09E-08 1.84E-07 1.06E-09 7.81E-09 5.20E-08 1.57E-07 1.45E-09 9.17E-09 5.33E-08 1.55E-07

Sr-89/Y-89m 3.18E-07 1.02E-06 7.15E-07 1.57E-07 1.75E-07 8.61E-07 8.85E-07 2.25E-07 2.22E-07 9.43E-07 8.44E-07 2.07E-07

Sr-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-91 5.81E-09 1.98E-08 1.80E-08 5.56E-09 3.05E-09 1.59E-08 2.12E-08 7.61E-09 3.89E-09 1.74E-08 2.03E-08 7.03E-09

Zr-95 1.45E-05 5.10E-05 5.33E-05 2.00E-05 7.28E-06 3.91E-05 6.01E-05 2.61E-05 9.46E-06 4.37E-05 5.87E-05 2.46E-05

Nb-95m 4.43E-08 1.55E-07 1.63E-07 6.08E-08 2.22E-08 1.19E-07 1.83E-07 7.95E-08 2.88E-08 1.33E-07 1.79E-07 7.48E-08

Nb-95 4.47E-05 1.99E-04 3.11E-04 1.29E-04 1.48E-05 1.22E-04 3.36E-04 1.67E-04 2.21E-05 1.46E-04 3.32E-04 1.57E-04

Mo-99/Tc-99m 3.25E-09 8.15E-12 1.77E-26 2.61E-46 3.85E-09 1.48E-11 4.70E-26 8.07E-46 3.60E-09 1.19E-11 3.30E-26 5.46E-46

Ru-103 1.04E-04 2.98E-04 1.20E-04 1.27E-05 6.40E-05 2.80E-04 1.66E-04 2.03E-05 7.90E-05 2.97E-04 1.53E-04 1.82E-05

Ru-106/Rh-106 3.51E-06 1.61E-05 5.91E-05 1.16E-04 1.23E-06 8.62E-06 4.65E-05 1.06E-04 1.70E-06 1.03E-05 4.84E-05 1.06E-04

Cd-113m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Cd-115m 2.07E-12 6.30E-12 3.41E-12 5.33E-13 1.19E-12 5.54E-12 4.41E-12 8.00E-13 1.50E-12 6.00E-12 4.16E-12 7.28E-13

Sb-125 9.19E-07 4.43E-06 1.92E-05 4.68E-05 2.95E-07 2.25E-06 1.43E-05 4.06E-05 4.25E-07 2.74E-06 1.53E-05 4.17E-05

Te-125m 3.97E-10 1.91E-09 8.29E-09 2.02E-08 1.27E-10 9.69E-10 6.19E-09 1.75E-08 1.83E-10 1.18E-09 6.59E-09 1.80E-08

Te-129m/Te-129 2.96E-06 7.73E-06 2.05E-06 1.25E-07 1.95E-06 7.81E-06 3.05E-06 2.16E-07 2.35E-06 8.11E-06 2.76E-06 1.88E-07

Te-132 2.70E-07 2.21E-09 1.22E-21 2.74E-38 3.19E-07 4.01E-09 3.25E-21 8.44E-38 2.98E-07 3.22E-09 2.28E-21 5.70E-38

Cs-134 2.20E-04 1.04E-03 4.35E-03 1.02E-02 5.14E-05 3.72E-04 2.29E-03 6.21E-03 3.44E-05 2.14E-04 1.15E-03 3.00E-03

Cs-136/Ba-136m 3.46E-06 3.46E-06 1.04E-08 1.68E-12 1.58E-06 2.42E-06 1.06E-08 2.01E-12 1.85E-06 2.44E-06 9.37E-09 1.70E-12

Cs-137/Ba-137m 1.37E-04 6.66E-04 3.15E-03 8.63E-03 4.44E-05 3.29E-04 2.28E-03 7.27E-03 6.35E-05 4.05E-04 2.45E-03 7.53E-03

Ba-140 5.41E-05 5.14E-05 1.21E-07 1.44E-11 5.48E-05 7.97E-05 2.76E-07 3.81E-11 5.64E-05 7.07E-05 2.14E-07 2.84E-11

La-140 3.32E-04 3.16E-04 7.45E-07 8.86E-11 3.37E-04 4.89E-04 1.70E-06 2.34E-10 3.47E-04 4.34E-04 1.31E-06 1.74E-10

Ce-141 5.66E-09 1.45E-08 3.49E-09 1.87E-10 3.80E-09 1.49E-08 5.26E-09 3.27E-10 4.60E-09 1.55E-08 4.78E-09 2.87E-10

Ce-144/Pr-144 1.31E-08 5.90E-08 2.00E-07 3.54E-07 4.71E-09 3.26E-08 1.62E-07 3.33E-07 6.60E-09 3.92E-08 1.70E-07 3.37E-07

Pr-143 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pr-144 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nd-147 7.38E-09 5.39E-09 3.73E-12 8.77E-17 7.09E-09 7.93E-09 8.06E-12 2.20E-16 8.01E-09 7.71E-09 6.83E-12 1.79E-16

Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pm-148m 1.46E-08 4.29E-08 1.95E-08 2.42E-09 3.39E-09 1.52E-08 1.01E-08 1.46E-09 7.15E-09 2.77E-08 1.61E-08 2.24E-09

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 1.34E-07 6.46E-07 2.98E-06 7.95E-06 3.70E-08 2.73E-07 1.86E-06 5.73E-06 2.19E-08 1.39E-07 8.25E-07 2.46E-06

Eu-155 9.47E-09 4.54E-08 2.05E-07 5.26E-07 2.74E-09 2.01E-08 1.33E-07 3.97E-07 4.31E-09 2.72E-08 1.57E-07 4.52E-07

Ta-182 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

ATTACHMENT G

RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

Page 8 of 20
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 55 of 70

Nuclide

FFTF 1

(decay time/delay time)

FFTF 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d

Mn-54 4.75E-05 1.93E-04 4.25E-04 4.95E-04 3.30E-05 1.46E-04 3.74E-04 4.78E-04

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 1.16E-04 3.75E-04 2.86E-04 8.24E-05 9.99E-05 3.53E-04 3.13E-04 9.88E-05

Co-60 1.86E-06 8.01E-06 2.29E-05 3.76E-05 9.05E-07 4.24E-06 1.41E-05 2.54E-05

Sr-89/Y-89m 1.16E-07 3.33E-07 1.47E-07 2.04E-08 1.20E-07 3.75E-07 1.92E-07 2.92E-08

Sr-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-91 2.64E-09 8.04E-09 4.60E-09 9.03E-10 2.71E-09 9.00E-09 5.98E-09 1.29E-09

Zr-95 1.10E-05 3.47E-05 2.29E-05 5.43E-06 1.12E-05 3.85E-05 2.95E-05 7.67E-06

Nb-95m 3.37E-08 1.06E-07 6.98E-08 1.65E-08 3.42E-08 1.17E-07 8.99E-08 2.34E-08

Nb-95 4.09E-05 1.52E-04 1.36E-04 3.51E-05 4.14E-05 1.68E-04 1.76E-04 4.96E-05

Mo-99/Tc-99m 2.85E-09 6.39E-12 8.73E-27 8.17E-47 2.81E-09 6.86E-12 1.09E-26 1.12E-46

Ru-103 2.12E-04 5.42E-04 1.37E-04 9.21E-06 2.11E-04 5.88E-04 1.73E-04 1.27E-05

Ru-106/Rh-106 1.08E-04 4.43E-04 1.03E-03 1.28E-03 7.93E-05 3.55E-04 9.56E-04 1.31E-03

Cd-113m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Cd-115m 1.95E-11 5.32E-11 1.81E-11 1.79E-12 1.88E-11 5.58E-11 2.21E-11 2.39E-12

Sb-125 1.85E-05 7.92E-05 2.16E-04 3.34E-04 1.16E-05 5.42E-05 1.72E-04 2.92E-04

Te-125m 7.98E-09 3.42E-08 9.33E-08 1.44E-07 4.99E-09 2.34E-08 7.42E-08 1.26E-07

Te-129m/Te-129 2.92E-08 6.84E-08 1.14E-08 4.43E-10 1.79E-08 4.57E-08 8.88E-09 3.77E-10

Te-132 2.47E-07 1.63E-09 3.44E-22 2.51E-39 2.55E-07 1.83E-09 4.48E-22 3.59E-39

Cs-134 1.97E-03 8.33E-03 2.20E-02 3.26E-02 5.68E-04 2.62E-03 8.04E-03 1.31E-02

Cs-136/Ba-136m 1.25E-04 1.12E-04 2.11E-07 2.18E-11 5.83E-05 5.69E-05 1.25E-07 1.41E-11

Cs-137/Ba-137m 7.95E-04 3.45E-03 1.03E-02 1.78E-02 4.31E-04 2.04E-03 7.04E-03 1.34E-02

Ba-140 3.72E-05 3.17E-05 4.71E-08 3.55E-12 3.82E-05 3.55E-05 6.12E-08 5.06E-12

La-140 2.29E-04 1.94E-04 2.89E-07 2.18E-11 2.35E-04 2.18E-04 3.76E-07 3.11E-11

Ce-141 4.27E-09 9.79E-09 1.49E-09 5.06E-11 4.37E-09 1.09E-08 1.93E-09 7.19E-11

Ce-144/Pr-144 2.35E-08 9.50E-08 2.03E-07 2.27E-07 1.86E-08 8.17E-08 2.03E-07 2.49E-07

Pr-143 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pr-144 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nd-147 6.02E-09 3.94E-09 1.72E-12 2.56E-17 5.95E-09 4.24E-09 2.15E-12 3.51E-17

Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pm-148m 1.40E-07 3.67E-07 1.05E-07 8.28E-09 7.15E-08 2.05E-07 6.81E-08 5.88E-09

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 2.40E-06 1.04E-05 3.02E-05 5.09E-05 8.53E-07 4.01E-06 1.36E-05 2.51E-05

Eu-155 7.93E-07 3.40E-06 9.67E-06 1.58E-05 4.02E-07 1.88E-06 6.20E-06 1.11E-05

Ta-182 8.62E-08 3.12E-07 4.01E-07 2.30E-07 8.41E-08 3.32E-07 4.96E-07 3.12E-07

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RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

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Nuclide

N Rx 1

(decay time/delay time)

N Rx 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d

Mn-54 6.96E-09 3.43E-08 1.26E-07 2.35E-07 1.38E-08 5.77E-08 1.69E-07 2.60E-07

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 2.02E-08 7.93E-08 1.01E-07 4.66E-08 2.06E-08 6.86E-08 7.00E-08 2.66E-08

Co-60 3.72E-08 1.93E-07 9.25E-07 2.42E-06 8.13E-08 3.60E-07 1.37E-06 2.97E-06

Sr-89/Y-89m 2.68E-07 9.34E-07 6.89E-07 1.53E-07 2.70E-07 8.02E-07 4.71E-07 8.63E-08

Sr-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-91 4.93E-09 1.82E-08 1.74E-08 5.45E-09 5.33E-09 1.67E-08 1.27E-08 3.30E-09

Zr-95 1.29E-05 4.90E-05 5.41E-05 2.04E-05 1.52E-05 4.91E-05 4.32E-05 1.35E-05

Nb-95m 3.92E-08 1.49E-07 1.65E-07 6.23E-08 4.62E-08 1.50E-07 1.32E-07 4.11E-08

Nb-95 3.81E-05 1.87E-04 3.13E-04 1.31E-04 5.47E-05 2.11E-04 2.56E-04 8.71E-05

Mo-99/Tc-99m 3.25E-09 8.81E-12 2.01E-26 3.01E-46 2.96E-09 6.84E-12 1.25E-26 1.54E-46

Ru-103 1.19E-04 3.67E-04 1.56E-04 1.67E-05 1.45E-04 3.81E-04 1.29E-04 1.14E-05

Ru-106/Rh-106 6.39E-06 3.18E-05 1.23E-04 2.44E-04 1.84E-05 7.80E-05 2.40E-04 3.94E-04

Cd-113m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Cd-115m 2.76E-12 9.10E-12 5.20E-12 8.21E-13 3.85E-12 1.08E-11 4.92E-12 6.43E-13

Sb-125 9.90E-07 5.20E-06 2.38E-05 5.86E-05 2.51E-06 1.11E-05 4.05E-05 8.25E-05

Te-125m 4.27E-10 2.25E-09 1.03E-08 2.53E-08 1.09E-09 4.81E-09 1.75E-08 3.56E-08

Te-129m/Te-129 3.41E-06 9.66E-06 2.71E-06 1.67E-07 4.08E-06 9.84E-06 2.19E-06 1.12E-07

Te-132 2.79E-07 2.48E-09 1.45E-21 3.26E-38 2.61E-07 1.97E-09 9.16E-22 1.71E-38

Cs-134 2.63E-05 1.35E-04 5.95E-04 1.41E-03 1.52E-04 6.60E-04 2.32E-03 4.54E-03

Cs-136/Ba-136m 3.83E-06 4.15E-06 1.31E-08 2.15E-12 7.24E-06 6.68E-06 1.68E-08 2.28E-12

Cs-137/Ba-137m 1.17E-04 6.12E-04 3.05E-03 8.45E-03 2.66E-04 1.19E-03 4.72E-03 1.08E-02

Ba-140 5.19E-05 5.35E-05 1.33E-07 1.60E-11 4.63E-05 4.06E-05 8.04E-08 7.98E-12

La-140 3.19E-04 3.29E-04 8.18E-07 9.81E-11 2.84E-04 2.49E-04 4.94E-07 4.90E-11

Ce-141 5.31E-09 1.47E-08 3.74E-09 2.03E-10 5.20E-09 1.23E-08 2.48E-09 1.11E-10

Ce-144/Pr-144 1.09E-08 5.31E-08 1.90E-07 3.39E-07 1.97E-08 8.22E-08 2.34E-07 3.45E-07

Pr-143 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pr-144 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nd-147 7.59E-09 6.01E-09 4.39E-12 1.04E-16 6.82E-09 4.59E-09 2.67E-12 5.23E-17

Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pm-148m 5.15E-09 1.64E-08 7.86E-09 9.85E-10 1.20E-08 3.24E-08 1.24E-08 1.28E-09

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

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Nuclide

N Rx 1

(decay time/delay time)

N Rx 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d

Eu-154 2.42E-08 1.26E-07 6.16E-07 1.65E-06 1.41E-07 6.29E-07 2.44E-06 5.43E-06

Eu-155 1.78E-08 9.24E-08 4.39E-07 1.14E-06 3.38E-08 1.50E-07 5.66E-07 1.22E-06

Ta-182 9.54E-12 4.19E-11 9.07E-11 8.33E-11 3.56E-11 1.33E-10 2.29E-10 1.74E-10

Nuclide

TRIGA 1

(decay time/delay time)

TRIGA 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d

Mn-54 3.52E-06 1.13E-05 2.22E-05 2.51E-05 8.80E-07 4.11E-06 1.47E-05 2.76E-05

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 2.52E-06 6.44E-06 4.41E-06 1.23E-06 1.91E-06 7.11E-06 8.82E-06 4.10E-06

Co-60 6.71E-05 2.28E-04 5.81E-04 9.26E-04 5.36E-06 2.65E-05 1.23E-04 3.25E-04

Sr-89/Y-89m 3.20E-07 7.26E-07 2.86E-07 3.86E-08 3.01E-07 9.97E-07 7.12E-07 1.60E-07

Sr-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-91 6.19E-09 1.49E-08 7.61E-09 1.45E-09 5.46E-09 1.91E-08 1.77E-08 5.61E-09

Zr-95 1.73E-05 4.30E-05 2.53E-05 5.83E-06 1.44E-05 5.19E-05 5.55E-05 2.12E-05

Nb-95m 5.29E-08 1.31E-07 7.72E-08 1.78E-08 4.38E-08 1.58E-07 1.69E-07 6.47E-08

Nb-95 6.42E-05 1.88E-04 1.51E-04 3.77E-05 4.25E-05 1.98E-04 3.22E-04 1.37E-04

Mo-99/Tc-99m 2.83E-09 5.00E-12 6.12E-27 5.56E-47 3.31E-09 8.52E-12 1.89E-26 2.85E-46

Ru-103 1.03E-04 2.08E-04 4.72E-05 3.07E-06 1.00E-04 2.95E-04 1.21E-04 1.31E-05

Ru-106/Rh-106 1.64E-05 5.32E-05 1.10E-04 1.33E-04 3.07E-06 1.45E-05 5.43E-05 1.09E-04

Cd-113m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Cd-115m 2.21E-12 4.75E-12 1.45E-12 1.39E-13 1.98E-12 6.21E-12 3.43E-12 5.48E-13

Sb-125 1.13E-05 3.80E-05 9.27E-05 1.39E-04 1.23E-06 6.06E-06 2.68E-05 6.68E-05

Te-125m 4.88E-09 1.64E-08 4.00E-08 5.99E-08 5.29E-10 2.61E-09 1.16E-08 2.88E-08

Te-129m/Te-129 2.82E-06 5.22E-06 7.80E-07 2.93E-08 2.86E-06 7.69E-06 2.08E-06 1.30E-07

Te-132 2.35E-07 1.36E-09 4.25E-22 5.84E-39 2.73E-07 2.31E-09 1.30E-21 2.98E-38

Cs-134 4.83E-04 1.61E-03 3.80E-03 5.46E-03 2.18E-06 1.06E-05 4.53E-05 1.08E-04

Cs-136/Ba-136m 5.68E-06 4.01E-06 6.75E-09 6.74E-13 1.22E-06 1.26E-06 3.84E-09 6.38E-13

Cs-137/Ba-137m 1.83E-03 6.27E-03 1.67E-02 2.81E-02 1.19E-04 5.95E-04 2.87E-03 8.04E-03

Ba-140 4.63E-05 3.11E-05 4.13E-08 3.02E-12 5.42E-05 5.30E-05 1.28E-07 1.55E-11

La-140 2.85E-04 1.91E-04 2.54E-07 1.85E-11 3.33E-04 3.25E-04 7.84E-07 9.52E-11

Ce-141 5.18E-09 9.37E-09 1.27E-09 4.19E-11 5.54E-09 1.46E-08 3.59E-09 1.96E-10

Ce-144/Pr-144 4.02E-08 1.28E-07 2.45E-07 2.66E-07 1.16E-08 5.38E-08 1.86E-07 3.36E-07

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Nuclide

TRIGA 1

(decay time/delay time)

TRIGA 2

(decay time/delay time)

10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d 10 d/30 d 40 d/30 d 180 d/30 d 1 y/30 d

Pr-143 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pr-144 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nd-147 6.74E-09 3.47E-09 1.35E-12 1.96E-17 7.87E-09 5.92E-09 4.18E-12 1.00E-16

Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pm-148m 2.49E-08 5.17E-08 1.32E-08 1.01E-09 1.50E-09 4.54E-09 2.11E-09 2.67E-10

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 5.29E-07 1.80E-06 4.69E-06 7.68E-06 1.83E-09 9.10E-09 4.30E-08 1.17E-07

Eu-155 1.32E-07 4.46E-07 1.13E-06 1.79E-06 1.93E-08 9.52E-08 4.38E-07 1.15E-06

Ta-182 5.18E-07 1.48E-06 1.71E-06 9.57E-07 4.19E-07 1.75E-06 3.66E-06 3.40E-06

Table G-4. Contributions to gross beta activity for an air or swipe sample.

Nuclide

ATR 1 ATR 2 ATR 3

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

Mn-54 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-60 8.21E-08 3.94E-07 1.79E-06 4.64E-06 2.68E-08 1.97E-07 1.31E-06 3.95E-06 3.65E-08 2.31E-07 1.34E-06 3.89E-06

Sr-89/Y-89m 2.69E-02 8.67E-02 6.06E-02 1.33E-02 1.48E-02 7.30E-02 7.50E-02 1.91E-02 1.88E-02 7.99E-02 7.16E-02 1.76E-02

Sr-90 4.01E-04 1.94E-03 9.17E-03 2.51E-02 1.30E-04 9.60E-04 6.66E-03 2.12E-02 1.85E-04 1.18E-03 7.16E-03 2.20E-02

Y-90 4.02E-04 1.94E-03 9.18E-03 2.51E-02 1.29E-04 9.61E-04 6.66E-03 2.12E-02 1.85E-04 1.18E-03 7.17E-03 2.20E-02

Y-91 3.22E-02 1.10E-01 9.95E-02 3.08E-02 1.69E-02 8.79E-02 1.17E-01 4.22E-02 2.15E-02 9.66E-02 1.12E-01 3.89E-02

Zr-95 3.48E-02 1.22E-01 1.28E-01 4.78E-02 1.74E-02 9.35E-02 1.44E-01 6.25E-02 2.26E-02 1.05E-01 1.40E-01 5.88E-02

Nb-95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Mo-99/Tc-99m 3.78E-03 9.46E-06 2.05E-20 3.03E-40 4.47E-03 1.71E-05 5.46E-20 9.36E-40 4.18E-03 1.38E-05 3.83E-20 6.33E-40

Ru-103 1.80E-02 5.15E-02 2.07E-02 2.19E-03 1.10E-02 4.83E-02 2.86E-02 3.51E-03 1.36E-02 5.14E-02 2.65E-02 3.13E-03

Ru-106/Rh-106 1.39E-03 6.39E-03 2.35E-02 4.61E-02 4.87E-04 3.42E-03 1.85E-02 4.21E-02 6.75E-04 4.09E-03 1.92E-02 4.22E-02

Cd-113m 7.23E-10 3.49E-09 1.63E-08 4.42E-08 2.88E-10 2.13E-09 1.46E-08 4.59E-08 1.04E-09 6.62E-09 3.97E-08 1.20E-07

Cd-115m 2.79E-06 8.49E-06 4.59E-06 7.19E-07 1.60E-06 7.47E-06 5.94E-06 1.08E-06 2.02E-06 8.09E-06 5.61E-06 9.81E-07

Sb-125 2.32E-05 1.12E-04 4.84E-04 1.18E-03 7.43E-06 5.66E-05 3.62E-04 1.02E-03 1.07E-05 6.92E-05 3.85E-04 1.05E-03

Te-129m/Te-129 6.00E-04 1.57E-03 4.17E-04 2.54E-05 3.97E-04 1.59E-03 6.19E-04 4.38E-05 4.78E-04 1.65E-03 5.60E-04 3.82E-05

Te-132 3.95E-03 3.24E-05 1.79E-17 4.01E-34 4.67E-03 5.87E-05 4.76E-17 1.24E-33 4.36E-03 4.72E-05 3.34E-17 8.35E-34

Cs-134 3.00E-04 1.42E-03 5.93E-03 1.39E-02 7.00E-05 5.06E-04 3.12E-03 8.46E-03 4.68E-05 2.91E-04 1.56E-03 4.09E-03

Cs-136/Ba-136m 8.78E-05 8.77E-05 2.62E-07 4.27E-11 4.01E-05 6.13E-05 2.69E-07 5.08E-11 4.70E-05 6.20E-05 2.38E-07 4.32E-11

Cs-137/Ba-137m 4.05E-04 1.96E-03 9.27E-03 2.54E-02 1.31E-04 9.69E-04 6.73E-03 2.14E-02 1.87E-04 1.19E-03 7.23E-03 2.22E-02

Ba-140 2.81E-02 2.67E-02 6.30E-05 7.49E-09 2.85E-02 4.14E-02 1.44E-04 1.98E-08 2.93E-02 3.67E-02 1.11E-04 1.48E-08

ATTACHMENT G

RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

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Nuclide

ATR 1 ATR 2 ATR 3

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

La-140 3.23E-02 3.08E-02 7.25E-05 8.63E-09 3.27E-02 4.77E-02 1.65E-04 2.28E-08 3.36E-02 4.23E-02 1.28E-04 1.70E-08

Ce-141 3.45E-02 8.84E-02 2.13E-02 1.14E-03 2.32E-02 9.08E-02 3.21E-02 2.00E-03 2.81E-02 9.47E-02 2.92E-02 1.75E-03

Ce-144/Pr-144 2.29E-02 1.03E-01 3.50E-01 6.20E-01 8.26E-03 5.70E-02 2.84E-01 5.83E-01 1.16E-02 6.87E-02 2.99E-01 5.90E-01

Pr-143 3.05E-02 3.20E-02 1.19E-04 2.61E-08 2.97E-02 4.77E-02 2.62E-04 6.63E-08 3.16E-02 4.38E-02 2.10E-04 5.12E-08

Pr-144 1.16E-02 5.22E-02 1.77E-01 3.13E-01 4.17E-03 2.88E-02 1.44E-01 2.94E-01 5.84E-03 3.47E-02 1.51E-01 2.98E-01

Nd-147 8.74E-03 6.39E-03 4.42E-06 1.04E-10 8.40E-03 9.40E-03 9.54E-06 2.60E-10 9.48E-03 9.13E-03 8.09E-06 2.13E-10

Pm-147 1.13E-03 5.75E-03 2.51E-02 6.09E-02 3.17E-04 2.90E-03 1.91E-02 5.36E-02 5.81E-04 4.22E-03 2.39E-02 6.48E-02

Pm-148m 1.70E-04 5.00E-04 2.27E-04 2.82E-05 3.94E-05 1.77E-04 1.18E-04 1.70E-05 8.33E-05 3.22E-04 1.87E-04 2.61E-05

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 9.45E-06 4.56E-05 2.11E-04 5.61E-04 2.61E-06 1.93E-05 1.31E-04 4.05E-04 1.55E-06 9.83E-06 5.82E-05 1.73E-04

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Ta-182 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Nuclide

FFTF 1 FFTF 2

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

Mn-54 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-60 4.69E-05 2.02E-04 5.76E-04 9.47E-04 2.28E-05 1.07E-04 3.54E-04 6.39E-04

Sr-89/Y-89m 9.82E-03 2.83E-02 1.24E-02 1.73E-03 1.01E-02 3.18E-02 1.63E-02 2.48E-03

Sr-90 8.10E-04 3.51E-03 1.05E-02 1.82E-02 4.41E-04 2.09E-03 7.21E-03 1.37E-02

Y-90 8.15E-04 3.51E-03 1.05E-02 1.82E-02 4.43E-04 2.09E-03 7.21E-03 1.37E-02

Y-91 1.46E-02 4.46E-02 2.55E-02 5.01E-03 1.50E-02 4.99E-02 3.31E-02 7.14E-03

Zr-95 2.64E-02 8.30E-02 5.48E-02 1.30E-02 2.69E-02 9.21E-02 7.06E-02 1.84E-02

Nb-95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Mo-99/Tc-99m 3.31E-03 7.42E-06 1.01E-20 9.49E-41 3.26E-03 7.96E-06 1.26E-20 1.30E-40

Ru-103 3.65E-02 9.35E-02 2.37E-02 1.59E-03 3.64E-02 1.02E-01 2.99E-02 2.20E-03

Ru-106/Rh-106 4.28E-02 1.76E-01 4.07E-01 5.08E-01 3.15E-02 1.41E-01 3.79E-01 5.19E-01

Cd-113m 2.11E-06 9.15E-06 2.70E-05 4.62E-05 8.57E-07 4.05E-06 1.38E-05 2.60E-05

Cd-115m 2.63E-05 7.18E-05 2.44E-05 2.41E-06 2.53E-05 7.52E-05 2.97E-05 3.22E-06

Sb-125 4.66E-04 2.00E-03 5.45E-03 8.43E-03 2.91E-04 1.37E-03 4.33E-03 7.35E-03

Te-129m/Te-129 5.94E-06 1.39E-05 2.32E-06 8.99E-08 3.64E-06 9.29E-06 1.80E-06 7.65E-08

Te-132 3.62E-03 2.39E-05 5.03E-18 3.67E-35 3.73E-03 2.68E-05 6.57E-18 5.26E-35

Cs-134 2.69E-03 1.13E-02 3.00E-02 4.44E-02 7.74E-04 3.57E-03 1.09E-02 1.78E-02

Cs-136/Ba-136m 3.18E-03 2.84E-03 5.36E-06 5.52E-10 1.48E-03 1.44E-03 3.16E-06 3.57E-10

Cs-137/Ba-137m 2.34E-03 1.01E-02 3.02E-02 5.25E-02 1.27E-03 6.00E-03 2.07E-02 3.95E-02

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Nuclide

FFTF 1 FFTF 2

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

Ba-140 1.93E-02 1.65E-02 2.45E-05 1.84E-09 1.99E-02 1.84E-02 3.18E-05 2.63E-09

La-140 2.22E-02 1.89E-02 2.82E-05 2.12E-09 2.28E-02 2.12E-02 3.66E-05 3.03E-09

Ce-141 2.61E-02 5.98E-02 9.07E-03 3.09E-04 2.67E-02 6.67E-02 1.18E-02 4.39E-04

Ce-144/Pr-144 4.12E-02 1.66E-01 3.55E-01 3.98E-01 3.25E-02 1.43E-01 3.55E-01 4.37E-01

Pr-143 1.80E-02 1.69E-02 3.98E-05 5.50E-09 1.82E-02 1.87E-02 5.10E-05 7.74E-09

Pr-144 2.08E-02 8.41E-02 1.80E-01 2.01E-01 1.64E-02 7.23E-02 1.79E-01 2.20E-01

Nd-147 7.13E-03 4.67E-03 2.03E-06 3.03E-11 7.05E-03 5.03E-03 2.55E-06 4.16E-11

Pm-147 5.06E-03 2.18E-02 5.94E-02 9.12E-02 3.45E-03 1.63E-02 5.16E-02 8.70E-02

Pm-148m 1.63E-03 4.27E-03 1.22E-03 9.64E-05 8.33E-04 2.39E-03 7.93E-04 6.85E-05

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 1.70E-04 7.32E-04 2.13E-03 3.60E-03 6.02E-05 2.83E-04 9.58E-04 1.77E-03

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Ta-182 1.54E-04 5.57E-04 7.16E-04 4.11E-04 1.50E-04 5.93E-04 8.86E-04 5.57E-04

Nuclide

FFTF 1 FFTF 2

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

Mn-54 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-60 9.37E-07 4.88E-06 2.33E-05 6.11E-05 2.05E-06 9.08E-06 3.45E-05 7.48E-05

Sr-89/Y-89m 2.27E-02 7.91E-02 5.84E-02 1.29E-02 2.29E-02 6.79E-02 3.99E-02 7.31E-03

Sr-90 3.11E-04 1.63E-03 8.14E-03 2.25E-02 6.61E-04 2.95E-03 1.17E-02 2.68E-02

Y-90 3.11E-04 1.64E-03 8.15E-03 2.25E-02 6.63E-04 2.95E-03 1.17E-02 2.68E-02

Y-91 2.73E-02 1.01E-01 9.66E-02 3.02E-02 2.95E-02 9.26E-02 7.06E-02 1.83E-02

Zr-95 3.08E-02 1.17E-01 1.29E-01 4.89E-02 3.63E-02 1.18E-01 1.03E-01 3.23E-02

Nb-95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Mo-99/Tc-99m 3.77E-03 1.02E-05 2.34E-20 3.49E-40 3.44E-03 7.94E-06 1.45E-20 1.78E-40

Ru-103 2.05E-02 6.34E-02 2.69E-02 2.88E-03 2.50E-02 6.58E-02 2.23E-02 1.97E-03

Ru-106/Rh-106 2.54E-03 1.26E-02 4.89E-02 9.68E-02 7.31E-03 3.09E-02 9.54E-02 1.56E-01

Cd-113m 3.35E-09 1.76E-08 8.66E-08 2.37E-07 4.60E-09 2.05E-08 8.06E-08 1.82E-07

Cd-115m 3.72E-06 1.23E-05 7.01E-06 1.11E-06 5.20E-06 1.46E-05 6.63E-06 8.66E-07

Sb-125 2.50E-05 1.31E-04 6.00E-04 1.48E-03 6.34E-05 2.81E-04 1.02E-03 2.08E-03

Te-129m/Te-129 6.93E-04 1.96E-03 5.50E-04 3.39E-05 8.29E-04 2.00E-03 4.46E-04 2.27E-05

Te-132 4.08E-03 3.63E-05 2.12E-17 4.78E-34 3.82E-03 2.89E-05 1.34E-17 2.50E-34

Cs-134 3.58E-05 1.83E-04 8.11E-04 1.91E-03 2.06E-04 8.99E-04 3.16E-03 6.18E-03

Cs-136/Ba-136m 9.71E-05 1.05E-04 3.32E-07 5.45E-11 1.84E-04 1.69E-04 4.25E-07 5.77E-11

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Nuclide

FFTF 1 FFTF 2

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

Cs-137/Ba-137m 3.43E-04 1.80E-03 8.99E-03 2.49E-02 7.85E-04 3.51E-03 1.39E-02 3.19E-02

Ba-140 2.70E-02 2.78E-02 6.92E-05 8.30E-09 2.40E-02 2.11E-02 4.18E-05 4.15E-09

La-140 3.10E-02 3.20E-02 7.97E-05 9.56E-09 2.76E-02 2.43E-02 4.81E-05 4.77E-09

Ce-141 3.24E-02 9.00E-02 2.29E-02 1.24E-03 3.17E-02 7.49E-02 1.51E-02 6.78E-04

Ce-144/Pr-144 1.90E-02 9.30E-02 3.33E-01 5.94E-01 3.46E-02 1.44E-01 4.10E-01 6.05E-01

Pr-143 2.79E-02 3.17E-02 1.25E-04 2.76E-08 2.43E-02 2.35E-02 7.39E-05 1.35E-08

Pr-144 9.61E-03 4.70E-02 1.68E-01 3.00E-01 1.75E-02 7.27E-02 2.07E-01 3.06E-01

Nd-147 9.00E-03 7.12E-03 5.20E-06 1.23E-10 8.07E-03 5.44E-03 3.16E-06 6.20E-11

Pm-147 1.22E-03 6.71E-03 3.09E-02 7.56E-02 2.56E-03 1.16E-02 4.21E-02 8.52E-02

Pm-148m 6.00E-05 1.91E-04 9.15E-05 1.15E-05 1.39E-04 3.77E-04 1.44E-04 1.49E-05

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 1.71E-06 8.91E-06 4.35E-05 1.17E-04 9.99E-06 4.44E-05 1.72E-04 3.83E-04

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Ta-182 1.70E-08 7.49E-08 1.62E-07 1.49E-07 6.36E-08 2.38E-07 4.09E-07 3.11E-07

Nuclide

TRIGA 1 TRIGA 2

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

Mn-54 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-60 1.69E-03 5.73E-03 1.46E-02 2.33E-02 1.35E-04 6.67E-04 3.09E-03 8.18E-03

Sr-89/Y-89m 2.71E-02 6.15E-02 2.42E-02 3.27E-03 2.55E-02 8.45E-02 6.03E-02 1.35E-02

Sr-90 5.26E-03 1.80E-02 4.79E-02 8.07E-02 3.48E-04 1.74E-03 8.37E-03 2.34E-02

Y-90 5.27E-03 1.80E-02 4.80E-02 8.07E-02 3.49E-04 1.74E-03 8.37E-03 2.34E-02

Y-91 3.43E-02 8.25E-02 4.22E-02 8.03E-03 3.03E-02 1.06E-01 9.82E-02 3.11E-02

Zr-95 4.15E-02 1.03E-01 6.07E-02 1.39E-02 3.44E-02 1.24E-01 1.33E-01 5.08E-02

Nb-95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Mo-99/Tc-99m 3.29E-03 5.81E-06 7.10E-21 6.45E-41 3.84E-03 9.89E-06 2.19E-20 3.31E-40

Ru-103 1.78E-02 3.59E-02 8.15E-03 5.29E-04 1.73E-02 5.10E-02 2.10E-02 2.27E-03

Ru-106/Rh-106 6.51E-03 2.11E-02 4.36E-02 5.27E-02 1.22E-03 5.75E-03 2.16E-02 4.32E-02

Cd-113m 2.75E-08 9.39E-08 2.47E-07 4.11E-07 1.93E-08 9.58E-08 4.58E-07 1.27E-06

Cd-115m 2.98E-06 6.41E-06 1.95E-06 1.88E-07 2.67E-06 8.36E-06 4.62E-06 7.39E-07

Sb-125 2.85E-04 9.59E-04 2.34E-03 3.50E-03 3.09E-05 1.53E-04 6.76E-04 1.68E-03

Te-129m/Te-129 5.74E-04 1.06E-03 1.58E-04 5.94E-06 5.80E-04 1.56E-03 4.23E-04 2.64E-05

Te-132 3.45E-03 2.00E-05 6.22E-18 8.54E-35 4.00E-03 3.38E-05 1.91E-17 4.36E-34

Cs-134 6.57E-04 2.19E-03 5.18E-03 7.44E-03 2.96E-06 1.44E-05 6.17E-05 1.47E-04

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Nuclide

TRIGA 1 TRIGA 2

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

Cs-136/Ba-136m 1.44E-04 1.02E-04 1.71E-07 1.71E-11 3.10E-05 3.19E-05 9.74E-08 1.62E-11

Cs-137/Ba-137m 5.39E-03 1.85E-02 4.91E-02 8.27E-02 3.51E-04 1.75E-03 8.45E-03 2.37E-02

Ba-140 2.41E-02 1.62E-02 2.15E-05 1.57E-09 2.81E-02 2.75E-02 6.63E-05 8.05E-09

La-140 2.76E-02 1.86E-02 2.47E-05 1.81E-09 3.23E-02 3.17E-02 7.64E-05 9.28E-09

Ce-141 3.16E-02 5.72E-02 7.76E-03 2.56E-04 3.38E-02 8.90E-02 2.19E-02 1.20E-03

Ce-144/Pr-144 7.04E-02 2.24E-01 4.29E-01 4.66E-01 2.03E-02 9.42E-02 3.26E-01 5.89E-01

Pr-143 2.58E-02 1.91E-02 4.03E-05 5.40E-09 3.09E-02 3.34E-02 1.28E-04 2.84E-08

Pr-144 3.56E-02 1.13E-01 2.17E-01 2.35E-01 1.02E-02 4.76E-02 1.65E-01 2.98E-01

Nd-147 7.98E-03 4.12E-03 1.60E-06 2.32E-11 9.32E-03 7.01E-03 4.95E-06 1.19E-10

Pm-147 1.13E-02 3.83E-02 9.31E-02 1.39E-01 1.30E-03 6.79E-03 3.02E-02 7.49E-02

Pm-148m 2.90E-04 6.02E-04 1.54E-04 1.18E-05 1.75E-05 5.28E-05 2.45E-05 3.11E-06

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 3.74E-05 1.27E-04 3.31E-04 5.42E-04 1.29E-07 6.42E-07 3.03E-06 8.25E-06

Eu-155 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Ta-182 9.25E-04 2.65E-03 3.06E-03 1.71E-03 7.48E-04 3.12E-03 6.53E-03 6.07E-03

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Table G-5. Contributions to gross gamma activity for an air or swipe sample.

Nuclide

ATR 1 ATR 2 ATR 3

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

Mn-54 1.91E-07 8.69E-07 3.04E-06 5.59E-06 6.77E-08 4.70E-07 2.41E-06 5.15E-06 9.30E-08 5.56E-07 2.49E-06 5.12E-06

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 2.76E-07 1.00E-06 1.21E-06 5.52E-07 1.05E-07 5.81E-07 1.04E-06 5.47E-07 2.75E-07 1.31E-06 2.04E-06 1.04E-06

Co-60 1.64E-07 7.89E-07 3.57E-06 9.28E-06 5.36E-08 3.94E-07 2.62E-06 7.90E-06 7.30E-08 4.62E-07 2.68E-06 7.79E-06

Sr-89/Y-89m 2.51E-06 8.06E-06 5.64E-06 1.24E-06 1.38E-06 6.79E-06 6.98E-06 1.78E-06 1.75E-06 7.43E-06 6.66E-06 1.63E-06

Sr-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-91 9.66E-05 3.29E-04 2.99E-04 9.25E-05 5.07E-05 2.64E-04 3.52E-04 1.26E-04 6.46E-05 2.90E-04 3.37E-04 1.17E-04

Zr-95 3.46E-02 1.21E-01 1.27E-01 4.76E-02 1.74E-02 9.31E-02 1.43E-01 6.22E-02 2.26E-02 1.04E-01 1.40E-01 5.85E-02

Nb-95 3.00E-02 1.44E-01 2.44E-01 1.02E-01 8.12E-03 8.19E-02 2.62E-01 1.33E-01 1.31E-02 1.01E-01 2.59E-01 1.25E-01

Mo-99/Tc-99m 4.06E-03 1.02E-05 2.20E-20 3.26E-40 4.81E-03 1.84E-05 5.87E-20 1.01E-39 4.49E-03 1.48E-05 4.12E-20 6.80E-40

Ru-103 1.68E-02 4.79E-02 1.93E-02 2.04E-03 1.03E-02 4.50E-02 2.66E-02 3.26E-03 1.27E-02 4.78E-02 2.46E-02 2.92E-03

Ru-106/Rh-106 2.38E-04 1.09E-03 4.01E-03 7.88E-03 8.33E-05 5.85E-04 3.16E-03 7.19E-03 1.15E-04 6.99E-04 3.29E-03 7.22E-03

Cd-113m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Cd-115m 6.19E-08 1.88E-07 1.02E-07 1.60E-08 3.56E-08 1.66E-07 1.32E-07 2.39E-08 4.47E-08 1.80E-07 1.24E-07 2.18E-08

Sb-125 1.98E-05 9.56E-05 4.14E-04 1.01E-03 6.37E-06 4.85E-05 3.10E-04 8.77E-04 9.17E-06 5.93E-05 3.30E-04 9.00E-04

Te-129m/Te-129 6.81E-05 1.78E-04 4.72E-05 2.88E-06 4.50E-05 1.80E-04 7.02E-05 4.97E-06 5.42E-05 1.87E-04 6.35E-05 4.34E-06

Te-132 3.63E-03 2.98E-05 1.65E-17 3.68E-34 4.30E-03 5.40E-05 4.38E-17 1.14E-33 4.01E-03 4.34E-05 3.07E-17 7.68E-34

Cs-134 6.68E-04 3.15E-03 1.32E-02 3.09E-02 1.56E-04 1.13E-03 6.94E-03 1.88E-02 1.04E-04 6.49E-04 3.49E-03 9.12E-03

Cs-136/Ba-136m 2.53E-04 2.53E-04 7.55E-07 1.23E-10 1.15E-04 1.77E-04 7.76E-07 1.46E-10 1.35E-04 1.78E-04 6.84E-07 1.24E-10

Cs-137/Ba-137m 3.64E-04 1.76E-03 8.34E-03 2.29E-02 1.18E-04 8.72E-04 6.05E-03 1.93E-02 1.68E-04 1.07E-03 6.50E-03 1.99E-02

Ba-140 1.13E-02 1.08E-02 2.54E-05 3.03E-09 1.15E-02 1.67E-02 5.80E-05 7.99E-09 1.18E-02 1.48E-02 4.48E-05 5.96E-09

La-140 6.95E-02 6.62E-02 1.56E-04 1.86E-08 7.03E-02 1.03E-01 3.56E-04 4.91E-08 7.25E-02 9.11E-02 2.75E-04 3.66E-08

Ce-141 1.66E-02 4.24E-02 1.02E-02 5.48E-04 1.11E-02 4.36E-02 1.54E-02 9.59E-04 1.35E-02 4.54E-02 1.40E-02 8.41E-04

Ce-144/Pr-144 1.55E-03 6.97E-03 2.36E-02 4.18E-02 5.57E-04 3.85E-03 1.92E-02 3.93E-02 7.80E-04 4.64E-03 2.01E-02 3.98E-02

Pr-143 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pr-144 2.97E-04 1.34E-03 4.55E-03 8.05E-03 1.07E-04 7.40E-04 3.69E-03 7.57E-03 1.50E-04 8.92E-04 3.88E-03 7.66E-03

Nd-147 1.75E-03 1.28E-03 8.82E-07 2.07E-11 1.68E-03 1.88E-03 1.91E-06 5.20E-11 1.89E-03 1.82E-03 1.62E-06 4.25E-11

Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pm-148m 5.74E-04 1.68E-03 7.66E-04 9.51E-05 1.33E-04 5.96E-04 3.98E-04 5.74E-05 2.81E-04 1.08E-03 6.32E-04 8.78E-05

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 1.54E-05 7.44E-05 3.44E-04 9.16E-04 4.27E-06 3.15E-05 2.14E-04 6.61E-04 2.53E-06 1.61E-05 9.51E-05 2.83E-04

Eu-155 1.13E-06 5.41E-06 2.44E-05 6.27E-05 3.26E-07 2.40E-06 1.59E-05 4.73E-05 5.13E-07 3.24E-06 1.87E-05 5.39E-05

Ta-182 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

ATTACHMENT G

RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

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Nuclide

FFTF 1 FFTF 2

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

Mn-54 1.17E-03 4.77E-03 1.05E-02 1.22E-02 8.15E-04 3.61E-03 9.23E-03 1.18E-02

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 9.69E-03 3.14E-02 2.40E-02 6.89E-03 8.36E-03 2.96E-02 2.62E-02 8.27E-03

Co-60 9.39E-05 4.04E-04 1.15E-03 1.89E-03 4.56E-05 2.14E-04 7.09E-04 1.28E-03

Sr-89/Y-89m 9.13E-07 2.63E-06 1.16E-06 1.61E-07 9.42E-07 2.96E-06 1.51E-06 2.30E-07

Sr-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-91 4.39E-05 1.34E-04 7.65E-05 1.50E-05 4.51E-05 1.50E-04 9.94E-05 2.14E-05

Zr-95 2.63E-02 8.27E-02 5.46E-02 1.29E-02 2.68E-02 9.17E-02 7.03E-02 1.83E-02

Nb-95 2.91E-02 1.13E-01 1.07E-01 2.80E-02 2.94E-02 1.25E-01 1.38E-01 3.95E-02

Mo-99/Tc-99m 3.55E-03 7.97E-06 1.09E-20 1.02E-40 3.50E-03 8.55E-06 1.36E-20 1.39E-40

Ru-103 3.40E-02 8.70E-02 2.21E-02 1.48E-03 3.39E-02 9.45E-02 2.78E-02 2.05E-03

Ru-106/Rh-106 7.31E-03 3.01E-02 6.96E-02 8.68E-02 5.38E-03 2.41E-02 6.49E-02 8.87E-02

Cd-113m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Cd-115m 5.85E-07 1.59E-06 5.42E-07 5.36E-08 5.62E-07 1.67E-06 6.60E-07 7.16E-08

Sb-125 3.99E-04 1.71E-03 4.67E-03 7.22E-03 2.49E-04 1.17E-03 3.71E-03 6.30E-03

Te-129m/Te-129 6.74E-07 1.58E-06 2.64E-07 1.02E-08 4.13E-07 1.05E-06 2.04E-07 8.67E-09

Te-132 3.33E-03 2.20E-05 4.63E-18 3.38E-35 3.43E-03 2.47E-05 6.04E-18 4.84E-35

Cs-134 5.99E-03 2.53E-02 6.68E-02 9.91E-02 1.73E-03 7.95E-03 2.44E-02 3.97E-02

Cs-136/Ba-136m 9.14E-03 8.18E-03 1.54E-05 1.59E-09 4.26E-03 4.15E-03 9.09E-06 1.03E-09

Cs-137/Ba-137m 2.11E-03 9.13E-03 2.72E-02 4.72E-02 1.14E-03 5.40E-03 1.87E-02 3.56E-02

Ba-140 7.81E-03 6.64E-03 9.88E-06 7.44E-10 8.02E-03 7.44E-03 1.28E-05 1.06E-09

La-140 4.78E-02 4.08E-02 6.07E-05 4.57E-09 4.91E-02 4.57E-02 7.89E-05 6.52E-09

Ce-141 1.25E-02 2.87E-02 4.35E-03 1.48E-04 1.28E-02 3.20E-02 5.64E-03 2.11E-04

Ce-144/Pr-144 2.78E-03 1.12E-02 2.40E-02 2.69E-02 2.19E-03 9.66E-03 2.40E-02 2.95E-02

Pr-143 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pr-144 5.35E-04 2.16E-03 4.62E-03 5.17E-03 4.22E-04 1.86E-03 4.61E-03 5.67E-03

Nd-147 1.42E-03 9.32E-04 4.06E-07 6.04E-12 1.41E-03 1.00E-03 5.08E-07 8.30E-12

Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pm-148m 5.49E-03 1.44E-02 4.12E-03 3.25E-04 2.81E-03 8.04E-03 2.67E-03 2.31E-04

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 2.77E-04 1.20E-03 3.48E-03 5.87E-03 9.84E-05 4.63E-04 1.56E-03 2.90E-03

Eu-155 9.45E-05 4.06E-04 1.15E-03 1.88E-03 4.79E-05 2.24E-04 7.39E-04 1.32E-03

Ta-182 2.25E-04 8.15E-04 1.05E-03 6.00E-04 2.19E-04 8.67E-04 1.29E-03 8.13E-04

ATTACHMENT G

RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

Page 18 of 20
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Nuclide

N Rx 1 N Rx 2

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

Mn-54 1.72E-07 8.46E-07 3.12E-06 5.79E-06 3.40E-07 1.42E-06 4.18E-06 6.42E-06

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 1.69E-06 6.64E-06 8.49E-06 3.90E-06 1.72E-06 5.75E-06 5.86E-06 2.22E-06

Co-60 1.87E-06 9.75E-06 4.66E-05 1.22E-04 4.10E-06 1.82E-05 6.91E-05 1.50E-04

Sr-89/Y-89m 2.11E-06 7.36E-06 5.43E-06 1.20E-06 2.13E-06 6.32E-06 3.71E-06 6.80E-07

Sr-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-91 8.20E-05 3.03E-04 2.90E-04 9.06E-05 8.86E-05 2.78E-04 2.12E-04 5.48E-05

Zr-95 3.07E-02 1.17E-01 1.29E-01 4.87E-02 3.62E-02 1.17E-01 1.03E-01 3.21E-02

Nb-95 2.52E-02 1.34E-01 2.46E-01 1.05E-01 3.86E-02 1.56E-01 2.02E-01 6.94E-02

Mo-99/Tc-99m 4.05E-03 1.10E-05 2.51E-20 3.75E-40 3.69E-03 8.53E-06 1.55E-20 1.92E-40

Ru-103 1.91E-02 5.90E-02 2.51E-02 2.68E-03 2.32E-02 6.12E-02 2.07E-02 1.83E-03

Ru-106/Rh-106 4.34E-04 2.16E-03 8.36E-03 1.66E-02 1.25E-03 5.29E-03 1.63E-02 2.67E-02

Cd-113m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Cd-115m 8.25E-08 2.72E-07 1.56E-07 2.46E-08 1.15E-07 3.24E-07 1.47E-07 1.92E-08

Sb-125 2.14E-05 1.12E-04 5.14E-04 1.27E-03 5.43E-05 2.41E-04 8.75E-04 1.78E-03

Te-129m/Te-129 7.85E-05 2.23E-04 6.23E-05 3.84E-06 9.40E-05 2.27E-04 5.05E-05 2.58E-06

Te-132 3.75E-03 3.34E-05 1.95E-17 4.40E-34 3.51E-03 2.66E-05 1.23E-17 2.30E-34

Cs-134 7.98E-05 4.09E-04 1.81E-03 4.27E-03 4.60E-04 2.00E-03 7.05E-03 1.38E-02

Cs-136/Ba-136m 2.80E-04 3.03E-04 9.56E-07 1.57E-10 5.29E-04 4.88E-04 1.22E-06 1.66E-10

Cs-137/Ba-137m 3.09E-04 1.62E-03 8.09E-03 2.24E-02 7.06E-04 3.16E-03 1.25E-02 2.87E-02

Ba-140 1.09E-02 1.12E-02 2.79E-05 3.35E-09 9.71E-03 8.51E-03 1.69E-05 1.67E-09

La-140 6.67E-02 6.89E-02 1.72E-04 2.06E-08 5.95E-02 5.23E-02 1.04E-04 1.03E-08

Ce-141 1.56E-02 4.32E-02 1.10E-02 5.94E-04 1.52E-02 3.59E-02 7.26E-03 3.25E-04

Ce-144/Pr-144 1.28E-03 6.27E-03 2.25E-02 4.01E-02 2.33E-03 9.71E-03 2.76E-02 4.08E-02

Pr-143 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pr-144 2.47E-04 1.21E-03 4.32E-03 7.71E-03 4.49E-04 1.87E-03 5.32E-03 7.85E-03

Nd-147 1.80E-03 1.42E-03 1.04E-06 2.46E-11 1.61E-03 1.09E-03 6.31E-07 1.24E-11

Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pm-148m 2.02E-04 6.43E-04 3.08E-04 3.87E-05 4.70E-04 1.27E-03 4.85E-04 5.03E-05

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 2.78E-06 1.46E-05 7.10E-05 1.91E-04 1.63E-05 7.25E-05 2.81E-04 6.26E-04

Eu-155 2.12E-06 1.10E-05 5.23E-05 1.36E-04 4.03E-06 1.78E-05 6.74E-05 1.45E-04

Ta-182 2.49E-08 1.09E-07 2.37E-07 2.17E-07 9.29E-08 3.47E-07 5.98E-07 4.54E-07

ATTACHMENT G

RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

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Nuclide

TRIGA 1 TRIGA 2

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

Mn-54 8.69E-05 2.79E-04 5.49E-04 6.20E-04 2.17E-05 1.01E-04 3.62E-04 6.81E-04

Fe-55 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Co-58 2.11E-04 5.39E-04 3.69E-04 1.03E-04 1.60E-04 5.95E-04 7.38E-04 3.43E-04

Co-60 3.38E-03 1.15E-02 2.93E-02 4.67E-02 2.70E-04 1.33E-03 6.17E-03 1.64E-02

Sr-89/Y-89m 2.52E-06 5.72E-06 2.25E-06 3.04E-07 2.37E-06 7.86E-06 5.61E-06 1.26E-06

Sr-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-90 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Y-91 1.03E-04 2.47E-04 1.27E-04 2.41E-05 9.08E-05 3.18E-04 2.95E-04 9.33E-05

Zr-95 4.13E-02 1.02E-01 6.04E-02 1.39E-02 3.43E-02 1.24E-01 1.32E-01 5.06E-02

Nb-95 4.57E-02 1.40E-01 1.19E-01 3.00E-02 2.80E-02 1.42E-01 2.52E-01 1.09E-01

Mo-99/Tc-99m 3.53E-03 6.24E-06 7.63E-21 6.93E-41 4.12E-03 1.06E-05 2.35E-20 3.55E-40

Ru-103 1.66E-02 3.34E-02 7.58E-03 4.93E-04 1.61E-02 4.74E-02 1.95E-02 2.11E-03

Ru-106/Rh-106 1.11E-03 3.61E-03 7.46E-03 9.00E-03 2.08E-04 9.83E-04 3.69E-03 7.39E-03

Cd-113m 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Cd-115m 6.61E-08 1.42E-07 4.34E-08 4.17E-09 5.92E-08 1.86E-07 1.03E-07 1.64E-08

Sb-125 2.44E-04 8.21E-04 2.00E-03 3.00E-03 2.64E-05 1.31E-04 5.79E-04 1.44E-03

Te-129m/Te-129 6.50E-05 1.20E-04 1.80E-05 6.74E-07 6.58E-05 1.77E-04 4.80E-05 2.99E-06

Te-132 3.17E-03 1.84E-05 5.72E-18 7.86E-35 3.68E-03 3.11E-05 1.76E-17 4.01E-34

Cs-134 1.46E-03 4.89E-03 1.15E-02 1.66E-02 6.60E-06 3.21E-05 1.37E-04 3.28E-04

Cs-136/Ba-136m 4.14E-04 2.92E-04 4.93E-07 4.92E-11 8.93E-05 9.19E-05 2.81E-07 4.66E-11

Cs-137/Ba-137m 4.85E-03 1.66E-02 4.42E-02 7.45E-02 3.16E-04 1.58E-03 7.61E-03 2.13E-02

Ba-140 9.72E-03 6.52E-03 8.67E-06 6.33E-10 1.14E-02 1.11E-02 2.68E-05 3.25E-09

La-140 5.95E-02 4.01E-02 5.33E-05 3.89E-09 6.96E-02 6.83E-02 1.65E-04 2.00E-08

Ce-141 1.52E-02 2.75E-02 3.72E-03 1.23E-04 1.62E-02 4.27E-02 1.05E-02 5.76E-04

Ce-144/Pr-144 4.75E-03 1.51E-02 2.89E-02 3.14E-02 1.37E-03 6.35E-03 2.20E-02 3.98E-02

Pr-143 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pr-144 9.14E-04 2.91E-03 5.57E-03 6.05E-03 2.63E-04 1.22E-03 4.24E-03 7.65E-03

Nd-147 1.59E-03 8.22E-04 3.20E-07 4.63E-12 1.86E-03 1.40E-03 9.89E-07 2.37E-11

Pm-147 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Pm-148m 9.78E-04 2.03E-03 5.19E-04 3.96E-05 5.89E-05 1.78E-04 8.26E-05 1.05E-05

Sm-151 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00

Eu-154 6.10E-05 2.08E-04 5.41E-04 8.85E-04 2.11E-07 1.05E-06 4.95E-06 1.35E-05

Eu-155 1.57E-05 5.32E-05 1.35E-04 2.13E-04 2.30E-06 1.13E-05 5.22E-05 1.37E-04

Ta-182 1.35E-03 3.87E-03 4.47E-03 2.50E-03 1.09E-03 4.56E-03 9.54E-03 8.86E-03

ATTACHMENT G

RADIONUCLIDE CONTRIBUTIONS TO SAMPLE ACTIVITY

Page 20 of 20
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 67 of 70

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. Considerations about parameter selection are included. 

In the examples below, one reactor case is chosen to work through the steps and mathematics. 

However, in dose reconstructions intakes and doses to the organ of concern have to be calculated for 

all the reactor cases to determine the one that maximizes the organ. This is necessary to justify use 

of the constant (upper bound) distribution. However, the user can choose to determine the intakes for 

one reactor case only and use the OTIB-0054 tool calculate the intakes (and organ doses) for the 

other reactor cases without having to use IMBA more than once.

Example 1. Gross beta bioassay data (with major chemical processing) from waste 

management worker.

In this example the bioassay data in Table H-1 were assumed for a (reactor) waste management 

worker employed from January 28, 1960, through January 25, 1971.

Table H-1. Gross beta urinalysis results 

and ATR-1 90Sr results (pCi/24 hr).

Date Gross betaa Sr-90b

07/23/1964 56 26.4

10/20/1964 42 19.8

08/13/1965 33 15.5

10/21/1965 41 19.3

02/16/1966 57 26.8

05/03/1966 31 14.6

08/11/1966 17 8.0

11/02/1966 31 14.6

a. Detection threshold of 20 pCi/24 hr.

b. Calculated by multiplying 1-year Sr-90 fraction, 

0.471, from Table 7-2 times the bioassay 

results.

Because the worker was involved in waste management activities, Table 5-3 indicates the use of a 1-

year fuel decay assumption for selection of intake parameters is appropriate. For gross beta 

urinalyses, 90Sr is the indicator radionuclide. 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, 1-year decay period, are used. There are nine options for the representative 

reactor cases and one must calculate nine separate intakes of 90Sr using the Table 7-2 fractions; 

however, calculating the intake once is sufficient if a tool automating the process is used to calculate 

the other eight cases. 

Using the ATR-1 reactor case in Table 7-2, the activity of 90Sr in each gross beta urinalysis result is 

0.471 times the gross beta result. These values are listed in Table H-1.

The 90Sr urinalysis estimates from Table H-1 were used to fit a chronic intake of 90Sr type F for the 

period from January 28, 1960, through January 25, 1971, and resulted in a calculated 90Sr intake rate 

of 76.0 pCi/d. Because all biological model parameters are the same for the other eight reactor 

cases, the 90Sr intake rates for the other eight reactor cases are the product of the ATR-1 intake rate 

ATTACHMENT H

EXAMPLE INTAKES ESTIMATED FROM BIOASSAY AND FROM AIR SAMPLE DATA

Page 1 of 4
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 68 of 70

and the ratio of the fraction of beta activity in urine for each reactor (from Table 7-2) to the ATR-1 

fraction. For example the ATR-2 90Sr intake rate is 

ATR-2 90Sr intake rate = (76.0 pCi/d)[(0.412)/(0.471)] = 66.5 pCi/d.

The intake rates from the other radionuclides in the ATR-1 source term were determined by 

multiplying the ATR-1 1-year 90Sr ratios in Table 7-3a by the 90Sr intake rate; the results are listed in 

Table H-2. Similarly, the intake rates of the other radionuclides in the other reactor source terms are 

determined using their corresponding ratios in Table 7-3.

Table H-2. ATR-1 radionuclide intake rates 

derived from 90Sr intake rate of 76.0 pCi/d.

Radionuclide 1-yr ratiosa pCi/d

Co-60 1.85E-4 1.41E-02

Sr-89 5.29E-1 4.02E+01

Sr-90 1.00 7.60E+01

Y-90 1.00 7.60E+01

Y-91 1.23 9.35E+01

Zr-95 1.90 1.44E+02

Nb-95 4.08 3.10E+02

Ru-103 8.73E-2 6.63E+00

Ru-106 9.17E-1 6.97E+01

I-131 6.33E-11 4.81E-09

Cs-134 5.52E-1 4.20E+01

Cs-137 1.01 7.68E+01

Ce-141 4.54E-2 3.45E+00

Ce-144 1.26E+1 9.58E+02

Pr-143 1.04E-6 7.90E-05

Pm-147 2.42 1.84E+02

Eu-154 2.23E-2 1.69E+00

Total daily intake - 2.08E+03

a. From Table 7-3a Sr-90 factors.

To determine the maximum dose to the organ of concern, the radionuclide intake rates for each of the 

nine representative reactor cases are used. Annual doses to the organ of concern are calculated 

assuming the absorption type for each radionuclide that maximizes the dose to the organ, with the 

exception of 89Sr and 90Sr, which are assumed to be type F, and 131I, which is assumed to be I2 vapor. 

Assigning 131I intake might not be appropriate for all cases (see Section 7.1.3)

Example 2. Gross beta bioassay data (with minor chemical processing) from worker involved 

in spent fuel storage operations.

Example 2 uses the bioassay data in Table H-1 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 5-3. For this example, it is assumed that the urinalysis did not 

include a chemical separation step, so the fractions of 90Sr in the urine sample come from Table 7-1a; 

the resulting 90Sr urine activity using the ATR-1 reactor case is listed in Table H-3.

ATTACHMENT H

EXAMPLE INTAKES ESTIMATED FROM BIOASSAY AND FROM AIR SAMPLE DATA

Page 2 of 4
Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 69 of 70

Table H-3. Gross beta urinalysis results 

and ATR-1 90Sr results (pCi/24 hr).

Date Urinalysisa Sr-90b

07/23/1964 56 1.2

10/20/1964 42 0.9

08/13/1965 33 0.7

10/21/1965 41 0.9

02/16/1966 57 1.2

05/03/1966 31 0.7

08/11/1966 17 0.4

11/02/1966 31 0.7

a. Detection threshold of 20 pCi/24 hr.

b. Calculated by multiplying 10-day ATR-1 Sr-90 

fraction, 2.14E-02, from Table 7-1a times the 

bioassay results.

The 90Sr data are used to fit a chronic intake of 90Sr type F for the period from January 28, 1960, 

through January 25, 1971, which resulted in a calculated 90Sr intake rate of 3.51 pCi/d. As discussed 

for Example 1, the 90Sr intake rates for the other eight representative reactor cases are given by the 

product of the ATR-1 intake rate and the ratio of the fraction of beta activity in urine for each reactor 

(from Table 7-1a) to the ATR-1 fraction. For example, the ATR-2 90Sr intake rate is 

ATR-2 90Sr intake rate = (3.51 pCi/d)[(0.0124)/(0.0214)] = 2.03 pCi/d.

The intake rates from the other radionuclides in the ATR-1 source term are determined by multiplying 

the ATR-1 10-day 90Sr ratios in Table 7-3a by the 90Sr intake rate; the results are listed in Table H-4. 

Similarly, the intake rates of the other radionuclides in the other reactor source terms are determined 

using their corresponding ratios in Table 7-3 

Table H-4. ATR-1 radionuclide intake rates 

derived from 90Sr intake rate of 3.51 pCi/d. 

Radionuclide 10-d ratiosa pCi/d

Co-60 2.05E-04 7.19E-04

Sr-89 6.72E+01 2.36E+02

Sr-90 1.00E+00 3.51E+00

Y-90 1.00E+00 3.52E+00

Y-91 8.04E+01 2.82E+02

Zr-95 8.67E+01 3.04E+02

Nb-95 7.49E+01 2.63E+02

Ru-103 4.50E+01 1.58E+02

Ru-106 1.74E+00 6.10E+00

I-131 1.21E+03 4.25E+03

Cs-134 7.48E-01 2.62E+00

Cs-137 1.01E+00 3.54E+00

Ce-141 8.62E+01 3.02E+02

Ce-144 2.89E+01 1.01E+02

Pr-143 7.60E+01 2.67E+02

Pm-147 2.81E+00 9.86E+00

Eu-154 2.36E-02 8.28E-02

Total daily intake - 6.19E+03

a. From Table 7-3a, ATR-1, Sr-90, 10-d decay.

ATTACHMENT H

EXAMPLE INTAKES ESTIMATED FROM BIOASSAY AND FROM AIR SAMPLE DATA

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Document No. ORAUT-OTIB-0054 Revision No. 01 Effective Date: 06/13/2013 Page 70 of 70

Example 3. Gross gamma 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 if 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 5-3. This example assumes monthly average gamma air 

concentrations in the dissolution area ranged from 1 × 10-14 to 1 × 10-10 μCi/mL during the period from 

1960 through 1971. It further assumes that the air concentrations were not tabulated, so no overall 

average value is available. The following calculation is used to bound radionuclide exposures using

the upper range of the monthly average air concentrations, 1 × 10-10 μCi/mL.

Each 180-day 137Cs fraction from Table 7-4b for the nine representative reactors is used to estimate 

the 137Cs bounding air concentrations by multiplying the fraction by 1 × 10-10 μCi/mL. The daily intake 

rates of 137Cs are given by

137Cs intake, pCi/d = (Table 7-4b gamma fraction)(air conc. in μCi/mL)(106 pCi/μCi)(106 mL/m3

)(2400 m3)/365 d. 

Or for a gamma-based air concentration of 1 x10-10 μCi/mL 

137Cs intake, pCi/d = (6.58 x102

)(Table 7-4b gamma fraction).

These 137Cs intake rates are used with the 180-day intake fractions in Table 7-3a to estimate air 

concentrations and inhalation intakes of associated radionuclides. The results for ATR-1 are listed in 

Table H-5. An ingestion intake must also be estimated (NIOSH 2004). Table H-5 summarizes the 

results of the air concentrations and intakes for ATR-1.

Table H-5. ATR-1 radionuclide air concentrations and intake rates derived from a gamma air 

concentration of 1 × 10-10 μCi/mL.

Radionuclide 180 da

Air concentration 

(μCi/mL)

Inhalation intake 

(pCi/d)

Ingestion intake 

(pCi/d) b

Co-60 1.93E-04 3.53E-16 2.32E-03 7.06E-05

Sr-89 6.54 1.20E-11 7.87E+01 2.39

Sr-90 9.89E-01 1.81E-12 1.19E+01 3.62E-01

Y-90 9.90E-01 1.81E-12 1.19E+01 3.62E-01

Y-91 1.07E+01 1.96E-11 1.29E+02 3.92

Zr-95 1.38E+01 2.53E-11 1.66E+02 5.05

Nb-95 2.63E+01 4.81E-11 3.16E+02 9.63

Ru-103 2.24 4.10E-12 2.70E+01 8.20E-01

Ru-106 1.27 2.32E-12 1.53E+01 4.65E-01

I-131 5.23E-04 9.57E-16 6.29E-03 1.91E-04

Cs-134 6.40E-01 1.17E-12 7.70E+00 2.34E-01

Cs-137 1.00 1.83E-12 1.20E+01 3.66E-01

Ce-141 2.30 4.21E-12 2.77E+01 8.42E-01

Ce-144 1.91E+01 3.50E-11 2.30E+02 6.99

Pr-143 1.29E-02 2.36E-14 1.55E-01 4.72E-03

Pm-147 2.71 4.96E-12 3.26E+01 9.92E-01

Eu-154 2.27E-02 4.15E-14 2.73E-01 8.31E-03

a. From Table 7-3a ATR-1 Cs-137 ratios.

b. From guidance in NIOSH 2004, Rev. 0.

ATTACHMENT H

EXAMPLE INTAKES ESTIMATED FROM BIOASSAY AND FROM AIR SAMPLE DATA

ATTACHMENT H

EXAMPLE INTAKES ESTIMATED FROM BIOASSAY AND FROM AIR SAMPLE DATA

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