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Technical Review: Near Field Flow and Transport for the 2020 Saltstone Disposal Facility Performance Assessment - Ga
	ML23017A086
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Site:	PROJ0734
Issue date:	04/18/2023
From:	George Alexander, Hans Arlt, Christianne Ridge; Division of Decommissioning, Uranium Recovery and Waste Programs
To:	;
	Alexander G, Ridge A
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ML23090A081	List: ML23017A083; ML23017A084; ML23017A085; ML23017A086; ML23017A087; ML23017A088; ML23017A089; ML23017A090; ML23017A113;
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Technical Review: Near Field Flow and Transport for the U.S. Department of Energy 2020 Performance Assessment for the Saltstone Disposal Facility at the Savannah River Site Date

((date:long Reviewers George Alexander, Risk Analyst, U.S. Nuclear Regulatory Commission Christianne Ridge, Sr. Risk Analyst, U.S. Nuclear Regulatory Commission Hans Arlt, Sr. Risk Analyst, U.S. Nuclear Regulatory Commission

1. Purpose and Scope The purpose of this U.S. Nuclear Regulatory Commission (NRC) staff Technical Review Report (TRR) is to document the NRC staff review of the near field flow and transport for the U.S.

Department of Energy (DOE) 2020 Performance Assessment (PA) for the Saltstone Disposal Facility (SDF) at the Savannah River Site. The NRC staff performed this review to support a future decision about whether the DOE has demonstrated that radioactive waste disposal activities at the SDF are in compliance with the performance objectives of Title 10 of the Code of Federal Regulations (10 CFR) Part 61, Subpart C. This technical review also supports NRC monitoring of the SDF under the following monitoring factors and monitoring areas in the NRC current plan for monitoring the SDF (NRCs Agencywide Documents Access and Management System under Accession No. ML13100A113):

Monitoring Factor (MF) 3.03 (Applicability of Laboratory Data to Field-Emplaced Saltstone) under Monitoring Area (MA) 3 (Waste Form Hydraulic Performance)
MF 4.01 (Waste Form Matrix Degradation) and MF 4.02 (Waste Form Macroscopic Fracturing) under MA 4 (Waste Form Physical Degradation)
MF 5.01 (Radionuclide Release from Field-Emplaced Saltstone), MF 5.02 (Chemical Reduction of Tc by Saltstone), MF 5.03 (Reducing Capacity of Saltstone), and MF 5.04 (Certain Risk-Significant Kd Values for Saltstone) under MA 5 (Waste Form Chemical Performance)
MF 6.01 (Certain Risk-Significant Kd Values in Disposal Structure Concrete), MF 6.03 (Performance of Disposal Structure Roofs and High-Density Polyethylene/Geosynthetic Clay HDPE/GCL Layers), MF 6.04 (Disposal Structure Concrete Fracturing), and MF 6.05 (Integrity of Non-cementitious Materials) under MA 6 (Disposal Structure Performance)
MF 10.05 (Moisture Characteristic Curves) under MA 10 (PA Model Revisions)

The scope of this review included the DOE assessment of near field flow and transport in the 2020 SDF PA. In this TRR, the near field is generally defined as the area encompassing the lower lateral drainage layer (LLDL), disposal structure, saltstone, and surrounding soils. That area was modeled by DOE using the vadose zone flow model (VZFM) and the vadose zone transport model (VZTM). This TRR does not address water percolation through the closure cap, Enclosure

which the NRC staff addressed in a separate TRR entitled Percolation Through and Potential Erosion near the closure cap (ML23017A083). This TRR focuses on risk-significant aspects of SDF performance related to the near field and changes made to the SDF PA since the 2009 PA that were not previously reviewed in other NRC staff TRRs, which are listed in Section 2.0 (Background).

2. Background

In 2012, the NRC staff issued a Technical Evaluation Report (TER) for the SDF (referred to as the 2012 TER) (ML121170309). In that TER, NRC staff evaluated the DOE design and assumed performance of the near field flow and transport, including sections on Waste Form, Source Term Release, and near field transport. Since the 2012 TER, the NRC staff conducted technical reviews on several of these areas:

November 7, 2013 - Solubility of Technetium Dioxides in Reducing Cementitious Material Leachates, a Thermodynamic Calculation (ML13304B159)
June 4, 2015 - Oxidation of Reducing Cementitious Waste Forms (ML15098A031)
August 1, 2016 - Quality Assurance Documentation for the Cementitious Barriers Partnership Toolbox (ML16196A179)
January 5, 2017 - Iodine Sorption Coefficients for Use in Performance Assessments for the SDF (ML16342C575)
March 23, 2017 - Saltstone Waste Form Hydraulic Performance (ML17018A137)
April 12, 2017 - Performance of the High-Density Polyethylene, High-Density Polyethylene/Geosynthetic Clay (HDPE/GCL) Liner, and the LLDL (ML17081A187)
May 22, 2018 - Update on Projected Technetium Release from Saltstone (ML18095A122)
May 28, 2019 - Saltstone Waste Form Physical Degradation (ML19031B221 In the 2020 SDF PA, the DOE incorporated several significant changes related to near field flow and transport since the 2009 SDF PA. Those changes included:
revised planned configuration from sixty-four 46-m (150-ft)1 diameter cylindrical disposal structures to six 150-ft diameter cylindrical disposal structures (i.e., Saltstone Disposal Structure (SDS) 2A, SDS 2B, SDS 3A, SDS 3B, SDS 5A, SDS 5B), seven 114-m (375-ft) diameter cylindrical disposal structures (i.e., SDS 6 through SDS 12), in addition to the existing rectangular disposal structures (i.e., SDS 1, SDS 4)
assumed performance of the HDPE/GCL composite barrier layer and sand drainage layers 1

Throughout this document, the cylindrical disposal structures are referred to as the 150-ft diameter and 375-ft diameter disposal structures for consistency with the DOEs terminology in the 2020 SDF PA.

initial saltstone hydraulic properties based on hydraulic testing of saltstone core samples
assumed saltstone physical degradation in the Compliance Case
inclusion of young-aged material as the initial condition of cementitious materials
revised Kd for Iodine-129 (I-129) and the use of a shrinking core model for I-129 release
solubility-controlled Technetium-99 (Tc-99) release with solubility limits based on measurements for field-emplaced saltstone The NRC staff review of the 2020 SDF PA near field flow and transport focuses on the: (1) significant changes since the 2009 SDF PA and 2012 TER, (2) risk-significant features, events, and processes (FEPs), and (3) risk-significant modeling results. The NRC staff review of aspects of the 2020 SDF PA near field flow and transport that have not changed substantially from the 2009 SDF PA will be incorporated by reference into this TRR.

For the development of the 2020 SDF PA, the DOE identified relevant FEPs, developed an interaction matrix of the system components and interactions between those components, and formulated a conceptual model and Central Scenario. The NRC staff reviewed that process in a TRR entitled Future Scenarios and Conceptual Models (ML23017A088). Then, the DOE developed the Compliance Case from the Central Scenario to evaluate the long-term performance of the SDF. The DOE developed three model cases as part of its Central Scenario: the Realistic Case, Compliance Case, and Pessimistic Case. The DOE described the Realistic Case as its best estimate, the Compliance Case as using the most probable and defensible input values, and the Pessimistic Case as biased toward increasing dose results and maximizing defensibility. The NRC staff used the same terms for those three cases throughout this TRR to facilitate comparison with cited DOE tables and figures. However, the use of those terms in this TRR does not indicate an NRC staff judgement regarding whether the parameter values are realistic, probable and defensible, or pessimistic.

3. Lower Lateral Drainage Layer, HDPE/GCL Composite Barrier, and GCL Layer As described in more detail in the 2020 SDF PA and the NRC staff TRR entitled Performance of the Composite Barrier Layers and Lateral Drainage Layers (ML23017A089), each of the 15 disposal structures will be overlain with a separate 0.6 meters (m) (2 feet (ft)) thick LLDL that extends 7.6 m (25 ft) past the edge of the disposal structure roofs. The LLDLs are sand layers that the DOE designed to shed water into a gap between disposal structures and prevent the buildup of head on top of the disposal structures. The LLDL will be underlain by a composite HDPE/GCL composite barrier layer that will rest on the disposal structure roofs. The LLDL and HDPE/GCL composite barrier acts as a redundant barrier to the upper lateral drainage layer (ULDL)2 and corresponding HDPE/GCL composite barrier.

The DOE assumed that over time colloidal clay will migrate with the water flux from the lower backfill layer to the underlying LLDL. The DOE modeled that water flux-driven clay as 2 The ULDL and corresponding HDPE/GCL composite barrier are located in the Closure Cap. These layers are described in more detail in the NRC staff TRR entitled Performance of the Composite Barrier Layers and Lateral Drainage Layers (ML23017A089).

accumulating in the LLDL from the bottom up. The thickness of the clay-filled portion was modeled as increasing with time, while the thickness of the unfilled portion was modeled as decreasing with time. Those changes will result in an overall decrease in the hydraulic conductivity and porosity of the LLDL; so that, after approximately 19,000 years, the DOE estimated that the hydraulic conductivity and porosity of the LLDL would be similar to those for the overlying backfill layer. The 2020 SDF PA described the composite barrier as essentially impermeable for thousands of years. Also described in the 2020 SDF PA and the NRC staff TRR entitled Performance of the Composite Barrier Layers and Lateral Drainage Layers (ML23017A089) are two additional barriers featuring a HDPE geomembrane layer or composite barrier layer. The 13 cylindrical disposal structures are designed to have an additional composite barrier layer encased within two layers of the underlying concrete mud mat. In addition, the six 150-ft diameter cylindrical disposal structures are designed to have a HDPE layer wrapping the exterior of the disposal structure walls. Those barriers are intended to reduce transfer of water, reduce inflow of carbon dioxide and oxygen; and reduce release of radionuclides.

4. Disposal Structure Design and Performance The 2020 SDF PA assumes that the final SDF configuration will include two existing rectangular disposal structures (SDS 1 and SDS 4, previously referred to as Vault 1 and Vault 4), six existing 150-ft diameter cylindrical disposal structures (SDS 2A, SDS 2B, SDS 3A, SDS 3B, SDS 5A, and SDS 5B), and 375-ft diameter cylindrical disposal structures (SDS 6, SDS 7, SDS 8, SDS 9, SDS 10, SDS 11, and SDS 12), resulting in a total of 15 disposal structures, as shown in Figure 1 below. Section 4.1 describes the engineered barriers common to all the disposal structures in and Section 4.2 describes the design and construction of each type of disposal structure.

Figure 1. SDF Layout (adapted from DOE document SRR-CWDA-2019-00001) 4.1. Design and Construction of the Disposal Structures 4.1.1. SDS 1 SDS 1 is a rectangular, reinforced concrete disposal structure approximately 180 m (600 ft) long by 30 m (100 ft) wide by 8.2 m (27 ft) high. A cross Section of SDS 1 is provided in Figure 2 below. The disposal structure is comprised of six cells (cells A through F), each 30 m by 30 m (100 ft by 100 ft). Cells A, B, and C have been filled with saltstone, clean capped with grout and are covered by a permanent roof. Cells D, E, and F are unfilled and have no roof. A conceptual diagram of SDS 1 materials is shown in Figure 2 below. The disposal structure was built and remains above grade but the roof will be completed with an HDPE/GCL composite layer, sand drainage layer and the disposal structure will be backfilled with soil and covered with a closure cap at the time of site closure. Additional details of SDS 1, including designs and concrete specifications are in Section 3.2.1 of the 2020 SDF PA. The DOE also evaluated a potential new disposal structure concrete formulation, which is discussed in more detail in Section 8 below.

Figure 2. SDS 1 Cross Section (adapted from DOE document SRR-CWDA-2019-00001) SDS 1 was built between 1986 and 1988 with cracking of the walls observed as early as 1988. In 1994, Cell A was filled with saltstone grout. However, rainwater migrated into Cell A, resulting in hydrostatic pressure, and contaminated water migrated through cracks in the wall of Cell A of SDS 1 (ESH-WPG-2006-00132 (ML101590241)). The DOE contained the leak, decontaminated the area to background levels, and installed a pipe to drain any remaining water. Roof repairs have limited further leakage. 4.1.2. SDS 4 SDS 4 is a rectangular, reinforced concrete disposal structure approximately 180 m (600 ft) long by 60 m (200 ft) wide by 9 m (30 ft) high. A cross Section of SDS 4 is provided in Figure 3 below. The disposal structure is comprised of 12 cells (cells A through L), each 30 m by 30 m (100 ft by 100 ft). Cells B through L each contain saltstone, whereas Cell A has drums of low-activity waste encapsulated by clean grout. Cells C and I contain saltstone to heights of ~4.6 m (15 ft) and ~4.0 m (13 ft), respectively. In addition, there are piles of several hundred plastic waste bags containing contaminated grout chips placed on top of the saltstone grout. Clean cap grout will be placed in Cells C and I to fill the remaining space. Additional discussion regarding the plastic bags is in NRC staff TRR entitled Site Stability [ML23017A114]. Roof joints allowed rainwater to also migrate into SDS 4 (ESH-WPG-2006-00132). The DOE determined that this resulted in wall cracking similar to SDS 1. The joints were subsequently sealed and sheet drains were installed in the remaining eight cells of SDS 4 (i.e., Cells B, D, E, F, H, J, K, L). The sheet drains consist of a polystyrene sheet with dimples on one side and a non-woven, needle-punched polypropylene filter fabric on the other side. By collecting drain water and condensate from the cells, the sheet drains limit head buildup between the walls and saltstone grout. The sheet drains also act as a barrier to separate the walls from moisture and the saltstone grout. Similar to SDS 1, SDS 4 was built and remains above grade, but the roof will be completed with an HDPE/GCL composite layer and sand drainage layer and the disposal structure will be backfilled with soil and covered with a closure cap at site closure. Additional details of SDS 4, including designs and concrete specifications are provided in Section 3.2.2 of the 2020 SDF PA.

Figure 3. SDS 4 Cross Section (adapted from DOE document SRR-CWDA-2019-00001) 4.1.3. 150-ft diameter disposal structures SDS 2A, SDS 2B, SDS 3A, SDS 3B, SDS 5A, and SDS 5B are cylindrical, reinforced concrete structures 150 ft in diameter with an interior height of 6.7 m (22 ft) (7.2 m [23.5 ft] at the center). A cross section of those disposal structures is provided in Figure 4 below. The floor slab, upper mud mat, roof, columns, and walls of those disposal structures were made of sulfate-resistant Type V concrete. The walls were cast on a steel diaphragm with reinforcing bars, pre-stressing wires, and shotcrete. The interlocking section of the steel diaphragm were filled with epoxy to form a water-tight barrier. The use of pre-stressing wires and backfilling the area surrounding the disposal structures with soil was designed to mitigate impacts from any hydraulic head and minimize leakage. The shotcrete was intended to protect the carbon-steel reinforcing bars and pre-stressing wires from corrosion. As described later in this TRR, the concrete for those disposal structures (other than the lower mud mats) also contain blast furnace slag to act as a chemical barrier in addition to a physical barrier to radionuclide release. Figure 4. 150-ft diameter Disposal Structure Cross Section (adapted from DOE document SRR-CWDA-2019-00001)

The 150-ft diameter disposal structures contain a drain water management system consisting of sheet drains, similar to the SDS 4 design, and a polyvinyl chloride or PVC system to collect free water and transfer the water back to be mixed with dry grout mix to form grout. After the operations are complete, the drain water system will be grouted to the extent practical and remain in place in the disposal structures. In addition to the cementitious barriers, the DOE utilized HDPE and HDPE/GCL composite layers in the 150-ft diameter disposal structures to limit advective and diffusive transport. Between the lower mud mat and upper mud mat, there is an HDPE/GCL composite layer, as shown in Figure 4. That layer was welded to an HDPE/GCL composite layer surrounding the walls up to a height of approximately 0.6 m (2 ft) above the upper mud mat. The HDPE liner extends the entire height of the disposal structure. During site closure, an HDPE/GCL composite layer will be installed on the roofs and overlain with a lateral sand drainage layer. The interior walls and floors of the 150-ft diameter disposal structures are coated with a mat-reinforced epoxy-novolac thermosetting lining to mitigate sulfate attack from exposure to saltstone bleed water. Additional details of the 150-ft diameter disposal structures, including designs and concrete specifications are provided in Section 3.2.3 of the 2020 SDF PA. 4.1.4. 375-ft diameter Disposal Structures SDS 6 through SDS 12 are cylindrical, reinforced concrete structures 375-ft in diameter with a minimum interior height of 13 m (43 ft). Those disposal structures were designed to provide additional capacity to support the increased feed rate from the Salt Waste Processing Facility. Cross sections of those disposal structures are provided in Figure 5 below. The floor slab, roof, and walls of those disposal structures were or will be made of sulfate-resistant Type V concrete. Fly ash and/or silica fume and blast furnace slag were or will be added to the disposal structure cementitious materials, other than the lower mud mats, to reduce the hydraulic conductivity. In addition, the blast furnace slag and potentially the fly ash provide chemically reducing conditions. The walls were or will be cast in place with reinforcing bars, vertical tensioning cables, horizontal pre-stressing wires, and covered with shotcrete. The roof will be supported by 208 61-cm (24-in) concrete columns made of Type II cement. The floor slab contains or will contain construction joints with water stops, as shown in Figures 3.2-25 and 3.2-28 in the 2020 SDF PA.

Figure 5. 375-ft Diameter Disposal Structure Cross sections for SDS 6 and SDS 7 through 12 (adapted from DOE document SRR-CWDA-2019-00001) In addition to the cementitious barriers, the DOE utilized HDPE and HDPE/GCL composite layers in the 375-ft diameter disposal structures to limit advective and diffusive transport. As for the other disposal structures, during site closure, an HDPE/GCL composite layer will be installed on the roofs and overlain by a lateral sand drainage layer. In addition to the composite barrier above the disposal structures, the DOE will also install composite barriers at the bottom of the 375-ft disposal structures. That is, between the lower mud mat and upper mud mat, the DOE will install an HDPE/GCL composite layer, as shown in Figure 5 above. Unlike the 150-ft diameter disposal structures, there is no HDPE on the exterior walls of the 375-ft diameter disposal structures. Also, unlike the 150-ft diameter disposal structures and eight of the SDS 4 cells, there are no sheet drains in 375-ft diameter disposal structures. The DOE included joints in the VZFM, which was implemented in the PORFLOW computer code, to represent the water stops and other engineered joints at the various SDS concrete interfaces. The joints were assigned the material properties of gravel. During construction of SDS 6, shrinkage cracks were observed in the roof and floor slab, as shown in Figure 6 below. Floor slab cracks were observed with widths between 0.07 to 0.5 mm (3 to 20 mils) (SRR-SDU-2017-00001 (ML20206L191)). Concrete core samples demonstrated that cracking extended through the entire floor slab thickness. Visible surface cracks were repaired with flowable epoxy. However, SDS 6 failed the initial hydrostatic leak test over the three days of the test and was suspected to be leaking through cracks in the floor slab or

construction joints. Water was observed between the upper and lower basemats and the slabs (SRR-SDU-2017-00001). Leakage between the floor slab and upper mud mat is shown in Figure 7 below. After leak testing, additional cracking in the floor slab was observed. An elastomer liner system was installed on the walls and floor to protect the concrete from the high pH, high-sulfate saltstone solutions and to prevent leakage. On the subsequent 10-day leak test, the DOE did not observe any signs of leakage. Figure 6. Cracks in the Floor Slab of SDS 6 (adapted from DOE document SRR-SDU-2017-00001) Figure 7. Leakage from the Hydrostatic Test (adapted from DOE document SRR-SDU-2017-00001)

Several months after the second leak test, water was observed in the East Sump of SDS 6 in March of 2017 (SRMC-CWDA-2022-00025 (ML22189A149)). The DOE observed that water would continue to refill the sump after the water was pumped out. The DOE initially assumed that the water was due to construction activities and rainwater migrating through an HDPE seam. However, in 2022, contamination from the saltstone grout was observed in the East Sump indicating the presence of a flow path from inside of SDS 6 to the sump (SRMC-UWMQE-2022-00001 (ML22189A147)). In addition to the materials described in the 2020 SDF PA, a 0.25 mm (10-mil) poly sheeting is or will be located in between the upper mud mat and the floor slab, as shown in Figure 7 above (SRR-SDU-2017-00001 (ML20206L191)). 4.2. Modeling of Disposal Structure Performance The physical properties assumed for the Compliance Case for SDS 1 and SDS 4 are in Table 1 below and the properties assumed for the 150-ft Disposal Structures are in Table 2 below. The properties for the 375-ft Disposal Structures are in SRR-CWDA-2018-00004, Rev.1 (ML20206L231). The walls of SDS 1 and 4 and the floor and roof of SDS 6 were modeled with increased hydraulic conductivities to account for observed cracks. The DOE evaluated three degradation phenomena: sulfate attack, carbonation-influenced steel corrosion, and decalcification using analytic solutions supported by numerical simulation codes provided through the Cementitious Barriers Partnership (CBP) Software Toolbox. The DOE also considered the effects of construction features and activities (e.g., anchor penetrations, exposure to bleed water) in the development of the degradation times. The combined effects of multiple phenomena were then considered to determine the time to complete degradation for each cementitious material as shown in Table 3 below. Table 3 also includes the assumed start time for degradation based on delays in degradation from other barriers. The floors, walls, and roofs of these disposal structures were assumed to degrade to the hydraulic properties of the surrounding soils (i.e., the materials do not impede flow). The assumed hydraulic properties between the initial and final states were represented by a geometric average of the values. Geometric averaging is described in more detail in Section 4.4.2.4 of the 2020 SDF PA and Sections 8.2.2 and 8.3.2 below. The DOE developed degradation time scales for the disposal structure cementitious materials and saltstone with conservative estimate, compliance value, and best estimate values. The focus of this TRR is on the compliance values used in the Compliance Case and several sensitivity cases. This is discussed further in the Evaluation section of this TRR.

Table 1. Modeling input values for the Compliance Case for SDS 1 and 4 (adapted from DOE document SRR-CWDA-2018-00004, Rev.1)

Table 2. Modeling Input Values for the Compliance Case for the 150-Foot Disposal Structures (adapted from DOE document SRR-CWDA-2018-00004, Rev.1)

Table 3. Assumed Start and End of Degradation of Cementitious Materials for the Compliance Case (adapted from Tables 4.4-45 in the 2020 SDF PA) In the 2020 SDF PA, the DOE explicitly represented the LLDL, HDPE/GCL composite barriers above the SDS roof, the HDPE/GCL composite barriers between the upper and lower mud mats (except for SDS 1 and SDS 4, which do not have that layer), and the vertical HDPE geomembrane layer wrapping the exterior walls of the 150-ft diameter disposal structure walls in the VZFM (see Figure 4.4-38 to 4.4-61 in the 2020 SDF PA). This section briefly summarizes key features of the model, which the NRC staff reviewed in greater detail in a separate TRR entitled Performance of the Composite Barrier Layers and Lateral Drainage Layers (ML23017A089). In Section 4.4.2.7 of the 2020 SDF PA, the DOE described the assumed degradation of the HDPE/GCL composite layers and HDPE geomembrane layers. The initial state was assumed to be the fully degraded version of the HDPE geomembrane in the Closure Cap Model (five 10-mm [0.4-in] holes per hectare). The model assumed that the HDPE geomembrane layer and composite barrier layer hydraulic properties decay to backfill, using geometric interpolation over time from 750 to 3,200 years (i.e., the logarithm of the hydraulic conductivity changes linearly in time).

5. Wasteform Properties 5.1. Description of Waste Form Saltstone is a cementitious material made of treated salt solution from the SRS F- and H- Tank Farms liquid waste storage tanks mixed with blast furnace slag, fly ash, and cement. The treated salt solution is comprised mostly of sodium nitrate, sodium hydroxide, sodium nitrite, sodium aluminum hydroxide, sodium carbonate, and sodium sulfate. After mixing, the slurry is pumped into the disposal structures and solidifies into an alkaline and chemically reducing grout with a microporous structure. The DOE also evaluated a potential new cement-free saltstone formulation, which is discussed in more detail in Section 7.8 below.

As discussed by the DOE (SRR-CWDA-2018-00004, Rev.1), the saltstone hydraulic properties were based on results from an SDS 2A core analysis and are shown in Table 4 below. The assumed initial hydraulic conductivity for saltstone in the Compliance Case was based on Dynamic Leach Method (DLM) testing of the SDS 2A core and is represented by the green line shown in Figure 8 below. The assumed initial effective diffusion coefficient was based on the maximum value for any analyte from the SDS 2A sample data. Table 4. Saltstone Hydraulic Properties (adapted from DOE document SRR-CWDA-2019-00001)

Figure 8. Saturated Hydraulic Conductivity Data for Saltstone Samples (adapted from DOE document SRR-CWDA-2018-00004, Rev.1) 5.2. Modeling of Waste Form Performance In the 2009 SDF PA, the DOE assumed that no physical degradation of saltstone was expected based on a thermodynamic and mass balance analysis (SRR-CWDA-2009-00017 (ML101590008) and WSRC-STI-2008-00236 (M L101600398)). Accordingly, the DOE Base Case3 (Case A) did not include a modeled representation of saltstone degradation, such as an increase in hydraulic conductivity or fracturing of the saltstone matrix. In the NRC 2012 SDF TER (ML121170309), the NRC indicated that the DOE assumption that saltstone will be hydraulically undegraded for 20,000 years was unrealistically optimistic. The NRC staff also indicated that the DOE assumption was inconsistent with observations of existing cracks in saltstone grout in SDS 4, although it was not clear how far those cracks extended into the saltstone grout, or the extent to which saltstone may be subject to additional cracking in the future. 3 In the DOE 2009 PA, the DOE referred to Case A as the Base Case, which was described as the scenario that the DOE most expected for the duration of the performance period. In both the DOE SDF FY 2013 and FY 2014 Special Analysis documents, the DOE did not use the term Base Case; but, instead developed an Evaluation Case by selecting parameter values that the DOE considered to be most probable and defensible. Although there were slight differences in how the DOE described the Base Case and Evaluation Case, the DOE used the Base Case and Evaluation Case results for comparison to the 10 CFR Part 61 performance objectives.

In response to the second set of NRC Request for Additional Information (RAI) Questions (ML103400571) on the DOE 2009 SDF PA, the DOE provided case K4 (SRR-CWDA-2011-00044, Rev. 1 (ML113320303.)), which modeled an ingrowth of fractures represented as a log-linear increase in hydraulic conductivity and diffusivity of saltstone grout with time. Complete physical degradation of saltstone was assumed to occur in 10,000 years with the hydraulic conductivity and diffusivity reaching values of 1.0 x 10-6 cm/s and 5.0 x 10-6 cm2/s, respectively. However, the NRC indicated in the 2012 SDF TER that there was significant uncertainty in the rate and extent of fracturing of cementitious materials over thousands of years. Consequently, the timing and magnitude of the projected dose from radionuclide releases from the SDF were uncertain. In both the Fiscal Year (FY) 2013 and the FY 2014 Special Analysis Documents, the DOE revised the approach to modeling saltstone degradation by assuming that decalcification5 via advective flow controls the degradation of saltstone (SRR-CWDA-2013-00062 (ML14002A069), SRR-CWDA-2014-00006 (ML15097A366), and SRR-CWDA-2013-00064 (ML14008A056)). As water migrates through the saltstone grout, calcium is assumed to leach from saltstone at a concentration equal to the pore solution concentration measured from a saltstone simulant (SRNL-STI-2013-00118, Rev. 1 (ML13189A205)). The DOE calculated the time to complete degradation according to the following equation:

Where:

 =                 ,
    =                                            ,     /
      =                                            ,    /

= ,

  =                  ,   /

The DOE then assumed that saltstone would degrade linearly with time from the initial hydraulic conductivity of 6.4 x 10-9 cm/s to that of the surrounding soil (SRR-CWDA-2014-00006). The diffusion coefficient of saltstone was also assumed to increase linearly from an initial value of 1.0 x 10-8 cm2/s to 5.3 x 10-6 cm2/s once the saltstone was assumed to be completely degraded. In the 2020 SDF PA, the DOE continued to assume that saltstone degradation is controlled by advective decalcification (SRNL-STI-2018-00077 Rev.1 (ML20206L165)). However, the DOE revised the degradation model from a linear increase in hydraulic conductivity with time to a geometric average. Calculation of the harmonic, geometric, and arithmetic means of the saturated hydraulic conductivity was provided in Section 4.4.2.4 of the 2020 SDF PA. Those different averages represent blends or spatial averages of intact and degraded saltstone 4 The DOE provided the NRC three cases related to Case K: Case K, K1, and K2. The only differences between those three cases were the Kd values used to represent Tc sorption in oxidizing and reducing cementitious materials (saltstone and disposal structure concrete). When that distinction is not important (e.g., when discussing hydraulic properties, which are the same in all three cases), the NRC staff used the term Case K to refer to all three cases. 5 The DOE assumed in the Evaluation Cases in both the FY 2013 and FY 2014 Special Analysis documents that degradation of the upper (i.e., 3.45 feet) of saltstone grout and clean cap in SDS 4 will be controlled by carbonation due to the presence of steel roof trusses.

hydraulic properties. A harmonic average may represent degradation proceeding as a horizontal moving front (i.e., flow perpendicular to the material layers) and has a p-averaging term of -1. Whereas an arithmetic average would be more representative of flow occurring parallel to the degraded materials with a p-averaging term of +1. A geometric average, which the DOE assumed in the 2020 SDF PA, lies in between harmonic and arithmetic means with a p-averaging term of 0. DOE provided additional information on the averaging of intact and degraded properties in Section 4.4.2.4 of the 2020 SDF PA. Additional discussion on the p-averaging term is also provided below in Section 8.3.2 Wasteform Degradation. Figure 9 below shows the assumed hydraulic conductivity for saltstone for a 375-ft Diameter SDS for the 2020 SDF PA Compliance Case. Figure 9. Comparison of the 2020 SDF PA Modeled Saturated Hydraulic Conductivities of Saltstone for a 375-ft Diameter Disposal Structure (adapted from DOE document SRR-CWDA-2021-00056) In the 2020 SDF PA, the DOE assumed that the moisture characteristic curves (MCCs) for cementitious materials varied in time to simulate degradation. However, the MCCs did not change much because the DOE assumed there would be very limited degradation of saltstone over tens of thousands of years, as discussed in further detail in Section 8.3 below. As shown in Figure 10 below, the cementitious materials modeled at the 100,000-to-200,000-year time interval have more restrictive MCCs. In other words, the cementitious materials are assumed to be less permeable at the same saturation level. For saltstone, the MCCs for the initial time interval (i.e., 0 to 50 years) and at 100,000 to 200,000 years are very similar. Consequently, the unsaturated hydraulic conductivity (i.e., the saturated hydraulic conductivity multiplied by the

relative permeability) for saltstone from 0 to 50 years is very similar to the unsaturated hydraulic conductivity from 100,000 to 200,000 years, as shown in Figure 11 below. 1.00E+00 Roof 0-50 yr Saltstone 0-50 yr Floor 0-50 yr Relative Permeability 1.00E-01 Roof 100k-200k yr Saltstone 100k-200k yr Floor 100k-200k yr 1.00E-02 1.00E-03 0.4 0.5 0.6 0.7 0.8 0.9 1 Saturation Figure 10. Relative Permeability of Cementitious Materials in the 2020 SDF PA (adapted from DOE PORFLOW Model Files) 1.00E-09 Unsaturated Hydraulic Conductivity, cm/s 1.00E-10 1.00E-11 1.00E-12 Saltstone 0-50 yr Saltstone 100k-200k yr 1.00E-13 0.4 0.5 0.6 0.7 0.8 0.9 1 Saturation Figure 11. Unsaturated Hydraulic Conductivity of Saltstone in SDS 7 (adapted from DOE PORFLOW Model Files)

6. Modeling of Near Field Flow In the 2020 SDF PA, the DOE relied upon a series of models, including numerical, analytical, and empirical models, to represent flow and transport in the SDF. The interfaces between each model were described in more detail in the NRC staff TRR entitled Model Integration (ML23017A090). In the 2020 SDF PA, the DOE used PORFLOW, which is a numerical software package, to estimate the flow fields within the lower backfill of the closure cap, the LLDL, the HDPE/GCL composite layer at the disposal structure roofs, the disposal structures, the saltstone within in the disposal structures, and backfill surrounding the disposal structures, and the unsaturated zone soils beneath the disposal structures. The VZFM represents the flow through each of those layers for each disposal structure at multiple modeled time intervals. The DOE used discrete time intervals (e.g., 0 to 50 years, 50 to 150 years) to model flow through materials with time-varying properties for the modeling period (e.g., 100,000 years). Additional information regarding the LLDL was provided in the NRC staff TRR entitled Performance of the Composite Barrier Layers and Lateral Drainage Layers (ML23017A089). The DOE modeled SDS 1, SDS 2 (for simulating the 150-ft diameter disposal structures), SDS 4, SDS 6, SDS 7 (for simulating SDS 7, SDS 8, SDS 10, SDS11, and SDS 12), and SDS 9 to develop flow fields for all of the disposal structures, as described in Section 4.4.4 of the 2020 SDF PA. Inputs into the VZFM varied with each modeled time interval as the assumed infiltration rates and hydraulic properties of the engineered materials change over time.

The projected flow fields from each modeled time interval for each of the modeled disposal structures from the VZFM were then input into a separate VZTM in PORFLOW to model flow and contaminant transport from the disposal structures, through the underlying unsaturated zone soils, and into the saturated zone at the water table. In addition, the VZFM results were also used as input to transport simulations using GoldSim to evaluate parameter uncertainty. For the development of the 2020 SDF PA, the DOE identified relevant FEPs, developed an interaction matrix of the system components and interactions between those components, and formulated a conceptual model and Central Scenario. The DOE developed the Compliance Case from the Central Scenario to evaluate the long-term performance of the SDF. As described in Section 2 of this TRR, the DOE also developed cases it referred to as the Realistic Case and Pessimistic Case as part of the Central Scenario to evaluate parametric uncertainty.

7. Radionuclide Release and Transport Modeling 7.1. Overview For the Compliance Case and other deterministic modeling cases, the DOE modeled radionuclide release and near field transport with the VZTM, which the DOE implemented in PORFLOW. The VZTM used flow rates from the VZFM with element-specific transport properties to project radionuclide release from saltstone and transport in the near field. The DOE also included an abstracted representation of radionuclide release and transport in the SDF GoldSim Model, which can be run probabilistically to provide risk insights. The NRC staff discussed benchmarking of the SDF GoldSim Model to VZTM results in a TRR entitled Model Integration.

Radionuclides in cementitious materials generally are immobile when they are associated with stationary phases (i.e., solids or immobile pore water) and mobile when they are associated with mobile pore water or pore gas. The VZTM represents radionuclide equilibria between

radionuclide concentrations in cementitious solids and pore water to approximate radionuclide transport in saltstone and disposal structure concrete. Because the DOE expects saltstone and disposal structure concrete to remain nearly 100 percent (%) saturated during the performance period, the VZTM only modeled the solid and aqueous phase radionuclides (i.e., it did not model gas-phase reactions). For most radionuclides, the relative concentrations in the solid and liquid phases depends on the pore water chemistry. In the 2020 SDF PA, the DOE modeled changes in pore water chemistry and the resulting changes in radionuclide mobility for 20,000 years after SDF closure (see Section 6.3 in this TRR). The DOE modeled the release of most elements from saltstone by assuming that a radionuclides concentration in the pore liquid was proportional to its concentration in (and on) the grout solids. The constant of proportionality is a called a sorption coefficient or Kd value. The only element for which the DOE represented liquid concentrations with something other than a sorption coefficient was technetium. For technetium, the DOE used a combination of solubility limits and a sorption coefficient (see Section 6.3 in this TRR). For most elements, the DOE developed Kd values from literature about cementitious materials. Because of the risk significance of iodine and technetium, the DOE conducted experiments with field-emplaced saltstone cores and saltstone simulants under chemically reduced conditions. The DOE used the results of those experiments as the bases for the transport properties for iodine and technetium in chemically reduced cementitious material. For chemically oxidized material, the DOE used literature values as the bases for sorption coefficients for all radionuclides, including iodine and technetium (see sections 6.5 through 6.7 in this TRR). After development of the 2020 SDF PA, the DOE evaluated the potential use of a new saltstone formulation. This TRR addresses the applicability of the measurements from the SDS 2A core samples and simulants to the new saltstone formulation in Section 6.8. 7.2. Release Models The DOE modeled radionuclide release from saltstone based on radionuclide diffusion and advection. The dominant process depended on the amount of water flowing through the saltstone at the time. The modeled radionuclide release also depended on the chemical conditions of the saltstone at the time. For elements other than iodine and technetium, the DOE modeled chemical transitions as occurring in whole components, such as an entire saltstone monolith, a disposal structure floor, or a disposal structure wall. The DOE modeled the chemical conditions in those components as changing when enough water had flowed through the whole component to replace all the pore water in that component a certain number of times. The DOE referred to the replacement of all the pore water in a component with new water as a pore volume exchange. Because of the risk significance and chemical sensitivities of iodine and technetium release, the DOE modeled chemical changes affecting those elements more gradually. Specifically, the DOE modeled chemical changes taking place in small sections of the saltstone and disposal structure concrete. The common modeling term for those small sections is finite element. The DOE tracked the amount of water projected to flow through each finite element and modeled the resulting chemical changes in each element individually. The DOE referred to this method of modeling the release and transport of iodine and technetium as a shrinking core model because the chemical conditions tended to change from the outside of each component (e.g., saltstone monolith, disposal structure floor), resulting in a shrinking core of the component that retained the original chemical conditions. The DOE used a shrinking core model for iodine and

technetium in saltstone and all parts of the disposal concrete that contain blast furnace slag (see Section 6.4 in this TRR). The blast furnace slag provides a chemical barrier to radionuclide release by providing chemical reducing capacity, which changes the transport properties of some radionuclides (see Section 6.3 in this TRR). The DOE made one additional modeling refinement for technetium. To accommodate limitations of the PORFLOW modeling environment, the DOE implemented the solubility limits for technetium with effective Kd values that mimicked the effect of a solubility limit. Because Kd values relate the aqueous concentration of an element to its concentration in the solid phase, the DOE set the initial effective Kd values to values that would result in technetium concentrations equal to the solubility limits given the expected concentration of technetium in saltstone grout. When a finite element in the VZTM transitioned from chemically reduced to chemically oxidized, the model changed the representation of technetium mobility from the effective Kd that mimicked a solubility limit to a Kd that represented linear portioning of technetium in oxidized cementitious material (i.e., 0.5 mL/g). However, the real sections of saltstone or concrete corresponding to finite elements would not transition instantaneously. To represent a gradual transition, the DOE lowered the Kd as a function of the oxidized fraction of the element to create a smooth transition between reduced and oxidized conditions. Figure 12 below shows the difference between the modeled iodine release with a shrinking core model (i.e., the Compliance Case) and the model the DOE used for most elements (i.e., the pore volume exchange model, also referred to a non-shrinking core model in Figure 12). The Figure shows that the flux of iodine to the water table projected by the shrinking core model was similar to the flux projected by the non-shrinking core model for approximately three thousand years after closure and approximately 30% smaller than the flux projected by the non-shrinking core model at 10,000 years after closure. Figure 12. Sensitivity of Iodine Release to the Release Model (from Figure 5.8-34 in the 2020 SDF PA) Figure 13 below, shows the same comparison for technetium except that it shows results for two different solubility limits for technetium: 4.5 x 10-7 moles per L (mole/L) and 9.7 x 10-7 mole/L.

Section 6.6 in this TRR describes the technical basis for those solubility limits. Figure 13 shows that the projections of the shrinking core model were generally bounded by the projections of the non-shrinking core model with the higher and lower solubility limits, except between approximately 500 and 1,500 years after closure, when the shrinking core model projected greater releases than the non-shrinking core model with either solubility limit. In the 2020 SDF PA, the DOE indicated that the shrinking core model results in greater releases at early times because some of the modeled finite elements in the shrinking core model become oxidized before the whole saltstone monolith would become oxidized and those elements release technetium faster than technetium is released from chemically reduced saltstone. In contrast, the non-shrinking core models represent the saltstone monolith as entirely chemically reduced for the entire performance period under Compliance Case infiltration rates. Figure 13. Technetium Flux Sensitivity to Solubility Limit and Type of Release Model (from Figure 5.8-33 in the 2020 SDF PA) 7.3. Chemical Transitions The DOE expects transport properties to change with time because water flowing into saltstone will change the pore water chemistry in saltstone and concrete in ways that affect radionuclide sorption and solubility. Specifically, the DOE expects that: (1) inflowing water will dissolve the solid phases that control the pore water pH and (2) oxygen in the water will consume the chemical reducing capacity provided by blast furnace slag and thereby affect the Eh. To determine when chemical transitions would occur, the DOE developed the Contaminant Release Model in Geochemists Workbench software. The Contaminant Release Model projected chemical transitions as a function of pore volume exchanges. The DOE then used the VZFM to relate pore volume exchanges to time. The age of cementitious materials is often characterized in terms of the pore water pH, with young material having a very high pH that gradually drops due to chemical reactions as water flows through the material. Consistent with common practice, the DOE used the terms Region I through Region IV to refer to concrete with different projected pH values. Region I conditions

represent cementitious material with a pH > 12.5, which is typically controlled by dissolution of cement phases containing alkali impurities (e.g., Na+ and K+). Region II conditions correspond to a pH around 12.5, which is buffered by portlandite. Region III conditions correspond to a pH between 10 and 12.5, with pH buffering around 10 buffered by dissolution of Calcium-Silica-Hydride phases. Finally, Region IV conditions correspond to pH values below 10 that depend on buffering from a variety of solid phases. The Contaminant Release Model projected that saltstone and disposal structure concrete would spend very little time in Region II. Therefore, the DOE modeled both materials as transitioning directly from Region I conditions to Region III conditions. Although the Contaminant Release Model also projected transitions to Region IV conditions, the VZFM projected that the number of pore volume exchanges needed to cause transitions to Region IV would not take place for hundreds of thousands of years after site closure. Therefore, this TRR does not address transitions to Region IV conditions any further. The DOE also use the Contaminant Release Model to project the oxidation state of saltstone and concrete as a function of pore water exchanges. In general, a material is oxidized if the Eh is greater than zero and reduced if the Eh is below zero. Saltstone and components of the structure concrete that contain blast furnace slag initially have chemically reduced conditions and gradually become oxidized as oxygen in the incoming water consumes reducing capacity. For both the saltstone and disposal structure concrete, the Contaminant Release Model projected that incoming water would change the pH more quickly than it would consume the reducing capacity from the slag. That is, by the time saltstone transitioned from chemically reduced to chemically oxidized, it would already be in Region III conditions. Therefore, the DOE projected the following chemical transitions for saltstone: Region I Reduced Region III Reduced Region III Oxidized The only exceptions to that progression of chemical states are the sections of the disposal structures that are modeled as initially oxidized either because they do not contain blast furnace slag or as a modeling decision (see Section 6.4 in this TRR). For those sections, the VZTM represented two chemical states during the performance period: Region I Oxidized Region III Oxidized Table 5 in this TRR shows the pore volume exchanges the DOE used to model chemical transitions in the VZTM. Table 5. Pore Volume Exchanges to Chemical Transitions in the VZTM (adapted from Tables 4.4-56 and 4.4-57 in the 2020 SDF PA) Cumulative Pore Volume Exchanges Material Transition Pessimistic Compliance Realistic pH Region III 1(a) 1(a) 1(a) Saltstone Eh to Oxidizing 600 850 1,100 Disposal Structure pH to Region III 1(a) 1(a) 1(a) Concrete Eh to Oxidizing 3,415 4,000 4,570 (a) The VZTM represented the transition from Region I to Region III for technetium after 6 pore volume exchanges.

The values the DOE used in the VZTM include two simplifications. First, as shown in Table 5 above, the VZTM models the chemical transition from Region I Reduced conditions to Region III Reduced conditions after one pore volume for all elements other than technetium. The DOE characterized that as a modeling simplification because the DOE only had data to support the timing of the transition for technetium and iodine. The Contaminant Transport Model projected that saltstone would transition from Region I to Region III conditions after six pore volume exchanges (SRNL-STI-2018-00586 (ML22297A156)). That transition time is consistent with the observed transition of technetium solubility in a core of field-emplaced saltstone (SRR-CWDA-2018-00046 (ML20206L243)). In contrast, iodine release from cores of field-emplace saltstone in DLM measurements showed the greatest change after one pore volume exchange (SRR CWDA201800045 (ML20206L242)). Second, in the Contaminant Transport Model the DOE represented the initial conditions of the disposal structure concrete, unlike saltstone, with chemically reduced Region III conditions. Therefore, the Contaminant Release Model did not provide a projection of the transition from Region I to Region III conditions in disposal structure concrete. The 2020 SDF PA and the supporting document about the Contaminant Release Model (SRNL-STI-2018-00586) did not provide a reason for beginning the Contaminant Release Model simulations for disposal structure concrete in Region III conditions. The DOE characterized using one pore volume exchange as the transition point from Region I to Region III conditions in disposal structure concrete in the VZTM for radionuclides other than technetium as a modeling simplification. The DOE converted the pore volume exchanges in Table 5, above, into years after site closure based on flow rates projected by the VZFM. Because parts of the disposal structure (e.g., roof, walls, floors, mud mats) and the saltstone wasteform in different disposal structures have different volumes and experience different flow rates, the amount of time it would take for a pore volume exchange differs in different disposal structures. Table 6 below shows the projected chemical transition times for the roofs, walls, and floors of SDS 9 through SDS 12 for the Central Scenario cases. Additional detail for SDS 9 through SDS 12 and transition times for other disposal structure designs are provided in Tables 4.4-59 through 4.4-64 in the 2020 SDF PA6. 6 Tables 4.4-59 through 4.4-64 in the 2020 SDF PA list times for the lower mud mats in the 150-foot and 375-foot disposal structures to transition from chemically reducing to chemically oxidizing. However, Table 4.4-58 in the 2020 SDF PA indicates that the DOE modeled the lower mud mats as initially chemically oxidized (i.e., oxidized at the time of site closure), which is consistent with the composition of the lower mud mats (i.e., not containing blast furnace slag).

Table 6. Projected Chemical Transitions for SDS 9 through SDS 12 (from Tables 4.4-62 through 4.4.-64 in the 2020 SDF PA) Time after Closure (years) (a) Material Transition Realistic Compliance Pessimistic Reduced RI Reduced RIII 8.12 x 105 6.04 x 104 1.46 x 104 Saltstone Reduced RIII Reduced RIII > 1.00 x 108 5.38 x 107 7.34 x 106 Oxidized RIII Reduced RIV > 1.00 x 10 8 8.88 x 10 7 1.71 x 107 Reduced RI Reduced RIII 5.94 x 104 5.14 x 103 2.66 x 103 Clean Grout Reduced RIII Reduced RIII 6.72 x 10 7 3.77 x 10 6 5.79 x 105 Oxidized RIII Reduced RIV 8.50 x 10 7 6.26 x 10 6 1.33 x 106 Reduced RI Reduced RIII 3.85 x 103 7.93 x 102 7.38 x 102 Roof Reduced RIII Reduced RIII 1.61 x 10 7 9.97 x 10 5 1.88 x 105 Oxidized RIII Reduced RIV 2.69 x 10 7 1.90 x 10 6 4.13 x 105 Reduced RI Reduced RIII 8.12 x 105 6.04 x 104 1.46 x 104 Floor Reduced RIII Reduced RIII > 1.00 x 10 8 5.38 x 10 7 7.34 x 106 Oxidized RIII Reduced RIV > 1.00 x 10 8 8.88 x 10 7 1.71 x 107 Reduced RI Reduced RIII 8.12 x 105 6.04 x 104 1.46 x 104 Upper Mud Reduced RIII Reduced RIII > 1.00 x 10 8 5.38 x 10 7 7.34 x 106 Mat Oxidized RIII Reduced RIV > 1.00 x 10 8 8.88 x 10 7 1.71 x 107 (a) The NRC staff reduced the values the DOE provided to three significant digits. To test the importance of the chemical transitions, the DOE performed sensitivity analyses in which the sorption coefficients for saltstone and disposal structure concrete were assigned initial values corresponding to different chemical states. Those sensitivity analyses did not affect the modeled technetium solubility because the analyses only adjusted sorption coefficients and the VZTM represents technetium mobility in chemically reduced materials with a solubility limit rather than a sorption coefficient, as previously described. Figure 14 below shows the effects of using sorption coefficients corresponding to reduced Region II, reduced Region III, or oxidized Region III chemical conditions. As shown the Figure, the Compliance Case resulted in greater projected doses than any of the alternative cases in which cementitious materials began in later chemical states. The DOE attributed that result to greater iodine sorption in reduced Region II, reduced Region III, and oxidized Region III conditions as compared to reduced Region I conditions.

Figure 14. Dose Sensitivity to Modeling Initial Sorption Coefficients for Alternative Chemical Conditions (from Figure 5.8-38 in the 2020 SDF PA) 7.4. Chemical Reducing Capacity As described in Section 6.2, the DOE included blast furnace slag in saltstone and most components of the disposal structure concrete to create chemically reducing conditions that limit the mobility of certain radionuclides, including technetium. The following SDF components contain or will contain slag:

roofs of 150-foot and 375-foot disposal structures
clean grout
saltstone
disposal structure walls
disposal structure roof support columns
disposal structure floors
upper mud mats The following SDF components will not contain slag:
roofs of SDS 1 and SDS 4
lower mud mats The DOE modeled slag-bearing components as beginning in a chemically reduced state, except for the walls of SDS 1 and SDS 4. Although the walls of SDS 1 and SDS 4 contain slag, the DOE modeled those components as initially chemically oxidized because of cracks in the SDS 1 and SDS 4 walls.

The modeled reducing capacities of saltstone and slag-bearing concrete affect the projected chemical transition times addressed in Section 6.3 in this TRR because the DOE model projects that the Eh transitions would occur when oxygen in water consumes the reducing capacity in the

modeled component. Therefore, greater chemical reducing capacity would delay the modeled Eh transition. Table 7 below shows the chemical reducing capacities the DOE used in the 2020 SDF PA. Table 7. Initial Reducing Capacities (adapted from Table 4.3-9 in the 2020 SDF PA) Modeling Case Saltstone Reducing Disposal (meq e-/g) Structure Concrete (meq e-/g) Realistic Case 0.65 0.239 Compliance Case 0.50 0.209 Pessimistic Case 0.35 0.178 For disposal structure concrete, the DOE based the Realistic Case and Pessimistic Case values of the reducing capacity on two cerium method measurements of a simulated concrete sample made with 10% Holcim slag (SRNL-STI-2009-00637 (ML100550015) and SRNS-STI-2008-00045, (ML090150234). The first study reported a measured reducing capacity of 0.178 electrons (meq e-) per gram (g) and the second reported a reducing capacity of 0.239 meq e-/g. In SRNL-STI-2009-00637, the DOE attributed the difference to one of three possible explanations: (1) sample variability, (2) subjectivity introduced by a colorimetric method of determining the titration endpoint, or (3) differences in the particle sizes used in crushed samples. In the 2020 SDF PA, the DOE assigned the lower measured value as the Pessimistic Case value and the higher measured value as the Realistic Case value. The DOE used the average of the two measurements as the Compliance Case value. For saltstone, the DOE based modeled reducing capacities on measurements of multiple simulated saltstone samples. The DOE document SRRCWDA201800048, provides measurements of reducing capacity in simulated saltstone samples made with two different sources of slag: (1) Holcim, which the DOE used in the saltstone currently in SDS 1, SDS 4, SDS 2A, SDS 2B, SDS 3A, and SDS 3B, and (2) Lehigh, which the DOE plans to use in the saltstone in SDS 6 through SDS 12. The measurements showed that the simulated saltstone samples made with Lehigh slag had about twice the reducing capacity of the samples made with Holcim slag. The document SRR-CWDA-2018-00048, also compared results from two different methods of measuring reducing capacity: the cerium method and the chromium method. The NRC staff compared the applicability of both methods in a 2018 TRR entitled Update on the Projected Technetium Release from Saltstone, (ML18095A122). In general, the DOE found that the measurements made with the cerium method were approximately twice as large as measurements made with the chromium method for samples of the same materials. Table 8 below provides the ranges the DOE measured for combinations of slag source and measurement method. The DOE reported that the average of all the measured values was 0.614 meq e-/g and the midpoint between the highest and lowest values was 0.465 meq e-/g. Table 8. Reducing Capacity of Simulated Saltstone (adapted from Table 1 through Table 3 in SRR-CWDA-2018-00048, Rev. 0) Cerium Method Chromium Method (meq e-/g) (meq e-/g) Holcim Slag 0.391 to 0.849 0.316 to 0.360 Lehigh Slag 1.60 not reported

To develop the modeled reducing capacities for saltstone for the realistic, Compliance, and Pessimistic cases, the DOE treated the collection of measurements with multiple steps: (1) identify the lowest, median, and mean values of all the measurements (2) identify the lowest, median, and mean values of the subset of measurements for samples made with Holcim slag and measured with the cerium method (3) determine the Best Estimate value from the average the means of the groups from steps (1) and (2) (4) determine the Compliance Case value from the average the medians of groups from steps (1) and (2) (5) determine the Pessimistic Case value from the average of the lowest values of the groups in steps (1) and (2) Steps three through five of that DOE process double count the measurements of the samples made with Holcim slag and measured with the cerium method. The DOE did not provide a basis for double counting those values or for the use of the means, medians, and lowest values of the data sets in steps three through five, above. Sensitivity cases the DOE provided in the DOE document SRNL-STI-2018-00586 demonstrated that the projected number of pore volumes required for either saltstone or disposal structure concrete to transition from reduced to oxidized were roughly linearly related to the modeled reducing capacity. For example, for saltstone, increasing the modeled reducing capacity from the Pessimistic Case value (0.35 meq e-/g) to the Realistic Case value (0.65 meq e-/g) increased the modeled number of pore volume exchanges before the oxidation state transition from approximately 600 to approximately 1,100. The DOE modeled a similar relationship between the number of pore volumes before the oxidation state transition and reducing capacity for disposal structure concrete. For example, for disposal structure concrete, the DOE projected that increasing the modeled reducing capacity from the Pessimistic Case value (178 meq e-/g) to the Realistic Case value (239 meq e-/g) would increase the projected pore volume exchanges before an Eh transition from 3,200 to 4,600. The DOE result that the oxidation state transition would require more pore volume exchanges in concrete than saltstone could appear to be counterintuitive because of the larger measured reducing capacity of saltstone compared to concrete. The DOE indicated the result is attributable the much lower porosity of disposal structure concrete (i.e., approximately 0.11) as compared to saltstone (i.e., approximately 0.66). The smaller porosity of the disposal structure concrete means that each pore volume contains less water and, therefore, less oxygen than a pore volume in saltstone. Therefore, it takes more pore volume exchanges in disposal structure concrete to consume the same amount of reducing capacity as a pore volume exchange could consume in saltstone. The DOE expects that iodine mobility will decrease under oxidizing conditions and technetium mobility will increase (see Sections 6.5 and 6.6 in this TRR). That opposing effect when the Eh changes from reduced to oxidized, in addition to the complexity of the shrinking core model, make the effects of the reducing capacity in saltstone and disposal structure concrete difficult to predict without running the VZTM. The DOE did not conduct deterministic sensitivity analyses for different assumed values of the reducing capacity of saltstone. However, the DOE analyzed

a sensitivity case in which the disposal structure concrete and a 30-cm (1-ft) thick outer layer of saltstone was fully oxidized (i.e., the oxidized rind case). That case shows the potential effect of a lower saltstone reducing capacity because it mimics the effect of faster oxidation. In the oxidized rind case, the projected contribution of Tc-99 to the peak dose to an offsite member of the public within 10,000 years of site closure increases by approximately 40%, as compared to the Compliance Case. In contrast, the peak contribution from I-129 decreases by approximately 50%. Because of the opposing effects of the redox state on technetium and iodine mobility in saltstone and concrete, the net effect of assuming an oxidized rind on the projected dose is small, as shown in Figure 15, below. Figure 15 also shows the results of a third sensitivity case referred to as the KdRe = 0.5 case. That case is more complex, as described further below. Figure 15. Projected Dose to an Offsite Member of the Public for Different Reducing Capacity Assumptions (from Figure 5.8-53 in the 2020 SDF PA) As described in Section 6.2 in this TRR, the DOE used effective Kd values to mimic the effect of implementing a solubility limit for technetium in chemically reduced saltstone and concrete. Because a Kd value is a proportionality constant between the concentration of an element in the solid phase and aqueous phase, the DOE could use an effective Kd value to set the concentration of an element in the aqueous phase based on its concentration in the solid phase. For example, in saltstone, the DOE chose the effective Kd value so that the aqueous concentration of technetium calculated based on the effective Kd value would equal the aqueous concentration of technetium determined by the solubility limit. However, for disposal structure concrete, the initial concentration of technetium in the solid phase is zero. Because of that initial zero concentration, the DOE could not develop an effective Kd value that would result in a non-zero aqueous concentration of technetium. Therefore, in the Compliance Case, the DOE modeled the sorption coefficient for technetium in reduced concrete with a value the DOE considered conservative and bounding (i.e., 0.01 mL/g).

However, upon review of the Compliance Case results, the DOE noted the use of an effective Kd of 0.01 mL/g to represent technetium mobility in chemically oxidized concrete led to the counterintuitive effect that the model projected faster technetium transport in chemically reduced concrete than in chemically oxidized concrete. The DOE conducted the TC KdRE=0.05 sensitivity case to determine the effect of assigning an effective Kd value for technetium in chemically reduced concrete equal to its value in chemically oxidized concrete (i.e., 0.05 mL/g). Figure 15 above shows that much of the difference between the Compliance Case results and the results of the Oxidized Rind sensitivity analysis can be attributed to assigning a sorption coefficient of 0.5 mL/g to technetium in the disposal structure concrete. 7.5. Iodine Sorption The DOE used different sorption coefficients to represent iodine sorption in saltstone and concrete. The DOE based the sorption coefficients for iodine in chemically reduced saltstone on measurements made with simulated saltstone. The DOE based sorption coefficients for iodine in oxidized saltstone, reduced concrete, and oxidized concrete on literature values. As described in Section 6.3 in this TRR, the DOE expects that only three combinations of oxidation state and age will occur in saltstone during the performance period: reduced Region I, reduced Region III, and oxidized Region III. The DOE expects iodine to take different chemical forms under reducing and oxidizing conditions. Under reducing conditions, the DOE expects iodine to exist only as iodide (I-). Under chemically oxidizing conditions in saltstone, the DOE expects iodine to exist as both iodide and iodate (IO3-). Because of that chemical change, the DOE expects iodine to have a large sorption coefficient in chemically oxidized saltstone than it does in chemically reduced saltstone. Table 9 below shows the Kd values the DOE used to represent iodine sorption in saltstone and concrete for those chemical conditions. Table 9. Modeled Iodine Sorption Coefficients in Saltstone and Concrete (adapted from Tables 4.3-5 through 4.3-7 in the 2020 SDF PA) Chemical Condition Central Scenario Reduced Reduced Oxidized Material Case Region I Region III Region III Realistic 0.16 2.77 4 Saltstone Compliance 0.07 0.71 4 Pessimistic 0.07 0.71 4 Disposal Realistic 0 0 4 Structure Compliance 0 0 4 Concrete Pessimistic 0 0 4 The DOE based the sorption coefficients for iodine in chemically reduced Region I and Region III saltstone on measured iodine concentrations in the effluent of DLM tests conducted with simulated saltstone samples. The NRC staff described the DLM in a TRR entitled Updated on the Projected Technetium Release from Saltstone. In brief, the DLM measures the concentration of a radionuclide in a flow of water passed through a grout sample under pressure. The DOE use a GoldSim optimization to fit sorption coefficients to plots of iodine released per pore volume exchanged (SRRCWDA201800045).

The DOE based the sorption coefficient for iodine in oxidized Region III saltstone and concrete on literature values of iodine sorption in concrete in which the best estimate values ranged from 1 mL/g to 8 mL/g (SRNL-STI-2009-00473, Rev.1 (ML113320386)). The DOE used expert judgement to select a value of 4 mL/g. In an RAI, the NRC staff requested additional information about the applicability of measurements made with concrete to iodine sorption in saltstone. Specifically, the NRC staff asked for additional information about the potential effect of the calcium to silicon ratio (Ca/Si ratio) and the lack of sand or aggregate in saltstone. In response, the DOE indicated that Kd values for iodine in chemical oxidized concrete were typically 8 to 40 times greater than Kd values in similar samples under chemically reduced conditions (SRR-CWDA-2021-00072, Rev.1 (ML21148A005)). The DOE RAI response indicated that the DOE regarded the Kd value of 4 mL/g as conservative because it was lower than the value would be if the DOE scaled the Kd value for iodine in reduced saltstone (i.e., 0.71 mL/g) up by a factor of 8 to 40. The DOE RAI response also indicated that the effect of the iodine Kd on iodine mobility in saltstone and disposal structure concrete under oxidized Region III conditions could be demonstrated by comparing the projections of the VZTM and the GoldSim model. The DOE indicated the two model projections bounded the effect of the oxidizing Region III Kd for iodine because the VZTM used a shrinking core model for iodine release. In the shrinking core model, some parts of the saltstone would reach oxidizing Region III conditions within 10,000 years of SDF closure. In contrast, the GoldSim model used a non-shrinking core model for iodine release, in which the entire saltstone monolith transitioned from reducing to oxidizing well after the end of the 10,000 year performance period. For the three Central Scenario cases, the DOE showed that the GoldSim projection of the dose attributable to I-129 was approximately a factor of 2 greater than the PORFLOW projection for the Compliance Case and the Pessimistic case, and a factor of 3 greater in the Realistic Case (see Table 10 below). Table 10. Peak Dose Contributions from I-129 within 10,000 Years of Closure (adapted from Table RR-2.5 in DOE document SRR-CWDA-2021-00072) 2020 SDF PA Model Using GoldSim Model Modeling Case the Vadose Zone Transport (mrem/yr)(a) Model(mrem/yr) (a) Realistic Case 0.1 0.3 Compliance Case 0.6 1.3 Pessimistic Case 6.6 12.5 (a) To convert millirem to millisievert, multiply by 100. Figure 16 below shows the DOE projected concentration of I-129 in SDS 7 at 10,000 years after site closure. In an RAI Question, the NRC staff asked for additional information to support the DOE projection of a high concentration of I-129 in the lower mud mat. The NRC staff indicated that comparison of the diagrams of total I-129 concentration and aqueous I-129 concentration in Figure 5.8-83 in the 2020 SDF PA showed that the high concentration of total I-129 in the lower mud mat corresponded to an area of low aqueous I-129 concentration, and therefore indicated that the high I-129 concentration occurred in the solid phase. The NRC staff questioned whether the conversion from reduced iodine, with a Kd value of 0 mL/g in concrete, to oxidized iodine, with a Kd value of 4 mL/g in concrete, could take place quickly enough so that the iodine would be converted as it flowed through the lower mud mat. In response (SRMC-CWDA-2022-00003 (ML22083A049)), the DOE described the oxidation states of iodine and indicated that the DOE expected oxidized iodine to have a lager sorption coefficient than reduced iodine. The DOE response did not address the expected speed of the chemical oxidation of iodine in disposal

structure concrete or compare that speed to the expected speed of water flow through the lower mud mats. Figure 16. Total I-129 Concentration in Part of SDS 7 at 10,000 years (adapted from Figure RR 1.1 in DOE document SRMC-CWDA-2022-00003) 7.6. Technetium Solubility and Sorption The VZTM represents aqueous concentrations of technetium in saltstone and cementitious material pore water as being solubility limited under chemically reducing conditions and controlled by a sorption coefficient in chemically oxidized conditions. Table 11 below shows the solubility limits and sorption coefficient the DOE used to represent technetium release under the chemical conditions the DOE expected to occur in saltstone within 10,000 years of SDF closure. Table 11. Solubility Limits and Sorption Coefficients for Technetium (adapted from Tables 4.3-5 and 4.3-8 in the 2020 SDF PA) Central Scenario Case Material Conditions Controlling Realistic Compliance Pessimistic Property Saltstone and Reduced Solubility 2.0 x 10-7 9.7 x 10-7 (a) 9.7 x 10-7 Concrete Region I (mole/L) Saltstone and Reduced Solubility 7.4 x 10-8 4.5 x 10-7 (b) 9.7 x 10-7 Concrete Region III (mole/L) Oxidized 0.8 0.8 0.8 Concrete Only Kd (mL/g) Region I Saltstone and Oxidized 0.5 0.5 0.5 Kd (mL/g) Concrete Region III (a) Implemented with an initial effective K of 20 mL/g (Table 1.4-1 in the 2020 SDF PA). d (b) Implemented with an initial effective K of 43 mL/g (Table 1.4-1 in the 2020 SDF PA). d The DOE based technetium solubility limits in chemically reduced saltstone on DLM tests conducted with samples of field-emplaced saltstone and saltstone simulants. The DOE used a GoldSim model to fit solubility limits a graph of measured aqueous technetium concentrations

as a function of the number of pore volume exchanges (SRRCWDA201800046). Based on that optimization, the DOE determined that the VZTM should use one solubility limit for reduced Region I material and a lower solubility limit when the material transitioned to reduced Region III conditions. The DOE analyzed the sensitivity of the projected dose to an offsite member of the public to the modeled technetium solubility using two sets of infiltration rates: (1) the infiltration rates for the Compliance Case in the 2020 SDF PA, and (2) the infiltration rates the DOE modeled in response to the NRC Request for Supplemental Information (RSI). As previously discussed, Figure 13 above shows the effect of the technetium release model and solubility limit on the projected technetium flux to the saturated zone for the Compliance Case. Figure 17 below shows similar information with the higher infiltration rate the DOE used in response to the NRC RSI. Figure 17. Sensitivity of Technetium Release to Solubility in the DOE Model for the DOE RSI Response (Figure SUA-2.10 in DOE document SRMC-CWDA-2022-00003) Both graphs in Figure 17 show that in the non-shrinking core model, assuming a higher solubility limit increases the projected technetium flux and choosing a lower solubility limit decreases the projected flux. Both graphs also show that the shrinking core model (i.e., labeled as Compliance Case in Figure 13 and SUA-2 Base Case in Figure 17) produces projections that fall between the high and low solubility non-shrinking core model projections for most of the time between closure and 10,000 years after closure.

7.7. Sorption Coefficients for Radionuclides Other Than Technetium and Iodine For most elements, the DOE applied the same sorption coefficients in saltstone grout and concrete. The exceptions are iodine, radium, and strontium, for which the DOE developed different sorption coefficients for saltstone grout and concrete. Table 12 below shows the sorption coefficients the DOE used for concrete in the 2020 SDF PA. The sorption coefficients for iodine, radium, and strontium in saltstone are provided in Table 13 below. Table 12. Sorption Coefficients (adapted from Table 4.3-5 in the 2020 SDF PA) Reducing Environment (Eh < 0 mV) Oxidizing Environment (Eh > 0 mV) Element (a) Region I Region II Region III Region I Region II Region III Actinium (Ac) 7,000 7,000 1,000 6,000 6,000 600 Aluminum (Al) 7,000 7,000 1,000 6,000 6,000 600 Americium (Am) 7,000 7,000 1,000 6,000 6,000 600 Carbon (C) 2,000 5,000 50 2,000 5,000 50 Californium (Cf) 7,000 7,000 1,000 6,000 6,000 600 Chlorine (Cl) 0 10 1 0 10 1 Curium (Cm) 7,000 7,000 1,000 6,000 6,000 600 Cobalt (Co) 5,000 5,000 1,000 4,000 4,000 400 Cesium (Cs) 2 20 10 2 20 10 Europium (Eu) 7,000 7,000 1,000 6,000 6,000 600 Iron (Fe) 7,000 7,000 1,000 6,000 6,000 600 Hydrogen (H) 0 0 0 0 0 0 Mercury (Hg) 500 500 100 300 300 100 Iodine (I) (b) 0 2 0 8 10 4 Potassium (K) 2 20 10 2 20 10 Niobium (Nb) 1,000 1,000 500 1,000 1,000 500 Nickel (Ni) 70 400 400 70 400 400 Neptunium (Np) 10,000 10,000 5,000 10,000 10,000 5,000 Protactinium (Pa) 10,000 10,000 5,000 10,000 10,000 5,000 Lead (Pb) 5,000 5,000 1,000 300 300 100 Platinum (Pt) 5,000 5,000 1,000 4,000 4,000 400 Plutonium (Pu) 10,000 10,000 2,000 10,000 10,000 2,000 Radium (Ra) (b) 6,000 6,000 600 200 100 200 Radon (Rn) 0 0 0 0 0 0 Antimony (Sb) 200 200 100 200 200 100 Selenium (Se) 300 300 200 3 3 3 Samarium (Sm) 7,000 7,000 1,000 6,000 6,000 600 Tin (Sn) 4,000 4,000 2,000 4,000 4,000 2,000 Strontium (Sr) (b) 1,000 1,000 100 90 20 90 Tc (c) NA NA NA 0.8 0.8 0.5 Thorium (Th) 10,000 10,000 2,000 10,000 10,000 2,000 Uranium (U) 5,000 5,000 5,000 1,000 5,000 5,000 Zirconium (Zr) 10,000 10,000 2,000 10,000 10,000 2,000 (a) This table includes elements corresponding to radionuclides included in the 2020 SDF PA. This table excludes chemical constituents included in Table 4.3-5 in the 2020 SDF PA that do not correspond to radionuclides included in the 2020 SDF PA.

(b) For these elements, the sorption coefficients in this table only apply to disposal structure concrete. Table 13 below provides sorption coefficients for these elements in saltstone. (c) In the 2020 SDF PA, the DOE represented the mobility of technetium in chemically reduced saltstone and concrete with solubility limits. Table 13. Sorption Coefficients for Strontium and Radium in Saltstone (adapted from Table 4.3-6 in the 2020 SDF PA) Reducing Environment (Eh < 0 mV) Oxidizing Environment (Eh > 0 mV) Element Region I Region II Region III Region I Region II Region III Iodine 0.07 0.71 0.71 8 10 4 Radium 6,000 6,000 600 6,000 6,000 600 Strontium 1,000 1,000 100 1,000 1,000 100 Because the NRC staff reviewed grout and concrete sorption coefficients in the 2012 NRC TER, this review focuses on the differences between the 2009 SDF PA and the 2020 SDF PA values and issues highlighted in the 2012 NRC TER. Table 14 below provides a comparison between the sorption coefficients for elements in concrete that the DOE used in the 2009 SDF PA and the 2020 SDF PA. The table focuses on the four chemical environments the DOE modeled in concrete in the 2020 SDF PA. Table 14. Comparison of Sorption Coefficients in SRS Concrete used in the 2009 SDF PA and the 2020 SDF PA Reduced Oxidized Ratio of 2020 Value to (Eh < 0 mV) (Eh < 0 mV) 2009 Value Region I Region III Region I Region III Cl, Hg, I, Ni, Cl, Hg, Ni, Cl, Co, Hg, 0.5 Cl, Ni, Pb, Se Sb, Tc Pu, Sb, Tc Ni, Pb, Se Ac, Al, Cf, Ac, Al, Cf, Ac, Al, Cf, Cm, Co, Cr, Cm, Co, Cs, Cm, Cr, Cs, Co, Nb, Pu, Cs, Eu, Fe,

          > 0.5 and  1                            Eu, Fe, K,                       Eu, Fe, I, K, Se, Sn                        Hg, I, K, Nb, Nb, Se, Sm,                       Nb, Sm, Sn, Pb, Pu, Sm, Sn                               Tc Sn, Tc Ac, Al, Am, Cf, Cm, Eu,                       Ra, Sb, Th,
           > 1 and  2                               Np, U                               Pu Fe, Pa, Sm,                            Zr Th, U, Zr Pa, Pb, Sr,
          > 2 and  10              Np, Pb                          Np, Pa, U      Ra, Sr, Th, Zr Th, Zr Np, Pa, Sb,
         > 10 and  100                C                Ra             C, Sr U
       > 100 and  1,000             none             none             none             none
     > 1,000 and  12,000           Sr, Ra            none             none             none Undefined Ratio: Zero Kd in the 2009 PA and            Cs (2)

C (50) C (50) non-zero Kd in the 2020 K (2) Pt (4,000) Pt (1,000) Pt (400) PA (2022 Kd values in Pt (5,000) parenthesis in mL/g) (a) Ratios calculated by NRC staff based on information in Table 4.3-5 in the 2020 SDF PA and Table 4.2-18 in the 2009 SDF PA

The DOE conducted a deterministic sensitivity analysis with the SDF GoldSim Model to provide information about the risk significance of sorption coefficients for the SDF. In that analysis, the DOE reduced all sorption coefficients in the model (i.e., including the disposal structure concrete, saltstone, and all soil types) by a factor of 100. That analysis showed that the reduction had a small effect on the projected dose to an offsite member of the public (Figure 18 below). The DOE found that most of the difference between the two lines in Figure 18 could be attributed to the change in the iodine sorption coefficients in soil. Therefore, Figure 18 shows that reducing sorption coefficients of radionuclides in saltstone or cementitious concrete by a factor of 100 would not significantly change the projections of the 2020 SDF PA. Figure 18. Deterministic Results of Reducing All Sorption Coefficients (Kd values) in the SDF GoldSim Model (from Figure 5.8-35 in the 2020 SDF PA) Table 14 above shows that the modeled Kd values of strontium and radium increased by a factor of more than 100 from the values the DOE used in the 2009 SDF PA. For those elements, the DOE basis for the revised sorption coefficients were high Kd values in experiments in cementitious materials containing slag. The DOE document that provided that technical basis (SRR-STI-2009-00473, Rev. 1) indicated that the DOE expected those elements to precipitate with sulfide in reduced cementitious material and that those precipitates would remain intact when the cementitious material oxidized. Table 14 above also shows the modeled Kd values for four elements increased from a zero value in the 2009 SDF PA to a non-zero value in the 2020 SDF PA: carbon, cesium, platinum, and potassium. For carbon, the DOE increased the Kd in reduced and oxidized Region III materials to 50 mL/g based on studies of carbon partitioning in concrete that were conducted while or slightly after the 2009 SDF PA was being prepared. For cesium, the DOE increased the modeled Kd value in reduced Region I material based on measurements of cesium sorption in cores of field-emplaced saltstone. For platinum, the DOE increased the modeled Kd value in cementitious material for all chemical conditions based on its chemical similarity to other metals for which the DOE observed precipitation with sulfides in cementitious material containing blast

furnace slag. For potassium, the DOE increased the modeled Kd value for reduced Region I conditions based on its chemical similarity to cesium. 7.8. Potential New Saltstone and Disposal Structure Concrete Formulations After the DOE prepared the 2020 SDF PA, the DOE evaluated the potential use of new formulations of both saltstone grout and disposal structure concrete (SRR-CWDA-2020-00064, Rev.1 (ML21232A639)). Table 15 below provides the proposed change to saltstone grout. For disposal structure concrete, the DOE has already approved multiple formulations. Table 16 below compares three formulations that have already been used at the SDF with a potential new formulation referred to as Mix 3B. Table 15. Saltstone Grout Dry Mix (from Table 2.3-5 in DOE document SRR-CWDA-2020-00064, Rev. 1) Material Formulation Evaluated in Cement-Free Formula the 2020 SDF PA (weight percent of dry mix) (weight percent of dry mix) Slag Grade 100 or 120 45 60 Class F Fly Ash 45 40 Ordinary Portland Cement 10 0 Table 16. Disposal structure Concrete Dry Mix (from Tables 2.3-1 through 2.3-3 in DOE document SRR-CWDA-2020-00064, Rev. 1) Material Quantity per Cubic Yard (a) Mix E 6000 Mix E 6000 Mix D 5000 Mix 3B PS-2-ABC PS-3-ABC S-2-AB (Wall, Column, (Floor) and Roof) Type I/II Cement 0 lbs 0 lbs 0 lbs 337 lbs Type V Cement 213 lbs 213 lbs 213 lbs 0 lbs Slag Grade 100 0 lbs 0 lbs 284 lbs 0 lbs Slag Grade 120 284 lbs 284 lbs 0 lbs 284 lbs Class F Fly Ash 163 lbs 163 lbs 163 lbs 0 lbs Silica Fume 50 lbs 50 lbs 50 lbs 0 lbs Sand 1091 lbs 1046 lbs 991 lbs 1020 lbs Metakaolin 0 lbs 0 lbs 0 lbs 89 lbs Aggregate (#4 0 lbs 435 lbs 0 lbs 0 lbs Stone) Aggregate (#8 0 lbs 0 lbs 0 lbs 481 lbs Stone) Aggregate (#57 0 lbs 0 lbs 0 lbs 1369 lbs Stone) Aggregate (#67 1750 lbs 1360 lbs 1850 lbs 0 lbs Stone) Air Entraining 5.5 oz 5.5 oz 3.91 oz 5.5 oz Admixture Water Reducing 0 oz 0 oz 42.6 oz 0 oz Admixture

High Range Water Reducing 56.8 oz 63.9 oz 34.1 oz 53.3 oz Admixture Shrinkage Reducing 96.0 oz 96.0 oz 0 oz 96.0 oz Admixture Potable Water 264 lbs 264 lbs 269 lbs 276.9 lbs (a) To convert pounds per cubic yard to kg per cubic meter, multiply by 0.593. To convert ounces per cubic yard to kg per cubic meter, multiply by 0.0371. To determine the effect changing the saltstone formulation would have on radionuclide release from saltstone, the DOE compared leaching test results from simulated saltstone made with the current saltstone premix and the cement-free formula (SREL Doc No. R-20-0002, Rev.1 (ML19303A022)). The DOE used the Environmental Protection Agency (EPA) Method 1315, which tests diffusive release, and the DLM, which tests advective release, to compare the materials. Table 17 below compares effective diffusivities measured for four chemical species in samples of field-emplaced saltstone, simulated saltstone with the current dry mix formula, and simulated saltstone made with the cement-free formula. Figure 19 below and Figure 20 below compare technetium and iodine leaching using the DLM method in simulated samples with the current premix (i.e., L45:45:10 in the figure legend) and the cement-free formula (i.e., L60:40 in the figure legend). Table 17. Effective Diffusivities in Saltstone Cores and Simulants (from Table 6 in DOE document SREL DOC No. R-20-0002, Rev. 1) Effective Diffusivity (cm2/s) Sample Type Technetium Nitrate Iodine Sodium Current Premix Simulant (2019 5.7 x 10-11 7.4 x 10-9 4.4 x 10-9 2.7 x 10-9 data) Cement-free Premix (2019 5.3 x 10-11 9.2 x 10-9 3.3 x 10-9 3.4 x 10-9 data) SDS 2A Sample A 6.4 x 10-11 1.3 x 10-8 1.0 x 10-8 NA(a) SDS 2A Sample B 5.8 x 10 -11 4.4 x 10 -9 2.5 x 10 -9 NA(a) SDS 2A Sample C 5.2 x 10 -11 5.5 x 10 -9 5.5 x 10 -9 NA(a) (a) Measurements were not available.

Figure 19. Cumulative Technetium Leached with the DLM in Duplicate Simulants using the Current Premix (L45:45:10) and the Cement-Free Formula (L60:40) (from Figure 9 in DOE document SREL DOC No. R-20-0002, Rev. 1). Figure 20. Cumulative Iodine Leached with the DLM in Duplicate Simulants using the Current Premix (L45:45:10) and the Cement-Free Formula (L60:40) (from Figure 11 in DOE document SREL DOC No. R-20-0002, Rev. 1). In addition to directly measuring leaching, the DOE also compared the reducing capacity of saltstone simulants made with the current premix and the cement-free saltstone formula. The DOE made one sample with each formula and split each into three subsamples for measurement. The DOE found the subsamples made with the current formula had had an average reducing capacity of 0.640 meq e-/g and subsamples made with the cement-free formula had an average reducing capacity of 0.625 meq e-/g.

Because the leaching measurements in SREL Doc. No. R-20-0002, Rev. 1 only reflect leaching in chemically reduced samples, the NRC staff asked RAI Questions about iodine and technetium sorption under chemically oxidized Region III conditions for cement-free saltstone. For iodine, the DOE based its response on the calcium (Ca) to silicon (Si) ratio in grout, which was previously shown to be important to iodine sorption (SRMC-CWDA-2022-00003). The DOE showed that the Ca/Si ratios for saltstone simulants made with either premix are more alike than they are to the Ca/Si ratio in a grout sample made without slag. Then, the DOE indicated that measured sorption coefficients in the grout made without slag were all greater than 4 mL/g. The DOE also indicated the slag-free grout had been kept in oxidizing conditions. Therefore, because the DOE expects: (1) the higher Ca/Si ratio in cement-free saltstone compared to slag-free grout should increase iodine sorption and (2) measured iodine sorption in slag-free grout exceeded 4 mL/g, the DOE determined that using a sorption coefficient of 4 mL/g to represent iodine sorption in oxidized Region III cement-free saltstone was a conservative choice. For technetium, the DOE indicated that the value applied in oxidized Region III saltstone (i.e., 0.5 mL/g) is so small that lowering the value further makes a small change to the projected dose. The DOE also indicated that setting the value to 0 mL/g would increase the technetium release rate under oxidized Region III conditions by a maximum of a factor of 1.7 (SRMC-CWDA-2022-00003).

8. NRC Evaluation 8.1. LLDL, HDPE/GCL Composite Barrier, and GCL Layer In the TRR entitled Performance of the Composite Barrier Layers and Lateral Drainage Layers (ML23017A089), the NRC staff provided a detailed technical review of the LLDL and HDPE/GCL composite barriers in the LLDL and between the upper and lower mud mats. In that TRR, the NRC staff concluded:

The NRC staff determined that the calculations and models used to calculate and simulate flow rates through the LLDL, and mud mat barriers is adequate for modeling the projected dose from the SDF for the purpose of the DOE demonstrating compliance with the 10 CFR 61.41 and 10 CFR 61.42 [performance objectives] in the 2020 SDF PA. The NRC staff made that determination because the technical basis and justification associated with those features is sufficient in relation to their significance to performance. As simulated and presented in the 2020 SDF PA, the ULDL barrier is the dominant barrier that reduces flow between the upper and lower portions of the closure cap (i.e., the input and output of the Closure Cap Model) by many orders of magnitudes [sic.]. As such, the LLDL and mud mat barriers are best seen as backup barriers; however, if the compliance cases should change to include significant flow through the LLDL barrier and significant surface area of saltstone being exposed to fast pathways through the disposal structures, then the adequacy of these barriers for modeling the projected dose from the SDF would need to be reassessed. In this TRR, the NRC staff reviewed the VZFM output files to evaluate the performance of the LLDL and HDPE/GCL composite barrier with respect to near field flow. The volumetric flow rates through the LLDL and disposal structure roofs are shown below in Table 18 for three different cases: the DOEs Compliance Case, the Soil-Only Case, and a case that the DOE developed in response to NRC question RSI-1 (ML20254A003). In addition, the NRC staff compared the ratio of flow entering the LLDL and roof. The comparison of water

entering these layers provides insight into the performance of the LLDL and HDPE/GCL Composite Barrier. For example, if all of the water that enters the LLDL enters the roof, then the LLDL is not a barrier to flow. If, however, a significant amount of water enters the LLDL and is diverted around the disposal structure roof, then the LLDL is acting as a significant barrier to flow. Note that the ratio for the Compliance Case is greater than 1, indicating that more water is flowing through the roof than the LLDL. This is because of lateral flow from the backfill and into the LLDL. Table 18. Modeled Performance of the LLDL and HDPE/GCL Composite Barrier For the Compliance Case, the LLDL and HDPE/GCL composite barrier do not act as a barrier to flow. Effectively, all of the water that reaches the LLDL reaches the roof (plus additional water from the adjacent backfill layer) at 100, 1,000, and 10,000 years after SDF closure. The NRC staff expects that result occurs because the LLDL is a redundant barrier to the ULDL. In other words, the LLDL can transmit all of the water that is transmitted through the ULDL without diverting it. For the Soil-Only Case, which is described in Sections 4.6.5 and 5.8.2.4 of the 2020 SDF PA, the DOE assumed that there is no closure cap and that the percolation rate equals the natural percolation rate of 41.8 cm/yr (16.5 in/yr). Although the DOE assumed that there was no Closure Cap in that case, including no ULDL and upper HDPE/GCL composite barrier, the LLDL and HDPE/GCL composite barrier was still assumed to perform. The ratios in Table 18 for the Soil-Only Case show that the LLDL and HDPE/GCL composite barrier shed the overwhelming majority of the percolation. Even at 10,000 years, the LLDL and HDPE/GCL composite barrier is assumed to shed almost 98% of the percolation. As discussed below in this section, the NRC staff determined that there is insufficient support for the performance of the drainage layers and

composite barriers. Because of that assumed performance and the limited support of the LLDL and HDPE/GCL composite barrier, the NRC staff did not rely upon the Soil-Only Case in evaluating SDF performance. In response to NRC RSI-1 (ML20254A003), the DOE provided a revised probabilistic analysis incorporating combined uncertainties with several key barriers. The key barriers included the ULDL and LLDL, upper and lower HDPE/GCL composite barriers, saltstone, and MCCs. One of the realizations from the DOE RSI-1 analysis (Realization 0757) showed that the LLDL and the HDPE/GCL composite barrier were not an impediment to flow. However, that realization, as well as many other realizations that encompassed the uncertainty in key barriers did not meet the performance objectives. Additional discussion regarding the DOE probabilistic analysis in response to RSI-1 is provided below in Section 8.4. 8.2. Disposal Structure Design and Performance In the 2013 NRC SDF Monitoring Plan (ML13100A113), the NRC staff discussed that the DOE assumed disposal structure performance was important for SDF performance for two primary reasons. The first reason was the significant diversion of water around the disposal structures due to the large difference in hydraulic conductivity between the LLDL and the disposal structure roofs and the HDPE/GCL composite layer. The second reason was that the DOE modeled release of some key radionuclides into the natural environment was significantly affected by retention in disposal structure concrete. In the DOE Cases K, K1, and K2 in the 2009 SDF PA, the peak dose from Tc-99 was significantly delayed and reduced by the assumed retention in disposal structure concrete. In the 2012 NRC TER, the NRC staff determined that this degree of Tc retention and unexpected reconcentration in disposal structure concrete was inconsistent with the presence of fast pathways through fractures in the concrete or joints between disposal structure components (e.g., floors, walls). The NRC staff expects that fast pathways could affect contaminant release due to increased flow rates and reduced sorption sites relative to an intact matrix. To address concerns related to disposal structure performance, the NRC staff identified five monitoring factors under Monitoring Area 6 (Disposal Structure Performance):

MF 6.01 (Certain Risk-Significant Kd Values in Disposal Structure Concrete)
MF 6.02 (Tc Sorption in Disposal Structure Concrete)
MF 6.03 (Performance of Disposal Structure Roofs and HDPE/GCL Layers)
MF 6.04 (Disposal Structure Concrete Fracturing)
MF 6.05 (Integrity of Non-cementitious Materials)

In Section 7.1.4 of the 2020 SDF PA and reproduced below in Figure 21 and Figure 22, the DOE provided graphs for the Compliance Case illustrating the release of I-129 and Tc-99 from saltstone, through the disposal structures, into the vadose zone soils, and then to the water table. The dose is directly related to the flux to the water table and the flux to the water table is related to the slope of the contaminant release curves. As shown in Figure 21 and Figure 22 below, the slope of the curves for saltstone, the disposal structures, and vadose zone soils are largely parallel. That shows that I-129 and Tc-99 are being released from the disposal structure and vadose zone at rates similar to their rate of release from the saltstone grout and are not building up in the disposal structures or vadose zone soils. Coupled with the flow data shown in Figure 28, which indicates that the disposal structures do not limit flow, the disposal structures and the vadose zone soils are not significantly impacting the magnitude of the dose.

Also in Figure 21 below, the black line for the release from saltstone to SDS 9 and the brown line for the release from SDS 9 to the vadose zone soils are in close proximity. If the disposal structure was not a barrier to flow or transport, then these lines would be overlapping, which was the case in the FY 2013 and FY 2014 SDF Special Analysis documents. The separation between the lines shows that the disposal structure and the vadose zone soils delay contaminant transport. For example, from closure of the SDF to 10,000 years, the disposal structure delays the I-129 release by up to 3,000 years. For Tc-99 during the same time period, the delay is approximately 1,000 years. Figure 21. Release and Transport of I-129 from SDS 9 for the Compliance Case (adapted from DOE document SRR-CWDA-2019-00001)

Figure 22. Release and Transport of Tc-99 from SDS 9 for the Compliance Case (adapted from DOE document SRR-CWDA-2019-00001) In the 2020 SDF PA, the DOE assumed that the disposal structures beneath the HDPE/GCL composite layer will not significantly affect the magnitude or timing of the peak dose. Accordingly, the NRC staff risk informed the review of disposal structure concrete performance by focusing on initial hydraulic properties, concrete degradation, and fast flow paths. 8.2.1. Disposal Structure Initial Properties The NRC staff found the values for the initial hydraulic conductivity and diffusivity values for disposal structure concrete to be acceptable, because of: (1) the projected limited risk significance of the disposal structure concrete performance, (2) the use of laboratory studies to support the assumed values, and (3) consistency with the range of values observed in the literature for intact concrete (CNWRA, 2008). The DOE projections of limited risk significance of the disposal structures are consistent with NRC staff expectations because of observations of disposal structure cracking at the SDF in a relatively short period of time (i.e., tens of years) and the expected impact of disposal structure cracking on SDF performance (ML121170309). In response to RAI Question SUA-2 (ML21341A551), the DOE provided a revised sensitivity case evaluating the initial saturated hydraulic conductivity of saltstone with revised parameter values for infiltration and engineered barriers in the vadose zone. That sensitivity case showed that the initial hydraulic conductivity of the disposal structures over a range of two orders of magnitude did not impact the projected doses. In the 2020 SDF PA, the saturated hydraulic conductivity of the disposal structure concrete were assumed by the DOE to increase by several orders of magnitude to that of the surrounding soils within hundreds to several thousand years, as shown in Table 1, Table 2, and Table 3 above.

During that period, when the initial saturated hydraulic conductivity of disposal structure concrete is low, the closure cap is projected by the DOE to be the primary barrier to flow and contaminant transport. Because the closure cap design has not been finalized, NRC staff should review the potential for early contaminant release if the NRC staff determines that a significant amount of water could percolate through the closure cap before the disposal structures are assumed to be fully degraded. Part of the DOEs limited projected risk significance of the disposal structure concrete is the inclusion of non-cementitious materials and fast pathways through the disposal structures in the 2020 SDF PA. The NRC staff found the DOE inclusion of fast pathways through the disposal structures in the VZFM and VZTM to be acceptable because the presence of fast pathways is consistent with observations of fast pathways and contaminant transport through the SDS 4 walls and hydrotests on the 150-ft and 375-ft diameter disposal structures. In the 2012 TER, the NRC staff was concerned about the potential for material interfaces of the disposal structures (e.g., waterstops, cold joints) to act as fast pathways. In the NRC 2013 SDF Monitoring Plan, the NRC staff described that the staff expected to close MF 6.05 (Integrity of non-cementitious materials) under PO §61.41 and PO §61.42 after NRC determines that support for the assumed performance of non-cementitious materials used in the disposal structures is adequate. As shown in Figure 28 and Figure 29 below, the disposal structure joints and columns act as preferential pathways for flow through the disposal structures. The risk significance of these features will depend on the performance of other key barriers (e.g., closure cap, LLDL and HDPE/GCL composite layer, saltstone). Additional information is provided in Section 8.4 of this report regarding flow through the disposal structures. As the DOE continues to develop information related to the closure cap, the NRC staff will continue to evaluate the impact of these features on SDF performance. Impacts due to preferential pathways include higher flow rates, earlier releases, and decreased attenuation. In contrast, preferential pathways could also result in potential bypass of water around the saltstone grout, which could reduce radionuclide release by diverting water from the radioactive inventory. The degradation processes the DOE modeled for the columns in the cylindrical disposal structures results in earlier modeled degradation for the columns than for saltstone. Accordingly, the columns can act as pathways for infiltrating water to migrate around the saltstone grout. Because of the potential for modeled bypass to affect the projected dose, the NRC staff will continue to review information related to the performance of non-cementitious materials and bypass flow. The NRC staff recommends keeping MF 6.05 open under PO §61.41 and PO §61.42 as a medium priority. In the 2020 SDF PA, the DOE assumed that contaminant release from the disposal structures is spread across the nodes beneath the footprint of the disposal structures. As shown in Figure 7 above, a 0.25 mm (10-mil) poly sheeting is located in between the upper mud mat and the floor slab (SRR-SDU-2017-00001). The DOE observed leakage at this discrete layer during the hydrotest of SDS 6 (SRMC-CWDA-2022-00025). Accordingly, it is not clear what impact certain layers or a combination of layers may have on the areal extent of contaminant release. In RAI Comment CBs&DLs-10 (ML21341A551), the NRC staff discussed the need for additional information to show the potential effects of a contaminant release of a relatively small areal extent from the disposal structure to the water table. In response (SRMC-CWDA-2022-00003), the DOE provided a sensitivity analysis using a single aquifer source node instead of the 172 nodes distributed beneath SDS 9 in the Compliance Case and other sensitivity analyses. That sensitivity analysis also included increased infiltration and flow through the HDPE/GCL composite layer relative to the Compliance Case. Relative to the Compliance Case, those sensitivity cases of focused contaminant release increased the contaminant concentration by as much as a factor of 40 in the case of Tc-99 and moved the timing of the peak dose from greater

than 10,000 years to approximately 850 years in the case of I-129. NRC staff notes that the modeled release of contaminants at one source node on a 7.6 m by 7.6 m (25 ft by 25 ft) node pattern is extreme for a 375-ft disposal structure. However, this sensitivity analysis demonstrates the risk significance of the assumption that contaminant release will be areally distributed. Accordingly, NRC staff recommends monitoring information related to localized contaminant release under a new MF under MA 6 (Disposal Structure Performance). NFFT-1 The NRC staff recommends monitoring information related to localized contaminant release from the saltstone disposal structures under a new Monitoring Factor entitled Localized Contaminant Release under Monitoring Area 6 (Disposal Structure Performance) as a medium priority. 8.2.2. Disposal Structure Degradation In the 2020 SDF PA, the DOE evaluated three disposal structure degradation mechanisms - sulfate attack, carbonation-induced steel corrosion, and decalcification. The DOE analysis indicated that sulfate attack would be the rate-determining degradation mechanism for disposal structure concrete. Degradation from sulfate attack was assumed to be caused by the dissolution and transport of sulfates from saltstone into the concrete pore solution and the formation of expansive phases, such as ettringite. Precipitation of these minerals then results in excess pressure in the pores of the concrete and damage to the concrete. The DOE modeling of sulfate attack was based on diffusive transport of ions and the formation of expansive mineral phases (e.g., ettringite, gypsum). The modeled damage was based on predicted volume changes due to expansive phase formation (see NRC staff TRR entitled QA Documentation for the CBP Toolbox (ML16196A179)). After filling pores in the cementitious materials, these expansive phases were assumed to lead to cracking and increased transport of deleterious species (e.g., chloride, sulfate). The NRC staff considers the diffusive transport of sulfate through intact cementitious materials and formation of ettringite to be well supported. However, the support for the basis and parameterization associated with the assumed volume change due to the formation of expansive mineral phases is more limited. In addition, support is also lacking for the subsequent damage-model components:

1. development of strain based on the predicted volume changes,

2. relationship between strain and the cracking of cementitious materials, and

3. the resultant effective diffusion for cracked cementitious materials.

The NRC staff determined that the PA modeling of sulfate attack is still in its early stages of development by the DOE. The initial diffusive transport of ions through intact cementitious materials is adequately support by laboratory data. However, support for the basis of degradation due to volume change from the formation of expansive mineral phases is inadequate. Furthermore, there is significant evidence that the volume change is not correlated with expansion and therefore it is not correlated with degradation (ML16196A179). Additional information is needed to support the use of volume increase based on equilibrium thermodynamic calculations as the basis for modeling damage due to sulfate attack. If support for the basis of volume increase due to calculated mineralogical changes can be developed for

relevant cementitious materials, then the NRC staff has additional concerns related to model parameterization based on the underlying research. The increase in hydraulic conductivity with time of the disposal structure concrete was based on a geometric average of the intact and degraded materials as function of time. The use of a geometric average results in a sublinear increase in hydraulic conductivity. The NRC staff determined that there is insufficient support for the use of a geometric average for averaging intact and degraded materials, which is discussed in more detail in Section 8.3.2 of this TRR. The use of different averaging schemes for the disposal structure concrete does not have as significant of an effect on overall SDF performance as it does for saltstone. However, if the DOE relies more heavily on disposal structure performance in the future, then the NRC staff will review the impacts of degradation averaging on overall SDF performance. In the NRC 2013 SDF Monitoring Plan, the NRC staff described that the staff expected to close MF 6.04 (Disposal Structure Concrete Fracturing) under PO §61.41 and PO §61.42 after NRC determines that support for the amount of fracturing of the disposal structure floor and walls expected to occur during the performance period is adequate or if NRC determines that the estimate that DOE uses in the PA model is conservative. The NRC staff recommends keeping MF 6.04 open under PO §61.41 and PO §61.42 with a medium priority. 8.3. Wasteform Properties In the 2013 Monitoring Plan, NRC staff discussed that the hydraulic properties of the wasteform (i.e., hydraulic conductivity and diffusivity) are important to SDF performance because of their effects on the rate of radionuclide release into the groundwater. Monitoring Area 3 primarily addressed the initial properties of saltstone, while Monitoring Area 4 addressed changes in the matrix properties and macroscopic fractures with time. In 2017 (ML17018A137), NRC staff recommended closing MF 3.01 (Hydraulic Conductivity of Field-Emplaced Saltstone), MF 3.02 (Variability of Field-Emplaced Saltstone, and MF 3.04 (Effect of Curing Temperature on Saltstone Hydraulic Properties) under PO §61.41 and PO §61.42. MF 3.03 (Applicability of Laboratory Data to Field-Emplaced Saltstone) remains open, but with a narrowed scope to understand the short-term changes in hydraulic conductivity between laboratory-prepared and field-emplaced saltstone samples. Also in the 2017, NRC staff recommended keeping MF 10.05 (Moisture Characteristic Curves) open under PO §61.41 and PO §61.42 based on potential future changes to modeling assumptions and parameterization resulting in MCCs becoming more risk-significant. In 2019 (ML19031B221), NRC staff recommended keeping open MF 4.01 (Waste Form Matrix Degradation), MF 4.02 (Waste Form Macroscopic Fracturing), and MF 10.05 (MCCs) under both PO §61.41 and PO §61.42. The NRC also recommended keeping MF 4.01 and MF 4.02 as high priority monitoring factors and increasing MF 10.05 from low to medium priority under both PO §61.41 and PO §61.42. During the review of the 2020 SDF PA, the NRC staff determined that it would be clearer to monitor MCCs with the other hydraulic properties of saltstone and disposal structure concrete. Therefore, NFFT-2 The NRC staff recommends closing MF 10.05 (Moisture Characteristic Curves) and replacing it with two medium-priority monitoring factors entitled Moisture Characteristic Curves for Saltstone in MA 4 (Waste Form Physical Degradation) and Moisture Characteristic Curves for Disposal Structure Concrete in MA 6 (Disposal

Structure Performance). The NRC staff recommends monitoring for the two new monitoring factors under §61.41 and §61.42. Similarly, during the review of the 2020 SDF PA the NRC staff determined that it would be clearer to monitor wasteform diffusivity under the existing high-priority MF 4.01 (Waste Form Matrix Degradation) rather than under the MF 10.03 (Diffusivity in Degraded Saltstone). Therefore, NFFT-3 The NRC staff recommends closing MF 10.03 (Diffusivity in Degraded Saltstone) and updating the text of MF 4.01 (Waste Form Matrix Degradation) to include degradation of saltstone diffusivity under §61.41 and §61.42. 8.3.1. Wasteform Initial Properties As shown in Figure 8, the DOE relied on laboratory analyses of SDS 2A core samples using the dynamic leaching methodology to determine the hydraulic conductivity of saltstone. The NRC staff determined that the modeled value for the initial hydraulic conductivity of saltstone (i.e., 5.0x10-10 cm/s) is reasonable for two reasons. First, analysis of saltstone core samples provides direct information and the use of the dynamic leaching method is reasonable. Second, in the 2020 SDF PA, the DOE assumed that saltstone will degrade. As such, the initial hydraulic conductivity of saltstone is less risk-significant than the assumed rate and extent of degradation. As discussed below, the initial saturated hydraulic conductivity of saltstone is still important to SDF performance depending on the DOEs assumed degradation. Accordingly, the NRC staff recommends keeping MF 3.03 (Applicability of Laboratory Data to Field-Emplaced Saltstone) open with a medium priority under both PO §61.41and PO §61.42 with the narrowed scope (i.e., understanding changes in the saltstone hydraulic conductivity in the first several pore volume exchanges7). NFFT-4 The NRC staff recommends changing the priority of MF 3.03 (Applicability of Laboratory Data to Field-Emplaced Saltstone) from high priority to medium under both PO §61.41and PO §61.42 and still with the narrowed scope (i.e., understanding changes in the saltstone hydraulic conductivity in the first several pore volume exchanges). In the TRR for Saltstone Waste Form Hydraulic Performance (ML17018A137), NRC staff determined that the assumed initial effective diffusion coefficient that the DOE used in the 2013 and 2014 Special Analyses (i.e., 1x10-8 cm2/s) was well supported based on diffusivity measurements from carefully controlled saltstone simulants and saltstone core samples from SDS 2A. In the 2020 SDF PA, the DOE assumed that the initial effective diffusion coefficient for saltstone was 1.3x10-8 cm2/s, based on the maximum value observed for any analyte from the SDS 2A sample data. Accordingly, NRC staff finds that the initial effective diffusion coefficient assumed in the 2020 SDF PA to be acceptable. 8.3.2. Wasteform Degradation In the 2020 SDF PA, the DOE assumed that the saltstone would not degrade until the disposal structure roofs are degraded for all disposal structures, except SDS 4. The NRC staff did not 7 As discussed by the DOE in Section 7.1 of the 2020 SDF PA, the Compliance Case projected that 0.32 pore-volumes of the saltstone grout will be exchanged within 10,000 years. NRC staff notes that the rate of pore-volume exchanges could increase significantly if the assumptions of key barriers, such as the closure cap and saltstone grout, are revised.

find the DOE assumed delay to saltstone degradation to be well supported. However, NRC staff found that the assumed delay to saltstone degradation would not have significantly impacted the Compliance Case results. The delays to degradation are shown in Table 3 above. The delays range from 853 to 1,552 years, depending on the disposal structure. For SDS 4, the DOE assumed that most of the saltstone would not degrade until both the roof and the top layer of the saltstone are degraded. For SDS 4, the delay for saltstone degradation is 10,499 years, because the DOE assumed that the upper layer of saltstone, which contains steel from the disposal structure roof, would have to degrade prior to the lower layers of saltstone. The NRC staff has two concerns regarding the delays in degradation. First, the delays to saltstone degradation are based on the roof degradation, which the DOE assumed to be controlled by sulfate attack. The NRC staff has concerns with the 2020 SDF PA assumptions regarding sulfate attack and degradation of the disposal structures, as discussed in Section 8.2.2 above. Second, the DOE assumption that saltstone will not degrade until the overlying materials are degraded is based on the assumption that advective flow and decalcification will control saltstone degradation. As discussed below and in the NRC staff TRR entitled Saltstone Waste Form Physical Degradation (ML19031B221), the NRC staff is concerned about the potential for additional degradation mechanisms to impact flow through saltstone. In other words, saltstone could begin degrading sooner than assumed in the 2020 SDF PA if saltstone is susceptible to degradation mechanisms other than decalcification (long-term drying shrinkage, expansive phase formation, microbial activity). Evidence of potential additional degradation mechanisms has already been observed in saltstone in SDS 4 with cracks in the saltstone observed, as discussed in the NRC 2012 TER (ML121170309). For the Compliance Case, the delay in wasteform degradation did not significantly impact the dose results because the DOE assumed that degradation does not appreciably increase the saturated hydraulic conductivity of saltstone until approximately 100,000 years after closure. However, changes to modeling assumptions in future analyses could increase the risk significance of early degradation, and therefore early release. Accordingly, the NRC staff recommends monitoring information related to the delay in degradation of saltstone under MA 4 (Waste Form Physical Degradation). NFFT-5 The NRC staff recommends expanding MF 4.01 (Waste Form Matrix Degradation) and 4.02 (Waste Form Macroscopic Fracturing) under Monitoring Area 4 (Waste Form Physical Degradation) to include monitoring information related to the delay in degradation of saltstone grout. The NRC staff found the inclusion of saltstone degradation in the 2020 SDF PA to be a significant improvement in the site conceptual model. However, the NRC staff found the DOE conceptual model of saltstone degradation being controlled by decalcification to not be well supported. In the 2020 SDF PA, the DOE assumed that the saltstone grout would degrade by decalcification when water flows through saltstone. With extremely limited water flow through the closure cap (i.e., infiltration is reduced by approximately three orders of magnitude from natural infiltration for over 10,000 years after SDF closure), the LLDL, and underlying composite barrier, the saltstone is projected not to degrade appreciably within 100,000 years of SDF closure. That is because of the combination of the long time to the projected complete degradation of saltstone (i.e., 17 million years) and the use of a geometric average to calculate the effective hydraulic conductivity of saltstone. The NRC staff is concerned that: (1) decalcification could occur more quickly than projected due to potentially greater-than-assumed infiltration as discussed in NRC staff TRR entitled

Percolation Through and Potential Erosion near the Closure Cap (ML23017A083); (2) additional and coupled degradation mechanisms could result in more rapid degradation of saltstone than the DOE assumed in the PA; and (3) the use of a geometric average is not adequately supported and could significantly underestimate the effective hydraulic conductivity of degraded saltstone. For the Compliance Case in the 2020 SDF PA, the DOE assumed that saltstone grout degradation would be controlled by advection-controlled decalcification. The DOE assumptions about saltstone from Section 2.7.6 of the 2020 SDF PA stated: The saltstone will be completely encapsulated within the concrete [disposal structures]. As such, no significant mechanical degradation is expected to influence the performance of saltstone. Similarly, due to the chemical characteristics of saltstone, it is not subject to sulfate attack or microbial induced degradation and, because saltstone has no rebar or steel embedded within it, it is also not subject to carbonation. Therefore, it is reasonable to assume that decalcification (i.e., dissolution and chemical leaching of calcium) is the primary mechanism of saltstone degradation. The NRC staff evaluated those characteristics in Appendix A of the NRC staff TRR entitled Saltstone Waste Form Physical Degradation (ML19031B221) and provided information on potential additional saltstone degradation mechanisms. In that TRR, the NRC staff described that mechanical degradation can still affect saltstone within the disposal structures due to mechanisms such as long-term drying shrinkage, settlement, and loading. With respect to chemical degradation, the NRC staff described the potential for expansive phase formation and the lack of the DOE support for excluding sulfate attack. The NRC staff also described an observation of microbial activity on cast stone, which is similar to saltstone. The alkalinity and high pH of saltstone do not appear to preclude microbial degradation, especially with successive pore volume flushes decreasing the alkalinity and salt content. Thermal degradation due to temporal and spatial thermal gradients also could occur. Furthermore, feedback between multiple degradation mechanisms could further increase the rate of degradation. The 2020 SDF PA and its supporting documents did not provide information to: (1) refute other plausible degradation mechanisms; (2) demonstrate that the assumed degradation rate due to decalcification represented or exceeded the potential rate of degradation due to additional and coupled degradation mechanisms; or (3) demonstrate the risk significance of saltstone degradation. In response to RSI-1 (ML20254A003), the DOE provided a probabilistic analysis addressing the impacts of increased deep infiltration and increased saltstone degradation, as discussed in more detail in Section 8.4 below. The NRC staff found the DOE use of a geometric average to not be adequately supported because it depends on the assumption that flow is perpendicular to degraded layers. If saltstone does not degrade uniformly from top-to-bottom creating a uniform, horizontal layer of degraded saltstone, then the geometric average will not yield a reasonable effective hydraulic conductivity. Degradation of saltstone is likely to be non-uniform and may follow the formation of preferential flow paths and localized decalcification or degradation caused by other mechanisms. Under a more-typical, non-uniform degradation front, degradation would tend to be parallel to the path of flow, which is better represented by an arithmetic average. Using an arithmetic average is consistent with what the DOE previously used in the FY 2014 DOE SDF Special Analysis Document (ML15097A366), which stated:

This [FY 2014 DOE SDF Special Analysis Document] applies the more conservative approach of linear averaging, in part to compensate for departures from flow and transport perpendicular to the uniform degradation front. In the 2020 SDF PA, the DOE modeled degradation of saltstone as the weighted geometric mean of the assumed initial hydraulic conductivity of saltstone grout (i.e., 5x10-10 cm/s) with the fully degraded saltstone grout, which is assumed to be equivalent to that of the surrounding soil (i.e., 4.1x10-5 cm/s). The proportion of intact versus degraded saltstone was time weighted. For example, the DOE assumed that saltstone, which the DOE assumed to be completely degraded at 17,000,000 years, would be 50% degraded at 8,500,000 years. The NRC staff generated the curves in Figure 23 below based on the assumed initial and degraded hydraulic conductivities and the formulas in Section 4.4.2.4 of the 2020 SDF PA. For reference, the DOE assumed infiltration rate for the Compliance Case was also included. As shown in Figure 23 below, the DOE assumed that saltstone does not appreciably degrade within 100,000 years after SDF closure with the geometric average, which was relied upon for the Compliance Case. The DOE assumed that at 8,500,000 years (i.e., 50% degraded), the effective saturated hydraulic conductivity would be 1.4x10-7 cm/s for a geometric average of intact and degraded saltstone. If the DOE assumed an arithmetic average, the hydraulic conductivity would have been 2.1x10-5 cm/s at 50% degradation. Figure 23. Hydraulic Conductivity of Saltstone Grout with Different Averaging Approaches Relative to the Assumed Infiltration Rate In Section 5.8.8.2 of the 2020 SDF PA, the DOE discussed a hypothetical fast flow paths sensitivity analysis to evaluate a series of FEPs, including seismic activity, differential

settlement, material shrinkage, chemical degradation of cementitious materials, and steel corrosion. However, in RAI Question CM&FSU-4 (ML22026A391), the NRC staff discussed that this sensitivity analysis did not include damage to either the closure cap or to the drainage and composite layers above the roof due to subsidence and seismic activities. Consequently, infiltration rates to the disposal structures and saltstone grout remained similar to those of the Compliance Case and the NRC staff needed additional information to evaluate the impact of these FEPs. In response to that RAI Question (SRMC-CWDA-2022-00016), the DOE discussed that the uncertainty evaluated in response to the NRCs RSI (SRR-CWDA-2021-00068), incorporated both an increase in infiltration through the closure cap, degradation of the drainage and composite layers above the roofs, and the effects of seismic damage to the saltstone wasteform. However, the results of that analysis and subsequent analyses, which are discussed further below, indicated that doses from this scenario could exceed the performance objectives. The DOE also discussed in the response that the closure cap would be designed to mitigate potential seismic effects and that seismic damage to the closure cap is expected have a very low probability. The impacts to the SDF from potential FEPs that could result in fast flow paths (e.g., seismic activity, differential settlement, material shrinkage) will be monitored under an updated monitoring factor (MF 9.01 Settlement due to Increased Overburden and Seismic Loading) based on the Site Stability TRR (ML23017A114) and Monitoring Factor 4.02 (Waste Form Macroscopic Fracturing). The NRC staff found that the values for the effective diffusion coefficient for degraded saltstone are not adequately supported. In the 2020 SDF PA, the DOE assumed that the saltstone effective diffusion coefficient degrades to that of backfill (i.e., 5.3x10-6 cm2/s). The DOE appears to have again used the geometric averaging with respect to the effective diffusion coefficient. Based on NRC staff review of the PORFLOW model files, the assumed effective diffusion coefficient of saltstone only increased from 1.30x10-8 cm2/s initially to 1.31x10-8 cm2/s at 20,000 years after closure. The NRC staff are not aware of field or laboratory studies (e.g., accelerated aging studies) or natural analog studies that could support this assumed level of performance. In the 2013 SDF Monitoring Plan, the NRC staff expected to close MF 4.01 under both §61.41 and §61.42 POs after the NRC determines that support for modeled changes in the saturated hydraulic conductivity and diffusivity during the performance period is sufficient. The NRC staff recommends keeping MF 4.01 open. Also, based on the expected risk significance of degradation to the rate of release of radionuclides, the NRC staff recommends that MF 4.01 remain prioritized as high priority under §61.41 and §61.42. In the NRC 2013 SDF Monitoring Plan, the NRC staff described that saltstone fracturing was important to site performance because it: (1) increases flow through the saltstone, (2) shortens the diffusive length for radionuclide release, and (3) provides additional surface area for the progression of saltstone oxidation, which increases Tc release. Also in the NRC 2013 SDF Monitoring Plan, the NRC staff described that they expected to close MF 4.02 under both §61.41 and §61.42 after the NRC determines that model support for the assumed formation of macroscopic fractures in saltstone during the performance period was sufficient. As discussed in the NRC staff TRR entitled Saltstone Waste Form Physical Degradation (ML19031B221), the DOE has not provided sufficient information to justify the assumed degradation mechanisms at the time the NRC staff issued that TRR (May 2019). The NRC staff also determined that the DOE did not provide additional support for the assumed formation of macroscopic fractures in saltstone in the 2020 SDF PA or the DOE response to the NRC staff RSI or RAI questions. As discussed above, the DOEs hypothetical fast flow paths sensitivity analysis described in Section 5.8.8.2 of the 2020 SDF PA does not provide sufficient information for NRC staff to evaluate the risk associated with macroscopic fracturing because of the extremely low

percolation assumed by the DOE through the Closure Cap. Accordingly, the timing, rate, and extent of saltstone fracturing is not clear. Based on the uncertainty in saltstone fracturing, the NRC staff recommends keeping MF 4.02 open under both §61.41 and §61.42. In the TRR entitled Saltstone Waste Form Physical Degradation (ML19031B221), the NRC staff also discussed the risk significance of saltstone fracturing. Regarding the risk significance of saltstone fracturing, recent research and reinterpretation of earlier research has reduced some of the NRC staff concerns associated with gas-phase transport of oxygen in unsaturated fractures (ML18095A122). Several lines of evidence support a conceptual model in which the projected SDF performance is governed equally, if not more, by releases from reduced Tc as it by oxidized Tc, including:

similar Tc releases from two cores of field-emplaced saltstone despite one being leached with deareated liquid and the other being leached with liquid equilibrated with laboratory air;
Tc releases from the field-emplaced core samples occurred at concentrations consistent with solid phases of reduced Tc even with measured Eh values suggesting Tc should be oxidized (i.e., Tc release did not appear to be sensitive to the presence of oxygen);
re-evaluation by the NRC staff of previous studies that had appeared to show the sensitivity of Tc to trace quantities of oxygen, resulting in a demonstration that the prior results were generally consistent with the projected release from chemically reduced Tc and that some were apparently due to experimental artifacts;
an independent analysis by NRC staff that showed greater releases from reduced as compared to oxidized saltstone under conditions of uniform flow in intact saltstone with a Tc concentration and Tc sorption as assumed in the 2014 Evaluation Case and Tc release from reduced saltstone consistent with recent results from studies that the DOE conducted with cores of field-emplaced saltstone.

In addition, if flow through chemically reduced saltstone grout is the risk-controlling source of radionuclide release, then saltstone fracturing could ultimately reduce radionuclide release as fractures may decrease the interaction between infiltrating water and the saltstone grout. However, fracturing could still result in increased radionuclide release relative to the DOE Compliance Case due to decreased diffusive lengths, oxidation of saltstone grout prior to contact with water, and increased flow through saltstone grout, depending on the type and extent of fracturing (e.g., large-scale network of small fractures). Therefore, based on the potential risk significance of fracturing on radionuclide release, the NRC staff recommends that MF 4.02 remain prioritized as high priority under both §61.41 and §61.42. The NRC staff found the support for the assumed MCCs for cementitious materials in the 2020 SDF PA to be adequate, because the assumed MCCs are similar to literature values and the MCCs are of limited risk significance in the 2020 SDF PA. In the 2020 SDF PA, the DOE assumed that the MCCs for cementitious materials changed with degradation (see Figure 10 above). That change in relative permeability with saturation is significantly less pronounced than in the 2009 SDF PA. For example, the relative permeability of intact saltstone and fractured saltstone in the 2009 SDF PA was approximately 1x10-2 and 2x10-7 at 99% saturation, respectively. Tables 4.4-91 to 4.4-94 in the 2020 SDF PA showed the assumed saturation of saltstone and the disposal structures. The saturation for saltstone grout ranged from 99.65% to 99.95%. At 0 to 50 and 100,000 to 200,000 years after closure, the DOE projected that the saltstone grout in the 375-ft diameter disposal structures would be 99.83% and 99.85% saturated. Based on the DOE assumed MCCs, the corresponding saltstone relative permeabilities would be approximately 80% of the saturated hydraulic conductivities. A relative

permeability value of 80% appears to be slightly optimistic relative to the literature values shown in Figure 2.7-3 in the NRC 2012 SDF TER. However, additional DOE analyses indicated that the assumed MCCs in the 2020 SDF PA are of limited risk significance, as discussed further below. In Section 5.8.8.3 of the 2020 SDF PA, the DOE described a sensitivity case that assumed a relative permeability of 1 for all cementitious materials, thereby negating all impacts from the MCCs. The doses from this sensitivity case and the Compliance Case were the same. The DOE indicated that this was because the flowrate through the disposal structures was limited by the infiltration rate and not the flowrate through the cementitious materials. However, the NRC staff discussed in the Preliminary Review of the 2020 SDF PA (ML20254A003) that the PA did not consider the full range of uncertainty in the performance of the closure cap or the engineered barriers above the disposal structures. Accordingly, the PA did not provide information on the risk significance of MCCs if the flow through the system was limited by the flow through the cementitious materials. As part of the Preliminary Review, the NRC staff requested a supplemental analysis to provide insight into the risk significance of MCCs of cementitious materials and fast pathways through the disposal structures if the closure cap, LLDL, and composite barrier above the disposal structures do not perform as expected. In response to RSI-5, the DOE selected 100 realizations as a subset of their probabilistic analysis. The RSI-5 analysis used the same settings as the RSI-1 analysis, which is discussed below in Section 8.4, with the exception of a relative permeability of 1 for all cementitious materials instead of MCCs. The statistical time history of the dose results for the RSI-1 and RSI-5 analyses are shown to be very similar in Figure 4.3-3 of the SRR-CWDA-2021-00066. Accordingly, the NRC staff found the MCCs to be of limited risk significance for the 2020 SDF PA Compliance Case as well as a degraded case. However, because future revisions to modeling assumptions could result in the MCCs becoming more risk-significant, the NRC staff recommends keeping MF 10.05 (Moisture Characteristic Curves) open but with a low priority under both PO §61.41and PO §61.42. The risk significance of MCCs in future analyses related to SDF performance would be more easily evaluated with a deterministic sensitivity analysis with the assumed MCCs and relative permeability equal to 1. The NRC staff agrees with the DOE that a relative permeability of 1 is unrealistic, but this approach provides a clearer understanding of the risk significance of MCCs because the effect of the MCCs could be obscured in a probabilistic analysis. Therefore, the NRC staff recommends keeping MF 10.05 (MCCs) open and with a low priority under both PO §61.41and PO §61.42. The DOE also provided sensitivity cases in Section 5.8.2.4 of the 2020 SDF PA to evaluate an accelerated saltstone degradation case and a case with an initial saltstone hydraulic conductivity of 2x10-9 cm/s for a scenario assuming a Soil-Only Closure Cap8. For the accelerated degradation case, the DOE assumed that the degradation of cementitious materials would occur twice as fast as in the Compliance Case (i.e., the time to reach complete degradation for saltstone would be 8,500,000 years versus 17,000,000 years). As shown in Figure 24 below, the assumed accelerated degradation with geometric averaging does not change the assumed saltstone hydraulic conductivity within 100,000 years. The results of the accelerated degradation case, which included accelerated degradation of other cementitious materials, did not affect the magnitude of the dose as shown in Figure 5.8-14 of the 2020 SDF PA. The sensitivity case with an initial saturated hydraulic conductivity for saltstone of 2x10-9 cm/s did increase the peak dose relative to the Soil-Only Closure Cap sensitivity case by 30-8 The DOE Soil-Only Closure Cap case still includes assumed performance of the lower lateral drainage layer and the HDPE/GCL composite layer.

40%. Because of concerns about the assumed performance of the drainage layers and HDPE/GCL composite layers and the assumed performance of the lower lateral drainage layer in the DOE analysis, The NRC staff requested additional information regarding the performance of the composite layers and saltstone grout (see Section 8.4 below). Figure 24. Assumed Hydraulic Conductivity for Saltstone Grout Sensitivity Cases (Graph Made by the NRC Staff Based on DOE Equations in the 2020 SDF PA) 8.4. Modeling of Near Field Flow The NRC staff found the DOE conceptual model for near field flow to be acceptable because key FEPs are represented in the VZFM. However, the NRC staff found that several of the values that the DOE used to model flow through the near field were not adequately supported. In the Preliminary Review of the 2020 SDF PA (ML20254A003), the NRC staff discussed concerns regarding uncertainty with the sand drainage layers, HDPE, HDPE/GCL composite barriers, saltstone degradation, and MCCs and requested an analysis demonstrating the effects of the combined uncertainty in RSI-1. In response to the NRC staff concerns, the DOE provided a new probabilistic analysis with 1,000 realizations conducted for a period of 100,000 years (SRR-CWDA-2021-00066). That analysis used inputs and model results developed from two DOE reports: Evaluation of the Uncertainties Associated with the SDF Closure Cap and Long-Term Infiltration Rates (SRR-CWDA-2021-00040) and Supplemental Information and Proposed Probabilistic Inputs Related to NRC RSI-4: Saltstone Degradation (SRR-CWDA-2021-00052). As the DOE described in document SRR-CWDA-2021-00066, the analysis included a revised:

infiltration rate
saturated hydraulic conductivity of sand in the LLDL
HDPE service life
degraded saturated hydraulic conductivity for the HDPE/GCL composite layer
initial saturated hydraulic conductivity of saltstone
initial degradation fraction of saltstone
p-averaging term The NRC staff found that the initial saturated hydraulic conductivity and effective diffusion coefficient shown in Figure 25 and Figure 26 below appear to be skewed low relative to the DOE laboratory results shown in SRR-CWDA-2021-00056. However, the DOE included higher initial saturated hydraulic conductivity and effective diffusion coefficients in their probabilistic analysis. The results of that analysis showed the potential risk significance of higher initial hydraulic properties. In addition, the DOE assumed in the analysis that the saturated hydraulic conductivity of saltstone and diffusivity of radionuclides in saltstone increase with saltstone degradation. The revised saturated hydraulic conductivity of saltstone for a 375-ft Diameter disposal structures for the probabilistic analysis is shown in Figure 9 above. Although the initial saturated hydraulic conductivity of saltstone was shown to have a significant effect on the degraded properties of saltstone within 10,000 years, as discussed in Section 6.2 of SRR-CWDA-2021-00056, the assumed p-averaging term had the most significant impact on the projected dose to a member of the public at the 100-m SDF boundary. The assumed HDPE Failure Condition had a significant impact several thousand years after SDF closure. Because of the lack of long-term data for these engineered materials, the NRC staff agrees with the DOE that there is significant uncertainty in the assumed p-averaging term and HDPE Failure Condition. Although there is significant uncertainty in the saturated hydraulic conductivity over time, the DOE probabilistic analysis provides risk information to understand what potential doses may occur in the future depending on the key modeling assumptions.

Figure 25. Probability Density Function for the Initial Saturated Hydraulic Conductivity of Saltstone the DOE used in Response to the NRC RSI (adapted from DOE document SRR-CWDA-2021-00056) Figure 26. Probability Density Function for the Initial Effective Diffusion Coefficient of Saltstone the DOE used in Response to the NRC RSI (adapted from DOE document SRR-CWDA-2021-00056) In Table 4.1-1 of SRR-CWDA-2021-00066, the projected peak of the mean dose to a member of the public at the 100-m boundary within 10,000 years of SDF closure for the DOE revised probabilistic analysis was 12 mrem/yr, 43 mrem/yr, and 518 mrem/yr for HDPE Failure Conditions 1, 2, and 39, respectively. However, the NRC staff was concerned that several of the parameters (e.g., HDPE degradation, infiltration rate, and the p-averaging term) had distribution ranges that were potentially optimistic. Because of concerns that the dose results from the DOE analysis in response to NRC question RSI-1 could be biased low, the NRC staff requested the 9 HDPE Failure Conditions were characterized by the DOE as 1 (no failure and infiltration only occurring due to initial defects in the HDPE), 2 (partial failure of the HDPE), and 3 (complete failure of the HDPE).

dose results from a subset of the 1,000 realizations from the revised probabilistic analysis for RSI-1. This was to better understand the risk significance of the barriers in the vadose zone flow and transport models. The requested parameter values/ranges for the subset of realizations were HDPE Failure Condition 2, Infiltration Rate greater than 1 in/year within 10,000 years, and a p-averaging term greater than 0 for the following reasons:

the DOE considered HDPE Failure Condition 2 to be the most likely and reasonable future condition (SRR-CWDA-2021-00057, SRR-CWDA-2021-00066)
an infiltration rate greater than 1 in/year within 10,000 years represents greater than an order of magnitude reduction in infiltration relative to natural background infiltration (the NRC staff discussed significant uncertainty in the performance of the closure cap over long periods of time (ML23017A083))
a p-averaging term greater than 0 represents an average that results in a non-zero amount of degradation of saltstone within 100,000 years The DOE developed a supplemental analysis of screened RSI-1 model results (SRR-CWDA-2021-00098 (ML21326A013)). Of the 1,000 realizations in the RSI-1 probabilistic analysis, 33 realizations met all three requested conditions. Figure 27, below, shows the projected peak dose within 10,000 years of SDF closure to a member of the public at the 100-m SDF boundary for those realizations. Of those 33 realizations, only 5 result in peak dose projections below 25 mrem/yr. The peak of the mean dose for these 33 realizations within 10,000 years of SDF closure is 336 mrem/yr and the median is 124 mrem/yr. The disparity between the mean and median peak doses indicates that several realizations with relatively high doses skew the statistics higher. As shown in Figure 27, below, several realizations resulted in peak dose projections of approximately 10 mSv (1 rem) or greater.

Figure 27. Peak Doses from Subset of RSI-1 Probabilistic Analysis Realizations (adapted from SRR-CWDA-2021-00098)

The DOE stated that limiting the analysis to these realizations with peak infiltration greater than 1 in/year within 10,000 years biases the dose results high because it excludes any barriers or conditions that limit higher infiltration rates from occurring. The NRC staff discussed uncertainty in the long-term performance of the closure cap and HDPE/GCL composite layer (ML23017A083, ML23017A089). The NRC staff understands that these barriers have not been constructed yet and will review additional information as it becomes available to better understand the long-term infiltration rates. In addition to concerns about the DOEs assumed HDPE performance, the NRC staff has concerns about the performance of the GCL layer beneath the HDPE (ML21341A551) In RAI CBs&DLs-9, NRC staff requested additional information regarding the impact of localized flow through HDPE defects on the underlying GCL performance (ML21341A551). In response to RAI CBs&DLs-9 (SRMC-CWDA-2022-00003), the DOE provided flow model results for both the Compliance Case, as shown in Figure 28 below, and for a revised case (referred to as Case CBs&DLs-9) for SDS 910, as shown in Figure 29 below. In the Compliance Case, the flows through the disposal structures and saltstone grout were limited by infiltration through the closure cap and the disposal structure. This can be seen in Figure 28 with the overlapping curves for several PORFLOW materials. The volumetric flow curves shown in Figure 29 are more distributed after 300 years indicating that the flows are generally not limited by the overlying layers after that time (i.e., before 300 years, all of the layers under the closure cap can transmit all of the water that flows through the closure cap). Flow through the joints and columns in the disposal structures is shown in Figure 28 and Figure 29 to represent between roughly 1% and 10% of the flow through saltstone when percolation through the closure cap was assumed to be low (i.e., less than 1 mm/yr 0.04 in/yr). Percolation was assumed by the DOE to be low throughout the 100,000-year modeling period shown in Figure 28 and for the first approximately 300 years in the CBs&DLs-9 Case shown in Figure 29. In the CBs&DLs-9 Case, flow through the joints and columns represented between one to two orders of magnitude more flow than through saltstone grout for much of the 100,000-year modeling period. 10 The DOE selected SDS 9 to evaluate various uncertainties because of several aspects (e.g., inventory, location within the SDF, proximity to the water table) that result in SDS 9 controlling the peak dose within 1,000 years and for a significant portion of 10,000 years after SDF closure (SRR-CWDA-2021-00057).

Figure 28. Volumetric flow rate through SDS 9 materials and infiltration for the Compliance Case (adapted from DOE documents SRR-CWDA-2022-00003 and SRR-CWDA-2021-00040)

Figure 29. Volumetric flow rate through SDS 9 materials for the CBs&DLs-9 Case (adapted from DOE documents SRMC-CWDA-2022-00003 and SRR-CWDA-2021-00040) The resultant projected peak dose within 20,000 years after SDF closure from the CBs&DLs-9 Case resulted in factors of increase in dose for I-129 and Tc-99 of 11 and 26 times, respectively. Although this case includes more degradation of the GCL layer, the assumed saltstone performance remains high. The initial hydraulic conductivity of saltstone was assumed to be 1.7x10-9 cm/s. Figure 29 shows that the saltstone hydraulic conductivity does not start to increase until approximately 10,000 years after closure and only increases by one order of magnitude by 40,000 years. As a counter example, out of the subset of 33 realizations from the DOEs RSI-1 probabilistic analysis described above, only 2 realizations had saturated hydraulic conductivity values of 1x10-9 cm/s by 10,000 years. In Appendix C of SRR-CWDA-2021-00098, the DOE showed that the mean saturated hydraulic conductivity of the 33 realizations was approximately 2x10-8 cm/s initially and almost 1x10-5 cm/s by 10,000 years. If the saltstone saturated hydraulic conductivity is greater than the DOE assumed (i.e., approximately 2x10-9 cm/s within 20,000 years), then the modeled doses would increase relative to the CBs&DLs-9 Case due to a consolidated release period and decreased bypass flow. Related to MF 4.01 described above, NRC staff remains concerned that there is very limited support for assuming very low saturated hydraulic conductivity values for saltstone over hundreds and thousands of years. The DOE stated in the response to the CBs&DLs-9 Case that that case represents complete and catastrophic failure of both the HDPE and the GCL at the disposal structure roof and between the disposal structure mud mats. In this Case, the HDPE and GCL are assumed to completely disappear once the service life of the HDPE is reached. However, the DOE expects that even once the service life is reached, the HDPE will continue providing a barrier to flow over most of the system and the GCL is expected to mitigate flow through those breaches. The

NRC staff agrees that the HDPE will continue to act as a barrier to flow in places even after its service life is reached. However, the NRC staff also notes that only a small fraction of the area of the HDPE may have to be breached to allow the majority of the infiltrating water to migrate through the HDPE. In other words, the HDPE may not have to completely disappear to permit most of the water through it. 8.5. Modeling of Radionuclide Release and Transport 8.5.1. Overview The NRC staff found the DOE approach to modeling radionuclide release from saltstone and transport in the near field to be acceptable because it used models that were suitable for the purpose with parameter values based on applicable experimental results and theoretical considerations. The NRC staff evaluated the Model Integration between the near field flow and transport models (i.e., the VZFM and VZTM) in a TRR entitled Model Integration (ML23017A090) and found it to be acceptable. In that TRR, the NRC staff also found the abstracted GoldSim model to be acceptably benchmarked to the results of the VZTM. 8.5.2. Release Models The NRC staff found the use of a shrinking core model for iodine and technetium release to be acceptable because it represented: (1) the spatial pattern of wasteform and concrete pH and Eh changes governed by the projected water flow and (2) the expected effect of the chemical changes on iodine and technetium mobility. The NRC staff found the use of a simpler pore volume exchange model for the other elements to be acceptable because of the lower risk significance of those elements. The NRC staff found the conceptual model of the consumption of reducing capacity by dissolved oxygen in water flowing into saltstone to be acceptable because it was consistent with the expected reaction of oxygen and reduced sulfur species supplied by the blast furnace slag. The NRC staff found the use of the modeled pH to represent the cementitious material age (e.g., Region I, Region III) to be acceptable because it is common modeling practice supported by empirical measurements. 8.5.3. Chemical Transitions As described in Section 6.4.3, the DOE used a combination of chemical modeling and empirical observations to develop values for the number of pore volume exchanges required for each chemical transition. The NRC staff found the general approach of using laboratory measurements for iodine and technetium and chemical modeling for the remaining radionuclides to be acceptable because of the larger risk significance of iodine and technetium. For the transition from Region I to Region III conditions, the NRC staff evaluated the radionuclides in three groups: (1) iodine and technetium; (2) chlorine, cesium, nickel, and uranium; and (3) all other radionuclides. The NRC staff identified different reasons why the DOE modeled pore volume exchanges to each transition for each group of radionuclides was acceptable, as described below. The NRC staff found the values for iodine and technetium to be acceptable because they were based on measured mobility changes for those radionuclides in DLM experiments with cores of field-emplaced saltstone and saltstone simulants. The NRC staff found the GoldSim models the DOE used to optimize the modeled pore volume exchanges, sorption coefficients for iodine, and

solubility limits for technetium to best fit the experimental DLM results to be fit for purpose, traceable, and transparent. Unlike for most radionuclides, the modeled mobility of chlorine, cesium, nickel, and uranium is greater under Region I conditions than under Region III conditions (see Table 12 in this TRR). For chlorine, cesium, and nickel, mobility decreases in Region III under both reduced and oxidized conditions. For uranium, mobility decreases in Region III only under oxidized conditions. For those radionuclides, modeling the transition from Region I conditions to Region III conditions after one pore volume exchange instead of six pore volume exchanges, as projected by the Contaminant Release Model, would tend to underestimate radionuclide mobility by shortening the time in Region I, when those radionuclides are more mobile. However, the NRC staff determined that the difference is unlikely to significantly affect dose projections. To bound the risk significance of an earlier change from Region I to Region III conditions for chlorine, cesium, nickel, and uranium, the NRC staff ran the DOE Compliance Case GoldSim model for SDS 9 and found that setting the Kd values under Region III conditions to match their values under Region I conditions did not make a detectable difference to the dose projections for an offsite member of the public at 10,000 years after closure. For the remaining radionuclides, the NRC staff found the DOE decision to model the change from Region I to Region III conditions after one pore volume exchange to be acceptable because it would tend to overestimate radionuclide mobility compared to using six pore volume exchanges, as projected by the DOE Contaminant Release Model. An earlier Region I to Region III transition would overestimate mobility because most radionuclides either increase or do not change mobility when the material changes from Region I to Region III conditions. Overestimating mobility would increase the projected dose to an offsite member of the public or an inadvertent intruder who drills a well near a disposal structure. Although increasing radionuclide mobility out of the wasteform and disposal structure would theoretically reduce the dose to an individual who inadvertently drills into a disposal structure, the DOE did not take credit for radionuclide leaching in that intrusion case so that case is not affected by the timing of the transition from Region I to Region III conditions. The NRC staff found the projected pore exchanges required to transition from chemically reduced to chemically oxidized conditions to be acceptable because reasonable alternatives did not have a large effect on the projected dose. In general, the NRC staff found the DOE approach of using a chemical model to project the necessary number of pore volumes to transition from reduced to oxidized cementitious material to be acceptable because it was consistent with the conceptual model of the material performance and expected chemical reactions occurring in slag-bearing cementitious material. However, as described in Section 7.4.4 of the TRR, the NRC staff did not find the values the DOE used to model the reducing capacity of saltstone and disposal structure concrete to be well supported. As that section also indicates, the relatively small modeled effect of changes in the reducing capacity (and, therefore, changes in the timing of the transition from reduced to oxidized material) depends in part on the compensating effects of a decrease in iodine mobility and increase in technetium mobility when the cementitious material in the model transition from being chemically reduced to chemically oxidized. Therefore, the acceptability of the projected pore volume exchanges to transition from chemically reduced to chemically oxidized cementitious material in the 2020 SDF PA depends on features specific to that model, such as the relative inventories of iodine and technetium, and is not generalizable to other models. 8.5.4. Chemical Reducing Capacity

The NRC staff found that the values the DOE used for the chemical reducing capacity of saltstone and disposal structure concrete were acceptable for use in the 2020 SDF PA because the values did not have a significant effect on the projected dose. However, like the transition time, the small size of the projected effect is due in part to the balancing effects of reduced iodine mobility and increased technetium mobility when cementitious material transitions from chemically reduced to chemically oxidized. For that reason, the low significance depends on the relative inventories of iodine and technetium and other aspects of the 2020 SDF PA models and is not generally applicable to other models. The NRC staff determined that the chemical reducing capacity the DOE assigned to saltstone grout and the disposal structure concrete would not be adequately supported if the risk significance changes in future SDF modeling for two reasons: (1) the DOE did not directly address the applicability of measurements made with the cerium method and (2) the DOE did not address the potential effects of sulfur leaching. In previous TRRs on saltstone oxidation (ML15098A031) and technetium release (ML18095A122), the NRC staff questioned the applicability of the cerium method, which uses an acidic step, to saltstone grout or disposal structure concrete, which have alkaline pore water. As discussed in Section 6.4.4 in this TRR, the DOE found that measurements of saltstone simulants yielded approximately twice the measured reducing capacity with the cerium method as they did with the chromium method. In addition, in a previous TRR on technetium release (ML18095A122), the NRC staff questioned whether leaching of sulfur from saltstone and disposal structure concrete could affect those materials reducing capacity, because the DOE indicated that the reducing capacity of saltstone and slag-bearing concrete is attributable to the sulfur in the blast furnace slag. In the 2020 SDF PA, the DOE reported multiple reducing capacity measurements in saltstone and concrete based on both the cerium and chromium methods (SRR-CWDA-2018-00048). Although the DOE acknowledged the comments the NRC staff made about applicability of the two methods, the DOE did not determine which method was more applicable. Instead, for concrete, the DOE used measurements made with the cerium method without indicating why it was applicable to the alkaline concrete environment. For saltstone, the DOE treated the collection of measurements made with both methods as equally applicable values. The NRC staff did not find the bases for either the concrete or saltstone modeled values to be sufficient because the DOE did not provide a basis for determining the cerium measurements were applicable to the alkaline environments of concrete or saltstone. In addition, the NRC staff did not find the basis for the values the DOE used for saltstone to be acceptable because: (1) the DOE did not provide a basis for double counting the measurements of samples made with Holcim slag and measured with the cerium method and (2) the DOE did not provide a basis for use of the mean, median, and lowest values of the data sets to represent the Best Estimate, Compliance Case, and Pessimistic Case. Although the NRC staff did not find the technical bases for the modeled reducing capacity of saltstone or slag-bearing concrete to be sufficient, the NRC staff found the modeled values to be acceptable because reasonable alternatives did not have a significant effect on the projected dose to an offsite member of the public (see Figure 15 above). The 2013 SDF NRC Monitoring Plan addresses the reducing capacity of saltstone under 10 CFR 61.41 and 10 CFR 61.42 in MF 5.03 (Reducing Capacity of Saltstone), which is a low priority monitoring factor. Because the NRC staff found the values to be acceptable based on sensitivity analyses rather than based on the technical justification for the modeled values, the acceptability could change if the model changes in the future. For example, the small effect of changing the reducing capacity on the modeled dose to an offsite member of the public

depended in part on the compensating effects of a decrease in iodine mobility and increase in technetium mobility when the cementitious material in the model transition from being chemically reduced to chemically oxidized. That result could change if the relative inventories of iodine and technetium change, or if the DOE changes the modeled sorption coefficients for iodine or technetium in oxidized cementitious materials. Therefore, the NRC staff does not recommend any changes to the status (open) or priority (low) of MF 5.03 under both §61.41 and §61.42 POs. The 2013 SDF NRC Monitoring Plan does not currently include a MF focused on the modeled reducing capacity of disposal structure concrete. As described in the previous NRC TRR on technetium release (ML18095A122), the DOE previously provided sensitivity analyses that demonstrated that the oxidation state of disposal structure concrete had a negligible effect on technetium release in the DOE FY 2013 Special Analysis (SRR-CWDA-2013-00062, Rev. 2). The results of the DOE sensitivity analysis described in Section 6.4 in this TRR support a finding that the modeled reducing capacity in disposal structure concrete in the 2020 SDF PA does not have a risk-significant impact on the projected dose to a member of the public at the 100-m SDF boundary. Therefore, the NRC staff does not recommend developing a new MF to specifically address the reducing capacity in disposal structure concrete. However, the NRC staff will continue to evaluate the contribution of the disposal structure physical and chemical barrier capabilities in future PA reviews. 8.5.5. Iodine Sorption The NRC staff found the sorption coefficients the DOE used to model iodine release from and transport in chemically reduced Region I and Region III saltstone to be acceptable because they were based on measurements of iodine mobility in representative saltstone simulants under applicable chemical conditions. As indicated in Section 7.4.3, the NRC staff found the GoldSim models the DOE used to develop iodine sorption coefficients based on the experimental data to be fit for purpose, traceable, and transparent. The NRC staff found the Kd value the DOE used to represent iodine sorption in Region III chemically oxidized saltstone and disposal structure concrete (i.e., 4 mL/g) to be acceptable because (1) it is consistent with the expected chemical speciation of iodine and (2) it was adequately supported by literature values for similar materials. Although the NRC staff initially questioned the applicability of the literature values to saltstone, the DOE response demonstrated that the value of 4 mL/g was likely to be conservative because the ratio of Kd values in oxidized and reduced cementitious materials in the literature would suggest a Kd value from two to ten times larger (SRR-CWDA-2021-00072, Rev. 1). As previously discussed in this TRR, using a smaller Kd value in oxidized materials is conservative in the 2020 SDF PA. In contrast, the NRC staff found the DOE did not provide support for the assumption that iodine would be chemically oxidized as it flowed through chemically oxidized sections of cementitious material. As described in Section 6.5 in this TRR, the NRC staff questioned whether chemically reduced iodine would be oxidized quickly enough to concentrate in the lower mud mat as shown in Figure 5.8-83 in the 2020 SDF PA. Although the DOE response described the expected mobility of different iodine species, it did not address the speed of iodine oxidation in cementitious materials compared to the expected amount of time iodine would be in the lower mud mat (SRMC-CWDA-2022-00003). In addition, the DOE response did not address whether reduced iodine would be oxidized if water flowed through fractures in the lower mud mat. Therefore, the DOE response did not provide a basis to determine whether the modeled concentration of iodine in the lower mud mats in Figure 16 in this TRR was realistic.

NFFT-6 The NRC staff recommends monitoring the development of information about the sorption of iodine in oxidized mud mats under MF 6.01 (Certain Risk-Significant Kd Values in Disposal Structure Concrete) under MA 6 (Disposal Structure Performance) under 10 CFR 61.41 and 10 CFR 61.42. The NRC staff should consider the potential for fast pathways that could limit the ability of the mud mats to chemically oxidize iodine as it flows through the mud mats. The NRC staff determined that the level of support for the potential effects of changing to a cement-free saltstone formula differed between diffusion-dominated and advection-dominated releases. The NRC staff found changing to a cement-free saltstone formula would be unlikely to significantly change iodine releases from diffusive transport because the measured effective diffusion coefficients of iodine in simulated cement-free saltstone samples were within the range diffusivities observed simulated saltstone made with the 45:45:10 formulation (SRR-CWDA-2020-00040). However, the NRC staff did not find adequate support for the use of a Kd value of 4 mL/g in oxidized cement-free saltstone because the literature values the DOE based the value on did not include cement-free samples. NFFT-7 Therefore, if the DOE chooses to use cement-free saltstone, the NRC staff recommends monitoring the development of information about iodine sorption in chemically oxidized cement-free saltstone under MF 5.04 (Certain Risk-Significant Kd Values for Saltstone) under MA 5 (Waste Form Chemical Performance) under 10 CFR 61.41 and 10 CFR 61.42. 8.5.6. Technetium Solubility and Sorption The NRC staff found the solubility limits and sorption coefficient used to represent technetium release and near field transport in the 2020 SDF PA to be acceptable because they are either consistent with applicable experimental results or represent conservative choices. The NRC staff found that modeling technetium release and transport with a combination of solubility limits in chemically reduced material and sorption coefficients in chemically oxidized material was consistent with theoretical expectations and empirical observations. For the solubility limits, the NRC staff found a limit of 9.7 x 10-7 mole/L in cementitious material with a pH greater than or equal to 11 and a limit of 4.5 x 10-7 mole/L in cementitious material with a pH less than 11 accurately represented results from DLM testing with samples of field-emplaced saltstone and saltstone simulants. For the sorption coefficients, the NRC staff found that the value of 0.5 mL/g for technetium in chemically oxidized cementitious material was consistent with studies in ordinary concrete. Although those studies were not specific to saltstone or disposal structure concrete, the NRC staff found that value to be acceptable because reasonable alternatives would make a small difference to the projected dose. The NRC staff found the VZTM implementation of solubility limits with effective Kd values to be acceptable because the effective Kd values result in the same modeled pore liquid concentrations of technetium as modeling a solubility limit would. Because a Kd represents a fixed ratio of the liquid and solid concentrations of a chemical species, the effective Kd must change to maintain a fixed liquid-phase concentration if the solid phase concentration changes. The NRC staff found that the DOE calculations for the effective Kd value appropriately accounted for the following features and processes, which affect the solid phase technetium concentration and the applicable solubility limits: (1) potential variation in the initial total technetium concentration in saltstone, (2) the change in technetium solubility when the material

pH changes from Region I to Region III conditions, and (3) the transition from chemically reduced to oxidized conditions. Similar to the staff determination for iodine, the NRC staff found the level of support for the transport properties of technetium in cement-free saltstone differed for diffusion-dominated and advection-dominated conditions. The NRC staff found adequate support for assuming that changing to a cement-free saltstone formula would not significantly change diffusive technetium releases because the measured effective diffusion coefficients of technetium in simulated cement-free saltstone samples were within the range diffusivities observed simulated saltstone made with the current formulation (i.e., the 45:45:10 formulation) (SRR-CWDA-2020-00040). The NRC staff also found the use of a Kd value of 0.5 mL/g for technetium in oxidized Region III saltstone to be acceptable for a cement-free formula because the DOE demonstrated that the value is sufficiently low that reasonable alternatives do not significantly affect the projected technetium flux. However, the NRC staff did not find adequate support for assuming the advective release of technetium from cement-free saltstone would be the same as advective release of technetium from the current formulation because the DOE did not report measurements of technetium solubility in cement-free saltstone. The NRC staff expects that technetium solubility will control its advective release under chemically reduced conditions. In probabilistic sensitivity analyses for the 2020 SDF PA, the DOE showed that technetium solubility in saltstone is one of the most risk-significant parameters for both an offsite member of the public and individual who inadvertently intrudes on the SDF. Therefore: NFFT-8 If the DOE chooses to use cement-free saltstone, the NRC staff recommends monitoring the development of information about technetium solubility in cement-free saltstone under MF 5.04 under MA 5 (Waste Form Chemical Performance) under 10 CFR 61.41 and 10 CFR 61.42. To reflect the use of a solubility limit to represent technetium transport in chemically reduced saltstone, the NRC staff recommends changing the name of the monitoring factor to include solubility limits. 8.5.7. Sorption Coefficients for Radionuclides Other Than Technetium and Iodine The NRC staff finds the sorption coefficients used for other radionuclides in the 2020 SDF PA to be acceptable for modeling release and near field transport in the 2020 SDF PA model because: (1) the NRC staff found the process the DOE used to assign Kd values to be reasonable and (2) the DOE sensitivity analyses demonstrated that reasonable changes in Kd values for radionuclides other than iodine and technetium would have an insignificant effect on the model projections. In the 2012 NRC TER, the NRC staff indicated that the presence of the salt solution and absence of sand or aggregate in saltstone (i.e., as compared to concrete) could cause differences between radionuclide sorption in saltstone grout and concrete. However, the NRC staff determined that applying the same sorption coefficients to saltstone and disposal structure concrete was acceptable for less risk-significant radionuclides. For the 2020 SDF PA, the NRC staff determined that technetium and iodine are the only elements for which the sorption coefficients in saltstone and concrete are risk-significant. Although isotopes of plutonium, tin, and americium contributed to dose in certain intrusion scenarios, the exposure pathways that led to dose from those radionuclides in those intrusion scenarios did not depend significantly on the radionuclide sorption in grout and concrete. The NRC staff found that few sorption coefficients had changed significantly from the values used in the DOE 2009 SDF PA (see Table 14 in this TRR), which the NRC staff reviewed in the

NRC 2012 TER. For elements with Kd values that increased by a factor of 100 or less, the DOE provided a sensitivity analysis that shows that there is little effect on the projected dose and most of that difference can be attributed to changes in the iodine Kd (see Figure 18 above). The NRC staff evaluated the remaining radionuclides in two groups: (1) Sr and Ra, and (2) C, Cs, K, and Pt. For Sr and Ra, the most significant increased occurred for chemically reduced Region I conditions. Under those conditions, the Kd for Sr increased by a factor of 2,000 and the Kd for Ra increased by a factor of 12,000. The NRC staff found that the DOE provided an adequate technical justification for changing the sorption coefficients because the revised values reflect literature results that show precipitation of those elements in slag-bearing cementitious materials and the DOE developed effective Kd values to mimic the effects of the expected precipitation. For C, Cs, K, and Pt, the modeled Kd increased from a value of 0 mL/g in the DOE 2009 DSF PA. For those radionuclides, the NRC staff found the increase to be acceptable because the revised values were based on applicable literature values. In addition, for Cs and K, the NRC staff does not expect the change to be risk-significant because the revised value is still low (i.e., 2 mL/g). In the 2013 version of the Monitoring Plan, the NRC staff evaluated sorption coefficients for certain risk significant radionuclides under MF 5.04 (for saltstone) and MF 6.01 (for disposal structure concrete). The NRC staff monitored sorption coefficients for radionuclides that had not been identified as having risk-significant assumptions related to sorption under MF 10.04 (for saltstone) and MF 10.06 (for disposal structure concrete). During the NRC staff review of the 2020 SDF PA, the NRC staff determined that it did not have specific concerns about sorption coefficients other than the issues described for iodine and Tc as described in Section 8.5.5 and 8.5.6 in this TRR. In addition, the NRC staff determined that it would reevaluate the risk significance of sorption coefficients for all radionuclides as a routine part of any PA review. Therefore, for clarity, NFFT-9 The NRC staff recommends closing MF 10.04 (Kd Values for Saltstone) and MF 10.06 (Kd Values for Disposal Structure Concrete). The NRC staff recommends changing the names of MF 5.04 and MF 6.01 to include all radionuclides, so that the name of MF 5.04 would be Sorption Coefficients in Saltstone and the name of MF 6.01 would be Sorption Coefficients in Disposal Structure Concrete. The NRC staff recommends updating the text of MF 5.04 and MF 6.01 to indicate the NRC staff should reevaluate the risk significance of sorption coefficients for all radionuclides in any PA review. 8.6. Monitoring Factor Update MF 3.03 Applicability of Laboratory Data to Field-Emplaced Saltstone The NRC staff determined that the modeled value for the initial hydraulic conductivity of saltstone (i.e., 5.0x10-10 cm/s) is reasonable for two reasons. First, analysis of saltstone core samples provides direct information and the use of the dynamic leaching method is reasonable. Second, in the 2020 SDF PA, the DOE assumed that saltstone will degrade. As such, the initial hydraulic conductivity of saltstone is less risk-significant than the assumed rate and extent of degradation. Because the initial saturated hydraulic conductivity of saltstone is still important to SDF performance depending on the DOEs assumed degradation, the NRC staff recommends changing the priority of MF 3.03 (Applicability of Laboratory Data to Field-Emplaced Saltstone) from high priority to medium under both PO §61.41and PO §61.42 and still with the narrowed scope (i.e., understanding changes in the saltstone hydraulic conductivity in the first several pore volume exchanges) as indicated in recommendation NFFT-3 in this TRR.

MF 4.01 Waste Form Matrix Degradation The NRC staff found the inclusion of saltstone degradation in the 2020 SDF PA to be a significant improvement in the site conceptual model. However, the NRC staff found the DOE conceptual model of saltstone degradation being controlled by decalcification to not be well supported. The NRC staff is concerned that: (1) decalcification could occur more quickly than projected due to potentially greater-than-assumed, (2) additional and coupled degradation mechanisms could result in more rapid degradation of saltstone than the DOE assumed in the PA, and (3) the use of a geometric average is not adequately supported and could significantly underestimate the effective hydraulic conductivity and effective diffusion coefficient of degraded saltstone. Based on the expected risk significance of degradation to the rate of release of radionuclides, the NRC staff recommends that MF 4.01 remain prioritized as high priority under 10 CFR 61.41 and 10 CFR 61.42. MF 4.02 Waste Form Macroscopic Fracturing In the NRC 2013 SDF Monitoring Plan, the NRC staff described that saltstone fracturing was important to site performance because it: (1) increases flow through the saltstone, (2) shortens the diffusive length for radionuclide release, and (3) provides additional surface area for the progression of saltstone oxidation, which increases Tc release. Also in the NRC 2013 SDF Monitoring Plan, the NRC staff described that they expected to close MF 4.02 under both §61.41 and §61.42 after the NRC determines that model support for the assumed formation of macroscopic fractures in saltstone during the performance period was sufficient. As discussed in the NRC staff TRR entitled Saltstone Waste Form Physical Degradation (ML19031B221), the DOE has not provided sufficient information to justify the assumed degradation mechanisms at the time the NRC staff issued that TRR (May 2019). The NRC staff also determined that the DOE did not provide additional support for the assumed formation of macroscopic fractures in saltstone in the 2020 SDF PA or the DOE response to the NRC staff RSI or RAI questions. As discussed above, the DOEs hypothetical fast flow paths sensitivity analysis described in Section 5.8.8.2 of the 2020 SDF PA did not provide sufficient information for NRC staff to evaluate the risk associated with macroscopic fracturing because of the extremely low percolation assumed by the DOE through the Closure Cap. Accordingly, the timing, rate, and extent of saltstone fracturing is not clear. Based on the uncertainty in saltstone fracturing, the NRC staff recommends keeping MF 4.02 open under both 10 CFR 61.41 and 10 CFR 61.42. In addition, if flow through chemically reduced saltstone grout is the risk-controlling source of radionuclide release, then saltstone fracturing could ultimately reduce radionuclide release as fractures may decrease the interaction between infiltrating water and the saltstone grout. However, fracturing could still result in increased radionuclide release relative to the DOE Compliance Case due to decreased diffusive lengths, oxidation of saltstone grout prior to contact with water, and increased flow through saltstone grout, depending on the type and extent of fracturing (e.g., large-scale network of small fractures). Therefore, based on the potential risk significance of fracturing on radionuclide release, the NRC staff recommends that MF 4.02 remain prioritized as high priority under both 10 CFR 61.41 and 10 CFR 61.42. In the 2020 SDF PA, the DOE assumed that contaminant release from the disposal structures is spread across the nodes beneath the footprint of the disposal structures. However, the DOE has observed localized leakage at a discrete layer during a hydrotest. Accordingly, it is not clear what impact certain layers or a combination of layers may have on the areal extent of contaminant release. In addition, the results of a sensitivity analysis conducted by the DOE demonstrated the risk significance of the assumed distribution of

contaminant release from the disposal structures. Accordingly, NRC staff recommends monitoring information related to localized contaminant release as indicated in recommendation NFFT-1 in this TRR. In the 2020 SDF PA, the DOE assumed that the saltstone would not degrade until the overlying disposal structure roofs were degraded. The NRC staff did not find the DOE assumed delay to saltstone degradation to be well supported, because of the potential for additional degradation mechanisms to impact saltstone. For the Compliance Case, the delay in wasteform degradation did not significantly impact the dose results because the DOE assumed that degradation does not appreciably increase the saturated hydraulic conductivity of saltstone until approximately 100,000 years after closure. However, changes to modeling assumptions in future analyses could increase the risk significance of early degradation, and therefore early release. Accordingly, the NRC staff recommends monitoring information related to the delay in degradation of saltstone. MF 5.01 Radionuclide Release from Field-Emplaced Saltstone The 2013 NRC SDF Monitoring Plan indicates that the NRC expects to close MF 5.01 (Radionuclide Release from Field-Emplaced Saltstone) under both §61.41 and §61.42 POs when the NRC determines that measurements of radionuclide release rates from field-emplaced saltstone used in the PA are reliable. In a 2018 TRR on technetium release, the NRC staff recommended that the NRC keep MF 5.01 open with high priority. In that TRR, the NRC staff indicated the expectation that DOE would (1) incorporate data from saltstone cores from SDS 2A and (2) provide technical justification for the projected duration of the conditions associated with young cementitious material. In the 2020 SDF PA, the DOE used sorption coefficients for iodine and technetium, as well as solubility limits for technetium, that the DOE based on experiments with both cores of field-emplaced saltstone and saltstone simulants. Those measurements applied to chemically reduced Region I and Region III conditions. The measurements also provided a basis for the projected duration of Region I (i.e., very young conditions). In contrast, comparable results were not available to support oxidized conditions. Although the DOE does not expect the entire saltstone monolith to become oxidized during the 10,000-year performance period in the Compliance Case, the VZTM projects an oxidized rind to develop around saltstone during the performance period. That rind affects the modeled release of iodine and technetium during the performance period because of the shrinking core model the DOE used to project iodine and technetium release. Although additional support is needed to support modeled iodine and technetium release under oxidized Region III conditions, the NRC staff found that the DOE adequately supported the chemical parameters affecting the release of iodine and technetium in chemically reduced Region I and Region III conditions. Because (1) Tc-99 and I-129 are the two radionuclides with the greatest dose contribution in the 2020 SDF PA, and (2) the DOE probabilistic sensitivity analysis identified technetium solubility and iodine sorption in reduced saltstone as two of the most risk-significant parameters affecting the projected dose to an offsite member of the public, the NRC staff determined that the remaining areas that need model support do not justify designation as a high priority MF. Therefore: NFFT-10 The NRC staff recommends reducing the priority from high to low for MF 5.01 (Radionuclide Release from Field-Emplace Saltstone) under MA 5 (Waste Form Chemical Performance) under 10 CFR 61.41 and 61.42.

MF 5.02 Chemical Reduction of Tc by Saltstone In a 2018 TRR on technetium release, the NRC staff indicated that the staff expected to close MF 5.02 when: (1) the DOE showed that the modeled technetium release was consistent with measurements from cores of field-emplaced saltstone and (2) the DOE provided physical evidence of Tc re-reduction in reduced cementitious material. As described in Section 7.4.6, the DOE provided a basis for the solubility limits in cores of field-emplaced saltstone under chemically reduced Region I and Region III conditions. Although the DOE did not provide evidence for re-reduction of saltstone when it flows from chemically oxidized areas through chemically reduced areas, the increased releases of technetium from chemically reduced areas (i.e., compared to the relative immobility of technetium in reduced saltstone modeled in the 2009 SDF PA) reduce the risk significance of risk significance of modeled re-reduction of technetium when flowing through reduced areas of saltstone and disposal structure concrete. Therefore: NFFT-11 The NRC staff recommends reducing the priority from medium to low for MF 5.02 (Chemical Reduction of Tc by Saltstone) under MA 5 (Waste Form Chemical Performance) under 10 CFR 61.41 and 61.42. MF 5.03 Reducing Capacity of Saltstone The 2013 NRC SDF Monitoring Plan indicates the NRC staff expects to close MF 5.03 under both §61.41 and §61.42 after the NRC determines that information for the initial reducing capacity of saltstone and the expected evolution of redox conditions over time is adequate. In a 2018 TRR on technetium release, The NRC staff recommended keeping MF 5.03 open under both §61.41 and §61.42 because the DOE did not directly address the implications of potential sulfur dissolution and release on the projected evolution of saltstone reducing capacity or the applicability of the cerium method of measuring reducing capacity to the species that reduce technetium in saltstone. However, based on sensitivity analyses described in that 2018 TRR, the NRC staff recommended reducing the priority of MF 5.03 under both §61.41 and §61.42 POs from medium to low. In the DOE document SRR-CWDA-2018-00048, Rev. 0, the DOE demonstrated that the cerium method resulted in significantly greater measured reducing capacity in simulated saltstone samples than the chromium method. However, in the 2020 SDF PA, the DOE did not address the NRC position in the 2018 TRR that the chromium method could be more applicable to saltstone and disposal structure concrete than the cerium method because the chromium method uses an alkaline environment that is more similar to the pore water conditions in saltstone and disposal structure concrete than the acidic environment used in the cerium method. Instead, the DOE grouped values from tests using the two measurement methods together and treated them as being equally applicable (see Section 7.4.4 in this TRR). In addition, the DOE did not address the NRC staff concern described in the NRC staff TRR on technetium release (ML18095A122) that the reducing capacity of saltstone and disposal structure concrete could be diminished by leaching of sulfur from those materials. In the 2020 SDF PA, the DOE provided sensitivity analyses that supported a low priority for the reducing capacity of saltstone, which is consistent with its current priority in the NRC Monitoring Plan. For that reason, in Section 7.4.4 in this TRR, the NRC staff did not recommend any changes to the status or priority of MF 5.03. MF 6.01 Certain Risk-Significant Kd Values in Disposal Structure Concrete As described in Section 7.4.5 in this TRR, the NRC staff determined there was not adequate support for the assumption that reduced iodine would become oxidized as it flowed through the matrix or fractures of oxidized disposal structure concrete. Because that modeling assumption led to projected reconcentration of iodine in the lower mud mats of some disposal structures, in

Section 7.4.5 in this TRR, the NRC staff recommended monitoring the development of information about the sorption of iodine in oxidized mud mats as indicated in monitoring recommendation NFFT-3 in this TRR. MF 6.04 Disposal Structure Concrete Fracturing In the NRC 2013 SDF Monitoring Plan, the NRC staff described that the staff expected to close MF 6.04 (Disposal Structure Concrete Fracturing) under PO §61.41 and PO §61.42 after NRC determines that support for the amount of fracturing of the disposal structure floor and walls expected to occur during the performance period is adequate or if NRC determines that the estimate that DOE uses in the PA model is conservative. In the 2020 SDF PA, the DOE assumed that sulfate attack would be the rate-determining degradation mechanism for disposal structure concrete. In addition, the DOE relied upon a geometric average of the intact and degraded materials as a function of time for determining the hydraulic conductivity. The NRC has concerns with the limited support for both of these assumptions. The use of different averaging schemes for the disposal structure concrete does not have as significant of an effect on overall SDF performance as it does for saltstone. However, if the DOE relies more heavily on disposal structure performance in the future, then the NRC staff will review the impacts of degradation averaging on overall SDF performance. Accordingly, the NRC staff recommends keeping MF 6.04 open under PO §61.41 and PO §61.42 with a medium priority. MF 6.05 Integrity of Non-cementitious Materials In the NRC 2013 SDF Monitoring Plan, the NRC staff described that the staff expected to close MF 6.05 (Integrity of non-cementitious materials) under PO §61.41 and PO §61.42 after NRC determines that support for the assumed performance of non-cementitious materials used in the disposal structures is adequate. The NRC staff found the DOE inclusion of fast pathways through the disposal structures in the VZFM and VZTM to be acceptable because the presence of fast pathways is consistent with observations of fast pathways and contaminant transport through the SDS 4 walls and hydrotests on the 150-ft and 375-ft diameter disposal structures. The risk significance of these features will depend on the performance of other key barriers (e.g., closure cap, LLDL and HDPE/GCL composite layer, saltstone). As the DOE continues to develop information related to the closure cap, the NRC staff will continue to evaluate the impact of these features on SDF performance. Impacts due to preferential pathways include higher flow rates, earlier releases, and decreased attenuation. In contrast, preferential pathways could also result in potential bypass of water around the saltstone grout, which could reduce radionuclide release by diverting water from the radioactive inventory. The degradation processes the DOE modeled for the columns in the cylindrical disposal structures results in earlier modeled degradation for the columns than for saltstone. Accordingly, the columns can act as pathways for infiltrating water to migrate around the saltstone grout. Because of the potential for modeled bypass to affect the projected dose, the NRC staff will continue to review information related to the performance of non-cementitious materials and bypass flow. The NRC staff recommends keeping MF 6.05 open under PO §61.41 and PO §61.42 as a medium priority. NFFT-10.05 Moisture Characteristic Curves The NRC staff found the support for the assumed MCCs for cementitious materials in the 2020 SDF PA to be adequate, because the assumed MCCs are similar to literature values and the MCCs are of limited risk significance in the 2020 SDF PA. However, because future revisions to modeling assumptions could result in the MCCs becoming more risk-significant, the NRC staff recommends keeping MF 10.05 (Moisture Characteristic Curves) open but with a low priority

under both PO §61.41and PO §61.42. Therefore, the NRC staff recommends keeping MF 10.05 (MCCs) open and with a low priority under both 10 CFR 61.41and 10 CFR 61.42.

9. Teleconference or Meeting There were no teleconferences or meetings with the DOE related to this TRR.

10. Follow-up Actions Except for NRC staff recommendations to revise the 2013 NRC SDF Monitoring Plan, there are no specific Follow-up Actions related to this TRR.

11. Conclusions The NRC staff determined that the near field flow and transport modeling in the 2020 SDF PA included several significant improvements from the 2009 SDF PA, including degradation of saltstone grout in the Compliance Case, initial properties of saltstone grout were based on field-emplaced saltstone core samples, a solubility-controlled release of Tc-99 based on laboratory analysis of saltstone core samples, an updated Kd value for I-129, a reduction in the assumed performance of the disposal structures, and MCCs that are less risk-significant and more consistent with the literature. However, the NRC staff determined that the information provided in the 2020 SDF PA and supporting documents did not provide an adequate basis for the NRC staff to fully assess near field flow and transport because the DOE indicated that the design and implementation plan for risk-significant features of the SDF (i.e., the closure cap) is not complete.

The DOE did provide additional information in response to requests for supplemental and additional information to address significant uncertainties in several key barriers (e.g., closure cap, lower lateral sand drainage layer and composite barrier, saltstone grout). Although those analyses did address the range of uncertainty in key barriers, many of the realizations exceeded the performance objectives. Accordingly, the NRC staff identified the following changes to the NRC SDF Monitoring Plan to assess the near field flow and transport of the SDF as the DOE finalizes its closure cap design and implementation plan: NFFT-1 The NRC staff recommends monitoring information related to localized contaminant release from the saltstone disposal structures under a new MF entitled Localized Contaminant Release) under Monitoring Area 6 (Disposal Structure Performance) under both 10 CFR 61.41and 10 CFR 61.42. NFFT-2 The NRC staff recommends changing the priority of MF 3.03 (Applicability of Laboratory Data to Field-Emplaced Saltstone) from high priority to medium under both PO §61.41and PO §61.42 and still with the narrowed scope (i.e., understanding changes in the saltstone hydraulic conductivity in the first several pore volume exchanges). NFFT-3 The NRC staff recommends monitoring information related to the delay in degradation of saltstone grout under both 10 CFR 61.41and 10 CFR 61.42. NFFT-4 The NRC staff recommends monitoring the development of information about the sorption of iodine in oxidized mud mats under MF 6.01 (Certain Risk-Significant Kd values in Disposal Structure Concrete) under MA 6 (Disposal Structure Performance)

under 10 CFR 61.41 and 10 CFR 61.42. The NRC staff should consider the potential for fast pathways that could limit the ability of the mud mats to chemically oxidize iodine as it flows through the mud mats. NFFT-5 If the DOE chooses to use cement-free saltstone, the NRC staff recommends monitoring the development of information about iodine sorption in chemically oxidized cement-free saltstone under MF 5.04 (Certain Risk-Significant Kd Values for Saltstone) under MA 5 (Waste Form Chemical Performance) under 10 CFR 61.41 and 10 CFR 61.42. NFFT-6 If the DOE chooses to use cement-free saltstone, the NRC staff recommends monitoring the development of information about technetium solubility in cement-free saltstone under MF 5.04 under MA 5 (Waste Form Chemical Performance) under 10 CFR 61.41 and 10 CFR 61.42. To reflect the use of a solubility limit to represent technetium transport in chemically reduced saltstone, the NRC staff recommends changing the name of the MF to Certain Risk-Significant Kd Values and Solubility Limits for Saltstone. NFFT-7 The NRC staff recommends reducing the priority from high to low for MF 5.01 (Radionuclide Release from Field-Emplace Saltstone) under MA 5 (Waste Form Chemical Performance) under 10 CFR 61.41 and 61.42. NFFT-8 The NRC staff recommends reducing the priority from medium to low for MF 5.02 (Chemical Reduction of Tc by Saltstone) under MA 5 (Waste Form Chemical Performance) under 10 CFR 61.41 and 61.42.

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___, SRR-CWDA-2021-00040, Rev.0, Evaluation of the Uncertainties Associated with the SDF Closure Cap and Long-Term Infiltration Rates, June 2021. ML21160A064 ___, SRR-CWDA-2020-00064, Rev. 1, FY 2020 Special Analysis for the Saltstone Disposal Facility at the Savannah River Site, April 2021. ML21232A639 ___, SRR-CWDA-2021-00052, Rev. 0, Memorandum: Supplemental Information and Proposed Probabilistic Inputs Related to NRC RSI-4: Saltstone Degradation, July 2021. ML21232A630 ___, SRR-CWDA-2021-00056, Rev. 0, Evaluation of the Uncertainties Associated with Long-Term Saltstone Degradation, July 2021. ML21217A081 ___, SRR-CWDA-2021-00057, Rev. 0, Memorandum: Selected PORFLOW Modeling Configurations to Support RSI-4 Analysis, July 2021. ML21232A633 ___, SRR-CWDA-2021-00066, Rev. 0, Evaluation of the Combined Uncertainties Associated with the Long-Term Performance of Saltstone Disposal Facility Flow Barriers, August 2021. ML21217A083 ___, SRR-CWDA-2021-00072, Rev. 1, Comment Response Matrix for the Second Set of U.S. NRC Request for Additional Information on the 2020 Performance Assessment for the Saltstone Disposal Facility at the Savannah River Site, November 2021. ML21148A005 ___, SRMC-CWDA-2022-00003, Rev. 1, Comment Response Matrix for the Third Set of U.S. NRC Request for Additional Information on the Performance Assessment for the Saltstone Disposal Facility at the Savannah River Site, March 2022. ML22083A049 ___, SRMC-CWDA-2022-00025, Rev. 3, Leak Rate Considerations Related to the SDU 6 Sumps, June 2022. ML22189A149 ___, SRR-CWDA-2021-00098, Rev. 0, Memorandum: Supplemental Analysis of Screened RSI-1 Model Results, November 2021. ML21326A013 ___, SRR-SDU-2017-00001, Rev. 0, Achieving Leak Tightness in New Savannah River Site Saltstone Disposal Units, WM2017 Conference, March 2017. ML20206L191 ___, WSRC-STI-2008-00236, Rev. 0, Thermodynamic and Mass Balance Analysis of Expansive Phase Precipitation in Saltstone. May 2008. ML101600398 U.S. Nuclear Regulatory Commission, Letter to T. Gutmann, DOE RE: Second RAI on the 2009 Performance Assessment for the Saltstone Disposal Facility at the Savannah River Site, December 2010. ML103400571 ___, TER for the Performance Assessment for the Saltstone Disposal Facility at the Savannah River Site, Rev. 1, April 2012. ML121170309 ___, NDAA WIR Monitoring Plan for the SRS Saltstone Disposal Facility, Rev.1, September 2013. ML13100A113 ___, Technical Review: Solubility of Technetium Dioxides, November 2013. ML13304B159

___, Technical Review: Oxidation of Reducing Cementitious Waste Forms, June 2015. ML15098A031 ___, Technical Review: QA Documentation for the CBP Toolbox, August 2016. ML16196A179 ___, Technical Review: Iodine Sorption Coefficients for Use in Performance Assessments for the Saltstone Disposal Facility, January 2017. ML16342C575 ___, Technical Review: Waste Form Hydraulic Properties, March 2017. ML17018A137 ___, Technical Review: Performance of the HDPE Layer, HDPE/GCL Composite Layer, and the LLDL, April 2017. ML17081A187 ___, Technical Review: Update on the Projected Technetium Release from Saltstone, May 2018. ML18095A122 ___, Technical Review: Saltstone Waste Form Physical Degradation, May 2019. ML19031B221 ___, Preliminary Review of the U.S. DOE 2020 Performance Assessment for the Savannah River Site Saltstone Disposal Facility, October 2020. ML20254A003 ___, Third Set of Request for Additional Information Questions for the U.S. Department of Energy 2020 Performance Assessment for the Savannah River Site Saltstone Disposal Facility, March 2022. ML21341A55}}

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