NUREG/CR-6567 

PNNL- 11659

Low-Level Radioactive 

Waste Classification, 

Characterization, and 

Assessment: Waste 

Streams and 

Neutron-Activated Metals 

Pacific Northwest National Laboratory 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Regulatory Research 

Washington, DC 20555-0001
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NUREG/CR-6567 

PNNL-11659 

Low-Level Radioactive 

Waste Classification, 

Characterization, and 

Assessment: Waste 

Streams and 

Neutron-Activated Metals 

Manuscript Completed: July 2000 

Date Published: August 2000 

Prepared by 

D. E. Robertson, C. W. Thomas, S. L. Pratt, 

E. A. Lepel, V. W. Thomas 

Pacific Northwest National Laboratory 

Richland, WA 99352 

P. R. Reed, NRC Project Manager 

Prepared for 

Division of Risk Analysis and Applications 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

NRC Job Code L1808
Disclaimer 

NUIREG/CR-6567 is not a substitute for NRC regulations and compliance is not required. The approaches and/or 

methods described in this NUREG/CR are provided for information only. Publication of this report does not 

necessarily constitute NRC approval or agreement with the information contained herein.
Abstract

This study was conducted to provide NRC, other 

federal and state regulatory agencies, nuclear 

utilities, and low-level radioactive waste (LLW) 

managers with further information on radionuclide 

source terms that will improve the health and safety 

aspects of LLW management and disposal. The 

goal of this work was to enhance the understanding 

of the occurrence, distribution, and assessment of 

the radiological impacts of radionuclides associated 

with neutron-activated metals and other LLW 

streams from commercial nuclear power stations. 

This study focused on identifying and characterizing 

a group of very long-lived radionuclides that are not 

specified in 10 CFR Part 61, but which are present 

in significant concentrations in various types of 

LLW materials generated at commercial nuclear 

power stations. The concentrations of '

0Be, 36

C1, 

93Mo, 93

mNb, 1OSmAg, 113m Cd, and.2

.m Sn (as well as 

the specified 10 CFR Part 61 radionuclides) have 

been measured in a variety of neutron-activated 

metal and spent primary demineralization resin 

LLW samples obtained from U. S. nuclear power 

stations. This work required the development of 

new or modified radiochemical separation, 

purification, and counting procedures to measure 

these radionuclides in the LLW materials of 

interest. Of this group of radionuclides, the l`Be, 36

C1, and 1°smAg appear to be present in some types 

of LLW materials in sufficient quantities to warrant 

further investigations to better assess their 

radiological and environmental impacts associated 

with LLW disposal. It is recommended that further 

studies be focused on: 1) providing an accurate 

assessment of the total quantities of these 

radionuclides in LLW from these sources, 2) 

determine the leaching characteristics of these LLW 

materials, 3) determine the migration behavior and 

environmental pathways of these radionuclides upon 

release from LLW disposal facilities, and 4) provide 

performance assessment modelers with the 

necessary radiological/geochemical information to 

better predict the potential impacts from disposal of 

this group of radionuclides. Activity scaling factors

have been developed for all of the additional 

radionuclides measured in this study. The use of 

activity scaling factors for estimating the 

concentration of a radionuclide in LLW is permitted 

by 10 CFR Part 61.55, and this method is desirable 

for radionuclides such as those listed above that are 

difficult to measure directly in LLW materials. 

Consistent scaling factors for a wide variety of 

LLW materials were determined for 93

Mo and 93'Nb 

relative to 6

"Co over a concentration range of four 

orders of magnitude. Generic scaling factors for the 

other long-lived radionuclides measured in this 

study showed much greater variability due to the 

unique composition of some of the materials. 

However, in some cases useful scaling factors were 

obtained when similiar LLW materials were 

considered as a group. In addition to these 

radionuclides, an assessment was conducted to 

provide an updated understanding of the occurrence, 

distribution, and activity scaling factors of `

4

C, 99Tc, 

and 1291, as well as an evaluation of the59 NV' Ni 

activity ratio in LLW. This assessment has shown 

that the concentrations and associated activity 

scaling factors of 99

Tc and 129 I are several orders of 

magnitude lower than values published earlier in 

nuclear power industry data bases, thereby 

significantly reducing the radiological/environmental 

impacts associated with their disposal in LLW. The 

assessment of the 59

Ni/63

Ni activity ratio in LLW 

indicated that early reported values of this ratio 

were too high by up to factors of 5-10. This was 

probably due to systematic and/or random 

uncertainties in their radiochemical analyses. More 

recent analyses and assessments indicate that this 

ratio should range from about 0.006 to 0.012, 

depending upon the neutron flux and irradiation 

time. Activity scaling factors for '°Be, 36CI, 93

Mo, 

93mib, losmAg, 113mCd, and"12"m Sn relative to60 Co 

were determined for a variety of activated metal 

and LLW samples. These scaling factors are 

believed to be the first published in the literature 

for these radionuclides.

ii NUREG/CR-6567
Contents 

Abstract . 111 

L ist of Figures . . . ......................................................... .... .. vii 

L ist of T ables . . . ............................................... ... ... ..... ... viii 

Executive Summary x.................................

Forew ord . . . . . . . ....... ....... ....... ....... ...... .. .. ... ... .. . . . . . . .. . .. ... . xv 

A cknow ledgm ents ............................................................. xvii 

1 Introduction 1................................

1.1 Low-Level Waste Classification and Characterization Requirements of 1 OCFR61 ............. 1 

1.2 Project Scope .......................................................... 2 

2 Radionuclide Production and Distribution in LLW Generated Nuclear Power Stations .............. 3 

2.1 Neutron Activation of Materials Inside Reactor Pressure Vessels 3........................3 

2.1.1 Composition of Reactor Internal Component Construction Materials ................. 6 

2.1.2 Long-Lived Neutron Activation Products 8................................

2.2 Neutron Activation of Corrosion Products Released to Reactor Coolant .................... 8 

3 Long-Lived Radionuclides Not Considered in 1 OCFR61 in Low-Level Waste .................... 11 

3.1 Beryllium- .10 .......................................................... 11 

3.2 Chlorine-36 ........................................................... 11 

3.3 M olybdenum -93 ........................................................ 13 

3.4 N iobium -93m .......................................................... 13 

3.5 Silver-108m ........................................................... 13 

3.6 Cadm ium -113m ........................................................ 13 

3.7 Tin-121m ............................................................ 13 

4 Sampling and Analyses of Neutron-Activated Metal and Primary Resin Low-Level Wastes .......... 15 

4.1 Sampling. ............................................................ 15 

4.1.1 Neutron-Activated Metal Wastes ......................................... 15 

4.1.2 Primary Demineralization Resin Wastes ................................... 18 

4.2 Summary of Radioanalytical Procedure Development for Additional Radionuclides ........... 18 

5 Concentrations of Additional Long-Lived Radionuclides in Low-Level Wastes ................... 21 

5.1 Neutron-Activated Metal W astes ............................................. 21 

*5.2 Primary Demineralization Resin Wastes ......................................... 26 

6 Impact on Low-Level Waste Classification ............................................. 35 

7 Determination and Review of Activity Scaling Factors ................................... 39 

7.1 Determination of Activity Scaling Factors for Additional Radionuclides .................. 39 

7.1.1 Neutron-Activated Metal Radionuclides ...................................... 39 

7.1.1.1 M olybdenum -93 .............................................. 39 

7.1.1.2 Niobium -93m ............................................... 43 

7.1.1.3 Silver-108m ................................................. 43 

v NUREG/CR-6567
Contents 

7.1.1.4 Cadm ium -113m .............................................. 43 

7.1.1.5 Tin-121m .................................................. 43 

7.1.2 Chlorine-36 ....................................................... 49 

7.1.3 B eryllium - .10 ...................................................... 49 

7.2 Review of Scaling Factors for 1

4

C, 99Tc, and129 I for Low-Level Waste ................... 49 

7.2.1 Recent Mass Spectrometric Measurements of 1291 and 99Tc in Low-Level Wastes ........ 53 

7.2.1.1 Sam pling ................................................... 53 

7.2.1.2 Radiochemical and Mass Spectrometric Analyses for 1291 and 99Tc in Low-Level 

W astes . ....... .. .. . ... .. .. .. .... .. .. .. . . . .. .. ... .. ... ..... 54 

7.2.1.3 Results of Mass Spectrometric Analyses of 12'I and 99Tc in Low-Level Wastes ... 54 

7.2.1.4 Correlation of "2'I and 99Tc with Gamma-Emitting Radionuclides ............ 55 

7.2.1.5 Comparison of PNNL and Industry Scaling Factors ...................... 55 

7.2.1.6 Conclusions Regarding 1291 and 99Tc Scaling Factors ..................... 61 

7.2.2 Carbon-14 Scaling Factors in Low-Level Wastes and Activated Metal .............. 61 

7.3 Review of Activity Ratios of 63Ni and 59Ni in Low-Level Wastes ........................ 63 

8 Summary, Conclusions, and Recommendations .......................................... 67 

9 R eferences . .. .... ... .. . ......... .. . . .. .. ... ........ .. .. . .. .. .. .. .. ... .. ... . 71 

Appendix A: Radiochemical Procedures for Additional Long-Lived Radionuclides .................. 75 

Appendix B: Environmental Considerations of Additional Long-Lived Radionuclides .............. .. 81

NUREG/CR-6567

vi
List of Figures

3.1 1"B (n,p) Cross Section as a Function of Neutron Energy ...................... 

5.1 Change in Radionuclide Concentration with Time in 304 SS from Duane Arnold Spent 

Control R od B lade ................................................ 

5.2 Change in Radionuclide Concentration with Time in Point Beach BPRA Inconel Springs 

5.3 Change in Radionuclide Concentration with Time in Point Beach RCCA Ag-Cd-In Alloy 

5.4 Change in Radionuclide Concentration with Time in Crystal River RWCU Resin ......

5.5 Change in Radionuclide Concentration with Time in WNP-2 EDRFDR+RWCU Resins . 

5.6 Decrease in Concentrations of Long-Lived Radionuclides in PWR Spent Ion Exchange 

R esins (RW CU ) ................................................. 

7.1 Correlation Between Mo-93 and Co-60 in Low-Level Wastes From Nuclear 

Pow er Stations .................................................. 

7.2 Correlation Between Nb-93m and Co-60 in Low-Level Wastes From Nuclear 

Pow er Stations .................................................. 

7.3 Correlation of 1-129 with Cs-137 (a) and 1-129 with Co-60 (b) in Commercial Nuclear 

Power Station Low-Level W astes ...................................... 

7.4 Correlation of Tc-99 with Co-60 (a) and Tc-99 with Cs-137 (b) in Commercial Nuclear 

Power Station Low-Level W astes ...................................... 

7.5 Comparison of Geometric Mean Values for the Battelle and Industry Activity Scaling 

Factors for (a) I-129/Cs-137 and (b) Tc-99/Cs-137 ...........................

. . . . . . . . . 12 

......... 24 

......... 27 

......... 28

29

.. ........ 32 

... ....... 33 

.......... 42 

.......... 46 

.......... 57 

.. . ... .... 59 

.......... 60

visi NUREG/CR-6567
List of Tables 

2.1 Summary of Typical Hardware Characteristics ........................................ 4 

2.2 Summary of Typical Spent Fuel Disassembly Hardware Characteristics ....................... 5 

2.3 Summary of Typical Composition Ranges for Metals Used in Reactor Internal Hardware ........... 7 

2.4 Long-Lived Neutron Activation Products in Structural Materials in Nuclear 

Pow er System s .............................................. .............. 9 

4.1 Neutron-Activated Metal Waste Samples for Radiological Characterization ................... 16 

4.2 Spent Ion Exchange Resin Samples for Radiological Characterization ........................ 17 

5.1 Long-Lived Neutron Activation Products in Irradiated Stainless Steel Low-Level Waste 

from Nuclear Power Stations ................................................... 22 

5.2 Long-Lived Neutron Activation Products in Irradiated Metal Low-Level Waste from 

Nuclear Pow er Stations ...................................................... 23 

5.3 Beryllium-l10 and 6

"Co Concentrations in Selected Boron-Containing Low-Level Waste 

Sam ples . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 

5.4 Long-Lived Neutron Activation Products in Spent Ion Exchange Resins from Nuclear 

Pow er Stations . ............................................................ 30 

6.1 Dose Conversion Factors for IOCFR61 and Additional Long-Lived Radionuclides 

M easured in Low-Level W aste ................................................. 36 

7.1 93Mo/60Co Scaling Factors for Low-Level Wastes ....................................... 40 

7.2 93mNb/60Co Scaling Factors for Low-Level Waste ..................................... 44 

7.3 Silver-108m and 6

°Co Concentrations in Selected Low-Level Waste Samples .................. 47 

7.4 Cadmium-1 13m and 60

Co Concentrations in Selected Low-Level Waste Samples ............... 48 

7.5 Tin-121m and 60

Co Concentrations in Selected Low-Level Waste Samples .................... 50 

7.6 Chlorine-36 and 60

Co Concentrations in Selected Low-Level Waste Samples .................. 51 

7.7 Beryllium- 10 and 6

°Co Concentrations in Selected Low-Level Waste Samples ................. 52 

7.8 Radionuclide Concentrations and Activity Scaling Factors for Low-Level Waste from 

Commercial Nuclear Power Stations .............................................. 56 

7.9 14

C/ 60Co Radwaste Scaling Factors ............................................... 62

NUREG/CR-6567 viii
7.10 Comparison of Calculated 59

Ni/ 63Ni Activity Ratios in Neutron Irradiated Low-Level 

W aste M aterial .......................................................... 65 

8.1 Summary of Measured Activity Scaling Factors for Additional Radionuclides ................. 66 

B. 1 Estimated Range of Distribution Coefficient (Kd) Values for Elements Having 

Long-Lived Radioiostopes . ................................................... 77

ix NUREG/CR-6567
Executive Summary

The technical requirements for land disposal of low

level radioactive waste ( LLW), as specified in 10 

CFR Part 61, necessitates that specific waste 

classification and characterization be performed on 

LLW before disposal. Waste classification involves 

the quantification of specified radionuclides in the 

waste materials to comply with disposal 

requirements and disposal site performance 

objectives. With the aging of commercial nuclear 

power stations, increasing quantities of neutron

activated metal LLW are being generated as 

replacement of hardware components within the 

reactor pressure vessel becomes necessary. In 

addition, relatively large amounts of spent ion 

exchange resins used for cleanup of reactor coolant 

water and other aqueous waste streams are being 

generated for disposal. 

The research work conducted in this project has 

addressed key issues associated with the 

characterization and disposal of all long-lived 

radionuclides of potential concern in neutron

activated metal wastes and other related waste 

streams. The objectives of this research were: 1) to 

identify all long-lived (half-lives greater than five 

years) activation products of potential concern (in 

addition to those specifically listed in 10 CFR Part 

61) that may be present in various types of LLW in 

significant amounts, 2) to develop new or revised 

radioanalytical procedures for the measurement of 

the additional long-lived radionuclides of interest, 3) 

to enhance NRC's understanding of the distribution 

and projected quantities of additional long-lived 

radionuclides in neutron-activated metal wastes and 

other related waste streams from commercial 

nuclear power stations, 4) to determine appropriate 

activity scaling factors for the difficult-to-measure 

radionuclides for use when the sample results are 

near or below the lower-limit-of-detection, and 5) to 

provide an assessment of the radiological and 

environmental impact of those additional 

radionuclides not specified in 10 CFR Part 61. 

Since the enactment of 10 CFR Part 61, other long

lived radionuclides have been identified (e.g., "

0Be, 

36ci, 93 Mo, lo 8 mAg, 1 3

m Cd, and 121

m Sn) which are not 

routinely measured, but which potentially may be of 

concern in certain types of LLW. These 

radionuclides are primarily associated with neutronactivated metal wastes (e.g., spent control rod 

assemblies and reactor internal components) and 

other types of LLW (e.g., spent ion exchange 

cleanup resins) being generated by the nuclear 

power industry for disposal at shallow-land LLW 

burial grounds. 

Of the group of additional radionuclides measured 

in this project, "0

Be. 361, and l°smAg appear to have 

the greatest potential impact with respect to 10 CFR 

Part 61. Although none of these radionuclides were 

present in the LLW examined in this study in 

quantities high enough to seriously impact the 

current LLW classification and shallow-land 

disposal criteria specified in 10 CFR Part 61, 

further radiological and environmental assessments 

may want to be considered to determine if they 

should be amended to 10 CFR Part 61. 

Beryllium-10 is of special interest because of its 

potential to be produced in relatively large 

quantities in nuclear power station LLW, originating 

in spent BWR and PWR control rods and in spent 

primary coolant demineralization resins. The '

0Be 

is produced by neutron activation of '

0B present as 

neutron absorbing materials, boron carbide and 

borosilicate glass in control rods, and as soluble 

borate added to PWR reactor coolant. The 

extremely long half-life of "Be (1.6E6 years), 

together with a relatively high ingestion dose 

conversion factor and a relatively low Kd value for 

adsorption onto sandy types of soil, create the 

potential for long-term migration and offsite 

radiation exposure of this radionuclide from LLW 

disposal facilities. This study has shown that *'Be 

could be the third most abundant radionuclide in 

spent control rods and in some PWR spent primary 

resins after a decay time of about 600 years. 

Therefore, it is recommended that further studies be 

conducted to: 1) provide an accurate assessment of 

the total quantities of '°Be contained in LLW from 

these sources, 2) determine the 1

"Be leaching 

characteristics from LLW, 3) determine the 

migration behavior and environmental pathways of 

'

0Be released from LLW disposal facilities, and 4) 

provide performance assessment modelers with the 

necessary radiological/geochernical information to 

evaluate the potential impacts from disposal of this 

radionuclide in LLW.

xi NUREG/CR-6567
Silver-108m is also of interest because of its 

production in large quantities in PWR spent RCCA 

control rods which contain 80% silver. Silver

108m has a 130 year half-life and a relatively high 

ingestion dose conversion factor due to its multiple 

high-energy gamma-ray emissions. Silver-108m 

will easily be the most abundant radionuclide in 

spent RCCA control rods for many hundreds of 

years. It is therefore of some concern from a waste 

disposal perspective. Fortunately, the stable 

metallic waste form and the expected high soil Kd 

values should minimize the leaching rate and offsite 

transport of this radionuclide during water 

infiltration at a LLW disposal facility. 

Nevertheless, it is recommended that further studies 

be conducted for losmAg, as described in the above 

paragraph for "°Be, to better understand the 

radiological, environmental, and health impacts 

associated with the disposal of neutron-activated 

metal components containing lsmAg. 

Chlorine-36 is the other very long-lived 

radionuclide measured in this study that deserves 

further assessment from a LLW disposal standpoint. 

Its very long half-life (3.01E5 years), relatively 

high ingestion dose conversion factor, and high 

environmental mobility combine to create the 

potential for long-term migration from a LLW 

disposal facility. Fortunately, the concentrations of 36

C1 in LLW from commercial nuclear power 

stations are quite low. Chlorine-36 is produced 

almost exclusively in nuclear power plants by 

neutron activation of soluble chloride impurities 

present in the reactor coolant water, and is then 

concentrated on primary demineralization resins 

used for reactor coolant cleanup. Because chloride 

ions accelerate the corrosion of stainless steel in the 

reactor primary loop, great care is taken to 

minimize the chloride concentrations in reactor 

coolant water, and typical concentrations are in the 

low ppb range. This is sufficient to produce about 

0.0001 to 0.002 uCilg of 36

C1 in spent primary 

resins. This equates to an estimated 0.2 to 5 Ci/y 

of 36C1 in primary resin LLW from all PWR nuclear 

power stations in the U. S. As discussed in Section 

6.1 of this report, LLW disposal site performance 

assessment modeling, using a source term 

containing only 0.1 Ci of 36

C1 was sufficient to 

predict that this radionuclide would be the major

contributor to the offsite radiation dose for the 

contaminated well/drinking water exposure scenario. 

Because of this potential for radiation exposure, it is 

recommended that further studies similar to those 

described in the above paragraph for "

0Be be 

conducted to obtain a better understanding of the 

radiological and environmental impacts from 

disposal of this radionuclide in LLW. 

The other additional radionuclides measured in this 

study, 93Mo, ll13mCd, and 121mSn, either have half

lives that are too short to be of consequence in 

LLW disposal, or in the case of 93

Mo, the 

concentrations do not appear to be high enough to 

contribute significantly to the potential radiation 

exposure. 

Activity scaling factors have been developed for all 

of the additional radionuclides measured in this 

study. The use of activity scaling factors for 

estimating the concentration of a radionuclide in 

LLW is permitted by 10 CFR Part 61.55, and this 

method is desirable for radionuclides such as those 

listed above that are difficult to measure directly in 

LLW materials. Consistent scaling factors for a 

wide variety of LLW materials were determined for 

93

Mo and 9SmNb relative to 60 Co over a 

concentration range of four orders of magnitude. 

Generic scaling factors for the other long-lived 

radionuclides measured in this study showed much 

greater variability due to the unique composition of 

some of the materials. However, in some cases 

useful scaling factors were obtained when similiar 

LLW materials were considered as a group. 

Recent measurements and assessments of 1291 and 

"99Tc in LLW from commercial nuclear power 

stations have shown that the true concentrations of 

these radionuclides are several orders of magnitude 

lower than previously reported values. Very 

sensitive mass spectrometric analytical methods 

were utilized to determine the concentrations and 

activity scaling factors for 129I and 99

Tc in a wide 

variety of LLW samples from a number of PWR 

and BWR stations in the U. S. (Robertson, et al., 

1997). These measurements provided 

concentrations and associated activity scaling factors 

(relative to '

37

Cs and6 6Co) for these radionuclides 

that were two to four orders of magnitude lower 

than the very conservative values published in

NUREG/CR-6567 xii
several nuclear power industry data bases. 

Geometric mean values of the '

291/137

Cs scaling 

factor for LLW from PWR and BWR stations 

measured in this study were 6.48 x 10' and 4.25 x 

10', respectively. The geometric means for 99

Tc/ 13 7CS in LLW from PWR and BWR stations 

were 7.37 x 107 and 3.21 x 10•, respectively. The 

correlations between 99

Tc and 6

0 Co were 

considerably better than the correlations between 

99

Tc and 137

Cs, and the geometric mean values of 

the 99

Tc/6 0Co ratios for PWR and BWR stations 

were 1.70 x 10.6 and 1.18 x 106, respectively. 

Because performance assessment modeling studies 

of LLW disposal facilities have indicated that 1291 

and 99Tc (as well as " C) will be the limiting 

radionuclides with regard to the total quantities of 

LLW permitted at a site, it is important to 

determine as accurately as possible the actual 

quantities of these radionuclides being disposed in 

LLW. The mass spectrometric measurements 

performed during this study have clearly shown that 

when the true concentrations and activity scaling 

factors are used for performance assessment 

modeling, the "I and 99Tc no longer become 

limiting radionuclides for LLW inventories at 

disposal facilities. Disposal facilities will, therefore, 

be able to accommodate larger quantities of LLW at 

each facility, thereby minimizing the expense of 

developing additional disposal sites. 

An assessment of the 63

Ni and 59

Ni production and 

activity scaling factors in LLW was conducted to 

evaluate the accuracy of past measurements of these 

radionuclides. This assessment indicated that early 

reported values of the 59

Ni/ 3Ni activity ratio were 

too high by up to factors of 5 to 10. More recent 

analyses and assessments indicate that this ratio 

should range between about 0.006 to 0.012, 

depending upon the neutron flux and the irradiation 

time of the exposed materials. In any event, this 

activity scaling factor does not appear to be a 

critical parameter with respect to 10 CFR Part 61 

waste classification criteria because the 63

Ni, rather 

than 59

Ni, will always be the controlling 

radionuclide of the two nickel radionuclides in 

LLW classification. 

This study required that a large number of 

radiochemical separation and purification 

procedures, as well as special radiation detection 

methods be developed to measure the additional

group of radionuclides addressed in this study. 

Appendix A of this report contains the 

radiochemical procedures that were developed 

during this study. Detailed procedures are presented 

for '

0Be, 36C1, 55 Fe, 59 Ni, 6' Ni, 90 Sr,93m Nb,9 4 Nb, 93

Mo, 9 9

Tc, 1

0 9

Cd, 113m Cd, 121 Sn, as well as Pu, Am, 

and Cm isotopes. Additional radioanalytical 

information used to help develop the radiochemical 

procedures given in Appendix A are published in: 

Thomas, C. W., Thomas, V. W., and Robertson, D. 

E., "Radioanalytical Technology for 10 CFR Part 

61 and Other Selected Radionuclides", NUREG/CR

6230, U. S. Nuclear Regulatory Commission, 

Washington, D. C., March, 1996. 

This study has provided a more comprehensive 

assessment of the spectrum of long-lived 

radionuclides that are present in the various types of 

LLW being generated at commercial nuclear power 

stations. This work will help insure that the health 

and safety aspects of LLW disposal are met in a 

judicious and compliant manner.

Xiii NUREG/CR-6567
Foreword

Increasing amounts of neutron activated metal low

level radioactive wastes (LLW) are being generated 

at commercial nuclear power stations. Most of 

these metals are being disposed of as LLW in 

commercial shallow-land burial facilities. These 

metals come from a variety of irradiated hardware 

from the internals of reactor pressure vessels, and 

this hardware contains a diverse spectrum of 

radionuclides depending upon the alloy composition. 

To insure that no unforeseen radiological hazards 

are associated with the disposal of these materials, a 

comprehensive examination of neutron-activated 

metal waste was conducted. In this project a 

number of long-lived radionuclides heretofore not 

measured in neutron activated metal wastes have 

been characterized to determine their occurrence 

and impacts with respect to 10CFR Part 61 

requirements.

These radionuclides include '

0

Be, 3 6C1, 9 3MO, 93mNb, 

losmAg, 1 13

m Cd, and 121

m Sn. In addition, this same 

group of radionuclides, plus 99

Tc and I291 and 

others commonly used for developing activity 

scaling factors, have also been measured in spent 

demineralization resins from commercial nuclear 

power stations. Where appropriate, activity scaling 

factors for these radionuclides have been developed 

for specific LLW types to facilitate the rapid 

estimation of the concentrations of these difficult

to-measure radionuclides. This comprehensive 

examination and assessment of these additional 

radionuclides measured in activated metal and 

spent ion exchange resin LLW has shown that no 

significant changes to 1OCFR Part 61 requirements 

would be necessary based on their observed 

radiations, concentrations, inventories, and 

environmental behaviors.

XV NUREG/CR-6567
Acknowledgements

The authors wish to express their appreciation to Dr. Phillip Reed (Project Manager) of the Radiation Protection, 

Environmental Risk, and Waste Management Branch, Division of Risk Analysis and Applications, Office of 

Nuclear Regulatory Research, U.S. Nuclear Regulatory Commission for helpful guidance and suggestions in 

conducting this work.

x'vii NUREG/CR-6567
1 Introduction

Since the enactment of the U. S. Nuclear Regulatory 

Commission (NRC) rule on "Licensing Requirements 

for Land Disposal of Radioactive Waste", 10 CFR 

Part 61 (Federal Register 1988), a great deal of 

experience and information has been developed on 

the radiological and physical properties and 

quantities of low-level waste (LLW) being generated 

and disposed by commercial nuclear power stations. 

Despite this abundance of information, there are a 

number of areas where additional data are needed to 

clarify ambiguities, present more accurate analytical 

results, and provide a more comprehensive 

assessment of the requirements of 

10 CFR Part 61. This project was initiated to 

conduct empirical measurements of long-lived 

radionuclides, particularly those not specified in 10 

CFR Part 61, but which could be present in some 

LLW types, including neutron activated metals, in 

significant concentrations. 

The reasons for conducting these studies include: 1) 

more neutron-activated metals are being disposed of 

as LLW, 2) the activated metal LLW is frequently 

not being adequately analyzed for all long-lived 

radionuclides, and 3) activated metal wastes are not 

included in performance assessment modeling of 

LLW disposal facilities. 

The relationship of these additional radionuclides 

with other 10 CFR Part 61 radionuclides, particularly 

those gamma-emitting radionuclides used for 

developing activity scaling factors, would help 

determine if further radiological, dosimetric, and 

environmental assessments are warranted. This work 

includes: 1) a literature assessment of the additional 

radionuclides [this task has been published as 

NUREG/CR-6230 (Thomas, Thomas and Robertson, 

1996)], 2) the development of radiochemical 

procedures for their measurement in LLW, 3) 

determination of their concentrations (as well as the 

specified 

10 CFR Part 61 radionuclides) in a variety of 

neutron-activated metal wastes and primary coolant 

cleanup resins, 4) an assessment of the activity 

scaling factors for the additional radionuclides and 

several 10 CFR Part 61 radionuclides which need 

clarification, and 5) an assessment of the relative 

concentrations of 59

Ni and 63

Ni in activated metal

waste to provide more accurate estimates of the 

activity ratio (scaling factor) for these radionuclides, 

and 6) an assessment of the impact of the additional 

radionuclides on waste classification and shallow

land disposal of LLW. 

1.1 Low-Level Waste 

Classification and 

Characterization 

Requirements of 

10 CFR Part 61 

The technical requirements for land disposal of LLW, 

as specified in 10 CFR Part 61, necessitates that 

specific waste classification and characterization be 

performed on LLW before disposal. Waste 

classification involves the quantification of specified 

radionuclides in the waste materials to comply with 

disposal requirements and disposal site performance 

objectives. Since the enactment of 10 CFR Part 61, 

other long-lived radionuclides have been identified 

(e.g., ' 0Be, "

6

C1, 93Mo, l°SmAg, 13m(Cd, and 121mSn) 

which are not routinely measured, but which 

potentially may be of concern in certain types of 

LLW. These radionuclides are primarily associated 

with neutron-activated metal wastes (e.g., spent 

control rod assemblies and reactor internal 

components) and perhaps other types of LLW which 

the nuclear power industry is generating in increasing 

amounts for disposal at shallow-land LLW burial 

grounds. 

A more comprehensive radionuclide analyses of the 

various types of neutron-activated metal wastes is 

therefore warranted to insure that these radionuclides 

do not pose a problem to public health and safety. In 

addition, the environmental fate of these long-lived 

radionuclides once they are disposed at LLW 

disposal facilities is of further concern. In order to 

predict the migration of these radionuclides from 

disposal sites, it is necessary to determine their 

leaching rates from the various neutron-activated 

metal matrices by simulated or actual waste trench 

solutions and/or groundwaters. The chemical forms 

of the resulting soluble radionuclides species are also 

important for predicting their movement through 

soils.

I NUREG/CR-6567
The concentration of a radionuclide in wastematerials 

is normally determined by direct measurements or 

indirect methods, such as the use of activity scaling 

factors which relate the inferred concentration of a 

difficult-to-measure radionuclide to another that is 

easily measured by gamma-ray spectrometry. For a 

number of radionuclides (e.g., `

4C, 99

Tc, and 1291), the 

activity scaling factors used in the nuclear power 

industry have been overly conservative, since the 

factors were based primarily on "less-than" values 

corresponding to the lower-limit-of-detection of the 

analytical methods used by the commercial analytical 

service laboratories performing the analyses. This 

methodology has not generally influenced the 

classification status of the waste, but does artificially 

increase the inventories of these radionuclides within 

waste containers, and subsequently the total 

inventories of these isotopes at LLW disposal sites. 

Recent site performance assessment modeling of 

proposed LLW disposal sites has indicated that the 

total disposal site inventories of these radionuclides 

are often the controlling factors for the total 

quantities of radioactive waste permitted in a LLW 

shallow-land burial facility (Vance 1988). Thus, 

overly conservative estimates of the concentrations of 

these isotopes in LLW could lead to premature 

closure of a disposal facility. It is therefore important 

that waste generators utilize the most accurate scaling 

factors possible, so that reliable estimates of the total

quantities of these nuclides can be made. 

1.2 Project Scope 

The research work conducted in this project has 

addressed key issues associated with the 

characterization and disposal of all long-lived 

radionuclides of potential concern in neutron

activated metal wastes and other related waste 

streams. The objectives of this research were: 1) to 

identify all long-lived (half-lives greater than five 

years) activation products of potential concern (in 

addition to those specifically listed in 10 CFR Part 

61) that may be present in various types of LLW in 

significant amounts, 2) to develop new or revised 

radioanalytical procedures for the measurement of 

the additional long-lived radionuclides of interest, 3) 

to enhance NRC's understanding of the distribution 

and projected quantities of additional long-lived 

radionuclides in neutron-activated metal wastes and 

other related waste streams from commercial nuclear 

power stations, 4) to determine appropriate activity 

scaling factors for the difficult-to-measure 

radionuclides for use when the sample results are 

near or below the lower-limit-of-detection, and 5) to 

resolve discrepancies between calculated versus 

measured concentrations of 59

Ni and 63

Ni in neutron

activated metal LLW.

NUREG/CR-6567 2
2 Radionuclide Production and Distribution in LLW Generated at Nuclear 

Power Stations

Radionuclides contained in nuclear power station 

LLW generally originate from four main sources: 1) 

leakage of fission products from defective fuel or 

from tramp uranium on the fuel assembly surfaces, 2) 

neutron activation of metal components contained 

within the reactor pressure vessel, 3) neutron 

activation of corrosion products leached from 

primary loop cdrhponents and transported by the 

reactor coolant to the high neutron flux region of the 

reactor, and 4) neutron activation of impurities in the 

reactor coolant and activation of the coolant itself 

These radionuclides are then redistributed throughout 

the nuclear power station during demineralization of 

the primary coolant or the treatment of other liquid 

waste streams, by spillage of radioactive liquids, and 

by spreading of solid radioactive corrosion products 

during station maintenance and component 

replacement. 

The focus of this project has been to examine the 

radionuclides produced primarily in neutron

activated metal waste components removed from the 

reactor pressure vessel (e.g., spent control rod 

assemblies, LPRM hardware, etc.) and performing an 

analysis/assessment of the long-lived radionuclides 

not normally measured in these waste components. 

Of particular interest is the characterization of those 

long-lived activation products not specified in 10 

CFR Part 61, but which may be present in some types 

of these wastes. Several samples of primary coolant 

demineralization resins were also included in this 

characterization since these resins concentrate the 

fission products and activation products which 

undergo corrosion and dissolution by the primary 

coolant.

2.1 Neutron Activation of 

Materials Inside Reactor 

Pressure Vessels 

This brief discussion pertains primarily to the types 

of neutron-activated metal wastes that originate 

within the neutron flux region within the reactor 

pressure vessel and which must be radiologically 

characterized before disposal as LLW. These waste 

types generally consist of expendable or defective 

reactor components which are replaced during 

normal reactor operations, and include such items as 

spent control rod blades and assemblies, spent 

instrument strings, thimble plug guide tubes, poison 

curtains, and flow channels. Also, some components 

associated with spent fuel disassembly hardware 

generated during fuel rod consolidation projects at 

nuclear power stations can be treated as LLW. Now 

that a number of older nuclear power plants have 

reached decommissioning status, a wide variety of 

neutron-activated reactor decommissioning wastes 

are beginning to be generated, including the reactor 

pressure vessel itself. 

A summary of the types, sizes, weights, surface 

areas, volumes, and materials of construction have 

previously been summarized and are provided in 

Tables 2.1 and 2.2 (Bedore, Levin, and Tuite 1987). 

Much of this radioactive material is close to being 

Greater-Than-Class-C (GTCC) waste because of its 

relatively high concentrations of long-lived neutron 

activation products, and careful radiological 

characterization of these waste components must be 

conducted prior to disposal to insure that the Class-C 

limit is not exceeded.

3 NUREG/CR-6567
w 

09

Source 

Assemblies 

Thimble Plug 

Guide Tubes 

BWR 

Components 

Control Rod 

Blades 

Poison Curtains 

Flow Channels 

Instrument 

String

Overall 

Length 

(cm) 

30

400 

400 

400

variable

440 

400 

420 

variable

Configuration 

Square array of 

plugs 

Square array of 

plugs and rods

Square array of 

plugs and rods

Square array of 

plugs and rods

Tubing 1 cm diam.

Cruciform 25x25

___________________________________ 1* Approx.

Weight (kg) 

5.5

Approx. Volume 

(cc)

570 304 SS

13.6 to 20.4 3700 to 5100

74.8

13.6 to 16

N/A

130

8500

4300

N/A

18000 (0.018m')

Flat plate 21.0x.2 16.4 1700

Rectangular box 14x14

Tubing 1.9 cm diam. LN/A N/A

Construction Materials of 

Areal Cross Section 

(cmxcm) 

15X15

304 SS Boron Carbide 

Borated SS 

Zircaloy 

304 SS 

304 SS

PWR 

Components 

Thimble Plug 

Assemblies 

Burnable Poison 

Assemblies 

Control Rod 

Assemblies

Table 2.1 Summary of Typical Hardware Characteristics (from Bedore, Levin and Tuite, 1987)

304 SS Inconel 

304 SS 

Inconel 

Borosilicate 

304 SS 

Inconel 

Ag/I[nCd 

304 SS 

Inconel 

Borosilicate and 

Source Material 

304 SS

5, 

0 

0 

U 

0

570

4800

30A
Table 2.2 Summary of Typical Spent Fuel Disassembly Hardware Characteristics 

(from Bedore, Levin and Tuite, 1987)

Overall Areal Cross 

PWR Length Section Weight Approx. Volume Materials of 

Components (cm) Configuration (cmxcm) (kg) (cc) Construction 

End Fittings 6 to 30 21X21 11.3 to 1400 to 2250 304 SS 

Square 18.1 

Guide Tubes variable Tubing 1 cm diam. 7.5 to 21.2 N/A Zircaloy 

Spacer Grids 7 Square 21.21 4.0 to 12.6 525 to 1660 Inconel 

Zircaloy 

Misc. 0.7 to 11.4 N/A 304SS 

(Springs, etc.) Inconel 

BWR 

Components 

End Fittings 17 Square 15x15 5.4 to 15.2 680 to 304 SS 

1900 

Guide Tubes variable Tubing 1 cm diam. 1.7 to 13.2 N/A Zircaloy

Spacer Grids 4 to 6 Square 15x15 1.1 to 1.6

1 _______ 4 I I N/A Inconel

Misc. 

(Springs, etcA

0.2 to 1.1

______________________ j __________________ .1 ________________ J __________________________ -

Inconel 

Zircaloy

tzj 

C) 

01

0 P

0 

r_ 

0,

145 to 210

N/A Inconel
Radionuclide Production and Distribution

2.1.1 Composition of Reactor 

Internal Component 

Construction Materials 

The radionuclide composition and quantities 

generated in neutron-activated metal LLW are 

directly dependent upon the major and trace element 

composition of the materials of construction, the 

neutron fluence, and the decay time after removal 

from the reactor. The types of construction materials 

predominantly used in reactor pressure vessel internal 

components and primary coolant loop materials are 

generally dominated by stainless steel, with smaller 

amounts of Inconel and Zircaloy alloys being used in 

specialized components (Roberts 1981). The 

elemental compositions of the two types of stainless 

steel (Types 304 and 316), the five types of Inconel 

(Types 600, 700, 710, 718, and 750), and the two 

types of Zircaloy (Types 2 and 4), the most 

commonly used alloys in nuclear reactor primary 

systems are given in Table 2.3 for those elements 

which produce long-lived neutron activation 

products. Also included is the main type of neutron

absorbing metal alloy (Ag-In-Cd alloy) used in PWR 

Rod Cluster Control Assemblies (RCCA) control 

rods. Table 2.3 provides the typical range of 

concentrations of the parent elements of long-lived 

neutron activation products that are produced in the 

types of metal components listed in Tables 2.1 and 

2.2. 

Although iron and chromium are major constituents 

in these structural materials, these elements do not 

produce long-lived neutron activation products, other 

than "Fe, a low-energy x-ray emitter with a 2.73

year half-life. Nickel is widely used in stainless steel 

and Inconel alloys, and produces two long-lived 

neutron activation products, 63

Ni (100 year half-life) 

and "Ni (76,000 year half-life). 

Great emphasis has been placed on reducing the 

cobalt content of metal alloys used in nuclear power 

systems because of the generation of 6

1Co (5.27 year 

half-life) and its accompanying high energy gamma 

radiation. Therefore, cobalt levels in the materials of

construction have been in the ppm concentration 

range, except for several Inconel alloys (Types 718 

and 750) which contain 1.0% cobalt to improve 

hardness. 

Niobium is of concern because of the potential for 

production of 94Nb (2.0E4 year half-life) in alloys 

containing added niobium. Niobium is present as a 

trace impurity in ppm concentrations in stainless 

steels, Zircaloy, and in some Inconels, but it is an 

additive to Inconel Types 710, 718, and 750, ranging 

from 1% to 5%. 

Molybdenum is also a minor constituent (2-3%) of 

stainless steel Type 316 and Inconel Types 700 (3%), 

710 (2.7%), and 718 (3%). Molybdenum is of 

concern because it is the parent element of both 9Tc 

(2.13E5 year half-life) and "Mo (3500 year half

life), which are environmentally mobile 

radionuclides. 

Tin is a minor additive to Zircaloy Types 2 and 4, 

being present at 1.2 to 3%, and is also present at 

0.3% in Inconel Type 750. Tin can produce "

2lmSn 

(55 year half-life) in these irradiated materials, 

although the production efficiency is quite low. 

Silver, indium, and cadmium are the major neutron

absorbing constituents of PWR Rod Cluster Control 

Assemblies (RCCA), being present at concentrations 

of 80, 15, and 5 %, respectively. The silver produces 

the long-lived l°smAg (130 year half-life) activation 

product, and the cadmium produces 113mCd (14.1 year 

half-life). There are no long-lived indium activation 

products. 

Chloride impurities in reactor structural materials are 

purposely kept to an absolute minimum to improve 

corrosion resistance of the metals. However, it is 

impossible to eliminate all traces of chloride from 

these alloys, and stainless steels reportedly have 

chloride contents ranging from <50 to 130 ppm 

(Evans, et al. 1984). There is little available 

information on the chloride contents of the other 

alloys commonly used in nuclear power

NUREG/CR-6567 6
Table 2.3 Summary of Typical Composition Ranges For Metals Used in Reactor Internal Hardware 

Stainless Steel Inconels 

Ag/In/Cd 

Element 304 316 600 700 710 718 750 Zircaloy-2,4 Alloy 

Fe (%) 68-76 64-65 6-14 7-18 18 18 7 1.5-2.2 <0.002 

Cr(%) 18-20 16-18 23 15 --- --- --- 0.1 ... 

Ni (%) 8.0-11 10-14 58-76 52-72 57 53 71 0.02-0.5 0.07-0.12 

Co (ppm) 230-2600 1300-1600 400-700 400-700 --- 10,000 10,000 10 0.006 

N (ppm) 400-1000 300-400 1000-2000 1000-2000 ... --- --- 60-100 ... 

Nb (ppm) 5-300 10-65 10-400 10-400 49,600 51,000 9,500 100-150 <0.0002 

Mo (ppm) 80-5500 20,000- 2000-3000 30,000 27,000 30,000 --- <200-800 0.0002

30,000 0.002 

Sn (ppm) <200 --- --- --- --- --- 3000 12,000- 0.0016

30,000 0.0020 

Ag (%) <0.0002 <0.0005 --- --- --- --- --- ... 80 

In --- -.-.-.-.-.-.-.-.-........ 15 

Cd ( ).... .............. . ...--- - 5 

C1 (ppm) <50-130 --- --- --- --- --- ---. 

Note: Adapted from Bedore, Levin and Tuite 1987; Roberts 1981; Evans, et al. 1984; Lepel, et al. 1995; Robertson, et al. 

1991)

0 

C,'

0 

(D 

0 

0 

5ý 

0~ 

0j
Radionuclide Production and Distribution

systems. The long-lived isotope 36

C1 (3.0 1E5 year) 

can be produced in neutron-irradiated metals in 

small, but perhaps significant amounts. 

The boron carbide used in BWR control rod blade 

assemblies and the borosilicate glass used in PWR 

Burnable Poison Rod Assemblies (BPRA) are used 

for neutron absorption for reactor control. The 1

°B 

isotope in these materials is the source of long-lived 

"

10

Be (I .6E6 year half-life) in the neutron poison 

materials, being produced by the reaction 1

1B (n,p) 

1

0

Be. 

2.1.2 Long-Lived Neutron 

Activation Products 

The long-lived neutron activation products produced 

in irradiated structural materials in nuclear power 

systems, together with their half-lives, primary 

production modes, and principal radiations are listed 

in Table 2.4. Included in the table are the 

radionuclides specifically listed in 10 CFR Part 61 

(except the TRU isotopes), plus a number of 

additional long-lived neutron activation products 

which are generated in a variety of reactor 

construction materials. Some of these additional 

radionuclides may be produced in activated metal 

and primary cleanup resin LLW in quantities that 

could require further environmental and dosimetric 

assessment with respect to their disposal in shallow

land burial facilities. For this reason it was deemed 

prudent to conduct empirical measurements of these 

radionuclides in a number of activated metal and 

primary cleanup resin LLW materials to provide a 

more reliable data base. These radionuclides are 

discussed individually in more detail in Section 3 of 

this report.

via two main mechanisms: 1) corrosion reactions 

between neutron-activated metal components within 

the reactor pressure vessel and the coolant water 

release soluble or particulate neutron activation 

products to the coolant, and 2) trace metals released 

from reactor construction materials outside of the 

neutron flux region of the reactor are carried to the 

reactor core in the cooling water where the metals are 

neutron-activated. 

The soluble radionuclide species and a fraction of the 

particulate radionuclides are removed from the 

primary coolant by two main mechanisms: 1) plate

out on the inner surfaces of the primary loop piping 

and other components, and 2) removal from the 

primary coolant by adsorption onto primary 

demineralization resins and by filtration onto primary 

filters used for cleanup of the primary coolant. These 

ion exchange resins and primary filters represent 

some of the most radioactive low-level waste types 

generated at nuclear power stations, and are the 

major contributors, along with neutron-activated 

metal waste, to the radionuclide inventory at low

level waste disposal facilities. Essentially the same 

radionuclides that are produced in the neutron

activated metal components are present in the spent 

primary resins and filters (see Table 2.4). Fission 

products are also released to the reactor coolant via 

recoil from the trace uranium contamination on fuel 

assemblies or escape of fission products through fuel 

cladding defects or failures. All of these 

radionuclides also enter a wide variety of other types 

of low-level waste, including sludges, dry active 

waste, secondary resins, and evaporator concentrates. 

The volumes, activities, and sources of these low

level waste types that are generated at commercial 

nuclear power stations have been reported in detail 

(Daloisio and Deltete 1988).

2.2 Neutron Activation of 

Corrosion Products Released 

to Reactor Coolant 

The reactor construction materials discussed in 

Section 2.1.1 represent the source-term for the long

lived radionuclides that are present in reactor coolant 

water. These radionuclides enter the primary coolant

NUREG/CR-6567 8
Table 2.4 Long-Lived Neutron Activation Products in Structural Materials in Nuclear Power Systems

Radionuclides Specified in 10CFR61

Principal 

Radionuclide Half-life (yr) Primary Production Mode(s) Radiations 

3

H 12.3 'Li(n,a) 3

H 18.6 keV 1 

14C 5730 14

N (n,p) 1

4

C 157 keV f3 

""_C(n,y) 14

C 

59Ni 7.6E4 5Ni(n,y) 59

Ni 6.93 keV x-ray 

63

Ni 100 62

Ni (n,y)63

Ni 66.9 keV • 

60Co 5.27 59

Co (n,y)°Co 318 keV P'; 

1173 keV and 1333 

keV y's 

9Sr-OY 29.1 235

U fission 546 keV and 2281 

keV j-s 

94Nb 2.0E4 93

Nb (n,Y) 94

Nb 473 keV P-; 

703 keV and 871 keV 

Y's 

9Tc 2.13E5 98

Mo (n,y) 'Mo-'Tc 292 keV P 235

U fission 

1291 1.57E7 235

U fission 150 keV P-; 

29.8 keV x-ray 

1

37Cs-137

Ba 30.17 235

U fission 514 keV P-; 

I_ 1 1 662 keV y

LA 

0'. 

--3

0 

0 

0 

0°
Table 2.4 (Continued) 

Additional Long-Lived Radionuclides Not Specified in 10CFR61

Principal 

Radionuclide Half-life (yr) Primary Production Mode(s) Radiations 

10Be 1.6E6 10

B (n,p)1 °Be 556 keV f" 

"•C1 3.01E5 35

C1 (n,y) 36C1 709 keV 1 

93

mmb 16.1 93

Nb (n,n') 93mNb 16.5 kev x-ray; 

30 kevy 

93

Mo 3500 92

Mo (n,y) 93

Mo 16.5 keV x-ray; 

30kev y 

l°smAg 130 1

07

Ag (n,y) l°smAg 434 keV, 614 keV, 

and 723 keV y;s 

113m Cd 14.1 12Cd (n,y) 113

m Cd 590 keV P* 

12lmSn 55 12 0

Sn (n,y) 12ImSn 354 keV P'; 26.3 keV 

x-ray

0' 

0 

0 

0

0 

0I

C
3 Long-Lived Radionuclides Not Considered in 10 CFR Part 61 in Low-Level 

Waste

A number of long-lived radionuclides, not 

specifically addressed in 10 CFR Part 61, are 

produced in neutron-activated metal components in 

significant concentrations. These radionuclides are 

generally beta emitters or low-energy photon emitters 

and are normally present in trace amounts in the 

usual types of low-level waste. However, in specific 

types uf low-level waste which are infrequently 

encountered in commercial nuclear power stations 

waste streams, these radionuclides can take on major 

proportions. Seven of these radionuclides are 

discussed below. 

3.1 Beryllium-10 

Beryllium- 10 is a very long-lived (1.6E6 year half

life) radionuclide which is a pure beta emitter (556 

keV max. B-). To the best of our knowledge, this 

radionuclide has never been measured in LLW from 

nuclear power stations prior to this study. Because of 

its very long half-life, it will, for all practical 

purposes, always remain with certain types of LLW. 

Therefore, it is important to empirically determine 

the concentrations of this radionuclide in the special 

types of waste in which it is generated. 

Beryllium- 10 can be produced by several nuclear 

reactions, including the following: 

9

Be (n, gamma) '°Be, 9

Be (d, p) '°Be, 1

0

B (n, p) '°Be, 

"B (n, d) '°Be, "IB (n, np) '

0Be, and 1

3

C (n, alpha) 

'°Be. The 9

Be (n, gamma) `

0Be reaction is probably 

of minimal importance because there are no 

significant sources of beryllium parent material 

present in components of nuclear power stations. 

The (n, d), the (n, np), and the (n, alpha) reactions 

have very high energy thresholds for these reactions 

and therefore will also be minimally important. 

For nuclear power station LLW, the "°B (n, p) "°Be 

reaction is probably the most important for producing 

'°Be because of its lack of an energy threshold for 

this reaction. The (n, p) cross section for this 

reaction is given in Figure 3.1. The figure shows this 

cross section to be several tens of millibams at 

thermal neutron energies and at neutron energies 

above about 1 meV. Therefore, this reaction does 

have high enough cross section values at appropriate

neutron energies for a fission neutron spectrum to 

produce significant amounts of '

0Be. Also, the parent 

isotope, '

0B, is present in spent BWR control rod 

blades as boron carbide (B4C), in spent PWR 

Burnable Poison Rod Assemblies as borosilicate 

glass, and is used as a chemical reactivity shim in 

PWR primary coolant as a soluble borate at 

concentrations of up to several thousand parts-per

million of boron. Therefore, "°Be can be produced 

from the `

0

B in spent control rod assemblies and in 

PWR primary coolant. Once the '

0

Be is produced in 

the primary coolant, it is capable of then entering into 

most of the LLW streams generated at nuclear power 

stations, e.g., primary demineralization resins, filters, 

sludges, evaporator bottoms, etc. 

3.2 Chlorine-36 

Chlorine-36 is a very long-lived (3.01 E5-year half

life) neutron activation product which primarily 

decays by emission of a 709 keV beta particle with 

no gamma radiations. Chlorine-36 is produced 

primarily by thermal neutron capture (43.6 barn cross 

section) by trace amounts of chloride impurities 

present in reactor construction materials and in 

reactor primary coolant via the reaction 35

C1 (n, 

gamma) 36

C1. The concentrations of Cl in these 

materials are kept as low as possible because of the 

deleterious corrosion effects on metals due to Cl ion. 

Stainless steel Type 304 has Cl concentrations 

ranging from less-than 50 to 130 ppm (see Table 

2.3), and reactor coolant Cl concentrations range 

from about0.6 to 7 ppb (Cline 1996). Because of 

these low Cl concentrations, the quantities of 36

CI 

produced in reactor materials and primary coolant are 

very low. Consequently, low-level wastes from 

nuclear power stations contain very low 

concentrations of 36C1, and it is not reported in waste 

classification manifests. 

A few measurements of 36

C1 have been previously 

made in reactor coolant and primary demineralization 

resins at several Swiss BWR and PWR reactors (Liu, 

et al. 1991). This study reported 36

C1/37

Cs activity 

scaling factors ranging from 1.4E-6 to 17.OE-6 for 

the coolant and resin samples

I1I NUREG/CR-6567
Long-Lived Radionuclides

8-10 (n,p) Cross Section

I-'1

L 

D 

b

14--' 1ID-4 1C-3 10-2 ID-1 UcP 11 ¶D' 1 1031' uý 10 1 D" 107 

El (EV) 

Figure 3.1 l°B (n,p) Cross Section as a Function of Neutron Energy

NUREG/CR-6567 12
Long-Lived Radionuclides

However, because of its very long half-life, its high 

degree of environmental mobility, and its relatively 

energetic beta particle emissions, 36

C1 is of concern 

since it will eventually migrate from a low-level 

waste disposal facility. Site performance assessment 

modeling has shown that less than Curie amounts of 

"

36

C1 disposed at a LLW disposal facility are enough 

for 36

C1 to become the most significant radiation dose 

contributor to the general public for the 

groundwater/domestic well transport-exposure 

scenario (Campbell 1995). 

3.3 Molybdenum-93 

Molybdenum-93 is a very long-lived (3500 year half

life) neutron activation product which is produced by 

thermal neutron capture by 92

Mo. Molybdenum is 

found in stainless steel at concentrations of about 

2600 + 1500 ppm, and Inconel contains about 30,000 

ppm of Mo. The 93

Mo decays through an 80% 

transition by electron capture to 13.6 year 93mNb. 

The 9

'"Nb decays with the emission of 16.5 keV Nb 

x-rays. Data from recent reactor hardware analyses 

gave 93

Mo/6

°Co activity scaling factors for stainless 

steel of about 2E-5 and for Inconel of about 2E-4 

(Robertson, Thomas, Wynhoff, and Haggard 1993; 

Lepel, Robertson, Thomas, Pratt, and Haggard 1995). 

After several thousand years, 93

Mo will be a 

significant radionuclide in activated metal wastes 

after the 'Co (5.27 year half-life) and 63

Ni (100 year 

half-life) have decayed. 

3.4 Niobium-93m 

Niobium-93m is produced by the reaction 93

Nb (n, 

n') 9 mNb, by beta decay of 93

Zr, and by the electron 

capture decay of 93

Mo. It is, therefore, produced by 

neutron activation of Nb, Zr, and Mo. Niobium is 

present as an impurity in stainless steel and some 

Inconel types at concentrations of 5 to 400 ppm, and 

is present as an additive in some types of Inconel at 

10,000 to 50,000 ppm (see Table 2.3). Niobium-93m 

has a half-life of 16.1 years, and decays with the 

emission of 16.5 keV Nb x-rays and a 30 keV 

gamma-ray. Because two of the 93

mNb parents, 93

Mo 

and 93

Zr, have extremely long half-lives, the 93

mNb 

will remain in some types of neutron-activated metal 

wastes for thousands of years. Analyses of neutron

irradiated stainless steel reactor hardware have

reported 93

mNb/ 6

OCo activity scaling factors of about 

3E-3, while Zircaloy and some Inconel alloys have 

ratios of about 1.0 (Robertson, Thomas, Wynhoff, 

and Haggard 1993; Lepel, Robertson, Thomas, Pratt, 

and Haggard 1995). 

3.5 Silver-108m 

Silver-I 08m is a 130-year half-life radionuclide 

produced by the thermal neutron activation reaction 

1

07

Ag (n, gamma) losmAg. Silver is present at a 

concentration of 80% in the Ag-Cd-In alloy used in 

PWR control rod assemblies. Therefore, losmAg is a 

major constituent of these types of spent control rod 

assemblies being replaced at PWR stations. The 

tOSrAg decays with the emission of several high

energy gamma rays (434, 614, and 723 keV) and 

therefore has a relatively high Curie-to-dose rate 

conversion factor. Silver-108m is generally not a 

major constituent of other types of low-level waste, 

e.g., DAW, spent ion exchange resins, and sludges. 

3.6 Cadmium-113m 

Cadmium-1 13m is a 14.1-year half-life radionuclide 

produced by thermal neutron capture by the 

following reaction, 1

12

Cd (n, gamma) 113

mCd. 

Cadmium is present at a concentration of 5% in Ag

Cd-In alloy used in PWR RCCA control rod 

assemblies. Therefore, "3

mCd is a major constituent 

of spent control rod assemblies of this type, but is not 

generally a major constituent of other types of low

level waste. Cadmium- 113m decays with the 

emission of a 590 keV beta particle with no 

accompanying gamma-ray emissions. 

3.7 Tin-121m 

Tin- 12 lm is a 55-year half-life radionuclide 

produced by thermal neutron activation of tin, 

according to the reaction, '

20

Sn (n, gamma) 12 mSn. 

Tin is added to Zircaloy metal components at 

concentrations ranging from 1 to 3%, and is, 

therefore, a major constituent of neutron irradiated 

Zircaloy, but is not a major constituent of other types 

of LLW. Tin-121m decays with the emission of a 

353 keV beta particle and an accompanying 26.3 keV 

x-ray.

13 NUREG/CR-6567
4 Sampling and Analyses of Neutron-Activated Metal and 

Primary Resin Low-Level Wastes

During this study, five types of neutron-activated 

metal wastes and two types of spent ion exchange 

resin wastes were collected from nuclear power 

stations and analyzed for 10 CFR Part 61 

radionuclides and other long-lived radionuclides not 

specifically mentioned in 10 CFR Part 61. This 

section describes the sampling and analyses of these 

highly radioactive specimens. 

4.1 Sampling 

This section describes the sampling of the activated 

metal wastes and the primary demineralization resins 

collected for the radiochemical analyses of the 10 

CFR Part 61 radionuclides and the additional long

lived radionuclides of interest. 

4.1.1 Neutron-Activated Metal Wastes 

Neutron-activated metal waste samples from nuclear 

power stations were difficult to obtain within the 

scope and budget of this project. Because of the very 

high gamma dose rates associated with these types of 

waste samples, it was difficult to obtain specimens 

that could be packaged, shipped, and handled without 

resorting to the use of very expensive handling 

procedures and highly-shielded shipping casks. 

Nevertheless, five specimens of neutron-activated 

metal waste materials were obtained for this study. 

These samples are listed in Table 4.1, and consist of 

three stainless steel (Type 304) specimens cut from 

two different BWR control rod blades and a cladding 

sample from a Low Power Range Monitor (LPRM); 

an Inconel-600 spring specimen from a PWR 

Burnable Poison Rod Assembly (BPRA); and a small 

piece of the Ag-In-Cd alloy used in a PWR Rod 

Control Cluster Assembly (RCCA). As evidenced by 

the discharge date of these samples from the reactors, 

they had been stored in the reactor's spent fuel pool 

for a number of years prior to disposal to allow 

shorter-lived radionuclides to decay and thus provide 

samples with manageable gamma dose rates.

The specimens of the LPRM and the control rod 

blade from Peach Bottom Station were cut from the 

whole components stored in the spent fuel pool at the 

reactor station and then shipped in shielded 55-gallon 

steel shipping drums to PNNL. The control rod 

blade from Duane Arnold Station and the BPRA and 

RCCA assemblies from Point Beach Station were 

shipped as whole components from the reactor 

stations to PNNL in special shipping casks as part of 

another program at PNNL sponsored by the DOE 

Office of Civilian Radioactive Waste Management. 

At PNNL all samples were unloaded into heavily 

shielded hot-cell facilities for further sub-sampling 

and preliminary cleaning of the small specimens. 

Following radiological analyses, all samples were 

disposed of as LLW at the Hanford DOE LLW 

disposal facility. 

Metal sub-samples were cut from the components, 

producing small pieces (several grams) which 

allowed for easier handling outside of the hot-cells 

and reduced the gamma dose rate from the sub

samples. The metal pieces were then very carefully 

cleaned in the hot-cell by repeatedly rinsing with a 

hot mixture of nitric/hydrochloric acid to 

decontaminate them from fission products and 

transuranic radionuclides that were present in the hot

cell in which the samples were prepared. The sub

samples were removed from the hot-cells and a 

second identical cleaning operation was conducted in 

an uncontaminated laboratory fume hood to insure 

complete decontamination from hot-cell radioactive 

material. Each metal sample was then either entirely 

dissolved or partially dissolved in appropriate acid 

mixtures and each stock solution was then brought to 

a total volume of 100 mL in volumetric flasks. The 

stock solutions were then used for radiochemical 

analyses of the radionuclides of interest listed in 

Table 4.2.

15 NUREG/CR-6567
Table 4.1 Neutron-Activated Metal Waste Samples for Radiological Characterization

Sample No.

#39 

#40

#43 

#41 

#42

Metal Waste Type 

Stainless steel cladding 

from LPRM(1) 

Stainless steel cladding 

from control rod blade 

Stainless steel cladding 

from control rod blade 

Inconel spring from 

BPRA(2 ) component 

Ag-Cd-In alloy from 

RCCA(') component

Discharge Date

1976 

3/89 

10/88 

11/75 

4/85

Reactor Station 

Peach Bottom 

Peach Bottom 

Duane Arnold 

Point Beach - 1 

Point Beach - 1

Materials Sampled

Stainless steel 

Type 304 

Stainless steel 

Type 304 

Stainless steel 

Type 304 

Inconel-600 

Ag (80%) 

Cd (5%) 

In (15%) alloy

(1) LPRM = Low Power Range Monitor 

(2) BPRA Burnable Poison Rod Assembly 

(3) RCCA = Rod Control Cluster Assembly

z 

C)

coo 

Cn 

Cd

ONl
Table 4.2 Spent Ion Exchange Resin Samples for Radiological Characterization

Sample Number 

Bead Resin 

#23 

Powdered Resin 

#22 

Bead Resin 

2-2

Discharge Date 

12/93 

3/93 

4/96

Reactor Station 

Crystal River 

Unit 3 

WNP-2 

Diablo Canyon

Resin Type 

Reactor Water Cleanup (RWCU) 

EDRFDR(') Composite 

+ RWCU(2 ) Composite 

Mixed-Bed Demin Resin 2-2

(1) EDRFDR = Equipment Drain Radioactive - Floor Drain Radioactive 

(2) RWCU = Reactor Water Clean Up

C) 

0

C.,'

C"
Sampling and Analysis

4.1.2 Primary Demineralization Resin 

Wastes 

The ion exchange resin LLW samples used for this 

study are listed in Table 4.2. 

The samples were packaged in shielded 5-gallon steel 

shipping cans at the nuclear power stations and 

shipped to PNNL. At PNNL the resin samples were 

transferred to clean laboratory fume hoods and 

prepared for radiochemical analyses. 

The representative aliquoting of the solid ion 

exchange resins for radiochemical analyses was a 

difficult task that required great care in minimizing 

inhomogeneities in the radionuclide content of the 

sub-samples for analyses. Each resin sample was 

thoroughly mixed, coned, and divided into separate 

aliquots for the various radionuclide analyses. Each 

of the aliquots was then gamma counted to determine 

the concentrations of the gamma-emitting 

radionuclides and to ascertain the representativeness 

of each aliquot. It was assumed that since the 

gamma-emitters included both fission and activation 

products, that the homogeneity of the gamma

emitters would be analogous for the non-gamma

emitters. 

This assessment showed that the aliquoting procedure 

provided separate aliquots that were very 

representative of one another. The major gamma

emitters in the powdered resin aliquots showed that 

the individual sub-samples varied from one another 

by only 1.2% to 11.5 % (percent standard deviation 

at the 95% confidence level), with the higher values 

due to relatively low counting statistics from some of 

the trace radionuclides. The bead resin sample 

provided a variance of only 3.3% to 5.0% (percent 

standard deviation at the 95% confidence level), 

indicating very representative sub-sampling of the 

resin. If the variance in the concentrations of the 

non-gamma-emitters is similar to that observed for 

the gamma-emitters, then it would be expected that 

the real variance in all radionuclides in the aliquoted 

sub-samples would not exceed the observed variance 

for the gamma-emitters, e.g., ranging from 

approximately 2% to 5% for most radionuclides. 

Thus, it appears that any slight inhomogeneities in 

radionuclide concentrations in the aliquots should not 

significantly bias the results of the radiochemical

measurements. 

The various resin aliquots were then used for 

radiochemical analyses of the radionuclides listed in 

Table 4.2. 

4.2 Summary of Radioanalytical 

Procedure Development for 

Additional Radionuclides 

This section provides a summary of the 

radioanalytical procedures that were developed for 

this project for the measurement of 10 CFR Part 61 

and other long-lived radionuclides present in various 

types of low-level waste from nuclear power stations. 

More detailed step-by-step radioanalytical procedures 

for these radionuclides are given in Appendix A. 

Beryllium-10 

Beryllium-10 is removed from the acid leach of 

resins and dissolution of metals by precipitating with 

NH4OH, dissolving the precipitate (ppt) in 6M HCl 

and passing the solution through an anion column 

which removes most of the transition metals. 

Beryllium-10 is reprecipitated with NH 4OH and the 

ppt dissolved in 0.1M HC1. Ultima Gold® 

scintillator is added and the "°Be measured in a 

Quantulus® beta spectrometer. Yields are 

determined by 7

Be tracer. 

Chlorine-36 

Chlorine-36 is separated from HC1 leach of the resin 

and the metal dissolution by adjusting the solution to 

1M HC1 and passing it through a cation resin column. 

Mercurous nitrate is added to the effluent of the 

column and Hg2Cl2 is precipitated. The ppt is 

washed, centrifuged, slurried with water, and 

transferred to a liquid scintillation vial. The slurry is 

suspended in Insta-Gel® scintillation cocktail and the 

36

C1 beta measured in a Quantulus® liquid 

scintillation beta spectrometer. Yields are 

determined using 38

C1 tracer. 

Iron-55 

Iron-55 is removed from the acid leach of the resin or

NUREG/CR-6567 18
Sampling and Analyses

the acid dissolution of the metal by successive 

precipitating with NH4OH. The ppt is dissolved in 

4M HCl and passed through an anion resin column 

where the iron is retained. Iron is then removed from 

the column by eluting with 0.01M HC1. The iron 

yield is determined gravimetrically and then the iron 

is transferred to a scintillation vial with Ultima 

Gold®, and the "Fe measured in a beta scintillation 

spectrometer. 

Nickel-59 and Nickel-63 

Nickel-59-63 is acid-leached from resin or metal 

samples are dissolved in appropriate acids. The acid 

solution is made basic after addition of iron carrier 

and nickel is scavenged by an iron hydroxide ppt 

with NH4OH. The solution is evaporated to just 

dryness, made just basic with NH4OH (pH 9), tartrate 

ion and NH4C1 solutions are added. Saturated 

dimethyl glyoxime in alcohol is added to ppt the Ni 

dimethyl glyoxime. The ppt is dissolved in nitric 

acid and evaporated to near dryness. HC1 is added 

with a drop of H2SO4 and evaporated to just dryness, 

pH 9 NH4OH is added, ammonium tartrate added, 

and Ni dimethyl glyoxime is precipitated. Once 

again the Ni precipitate is dissolved in nitric acid and 

converted to the chloride. The residue is dissolved in 

water which is divided into two equal parts. One part 

is placed in a scintillation vial, Ultima Gold® 

scintillator added, and the sample beta counted for 

"

36

Ni in a liquid scintillation spectrometer. The other 

portion is put in a beaker and sulfuric acid and 

thymol blue indicator added. The solution is adjusted 

with NH4OH until a gold-green end-point and the 

nickel is electroplated on a tared disk. Yield is 

determined gravimetrically and the electroplated 

nickel is counted in a intrinsic germanium low 

energy gamma spectrometer for measurement of 5

•Ni 

x-rays. 

Niobium-93m 

Niobium-95 tracer, Ni, Sr, Fe, Co, Sb, and Nb 

carriers are added to an aliquot of acid leachate or 

dissolution sample. The sample is then made basic 

with carbonate-free sodium hydroxide, and the 

resulting ppt is centrifuged. The precipitate is 

dissolved in hydrofluoric acid, and Co and Sb 

carriers are again added. The sample is transferred to 

a beaker containing concentrated nitric acid and

heated with stirring until the volume of acid is 

reduced to half. The sample is cooled and transferred 

to a centrifuge tube and the resulting ppt centrifuged. 

The precipitate is washed three times with water and 

then redissolved in HF. The sample is again 

transferred to a beaker containing concentrated nitric 

acid and heated with stirring until the volume is 

reduced to half. The precipitate is then washed twice 

with water and transferred to a 2-inch diameter 

planchet and dried. The planchets are covered with 

two layers of thin Mylar® plastic sheets and counted 

on the Low Energy Photon Spectrometer (LEPS) 

detector to determine the 93 MNb. The samples are 

then counted on the gamma detector for 1000 

minutes to determine the 9NB,and the 9Nb yield 

tracer. 

Strontium-90 

Cobalt and cesium carriers and "5

Sr tracer are added 

to an aliquot of each acid leachate or dissolution 

sample. The samples are brought to dryness, wet 

ashed, and dissolved in 8M nitric acid. Sr-Spec® 

columns are conditioned with 8M nitric acid and the 

samples loaded onto the columns. After the samples 

have passed through the columns, they are washed 

with 50 mL of 8M nitric acid. The columns are then 

heat-sealed in plastic and counted on an intrinsic 

germanium gamma detector to determine whether or 

not any 6°Co or '

37

Cs remained in the sample. If no 6

°Co or "

39

Cs is detected, five more mL of acid is 

passed through the columns. The time of the last 

wash is recorded. The Sr is then eluted with 10 mL 

of water and collected in a liquid scintillation vial. 

The samples are counted on the gamma detector to 

determine the "5

Sr recovery. They were then 

evaporated to dryness under a heat lamp and 15 mL 

of Ultima Gold® liquid scintillation cocktail is added 

to each vial. The samples are counted in the 

Quantulus® liquid scintillation spectrometer to 

determine the 1Sr content. 

Technetium-99 and Molybdenum-93 

The "Tc and 93

Mo are separated from an aliquot of 

the acid leach of the resin or the acid dissolution of 

the metal by adjusting to IM HCl, and passing the 

solution through a cation resin column. The effluent 

is passed through a TEVA-Spec® column where 

both the Tc and Mo are removed. The TEVA-Spec®

19 NUREG/CR-6567
Sampling and Analysis

column is washed and the resin extruded into a 

preweighed plastic cup, dried, weighed, and covered 

with Mylar® plastic film and counted Mylar®-side 

down on a LEPS detector for determining the yield of 

the 99Mo tracer. After the 67-hour 'Mo tracer has 

decayed away, the sampled is recounted on the LEPS 

for measuring 93Mo x-rays. The sample is then 

transferred to a scintillation vial, slurried with water, 

suspended in Insta-Gel® scintillation cocktail, and 

counted in a Quantulus® beta scintillation 

spectrometer for 99

Tc. 

Cadmium-109 and Cadmium-I 13m 

The •'0 Cd and " 3

mCd are purified by scavenging 

interfering radionuclides from aliquots of the acid 

leachate of the resin and the acid dissolution of the 

metal by precipitating AgCl. The ppt is filtered and 

the supernate containing the Cd is then carried with 

Fe(OH)3 from a slightly basic NaOH solution. The 

precipitate is dissolved in IM HC1 and passed 

through an anion resin column where the Cd is 

retained. After washing the column, the Cd is eluted 

with NH4OH. The ammonium salts are destroyed 

and the residue is dissolved in 0.1M HC1, made pH 8 

with NaOH, and thioacetamide is added to precipitate 

CDs. The CDs is counted in a gamma spectrometer 

to measure the 1

`9Cd and the "

5

1Cd tracer. The CDs is 

then evaporated to dryness with HNO3 to destroy the 

sulfide, and the Cd converted to the chloride by 

evaporating to dryness with HCI. The residue is 

dissolved in 0.1M HC1 and transferred to a 

scintillation vial where Ultima Gold® scintillator is 

added and the sample beta counted in a liquid 

scintillation spectrometer for I 3

mCd.

through a polysulfone filter in which a solution 

containing 100 ug of NdF3 has been passed through. 

The filters are mounted and counted in an alpha 

energy analyzer for the tracers and Pu, Am, Cm, and 

U isotopes. 

Tin-121m 

Am aliquot of each sample is placed in a beaker 

along with the following hold back carriers: Sn (II), 

Co, Ni, Nb and Cs. Then i.7mSn tracer and 2-3 mL of 

conc. HC1 are added and the samples are evaporated 

to dryness. Sulfuric acid and hydrogen peroxide are 

added and the samples taken to fumes. The samples 

are cooled, water is added, and the samples are 

chilled in an ice bath. Potassium iodide is then added 

to each sample and the tin is extracted twice into 

toluene. The tin is then back-extracted into a weak 

sulfuric acid solution, neutralized with sodium 

hydroxide, and precipitated as the sulfide using 

thioacetamide. The sulfide precipitate is destroyed 

with concentrated nitric acid, which results in tin 

oxide being formed. This oxide is washed with hot 

water, centrifuged, and slurried with a small amount 

of water. The samples are first counted on an 

intrinsic germanium gamma detector to determine the 

yield of the llmSn, and then a liquid scintillation 

cocktail is added and the samples are counted in the 

Quantulus® liquid scintillation counter to determine 

the 1

2

lmSn content.

Plutonium-Americium-Curium-Uranium

Aliquots of the acid leach of the resin or acid 

dissolution of the metal are traced with 242Pu, 243

Am, 

232

U, and evaporated to dryness. The residue is 

dissolved in 3M HNO3, made 0.1M with ferrous 

sulfamate and 0. 1M with ascorbic acid, and passed 

through a TRU-Spec® column. The column is 

washed and the Am-Cm eluted with 4M HCL. The 

Pu is then eluted with 4M HC1 which contains TiCl3, 

and the U is eluted with 0. 1M ammonium oxalate 

solution. Each fraction is evaporated to dryness, 

dissolved in 10 mL of IM HC1, 100 ug Nd carrier 

added, and 1 mL HF added. Each fraction is filtered

NUREG/CR-6567 20
5 Concentrations of Additional Long-Lived Radionuclides in Low-Level 

Wastes

This section provides the radioanalytical results of 

the analyses of the five neutron-activated metal and 

three spent ion exchange resin samples. These 

analyses include the 10 CFR Part 61 radionuclides, 

plus the additional long-lived radionuclides discussed 

in Section 3. These long-lived radionuclides are of 

interest because new LLW disposal facilities will be 

designed to contain the wastes for hundreds of years 

before the facilities are breeched and infiltrated with 

groundwater. Therefore, those radionuclides having 

half-lives long enough to remain for hundreds of 

years will be those of greatest radiological concern to 

the surrounding public. 

5.1 Neutron-Activated Metal 

Wastes 

The radionuclide measurements of the five neutron

activated metal samples are given in Tables 5.1 and 

5.2. Table 5.1 gives the neutron activation products 

measured in three samples of Type-304 stainless steel 

from two spent control rod blades and a spent Low 

Power Range Monitor (LPRM). These data provide 

not only 10 CFR Part 61 radionuclide analyses, but 

also measurements of a number of other long-lived 

radionuclides that are present in some types of 

activated metal components from nuclear power 

stations. For the stainless steel specimens, the 

concentrations of most of the additional long-lived 

radionuclides were very low or non-detectable. 

The radionuclide concentrations for sample #43, the 

Duane Arnold Control Rod Blade (CRB) at the time 

of discharge from the reactor and their change in 

relative abundance with time are shown in Figure 

5.1. As shown in the figure, 'Co is initially the most 

abundant radionuclide but decays relatively rapidly. 

The 63

Ni becomes the most abundant radionuclide 

after 5 to 10 years and remains the most abundant for 

many hundreds of years. At about 700 years after 

discharge from the reactor, the concentrations of 63

Ni 

and 59

Ni become about equal and about two orders of 

magnitude greater than the next most abundant 

radionuclide, "

0

Be.

These measurements of "

0Be are believed to be the 

first ever made for LLW materials. The "°Be is 

produced by neutron activation of the boron in the 

boron carbide and borosilicate glass neutron-absorber 

materials in the Duane Arnold CRB and the Point 

Beach Burnable Poison Rod Assembly (BPRA), 

respectively, by the reaction, `°B (n,p) 1

°Be. Table 

5.3 shows the "0

Be concentrations measured in these 

materials, together with the 60

Co concentrations for 

reference. For these types of boron-containing 

control rod assemblies, 10

Be will be the second most 

abundant activation product (next to 59

Ni) after about 

1400 years of decay, and because of it's relatively 

energetic beta emissions (556 keV) it will be a major 

contributor to the radiation dose from these control 

rod components. 

Of the additional long-lived radionuclides measured 

in the stainless steel specimens, only the 9

SmNb was 

detectable in all samples. As shown in Table 2.4, the 

93mNb is produced by the reaction, 93

Nb (n, n') 93mNb 

on the stable niobium present in Type-304 stainless 

steel at concentrations ranging from 5 to 300 ppm. 

Although the concentrations of the 93'Nb were 

initially about 70 times higher than the 

concentrations of the 4Nb, after about 100 years the 

concentrations of the two radionuclides will be about 

equal, due to the much shorter half-life of 9

3mNb. 

After about 200 years, the 9"'Nb will become 

negligible in low-level waste. 

The stainless steel from the Duane Arnold CRB 

contained a detectable amount of lI3mCd. After about 

175 years the "3

'Cd will be about the fifth most 

abundant activation product in this CRB. Although 

`Cd was also detected in this sample, it was much 

less abundant than the tI3mCd and it's shorter half-life 

would not make it a significant contributor to the 

long-lived radionuclide inventory in CRBs. 

A small amount of 93

Mo was detected in sample #43, 

but would never be a major radionuclide constituent 

or contributor to the radiation dose.

21 NUREG/CR-6567
Concentrations 

Table 5.1 Long-Lived Neutron Activation Products in Irradiated Stainless Steel Low-Level Waste from Nuclear Power Stationne

Concentration (microcuries/gram of metal)* 

#39 

304 Stainless Steel #43 #40 

LPRM0 ) 304 Stainless Steel CRB(2 ) 304 Stainless Steel CRBa2 ) 

Radionuclide Peach Bottom Duane Arnold Peach Bottom 

1OCFR61 Radionuclides 

53

Fe 1.58±0.03E-2 6.83±0.21E2 2.10-4.06E0 

59

Ni 1.00±0.17E-4 3.26-0.23E0 2.7±1.1E-3 

6Ni 1.07±0.08E-2 4.54+0.09E2 2.7±0.15E-1 

°Co 4.83±0.41E-2 1.49±0.043E3 3.61±0.04E-1 

'Sr <2.OE-6 <2.9E-4 <2_1E-5 

94Nb <1.3E-5 6.5±2.7E-3 <3.5E-4 

'Tc <2.9E-6 4.08+0-33E-3 <L-8E-4 

1291 

137s <2.OE-4 <9.2E-1 <6.3E-4 

24U <1.6E-7 <l.lE-6 <1.7E-7 

2U <8.1E-7 <1.9E-6 <8.8E-7 

<1.7E-7 cU <1.8E-6 <2.2E-7 

3

8 Pu <3.5E-8 <1.2E-5 <2.3E-4 

29'24°pu 1.72±0.95E-7 2.9t1.1E-5 3-5±1.9E-5 

24 Pu <4.4E-5 <1.6E-3 <4.9E-5 

"24]Amn <2.6E-7 <1.8E-5 <6.7E-7 

244

Cm <3.8E-7 <2.7E-5 <1.OE-6 

Additional Long-Lived Radionuclides 

"°Be <5.3E-5 <2.2E-3 <8.2E-5 

36C1 <3.7E-5 <1.8E-3 <7.2E-5 

'Mo <2.5E-6 5.22+0.20E-4 <7.lE-6 

93.Nb 1.23±0.55E-4 4.48±0.32E-1 2.0±1.OE-3 

l°8iAg <2.4E-4 <1.4E0 <6.5E-4 

liiCd <4.OE-5 6.4±2.2E-3 <1.2E-4 

"113mCd <5.8E-5 9.3+1.8E-2 

_2__Sn <2.1E-4

0) JruiUvI = Low Power Rtange lVlomtor; flJ4 Stamless steel cladding; discharged 1976. 

(2) CRB = Control Rod Blade; 304 Stainless steel cladding; discharged 10/88. 

* Concentration on June, 1996 

NUREG/CR-6567 22
Concentrations 

Table 5.2 Long-Lived Neutron Activiation Products in Irradiated Metal Low-Level Waste from Nuclear Power 

Stations 

Concentration (microcurieslgram of metal)* 

#41 #42 

Ineonel Ag-Cd-In Alloy 

BPRA Springa•) RCCA (2) 

Radionuclide Point Beach Point Beach 

10 CFR61 Radionuclides 

.

5 Fe 1.14±0.03E-1 6.76±0.20E-3 

59

Ni 2.15±0.17E-1 1.39±0.62E-3 

6'Ni 2.77±0.11E1 1.16±0.005E-1 

°Co 1.95-0.1OEO 2.00-0.10E1 

90Sr <3.2E-5 <2.3E-4 

94Nb 1.56±0.09E-3 <3.7E-4 

9

'Tc <1.4E-5 2.92*0.37E-4 

1

37

Cs <4.6E-3 <2.OE-1 

24u <4.1E-7 <1.2E-6 

23SU <5.9E-7 <7.1E-7 

23SU <1.4E-7 <4.6E-7 

w8

Pu <9.3E-8 3.7:-2.0E-7 

239'24°pu 1-6±0.5E-6 6.3-+2.5E-7 

241Pu <7.2E-5 <5.3E-4 

241AM <5.4E-7 <3.3E-6 

2 4

4Cm <8.1E-7 <4.5E-6 

Additional Long-Lived Radionuclides 

"

0

Be 1.2-±0.5E-4 <1.6E-4 

3

MCI 3.95±0.95E-4 

93

Mo 1.62±0.63E-4 <1.8E-5 

93.Nb 1.31:L0.06E- 1 3.5±1l.6E-3 

1°8'Ag <5.1E-3 6.37•-0.19E2 

1'9Cd 1.4:0.3E-3 3.23--O.07EO 

13.Cd 1.28-0.03E-2 2.07±O.01E1 

12ISn <9.9E-5 <1.5E-3 

(1) BPRA = Burnable Poison Rod Assembly; discharged 11/75 

(2) RCCA = Rod Cluster Control Assembly; discharged 4/85 

* Concentration on June, 1996

23 NUREG/CR-6567
Concentrations 

104 

103 

E 12 N-63 

E 101 

1 Ni-59 

z 0.- 10 0 

I-- - 1 

Z 10 Be-10 uLJ\ 

U 

z1 Nb-94 

uJ Tc-99 S 10-3 

Mo-93 z 10 0 10 Cd-113m 

10-~ f

Fe-55 Co-60 N•3: 

10" 6- 1 1 1 1 

0 200 400 600 800 1000 

DECAY TIME (YEARS) 

Figure 5.1 Change in Radionuclide Concentration with Time in 304 SS From Duane Arnold 

Spent Control Rod Blade

NUREG/CR-6567 24
Concentrations

Table 5.3 Beryllium-10 and 6 0

Co Concentrations in Selected Boron-Containing LLW 

Samples 

Activity 

Concentration (uCi/g)* Ratio 

Sample Material Source 10

Be 6

°Co '

0Be/I°Co 

DER-1 boron Spent BWR 3.68 h 0.04E-2 10.2 3.61E-3 

carbide Control Rod 

Blade 

Duane Arnold 

DER-2 boron Spent BWR 3.91 4- 0.04E-2 5.47 7.15E-3 

carbide Control Rod 

Blade 

Duane Arnold 

DER-4 borosilicate Spent PWR 2.72± 0.08E-3 5.12 5.31E-4 

glass BPRA Rod () 

Point Beach 

2-2 ion RWCU Resin() 8.22 ± 0.36E-3 8.70 9.45E-4 

exchange Diablo Canyon 

resin 

* Concentrations decay corrected to 5 years after discharge of components or materials from 

reactors 

(a) BPRA = Burnable Poison Rod Assembly 

(b) RWCU = Reactor Water Clean-Up

25 NUREG/CR-6567
Concentrations

Tin-12 1 m was not a significant constituent in 

spentCRB stainless steel, although it would be 

significant in neutron-irradiated Zircaloy materials 

used for fuel cladding and other reactor internal 

components. 

Table 5.2 gives the radionuclide concentrations 

measured in two types of specialty alloys used in 

control rod assemblies. The Point Beach BPRA 

control rod uses springs made of an Inconei alloy. 

This specimen (sample #39) contains relatively high 

concentrations of 63

Ni and 59

Ni due to nickel being a 

major constituent (52 to 76%) in a variety of Inconel 

alloys. These two nickel radioisotopes will dominate 

the activation product abundances in Inconel alloys 

for many hundreds of years (see Figure 5.2). 

Niobium is also an added constituent (up to about 

5%) in some Inconel alloys (see Table 2.3), and in 

sample #39, 93mNb and 94

Nb are major radionuclide 

constituents. After about 1500 years, the 94

Nb will be 

the second most abundant radionuclide in this 

Inconel, and because of it's relatively high-energy 

beta and gamma radiations, it will be a major 

contributor to the radiation dose from neutron

activated Inconel alloys containing niobium. 

Molybdenum-93 will also be a major long-lived 

radionuclide constituent, being produced by neutron 

activation of the percent levels of molybdenum 

present in Inconel alloys. Technetium-99 is also 

produced by neutron activation of molybdenum, and 

is present at about 10-fold lower concentration 

compared to the 93

Mo. Nevertheless, 9Tc will be 

about the fifth most abundant radionuclide after 

about 300 years. The cadmium and tin radionuclides 

will never be major constituents of neutron-irradiated 

Inconel. 

Another speciality metal associated with RCCA 

control rods is the silver (80%)-indium (15%)

cadmium (5%) alloy used as a neutron-absorber 

material. This material has a unique radionuclide 

composition following neutron irradiation. Silver

108m will be the dominant neutron activation 

product for many hundreds of years, and because of 

it's relatively high energy beta and gamma 

radiations, it will be the major dose contributor for 

this material (see Figure 5.3). Shortly after 

discharge, the cadmium radionuclides, 13

'mCd and

1

. 9Cd, will be the next most abundant radionuclides, 

but the 'O9

Cd rapidly decays to insignificance. The 

"m3mCd will be the second most abundant radionuclide 

up to about 150 years and then the 63

Ni becomes the 

second most abundant. 

5.2 Primary Demineralization Resin 

Wastes 

Next to neutron-activated metal wastes, primary 

coolant demineralization resin wastes from nuclear 

power stations contain the highest concentrations of 

radionuclides. These radionuclides are a mixture of 

neutron activation products and fission products. 

Table 5.4 provides the concentrations of 10 CFR Part 

61 radionuclides, plus the additional long-lived 

radionuclides measured in this study in samples of 

PWR (sample #23) and BWR (sample #22) primary 

coolant demineralization resins. 

The PWR resin (sample #23), a RWCU resin from 

Crystal River station contains predominantly 137

Cs, 

indicating minor release of fission products from the 

fuel. The change in radionuclide concentrations with 

time for this material is shown in Figure 5.4. After 

about 100 years, the 137

Cs and the 'Co will have 

decayed below 63

Ni in concentration and the long

lived nickel radionuclides will be the most abundant 

for many hundreds of years. After about 1500 years, 

the 59

Ni will be the most abundant and 36

C1 will be 

the second most abundant radionuclide is this resin 

waste. The additional long-lived radionuclides 

measured in this study, except for the 36

C1 will never 

be major contributors to the radionuclide 

concentration or radiation dose from this material. 

The BWR resin (sample #22), a mixture of RWCU 

and EDRFDR resins from WNP-2 nuclear power 

station is dominated by 6

"Co for the first few decades. 

Nickel-63 and '

37

Cs then become the most abundant 

radionuclides after about 30 to 150 years. Thereafter, 

the 63

Ni and I

9

Ni will be predominant, followed by 

"

49

Nb, 93

Mo, and 9Tc which are all close in 

concentration. The additional long-lived 

radionuclides measured in this study will not be 

significant contributors to the concentration or 

radiation dose for this material.

NUREG/CR-6567 26
Concentrations

0 200 400 600 800 

DECAY TIME (YEARS) 

Change in Radionucide Concentration with Time in Point Beach BPRA 

Inconel Spring

1000

NUREG/CR-6567

E 

03 

z 

0 

I

z 

LU 

C.-

z 

0 

C, 0 

uLJ 

0 

z 

0 

0 

Wi

102 

101 

100 

10" 

10" 

10

10-' 

10-' 

10

10

10-1 

10-.

Figure 5.2

27
Concentrations

0 200 400 600 800

DECAY TIME (YEARS)

Change in Radionuclide Concentration with Time in Point Beach RCCA 

Ag-Cd-In Alloy

NUREG/CR-6567

103

10

2

101 

100 

10" 

10"

E 

z 

0 

I

0 

z 

0 

0 

LU 

z 

0 

--I 

0 

0 

0•

10" 

10" 

10" 

10" 

10"

Figure 5.3

1000

28
Concentrations

200 400 600 800 1 

DECAY TIME (YEARS) 

Change in Radionuclide Concentration with Time in Crystal River RWCU 

Resin

000

NUREG/CR-6567

C.) 

z 

0 

z 

LIi 

0 

z 

0 

0 

,-J 

i

z 

0

103 

102 

101 

100 

10" 

10" 

10" 

10" 

10" 

10"

0

Figure 5.4

29
Concentrations 

Table 5.4 Long-Lived Neutron Activiation Products in Spent Ion Exchange Resins from Nuclear Power Stations

Uoncentration on January, 196 ** Measured by thermal emission mass spectrometry 

(1) RWCU = Reactor Water Clean Up; discharged 12/93 

(2) EDRFDR = Equipment Drain Radioactive-Floor Drain Radioactive; discharged 3/93 

NUREG/CR-6567 30

Concentration (microcuries/gram of resin)* 

#23. #22 

Bead Resin Powdered Resin 

RWCUa') EDRFDR(2)+RWCU Resins 

Radionuclide Crystal River WNP-2 

10 CFR61 Radionuclides 

5'Fe 4 48±0.04E- 1 1.64±0.05E-1 

59Ni 1.91:E0.09E-1 9.3±1.2E-3 

63Ni 2.25-0.07El 8.90±0.23E-1 

GOCo 6.40±0.14E0 2.777±0.08E1 

90Sr 4.48±0.39E-1 9.8±1.9E-3 

94Nb <7.4E-4 <3.5E-4 

99Tc 9.7±3.3E-5 <2.2E-4 

1291** 6.51±0.07E-5 1.64±0.01E-7 

137

Cs 2.23±0.09E2 6.17±0.19E-1 

234u <2.4E-5 <2.1E-6 

23u <9.9E-6 <LIE-5 

238U <3.3E-6 <6.4E-6 

2

8

Pu 1.07±0.19E-5 <2.5E-6 

23 9

,

2 4 0

PU 7.0±1.4E-6 7.9±5.4E-6 

24

'Pu <6.9E-4 

24Am 2.65±0.70E-5 <6.3E-6 

2 4

4Cm 9.3±4.1E-6 <9.9E-6 

Additional Long-Lived Radionuclides 

1°Be <1.08E-4 <7.9E-4 

36

C1 2.28±1.06E-3 <1.4E-2 

93

Mo <1.6E-5 2.6±1.OE-4 

"q3 9 Nb 3.7±1.6E-3 3.9±1.5E-3 

losmAg <5.OE-2 <1_3E-2 

109Cd 5.2±3.8E-4 <2.1E-2 

l

13

mCd 5.8±l.3E-3 1.71±0.58E-2 

_2__Sn <3.9E-4 <5.2E-3
Concentrations

Another RWCU resin from Diablo Canyon nuclear 

power station that was prepared for disposal was also 

analyzed for several of the additional long-lived 

radionuclides, including "

0

Be and 36C1 (see Figure 

5.6). The concentration units reported for this resin 

are in total mCi of each radionuclide in the entire 

resin shipment. Initially, 6

°Co is the most abundant 

radionuclide, but after a few decades `

4

C becomes, by 

far, the most abundant activation product, followed 

by 63

Ni, 59

Ni, 1

"Be, and 4Nb which will be the 

dominant radionuclides in this waste for thousands of 

years. The 36

C1 is a rather minor constituent of this 

resin, as are the fission product radionuclides, 

indicating very good fuel cladding integrity at Diablo 

Canyon station. This resin does illustrate that "°Be 

will be one of the major long-lived radionuclides in 

spent RWCU resin wastes from reactor stations using 

high concentrations of boron in their primary coolant 

as a reactivity shim.

NUREG/CR-6567 31
Concentrations

0 200 400 600 800

DECAY TIME (YEARS)

Change in Radionuclide Concentration with Time in WNP-2 

EDRFDR+RWCU Resins

NUREG/CR-6567

f

.m 

z 

0 

I.

z 

U 

z 

0 0 

LU. 

0 

-.J 

z 

0

102 

101 

100 

10" 

10" 

10" 

10" 

10" 

10" 

10" 

10"

Figure 5.5

1000

32
Concentrations

105 

C-14 

104 

1013 

Ni-5 

Ni-5 z 0 Be-lO0 

o 10-1 

<C) z 100 

-J 

U. Nb-94 

Z 1 0 10" 

tu CI-36 

10" 37 1s00 

m Pu-239 

z lO" 3. ' 

0 

S10" 1-129 

10-6 C 

0 1 000 2000 3000 4000 5000 

DECAY TIME (YEARS) 

Figure 5.6 Decrease in Concentrf;no- •-f' . T Z D-3 _..... in........ .

Exchange Resin (RWCU) .... vu Hamonucliaes i .WR Spent Ion

33 NUREG/CR-6567
6 Impact on Low-Level Waste Classification

The classification and disposal requirements of LLW 

are based, among other things, upon the 

concentrations of specified long-lived radionuclides 

in the waste. These radionuclides were selected 

because of their long half-lives, high environmental 

mobility, and potential to be present in wastes in 

significant concentrations. In addition to the 

specified radionuclides, a number of other long-lived 

radionuclides not included in 10 CFR Part 61 may be 

present in some types of wastes generated at 

commercial nuclear power stations. This study has 

identified a number of these additional radionuclides 

and this section provides an assessment of their 

impact upon 10 CFR Part 61 requirements. 

The additional long-lived radionuclides identified in 

this examination of LLW materials, including 

neutron-activated metal wastes and primary resin 

wastes, appears to have minimal impact upon the 

requirements of 10 CFR Part 61. Of the additional 

radionuclides studied in this project, only JOSmAg, 

"°Be, and 36

CI would appear to be of any significance 

from a low-level waste disposal and radiation 

dosimetric standpoint. The other long-lived 

radionuclides measured in this study ( 3

Mo, 93mNb, 

[I3

mCd, 1

z"rSn) are present in sufficiently low 

concentrations and have relatively low dose 

conversion factors and environmental mobility, and 

thus do not appear to represent any significant 

potential hazard in the disposal of these LLW 

materials. 

The relative concentrations of these radionuclides do 

not appear to be high enough to affect the waste 

classification criteria specified in 10 CFR Part 61. 

The potential radiation dose to the surrounding 

offsite general population is the primary concern of 

any additional radionuclides measured in LLW from 

nuclear power stations. 

Table 6.1 lists the radiation dose conversion factors 

for the existing 10 CFR Part 61 radionuclides, plus 

the additional long-lived radionuclides measured in 

this study (Kennedy and Strenge, 1994). Of the 

group of additional radionuclides, 'OSmAg has the 

highest external dose conversion factor because of its 

three high energy, high abundance gamma-rays (see 

Table 2.4). Fortunately, the only major source of this 

radionuclide in LLW from nuclear power stations is

from spent Rod Cluster Control Assemblies (RCCA) 

from PWR stations. As discussed in Section 4, these 

control rod assemblies contain 80% silver, which is 

the parent material for the l°smAg. Thus, the 

potential LLW source term for this radionuclide will 

be somewhat limited. Nevertheless, it may be 

prudent to conduct a detailed assessment of l°SmAg to 

determine if amendment into 10 CFR Part 61 would 

be warranted. 

Beryllium- 10 is produced from boron present in 

PWR primary coolant and in BWR and PWR control 

rods and will be continually present in LLW 

containing primary demineralization resins and spent 

control rods. Beryllium-10 has a relatively low 

external dose conversion factor because it is a pure 

beta emitter, but has moderately high inhalation and 

ingestion dose conversion factors. Because of its 

potential to be present in significant concentrations in 

LLW from nuclear power stations, "0

Be should be 

further evaluated to determine if it should be 

amended to the list of specified radionuclides in 10 

CFR Part 61. 

The external dose conversion factor for 36

C1 is also 

relatively low because it is a pure beta emitter, but 

the ingestion and inhalation dose conversion factors 

are moderately high. Because the main source of this 

radionuclide in LLW from nuclear power stations is 

the reactor primary coolant demineralizer resins, it 

will be ubiquitous in LLW in relatively small 

quantities. However, the high environmental 

mobility of this radionuclide will allow it to migrate 

from a LLW disposal site that has been infiltrated 

with groundwater. For this reason, it may be prudent 

to conduct a detailed assessment of 36

C1 to determine 

if it should be included in 1 OCFR6 1. An estimate of 

the yearly quantities of 36

C1 contained in spent 

primary resin LLW from all PWR stations in the U. 

S. can be estimated as follows. About 3.4E7 

grams/year per PWR station of spent primary resins 

are generated for disposal. Using the range of 36C1 

concentrations measured in the two primary resins in 

this study (9.3E-5 to 3.4E-3 uCi/g), approximately 

3200 to 78,000 uCi/year per PWR station are 

produced. Multiplying by the number of PWR 

stations in the U. S. (70) gives an estimated 0.2 to 5.5 

Ci/year of 36

C1 contained in LLW from PWR stations. 

In a LLW site performance assessment modeling

35 NUREG/CR-6567
Low-Level Waste

Dose Conversion Factors for 1OCFR61 and Additional Long-Lived Radionuclides Measured 

in Low-Level Waste (from Kennedy and Strenge, 1994)

External (surface) 

Sv/d 

Radionuclide Ingestive (Sv/Bq) Inhalation (Sv/Bq) Bq/m2 

1OCFR61 Radionuclides 

14

C 5.64E- 10 5.64E- 10 1.39E-15 

'

9

Ni 5.67E-11 7.31E-10 0 

63

Ni 1.56E-10 1.70E-9 0 

6 0

Co 7.28E-9 5.91E-8 2.03E-10 

90

Sr 3.85E-8 3.5 1E-7 2.46E-14 

90y 2.91E-9 2.28E-9 4.60E- 13 

94

Nb 1.93E-9 1.12E-7 1.32E-10 

99

Tc 3.95E-10 2.25E-9 6.73E-15 

1291 7.46E-8 4.69E-8 2.23E-12 

1

3 7

Cs 1.35E-8 8.63E-9 2.46E- 14 

238Pu 8.65E-7 1.06E-4 7.23E-14 

239pu 9.56E-7 1.16E-4 3.17E-14 

240pu 9.56E-7 1.16E-4 6.93E-14 

24'Am 9.84E-7 1.20E-4 2.37E- 12 

Additional Radionuclides 

1

°Be 1.26E-9 9.58E-8 3.56E-14 

36

C1 8.18E-10 5.93E-9 5.81E-14 

93

Mo 3.64E-10 7.68E-9 4.61E-13 

93mNb 1.41E-10 7.90E-9 8.11E-14 

"l°smAg 2.09E-9 4.18E-9 1.60E-10 

13mCd 4.35E-8 4.13E-7 2.27E-14

To convert Sv/Bq to Rem/Ci or Sv/d to Rem/d multiply values in table by 3.70E12 

Bq/m2 Ci/nm

NUREG/CR-6567

Table 6.1

36
Low-Level Waste

exercise conducted by the NRC, modeling results, 

using a 0.1 Ci 36

CI source term, indicated that this 

radionuclide would be the major contributor to the 

offsite ingestion radiation dose for the contaminated 

well/drinking water exposure scenario (Campbell, 

1994). Thus, even though the relative quantities of 

"

36

C1 in LLW are small compared to other 

radionuclides, it can be a major contributor to offsite 

radiation doses, and deserves further study to gain a 

better understanding of its radiological and 

environmental impacts from LLW disposal. 

Cadmium-113m has relatively high ingestion and 

inhalation dose conversion factors. However, 

because of its relatively short half-life (14.1 years) 

and the fact that it is present in LLW only in spent 

PWR Reactor Control Cluster Assemblies, further 

consideration for inclusion in 10 CFR Part 61 may 

not be warranted.

37 NUREG/CR-6567
7 Determination and Review of Activity Scaling Factors

A common practice in the nuclear power industry is 

to use activity scaling factors for estimating the 

concentrations of difficult-to-measure radionuclides 

and determining their waste classification in LLW 

packages being prepared for disposal. Typically, a 

nuclear power station will periodically sample a 

variety of LLW types and submit them to service 

analytical laboratories for comprehensive 10 CFR 

Part 61 radionuclide analyses. Then the station will 

develop activity scaling factors for the difficult-to

measure radionuclides (pure beta emitters, low

energy photon emitters, and alpha emitters) relative 

to the most abundant long-lived gamma-emitting 

radionuclides present in the waste, usually 'Co and 

137

Cs, which are easily measured by gamma-ray 

spectrometry. A data base of scaling factors is then 

generated for the various sample types and used in 

conjunction with gamma-ray spectrometry to 

quantify the 1 OCFR61 radionuclides and determine 

their waste classification. Several data bases and 

assessments of activity scaling factors for the 

radionuclides specified in 10 CFR Part 61 in LLW 

from nuclear power stations have been published 

(Best and Miller, 1985, 1987; Cline, Noyce, Coe and 

Wright, 1985; Cline, Wright and Hollcroft, 1980; 

James and Vance, 1992; Robertson, et al., 1989; 

Vance, 1992; Vance, 1993; Vance, Helmholz and 

Sutherland, 1988; Vance, Holderness, James, and 

Lee, 1992). 

A number of difficult-to-measure radionuclides lend 

themselves well to this activity scaling factor 

methodology, particularly neutron activation 

products (e.g., "Fe, 63

Ni, 59

Ni) which have similar 

production modes and similar physicochemical 

behavior in the reactor coolant and related waste 

streams compared to 6

"Co. Other long-lived fission 

product radionuclides, such as 1291, can be generically 

scaled to '

37

Cs, whereas others are only marginally 

adapted to the use of scaling factors. 

7.1 Determination of Activity 

Scaling Factors for 

Additional Radionuclides

There are a number of other long-lived difficult-to

measure radionuclides present in various types of 

LLW that are not specifically mentioned in 

I OCFR6 1. These radionuclides are usually (but not 

always) present as minor radionuclide constituents in 

LLW. For example, l""mAg and I 3

mCd are major 

radionuclide constituents in spent RCCA control rods 

composed of Ag-Cd-In neutron-absorbing alloy 

material, and 9mNb is a major constituent in neutron

irradiated high-niobium Inconel alloys. This section 

evaluates the potential use of activity scaling factors 

for additional long-lived radionuclides observed in 

various waste types. 

7.1.1 Neutron-Activated Metal 

Radionuclides 

7.1.1.1 Molybdenum-93 

Molybdenum-93, which is produced by neutron 

activation of molybdenum present in stainless steel 

and some Inconel alloys, can generally be correlated 

rather well with 6

"Co. However, the data base for this 

radionuclide in activated metals is quite limited, 

being confined to this study and to data published in 

NUREG/CR-6390 (Lepel, et al., 1995). Table 7.1 

provides the measured concentrations of 9"Mo and 

"Co in the neutron-activated metal samples 

determined in this study, plus data taken from 

NUREG/CR-6390 (Lepel, et al., 1995). Although the 

absolute concentrations of these radionuclides vary 

by many orders of magnitude for the various 

activated-metal samples examined, the activity 

scaling factors for the 'Mo/"Co ratios were 

reasonably consistent, with a geometric mean of 

1.62E-6. Figure 7.1 graphically shows the 

correlation of these two radionuclides for the samples 

listed in Table 7.1. This correlation is quite good, 

with only a few outlying data points, considering the 

wide range of sample types examined. Thus, the use 

of a generic activity scaling factor for nuclear power 

station LLW appears to be reasonable for this 

radionuclide pair.

39 NUREG/CR-6567
Determination and Review

Table 7.1 93Mo/6°Co Scaling Factors for Low-Level Waste

Scaling 

Concentration (gCi/g)(a) Factor 

ID No. Sample Material 93

Mo 60

Co 93Mo/ 60Co 

DER-1 Duane Arnold SS 1.49E-2 1.08E-4 1.38E-6 

CRB") 

DER-1Z SS 8.46E-3 8.37E-3 1.01E-6 

DER-2 SS 7.82E-4 1.61E-3 4.86E-7 

DER-2Z SS 5.40E-3 4.79E-3 1.13E-6 

DER-3 SS <4.2E-4 2.03E-2 2.07E-6 

DER-3Z SS <3.6E-4 1.14E-2 3.16E-6 

DER-4(out) Point Beach SS 5.17E-3 1.46E-4 3.54E-7 

BPRA~c) 

DER-4(in) SS 1.92E-2 1.80E-3 1.07E-5 

DER-5(out) SS 7.42E-3 1.79E-4 4.15E-7 

DER-5(in) SS 1.22E-2 2.20E-3 5.55E-6 

DER-6(out) SS 2.53E-3 1.70E-4 1.49E-7 

DER-6(in) SS 9.89E-3 2.01E-3 4.92E-6 

DER-10A Point Beach SS 2.07E-2 4.54E-4 4.56E-7 

RCCA("d) 

DER-10B SS 2.31E-2 5.69E-4 4.06E-7 

DER-8 SS 1.44E-2 9.85E-3 1.46E-6 

#39 Peach Bottom SS <2.5E-6 3.71E-1 <6.90E-6 

LPRM('e) 

#43 Duane Arnold SS 5.22E-4 2.12E-3 2.46E-7 

CRB 

#40 Peach Bottom SS <7.lE-6 4.84E-1 <1.48E-5 

CRB 

#41 Point Beach Inconel 1.62E-4 1.52E-1 1.07E-5 

BPRA 

(Spring)

NUREG/CR-6567 40
Determination and Review

Table 7.1 (continued)

Scaling 

Concentration (pCi/g)(a) Factor 

ID No. Sample Material 93

Mo 60

Co 9 3

Mo/60Co 

#42 Point Beach Ag-Cd-In <1.81E-5 4.50E-1 <4.OOE-7 

RCCA 

#23 Cyrstal River Resin <1.6E-5 4.36E-0 <3.67E-6 

RWCU 

Resin M 

#22 WNP-2 Resin Resin 2.6E-4 2.08E-1 1.25E-5 

Mean • 3.76E-6 

Standard 4.52E-6 

Deviation 

Geometric 1.62E-6 Mean * * 

(a) Concentrations decay-corrected to 5 years after discharge of components or materials from the 

reactor. It was necessary to sandardize the decay time after discharge from the reactors because 

some of the materials were quite old and the 'Co had undergone substantial radioactive decay. 

(b) CRB = Control Rod Blade 

(c) BPRA = Burnable Poison Rod Assembly 

(d) RCCA = Rod Cluster Control Assembly 

(e) LPRM = Low Power Range Monitor 

(f) RWCU = Reactor Water Clean-Up 

Note: All DER sample data from NUREG/CR-6390 (Lepel, et al., 1995) 

* Arithmetic mean of all 22 values 

* * Geometric mean of all 22 values

41 NUREG/CR-6567
Determination and Review

105

Co-60 CONCENTRATION (uCi/g)

Correlation Between Mo-93 and Co-60 in Low-Level Wastes from Nuclear 

Power Stations

NUREG/CR-6567

10" 

10"

10"

z 

0) 

z 

0 

i

z z 

0 

C., 

0 

0 6

10"

10"

10"

Figure 7.1

42
Determination and Review

7.1.1.2 Niobium-93m 

The NRC rule 10 CFR Part 61 specifically identifies 

"

94

Nb as a reportable long-lived radionuclide in LLW. 

However, 10 CFR Part 61 does not mention 93 MNb, 

despite the fact that its concentration is over two 

orders of magnitude higher than 9Nb in neutron

activated metal waste. The reason 10 CFR Part 61 is 

more concerned about 9Nb is because of its longer 

half-life (20,000 years) compared to 9mNb (half-life 

= 16.1 years), and the higher energy gamma-ray 

emissions of 9Nb (871 and 703 keV). Niobium-93m 

decays by emission of a 30-keV gamma-ray which is 

essentially all internally converted. The predominant 

external radiation emitted by 93

mNb is due to the 16.5

keV Nb x-rays. Previous calculations of 94

Nb and 

9mNb in the neutron-activated shroud from a 

reference BWR estimated that the 93

mNb/ 94

Nb ratio 

would be 0.09 (Oak, Holder, Kennedy, and Konzek 

1980). However, more recent empirical 

measurements have shown that the average value of 

this ratio for irradiated stainless steel spent fuel 

assembly hardware is 158 (Robertson, Thomas, 

Wynhoff, and Hetzer 1991). Thus, the actual 

measured concentrations of 9

3mNb in neutron

activated stainless steel is 1800 times higher than 

previously calculated. 

Table 7.2 lists the measured activity scaling factors 

for the 93

mNb/ 6

°Co ratio in samples measured in this 

study, plus data from NUREG/CR-6390 (Lepel, et 

al., 1995). The geometric mean of this scaling factor 

for twenty different activated metal samples and two 

samples of spent ion exchange resins was 6.9 IE-4. 

Figure 7.2 shows a reasonable correlation between 

"

93

mNb and 'Co concentrations in the LLW materials 

listed in Table 7.2. Thus, a generic activity scaling 

factor for this radionuclide may be justified for 

nuclear power station LLW. 

7.1.1.3 Silver-108m 

Silver-108m is primarily produced in the Ag-Cd-In 

alloy used for neutron absorption in RCCA control 

rods from PWR stations. Silver-108m is generally 

not detectable in other types of LLW generated at 

nuclear power stations. As shown in Table 7.3 the 

1°osAg/ 6

oCo activity scaling factors for all types of 

LLW except the RCCA control rods are "less-than" 

values ranging from <1.1 E-2 to <3.4E-4. The AgCd-In alloy from the RCCA control rod measured in 

this study, together with earlier analyses published in 

NUREG/CR-6390 (Lepel, et al., 1995), show a rather 

consistent l°OmAeg6Co activity scaling factor 

averaging 8.45. As discussed in Section 5.1, the 

"IOSmAg is the most abundant radionuclide in this type 

of neutron activated material, and because it decays 

with the emission of several relatively high energy 

gamma rays it is quite easy to measure by direct 

gamma-ray spectrometry. Therefore, an activity 

scaling factor may not be necessary for this 

radionuclide because it can be measured directly. 

7.1.1.4 Cadmium-113m 

Cadmium- 113m is likewise produced in the Ag-Cd

In alloy used in RCCA control rods. Table 7.4 lists 

the I ICd/ 6

OCo activity scaling factors for a variety 

of LLW from nuclear power stations. The scaling 

factors for non-RCCA type materials are low and 

range from <5.OE-5 to 1.8E-3. Even the scaling 

factors for the Ag-Cd-In alloy materials are quite 

variable, and range from 0.226 to 5.69. This large 

variability may be due to either large ranges in cobalt 

impurities in the alloy, and/or differences in the 

neutron capture cross section versus neutron energy 

for the two parent isotopes. In any event, there does 

not appear to be a useable generic activity scaling 

factor (relative to 'Co) for this radionuclide based on 

the limited data base of measurements in LLW. A 

more useful scaling factor for the Ag-Cd-In alloy 

material appears to be the lI3

mCd/losmAg ratio, since 

the losmAg can be measured easily by direct gamma

ray spectrometry in this material. 

7.1.1.5 Tin-121m 

Tin- 121 im was not detectable in any of the activated 

metal or spent resin samples measured in this study 

(see Table 7.5). For the five LLW sample types 

analyzed, an average 12mSn/6OCo activity scaling 

factor of<5.OE-5 was determined. However, 

because of the very low concentrations of 12lmSn in 

LLW, the need for a generic activity scaling factor 

would be very limited.

43 NUREG/CR-6567
Determination and Review

Table 7.2 93

mNb/°Co Scaling Factors for Low-Level Waste

Scaling 

Concentration (ICi/g)(a) Factor 

ID No. Sample Material 93mlb 60

Co 9

3mNb/ 6OCo 

DER-1 Duane Arnold SS 45.6 1.08E-4 4.22E-3 

CRBýb) 

DER-1Z SS 73.6 8.37E-3 8.75E-3 

DER-2 SS 1.32 7.87E-3 1.68E-4 

DER-2Z SS 3.17 4.79E-3 6.62E-4 

DER-3 SS 2.94E-2 2.03E-2 1.44E-4 

DER-3Z SS 4.74E-2 1.14E-2 4.16E-4 

DER-4(out) Point Beach SS 2.31 1.49E-4 1.55E-4 

BPRA(c) 

DER-4(mn) SS 5.13 1.80E-3 2.85E-3 

DER-5(out) SS 2.86 1.79E-4 1.60E-4 

DER-5(in) SS 6.26 2.20E-3 2.85E-3 

DER-6(out) SS 2.84 1.70E-4 1.67E-4 

DER-6(in) SS 5.66 2.01E-3 2.82E-3 

DER-10A Point Beach SS 41.4 4.54E-4 9.12E-4 

RCCA(d) 

DER-10B SS 24.0 5.69E-4 4.22E-4 

DER-8 SS 1.37 9.85E-3 1.39E-4 

#39 Peach Bottom SS 2.40E-4 0.371 6.47E-4 

LPRM(e) 

#43 Duane Arnold SS 0.503 2.12E-3 2.37E-4 

CRB 

#40 Peach Bottom SS 2.20E-3 0.484 4.55E-3 

CRB 

#41 Point Beach Inconel 0.256 15.2 1.68E-2 

BPRA 

(Spring)

NUREG/CR-6567 44
Determination and Review

Table 7.2 (continued)

Scaling 

Concentration (pCi/g)(a) Factor 

ID No. Sample Material 93

mNb 6 0

Co 9

3mNb/6OCo 

#42 Point Beach Ag-Cd-In 4.56E-3 45.0 1.01E-4 

RCCA 

#23 Cyrstal River Resin 4.20E-3 4.36 9.63E-4 

RWCU 

Resin (f 

#22 WNP-2 Resin Resin 4.28E-3 20.8 2.06E-4 

Mean 2.20E-3 

Standard 3.90E-3 

Deviation 

Geometric 6.91E-4 

Mean * 

(a) Concentrations decay-corrected to 5 years after discharge of components or materials from the 

reactor. It was necessary to sandardize the decay time after discharge from the reactors because 

some of the materials were quite old and the 'Co had undergone substantial radioactive decay. 

(b) CRB = Control Rod Blade 

(c) BPRA = Burnable Poison Rod Assembly 

(d) RCCA = Rod Cluster Control Assembly 

(e) LPRM = Low Power Range Monitor 

(f) RWCU = Reactor Water Clean-Up 

Note: All DER sample data from NUREG/CR-6390 (Lepel, et al., 1995) 

, Arithmetic mean of all 22 samples 

, * Geometric mean of all 22 samples

NUREG/CR-6567 45
Determination and Review

10 2 

y = 1.6692e-3 xA0.90840 RA2 = 0.816 2 

[]l 

100 

S0 

n-n 

10" 

z 0 0 

10 z 

LU 

z 

0 

1 0 " 

E 

z 

10-"3 

10-41 

1 0- 1' ' '"1 0 0 , ' ,,,1 01 ' '"1 02 ' ' ""1 03 ' ""1 04 ' " "1 05 

Co-60 CONCENTRATION (uCilg) 

Figure 7.2 Correlation Between Nb-93m and Co-60 in Low-Level Wastes from Nuclear

Power Stations

NUREG/CR-6567 46
Determination and Review 

Table 7.3 Silver-108m and "0Co Concentrations in Selected LLW Samples 

Scaling 

Concentration(ICi/g)* Factor 

Sample Material Source losmAg 60Co 10smAg/60 0o 

#39 stainless steel LPRM Cladding(a <2.4E-4 3.71E-1 <6.5E-4 

Peach Bottom 

#40 stainless steel Control Rod Blade <6.5E-4 4.84E-1 <1.34E-3 

Peach Bottom 

#43 stainless steel Control Rod Blade <1.4 2.12E-3 <6.6E-4 

Duane Arnold 

#41 Inconel spring BPRA Control <5.1E-3 1.52E-1 <3.4E-4 Rod•) 

Point Beach 

#42 Ag-Cd-In Alloy RCCA Control 6.37 ± 0.19E2 4.50E-1 14.2 

Rod(') 

Point Beach 

DER-7 Ag-Cd-In Alloy RCCA Control Rod 5.44E-3 5.75E-4 9.46 

Point Beach 

DER-8 Ag-Cd-In Alloy RCCA Control Rod 1.12E-2 1.46E-3 7.67 

Point Beach 

DER-9 Ag-Cd-In Alloy RCCA Control Rod 1.11 1.06E-1 10.5 

Point Beach 

DER-10A Ag-Cd-In Alloy RCCA Control Rod 2.14E-4 2.52E-3 8.49 

Point Beach 

DER-10B Ag-Cd-In Alloy RCCA Control Rod 2.73E-4 4.46E-3 6.12 

Point Beach 

#23 ion exchange resin RWCU Resin(d) <5.0E-2 4.36E-0 <1.1E-2 

Crystal River 

#22 ion exchange resin EDRFDR + <1.3E-2 2.08E-1 <6.3E-4 

RWCU(e) 

WNP-2 

* Concentrations decay corrected to 5 years after discharge of components or materials from reactors 

(a) LPRM = Low Power Range Monitor 

(b) BPRA = Burnable Poison Rod Assembly 

(c) RCCA = Rod Cluster Control Assembly 

(d) RWCU = Reactor Water clean-Up 

(e) EDRFDR = Equipment Drain Radioactive/Floor Drain Radioactive

47 NUREG/CR-6567
Determination and Review

Table 7.4 Cadmium-113m and "°Co Concentrations in Selected LLW Samples 

Scaling 

Concentration(gCi/g)* Factor 

Sample Material Source 113mCd 60Co ll 3

mCd/f°Co 

#39 stainless LPRM Claddingwa <1.2E-4 3.71E-1 <3.3E-4 

steel Peach Bottom 

#43 stainless Control Rod Blade 1.06E-1 2.12E-3 5.OE-5 

steel Duane Arnold 

#41 Inconel BPRA Control Rod() 2.76E-2 1.52E-1 1.8E-3 

spring Point Beach 

#42 Ag-Cd-In RCCA Control Rod~c) 28.0 ± 0.1 4.5E-1 6.23E-1 

Alloy Point Beach 

DER-9 Ag-Cd-In RCCA Control Rod() 0.603 1.06E-1 5.69 

Alloy Point Beach 

DER- Ag-Cd-In RCCA Control Rod(c) 577 2.55E-3 0.226 

10A Alloy Point Beach 

DER- Ag-Cd-In RCCA Control Rod(c) 1020 4.51E-3 0.226 

10B Alloy Point Beach 

#23 ion RWCU Resin(d) 4.9 + 1.OE-3 4.36 1.1E-3 

exchange Crystal River 

resin 

#22 ion EDRFDR + RWCU(e) 1.54 + 0.5E-2 2.08E-1 7.40E-4 

exchange WNP-2 

resin 

Concentrations decay corrected to 5 years after discharge of components or materials from 

reactors 

(a) LPRM = Low Power Range Monitor 

(b) BPRA = Burnable Poison Rod Assembly 

(c) RCCA = Rod Cluster Control Assembly 

(d) RWCU = Reactor Water Clean-Up 

(e) EDRFDR = Equipment Drain Radioactive/Floor Drain Radioactive

NUREG/CR-6567 48
Determination and Review

7.1.2 Chlorine-36

Chlorine-36 was detectable in only the spent ion 

exchange resin samples from the two PWR nuclear 

power stations (see Table 7.6). The average 36

CI/6VCo 

activity scaling factor for these two samples was 

2.65E-4. The 3

6

C1/' 37

Cs activity scaling factor ranged 

from 1.43E-6 to 1.02E-5 and averaged 5.8E-6. This 

value is in good agreement with the limited data 

reported for spent primary resin samples from Swiss 

PWR and BWR nuclear power stations (Liu, et al., 

1991). Although the 36

C1 concentrations in spent 

primary resins is quite low, its very long half-life, 

high environmental mobility, and relatively high 

ingestion dose conversion factor may require that a 

more detailed radiological performance assessment 

for its disposal in LLW be conducted. 

7.1.3 Beryllium-lO 

Beryllium-10 is produced in LLW from nuclear 

power stations as a by-product from neutron 

activation of 1

0

B present in control rods (as B4C and 

borosilicate glass) and in primary coolant (as soluble 

borate). Therefore, "

0

Be is a significant contaminant 

in spent control rods and primary coolant 

demineralization resins. The concentration of soluble 

boron in PWR primary coolant can range from 10 to 

2300 ppm B depending on the burnup in the fuel 

cycle. Therefore, the production of '

0Be from 

neutron activation of `

0B can vary by a factor of up to 

230, and 1

0

Be concentrations in primary coolant (and 

primary demineralization resins) would also be 

expected to be quite variable. 

Table 7.7 lists the '

0

Be/6'Co activity scaling factors 

for activated metal and spent resin LLW samples. 

Beryllium- 10 concentrations in the activated metal 

samples were non-detectable, except for a trace 

amount in the Inconel speciman (Sample #41). Of 

the resin samples, only the Diablo Canyon sample 

(#2-2) contained detectable "

0

Be. As shown in Figure 

5.6, the "°Be in this spent resin will be the third most 

abundant radionuclide (behind `'C and 59Ni) after a 

disposal period of about 600 years. The '

0

Be/°Co 

activity scaling factor for the Diablo Canyon resin 

sample was 9.45E-4. However, this scaling factor 

would be expected to be quite variable depending on 

the burnup in the fuel cycle, and should not be

considered representative of a generic scaling factor 

for spent resin LLW. 

The '°Be/°Co scaling factors for the boron carbide 

present in the two samples of the spent Duane Arnold 

control rod blade were less variable and averaged 

5.38E-3. The scaling factor for the borosilicate glass 

in the BPRA rod from Point Beach was 5.3 1E-4. 

These scaling factors are probably fairly 

representative of these materials, and may be useful 

for estimating "

0

Be concentrations in these materials 

by gamma counting for 'Co. These are the only 

known measured '

0

Be/60Co scaling factors in 

existence. 

Because of the very long half-life of "

0

Be (1.6E6 

years), together with its relatively low soil adsorption 

in sandy types of soil and its relatively high ingestion 

dose conversion factor, it is recommended that 

further radiological assessment be conducted to better 

predict its potential radiation dose to man from 

disposal in LLW. 

7.2 Review of Scaling Factors for 

14

C, 99

Tc, and 1291 for Low

Level Waste 

Recent assessments of shipping manifest data for 

LLW being disposed from commercial nuclear power 

stations have shown that the concentrations of 1291 

and 9Tc have been conservatively overestimated by 

several orders of magnitude. Anomalies in the `

4C 

data base were also observed. Bhattacharyya, Janati, 

and Shearer (1987) were the first to point out that the 

nuclear power industry's radionuclide source term 

data base is overly conservative with respect to the 

concentrations of I

29

I in LLW generated by nuclear 

power stations. The shipping manifests for LLW 

generated at nuclear power stations normally report 

"less-than" values for 1291 and 19Tc which are supplied 

by service analytical laboratories who perform the 

required analyses for the utilities to meet 10 CFR Part 

61 requirements for waste classification. The 

reported "less-than" values are simply the lower

limit-of-detection of the radiometric methods used by 

the analytical laboratories. Iodine-129 is a low

energy beta and photon emitter, and 9Tc and I4C are

49 NUREG/CR-6567
Determination and Review

Table 7.5 Tin-121m and 6

°Co Concentrations in Selected LLW Samples 

Scaling 

ConcentrationItCi/g)* Factor 

Sample Material Source l

2

lmSn 6°Co 1l2mSnI°Co 

#40 stainless Control Rod <2.1E-4 4.84E-1 <4.3E-4 

steel Blade 

Peach Bottom 

#41 Inconel BPRA Control <9.9E-5 1.52E-1 <6.5E-6 

spring Rod(a) 

Point Beach 

#42 Ag-Cd-In RCCA Control <1.5E-3 4.50E-1 <3.3E-5 

Alloy Rod.) 

Point Beach 

#23 ion RWCU Resin(c) <3.9E-4 4.36E-0 <8.9E-5 

exchange Crystal River 

resin 

#22 ion EDRFDR + <5.2E-3 2.08E-1 <2.5E-4 

exchange RWCU(d) 

resin WNP-2 

* Concentrations decay corrected to 5 years after discharge of components or materials 

from reactors 

(a) BPRA = Burnable Poison Rod Assembly 

(b) RCCA = Rod Cluster Control Assembly 

(c) RWCU = Reactor Water Clean-Up 

(d) EDRFDR = Equipment Drain Radioactive/Floor Drain Radioactive

NUREG/CR-6567 50
Determination and Review

Table 7.6 Chlorine-36 and 6

°Co Concentrations in Selected LLW Samples 

Scaling 

Concentration(zCi/g)* Factor 

Sample Material Source 36C1 'Co 36CI/6

°Co 

2-2 ion RWCU resin(a) 9.3 ± 2.2E-5 8.70 1.1E-5 

exchange Diablo Canyon 

resin 

#23 ion RWCU resin 2.3 1 1.1E-3 4.36 5.3E-4 

exchange Crystal River 

resin 

#22 ion EDRFDR + <7.9E-4 20.8 <3.8E-5 

exchange RWCU'b) resin WNP-2 

#39 stainless LPRM <3.7E-5 3.71E-1 <9.9E-5 

steel Claddinge• 

Peach Bottom 

#40 stainless Control Rod <7.2E-5 4.84E-1 <1.5E-4 

steel Blade 

Peach Bottom 

#43 stainless Control Rod <1.8E-3 2.23E-3 <8.1E-7 

steel Blade 

Duane Arnold 

#41 Inconel BPRA Control 3.9 ± 0.9E-4 1.52E-1 2.6E-5 

spring Rod(d) 

Point Beach 

* Concentrations decay correted to 5 years after discharge of components or materials 

from reactors

(a) 

(b) 

(c) 

(d)

RWCU = Reactor Water Clean-Up 

EDRFDR = Equipment Drain Radioactive/Floor Drain Radioactive 

LPRM = Low Power Range Monitor 

Burnable Poison Rod Assembly

51 NUREG/CR-6567
Determination and Review 

Table 7.7 Beryllium-10 and "0

Co Concentrations in Selected LLW Samples 

Concentration(gCi/g)* Scaling Factor 

Sample Material Source 1

OBe 60Co '

0Be/6 0Co 

DER-1 boron Spent BWR 3.68 ± 10.2 3.61E-3 

carbide Control Rod Blade 0.04E-2 

Duane Arnold 

DER-2 boron Spent BWR 3.91 ± 5.47 7.15E-3 

carbide Control Rod Blade 0.04E-2 

Duane Arnold 

DER-4 borosilicate Spent PWR BPRA 2.72 + 5.12 5.31E-4 

glass Rod(a) 0.08E-3 

Point Beach 

2-2 ion RWCU Resincb) 8.22 + 8.70 9.45E-4 

exchange Diablo Canyon 0.36E-3 

resin 

#23 ion RWCU Resin <1.08E-4 4.36 <2.5E-5 

exchange Crystal River 

resin 

#22 ion EDRFDR + <7.9E-4 20.8 <3.8E-5 

exchange RWCU(c) 

resin WNP-2 

#39 stainless LPRM <5.3E-5 3.71E-1 <1.4E-4 

steel Claddinged) Peach 

Bottom 

#40 stainless Control Rod Blade <8.2E-5 4.84E-1 <1.7E-4 

steel Peach Bottom 

#43 stainless Control Rod Blade <2.2E-3 2.23E-3 <9.9E-7 

steel Duane Arnold 

#41 Inconel BPRA Control Rod 1.20 + 15.2 7.9E-6 

spring Point Beach 0.05E-4 

#42 Ag-Cd-In RCCA Control <1.6E-4 45.0 <3.6E-6 

Alloy Rodle) 

Point Beach 

* Concentrations decay corrected to 5 years after discharge of components or materials 

from reactors 

(a) BPRA + Burnable Poison Rod Assembly 

(b) RWCU = Reactor Water Clean-Up 

(c) EDRFDR = Equipment Drain Radioactive/Floor Drain Radioactive 

(d) LPRM = Low Power Range Monitor 

(e) RCCA = Rod Cluster Control Assembly 

NUREG/CR-6567 52
Determination and Review

pure beta emitters. Their analyses require 

radiochemical purification followed by either beta or 

low-energy photon counting. 

The usual practice by the nuclear power industry is 

to send a limited number of selected LLW samples, 

on an annual or semi-annual basis, to the service 

analytical laboratories for comprehensive 10 CFR 

Part 61 radionuclide analyses. Then each nuclear 

power station generates an updated, station-specific 

data base from which activity scaling factors for the 

difficult-to-measure radionuclides are calculated 

relative to easily measured gamma-emitting 

radionuclides (i.e., 137Cs and 60

Co) usually present in 

the LLW. Then, the concentrations of the difficult

to-measure radionuclides can be estimated in batches 

of LLW by measuring the gamma-emitting 

radionuclides and applying the activity scaling 

factors developed for various waste types at each 

station. This procedure is entirely adequate for 

waste classification, and the reported "less than" 

concentrations are generally orders of magnitude 

below the least-restrictive Class A waste category. 

However, the true concentrations of "291 and 9Tc in 

the LLW are generally several orders of magnitude 

below the "less-than" values reported by the 

analytical laboratories. 

The problem associated with this methodology came 

later, when performance assessment modeling for 

future LLW disposal sites, using the highly 

conservative industry data base consisting primarily 

of "less-than" values for 129I in LLW, indicated that 

the 1291 releases from the site would exceed the dose 

limits specified by 10 CFR Part 61, and thereby limit 

the quantities of LLW that could be disposed at a 

shallow-land burial facility. The obvious need of the 

nuclear power industry is to determine the actual 

concentrations of 1291 and 99

Tc in LLW by an 

analytical method, such as mass spectrometry, which 

possesses adequate sensitivity to measure the actual 

concentrations of these nuclides. Using a more 

accurate data base in performance assessment 

modeling would eliminate these two radionuclides as 

major contributors to the offsite dose from a LLW 

disposal facility.

7.2.1 Recent Mass Spectrometric 

Measurements of 1291 and 99Tc in 

LLW 

During the past several years PNNL has conducted a 

number of programs for EPRI, NRC, and other 

clients (Vance and Associates, as well as a number of 

nuclear utilities) during which a variety of LLW 

samples from commercial nuclear power stations 

have been analyzed for 1291 and 9Tc by extremely 

sensitive mass spectrometric methods. In addition, 

the gamma-emitting radionuclides that are used for 

developing activity scaling factors were measured in 

the same samples, i.e., 13'Cs and 'Co. These 

measurements have provided the most accurate data 

base to date of these activity scaling factors. Because 

of the large differences between the two data bases, 

the dissemination of the accurate mass spectrometric 

data would benefit the nuclear utilities, state and 

federal regulators, performance assessment modelers, 

radioactive waste managers, and LLW disposal 

facility operators. 

7.2.1.1 Sampling 

A variety of LLW sample types from a number of 

commercial nuclear power stations were collected for 

analysis over the past several years. Because the 

reactor primary coolant is the main source of 129I, 

9Tc, and other fission and activation products in 

LLW, a concerted effort was made to characterize the 

concentrations of these radionuclides in mixed-bed 

ion exchange resins used for demineralization of the 

primary coolant. To more carefully track the 

behavior of these radionuclides on mixed-bed ion 

exchange resin, scaled-down versions of the reactor 

primary coolant demineralizer systems were 

constructed, installed, and operated at PWR and 

BWR stations to sample the primary coolant. 

For PWR stations, the scaled-down primary 

purification demineralizer systems were placed in

line in the primary coolant sampling line, and 

primary coolant was passed through the ion exchange 

resin column for periods ranging from several days to 

up to six weeks to simulate the cleanup operations 

and resin loading at a commercial PWR station. The 

test column consisted of a 40-cm long by 2.5-cm 

inside diameter Plexiglass tube with stainless steel

53 NUREG/CR-6567
Determination and Review

screens placed on each end to contain the resin. The 

column was filled with about 30 cm of the standard 

mixed-bed resin routinely used at each station for 

cleanup of the primary coolant. The primary coolant 

flow rate through the column was scaled-down 

proportionally to the size of the column and typically 

was about 80 mL/min. Columns were installed at a 

number of stations to obtain an adequate data base 

for determining the concentrations and distributions 

of the isotopes on the resin and for computing 

reliable activity scaling factors. After processing the 

primary coolant, the column was disconnected and 

carefully packaged for shipping to PNNL for 

radiochemical and mass spectrometric analysis of the 

resin samples. 

For BWR stations, the typical powdered resin used 

for primary coolant cleanup was utilized in a scaled

down, in-line filter assembly to simulate the coolant 

cleanup process used at BWR stations. A 2-cm 

diameter Millipore stainless steel in-line filter holder 

was loaded with about 1.2 g of powdered resin to 

form a 0.6-cm thick bed, and reactor coolant was 

pumped through the resin bed at a flow rate of 42 

mL/min for 48 hours. The resin holder was then 

disconnected, packaged, and shipped to PNNL for 

radionuclide analysis of the resin. 

In addition to the test resin samples, a variety of 

LLW samples were obtained from nuclear power 

stations, including primary and secondary resins, 

primary coolant water, dry active waste (DAW), 

primary coolant purification filters, charcoal, soil, 

and spent oil, for a total of 43 samples. 

7.2.1.2 Radiochemical and Mass 

Spectrometric Analyses for "29I and "Tc 

in LLW 

The samples were first homogenized and appropriate 

aliquots taken for direct gamma-ray spectrometric 

analyses to measure the gamma-emitting 

radionuclides. Then the gamma-counted aliquots 

were subjected to destructive radiochemical analyses 

of 1291 and 9Tc, and quantified by two mass 

spectrometric methods. 

The 1291 was separated from the sample matrix by 

controlled combustion in oxygen, trapping the 

volatilized iodine on charcoal, and further purifying

the iodine prior to loading the iodine (as silver 

iodide) on a lanthanum hexaboride-coated rhenium 

filament for thermal emission mass spectrometric 

analysis (Strebin, Brauer, Kaye, Rapids, and Stoffels 

1988). The detection limit afforded by this method is 

about 4 x 10" pzCi of 29I, which provides adequate 

sensitivity to accurately measure "29I in essentially 

any type of LLW sample from the nuclear power 

stations. 

The 99

Tc was separated from the sample matrix by 

leaching with hot 10 M nitric acid, evaporating the 

acid to near dryness, dissolving the residue in 0.01 M 

nitric acid, and purifying the Tc by an ion exchange 

procedure (Robertson, Smith, Koppenaal 1990). The 

final separated Tc was dissolved in 1.0 mL of 1 M 

nitric acid which was then submitted for 99

Tc analysis 

by inductively coupled plasma mass spectrometry 

(Smith, Wyse, and Koppenaal 1992). This method 

provided a detection limit of about I x 10.8 uCi of 99

Tc. 

7.2.1.3 Results of Mass Spectrometric Analyses 

of 1291 and 19Tc in LLW 

The results of the gamma-spectrometric and mass 

spectrometric analyses of the test resin samples and 

the LLW samples are presented in Table 7.8, which 

lists the concentrations of 129I, 99

Tc, and the gamma

emitters "'Cs and 6

"Co that are used for developing 

activity scaling factors. Also listed are the computed 

activity scaling factors for 129I and 9Tc relative to the 

gamma-emitters. Table 7.8 was compiled 

frommeasurements made during this study, from 

measurements published in Vance, et al. 1992, in 

Robertson, et al. 1990, 1997, and from previously 

unpublished measurements made at PNNL. 

Because of the diverse nature of the LLW samples 

from a number of PWR and BWR stations which 

were analyzed, the concentrations of these nuclides 

ranged over four to seven orders of magnitude. The 

highest concentrations were generally observed in 

primary demineralization resins which remove these 

radionuclides from the reactor coolant, and the 

concentrations of 129I and 99Tc varied by about three 

orders of magnitude in these resins. The primary 

coolant cleanup filters contained the highest 

concentrations of 'Co and other activation products, 

but not necessarily the soluble fission products, such

NUREG/CR-6567 54
Determination and Review

as 1291 and 9Tc. 

7.2.1.4 Correlation of 1291 and "Tc with 

Gamma-Emitting Radionuclides 

The radionuclide concentrations generated in the 

reactor primary coolant (the source-term for LLW 

radionuclides) are dependant on a number of factors, 

including the amount of tramp uranium on the fuel 

element surfaces, the degree of fuel element integrity, 

the elemental composition of the materials of 

construction of the primary loop system, the 

corrosion rates of primary loop materials, and the 

reactor coolant chemistry. Because these factors can 

be highly variable from station to station, the relative 

concentrations of fission and activation products in 

the primary coolant can also vary significantly. 

Iodine-129 and "'Cs are strictly fission products 

which are relatively soluble in~the reactor primary 

coolant, and 6°Co is strictly an activation product of 

limited solubility. For this reason, and despite the 

fact that 129I is present in reactor coolant as an anion 

and '"Cs is present as a cation, 129! is usually 

correlated with "'Cs to provide activity scaling 

factors for estimating the quantities of 129I from the 

117Cs concentrations in LLW measured by gamma

ray spectrometry. 

In addition to being a fission product, 99

Tc is also 

produced by thermal neutron activation of 

molybdenum, and enters the primary coolant by 

corrosion/dissolution of the molybdenum-containing 

construction materials (e.g., stainless steel, some 

Inconel and titanium alloys) present in the primary 

coolant system. At some nuclear power stations the 

quantities of 9Tc produced by neutron activation of 

molybdenum is quite significant relative to fission 

product 99

Tc. 

Figures 7.3a and 7.3b show the correlations of 129I 

with 1'"Cs and with 6

"Co for all resin and LLW 

samples. These two radionuclides are the most 

commonly used gamma-emitters utilized for 

computing scaling factors. As the scatter diagrams 

show, the correlation between 1291 and '"Cs is quite 

good, even for diverse sample types having 

radionuclide concentrations ranging over eight orders 

ofinagnitude. The regression analysis for this data 

set gave an R2 value of 0.852. Thus, the use of

1

29

1/13 7

Cs scaling factors for all types of LLW appears 

to be a reasonable method for estimating the 

concentrations of 1291. An assessment of the 

correlations with respect to sample type, reactor 

station type (PWR and BWR), and sample activity 

levels showed no definite trends, except for the 

observation that the 1291 and "'Cs correlations were 

much better for PWR stations compared to BWR 

stations. The reason for this anomaly may be due to 

the fact that in BWR stations approximately two

thirds of the reactor coolant iodine can be transported 

by the steam to the condensate resin while most of 

the cesium would be expected to be removed by the 

primary demineralization system (Vance, Helmholz, 

and Sutherland 1988). As might be expected, the 

correlation of 129I with 'Co was quite poor due to 

their different production modes and physico

chemical behaviors in reactor coolant. 

Figures 7.4a and 7.4b show the correlations of 9Tc 

with 60

Co and "'Cs. These data show a reasonable 

correlation of 99

Tc with 6

°Co (R2 value of 0.631), but 

essentially no correlation with 1'"Cs. These 

observations are not entirely unexpected, and suggest 

that the most significant production mode for 9Tc at 

many nuclear power stations may be by neutron 

activation of molybdenum. Technetium has normally 

been assumed to behave as the soluble Tc(VII) 

pertechnetate anion in reactor coolant. However, this 

assumption may not be entirely correct, particularly 

in view of its apparent correlation with 'Co, which 

suggests that much of the technetium may be 

presence in reactor coolant in a less-soluble reduced 

oxidation state. For those nuclear power stations 

experiencing a relatively high rate of fuel cladding 

failures, the 9Tc/137

Cs scaling factor would be 

expected to show less variability. 

7.2.1.5 Comparison of PNNL and Industry 

Scaling Factors 

The activity scaling factors for 1291 and 99Tc 

computed from the PNNL data base in which highly 

sensitive and selective mass spectrometric 

measurement techniques were utilized were much 

lower than the scaling factors developed from the 

nuclear power industry data base. The industry data 

bases contain LLW radionuclide concentrations and 

activity scaling factors for most of the nuclear power 

stations in the U. S., and have been published in the

55 NUREG/CR-6567
Table 7.8 Radionuclide Concentrations and Activity Scaling Factors for LLW from Commercial Nuclear Power Stations 

SAMPLETYPE IDNUMBER SAMPLE DATE 1-129 Bq/g Tc-99 Bq/ __ -13 0-60 Bq i-129/c- T-129/Co-60 s-137 Tc-99/C0-60 

s prmry resin _PWR Sta. A 7-Dec-59 1,025+01t_ 1.516+01 1.37+08 3.07E+05 7.47E-08 1.I0E-07 3.33E-05 4.90.E-05 tA _ primay resin PWR Ste. B 5-Nov-90 2,46E+00 9.77E-01 4.26E+07 5.51E+05 5.76E-08 2,29E-08 4.47E-06 12785-06 

-prmrym resin PWRiSla. C 13-Feb-90 9.106-03 1.55E-.01 8.66E+04 6.77E+05 1.05E-07_ . _.79E-06- 1,_I_35E-08 2.29E-07 . 

primary resin PWR Sta. D 13-Feb-90 I*IOE+00 2.80E-01 8.29E+06 3,23E+04 __1.33E-07 . . 3.38E-08 3.4 1E-05 8.67E-06_ 

i__prmary resin PWR Sta. E 21-Jan-90 l.09E-01 6.36E-02 2.56E+06 6.96E+03 4.26E-08 2.48E-08 1.575-05 9.15E-06 

prima!ry resin PWR Sta. F 19-Au -90 1.79E-01 1.62E-01 3.33E+05 _ 7.66E+03 5.38E-07 4.85E-07 2.34E-05 2.11 E-05 

__prmmary resin PWR Sta. G 6-Jul-92 5.40E-03 2.33E-02 __5.03E+04 __5.40E+03 1.07E-07 4.63E-07_ 1.ooE-06 4.32E-06 

-. primary resin PWR Sta. B 5-Feb-92 3.25E-02 9,62E-03 2.70E+06 2.33E+03 1.20E-08 3.56E-09 1.40E-05 4.13E-06 

.. prma_•y resin PWRSa. B. 30-Jan-92 2.01E-02 1.78E-02 2.18E+06 1.41E+03 9.19E-09 . 8.14E-09• . .43E-05 1.26E-05 - qpTrimaryesin PWR Sta. C 15-Jan-92 1.47E-02 1.96E-02 3.30E+06 5.40E+04 4.44E-09 5.94E-09 2.71E-07 3.635-07 

prmtiry resin PWR Sta, D 15-Apr-92 7.55E-02 1.70E-02 4.92E+04 3.04E+04 1.53E-06 3.46E-07 2.48E-06 5.60E-07 

p rmry _resin PWR Sta. H 9-Nov-90 7.81E-03 3.66E-02 6.81 E+05 6.55E+04 1.15E-08 5.38E-08 1,19E-07 5.59E-07 

__prm _resin ___PWRSta.i 29-Nov-90 3.52E-02 7.03E-02 9.07E+04 4.44E+04 ----- 3.89E-07 7.76E-07 7.93E-07 1.58E-06 

primary resin PWR Sta. J 14-Dec-90 1.49E-02 4.44E-02 3.92.+06 5.66E+04 3.80E-09 1.13E-08 2_63E-07 7,84E07 ____prima. resin PWR Sta. K 10-Apr-96 1.18E+00 2.66E.-01 2.37E+07 6.07E+06 ... 4.98E-08 1.12E-08 1.95E-07 4.39E-08 

-. prlmary resin PWR Sta. K 1-Jan-96 4.26E-03 5.92E+00 1.48E+04 6.77E+06 2.88E-07 4.OOE-v4 6.28E-10 8.74E-07 

- _p__r!mary res!n PWRSta. D_ 22-Jan-86 3.32E-02 0.OOE+00 6.51E+06 5.07E+05 5.09E-09 6.54E-08 

___pri.ma rrrry r esin WR StO. I 12-Apr-90 1.305-01 1.98E+01 6.59E+05. . 2.21E+06 1.97E-07 3.01 E-05 5.85E-08 8.95E-.06 

__prirm ary resin BWR Sta. 2 21-Mar-90 3.64E-01 1.49E+01 1.75E+06 1.45E+07 2.08E-07 8.545-06 2.51E-08 1,03E-06 

__ pqprimay resin WR SIa. 3 26-Feb-90 7.40E-02 1.31E+00 1.80E+04 3.16E+06 4.115-06 7.27E-05 2.34E-08 4.1 4E-07 ___pr!mary resi n BWARSta.4 26-Feb-90 7.40E-02 1.11E+00 3.41E+06 5.13E-05 7.44E+03 7.725-04 2.17E-08 3.27E-07 

, •prirm_.a resin BWR Sta. 5 _ -Jul-92 2.01 E-04 4.445-01 0.70E+02 4.92E+05 2.31"E-07 5.11E-04 4.086-10 9.021-07 __primary resin BWR Sta .6 24-Jan-92 2,80E-04 1.866+01 1.04E+03 1.035+06 2.70E-07 1.80E-02 2,72E-10 1,816-05 

primary9sln . BWR Sta. 7 13-Sep-90 4.81E-04_ __9.18E+00 1.41E+03 1.09E+06 3.42E-07 6.53E-03 4.42E-10 8.446-06 

primary resin DWR Sta. 8 26-Oct-90 3.24E-04 2.615+00 1.55E+03 1.91E+06 2.08E-07 1.661E-03 1.69E-10 1.36E-06 

___-__pmary resin BWR Sta-.2 22-Jan-92 3.19E-03 _ 3.07E+00 1.26E+04 5.07E+06 2.53E-07 2.44E-04 6.286-10 6.06E-07 

_ primaryresin Ste. .WR 5 6-Jul-92 3.32E-04 2.81 E-01 8.51E+03 7.40E+06 3.90E-08____ 3.30E-05 4.48E-- 11 3.80E-08 

beboratgdemIn. resin PWR Sta. K 1-Jan-96 4.77E-04 4.815+00 5,55E+03 1.16E+07 8.60E-08 8.67E-04 4.12E-1I 4.15E-07 

CVCS mixed demln. resin PWR Sta. K 1-Jan-96 __ 4.26E-03 _ 5.92E+00 1.48E+04 6.73E+06 2.88E07- 4.00-E04 6.32E- 10 8.791-07u 

____dry_ active waste PWR Sta. L 12-Jun-09 1.23E-04 0.00+E00 1.45E+03 6.25E+03 8.52E-08 1.97E-08 

dry_ active waste PWRSiaM 12-Jun-09__, 1.46E-04 0.00E+00 1.24E+03 3.896+02 1.17E-07 3.75E-07 

____dry active waste PWR Sta. C 12-Jun-09 1.15E-04 0.00r+00 2,29E+03 3.02E+03 5.01 E-08 3,80E-08 

___dry active waste BWR Sta. 9 12-Jun-09 3.40E-03 0.0011+00 3.186+03 5.92E+03 1.07E-06 5.74E-07 

dry. active waste SWR Sta, 10 12-Jun-09 6.66E-05 0.00E+00 8.84E+01 1.17E+03 7.53E-07 5.71E-08 

CVCS Filter PWR Sta, D 12-Jun-09 5155E-02 0.00E+00 1.97E+06 9.36E+08 2.82E-08 5.936-11 

RCS letdown tilter PWR Sta. K 12-Apr-96 1. 11E-02 1.11E+00 1.26E+05 4.40E+07 8.856-08 8.82E-06 2.52E-10 2.516-08 

RF water pu-t, tiller PWRSa,K . 25.Apt-_96 1.965-03 8.14E+00 1.04E+05 9.62E+06 1.89E-08 7.86E-05 2.036-10 8.46E-07 

CCS tilter PWR Sta, K 25-My_-9_5 7,55E-04 5.92E+01 9.25E+02 8.586+07 8.16E-07 6.40E-02 8.79E-12 6.906-07 

charcoal PWRSta. C 12-Jun-09 3.89E-03 1.10E+05 2.09E+06 3,52E-08 1.86E-09 

secondary mixedresin PWR Sta. N 12-Jun-09 1.22E-04 1.10E+03 6.99E+03 1.11E-07 1.74E-08 

soil BWR Ste. 11 _12-Jun-09- 2.81E-06 1.38E+01 1.20E+01 2.033E-07 ... .. 2.34E-07 

spent lubricating oil BWR Sta. 10 12-Jun-09 3.25&-04 2.066+03 1.816+03 1,58E-07. 1.80E-07 

reactor coolant PWRSta.tC 26-May-88 4.18E-07 1.41E+01 3,62E+00 2.97E-08 1.152-07 

RWCU prLmaý_yesin BWA Sta. 2 20-Apr-93 6.07E-02 2.286+05 2.74E-07 

primay demln. resin PWR Sta. D 10-Nov-92 2.41 E+01 _____ __ 8.256+07 2.92E-07 

aeometrc mean 1A20-07 9.231-06 3762-08 1.67-06 ~ ~o tk . . .. ... . ..... .. . . .. . .. . . .
Determination and Review

Cs-137 CONCENTRATION (Bq/g)

102 

101

Figures 7.3a, 7.3b

Co-6 CONCENTRATION (Bq/g) 

Correlation of 1-129 with Cs-137 (a) and 1-129 with Co-60 (b) in 

Commercial Nuclear Power Station Low-Level Wastes 

57 NUREG/CR-6567

z 

0 

F 

z 

o L

0 

z 

0 U9

z 

0, 

0* 

z 

Io• 

z 

0

100 

10-1. 

10"2. 

10"3. 

10°4. 

10-5. 

10, 6 

10-7,
Determination and Review

following reports: NUREG-4101 (Cline, Noyce, Coe, 

and Wright 1985), NP-4037 (Best and Miller 1985), 

NP-5077 (Best and Miller 1987), and TR-100641 

(James and Vance, 1992). The industry data bases 

for "291 and 99Tc largely represent detection limits for 

the LLW samples analyzed and reported by the 

service analytical laboratories using radiometric 

methods of analysis. As mentioned earlier, these 

"less than" values were entirely acceptable for 

classification of the radioactive waste, but very 

conservatively overestimate the actual quantities of 

these radionuclides being sent to LLW shallow-land 

burial facilities. 

Figures 7.5a and 7.5b show the geometric means for 

measurements made at PNNL and industry scaling 

factors for 129I and 9Tc for all PWR and BWR data. 

The PNNL 2 9

1/137Cs scaling factor for PWR stations 

of 6.48 x 10' is 26,000 and 4,800 times lower than 

the scaling factors reported in NUREG-4 101 and NP

5077, respectively. For BWR stations, the PNNL 

scaling factor of 4.25 x 10-7 is 1200 and 200 times 

lower than those reported in NUREG-4101 and NP

5077. The PNNL scaling factors are in very close 

agreement with the calculated PWR and BWR fuel 

source term isotopic ratios of 129I and '

37Cs computed 

by the ORIGEN-2 isotope generation code. Best and 

Miller (1985) have published ORIGEN-calculated 

isotopic ratios for the 10 CFR Part 61 radionuclides 

as a function of fuel bumup, and Bhattacharyya, 

Janati, and Shearer (1987) have conducted a similiar 

assessment for .29I and "

37

Cs. The 1

2 1/137Cs activity 

ratio varies only slightly with burnup, and after a 500 

day irradiation period the isotopic ratios for PWR and 

BWR stations are 2.8 x 10' and 2.9 x 107, 

respectively (Best and Miller 1985). The ratios 

reported by Bhattacharyya, Janati, and Shearer 

(1987) are essentially identical. In addition, more 

recent measurements of the 1

29

I/' 37Cs scaling factor 

for French reactor LLW have shown values very 

similiar to those measured in the PNNL study 

(Leuthrot, Ridoux, and Harrer, 1997). Thus, it 

appears that the iodine and cesium are being leached 

from the irradiated fuel at about the same rate into 

the reactor primary coolant and that major 

fractionation of the two radionuclides does not occur 

during their transport throughout the various waste 

streams at the nuclear power stations. For these 

reasons, it may be appropriate to consider a generic 

1

29

1/ 137Cs industry scaling factor for commercial

nuclear power stations, particularly PWR stations. 

James (1997) has recently reviewed the use of scaling 

factors for 99

Tc and 1291, and concluded that useful 

values based on recent, more accurate measurements 

can be used for estimating the concentrations of these 

radionuclides in LLW. 

For comparison purposes with the industry data base, 

a PNNL 9Tc/' 37

Cs scaling factor was computed even 

though the correlation of these two radionuclides was 

very poor. For PWR stations the PNNL scaling 

factor of 1.70 x 10.6 is 2000 and 460 times lower than 

the scaling factors reported in NUREG-4 101 and NP

5077, respectively. For BWR stations, the PNNL 

scaling factor of 1.18 x 10.6 is 1300 and 290 times 

lower than those reported in NUREG-4 101 and NP

5077, respectively. The industry data base 

assessment in NP-5077 also recognized that the 

correlation between 99

Tc and 'Co was much better 

than the correlation with '

37

Cs, and this report 

provided a geometric mean value for the 9Tc/60Co 

scaling factor for PWR stations of 2.3 x 10-". This 

value is still about 180 times higher than the PNNL 

scaling factor of 1.29 x 10.6 for PWR station LLW. 

The ORIGEN-calculated 9

"Tc/1'37Cs activity ratios for 

500-day irradiated fuel at PWR and BWR stations are 

1.31 x 10' and 1.33 x 10', respectively (Best and 

Miller 1985), and these values are about two orders 

of magnitude higher than the measured PNNL ratios. 

The most plausible explanation for this discrepancy 

is that the technetium leach rate from irradiated fuel 

is much lower than that for cesium. Studies of PWR 

spent fuel dissolution by simulated groundwater have 

shown that the initial leach rate of Cs from the spent 

fuel was much greater than for Tc (Wilson 1990). 

This study also showed that under the proper 

conditions leached Tc could be reduced, presumably 

to the +IV oxidation state, and removed from 

solution on hydrous iron oxides. If this behavior was 

occurring in reactor coolant water, it could explain 

the low 99Tc/ 137

Cs ratios observed in the PNNL data, 

as well as the much better correlations observed 

between 9Tc and 'Co in LLW. 

Similiar behavior is noted for 239.240pu relative to 

"

37Cs. The ORIGEN-2 code predicts a 239

,

24 pu/137

Cs 

ratio of 1.17 x 10-2 for PWR fuel irradiated for 500 

days, whereas the industry PWR data base for LLW 

gives a geometric mean value of 4.0 x 10'. This

NUREG/CR-6567 58
Determination and Review

Co-60 CONCENTRATION (Bq/g)

a

D 

0

0

0 0

a

a

0 0

0 0

00 

a 

00 

0B

(b)

- C-73 14 "7 C oE A T 6 . . 7 ... 08 

CS-137 CONCENTRATION (Bo,/g)

0

0 

0

Figures 7.4a, 7.4b Correlation of Tc-99 with Co-60 (a) and Tc-99 with Cs-137 (b) in 

Commercial Nuclear Power Station Low-Level Wastes

NUREG/CR-6567

102

S 

0 

0 

C.. 

z 

0 I-. 

0 

I--

102

101 

100

02 

0 

z 

0 

I.

z 

0 

z• 

I--

10-2.

.0 12

a

a

59
Determination and Review 

10" ,: 

(a) 

10"

c

0 

LL 

0 

CN 

,z, 

o3 

L10" 

10" 

10"

10" 

10

10"1

cc 

0 I

C.) 

La

1 

z 

-j 

C.) 

Co 

1c

U, 

I10

10

10-

10"

NUCLEAR POWER STATION LOW-LEVEL WASTE

NUCLEAR POWER STATION LOW-LEVEL WASTE

Figures 7.5a, 7.5b 

NUREG/CR-6567

Comparison of Geometric Mean Values for the PNNL and Industry 

Activity Scaling Factors for (a) 1-129/Cs-137 and (b) Tc-99/Cs-137 

60
Determination and Review

difference of a factor of 30 between the predicted 

fuel source term 3

9'24°Pu/1 37

Cs activity ratio versus 

the measured ratio in LLW is reasonably similiar to 

the factor of 77 difference between the predicted fuel 

source term 99Tc/13 7

Cs ratio versus the measured ratio 

in LLW. Thus, if a major fraction of the 9Tc in 

reactor coolant is present as a relatively insoluble 

reduced species, it would behave similiar to other 

relatively insoluble radionuclides such as 239

.

240

Pu and 

'Co. 

7.2.1.6 Conclusions Regarding 129I and "Tc 

Scaling Factors 

The generic activity scaling factors for "291 and 99

Tc 

developed in this study are believed to be the most 

accurate data base for these scaling factors, and are 

the ones recommended for use by the nuclear power 

industry. Based on the results of the ultra-sensitive 

mass spectrometric analyses of 1291 and 99Tc 

conducted in this study, the concentrations of these 

radionuclides and their associated scaling factors in 

LLW have been shown to be two to four orders of 

magnitude lower than the corresponding values in the 

nuclear power industry data bases. This difference is 

primarily due to the fact that the concentrations of 

these radionuclides reported in the industry data base 

are mostly lower limits of detection of the 

radiometric measurement methods used for their 

determination by service analytical laboratories. 

The quantities of "29I and 9Tc actually deposited in 

existing LLW disposal facilities has undoubtedly 

been over-estimated by several orders of magnitude. 

In addition, site performance assessment modeling of 

future LLW disposal facilities using the industry data 

bases to compute the radionuclide inventories at 

these disposal sites will likewise overestimate the 

quantities of 129I and 9Tc disposed at a facility by 

several orders of magnitude. If the true quantities of 

these two radionuclides were used in the site 

performance assessment modeling, the "29I and 99Tc 

would no longer be the limiting radionuclides 

controlling the total quantity of radioactive material 

allowed at a LLW disposal facility. 

The measured '

29

I/' 37Cs scaling factor is in good 

agreement with the ratio of these radionuclides in the 

ORIGEN-2 calculated reactor fuel source term.

However, the measured 99

Tc/' 37Cs scaling factor is* 

about 77 times lower than the ORIGEN-2 predicted 

ratio for the reactor fuel source term. The reason for 

this difference should be investigated by determining 

the oxidation state and physicochemical behavior of 

Tc in the reactor coolant and in LLW. This 

information is critical to a better understanding of the 

ultimate fate of 99Tc during the disposal of LLW in 

shallow-land burial facilities. 

It is recommended that the nuclear utilities include a 

few selected samples of LLW each year to be 

analyzed for 129I and 99Tc by ultra-sensitive mass 

spectrometry in their 10 CFR Part 61 compliance 

radioanalytical program. Although this small 

number of analyses is far from ideal, the data base 

that would be generated would be far superior to the 

overly-conservative scaling factors that are currently 

being developed and used by the nuclear utilities. An 

alternative method for estimating more accurate 

scaling factors would be to utilize one of the 

computer codes, such as RADSOURCE (Vance, 

Holderness, James, and Lee 1992) or 3R-STAT 

(Vance and Associates 1993), for predicting activity 

scaling factors from reactor coolant isotopic data and 

other operational parameters. However, it is still 

recommended that these computer code methods be 

validated by occasional mass spectrometric analyses 

of the 1291 and 99Tc concentrations in LLW from each 

station. 

7.2.2 Carbon-14 Scaling Factors in LLW and 

Activated Metal 

Similiar problems to those observed for 129I have 

been documented for "

4C values in the industry data 

base (Vance, Helmholz, and Sutherland 1988). 

Carbon-14 concentrations are frequently estimated in 

LLW using 14

C/°Co scaling factors, and there are 

discrepancies in these factors. As shown in Table 

7.9, the 1

4C/ICo scaling factors reported in 

NUREG/CR-4101 (Cline, et al. 1985) for the BWR 

and PWR full data bases are 47 and 5.4 times higher, 

respectively, than the scaling factors reported in 

EPRI NP-5077 (Best and Miller 1987). In addition, 

the scaling factors reported in EPRI NP-5077 

systematically decreased by two orders of magnitude 

with increasing sample activity. This is probably a 

reflection of reporting lower-limit-of-detection 

(LLD) values or values very near the detection limit

61 NUREG/CR-6567
Determination and Review

Table 7.9 1

4CI/Co Radwaste Scaling Factors (from Vance, et al., 1988)

Reference Sample Type BWR PWR 

NUREG/CR-4101(a) Full Data Base 3.5E-2 6.5E-2 

NP-5077(b Full Data Base 7.4E-4 1.2E-2 

Low Activity 3.8E-3 5.4E-2 

Medium Activity 7.6E-4 3.6E-3 

High Activity 3.1E-5 9.9E-4

(a) Clime, Noyce, Coe and Wright, 1985 

(b) Best and Miller, 1987

NUREG/CR-6567

14

Cf 0Co Scaling Factor

62
Determination and Review

for "

4

C in the low and medium activity samples that 

are one to two orders of magnitude higher than the 

actual "

4C concentrations in the samples. The service 

analytical laboratories who perform the 10 CFR Part 

61 radionuclide analyses generally use very small 

samples of LLW materials for analysis of "

4

C because 

a total destruction of the sample by combustion or 

oxidative acid dissolution is used, and excessively 

large quantities of liberated CO2 would be generated 

if large amounts of organic matter were present in the 

samples. Thus, many of the "

4

C measurements, 

especially for the low- and medium-activity samples, 

are made very near the detection limit of the 

analytical method, and any small variations in 

instrument background, noise, or statistical 

fluctuations could result in a false positive signal 

construed to be due to "

4

C. Also, the final 

measurement of the separated 14

C is usually done by 

nonspecific liquid scintillation counting, and any 

cross-contamination with other more abundant 

radionuclides in the LLW materials would are result 

in false positive values for "

4C. Thus, the correlation 

between "

4C and 'Co is rather poor, but this is not 

surprising since the two radionuclides are produced 

by entirely different mechanisms. Most of the "

4

C is 

produced in the reactor coolant by neutron activation 

(n, alpha) of the 1"0 present in the water molecules of 

the reactor coolant. The 'Co is produced by neutron 

activation of cobalt impurities in metal components 

used in the reactor pressure vessel (and internals) and 

in primary loop construction materials. 

Very few 'T4

C measurements have been made in 

neutron-activated metal wastes. Three different types 

of spent fuel assembly hardware were analyzed for 

1

4

C and other long-lived activation products 

(Robertson, Thomas, Wynhoff, and Haggard 1993). 

This work, which examined stainless steel, Inconel, 

and Zircaloy alloys, provided '

4

C/°Co activity 

scaling factors that varied from 3.67E-5 to 5.49E-3, 

with a geometric mean of 5.39E-4 for the full data 

set. This geometric mean is in reasonably good 

agreement with the industry data bases for high 

activity wastes from PWR and BWR stations. The 

concentrations of 1

4

C were dependent upon the 

concentration of the parent element (nitrogen) in the 

alloy and the distance from the center of the fueled 

region of the assemblies to the location of the 

samples. The '

4

C in neutron-activated metal 

components is produced primarily by the reaction

'

4

N(n,p)"4C with the nitrogen being present in the 

metals as the nitride or as interstitial nitrogen gas 

(Davis 1979). The nitrogen contents of the metal 

alloys of construction are often not well known, and 

when specifications are given the nitrogen contents 

have a considerable range. For this reason, it is 

difficult to conduct accurate neutron activation 

modeling estimates of 14

C concentrations in irradiated 

metals, and empirical measurements have provided 

the most accurate activity scaling factor data. 

7.3 Review of Activity Ratios of 

"36

Ni and 59

Ni in Low-Level 

Waste 

The nickel radionuclides, 63

Ni and 59

Ni, are co

produced by neutron activation of the stable nickel 

isotopes, 62

Ni and 58

Ni, respectively. The major 

source materials for the production of these 

radionuclides are stainless steel and Inconel metals 

used in reactor construction (see Table 2.3). Because 

the 63

Ni and gNi emit very low energy radiations 

during their decay, they are quite difficult to measure 

and require very clean radiochemical separations 

prior to specialized counting methods. In addition, 

adequate calibration standards of 59

Ni have not been 

commercially available for calibrating counting 

equipment. Because of these problems, there have 

been substantial differences in the relative amounts 

of these two radionuclides reported by various 

analytical service laboratories. 

When 10 CFR Part 61 was first enacted, a large 

number of analyses of these radionuclides were 

conducted by service analytical laboratories for a 

wide variety of LLW samples (Cline, et al., 1980; 

Cline et al., 1985). These initial measurements 

provided mean 59

Ni/63

Ni activity ratios of 0.0744 for 

all BWR waste samples and 0.0355 for all PWR 

waste samples analyzed. Calculated values of this 

ratio indicated that the ratio should be much lower, 

ranging from 0.0075 to 0.0116 depending on the total 

neutron fluence during irradiation of the neutron

activated material (Evans, et al., 1984; Vance, 

Helmholz and Sutherland, 1988). Niese and Boden 

(1995) calculated that the 59

Ni/ 63

Ni activity ratios in 

neutron irradiated materials would vary between 

0.0054 to 0.0080, depending on the neutron flux and

NUREG/CR-6567 63
Determination and Review

the irradiation interval. Table 7.10 shows the results 

of two sets of these calculations. Further 

assessments showed major discrepancies between 

calculated and measured values of the 59

Ni/63Ni 

activity ratios in LLW materials (Bedore, Levin and 

Tuite, 1987; Luksic, 1989; Robertson, et al., 1991). 

It now appears that the early measurements of 63

Ni 

and 59Ni in LLW reported by Cline et al (1980; 1985) 

had significantly higher 59

Ni/63

Ni activity ratios 

(perhaps by as much as a factor of 10) compared to 

more recent values. An examination of individual 

LLW sample analyses suggests that both systematic 

and random analytical errors resulted in large 

uncertainties in the measurement of 63

Ni and 59Ni. It 

also appears that some of the calculated ratios made 

in other investigations (Luksic, 1989; Robertson, et 

al., 1991) were in error due to insufficient neutron 

absorption cross-section data as a function of neutron 

energy. 

The bounding calculations of TNi/i63

Ni activity ratios 

shown in Table 7.10 should cover the true range of 

values for these radionuclides in LLW from nuclear 

power stations, and ratios ranging from 0.005 to 

0.0 12 would be possible. In any case, the 63

Ni, rather 

than the 59

Ni, would always be the nickel 

radionuclide that would control the waste 

classification criteria specified in 10 CFR Part 61.

NUREG/CR-6567 64
Determination and Review

Table 7.10 Comparison of Calculated 59

Ni/ 3

Ni Activity Ratios in Neutron-Irradiated 

LLW Material

Vance, Helmholz and Sutherland (1988) 

Thermal Neutron Flux Irradiation Period Activity Ratio 

(nlcm%/s) (years) (Ni-59/Ni-63) 

1.3E11 1 0.0076 

3 0.0077 

10 0.0079 

30 0.0084 

1.3E13 1 0.0116 

3 0.0114 

10 0.0108 

30 0.0094 

Niese and Bolen (1995) 

Thermal Neutron Flux Irradiation Period Activity Ratio 

(n/cm2ls) (years) (Ni-59/Ni-63) 

1.0E12 1 0.0079 

2 0.0080 

5 0.0080 

1.0E13 1 0.0079 

2 0.0078 

5 0.0077 

1.0E14 1 0.0072 

2 0.0067 

5 0.0054

65 NUREG/CR-6567
8 Summary, Conclusions, and Recommendations

This project has examined the occurrence and some 

of the potential radiological and environmental 

impacts of a number of very long-lived radionuclides 

produced in LLW generated at commercial nuclear 

power stations. The study has focused on a group of 

additional radionuclides that are not specified in 

1 OCFR6 1, but that are produced in neutron activated 

metal components of reactor internal hardware and in 

primary coolant demineralization resins that make up 

a significant part of the LLW streams from 

commercial nuclear power stations. These 

radionuclides include 1

0Be, 36

C1, 93

Mo, losmAg, 113mCd, 

and 1 1

mmSn. In addition to this group of radionuclides, 

the specified 10 CFR Part 61 radionuclides were also 

measured in the selected samples. 

Of the group of additional radionuclides measured in 

this project, l"Be, 36

C1, and l°smAg appear to have the 

greatest potential impact with respect to 10 CFR Part 

61. Although none of these radionuclides are present 

in LLW in quantities high enough to seriously impact 

the current LLW classification and shallow-land 

disposal criteria specified in 10 CFR Part 61, further 

radiological and environmental assessments should 

be considered to determine if they should be 

amended to 10 CFR Part 61. It is recommended that 

further studies be focused on: I) providing an 

accurate assessment of the total quantities of these 

radionuclides in LLW from these sources, 2) 

determine the leaching characteristics of these LLW 

materials, 3) determine the migration behavior and 

environmental pathways of these radionuclides upon 

release from LLW disposal facilities, and 4) provide 

performance assessment modelers with the necessary 

radiological/geochemical information to better 

predict the potential impacts from disposal of this 

group of radionuclides. 

Beryllium- 10 is of special interest because of its 

potential to be produced in relatively large quantities 

in nuclear power station LLW, originating in spent 

BWR and PWR control rods and in spent primary 

coolant demineralization resins. The "OBe is 

produced by neutron activation of 1

0

B present as 

neutron absorbing materials, boron carbide and 

borosilicate glass in control rods, and as soluble 

borate added to PWR reactor coolant. The extremely 

long half-life of '"Be (1.6E6 years), together with a 

relatively high ingestion dose conversion factor and a

relatively low Kd value for adsorption onto sandy 

types of soil, create the potential for long-term 

migration and offsite radiation exposure of this 

radionuclide from LLW disposal facilities. This 

study has shown that "oBe could be the third most 

abundant radionuclide in spent control rods and in 

some PWR spent primary resins after a decay time of 

about 600 years. Therefore, it is recommended that 

further studies be conducted to: 1) provide an 

accurate assessment of the total quantities of "OBe 

contained in LLW from these sources, 2) determine 

the 'OBe leaching characteristics from LLW, 3) 

determine the migration behavior and environmental 

pathways of 'oBe released from LLW disposal 

facilities, and 4) provide performance assessment 

modelers with the necessary 

radiological/geochemical information to evaluate the 

potential impacts from disposal of this radionuclide 

in LLW. 

Silver- 1 08m is also of interest because of its 

production in large quantities in PWR spent RCCA 

control rods which contain 80% silver. Silver- 108m 

has a 130 year half-life and a relatively high 

ingestion dose conversion factor due to its multiple 

high-energy gamma-ray emissions. Silver-108m will 

easily be the most abundant radionuclide in spent 

RCCA control rods for many hundreds of years. It is 

therefore of some concern from a waste disposal 

perspective. Fortunately, the stable metallic waste 

form and the expected high soil Kd values should 

minimize the leaching rate and offsite transport of 

this radionuclide during water infiltration at a LLW 

disposal facility. Nevertheless, it is recommended 

that further studies be conducted for l°smAg, as 

described in the above paragraph for "°Be, to better 

understand the radiological, environmental, and 

health impacts associated with the disposal of 

neutron-activated metal components containing lOSmAg. 

Chlorine-36 is the other very long-lived radionuclide 

measured in this study that deserves further 

assessment from a LLW disposal standpoint. Its very 

long half-life (3.01E5 years), relatively high 

ingestion dose conversion factor, and high 

environmental mobility combine to create the 

potential for long-term migration from a LLW 

disposal facility. Fortunately, the concentrations of

67 NUREG/CR-6567
Summary

"

63

C1 in LLW from commercial nuclear power stations 

are quite low. Chlorine-36 is produced almost 

exclusively in nuclear power plants by neutron 

activation of soluble chloride impurities present in 

the reactor coolant water, and is then concentrated on 

primary demineralization resins. Because chloride 

ions accelerate the corrosion of stainless steel in the 

reactor primary loop, great care is taken to minimize 

the chloride concentrations in reactor coolant water, 

and typical concentrations are in the low ppb range. 

This is sufficient to produce about 0.0001 to 0.002 

uCeig of 36

C1 in spent primary resins. This equates to 

an estimated 0.2 to 5 Ci/y of 36

C1 in primary resin 

LLW from all PWR nuclear power stations in the U. 

S. As discussed in Section 6.1 of this report, LLW 

disposal site performance assessment modeling, 

using a source term containing only 0.1 Ci of 36

C1 

was sufficient to predict that this radionuclide would 

be the major contributor to the offsite radiation dose 

for the contaminated well/drinking water exposure 

scenario. Because of this potential for radiation 

exposure, it is recommended that further studies 

similar to those described in the above paragraph for 

"10Be be conducted to obtain a better understanding of 

the radiological and environmental impacts from 

disposal of this radionuclide in LLW. 

The other additional radionuclides measured in this 

study, 93

Mo, lI3

mCd, and 1

2lmSn, either have half-lives 

that are too short to be of consequence in LLW 

disposal, or in the case of 93

Mo, the concentrations do 

not appear to be high enough to contribute 

significantly to the potential radiation exposure. 

Activity scaling factors have been developed for all 

of the additional radionuclides measured in this 

study. The use of activity scaling factors for 

estimating the concentration of a radionuclide in 

LLW is permitted by 10 CFR Part 61.55, and this 

method is desirable for radionuclides such as those 

listed above that are difficult to measure directly in 

LLW materials. Consistent scaling factors for a wide 

variety of LLW materials were determined for 93

Mo 

and 9

`mNb relative to 'Co over a concentration range 

of four orders of magnitude. Generic scaling factors 

for the other long-lived radionuclides measured in 

this study showed much greater variability due to the 

unique composition of some of the materials. 

However, in some cases useful scaling factors were 

obtained when similiar LLW materials were

considered as a group. Because of the potential for 

offsite migration of 36C1 from a LLW disposal 

facility, it is recommended that more analyses of 

spent ion exchange cleanup resins ( the major source 

of 36

C1 in LLW from nuclear power stations) to 

develop a better data base of activity scaling factors. 

Recent measurements and assessments of 1291 and 

9Tc in LLW from commercial nuclear power stations 

have shown that the true concentrations of these 

radionuclides are several orders of magnitude lower 

than previously reported values. Very sensitive mass 

spectrometric analytical methods were utilized to 

determine the concentrations and activity scaling 

factors for 1291 and 99

Tc in a wide variety of LLW 

samples from a number of PWR and BWR stations in 

the U. S. (Robertson, et al., 1997). These 

measurements provided concentrations and 

associated activity scaling factors (relative to t3

VCs 

and 'Co) for these radionuclides that were two to 

four orders of magnitude lower than the very 

conservative values published in several nuclear 

power industry data bases. Geometric mean values 

of the 1

29

1/137Cs scaling factor for LLW from PWR 

and BWR stations measured in this study were 6.48 x 

10-8 and 4.25 x 107, respectively. The geometric 

means for 9Tc/137

Cs in LLW from PWR and BWR 

stations were 7.37 x 10` and 3.21 x 10 ., respectively. 

The correlations between 99

Tc and 6

"Co were 

considerably better than the correlations between 99

Tc 

and 137

Cs, and the geometric mean values of the 

99

Tc/6Co ratios for PWR and BWR stations were 

1.70 x 10-6 and 1.18 x 10', respectively. Due to the 

dual production modes for 99

Tc (neutron activation of 

molybdenum and fission of 23.U), it is recommended 

that a closer examination of the 9Tc activity scaling 

factors be conducted to develop more accurate 

scaling factors for the specific modes of production 

at reactor stations that are dominated by one or the 

other production modes. 

Because performance assessment modeling studies of 

LLW disposal facilities have indicated that 1291 and 

"

9

Tc (as well as '

4C) will be the limiting radionuclides 

with regard to the total quantities of LLW permitted 

at a site, it is important to determine as accurately as 

possible the actual quantities of these radionuclides 

being disposed in LLW. The mass spectrometric 

measurements performed during this study have 

clearly shown that when the true concentrations and

NUREG/CR-6567 68
Summary

activity scaling factors are used for performance 

assessment modeling, the "29I and 9Tc no longer 

become limiting radionuclides for LLW inventories 

at disposal facilities. Disposal facilities will, 

therefore, be able to accommodate larger quantities 

of LLW at each facility, thereby minimizing the 

expense of developing additional disposal sites. 

An assessment of the 63

Ni and 59

Ni production and 

activity scaling factors in LLW indicates that early 

reported values of the 59

Ni/63

Ni activity ratio were too 

high by up to factors of 5 to 10. More recent 

analyses and assessments indicate that this ratio 

should range between about 0.006 to 0.0 12, 

depending upon the neutron flux and the irradiation 

time of the exposed materials. In any event, this 

activity scaling factor does not appear to be a critical 

parameter with respect to 10 CFR Part 61 waste 

classification criteria because the "

3

Ni, rather than 

"

95

Ni, will always be the controlling radionuclide of 

the two nickel radionuclides in LLW classification. 

Because all of the above radionuclides measured in 

this study do not have energetic gamma radiations and 

are very difficult to measure, it is desirable to apply 

activity scaling factors (developed from the empirical 

measurements) for estimating their concentrations in

LLW materials. Table 8.1 presents a summary of the 

scaling factors developed during this study and from 

previous analyses of LLW materials conducted at 

PNNL. The 93

Mo/6Co, 93mNb/6

OCo, 99

Tc/0Co, and 

1

29

1/137

Cs activity scaling factors are quite consistent 

over a wide range of LLW materials and 

concentrations, and generic scaling factors can be 

applied for these radionuclides. These radionuclides 

have been very carefully measured by either ultra

sensitive mass spectrometric methods or by state-of

the-art radiochemical separations and counting 

techniques. They are, therefore, considered to be the 

most accurate activity scaling factors in existence for 

these radionuclides. For the other radionuclides 

measured in this study, useful scaling factors can be 

developed for specific types of LLW materials, and 

these scaling factors are shown in Table 8.1. A 

number of these specific scaling factors have been 

developed using "less-than" concentration values for 

samples where the actual concentrations were below 

the analytical limit of detection. These specific 

scaling factors should be useful for estimating 

concentrations, or lower-limits-of-detection, of these 

radionuclides in various types of LLW materials as 

shown in Table 8.1.

69 NUREG/CR-6567
Summary 

Table 8.1 Summary of Measured Activity Scaling Factors for 

Additonal Radionuclides 

Number Activity Scaling of Scaling Factor 

Radionuclide Radionuclides Sample Type Samples Geometric Mean 

93Mo 93

Mo/ 6 °Co All (Table 7.1) 22 1.6E-6* 

93m Nb 93mNb/6°Co All (Table 7.2) 22 6.9E-4 

103e 10

Be/ 6°Co Control RodsA 3 2.4E-3 

Spent ResinB 3 2.lE-4* 

Metalsc 5 9.0E-6* 

36C1 36CI/ 6 0Co Spent ResinB 3 6.OE-5 

Metalsc 4 2.4E-5* 

ll 3

mCd 113Cd/ 69 Co Spent Resin3 2 8.8E-4 

Metals' 3 3.1E-4 

Cd Alloy' 4 0.65 

108mAg lOBmAg/6°Co Spent Resin3 2 6.6E-4* 

Metalsc 4 2.6E-3* 

Ag AlloyE 6 9.1 

12 mSn 12 1

m sn/BOCo Spent ResinB 2 4.5E-5* 

Metalsc 3 1.5E-4* 

1291 129 1/1 37Cs All (Table 7.8) 45 1.20E-7 

99Tc 9 9

Tc/1 37Cs All (Table 7.8) 31 5.23E-6 99Tc/ 6 °Co All (Table 7.8) 31 1.67E-6

* Scaling factor comprized 

A. Control rod materials 

B. Spent reactor coolant 

cleanup resin.

of two or more "less-than" values. 

containing 1'B4C and 1

`B enriched glass. 

cleanup resin and liquid radwaste

C. Stainless steel and inconel hardware from control rod 

assemblies. 

D. 5% Cd alloy used in RCCA control rod assemblies. 

E. 80% Ag alloy used in RCCA control rod assemblies.

NUREG/CR-6567 70
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CA, 1990. 

Serne, R. J., et al., "Solid-Waste Leach 

Characteristics and Contaminant-Sediment 

Interactions", PNL-8889, Vol. 1, Pacific Northwest 

Laboratory, Richland, WA, October, 1993. 

Smith, M. R., E. J. Wyse, and D. W. Koppenaal, 

"Radionuclide Detection by Inductively Coupled 

Plasma Mass Spectrometry: a Comparison of Atomic 

and Radiation Detection Methods". J. Radioanal. 

Nucl. Chem., Articles 160, 341-354, 1992. 

Strebin, R. S., F. P. Brauer, J. H. Kaye, M. R. Rapids 

and J. J. Stoffels, " Neutron Activation and Mass 

Spectrometric Measurement of 1 9

I", J. Radioanal. 

Nucl. Chem., Letters 127,59-73, 1988.

NUREG/CR-6567 72
References

Thomas, C. W., V. W. Thomas and D. E. Robertson, 

"Radioanalytical Technology for 10 CFR Part 61 and 

Other Selected Radionuclides", NUREG/CR-6230, 

U. S. Nuclear Regulatory Commission, Washington, 

D. C., March, 1996. 

Vance, J. N., "Generic Scaling Factors for Dry 

Active Wastes", TR- 100740, Electric Power 

Research Institute, Palo Alto, CA, September, 1992. 

Vance, J. N. "Topical Report for 3R-STAT: a Tc-99 

and 1-129 Release Analysis Computer Code", Vance 

and Associates, Ruidoso, NM, 1993.

Vance, J. N., H. R. Helmholz and A. Sutherland, 

"Assessing the Impact of NRC Regulation 10CFR61 

on the Nuclear Industry", NP-5983, Electric Power 

Research Institute, Palo Alto, CA, 1988. 

Vance, J. N., J. H. Holderness, D. W. James and C. 

E. Lee, "RADSOURCE, Vol. 1, Part 1: A Scaling 

Factor Prediction Computer Program: Technical 

Manual and Code Validation", TR- 101960, Electric 

Power 

Research Institute, Palo Alto, CA, 1992. 

Wilson, C. N., "Results from NNWSI Series 3 Spent 

Fuel Dissolution Tests", PNL-7170, UC-802, Pacific 

Northwest Laboratory, Richland, WA, 1990.

73 NUREG/CR-6567
Appendix A 

Radiochemical Procedures for Additional Long-Lived Radionuclides

75 NUREG/CR-6567
Appendix A

A.1 Pre-Treatment of Ion 

Exchange Resin, Neutron

Activated Metal, and solid 

LLW Samples Prior to 

10CFR Part 61 Radionuclide 

Analyses 

Ion Exchange Resins and Dry Active Waste 

Sample aliquots of spent ion exchange resin and solid 

dry active waste (DAW) were leached in acid to 

solubilize the radionuclides of interest in preparation 

for lOCFR Part 61 radiochemical analyses. Sample 

sizes ranging from 0.1 to 50 grams, depending upon 

the activity of the material, were contacted with hot 

(100I C)6N HC1 for 20 minutes. This leaching was 

repeated twice more. The combined acid leachate 

was cooled and brought to a standard volume with 

6N HCl. This served as the stock solution for further 

aliquoting and radiochemical analyses. 

Neutron-Activated Metal Specimens 

Because of the very high activities and gamma dose 

rates, most samples of neutron-activated metal were 

cut into small pieces inside of the PNNL hot cell 

facilities. To eliminate any fission product and TRU 

contamination of the metal specimens which could 

have occurred during operations in the hot cell, the 

metal samples were repeatedly (3-5 times) leached 

with hot acid in a clean fume hood to remove any hot 

cell contamination prior to dissolving the metal 

specimens for analyses. The stainless steel specimens 

were pre-cleaned by leaching with hot 6N HCI. 

Inconel and zircaloy samples were leached with hot 

6N HCl followed by hot 6N HNO 3 containing one 

mL of HF. The Ag-Cd-In alloy was leached with hot 

6N HNO3 . The cleaned metal specimens were then 

dissolved in hot 6N acid of the same types used for 

the leaching. The solutions of acid-dissolved metal 

were then cooled and brought to a standard volume 

with the appropriate acids. These acid solutions 

served as the stock solutions for further sample 

aliquoting for radionuclide analyses.

A.2 Detailed Radiochemical 

Procedures 

Beryllium- 10 Procedures 

I. To an aliquot of resin acid leach or metal acid 

dissolution, add 20 mg Be carrier and 7

Be tracer. 

2. Add 5 mg each of Co, Ni, Sr, and Cs carriers and 

make basic with NH4OH. Let sample stand for 1 

hour. Centrifuge, discard supemate, wash 

precipitate (ppt) with water, discarding wash. 

3. Dissolve ppt in a few mL of IN HCI and repeat 

step 2. 

4. Repeat step 3. 

5. Dissolve ppt in IM HCI, add 10 mg Eu carrier 

and 1 mL of saturated oxalic acid. Adjust 

solution to pH 3-4 with NH4OH. Heat and filter, 

discarding ppt. 

6. Adjust solution to IOM HCI and pass through an 

anion resin column, Cl- form, 100-200 mesh, that 

has been equilibrated with IOM HCl. Save 

effluent and wash column with 5 mL 10 M HC1, 

adding wash to effluent. Evaporate to about 5 

mL. 

7. Make solution pH 10 with NH4OH. Centrifuge 

and discard supemate. Wash ppt twice with 

water, discarding washes. Determine yield by 

measuring the volume of Be(OH), ppt and 

comparing to volume of 20 mg Be carrier 

precipitated as Be(OH)2 . Dissolve ppt with 5 mL 

of 0.1M HCI and transfer solution into a 

scintillation vial, add Ultima Gold® scintillator 

and beta count in the Quantilus® liquid 

scintillation. 

Chlorine-36 Procedure 

1. To sample aliquot of the HC1 acid dissolution or 

leach which contains up to 100 mg of CI ion per 

mL of solution, adjust sample acidity to IM with 

water and add 38

C1 tracer.

NUREG/CR-6567 76
Appendix A

2. Prepare a cation resin column, +H form, 100-200 

mesh (lcm x 15cm). Pass sample aliquot through 

column, saving effluent. Wash column with 5ml 

of 0.01 M Chl. Combine wash and effluent. 

Discard resin. 

3. To solution, add stoichiometric amount of silver 

nitrate solution, stir, centrifuge, check for 

complete precipitation of AgCI and discard 

supemate. Wash ppt twice with 10 mL of 2M 

HNO3. Discard washes. Wash ppt twice with 

water, discarding washes. 

4. Slurry ppt with 5 mL of water and transfer to a 

scintillation vial. Gamma count for yield. 

Suspend ppt in Insta-Gel® and count in 

Quantilus® liquid scintillation spectrometer to 

determine 36

C1. 

Iron-55 Procedure 

1. To sample aliquot in a 50 mL beaker, add 

enough Fe carrier to give 5 mg. Add 10 mg each 

Cs, Ni, Co, Sr. Make basic (pH 10-12) with 

fresh NHIOH. Stir for about 5 minutes. Let 

stand for 1 hr. 

2. Transfer to centrifuge cone, centrifuge and 

discard supemate. 

3. Dissolve ppt in several mL of dilute HNO3, add 

10 mg each Cs, Ni, Co, Sr carriers. Make basic 

(pH 10-12) with fresh HN4OH. Stir for about 5 

minutes. Let stand for I hr. 

4. Transfer to centrifuge cone, centrifuge, discard 

supernate. 

5. Dissolve ppt in 20 ml of 4M HC1, add 10 grams 

of anion resin that has been equilibrated in 4M 

HC1. Stir for 15 minutes, filter, wash with 10 ml 

of 4M HCI. Discard filtrate and wash. 

6. Elute Fe by passing three 10 mL portions of 

0.01M HC1 through the resin filter bed. Save 

eluate and discard resin. 

7. Evaporate eluant, transfer to a tared planchet,

and gravimetrically determine yield. Dissolve 

Fe with 5 mL 0.1 M HC1, transfer to scintillation 

vial and add Ultima Gold® and count in 

Quantilus® liquid scintillation spectrometer for 55

Fe. 

Nickel-59,63 Procedure 

1. To an aliquot of resin acid leach or acid 

dissolution of metal, add 20 mg each of Ni, Co, 

and Fe carriers. Evaporate to just dryness. 

2. Dissolve in small amount of 1M HCI, transfer to 

centrifuge cone and make basic with NH4OH 

(pH 9). Centrifuge and discard ppt. To 

supernate, add 1 mL of I M ammonium tartrate 

and 1 mL 6M NH4Cl. Add 5 mL of saturated 

dimethyl glyoxime (DMG), stir, and let stand for 

30 minutes. 

3. Centrifuge, discard supernate, dissolve ppt with 

8 M HNO3, transfer to small beaker, and 

evaporate to near dryness. Add 5 mL conc. HC1 

and evaporate to near dryness. Add 5 mL conc. 

HCI, 1 drop of 4M H2S0 4, and evaporate to near 

dryness. 

4. Add 5 mL IM HC1, 1 mL IM ammonium 

tartrate, I mL 6M NH4C1, and make basic with 

NH4OH to pH 9. Add 5 mL of saturate DMG, 

stir, and let stand for 30 minutes. 

5. Centrifuge, discard supemate, dissolve ppt with 

8M HNO3, transfer to a small beaker and 

evaporate to near dryness. Add 5 mL conc. HC1 

and heat to near boiling, add 0.1 mL of 4M 

H2SO4, and heat to near dryness. Add conc. HCI 

and evaporate to dryness. 

6. Dissolve reside in 10 mL of water, divide sample 

in two equal parts, placing one portion in a 

scintillation vial and the other in a clean beaker. 

7. To the portion in the scintillation vial, add 

Ultima Gold® and beta count in a Quantulus® 

liquid scintillation counter for 63

Ni. 

8. To portion in the beaker, add 1 mL of H2S0 4,

77 NUREG/CR-6567
Appendix A

thymol blue and add NH4OH until color is gold

green. Electroplate for 1-2 hours on a tared 

stainless steel disc. Wash disc with alcohol, dry 

and place in drying.oven at 100°C for 30 

minutes. Weigh, cover with Mylar®, and count 

on Low Energy Photon Spectrometer (LEPS) for 

100 minutes for 59

Ni. Calculate yield 

gravimetrically. 

Strontium-90 Procedure 

1. Place an appropriate aliquot of sample in a 50 

mL beaker. Prepare reagent blanks and a 10 mL 

standard of 8

"Sr. 

2. Add 5 mg each of Co and Cs carriers and about 

150 dpm 8'

5 r tracer to each sample. Evaporate 

the samples to dryness on a hot plate. Dissolve 

the residue in 5 mL 8M HNO3 . 

3. Prepare a Sr-Spec® column for each sample by 

removing the bottom plug and cap, press the top 

frit down to the resin surface and let the water 

drain out. Add 5 mL 8M HNO3 and let the 

solution drain through the column. 

4. Transfer the sample solution to the reservoir of 

the column. 

5. Rinse the beaker three times with 3 mL portions 

of 8M HNO 3 (let one solution pass completely 

through the column before adding the next and 

rinse the column reservoir well with each 

addition). Wash the column with an additional 

40 mL 8M HNO3. Discard the washes. 

6. Cap the column on both ends and heat-seal in 

plastic. Perform a quick gamma count to 

determine whether or not any 6

°Co or '

37

Cs 

remain on the column. If no 6°Co or 1

37

Cs are 

present, retum the column to the column stand, 

remove the cap and plug, and wash the column 

with 5 mL of 8M HNO 3. Record the time of the 

last wash. If there is 'Co or 137

Cs remaining on 

the column, wash the column with additional 8M 

HNO3, recording the time of the last wash. 

7. Elute the Sr with 10 mL H20 into a liquid 

scintillation vial. Count this fraction on a 

gamma detector to determine the `Sr yield.

Compare with the 10 mL 85

Sr standard prepared 

at the beginning of the procedure to determine 

yield. 

8. After the gamma count, evaporate the samples to 

dryness in the liquid scintillation vial under a 

Niobium-93m, 94 Procedure 

I. To an aliquot of sample, add 5 mg of each of the 

following carriers, Ni, Sr, Fe, Co, 20 mg Sb 

carrier and 50 mg Nb carrier. Also, add a 9Nb 

tracer to each sample. 

2. Make the samples just basic (pH 8) with 

carbonate-free NaOH. Centrifuge the sample. 

Discard the supernate. 

3. Dissolve the precipitate resulting from step two 

in 2M HNO3. Make the sample basic with 

NH4OH. Centrifuge. Discard the supernate. 

4. Dissolve the precipitate from step 3 with 5 mL 

conc. HF. Add 5 mg Co carrier and 20 mg Sb 

carrier to the sample. Mix the sample well. 

Transfer the sample to a beaker containing 150 

mL conc. HNO3. 

5. Heat the solution on a hot plate using a Teflon 

covered stir bar. Continue heating and stirring 

until the sample volume is reduced to at least 

half of the original volume. Cool the solution 

and transfer to a centrifuge tube with conc. 

HNO3. Centrifuge and discard the supernate. 

6. Rinse the ppt three times with water, 

centrifuging and discarding the wash each time. 

7. Dissolve the ppt in 5 mL conc. HF, repeating 

steps 4, 5 and 6. A very fine white precipitate 

will be evident. 

8. Transfer the washed ppt to a 2-inch diameter 

planchet. Dry under a heat lamp and cover with 

two layers of Mylar®. 

9. Count the 93"Nb on the Low Energy Photon 

Spectrometer (LEPSO). Following the LEPS 

count, count each sample on a gamma detector 

for a 1000 minute count of the 94

Nb, and "Nb

NUREG/CR-6567 78
Appendix A

tracer for yield determination. 

Technetium-99 and Molybdenum-93 Procedure 

1. To sample aliquot add 1 mg Re carrier, 99

Mo 

tracer/carrier, adjust to 1 M HCl, add I drop of a 

5% NaOC1 solution, and pass through a prepared 

cation resin (+H, 100-200 mesh) column. Save 

effluent, wash column with 5 mL of 0.01 M HC1, 

adding wash to effluent and discarding resin. 

2. Pass supernate through a TEVA-Spec® column. 

Discard effluent. The Mo and Tc both stay on 

the column. Wash column with pH 8 NH4OH, 

discarding wash. 

3. Remove top flit from column, extrude resin from 

the column into a preweighed plastic cup and dry 

at 100 °C for 1 hour. Reweigh, cover with 

Mylar® and count Mylar® side toward LEPS 

detector for yield determination. 

4. Allow 67-hour 'Mo tracer to decay away, then 

recount on the LEPS detector for the 16-key x rays of 93

Mo. Correct 93

Mo for both yield and 

self-absorption due to the TEVA-Spec material. 

5. Slurry resin with 5 mL of water and transfer to a 

scintillation vial. Add Insta Gel® scintillation 

cocktail, shake, and count in the Quantilus® 

liquid scintillation spectrometer for 99Tc 

determination. 

Cadmium-109, 113m Procedure 

1. To aliquot of resin acid leach or the acid 

dissolution of the metal, add ..5

Cd tracer. Add 5 

mg Co and 20 mg Ag carriers. 

2. Dilute sample to 15 mL with 0. 1 M HNO3. Add 

sufficient HC1 to ppt the silver present. Heat to 

coagulate ppt, filter and discard filter. 

3. To filtrate, add 20 mg of Ag carrier. If 

necessary, add additional HCI to ppt the AgCI. 

Heat, let stand and filter, discarding filter.

4. Add Fe carrier, co-precipitate Cd on Fe(OH)3, making first basic with NaOH. Centrifuge, 

discarding supernate. Dissolve ppt with IM HCL. 

5. Pass solution through an anion resin column 

previously equilibrated with 1M HCL. Wash 

column with 1 M HCl, discarding both effluent 

and wash. 

6. Elute Cd with 7.5M NH4OH. Evaporate to 

dryness to eliminate NH3 salts. 

7. Dissolve residue to 0.1 M HCl, make pH 8 with 

NaOH, add 10 mL IM thioacetamide and heat to 

form CDS ppt. 

8. Centrifuge, transfer ppt to a scintillation vial and 

gamma count for yield and 1 9Cd determination. 

9. Transfer CDs ppt to a beaker with conc. HNO3 . Evaporate to dryness, ad conc. HNO3 and 

evaporate to dryness. Add 6M HCl to beaker 

and evaporate to dryness. 

10. Dissolve CdCl2 residue with 5 mL of 0.IM HCl 

and transfer to a scintillation vial. After "Cd 

has decayed, add Ultima Gold® scintillator and 

count in a Quantilus® liquid scintillation beta 

spectrometer for 11m

3 Cd determination. 

Tin- 121 Procedure 

1. To a sample aliquot, add 20 mg each of Sn(II), 

Co, Ni, Nb and Cs carriers. Add lI mSn tracer 

and 2-3 mL of HCI and evaporate to just 

dryness. 

2. Add 4.5 mL H2SO4 and a few drops of H20 2 and 

heat to fumes. Allow the sample to cool. 

3. Dissolve the sample in 12 mL water and chill in 

an ice bath for at least a half hour. Add 4.0 mL 

5M KI to each sample and mix well. Pour the 

sample into a separatory funnel, along with 20 

mL toluene, and shake vigorously for several 

minutes. Allow the phases to separate and drain 

off the aqueous phase into a beaker. Drain the 

toluene into another clean beaker and then pour

79 NUREG/CR-6567
Appendix A

the aqueous fraction back into the separatory 

funnel. Add 20 mL of fresh toluene to the 

separatory funnel and shake well. 

4. Allow the phases to separate and drain the 

aqueous phase from the separatory funnel. 

Return the first toluene fraction to the separatory 

funnel containing the toluene from the second 

extraction. 

5. Wash the toluene fractions twice with 15 mL of 

3.6M H2SO 4 - 1M KI. Discard thee washes. 

6. Back extract the Sn from the toluene by placing 

10 mL of 0.9M H2SO into the separatory funnel 

and shaking vigorously for several minutes. 

Allow time for the two phases to separate. 

7. Drain the aqueous phase into a 50-mL centrifuge 

tube. Make the solution pH 8 with NaOH. Add 

10 mL of thioacetamide, stirring well. Heat the 

tubes in a hot water bath and cool. Centrifuge 

the samples discarding the supemate. 

8. Transfer the ppt to a beaker with 5-6 mL conc. 

HNO3 . Heat to destroy the sulfide and to form 

SnO2. Dilute the precipitate with 15 mL of hot 

water. Cool and centrifuge the sample, 

discarding the supernate. 

9. Slurry the precipitate with 5.0 mL of water and 

transfer to a liquid scintillation vial. Prepare the 

samples appropriately for gamma counting and 

count the l

7

mSn to determine the yield. 

10. Suspend the slurried precipitate in Insta-Gel® 

liquid scintillation cocktail and count in the 

Quantulus® liquid scintillation spectrometer to 

determine the amount of '

21

Sn in the sample. 

Pu. Am-Cm and U Procedure

3. Pass solution through a TRU-Spec® column that 

has been conditioned by passing through 5 mL 

of 2-3M HNO3. Wash column with 10 ml of 2

3M HNO3 to which 0.1 ml of saturated NaNO 2 

has been added. Discard wash. 

4. Elute Am with 15 mL of 4M HC1. 

5. Elute Pu with 15 mL of 4M HCI to which 0.3 

mL of a 30% TiC13 solution has been added. 

6. Elute U with 15 mL of a 0.1M ammonium 

oxalate solution. 

Co-ppt Method for Alpha Energy Analysis (AEA) 

1. Take 10 mL of IM HC1 and add 100 ug Nd3÷, 

place in a plastic beaker, and add I mL of 48% 

HF, mix and let stand for 30 minutes. Filter 

through a 25 mm polysulfone filter. Repeat this 

step to make 3 thin pads of NdF3 ppt. Save. 

2. To eluant from Step 4, evaporate to small 

volume, and adjust to about 1M HCL. Add 100 

ug Nd3+ and I mL of 48% HF and let stand for 

30 minutes. Filter through one of the beds 

prepared in Step 1. Mount, dry and count on a 

SiLi alpha energy analyzer. 

3. To eluant from Step 5, treat like Step 2 for 

determination of Pu Isotopes. 

4. To eluant from Step 6, treat like Step 2 for 

determination of U isotopes.

1. To sample aliquot add 242pu, 243Am, and 212U 

tracers. Evaporate to dryness. 

2. Dissolve residue in 2-3M H4NO 3 , make 0.1M in 

ferrous sulfamate and 0.1M in ascorbic acid and 

stir for 10 minutes.

NUREG/CR-6567 so
Appendix B

Environmental Considerations of Additional Long-Lived Radionuclides

B.1 Environmental 

Considerations of Additional 

Long-Lived Radionuclides 

Of the additional long-lived radionuclides examined 

in low-level waste in this study, only l°smAg, "°Be, 

"

36

C1, and perhaps "3

'mCd appear to be of sufficient 

concentrations in some types of low-level wastes to 

be of a potential environmental concern during LLW 

disposal. The two most important processes affecting 

environmental mobility are: 1) the leachability of the 

radionuclides from the wastes by infiltrating 

groundwater, and 2) the soil retention properties of 

the radionuclides, i. e., the distribution coefficient, 

Kd. 

The expected leachability of this group of 

radionuclides is diverse. The l°smAg and 1l3

mCd 

leachabilities are expected to be low because they 

will be present in activated metal having relatively 

low corrosivity. The 1

0

Be present in primary coolant 

demineralization resins will be more readily available 

for leaching, but the '

0

Be present in reactor control 

rods will be in boron carbide and borosilicate glass 

which should have low leachabilities. The 3

aC1 

present in primary coolant demineralization resins 

will probably be readily leachable. However, no 

quantitative information is available to assess the 

rates and magnitude of the leachability of these 

radionuclides.

The soil adsorption properties for the additional long

lived radionuclides are listed as ranges of distribution 

coefficients (Kd) in Table 6.2. For all elements 

except Cl and Mo, the soil Kd values are relatively 

high and indicate that soil retention will be 

sufficiencly high to efficiently retard their migration 

when water infiltration and leaching of the waste 

occurs. Beryllium- 10 could have relatively low Kd 

values in some types of sandy soils, and together with 

its very long half-life (1.6 million years), the 

potential for offsite migration is considerable over 

very long periods of time. 

The 36

C1 and the 93

Mo will be leached as anions 

which will migrate readily in 

groundwater. Chlorine-36 has always been viewed 

as a conservative radionuclide in the 

soil/groundwater environment and is, in fact, used for 

age dating groundwaters because of its minimal 

retardation characteristics. Because of its high 

leachability and environmental mobility, even the 

small quantities of 36

C1 expected in LLW could 

significantly contribute to the offsite radiation dose 

of the contaminated groundwater ingestion pathway 

(Campbell, 1994). 

The generally low 93

Mo concentrations in most 

activated metal LLW, together with the expected low 

leachability of the host metals (stainless steel and 

Inconel) should release relatively small quantities of 

this radionuclide to the offsite environs.

81 NUREG/CR-6567
Table B.2 Estimated Range of Distribution Coefficient (Kd) Values for Elements Having Long

Lived Radioisotopes

Element Range of 

Distribution Coefficients-Kd (mL/g) 

Be 50-700 (sandy topsoil) (a) 

800-12,000 (sandy loam topsoil)(a) 

Mo -23 (allurium)(1 ) 

Cl 04a) 

Nb 530-5100 (allurium)(b) 

180-2900 (bentonite)@) 

Sn 1000-3000 (sandy topsoil)(a) 

4400-11,000 (sandy loam topsoil)(a) 

Ag 

Cd 1300-16.900(c)

(a) Semne, et al., 1993 

(b) Erdal, et al., 1977 

(c) Del Debbio, 1989

NUREG/CR-6567

Appendix B

82
NRC FORM 335 U.S. NUCLEAR REGULATORY COMMISSION I. REPORT NUMBER (RC2 FORM9 3 (Assigned by NRC. Add Vol., Suop., Rev., 12-91 and Addendum Numbers, If any.) 3201.R202 BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse) NUREG/CR-6567 

2. TITLE AND SUBTITLE 

Low-Level Radioactive Weste Classification, Characterization, 3. DATE REPORTPUBLISHED 

and Assessment: Waste Streams and Neutron-Activated Metals MONTH YEAR 

August 2000 

4. FIN OR GRANT NUMBER 

L1808 

5. AUTHOR(S) 6. TYPE OF REPORT 

Technical 

D. E. Robertson, C. W. Thomas, S. L. Pratt, E. L. Lepel, and V. W Thomas 

7. PERIOD COVERED (lnclusiee Dares) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC. provide Oivion, Office or Region, U.S. Nuclear Regulatory Commission, and mailing address, if contractor, provide 

name and mailing addres)

Pacific Northwest National Laboratory 

Richland, WA 99352

9. SPONSORING ORGANIZATION - NAME AND ADDRESS fIf NRC, type '$Sre as aboce " if contractor. proeide NRC Oivision. Office or Region. U.S. Nuclear Regulatory Commisson, 

and mailing address) 

Division of Regulatory Applications 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001

10. SUPPLEMENTARY NOTES 

P. R. Reed, NRC Project Manager 

11. ABSTRACT (200 words or less) 

This study focused on identifying and characterizing a group of very long-lived radionuclides that are not specified in 10 CFR Part 61, but 

which are present in significant concentrations in various types of 

LLW materials generated at commercial nuclear power, stations. The 

concentrations of 10Be, 3

.

6

CI, 93

Mo, 1oSmAg, 113mCd, and l"lSn (as well as 

the specified 10 CFR Part 61 radionuclides) have been measured in a 

variety of neutron-activated metal and spent primary demineralization 

resin LLW samples obtained from U.S. nuclear power stations. Of this 

group of radionuclides, the '0Be, 3 C1, and l°smAg appear to be present 

in some types of LLW materials in sufficient quantities to warrant 

further investigations to better assess their radiological and 

environmental impacts associated with LLW disposal. In addition to 

these radionuclides, an assessment was conducted to provide an updated understanding of the occurrence, distribution, and activity scaling 

factors of "

4C, 99Tc, and 1291, as well as an evaluation of the 59Ni/ 6 3Ni 

activity ratio in LLW.

12. KEY WORDS/DESCR!PTORS (List words or phrases that /il assist researchers in locating the report. ) 13. AVAILABILITY STATEMENT 

low-level waste unlimited 

act ivated metals 14. SECURITY CLASSIFICATION 

waste (This Page) classification Unclassified 

long-lived radionuclides (This RepoT) 

beryllium- 10 Unclassified 

chlorine-3 6 

silver-108m 15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 (2-89)
Federal Recycling Program
NUREG/CR-6567 AUGUST 2000 LOW-LEVEL RADIOACTIVE WAbIt CLAbbItICA"iUN, LtIAI(AtAKLAIIUN, A_"D 

ASSESSMENT: WASTE STREAMS AND NEUTRON-ACTIVATED METALS