October 23, 2009 

MEMORANDUM TO: Gregory Suber, Chief 

 Low-Level Waste Branch 

 Environmental Protection 

 and Performance Assessment Directorate 

 Division of Waste Management 

 and Environmental Protection 

 Office of Federal and State Materials 

 and Environmental Management Programs 

THRU: Christepher McKenney, Chief /RA/ 

 Performance Assessment Branch 

 Environmental Protection 

 and Performance Assessment Directorate 

 Division of Waste Management 

 and Environmental Protection 

 Office of Federal and State Materials 

 and Environmental Management Programs 

FROM: Karen Pinkston, Systems Performance Analyst 

 Performance Assessment Branch 

 Environmental Protection 

 and Performance Assessment Directorate 

 Division of Waste Management 

 and Environmental Protection 

 Office of Federal and State Materials 

 and Environmental Management Programs 

SUBJECT: TECHNICAL REVIEW: SALTSTONE AND CONCRETE 

INTERACTIONS WITH RADIONUCLIDES: SORPTION (Kd), 

DESORPTION, AND REDUCTION CAPACITY 

MEASUREMENTS 

 On January 9, 2009, the U.S. Department of Energy, Savannah River Operations Office, 

provided the subject report for review by U.S. Nuclear Regulatory Commission staff (NRC) 

pursuant to Section 3116(b) of the Ronald W. Reagan National Defense Authorization Act for 

Fiscal Year 2005. The subject report is available on NRC’s Agencywide Documents Access 

and Management System (ADAMS) at accession number ML090150234. This report was 

reviewed in accordance with monitoring activities described in “U.S. Nuclear Regulatory 
G. Suber 2 

Commission Plan for Monitoring the U.S. Department of Energy Salt Waste Disposal at the 

Savannah River Site in Accordance with the National Defense Authorization Act for Fiscal 

Year 2005” (ML070730363). The staff’s technical review summary is attached for your use. 

Docket No: PROJ0734 

Enclosure: Technical Review Summary 
G. Suber 2 

Commission Plan for Monitoring the U.S. Department of Energy Salt Waste Disposal at the 

Savannah River Site in Accordance with the National Defense Authorization Act for Fiscal 

Year 2005” (ML070730363). The staff’s technical review summary is attached for your use. 

Docket No: PROJ0734 

Enclosure: Technical Review Summary 

DISTRIBUTION: 

NDevasar DEsh CGrossman ARidge LHoward/CNWRA 

ML092890633 

OFC DWMEP DWMEP DWMEP:BC 

NAME K.Pinkston A. Walker-Smith C.McKenney 

DATE 10/19 /09 10/19/09 10/23/09 

OFFICIAL RECORD COPY
 

Enclosure 

Technical Review Summary: TECHNICAL REVIEW: “SALTSTONE AND CONCRETE 

INTERACTIONS WITH RADIONUCLIDES: SORPTION (Kd), DESORPTION, AND 

REDUCTION CAPACITY MEASUREMENTS”

Review Completed: October 2009 

Reviewer(s): K. Pinkston, D. Pickett 

Document(s): Kaplan, D.I., Roberts, K., Coates, J., Siegfried, M., Serkiz, S., 

Saltstone and Concrete Interactions with Radionuclides: Sorption (Kd),

Desorption, and Reduction Capacity Measurements, SRNS-STI-2008-

00045, Savannah River Nuclear Solutions. October, 2008. ADAMS 

Accession # ML090150234 

Evaluation

The paper “Saltstone and Concrete Interactions with Radionuclides: Sorption (Kd), Desorption, 

and Reduction Capacity Measurements” by Kaplan et al. describes the results of experiments 

performed to provide additional data and model support for the performance assessment for the 

Saltstone Disposal Facility (SDF). This research includes the measurement of Kd values for the 

sorption of radionuclides on simulated saltstone and Vault 2 concrete, the adsorption and 

desorption rates, and the measurement of the reduction capacity of simulated saltstone. 

Nuclear Regulatory Commission (NRC) staff had the opportunity to discuss and observe these 

experiments during an onsite observation visit to the Savannah River Site (SRS) on July 31, 

2008. The onsite observation report for this can be found at NRC’s Agencywide Documents 

Access and Management System (ADAMS) Accession # ML082530057. 

The simulated saltstone material was made with a non-radiologically contaminated feed 

solution. In most cases, a simulated waste stream from the Deliquification, Dissolution, and 

Adjustment (DDA) process was used to create this material because this waste stream was 

believed to be most representative of the waste stream that is going to be disposed of at the 

SDF. Different leaching solutions were used to simulate different stages of the cement aging 

process. A portlandite saturated solution was used to simulate the first two stages of cement 

aging, and a calcite saturated solution was used to simulate the third stage of cement aging. 

The first stage of cement aging occurs immediately after the cement hardens and is 

characterized by high pH (>12.5), high ionic strength, the dissolution of alkali impurities, and the 

formation of CSH gels. The second stage of cement occurs after all of the alkali salts have 

been dissolved and the pH is controlled to a value of 12.5 by portlandite. In the third stage, all 

of the portlandite has been dissolved, the CSH gels begin to dissolve, and the pH begins 

decreasing. 

This research was intended to gain understanding about how the radionuclides will interact with 

the saltstone material and vault concrete once the radionuclides have leached out of the 

saltstone. These studies do not address how the radionuclides are bound in the saltstone 

during curing, and these studies do not provide information on how the radionuclides will be 

leached from saltstone. 
 

 2

Kd Value Measurements 

The solid-water distribution ratios, or Kd values, were measured for both simulated saltstone and 

Vault 2 concrete using batch sorption studies for I-125, Np-237, Se-75, Pa-231, Pu-238, Tc-99, 

U-233, and a suite of radionuclides that can be measured with gamma spectroscopy that 

included Am-241, Cd-109, Ce-139, Co-60, Cs-137, Hg-230, Sn-113, Sr-89, and Y-88. 

Measurements were made with leaching solutions that simulated both cement early in the aging 

process and aged cement. Additionally, measurements were made for DDA simulated saltstone 

under both oxidizing and reducing conditions. 

The results of these experiments are presented in Tables 2 and 3. Data were not obtained for 

uranium because of analytical problems associated with the measurements for uranium. 

Additionally, a problem with precipitation of the radionuclides occurred in the no solids control 

sample for the suite of gamma-emitting radionuclides. The no solids control sample contained 

either the portlandite or calcite saturated solution and the radionuclides without the vault 

concrete or simulated saltstone solids. The purpose of the no solids control sample was to 

determine if there was a loss of radionuclides through a mechanism other than sorption onto the 

cementitious material, such as precipitation or sorption onto the vial. 

The Kd values measured for Se, I, and Tc were lower than those reported in the literature. The 

lower Kd for Se was attributed to the activity of Se being near the detection limit. The measured 

Tc Kd values were several orders of magnitude lower than the literature value. This indicates 

that the Tc in these experiments was not reduced. The measured Kd values for Np and Pu were 

greater than previous measurements. Updated Kd values for I, Np, and Pu were recommended 

based on these results. The recommended Kd for iodine was changed from 2-10 mL/g to 5-9 

mL/g for cement in the first two stages of aging, and from 4 mL/g to 0 mL/g for aged cement 

(i.e., stage 3). The recommended Kd values for Np and Pu were increased from the previous 

literature value. The recommended Kd value for Tc was not changed from the previous 

literature value of 5,000 mL/g. 

NRC staff does not believe that the data from these experiments supports the continued use of 

the Kd value of 5,000 mL/g for Tc. Additionally, it is not clear if other literature on the Kd of Tc 

supports the use of this value for reducing grout. In the report by Bradbury and Sarott (1995), 

which was cited as the basis for the Tc Kd value in Kaplan and Coates (2007), a Kd value of 1 

m3

/kg (1,000 mL/g) is recommended in Table 4. In section 5.2.10 of Bradbury and Sarott, work 

done by Bayliss et al. (1991) is cited in which a Tc Kd value of 5 m3

/kg (5,000 mL/g) was 

measured. However, this Kd value was measured in the presence of the sodium dithionite 

reducing agent, which is not present in saltstone, so it is not clear if this Kd value is applicable to 

this system. Understanding and correctly modeling the behavior of Tc is important because it 

significantly affects the performance and expected dose from this disposal system. As 

described in more detail below, the research at SRS on the Kd for Tc is ongoing. NRC staff 

believes that this research will be useful in addressing this issue. 

Adsorption/Desorption Experiments 

The rates of adsorption and desorption were measured to determine if the steady-state 

assumption was valid and to compare adsorption and desorption rates. The Kd approach 
3 

commonly used in modeling environmental transport assumes that both rates are same. 

The rate of adsorption of the radionuclides was measured for simulated vault 2 concrete and 

DDA simulant saltstone in batch experiments. As was true for the Kd experiments, the 

experiments with the gamma emitting suite of radionuclides had problems with precipitation of 

the radionuclides in the no solids controls. The rates of desorption were measured using 

columns packed with the Vault 2 concrete and saltstone material from the adsorption 

experiments. The concentrations of radionuclides in the column effluent were measured with a 

gamma detector. 

The results of these experiments indicated that the rate of desorption was much slower than the 

rate of adsorption for the simulated Vault 2 concrete. However, NRC staff believes that the use 

of the solid samples (in which there may have been problems with precipitation) to measure the 

desorption rates may not have been appropriate because the measured rates of desorption may 

be attributable to dissolution rather than desorption. Additionally, the precipitation observed in 

the no solid control samples may be due to an artifact of the experimental design and may not 

represent a phenomenon that would occur in the actual disposal system. 

The rates of sorption and desorption was also measured for Tc on DDA simulant saltstone and 

Vault 2 concrete in a calcite saturated solution. In these experiments, the solids and calcite 

solution were brought to equilibrium, and then 99TcO4

-

 was added. Samples were removed and 

analyzed over time. After two weeks, the liquid phase was removed and replaced with a 

solution that did not contain Tc-99. Samples were analyzed at different times over 26 days to 

measure the rate of desorption. 

In these experiments, very little Tc-99 was sorbed by either the Vault 2 concrete or DDA 

simulant saltstone. The measured Kd values in this portion of the experiment were 13 mL/g for 

simulated saltstone and 28.1 mL/g for Vault 2 concrete. These values are in the same range as 

the Kd value of 6.5 mL/g measured in the Kd experiments described above. These results 

indicate that Tc(VII) was not reduced to Tc(IV) because, if reduction had occurred, the Tc would 

have precipitated and the apparent Kd value measured would have been much higher. These 

results are unexpected because the saltstone and Vault 2 concrete had large reduction 

capacities. It was hypothesized by the researchers that the length of the experiment was not 

long enough for reduction to occur. 

These experiments also showed that Tc sorption occurred quickly, but the rate of desorption 

was much slower. The paper concludes that this fact implies that if the groundwater flows along 

a crack and by-passes the reduction capacity of Saltstone, the oxygenated water will not 

promote Tc desorption as much as is predicted based on the Kd value. 

NRC staff believes that it is important to understand the behavior of Tc in saltstone because as 

stated above the behavior of this radionuclide can significantly affect the dose from this disposal 

system. NRC staff also believe that additional research is needed on the relative adsorption 

and desorption rates of Tc. Because the Tc was not reduced in these experiments, the rates of 

adsorption and desorption may not be applicable to a system in which it is reduced. 
4 

Reduction Capacity Measurements 

The reduction capacity and redox conditions of saltstone over time is an important factor in the 

performance of the system. The environmental mobility of some radionuclides, such as Tc-99, 

is strongly dependent on its oxidation state. A previous long-term lysimeter study at SRS 

showed that the addition of blast furnace slag, which contains reductants iron(II) and sulfide, to 

saltstone limited the leaching of Tc-99. 

In the research described in this paper, the reduction capacities were measured by a titration 

method for a variety of materials including: DDA simulant saltstone, Vault 2 concrete, blast 

furnace slag, subsurface sediments, and 50 year old SRS concrete. The measured reduction 

capacities of the DDA simulant saltstone and Vault 2 concrete were higher than expected based 

on the measured reduction capacity of the blast furnace slag. The measured reduction capacity 

of the blast furnace slag was 832.4 μeq/g, and the measured reduction capacity of the 

simulated saltstone, which only contains 23% blast furnace slag, was 821.8 μeq/g. Similarly, 

the reduction capacity of the vault concrete was 240 μeq/g, which was higher than expected 

based on the vault concrete containing 10% blast furnace slag by weight. Several hypotheses 

were proposed for the reduction capacity in the simulated saltstone and vault concrete being 

higher than expected including: the higher pH of saltstone than blast furnace slag causing more 

of the matrix to dissolve and more of the reducing agents to be available for reaction, the 

saltstone could be behaving like a semiconductor, the minerals in the fly ash could be adding 

additional reducing capacity, and the microporosity of saltstone could be higher than blast 

furnace slag, which would make a larger surface area available for reaction. These results 

could also be attributable to the cerium used in the titration method sorbing more strongly to the 

saltstone than the blast furnace slag, but this was thought to be a less likely explanation. 

In addition to the reduction capacity, the redox state of the system was also investigated using a 

Eh probe. In this experiment, the pH and Eh were measured in the effluent from a column 

containing DDA simulant saltstone that had an air-sparged calcite solution passed through it. It 

was found that the pH value decreased with the volume of water that flowed through the 

system. The Eh value increased quickly to around 0 mV as the oxygenated water flowed 

through the column. Additional experiments showed that the Eh value decreased slightly after 

the flow through the column had been stopped for a period of 4-6 hours. This was attributed to 

the system not being at steady state. The flow rate of oxygenated water through the column 

was much faster than would be expected in the disposal system, and the kinetics of the 

reducing reactions may have been too slow to create reducing conditions in the column effluent. 

The ability of the saltstone in the actual system to maintain reducing conditions may be better 

than was seen in the column experiments. 

NRC staff believes that the reduction capacity and redox state of the saltstone is an important 

factor in the long-term performance of the saltstone, and the research described in the subject 

paper is extremely useful to providing model support for the performance assessment. NRC 

staff noted that even though the measured reduction capacity of saltstone was high, the Eh was 

not poised to reducing conditions in the column experiments. NRC staff agree that this was 

likely attributable to the high flow rate through the columns, but it would be prudent for additional 

experiments to be performed to confirm that the Eh will be poised to reducing conditions in 

saltstone that has oxygenated water flowing through it. 
5 

Teleconferences and Meetings

NRC staff discussed the research and observed some of the experiments described in the 

subject report during a meeting on July 31, 2008 at SRS. This discussion is summarized in an 

onsite observation report (ADAMS accession # ML082530057). 

The subject report was also discussed with representatives of the Department of Energy and its 

contractors on March 25-26, 2009 at the onsite observation conducted at the Savannah River 

Site (SRS). A summary of the discussion is provided in the onsite observation report (ADAMS 

Accession No. ML091320439). The follow-up actions as a result of the discussion were: 

1. Explain what measures were taken to ensure that experiments with technetium 

were not affected by experimental losses, such as technetium holdup in labware, 

resulting in underestimates of technetium concentration. 

2. Clarify the pH of the calcite solution used in these experiments (page 9 and 16 

state the pH=10; page 7 states that solution pH = 8.3). 

3. Clarify the selenium Kd value reported in Table 5, which is different than the value 

reported previously in the report. 

SRS staff addressed these follow-up actions in a teleconference on August 5, 2009. The 

meeting summary for this teleconference and supporting documents are located in ADAMS at 

Accession # ML092650394. The response to action 1 was that non-sediment control samples 

were used as an internal control for experimental losses. Actions 2 and 3 were attributed to 

transcription errors. The correct pH value referred to in action 2 is 8.3, and the correct value for 

the selenium Kd for a Stage 3 concrete referred to in action 3 is 150 mL/g. 

NRC staff found these responses to be acceptable, and these three follow-up actions were 

closed. However, NRC staff identified a new follow-up action based on the response to 

action 1: 

ML092650394-001. Provide details on the amount of variability observed in the 

measured concentration of the non-sediment control sample 

(i.e. Co in Attachment 3). 

Open Issues

An open issue was identified related to the research described in the subject report during the 

March 25-26, 2009 onsite observation conducted at SRS (ML091320439): 

Open Issue 2009-1 

At the SRS Saltstone Facility, DOE should demonstrate that (1) technetium-99 in salt 

waste is converted to its reduced chemical form in saltstone grout during the curing of 

saltstone grout, and is thereby strongly retained in saltstone grout, and (2) the sorption 

of dissolved technetium-99 onto saltstone grout and vault concrete is consistent with Kd

values for technetium-99 that were assumed in the performance assessment. 
6 

SRS staff discussed their proposed approach to close this open issue during the August 5, 2009 

teleconference. Their proposed approach involves performing additional testing on the 

saltstone reducing capacity and the Kd value of Tc-99. NRC staff agreed that the proposed 

approach seemed appropriate. 

Conclusion

The research described in the subject report provides valuable information to support key 

parameters in the performance assessment. NRC staff also believes that the additional 

research being performed by SRS for the oxidation state and Kd value of Tc-99 in saltstone is 

appropriate and will provide useful information on the behavior of this radionuclide in this 

system. 

References

Bradbury, M.H., Sarott, F.A., Sorption Databases for the Cementitious Near-Field of a L/ILW 

Repository for Performance Assessment, ISSN 1019-0643, PSI-Bericht Number 95-06. Paul 

Scherrer Institut. March, 1995. 

Kaplan, D.I., Coates, J.M., Partitioning of Dissolved Radionuclides to Concrete under Scenarios 

Appropriate for Tank Closure Performance Assessments, WSRC-STI-2007-00640, Washington 

Savannah River Company. December, 2007.