Screening Analysis of Criticality Features, Events, and Processes for License Application Rev 00, ICN 00

ANL-EBS-NU-000008

February 2004
 
1. PURPOSE 
The purpose and scope of this analysis report is to establish and justify the input parameters used 
in the screening analysis of postclosure criticality related features, events, and processes (FEPs) 
for use in the license application, to perform the screening analysis, and to document the analysis 
results. The results of the analysis are to be used to support criticalitys inclusion into, or 
exclusion from, the Total System Performance Assessment for the License Application 
(TSPA-LA). 
The analysis calculates the probability of criticality resulting from degradation processes (inpackage 
and external) as well as disruptive events (i.e., seismic, igneous, and rockfall). 
Probability evaluations are performed utilizing the configuration generator model described in 
Configuration Generator Model for In-Package Criticality (BSC 2003 [DIRS 165629]), a 
component of the methodology from Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). The configuration generator model is utilized for the probability 
evaluation of the base case in-package degradation process, the seismic disruptive event and the 
rockfall disruptive event. Probability evaluations of the base case external processes and the 
igneous disruptive event are performed using analytical arguments. The probability analysis 
results are then compared to the regulatory probability criterion (10 CFR 63.114(d) [DIRS 
156605]) for including or excluding FEPs from evaluation in the TSPA-LA. This comparison is 
the basis of the screening recommendation for the criticality FEPs. 
The limitations of the analysis are: 
 Only specific information and data for the 21-PWR with Absorber Plates, 12-PWR 
Long, 44-BWR, and 24-BWR waste package types were utilized. Assumptions (which 
require confirmation) were utilized to extend the probability evaluation to the 21-PWR 
with Control Rods and DOE SNF waste package types. 
 To date, model reports necessary to support the criticality FEPs screening analysis have 
been developed and validated only for PWR commercial SNF. Therefore, it is not 
possible to extend these models to other waste forms (including BWR commercial SNF) 
and, at the same time, strictly adhere to the criticality analysis methodology outlined in 
Disposal Criticality Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). 
 The inputs used to establish probabilities for this analysis are based on information and 
data for the TSPA-LA, where available. Information and data for the Total System 
Performance Assessment for the Site Recommendation (TSPA-SR) are used where the 
TSPA-LA data are not available. In addition, modification of any of the TSPA-LA 
information used in the development of this analysis could necessitate an update to the 
criticality FEPs screening analysis. 
 The current probability evaluation is extended only to the point of waste package 
flooding as developed in the configuration generator model. If necessary to gain further 
reductions in the total probability of criticality, it is possible to extend the evaluation 
February 2004 15 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
beyond this point to include the probability of configuration class formation and 
configuration class criticality potential. 
The activity of developing this screening analysis is defined in Technical Work Plan for: Risk 
and Criticality Department (BSC 2003 [DIRS 165559]). There were no deviations from this 
plan. 
February 2004 16 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
2. QUALITY ASSURANCE 
Technical Work Plan for: Risk and Criticality Department (BSC 2003 [DIRS 165559], 
Section 8) determined that the development of this analysis report and the associated activities 
are subject to Quality Assurance Requirements and Description (DOE 2003 [DIRS 162903]). 
This report contributes to the analysis and modeling used to support performance assessment. 
This analysis report investigates the performance of the following natural and engineered barriers 
that are important to waste isolation: 
 Commercial Spent Nuclear Fuel Cladding 
 DOE and Commercial Waste Packages 
 Emplacement Drift Invert 
 Drip Shield 
 Saturated Zone (between the repository and the accessible environment) 
 Surface Topography, Soils and Bedrock 
 Unsaturated Zone above the Repository 
 Unsaturated Zone below the Repository 
 Waste Form 
Although these barriers are categorized as Safety Category in Q-List (BSC 2003 [DIRS 
165179]), the evaluations and conclusions do not directly impact the features important to safety, 
defined in AP-2.22Q, Classification Analyses and Maintenance of the Q-List [DIRS 164786]. 
The methods used to control the electronic management of data as required by AP-SV.1Q, 
Control of the Electronic Management of Information [DIRS 165687], are identified in 
Technical Work Plan for: Risk and Criticality Department (BSC 2003 [DIRS 165559], 
Section 8). 
Also in accordance with Technical Work Plan for: Risk and Criticality Department (BSC 2003 
[DIRS 165559], Table 1), development of this analysis was controlled by AP-SIII.9Q, Scientific 
Analyses [DIRS 164456]. 
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INTENTIONALLY LEFT BLANK 
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3. USE OF SOFTWARE 
3.1 QUALIFIED AND BASELINE SOFTWARE 
3.1.1 MCNP 
 Title: MCNP 
 Version/Revision number: Version 4B2LV 
 Software Tracking Number (STN): 30033-V4B2LV 
 Status/Operating System: Qualified/HP-UX B.10.20 
 Computer type: Hewlett Packard (HP) 9000 Series Workstations 
 Computer processing unit number: CRWMS M&O Tag 700887 
Input and output files for the various MCNP calculations are provided in Attachment VII. The 
MCNP software is: (1) appropriate for the application of keff calculations; (2) used only within 
the range of validation as documented throughout MCNP-A General Monte Carlo N-Particle 
Transport Code (Briesmeister 1997 [DIRS 103897]), Software Qualification Report for MCNP 
Version 4B2, A General Monte Carlo N-Particle Transport Code (CRWMS M&O 1998 [DIRS 
102836]), Software Code: MCNP (CRWMS M&O 1998 [DIRS 154060]); and (3) obtained from 
Software Configuration Management in accordance with AP-SI.1Q, Software Management 
[DIRS 165023]. 
Input and output files for the MCNP calculations may be found in Attachment VII (a CD-ROM). 
The input files in Attachment VII allow an independent reproduction of the calculations. 
3.1.2 SAPHIRE 
 Title: SAPHIRE 
 Version/Revision number: 7.18 
 Software Tracking Number (STN): 10325-7.18-00 
 Status/Operating System: Microsoft Windows 2000 Professional 
 Computer Type: DELL Latitude C640 Laptop PC 
 Computer processing unit number: CRWMS M&O Tag number 501215 
The software code SAPHIRE V7.18 (BSC 2002 [DIRS 160873]) was used to develop and 
quantify event trees and fault trees in this analysis. SAPHIRE (Systems Analysis Programs for 
Hands-on Integrated Reliability Evaluations) is a state-of-the-art probabilistic risk analysis 
software program that utilizes an integrated event tree/fault tree methodology to develop and 
analyze the logical interactions that may occur between systems and components to determine 
the probability or frequency of an events occurrence. 
SAPHIRE is qualified software that was obtained from Software Configuration Management. It 
is appropriate for use in the present analysis, and is used only within its range of validation, in 
accordance with AP-SI.1Q [DIRS 165023]. 
February 2004 19 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The event trees, fault trees, and logic rules developed for the SAPHIRE calculations are 
documented in Attachment II. All of the electronic files necessary for the performance of the 
SAPHIRE calculation may be found in Attachment VII (a CD-ROM). The input files in 
Attachment VII allow an independent reproduction of the calculations. 
3.2 CONTROLLED SOFTWARE 
3.2.1 EXCEL 
 Title: EXCEL 
 Version/Revision number: Microsoft Excel 97 SR-2 
 Status/Operating System: Microsoft Windows 2000 Professional 
 Computer Type: DELL OptiPlex GX260 PC 
 Computer processing unit number: CRWMS M&O Tag number 152855 
Microsoft Excel for Windows, Version 97 SR-2, is used in this analysis to manipulate the inputs 
using standard mathematical expressions and operations. It is also used to tabulate and chart 
results. The user-defined formulas, inputs, and results are documented in sufficient detail to 
allow an independent repetition of computations. Thus, Microsoft Excel is used only as a 
worksheet and not as a software routine. Microsoft Excel 97 SR-2 is controlled under the 
Software Configuration Management, but is not required to be qualified as specified in 
Sections 2.1.1 and 2.1.6 of AP-SI.1Q [DIRS 165023]. 
Electronic files of the EXCEL calculations used in this analysis may be found in Attachment VII 
(a CD-ROM). The input files in Attachment VII allow an independent reproduction of the 
calculations. 
3.2.2 Mathcad 
 Title: Mathcad 
 Version/Revision number: Mathsoft Engineering and Education, Inc. Mathcad 2001i 
Professional 
 Status/Operating System: Microsoft Windows 2000 Professional 
 Computer Type: DELL OptiPlex GX260 PC 
 Computer processing unit number: CRWMS M&O Tag number 152369 
Mathcad for Windows 2000, Version 2001i Professional, is a problem-solving environment 
used in calculations and analysis. It is also used to tabulate and chart results. The user-defined 
expressions, inputs, and results are documented in sufficient detail to allow an independent 
repetition of computations. Thus, Mathcad is used as a worksheet and not as a software routine. 
Mathcad is controlled under the Software Configuration Management, but is not required to be 
qualified as specified in Sections 2.1.1 and 2.1.6 of AP-SI.1Q [DIRS 165023]. 
Input and output files for the various Mathcad calculations are documented in Attachments III, 
IV, and V. The electronic files of these calculations may be found in Attachment VII (a CDROM). 
The input files in Attachment VII allow an independent reproduction of the calculations. 
February 2004 20 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
4. INPUTS 
4.1 DIRECT INPUTS 
The following sections present the data, parameters, and technical information used to perform 
the criticality FEPs screening analysis. 
4.1.1 Data 
The following data were utilized in the development of this analysis. Use of these data is 
justified as they come from qualified project sources and their application is compatible with 
their developed purpose and limitations. 
Corrosion Rate Data for Stainless Steel Type 316 4.1.1.1 
The corrosion rate for Stainless Steel Type 316 was used to determine the corrosion rate for 
Neutronit A978. The corrosion rates for Stainless Steel Type 316 are from 
DTN: MO0401SPAMCRAE.000 ([DIRS 166801]). The Stainless Steel Type 316 corrosion 
rates based on J-13 well water are used to represent the corrosion rate of Neutronit A978. 
However, the corrosion rate for Neutronit A978 can be greater than Stainless Steel Type 316 
because of the boron added; therefore, the corrosion rate will be enhanced by a factor of 1.5 (see 
Assumption 5.2.3). The corrosion rate data listed in Table 4.1-1, along with the corrosion rate 
data multiplied by the enhancement factor of 1.5, was used to fit a Weibull distribution. These 
data are used in Attachment V. 
Table 4.1-1. Stainless Steel Type 316 Corrosion Rate Data and 1.5 Times the Corrosion Rate 
February 2004 
1.5 times Stainless Steel 
Type 316 Corrosion Rates 
(m/yr) 
0.037 0.055 
0.102 0.153 
0.109 0.164 
0.152 0.228 
0.154 0.231 
0.178 0.267 
0.203 0.305 
0.229 0.344 
0.229 0.344 
0.254 0.381 
0.254 0.381 
0.254 0.381 
0.279 0.419 
Source: a DTN: MO0401SPAMCRAE.000 ([DIRS 
166801]), aqueous-316L.xls 
21 of 176 
SS Type 316 
Corrosion Ratesa 
(m/yr) 
ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
4.1.2 Parameters 
The following parameters were used to develop this analysis. Use of these parameters is justified 
because they come from qualified project sources and their application is compatible with their 
developed purpose and limitations. 
Configuration Generator Model 4.1.2.1 
This report utilizes the configuration generator model from Configuration Generator Model for 
In-Package Criticality (BSC 2003 [DIRS 165629], Attachment XII) to perform event tree / fault 
tree probability evaluations to support the criticality FEPs screening analysis. Information and 
parameters necessary to define the configuration generator model inputs are presented 
throughout Section 4.1. The configuration generator model inputs are developed and evaluated 
in Sections 6.3 through 6.5. Documentation of the configuration generator model used in the 
criticality FEPs screening analysis is provided in Section 6.2 and Attachment II. 
Seepage Rate Information 4.1.2.2 
The seepage rate is determined from the inputs discussed in the Abstraction of Drift Seepage 
(BSC 2003 [DIRS 165564], Section 6.7.1). The seepage flux is a function of three parameters: 
capillary strength (1/ ) permeability (k), and adjusted percolation flux (qperc,ff). The values for 
each these parameters will be discussed. 
Capillary strength (1/ ) is developed into two separate distributions, one to account for spatial 
variability and the second to account for uncertainty. The spatial variability follows a uniform 
distribution with a mean of 591 Pa, a lower bound of 402 Pa, and an upper bound of 780 Pa. The 
uncertainty ( .1/ ) is represented by a triangular distribution with a mean of 0.0 Pa, a lower 
bound of -105 Pa, and an upper bound of 105 Pa. These distributions are applicable for all 
geologic repository zones. 
Permeability (k) is developed into two separate distributions, one to account for spatial 
variability and the other to account for uncertainty. The spatial variability for permeability was 
statistically analyzed using log-transformed data and found to follow a lognormal distribution (in 
log 10) (BSC 2003 [DIRS 165564], Section 6.6.2.1). The uncertainty ( .k) follows a triangular 
distribution. Depending on the geologic repository zone, there are different values for the 
lognormal distribution and the triangular distribution. 
Lithophysal zone: 
Lognormal distribution mean is -11.5 and standard deviation is 0.47 (in log 10). 
Triangular distribution mean is 0.0, lower bound is -0.92, and upper bound is 0.92. 
Nonlithophysal zone: 
Lognormal distribution mean is -12.2 and standard deviation is 0.34 (in log 10). 
Triangular distribution mean is 0.0, lower bound is -0.68, and upper bound is 0.68. 
The percolation flux for the glacial transition climate used in this analysis is from 
DTN: LB0310AMRU0120.002 ([DIRS 166116]) and is based on the percolation in the 
repository area only (BSC 2003 [DIRS 165564], Figure 6.6-10). The percolation flux for the 
February 2004 22 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
glacial transition climate is described using three different scenarios (i.e., lower-bound, mean, 
and upper-bound), which are used in this analysis. The probability associated with the three 
different percolation flux scenarios are 0.24, 0.41, and 0.35 for the lower-bound, mean, and 
upper-bound, respectively (BSC 2003 [DIRS 165991], Section 7, Table 7-1). These probabilities 
are based on the glacial transition climate excluding the contingency area. 
The final input for determining the seepage rate at the drift is the seepage rates, which are 
developed from lookup tables based on the three key parameters discussed above. The seepage 
rates are obtained through interpolation given a capillary strength (1/ ), permeability (k), and 
adjusted percolation flux (qperc,ff). The seepage rates are from DTN: LB0304SMDCREV2.002 
([DIRS 163687]) for nondegraded drifts and DTN: LB0307SEEPDRCL.002 ([DIRS 164337]) 
for degraded drifts. The seepage rates are adjusted to account for uncertainty, which follows a 
uniform distribution with a mean of 0.0, and lower-bound and upper-bound values of -1.7321 
and 1.7321, respectively. 
This information is used in Section 6.4.1.1.1 and Attachment IV. 
Mean Annual Seismic Exceedance Frequency Range and Time of Seismic Event 4.1.2.3 
The range of mean annual seismic exceedance frequencies is based on 
DTN: MO0308SPACALSS.002 [DIRS 164822], which follows a uniform distribution. The 
mean annual seismic exceedance frequency ranges from 10-8 to 10-4 per year. The time of 
occurrence of a seismic event ranges from repository closure to the performance period. This 
range is uniformly distributed from 1 year to 10,000 years (DTN: MO0308SPACALSS.002 
[DIRS 164822]). This information is used in Section 6.4.1.1 and Attachment III. 
Seismic Peak Ground Velocity 4.1.2.4 
The horizontal peak ground velocity (PGV) is related to the mean annual seismic exceedance 
frequency. This relationship was developed in Seismic Consequence Abstraction (BSC 2003 
[DIRS 161812], Section 6.4). The relationship between the PGV values and the mean annual 
seismic exceedance frequency was developed by scaling the PGV values at the monitored 
geologic repository (MGR) surface down to the drift. Based on this relationship scaled to the 
drift, the PGV values and their related mean annual seismic exceedance frequencies are listed in 
Table 4.1-2 (DTN: MO0308SPACALSS.002 [DIRS 164822]). This information is used in 
Section 6.4.1.1 and Attachment III. 
February 2004 23 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 4.1-2. Mean Annual Exceedance Frequency and Corresponding Peak Ground Velocity 
4.1.2.5 
Mean Annual Exceedance 
Frequency (1/yr) 
Drip Shield Failure from Seismic Event 
Damage to the drip shield can occur due to vibratory ground motion, which has the potential to 
allow advective flow to reach the waste package. The percent damaged area to the drip shield 
from a seismic event follows a uniform distribution. The lower bound of the uniform 
distribution for the percent damaged area is based on linear interpolation below a PGV value of 
5.35 m/s and linear extrapolation for PGV values above 5.35 m/s. The lower bound values are 
shown in Table 4.1-3 (DTN: MO0308SPACALSS.002 [DIRS 164822]). 
6.26  10-4 
2.78  10-4 
9.30  10-5 
1.84  10-5 
3.07  10-6 
2.28  10-7 
8.15  10-8 
2.60  10-8 
6.56  10-9 
Source: DTN: MO0308SPACALSS.002 [DIRS 164822] 
Peak Ground Velocity 
(m/s) 
0.159 
0.239 
0.398 
0.796 
1.59 
3.98 
5.57 
7.96 
11.9 
Table 4.1-3. Lower-Bound Percent Damaged Area to Drip Shield Due to Seismic Event 
PGV Value 
(m/s) 
ANL-EBS-NU-000008 REV 00 
0.00 0.0 
2.44 0.0 
5.35 10.0 
Source: DTN: MO0308SPACALSS.002 
[DIRS 164822] 
The upper bound of the uniform distribution is also correlated to the PGV value. Table 4.1-4 
provides the upper-bound percent damaged area of the drip shield based on PGV value. The 
upper-bound value can be interpolated for PGV values not directly listed. The input values 
(DTN: MO0308SPACALSS.002 [DIRS [DIRS 164822]) are listed in Table 4.1-4. 
Damaged Area to Drip Shield 
(percent) 
24 of 176 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
4.1.2.6 
4.1.2.7 
4.1.2.8 
Table 4.1-4. Upper-Bound Percent Damaged Area to Drip Shield Due to Seismic Event 
Damaged Area to Drip Shield 
(percent) 
0.00 0.00 
0.535 0.00 
2.44 2.68 
5.35 50.0 
20.0 50.0 
Source: DTN: MO0308SPACALSS.002 [DIRS 164822] 
This information is used in Section 6.4.1.1 and Attachment III. 
Drip Shield Failure from Seismically Induced Rockfall 
Damage to a drip shield from seismically induced rockfall uses a log triangular distribution. The 
minimum percent damaged area of the drip shield is 0.001 and the maximum percent damaged 
area of the drip shield is 100 (DTN: MO0308SPACALSS.002 [DIRS 164822]). This 
information is used in Section 6.4.1.1 and Attachment III. 
Damage to the waste package can occur due to vibratory ground motion, which can allow the 
infiltration of advective flow. The percent damaged area to the waste package from a seismic 
event follows a uniform distribution. The minimum percent damaged area is 0.0 
(DTN: MO0308SPACALSS.002) [DIRS 164822]. The upper bound of the uniform distribution 
for the percent damaged area of the waste package is correlated to the PGV value. This 
information is used in Section 6.4.1.1 and Attachment III. 
Waste Package and Drip Shield Fabrication Error Probabilities 
Waste package and drip shield fabrication and closure process error probabilities have been 
obtained from Analysis of Mechanisms for Early Waste Package/Drip Shield Failure (BSC 2003 
[DIRS 164475]). The waste package and drip shield fabrication and closure process error 
probabilities used in this analysis are presented in Table 4.1-5 and have been obtained from 
Table 20 of Analysis of Mechanisms for Early Waste Package/Drip Shield Failure (BSC 2003 
[DIRS 164475]). 
The waste package information of Table 4.1-5 is used in Section 6.3.3.3.4 and the drip shield 
information is used in Section 6.3.3.2.5. 
25 of 176 
PGV Value 
(m/s) 
Waste Package Failure 
ANL-EBS-NU-000008 REV 00 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 4.1-5. Defect Types to Consider for Waste Package and Drip Shield Performance 
Waste Package Defect Type 
Weld flaws 
Evaluation of Probability per Waste Package 
See Table 11 thru Table 13 of BSC 2003 [DIRS 164475] 
error factor = 15 
Lognormal distribution: 
Improper heat treatment grouped with Median = 7.2  10-6 per waste package 
improper laser peening and waste Mean = 2.8  10-5 per waste package 
package damaged by mishandling 
upper truncation value = 7.44213  10-3 per waste package 
Drip Shield Defect Type Main Characteristics 
Weld flaws 
Base metal flaws 
Improper heat treatment 
Damage by mishandling 
Mean number of flaws: 4.1 per drip shield 
Mean size of flaw: 1.3 mm 
See Table 18 of BSC 2003 [DIRS 164475] 
Mean probability: 1.3  10-5 per drip shield 
Mean probability: 4.8  10-7 per drip shield 
Source: BSC 2003 [DIRS 164475], Table 20 
It should be noted that one of the recommendations for modeling waste package damage due to 
improper heat treatment (grouped with improper laser peening and waste package damaged by 
mishandling) is to consider the entire waste package surface to be affected (BSC 2003 [DIRS 
164475], Section 6.4.8). This information is used in Section 6.3.3.3.4. 
Additionally, drip shield emplacement errors are calculated to be probable (BSC 2003 [DIRS 
164475, Section 6.3.7), but do not result in an advective flow path through the drip shield and 
onto the waste package (BSC 2003 [DIRS 164475, Section 6.4.7). This information is used in 
Section 6.3.3.2.4. 
Emplacement Drift Information 4.1.2.9 
Emplacement drift information is required to perform the criticality rockfall disruptive event 
FEPs analysis (Section 6.5), as it is important to account for the total number of drip shields 
available and how many drip shields are emplaced in the two geological zones  lithophysal and 
nonlithophysal. The total number of drip shields to be emplaced can be estimated by dividing 
the total emplacement drift length by the average length of a drip shield. The lithophysal and 
nonlithophysal fractional areas can be calculated by dividing the emplacement drift area of both 
geological zones by the total drift area. 
The total emplacement drift length can be calculated by summing the subtotals of the available 
emplacement drift lengths of each of the four panels presented in Tables 4 through 7 of RDP/PA 
IED Subsurface Facilities (BSC 2003 [DIRS 164490]). This information is summarized in 
Table 4.1-6 and the results used in Section 6.5.1. The drift emplacement area by geological unit 
is found in Table 9 of Repository Design Project, Repository/PA IED Subsurface Facilities IED 
(BSC 2003 [DIRS 164491]). This information is summarized in Table 4.1-7 and the results used 
in Sections 6.2 and 6.5.1. 
February 2004 26 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 4.1-6. Available Emplacement Drift Length 
Reference Panel Number Available Emplacement Drift Length 
(meters) 
4,092 
18,850 
24,000 
BSC 2003 [DIRS 164490], Table 4 
BSC 2003 [DIRS 164490], Table 5 
BSC 2003 [DIRS 164490], Table 6 
BSC 2003 [DIRS 164490], Table 7 17,003 
1 
2 
3 
4 
TOTAL 63,945 sum of rows 1 through 4 
Source: BSC 2003 [DIRS 164490], Tables 4 through 7 
Table 4.1-7. Drift Emplacement Area by Geological Unit 
Reference Geological Unit Drift Emplacement Area 
(square meters) 
BSC 2003 [DIRS 164491], Table 9 
BSC 2003 [DIRS 164491], Table 9 
BSC 2003 [DIRS 164491], Table 9 
BSC 2003 [DIRS 164491], Table 9 
sum of rows 1 and 3 
sum of rows 2 and 4 
Tptpul (lithophysal) 224,398 
Tptpmn (nonlitophysal) 616,003 
Tptpll (lithophysal) 4,013,268 
Tptpln (nonlithophysal) 129,483 
Total Lithophysal 4,237,666 
Total Nonlithophysal 745,486 
TOTAL 4,983,152 sum of rows 5 and 6 
Source: BSC 2003 [DIRS 164491], Table 9 
The average drip shield length is given as 5,805 mm (5.805 m) in D&E / PA/C IED Interlocking 
Drip Shield and Emplacement Pallet (BSC 2004 [DIRS 167309], Table 1). This value is used in 
Section 6.5.1. 
4.1.2.10 Waste Package Population 
Table 4.1-8 presents the percent breakdown of waste package by type for 70,000 metric tons of 
heavy metal (MTHM) currently proposed for disposal in the MGR. This information is obtained 
from D&E/PA/C IED Typical Waste Package Components Assembly (BSC 2004 [DIRS 167207], 
Table 11). It is used in Attachment II as the basis for the assignment of the basic event values for 
the waste form and waste package type fractions of event tree WP_TYPE. 
ANL-EBS-NU-000008 REV 00 February 2004 27 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 4.1-8. Breakdown of 70,000 MTHM Emplacement Inventory by Waste Package Type 
Nominal Waste 
Package Inventory 
for LA (%) Waste Package Design 
Nominal Waste 
Package 
Inventory for LA 
4299 
95 
163 
21-PWR with Absorber Plates 38.4 
21-PWR with Control Rods 0.8 
12-PWR Long 1.5 
44-BWR 2831 25.3 
24-BWR 84 0.8 
5-DHLW/DOE SNF Short (nonnaval spent nuclear fuel) 1147 10.3 
5-DHLW/DOE SNF Long (nonnaval spent nuclear fuel) 2116 a 18.9 
2-MCO/2-DHLW Long 149 1.3 
Naval SNF Short 144 1.3 
Naval SNF Long 156 1.4 
Total 11184 100.0 
Note: a includes waste package quantity for 5 HLW Long/1 DOE SNF Short (31) and 5 HLW Long Only (679) 
waste package configurations 
Source: BSC 2004 [DIRS 167207, Table 11 
4.1.2.11 Configuration Generator Model Input Parameters 
Table 4.1-9 documents the input sources for the basic event input values used in the SAPHIRE 
probability calculations. 
February 2004 28 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Event Tree Top 
Events 
Sufficient water 
reaches drift 
MS-IC-1 
Drip shield barrier 
penetration 
MS-IC-2 
Waste package 
barrier penetration 
MS-IC-3 
4.1.2.12 Magma Composition Parameters 
The average composition of magma predicted to enter the MGR drifts is provided in 
Table 4.1-10. This information is used as input to the MCNP calculations presented in 
Attachment VII. The information contained in Table 4.1-10 was obtained from Table 6 of 
Characterize Eruptive Processes at Yucca Mountain, Nevada (BSC 2003 [DIRS 166407]). 
ANL-EBS-NU-000008 REV 00 
Table 4.1-9. Configuration Generator Model Input Sources 
Basic Event Input 
Parameter 
Minimum seepage rate 
Drip shield failure due 
to stress corrosion 
cracking 
Drip shield failure due 
to pitting and crevice 
corrosion 
Drip shield failure due 
to emplacement error 
Drip shield failure due 
to fabrication errors 
Drip shield failure due 
to floor heave 
Drip shield failure due 
to thermal expansion 
Drip shield failure due 
to rockfall 
Drip shield failure due 
to a seismic event 
Waste package failure 
due to a seismic event 
Waste package failure 
due to fabrication 
errors 
Reference Document(s) Section 
Used 
 Abstraction of Drift Seepage (BSC 2003 [DIRS 
165564], Section 6.7.1 and Figure 6.6-10) 
 Analysis of Infiltration Uncertainty (BSC 2003 [DIRS 
Sections 
6.3, 6.4, 
and 6.5 
165991], Section 7, Table 7-1) 
 Stress Corrosion Cracking of the Drip Shield, the 
Waste Package Outer Barrier, and the Stainless Steel 
Structural Material (BSC 2003 [DIRS 161234], 
Section 6.3.7) 
 General Corrosion and Localized Corrosion of the Drip 
Shield (BSC 2003 [DIRS 161236], Section 6.4.3) 
 Analysis of Mechanisms for Early Waste Package/Drip 
Shield Failure, (BSC 2003 [DIRS 164475], 
Sections 6.3.7 and 6.4.7) Sections 
6.3, 6.4, 
and 6.5 
 Analysis of Mechanisms for Early Waste Package/Drip 
Shield Failure, (BSC 2003 [DIRS 164475], Tables 11, 
13, and 20) 
 Engineered Barrier System Features, Events, and 
Processes (BSC 2003 [DIRS 166464], Section 6.2.30) 
 EBS Radionuclide Transport Abstraction (BSC 2003 
[DIRS 166466], Section 6.3.1.3) 
 Seismic Consequence Abstraction (BSC 2003 [DIRS 
161812], Sections 6.6.2 and 6.6.1.1) 
 Seismic Consequence Abstraction (BSC 2003 [DIRS 
161812], Section 6.6.3) 
 Seismic Consequence Abstraction (BSC 2003 [DIRS 
161812], Section 6.6.3) 
 Analysis of Mechanisms for Early Waste Package/Drip 
Shield Failure, (BSC 2003 [DIRS 164475], Tables 18 
and 20 and Section 6.4.8) 
Sections 
6.3, 6.4, 
and 6.5 
 WAPDEG Analysis of Waste Package and Drip Shield 
Degradation (BSC 2003 [DIRS 161317], Tables 46 
and 47) 
February 2004 29 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
4.1.3.1 
10 CFR 63.114 
(d) 
4.1.3.2 
Table 4.1-10. Magma Composition Weight Percents 
Mineral Wt. % 
SiO2 48.50 
TiO2 1.93 
Al2O3 16.74 
Fe2O3 1.74 
FeO 8.90 
MnO 0.17 
MgO 5.83 
CaO 8.60 
Na2O 3.53 
K2O 1.84 
P2O5 1.22 
Source: BSC 2003 [DIRS 166407], Table 6 
4.1.3 Technical Information 
The following technical information was utilized in the development of this analysis and is 
justified, since it comes from approved sources and its application is compatible with its 
developed purpose and limitations. 
Regulatory Probability Criterion 
The following criterion has been utilized in the screening analysis as the basis for excluding 
criticality FEPs from TSPA-LA evaluations on low probability. This criterion is also the basis 
for some of the project performance assessment criteria listed in Table 4.2-1. The regulatory 
probability criterion is cited from 10 CFR Part 63 [DIRS 156605]. 
Requirements for Performance Assessment 
Any performance assessment used to demonstrate compliance with 63.113 must: 
Consider only events that have at least one chance in 10,000 of occurring 
over 10,000 years. 
The regulatory probability criterion is used for the criticality FEPs screening decisions in 
Section 6.8. 
Technical Information Used to Determine Boron Loss in the Commercial SNF 
Waste Packages 
Except for the information obtained from DAgostino and Stephens (1986 [DIRS 160320]), the 
technical information listed in Table 4.1-11 is required to calculate the amount of boron in the 
waste packages containing Neutronit plates. The sources for the required parameters are noted in 
Table 4.1-11. This information is utilized in Section 6.4.1.1.1, Tables 6.4-3 through 6.4-6, and 
Attachment V. 
30 of 176 ANL-EBS-NU-000008 REV 00 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The information obtained from DfAgostino and Stephens (1986 [DIRS 160320]) is utilized in 
Attachment V for the performance of a statistical test to determine the acceptability of the 
Weibull distributionfs fit of the Neutronit corrosion information. 
Source: Weast 1985 [DIRS 111561], pages B-68 through B-161 
Table 4.1-11. Technical Information Used to Calculate Amount of Boron Still Remaining Inside the 
Commercial SNF Waste Packages 
Source) 
Variable(s) Description 
4.1.3.3 
4.1.3.4 
316 
Stainless Steel Type 
316N Grade density 
AW Atomic Weight of 
boron 
bwf 
Boron content in 
Neutronit 
n and  5% significance level 
 for n*D Case 3 
Value 
ASM 1980 [DIRS 104317], p. 34, Table 12 
Parrington et al. 1996 [DIRS 103896], p. 63 
ASTM A 887-89 [DIRS 154062], Table 1, Type 
304B3 
ASTM A 887-89 [DIRS 154062], Table 1, Type 
304B4 
8.00 g/cm3 
10.811 g/mol 
0.75 - 0.99 wt% 
1.00 - 1.24 wt% 
1.50 - 1.74 wt% ASTM A 887-89 [DIRS 154062], Table 1, Type 
304B6 
DfAgostino and Stephens 1986 [DIRS 160320], 
Table 4.18, p. 148 
10 0.819 
20 0.843 
50 0.856 
 0.874 
External Criticality Information 
The uranium loading per fuel assembly used in the external criticality evaluations is 0.469 MTU 
(Punatar 2001 [DIRS 155635], Table 3-1). This value is appropriate as it is representative of a 
commercial SNF assemblyfs uranium loading. This information is used in Section 6.3.2. 
Magma Composition Information 
The theoretical density of the minerals comprising the magma predicted to enter the MGR drifts 
is given in Table 4.1-12. Atomic information for the elemental constituents of the magma 
composition is presented in Table 4.1-13. Sources for water and SNF information used in the 
igneous MCNP calculations are provided in Table 4.1-14. The MCNP calculations support 
Section 6.6 and the calculation inputs and outputs are contained in Attachment VII. 
February 2004 
Table 4.1-12. Magma Composition Theoretical Densities 
Mineral 
ANL-EBS-NU-000008 REV 00 
Theoretical Density (g/cc) b 
SiO2 2.32 
TiO2 3.84 
Al2O3 3.965 
Fe2O3 5.24 
FeO 5.70 
MnO 5.46 
MgO 3.58 
CaO 3.38 
Na2O 2.27 
K2O 2.32 
P2O5 2.39 
31 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 4.1-13. Elemental Composition of Magma with 0.5 Weight Percent Water 
Element MCNP ID b Molar Mass 
(g/mol) a 
H 1.01 1001.50C 
O 16.00 8016.50C 
Si 28.09 14000.50C 
Al 26.98 13027.50C 
Fe 55.85 26000.55C 
Mg 24.31 12000.50C 
Ca 40.08 20000.50C 
Na 22.99 11023.50C 
K 39.10 19000.50C 
Ti 47.87 22000.50C 
P 30.97 15031.50C 
Mn 54.94 25055.50C 
Total NA NA 
Sources: a Parrington et al. 1996 [DIRS 103896] 
b Briesmeister 1997 [DIRS 103897], Appendix G 
Table 4.1-14. MCNP Input Data for Criticality FEPs Igneous Evaluations 
Source of Value(s) Units Value(s) Parameter Description 
Fuel pellet radius cm 0.47 DOE 1987 [DIRS 132333], p. 2A-34 
Fuel pellet length cm 1.1 DOE 1987 [DIRS 132333], p. 2A-34 
Molar mass of U-235 
g/mol Parrington et al. 1996 [DIRS 103896] 
235.043922 
238.050785 
15.9949146 
U-238 
O-16 
1.0 g/cc Batchelor 1967 [DIRS 103289], p. 596 Water theoretical 
density 
4.2 CRITERIA 
This section lists the criteria and requirements addressed by this analysis report. Table 4.2-1 lists 
the applicable project requirements from Project Requirements Document (Canori and Leitner 
2003 [DIRS 166275]). Tables 4.2-2 and 4.2-3 list the applicable Yucca Mountain Review Plan 
(NRC 2003 [DIRS 163274]) acceptance criteria. Section 7.4 presents how these criteria and 
requirements have been addressed in this analysis. 
ANL-EBS-NU-000008 REV 00 February 2004 32 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 4.2-1. Applicable Project Requirements 
Requirement Text 
Requirement 
Number and 
Title 
PRD-002/T-015a; For complete requirement text, 
Requirements for see 10 CFR 63.114 [DIRS 
Performance 156605] 
Assessment 
PRD-002/T-034b; For complete requirement text, 
Limits on see 10 CFR 63.342 [DIRS 
Performance 156605] 
Assessments 
PRD-013/T-016c; The methodology defined in the 
DOE SNF 
Canister 
Criticality 
Potential 
Postclosure 
Disposal Criticality Analysis 
Methodology Topical Report (YMP 
2003 [DIRS 165505]) shall be 
used to demonstrate acceptable 
criticality control for canisters and 
the waste packages in which they 
are disposed. 
PRD-013/T-023d; The methodology defined in the 
Naval SNF 
Canister 
Criticality 
Potential 
Postclosure 
NNPP addendum (Mowbray 1999 
[DIRS 149585]) to the Disposal 
Criticality Analysis Methodology 
Topical Report (YMP 2003 [DIRS 
165505]) shall be used to 
demonstrate acceptable criticality 
control for canisters and the waste 
packages in which they are 
disposed. 
The methodology defined in the 
Disposal Criticality Analysis 
Methodology Topical Report (YMP 
2003 [DIRS 165505]) shall be 
used to demonstrate acceptable 
criticality control for canisters and 
the waste packages in which they 
are disposed. 
PRD-013/T- 
038e;Disposable 
Commercial- 
Origin DOE SNF 
Canister 
Criticality 
Potential 
Postclosure 
Source: Canori and Leitner 2003 [DIRS 166275] 
NOTES: a Requirement basis is 10 CFR 63.114 and 63.113 [DIRS 156605] & YMP-RD 3.3.4.19 (YMP 2001). 
b Requirement basis is 10 CFR 63.342 [DIRS 156605] (40 CFR 197.36 [DIRS 155238]) 
c Requirement basis is WASRD 4.3.12.B (DOE 2002 [DIRS 158873]) 
d Requirement basis is WASRD 4.4.13.B (DOE 2002 [DIRS 158873]) 
e Requirement basis is WASRD 4.5.13.B (DOE 2002 [DIRS 158873]) 
ANL-EBS-NU-000008 REV 00 
Rationale for Requirement 
Regulation 10 CFR 63.114 [DIRS 156605] specifies technical 
requirements to be used in a performance assessment to 
demonstrate compliance to 10 CFR 63.113. It includes 
requirements for calculations, including data related to site 
geology, hydrology, and geochemistry; the need to account for 
uncertainties and variabilities in model parameters; the need to 
consider alternative conceptual models; and technical bases for 
inclusion or exclusion of specific features, events, and 
processes (FEPs); deterioration or degradation processes of 
engineered barriers; and all the models used in the 
performance assessment. The Performance Assessment 
organization is responsible for developing and using TSPA 
calculations, methods, models, and processes that comply with 
the requirements of this section. 
This section states that the license applicant's performance 
assessments should not include very unlikely FEPs, defined as 
those that are estimated to have less than one chance in 
10,000 of occurring within 10,000 years of disposal. 
Furthermore, this section states that the performance 
assessments need not evaluate the impacts of sequences of 
FEPs with a higher chance of occurrence if the results of the 
earlier performance assessments would not be changed 
significantly. The Performance Assessment organization is 
responsible for incorporating these limits on performance 
assessments into its analytical models, methods, and activities. 
This requirement specifies the method by which acceptable 
criticality control is demonstrated for the canisters and the 
waste packages for postclosure. 
The methodology in the NNPP addendum demonstrates the 
method by which acceptable postclosure criticality control is 
demonstrated for the waste packages with NNPP canisters. 
NNPP is directly responsible for completing the postclosure inpackage 
criticality analysis for naval SNF waste packages and 
supplying the results to DOE. NNPP will also provide the 
results of the fissile material loss from waste packages source 
term calculations to the DOE for any out-of-package criticality 
analyses that may ne needed. 
The methodology in the Topical Report demonstrates the 
method by which acceptable postclosure criticality control is 
demonstrated for canisters and waste packages in a repository. 
February 2004 33 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 4.2-2. Acceptance Criteria for Scenario Identification and Screening 
Acceptance Criteria Name 
Acceptance Criterion 1: 
The Identification of a List of 
Features, Events, and Processes 
Is Adequate 
Acceptance Criterion 2: 
Screening of the List of Features, 
Events, and Processes Is 
Appropriate 
Source: NRC 2003 [DIRS 163274], Section 2.2.1.2.1.3 
ANL-EBS-NU-000008 REV 00 
Description 
(1)The Safety Analysis Report contains a complete list of features, events 
and processes, related to the geologic setting or the degradation, 
deterioration, or alteration of engineered barriers (including those processes 
that would affect the performance of natural barriers) that have the potential 
to influence repository performance. The list is consistent with the site 
characterization data. Moreover, the comprehensive features, events, and 
processes list includes, but is not limited to, potentially disruptive events 
related to igneous activity (extrusive and intrusive); seismic shaking (highfrequency-
low-magnitude, and rare large-magnitude events); tectonic 
evolution (slip on existing faults and formation of new faults); climatic 
change (change to pluvial conditions); and criticality. 
(1) The U.S. Department of Energy has identified all features, events, and 
processes related to either the geologic setting or to the degradation, 
deterioration, or alteration of engineered barriers (including those 
processes that would affect the performance of natural barriers) that 
have been excluded; 
(2) The U.S. Department of Energy has provided justification for those 
features, events, and processes that have been excluded. An 
acceptable justification for excluding features, events, and processes is 
that either the feature, event, and process is specifically excluded by 
regulation; probability of the feature, event, and process (generally an 
event) falls below the regulatory criterion; or omission of the feature, 
event, and process does not significantly change the magnitude and 
time of the resulting radiological exposures to the reasonably maximally 
exposed individual, or radionuclide releases to the accessible 
environment; and 
(3) The U.S. Department of Energy has provided an adequate technical 
basis for each feature, event, and process, excluded from the 
performance assessment, to support the conclusion that either the 
feature, event, or process is specifically excluded by regulation; the 
probability of the feature, event, and process falls below the regulatory 
criterion; or omission of the feature, event, and process does not 
significantly change the magnitude and time of the resulting radiological 
exposures to the reasonably maximally exposed individual, or 
radionuclide releases to the accessible environment. 
February 2004 34 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 4.2-3. Acceptance Criteria for Uncertainty in Event Probability 
Acceptance Criterion Number 
and Title 
Acceptance Criterion 1: 
Events Are Adequately Defined 
Acceptance Criterion 2: 
Probability Estimates for Future 
Events Are Supported by 
Appropriate Technical Bases 
Acceptance Criterion 3: 
Probability Model Support Is 
Adequate 
Acceptance Criterion 4: 
Probability Model Parameters 
Have Been Adequately 
Established 
Acceptance Criterion 5: 
Uncertainty in Event Probability 
Is Adequately Evaluated 
Source: NRC 2003 [DIRS 163274], Section 2.2.1.2.2.3 
ANL-EBS-NU-000008 REV 00 
Description 
(1) Events or event classes are defined without ambiguity and used 
consistently in probability models, such that probabilities for each event 
or event class are estimated separately; and 
(2) Probabilities of intrusive and extrusive igneous events are calculated 
separately. Definitions of faulting and earthquakes are derived from the 
historical record, paleoseismic studies, or geological analyses. 
Criticality events are calculated separately by location. 
(1) Probabilities for future natural events have considered past patterns of 
the natural events in the Yucca Mountain region, considering the likely 
future conditions and interactions of the natural and engineered 
repository system. These probability estimates have specifically 
included igneous events, faulting and seismic events, and criticality 
events. 
(1) Probability models are justified through comparison with output from 
detailed process level models and/or empirical observations (e.g., 
laboratory testing, field measurements, or natural analogs, including 
Yucca Mountain site data). Specifically: 
(a) For infrequent events, the U.S. Department of Energy justifies, to 
the extent appropriate, proposed probability models with data from 
reasonably analogous systems. Analog systems should contain 
significantly more events than the Yucca Mountain system, to 
provide reasonable evaluations of probability model performance; 
(b) The U.S. Department of Energy justifies, to the extent appropriate, 
the ability of probability models to produce results consistent with 
the timing and characteristics (e.g., location and magnitude) of 
successive past events in the Yucca Mountain system; and 
(c) The U.S. Department of Energy probability models for natural 
events use underlying geologic bases (e.g., tectonic models) that 
are consistent with other relevant features, events, and processes 
evaluated, using Section 2.2.1.2.1. 
(1) Parameters used in probability models are technically justified and 
documented by the U.S. Department of Energy. Specifically: 
(a) Parameters for probability models are constrained by data from the 
Yucca Mountain region and engineered repository system to the 
extent practical; 
(b) The U.S. Department of Energy appropriately establishes 
reasonable and consistent correlations between parameters; and 
(c) Where sufficient data do not exist, the definition of parameter values 
and conceptual models is based on appropriate use of other 
sources, such as expert elicitation conducted in accordance with 
appropriate guidance. 
(1) Probability values appropriately reflect uncertainties. Specifically: 
(a) The U.S. Department of Energy provides a technical basis for 
probability values used, and the values account for the uncertainty 
in the probability estimates; and 
(b) The uncertainty for reported probability values adequately reflects 
the influence of parameter uncertainty on the range of model results 
(i.e., precision) and the model uncertainty, as it affects the timing 
and magnitude of past events (i.e., accuracy). 
February 2004 35 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
4.3 CODES AND STANDARDS 
The following codes have been cited in this analysis: 
 40 CFR 197 [DIRS 155238]. 2001. Protection of Environment: Public Health and 
Environmental Radiation Protection Standards for Yucca Mountain, Nevada. 
 10 CFR 63 [DIRS 156605]. Energy: Disposal of High-Level Radioactive Wastes in a 
Geologic Repository at Yucca Mountain, Nevada. 
The following standards are applicable to criticality FEPs screening evaluations for the 
repository: 
 ASM 1980 [DIRS 104317]. Properties and Selection: Stainless Steels, Tool 
Materials and Special-Purpose Metals. Volume 3 of Metals Handbook. 
 ASTM A 887-89 [DIRS 154062] (Reapproved 2000) 2000. Standard Specification 
for Borated Stainless Steel Plate, Sheet, and Strip for Nuclear Application. 
February 2004 36 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
5. ASSUMPTIONS 
5.1 GENERAL CRITICALITY FEPS ANALYSIS ASSUMPTIONS 
5.1.1 No Waste Package Corrosion Failures 
Assumption: It is assumed that the TSPA-LA results will show that there are no corrosion 
failures of the waste package before 10,000 years. 
Rationale: Based on results from WAPDEG Analysis of Waste Package and Drip Shield 
Degradation (CRWMS M&O 2000 [DIRS 151566], Section 6.5.1 and BSC 2003 [DIRS 
161317], Section 6.6.2, Figures 36, 37 and 48), the earliest failure due to waste package 
corrosion mechanisms (general corrosion, stress corrosion cracking, and localized corrosion) was 
beyond the 10,000-year performance period. However, these results cannot be referenced as 
they require confirmation by TSPA-LA (BSC 2003 [DIRS 161317], Section 1). If the 
environmental conditions of the TSPA-LA analysis do not exclude corrosion of the waste 
package, the probability evaluations of the criticality FEPs must be updated to account for the 
increased waste package damage area resulting from these failure mechanisms. 
Confirmation Status: Confirmation of this assumption will be required when the waste package 
corrosion failure results from the TSPA-LA are available. 
Use in the Analysis: This assumption is used in Sections 6.3.3 and 7.3.2. 
5.1.2 Boron Loss in Commercial SNF Waste Packages 
Assumption: It is assumed that the updated commercial SNF waste package design parameters 
will result in similar boron loss probability results as the previous design parameters. 
Rationale: Repository Design Project, RDP/PA IED Typical Waste Package Components 
Assembly (2) (BSC 2003 [DIRS 163855]) has been superceded by D&E/PA/C IED Typical 
Waste Package Components Assembly (BSC 2004 [DIRS 167207]). However, D&E/PA/C IED 
Typical Waste Package Components Assembly (BSC 2004 [DIRS 167207]) does not contain the 
necessary information required to update the product output as utilized from 
DTN: MO0210MWDEXC01.008 ([DIRS 163531]), which is generated by Boron Loss from 
CSNF Waste Packages (BSC 2003 [DIRS 165890]). Revisions of the information obtained from 
Boron Loss from CSNF Waste Packages (BSC 2003 [DIRS 165890]) are necessary based on the 
updated information that is provided in D&E/PA/C IED Typical Waste Package Components 
Assembly (BSC 2004 [DIRS 167207]). Once D&E/PA/C IED Typical Waste Package 
Components Assembly (BSC 2004 [DIRS 167207]) is revised to incorporate the required 
information, a future revision of this analysis will reflect this new information. 
The parameters listed in Table 5.1-1 reflect the previous waste package design and are used in 
this analysis to determine the amount of boron remaining in the waste packages containing 
Neutronit plates during degradation. The sources for the required parameters are noted in 
Table 5.1-1. 
February 2004 37 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 5.1-1. Parameters Used to Calculate Amount of Boron Still Remaining Inside the Commercial SNF 
Waste Packages 
SA 
Vr 
Fuel Basket A-Plate 8 plates per waste package 
Fuel Basket B-Plate 8 plates per waste package 
Fuel Basket C-Plate 16 plates per waste package 44,000 g 
SA 
Vr 
Fuel Basket A-Plate 4 plates per waste package 
Fuel Basket B-Plate 4 plates per waste package 
Fuel Basket C-Plate 16 plates per waste package 34,000 g 
SA 
Vr 
Fuel Basket A-Plate 4 plates per waste package 
Fuel Basket B-Plate 4 plates per waste package 
Fuel Basket C-Plate 16 plates per waste package 15,000 g 
Fuel Basket D-Plate 16 plates per waste package 44,000 g 
Fuel Basket E-Plate 16 plates per waste package 44,000 g 
SA 
Vr 
Fuel Basket A-Plate 4 plates per waste package 
Fuel Basket B-Plate 4 plates per waste package 
Fuel Basket C-Plate 8 plates per waste package 
Fuel Basket D-Plate 8 plates per waste package 
Fuel Basket E-Plate 16 plates per waste package 30,000 g 
ANL-EBS-NU-000008 REV 00 
Variable(s) Description Value 
21-PWR Absorber Plate Waste Package (Parameters used in Section 6.4.1.1.1, Table 6.4-3) 
Surface area of Neutronit 
Plates 
Void Volume of waste 
package 
12-PWR Absorber Plate Waste Package (Parameters used in Section 6.4.1.1.1, Table 6.4-4) 
Surface area of Neutronit 
Plates 
Void Volume of waste 
package 
44-BWR Absorber Plate Waste Package (Parameters used in Section 6.4.1.1.1, Table 6.4-5) 
Surface area of Neutronit 
Plates 
Void Volume of waste 
package 
24-BWR Absorber Plate Waste Package (Parameters used in Section 6.4.1.1.1, Table 6.4-6) 
Surface area of Neutronit 
Plates 
Void Volume of waste 
package 
DTN: MO0210MWDEXC01.008 [DIRS 163531], 
5.29E+05 cm2 spreadsheet CSNF WP Model Abstraction.xls, 
4,685 L 
85,000 g 
85,000 g 
DTN: MO0309SPABRNAM.001 [DIRS 
3.19E+05 cm2 165892], spreadsheet 12 PWR Long WP.xls, 
3,280 L 
76,000 g 
76,000 g 
DTN: MO0309SPABRNAM.001 
9.55E+05 cm2 [DIRS 165892], spreadsheet 44 BWR WP.xls, 
4,850 L 
63,000 g 
63,000 g 
DTN: MO0309SPABRNAM.001 
6.93E+05 cm2 [DIRS 165892], spreadsheet 24 BWR WP.xls, 
2,700 L 
89,000 g 
89,000 g 
90,000 g 
90,000 g 
38 of 176 
Source 
sheet Volumes and Surface Areas 
DTN: MO0210MWDEXC01.008 [DIRS 163531], 
spreadsheet CSNF WP Model Abstraction.xls, 
sheet Volumes and Surface Areas, referred to 
as normalization factor and liters of void 
volume 
BSC 2003 [DIRS 163855], Table 2 
sheet Volumes and Surface Areas 
DTN: MO0309SPABRNAM.001 
[DIRS 165892], spreadsheet 12 PWR Long 
WP.xls, sheet Volumes and Surface Areas 
BSC 2003 [DIRS 163855], Table 8 
sheet Volumes and Surface Areas 
DTN: MO0309SPABRNAM.001 
[DIRS 165892], spreadsheet 44 BWR WP.xls, 
sheet Volumes and Surface Areas 
BSC 2003 [DIRS 163855], Table 3 
sheet Volumes and Surface Areas 
DTN: MO0309SPABRNAM.001 
[DIRS 165892], spreadsheet 24 BWR WP.xls, 
sheet Volumes and Surface Areas 
BSC 2003 [DIRS 163855], Table 9 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Confirmation Status: Confirmation of this assumption will be required once D&E/PA/C IED 
Typical Waste Package Components Assembly (BSC 2004 [DIRS 167207]) is revised to 
incorporate the required waste package design information. 
Use in the Analysis: This information is used in Tables 6.4-3 through 6.4-6 of Section 6.4.1.1.1 
and Attachment III. 
5.1.3 Waste Packages Emplaced As Designed 
Assumption: It is assumed that the initial neutron absorber material mass and location and the 
waste package internal structural capabilities are as-designed. 
Rationale: This assumption is necessary to account for the possibility that, through human error 
either during manufacturing or waste package loading, the waste package internals placed in the 
waste package may not be as-designed. An example of these human errors include selecting a 
21-PWR with Control Rods Waste Package for the insertion of commercial SNF intended to be 
placed into a 21-PWR with Absorber Plates Waste Package. This would result in the waste 
package basket containing no neutron absorber material. Although the possibility exists that the 
waste package will not be emplaced as designed, it is not possible at this time to assess the 
probability of its occurrence. 
Confirmation Status: Removal of this assumption cannot occur until the probability that a waste 
package will be emplaced in a not as-designed condition is calculated. Performance of this 
calculation cannot occur until the manufacturing processes and surface facility procedures for 
selecting, loading, closing and emplacing the waste packages are established. 
Use in the Analysis: This assumption is used in Sections 6.3.3 and 7.3.3. 
5.1.4 Corrosion Rate of Neutronit A978 Neutron Absorber Material 
Assumption: It is assumed that the corrosion rate of the Neutronit A978 neutron absorber 
material is not in excess of 1.5 times that of Stainless Steel Type 316. 
Rationale: Although no numerical corrosion rates are presented in the literature available from 
the manufacturer of Neutronit A978 (Kgler 1997 [DIRS 134327]), the information that is 
available does state that Neutronit A976 has a corrosion resistance similar to Stainless Steel Type 
321. However, the repository currently plans to utilize Neutronit A978 for the manufacturer of 
neutron absorber baskets in most commercial SNF waste packages (only the 21-PWR with 
Control Rods Waste Package does not contain Neutronit in its basket assembly design). 
Neutronit A978 is similar to a stainless steel type that contains molybdenum. Stainless steel 
alloys containing molybdenum, such as Stainless Steel Type 316, have a higher corrosion 
resistance than stainless steel types that do not contain molybdenum, such as Stainless Steel 
Type 321. It is reasonable to expect, therefore, that Neutronit A978 would have a corrosion rate 
similar to that of Stainless Steel Type 316. For uncertainty considerations, the stainless steel 316 
corrosion rate is increased by 50 percent for use in Neutronit A978 degradation evaluations. 
Confirmation Status: This assumption requires confirmation through testing. 
February 2004 39 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Use in the Analysis: This assumption is used in Section 6.4.1, Section 7.3.4, and Attachment III. 
5.1.5 Boron Loss from Waste Packages with Absorber Plates 
Assumption: It is assumed that up to 90 percent of the boron can be removed from a degraded 
waste package with absorber plates containing typical PWR fuel assemblies without a criticality 
concern. It is further assumed that up to 50 percent of the boron can be removed from a 
degraded waste package with absorber plates containing typical BWR fuel assemblies without a 
criticality concern. 
Rationale: Since the corrosion rate of the basket component materials not carrying boron (i.e., 
carbon steel) are greater than the material carrying boron (i.e., stainless steel), corrosion products 
will accumulate faster than boron is lost. These corrosion products will displace water (a neutron 
moderator) and reduce the criticality potential of the degraded waste package. The percentages 
are estimates based on preliminary scoping analyses of when the boron loss effect will exceed 
the corrosion product build-up effect. The difference in allowable boron loss rates between the 
PWR and BWR waste packages is due to the BWR SNF assemblies being disposed of with their 
channels. These channels inhibit the accumulation of corrosion products within the assembly 
matrix. 
Confirmation Status: This assumption requires confirmation by analysis. 
Use in the Analysis: This assumption is used in Section 6.4.1, Section 7.3.4, and Attachment III. 
5.1.6 Degradation of Neutron Absorber Material in the 21-PWR with Control Rods 
Waste Package 
Assumption: It is assumed that the neutron absorber material in the 21-PWR with Control Rods 
Waste Package will not degrade during the performance period. 
Rationale: The 21-PWR with Control Rods Waste Package is designed for PWR commercial 
SNF having a higher assembly k than is acceptable for placement in the 21-PWR with Absorber 
Plates Waste Package. The 21-PWR with Control Rods Waste Package uses zirconium clad, 
boron carbide (B4C) control rods for reactivity control (CRWMS M&O 1997 [DIRS 100224], 
Section 7.3.2). These control rods are inserted into each assembly guide tube location. The 
zirconium cladding of the control rods is the same as the Zircaloy used for the manufacture of 
fuel rod cladding. Under normal conditions, Zircaloy-clad fuel rods will be intact beyond the 
performance period because Zircaloy cladding is highly resistant to corrosion (Hillner et al. 1998 
[DIRS 100455], Abstract). Because the zirconium cladding of the control rods will be 
unirradiated and will be thicker than the fuel rod cladding, its durability and corrosion resistance 
is expected to be even greater than that of the Zircaloy cladding of the fuel rods. In addition, 
because the zirconium control rod cladding is thicker and because the controls are protected by 
the fuel assembly guide tubes, it is unlikely the control rod cladding will be damaged during 
seismic events. Therefore, it is assumed that the neutron absorber materials of the 21-PWR with 
Control Rods Waste Package cannot be flushed from the waste package during the performance 
period. 
Confirmation Status: This assumption requires confirmation by analysis. 
February 2004 40 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Use in the Analysis: This assumption is used in Sections 6.3.3, 6.4.1, 7.3.1, and 7.3.4. 
5.1.7 Degradation of Neutron Absorber Material in the DOE SNF Waste Packages 
Assumption: It is assumed that the loss rate of the neutron absorber material from the DOE 
standardized SNF canisters contained in the DOE SNF waste package types is no greater than 
that of the 21-PWR with Absorber Plates Waste Package. 
Rationale: The current design of the DOE-standardized SNF canisters (e.g., CRWMS M&O 
2000 [DIRS 151742] or CRWMS M&O 2000 [DIRS 147650]) is for the inclusion of neutron 
absorber materials contained in a material (Alloy 22) with long term performance characteristics 
(i.e., low corrosion rate) that are greater than that of the neutron absorber materials contained in 
the material (Stainless Steel Type 316) of the 21-PWR with Absorber Plates Waste Package 
(DTN: MO0401SPAMCRAE.000 [DIRS 166801]). 
Confirmation Status: This assumption requires confirmation by analysis. 
Use in the Analysis: This assumption is used in Sections 6.3.3, 7.3.1 and 7.3.4. 
5.2 SEISMIC CRITICALITY FEPS ANALYSIS ASSUMPTIONS 
5.2.1 Seismic Damage to Drip Shield and Waste Package 
Assumption: It is assumed that all seismically induced damage is located on the top of the drip 
shield and waste package. 
Rationale: Damage at the top of the drip shield and waste package allows advective flow to 
penetrate the waste package and create a bathtub configuration. Bathtub configurations are the 
most critical because commercial SNF assemblies are in a core-like geometry and, with no 
neutron absorber materials present, have near-ultimate neutron moderation. This assumption is 
conservative. 
Confirmation Status: This assumption does not require further confirmation by testing, design, 
or analysis. 
Use in the Analysis: This assumption is used in Sections 6.3.3 and 6.4.1. 
5.2.2 Time of Water Ingress into a Damaged Waste Package 
Assumption: It is assumed that water can not penetrate a damaged commercial SNF waste 
package prior to 700 years after closure of the repository. 
Rationale: This assumption is used in order to determine the start of a time period for advective 
flow into a damaged commercial SNF waste package to degrade and flush out the Neutronit. 
Advective flow into a damaged waste package is assumed improbable prior to 700 years after 
repository closure due to the dryout of the drift from the thermal pulse (BSC 2003 [DIRS 
166463], Table 6.3-5). 
February 2004 41 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Confirmation Status: This assumption does not require further confirmation by testing, design, 
or analysis. 
Use in the Analysis: This assumption is used in Section 6.3.2, Section 6.4.1 and Attachment III. 
5.2.3 Major Assumptions in Supporting Calculations 
The following are major assumptions that are listed in the supporting documents. A synopsis of 
the assumptions is listed here for brevity. For more detailed information, refer to the referenced 
document. 
5.2.3.1 Supporting Assumptions used by Seismic Consequence Abstraction (BSC 2003 
[DIRS 161812], Section 5) 
The following assumptions from Seismic Consequence Abstraction (BSC 2003 [DIRS 161812], 
Section 5) are important to the development of the damaged area(s) of the drip shield and waste 
package along with calculations used in the seismic probability calculation. 
1) 
2) 
3) 
4) 
ANL-EBS-NU-000008 REV 00 
The affected area(s) where the residual stress from mechanical damage exceeds the 
residual stress of the barrier is assumed to have failed as a barrier to flow and transport. 
The rationale is based on using this residual stress threshold as the failure criterion. This 
is a nonmechanistic criterion because detailed calculations of the actual rates of general 
corrosion, stress corrosion cracking, or localized corrosion are not being used to 
determine the actual failure time after a seismic event. Rather, it is acknowledged the 
potential exists for one or several of these processes to occur with such rapidity that the 
entire damaged area ceases to function as an effective barrier to flow and transport. 
Seismic events occur in a random manner, following a Poisson process. The rationale for 
this assumption (the behavior of the earth is generally random [Poisson process]) is the 
underlying assumption in all probabilistic hazard analyses. In other words, all 
earthquakes are considered as independent events with regard to magnitude, time, and 
location. Although there may be cases where sufficient data and information exists to 
depart from this assumption, the Poisson process is generally an effective representation 
of nature and represents a compromise between the complexity of natural processes, 
availability of information, and sensitivity of results of engineering relevance. 
No damage occurs to EBS components until the repository experiences ground motions 
larger than those for the 10-4 per year annual exceedance frequency. Damage to EBS 
components from vibratory ground motion is assumed to become nonzero between 10-4 
and 10-5 per year annual exceedance frequencies. The rationale for this assumption is 
based on structural analyses performed at vibratory ground motions for an annual 
exceedance frequency of 5  10-4 per year. These results showed no residual stresses 
greater than the residual stress of the barriers. The confirmation of this assumption has 
an associated TBV (TBV-5106) (BSC 2003 [DIRS 161812], Section 5.1). 
The fault displacement hazard curves for the Pagany Wash and Sevier Wash faults are 
identical to the fault displacement hazard curve for the Drill Hole Wash fault. The 
February 2004 42 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
rationale is based on: (1) the Drill Hole Wash fault provides the best field data for the 
three faults; (2) none of the faults suggest displacement in Quaternary alluvial terraces 
(the scale of the cumulative vertical displacement is less that 5 to 10 meters); (3) the total 
fault length is similar for the three faults; and (4) previous geologic studies have 
consolidated the three faults based on similar characteristics. Therefore, the three faults 
can be treated in a similar manner with regard to the potential seismic hazard. 
5) The derivation of the mean dose formula is assumed to be a function of the time of 
occurrence and the amplitude of the PGV for the seismic hazard. The dose time histories 
for the reasonably maximally exposed individual at time () from a seismic event 
occurring at a time (t) prior to  depends only on the time of occurrence of the event and 
on PGV at the waste emplacement drifts associated with the seismic event. The rationale 
for this assumption is based on the simplification of the mathematical equation, which 
derives the mean dose calculation. 
6) These supporting calculations include several major assumptions not directly used in the 
abstraction process, but noteworthy enough to deserve repeating here. 
(a) The structural response calculations for the waste package and drip shield incorporate 
assumptions for structural thickness and for material properties for Alloy 22 and of 
Titanium Grade 7. The thicknesses of the drip shield plates and the waste package 
outer shell have been reduced by 2 mm in these calculations to represent the potential 
degradation of these structures by general corrosion over the first 10,000 years after 
repository closure. The material properties of Alloy 22 and of Titanium Grade 7 
have been evaluated at an elevated temperature (150oC) that provides conservative 
values for mechanical properties over most (97 percent) of the 10,000-year duration. 
(b) Rockfall calculations for the lithophysal and nonlithophysal zones also make several 
key assumptions. In the lithophysal zone, block size distribution is assumed to be a 
function of the inter-lithophysal fracture density and the lithophysae spacing. This 
assumption is relevant to the abstraction process because it limits the potential 
damage to the drip shield from tunnel collapse in the lithophysal zone, as discussed in 
Section 6.6.2 of Seismic Consequence Abstraction (BSC 2003 [DIRS 161812]). 
5.2.3.2 Supporting Assumptions used by Abstraction of Drift Seepage (BSC 2003 [DIRS 
165564], Section 5) 
There are two primary assumptions listed in Abstraction of Drift Seepage (BSC 2003 [DIRS 
165564], Section 5). These assumptions are important to the development of the seepage 
parameters used in the seepage development. 
1) The first assumption discusses the capillary diversion depending upon the difference in 
capillary strength (1/) between the interior of the drift and the rock surrounding the 
drift. This assumption assumes capillary strength of the rubble rock material for a 
collapsed drift to be 100 Pa. The rationale for this assumption is based on the porosity of 
the rubble rock material being much greater than that of intact rock, because of the large 
voids between chunks of fragmented rock. The resulting capillary strength of the rubble- 
February 2004 43 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
filled drift is therefore, much weaker than that of the intact surrounding rock. The value 
of 100 Pa is therefore chosen as a conservative, nonzero value to represent the effective 
capillary strength of the rubble-filled drift with an air gap forming at the ceiling. 
2) The second assumption discusses thermal-hydrological simulations for collapsed drifts 
requiring knowledge about the thermal hydraulic properties of the rubble rock material 
filling the drift. The thermal conductivity is set to that of air; the heat capacity is set to 
zero. The interface area between void continuum and the fragmented rock continuum, 
important for the fluid and heat exchange between the two media, is estimated from a 
simple geometry model, calculating the surface area of spherical rock blocks with a 
0.1-m diameter. The contact area for flow and heat transport between individual rock 
fragments was assumed to follow two alternative cases. The two cases are a geometric 
interface area between grid elements reduced by a factor of (1-0.231) where 0.231 refers 
to the volume fraction of the voids in the rubble material and one-half of this value. 
Its rationale states that the thermal hydraulic properties of the lithophysal matrix rock are 
appropriate for the fragmented rock blocks, because they are formed from chunks of 
lithophysal matrix rock that fell into the drift. In addition, a sensitivity analysis was 
conducted with a one order of magnitude variation to permeability. That analysis 
demonstrated that the general conclusions for seepage abstraction are not affected by this 
parameter variation. The two cases showed similar results. 
5.3 ROCKFALL CRITICALITY FEPS ANALYSIS ASSUMPTIONS 
No additional assumptions are required to evaluate the rockfall disruptive event criticality FEP. 
5.4 IGNEOUS CRITICALITY FEPS ANALYSIS ASSUMPTIONS 
5.4.1 System for Igneous Event Commercial SNF Criticality Analyses 
Assumption: It is assumed that the system modeled for an igneous event criticality is infinite. 
The waste form utilized is 5 weight percent enriched uranium-235 commercial SNF pellets and is 
surrounded by a cubic lattice of magma, which serves as the neutron moderator. 
Rationale: This is a conservative approach because (1) an infinite system experiences zero 
neutron leakage; (2) 5 weight percent is the current upper limit for commercial SNF enrichment 
(BSC 2003 [DIRS 165732], Section 6); and (3) the fuel pellet is completely surrounded by a 
neutron moderator. 
Confirmation Status: This assumption does not require confirmation by testing, design, or 
analysis. 
Use in the Analysis: This assumption is used in Section 6.6.2. 
5.4.2 Separation of Neutron Absorber Materials 
Assumption: For commercial SNF, it is assumed that the fissile material becomes separated from 
the neutron absorber material following the destruction of the waste package by igneous 
February 2004 44 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
intrusion. This assumption will be utilized even though there is no identified mechanism during 
an igneous event by which this separation would occur. 
Rationale: This is a conservative approach. The loss of the neutron absorber will increase the 
criticality potential (i.e., the neutron multiplication factor) of the system. 
Confirmation Status: This assumption does not require confirmation by testing, design, or 
analysis. 
Use in the Analysis: This assumption is used in Section 6.6.2. 
5.4.3 System for Igneous Event DOE SNF Criticality Analyses 
Assumption: It is assumed that configurations formed due to igneous events in or out of waste 
packages containing DOE SNF canisters will not result in the formation of critical systems. 
Rationale: Igneous events will not result in critical configurations due to the limited magma 
moderation (less than 0.5 weight percent water), the less than optimum geometrical 
configurations that will be formed, and the mixing of neutron absorbers and fissile materials. 
Confirmation Status: This assumption requires confirmation by analysis. 
Use in the Analysis: This assumption is used in Sections 6.6.3 and 7.3.5. 
5.4.4 Critical Limit for Igneous Event Criticality Analyses 
Assumption: It is assumed that the critical limit for configurations formed due to igneous events 
in or out of commercial SNF waste packages is higher than 0.82. 
Rationale: The lowest calculated keff value for the solution benchmark experiments evaluated in 
Summary Report of Laboratory Critical Experiment Analyses Performed for the Disposal 
Criticality Analysis Methodology (CRWMS M&O 1999 [DIRS 157731], Section 4.2) is higher 
than 0.96. Using appropriate benchmark experiments (mainly solution) for configurations 
formed due to igneous events and the methodology to be developed in the external criticality 
model report is expected to result in the calculation of a critical limit well above 0.82. 
Confirmation Status: This assumption requires confirmation by analysis. 
Use in the Analysis: This assumption is used in Sections 6.6.3 and 7.3.5. 
5.4.5 UO2 Density Used in Igneous Criticality Analyses 
Assumption: It is assumed that the UO2 density for MCNP calculations is 10.41 g/cc, which is 
95 percent of the theoretical density of 10.96 g/cc (DTN: MO9906RIB00048.000 
[DIRS 147618]). 
Rationale: Using a UO2 fractional density is common practice. 
February 2004 45 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Confirmation Status: This assumption does not require further confirmation by testing, design, 
or analysis. 
Use in the Analysis: This assumption is used in Section 6.6.2. 
5.4.6 Magma Water Content 
Assumption: It is assumed that the magma water content is 0.5 weight percent at a temperature 
of 1150C. 
Rationale: Magma water content is presented as an assumption in Igneous Intrusion Impacts on 
Waste Package and Waste Form (BSC 2003 [DIRS 165002], Section 5.1.2), and discussed in 
detail in Characterize Eruptive Processes at Yucca Mountain, Nevada (BSC 2003 [DIRS 
166407], Section 6.3.2.2). 
Confirmation Status: This assumption does not require further confirmation by testing, design, 
or analysis. 
Use in the Analysis: This assumption is used in Section 6.6.2. 
February 2004 46 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
6. SCIENTIFIC ANALYSIS DISCUSSION 
The following sections discuss the criticality FEPs analyses. Section 6.1 discusses the methods 
and approach used for the FEPs process as it applies to the criticality FEPs, as well as changes in 
the criticality FEPs from the TSPA-SR to the TSPA-LA FEPs list. Section 6.1 of this analysis 
report identifies the source of the criticality FEPs, describes the FEPs screening process, and 
provides documentation related to consideration of generic issues such as uncertainty, alternative 
conceptual models, and model and software issues. Section 6.2 discusses the SAPHIRE model 
used to establish the technical basis for the criticality FEPs screening, other than the FEPs related 
to igneous events which is presented in Section 6.6. The SAPHIRE model summarized in 
Section 6.2 was developed specifically for this purpose in Configuration Generator Model for 
In-Package Criticality (BSC 2003 [DIRS 165629]) and is consistent with the TSPA approach to 
satisfy the regulatory probability criterion and performance objectives. Additionally, these 
analyses are also appropriate because they address the NRCs acceptance criteria in Yucca 
Mountain Review Plan (NRC 2003 [DIRS 163274]) as previously discussed in Section 4.2, 
which are applicable to the FEPs discussions provided in Sections 6.3 through 6.8 of this 
analysis report. Section 6.3 provides the details and results of the base case criticality FEPs 
screening analysis. Section 6.4 provides the details and results of the seismic disruptive event 
criticality FEPs screening analysis. Section 6.5 provides the details and results of the rockfall 
disruptive event criticality FEPs screening analysis. Section 6.6 provides the details and results 
of the igneous disruptive event criticality FEPs screening analysis. Section 6.7 summarizes the 
results of Sections 6.3 through 6.6. Section 6.8 provides the screening discussions for the 
criticality FEPs. 
6.1 SCIENTIFIC APPROACH AND TECHNICAL METHODS 
The methods and approach for FEPs screening for TSPA-LA are provided in generic form in The 
Enhanced Plan for Features, Events, and Processes (FEPs) at Yucca Mountain (BSC 2002 
[DIRS 158966]) and KTI Letter Report Response to Additional Information Needs on 
TSPAI 2.05 and TSPAI 2.06 (Freeze 2003 [DIRS 165394]). The YMP has chosen to satisfy the 
regulatory probability criterion and performance objectives by adopting a FEP analysis and 
scenario development process. The first step of the FEP analysis process is the identification of 
the FEPs potentially relevant to the performance of the MGR. A review of FEPs analysis and 
scenario development in other radioactive waste disposal programs is provided in Section 2 of 
The Enhanced Plan for Features, Events, and Processes (FEPs) at Yucca Mountain (BSC 2002 
[DIRS 158966]) and includes a discussion of alternative FEP identification lists and scenario 
development processes. Regardless of the specific approach chosen to perform the screening, 
the screening process is, in essence, a comparison of the FEP against the regulatory probability 
criterion and performance objectives specified in 10 CFR 63.114(d), (e), and (f) [DIRS 156605] 
regarding the inclusion or exclusion of FEPs into the TSPA-LA evaluation. 
6.1.1 Criticality FEPs Origin and Identification 
The development of a comprehensive list of FEPs potentially relevant to postclosure 
performance of the potential Yucca Mountain repository is an ongoing, iterative process based 
on site-specific information, design, and regulations. The approach for developing an initial list 
February 2004 47 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
of FEPs in support of TSPA-LA was documented in Freeze et al. (2001 [DIRS 154365]). The 
initial FEPs list contained 328 FEPs, of which 176 were included in TSPA-SR models (CRWMS 
M&O 2000 [DIRS 153246], Tables B-9 through B-17). Each FEP was assigned a unique YMP 
FEP database number, based on the Nuclear Energy Agency (NEA 1992 [DIRS 100479]) 
categories. The database number is the primary method for identifying FEPs, and consists of an 
eight-digit number having a format x.x.xx.xx.xx. A similar numbering system is used for the 
TSPA-LA FEPs list to provide a unique identifier for each FEP. In general, TSPA-SR FEPs with 
numbers ending in .00 were converted to TSPA-LA FEPs with numbers ending in .0A. New 
TSPA-LA FEPs (created when splitting TSPA-SR FEPs) are further designated with a sequential 
suffix (.0B, .0C, etc.) to ensure traceability. 
Twenty-two of the TSPA-SR FEPs in the YMP FEPs database (Freeze et al. 2001 [DIRS 
154365]) were identified as criticality related, two of which were redundant and subsequently 
deleted. The remaining 20 FEPs form the initial TSPA-LA criticality FEPs list. Table 6.1-1 lists 
the two deleted FEPs. 
Table 6.1-1. Deleted Criticality FEPs per the Enhanced FEPs Plan 
TSPA-SR 
FEP 
Name 
TSPA-SR 
FEP 
Number 
2.1.14.01.00 
Criticality 
in waste 
and EBS 
Critical 
assembly 
2.2.14.01.00 forms 
away from 
repository 
Source: Freeze et al. 2001 [DIRS 154365], Appendix B 
Additionally, the description of TSPA-SR FEP 2.1.14.14.00 was expanded for TSPA-LA to 
address multiple disruptive criticality initiating events (seismic, igneous, and rockfall). For 
TSPA-SR, this FEP only addressed an igneous initiating event. This change is shown in 
Table 6.1-2. 
TSPA-SR FEP Description 
Nuclear criticality refers to a self-sustaining fission chain reaction 
that requires a sufficient concentration and localized (critical) mass 
of fissionable isotopes (e.g., U-235, Pu-239). Thermal criticality 
requires the additional presence of neutron-moderating materials 
(e.g., water) in a suitable geometry. Fast criticality can occur 
without moderator, but generally requires a much larger critical 
mass than thermal criticality. Criticality can be prevented by the 
presence of neutron absorbing elements (e.g., boron, gadolinium). 
Within the waste and EBS, a critical mass may occur within the 
waste package (in situ) or out of the waste package and in the drift 
(near-field). This FEP aggregates all mechanisms for in situ and 
near-field criticality into a single category. Specific processes that 
could produce criticality are discussed in FEPs 2.1.14.03.00 
through 2.1.14.08.00 (for in situ) and in FEPs 2.1.14.09.00 through 
2.1.14.14.00 (for out-of-container). 
Nuclear criticality requires a sufficient concentration and localized 
(critical) mass of fissile isotopes (e.g., U-235, Pu-239) and also the 
presence of neutron-moderating materials (e.g., water) in a suitable FEP Deleted. 
geometry. Criticality is liable to be damped by the presence of 
neutron absorbing isotopes (e.g., Pu-240). Far-field criticality can 
occur if fissile material is transported away from the repository and 
then a critical mass accumulates in the presence of water. This 
FEP aggregates all mechanisms for far-field criticality into a single FEPs. 
category. Specific processes that could produce far-field criticality 
are discussed in FEPs 2.2.14.02.00 through 2.2.14.08.00. 
Basis for 
Deletion From 
TSPA-LA 
FEP Deleted. 
Redundant to 
other 
2.1.14.0x.0A 
TSPA-LA 
FEPs. 
Redundant to 
other 
2.1.14.0x.0A 
TSPA-LA 
ANL-EBS-NU-000008 REV 00 48 of 176 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 6.1-2. Changes to the Criticality FEPs per the Enhanced FEPs Plan 
TSPA-SR 
FEP Number 
TSPA-SR FEP Name, 
and Descriptiona 
2.1.14.14.00 Out-of-package 
criticality, fuel/magma 
mixture 
Interaction between fuel 
and magma dilutes 
fissile material, 
excludes water, and 
minimizes its return. For 
criticality to occur, 
neutron absorbers must 
also be removed. 
Source: a Freeze et al. 2001 [DIRS 154365], Appendix B 
b DTN: MO0307SEPFEPS4.000 [DIRS 164527] 
An initial criticality FEPs list for TSPA-LA was then developed based on the deletion and 
expansion of individual FEPs as discussed above. The initial TSPA-LA criticality FEPs have 
been documented in LA FEP List (DTN: MO0307SEPFEPS4.000 [DIRS 164527]) and are listed 
in Table 6.1-3. 
Table 6.1-3. Listing of TSPA-LA Criticality Features, Events, and Processes 
FEP Number 
2.1.14.02.0A 
2.1.14.03.0A 
2.1.14.04.0A 
FEP Name 
Criticality in situ, 
nominal 
configuration, top 
breach 
Criticality in situ, 
WP internal 
structures degrade 
faster than waste 
form, top breach 
Criticality in situ, 
WP internal 
structures degrade 
at same rate as 
waste form, top 
breach 
FEP Description 
The waste package internal structures and the waste form remain 
intact (nominal configuration). There is a breach near the top of the 
waste package, which allows water to collect in the waste package. 
Criticality then occurs in situ. 
The waste package internal structures degrade, but not the waste form. 
There is a breach near the top of the waste package, which allows 
standing water to collect in the waste package. Significant amounts of 
the neutron absorber are flushed out the top of the waste package and 
criticality occurs in situ. 
The waste package internal structures degrade at the same rate as the 
waste form. There is a breach near the top of the waste package, 
which allows water to collect in the waste package. Significant amounts 
of the neutron absorber are flushed out the top of the waste package. A 
slurry with insufficient neutron absorbing material forms at the waste 
package bottom and criticality occurs in situ. 
ANL-EBS-NU-000008 REV 00 
TSPA-LA FEP Number, Name, and 
Description 
2.1.14.14.0A  Criticality resulting from 
disruptive events 
Nuclear criticality refers to a self-sustaining 
fission chain reaction that requires sufficient 
concentration and localized (critical) mass of 
isotopes (e.g., U-235, Pu-239). This can 
include thermal criticality, which requires the 
additional presence of neutron-moderating 
materials (e.g., water) in a suitable 
geometry. Fast criticality can occur without 
moderator, but generally requires a much 
larger critical mass than thermal criticality. 
The repository will house a variety of nuclear 
waste types and configurations (e.g., CSNF 
and DSNF). A disruptive event such as 
seismic ground motion, rockfall, or igneous 
intrusion could lead to damaged packages 
and allow water (a moderator) to enter the 
packages. They could also lead to 
destruction of the internal configuration of 
the packages; release and distribution of the 
waste exterior to package; or in the case of 
an igneous intrusion drastically change the 
chemical environment and/or mix with the 
waste. Thereby, disruptive events could be 
a criticality initiating event. 
49 of 176 
Basis of Change for 
TSPALAb 
This FEP was 
expanded to include all 
disruptive criticality 
initiating events, not 
just an igneous 
intrusion resulting in a 
fuel magma mixture. 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 6.1-3. Listing of TSPA-LA Criticality Features, Events, and Processes (Continued) 
FEP Number 
2.1.14.05.0A 
2.1.14.06.0A 
FEP Name 
Criticality in situ, 
WP internal 
structures degrade 
slower than waste 
form, top breach 
Criticality in situ, 
waste form 
degrades in place 
and swells, top 
breach 
2.1.14.07.0A 
Criticality in situ, 
bottom breach 
allows flow through 
WP, fissile material 
collects at bottom 
of WP 
2.1.14.08.0A 
2.1.14.09.0A 
Criticality in situ, 
bottom breach 
allows flow through 
WP, waste form 
degrades in place 
Near-field 
criticality, fissile 
material deposited 
in near-field pond 
2.1.14.10.0A 
Near-field 
criticality, fissile 
solution flows into 
drift low point 
2.1.14.11.0A 
2.1.14.12.0A 
2.1.14.13.0A 
Near-field 
criticality, fissile 
solution is 
adsorbed or 
reduced in invert 
Near-field 
criticality, filtered 
slurry or colloidal 
stream collects on 
invert surface 
Near-field criticality 
associated with 
colloidal deposits 
ANL-EBS-NU-000008 REV 00 
FEP Description 
The waste package internal structures degrade slower than waste 
form. There is a breach near the top of the waste package, which 
allows water to collect in the waste package. The waste form degrades, 
separating from the neutron absorbers. A slurry forms at the waste 
package bottom and criticality occurs in situ. 
The waste package internal structures remain intact while the waste 
form degrades. There is a breach near the top of the waste package, 
which allows water to collect in the waste package. The waste form 
degrades in place, but swells into a more reactive configuration, which 
may overwhelm the in-place neutron absorbing material. Criticality 
occurs in situ. 
There is a breach at the bottom of the waste package, which does not 
allow water to collect in the waste package. Moderation is provided by 
water retained in clay or hydrated metal corrosion products 
accumulating in the bottom of the waste package with the fissile 
material. Significant amounts of the neutron absorber are either flushed 
from the waste package or remain distributed throughout the waste 
package, while fissile material collects at bottom of the waste package. 
Criticality occurs in situ. 
There is a breach at the bottom of the waste package, which does not 
allow water to collect in the waste package. Moderation is provided by 
water trapped in the clay or oxides. The waste form degrades in place 
and the neutron absorbing material mobilizes away from the waste 
form. Criticality occurs in situ. 
Fissile material-bearing solution or intact fissile material is deposited in 
a near-field pond. Fissile material may migrate due to bottom-only 
breach of cask or due to massive structural failure of waste package. 
Near-field criticality can result if fissile material geometry represents 
critical configuration and sufficient water is present in pond. 
Near-field criticality results when fissile material-bearing solution flows 
into a drift low point. The poison has already been separated from the 
solution carrying the fissile material, either due to retention in intact 
components within the waste package or prior removal by flow-through 
leaching within the waste package. 
Near-field criticality results from fissile solution adsorbed or reduced in 
invert (concrete and crushed tuff). The geometry of the invert allows 
zonal precipitation (under the influence of gravity) wherein the fissile 
and non fissile species may precipitate at different places within the 
invert. 
Near-field criticality results when slurry or colloidal stream is filtered 
(i.e., neutron absorbers are removed) by waste package corrosion 
products and collect on top of invert surface. 
Near-field criticality could result from colloids deposited in fractured or 
degraded concrete, from colloids filtered in the invert, or from colloids 
deposited in dead-ends of stress-relief cracks in the surrounding 
tunnel. 
February 2004 50 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 6.1-3. Listing of TSPA-LA Criticality Features, Events, and Processes (Continued) 
FEP Number FEP Name 
Criticality resulting 
from disruptive 
events 
2.1.14.14.0A 
2.2.14.02.0A 
2.2.14.03.0A 
2.2.14.04.0A 
2.2.14.05.0A 
2.2.14.06.0A 
2.2.14.07.0A 
2.2.14.08.0A 
Far-field criticality, 
precipitation in 
organic reducing 
zone in or near 
water table 
Far-field criticality, 
sorption on 
clay/zeolite in TSbv 
Far-field criticality, 
precipitation 
caused by 
hydrothermal 
upwell or redox 
front in the 
saturated zone 
Far-field criticality, 
precipitation in 
perched water 
above TSbv 
Far-field criticality, 
precipitation in 
fractures of TSw 
rock 
Far-field criticality, 
dryout produces 
fissile salt in a 
perched water 
basin 
Far-field criticality 
associated with 
colloidal deposits 
DTN: MO0307SEPFEPS4.000 [DIRS 164527] 
FEP Description 
Nuclear criticality refers to a self-sustaining fission chain reaction that 
requires sufficient concentration and localized (critical) mass of 
isotopes (e.g., U-235, Pu-239). This can include thermal criticality, 
which requires the additional presence of neutron-moderating materials 
(e.g., water) in a suitable geometry. Fast criticality can occur without 
moderator, but generally requires a much larger critical mass than 
thermal criticality. The repository will house a variety of nuclear waste 
types and configurations (e.g., CSNF and DSNF). A disruptive event 
such as seismic ground motion, rockfall, or igneous intrusion could lead 
to damaged packages and allow water (a moderator) to enter the 
packages. They could also lead to destruction of the internal 
configuration of the packages; release and distribution of the waste 
exterior to package; or in the case of an igneous intrusion drastically 
change the chemical environment and/or mix with the waste. Thereby, 
disruptive events could be a criticality initiating event. 
Fissile material is transported to an organic reducing zone and 
precipitates in a geometrically favorable configuration in or near water 
table. 
Fissile material is transported to Topopah Spring unit where it sorbs 
onto the clays and zeolites of the basal vitrophyre in a geometrically 
favorable configuration. 
Fissile material is transported to the saturated zone where it 
encounters hydrothermal upwelling or a redox front and precipitates in 
a geometrically favorable configuration in the saturated zone. 
Fissile material is transported to the perched water above the Topopah 
Spring basal vitrophyre, where chemical change causes it to precipitate 
in a geometrically favorable configuration. 
Fissile material is transported to Topopah Spring welded unit where it 
precipitates in a geometrically favorable configuration within the 
fractures. 
Fissile material is transported to a perched water basin. Dryout 
(evaporation exceeds infiltration) of the basin and the solution 
containing fissile material results in a fissile salt in a geometrically 
favorable configuration in the basin. 
Far-field criticality could result from colloids deposited in clays/zeolites 
in TSbv or deposited in perched water above the relatively 
impermeable TSbv. 
ANL-EBS-NU-000008 REV 00 February 2004 51 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Subsequent to their documentation in LA FEP List (DTN: MO0307SEPFEPS4.000 
[DIRS 164527]), the initial criticality FEPs for TSPA-LA were re-evaluated. To facilitate the 
TSPA-LA screening analysis, the criticality FEPs were reclassified for greater clarification and 
transparency into the following categories: 
1. Base Case (Nominal Scenario Class) FEPs 
A. Internal to the Waste Package 
1. Intact Configuration 
2. Degraded Configurations 
B. External to the Waste Package 
1. Near-Field Configurations 
2. Far-Field Configurations 
2. Seismic Disruptive Event FEPs 
A. Internal to the Waste Package 
Intact Configuration 
Degraded Configurations 
B. External to the Waste Package 
Near-Field Configurations 
Far-Field Configurations 
1. 
2. 
1. 
2. 
A. 
3. Rockfall Disruptive Event FEPs 
Internal to the Waste Package 
Intact Configuration 
Degraded Configurations 
B. External to the Waste Package 
Near-Field Configurations 
Far-Field Configurations 
1. 
2. 
1. 
2. 
4. 
A. 
Igneous Disruptive Event FEPs 
Internal to the Waste Package 
Intact Configuration 
Degraded Configurations 
B. External to the Waste Package 
Near-Field Configurations 
Far-Field Configurations 
1. 
2. 
1. 
2. 
The Internal to the Waste Package degraded configuration FEPs encompass the configuration 
classes identified in Figures 3-2a and 3-2b of Disposal Criticality Analysis Methodology Topical 
Report (YMP 2003 [DIRS 165505]). Figure 3-2a defines in-package bathtub configuration 
classes (hole at top of waste package and waste package flooded). Figure 3-2b defines inpackage 
flow-through configuration classes (hole at top and bottom of waste package and water 
flowing through and over waste package internals and waste form). 
February 2004 52 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The External to the Waste Package near-field configuration FEPs encompass the degraded 
configuration classes identified in Figure 3-3a of Disposal Criticality Analysis Methodology 
Topical Report (YMP 2003 [DIRS 165505]) and the far-field configuration FEPs encompass the 
configuration classes of Figure 3-3b. The near-field environment is defined as external to the 
waste package and inside the drift wall (including any drift liner and the invert). The far-field 
environment is defined as the area beyond the drift wall (i.e., in the host rock of the repository). 
Table 6.1-4 presents the revised list of 16 criticality FEPs for TSPA-LA that resulted from this 
reclassification. Because the revised list of 16 criticality FEPs and the initial list of 20 criticality 
FEPs (refer to Table 6.1-3) were derived from the configuration classes identified in Disposal 
Criticality Analysis Methodology Topical Report (YMP 2003 [DIRS 165505], Figures 3.2a, 3.2b, 
3.3a, and 3.3b), the revised list completely replaces and supercedes the initial list. Table 6.1-4 
provides a cross-reference to confirm that all of the initial criticality FEPs have been 
incorporated into the revised list of criticality FEPs for TSPA-LA. 
Table 6.1-4. Criticality FEPs List to be Utilized in Criticality Screening Analysis 
FEP Name FEP 
Number 
2.1.14.15.0A 
In-package 
criticality 
(intact 
configuration) 
2.1.14.16.0A 
In-package 
criticality 
(degraded 
configurations) 
2.1.14.17.0A Near-field 
criticality 
2.2.14.09.0A Far-field 
criticality 
FEP Description 
Base Case FEPs 
The waste package internal structures and the waste 
form remain intact. A breach (or breaches) in the waste 
package allow(s) water to either accumulate or flowthrough 
the waste package. Criticality then occurs in 
situ. In-package criticality resulting from disruptive 
events is addressed in separate FEPs. 
The waste package internal structures and the waste 
form degrade. A critical configuration (sufficient fissile 
material and neutron moderator, lack of neutron 
absorbers) develops and criticality occurs in situ. 
Potential in situ critical configurations are defined in 
Figures 3.2a and 3.2b of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 165505]). 
In-package criticality resulting from disruptive events is 
addressed in separate FEPs. 
Near-field criticality occurs when fissile material-bearing 
solution from the waste package is transported into the 
drift and the fissile material is precipitated into a critical 
configuration. Potential near-field critical configurations 
are defined in Figure 3.3a of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 165505]). 
In-package criticality resulting from disruptive events is 
addressed in separate FEPs. 
Far-field criticality occurs when fissile material-bearing 
solution from the waste package is transported beyond 
the drift and the fissile material is precipitated into a 
critical configuration. Potential far-field critical 
configurations are defined in Figure 3.3b of Disposal 
Criticality Analysis Methodology Topical Report (YMP 
2003 [DIRS 165505]). In-package criticality resulting 
from disruptive events is addressed in separate FEPs. 
Cross-Reference 
to Initial LA FEP 
List 
2.1.14.02.0A 
2.1.14.03.0A 
2.1.14.04.0A 
2.1.14.05.0A 
2.1.14.06.0A 
2.1.14.07.0A 
2.1.14.08.0A 
2.1.14.09.0A 
2.1.14.10.0A 
2.1.14.11.0A 
2.1.14.12.0A 
2.1.14.13.0A 
2.2.14.02.0A 
2.2.14.03.0A 
2.2.14.04.0A 
2.2.14.05.0A 
2.2.14.06.0A 
2.2.14.07.0A 
2.2.14.08.0A 
ANL-EBS-NU-000008 REV 00 53 of 176 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 6.1-4. Criticality FEPs List to be Utilized in Criticality Screening Analysis (Continued) 
FEP Name FEP 
Number 
2.1.14.18.0A 
2.1.14.19.0A 
In-package 
criticality 
resulting from 
a seismic 
event (intact 
configuration) 
In-package 
criticality 
resulting from 
a seismic 
event 
(degraded 
configurations) 
2.1.14.20.0A 
Near-field 
criticality 
resulting from 
a seismic 
event 
Far-field 
criticality 
2.2.14.10.0A resulting from 
a seismic 
event 
2.1.14.22.0A 
In-package 
criticality 
2.1.14.21.0A resulting from 
rockfall (intact 
configuration) 
In-package 
criticality 
resulting from 
rockfall 
(degraded 
configurations) 
2.1.14.23.0A 
Near-field 
criticality 
resulting from 
rockfall 
ANL-EBS-NU-000008 REV 00 
FEP Description 
Seismic Disruptive Event FEPs 
The waste package internal structures and the waste 
form remain intact either during or after a seismic 
disruptive event. A breach (or breaches) in the waste 
package allow(s) water to either accumulate or flowthrough 
the waste package. Criticality then occurs in 
situ. 
Either during, or as a result of, a seismic disruptive 
event, the waste package internal structures and the 
waste form degrade. A critical configuration develops 
and criticality occurs in situ. Potential in situ critical 
configurations are defined in Figures 3.2a and 3.2b of 
Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). 
Either during, or as a result of, a seismic disruptive 
event, near-field criticality occurs when fissile materialbearing 
solution from the waste package is transported 
into the drift and the fissile material is precipitated into a 
critical configuration. Potential near-field critical 
configurations are defined in Figure 3.3a of Disposal 
Criticality Analysis Methodology Topical Report (YMP 
2003 [DIRS 165505]). 
Either during, or as a result of, a seismic disruptive 
event, far-field criticality occurs when fissile materialbearing 
solution from the waste package is transported 
beyond the drift and the fissile material is precipitated 
into a critical configuration. Potential far-field critical 
configurations are defined in Figure 3.3b of Disposal 
Criticality Analysis Methodology Topical Report (YMP 
2003 [DIRS 165505]). 
Rockfall Disruptive Event FEPs 
The waste package internal structures and the waste 
form remain intact either during or after a rockfall event. 
A breach (or breaches) in the waste package allow(s) 
water to either accumulate or flow-through the waste 
package. Criticality then occurs in situ. 
Either during, or as a result of, a rockfall event, the waste 
package internal structures and the waste form degrade. 
A critical configuration develops and criticality occurs in 
situ. Potential in situ critical configurations are defined in 
Figures 3.2a and 3.2b of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 165505]). 
Either during, or as a result of, a rockfall event, near-field 
criticality occurs when fissile material-bearing solution 
from the waste package is transported into the drift and 
the fissile material is precipitated into a critical 
configuration. Potential near-field critical configurations 
are defined in Figure 3.3a of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 165505]). 
54 of 176 
Cross-Reference 
to Initial LA FEP 
List 
2.1.14.14.0A 
2.1.14.14.0A 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 6.1-4. Criticality FEPs List to be Utilized in Criticality Screening Analysis (Continued) 
FEP Name FEP 
Number 
2.2.14.11.0A 
Far-field 
criticality 
resulting from 
rockfall 
2.1.14.24.0A 
In-package 
criticality 
resulting from 
an igneous 
event (intact 
configuration) 
In-package 
criticality 
resulting from 
2.1.14.25.0A an igneous 
event 
(degraded 
configurations) 
Near-field 
criticality 
2.1.14.26.0A resulting from 
an igneous 
event 
Far-field 
criticality 
2.2.14.12.0A resulting from 
an igneous 
event 
6.1.2 FEPs Screening Process 
The first step in the FEP analysis process, described in Section 6.1.1, was the identification of 
FEPs. The second step in the FEP analysis process is the screening of each FEP against the 
project screening criteria. The NRC requires the consideration and evaluation of FEPs as part of 
the performance assessment activities. More specifically, the NRC regulations allow the 
exclusion of FEPs from the TSPA if they can be shown to be of low probability or of low 
consequence. The specified criteria can be summarized in the form of two FEP screening 
statements as follows. 
1. The event has at least one chance in 10,000 of occurring over 10,000 years (10 CFR 
63.114(d) [DIRS 156605]). 
ANL-EBS-NU-000008 REV 00 
FEP Description 
Either during, or as a result of, a rockfall event, far-field 
criticality occurs when fissile material-bearing solution 
from the waste package is transported beyond the drift 
and the fissile material is precipitated into a critical 
configuration. Potential far-field critical configurations 
are defined in Figure 3.3b of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 165505]). 
Igneous Disruptive Event FEPs 
The waste package internal structures and the waste 
form remain intact either during or after an igneous 
disruptive event. A breach (or breaches) in the waste 
package allow(s) water to either accumulate or flowthrough 
the waste package. Criticality then occurs in 
situ. 
Either during, or as a result of, an igneous disruptive 
event, the waste package internal structures and the 
waste form degrade. A critical configuration develops 
and criticality occurs in situ. Potential in situ critical 
configurations are defined in Figures 3.2a and 3.2b of 
Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). 
Either during, or as a result of, an igneous disruptive 
event, near-field criticality occurs when fissile materialbearing 
solution from the waste package is transported 
into the drift and the fissile material is precipitated into a 
critical configuration. Potential near-field critical 
configurations are defined in Figure 3.3a of Disposal 
Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). 
Either during, or as a result of, an igneous disruptive 
event, far-field criticality occurs when fissile materialbearing 
solution from the waste package is transported 
beyond the drift and the fissile material is precipitated 
into a critical configuration. Potential far-field critical 
configurations are defined in Figure 3.3b of Disposal 
Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). 
55 of 176 
Cross-Reference 
to Initial LA FEP 
List 
2.1.14.14.0A 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
2. The magnitude and time of the resulting radiological exposure to the reasonably 
maximally exposed individual, or radionuclide release to the accessible environment, 
would be significantly changed by its omission (10 CFR 63.114(e) and (f) [DIRS 
156605]). 
Additionally, the Acceptance Criteria in Yucca Mountain Review Plan (NRC 2003 [DIRS 
163274], Section 2.2.1.2.1.3) requires evaluating FEPs based on the regulations. This criterion 
can be summarized in the form of a third FEP screening statement. 
3. The FEP is not excluded by regulation. 
Evaluating FEPs against these screening statements may be done in any order. If there are 
affirmative conditions for all three screening criteria, the FEP is Included in the TSPA-LA 
model. If there is a negating condition in any of the three screening criteria, the FEP is 
Excluded from the TSPA-LA model. 
The approach taken in this analysis is to evaluate the probability of the criticality FEPs based on 
the interactions between the natural and engineered barrier systems over the performance period 
to determine if this event can be excluded from evaluation in the TSPA-LA based on low 
probability of occurrence. 
6.1.3 Background, Technical Information, and Literature Searches 
Data and technical information sources used for the FEPs evaluation are cited within each FEP 
discussion, and use of this information and data has been documented per YMP procedural 
requirements. Where possible, the technical information used in this analysis report to support 
the screening decisions was obtained from controlled source documents and references using the 
appropriate document identifiers or records system accession numbers. Sources of such data 
include, but are not limited to, analyses, models, technical reports, and other YMP documents 
and databases. As needed, alternative and corroborative information and data were obtained 
from literature searches of peer-reviewed journals, other widely recognized scientific periodicals, 
results of review of YMP documents by external organizations, and other appropriate sources 
such as technical handbooks and textbooks. A listing of the corroborating data, models, product 
input, or technical information used to support the criticality FEPs screening decisions are 
provided in Table 6.1-5. 
Table 6.1-5. Corroborating Data, Models, Product Input or Technical Information 
Reference Description 
Evaluation of Codisposal Viability for MOX 
(FFTF) DOE-Owned Fuel (CRWMS M&O 
1999 [DIRS 125206]) 
Evaluation of Codisposal Viability for UzrH 
(TRIGA) DOE-Owned Fuel (CRWMS M&O 
2000 [DIRS 147650]) 
Evaluation of Codisposal Viability for HEU 
Oxide (Shippingport PWR) DOE-Owned Fuel 
(CRWMS M&O 2000 [DIRS 147651]) 
21-PWR Waste Package with Absorber Plates 
Loading Curve Evaluation (BSC 2003 [DIRS 
166610]) 
ANL-EBS-NU-000008 REV 00 
Section Used in Input Description 
criticality analyses 
February 2004 
Reference 
Sections Used 
Reference to FFTF Entire Table 6.2-1 criticality evaluations 
Reference to TRIGA Entire Table 6.2-1 criticality analyses 
Reference to 
Entire Table 6.2-1 Shippingport PWR 
56 of 176 
Reference to loading Entire Section 6.2 curve analyses Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 6.1-5. Corroborating Data, Models, Product Input or Technical Information (Continued) 
Reference Description 
Evaluation of Codisposal Viability for U-Zr/UMo 
Alloy (Enrico Fermi) DOE-Owned Fuel 
(CRWMS M&O 2000 [DIRS 151742]) 
Evaluation of Codisposal Viability for Th/U 
Oxide (Shippingport LWBR) DOE-Owned Fuel 
(CRWMS M&O 2000 [DIRS 151743]) 
Evaluation of Codisposal Viability for U-Metal 
(N Reactor) DOE-Owned Fuel (CRWMS M&O 
2001 [DIRS 154194]) 
Yucca Mountain Science and Engineering 
Report (DOE 2002 [DIRS 155943]) 
Evaluation of Codisposal Viability for Melt and 
Dilute DOE-Owned Fuel (BSC 2001 [DIRS 
157733]) 
Evaluation of Codisposal Viability for Th/U 
Carbide (Fort Saint Vrain HTGR) DOE-Owned 
Fuel (BSC 2001 [DIRS 157734]) 
External Accumulation of Fissile Material from 
DOE Co-Disposal Waste Packages (BSC 
2002 [DIRS 159913]) 
44 BWR Waste Package Loading Curve 
Evaluation (BSC 2001 [DIRS 161125]) 
Disposal Criticality Analysis Methodology 
Topical Report (YMP 2003 [DIRS 165505]) 
Criticality Model Report (BSC 2003 [DIRS 
165733]) 
WAPDEG Analysis of Waste Package and 
Drip Shield Degradation (CRWMS M&O 2000 
[DIRS 151566]) 
WAPDEG Analysis of Waste Package and 
Drip Shield Degradation (CRWMS M&O 2000 
[DIRS 151566]) 
WAPDEG Analysis of Waste Package and 
Drip Shield Degradation (BSC 2003 [DIRS 
161317]) 
Geochemistry Model Abstraction and 
Sensitivity Studies for the 21 PWR CSNF 
Waste Packages (BSC 2002 [DIRS 160638]) 
Characterize Framework for Igneous Activity 
at Yucca Mountain, Nevada (BSC 2003 [DIRS 
163769]) 
Igneous Intrusion Impacts on Waste Package 
and Waste Form (BSC 2003 [ DIRS 165002]) 
Thermodynamics (Wark 1983 [DIRS 157283]) 
ANL-EBS-NU-000008 REV 00 
Input Description Section Used in Reference 
Sections Used 
Reference to Enrico Entire Table 6.2-1 Fermi criticality analyses 
Reference to 
Entire Table 6.2-1 Shippingport LWBR 
criticality analyses 
Reference to N Reactor Entire Table 6.2-1 criticality analyses 
Section 6.6 Table 3-3 
Section 6.6 Section 
4.3.3.2.1 
Percentage of MTHM by 
waste package design 
Possibility of criticality 
due to an igneous event 
Reference to Melt and Entire Table 6.2-1 Dilute criticality analyses 
Reference to Fort St. Entire Table 6.2-1 Vrain criticality analyses 
Section 6.3.2 Section 5.2.2 Plume description cited to 
support conclusions. 
Reference to loading Entire Section 6.2 curve analyses 
Sections 6.2 and Section 3.3 Master Scenario List 
Section 6.2.1 
6.3 
Section 6.2 
Sections 6.2 and Entire 6.3 
Section 6.3.3.3.2 Section 6.5.1 
Definition of criticality limit 
Reference to criticality 
model 
Waste package failure 
due to pitting and crevice 
corrosion 
Formation and generation Entire Section 6.3.3.4 of bathtub configurations 
37, and 38 
General corrosion failure 
Section 6.6.2 Sections 6.3.3.2.1, of the waste package and 
and Figures 36, 6.3.3.3.1, 6.3.3.3.3, drip shield and stress 
and 6.3.3.3.4 
Section 6.3.2 Sections 4.1.2.1 
and 4.1.2.2 
corrosion cracking failure 
of the waste package 
Information on the 
retention and release of 
fissile material from a 
degraded waste package 
Table 22 Sections 6.2 and 
6.6 
Annual frequency of an 
intrusive igneous event 
Section 6.6.1 Section 1 
Sections 6.6.1 and 
6.6.2 
Sections 6.5.2 
and 6.7.2 
Igneous impacts on 
waste package in Zone 1 
and Zone 2; Zone 2 
impacts due to volatile 
gases 
Section 6.6.3 Table A-12 Saturation temperature at 
7.5 MPa 
February 2004 57 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
6.1.4 Assumptions and Simplifications, Alternative Conceptual Models, and 
Consideration of Uncertainty in FEPs Screening 
The generic assumptions used in the FEPs screening for the criticality FEPs are provided in 
Section 5, along with their rationale, confirmation status, and where used in this document. 
Specific guidance and criteria for the consideration of alternative conceptual models (and their 
relationship to FEPs) and the treatment of uncertainty are provided in Section 5 and 
Appendices A and C of Scientific Processes Guidelines Manual (BSC 2002 [DIRS 160313]). 
Alternative conceptual models and uncertainty issues are addressed in supporting documentation 
cited as part of the FEPs evaluation. 
6.1.5 Alternative Approaches, Mathematical Formulations, and Units of Measure 
Alternative approaches to the configuration generator model used to perform most of the FEP 
screening arguments are discussed in Configuration Generator Model for In-Package Criticality 
(BSC 2003 [DIRS 165629]). The current approach was selected because its results are more 
transparent. 
In general, FEPs screening involves the comparison of the probability of occurrence of some 
feature, event, or process to some threshold level of probability, or some other threshold measure 
that defines the onset of consequence to repository performance. Mathematical and numerical 
formulations typically are used to define the measure of the FEP of interest (e.g., development of 
the seismic hazard curves) and its probability of occurrence. Any mathematical formulations, 
equations, or algorithms used in the criticality FEP evaluations are discussed in the appropriate 
section. 
Depending on the criticality FEPs being evaluated, the units of measure of the source 
information may vary. In all cases, the units as they appeared in the cited source are provided to 
allow traceability, and metric equivalents, if necessary, are provided in parentheses for 
consistency and transparency. 
6.1.6 Model and Software Issues for Previously Developed and Validated Models 
The configuration generator model used to determine the probability of potentially critical 
configurations for the FEPs screening analysis uses the SAPHIRE code. The validation of the 
model is discussed in a separate model report (BSC 2003 [DIRS 165629]). The SAPHIRE code 
is discussed in Section 3.1.2. 
6.1.7 Intended Use and Limitations 
The intended use of this analysis report is to provide criticality FEPs screening information for 
the project-specific FEPs database, and to promote traceability and transparency regarding FEPs 
disposition. This analysis report presents the source documentation to provide the technical 
basis, and to provide the supporting arguments for exclusion of criticality FEPs from the TSPALA 
model. Accordingly, this analysis report may be of use to reviewers during the licensing 
review process. 
February 2004 58 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
6.2 CRITICALITY FEPS SCREENING ANALYSIS MODEL 
The criticality FEPs screening analysis utilizes the event tree/fault tree process model developed 
in Configuration Generator Model for In-Package Criticality (BSC 2003 [DIRS 165629], 
Attachment XII) for the evaluation of the overall probability of criticality. The configuration 
generator model identifies the possible pathways required for the development of internal waste 
package configuration classes (degraded or otherwise), evaluates the probability of occurrence 
for the configuration classes, and provides the configuration class associated parameter ranges to 
determine the criticality potential of each configuration class. Although currently only validated 
for application with 21-PWR with Absorber Plates Waste Packages, the methods provided in this 
model are considered to be applicable to all waste forms and waste package types with adequate 
justification. 
In addition, as documented, the configuration generator model allows for a probability evaluation 
of the potentially critical configurations to be performed at several points during the 
development of the configuration classes. The first logical point of evaluation would be the 
probability that water will enter a failed waste package and the waste package would remain in a 
flooded condition during the postclosure performance period (i.e., during the first 10,000 years 
after the permanent closure of the repository). This is a logical evaluation point because unless 
water enters and floods the waste package, the waste package internals cannot degrade and allow 
the neutron absorber material to be removed from the waste package. The loss of neutron 
absorber materials and the presence of a neutron moderator are required to achieve an in-package 
criticality during the postclosure performance period. 
The second logical evaluation point is the probability of configuration class formation. These 
configuration classes are defined by the Master Scenario List in Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 165505], Section 3.3) and are discussed in 
Section 6.2 of Configuration Generator Model for In-Package Criticality (BSC 2003 [DIRS 
165629]). These configuration classes, representing waste package flooded and flow-through 
configurations, represent the physical characteristics of the waste package, waste package 
internals, and waste forms that are possible during the waste package/waste form degradation 
processes. It should be noted that the criticality FEPs were derived from the configuration 
classes defined by the Master Scenario List in Disposal Criticality Analysis Methodology Topical 
Report (YMP 2003 [DIRS 165505], Section 3.3). It is necessary to degrade the waste 
package/waste form in some manner to achieve a potentially critical configuration. This is 
because intact, fully loaded, fully flooded waste package conditions are precluded from 
achieving criticality by design to satisfy a preclosure operations requirement that the MGR 
provide means to ensure criticality control during SNF/HLW handling operations, including 
waste package loading (Siddoway 2003 [DIRS 163904], Requirement 1.1.6-4). If this 
requirement is satisfied, in situ criticality in an intact configuration (criticality 
FEPs 2.1.14.15.0A, 2.1.14.18.0A, 2.1.14.21.0A, or 2.1.14.24.0A) cannot occur. 
It has been previously demonstrated that through loading curve analyses for the 21-PWR with 
Absorber Plates and the 44-BWR waste package types that an intact, fully flooded waste package 
configuration cannot achieve criticality (BSC 2003 [DIRS 166610] and BSC 2001 [DIRS 
161125], respectively). To satisfy Requirement 1.1.6-4 (Siddoway 2003 [DIRS 163904]), 
similar analyses must be performed for the remaining commercial SNF waste packages. 
February 2004 59 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Analyses have also been previously performed for eight of the nine representative DOE SNF 
waste forms that demonstrate subcriticality of these waste package types for intact, fully flooded 
conditions. The references for these analyses are listed in Table 6.2-1. 
Table 6.2-1. DOE SNF Intact Configuration Criticality Analysis References 
Intact Configuration Criticality Analysis 
Reference 
CRWMS M&O 1999 [DIRS 125206] 
CRWMS M&O 2000 [DIRS 147650] 
CRWMS M&O 2001 [DIRS 154194] 
CRWMS M&O 2000 [DIRS 147651] 
CRWMS M&O 2000 [DIRS 151743] 
BSC 2001 [DIRS 157734] 
BSC 2001 [DIRS 157733] 
CRWMS M&O 2000 [DIRS 151742] 
Not available. 
The final logical evaluation point is to assess the criticality probability for each configuration 
class. Using the criticality model (BSC 2003 [DIRS 165733]) and the configuration class 
characteristics defined by the configuration generator model, detailed criticality analyses are 
performed to determine the effective neutron multiplication factor (keff) for the range of 
parameters associated with each configuration class. If the calculated keff is below a prescribed 
critical limit (BSC 2003 [DIRS 165733], Section 6.2.1) for the entire range of parameters, the 
configuration class has no criticality potential. If the calculated keff is above a prescribed critical 
limit for some or all of the range of parameters, then the probability of achieving these parameter 
ranges is assessed. The probability of achieving these parameter ranges is the configuration 
classs criticality probability. 
It is the intent of the criticality FEPs screening analysis to perform a probability evaluation of 
only the potential for water to enter the waste package and maintain a flooded condition during 
the postclosure performance period. The portions of the configuration generator event tree/fault 
tree process model to be exercised in this analysis are as follows. 
The first event tree developed in the configuration generator model defines the fractional 
breakdown of the waste forms and waste package types proposed for disposal in the repository. 
This event tree (Figure 6.2-1) has been modified slightly to indicate the fraction of total waste 
package inventory for each waste package type, including naval waste package types. The 
updated inventory fractions are based on the information provided in Table 4.1-8. Although the 
Naval Nuclear Propulsion Program is responsible for assessing criticality potential of the naval 
waste package types in accordance with an addendum (Mowbray 1999 [DIRS 149585]) to 
Disposal Criticality Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]), these 
waste package types are presented on this event tree for completeness. 
February 2004 
DOE SNF Representative Waste Form 
1. Fast Flux Test Facility (FFTF) 
2. Training, Research, Isotopes, General Atomics (TRIGA) 
3. N Reactor 
4. Shippingport PWR 
5. Shippingport LWBR 
6. Fort St. Vrain 
7. Melt and Dilute 
8. Enrico Fermi 
9. Three Mile Island II 
60 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Waste Package 
Fraction 
Waste Package Type 
Percent Breakdown 
Waste Form 
Sub-Type 
Percentages 
Waste Form Type 
Percentages 
by Originator 
WP-SOURCE WP WP-TYPE WP-IND 
PWR 
(40.7% of total inventory) 
Commercial CSNF 
(66.8% of total inventory) 
BWR 
(26.1% of total inventory) 
Waste Package Fraction 
DOE SNF 
(30.5% of total inventory) 
Naval SNF 
(2.7% of total inventory) 
Source: Criticality FEPs Analysis SAPHIRE Model (Attachment VII) 
Figure 6.2-1. Waste Form and Waste Package Types Proposed for Disposal at the Repository 
Eight commercial and DOE SNF waste package types (Figure 6.2-1) have been utilized as the 
initiating event in a new event tree, which only transfers to another event tree that directs the 
evaluation of a waste packages criticality potential resulting from the four criticality FEPs cases. 
An example of the Waste Package Type event tree is presented in Figure 6.2-2. The Waste 
Package Type event tree of Figure 6.2-2 automatically transfers to a third event tree. The 
transfer is indicated by the T after the event tree sequence number in the # column. The 
CASE end state name in the END_STATE_NAMES column of this event tree indicates the 
name of the event tree to which the transfer occurs. 
The CASE event tree is presented in Figure 6.2-3. This event tree directs the evaluation of the 
four criticality FEPs cases  (1) Base Case, (2) Seismic Disruptive Event, (3) Rockfall 
Disruptive Event, and (4) Igneous Disruptive Event. These cases are represented as branches in 
the event tree top event FEP. As shown in Figure 6.2-3, the Igneous Disruptive Event ends in an 
OK end state that indicates SAPHIRE processing and no further event tree/fault tree evaluation 
of this initiating event is performed. An evaluation of the igneous disruptive event is provided in 
Section 6.6 of this report. The probabilities of occurrence assigned to each of the four basic 
events representing these four criticality FEPs cases are as follows: 
61 of 176 ANL-EBS-NU-000008 REV 00 February 2004 
# END-STATE-NAMES 
21-PWR Absorber Plate 
(38.4% of total inventory) 
1 21PWR-AP-WP 
21-PWR Control Rod 
(0.8% of total inventory) 
2 21PWR-CR-WP 
12-PWR Long 
(1.5% of total inventory) 
3 12PWR-WP 
44-BWR 
(25.3% of total inventory) 
4 44BWR-WP 
24-BWR 
(0.8% of total inventory) 
5 24BWR-WP 
DOE Long 
(10.3% of total inventory) 
6 DOE-LONG-WP 
DOE Short 
(18.9% of total inventory) 
7 DOE-SHORT-WP 
DOE MCO 
(1.3% of total inventory) 
8 DOE-MCO-WP 
Naval Short 
(1.3% of total inventory) 
9 NAVAL-SHORT-WP 
Naval Long 
(1.4% of total inventory) 
10 NAVAL-LONG-WP 
FRACTION 
3.840E-001 
8.000E-003 
1.500E-002 
2.53E-001 
8.000E-003 
1.030E-001 
1.890E-001 
1.300E-002 
1.300E-002 
1.400E-002 
Screening Analysis for Criticality Features, Events, and Processes for License Application 
FEP_BASECASE = 1.000E-000 (always possible) 
FEP_SEISMIC = 1.000E-000 (always possible) 
FEP_ROCKFALL = 1.000E-000 (always possible) 
FEP_IGNEOUS = 1.700E-004 (event frequency of 1.7E-8/year  10,000 years) 
(BSC 2003 [DIRS 163769], Table 22) 
Initiate Evaluation of 
21-PWR with Absorber Plates 
Waste Package Type 
# <PASS> WP01-21-PWR-AP 
1 T 
Source: Criticality FEPs Analysis SAPHIRE Model (Attachment VII) 
Figure 6.2-2. Waste Package Type Event Tree 
The seismic disruptive event has been assigned a value of 1.0 because the probability of its 
occurrence is integral to the calculation of the minimum required seepage. This calculation is 
documented in Section 6.4 and Attachments III-V. Assigning the seismic disruptive event a 
value other than 1.0 would result in double counting of the probability of its occurrence. The 
rockfall disruptive event also has been assigned a value of 1.0. Rockfall is the result of natural 
drift degradation phenomena and is expected to occur throughout the postclosure period without 
any predictable frequency. The evaluation of the rockfall disruptive event is presented in 
Section 6.5. The rockfall disruptive event is differentiated from rockfall that may occur during a 
seismic disruptive event. Damage resulting from seismic induced rockfall is accounted for in the 
Section 6.4. 
62 of 176 ANL-EBS-NU-000008 REV 00 
END_STATE_NAMES 
CASE 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Incoming Waste 
Package Type 
Identifier 
IE1
WP Type 
Source: Criticality FEPs Analysis SAPHIRE Model (Attachment VII) 
A second top event, DRIFT, is presented in the CASE event tree of Figure 6.2-3. This top 
event splits the evaluation between the two geological zones of the drifts  lithophysal and 
nonlithophysal. Based on total drift area information (BSC 2003 [DIRS 164491], Table 9), of 
the 4,983,152 m2 of total emplacement drift, 745,486 m2 reside in the nonlithophysal geological 
zone. This results in a top event split fraction of 0.15 for nonlithophysal (745,486/4,983,152) 
and 0.85 for lithophysal. These values are applied to the event tree evaluation by setting the 
probability of basic event DRIFT to 0.85. As will be discussed in Sections 6.4 (seismic 
disruptive event criticality FEPs) and 6.5 (rockfall disruptive event criticality FEPs), it is 
important to distinguish between the two geological units during rockfall impact evaluations. 
The CASE event tree of Figure 6.2-3 automatically transfers to a fourth event tree. The 
transfer is indicated by the T after the first six event tree sequence numbers in the # column. 
The FEPS end state name in the END-STATE column of this event tree indicates the name 
of the event tree to which the transfer occurs. The FEPS event tree (shown in Figure 6.2-4) 
performs the probability evaluation for water ingress into a waste package, waste package 
flooding, and maintenance of a flooded configuration (i.e., bathtub) for the duration of the 
postclosure performance period. 
ANL-EBS-NU-000008 REV 00 
Figure 6.2-3. Event Tree for Processing Criticality FEPs Cases 
63 of 176 
Rock Type 
of Drift 
FEPs Initiating 
Event 
FEP
Base Case 
Probability = 1.0 
Seismic Disruptive Event 
Probability = 1.0 
Rock Fall Disruptive Event 
Probability = 1.0 
Igneous Disruptive Event 
Probability = 1.7E-4 
DRIFT 
Nonlithophysal (15%) 
Lithophysal (85%) 
Nonlithophysal 
Lithophysal 
Nonlithophysal 
Lithophysal 
END-STATE # 
FEPS 1 T 
FEPS 2 T 
FEPS 3 T 
FEPS 4 T 
FEPS 5 T 
FEPS 6 T 
7 OK 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
As presented in Figure 6.2-4, only four top events are necessary to define the processes to assess 
the probability of the formation of a waste package bathtub configuration. The purpose of the 
first top event (MS-IC-1) is to evaluate the probability of the water flux necessary to penetrate a 
waste package in order to degrade waste package internals and remove the neutron absorber 
material (creating a potentially critical configuration) reaching the drift. If sufficient water 
reaches the drift, then top event MS-IC-2 is queried. However, if sufficient water does not reach 
the drift (i.e., the branching goes up), then the analysis ends at this point because there is not 
sufficient water to create a potentially critical configuration. 
Top event MS-IC-2 evaluates the probability that, given sufficient water flux in the drift, the drip 
shield is failed in such a manner to allow water to pass through to the waste package. If the drip 
shield is not failed, then the analysis is halted as indicated by the up branch of this top event. 
However, if the drip shield is failed and sufficient water is allowed to pass through, then top 
event MS-IC-3 is queried. 
Top event MS-IC-3 evaluates the probability that, given sufficient water flux passes through the 
drip shield, the waste package is failed in such a manner to allow the water to enter in sufficient 
quantity to support the generation of a potentially critical configuration. If the waste package is 
not failed, then the analysis is halted. However, if the waste package is failed and sufficient 
water is allowed to pass through, then top event MS-IC-4 is queried. 
February 2004 64 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Sufficient Water 
Reaches Drift 
Drip Shield 
Fails 
Waste Package 
Fails 
Transfer from 
Initiating Event Tree 
MS-IC-1 MS-IC-2 IE3 
Insufficient Water 
Reaches the Drift 
Water Does Not 
Drip on Waste Package 
Sufficient Water 
Reaches the Drift 
Water Drips on 
Waste Package 
Source: Criticality FEPs Analysis SAPHIRE Model (Attachment VII) 
Figure 6.2-4. Criticality FEPs Screening Analysis Event Tree 
Top event MS-IC-4 evaluates the probability that, given sufficient water flux passes into the 
waste package, water accumulates in the waste package for a sufficient duration to allow for the 
degradation of the waste package internals and the removal of the neutron absorbing materials, 
creating a potentially critical configuration. The accumulation and retention of water in the 
waste package is referred to as a bathtub configuration. It is also possible for sufficient water to 
enter the waste package, but the water does not accumulate. Rather, a breach in the waste 
package bottom could allow for water to flow out. This condition is referred to as a flowthrough 
configuration. The waste package internals could degrade under such conditions and the 
neutron absorber materials removed, but there would be insufficient water retained in the waste 
package to allow for neutron moderation, a necessary component to support in-package 
criticality. The analysis is halted at this point. 
Another possible configuration is one in which a breach in the top and bottom of the waste 
package exists, but that the bottom hole is much smaller than the top hole, so more water can 
enter the waste package through the top than can exit through the bottom. This configuration is 
not explicitly considered in this analysis because the low seepage rates predicted in the 
repository would preclude this configuration from occurring and because the bottom breach 
ANL-EBS-NU-000008 REV 00 65 of 176 
Water Accumulates 
in Waste Package 
MS-IC-3 END-STATE-NAMES MS-IC-4 # 
NO-CRITICALITY 1 
NO-CRITICALITY 2 
Water Does Not 
Penetrate Waste Package 
NO-CRITICALITY 3 
Water Flows Through 
Waste Package FLOW-THRU-CONFIG 4 
Water Penetrates Top 
of Waste Package 
Water Accumulates 
in Waste Package 
BATHTUB-CONFIG 5 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
would rapidly ablate due to accelerated corrosion resulting from the pressurized flow through the 
bottom hole. In addition, this waste package configuration can be considered a subset of the 
bathtub configuration. 
If any of the first three top events branch upward (nonoccurrence of the event), the analysis stops 
because the conditions required to achieve a potentially critical configuration are not met. This 
is noted in the END-STATE-NAMES column through the assignment of the NOCRITICALITY 
end state. The evaluation of top event MS-IC-4 results in the generation of two 
end states indicating the probability of generating a waste package bathtub configuration (end 
state BATHTUB-CONFIG) or a waste package flow through configuration (end state FLOWTHRU-
CONFIG). 
Each of the top events of Figure 6.2-4 is supported with fault trees. In turn, each fault tree is 
quantified through the evaluation of logic that defines the relationship between dependent and 
independent events (referred to as basic events). The supporting fault trees are presented in 
Attachment II, Figures II-5 through II-8. Sections 6.3 through 6.5 quantify the fault tree basic 
events for the three criticality FEP cases evaluated using SAPHIRE (i.e., Base Case, Seismic 
Disruptive Event, and Rockfall Disruptive Event), respectively. 
6.3 ANALYSIS OF BASE CASE CRITICALITY FEPS 
This screening analysis of the base case postclosure criticality FEPs is based on the probability 
that sufficient water to degrade and subsequently flush out the neutron absorber material reaches 
the waste form during the period of performance (10,000 years after repository closure). 
For a criticality event to occur, the proper combination of materials (neutron moderators, neutron 
absorbers, fissile) and geometric configuration must exist. A critical system for the geological 
repository is defined as one having an effective neutron multiplication factor (keff), larger than 
the critical limit. The critical limit is the value of keff at which a system (configuration of fissile 
material) is considered critical as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
All base case (i.e., nominal scenario class, no disruptive events) postclosure criticality FEPs, 
internal and external, require water infiltration to degrade the waste package internals. Neutron 
absorber material loss and a flooded waste package condition for neutron moderation is the most 
likely scenario that could result in a potentially critical configuration in any of the in situ 
criticality FEPs. External criticality FEPs (near-field and far-field) also require the separation of 
neutron absorber materials from the waste form, transport of fissile material from the waste 
package, and accumulation of the fissile material within the drift invert or beyond. 
Water, silica, and carbon are the only potential moderating materials for internal and external 
configurations. Water is the most effective neutron-moderating material, which can enter the 
waste package as percolation flow or be present in the pores of the rock. Silica is present in 
appreciable quantities in the high-level radioactive waste glass canisters and in the rock. Silica 
can also be introduced into the waste package through precipitation from the percolation flow. 
Carbon is present in only limited amounts in select DOE SNF types and, therefore, has a limited 
impact on the potential for criticality. The loading of the DOE-standardized SNF canisters, the 
February 2004 66 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
design of the basket structure inside the canisters, and the addition of neutron absorbers take into 
account the presence and effect of glass in DOE SNF codisposal waste packages. Silica from the 
high-level radioactive waste glass canisters has no impact on the potential for criticality in DOE 
SNF codisposal waste packages. Silica is a much less effective moderator than water and its 
introduction into commercial SNF waste packages will displace water and effectively reduce the 
reactivity of the system, thus reducing the potential for criticality. 
In addition, criticality without water infiltration is unlikely for the geologic repository because 
the waste package is designed such that a criticality event in an intact waste package 
configuration is not possible (refer to intact waste package configuration discussion in 
Section 6.2). 
Some of the DOE SNF waste forms are highly enriched and could potentially support 
unmoderated (fast) criticality if (1) the (fissile) material is concentrated beyond its design 
concentration in the waste form and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration could result from either 
waste form degradation or a disruptive event. However, removal of the neutron absorber 
materials from a DOE SNF waste package would require a breach of the waste package and a 
removal mechanism. The most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal components, the 
dissolution of the neutron absorber material, and flushing of the material from the waste package. 
6.3.1 Internal (In Situ) Criticality 
Water entering a failed waste package may occur from two primary pathways: (1) water dripping 
from the drift crown, and (2) water dripping from the underside of the drip shield due to 
evaporation and condensation. The first pathway can occur if the drip shield fails to divert 
dripping water from the drift crown into a failed waste package. The second pathway can occur 
if water vapor condenses on the underside of the drip shield and falls onto and enters a failed 
waste package. Only the first pathway is discussed for top event MS-IC-1 in Section 6.3.3.1 as 
the model for calculating the quantity of condensation available to enter a failed waste package is 
not yet available. The availability of water, drip shield and waste package failure, and the 
formation of a bathtub configuration are associated with top events MS-IC-1, MS-IC-2, MS-IC-3 
and MS-IC-4 of the criticality FEPs screening analysis event tree. 
The internal criticality FEPs probability analysis is performed only to the point of waste package 
flooding because a flooded waste package is necessary to obtain appreciable degradation of the 
waste package internals for the removal of the neutron absorber materials and to provide 
effective neutron moderation. The intact, fully flooded configuration of FEPs 2.1.14.15.0A, 
2.1.14.18.0A, 2.1.14.21.0A, and 2.1.14.24.0A has been discussed previously in Section 6.2. 
Criticality is precluded by design for this configuration. Table 6.3-1 presents the list of base case 
internal (in situ) criticality FEPs. 
February 2004 67 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
FEP Number 
2.1.14.15.0A 
2.1.14.16.0A 
Source: Table 6.1-4 
Table 6.3-1. Base Case Configurations: Internal (In Situ) Criticality FEPs 
FEP Name 
In-package criticality 
(intact configuration) 
In-package criticality 
(degraded 
configurations) 
6.3.2 External (Near-Field and Far-Field) Criticality 
The probability of external criticality is less than the probability of water entering a failed waste 
package. If the probability of water entering a failed waste package (in either a bathtub or flowthrough 
configuration) during the performance period is calculated to be below the regulatory 
probability criterion (less than one chance in 10,000 of occurring over 10,000 years 
(10 CFR 63.114(d) [DIRS 156605]), then the probability of an external criticality would be even 
lower. This is because, in addition to the events evaluated to calculate the probability of water 
entering a failed waste package, the probability of the following events must be considered for 
external criticality: 
 Waste form degradation over the performance period; 
 Separating the fissile materials from the degraded waste form; 
 Removing the fissile materials from the waste package; 
 Accumulating sufficient fissile material into a potentially critical configuration in the 
near-field or far-field environments; and 
 Having sufficient neutron moderator available. 
The base case external criticality FEPs are presented in Table 6.3-2. The external FEPs define 
criticality configurations that begin with source terms resulting from the transport of fissile 
materials from the waste package in a form (either as solutes, colloids, or slurry of fine 
particulate) that can be transported into the drift invert (near-field) and beyond (far-field). 
ANL-EBS-NU-000008 REV 00 
FEP Description 
The waste package internal structures and the waste form remain 
intact. A breach (or breaches) in the waste package allow(s) water to 
either accumulate or flow-through the waste package. Criticality then 
occurs in situ. In-package criticality resulting from disruptive events is 
addressed in separate FEPs. 
The waste package internal structures and the waste form degrade. A 
critical configuration (sufficient fissile material and neutron moderator, 
lack of neutron absorbers) develops and criticality occurs in situ. 
Potential in situ critical configurations are defined in Figures 3.2a and 
3.2b of Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). In-package criticality resulting from 
disruptive events is addressed in separate FEPs. 
68 of 176 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
FEP Description 
Table 6.3-2. Base Case Configurations: External (Near-Field and Far-Field) Criticality FEPs 
Near-field criticality occurs when fissile material-bearing solution from 
the waste package is transported into the drift and the fissile material 
is precipitated into a critical configuration. Potential near-field critical 
configurations are defined in Figure 3.3a of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). 
In-package criticality resulting from disruptive events is addressed in 
separate FEPs. 
Far-field criticality occurs when fissile material-bearing solution from 
the waste package is transported beyond the drift and the fissile 
material is precipitated into a critical configuration. Potential far-field 
critical configurations are defined in Figure 3.3b of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). 
In-package criticality resulting from disruptive events is addressed in 
separate FEPs. 
FEP Number 
Source: Table 6.1-4 
February 2004 69 of 176 
FEP Name 
2.1.14.17.0A Near-field criticality 
2.2.14.09.0A Far-field criticality 
Retention and release of key isotopes from degrading commercial SNF waste packages is 
discussed in Geochemistry Model Abstraction and Sensitivity Studies for the 21 PWR CSNF 
Waste Packages (BSC 2002 [DIRS 160638], Sections 4.1.2.1 and 4.1.2.2). The process of 
determining external criticality potential starts with a source term, which is the waste package 
effluent water containing soluble radionuclides including fissile material. A significant fraction 
of the fissile material in the source term can be precipitated in the near-field and far-field by 
mixing the effluent water with diverted water at near-neutral pH levels that has flowed around 
the drift. This requires a simultaneous large flow rate around the waste package and a nominal 
flow rate through the waste package. 
The results of Geochemistry Model Abstraction and Sensitivity Studies for the 21 PWR CSNF 
Waste Packages (BSC 2002 [DIRS 160638]) show that the commercial SNF waste form is very 
durable being well protected by the Zircaloy cladding and very insoluble in the relatively weak 
chemistry of commercial SNF waste packages. Assuming an early waste package failure results 
in an advective flow path into the waste package, it is expected that a very insignificant amount 
(less than 0.5 percent) of the uranium and plutonium in commercial SNF waste packages will be 
released into solution and removed from the waste package during the performance period. This 
release value is applicable to either bathtub or flow-through waste package configurations. In 
addition, this result was calculated assuming that up to 10 percent of the fuel is exposed (failed) 
at the time of waste package breach. If 100 percent fuel exposure is assumed, the waste package 
is predicted to fill with corrosion products in less than 1,000 years, thereby preventing the 
migration of fissile material external to the waste package. 
Based on 0.469 MTU per assembly (Punatar 2001 [DIRS 155635], Table 3-1), 0.5 percent 
corresponds to less than 50 kg of low enriched uranium (less than 5 weight percent uranium-235) 
for a fully loaded 21-PWR waste package. Due to the large plume (both in depth and volume) 
associated with the accumulation of uranium-oxide fuels in the near-field and far-field 
environments (BSC 2002 [DIRS 159913], Section 5.2.2) and the presence of tuff throughout the 
accumulation zone, 50 kg of low enriched uranium cannot form a critical configuration in the 
external environment. 
ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
6.3.3 SAPHIRE Basic Event Probabilities 
The criticality FEPs screening analysis is an application of the Configuration Generator Model 
for In-Package Criticality (BSC 2003 [DIRS 165629]). Potential critical configurations are a 
result of waste package bathtub configurations in which the waste package internals degrade 
faster than the waste form and the neutron absorber material is flushed from the waste package. 
The four events and processes required to achieve a bathtub configuration are listed as top events 
in Figure 6.2-3. These four events are: (1) the probability that sufficient seepage flux is available 
to degrade the waste package internals and flush out the neutron absorber materials (top event 
MS-IC-1); (2) the probability of drip shield failure (top event MS-IC-2); (3) the probability of 
waste package failure (top event MS-IC-3); and (4) the probability of achieving and maintaining 
a bathtub configuration during the performance period (top event MS-IC-4). 
Each of these top events is quantified by a corresponding fault tree. Figures of these fault trees 
are provided in Attachment II (Figures II-5 through II-8). Each fault tree has several basic 
events connected by logic gates to accurately account for the relationships between the events in 
order to calculate the probability of occurrence of that top event. The following sections present 
the justification of the probability assignments to the basic events necessary to evaluate the base 
case internal criticality FEPs. 
6.3.3.1 Top Event MS-IC-1 
Water reaching the drift is an important factor in waste package degradation and criticality 
potential. It is associated with top event MS-IC-1 of the criticality FEPs analysis event tree 
(Figure II-4) and its associated fault tree inputs (Figures II-5). Two parameters characterize the 
seepage into the emplacement drifts  the seepage fraction (location within the drifts that see 
seepage) and the seepage flux rate (the volume of water entering the drift on an annual basis). 
The process for evaluating these parameters is presented in Attachment IV. The seepage 
evaluation incorporates the seepage fraction parameter into the calculation of determining the 
probability of achieving the minimum (or greater) seepage flux required. The seepage basic 
event and its value used to quantify fault tree MS-IC-1 is presented in Table 6.3-3. The 
justification for its value assignment is discussed in the Section 6.3.3.1.1. 
Table 6.3-3. Base Case SAPHIRE Basic Event Assignment for Fault Tree MS-IC-1 
Input Data Description 
Input Data Probability 
(per waste package for all 
waste package types) 
70 of 176 
0.0 
Justification 
Section 6.3.3.1.1 
Assumption 5.1.6 
Assumption 5.1.7 
February 2004 
Sufficient water (infiltrate/condensate) 
reaches drift by 10,000 years 
BE-SEEPAGE-10K 
Minimum Seepage Flux 6.3.3.1.1 
The seepage flux rate entering the drift is based on the Weibull distributions calculated in 
Attachment IV. This seepage flux is independent of both the waste form and waste package. 
Therefore, the seepage flux can be used for all waste forms being analyzed and are applicable for 
all waste forms configuration classes. 
ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Although the seepage rate into the drift is the same for all waste forms, the probability values 
assigned to the minimum required seepage for the various waste forms will differ. This is 
because these probabilities are determined from the requirement that the seepage rates be 
sufficient to permit the development of particular configuration classes (degradation of the waste 
package internals and removal of the neutron absorber materials from the waste package) within 
a specified time (i.e., 10,000 years). The minimum required seepage rate is also dependent upon 
the size and location of the flow path from the drift overhead into the waste package. This 
requires evaluation of drip shield and waste package failure mechanisms and their resulting 
failure areas. For the calculation of the minimum required seepage, the neutron absorber 
materials present in the evaluated waste packages are at their specified design mass 
(Assumption 5.1.3). 
The probability of having sufficient seepage into a waste package is based on many factors (time 
step of interest, breach size of the drip shield and waste package, evaporation rate, and 
condensation rate), all of which play an important part in determining the probability of the 
seepage flux. 
However, the probability of attaining the minimum seepage flux for the base case criticality 
FEPs is calculated to be zero because the probability of drip shield failure is determined to be 
zero for base case conditions (refer to Sections 6.3.3.2). For base case criticality FEPs, this 
result is applicable for all waste form/waste package types. Basic event BE-SEEPAGE-10K in 
the MS-IC-1A fault tree will be utilized for all waste package types and will be assigned a value 
of 0.0. 
6.3.3.2 Top Event MS-IC-2 
Water passing through the drip shield to the waste package is an important factor in waste 
package degradation and criticality and is associated with top event MS-IC-2 of the criticality 
FEP analysis event tree (Figure II-4) and its associated fault tree (Figure II-6). Water can reach 
the waste package along two primary pathways  water dripping from the drift crown through a 
failed drip shield and water dripping from the underside of a drip shield due to 
evaporation/condensation. The first pathway can occur if the dripping water from the drift 
overhead passes through a failed area on the drip shield onto the waste package. The second 
pathway can occur if water condenses on the underside of a drip shield and falls onto the waste 
package. However, as discussed previously in Section 6.3.1, evaporation and condensation are 
not considered in this evaluation as the TSPA-LA condensation model is not yet available. 
Water pathways through the drip shield can be created by corrosion and/or gaps caused by the 
drip shield response to events such as seismic activity and emplacement errors. Drip shield 
failures can be categorized as being caused by either time-dependent or time-independent 
mechanisms. Corrosion failure mechanisms are time-dependent and may be active or inactive 
during the performance evaluation period. 
Time-independent drip shield failure mechanisms are defined as those failure mechanisms that 
can occur randomly from the time of initial emplacement. Drip shield emplacement errors, 
rockfall, or seismic events are types of time-independent failure mechanisms that can potentially 
result in immediate creation of an advective pathway through the drip shield. In certain cases, 
February 2004 71 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
such as fabrication errors, the failure mechanism is an initiator that exacerbates corrosion (a 
time-dependent mechanism). 
The drip shield failure mechanisms identified in Configuration Generator Model for In-Package 
Criticality (BSC 2003 [DIRS 165629], Figures II-2 and II-3) are discussed in the remainder of 
this section. The intent of these discussions is to justify the basic event probability values used 
in the evaluation of the base case criticality FEPs. Drip shield failure is defined as those drip 
shield damage mechanisms that can result in an advective flow path through the drip shield and 
onto the waste package surface. Drip shield failure could be the result of a crack in the drip 
shield surface or from the catastrophic failure of the complete drip shield. As will be discussed, 
not all drip shield damage results in the failure of the drip shields primary function. 
The drip shield failure basic event probability values are presented in Table 6.3-4. The list of 
failure mechanisms developed for the configuration generator model is based on the available 
information from TSPA-SR. It should be noted that drip shield failure due to floor heave 
identified in Configuration Generator Model for In-Package Criticality (BSC 2003 [DIRS 
165629], Figure II-2) is not considered in this analysis. Floor heave was discredited as a drip 
shield failure mechanism in Engineered Barrier System Features, Events, and Processes (BSC 
2003 [DIRS 166464], Section 6.2.30). The BE-DS-FLOOR-HEAVE basic event has therefore 
been removed from the MS-IC-2 fault tree. 
Drip shield degrades due to general corrosion within 
10,000 years. 
BE-DS-GENCOR-10K 
Drip shield degrades due to crevice corrosion within 
10,000 years (form of localized corrosion). 
BE-DS-CREVICE-10K 
Drip shield degrades due to pitting corrosion within 
10,000 years (form of localized corrosion). 
BE-DS-PITTING-10K 
Drip shield degrades due to stress corrosion cracking 
within 10,000 years. 
BE-DS-SCC-10K 
Drip shield fails due to thermal expansion. 
BE-DS-THERM-EXPAN 
Drip shield failure due to seismic event. 
BE-DS-SEISMIC 
Table 6.3-4. SAPHIRE Basic Event Assignment for Fault Tree MS-IC-2 
Input Data Description Input Data Probability (per drip 
shield for all waste package types) 
0.0 Section 6.3.3.2.1 
0.0 Section 6.3.3.2.2 
0.0 Section 6.3.3.2.2 
0.0 Section 6.3.3.2.3 
0.0 Section 6.3.3.2.4 
0.0 Section 6.3.3.2.5 
0.0 Section 6.3.3.2.6 
0.0 Section 6.3.3.2.7 
0.0 Section 6.3.3.2.8 
Justification 
February 2004 
Drip shield emplacement error. 
BE-DS-EMPLACEMENT 
Drip shield fails due to fabrication error (early failure). 
BE-DS-FABRICATION 
72 of 176 
Drip shield failure due to rockfall of sufficient size. 
BE-DS-ROCK-FALL 
General Corrosion Failure of the Drip Shield 6.3.3.2.1 
This is a time-dependent drip shield failure mechanism. As stated in WAPDEG Analysis of 
Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317], Section 6.6.2 and 
ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Figure 37), the earliest failure of the drip shield due to general corrosion does not occur until 
after 10,000 years (approximately 47,500 years). However, these results cannot be referenced as 
they require confirmation by TSPA-LA (BSC 2003 [DIRS 161317], Section 1). 
It is assumed that TSPA-LA will show there are no general corrosion failures of the drip shield 
before 10,000 years (Assumption 5.1.1) and, therefore, the probability of drip shield failure due 
to general corrosion during the performance period is zero. The probability of basic event BEDS-
GENCOR-10K is, therefore, set to 0.0. 
Localized Corrosion Failure of the Drip Shield 6.3.3.2.2 
This is a time-dependent drip shield failure mechanism. As discussed in Section 6.3.3.2.5, drip 
shield fabrication errors can result in localized corrosion. General Corrosion and Localized 
Corrosion of the Drip Shield (BSC 2003 [DIRS 161236], Section 6.4.3) states that Localized 
corrosion of Titanium Grade 7 would not initiate in a repository-relevant environment 
Therefore, the probabilities of basic events BE-DS-PITTING-10K and BE-DS-CREVICE-10K 
are set to 0.0. 
Stress Corrosion Cracking Failure of the Drip Shield 6.3.3.2.3 
This is a time-dependent drip shield failure mechanism. As discussed in Section 6.3.3.2.5, drip 
shield fabrication errors can result in the formation of stress corrosion cracks. Stress Corrosion 
Cracking of the Drip Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural 
Material (BSC 2003 [DIRS 161234], Section 6.3.7) states that stress corrosion cracks are 
expected to fill with corrosion products or be plugged with precipitates such as carbonate. Stress 
corrosion cracks are expected to be sealed within a few hundred years if water flows through the 
cracks at the expected very low film flow rate. If the cracks are bridged by water, the sealing 
process may take several thousand years, but no flow occurs. Because of the high density of the 
crack plugging materials and the lack of a pressure gradient to drive water through the crack, the 
probability of flow through the plugged crack approaches zero. 
Given the very low flow rates through a stress corrosion crack in the drip shield for at most a few 
hundred years, it is concluded that stress corrosion cracking does not prevent the drip shield from 
fulfilling its primary role to keep water from contacting the waste packages. The probability of 
basic event BE-DS-SCC-10K is set to 0.0. 
Emplacement Error Failure of the Drip Shield 6.3.3.2.4 
This is a time-independent drip shield failure mechanism. The probability of a drip shield 
emplacement error is calculated in Analysis of Mechanisms for Early Waste Package/Drip Shield 
Failure, (BSC 2003 [DIRS 164475], Section 6.3.7) as having a median value of 6.0  10-6 per 
drip shield with an error factor of 4.7. The 5th percentile, the 95th percentile, and the mean 
values are calculated to be 1.3  10-6, 2.8  10-5, and 9.3  10-6, respectively. 
However, Analysis of Mechanisms for Early Waste Package/Drip Shield Failure (BSC 2003 
[DIRS 164475], Section 6.4.7) goes on to state that there are no credible consequences of a drip 
shield emplacement error. Because the gap between two adjacent drip shields improperly 
February 2004 73 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
interlocked is expected to be small, water from the drift is not expected to fall directly onto an 
underlying waste package. Because of the drip shield interlock geometry, water will most likely 
first hit the lower drip shields connecting plate and be diverted from the waste package surface. 
Therefore, although a drip shield emplacement error is probable, the drip shield failure area due 
to such an emplacement error is zero. Because the primary function of the drip shield (to prevent 
advective flow onto the waste package) is not compromised, the value of basic event BE-DSEMPLACEMENT 
is set to 0.0. 
Fabrication Error Failure of the Drip Shield 6.3.3.2.5 
This is a time-independent drip shield failure mechanism. Four drip shield fabrication errors are 
identified in Analysis of Mechanisms for Early Waste Package/Drip Shield Failure, (BSC 2003 
[DIRS 164475], Table 20) as having the potential to increase the susceptibility of the drip shield 
to stress corrosion cracking or localized corrosion. These fabrication errors are weld flaws, base 
metal flaws, improper heat treatment, and damage by mishandling. 
However, General Corrosion and Localized Corrosion of the Drip Shield (BSC 2003 [DIRS 
161236], Section 6.4.3) states that Localized corrosion of Titanium Grade 7 would not initiate 
in a repository-relevant environment In addition, Stress Corrosion Cracking of the Drip 
Shield, the Waste Package Outer Barrier, and the Stainless Steel Structural Material (BSC 2003 
[DIRS 161234], Section 6.3.7) states that stress corrosion cracks are expected to fill with 
corrosion products or be plugged with precipitates such as carbonate. Stress corrosion cracks are 
expected to be sealed within a few hundred years if water flows through the cracks at the 
expected very low film flow rate. If the cracks are bridged by water, the sealing process may 
take several thousand years, but no flow occurs. Because of the high density of the crack 
plugging materials and the lack of a pressure gradient to drive water through the crack, the 
probability of flow through the plugged crack approaches zero. 
Since neither localized corrosion or stress corrosion cracking will result in an advective flow area 
through the drip shield, the drip shield failure area associated with drip shield fabrication errors 
is zero. Therefore, because the primary function of the drip shield is not compromised the value 
of basic event BE-SE-FABRICATION is set to 0.0. 
Thermal Expansion Failure of the Drip Shield 6.3.3.2.6 
This is a time-independent drip shield failure mechanism. As stated in EBS Radionuclide 
Transport Abstraction (BSC 2003 [DIRS 166466], Section 6.3.1.3), Thermal expansion will 
produce minor structural response in relation to the potential slippage or overlap between 
adjacent drip shields for the as-emplaced drip shield configuration. This mechanism has 
therefore been screened out from the TSPA-LA Therefore, the probability of basic event BEDS-
THERM-EXPAN is set to 0.0. 
Seismic Failure of the Drip Shield 6.3.3.2.7 
This is a time-independent drip shield failure mechanism. Seismic failures of the drip shield are 
not considered during the base case criticality FEPs analysis. This failure mechanism is only 
February 2004 74 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
considered during the evaluation of the seismic disruptive event criticality FEPs analysis 
(Section 6.4). Therefore, the probability of basic event BE-DS-SEISMIC is set to 0.0. 
Rockfall Failure of the Drip Shield 6.3.3.2.8 
This is a time-independent drip shield failure mechanism. Rockfall failures of the drip shield are 
not considered during the base case criticality FEPs analysis. This failure mechanism is only 
considered during the evaluation of the rockfall disruptive event criticality FEPs (Section 6.5). 
Therefore, the probability of basic event BE-DS-ROCK-FALL is set to 0.0. 
It should be noted that rockfall damage to the drip shield due to a seismic event is accounted for 
in the BE-DS-SEISMIC1 basic event during the seismic initiating event evaluation presented in 
Section 6.4. 
6.3.3.3 Top Event MS-IC-3 
The ability for water to enter a failed waste package is an important factor in waste package 
degradation and criticality and is associated with top event MS-IC-3 of the criticality FEP 
analysis event tree (Figure II-4) and its associated fault tree (Figures II-7). Water pathways into 
the waste package can be created by corrosion and/or failures caused by the waste package 
response to events such as seismic activity and fabrication errors. Waste package failures can be 
categorized as being caused by either time-dependent or time-independent mechanisms. 
Corrosion failure mechanisms are time-dependent and may be active or inactive during the 
performance evaluation period. 
Time-independent waste package failure mechanisms are defined as those failure mechanisms 
that can occur randomly from the time of initial emplacement. A seismic event is a type of timeindependent 
failure mechanism that can potentially result in immediate creation of an advective 
pathway into the waste package. In certain cases, such as fabrication errors, the failure 
mechanism is an initiator that exacerbates corrosion (a time-dependent mechanism). 
The waste package failure mechanisms identified in Configuration Generator Model for In- 
Package Criticality (BSC 2003 [DIRS 165629], Figure II-4) are discussed in the remainder of 
this section. The intent of these discussions is to justify the basic event probability values used 
in the evaluation of the base case criticality FEPs. Waste package failure is defined as those 
waste package damage mechanisms that can result in an advective flow path into the waste 
package. Waste package failure could be the result of a crack in the waste package surface or 
from the catastrophic failure of the complete waste package. As will be discussed, not all waste 
package damage results in the failure of the waste packages primary function. The basic event 
probability values for waste package failure are presented in Table 6.3-5. 
0.0 Section 6.3.3.3.1 
Table 6.3-5. SAPHIRE Basic Event Assignment for Fault Tree MS-IC-3 
Input Data Description 
Waste package degrades due to general corrosion 
Input Data Probability (per waste 
package for all waste package types) 
75 of 176 
within 10,000 years. 
BE-WP-GENCOR-10K 
ANL-EBS-NU-000008 REV 00 
Justification 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Input Data Description 
Waste package degrades due to crevice corrosion 
within 10,000 years (form of localized corrosion). 
BE-WP-CREVICE-10K 
Waste package degrades due to pitting corrosion 
within 10,000 years (form of localized corrosion). 
BE-WP-PITTING-10K 
Waste package degrades due to stress corrosion 
cracking within 10,000 years. 
BE-WP-SCC-10K 
Early failure of waste package. 
BE-WP-EARLY-F 
Justification 
Section 6.3.3.3.4 
February 2004 
Input Data Probability (per waste 
package for all waste package types) 
0.0 Section 6.3.3.3.2 
0.0 Section 6.3.3.3.2 
0.0 Section 6.3.3.3.3 
2.8  10-5 
0.0 Section 6.3.3.3.5 
76 of 176 
Waste package failure due to seismic event. 
BE-WP-SEISMIC 
General Corrosion Failure of the Waste Package 6.3.3.3.1 
This is a time-dependent waste package failure mechanism. As stated in WAPDEG Analysis of 
Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317], Section 6.6.2 and 
Figure 36), the earliest patch failure of the waste package due to general corrosion does not occur 
until after 10,000 years (approximately 120,000 years). However, this information cannot be 
referenced as it requires confirmation by TSPA-LA (BSC 2003 [DIRS 161317], Section 1). 
It is assumed that TSPA-LA will show that there are no general corrosion failures of the waste 
package before 10,000 years (Assumption 5.1.1) and, therefore, the probability of waste package 
failure due to general corrosion during the performance period is zero. The probability of basic 
event BE-WP-GENCOR-10K is, therefore, set to 0.0. 
Localized Corrosion Failure of the Waste Package 6.3.3.3.2 
This is a time-dependent waste package failure mechanism. As stated in WAPDEG Analysis of 
Waste Package and Drip Shield Degradation (CRWMS M&O 2000 [DIRS 151566], 
Section 6.5.1), localized corrosion does not initiate for the waste package because the exposure 
conditions on the waste package surface are not severe enough. However, because this 
information was initially developed for TSPA-SR, it cannot be referenced as it requires 
confirmation by TSPA-LA. 
TSPA-LA results will show that there are no localized corrosion failures of the waste package 
before 10,000 years (Assumption 5.1.1) and, therefore, the probability of waste package failure 
due to localized corrosion during the performance period is zero. The probabilities of basic 
events BE-WP-PITTING-10K and BE-WP-CREVICE-10K are therefore set to 0.0. 
Stress Corrosion Cracking Failure of the Waste Package 6.3.3.3.3 
This is a time-dependent waste package failure mechanism. As stated in WAPDEG Analysis of 
Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317], Section 6.6.2 and 
Figure 38), the earliest crack failure of the waste package due to stress corrosion cracking does 
not occur until after 10,000 years (approximately 120,000 years). However, this information 
ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
cannot be referenced as it requires confirmation by TSPA-LA (BSC 2003 [DIRS 161317], 
Section 1). 
It is assumed that TSPA-LA will show that there are no stress corrosion cracking failures of the 
waste package before 10,000 years (Assumption 5.1.1) and, therefore, the probability of waste 
package failure due to stress corrosion cracking during the performance period is zero. The 
probability of basic event BE-WP-SCC-10K is, therefore, set to 0.0. 
Early Failure of the Waste Package 6.3.3.3.4 
This is a time-independent waste package failure mechanism. Four waste package fabrication 
errors are identified in Analysis of Mechanisms for Early Waste Package/Drip Shield Failure, 
(BSC 2003 [DIRS 164475], Table 20) as having the potential to increase the susceptibility of the 
waste package to stress corrosion cracking or localized corrosion. These fabrication errors are 
weld flaws, improper heat treatment, improper laser peening, and damage by mishandling. 
After ultrasonic testing inspection, the mean probability of the occurrence of one or more weld 
flaws in the upper and middle closure lids is 0.18 and 0.20, respectively (BSC 2003 [DIRS 
164475], Table 13). For the waste package seam weld, the mean probability increases to 0.46 
(BSC 2003 [DIRS 164475], Table 13). The residual stresses/stress intensity factors resulting 
from weld flaws may induce stress corrosion cracking. However, as noted in Section 6.3.3.3.3, 
the earliest crack failure of the waste package due to stress corrosion cracking is not predicted to 
occur until after 10,000 years (approximately 120,000 years) (BSC 2003 [DIRS 161317], 
Figure 38). As noted in Section 6.3.3.3.3, it is assumed that TSPA-LA will confirm the results of 
evaluations of waste package failures due to stress corrosion cracking as required by WAPDEG 
Analysis of Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317], Section 1) 
(Assumption 5.1.1). 
The probability of improper heat treatment has been combined with the probabilities of improper 
laser peening and damage by mishandling. From the information presented in Table 4.1-5, this 
event has been calculated to have a median value of 7.2  10-6 per waste package with an error 
factor of 15. The mean value has been calculated to be 2.8  10-5 per waste package. The 
probability of having at least one waste package early failure in the repository due to improper 
heat treatment has been calculated to be 0.17 (BSC 2003 [DIRS 161317], Table 46). An average 
of 1.8 waste package failures are calculated from the information provided in Table 47 of 
WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317]). 
Recommendations from Analysis of Mechanisms for Early Waste Package/Drip Shield Failure 
(BSC 2003 [DIRS 164475], Section 6.4.8) include that the entire waste package surface should 
be considered affected by a waste package early failure due to an improper heat treatment. 
Based on the information above, the basic event value for waste package early failure, BE-WPEARLY-
F, is set to 2.8  10-5. This is appropriate since the SAPHIRE event tree evaluations are 
performed on a per waste package basis. 
February 2004 77 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seismic Failure of the Waste Package 6.3.3.3.5 
This is a time-independent waste package failure mechanism. Seismic failures of the waste 
package are not considered during the base case criticality FEPs analysis. This failure 
mechanism is only considered during the evaluation of the seismic disruptive event criticality 
FEPs (Section 6.4). Therefore, the probability of basic event BE-WP-SEISMIC is set to 0.0. 
6.3.3.4 Top Event MS-IC-4 
Water accumulating in a waste package is associated with top event MS-IC-4 of the criticality 
FEPs analysis event tree (Figure II-4) and its associated fault tree (Figure II-8). The parameters 
associated with the formation of a waste package bathtub configuration are the likelihood that the 
waste package failure locations will support a bathtub formation and that the waste package 
degradation processes will maintain this configuration for a sufficient time period to degrade the 
waste package internals, flush the neutron absorber materials from the waste package, and allow 
for the generation of a potentially critical configuration. The process for evaluating these 
parameters is presented in Configuration Generator Model for In-Package Criticality (BSC 2003 
[DIRS 165629], Section 6.6.6). The basic events necessary to quantify fault tree MS-IC-4 are 
summarized in Table 6.3-6. The justification for their value assignment is discussed in the 
remainder of this section. 
Probability of being in a bathtub at 10,000 years for 
general corrosion waste package failures 
BE-BATHTUB-10K 
Probability of being in a bathtub at 10,000 years for 
nongeneral corrosion waste package failures 
BE-BATHTUB-10K 
Bathtub configuration formed for general corrosion 
BE-MS-IC-4 
Table 6.3-6. SAPHIRE Basic Event Assignment for Fault Tree MS-IC-4 
Input Data Description Input Data Probability (per waste 
package for all waste package types) 
0.0 Section 6.3.3.4.1 
1.0 Section 6.3.3.4.1 
0.0 Section 6.3.3.4.2 
1.0 Section 6.3.3.4.2 
78 of 176 
Justification 
February 2004 
Bathtub configuration formed for nongeneral 
corrosion waste package failures 
BE-MS-IC-4 
Duration of Bathtub Configuration 6.3.3.4.1 
A bathtub configuration is when a breach, or failure, on the top part of the waste package occurs 
prior to a breach or failure on the bottom part. The duration of flooding conditions, which last as 
long as the bottom surface is intact, is a function of the waste package failure mechanisms, 
evaluated as: 
 Start time for accumulating water = First failure time at top of outer barrier 
 Start time for flow-through geometry = First failure time at bottom of outer barrier 
 Bathtub Duration Time = Start time for flow-through geometry  Start time for 
accumulating water. 
ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Possible closure of the bottom failures converting a flow-through geometry into a bathtub 
arrangement are considered to be very unlikely since a second patch failure is likely to occur 
within a relatively short period (CRWMS M&O 2000 [DIRS 151566], Figure 24). 
Bathtub configuration duration starts at the time of first breach of the waste package and stops at 
the occurrence of a breach of the waste package bottom or at the end of the performance period, 
whichever comes first. For waste package failures resulting from general corrosion, the earliest 
failure of the waste package occurs at 120,000 years (BSC 2003 [DIRS 161317], Section 6.6.2 
and Figure 36). Since the first waste package failure does not occur until substantially after the 
end of the performance period, the probability of maintaining a bathtub configuration during the 
performance period for general corrosion waste package failures is 0.0. This value is assigned to 
basic event BE-BATHTUB-10K. It should be noted that, as stated in Section 6.3.3.3.1, the 
information from WAPDEG Analysis of Waste Package and Drip Shield Degradation (BSC 
2003 [DIRS 161317]) requires confirmation by TSPA-LA (BSC 2003 [DIRS 161317], 
Section 1). It is assumed (Assumption 5.1.1) that TSPA-LA will confirm general corrosion of the 
waste package outer barrier does not occur until well after 10,000 years. 
However, no information is available for the duration of bathtub configurations resulting from 
any waste package failure mechanism other than general corrosion. It is conservative to assume 
that if a bathtub configuration is formed, it will endure for the remainder of the performance 
period (Assumption 5.2.1) as this will maximize the BATHTUB-CONFIG end state probability 
of Figure 6.2-3. Therefore, for nongeneral corrosion waste package failures, basic event BEBATHTUB-
10K is assigned a value of 1.0. 
Probability of Bathtub Configuration Formation 6.3.3.4.2 
The first failure location on the waste package can occur either on its top or bottom. Only waste 
package top failures can result in the formation of a bathtub configuration. The probability of 
bathtub configuration formation has been previously calculated based on the results of WAPDEG 
Analysis of Waste Package and Drip Shield Degradation (CRWMS M&O 2000 [DIRS 
151566]). But this evaluation is only applicable to waste package failures due to general 
corrosion. However, since the currently calculated mean first failure of the waste package due to 
general corrosion does not occur until 120,000 years (BSC 2003 [DIRS 161317], Figure 36), the 
probability of forming a bathtub configuration during the performance period as a result of 
general corrosion is 0.0. Therefore, basic event BE-MS-IC-4 is assigned a value of 0.0 for 
general corrosion waste package failure mechanisms. 
It should be noted that, as stated in Section 6.3.3.3.1, the information from WAPDEG Analysis of 
Waste Package and Drip Shield Degradation (BSC 2003 [DIRS 161317]) requires confirmation 
by TSPA-LA (BSC 2003 [DIRS 161317], Section 1). It is assumed (Assumption 5.1.1) that 
TSPA-LA will confirm general corrosion of the waste package outer barrier does not occur until 
well after 10,000 years. 
No information is available for the formation of bathtub configurations resulting from any waste 
package failure mechanism other than general corrosion. It is conservative to assume that if an 
event, such as a seismic event, results in damage to the waste package, this damage occurs on the 
top of the waste package (Assumption 5.2.1) and a bathtub configuration is formed. This 
February 2004 79 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
assumption will maximize the BATHTUB-CONFIG end state probability of Figure 6.2-3. 
Therefore, basic event BE-MS-IC-4 is assigned a value of 1.0 for all waste package failure 
mechanisms other than general corrosion. 
6.3.4 Base Case Criticality FEPs Analysis Results 
6.3.4.1 In-Package Results 
The probabilities of waste package flooding at the end of the performance period are shown in 
Table 6.3-7. These probability results have been generated to address in-package criticality 
FEPs. Because there is no mechanism to breach the drip shield for these base case criticality 
FEPs during the performance period, there is no probability of water entering the waste package 
and generating a bathtub configuration. Therefore, the probability of waste package flooding is 
zero and there is no probability of criticality for the in situ criticality FEPs presented in 
Table 6.3-1. 
Table 6.3-7. Per Waste Package Flooding Probabilities for Base Case Criticality FEPs 
Per Waste Package Flooding 
Probabilityb 
0.00E+00 
0.00E+00 
0.00E+00 
0.00E+00 
0.00E+00 
3,412c 0.00E+00 
b SAPHIRE V7.18 (BSC 2002 [DIRS 160873]) analysis results (Attachment II, p. II-20) 
c Sum of 5-DHLW/DOE SNF Short, 5DHLW/DOE SNF Long and 2-MCO/2-DHLW Long values from Table 4.1-8 
External criticality FEPs can originate from either a bathtub or flow-through waste package 
configuration. However, because there is no mechanism to breach the drip shield during the 
performance period under base case conditions, there is no probability of water entering the 
waste package. Therefore, there is no probability of criticality for the external criticality FEPs 
presented in Table 6.3-2. 
6.4 ANALYSIS OF SEISMIC DISRUPTIVE EVENT CRITICALITY FEPS 
The seismic disruptive event criticality FEPs are presented in Table 6.4-1. 
February 2004 
Number of Waste Packagesa 
4,299 
95 
163 
2,831 
84 
External Probability Results 
80 of 176 
Waste Package Type 
21-PWR with Absorber Plates 
21-PWR with Control Rods 
12-PWR Long 
44 BWR 
24 BWR 
DOE SNF 
Source: a Values from Table 4.1-8 
6.3.4.2 
ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
FEP Number 
2.1.14.18.0A 
2.1.14.19.0A 
2.1.14.20.0A 
2.2.14.10.0A 
Source: Table 6.1-4 
Table 6.4-1. Seismic Disruptive Event Criticality FEPs 
FEP Title 
In-package 
criticality resulting 
from a seismic 
event (intact 
configuration) 
In-package 
criticality resulting 
from a seismic 
event (degraded 
configurations) 
Near-field 
criticality resulting 
from a seismic 
event 
Far-field criticality 
resulting from a 
seismic event 
Vibratory ground motion and rockfall induced by a seismic event can cause damage to the drip 
shield and waste package allowing the influx of seepage into the waste package, which has the 
potential to cause a criticality. A seismic event can also induce fault displacement, which can 
lead to damage of the drip shield and waste package that can also allow advective flow into the 
waste package and lead to a potential criticality. Additionally, new fractures that intersect the 
drift segments and the collapsing of the drift as a result of a seismic event would have an affect 
on the seepage water. This change in seepage water onto damaged waste packages may increase 
their potential for criticality. Table 6.4-1 presents the seismic disruptive event criticality 
FEPs 2.1.14.18.0A, 2.1.14.19.0A, 2.1.14.20.0A, and 2.2.14.10.0A, which may initiate a sequence 
of events that can lead to a potential critical event. The direct and indirect effects of seismic 
activities on in-package criticality, near field criticality, and far field criticality are analyzed in 
this section. 
Because uncertainty is an important part of any analysis, it is included in the seismic evaluation 
of potential in-package criticality. Uncertainty is included throughout the evaluation by the 
development of probability distributions sampled via a Latin Hypercube Sampling method. The 
principle of Latin Hypercube Sampling is provided by Modarres (1993 [DIRS 104667], p. 244). 
The developed probability distributions represent the epistemic uncertainty for the parameters of 
interest. An example is the damaged area of a drip shield depending upon the PGV of the 
seismic event. The analysis will develop a probability distribution representing the epistemic 
uncertainty about the damaged area of the drip shield and then sample this distribution to obtain 
the damaged area based on the seismic event. 
The developed Latin Hypercube Sampling method evaluates the epistemic uncertainty of all 
input parameters either developed within the evaluation (e.g., probability distribution for 
ANL-EBS-NU-000008 REV 00 
The waste package internal structures and the waste form remain intact 
either during or after a seismic disruptive event. A breach (or breaches) in 
the waste package allow(s) water to either accumulate or flow-through the 
waste package. Criticality then occurs in situ. 
Either during, or as a result of, a seismic disruptive event, the waste 
package internal structures and the waste form degrade. A critical 
configuration develops and criticality occurs in situ. Potential in situ critical 
configurations are defined in Figures 3.2a and 3.2b of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). 
Either during, or as a result of, a seismic disruptive event, near-field 
criticality occurs when fissile material-bearing solution from the waste 
package is transported into the drift and the fissile material is precipitated 
into a critical configuration. Potential near-field critical configurations are 
defined in Figure 3.3a of Disposal Criticality Analysis Methodology Topical 
Report (YMP 2003 [DIRS 165505]). 
Either during, or as a result of, a seismic disruptive event, far-field criticality 
occurs when fissile material-bearing solution from the waste package is 
transported beyond the drift and the fissile material is precipitated into a 
critical configuration. Potential far-field critical configurations are defined in 
Figure 3.3b of Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). 
81 of 176 
FEP Description 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
damaged area of drip shield based on PGV of seismic event) or based on other reports. An 
example of an external parameter with its epistemic uncertainty accounted for in the Latin 
Hypercube Sampling method would be the seepage rate. By using the Latin Hypercube 
Sampling method, uncertainty is handled in the seismic evaluation of in-package criticality. 
It should be noted, in addition to the assumption referenced directly by this section, several 
assumptions from the supporting documents used in the development of the seismic disruptive 
event criticality FEPs analysis are listed in Section 5.2.3. Although not referenced directly by 
this analysis, these assumptions may influence the reported results. 
6.4.1 Seismic Ground Motion Effects on In-Package Criticality Evaluations 
A seismic event has the potential to lead to a critical event by causing damage to the drip shield 
and waste package, which can allow advective flow to penetrate the damaged waste package. 
Water moderation is an important factor that is required for criticality. This section will use the 
information from Seismic Consequence Abstraction (BSC 2003 [DIRS 161812]), Boron Loss 
from CSNF Waste Packages (BSC 2003 [DIRS 165890]) and Abstraction of Drift Seepage (BSC 
2003 [DIRS 165564]) to evaluate the probability of sufficient seepage penetrating a damaged 
waste package, which can lead a criticality. 
Base Case Seismic Evaluation 6.4.1.1 
The following sections are based on the nominal design of the commercial SNF waste packages. 
This part of the analysis is deemed base case. 
Seismic Effects in the Lithophysal Zone 6.4.1.1.1 
A seismic event can affect the drip shield, waste package, and cladding from vibratory ground 
motion and induced rockfall. A seismic event can also affect the seepage of water into the drift 
from new fractures or the collapsing of the drift. Seepage is an important factor that can lead to a 
waste package criticality. Seepage is required to degrade and flush out the neutron absorbing 
material and provide moderation. In order for seepage to penetrate the waste package and 
potentially lead to a criticality, multiple barriers must be breached. A seismic event can breach 
these barriers, which are the drip shield and the waste package outer barrier. Breaching (i.e., 
percent damaged area) of a drip shield and waste package due to vibratory ground motion will be 
analyzed below. In addition, the development and use of seepage rate distributions along with 
the evaluation of the degradation and flushing of the neutron absorbing material from a damaged 
waste package. 
The drip shield failure for this analysis is divided into two separate repository geological zones- 
nonlithophysal and lithophysal-require a separate evaluation because of the different effects 
induced by a seismic event on a drip shield. In addition, seepage water is affected within each of 
these zones because of drift fracturing or collapse. The lithophysal zone represents 
approximately 85 percent of the total repository drift area (refer to Section 6.2). 
The damage to a waste package is the same for both repository drift zones because rockfall 
cannot impact the waste package, since the drip shield protects it. Therefore, waste package 
damage is independent of location within the repository. 
February 2004 82 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Drip Shield and Waste Package Damage 
Drifts in the lithophysal zone are expected to collapse during a seismic event and the void area 
between the drip shield and the drift area to become filled. The collapse of the drift in this area 
does not damage the drip shield because the rock type is very low in compressive strength and is 
permeated with void spaces (BSC 2003 [DIRS 161812], Section 6.6.2). This weak rock is 
expected to collapse into small fragments under the load imposed by the vibratory ground 
motion. Any damage to the drip shield in this zone is expected to occur only from the vibratory 
ground motion. 
To account for the damage to the drip shield, Seismic Consequence Abstraction (BSC 2003 
[DIRS 161812], Section 6.6.3) developed an abstraction used to calculate the percent damaged 
area of the drip shield. A Mathcad spreadsheet is developed (Attachment III) to use the seismic 
inputs from Section 4.1.2 and account for the uncertainty in the percent damaged area of the drip 
shield by performing a Latin Hypercube Sampling process. 
The process to calculate the percent damaged area on the drip shield is outlined in Seismic 
Consequence Abstraction (DTN: MO0308SPACALSS.002 [DIRS 164822]). The process 
discusses sampling the mean annual exceedance frequency for a seismic event to obtain a PGV 
value. The mean annual seismic exceedance frequency follows a uniform distribution between 
1.0  10-8 to 1.0  10-4. The sampled mean annual seismic exceedance frequency is used to 
obtain the corresponding PGV value from log-linear interpolation of the lookup table (refer to 
Table 4.1-2). The interpolated PGV value is used to determine the upper and lower bounds of 
the uniform distribution representing the percent damaged area of the drip shield. 
The upper-bound percent damaged area is interpolated from Table 4.1-4 based on the sampled 
PGV value. This interpolated upper-bound value is input into a uniform distribution for the 
percent damaged area of the drip shield. The lower bound of the uniform distribution is also 
determined by interpolation. The lower-bound percent damaged area uses the lookup table 
shown in Table 4.1-3. The lower-bound percent damaged area is based on the same sampled 
PGV value. Once the lower and upper bounds of the uniform distribution are obtained, this 
distribution is then sampled to calculate the percent damaged area of the drip shield for that 
particular seismic event. This percent damaged area value is stored within the Mathcad 
spreadsheet. 
The process then repeats with a newly sampled mean seismic exceedance frequency. This newly 
sampled mean seismic exceedance frequency leads to a new percent damaged area of the drip 
shield based on that seismic event. This process is continued for 20,000 realizations. From these 
(refer to Attachment III, p. III-7), the mean fraction of damaged area on the drip shield (i.e., 
percent damage divided by 100) from sampled vibratory ground motions is 2.23  10-3 and the 
5th and 95th are 0.0 and 6.49  10-3, respectively. 
The waste package is also damaged due to vibratory ground motion in the lithophysal zone. The 
damage to the waste package is calculated in the same manner as that done for the drip shield. 
The only difference between these two calculations is how the upper-bound value for the 
uniform distribution representing the percent damaged area of the waste package is calculated. 
The upper-bound value for the uniform distribution is calculated using Equation 6.4-1. 
February 2004 83 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
= . 0 305 (Eq. 6.4-1) PGV 0.436  WPupperbound - 
The upper-bound value is calculated by inputting the sampled PGV value, which is obtained 
from log-linear interpolation of the lookup table (Table 4.1-2) based on the sampled mean annual 
seismic exceedance frequency. The lower-bound value for the uniform distribution is set to zero. 
This uniform distribution is then sampled to obtain the percent damaged area of the waste 
package due to that seismic event. As discussed for the drip shield, the process starts all over 
again by sampling a new mean annual seismic exceedance frequency, which is used to calculate 
a PGV value. This newly calculated PGV value is used to calculate the upper-bound value for 
the uniform distribution that represents the percent damaged area of the waste package. This 
newly created uniform distribution is then sampled to obtain the new percent damaged area of 
the waste package for that seismic event. This process continues for 20,000 realizations. From 
these, the mean fraction of damaged area on the waste package outer barrier (i.e., percent 
damage divided by 100) is 2.58  10-4 and the 5th and 95th percentiles are 0 and 1.47  10-3, 
respectively (Attachment III, p. III-9). 
The fraction of damaged area (i.e., percent damage divided by 100) calculated above occurs on 
all of the drip shields and waste packages in the lithophysal zone of the repository; therefore, the 
probability that a drip shield or waste package is damaged due to a seismic event is 1.0. It is 
assumed that this damaged area occurs on the top of the drip shield and waste package allowing 
advective flow through the drip shield and into the waste package, forming a bathtub 
configuration (Assumption 5.2.1). This probability is input into basic events BE-DS-SEISMIC1 
and BE-WP-SEISMIC1, which are substituted for basic events BE-DS-SEISMIC and BE-WPSEISMIC 
of fault trees MS-IC-2 and MS-IC-3, respectively. These fault trees are used in the 
SAPHIRE evaluation of the seismic disruptive event criticality FEPs. 
Although it is possible to have seismic induced damage on a waste package already damaged due 
to an early failure event, the probability of damage due to a seismic event is greater than that of 
an early failure. Therefore, the probability of damage from a seismic event (i.e., probability of 
1.0) overwhelms the probability (i.e., probability of 2.8  10-5) of damage occurring from an 
early failure. The damaged area on a waste package from a seismic event is smaller (BSC 2003 
[DIRS 161812], Section 6.6.3) than that recommended to be modeled for an early failure 
(BSC 2003 [DIRS 164475], Section 6.4.8), however, by including the probability of each event, 
damage to the waste package from early failure becomes negligible. For these reasons, basic 
event BE-WP-EARLY-F is replaced by basic event BE-WP-EARLY-F1 during the seismic 
initiating event evaluations. The probability of BE-WP-EARLY-F1 is set to 0.0. 
Seepage Rate Probability Distribution for Lithophysal 
The process used to determine the seepage rate distribution, which is used to calculate the 
seepage probability, follows process steps discussed in Abstraction of Drift Seepage (BSC 2003 
[DIRS 165564], Section 6.7.1). The determination of the seepage rate distribution is discussed 
below and presented in Attachment IV. 
In order to determine seepage rate distribution, a Latin Hypercube Sampling method was 
developed to handle spatial variability and uncertainty. The routine sampled each input for 
February 2004 84 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
20,000 realizations to ensure sufficient coverage of the parameter range. Three key parameters 
are sampled to determine the seepage rate distribution. 
The first parameter, capillary strength (1/ ), is determined to have a spatial variability that is 
uniformly distributed with a range between 402 Pa to 780 Pa, and a mean of 591 Pa. The 
uncertainty about the capillary strength, .1/ , follows a triangular distribution with a lower 
bound of .105 Pa, upper bound of +105 Pa, and a mean of 0.0. These distributions are identical 
for all geological zones. The Latin Hypercube Sampling method samples a capillary strength 
value from the spatial variability and adds it to the sampled capillary strength value from the 
uncertainty distribution. This calculated capillary strength is used in the interpolation process 
along with the other sampled key parameters to determine the seepage rate. This sampling 
process is performed for 20,000 realizations. 
The next key parameter for the lithophysal zone, permeability (k), is determined to have a spatial 
variability distribution that is lognormal with a mean of -11.5 (in log 10) and a standard 
deviation of 0.47(in log 10). The mean and standard deviation of permeability was determined 
from statistical analysis on the log-transformed data (BSC 2003 [DIRS 165564], 
Section 6.6.2.1). The permeability uncertainty ( .k) follows a triangular distribution with a lower 
bound of .0.92, upper bound +0.92, and a mean of 0.0. These distributions are for the 
lithophysal zone only. The Latin Hypercube Sampling method samples a permeability value 
from the spatial variability and adds it to the sampled permeability value from the uncertainty 
distribution. This calculated permeability is used in the interpolation process along with the 
other sampled key parameters to determine the seepage rate. This sampling process is performed 
for 20,000 realizations. 
Percolation flux is sampled from the percolation flux information that represents the repository 
area (BSC 2003 [DIRS 165564], Figure 6.6-10). The sampling process uses glacial transition 
climate percolation flux information, which occurs 2,000 years after repository closure and lasts 
through the regulatory period of 10,000 years (USGS 2001 [DIRS 158378], Section 6.6.1). The 
percolation flux uncertainty is expressed by three different scenarios (lower-bound, mean, and 
upper-bound). Since there are three different scenarios that are used to represent the uncertainty, 
three different final seepage rate distributions are obtained (one for each scenario). 
The percolation flux is adjusted for intermediate-scale heterogeneity by using flow focusing 
factors (Equation 6.4-2) (DTN: LB0104AMRU0185.012 [DIRS 163906]), which is sampled and 
multiplied by the sampled percolation flux. Equation 6.4-2 is the cumulative distribution 
function for the flow focusing factors where the variable x represents the flow focusing factor. 
2 3 4 434 ff = .0.3137x + 5.4998x . 35.66x +102.3x .11. (Eq. 6.4-2) 
The seepage rate for each of the uncertainty scenarios (i.e., lower-bound, mean, and upperbound) 
is determined using a sampling routine (refer to Attachment IV). The sampled value 
from the three key parameters (i.e., capillary strength, permeability, adjusted percolation flux) is 
used to interpolate the mean seepage rate and seepage rate standard deviation from the lookup 
table for the degraded drift (DTN: LB0307SEEPDRCL.002 [DIRS 164337]). The standard 
deviation is adjusted to account for uncertainty by creating a uniform distribution with a lower 
bound of .1.7321 times the sampled standard deviation and an upper bound of 1.7321 times the 
sampled standard deviation. The uniform distribution to account for uncertainty is sampled and 
February 2004 85 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
added to the interpolated mean seepage rate. This process is performed for 20,000 realizations 
(refer to Attachment IV). 
The resulting seepage rate values are adjusted prior to being used to determine the seepage flux 
probability by (1) setting seepage rates less than 0.1 kg/yr per waste package to zero (since these 
small values are the result of interpolation (BSC 2003 [DIRS 165564], p. 173)), and (2) setting 
calculated seepage rates greater than 100 percent to 100. 
Seepage rates are then filtered in order to develop a distribution that represents the seepage rate 
values by discarding all seepage rates with a zero value. The remaining nonzero seepage rates 
are then used to develop a Weibull distribution for each of the scenarios (i.e., lower-bound, 
mean, and upper-bound) to represent the seepage rate at the drift (refer to Attachment IV). These 
Weibull distributions are used to calculate the probability of having sufficient seepage in order to 
degrade and flush out the Neutronit from a seismically damaged waste package. In addition, to 
calculate the fraction of waste package locations with seepage, the number of nonzero seepage 
rates is divided by the total number of realizations (seepage fraction). The Weibull parameters, 
scale and shape ( and , respectively), and seepage fraction for each scenario are listed in 
Table 6.4-2. 
Table 6.4-2. Weibull Parameters and Seepage Fraction (Lithophysal Zone) 
Value 
Lower Bound (Drift Collapse) 
8.87E.03a (m3/yr) 
5.20E.01a 
Mean (Drift Collapse) 
1.46E.01c (m3/yr) 
4.68E.01c 
Weibull Parameters 
 (scale) 
 (shape) 
 (scale) 
 (shape) 
 (scale) 
 (shape) 
Source: a Attachment IV, p. IV-15 
ANL-EBS-NU-000008 REV 00 
Upper Bound (Drift Collapse) 
b Attachment IV, p. IV-14 
c Attachment IV, p. IV-28 
d Attachment IV, p. IV-27 
e Attachment IV, p. IV-41 
f Attachment IV, p. IV-40 
Boron Loss from Commercial SNF Waste Packages 
The loss of Neutronit from a commercial SNF waste package depends on the specific 
commercial SNF parameters (Assumption 5.1.2) and the time a seismic event occurs. The 
specific commercial SNF waste package parameters are input into the boron loss equation (i.e., 
loss of Neutronit), which is used to calculate the required drip rate (i.e., seepage rate m3/yr). The 
calculated required drip rate is based on the time available for Neutronit degradation and 
flushing. Since the time when a seismic event occurs is random, the boron loss equation is 
solved for each of the 20,000 realizations. Therefore, some of the commercial SNF inputs are 
not constant and vary based on the randomness of the seismic event. In addition, the time 
required to degrade and flush out the Neutronit is limited based on when water can first penetrate 
3.83E.01e (m3/yr) 
4.94E.01e 
86 of 176 
Seepage Fraction 
1.94E.01b 
5.14E.01d 
6.37E.01f 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
a damaged waste package, assumed to be 700 years after repository closure (Assumption 5.2.2). 
The boron loss equation used in this analysis to calculate the required drip rate is: 
( ) .t t f V D V 
f t .  
VR R R . N t t ) = (Eq. 6.4-3) exp 1 exp f B . . . 
.
. 
. 
where: 
( > 
and the corrosion rate distribution. 
   
.
. 
NB = boron in waste package (moles) 
t = time (years) (maximum time is the regulatory period of 10,000 years) 
tf = time when the Neutronit has fully degraded (years) 
VR = waste package void volume (liters) 
D = boron released from Neutronit (moles/year) 
 = volumetric flow rate (m3/year) 
Equation 6.4-3 is solved for , which is the minimum drip rate (i.e., seepage rate m3/yr) required 
to degrade and flush out the Neutronit from a commercial SNF waste package. The input 
parameters used in calculating the minimum drip rate (i.e., seepage rate m3/yr) depends on which 
commercial SNF waste package is analyzed. The commercial SNF waste package input 
parameters are discussed below. 
The input parameters (constant or random) for the 21-PWR with Absorber Plates Waste Package 
are listed in Table 6.4-3. Random parameters are based on the time when a seismic event occurs 
. 
.

87 of 176 February 2004 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 6.4-3. 21-PWR with Absorber Plates Waste Package Neutronit Degradation Parameters 
Variable Description 
t 
n 
b 
Bi 
VR 
k 
tdeg 
D 
ANL-EBS-NU-000008 REV 00 
Neutronit plate thickness 
Neutronit initially in waste 
package 
Boron content in Neutronit 
Initial boron in waste package 2.38E+03 moles Bi = (n ~ b)/AW (AW in Table 4.1-11) 
Void volume of waste 
package 
Degradation rate 
Time required to degrade all 
of the Neutronit 
Moles of boron released from 
Neutronit 
Value 
7 mm 
2.06E+06 grams 
1.245E.02 
weight fraction 
BSC 2003 [DIRS 163855], Table 2 
From Table 5.1-1 (Assumption 5.1.2) 
[(g per A-plate ~ A plates per waste package) + 
(g per B-plate ~ B plates per waste package) + 
(g per C-plate ~ C plates per waste package)] 
Average of lowest boron content (0.75 wt%) and 
highest content (1.74 wt%) from Table 4.1-11, 
divided by 100 
4.685E+03 From Table 5.1-1 (Assumption 5.1-2) 
k = crn ~  ~ conv 
(where: 
crn = corrosion rate of Neutronit (follows a Weibull 
Varies 
(g/cm2 ~ yr) 
distribution (see Assumption 5.1.4) 
 = density of Stainless Steel Type 316 
varies 
(years) 
(Table 4.1-11) 
conv = conversion factor 1.0E-04 cm/m) 
tdeg = (n/k ~ SA) + Ts 
(where: 
SA = surface area (Table 5.1-1 [Assumption 5.1.2]) 
Ts = time to seismic event greater than 700 
years [700 years based on Assumption 5.2.2]) 
D = Bi/tdeg varies 
(moles/year) 
88 of 176 
Source 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The input parameters (constant or random) for the 12-PWR Long Waste Package are listed in 
Table 6.4-4. Random parameters are based on the time when a seismic event occurs and the 
Value 
7 mm 
1.15E+06 grams 
1.245E.02 
weight fraction 
BSC 2003 [DIRS 163855], Table 8 
From Table 5.1-1 (Assumption 5.1.2) 
[(g per A-plate ~ A plates per waste package) + 
(g per B-plate ~ B plates per waste package) + 
(g per C-plate ~ C plates per waste package)] 
Average of lowest boron content (0.75 wt%) and 
highest content (1.74 wt%) from Table 4.1-11, 
divided by 100 
3.28E+03 From Table 5.1-1 (Assumption 5.1.2) 
k = crn ~  ~ conv 
(where: 
crn = corrosion rate of Neutronit (follows a Weibull 
varies 
(g/cm2 ~ yr) 
distribution (see Assumption 5.1.4) 
 = density of Stainless Steel Type 316 
(Table 4.1-11) 
varies 
(years) 
conv = conversion factor 1.0E-04 cm/m) 
tdeg = (n/k ~ SA) + Ts 
(where: 
SA = surface area (Table 5.1-1, Assumption 5.1.2) 
Ts = time to seismic event greater than 700 
years [700 years based on Assumption 5.2.2]) 
D = Bi/tdeg varies 
(moles/year) 
Source 
corrosion rate distribution. 
Variable Description 
t 
n 
b 
Bi 
VR 
k 
tdeg 
D 
Table 6.4-4. 12-PWR Long Waste Package Neutronit Degradation Parameters 
Neutronit plate thickness 
Neutronit initially in waste 
package 
Boron content in Neutronit 
Initial boron in waste package 1.33E+03 moles Bi = (n ~ b)/AW (AW in Table 4.1-11) 
Void volume of waste 
package 
Degradation rate 
Time required to degrade all 
of the Neutronit 
Moles of boron released from 
Neutronit 
ANL-EBS-NU-000008 REV 00 89 of 176 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The input parameters (constant or random) for the 44-BWR Waste Package are listed in Table 
6.4-5. Random parameters are based on the time when a seismic event occurs and the corrosion 
Value 
5 mm 
2.15E+06 grams 
1.245E.02 
weight fraction 
4.85E+03 
varies 
(g/cm2 ~ yr) 
varies 
(years) 
varies 
(moles/year) 
Source 
BSC 2003 [DIRS 163855], Table 3 
From Table 5.1-1 (Assumption 5.1.2) 
[(g per A-plate ~ A plates per waste package) + 
(g per B-plate ~ B plates per waste package) + 
(g per C-plate ~ C plates per waste package) 
(g per D-plate ~ D plates per waste package) 
(g per E-plate ~ E plates per waste package)] 
Average of lowest boron content (0.75 wt%) and 
highest content (1.74 wt%) from Table 4.1-11, 
divided by 100 
From Table 5.1-1 (Assumption 5.1.2) 
k = crn ~  ~ conv 
(where: 
crn = corrosion rate of Neutronit (follows a Weibull 
distribution (see Assumption 5.1.4) 
 = density of Stainless Steel Type 316 
(Table 4.1-11) 
conv = conversion factor 1.0E.04 cm/m) 
tdeg = (n/k ~ SA) + Ts 
(where: 
SA = surface area (Table 5.1-1, Assumption 5.1.2) 
Ts = time to seismic event greater than 700 
years [700 years based on Assumption 5.2.2]) 
D = Bi/tdeg 
rate distribution. 
Variable Description 
t 
n 
b 
Bi 
VR 
k 
tdeg 
D 
Table 6.4-5. 44-BWR Waste Package Neutronit Degradation Parameters 
Neutronit plate thickness 
Neutronit initially in waste 
package 
Boron content in Neutronit 
Initial boron in waste package 2.48E+03 moles Bi = (n ~ b)/AW (AW in Table 4.1-11) 
Void volume of waste 
package 
Degradation rate 
Time required to degrade all 
of the Neutronit 
Moles of boron released from 
Neutronit 
ANL-EBS-NU-000008 REV 00 90 of 176 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The input parameters for the 24-BWR Waste Package are listed in Table 6.4-6. The parameters 
are either constant or random. The random parameters are based on the time when a seismic 
event occurs and the corrosion rate distribution. 
t 
n 
b 
Bi 
VR 
k 
tdeg 
Table 6.4-6. 24-BWR Waste Package Neutronit Degradation Parameters 
Value 
10 mm 
2.63E+06 grams 
1.245E.02 
weight fraction 
2.70E+03 
varies (g/cm2yr) 
varies 
(years) 
varies 
(moles/year) 
Variable Description 
Neutronit plate thickness 
Neutronit initially in waste 
package 
Boron content in Neutronit 
package 
Degradation rate 
Time required to degrade all 
of the Neutronit 
Moles of boron released from 
Neutronit 
Source 
BSC 2003 [DIRS 163855], Table 9 
From Table 5.1-1 (Assumption 5.1.2) 
[(g per A-plate * A plates per waste package) + 
(g per B-plate * B plates per waste package) + 
(g per C-plate * C plates per waste package) 
(g per D-plate * D plates per waste package) 
(g per E-plate * E plates per waste package)] 
Average of lowest boron content (0.75 wt%) and 
highest content (1.74 wt%) from Table 4.1-11, 
divided by 100 
Initial boron in waste package 3.03E+03 moles Bi = (n * b)/AW (AW in Table 4.1-11) 
Void volume of waste From Table 5.1-1 (Assumption 5.1.2) 
k = crn *  * conv 
(where: 
crn = corrosion rate of Neutronit (follows a Weibull 
distribution (see Assumption 5.1.4) 
 = density of Stainless Steel Type 316 (Table 4.1- 
11) 
conv = conversion factor 1.0E-04 cm/m) 
tdeg = (n/k * SA) + Ts 
(where: 
SA = surface area (Table 5.1-1, Assumption 5.1.2) 
Ts = time to seismic event greater than 700 
years [700 years based on Assumption 5.2.2]) 
February 2004 
D = Bi/tdeg 
Using the input parameters, the boron loss equation is solved to determine the minimum drip rate 
(i.e., seepage rate m3/yr) required to degrade and flush out the Neutronit. The minimum drip rate 
is based on the sampled corrosion rate of Neutronit and time when the seismic event occurred 
after closure. The boron loss equation is solved for the drip rate (i.e., seepage rate m3/yr) when 
90 percent of the boron has been flushed from a damaged PWR commercial SNF waste package. 
For the damaged BWR commercial SNF waste package, the boron loss equation is solved for the 
drip rate (i.e., seepage rate m3/yr) when 50 percent of the boron has been flushed out. The 
amount of boron flushed from the commercial SNF waste packages is based on 
Assumption 5.1.5. The minimum required drip rate (i.e., seepage rate m3/yr), as calculated by 
the boron loss equation, is fed into the equations to determine the required seepage rate at the 
drift, as discussed in the next section. 
The seepage rate required to reach the drift is calculated using the parameters that were in the 
previous sections. The seepage rate required to reach the drift is calculated by using the 
91 of 176 
D 
Minimum Seepage Rate Probability 
ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
calculated required drip rate and the damaged area to the drip shield and waste package from a 
seismic event. The required seepage rate at the drift is calculated as: 
Q Qreq seepage 
=
F(  = P X seep 
where: 
Qseepage = seepage rate required to reach the drift (m3/yr) 
Qreq = minimum required drip rate to degrade and flush the Neutronit based on the 
solution of the boron loss equation (m3/yr) (Section 6.4.1.1.3) 
DSfr = fraction of damaged drip shield area due to a seismic event 
WPfr = fraction of damaged waste package area due to a seismic event 
Qcond = amount of condensation rate that can enter the waste package (m3/yr). 
Qcond = 0.0 as long as the temperature of the drip shield is greater than the temperature of 
the drift invert. Otherwise, Qcond is equal to the evaporation rate in the drift. 
Equation 6.4-4, based on EBS Radionuclide Transport Abstraction (BSC 2003 [DIRS 166466], 
Section 6.3.1.1), back calculates the seepage rate required to reach the drift by starting with the 
drip rate required to penetrate the damaged waste package in order to degrade and flush out 
Neutronit. The required drip rate to penetrate a damaged waste package is divided by the 
fraction of damaged area of the waste package and drip shield to account for the fraction of 
seepage rate that can reach the damaged waste package. The remaining seepage rate will flow 
around the drip shield and waste package damaged areas and into the invert. Therefore, the 
calculated seepage rate that is required to reach the drift accounts for the amounts of seepage rate 
that will penetrate the damaged waste package and that will flow into the invert. 
The minimum required seepage rate at the drift, as determined from Equation 6.4-4, is fed into 
Equation 6.4-5 (Walpole et al. 1998 [DIRS 152180], Section 3.3) to calculate the probability of 
having at least the minimum seepage flux reaching the drift (based on solution of 
Equation 6.4-4). 
where: 
Pseep = probability of having the required seepage rate or greater reaching the drift 
f(x) = probability distribution of the seepage rate into the drift (Attachment IV) 
Seepage rate distributions (i.e., low, mean and upper scenario), f(x) in Equation 6.4-5, were 
determined in Section 6.4.1.1.2 and discussed in Attachment IV and are based on the glacial 
transition climate, which is expected to last from roughly 2,000 to 10,000 years after closure 
(USGS 2001 [DIRS 158378], Section 6.6.1). The seepage rate distributions associated 
probabilities are 0.24, 0.41, and 0.35 for the low, mean, and upper scenario cases, respectively 
(BSC 2003 [DIRS 165991], Section 7, Table 7-1). In order to determine which scenario case is 
used, a random number is generated and tested against the probabilities. If the random number is 

 
seepage 
. .. 
. 
ANL-EBS-NU-000008 REV 00 
1 . Qcond DS fr WPfr (Eq. 6.4-4) 
. 
x f ) ( 
. .. 
. 
. 
Qseepage dx 
Q (Eq. 6.4-5) 
) = 
92 of 176 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
(1) less than 0.24, then the low seepage rate distribution is used to determine the minimum 
seepage rate probability; (2) if between 0.24 and 0.65, then the mean seepage rate distribution is 
used to determine the minimum seepage rate probability; or (3) greater than 0.65, then the upper 
seepage rate distribution is used to determine the minimum seepage rate probability. 
The seepage rate probability is then fed into Equation 6.4-6 (BSC 2003 [DIRS 161812], 
Attachment VIII, Equation VIII-2.11) to calculate the mean probability of seepage water 
penetrating a damaged waste package given a seismic event. The calculated mean probability is 
based on the minimum seepage required to reach the drift, which is required to penetrate the 
damaged waste package and degrade and flush out some percentage of the boron (i.e., 90 percent 
from PWR waste packages and 50 percent from BWR waste packages). The minimum seepage 
is based on the Neutronit corrosion rate which is the second probability term in Equation 6.4-6 
(i.e., minimum Neutronit corrosion rate or greater). The equation takes into account the time of 
seismic event. 
. . ( ) IE )] n [(T T l u l u  . . (Eq. 6.4-6) ( | ) ( cr ) P Seepage Seismic P sf P X cr n j i , seep j i 
IE 
n i 1 
= 
interpolated PGV value from Step 2. 
= 
where: 
P(Seepage|Seismic) = mean probability of seepage to penetrate a waste package given a 
seismic event 
Tu = upper bound of time sampling, regulatory period of 10,000 years 
Tl = lower bound of time sampling, time after closure of 1 year 
IEu = upper bound of seismic exceedance frequencies (1.0  10.4) 
IEl = lower bound of seismic exceedance frequencies (1.0  10.8) 
n = number of realizations (20,000) 
Pseep = probability of ith minimum seepage rate or greater depending on which scenario, j, 
(i.e., low, mean, upper) 
sfj = seepage fraction based on which scenario (i.e., low, mean, or upper) 
Pcr(X  cr) = probability of Neutronit corrosion rate or greater 
The sampling routine developed (to determine the damaged area of the drip shield and waste 
package along with how these damaged areas are used to calculate the mean probability of 
sufficient seepage) is presented in Attachment III and summarized as follows: 
1. Sample a mean annual seismic exceedance frequency from its uniform distribution. 
2. Sample the time of the seismic event from its uniform distribution. 
3. Interpolate, via log-linear interpolation, the PGV value based on the sampled mean 
annual exceedance frequency from the lookup table. 
4. Interpolate the respective lookup table to obtain the lower- and upper-bound values of 
the uniform distribution representing the damaged area of the drip shield using the 
. 
93 of 176 February 2004 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
5. Calculate the upper-bound damaged area of the waste package using the interpolated 
PGV value from Step 2. 
6. Sample from the developed uniform distributions for the damaged area of the drip 
shield and waste package to obtain the respective damaged area for that specific 
seismic event (sampled event). 
7. Sample a corrosion rate for Neutronit from its developed Weibull distribution. 
8. Calculate the time required to degrade Neutronit plates within the specific commercial 
SNF waste package based on time of the seismic event and sampled corrosion rate for 
Neutronit. This time to degrade the Neutronit assumes no water can penetrate a 
damaged waste package prior to 700 years after repository closure (see 
Assumption 5.2.2). 
9. Calculate the required drip rate into the specific damaged commercial SNF waste 
package to degrade and flush out some of the boron (i.e., 90 percent from a PWR 
waste package and 50 percent from a BWR waste package). 
10. Input the required drip rate from step 9 along with the damaged area for the drip shield 
and waste package into Equation 6.4-4 to calculate the required seepage rate at the 
drift. 
11. Input the calculated minimum seepage rate at the drift into Equation 6.4-5 to calculate 
the probability of having at least that minimum seepage rate. 
12. A random number is tested against the probability of being in the lower, mean, or 
upper scenario of the glacial transition climate. The lower scenario has a probability 
of 0.24, the mean scenario has a probability of 0.41, and the upper has a probability of 
0.35. If the random number is less than 0.24, then the probability of having that 
minimum seepage rate or greater is based on the lower scenario case. This probability 
is calculated using Equation 6.4-5 and is multiplied to its respective seepage fraction 
and this probability is carried forward for that particular realization. If the random 
number is between 0.24 and 0.65, then the probability of having that minimum 
seepage rate or greater is based on the mean scenario case. This probability is 
calculated using Equation 6.4-5 and is multiplied to its respective seepage fraction and 
this probability is carried forward for that particular realization. If the random number 
is greater than 0.65, then the probability of having that minimum seepage rate or 
greater is based on the upper scenario case. This probability is calculated using 
Equation 6.4-5 and is multiplied to its respective seepage fraction and carried forward 
for that realization. 
13. The calculated probability times its respective seepage fraction (Step 12) is input into 
Equation 6.4-6, which calculates the mean probability based on the 20,000 
realizations. The mean probability is then input into SAPHIRE. 
The mean probability from the 20,000 realizations varied depending upon which commercial 
SNF waste package was evaluated. The calculated probability of sufficient seepage water 
February 2004 94 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
available to penetrate the damaged waste package in order to degrade and flush out the Neutronit 
is fed into the SAPHIRE model. The calculated probabilities for the commercial SNF waste 
packages are presented as 0.0 for the 21-PWR with Absorber Plates, 12-PWR Long, and 24- 
BWR waste package types (Attachment III, p. 21). The 44-BWR Waste Package had a 
calculated probability of 1.53  10-6. 
In the SAPHIRE evaluation of seismic events for the lithophysal zone, basic event 
BE-SEEPAGE-10K is substituted by basic event BE-SEEPAGE-WP1-L to represent the 
probability for the 21-PWR with Absorber Plates Waste Package. Similarly, BE-SEEPAGE- 
10K will be substituted by basic events BE-SEEPAGE-WP3-L, BE-SEEPAGE-WP4-L, and BESEEPAGE-
WP5-L to represent waste package type 12-PWR Long, 44-BWR, and 24-BWR 
probabilities, respectively. 
For the 21-PWR Control Rod Waste Package, the probability of attaining the minimum required 
seepage is always zero because zirconium cladding neutron absorber material will not degrade 
during the performance period (Assumption 5.1.6). For the evaluation of seismic criticality 
FEPs, the probability of attaining the minimum required seepage for a 21-PWR with Control 
Rods Waste Packages in the lithophysal zone will be defined by basic event BE-SEEPAGEWP2-
L. Basic event BE-SEEPAGE-WP2-L will be assigned a value of 0.0. 
The calculated probability of attaining the minimum seepage rate for the 21-PWR with Absorber 
Plates Waste Package is extended to the DOE SNF waste package types assuming DOE SNF 
waste package neutron absorber material degradation is at least the same as that of the 21-PWR 
with Absorber Plates Waste Package (Assumption 5.1.6). The probability of attaining the 
minimum required seepage for the DOE SNF waste packages in the lithophysal zone will be 
defined by basic events BE-SEEPAGE-WP6-L, BE-SEEPAGE-WP7-L, and BE-SEEPAGEWP8-
L for the DOE SNF Long, DOE SNF Short, and DOE SNF MCO waste package types, 
respectively. 
As noted earlier, since the probability of the seismic event is incorporated into these basic events, 
the probability of a seismic disruptive event on the FEPS event tree has been set to 1.0. 
Seismic Effects in the Nonlithophysal Zone 6.4.1.1.2 
The nonlithophysal zone is analyzed separately because the drip shield can be damaged from 
both rock blocks ejected from the drift and vibratory ground motion. Drip shield damage due to 
vibratory ground motion has already been calculated and discussed in Section 6.4.1.1.1. 
Drip Shield and Waste Package Damage 
The calculation to obtain the percent damaged area to the drip shield from rockfall uses two 
separate equations. The first equation determines if damage to the drip shield occurred due to a 
rockfall and the second calculates the mode percent damaged area. These equations are utilized 
as outlined in Seismic Consequence Abstraction (BSC 2003 [DIRS 161812], Section 6.6.1.4 and 
6.10). Based on the process, each realization constitutes a rockfall that impacts the drip shield. 
February 2004 95 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The first equation (Equation 6.4-7) is evaluated based on the sampled PGV value (to determine if 
the rock block can damage the drip shield) by generating and comparing a random number to the 
value obtained from: 
601. . 0 DS nodamage 
DSmode 
= 
If the random number is less than DSnodamage, then the rock block caused no damage to the drip 
shield. However, if the random number is larger than or equal to DSnodamage, then damage has 
occurred to the drip shield from the rock block. The amount of damage is then calculated using a 
log-triangular distribution (Equation 6.4-8) with a minimum of 0.001 and a maximum of 
100 percent damage. The mode of the log-triangular distribution varies based on the sampled 
PGV value and is calculated as: 
. 0 00204 . 
96 of 176 
= 
The log-triangular distribution is then sampled each time the random number is greater than 
DSnodamage to determine the percent damaged area to the drip shield due to rockfall. This percent 
damaged area is then added to the percent damaged area sampled due to vibratory ground 
motion. The mean percent damaged area is converted to the mean fraction of damaged area by 
dividing by 100. The calculated mean fraction of damaged area to the drip shield from rockfall 
and vibratory ground motion is 4.74  10.3. The calculated 5th and 95th percentiles are 
9.92  10.9 and 1.06  10.2, respectively (Attachment III, p. III-8). 
The fraction of damaged area to the waste package is the same for the nonlithophysal as that 
calculated for the lithophysal, since the drip shield will still be intact and will deflect rock blocks 
from hitting a waste package. 
Seepage Rate Probability Distribution for Nonlithophysal 
The process used to determine the seepage rate distribution for the nonlithophysal zone, which is 
used to calculate the seepage probability, follows the process steps discussed in Abstraction of 
Drift Seepage (BSC 2003 [DIRS 165564], Section 6.7.1) and Section 6.4.1.1.1. The only 
difference is the permeability, k, and the look-up table for seepage rate and seepage rate standard 
deviation. 
Capillary strength (1/ ) is the same for the nonlithophysal zone as it is for the lithophysal zone. 
The nonlithophysal zone permeability (k) is determined to have a spatial variability distribution 
that is lognormal with a mean of -12.2 (in log 10) and a standard deviation of 0.34 (in log 10). 
The mean and standard deviation of permeability was determined from statistical analysis on the 
log-transformed data (BSC 2003 [DIRS 165564], Section 6.6.2.1).. The permeability uncertainty 
( .k) follows a triangular distribution with a lower bound of .0.68, upper bound +0.68, and a 
mean of 0.0. These distributions are for the nonlithophysal zone only. 
The percolation flux representing the repository area (BSC 2003 [DIRS 165564], Figure 6.6-10) 
is same for the nonlithophysal as for the lithophysal. The percolation flux uncertainty is 
expressed by three different scenarios for the spatial flux distributions (lower-bound, mean, and 
upper-bound). Since three different scenarios are used to represent the uncertainty, three seepage 
rate distributions (one for each scenario) are obtained for the nonlithophysal zone. 
ANL-EBS-NU-000008 REV 00 
. 0 735 PGV (Eq. 6.4-7) .
. 3 7767 (Eq. 6.4-8) PGV 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The percolation flux is adjusted for intermediate-scale heterogeneity by using flow-focusing 
factors (Equation 6.4-2), which are sampled and multiplied by the sampled percolation flux. 
However, for the nonlithophysal zone, the interpolated seepage rate is increased by 20 percent to 
account for rock bolts and drift degradation (BSC 2003 [DIRS 165564], Section 6.7.1). 
The seepage rate at the drift for each of the uncertainty scenarios (i.e., lower-bound, mean, and 
upper-bound) is determined using the same sampling process discussed in Section 6.4.1.1.1. The 
sampled value from the three key parameters (i.e., capillary strength, permeability, adjusted 
percolation flux) is used to interpolate the mean seepage rate and seepage rate standard deviation 
using the lookup table for the nondegraded drift (DTN: LB0304SMDCREV2.002 
[DIRS 163687]). The standard deviation is adjusted to account for uncertainty by creating a 
uniform distribution with a lower bound of .1.7321 times the sampled standard deviation and an 
upper bound of 1.7321 times the sampled standard deviation. The uniform distribution to 
account for uncertainty is sampled and then added to the interpolated mean seepage rate. This 
process is performed for 20,000 realizations (refer to Attachment IV). 
The resulting seepage rate values are adjusted prior to being used to determine the seepage rate 
probability by (1) setting seepage rates less than 0.1 kg/yr per package to zero (since these small 
values are the result of interpolation (BSC 2003 [DIRS 165564], Section 6.7.1)), and (2) setting 
calculated seepage rates greater than 100 percent to 100. 
Seepage rates are then filtered in order to develop a distribution that represents the seepage rate 
values by discarding all seepage rates with a zero value. The remaining nonzero seepage rates 
are then used to develop a Weibull distribution for each of the scenarios (i.e., lower-bound, 
mean, and upper-bound) to represent the seepage rate at the drift (refer to Attachment IV). These 
Weibull distributions are used to calculate the probability of having sufficient seepage in order to 
degrade and flush out the Neutronit from a seismically damaged waste package. In addition, to 
calculate the fraction of waste package locations that can see seepage, the number of nonzero 
seepage rates is divided by the total number of realizations (seepage fraction). The Weibull 
parameters, scale and shape ( and , respectively), and seepage fraction for each scenario are 
listed in Table 6.4-7. 
Table 6.4-7. Weibull Parameters and Seepage Fraction (Nonlithophysal Zone) 
Values 
Lower Bound (Drift Collapse) 
4.95E.03a (m3/yr) 
5.36E.01a 
Mean (Drift Collapse) 
8.56E.02c (m3/yr) 
4.73E.01c 
Weibull Parameters 
 (scale) 
 (shape) 
 (scale) 
 (shape) 
 (scale) 
 (shape) 
ANL-EBS-NU-000008 REV 00 
Upper Bound (Drift Collapse) 
Source: a Attachment IV, p. IV-54 
b Attachment IV, p. IV-53 
c Attachment IV, p. IV-67 
d Attachment IV, p. IV-66 
e Attachment IV, p. IV-80 
f Attachment IV, p. IV-79 
2.25E.01e (m3/yr) 
5.01E.01e 
97 of 176 
Seepage Fraction 
1.54E.01b 
5.24E.01d 
6.72E.01f 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Minimum Seepage Rate Probability 
The same process for determining the mean seepage probability discussed for the lithophysal 
zone is performed for the nonlithophysal zone. Equation 6.4-6 is used to calculate the mean 
seepage probability for nonlithophysal zone with different values for input seepage rate 
distributions and damaged area of the drip shield. The seepage rate distributions are different 
because they use different permeability parameters for the nonlithophysal zone and a different 
lookup table for the seepage rate. The other difference is the fraction of damaged area to the drip 
shield is larger because of the additional rockfall damage. Mathcad spreadsheets in 
Attachment III (p. 21) list probability evaluations for waste package types with at least this 
seepage. 
The mean probability from the 20,000 realizations varied depending upon which commercial 
SNF waste package was evaluated. The calculated mean probability of sufficient seepage water 
to penetrate the damaged waste package in order to degrade and flush out the Neutronit is fed 
into the SAPHIRE model. The calculated probabilities for the commercial SNF waste packages 
are 0.0 for the 21-PWR with Absorber Plates, 12-PWR Long, and 24-BWR waste package types. 
The 44-BWR Waste Package had a calculated probability of 9.74  10-7. 
In the SAPHIRE evaluation of seismic events for the nonlithophysal zone, basic event 
BE-SEEPAGE-10K is substituted by basic event BE-SEEPAGE-WP1-NL to represent the 
probability for the 21-PWR with Absorber Plates Waste Package. Similarly, 
BE-SEEPAGE-WP1-NL will be substituted by basic events BE-SEEPAGE-WP3-NL, 
BE-SEEPAGE-WP4-NL, and BE-SEEPAGE-WP5-NL to represent waste package type 12-PWR 
Long, 44-BWR, and 24-BWR probabilities, respectively. 
For the 21-PWR Control Rod Waste Package, the probability of attaining the minimum required 
seepage is always zero because zirconium cladding neutron absorber material will not degrade 
during the performance period (Assumption 5.1.6). For the evaluation of seismic criticality 
FEPs, the probability of attaining the minimum required seepage for the 21-PWR with Control 
Rods Waste Package in the nonlithophysal zone will be defined by basic event BE-SEEPAGEWP2-
NL. Basic event BE-SEEPAGE-WP2-NL will be assigned a value of 0.0. 
The calculated probability of attaining the minimum seepage rate for the 21-PWR with Absorber 
Plates Waste Package is extended to the DOE SNF waste package types assuming DOE SNF 
waste package neutron absorber material degradation is at least the same as that of the 21-PWR 
with Absorber Plates Waste Package (Assumption 5.1.6). The probability of attaining the 
minimum required seepage for the DOE SNF waste packages in the nonlithophysal zone will be 
defined by basic events BE-SEEPAGE-WP6-NL, BE-SEEPAGE-WP7-NL, and BE-SEEPAGEWP8-
NL for the DOE SNF Long, DOE SNF Short, and DOE SNF MCO waste package types, 
respectively. 
As noted earlier, since the probability of the seismic event is incorporated into these basic events, 
the probability of a seismic initiating event on the FEPs event tree has been set to 1.0. 
February 2004 98 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Sensitivity Case Seismic Evaluation 6.4.1.2 
The following sections describe a sensitivity case based on an increase in the Neutronit plate 
thickness for the 44-BWR Waste Package and uses the same base case analysis process and 
inputs, except Neutronit plate thickness is increased from 5 mm (refer to Table 6.4-5) to 7 mm. 
Sensitivity Case of the Seismic Effects in the Lithophysal Zone 6.4.1.2.1 
The same process for calculating the mean seepage probability for the base case was performed 
for this sensitivity case. The increase in the Neutronit plate thickness changed boron loss 
equation input parameters (mass of Neutronit, from 2.15  106 grams to 3.152  106 grams; and 
surface area, from 9.55  105 cm2 to 9.837  105 cm2) (Attachment III, p. III-23). By making 
this adjustment, the boron loss equation was solved again to determine the minimum drip rate 
(i.e., seepage rate m3/yr) required to degrade and flush out the Neutronit. This new drip rate was 
input into Equation 6.4-4 to calculate the required seepage rate at the drift. Using this newly 
calculated seepage rate, the mean probability from Equation 6.4-6 is calculated. The probability 
evaluation for having this amount of seepage or more is listed in the Mathcad spreadsheet 
documentation of Attachment III. 
The mean probability from the 20,000 realizations showed a significant decrease in the mean 
probability based on the increase in Neutronit plate thickness. This decrease is understandable, 
since the degradation and flushing would take longer because of the thicker Neutronit plates. 
The calculated mean probability for this sensitivity case is 1.24  10-10 (Attachment III, 
p. III-25). This calculated mean probability is input into the SAPHIRE model for the evaluation 
of this sensitivity case. The probabilities for the other commercial SNF waste packages remain 
the same since no adjustment was made to their design. 
Sensitivity Case of the Seismic Effects in the Nonlithophysal Zone 6.4.1.2.2 
The same process for calculating the mean seepage probability for the base case was performed 
for this sensitivity case. The only difference, as noted, is the increase in thickness of the 
Neutronit plate for the 44-BWR Waste Package. By making this adjustment, the boron loss 
equation was solved again to determine the minimum drip rate (i.e., seepage rate m3/yr) required 
to degrade and flush out the Neutronit. This new drip rate was fed into Equation 6.4-4, using the 
increased drip shield damaged area due to rockfall to calculate the required seepage rate at the 
drift. Using this newly calculated seepage rate, the mean probability from Equation 6.4-6 is 
calculated. The probability evaluation for having this amount of seepage or more is listed in the 
Mathcad spreadsheet documentation of Attachment III. 
The results from the 20,000 realizations showed a significant decrease in the mean probability 
was obtained by increasing the Neutronit plate thickness. The calculated mean probability for 
this sensitivity case in the nonlithophysal zone is 9.92  10-12 (Attachment III, p. 25) and is input 
into the SAPHIRE model. The probabilities for the other commercial SNF waste packages 
remain the same since no adjustment was made to their design. 
February 2004 99 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
6.4.2 SAPHIRE Basic Event Probability Modifications for Seismic Analysis 
Based on the calculations in the above sections, the following basic events are modified from the 
base case criticality FEPs SAPHIRE analysis for the seismic disruptive event base case and 
sensitivity case evaluations. 
Seismic Base Case Basic Event Probability Modifications 6.4.2.1 
Based on the calculations in the above sections, the following basic events are modified from the 
base case SAPHIRE criticality FEPs analysis for the seismic disruptive event base case 
evaluation. The basic event modifications for the lithophysal and nonlithophysal zones of the 
drift are listed in Table 6.4-8. 
Table 6.4-8. Seismic Base Case SAPHIRE Basic Event Assignment 
Input Data Description 
Sufficient water 
(infiltrate/condensate) reaches 
lithophysal drift by 10,000 years 
BE-SEEPAGE-WP1-L 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
nonlithophysal drift by 10,000 years 
BE-SEEPAGE-WP1-NL 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
lithophysal drift by 10,000 years 
BE-SEEPAGE-WP2-L 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
nonlithophysal drift by 10,000 years 
BE-SEEPAGE-WP2-NL 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
lithophysal drift by 10,000 years 
BE-SEEPAGE-WP3-L 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
nonlithophysal drift by 10,000 years 
BE-SEEPAGE-WP3-NL 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
lithophysal drift by 10,000 years 
BE-SEEPAGE-WP4-L 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
nonlithophysal drift by 10,000 years 
BE-SEEPAGE-WP4-NL 
(fault tree MS-IC-1) 
Input Data Probability 
Waste Package Type Probability 
21-PWR with Absorber Plates 0.0 
21-PWR with Absorber Plates 0.0 
0.0 21-PWR with Control Rods 
0.0 21-PWR with Control Rods 
0.0 12-PWR Long 
0.0 12-PWR Long 
44-BWR Absorber Plate 1.53E-6 
44-BWR Absorber Plate 9.74E-7 
Justification 
Section 6.4.1.1.1 
Section 6.4.1.1.2 
Section 6.4.1.1.1 
Section 6.4.1.1.2 
Section 6.4.1.1.1 
Section 6.4.1.1.2 
Section 6.4.1.1.1 
Section 6.4.1.1.2 
ANL-EBS-NU-000008 REV 00 100 of 176 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 6.4-8. Seismic Base Case SAPHIRE Basic Event Assignment 
Input Data Description 
Sufficient water 
(infiltrate/condensate) reaches 
lithophysal drift by 10,000 years 
BE-SEEPAGE-WP5-L 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
nonlithophysal drift by 10,000 years 
BE-SEEPAGE-WP5-NL 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
lithophysal drift by 10,000 years 
BE-SEEPAGE-WP6-L 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
nonlithophysal drift by 10,000 years 
BE-SEEPAGE-WP6-NL 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
lithophysal drift by 10,000 years 
BE-SEEPAGE-WP7-L 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
nonlithophysal drift by 10,000 years 
BE-SEEPAGE-WP7-NL 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
lithophysal drift by 10,000 years 
BE-SEEPAGE-WP8-L 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
nonlithophysal drift by 10,000 years 
BE-SEEPAGE-WP8-NL 
(fault tree MS-IC-1) 
Drip shield failure due to seismic 
event 
BE-DS-SEISMIC1 
(fault tree MS-IC-2) 
Waste package failure due to 
seismic event 
BE-WP-SEISMIC1 
(fault tree MS-IC-3) 
Waste package failure due to early 
failures 
BE-WP-EARLY-F1 
(fault tree MS-IC-3) 
ANL-EBS-NU-000008 REV 00 
Probability 
0.0 
0.0 
0.0 
0.0 
0.0 
0.0 
0.0 
0.0 
1.0 
1.0 
0.0 
Input Data Probability 
Waste Package Type 
24-BWR Absorber Plate 
24-BWR Absorber Plate 
DOE SNF Long 
DOE SNF Long 
DOE SNF Short 
DOE SNF Short 
DOE SNF MCO 
DOE SNF MCO 
All waste package types 
All waste package types 
All waste package types 
101 of 176 
Justification 
Section 6.4.1.1.1 
Section 6.4.1.1.2 
Section 6.4.1.1.1 
Section 6.4.1.1.2 
Section 6.4.1.1.1 
Section 6.4.1.1.2 
Section 6.4.1.1.1 
Section 6.4.1.1.2 
Section 6.4.1.1.1 
Section 6.4.1.1.1 
Section 6.4.1.1.1 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seismic Sensitivity Case Basic Event Probability Modifications 6.4.2.2 
Based on the calculations in the above sections, the following basic events in Table 6.4-9 are 
modified from the seismic disruptive event base case for the seismic disruptive event sensitivity 
Input Data Probability 
Justification 
Waste Package Type Probability 
Section 6.4.1.2.2 44-BWR 1.24E-10 
44-BWR Section 6.4.1.2.2 9.92E-12 
case evaluation. 
Table 6.4-9. Seismic Sensitivity Case SAPHIRE Basic Event Assignment 
Input Data Description 
Sufficient water 
(infiltrate/condensate) reaches 
lithophysal drift by 10,000 years 
BE-SEEPAGE-WP4-L 
(fault tree MS-IC-1) 
Sufficient water 
(infiltrate/condensate) reaches 
nonlithophysal drift by 10,000 years 
BE-SEEPAGE-WP4-NL 
(fault tree MS-IC-1) 
6.4.3 Seismic Criticality FEPs Analysis Results 
The waste package flooding probabilities resulting from the quantification of the SAPHIRE 
seismic disruptive event analyses are presented below. 
Seismic Base Case Criticality FEPs Analysis Results 6.4.3.1 
The waste package flooding probabilities resulting from the quantification of the SAPHIRE 
seismic disruptive event base case are presented in Table 6.4-10. These results include the 
flooding probabilities for waste packages in the lithophysal and nonlithophysal zones of the 
drifts. 
Table 6.4-10. Per Waste Package Flooding Probabilities for Seismic Base Case Criticality FEPs 
Waste Package Type 
21-PWR with Absorber Plates 
21-PWR with Control Rods 
12-PWR Long 
44 BWR 
24 BWR 
DOE SNF 
Source: a Values from Table 4.1-8 
b SAPHIRE V7.18 (BSC 2002 [DIRS 160873]) analysis results (Attachment II, p. II-20) 
c Sum of 5-DHLW/DOE SNF Short, 5-DHLW/DOE SNF Long and 2-MCO/2-DHLW Long values from 
Table 4.1-8 
Per Waste Package Flooding 
Probabilityb Number of Waste Packagesa 
Lithophysal Nonlithophysal 
0.00E+00 0.00E+00 4,299 
0.00E+00 0.00E+00 95 
0.00E+00 0.00E+00 163 
2,831 1.30E-06 1.46E-07 
0.00E+00 0.00E+00 84 
3,412c 0.00E+00 0.00E+00 
ANL-EBS-NU-000008 REV 00 February 2004 102 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seismic Sensitivity Case Criticality FEPs Analysis Results 6.4.3.2 
The waste package flooding probabilities resulting from the quantification of the SAPHIRE 
seismic disruptive event sensitivity case are presented in Table 6.4-11. These results include the 
flooding probabilities for waste packages in the lithophysal and nonlithophysal zones of the 
drifts. 
Table 6.4-11. Per Waste Package Flooding Probabilities for Seismic Sensitivity Case Criticality FEPs 
Waste Package Type 
21-PWR with Absorber Plates 
21-PWR with Control Rods 
12-PWR Long 
44 BWR 
24 BWR 
DOE SNF 
Source: a Values from Table 4.1-8 
b SAPHIRE V7.18 (BSC 2002 [DIRS 160873]) analysis results (Attachment II, p. II-20) 
c Sum of 5-DHLW/DOE SNF Short, 5DHLW/DOE SNF Long and 2-MCO/2-DHLW Long values from 
Table 4.1-8 
Total Seismic Criticality FEPs Analysis Results 6.4.3.3 
Number of Waste 
Packagesa 
Per Waste Package Flooding Probabilityb 
Lithophysal Nonlithophysal 
4,299 
95 
163 
2,831 
84 
3,412c 0.00E+00 0.00E+00 
0.00E+00 0.00E+00 
0.00E+00 0.00E+00 
0.00E+00 0.00E+00 
1.05E-10 1.39E-12 
0.00E+00 0.00E+00 
Table 6.4-12 summarizes the SAPHIRE seismic disruptive event results for the seismic base case 
and sensitivity case evaluations. The probabilities for the lithophysal and nonlithophysal zones 
from Tables 6.4-10 and 6.4-11 are combined to provide the total probability for each of these 
cases. 
Table 6.4-12. Total Per Waste Package Flooding Probabilities for Seismic Criticality FEPs 
Waste Package Type 
21-PWR with Absorber Plates 
21-PWR with Control Rods 
12-PWR Long 
44 BWR 
24 BWR 
DOE SNF 
Source: a Values from Table 4.1-8 
b SAPHIRE V7.18 (BSC 2002 [DIRS 160873]) analysis results (Attachment II, p. II-20) 
c Sum of 5-DHLW/DOE SNF Short, 5DHLW/DOE SNF Long and 2-MCO/2-DHLW Long values from 
Table 4.1-8 
d Due to roundoff, the sum of the lithophysal and nonlithophysal results probabilities reported in 
Tables 6.4-10 and 6.4-11 are different than this value. 
Number of Waste Packagesa 
4,299 
95 
163 
2,831 
84 
3,412c 0.00E+00 0.00E+00 
Total Per Waste Package Flooding 
Probabilityb 
Sensitivity Case Base Case 
0.00E+00 0.00E+00 
0.00E+00 0.00E+00 
0.00E+00 0.00E+00 
1.44E-06 1.07E-10d 
0.00E+00 0.00E+00 
ANL-EBS-NU-000008 REV 00 103 of 176 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
6.4.4 Seismic Ground Motion Effects on Near-Field and Far-Field Criticality 
In order for a criticality to occur in the near-field or far-field, sufficient quantity of the waste 
forms fissile inventory must be removed from the waste package and accumulate in void spaces 
within the host rock. This requires sufficient damage to the waste package to allow adequate 
seepage to degrade waste form and flush the fissile material from the waste package. The 
probability evaluation for in-package criticality due to a seismic event can be used as the starting 
point for the additional sequence of events required to cause near-field or far-field criticality. 
The remaining probability of events must be evaluated to determine the probability of an external 
criticality due to a seismic event include: (1) the waste form will degrade within 10,000 years; 
(2) fissile material will be flushed from the waste package; and (3) fissile material accumulates 
in a sufficient quantity and geometry to allow for criticality. Therefore, once the probabilities of 
these events are determined, the probability of a near-field or far-field criticality due to a seismic 
event will be lower than that calculated for in-package criticality. 
6.5 ANALYSIS OF ROCKFALL DISRUPTIVE EVENT CRITICALITY FEPS 
Rockfall disruptive event criticality FEPs are presented in Table 6.5-1. 
Table 6.5-1. Rockfall Disruptive Event Criticality FEPs 
FEP Number 
2.1.14.21.0A 
FEP Title 
In-package 
criticality resulting 
from rockfall (intact 
configuration) 
In-package 
criticality resulting 
2.1.14.22.0A from rockfall 
(degraded 
configurations) 
Near-field criticality 
2.1.14.23.0A resulting from 
rockfall 
Far-field criticality 
2.2.14.11.0A resulting from 
rockfall 
Source: Table 6.1-4 
The following sections discuss quantification of SAPHIRE basic events required to be modified 
to perform the probabilistic evaluation of the rockfall disruptive event criticality FEPs and 
presents the SAPHIRE evaluation results. 
FEP Description 
The waste package internal structures and the waste form remain intact 
either during or after a rockfall event. A breach (or breaches) in the waste 
package allow(s) water to either accumulate or flow-through the waste 
package. Criticality then occurs in situ. 
Either during, or as a result of, a rockfall event, the waste package internal 
structures and the waste form degrade. A critical configuration develops 
and criticality occurs in situ. Potential in situ critical configurations are 
defined in Figures 3.2a and 3.2b of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 165505]). 
Either during, or as a result of, a rockfall event, near-field criticality occurs 
when fissile material-bearing solution from the waste package is 
transported into the drift and the fissile material is precipitated into a critical 
configuration. Potential near-field critical configurations are defined in 
Figure 3.3a of Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). 
Either during, or as a result of, a rockfall event, far-field criticality occurs 
when fissile material-bearing solution from the waste package is 
transported beyond the drift and the fissile material is precipitated into a 
critical configuration. Potential far-field critical configurations are defined in 
Figure 3.3b of Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). 
February 2004 
6.5.1 Rockfall Drip Shield Failure Probability 
Rockfall disruptive event criticality FEPs 2.1.14.21.0A, 2.1.14.22.0A, 2.1.14.23.0A, and 
2.2.14.11.0A require an assessment of the probability of criticality due to rockfall. A rockfall 
ANL-EBS-NU-000008 REV 00 104 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
event can occur as result of normal drift degradation, as well as the result of a seismic event. 
Because the frequency of rockfall due to static drift degradation cannot be readily predicted, a 
probability of 1.0 is assigned to this disruptive event. A rockfall event could potentially result in 
drip shield damage depending on the size of the rockfall, the impact velocity and drip shield 
impact location. Because the drip shield covers the waste package, no waste package damage is 
predicted due to a rockfall event. 
Although the rockfall disruptive event is considered to be a static event (i.e., the rocks drop from 
the drift overhead due to static drift degradation, not as a result of any external initiating event 
such as an seismic event), the nonstatic results of Seismic Consequence Abstraction (BSC 2003 
[DIRS 161812]) are used in this evaluation to bound the analysis results. This approach is 
believed to be bounding because more blocks would be expected to fall during a seismic event 
and the rock-to-drip-shield impact velocity is greater than would be expected under static 
conditions. 
The probability of drip shield damage due to rockfall is based on the information contained in 
Seismic Consequence Abstraction (BSC 2003 [DIRS 161812]). The drip shield rockfall damage 
results are based on the impact of 279 and 380 blocks for seismic event annual exceedance 
frequencies of 10-6 and 10-7, respectively (BSC 2003 [DIRS 161812], Section 6.6.1.1). These 
blocks occur in the nonlithophysal zone of the drifts, which comprises only 15 percent of the 
total drift area (745,486 m2 of drift area resides in the nonlithophysal geological unit out of a 
total drift area of 4,983,152 m2 (BSC 2003 [DIRS 164491], Table 9]). The drip shield is not 
damaged by rockfall impacts in the lithophysal zone (BSC 2003 [DIRS 161812], Section 6.6.2). 
The total available emplacement drift length is 63,945 m (BSC 2003 [DIRS 164490], Tables 4 
through 7). If the average drip shield length is 5.805 m (BSC 2004 [DIRS 167309], Table 1), 
then the total number of drip shields in the repository is calculated to be 11,016 (63,945 m / 
5.805 m per drip shield). Therefore, the total number of drip shields available for rockfall impact 
damage in the nonlithophysal zone is calculated to be 1,653 (15 percent of 11,016). 
Although Seismic Consequence Abstraction (BSC 2003 [DIRS 161812]) calculates drip shield 
damage due to single block impacts as well as multiple rock impacts, the determination of basic 
event probability for drip shield damage conservatively calculates that each rock block is 
available to fall on a unique drip shield. This results in a greater number of drip shields being 
impacted by rockfall. This calculation will also use 380 rock blocks (the number of rock blocks 
from the 10-7 annual exceedance frequency) as this will also result in a greater number of 
impacted drip shields. In the seismic calculation of Attachment III, the rockfall evaluation 
presented on page III-9 resulted in the calculation of drip shield damage occurring 14.4 percent 
of the time. This is based on 20,000 realizations in which each assumes a rock block hits a drip 
shield. 
February 2004 105 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Utilizing the above information, the probability of drip shield damage due to rockfall can be 
determined by multiplying the fraction of drip shields hit by rocks and the probability that the 
drip shield will be damaged by the impact. This calculation is presented as: 
Probdsd = Nrf / Nds  DSdf (Eq. 6.5-1) 
where: 
Probdsd = probability of drip shield damage 
Nrf = number rock blocks available to fall (380) 
Nds = number of drips shields available for the rocks to fall upon (1,653) 
DSdf = fraction of drip shields damaged due to rockfall (0.144). 
Inserting these values into the above equation results in a probability of drip shield damage due 
to rockfall (Probdsd) of 3.31  10-2. 
However, for the rockfall disruptive event, the probability of drip shield damage does not 
correlate to the probability of drip shield failure. Drip shield failure is defined as the failure of 
the drip shield to perform its primary function  to prevent advective flow from contacting the 
waste package. Drip shield failure may be the result of a stress corrosion crack or complete 
structural failure. Although rockfall will result in stress corrosion cracking of the drip shield, the 
resulting cracks are predicted to be plugged with corrosion products or precipitates (BSC 2003 
[DIRS 161234], Section 6.3.7) causing the probability of advective flow through the cracks to 
approach zero (BSC 2003 [DIRS 161234], Section 6.3.7). 
Therefore, the probability of drip shield failure resulting from a rockfall disruptive event is 0.0 
and is the value assigned to default basic event BE-DS-ROCK-FALL of the drip shield fault tree 
MS-IC-2. This is the failure probability that will be utilized in both the lithophysal and 
nonlithophysal geological zone analyses of the rockfall disruptive event SAPHIRE evaluation. 
6.5.2 Minimum Seepage Probability for Rockfall Disruptive Event 
The probability of attaining the minimum seepage flux resulting from a rockfall disruptive event 
is calculated to be zero because of a zero-probability function for drip shield damage area for the 
rockfall disruptive event (i.e., no drip shield failures). Rockfall in the nonlithophsal geological 
zone does not cause a drip shield failure, but produces areas of stress in the drip shield surface 
that may result in stress corrosion cracking. However, rockfall induced stress corrosion cracks 
are predicted to be plugged by corrosion products or be sealed by precipitants that will prevent 
advective flow onto the waste package (BSC 2003 [DIRS 161234], Section 6.3.7). 
It should be noted that, as previously stated, drip shield damage due to rockfall occurs only in the 
nonlithophysal geological zone. No drip shield damage due to rockfall is predicted to occur in 
the lithophysal geological zone; therefore, the damage probability function is zero for drip 
shields in the lithophysal geological zone. 
Since rockfall does not impact the waste package, the only viable waste package failure 
mechanism during the rockfall disruptive event results from fabrication errors. As discussed in 
February 2004 106 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Section 6.3.3.3.4, improper heat treatment of the waste package results in a probability of early 
waste package failure of 2.8  10-5 per waste package. 
However, because there are no drip shield failures to allow advective flow onto any failed waste 
packages, the probability of sufficient seepage water to degrade and flush out the neutron 
absorbing material resulting from a rockfall disruptive event is 0.0. This value is assigned to the 
default basic event (BE-SEEPAGE-10K) of the seepage fault tree MS-IC-1 and utilized in the 
lithophysal and nonlithophysal geological zone analyses. 
6.5.3 SAPHIRE Basic Event Probability Modifications for Rockfall Analysis 
Based on the information presented in the sections above, no basic events must be modified from 
the base case SAPHIRE analysis for the rockfall disruptive event SAPHIRE analysis. 
6.5.4 Rockfall Criticality FEPs Analysis Results 
The quantification of the SAPHIRE rockfall disruptive event resulted in the calculation of the 
waste package fractional probabilities presented in Table 6.5-2. Because the probability of 
attaining the minimum required seepage to degrade the waste package internal components and 
flush out the neutron absorber material is zero, the fractional probability of flooding the waste 
package types is also zero. This result is applicable to all rockfall criticality FEPs regardless of 
analysis location (internal or external to the waste package) or waste form/waste package type. 
Table 6.5-2. Per Waste Package Flooding Probabilities for Rockfall Criticality FEPs 
Per Waste Package Flooding 
Probabilityb 
0.00E+00 
0.00E+00 
0.00E+00 
0.00E+00 
0.00E+00 
February 2004 
Waste Package Type 
21-PWR with Absorber Plates 
21-PWR with Control Rods 
12-PWR Long 
44 BWR 
24 BWR 
DOE SNF 
Source: a Values from Table 4.1-8 
b SAPHIRE V7.18 (BSC 2002 [DIRS 160873]) analysis results (Attachment II, p. II-20) 
c Sum of 5-DHLW/DOE SNF Short, 5DHLW/DOE SNF Long and 2-MCO/2-DHLW Long values from Table 4.1-8 
6.5.5 Rockfall External Criticality Probability 
External criticality FEPs can originate from either a bathtub or flow-through waste package 
configuration. However, because there is no mechanism to breach the drip shield during the 
performance period for a rockfall disruptive event, there is no probability of water entering the 
waste package. Therefore, there is no probability of criticality for the rockfall external criticality 
FEPs presented in Table 6.5-1. 
6.6 ANALYSIS OF IGNEOUS DISRUPTIVE EVENT CRITICALITY FEP 
The igneous disruptive event criticality FEPs are presented in Table 6.6-1. 
ANL-EBS-NU-000008 REV 00 
Number of Waste Packagesa 
4299 
95 
163 
2831 
84 
3412c 0.00E+00 
107 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 6.6-1. Igneous Disruptive Event Criticality FEPs 
FEP Number FEP Name 
In-package 
criticality resulting 
2.1.14.24.0A from an igneous 
event (intact 
configuration) 
In-package 
criticality resulting 
2.1.14.25.0A from an igneous 
event (degraded 
configurations) 
2.1.14.26.0A 
Near-field 
criticality resulting 
from an igneous 
event 
Far-field criticality 
2.2.14.12.0A resulting from an 
igneous event 
Source: Table 6.1-4 
Characterize Framework for Igneous Activity at Yucca Mountain, Nevada (BSC 2003 [DIRS 
163769], Table 22), estimates the frequency of an igneous intrusion event to be 1.710-8 per 
year, which is the frequency of a basaltic dike intersecting the subsurface area of the repository 
(intrusive scenario). If a dike intersects the repository, there is about a 78 percent chance that at 
least one or more eruptive centers would be located within the repository for an annual frequency 
of 1.310-8 per year (BSC 2003 [DIRS 163769], Table 22). An igneous disruptive event could 
lead to the destruction of waste packages and their waste forms. Therefore, the possibility of a 
waste form undergoing a criticality event must be examined. 
In the event of igneous intrusion, moderating materials (primarily consisting of silicon dioxide 
and water) may be present. As a result of igneous intrusion into the drifts, a criticality event may 
be possible due to (DOE 2002 [DIRS 155943], Section 4.3.3.2.1): 
1. Immediate breach of the waste package; 
2. Separation of a significant fraction of the fissile material from the neutron absorber by 
magma transport; or 
3. Accumulation of a critical mass of fissile material from, or within, the transporting 
magma. 
6.6.1 Igneous Intrusion Impacts on Waste Packages and Waste Forms 
The TSPA-LA approach to implementing the models for waste package and waste form response 
during igneous intrusion considers two impact regions: (1) Zone 1, which includes the 
emplacement drift intruded by the basalt dike; and (2) Zone 2, which includes the emplacement 
drift adjacent to the intruded drift (BSC 2003 [DIRS 165002], Section 1). 
ANL-EBS-NU-000008 REV 00 
FEP Description 
The waste package internal structures and the waste form remain intact 
either during or after an igneous disruptive event. A breach (or breaches) in 
the waste package allow(s) water to either accumulate or flow-through the 
waste package. Criticality then occurs in situ. 
Either during, or as a result of, an igneous disruptive event, the waste 
package internal structures and the waste form degrade. A critical 
configuration develops and criticality occurs in situ. Potential in situ critical 
configurations are defined in Figures 3.2a and 3.2b of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). 
Either during, or as a result of, an igneous disruptive event, near-field 
criticality occurs when fissile material-bearing solution from the waste 
package is transported into the drift and the fissile material is precipitated 
into a critical configuration. Potential near-field critical configurations are 
defined in Figure 3.3a of Disposal Criticality Analysis Methodology Topical 
Report (YMP 2003 [DIRS 165505]). 
Either during, or as a result of, an igneous disruptive event, far-field criticality 
occurs when fissile material-bearing solution from the waste package is 
transported beyond the drift and the fissile material is precipitated into a 
critical configuration. Potential far-field critical configurations are defined in 
Figure 3.3b of Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). 
February 2004 108 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Zone 1 
It is expected that the drip shields, invert, and waste packages in Zone 1 will be compressed and 
damaged, allowing magma to occupy the entire emplacement drift. The igneous intrusion 
temperature may be as high as 1,169C (BSC 2003 [DIRS 166407], Table 38). The melting 
points of waste packages made of Alloy C-22 and Stainless Steel Type 316NG are approximately 
1,357C and 1,375C, respectively (CRWMS M&O 1999 [DIRS 121300], Section 5.1). 
Although the intrusive igneous temperature is lower than the melting points of steel and alloy, 
these engineered materials could be severely damaged at the intrusive temperature (e.g., through 
softening, creeping and breaking down) combined with the shear forces of the viscous magma 
moving at the assumed velocity. 
When the waste packages are damaged, the waste forms will be exposed to and are likely to be 
enveloped and fused by the flowing magma. The fuel assemblies will be crushed and 
fragmented, introducing different size fragments and granules of UO2 pellets/cladding, neutron 
absorber, and control rods. The crushed material may form radionuclide-bearing minerals by 
incorporating crystallizing silicate minerals. 
The igneous intrusion scenario shows a range of consequences, extending from virtually no 
impact up to an impact upon all waste packages in the repository. The 50th percentile value 
indicates approximately 3,160 waste packages impacted, out of over 11,000 (BSC 2003 [DIRS 
161851], Section 7.2). 
Zone 2 
Analyses of possible impacts from thermal and volatile gases are conducted on Zone 2 drip 
shields, waste packages, and cladding to determine the potential for elevated corrosion rates due 
to deleterious environment, marked by the conducting heat and diffusing volatile gases evolving 
from the basalt magma intruded into Zone 1 emplacement drifts (BSC 2003 [DIRS 165002], 
Section 6.5.2). From the spatial and temporal heat conduction simulations and analysis, the high 
temperatures after a magma event attenuate rapidly with distance. The maximum temperature 
rise in an adjacent drift is small (less than 10C), and the rock provides effective thermal 
insulation to the impacts of high temperature. From the gas transport simulations, the maximum 
gas concentrations entering the Zone 2 emplacement drifts are extremely low. It is concluded 
that there are no impacts from thermal or volatile gases on waste packages and waste forms in 
Zone 2 (BSC 2003 [DIRS 165002], Section 6.7.2). 
6.6.2 Configurations Resulting from Complete Waste Package Destruction 
The drip shields, waste packages, and fuel cladding in Zone 2 remain intact with no impacts 
resulting from the heat or volatile gases released during an igneous intrusion. Therefore, 
criticality evaluation of the waste packages and waste forms in Zone 2 are not required as these 
would be enveloped by the base case analysis of Section 6.3. 
However, in Zone 1, after a postulated waste package destruction, turbulent magma could move 
the waste form away from the neutron absorber materials that are placed into the waste packages 
to inhibit criticality. 
February 2004 109 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
A conservative evaluation of the possibility of criticality after an igneous intrusion, is based on 
the following elements: 
1. Fissile material (commercial SNF pellet) is surrounded by a cubic lattice of magma 
which serves as the moderator (Assumption 5.4.1), 
2. Fissile material becomes separated from the neutron absorber material 
(Assumption 5.4.2), and 
3. Magma water content is 0.5 weight percent (Assumption 5.4.6). 
The post-Miocene data on water contents of basaltic magma in the Yucca Mountain region is 
sparse, making it difficult to define rigorously a probability distribution function for water 
content for use in the Performance Assessment. As such, it is recommended to use between 3 
and 4 weight percent dissolved water. The probability decreases linearly so that it is zero at 4 
weight percent, representing the expectation that at about 4 weight percent, basaltic magma will 
crystallize underground rather than erupt (BSC 2003 [DIRS 166407], Section 6.3.2.2). For 
intrusive dike impacts, the upper bound of 4 weight percent initial dissolved water content in the 
magma is considered. As the magma ascends, it decompresses at low pressures to cause magma 
crystallization and water vapor exsolving, and as a result, the initial dissolved mass of water in 
the magma will reduce. At the repository level and the corresponding pressure, the dissolved 
water in the magma is expected to reduce to roughly 0.5 weight percent at a temperature of 
1,150C (Assumption 5.4.6). 
6.6.3 Complete Waste Package Destruction: Internal or Near-Field Criticality 
In the event of igneous intrusion, the enrichment and burnup of the fissile materials will not 
change due to the crush of waste packages and the mixing of waste forms and basalt/magma. 
The criticality concerns would be the possibilities of separation of neutron absorber from the fuel 
assemblies, the presence of water and silica as potential neutron moderators, and the 
convergence of fuel assemblies into a critical mass. 
As mentioned earlier in Section 6.6, criticality during an igneous intrusion depends on critical 
mass and critical configuration, which includes separation of the neutron absorber (control rods 
and boron plates) from the fissile material, and the presence of sufficient amounts of a neutron 
moderator. Although silica is abundant in the basalt/magma, its neutron moderation is much less 
than that of water. The initial water content in magma at the repository level is assumed to be 
0.5 weight percent (Assumption 5.4.6). Considering igneous intrusion high temperature (up to 
1,169C) (BSC 2003 [DIRS 166407], Table 38), and high pressure (approximately 7.5 MPa) 
(CRWMS M&O 2000 [DIRS 151552], Section 6.3.2) conditions, the remaining water is 
expected to vaporize rapidly since the saturation temperature of water at 7.5 MPa is only 290C 
(Wark 1983 [DIRS 157283], Table A-12), well below the magma temperature. 
Criticality calculations of disassembled waste packages were performed previously for igneous 
scenarios in Probability of Criticality Before 10,000 Years (CRWMS M&O 2000 [DIRS 
149939], Section 6.2.2) using MCNP (Monte Carlo N-Particle Transport Code System). MCNP 
is widely accepted software used to perform criticality analysis of waste packages and waste 
February 2004 110 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
forms configuration. A PWR commercial SNF waste package with 3.5 weight percent 
enrichment, 10.0 GWd/MTU burnup, and 5 years decay was used in that report. The lattice 
spacing of the commercial SNF pellets (or total volume surrounding the pellets) was varied in 
order to determine the optimum volume favorable to criticality. However, in real operation, 
there would be no pattern of loading the same PWR waste packages together. For the sake of 
conservatism and simplicity, that report (CRWMS M&O 2000 [DIRS 149939]) modeled fuel 
pellets from seven commercial SNF waste packages, spread out in a cubic lattice that was filled 
with magma and reflected by tuff in a spherical geometry. The maximum estimated water 
content of 5.0 weight percent was used in the magma composition. The maximum calculated 
value of keff from these analyses was 0.769. 
It should be noted that although an igneous event may result in crushing and fragmenting the 
waste package and its contents, previous and current criticality evaluations assume the fuel pellet 
remains intact within the magma system. This treatment is appropriate because for low-enriched 
uranium systems, such as commercial SNF, a heterogeneous lattice configuration (e.g., intact 
fuel pellets aligned in a set order) is more reactive (yields a higher keff) than a homogeneous 
configuration with granulated fuel (Duderstadt and Hamilton 1976 [DIRS 106070], pp. 403 
to 405). 
New evaluations were performed using the estimated 0.5 weight percent water content in magma 
at the repository level (Assumption 5.4.6). The elemental composition and atom densities of 
magma based on this water content level are listed in Tables 4.1-10 and 4.1-12. A cylindrical 
PWR fuel pellet (0.47 cm round and 1.1 cm long) is imbedded in the magma cube (DOE 1987 
[DIRS 132333], pp. 2A to 34). The elemental composition and mass densities of UO2 with 5.0 
weight percent enrichment (fresh fuel) can be found in Table 4.1-13. The atom densities input 
into the MCNP analyses are presented in Table 6.6-2. The element atom densities are calculated 
in the Microsoft EXCEL file FepIgn1.xls of Attachment VII (based on the information presented 
in Tables 4.1-10, 4.1-12, and 4.1-13). 
Total 
Source: Microsoft Excel spreadsheet 
FepIgn1.xls results (Attachment VII) 
Table 6.6-2. Atom Densities of Magma with 0.5 Weight Percent Water 
Element 
H 
O 
Si 
Al 
Fe 
Mg 
Ca 
Na 
K 
Ti 
P 
Mn 
ANL-EBS-NU-000008 REV 00 
Atom Density 
(atoms/b-cm) c 
9.53E-04 
4.74E-02 
1.39E-02 
5.64E-03 
2.50E-03 
2.48E-03 
2.63E-03 
1.96E-03 
6.71E-04 
4.15E-04 
2.95E-04 
4.11E-05 
7.88E-02 
111 of 176 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The magma cube surfaces are designated to be reflective to simulate an infinite system. The size 
of the magma cube is varied to search for the bounding configuration - that which results in the 
maximum keff value. These data and geometric configurations are applied in running the MCNP 
code to calculate keff. The results of this evaluation are presented in Table 6.6-3. 
Table 6.6-3 contains values for keff and width of the magma cube. These values are plotted in 
Figure 6.6-1. The associated calculated uncertainty values, sigma (), are also presented in 
Table 6.6-3 for each keff value. The 95th percentile keff is calculated by adding two times the 
sigma value to the keff value. This result is accounted for in the last column of the table. From 
Table 6.6-3 it is seen that the keff+2 value peaks at 0.813 for a magma cube width of 2.46 cm. 
Since the evaluated configurations are infinite systems, these results indicated that no matter how 
many waste packages are disassembled by an igneous event, the systems would remain 
subcritical. Therefore, no probability evaluations of these configurations are required. 
The value of keff+2 is expected to be lower than the calculated critical limit for external systems 
moderated by silica (tuff or magma) (Assumption 5.4.4). Although the above evaluation was 
performed for PWR commercial SNF fuel pellet, there are no appreciable differences between 
PWR and BWR fuel pellets and the results are, therefore, applicable to both. However, given the 
differences between commercial SNF and some highly enriched DOE SNF, results from the 
commercial SNF igneous evaluation cannot be directly abstracted to DOE SNF. Therefore, 
pending further evaluation, it is assumed that configurations resulting from an igneous event 
involving DOE SNF will not result in a critical system (Assumption 5.4.3). Therefore, the 
probability of criticality due to an igneous disruptive event is set to zero. 
Table 6.6-3. Summary of Criticality Calculations with Infinite System 
Keff Case Name (cm3) 
Width of Cube Volume of Cube 
(cm) 
112 of 176 ANL-EBS-NU-000008 REV 00 
cpinf08 1.12 1.40 0.772 5.0 ~ 10.4 
cpinf09 1.16 1.56 0.771 6.0 ~ 10.4 
cpinf10 1.2 1.73 0.767 6.1 ~ 10.4 
cpinf06 1.26 2.00 0.763 6.1 ~ 10.4 
cpinf01 1.66 4.57 0.758 6.7 ~ 10.4 
cpinf07 2.06 8.74 0.794 7.3 ~ 10.4 
cpinf02 2.46 14.9 0.811 6.6 ~ 10.4 
cpinf03 2.86 23.4 0.774 8.6 ~ 10.4 
cpinf04 3.66 49.0 0.600 6.5 ~ 10.4 
cpinf05 4.06 66.9 0.510 5.7 ~ 10.4 
Source: CRWMS M&O 1998 [DIRS 154060] analysis results (Attachment VII) 
Sigma () keff + 2  
0.773 
0.772 
0.769 
0.764 
0.759 
0.796 
0.813 
0.775 
0.602 
0.511 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
0.9 
Keff 
0.8 
0.7 
0.6 
0.5 
0.4 
0.3 
0.2 
0.1
0 
6.7 CRITICALITY FEPS RESULTS 
Evaluation of SAPHIRE event trees for the base case events, seismic disruptive event, and 
rockfall disruptive event resulted in the per waste package probabilities presented in Table 6.7-1. 
Table 6.7-1 summarizes the SAPHIRE analysis results presented in Tables 6.3-8, 6.4-12 (both 
seismic base case [as-designed 44-BWR Waste Package absorber plate thickness as designed of 
5 mm] and sensitivity case [44-BWR Waste Package absorber plate thickness modified to 7 mm] 
results) and 6.5-2. Additionally, the results of the igneous disruptive event criticality FEPs 
evaluation of Section 6.6 have been added to this table. The total per waste package flooding 
probability results of Table 6.7-1 is the sum of the initiating event per waste package flooding 
probabilities for each waste package type (i.e., Total = Base Case + Seismic + Rockfall + 
Igneous). 
ANL-EBS-NU-000008 REV 00 
Figure 6.6-1. Criticality Calculations with Infinite System 
113 of 176 
0 0.5 1 1.5 2 2.5 3 3.5 4 4.5 
Width of Cube (cm) 
Source: Table 6.6-3 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Total 
= 
Table 6.7-1. Per Waste Package Flooding Probability per Waste Package Type 
Per Waste Package Flooding Probability Number of 
Waste 
Packages Base Case Seismic Rockfall Igneous 
3,412 
21-PWR with Absorber Plates 4,299 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
21-PWR with Control Rods 95 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
12-PWR Long 169 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
44-BWR (Seismic Base Casea) 2,831 0.00E+00 1.44E-6 0.00E+00 0.00E+00 1.44E-6 
44-BWR (Seismic Sensitivity Caseb) 2,831 0.00E+00 1.07E-10 0.00E+00 0.00E+00 1.07E-10 
24-BWR 84 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
DOE SNF 0.00E+00 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
Source: SAPHIRE V7.18 (BSC 2002 [DIRS 160873]) analysis results (Attachment II, p. II-20) 
Notes: a 44-BWR Waste Package Absorber Plate thickness is as designed at 5 mm. 
b 44-BWR Waste Package absorber plate thickness is modified to 7 mm. 
Using the binomial distribution equation (Equation 6.7-1) (Walpole et al. 1998 [DIRS 152180], 
Section 5.3), the total probability of waste package flooding for each waste package type can be 
n.x (Eq. 6.7-1) ( , , p n x b ) 
x 
p (1. p) 
x
n . .
.
. 
February 2004 
 
.
. 
where: 
x = number of waste packages flooded (varied between 0 and 3) 
n = number of the waste package type being evaluated (Table 6.7-1, Column 2) 
p = per waste package flooding probability of the waste package type being evaluated 
(Table 6.7-1, Column 7) 
The total probability of flooding between one and four waste packages is presented in 
Table 6.7-2 for each of the waste package types presented in Table 6.7-1. As presented in this 
table, for the 21-PWR with Absorber Plates Waste Package, the analyses show that the waste 
package flooding probability is always below the regulatory probability criterion of less than one 
chance in 10,000 of occurring over 10,000 years (10 CFR 63.114(d) [DIRS 156605]). However, 
the binomial distribution analysis of all waste package types using the seismic base case results 
(as-designed 44-BWR Waste Package absorber plate thickness of 5 mm) indicates that the 
individual and total probability of flooding at least one waste package is greater than the 
regulatory probability criterion. These table cells have been highlighted. The individual and 
total waste package flooding probabilities drop below the regulatory probability criterion for two 
or more flooded waste packages. 
If the seismic sensitivity case results are considered (44-BWR Waste Package absorber plate 
thickness modified to 7 mm), the binomial distribution results for all waste package types 
indicate the individual and total waste package flooding probabilities are always below the 
regulatory probability criterion. Based on these results, it is expected that the absorber plate 
thickness of the 44-BWR Waste Package will be modified to increase its thickness to 7 mm. The 
criticality FEPs screening decisions of Section 6.8 will be based on the results from the 44-BWR 
Waste Package evaluation with the 7-mm absorber plate thickness. 
114 of 176 
Waste Package Type 
calculated as: 
ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Notes: 
Name: 
Table 6.7-2. Binomial Distribution Waste Package Flooding Probability 
Number of 
Waste 
Packages 
4,299 
95 
169 
10,890 
10,890 
Waste Package Type 
Total (all waste package types, 44-BWR 
Seismic Sensitivity Caseb) 
a
b 
6.8.1 FEP 2.1.14.15.0A Screening Discussion 
2.1.14.15.0A Number: 
Description: 
ANL-EBS-NU-000008 REV 00 
Waste Package Flooding Probability 
Number of Flooded Waste Packages 
1 2 3 4 
21-PWR with Absorber Plates 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
21-PWR with Control Rods 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
12-PWR Long 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
44-BWR (Seismic Base Casea) 2,831 4.06E-03 8.27E-06 1.12E-08 1.14E-11 
44-BWR (Seismic Sensitivity Caseb) 2,831 3.03E-07 4.59E-14 4.63E-21 3.50E-28 
24-BWR 84 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
DOE SNF 3,412 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
Total (21-PWR with Absorber Plates) 4,299 0.00E+00 0.00E+00 0.00E+00 0.00E+00 
Total (all waste package types, 44-BWR 4.06E-03 Seismic Base Casea) 8.27E-06 1.12E-08 1.14E-11 
3.03E-07 4.59E-14 4.63E-21 3.50E-28 
Source: Microsoft Excel spreadsheet Binom Dist.xls results (Attachment VII) 
44-BWR Waste Package Absorber Plate thickness is as designed at 5 mm. 
44-BWR Waste Package absorber plate thickness is modified to 7 mm. 
6.8 CRITICALITY FEPS SCREENING DECISIONS 
A discussion of the screening decisions for each of the sixteen criticality FEPs follows. 
In-package criticality (intact configuration) 
The waste package internal structures and the waste form remain intact. 
A breach (or breaches) in the waste package allow(s) water to either 
accumulate or flow-through the waste package. Criticality then occurs 
in situ. In-package criticality resulting from disruptive events is 
addressed in separate FEPs. 
Descriptor Phrases: Criticality (in waste package) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
February 2004 115 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Waste form criticality analyses demonstrate that an intact, fully flooded 
with water (a neutron moderator), waste package configuration cannot 
achieve criticality (CRWMS M&O 1999 [DIRS 125206], CRWMS 
M&O 2000 [DIRS 147650], CRWMS M&O 2000 [DIRS 147651], BSC 
2003 [DIRS 166610], CRWMS M&O 2000 [DIRS 151742], CRWMS 
M&O 2000 [DIRS 151743], CRWMS M&O 2001 [DIRS 154194], BSC 
2001 [DIRS 157733], BSC 2001 [DIRS 157734], BSC 2001 [DIRS 
161125]). Additionally, intact, fully loaded, fully flooded waste 
packages are precluded from achieving criticality by design to satisfy a 
preclosure operations requirement that the MGR provide means to 
ensure criticality control during SNF/HLW handling operations, 
including waste package loading (Siddoway 2003 [DIRS 163904], 
Requirement 1.1.6-4). 
Therefore, the probability of criticality for a nominal waste package 
configuration is zero (refer to Section 6.2 of this report). This result is 
applicable for all waste form / waste package types 
Not Applicable TSPA Disposition: 
Supporting Reports:  21 PWR Waste Package with Absorber Plates Loading Curve 
Evaluation (BSC 2003 [DIRS 166610]) 
 44-BWR Waste Package Loading Curve Evaluation (BSC 2001 
[DIRS 161125]) 
 Evaluation of Codisposal Viability for MOX (FFTF) DOE-Owned 
Fuel (CRWMS M&O 1999 [DIRS 125206]) 
 Evaluation of Codisposal Viability for UzrH (TRIGA) DIE-Owned 
Fuel (CRWMS M&O 2000 [DIRS 147650]) 
 Evaluation of Codisposal Viability for HEU Oxide (Shippingport 
PWR) DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 147651]) 
 Evaluation of Codisposal Viability for U-Zr/U-Mo (Enrico Fermi) 
DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 151742]) 
 Evaluation of Codisposal Viability for Th/U Oxide (Shippingport 
LWBR) DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 151743]) 
 Evaluation of Codisposal Viability for U-Metal (N Reactor) DOEOwned 
Fuel (CRWMS M&O 2001 [DIRS 154194]) 
 Evaluation of Codisposal Viability for Melt and Dilute DOE-Owned 
Fuel (BSC 2001 [DIRS 157733]) 
 Evaluation of Codisposal Viability for Th/U Carbide (Fort Saint 
Vrain HTGR) DOE-Owned Fuel (BSC 2001 [DIRS 157734]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
February 2004 116 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
6.8.2 FEP 2.1.14.16.0A Screening Discussion 
In-package criticality (degraded configuration) Name: 
2.1.14.16.0A Number: 
Description: The waste package internal structures and the waste form degrade. A 
critical configuration (sufficient fissile material and neutron moderator, 
lack of neutron absorbers) develops and criticality occurs in situ. 
Potential in situ critical configurations are defined in Figures 3.2a and 
3.2b of Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]). In-package criticality resulting from 
disruptive events is addressed in separate FEPs. 
Descriptor Phrases: Criticality (in waste package) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
All base case (i.e., nominal scenario class, no disruptive events) 
postclosure criticality FEPs, internal and external, require water 
infiltration to degrade the waste package internals and waste form. 
Neutron absorber material loss and a flooded waste package condition 
for neutron moderation is the most likely scenario that could result in a 
potentially critical configuration in any of the in situ criticality FEPs. 
External criticality FEPs (near-field and far-field) also require the 
separation of neutron absorber materials from the waste form, transport 
of fissile material from the waste package, and accumulation of the 
fissile material within the drift invert or beyond. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can enter the waste package as 
percolation flow or be present in the pores of the rock. Silica is present 
in appreciable quantities in the high-level radioactive waste glass 
canisters and in the rock. Silica can also be introduced into the waste 
package through precipitation from the percolation flow. Carbon is 
present in only limited amounts in select DOE SNF types and, therefore, 
has a limited impact on the potential for criticality. The loading of the 
DOE-standardized SNF canisters, the design of the basket structure 
February 2004 117 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
inside the canisters, and the addition of neutron absorbers take into 
account the presence and effect of glass in DOE SNF codisposal waste 
packages. Silica from the high-level radioactive waste glass canisters 
has no impact on the potential for criticality in DOE SNF codisposal 
waste packages. Silica is a much less effective moderator than water 
and its introduction into commercial SNF waste packages will displace 
water and effectively reduce the reactivity of the system, thus reducing 
the potential for criticality. 
In addition, criticality without water infiltration is unlikely for the 
geologic repository because the waste package is designed so a 
criticality event in an intact waste package configuration is not possible. 
This satisfies a preclosure operations requirement that the MGR provide 
means to ensure criticality control during SNF/HLW handling 
operations, including waste package loading (Siddoway 2003 [DIRS 
163904], Requirement 1.1.6-4). 
Some of the DOE SNF waste forms have highly enriched fuel or a waste 
form that could potentially support unmoderated (fast) criticality if (1) 
the (fissile) material is concentrated beyond its design concentration in 
the waste form, and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration 
could result from either the degradation of the waste form resulting from 
water infiltration or a disruptive event. However, removal of the 
neutron absorber materials from a DOE SNF waste package would 
require a breach of the waste package and a removal mechanism. The 
most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal 
components, dissolving of the neutron absorber material in the water, 
and flushing of the material from the waste package. Since water 
infiltration does not occur, fast criticality can be excluded based on low 
probability. 
The probability of criticality estimate accounts for factors such as early 
failures, manufacturing defects, fuel assembly misloads, etc. However, 
it may not be necessary to directly account for these factors if the total 
probability of criticality is calculated to be sufficiently below the 
regulatory probability criterion without utilizing them. For example, if 
the waste package flooding calculated probability is below the 
regulatory probability criterion, incorporation of the probability of a 
waste package misload would only result in a lower probability than has 
already been calculated. 
Fuel assembly misloads (enrichment and/or burnup) could result in more 
(or less) fissile material being loaded into the waste package than 
permitted by the design loading curves. Additional fissile material in 
the waste package results in a higher criticality potential of the in- 
February 2004 118 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
package degraded configurations. However, because no water (i.e., 
neutron moderator) enters the waste package for base case criticality 
FEP conditions, the additional fissile material that could result from a 
fuel assembly misload cannot result in the formation of a critical 
configuration. 
Because the probability function for drip shield damage area is zero for 
the base case (i.e., no drip shield failures [refer to Section 6.3.4.1]), 
there is no advective flow path into the waste package. Waste package 
failures result only from early waste package failure mechanisms (refer 
to Section 6.3.3.3 of this report). Therefore, the probability of waste 
package flooding is zero and the probability of criticality for this base 
case FEP is zero (refer to Table 6.7-2 of this report). Based on 
assumptions requiring confirmation, this result is applicable for all waste 
form / waste package types. 
Not Applicable TSPA Disposition: 
Supporting Reports:  Configuration Generator Model for In-Package Criticality (BSC 
2003 [DIRS 165629]) 
 Stress Corrosion Cracking of the Drip Shield, the Waste Package 
Outer Barrier, and the Stainless Steel Structural Material (BSC 
2003 [DIRS 161234]) 
 General Corrosion and Localized Corrosion of the Drip Shield 
(BSC 2003 [DIRS 161236]) 
 Analysis of Mechanisms for Early Waste Package/Drip Shield 
Failure (BSC 2003 [DIRS 164475]) 
 WAPDEG Analysis of Waste Package and Drip Shield Degradation 
(BSC 2003 [DIRS 161317]) 
 EBS Radionuclide Transport Abstraction (BSC 2003 [DIRS 
166466]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
6.8.3 FEP 2.1.14.17.0A Screening Discussion 
Near-field criticality Name: 
2.1.14.17.0A Number: 
Description: Near-field criticality occurs when fissile material-bearing solution from 
the waste package is transported into the drift and the fissile material is 
precipitated into a critical configuration. Potential near-field critical 
February 2004 119 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
configurations are defined in Figure 3.3a of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). Inpackage 
criticality resulting from disruptive events is addressed in 
separate FEPs. 
Descriptor Phrases: Criticality (in drift) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
All base case (i.e., nominal scenario class, no disruptive events) 
postclosure criticality FEPs, internal and external, require water 
infiltration to degrade the waste package internals and waste form. 
Neutron absorber material loss and a flooded waste package condition 
for neutron moderation is the most likely scenario that could result in a 
potentially critical configuration in any of the in situ criticality FEPs. 
External criticality FEPs (near-field and far-field) also require the 
separation of neutron absorber materials from the waste form, transport 
of fissile material from the waste package, and accumulation of the 
fissile material within the drift invert or beyond. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can enter the waste package as 
percolation flow or be present in the pores of the rock. Silica is present 
in appreciable quantities in the high-level radioactive waste glass 
canisters and in the rock. Silica can also be introduced into the waste 
package through precipitation from the percolation flow. Carbon is 
present in only limited amounts in select DOE SNF types and, therefore, 
has a limited impact on the potential for criticality. The loading of the 
DOE-standardized SNF canisters, the design of the basket structure 
inside the canisters, and the addition of neutron absorbers take into 
account the presence and effect of glass in DOE SNF codisposal waste 
packages. Silica from the high-level radioactive waste glass canisters 
has no impact on the potential for criticality in DOE SNF codisposal 
waste packages. Silica is a much less effective moderator than water 
and its introduction into commercial SNF waste packages will displace 
water and effectively reduce the reactivity of the system, thus reducing 
the potential for criticality. 
February 2004 120 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
In addition, criticality without water infiltration is unlikely for the 
geologic repository because the waste package is designed such that a 
criticality event in an intact waste package configuration is not possible. 
This satisfies a preclosure operations requirement that the MGR provide 
means to ensure criticality control during SNF/HLW handling 
operations, including waste package loading (Siddoway 2003 [DIRS 
163904], Requirement 1.1.6-4). 
Some of the DOE SNF waste forms have highly enriched fuel or waste 
form that could potentially support unmoderated (fast) criticality if (1) 
the (fissile) material is concentrated beyond its design concentration in 
the waste form and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration 
could result from either the degradation of the waste form resulting from 
water infiltration or a disruptive event. However, removal of the 
neutron absorber materials from a DOE SNF waste package would 
require a breach of the waste package and a removal mechanism. The 
most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal 
components, dissolving of the neutron absorber material in the water, 
and flushing of the material from the waste package. Since water 
infiltration does not occur, fast criticality can be excluded based on low 
probability. 
The probability of criticality estimate accounts for factors such as early 
failures, manufacturing defects, fuel assembly misloads, etc. However, 
it may not be necessary to directly account for these factors if the total 
probability of criticality is calculated to be sufficiently below the 
regulatory probability criterion without utilizing them. For example, if 
the calculated probability of waste package flooding is below the 
regulatory probability criterion, incorporation of the probability of a 
waste package misload would only result in a lower probability than has 
already been calculated. 
Fuel assembly misloads (enrichment and/or burnup) could result in more 
(or less) fissile material being loaded into the waste package than 
permitted by the design loading curves. Additional fissile material in 
the waste package results in a higher criticality potential of the inpackage 
degraded configurations. However, because no water (i.e., 
neutron moderator) enters the waste package for the base case criticality 
FEPs conditions, the additional fissile material that could result from a 
fuel assembly misload cannot result in the formation of a critical 
configuration. 
It then follows that the probability of external criticality must be less 
than the probability of waste package flooding. This is because, in 
addition to the events evaluated to calculate the probability of water 
February 2004 121 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
entering a failed waste package, the probability of the following events 
must also be considered for external criticality: 
 Waste form degradation over the performance period; 
 Separating the fissile materials from the degraded waste form; 
 Removing the fissile materials from the waste package; 
 Accumulating sufficient fissile material into a potentially critical 
configuration in the near-field or far-field environments; and 
 Having sufficient neutron moderator available. 
Because the probability function for drip shield damage area is zero for 
the base case (i.e., no drip shield failures [refer to Section 6.3.4.1 of this 
report]), thereby preventing an advective flow path into the waste 
package, the probability of waste package flooding is zero. Since no 
water can enter the waste package to degrade the waste package 
internals or waste form, no fissile material can be transported from the 
waste package and into the near-field environment. Therefore, the 
probability of criticality for this base case FEP is zero (refer to 
Section 6.3.4.2 of this report). Based on assumptions requiring 
confirmation, this result is applicable for all waste form / waste package 
types. 
Not Applicable TSPA Disposition: 
Supporting Reports:  Geochemistry Model Abstraction and Sensitivity Studies for the 21 
PWR CSNF Waste Packages (BSC 2002 [DIRS 160638]) 
 External Accumulation of Fissile Material from DOE Co-Disposal 
Waste Packages (BSC 2002 [DIRS 159913]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report (YMP 
2003 [DIRS 165505]) 
6.8.4 FEP 2.2.14.09.0A Screening Discussion 
Far-field criticality Name: 
2.2.14.09.0A Number: 
Description: Far-field criticality occurs when fissile material-bearing solution from 
the waste package is transported beyond the drift and the fissile material 
is precipitated into a critical configuration. Potential far-field critical 
configurations are defined in Figure 3.3b of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). Inpackage 
criticality resulting from disruptive events is addressed in 
February 2004 122 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
separate FEPs. 
Descriptor Phrases: Criticality (in the geosphere) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
All base case (i.e., nominal scenario class, no disruptive events) 
postclosure criticality FEPs, internal and external, require water 
infiltration to degrade the waste package internals and waste form. 
Neutron absorber material loss and a flooded waste package condition 
for neutron moderation is the most likely scenario that could result in a 
potentially critical configuration in any of the in situ criticality FEPs. 
External criticality FEPs (near-field and far-field) also require the 
separation of neutron absorber materials from the waste form, transport 
of fissile material from the waste package, and accumulation of the 
fissile material within the drift invert or beyond. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can enter the waste package as 
percolation flow or be present in the pores of the rock. Silica is present 
in appreciable quantities in the high-level radioactive waste glass 
canisters and in the rock. Silica can also be introduced into the waste 
package through precipitation from the percolation flow. Carbon is 
present in only limited amounts in select DOE SNF types and, therefore, 
has a limited impact on the potential for criticality. The loading of the 
DOE-standardized SNF canisters, the design of the basket structure 
inside the canisters, and the addition of neutron absorbers take into 
account the presence and effect of glass in DOE SNF codisposal waste 
packages. Silica from the high-level radioactive waste glass canisters 
has no impact on the potential for criticality in DOE SNF codisposal 
waste packages. Silica is a much less effective moderator than water 
and its introduction into commercial SNF waste packages will displace 
water and effectively reduce the reactivity of the system, thus reducing 
the potential for criticality. 
In addition, criticality without water infiltration is unlikely for the 
geologic repository because the waste package is designed such that a 
February 2004 123 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
criticality event in an intact waste package configuration is not possible. 
This satisfies a preclosure operations requirement that the MGR provide 
means to ensure criticality control during SNF/HLW handling 
operations, including waste package loading (Siddoway 2003 [DIRS 
163904], Requirement 1.1.6-4). 
Some of the DOE SNF waste forms have highly enriched fuel or waste 
form that could potentially support unmoderated (fast) criticality if (1) 
the (fissile) material is concentrated beyond its design concentration in 
the waste form and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration 
could result from either the degradation of the waste form resulting from 
water infiltration or a disruptive event. However, removal of the 
neutron absorber materials from a DOE SNF waste package would 
require a breach of the waste package and a removal mechanism. The 
most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal 
components, dissolving of the neutron absorber material in the water, 
and flushing of the material from the waste package. Since water 
infiltration does not occur, fast criticality can be excluded based on low 
probability. 
The probability of criticality estimate accounts for factors such as early 
failures, manufacturing defects, fuel assembly misloads, etc. However, 
it may not be necessary to directly account for these factors if the total 
probability of criticality is calculated to be sufficiently below the 
regulatory probability criterion without utilizing them. For example, if 
the calculated probability of waste package flooding is below the 
regulatory probability criterion, incorporation of the probability of a 
waste package misload would only result in a lower probability than has 
already been calculated. 
Fuel assembly misloads (enrichment and/or burnup) could result in more 
(or less) fissile material being loaded into the waste package than 
permitted by the design loading curves. Additional fissile material in 
the waste package results in a higher criticality potential of the inpackage 
degraded configurations. However, because no water (i.e., 
neutron moderator) enters the waste package for the base case criticality 
FEPs conditions, the additional fissile material that could result from a 
fuel assembly misload cannot result in the formation of a critical 
configuration. 
It then follows that the probability of external criticality must be less 
than the probability of waste package flooding. This is because, in 
addition to the events evaluated to calculate the probability of water 
entering a failed waste package, the probability of the following events 
must also be considered for external criticality: 
February 2004 124 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
 Waste form degradation over the performance period; 
 Separating the fissile materials from the degraded waste form; 
 Removing the fissile materials from the waste package; 
 Accumulating sufficient fissile material into a potentially critical 
configuration in the near-field or far-field environments; and 
 Having sufficient neutron moderator available. 
Because the probability function for drip shield damage areas is zero for 
the base case (i.e., no drip shield failures [refer to Section 6.3.4.1 of this 
report]), thereby preventing an advective flow path into the waste 
package, the probability of waste package flooding is zero [refer to 
Section 6.3.4.2 of this report]. Since water cannot enter the waste 
package to degrade the waste package internals and waste form, fissile 
material cannot be transported from the waste package and into the farfield 
environment. Therefore, the probability of criticality for this base 
case FEP is zero. Based on assumptions requiring confirmation, this 
result is applicable for all waste form / waste package types. 
Not Applicable TSPA Disposition: 
Supporting Reports:  Geochemistry Model Abstraction and Sensitivity Studies for the 21 
PWR CSNF Waste Packages (BSC 2002 [DIRS 160638]) 
 External Accumulation of Fissile Material from DOE Co-Disposal 
Waste Packages (BSC 2002 [DIRS 159913]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
6.8.5 FEP 2.1.14.18.0A Screening Discussion 
Name: In-package criticality resulting from a seismic event (intact 
configuration) 
2.1.14.18.0A Number: 
Description: The waste package internal structures and the waste form remain intact 
either during or after a seismic disruptive event. A breach (or breaches) 
in the waste package allow(s) water to either accumulate or flowthrough 
the waste package. Criticality then occurs in situ. 
Descriptor Phrases: Criticality (in waste package), Criticality (from a seismic event) 
February 2004 125 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
Waste form criticality analyses demonstrate that an intact, fully flooded 
with water (a neutron moderator), waste package configuration cannot 
achieve criticality (CRWMS M&O 1999 [DIRS 125206], CRWMS 
M&O 2000 [DIRS 147650], CRWMS M&O 2000 [DIRS 147651], BSC 
2003 [DIRS 166610], CRWMS M&O 2000 [DIRS 151742], CRWMS 
M&O 2000 [DIRS 151743], CRWMS M&O 2001 [DIRS 154194], BSC 
2001 [DIRS 157733], BSC 2001 [DIRS 157734], BSC 2001 [DIRS 
161125]). Additionally, intact, fully loaded, fully flooded waste 
packages are precluded from achieving criticality by design to satisfy a 
preclosure operations requirement that the MGR provide means to 
ensure criticality control during SNF/HLW handling operations, 
including waste package loading (Siddoway 2003 [DIRS 163904], 
Requirement 1.1.6-4). 
Therefore, the probability of criticality for a nominal waste package 
configuration for the seismic disruptive event is zero. This result is 
applicable for all waste form / waste package types 
Not Applicable TSPA Disposition: 
Supporting Reports:  21 PWR Waste Package with Absorber Plates Loading Curve 
Evaluation (BSC 2003 [DIRS 166610]) 
 44-BWR Waste Package Loading Curve Evaluation (BSC 2001 
[DIRS 161125]) 
 Evaluation of Codisposal Viability for MOX (FFTF) DOE-Owned 
Fuel (CRWMS M&O 1999 [DIRS 125206]) 
 Evaluation of Codisposal Viability for UzrH (TRIGA) DIE-Owned 
Fuel (CRWMS M&O 2000 [DIRS 147650]) 
 Evaluation of Codisposal Viability for HEU Oxide (Shippingport 
PWR) DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 147651]) 
 Evaluation of Codisposal Viability for U-Zr/U-Mo (Enrico Fermi) 
DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 151742]) 
 Evaluation of Codisposal Viability for Th/U Oxide (Shippingport 
LWBR) DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 151743]) 
 Evaluation of Codisposal Viability for U-Metal (N Reactor) DOEOwned 
Fuel (CRWMS M&O 2001 [DIRS 154194]) 
February 2004 126 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
 Evaluation of Codisposal Viability for Melt and Dilute DOE-Owned 
Fuel (BSC 2001 [DIRS 157733]) 
 Evaluation of Codisposal Viability for Th/U Carbide (Fort Saint 
Vrain HTGR) DOE-Owned Fuel (BSC 2001 [DIRS 157734]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
6.8.6 FEP 2.1.14.19.0A Screening Discussion 
Name: In-package criticality resulting from a seismic event (degraded 
configuration) 
2.1.14.19.0A Number: 
Description: Either during, or as a result of, a seismic disruptive event, the waste 
package internal structures and the waste form degrade. A critical 
configuration develops and criticality occurs in situ. Potential in situ 
critical configurations are defined in Figures 3.2a and 3.2b of Disposal 
Criticality Analysis Methodology Topical Report (YMP 2003 [DIRS 
165505]). 
Descriptor Phrases: Criticality (in waste package), Criticality (from a seismic event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
All base case (i.e., nominal scenario class, no disruptive events) 
postclosure criticality FEPs, internal and external, require water 
infiltration to degrade the waste package internals and waste form. 
Neutron absorber material loss and a flooded waste package condition 
for neutron moderation is the most likely scenario that could result in a 
potentially critical configuration in any of the in situ criticality FEPs. 
External criticality FEPs (near-field and far-field) also require the 
separation of neutron absorber materials from the waste form, transport 
February 2004 127 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
of fissile material from the waste package, and accumulation of the 
fissile material within the drift invert or beyond. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can enter the waste package as 
percolation flow or be present in the pores of the rock. Silica is present 
in appreciable quantities in the high-level radioactive waste glass 
canisters and in the rock. Silica can also be introduced into the waste 
package through precipitation from the percolation flow. Carbon is 
present in only limited amounts in select DOE SNF types and, therefore, 
has a limited impact on the potential for criticality. The loading of the 
DOE-standardized SNF canisters, the design of the basket structure 
inside the canisters, and the addition of neutron absorbers take into 
account the presence and effect of glass in DOE SNF codisposal waste 
packages. Silica from the high-level radioactive waste glass canisters 
has no impact on the potential for criticality in DOE SNF codisposal 
waste packages. Silica is a much less effective moderator than water 
and its introduction into commercial SNF waste packages will displace 
water and effectively reduce the reactivity of the system, thus reducing 
the potential for criticality. 
In addition, criticality without water infiltration is unlikely for the 
geologic repository because the waste package is designed such that a 
criticality event in an intact waste package configuration is not possible. 
This satisfies a preclosure operations requirement that the MGR provide 
means to ensure criticality control during SNF/HLW handling 
operations, including waste package loading (Siddoway 2003 [DIRS 
163904], Requirement 1.1.6-4). 
Some of the DOE SNF waste forms have highly enriched fuel or waste 
form that could potentially support unmoderated (fast) criticality if (1) 
the (fissile) material is concentrated beyond its design concentration in 
the waste form and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration 
could result from either the degradation of the waste form resulting from 
water infiltration or a disruptive event. However, removal of the 
neutron absorber materials from a DOE SNF waste package would 
require a breach of the waste package and a removal mechanism. The 
most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal 
components, dissolving of the neutron absorber material in the water, 
and flushing of the material from the waste package. Since water 
infiltration does not occur, fast criticality can be excluded based on low 
probability. 
February 2004 128 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The probability of criticality estimate accounts for factors such as early 
failures, manufacturing defects, fuel assembly misloads, etc. However, 
it may not be necessary to directly account for these factors if the total 
probability of criticality is calculated to be sufficiently below the 
regulatory probability criterion without utilizing them. For example, if 
the calculated probability of waste package flooding is below the 
regulatory probability criterion, incorporation of the probability of a 
waste package misload would only result in a lower probability than has 
already been calculated. 
Fuel assembly misloads (enrichment and/or burnup) could result in more 
(or less) fissile material being loaded into the waste package than 
permitted by the design loading curves. Additional fissile material in 
the waste package results in a higher criticality potential of the inpackage 
degraded configurations. However, because the probability of 
removing sufficient neutron absorber material from the waste package is 
below the regulatory probability threshold for the seismic criticality 
FEPs, the additional fissile material that could result from a fuel 
assembly misload cannot result in the formation of a critical 
configuration. 
A seismic event results in a probability of drip shield and waste package 
failure of 1.0, allowing an advective flow to enter the waste package. 
However, the probability of sufficient seepage flux penetrating at least 
one waste package, degrading the waste package internals, and flushing 
the neutron absorber material from the waste package is only 3.03  10-7 
during the performance period (refer to Table 6.7-2 of this report). 
Based on assumptions requiring confirmation, this result is applicable to 
all waste form / waste package types. However, it should be noted that 
the 44-BWR waste package type is the only nonzero contributor to this 
total probability. 
Not Applicable TSPA Disposition: 
Supporting Reports:  Configuration Generator Model for In-Package Criticality (BSC 
2003 [DIRS 165629]) 
 Stress Corrosion Cracking of the Drip Shield, the Waste Package 
Outer Barrier, and the Stainless Steel Structural Material (BSC 
 General Corrosion and Localized Corrosion of the Drip Shield 
 Analysis of Mechanisms for Early Waste Package/Drip Shield 
 WAPDEG Analysis of Waste Package and Drip Shield Degradation 
 EBS Radionuclide Transport Abstraction (BSC 2003 [DIRS 
2003 [DIRS 161234]) 
(BSC 2003 [DIRS 161236]) 
Failure (BSC 2003 [DIRS 164475]) 
(BSC 2003 [DIRS 161317]) 
166466]) 
February 2004 129 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
 Abstraction of Drift Seepage (BSC 2003 [DIRS 165564]) 
 Analysis of Infiltration Uncertainty (BSC 2003 [DIRS 165991]) 
 Seismic Consequence Abstraction (BSC 2003 [DIRS 161812]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
6.8.7 FEP 2.1.14.20.0A Screening Discussion 
Near-field criticality resulting from a seismic event Name: 
2.1.14.20.0A Number: 
Description: Either during, or as a result of, a seismic disruptive event, near-field 
criticality occurs when fissile material-bearing solution from the waste 
package is transported into the drift and the fissile material is 
precipitated into a critical configuration. Potential near-field critical 
configurations are defined in Figure 3.3a of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). 
Descriptor Phrases: Criticality (in drift), Criticality (from a seismic event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
All base case (i.e., nominal scenario class, no disruptive events) 
postclosure criticality FEPs, internal and external, require water 
infiltration to degrade the waste package internals and waste form. 
Neutron absorber material loss and a flooded waste package condition 
for neutron moderation is the most likely scenario that could result in a 
potentially critical configuration in any of the in situ criticality FEPs. 
External criticality FEPs (near-field and far-field) also require the 
separation of neutron absorber materials from the waste form, transport 
of fissile material from the waste package, and accumulation of the 
fissile material within the drift invert or beyond. 
February 2004 130 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can enter the waste package as 
percolation flow or be present in the pores of the rock. Silica is present 
in appreciable quantities in the high-level radioactive waste glass 
canisters and in the rock. Silica can also be introduced into the waste 
package through precipitation from the percolation flow. Carbon is 
present in only limited amounts in select DOE SNF types and, therefore, 
has a limited impact on the potential for criticality. The loading of the 
DOE-standardized SNF canisters, the design of the basket structure 
inside the canisters, and the addition of neutron absorbers take into 
account the presence and effect of glass in DOE SNF codisposal waste 
packages. Silica from the high-level radioactive waste glass canisters 
has no impact on the potential for criticality in DOE SNF codisposal 
waste packages. Silica is a much less effective moderator than water 
and its introduction into commercial SNF waste packages will displace 
water and effectively reduce the reactivity of the system, thus reducing 
the potential for criticality. 
In addition, criticality without water infiltration is unlikely for the 
geologic repository because the waste package is designed such that a 
criticality event in an intact waste package configuration is not possible. 
This satisfies a preclosure operations requirement that the MGR provide 
means to ensure criticality control during SNF/HLW handling 
operations, including waste package loading (Siddoway 2003 [DIRS 
163904], Requirement 1.1.6-4). 
Some of the DOE SNF waste forms have highly enriched fuel or waste 
form that could potentially support unmoderated (fast) criticality if (1) 
the (fissile) material is concentrated beyond its design concentration in 
the waste form and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration 
could result from either the degradation of the waste form resulting from 
water infiltration or a disruptive event. However, removal of the 
neutron absorber materials from a DOE SNF waste package would 
require a breach of the waste package and a removal mechanism. The 
most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal 
components, dissolving of the neutron absorber material in the water, 
and flushing of the material from the waste package. Since water 
infiltration does not occur, fast criticality can be excluded based on low 
probability. 
The probability of criticality estimate accounts for factors such as early 
failures, manufacturing defects, fuel assembly misloads, etc. However, 
it may not be necessary to directly account for these factors if the total 
probability of criticality is calculated to be sufficiently below the 
regulatory probability criterion without utilizing them. For example, if 
February 2004 131 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
the calculated probability of waste package flooding is below the 
regulatory probability criterion, incorporation of the probability of a 
waste package misload would only result in a lower probability than has 
already been calculated. 
Fuel assembly misloads (enrichment and/or burnup) could result in more 
(or less) fissile material being loaded into the waste package than 
permitted by the design loading curves. Additional fissile material in 
the waste package results in a higher criticality potential of the inpackage 
degraded configurations. However, because the probability of 
removing sufficient neutron absorber material from the waste package is 
below the regulatory probability threshold for the seismic criticality 
FEPs conditions, the additional fissile material that could result from a 
fuel assembly misload cannot result in the formation of a critical 
configuration. 
It then follows that the probability of external criticality must be less 
than the probability of waste package flooding. This is because, in 
addition to the events evaluated to calculate the probability of water 
entering a failed waste package, the probability of the following events 
must also be considered for external criticality: 
 Waste form degradation over the performance period; 
 Separating the fissile materials from the degraded waste form; 
 Removing the fissile materials from the waste package; 
 Accumulating sufficient fissile material into a potentially critical 
configuration in the near-field or far-field environments; and 
 Having sufficient neutron moderator available. 
The probability of waste package flooding for a seismic disruptive event 
is below the regulatory probability criterion. Given the considerations 
listed above, the probability of criticality in the near-field environment 
will be even smaller. Therefore, this seismic disruptive event FEP can 
be excluded based on low probability (refer to Section 6.4.4 of this 
report). Based on assumptions requiring confirmation, this result is 
applicable for all waste form / waste package types. 
Not Applicable TSPA Disposition: 
Supporting Reports:  Geochemistry Model Abstraction and Sensitivity Studies for the 21 
PWR CSNF Waste Packages (BSC 2002 [DIRS 160638]) 
 External Accumulation of Fissile Material from DOE Co-Disposal 
Waste Packages (BSC 2002 [DIRS 159913]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
February 2004 132 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
6.8.8 FEP 2.2.14.10.0A Screening Discussion 
Far-field criticality resulting from a seismic event Name: 
2.2.14.10.0A Number: 
Description: Either during, or as a result of, a seismic disruptive event, far-field 
criticality occurs when fissile material-bearing solution from the waste 
package is transported beyond the drift and the fissile material is 
precipitated into a critical configuration. Potential far-field critical 
configurations are defined in Figure 3.3b of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). 
Descriptor Phrases: Criticality (in the geosphere), Criticality (from a seismic event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
All base case (i.e., nominal scenario class, no disruptive events) 
postclosure criticality FEPs, internal and external, require water 
infiltration to degrade the waste package internals and waste form. 
Neutron absorber material loss and a flooded waste package condition 
for neutron moderation is the most likely scenario that could result in a 
potentially critical configuration in any of the in situ criticality FEPs. 
External criticality FEPs (near-field and far-field) also require the 
separation of neutron absorber materials from the waste form, transport 
of fissile material from the waste package, and accumulation of the 
fissile material within the drift invert or beyond. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can enter the waste package as 
percolation flow or be present in the pores of the rock. Silica is present 
in appreciable quantities in the high-level radioactive waste glass 
canisters and in the rock. Silica can also be introduced into the waste 
package through precipitation from the percolation flow. Carbon is 
present in only limited amounts in select DOE SNF types and, therefore, 
has a limited impact on the potential for criticality. The loading of the 
DOE-standardized SNF canisters, the design of the basket structure 
inside the canisters, and the addition of neutron absorbers take into 
February 2004 133 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
account the presence and effect of glass in DOE SNF codisposal waste 
packages. Silica from the high-level radioactive waste glass canisters 
has no impact on the potential for criticality in DOE SNF codisposal 
waste packages. Silica is a much less effective moderator than water 
and its introduction into commercial SNF waste packages will displace 
water and effectively reduce the reactivity of the system, thus reducing 
the potential for criticality. 
In addition, criticality without water infiltration is unlikely for the 
geologic repository because the waste package is designed such that a 
criticality event in an intact waste package configuration is not possible. 
This satisfies a preclosure operations requirement that the MGR provide 
means to ensure criticality control during SNF/HLW handling 
operations, including waste package loading (Siddoway 2003 [DIRS 
163904], Requirement 1.1.6-4). 
Some of the DOE SNF waste forms have highly enriched fuel or waste 
form that could potentially support unmoderated (fast) criticality if (1) 
the (fissile) material is concentrated beyond its design concentration in 
the waste form and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration 
could result from either the degradation of the waste form resulting from 
water infiltration or a disruptive event. However, removal of the 
neutron absorber materials from a DOE SNF waste package would 
require a breach of the waste package and a removal mechanism. The 
most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal 
components, dissolving of the neutron absorber material in the water, 
and flushing of the material from the waste package. Since water 
infiltration does not occur, fast criticality can be excluded based on low 
probability. 
The probability of criticality estimate accounts for factors such as early 
failures, manufacturing defects, fuel assembly misloads, etc. However, 
it may not be necessary to directly account for these factors if the total 
probability of criticality is calculated to be sufficiently below the 
regulatory probability criterion without utilizing them. For example, if 
the calculated probability of waste package flooding is below the 
regulatory probability criterion, incorporation of the probability of a 
waste package misload would only result in a lower probability than has 
already been calculated. 
Fuel assembly misloads (enrichment and/or burnup) could result in more 
(or less) fissile material being loaded into the waste package than 
permitted by the design loading curves. Additional fissile material in 
the waste package results in a higher criticality potential of the inpackage 
degraded configurations. However, because the probability of 
February 2004 134 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
removing sufficient neutron absorber material from the waste package is 
below the regulatory probability threshold for the seismic criticality 
FEPs conditions, the additional fissile material that could result from a 
fuel assembly misload cannot result in the formation of a critical 
configuration. 
It then follows that the probability of external criticality must be less 
than the probability of waste package flooding. This is because, in 
addition to the events evaluated to calculate the probability of water 
entering a failed waste package, the probability of the following events 
must also be considered for external criticality: 
 Waste form degradation over the performance period; 
 Separating the fissile materials from the degraded waste form; 
 the probability of removing the fissile materials from the waste 
package; 
 Accumulating sufficient fissile material into a potentially critical 
configuration in the near-field or far-field environments; and 
 Having sufficient neutron moderator available. 
The probability of waste package flooding for a seismic disruptive event 
is below the regulatory probability criterion. Given the considerations 
listed above, the probability of criticality in the far-field environment 
will be even smaller. Therefore, this seismic disruptive event FEP can 
be excluded based on low probability (refer to Section 6.4.4 of this 
report). Based on assumptions requiring confirmation, this result is 
applicable for all waste form / waste package types. 
Not Applicable TSPA Disposition: 
Supporting Reports:  Geochemistry Model Abstraction and Sensitivity Studies for the 21 
PWR CSNF Waste Packages (BSC 2002 [DIRS 160638]) 
 External Accumulation of Fissile Material from DOE Co-Disposal 
Waste Packages (BSC 2002 [DIRS 159913]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
February 2004 135 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
6.8.9 FEP 2.1.14.21.0A Screening Discussion 
In-package criticality resulting from rockfall (intact configuration) Name: 
2.1.14.21.0A Number: 
Description: The waste package internal structures and the waste form remain intact 
either during or after a rockfall event. A breach (or breaches) in the 
waste package allow(s) water to either accumulate or flow-through the 
waste package. Criticality then occurs in situ. 
Descriptor Phrases: Criticality (in waste package), Criticality (from a rockfall event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
Waste form criticality analyses demonstrate that an intact, fully flooded 
with water (a neutron moderator), waste package configuration cannot 
achieve criticality (CRWMS M&O 1999 [DIRS 125206], CRWMS 
M&O 2000 [DIRS 147650], CRWMS M&O 2000 [DIRS 147651], BSC 
2003 [DIRS 166610], CRWMS M&O 2000 [DIRS 151742], CRWMS 
M&O 2000 [DIRS 151743], CRWMS M&O 2001 [DIRS 154194], BSC 
2001 [DIRS 157733], BSC 2001 [DIRS 157734], BSC 2001 [DIRS 
161125]). Additionally, intact, fully loaded, fully flooded waste 
packages are precluded from achieving criticality by design to satisfy a 
preclosure operations requirement that the MGR provide means to 
ensure criticality control during SNF/HLW handling operations, 
including waste package loading (Siddoway 2003 [DIRS 163904], 
Requirement 1.1.6-4). 
Therefore, the probability of criticality for a nominal waste package 
configuration during a rockfall disruptive event is zero. This result is 
applicable for all waste form / waste package types 
Not Applicable TSPA Disposition: 
Supporting Reports:  21 PWR Waste Package with Absorber Plates Loading Curve 
Evaluation (BSC 2003 [DIRS 166610]) 
 44-BWR Waste Package Loading Curve Evaluation (BSC 2001 
[DIRS 161125]) 
February 2004 136 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
 Evaluation of Codisposal Viability for MOX (FFTF) DOE-Owned 
Fuel (CRWMS M&O 1999 [DIRS 125206]) 
 Evaluation of Codisposal Viability for UzrH (TRIGA) DIE-Owned 
Fuel (CRWMS M&O 2000 [DIRS 147650]) 
 Evaluation of Codisposal Viability for HEU Oxide (Shippingport 
PWR) DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 147651]) 
 Evaluation of Codisposal Viability for U-Zr/U-Mo (Enrico Fermi) 
DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 151742]) 
 Evaluation of Codisposal Viability for Th/U Oxide (Shippingport 
LWBR) DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 151743]) 
 Evaluation of Codisposal Viability for U-Metal (N Reactor) DOEOwned 
Fuel (CRWMS M&O 2001 [DIRS 154194]) 
 Evaluation of Codisposal Viability for Melt and Dilute DOE-Owned 
Fuel (BSC 2001 [DIRS 157733]) 
 Evaluation of Codisposal Viability for Th/U Carbide (Fort Saint 
Vrain HTGR) DOE-Owned Fuel (BSC 2001 [DIRS 157734]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
6.8.10 FEP 2.1.14.22.0A Screening Discussion 
In-package criticality resulting from rockfall (degraded configuration) Name: 
2.1.14.22.0A Number: 
Description: Either during, or as a result of, a rockfall event, the waste package 
internal structures and the waste form degrade. A critical configuration 
develops and criticality occurs in situ. Potential in situ critical 
configurations are defined in Figures 3.2a and 3.2b of Disposal 
Criticality Analysis Methodology Topical Report (YMP 2003 [DIRS 
165505]). 
Descriptor Phrases: Criticality (in waste package), Criticality (from a rockfall event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
February 2004 137 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
All base case (i.e., nominal scenario class, no disruptive events) 
postclosure criticality FEPs, internal and external, require water 
infiltration to degrade the waste package internals and waste form. 
Neutron absorber material loss and a flooded waste package condition 
for neutron moderation is the most likely scenario that could result in a 
potentially critical configuration in any of the in situ criticality FEPs. 
External criticality FEPs (near-field and far-field) also require the 
separation of neutron absorber materials from the waste form, transport 
of fissile material from the waste package, and accumulation of the 
fissile material within the drift invert or beyond. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can enter the waste package as 
percolation flow or be present in the pores of the rock. Silica is present 
in appreciable quantities in the high-level radioactive waste glass 
canisters and in the rock. Silica can also be introduced into the waste 
package through precipitation from the percolation flow. Carbon is 
present in only limited amounts in select DOE SNF types and, therefore, 
has a limited impact on the potential for criticality. The loading of the 
DOE-standardized SNF canisters, the design of the basket structure 
inside the canisters, and the addition of neutron absorbers take into 
account the presence and effect of glass in DOE SNF codisposal waste 
packages. Silica from the high-level radioactive waste glass canisters 
has no impact on the potential for criticality in DOE SNF codisposal 
waste packages. Silica is a much less effective moderator than water 
and its introduction into commercial SNF waste packages will displace 
water and effectively reduce the reactivity of the system, thus reducing 
the potential for criticality. 
In addition, criticality without water infiltration is unlikely for the 
geologic repository because the waste package is designed such that a 
criticality event in an intact waste package configuration is not possible. 
This satisfies a preclosure operations requirement that the MGR provide 
means to ensure criticality control during SNF/HLW handling 
operations, including waste package loading (Siddoway 2003 [DIRS 
163904], Requirement 1.1.6-4). 
Some of the DOE SNF waste forms have highly enriched fuel or waste 
form that could potentially support unmoderated (fast) criticality if (1) 
the (fissile) material is concentrated beyond its design concentration in 
the waste form and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration 
could result from either the degradation of the waste form resulting from 
February 2004 138 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
water infiltration or a disruptive event. However, removal of the 
neutron absorber materials from a DOE SNF waste package would 
require a breach of the waste package and a removal mechanism. The 
most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal 
components, dissolving of the neutron absorber material in the water, 
and flushing of the material from the waste package. Since water 
infiltration does not occur, fast criticality can be excluded based on low 
probability. 
The probability of criticality estimate accounts for factors such as early 
failures, manufacturing defects, fuel assembly misloads, etc. However, 
it may not be necessary to directly account for these factors if the total 
probability of criticality is calculated to be sufficiently below the 
regulatory probability criterion without utilizing them. For example, if 
the calculated probability of waste package flooding is below the 
regulatory probability criterion, incorporation of the probability of a 
waste package misload would only result in a lower probability than has 
already been calculated. 
Fuel assembly misloads (enrichment and/or burnup) could result in more 
(or less) fissile material being loaded into the waste package than 
permitted by the design loading curves. Additional fissile material in 
the waste package results in a higher criticality potential of the inpackage 
degraded configurations. However, because no water (i.e., 
neutron moderator) enters the waste package for the rockfall criticality 
FEPs conditions, the additional fissile material that could result from a 
fuel assembly misload cannot result in the formation of a critical 
configuration. 
The rockfall disruptive event does not result in the formation of an 
advective flow path for water through the drip shield and into the waste 
package. This is because there is no mechanism for drip shield failure 
during this event (refer to Section 6.5.1 of this report). Waste package 
failures result only from early waste package failure mechanisms (refer 
to Section 6.3.3.3 of this report). Therefore, the screening argument for 
the rockfall disruptive event is the same as for the base case criticality 
FEP 2.1.14.16.0A. The probability of waste package flooding is zero 
and the probability of criticality for this rockfall disruptive event FEP is 
zero (refer to Table 6.7-2 of this report). Based on assumptions 
requiring confirmation, this result is applicable for all waste form / 
waste package types. 
Not Applicable TSPA Disposition: 
February 2004 139 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Supporting Reports:  Configuration Generator Model for In-Package Criticality (BSC 
2003 [DIRS 165629]) 
 Stress Corrosion Cracking of the Drip Shield, the Waste Package 
Outer Barrier, and the Stainless Steel Structural Material (BSC 
2003 [DIRS 161234]) 
 General Corrosion and Localized Corrosion of the Drip Shield 
(BSC 2003 [DIRS 161236]) 
 Analysis of Mechanisms for Early Waste Package/Drip Shield 
Failure (BSC 2003 [DIRS 164475]) 
 WAPDEG Analysis of Waste Package and Drip Shield Degradation 
(BSC 2003 [DIRS 161317]) 
 EBS Radionuclide Transport Abstraction (BSC 2003 [DIRS 
166466]) 
 Seismic Consequence Abstraction (BSC 2003 [DIRS 161812]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report (YMP 
2003 [DIRS 165505]) 
6.8.11 FEP 2.1.14.23.0A Screening Discussion 
Near-field criticality resulting from rockfall Name: 
2.1.14.23.0A Number: 
Description: Either during, or as a result of, a rockfall event, near-field criticality 
occurs when fissile material-bearing solution from the waste package is 
transported into the drift and the fissile material is precipitated into a 
critical configuration. Potential near-field critical configurations are 
defined in Figure 3.3a of Disposal Criticality Analysis Methodology 
Topical Report (YMP 2003 [DIRS 165505]). 
Descriptor Phrases: Criticality (in drift), Criticality (from a rockfall event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
February 2004 140 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
All base case (i.e., nominal scenario class, no disruptive events) 
postclosure criticality FEPs, internal and external, require water 
infiltration to degrade the waste package internals and waste form. 
Neutron absorber material loss and a flooded waste package condition 
for neutron moderation is the most likely scenario that could result in a 
potentially critical configuration in any of the in situ criticality FEPs. 
External criticality FEPs (near-field and far-field) also require the 
separation of neutron absorber materials from the waste form, transport 
of fissile material from the waste package, and accumulation of the 
fissile material within the drift invert or beyond. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can enter the waste package as 
percolation flow or be present in the pores of the rock. Silica is present 
in appreciable quantities in the high-level radioactive waste glass 
canisters and in the rock. Silica can also be introduced into the waste 
package through precipitation from the percolation flow. Carbon is 
present in only limited amounts in select DOE SNF types and, therefore, 
has a limited impact on the potential for criticality. The loading of the 
DOE-standardized SNF canisters, the design of the basket structure 
inside the canisters, and the addition of neutron absorbers take into 
account the presence and effect of glass in DOE SNF codisposal waste 
packages. Silica from the high-level radioactive waste glass canisters 
has no impact on the potential for criticality in DOE SNF codisposal 
waste packages. Silica is a much less effective moderator than water 
and its introduction into commercial SNF waste packages will displace 
water and effectively reduce the reactivity of the system, thus reducing 
the potential for criticality. 
In addition, criticality without water infiltration is unlikely for the 
geologic repository because the waste package is designed such that a 
criticality event in an intact waste package configuration is not possible. 
This satisfies a preclosure operations requirement that the MGR provide 
means to ensure criticality control during SNF/HLW handling 
operations, including waste package loading (Siddoway 2003 [DIRS 
163904], Requirement 1.1.6-4). 
Some of the DOE SNF waste forms have highly enriched fuel or waste 
form that could potentially support unmoderated (fast) criticality if (1) 
the (fissile) material is concentrated beyond its design concentration in 
the waste form and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration 
could result from either the degradation of the waste form resulting from 
water infiltration or a disruptive event. However, removal of the 
neutron absorber materials from a DOE SNF waste package would 
require a breach of the waste package and a removal mechanism. The 
February 2004 141 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal 
components, dissolving of the neutron absorber material in the water, 
and flushing of the material from the waste package. Since water 
infiltration does not occur, fast criticality can be excluded based on low 
probability. 
The probability of criticality estimate accounts for factors such as early 
failures, manufacturing defects, fuel assembly misloads, etc. However, 
it may not be necessary to directly account for these factors if the total 
probability of criticality is calculated to be sufficiently below the 
regulatory probability criterion without utilizing them. For example, if 
the calculated probability of waste package flooding is below the 
regulatory probability criterion, incorporation of the probability of a 
waste package misload would only result in a lower probability than has 
already been calculated. 
Fuel assembly misloads (enrichment and/or burnup) could result in more 
(or less) fissile material being loaded into the waste package than 
permitted by the design loading curves. Additional fissile material in 
the waste package results in a higher criticality potential of the inpackage 
degraded configurations. However, because no water (i.e., 
neutron moderator) enters the waste package for the rockfall criticality 
FEPs conditions, the additional fissile material that could result from a 
fuel assembly misload cannot result in the formation of a critical 
configuration. 
It then follows that the probability of external criticality must be less 
than the probability of waste package flooding. This is because, in 
addition to the events evaluated to calculate the probability of water 
entering a failed waste package, the probability of the following events 
must also be considered for external criticality: 
 Waste form degradation over the performance period; 
 Separating the fissile materials from the degraded waste form; 
 Removing the fissile materials from the waste package; 
 Accumulating sufficient fissile material into a potentially critical 
configuration in the near-field or far-field environments; and 
 Having sufficient neutron moderator available. 
Because the probability function for drip shield damage areas is zero for 
the rockfall disruptive event (i.e., no drip shield failures [refer to 
Section 6.5.1 of this report]), the probability of waste package flooding 
is zero for the rockfall disruptive event (refer to Table 6.7-2 of this 
report). Since no water can enter the waste package to degrade the 
waste package internals and waste form, no fissile material can be 
transported from the waste package and into the near-field environment. 
February 2004 142 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Therefore, the probability of criticality for this rockfall disruptive event 
FEP is zero (refer to Section 6.5.5 of this report). Based on assumptions 
requiring confirmation, this result is applicable for all waste form / 
waste package types. 
Not Applicable TSPA Disposition: 
Supporting Reports:  Geochemistry Model Abstraction and Sensitivity Studies for the 21 
PWR CSNF Waste Packages (BSC 2002 [DIRS 160638]) 
 External Accumulation of Fissile Material from DOE Co-Disposal 
Waste Packages (BSC 2002 [DIRS 159913]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
6.8.12 FEP 2.2.14.11.0A Screening Discussion 
Far-field criticality resulting from rockfall Name: 
2.2.14.11.0A Number: 
Description: Either during, or as a result of, a rockfall event, far-field criticality 
occurs when fissile material-bearing solution from the waste package is 
transported beyond the drift and the fissile material is precipitated into a 
critical configuration. Potential far-field critical configurations are 
defined in Figure 3.3b of Disposal Criticality Analysis Methodology 
Topical Report (YMP 2003 [DIRS 165505]). 
Descriptor Phrases: Criticality (in the geosphere), Criticality (from a rockfall event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
All base case (i.e., nominal scenario class, no disruptive events) 
postclosure criticality FEPs, internal and external, require water 
infiltration to degrade the waste package internals and waste form. 
February 2004 143 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Neutron absorber material loss and a flooded waste package condition 
for neutron moderation is the most likely scenario that could result in a 
potentially critical configuration in any of the in situ criticality FEPs. 
External criticality FEPs (near-field and far-field) also require the 
separation of neutron absorber materials from the waste form, transport 
of fissile material from the waste package, and accumulation of the 
fissile material within the drift invert or beyond. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can enter the waste package as 
percolation flow or be present in the pores of the rock. Silica is present 
in appreciable quantities in the high-level radioactive waste glass 
canisters and in the rock. Silica can also be introduced into the waste 
package through precipitation from the percolation flow. Carbon is 
present in only limited amounts in select DOE SNF types and, therefore, 
has a limited impact on the potential for criticality. The loading of the 
DOE-standardized SNF canisters, the design of the basket structure 
inside the canisters, and the addition of neutron absorbers take into 
account the presence and effect of glass in DOE SNF codisposal waste 
packages. Silica from the high-level radioactive waste glass canisters 
has no impact on the potential for criticality in DOE SNF codisposal 
waste packages. Silica is a much less effective moderator than water 
and its introduction into commercial SNF waste packages will displace 
water and effectively reduce the reactivity of the system, thus reducing 
the potential for criticality. 
In addition, criticality without water infiltration is unlikely for the 
geologic repository because the waste package is designed such that a 
criticality event in an intact waste package configuration is not possible. 
This satisfies a preclosure operations requirement that the MGR provide 
means to ensure criticality control during SNF/HLW handling 
operations, including waste package loading (Siddoway 2003 [DIRS 
163904], Requirement 1.1.6-4). 
Some of the DOE SNF waste forms have highly enriched fuel or waste 
form that could potentially support unmoderated (fast) criticality if (1) 
the (fissile) material is concentrated beyond its design concentration in 
the waste form and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration 
could result from either the degradation of the waste form resulting from 
water infiltration or a disruptive event. However, removal of the 
neutron absorber materials from a DOE SNF waste package would 
require a breach of the waste package and a removal mechanism. The 
most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal 
components, dissolving of the neutron absorber material in the water, 
February 2004 144 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
and flushing of the material from the waste package. Since water 
infiltration does not occur, fast criticality can be excluded based on low 
probability. 
The probability of criticality estimate accounts for factors such as early 
failures, manufacturing defects, fuel assembly misloads, etc. However, 
it may not be necessary to directly account for these factors if the total 
probability of criticality is calculated to be sufficiently below the 
regulatory probability criterion without utilizing them. For example, if 
the calculated probability of waste package flooding is below the 
regulatory probability criterion, incorporation of the probability of a 
waste package misload would only result in a lower probability than has 
already been calculated. 
Fuel assembly misloads (enrichment and/or burnup) could result in more 
(or less) fissile material being loaded into the waste package than 
permitted by the design loading curves. Additional fissile material in 
the waste package results in a higher criticality potential of the inpackage 
degraded configurations. However, because no water (i.e., 
neutron moderator) enters the waste package for the rockfall criticality 
FEPs conditions, the additional fissile material that could result from a 
fuel assembly misload cannot result in the formation of a critical 
configuration. 
It then follows that the probability of external criticality must be less 
than the probability of waste package flooding. This is because, in 
addition to the events evaluated to calculate the probability of water 
entering a failed waste package, the probability of the following events 
must also be considered for external criticality: 
 Waste form degradation over the performance period; 
 Separating the fissile materials from the degraded waste form; 
 Removing the fissile materials from the waste package; 
 Accumulating sufficient fissile material into a potentially critical 
configuration in the near-field or far-field environments; and 
 Having sufficient neutron moderator available. 
Because the probability function for drip shield damage areas is zero for 
the rockfall disruptive event (i.e., no drip shield failures [refer to 
Section 6.5.1 of this report]), the probability of waste package flooding 
is zero for the rockfall disruptive event (refer to Table 6.7-2 of this 
report). Since water cannot enter the waste package to degrade the 
waste package internals and waste form, no fissile material can be 
transported from the waste package and into the far-field environment. 
Therefore, the probability of criticality for this rockfall disruptive event 
FEP is zero (refer to Section 6.5.5 of this report). Based on assumptions 
requiring confirmation, this result is applicable for all waste form / 
waste package types. 
February 2004 145 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Not Applicable TSPA Disposition: 
Supporting Reports:  Geochemistry Model Abstraction and Sensitivity Studies for the 21 
PWR CSNF Waste Packages (BSC 2002 [DIRS 160638]) 
 External Accumulation of Fissile Material from DOE Co-Disposal 
Waste Packages (BSC 2002 [DIRS 159913]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
6.8.13 FEP 2.1.14.24.0A Screening Discussion 
Name: In-package criticality resulting from an igneous event (intact 
configuration) 
2.1.14.24.0A Number: 
Description: The waste package internal structures and the waste form remain intact 
either during or after an igneous disruptive event. A breach (or 
breaches) in the waste package allow(s) water to either accumulate or 
flow-through the waste package. Criticality then occurs in situ. 
Descriptor Phrases: Criticality (in waste package), Criticality (from an igneous event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
Waste form criticality analyses demonstrate that an intact, fully flooded 
with water (a neutron moderator), waste package configuration cannot 
achieve criticality (CRWMS M&O 1999 [DIRS 125206], CRWMS 
M&O 2000 [DIRS 147650], CRWMS M&O 2000 [DIRS 147651], BSC 
2003 [DIRS 166610], CRWMS M&O 2000 [DIRS 151742], CRWMS 
M&O 2000 [DIRS 151743], CRWMS M&O 2001 [DIRS 154194], BSC 
2001 [DIRS 157733], BSC 2001 [DIRS 157734], BSC 2001 [DIRS 
161125]). Additionally, intact, fully loaded, fully flooded waste 
February 2004 146 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
packages are precluded from achieving criticality by design to satisfy a 
preclosure operations requirement that the MGR provide means to 
ensure criticality control during SNF/HLW handling operations, 
including waste package loading (Siddoway 2003 [DIRS 163904], 
Requirement 1.1.6-4). 
For the igneous disruptive event, waste packages have been segregated 
into two zones defined by the impact of the igneous event (refer to 
Section 6.6.1 of this report). In Zone 1, the waste packages are 
completely disassembled. In Zone 2, the waste packages remain intact 
(nominal waste package configuration). Therefore, for Zone 2 waste 
packages, the probability of criticality for this igneous disrupt event FEP 
is zero. This result is applicable for all waste form / waste package 
types 
Not Applicable TSPA Disposition: 
Supporting Reports:  21 PWR Waste Package with Absorber Plates Loading Curve 
Evaluation (BSC 2003 [DIRS 166610]) 
 44-BWR Waste Package Loading Curve Evaluation (BSC 2001 
[DIRS 161125]) 
 Evaluation of Codisposal Viability for MOX (FFTF) DOE-Owned 
Fuel (CRWMS M&O 1999 [DIRS 125206]) 
 Evaluation of Codisposal Viability for UzrH (TRIGA) DIE-Owned 
Fuel (CRWMS M&O 2000 [DIRS 147650]) 
 Evaluation of Codisposal Viability for HEU Oxide (Shippingport 
PWR) DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 147651]) 
 Evaluation of Codisposal Viability for U-Zr/U-Mo (Enrico Fermi) 
DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 151742]) 
 Evaluation of Codisposal Viability for Th/U Oxide (Shippingport 
LWBR) DOE-Owned Fuel (CRWMS M&O 2000 [DIRS 151743]) 
 Evaluation of Codisposal Viability for U-Metal (N Reactor) DOEOwned 
Fuel (CRWMS M&O 2001 [DIRS 154194]) 
 Evaluation of Codisposal Viability for Melt and Dilute DOE-Owned 
Fuel (BSC 2001 [DIRS 157733]) 
 Evaluation of Codisposal Viability for Th/U Carbide (Fort Saint 
Vrain HTGR) DOE-Owned Fuel (BSC 2001 [DIRS 157734]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
February 2004 147 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
6.8.14 FEP 2.1.14.25.0A Screening Discussion 
Name: In-package criticality resulting from an igneous event (degraded 
configuration) 
2.1.14.25.0A Number: 
Description: Either during, or as a result of, an igneous disruptive event, the waste 
package internal structures and the waste form degrade. A critical 
configuration develops and criticality occurs in situ. Potential in situ 
critical configurations are defined in Figures 3.2a and 3.2b of Disposal 
Criticality Analysis Methodology Topical Report (YMP 2003 [DIRS 
165505]). 
Descriptor Phrases: Criticality (in waste package), Criticality (from an igneous event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
All base case (i.e., nominal scenario class, no disruptive events) 
postclosure criticality FEPs, internal and external, require water 
infiltration to degrade the waste package internals and waste form. 
Neutron absorber material loss and a flooded waste package condition 
for neutron moderation is the most likely scenario that could result in a 
potentially critical configuration in any of the in situ criticality FEPs. 
External criticality FEPs (near-field and far-field) also require the 
separation of neutron absorber materials from the waste form, transport 
of fissile material from the waste package, and accumulation of the 
fissile material within the drift invert or beyond. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can enter the waste package as 
percolation flow or be present in the pores of the rock. Silica is present 
in appreciable quantities in the high-level radioactive waste glass 
canisters and in the rock. Silica can also be introduced into the waste 
package through precipitation from the percolation flow. Carbon is 
present in only limited amounts in select DOE SNF types and, therefore, 
has a limited impact on the potential for criticality. The loading of the 
DOE-standardized SNF canisters, the design of the basket structure 
February 2004 148 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
inside the canisters, and the addition of neutron absorbers take into 
account the presence and effect of glass in DOE SNF codisposal waste 
packages. Silica from the high-level radioactive waste glass canisters 
has no impact on the potential for criticality in DOE SNF codisposal 
waste packages. Silica is a much less effective moderator than water 
and its introduction into commercial SNF waste packages will displace 
water and effectively reduce the reactivity of the system, thus reducing 
the potential for criticality. 
In addition, criticality without water infiltration is unlikely for the 
geologic repository because the waste package is designed such that a 
criticality event in an intact waste package configuration is not possible. 
This satisfies a preclosure operations requirement that the MGR provide 
means to ensure criticality control during SNF/HLW handling 
operations, including waste package loading (Siddoway 2003 [DIRS 
163904], Requirement 1.1.6-4). 
Some of the DOE SNF waste forms have highly enriched fuel or waste 
form that could potentially support unmoderated (fast) criticality if (1) 
the (fissile) material is concentrated beyond its design concentration in 
the waste form and (2) the neutron absorber materials are removed. 
Concentration of the fissile material beyond its design concentration 
could result from either the degradation of the waste form resulting from 
water infiltration or a disruptive event. However, removal of the 
neutron absorber materials from a DOE SNF waste package would 
require a breach of the waste package and a removal mechanism. The 
most likely neutron absorber material removal mechanism is through 
water infiltration resulting in degradation of the waste package internal 
components, dissolving of the neutron absorber material in the water, 
and flushing of the material from the waste package. Since water 
infiltration does not occur, fast criticality can be excluded based on low 
probability. 
The probability of criticality estimate accounts for factors such as early 
failures, manufacturing defects, fuel assembly misloads, etc. However, 
it may not be necessary to directly account for these factors if the total 
probability of criticality is calculated to be sufficiently below the 
regulatory probability criterion without utilizing them. For example, if 
the calculated probability of waste package flooding is below the 
regulatory probability criterion, incorporation of the probability of a 
waste package misload would only result in a lower probability than has 
already been calculated. 
Fuel assembly misloads (enrichment and/or burnup) could result in more 
(or less) fissile material being loaded into the waste package than 
permitted by the design loading curves. Additional fissile material in 
the waste package results in a higher criticality potential of the in- 
February 2004 149 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
package degraded configurations. However, because of the waste 
package / waste form configuration resulting from the igneous 
disruptive event, the additional fissile material that could result from a 
fuel assembly misload cannot result in the formation of a critical 
configuration. 
For the igneous disruptive event, waste packages have been segregated 
into two zones defined by the impact of the igneous event (refer to 
Section 6.6.1 of this report). In Zone 1, the waste packages are 
completely disassembled. For Zone 1 waste packages, no in-package 
criticality event is possible since the waste package has been 
disassembled. The screening argument of Zone 1 waste packages is 
provided in FEP 2.1.14.26.0A. In Zone 2, the waste packages remain 
intact. For those waste packages in Zone 2, the screening argument of 
FEP 2.1.14.24.0A applies. The probability of criticality for this igneous 
disruptive event FEP is zero. This result is applicable for all waste form 
/ waste package types. 
Not Applicable TSPA Disposition: 
Supporting Reports:  Dike Propagation Near Drifts (CRWMS M&O 2000 [DIRS 
151552]) 
 Waste Package Behavior in Magma (CRWMS M&O 1999 [DIRS 
121300]) 
 Characterize Eruptive Processes at Yucca Mountain, Nevada (BSC 
2003 [DIRS 166407]) 
 Igneous Intrusion Impacts on Waste Package and Waste Forms 
(BSC 2003 [DIRS 165002]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report 
(YMP 2003 [DIRS 165505]) 
6.8.15 FEP 2.1.14.26.0A Screening Discussion 
Near-field criticality resulting from an igneous event Name: 
2.1.14.26.0A Number: 
Description: Either during, or as a result of, an igneous disruptive event, near-field 
criticality occurs when fissile material-bearing solution from the waste 
package is transported into the drift and the fissile material is 
precipitated into a critical configuration. Potential near-field critical 
configurations are defined in Figure 3.3a of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). 
February 2004 150 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Descriptor Phrases: Criticality (in drift), Criticality (from an igneous event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
For the igneous disruptive event, waste packages have been segregated 
into two zones defined by the impact of the igneous event (refer to 
Section 6.6.1 of this report). In Zone 1, the waste packages are 
disassembled. In Zone 2, the waste packages remain intact. It is 
expected that the drip shields, invert and waste packages in Zone 1 will 
be compressed and damaged, allowing the magma to occupy the entire 
emplacement drift. The igneous intrusion temperature may be as high as 
1169C (BSC 2003 [DIRS 166407], Table 38). The melting points of 
waste packages made of Alloy C-22 and Stainless Steel Type 316NG 
are approximately 1,357C and 1,375C, respectively (CRWMS M&O 
1999 [DIRS 121300], Section 5.1). Although the intrusive igneous 
temperature is lower than the melting points of steel and alloy, these 
engineered materials could be severely damaged at the intrusive 
temperature (e.g., through softening, creeping and breaking down) and 
in combination with the shear forces of the viscous magma moving at 
the assumed velocity. 
When the waste packages are damaged, the waste forms will be exposed 
and are likely to be enveloped and fused by the flowing magma. The 
fuel assemblies will be crushed and fragmented, introducing different 
size fragments and granules of UO2 pellets/cladding, neutron absorber, 
and control rods. The crushed material may form radionuclide-bearing 
minerals by incorporating crystallizing silicates minerals. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can be present in the pores of the 
rock or within the magma. Carbon is present in only limited amounts in 
select DOE SNF types and, therefore, has a limited impact on the 
potential for criticality. Silica is present in appreciable quantities in the 
high-level radioactive waste glass canisters, in the rock, or, in the case 
of an igneous event, in the magma. The loading of the DOEstandardized 
SNF canisters, the design of the basket structure inside the 
canisters, and the addition of neutron absorbers take into account the 
February 2004 151 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
presence and effect of glass in DOE SNF codisposal waste packages. 
Silica from the high-level radioactive waste glass canisters has no 
impact on the potential for criticality in DOE SNF codisposal waste 
packages. Silica is a much less effective moderator than water and its 
introduction into commercial SNF waste packages will displace water 
and effectively reduce the reactivity of the system, thus reducing the 
potential for criticality. 
Criticality evaluations have been performed to determine the keff of an 
assumed igneous event configuration in the near-field environment. 
From these criticality evaluations, the probability of criticality for this 
igneous disruptive event FEP is determined to be sufficiently below the 
regulatory probability criterion that it is negligible and therefore set to 
zero (refer to Section 6.6.3 of this report). Through the use of 
assumptions requiring confirmation, this result is applicable for all waste 
form / waste package types. 
Not Applicable TSPA Disposition: 
Supporting Reports:  Geochemistry Model Abstraction and Sensitivity Studies for the 21 
PWR CSNF Waste Packages (BSC 2002 [DIRS 160638]) 
 External Accumulation of Fissile Material from DOE Co-Disposal 
Waste Packages (BSC 2002 [DIRS 159913]) 
 Dike Propagation Near Drifts (CRWMS M&O 2000 [DIRS 
151552]) 
 Waste Package Behavior in Magma (CRWMS M&O 1999 [DIRS 
121300]) 
 Characterize Eruptive Processes at Yucca Mountain, Nevada (BSC 
2003 [DIRS 166407]) 
 Igneous Intrusion Impacts on Waste Package and Waste Forms 
(BSC 2003 [DIRS 165002]) 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report (YMP 
2003 [DIRS 165505]) 
6.8.16 FEP 2.2.14.12.0A Screening Discussion 
Far-field criticality resulting from an igneous event Name: 
2.2.14.12.0A Number: 
Description: Either during, or as a result of, an igneous disruptive event, far-field 
criticality occurs when fissile material-bearing solution from the waste 
February 2004 152 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
package is transported beyond the drift and the fissile material is 
precipitated into a critical configuration. Potential far-field critical 
configurations are defined in Figure 3.3b of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 [DIRS 165505]). 
Descriptor Phrases: Criticality (in the geosphere), Criticality (from an igneous event) 
Screening Decision: Excluded based on low probability. 
Screening Argument: For a criticality event to occur, the proper combination of materials 
(neutron moderators, neutron absorbers, fissile) and geometric 
configuration must exist. A critical system for the geological repository 
is defined as one having an effective neutron multiplication factor (keff), 
larger than the critical limit. The critical limit is the value of keff at 
which a system (configuration of fissile material) is considered critical 
as characterized by statistical tolerance limits (BSC 2003 [DIRS 
165733], Section 6.2.1). 
For the igneous disruptive event, waste packages have been segregated 
into two zones defined by the impact of the igneous event (refer to 
Section 6.6.1 of this report). In Zone 1, the waste packages are 
disassembled. In Zone 2, the waste packages remain intact. It is 
expected that the drip shields, invert and waste packages in Zone 1 will 
be compressed and damaged, allowing the magma to occupy the entire 
emplacement drift. The igneous intrusion temperature may be as high as 
1169C (BSC 2003 [DIRS 166407], Table 38). The melting points of 
waste packages made of Alloy C-22 and Stainless Steel Type 316NG 
are approximately 1,357C and 1,375C, respectively (CRWMS M&O 
1999 [DIRS 121300], Section 5.1). Although the intrusive igneous 
temperature is lower than the melting points of steel and alloy, these 
engineered materials could be severely damaged at the intrusive 
temperature (e.g., through softening, creeping and breaking down) and 
in combination with the shear forces of the viscous magma moving at 
the assumed velocity. 
When the waste packages are damaged, the waste forms will be exposed 
and are likely to be enveloped and fused by the flowing magma. The 
fuel assemblies will be crushed and fragmented, introducing different 
size fragments and granules of UO2 pellets/cladding, neutron absorber, 
and control rods. The crushed material may form radionuclide-bearing 
minerals by incorporating crystallizing silicates minerals. 
Water, silica, and carbon are the only potential moderating materials for 
both internal and external configurations. Water is the most effective 
neutron-moderating material, which can be present in the pores of the 
rock or in the magma. Carbon is present in only limited amounts in 
select DOE SNF types and, therefore, has a limited impact on the 
February 2004 153 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
potential for criticality. Silica is present in appreciable quantities in the 
high-level radioactive waste glass canisters, in the rock, or, in the case 
of an igneous event, in the magma. The loading of the DOEstandardized 
SNF canisters, the design of the basket structure inside the 
canisters, and the addition of neutron absorbers take into account the 
presence and effect of glass in DOE SNF codisposal waste packages. 
Silica from the high-level radioactive waste glass canisters has no 
impact on the potential for criticality in DOE SNF codisposal waste 
packages. Silica is a much less effective moderator than water and its 
introduction into commercial SNF waste packages will displace water 
and effectively reduce the reactivity of the system, thus reducing the 
potential for criticality. 
The probability of criticality in the far-field environment must be less 
than the probability of accumulating sufficient fissile material in the 
near-field environment. This is because, once the fissile material is in 
the near-field environment, the probability of the following events must 
be considered for far-field criticality: 
 The probability of transporting and accumulating sufficient fissile 
material into a potentially critical configuration in the far-field 
environments; and 
 The probability of having sufficient neutron moderator available. 
Because the probability of near-field criticality is below the regulatory 
probability criterion for an igneous disruptive event (refer to 
Section 6.6.3 of this report), given the considerations listed above, the 
probability of criticality in the far-field environment will be even 
smaller. Therefore, the probability of criticality for this igneous 
disruptive event FEP is considered negligible and is set to zero. Based 
on assumptions requiring confirmation, this result is applicable for all 
waste form / waste package types. 
Not Applicable TSPA Disposition: 
Supporting Reports:  Geochemistry Model Abstraction and Sensitivity Studies for the 21 
 External Accumulation of Fissile Material from DOE Co-Disposal 
 Dike Propagation Near Drifts (CRWMS M&O 2000 [DIRS 
 Waste Package Behavior in Magma (CRWMS M&O 1999 [DIRS 
 Characterize Eruptive Processes at Yucca Mountain, Nevada (BSC 
 Igneous Intrusion Impacts on Waste Package and Waste Forms 
PWR CSNF Waste Packages (BSC 2002 [DIRS 160638]) 
Waste Packages (BSC 2002 [DIRS 159913]) 
151552]) 
121300]) 
2003 [DIRS 166407]) 
(BSC 2003 [DIRS 165002]) 
February 2004 154 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
 Criticality Model Report (BSC 2003 [DIRS 165733]) 
 Project Functional and Operational Requirements (Siddoway 2003 
[DIRS 163904]) 
 Disposal Criticality Analysis Methodology Topical Report (YMP 
2003 [DIRS 165505]) 
February 2004 155 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
INTENTIONALLY LEFT BLANK 
February 2004 156 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
7. CONCLUSIONS 
Using the geologic repository and engineered barrier systems information currently available and 
several assumptions requiring confirmation (including the modification of the 44-BWR Waste 
Package absorber plate thickness to 7 mm), criticality can be screened from further consideration 
in TSPA on the sole basis of the probability of waste package flooding and neutron absorber 
material removal. The results of this analysis indicate that, for all waste package types, the 
calculated total probability of waste package flooding and neutron absorber material removal is 
below the regulatory probability criterion (10 CFR 63.114(d) [DIRS 156605]). This conclusion 
is based on the binomial distribution evaluation of the calculated probability results (refer to 
Table 6.7-2). 
Information currently being updated that could influence the results of this analysis is the failure 
potential of the drip shield and waste package due to the various corrosion mechanisms (i.e., 
general, localized, and stress corrosion cracking). Although the models for these failure 
mechanisms have been developed, evaluation of these models is dependent on the drift 
environment to be modeled in the TSPA-LA analyses. Future results in these areas have the 
potential to impact the analysis results. Also having the potential to impact the analysis results 
are updates to the qualified data, product outputs and technical information used in this analysis 
and the results from the testing, design and analysis required for the confirmation of assumptions 
in Section 5. 
7.1 SUMMARY 
The safety strategy for the monitored geologic repository relies on a multiple barrier system for 
the long-term isolation of the emplaced waste packages from the general environment. Over 
time, waste packages emplaced in the geologic repository as part of the engineered barrier 
systems can undergo various degradation processes that modify the waste package structural and 
mineral content and, thus, affect the potential for a criticality event. These degradation processes 
have major effects on the waste packages isotopic content (through flushing) and spatial 
distribution of the waste form within the affected waste package (through component 
degradation). Separation of neutron absorbers from fissile material, volume changes, shape 
changes, loss of fissile and/or absorber material from the waste package, and rearrangement of 
degraded components are potential effects of the degradation processes. 
For a criticality to occur, multiple changes in conditions must occur (waste package breach, 
water intrusion, water retention, and removal of neutron absorbers). Should a criticality occur, 
geological and engineered barriers prevent and reduce the release of energy and the rate of 
radionuclide transport to the accessible environment. 
February 2004 157 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
This screening analysis report: 
1. Contributes to the Yucca Mountain scenario development methodology by screening 
the FEPs related to criticality. 
2. Develops screening arguments for these FEPs. 
3. Provides information for the YMP FEP database and guidance to Total System 
Performance Assessment for the License Application analyses applicable to the license 
application document. 
Screening decisions reached in this report are summarized in Table 7.2-1. 
7.2 CRITICALITY FEPS SCREENING RECOMMENDATIONS FOR THE LICENSE 
APPLICATION 
Recommendations for the criticality FEPs and their reference section are provided in 
Table 7.2-1. These recommendations for the base case in situ and external criticality FEPs 
evaluations are applicable to all waste package/waste form combinations. This is because the 
probability of water entering any waste package during the performance period is calculated to 
be 0.0 for all base case criticality FEPs because there are no drip shield failures predicted during 
the performance period. 
The evaluation of the rockfall and igneous disruptive event criticality FEPs are applicable to all 
waste package/waste form combinations. This is because the probability of water entering any 
waste package during the performance period is 0.0 for the rockfall disruptive event (no drip 
shield failures) and because it is improbable that a critical configuration could be formed during 
an igneous disruptive event. 
For the evaluation of the seismic disruptive event criticality FEPs, it is necessary to calculate the 
analysis results for the individual commercial and DOE SNF waste package types. This is 
because of the differences in the waste package types internal configurations and compositions 
that degrade at different rates. Additionally, the seismic criticality results are based on the 
increase of the 44-BWR waste package basket thickness to 7 mm from 5 mm to allow sufficient 
retention of boron in the waste package during the performance period. The result of the FEPs 
evaluation is the calculation of a total probability of waste package flooding and the neutron 
absorber material removal below the regulatory probability criterion (10 CFR 63.114(d) [DIRS 
156605]). Because the total probability of flooding and degrading the waste package internals is 
below the regulatory probability criterion, the total probability of criticality is also below the 
regulatory probability criterion. 
The Naval Nuclear Propulsion Program is responsible for the assessment of criticality potential 
of the naval SNF Short and naval SNF Long waste package types in accordance with an 
Addendum (Mowbray 1999 [DIRS 149585]) to the Disposal Criticality Analysis Methodology 
Topical Report (YMP 2003 [DIRS 165505]). 
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FEP Number 
2.1.14.15.0A 
2.1.14.16.0A 
2.1.14.17.0A 
2.2.14.09.0A 
2.1.14.18.0A 
2.1.14.19.0A 
ANL-EBS-NU-000008 REV 00 
Table 7.2-1. Screening Decisions for Criticality FEPs 
FEP Name 
In-package 
criticality 
(intact 
configuration) 
In-package 
criticality 
(degraded 
configurations) 
Near-field 
criticality 
Far-field 
criticality 
In-package 
criticality 
resulting from 
a seismic 
event (intact 
configuration) 
In-package 
criticality 
resulting from 
a seismic 
event 
(degraded 
configurations) 
FEP Description 
The waste package internal structures and the 
waste form remain intact. A breach (or breaches) 
in the waste package allow(s) water to either 
accumulate or flow-through the waste package. 
Criticality then occurs in situ. In-package criticality 
resulting from disruptive events is addressed in 
separate FEPs. 
The waste package internal structures and the 
waste form degrade. A critical configuration 
(sufficient fissile material and neutron moderator, 
lack of neutron absorbers) develops and criticality 
occurs in situ. Potential in situ critical 
configurations are defined in Figures 3.2a and 3.2b 
of Disposal Criticality Analysis Methodology Topical 
Report (YMP 2003 [DIRS 165505]). In-package 
criticality resulting from disruptive events is 
addressed in separate FEPs. 
Near-field criticality occurs when fissile materialbearing 
solution from the waste package is 
transported into the drift and the fissile material is 
precipitated into a critical configuration. Potential 
near-field critical configurations are defined in 
Figure 3.3a of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 
165505]). In-package criticality resulting from 
disruptive events is addressed in separate FEPs. 
Far-field criticality occurs when fissile materialbearing 
solution from the waste package is 
transported beyond the drift and the fissile material 
is precipitated into a critical configuration. Potential 
far-field critical configurations are defined in Figure 
3.3b of Disposal Criticality Analysis Methodology 
Topical Report (YMP 2003 [DIRS 165505]). Inpackage 
criticality resulting from disruptive events 
is addressed in separate FEPs. 
The waste package internal structures and the 
waste form remain intact either during or after a 
seismic disruptive event. A breach (or breaches) in 
the waste package allow(s) water to either 
accumulate or flow-through the waste package. 
Criticality then occurs in situ. 
Either during, or as a result of, a seismic disruptive 
event, the waste package internal structures and 
the waste form degrade. A critical configuration 
develops and criticality occurs in situ. Potential in 
situ critical configurations are defined in Figures 
3.2a and 3.2b of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 
165505]). 
159 of 176 
Section 
Screening 
Addressed 
TSPA-LA 
Screening 
Decision 
Section 
6.8.1 
Excluded  
Low 
Probability 
Section 
6.8.2 
Excluded  
Low 
Probability 
Section 
6.8.3 
Excluded  
Low 
Probability 
Section 
6.8.4 
Excluded  
Low 
Probability 
Section 
6.8.5 
Excluded  
Low 
Probability 
Section 
6.8.6 
Excluded  
Low 
Probability 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 7.2-1. Screening Decisions for Criticality FEPs (Continued) 
FEP Name FEP Number 
2.1.14.20.0A 
Near-field 
criticality 
resulting from 
a seismic 
event 
2.2.14.10.0A 
Far-field 
criticality 
resulting from 
a seismic 
event 
2.1.14.21.0A 
In-package 
criticality 
resulting from 
rockfall (intact 
configuration) 
2.1.14.22.0A 
In-package 
criticality 
resulting from 
rockfall 
(degraded 
configurations) 
2.1.14.23.0A 
Near-field 
criticality 
resulting from 
rockfall 
2.2.14.11.0A 
Far-field 
criticality 
resulting from 
rockfall 
2.1.14.24.0A 
In-package 
criticality 
resulting from 
an igneous 
event (intact 
configuration) 
ANL-EBS-NU-000008 REV 00 
FEP Description 
Either during, or as a result of, a seismic disruptive 
event, near-field criticality occurs when fissile 
material-bearing solution from the waste package 
is transported into the drift and the fissile material is 
precipitated into a critical configuration. Potential 
near-field critical configurations are defined in 
Figure 3.3a of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 
165505]). 
Either during, or as a result of, a seismic disruptive 
event, far-field criticality occurs when fissile 
material-bearing solution from the waste package 
is transported beyond the drift and the fissile 
material is precipitated into a critical configuration. 
Potential far-field critical configurations are defined 
in Figure 3.3b of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 
165505]). 
The waste package internal structures and the 
waste form remain intact either during or after a 
rockfall event. A breach (or breaches) in the waste 
package allow(s) water to either accumulate or 
flow-through the waste package. Criticality then 
occurs in situ. 
Either during, or as a result of, a rockfall event, the 
waste package internal structures and the waste 
form degrade. A critical configuration develops and 
criticality occurs in situ. Potential in situ critical 
configurations are defined in Figures 3.2a and 3.2b 
of Disposal Criticality Analysis Methodology Topical 
Report (YMP 2003 [DIRS 165505]). 
Either during, or as a result of, a rockfall event, 
near-field criticality occurs when fissile materialbearing 
solution from the waste package is 
transported into the drift and the fissile material is 
precipitated into a critical configuration. Potential 
near-field critical configurations are defined in 
Figure 3.3a of Disposal Criticality Analysis 
Methodology Topical Report (YMP 2003 [DIRS 
165505]). 
Either during, or as a result of, a rockfall event, farfield 
criticality occurs when fissile material-bearing 
solution from the waste package is transported 
beyond the drift and the fissile material is 
precipitated into a critical configuration. Potential 
far-field critical configurations are defined in Figure 
3.3b of Disposal Criticality Analysis Methodology 
Topical Report (YMP 2003 [DIRS 165505]). 
The waste package internal structures and the 
waste form remain intact either during or after an 
igneous disruptive event. A breach (or breaches) 
in the waste package allow(s) water to either 
accumulate or flow-through the waste package. 
Criticality then occurs in situ. 
160 of 176 
Section 
Screening 
Addressed 
TSPA-LA 
Screening 
Decision 
Section 
6.8.7 
Excluded  
Low 
Probability 
Section 
6.8.8 
Excluded  
Low 
Probability 
Section 
6.8.9 
Excluded  
Low 
Probability 
Section 
6.8.10 
Excluded  
Low 
Probability 
Section 
6.8.11 
Excluded  
Low 
Probability 
Section 
6.8.12 
Excluded  
Low 
Probability 
Section 
6.8.13 
Excluded  
Low 
Probability 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 7.2-1. Screening Decisions for Criticality FEPs (Continued) 
FEP Name FEP Number 
In-package 
criticality 
resulting from 
2.1.14.25.0A an igneous 
event 
(degraded 
configurations) 
2.1.14.26.0A 
Near-field 
criticality 
resulting from 
an igneous 
event 
2.2.14.12.0A 
Far-field 
criticality 
resulting from 
an igneous 
event 
FEP Description 
Either during, or as a result of, an igneous 
disruptive event, the waste package internal 
structures and the waste form degrade. A critical 
configuration develops and criticality occurs in situ. 
Potential in situ critical configurations are defined in 
Figures 3.2a and 3.2b of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 
[DIRS 165505]). 
Either during, or as a result of, an igneous 
disruptive event, near-field criticality occurs when 
fissile material-bearing solution from the waste 
package is transported into the drift and the fissile 
material is precipitated into a critical configuration. 
Potential near-field critical configurations are 
defined in Figure 3.3a of Disposal Criticality 
Analysis Methodology Topical Report (YMP 2003 
[DIRS 165505]). 
Either during, or as a result of, an igneous 
disruptive event, far-field criticality occurs when 
fissile material-bearing solution from the waste 
package is transported beyond the drift and the 
fissile material is precipitated into a critical 
configuration. Potential far-field critical 
configurations are defined in Figure 3.3b of 
Disposal Criticality Analysis Methodology Topical 
Report (YMP 2003 [DIRS 165505]). 
TSPA-LA 
Screening 
Decision 
7.3 UNCERTAINTIES AND RESTRICTIONS 
This document may be affected by technical product input information that requires 
confirmation. Any changes to the document that may occur as a result of completing the 
confirmation activities will be reflected in subsequent revisions. The status of the input 
information quality may be confirmed by review of the Document Input Reference System 
database. 
7.3.1 Restriction #1: This Screening Analysis Is a Draft Demonstration of the Screening 
Methodology 
Waste package specific information has been utilized for the evaluation of 12-PWR Long, 
44-BWR, and 24-BWR Waste Package types. However, the necessary model reports have not 
been developed or validated for use with BWR fuel and, therefore, strict adherence to the 
methodology outlined in Disposal Criticality Analysis Methodology Topic Report (YMP 2003 
[DIRS 165505]) was not possible. 
Although an assumption has been made extending the 21-PWR with Absorber Plates Waste 
Package inputs to the DOE SNF waste package types (Assumption 5.1.7) and an additional 
assumption has been made regarding the 21-PWR with Control Rods Waste Package type 
(Assumption 5.1.6), both of these assumptions require confirmation through additional analysis. 
ANL-EBS-NU-000008 REV 00 161 of 176 
Section 
Screening 
Addressed 
Section 
6.8.14 
Excluded  
Low 
Probability 
Section 
6.8.15 
Excluded  
Low 
Probability 
Section 
6.8.16 
Excluded  
Low 
Probability 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
7.3.2 Restriction #2: Time-Dependent Corrosion Will Not be Available Until the TSPALA 
Is Performed 
Because of the use of corrosion-resistant materials, it is important to assume for this screening 
analysis, corrosion damage to the drip shields and the waste packages is caused only by an early 
failure mechanism (improper heat treatment) and not by the time-dependent corrosion 
mechanisms typically resulting from water dripping onto the drip shield and the waste package 
(Assumption 5.1.1). Additionally, the detailed time-dependent corrosion information for (1) 
general corrosion, (2) localized corrosion (crevice corrosion and pitting corrosion), and (3) stress 
corrosion cracking, will not be available until the TSPA-LA is performed. 
This assumption requires further verification and confirmation when the TSPA-LA calculations 
are published. 
7.3.3 Restriction #3: Evaluation Does Not Account for Variations In Waste Package 
Design. 
The FEPs screening analyses assume the waste package is fabricated, loaded, closed, and 
emplaced as designed (Assumption 5.1.3). Once fabrication and operational processes and 
procedures are developed and approved, evaluations of off-normal waste package configurations 
may be performed. 
7.3.4 Restriction #4: Assumptions Regarding Neutron Absorber Material Corrosion 
Rate and Retention in Waste Package 
The corrosion rate of the waste package neutron absorber material (Assumptions 5.1.4, 5.1.6, and 
5.1.7) and the amount of neutron absorber material that must be retained in a degraded waste 
package to prevent a criticality event (Assumption 5.1.5) must be confirmed through 
measurements and analyses, respectively. These parameters are important as they determine the 
rate of neutron absorber material loss from a breached, flooded waste package. A slow corrosion 
rate or low required retention amount will possibly prolong a subcritical condition within the 
waste package to well beyond the performance period. 
7.3.5 Restriction #5: Assumptions Regarding Igneous Event Configuration and External 
Critical Limit 
The high fissile material enrichment of some DOE SNF waste forms could facilitate the 
formation of critical configurations in low or no neutron moderator environments such as is 
expected during an igneous disruptive event. It is assumed that configurations formed due to 
igneous events involving DOE SNF waste packages will not result in the formation of a critical 
system (Assumption 5.4.3). 
eff 
In addition, a critical limit will need to be determined using the methodology to be developed in 
the external criticality model for comparison to the calculated igneous event configuration k 
values (Assumption 5.4.4). This is necessary for assessing the criticality potential of the 
configurations resulting from the igneous disruptive event. 
Both of these assumptions require confirmation by analysis. 
February 2004 162 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
7.3.6 Restriction #6: Evaluation Does Not Account for Condensation 
Water entering a failed waste package may occur from two primary pathways: (1) water 
dripping from the drift crown through a failed drip shield and into a failed waste package, and 
(2) water dripping from the underside of a drip shield due to evaporation and condensation into a 
failed waste package. The first pathway is evaluated in the FEPs screening analysis. However, 
the model to account for the second pathway is not yet available. Evaluation of this pathway will 
be necessary once this model is available. 
7.4 ADDRESSING OF REQUIREMENTS AND CRITERIA 
This section discusses how the criteria and requirements listed in Section 4.2 were addressed in 
this analysis report. Table 7.4-1 discusses how the applicable project requirements from Project 
Requirements Document (Canori and Leitner 2003 [DIRS 166275]) were addressed. 
Tables 7.4-2 and 7.4-3 address how the applicable Yucca Mountain Review Plan (NRC 2003 
[DIRS 163274]) acceptance criteria were addressed. 
February 2004 163 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Requirement Number 
and Title 
PRD-002/T-015a; 
Requirements for 
Performance Assessment 
PRD-002/T-034b; Limits on For complete requirement text, see 
Performance Assessments 10 CFR 63.342 [DIRS 156605] 
PRD-013/T-016c; 
Postclosure [of DOE SNF 
Canister Criticality 
Potential] 
PRD-013/T-023d; 
Postclosure [of Naval SNF 
Canister Criticality 
Potential] 
PRD-013/T-038e; 
Postclosure.[of 
Commercial- Origin DOE 
SNF Canister Criticality 
Potential] 
Source: Canori and Leitner 2003 [DIRS 166275] 
NOTES: a Requirement basis is 10 CFR 63.114 and 63.113 [DIRS 156605] & YMP-RD 3.3.4.19 (YMP 2001). 
b Requirement basis is 10 CFR 63.342 [DIRS 156605] (Reference 40 CFR 197.36 [DIRS 155238]) 
c Requirement basis is WASRD 4.3.12.B (DOE 2002 [DIRS 158873]) 
d Requirement basis is WASRD 4.4.13.B (DOE 2002 [DIRS 158873]) 
e Requirement basis is WASRD 4.5.13.B (DOE 2002 [DIRS 158873]) 
ANL-EBS-NU-000008 REV 00 
Table 7.4-1. Addressing Project Requirements 
Requirement Text 
For complete requirement text, see 
10 CFR 63.114 [DIRS 156605] 
The methodology defined in the 
Disposal Criticality Analysis 
Methodology Topical Report (YMP 
2003 [DIRS 165505]) shall be used 
to demonstrate acceptable criticality 
control for canisters and the waste 
packages in which they are 
disposed. 
The methodology in the NNPP 
addendum (Mowbray 1999 [DIRS 
149585]) to the Disposal Criticality 
Analysis Methodology Topical 
Report (YMP 2003 [DIRS 165505]) 
shall be used to demonstrate 
acceptable criticality control for 
canisters and the waste packages in 
which they are disposed. 
The methodology defined in the 
Disposal Criticality Analysis 
Methodology Topical Report (YMP 
2003 [DIRS 165505]) shall be used 
to demonstrate acceptable criticality 
control for canisters and the waste 
packages in which they are 
disposed. 
164 of 176 
How Requirement Addressed 
This report provides the technical bases for 
excluding criticality FEPs. The technical basis 
is provided in Section 6. 
This report provides the screening so as to 
not include very unlikely FEPs, defined as 
those that are estimated to have less than 
one chance in 10,000 of occurring within 
10,000 years of disposal, in the performance 
assessments. The screening is provided in 
Section 6. 
This report documents the partial 
implementation of the methodology from the 
from Disposal Criticality Analysis Methodology 
Topical Report (YMP 2003 [DIRS 165505]). 
The complete methodology is not used 
because the partial application excludes 
criticality. The implementation is provided in 
Section 6. 
This report does not effect how acceptable 
criticality control is demonstrated in the NNPP 
addendum (Mowbray 1999 [DIRS 149585]). 
The external configurations which would 
require isotopic concentrations for naval SNF 
from the NNPP are screen out in Section 6.3. 
This report documents the partial 
implementation of the methodology from 
Disposal Criticality Analysis Methodology 
Topical Report (YMP 2003 [DIRS 165505]). 
The complete methodology is not used 
because the partial application excludes 
criticality. The implementation is provided in 
Section 6. 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 7.4-2. Addressing Acceptance Criteria for Scenario Identification and Screening 
Acceptance Criteria 
Name 
Acceptance Criterion 1: 
The Identification of a List 
of Features, Events, and 
Processes Is Adequate. 
Acceptance Criterion 2: 
Screening of the List of 
Features, Events, and 
Processes Is Appropriate. 
Source: NRC 2003 [DIRS 163274], Section 2.2.1.2.1.3 
How Acceptance Criteria is Addressed 
(1) The report contains a complete list of criticality features, events and processes 
that have the potential to influence repository performance. The list is 
consistent with the site characterization data and includes potentially 
disruptive events (i.e., igneous; seismic, and rockfall); and criticality. This list 
is provided in Table 6.1-3. 
(1) The report identifies all the criticality features, events, and processes that have 
been excluded in Section 6.8, 
(2) The report provides justification for excluding the criticality features, events, 
and processes in Section 6.8; and 
(3) The report provides an adequate technical basis for excluding each criticality 
feature, event, and process based on low-probability in Sections 6.3 through 
6.6. 
ANL-EBS-NU-000008 REV 00 February 2004 165 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table 7.4-3. Addressing Acceptance Criteria for Uncertainty in Event Probability 
Acceptance Criterion Number 
and Title 
Acceptance Criterion 1: 
Events Are Adequately Defined. 
Acceptance Criterion 2: 
Probability Estimates for Future 
Events Are Supported by 
Appropriate Technical Bases. 
Acceptance Criterion 3: 
Probability Model Support Is 
Adequate. 
Acceptance Criterion 4: 
Probability Model Parameters 
Have Been Adequately 
Established. 
Acceptance Criterion 5: 
Uncertainty in Event Probability 
Is Adequately Evaluated. 
Source: NRC 2003 [DIRS 163274], Section 2.2.1.2.2.3 
How Acceptance Criteria is Addressed 
(1) The report identifies criticality events and estimates the probabilities of 
each criticality event separately, so that criticality events are defined 
without ambiguity and used consistently in probability models. The 
criticality events and their correlation with individual FEPs are provided 
in Sections 6.3 through 6.6 of the report; and 
(2) The probabilities of criticality events are calculated separately by 
location (internal and external) as appropriate in Sections 6.3 through 
6.6. 
(1) The inputs used in the report to estimate the probabilities for future 
criticality events come from models which consider past patterns of the 
natural events in the Yucca Mountain region (e.g., seismic) in 
estimating the likely future conditions and interactions of the natural and 
engineered repository system. The inputs and their sources are listed 
in Sections 4.1.2 and 4.1.3. 
(1) The reports probability models use the outputs from validated, detailed 
process level models as inputs. These inputs are discussed in 
Sections 4.1.2 and 4.1.3 The specific means of validating the models 
are described in the separate reports. 
(1) The report provides the technical justification for the parameters used in 
the probability models in Sections 4.1.2, 4.1.3 and 6.3 through 6.6. 
Specifically: 
(a) Parameters used in the probability models are from the outputs from 
models of the natural and engineered systems for a repository at Yucca 
Mountain (Sections 4.1.2 and 4.1.3); 
(b) The report establishes reasonable and consistent correlations between 
parameters (discussed in Sections 6.3 through 6.6) ; and 
(c) Where updated models of natural and engineered systems are not 
available to generate parameters for the probability models, other 
appropriate sources are noted and confirmation of this appropriateness 
is noted in Section 5. 
(1) The report addresses uncertainty in probability values by accounting for 
uncertainty in the model outputs used to develop the probability values. 
The uncertainty in model outputs used to develop probabilities is 
discussed in Section 6.4. Specifically: 
(a) The report provides a technical basis for probability values used 
(Sections 6.4 and Attachment III), and the values account for the 
uncertainty in the probability estimates; and 
(b) The uncertainties are not reported separately for probability values. 
The probability values are based on results that incorporate the 
parameter uncertainty from the model results and model uncertainty. 
ANL-EBS-NU-000008 REV 00 February 2004 166 of 176 Screening Analysis for Criticality Features, Events, and Processes for License Application 
8. INPUTS AND REFERENCES 
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00202-000-00B. Las Vegas, Nevada: Bechtel SAIC Company. ACC: 
ENG.20040202.0010. 
166275 Canori, G.F. and Leitner, M.M. 2003. Project Requirements Document. TER-MGRMD-
000001 REV 02. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: DOC.20031222.0006. 
February 2004 169 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
100224 CRWMS M&O (Civilain Radioactive Waste Management System Management & 
Operating Contractor) 1997. Determination of Waste Package Design Configurations. 
BBAA00000-01717-0200-00017 REV 00. Las Vegas, Nevada: CRWMS M&O. 
ACC: MOL.19970805.0310. 
102836 CRWMS M&O 1998. Software Qualification Report for MCNP Version 4B2, A 
General Monte Carlo N-Particle Transport Code. CSCI: 30033 V4B2LV. DI: 
30033-2003, Rev. 01. Las Vegas, Nevada: CRWMS M&O. 
ACC: MOL.19980622.0637. 
121300 CRWMS M&O 1999. Waste Package Behavior in Magma. CAL-EBS-ME-000002 
REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19991022.0201. 
125206 CRWMS M&O 1999. Evaluation of Codisposal Viability for MOX (FFTF) DOEOwned 
Fuel. BBA000000-01717-5705-00023 REV 00. Las Vegas, Nevada: CRWMS 
M&O. ACC: MOL.19991014.0235. 
157731 CRWMS M&O 1999. Summary Report of Laboratory Critical Experiment Analyses 
Performed for the Disposal Criticality Analysis Methodology. B00000000-01717-5705- 
00076 REV 02. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.19990920.0167. 
144180 CRWMS M&O 2000. Features, Events, and Processes: System-Level and Criticality. 
ANL-WIS-MD-000019 REV 00. Las Vegas, Nevada: CRWMS M&O. 
ACC: MOL.20010108.0012. 
147650 CRWMS M&O 2000. Evaluation of Codisposal Viability for UZrH (TRIGA) DOEOwned 
Fuel. TDR-EDC-NU-000001 REV 00. Las Vegas, Nevada: CRWMS M&O. 
ACC: MOL.20000207.0689. 
147651 CRWMS M&O 2000. Evaluation of Codisposal Viability for HEU Oxide 
(Shippingport PWR) DOE-Owned Fuel. TDR-EDC-NU-000003 REV 00. Las Vegas, 
Nevada: CRWMS M&O. ACC: MOL.20000227.0240. 
149939 CRWMS M&O 2000. Probability of Criticality Before 10,000 Years. CAL-EBS-NU- 
000014 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20001107.0303. 
151552 CRWMS M&O 2000. Dike Propagation Near Drifts. ANL-WIS-MD-000015 REV 00 
ICN 1. Las Vegas, Nevada: CRWMS M&O. ACC: MOL.20001213.0061. 
151566 CRWMS M&O 2000. WAPDEG Analysis of Waste Package and Drip Shield 
Degradation. ANL-EBS-PA-000001 REV 00 ICN 01. Las Vegas, Nevada: CRWMS 
M&O. ACC: MOL.20001208.0063. 
151742 CRWMS M&O 2000. Evaluation of Codisposal Viability for U-Zr/U-Mo Alloy (Enrico 
Fermi) DOE-Owned Fuel. TDR-EDC-NU-000002 REV 00. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL.20000815.0317. 
February 2004 170 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
151743 CRWMS M&O 2000. Evaluation of Codisposal Viability for Th/U Oxide 
(Shippingport LWBR) DOE-Owned Fuel. TDR-EDC-NU-000005 REV 00. Las Vegas, 
Nevada: CRWMS M&O. ACC: MOL.20001023.0055. 
153246 CRWMS M&O 2000. Total System Performance Assessment for the Site 
Recommendation. TDR-WIS-PA-000001 REV 00 ICN 01. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL.20001220.0045. 
154194 CRWMS M&O 2001. Evaluation of Codisposal Viability for U-Metal (N Reactor) 
DOE-Owned Fuel. TDR-EDC-NU-000004 REV 00. Las Vegas, Nevada: CRWMS 
M&O. ACC: MOL.20010314.0004. 
160320 D'Agostino, R.B. and Stephens, M.A., eds. 1986. Goodness-Of-Fit Techniques. 
Statistics, Textbooks and Monographs Volume 68. New York, New York: Marcel 
Dekker. TIC: 253256. 
132333 DOE 1987. Appendix 2A. Physical Descriptions of LWR Fuel Assemblies. Volume 3 
of Characteristics of Spent Fuel, High-Level Waste, and Other Radioactive Wastes 
Which May Require Long-Term Isolation. DOE/RW-0184. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: HQX.19880405.0024. 
155943 DOE 2002. Yucca Mountain Science and Engineering Report. DOE/RW-0539, Rev. 1. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: MOL.20020404.0042 
158873 DOE 2002. Waste Acceptance System Requirements Document. DOE/RW-0351, Rev. 
4. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive 
Waste Management. ACC: MOL.20020326.0056. 
162903 DOE 2003. Quality Assurance Requirements and Description. DOE/RW-0333P, Rev. 
13. Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive 
Waste Management. ACC: DOC.20030422.0003. 
106070 Duderstadt, J.J. and Hamilton L.J. 1976. Nuclear Reactor Analysis. New York, New 
York: John Wiley & Sons. TIC: 245454. 
154365 Freeze, G.A.; Brodsky, N.S.; and Swift, P.N. 2001. The Development of Information 
Catalogued in REV00 of the YMP FEP Database. TDR-WIS-MD-000003 REV 00 
ICN 01. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20010301.0237. 
165394 Freeze, G. 2003. KTI Letter Report, Response to Additional Information Needs on 
TSPAI 2.05 and TSPAI 2.06. REG-WIS-PA-000003 REV 00 ICN 04. Las Vegas, 
Nevada: Bechtel SAIC Company. ACC: DOC.20030825.0003. 
100455 Hillner, E.; Franklin, D.G.; and Smee, J.D. 1998. The Corrosion of Zircaloy-Clad Fuel 
Assemblies in a Geologic Repository Environment. WAPD-T-3173. West Mifflin, 
Pennsylvania: Bettis Atomic Power Laboratory. TIC: 237127. 
February 2004 171 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
134327 Kgler, A. 1997. Sheet and Plate for Nuclear Engineering, Bhler Neutronit A976. 
Houston, Texas: Bhler Bleche GmbH. TIC: 246410. 
104667 Modarres, M. 1993. What Every Engineer Should Know About Reliability and Risk 
Analysis. New York, New York: Marcel Dekker. TIC: 238168. 
149585 Mowbray, G.E. 1999. Transmittal of the Naval Nuclear Propulsion Program 
Addendum to the Yucca Mountain Site Characterization Office "Disposal Criticality 
Analysis Methodology Topical Report." Letter from G.E. Mowbray (Department of the 
Navy) to C.W. Reamer (NRC), October 29, 1999. ACC: MOL.20000316.0531. 
163274 NRC (U.S. Nuclear Regulatory Commission) 2003. Yucca Mountain Review Plan, 
Final Report. NUREG-1804, Rev. 2. Washington, D.C.: U.S. Nuclear Regulatory 
Commission, Office of Nuclear Material Safety and Safeguards. TIC: 254568. 
100479 Nuclear Energy Agency 1992. Systematic Approaches to Scenario Development: A 
Report of the NEA Working Group on Identification and Selection of Scenarios for 
Performance Assessment of Radioactive Waste Disposal. Paris, France: Nuclear 
Energy Agency, Organisation for Economic Co-operation and Development. 
TIC: 8083. 
103896 Parrington, J.R.; Knox, H.D.; Breneman, S.L.; Baum, E.M.; and Feiner, F. 1996. 
Nuclides and Isotopes, Chart of the Nuclides. 15th Edition. San Jose, California: 
General Electric Company and KAPL, Inc. TIC: 233705. 
155635 Punatar, M.K. 2001. Summary Report of Commercial Reactor Criticality Data for 
Crystal River Unit 3. TDR-UDC-NU-000001 REV 02. Las Vegas, Nevada: Bechtel 
SAIC Company. ACC: MOL.20010702.0087. 
163904 Siddoway, D.W. 2003. Project Functional and Operational Requirements. TDRMGR-
ME-000003 REV 01. Las Vegas, Nevada: Bechtel SAIC Company. 
ACC: ENG.20030630.0001. 
158378 USGS (U.S. Geological Survey) 2001. Future Climate Analysis. ANL-NBS-GS- 
000008 REV 00 ICN 01. Denver, Colorado: U.S. Geological Survey. 
ACC: MOL.20011107.0004. 
152180 Walpole, R.E.; Myers, R.H.; and Myers, S.L. 1998. Probability and Statistics for 
Engineers and Scientists. 6th Edition. Upper Saddle River, New Jersey: Prentice Hall. 
TIC: 242020. 
157283 Wark, K. 1983. Thermodynamics. 4th Edition. New York, New York: McGraw-Hill. 
TIC: 243020. 
111561 Weast, R.C., ed. 1985. CRC Handbook of Chemistry and Physics. 66th Edition. 
Boca Raton, Florida: CRC Press. TIC: 216054. 
February 2004 172 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
156713 YMP 2001. Yucca Mountain Site Characterization Project Requirements Document 
(YMP-RD). YMP/CM-0025, Rev. 4, DCN 02. Las Vegas, Nevada: Yucca Mountain 
Site Characterization Office. ACC: MOL.20010322.0491; MOL.20011107.0002. 
165505 YMP 2003. Disposal Criticality Analysis Methodology Topical Report. YMP/TR- 
004Q, Rev. 02. Las Vegas, Nevada: Yucca Mountain Site Characterization Office. 
ACC: DOC.20031110.0005. 
8.2 CODES, STANDARDS, REGULATIONS, AND PROCEDURES 
156605 10 CFR 63. Energy: Disposal of High-Level Radioactive Wastes in a Geologic 
Repository at Yucca Mountain, Nevada. Readily available. 
155238 40 CFR 197. 2001. Protection of Environment: Public Health and Environmental 
Radiation Protection Standards for Yucca Mountain, Nevada. Readily available. 
165023 AP-SI.1Q, Rev. 5, ICN 2. Software Management. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20030902.0003. 
166252 AP-SIII.9Q, Rev. 1, ICN 2. Scientific Analyses. Washington, D.C.: U.S. Department of 
Energy, Office of Civilian Radioactive Waste Management. 
ACC: DOC.20031126.0001. 
165687 AP-SV.1Q, Rev. 1, ICN 0. Control of the Electronic Management of Information. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: DOC.20030929.0004. 
164786 AP-2.22Q, Rev. 1, ICN 0. Classification Analyses and Maintenance of the Q-List. 
Washington, D.C.: U.S. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: DOC.20030807.0002. 
154062 ASTM A 887-89 (Reapproved 2000). 2000. Standard Specification for Borated 
Stainless Steel Plate, Sheet, and Strip for Nuclear Application. West Conshohocken, 
Pennsylvania: American Society for Testing and Materials. TIC: 249544. 
8.3 SOURCE DATA, LISTED BY DATA TRACKING NUMBER 
163906 LB0104AMRU0185.012. Section 6.4.2 Focusing and Discrete Flow Paths in the TSW - 
Data Summary. Submittal date: 05/15/2001. 
163687 LB0304SMDCREV2.002. Seepage Modeling for Performance Assessment, Including 
Drift Collapse: Summary Plot Files and Tables. Submittal date: 04/11/2003. 
164337 LB0307SEEPDRCL.002. Seepage Into Collapsed Drift: Data Summary. Submittal 
date: 07/21/2003. 
166116 LB0310AMRU0120.002. Mathcad 11 Spreadsheets for Probabilistic Seepage 
Evaluation. Submittal date: 10/23/2003. 
February 2004 173 of 176 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
163531 MO0210MWDEXC01.008. CSNF Results in Excel Spreadsheets - 
CSNF_Spreadsheets. Submittal date: 10/22/2002. 
166801 MO0401SPAMCRAE.000. Materials Corrosion Rates in Aqueous Environments. 
Submittal date: 01/16/2004. 
164527 MO0307SEPFEPS4.000. LA FEP List. Submittal date: 07/31/2003 
164822 MO0308SPACALSS.002. Computational Algorithm for the Seismic Scenario for 
TSPA. Submittal date: 8/13/2003. 
165892 MO0309SPABRNAM.001. Calculated Parameters for Boron Abstraction Model for 
CSNF Waste Packages. Submittal date: 09/25/2003. 
147618 MO9906RIB00048.000. Waste Package Material Properties: Waste Form Materials. 
Submittal date: 6/9/1999. 
8.4 SOFTWARE CODES 
154060 CRWMS M&O 1998. Software Code: MCNP. V4B2LV. HP, HPUX 9.07 and 10.20; 
PC, Windows 95; Sun, Solaris 2.6. 30033 V4B2LV. 
160873 BSC 2002. Software Code: SAPHIRE. V7.18. PC - Windows 2000/NT 4.0. 10325- 
7.18-00. 
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ATTACHMENT Title 
I. 
II. 
III. 
IV. 
V. 
VI. 
VII. 
ANL-EBS-NU-000008 REV 00 
9. ATTACHMENTS 
Glossary 
SAPHIRE Model Used for Criticality FEPs Screening Analysis 
Seismic Calculations Spreadsheets (Output from MATHCAD Files) 
Seepage Analysis Spreadsheets (Output from MATHCAD Files) 
Neutronit Corrosion Spreadsheets (Output from MATHCAD Files) 
Listing of Files on CD-ROM 
Read-Only Compact Disc (CD-ROM) 
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ATTACHMENT I 
GLOSSARY 
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Absorption 
Advection 
Aleatory 
Burnup1 
Chain reaction1 
Cladding1 
Critical condition 
Critical limit 
Criticality1 
ANL-EBS-NU-000008 REV 00 
ATTACHMENT I - GLOSSARY 
(1) To take in and make part of an existent whole. 
(2) To receive without recoil. 
(1) The usually horizontal movement of a mass of fluid (as air or an 
ocean current). 
(2) The process in which solutes are transported by groundwater 
movement. 
Having a random character, in the sense that the likelihood of taking 
place over various intervals of time can be estimated, but it is not 
possible to determine whether or not, they will actually occur. See 
epistemic. 
A measure of nuclear reactor fuel consumption expressed either as the 
percentage of fuel atoms that have undergone fission or as the amount of 
energy produced per unit weight of fuel. 
A continuing series of nuclear fission events. Neutrons produced by a 
split nucleus collide with and split other nuclei causing a chain of fission 
events. 
The metal outer sheath of a fuel rod generally made of a zirconium alloy, 
and in the early nuclear power reactors, of stainless steel. Intended to 
protect the uranium dioxide pellets, which are the nuclear fuel, from 
dissolution by exposure to high temperature water under operating 
conditions in a reactor. 
A self-sustaining nuclear fission chain reaction: When the number of 
neutrons resulting from fission in each generation equals the number of 
neutrons lost by both absorption and leakage in the preceding generation. 
In this circumstance the effective neutron multiplication factor equals 
one (keff= 1). 
The value of keff at which a configuration is considered potentially 
critical, as characterized by statistical tolerance limits. 
(1) A condition that would require the original waste form, which is part 
of the waste package, to be exposed to degradation, followed by 
conditions that would allow concentration of sufficient nuclear fuel, the 
presence of neutron moderators, the absence of neutron absorbers, and 
favorable geometry. (2) The condition in which a fissile material 
sustains a chain reaction. It occurs when the number of neutrons present 
in one generation cycle equals the number generated in the previous 
cycle. The state is considered critical when a self-sustaining nuclear 
chan reaction is ongoing. 
February 2004 I-3 of I-8 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Criticality analysis 
Criticality control 
Criticality, fast 
Criticality, thermal 
Disposal2 
Disruptive event1 
Drift1 
Effective neutron 
multiplication factor 
Engineered barrier system2 The waste packages, including engineered components and systems other 
Epistemic 
ANL-EBS-NU-000008 REV 00 
A mathematical analysis, usually performed with a computer, of the 
neutron multiplication factor of a system or configuration that contains 
material capable of undergoing a self-sustaining chain reaction. 
The suite of measures taken to control the occurrence of self-sustaining 
nuclear chain reactions in fissionable materials, including spent nuclear 
fuel. For postclosure disposal applications, criticality control is ensuring 
that the probability of a criticality event is so small that the occurrence is 
unlikely, and the risk that any criticality will violate repository 
performance objectives is negligible. 
A critical condition where fast (high-energy) neutrons sustain the fission 
process. 
A critical condition where thermal (low-energy) neutrons sustain the 
fission process. 
The emplacement of radioactive waste in a geological repository with the 
intent of leaving it in there permanently. 
An off-normal event that, in the case of the repository, includes volcanic 
activity, seismic activity, and nuclear criticality. Disruptive events have 
two possible effects: (1) direct release of radioactivity to the surface, or 
(2) alteration of the nominal behavior of the system. For the purposes of 
screening features, events, and processes for total system performance 
assessment, a disruptive event is defined as an event that has a significant 
effect on the expected annual dose and that has a probability of 
occurrence during the 10,000 year period of performance less than 1.0, 
but greater than a cutoff of 0.0001. 
From mining terminology, a horizontal, underground, passage. The 
nearly horizontal underground passageways from the shaft(s) to the 
alcoves and rooms. Drifts includes excavations for emplacement 
(emplacement drifts) and access (access mains). 
See critical condition. 
than the waste package (e.g., drip shields), and the underground facility. 
Refers to the state of knowledge about a parameter because the data may 
be limited or because there may be alternative interpretations of the 
available data. The state of knowledge about the exact value of the 
parameter can increase through testing and data collection such that the 
uncertainty is reducible. See aleatory. 
February 2004 I-4 of I-8 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Events1 
Far-field 
Far-field for criticality 
Features1 
Fissile materials 
Fissionable materials 
High-level waste 
High-level radioactive waste (1) The highly radioactive material resulting from the reprocessing of 
spent nuclear fuel, including liquid waste produced directly in 
reprocessing, and any solid material derived from such liquid waste 
that contains fission products in sufficient concentration. 
(2) Other highly radioactive materials that the U.S. Nuclear Regulatory 
Initiating event2 
keff 
License application1 
ANL-EBS-NU-000008 REV 00 
(1) Occurrences that have a specific starting time and, usually, a duration 
shorter than the time being simulated in a model. (2) Uncertain 
occurrences that take place within a short time relative to the time frame 
of the model. For the purposes of screening features, events, and 
processes for total system performance assessment, an event is defined to 
be a natural or human-caused phenomenon that has a potential to affect 
disposal system performance and that occurs during an interval that is 
short compared with the period of performance. 
With reference to processes, those occurring at the scale of the mountain. 
The area of the geosphere and biosphere far enough away from the 
geological repository that, when numerically modeled, represents 
releases from the geological repository as a homogeneous, single-source 
effect. 
Far-field for criticality is defined as the space beyond the drift wall (i.e., 
in the host rock of the geological repository). 
Physical, chemical, thermal or temporal characteristics of the site or 
repository system. For the purpose of screening features, events, and 
processes for total system performance assessment, a feature is defined 
to be an object, structure or condition that has a potential to affect 
disposal system performance. 
Fissile materials are those materials that will undergo fission with 
thermal (slow) neutrons. The three primary fissile materials are 
uranium-233, uranium-235, and plutonium-239. 
Fissionable materials are those materials that will undergo fission by 
neutrons with sufficient energy. Note that while all fissile materials are 
fissionable, the reverse is not true. Although fissile, rather than 
fissionable, is used in most places in this report, fissionable may be 
applicable in some configurations. 
See high-level radioactive waste. 
Commission, consistent with existing law, determines by rule require 
permanent isolation. 
A natural or human induced event that causes an event sequence. 
Effective neutron multiplication factor. 
An application to the U.S. Nuclear Regulatory Commission for a license 
to construct and operate a repository. 
February 2004 I-5 of I-8 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Lithophysae 
Lithophysal 
Near-field1 
Near-field for criticality 
Neutron, fast 
Neutron, thermal 
Neutron leakage 
Neutron moderator 
Nuclear fission 
Performance assessment2 
Period of performance 
Permanent closure2 
ANL-EBS-NU-000008 REV 00 
Voids having concentric shells of finely crystalline alkali feldspar, 
quartz, and other materials that were formed by entrapped gas that later 
escaped. 
Pertaining to tuff units with lithophysae. 
The area and conditions within the geological repository including the 
drifts and waste packages and the rock immediately surrounding the 
drifts. The region around the repository where the natural hydrogeologic 
system has been significantly impacted by the excavation of the 
repository and the emplacement of waste. 
The area outside the waste package and inside the drift wall (including 
the drift liner and invert). 
A neutron with kinetic energy greater than its surroundings when 
released during fission. 
A neutron that has (by collision with other particles) been slowed to an 
energy state equal to that of its surroundings, typically on the order of 
0.025 eV (electron volts) and having a velocity of approximately 
2,200 m/s. 
The fraction of neutrons lost as result of escape from a fissile system. 
A material such as ordinary water, heavy water, or graphite that is used 
to slow down fast (high-energy) neutrons to thermal (low-energy) 
neutrons, thus increasing the likelihood of fission. 
The act of splitting a nucleus into two or more nuclei, resulting in the 
release of two or more neutrons and a relatively large amount of energy. 
A probabilistic analysis that: 
(1) Identifies the features, events, and processes that might affect the 
performance of the geological repository; 
(2) Examines the effects of those features, events, and processes on the 
performance of the geological repository; and 
(3) Estimates the consequences (e.g., radiological exposures to the 
reasonably maximally exposed individual, radionuclide releases to 
the accessible environment) of releases from the geologic repository. 
10,000 years after permanent closure of the geologic repository. 
Final back-filling of the underground facility, if appropriate, and the 
sealing of shafts, ramps, and boreholes. 
February 2004 I-6 of I-8 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Probabilistic1 
Probability1 
Probability distribution1 
Processes1 
Pyroclastic 
Safety analysis, preclosure2 A systematic examination of the site; the design; and the potential 
Saturated zone2 
Scenario1 
Scenario class1 
ANL-EBS-NU-000008 REV 00 
(1) Based on or subject to probability. (2) Involving a variate, such as 
temperature or porosity. At each instance of time, the variate may take 
on any of the values of a specified set with a certain probability. Data 
from a probabilistic process are an ordered set of observations, each of 
which is one item from a probability distribution. 
The chance that an outcome will occur from the set of possible 
outcomes. Statistical probability examines actual events and can be 
verified by observation or sampling. Knowledge of the exact probability 
of an event is usually limited by the inability to know, or compile, the 
complete set of possible outcomes over time or space, a degree of belief. 
The set of outcomes (values) and their corresponding probabilities for a 
random variable. 
Phenomena and activities that have gradual, continuous interactions with 
the system being modeled. For purposes of screening features, events, 
and processes for total system performance assessment, a process is 
defined as a natural or human-caused phenomenon that has a potential to 
affect disposal system performance and that operates during all or a 
significant part of the period of performance. 
Of or relating to individual particles or fragments of clastic rock material 
of any size formed by volcanic explosion or ejected from a volcanic vent. 
hazards, initiating events, and event sequences and their consequences 
(e.g., radiological exposures to workers and the public). The analysis 
identifies structures, systems, and components important to safety. 
That part of the earths crust beneath the regional water table in which all 
voids, large and small, are ideally filled with water under pressure 
greater than atmospheric. See also unsaturated zone. 
A well-defined, connected sequence of features, events, and processes 
that can be thought of as an outline of a future condition of the repository 
system. Scenarios can be undisturbed, in which case the performance 
would be expected, or nominal, behavior for the system. Scenarios can 
also be disturbed, if altered by disruptive events such as human intrusion 
or natural phenomena such as volcanism or nuclear criticality. 
A set of related scenarios that share sufficient similarities that they can 
usefully be aggregated for the purposes of screening or analysis. The 
number and breadth of scenario classes depends on the resolution at 
which scenarios have been defined. Coarsely defined scenarios result in 
fewer, broad scenario classes, whereas narrowly defined scenarios result 
in many narrow scenario classes. Scenario classes (and scenarios) 
should be aggregated at the coarsest level at which a technically sound 
argument can be made, while still retaining adequate detail for the 
purposes of analysis. 
February 2004 I-7 of I-8 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seepage1 
Seismic1 
Spent nuclear fuel1 
Uncertainty1 
Unsaturated zone2 
Variability (statistical)1 
Waste form2 
Water table2 
1 Definition cited from glossary of Yucca Mountain Review Plan (NRC 2003). 
2 Definition cited from 10 CFR 63.2 (10 CFR 63). 
ANL-EBS-NU-000008 REV 00 
The inflow of groundwater moving in fractures or pore spaces of 
permeable rock to an open space in the rock such as a drift. Seepage rate 
is the percolation flux that enters the drift. Seepage is an important 
factor in waste package degradation and mobilization and migration of 
radionuclides out of the repository. 
Pertaining to, characteristic of, or produced by earthquakes or earth 
vibrations. 
Fuel that has been withdrawn from a nuclear reactor following 
irradiation, the constituent elements of which have not been separated by 
reprocessing. Spent fuel that has been burned (irradiated) in a reactor to 
the extent that it no longer makes an efficient contribution to a nuclear 
chain reaction. This fuel is more radioactive than it was before 
irradiation, and releases significant amounts of heat from the decay of its 
fission product radionuclides. See burnup. 
A measure of how much a calculated or estimated value varies from the 
unknown true value. 
The zone between the land surface and the regional water table. 
Generally, fluid pressure in this zone is less than atmospheric pressure, 
and some of the voids may contain air or other gases at atmospheric 
pressure. Beneath flooded areas or in perched waster bodies, the fluid 
pressure locally may be greater than atmospheric. 
A measure of how a quantity varies over time or space. 
The radioactive waste materials and any encapsulating or stabilizing 
matrix. 
That surface in a groundwater body, separating the unsaturated zone 
from the saturated zone, at which the water pressure is atmospheric. 
February 2004 I-8 of I-8 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ATTACHMENT II 
SAPHIRE MODEL USED FOR CRITICALITY FEPS SCREENING ANALYSIS 
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ATTACHMENT II - SAPHIRE MODEL USED FOR CRITICALTY 
FEPS SCREENING ANALYSIS 
The SAPHIRE model used for the evaluation of the criticality FEPs screening analysis is based 
on the configuration generator model (BSC 2003 [DIRS 165629]). The event trees and fault 
trees used in the SAPHIRE criticality FEPs analysis are presented and discussed in Section II.1. 
The logic rules used to assign the basic event probabilities and direct the evaluation of the event 
trees are presented in Sections II.2 through II.5. The basic event values used in this analysis are 
presented in Section II.6 and the SAPHIRE-calculated end-state results are presented in 
Section II.7. 
II.1 SAPHIRE EVENT TREES AND FAULT TREES 
Figures II-1 through II-4 present the event trees used in the criticality FEPs screening analysis. 
The fault trees supporting the top events of the FEPS event trees are presented in Figures II-5 
through II-8. 
Figure II-1 presents the WP_TYPE event tree used for determining the waste form and waste 
package type inventory fraction. Figure II-2 presents the 21-PWR with Absorber Plates Waste 
Package, an example of the eight waste package type event trees. These event trees are used to 
initiate the calculation of the waste package flooding probabilities on a per waste package type 
basis. Figure II-3 presents the CASE event tree for directing the SAPHIRE evaluation of the 
criticality FEPs cases  base case, seismic disruptive event, and rockfall disruptive event (the 
igneous disruptive event is not evaluated by SAPHIRE). Finally, Figure II-4 presents the FEPS 
event tree for processing the events that determine the probability of waste package flooding. 
Figure II-5 through II-8 present the fault trees and their supporting basic events used to quantify 
top events MS-IC-1, MS-IC-2, MS-IC-3, and MS-IC-4, respectively, of the FEPS event tree. 
Fault tree MS-IC-1 contains the basic event defining the probability of achieving the minimum 
required seepage rate. Fault tree MS-IC-2 contains the basic events and their relational logic to 
define the probability of drip shield failure. Fault tree MC-IC-3 contains the basic events and 
their relational logic to define the probability of waste package failure. Finally, fault tree 
MS-IC-4 contains the basic events and their relational logic to define the probability of bathtub 
configuration forming and enduring over the period of performance. 
February 2004 II-3 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Waste Package 
Fraction 
Waste Package Type 
Percent Breakdown 
Waste Form 
Sub-Type 
Waste Form Type 
Percentages 
Percentages by Originator 
WP WP-SOURCE WP-IND FRACTION # END-STATE-NAMES WP-TYPE 
21-PWR Absorber Plate 
(38.4% of total inventory) 
1 CSNF-21PWR-ABS-WP 3.840E-001 
21-PWR Control Rod 
(0.8% of total inventory) 
PWR 
(40.7% of total inventory) 
8.000E-003 2 CSNF-21PWR-CR-WP 
12-PWR Long 
(1.5% of total inventory) Commercial CSNF 
(66.8% of total inventory) 3 CSNF-12PWR-LONG-WP 1.500E-002 
44-BWR 
(25.3% of total inventory) 
2.53E-001 4 CSNF-44BWR-WP 
BWR 
(26.1% of total inventory) 
24-BWR 
(0.8% of total inventory) 
8.000E-003 5 CSNF-24BWR-WP 
DOE Long 
(10.3% of total inventory) 
1.030E-001 6 DOESNF-LONG-WP Waste Package Fraction 
DOE Short 
(18.9% of total inventory) 
DOE SNF 
(30.5% of total inventory) 
1.890E-001 7 DOESNF-SHORT-WP 
DOE MCO 
(1.3% of total inventory) 
1.300E-002 8 DOESNF-MCO-WP 
Naval Short 
(1.3% of total inventory) 
1.300E-002 9 NAVAL-SHORT-WP Naval SNF 
(2.7% of total inventory) 
Naval Long 
(1.4% of total inventory) 
1.400E-002 10 NAVAL-LONG-WP 
Figure II-1. Waste Form and Waste Package Type Inventory Fraction Event Tree 
February 2004 II-4 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Initiate Evaluation of 
21-PWR with Absorber Plates 
Waste Package Type 
END_STATE_NAMES # <PASS> WP01-21-PWR-AP 
CASE 1 T 
Figure II-2. Waste Package Type Event Tree 
February 2004 II-5 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Rock Type 
of Drift 
FEPs Initiating 
Event 
Incoming Waste 
Package Type 
Identifier 
IE1 FEP END-STATE DRIFT # 
Nonlithophysal (15%) 
FEPS 1 T Base Case 
Probability = 1.0 
Lithophysal (85%) 
FEPS 2 T 
Nonlithophysal FEPS 3 T Seismic Disruptive Event 
Probability = 1.0 
Lithophysal WP Type FEPS 4 T 
Nonlithophysal FEPS 5 T Rock Fall Disruptive Event 
Probability = 1.0 
Lithophysal FEPS 6 T 
Igneous Disruptive Event 
Probability = 1.6E-4 
OK 7 
Figure II-3. Criticality FEPs Case Assignment Event Tree 
February 2004 II-6 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Transfer from 
Initiating Event Tree 
Sufficient Water 
Reaches Drift 
Drip Shield 
Fails 
Waste Package 
Fails 
Water Accumulates 
in Waste Package 
IE3 MS-IC-1 MS-IC-2 MS-IC-3 END-STATE-NAMES MS-IC-4 # 
Insufficient Water 
Reaches the Drift 
NO-CRITICALITY 1 
Water Does Not 
Drip on Waste Package 
NO-CRITICALITY 2 
Sufficient Water 
Reaches the Drift 
Water Does Not 
Penetrate Waste Package 
NO-CRITICALITY 3 
Water Drips on 
Waste Package 
Water Flows Through 
Waste Package FLOW-THRU-CONFIG 4 
Water Penetrates Top 
of Waste Package 
Water Accumulates 
in Waste Package 
BATHTUB-CONFIG 5 
Figure II-4. Probability of Waste Package Flooding Event Tree 
February 2004 II-7 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Time-Independent 
Drip Shield 
Failures 
MS-IC-2-2 
Early Failure 
of Drip Shield 
Drip Shield 
Failure Due to 
Seismic Event 
BE-DS-FABRICATION BE-DS-SEISMIC 
Drip Shield 
Failure Due to 
Improper installation 
Drip Shield 
Failure Due to 
Thermal Expansion 
BE-DS-THERM-EXPAN BE-DS-EMPLACEMENT 
Sufficient Water 
Reaches Drift by 
10,000 years 
MS-IC-1 
Water Reaches 
Drift at 
10,000 years 
BE-SEEPAGE-10K 
Probability of 
Drip Shield 
Failure 
MS-IC-2 
Time-Dependent 
Drip Shield 
Failures 
MS-IC-2-1 
Drip Shield 
Failure Due to 
Rock Fall 
Drip Shield 
Degrades to 
Localized Corrosion 
MS-IC-2-4 BE-DS-ROCK-FALL 
Figure II-5. MS-IC-1 Fault Tree (Minimum Required Seepage Rate) 
Figure II-6. MS-IC-2 Fault Tree (Drip Shield Failure) 
Drip Shield 
Degrades Due to 
General Corrosion 
BE-DS-GENCOR-10K
Drip Shield 
Degrades Due to 
Crevice Corrosion 
BE-DS-CREVICE-10K 
ANL-EBS-NU-000008 REV 00 
Drip Shield Degrades 
Due to Stress 
Corrosion Cracking 
BE-DS-SCC-10K 
Drip Shield 
Degrades Due to 
Pitting Corrosion 
BE-DS-PITTING-10K 
February 2004 II-8 of II-18 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Time-Independent 
Waste Package 
Failures 
MS-IC-3-2 
Failure of Waste 
Package Due to 
Seismic Event 
Early Failure 
of Waste 
Package 
BE-WP-SEISMIC BE-WP-EARLY-F 
Figure II-7. MS-IC-3 Fault Tree (Waste Package Failure) 
Liquid Accumulates 
In Waste Package 
BE-MS-IC-4 
Figure II-8. MS-IC-4 Fault Tree (Bathtub Configuration) 
Waste Package 
Penetration at 
Top Surface 
MS-IC-3
Waste Package 
Degrades Due to 
SCC 
BE-WP-SCC-10K 
Liquid accumulates 
in waste package 
MS-IC-4 
Time-dependent 
Waste Package 
Failures 
MS-IC-3-3 
Waste Package 
Degrades Due to 
Localized Corrosion 
MS-IC-3-4 BE-WP-GENCOR-10K 
Waste Package 
Degrades Due to 
Crevice Corrosion 
Waste Package 
Degrades Due to 
Pitting Corrosion 
BE-WP-PITTING-10K BE-WP-CREVICE-10K 
Probability of 
being in a bathtub 
at 10,000 yrs 
BE-BATHTUB-10K 
ANL-EBS-NU-000008 REV 00 II-9 of II-18 
Waste Package 
Degrades Due to 
General Corrosion 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
II.2 LINKAGE RULES FOR THE WP_TYPE EVENT TREE (FIGURE II-1) 
The following linkage rules are used to assign the basic event values representing the percentage 
of total waste package inventory for the various waste form types, waste form subtypes, and 
waste package types. 
IF ALWAYS THEN 
/WP-SOURCE = WP-SOURCE-CSNF; 
WP-SOURCE[1] = WP-SOURCE-DOE; 
WP-SOURCE[2] = WP-SOURCE-NAVAL; 
ENDIF; 
| 
IF /WP-SOURCE THEN 
/WP-IND = WP-PWR; 
WP-IND = WP-BWR; 
ENDIF; 
| 
IF (/WP-SOURCE * /WP-IND) THEN 
/WP-TYPE = WP-TYPE-21ABS; 
WP-TYPE[1] = WP-TYPE-21CR; 
WP-TYPE[2] = WP-TYPE-12LONG; 
ENDIF; 
| 
IF (/WP-SOURCE * WP-IND) THEN 
/WP-TYPE = WP-TYPE-44BWR; 
WP-TYPE = WP-TYPE-24BWR; 
ENDIF; 
| 
IF WP-SOURCE[1] THEN 
/WP-TYPE = WP-DOE-LONG; 
WP-TYPE[1] = WP-DOE-SHORT; 
WP-TYPE[2] = WP-DOE-MCO; 
ENDIF; 
| 
IF WP-SOURCE[2] THEN 
/WP-TYPE = WP-NAVAL-SHORT; 
WP-TYPE = WP-NAVAL-LONG; 
ENDIF; 
February 2004 II-10 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
II.3 LINKAGE RULES FOR THE CASE EVENT TREE (FIGURE II-3) 
The following linkage rules are used to substitute the basic event value for the four criticality 
FEPs cases considered in the SAPHIRE analysis  (1) Base Case (probability =1.0), (2) Seismic 
Disruptive Event (probability = 1.0); (3) Rock Fall Disruptive Event (probability = 1.0), , and 
Igneous Disruptive Event (probability = 1.7E-4). 
| 
| SET FEP INITIATING EVENTS 
| 
IF ALWAYS THEN 
/FEP = FEP_NOMINAL; 
FEP[1] = FEP_SEISMIC; 
FEP[2] = FEP_ROCKFALL; 
FEP[3] = FEP_IGNEOUS; 
ENDIF 
February 2004 II-11 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
II.4 PROJECT RECOVERY RULES 
The following recovery rules are used to substitute basic events in the event tree sequences and 
to prevent mutually exclusive events from occurring within a sequence. 
| 
| Basic event substitution rules for MS-IC-1 Fault Tree 
| 
if (FEP_SEISMIC * BE-SEEPAGE-10K * WP01-21-PWR-AP * DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP1-L; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP01-21-PWR-AP * /DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP1-NL; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP02-21-PWR-CR * DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP2-L; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP02-21-PWR-CR * /DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP2-NL; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP03-12-PWR-LONG * DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP3-L; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP03-12-PWR-LONG * /DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP3-NL; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP04-44-BWR * DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP4-L; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP04-44-BWR * /DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP4-NL; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP05-24-BWR * DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP5-L; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP05-24-BWR * /DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP5-NL; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP06-DOE-LONG * DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP6-L; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP06-DOE-LONG * /DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP6-NL; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP07-DOE-SHORT * DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP7-L; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP07-DOE-SHORT * /DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP7-NL; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP08-DOE-MCO * DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP8-L; 
elsif (FEP_SEISMIC * BE-SEEPAGE-10K * WP08-DOE-MCO * /DRIFT) then 
DeleteEvent = BE-SEEPAGE-10K; 
AddEvent = BE-SEEPAGE-WP8-NL; 
endif; 
February 2004 II-12 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
| 
| Basic event substitution rules for MS-IC-2 Fault Tree 
| 
if (BE-DS-SEISMIC * FEP_SEISMIC) then 
DeleteEvent = BE-DS-SEISMIC; 
AddEvent = BE-DS-SEISMIC1; 
endif; 
| 
| Basic event substitution rules for MS-IC-3 Fault Tree 
| 
if (BE-WP-EARLY-F * FEP_SEISMIC) then 
DeleteEvent = BE-WP-EARLY-F; 
AddEvent = BE-WP-EARLY-F1; 
elsif (BE-WP-SEISMIC * FEP_SEISMIC) then 
DeleteEvent = BE-WP-SEISMIC; 
AddEvent = BE-WP-SEISMIC1; 
endif; 
| 
| Recovery Rules for MS-IC-4 Fault Tree 
| 
if (BE-WP-EARLY-F + BE-WP-SEISMIC1) * BE-MS-IC-4 then 
DeleteEvent = BE-MS-IC-4; 
endif; 
February 2004 II-13 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
II.5 PROJECT PARTITION RULES 
The following partition rules are used to create encoded end states for the FEPS event tree 
sequences that result in either a bathtub or flow-through configuration. These encoded end states 
represent sequences for the three SAPHIRE evaluated criticality FEPs cases (igneous disruptive 
event not evaluated in SAPHIRE), eight of the ten waste package types (naval waste package 
types are not considered in this evaluation), and the two geological zones (lithophysal and 
nonlithophysal) considered in this analysis. 
| 
| DEFINE VARIABLES 
| 
_FEPIE1_ = /FEP_BASECASE; 
_FEPIE2_ = FEP_ROCKFALL; 
_FEPIE3_ = FEP_SEISMIC; 
_BATHTUB_ = ~/MS-IC-4; 
_NOBATHTUB_ = /MS-IC-4; 
_LITH_ = DRIFT; 
_NONLITH_ = /DRIFT; 
_SORT_ = (~/MS-IC-1 * ~/MS-IC-2 * ~/MS-IC-3 * (DRIFT + /DRIFT)); 
| 
| 
if _FEPIE1_ * _SORT_ then 
GlobalPartition="N-??????????"; 
elsif _FEPIE2_ * _SORT_ then 
GlobalPartition="R-??????????"; 
elsif _FEPIE3_ * _SORT_ then 
GlobalPartition="S-??????????"; 
endif; 
| 
if init(WP01-21-PWR-AP) * _SORT_ then 
GlobalPartition="??PWR1??????"; 
elsif init(WP02-21-PWR-CR) * _SORT_ then 
GlobalPartition="??PWR2??????"; 
elsif init(WP03-12-PWR-LONG) * _SORT_ then 
GlobalPartition="??PWR3??????"; 
elsif init(WP04-44-BWR) * _SORT_ then 
GlobalPartition="??BWR1??????"; 
elsif init(WP05-24-BWR) * _SORT_ then 
GlobalPartition="??BWR2??????"; 
elsif init(WP06-DOE-LONG) * _SORT_ then 
GlobalPartition="??DOE0??????"; 
elsif init(WP07-DOE-SHORT) * _SORT_ then 
GlobalPartition="??DOE0??????"; 
elsif init(WP08-DOE-MCO) * _SORT_ then 
GlobalPartition="??DOE0??????"; 
endif; 
| 
if _BATHTUB_ * _SORT_ then 
GlobalPartition="??????-BT???"; 
elsif _NOBATHTUB_ * _SORT_ then 
GlobalPartition="??????-FT???"; 
endif; 
| 
if _LITH_ * _SORT_ then 
GlobalPartition="?????????-LI"; 
elsif _NONLITH_ * _SORT_ then 
GlobalPartition="?????????-NL"; 
endif; 
II-14 of II-18 ANL-EBS-NU-000008 REV 00 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
II.6 BASIC EVENTS FOR SAPHIRE ANALYSIS 
Table II-1 lists the basic event values used in the Criticality FEPs evaluations. 
Table II-1. Basic Events For SAPHIRE Criticality FEPs Analysis 
Basic Event Name 
BASIC EVENTS FOR THE MS-IC-1 TOP EVENT OF THE FEPS EVENT TREE 
BE-SEEPAGE-10K 
BE-SEEPAGE-WP1-L 
BE-SEEPAGE-WP1-NL 
BE-SEEPAGE-WP2-L 
BE-SEEPAGE-WP2-NL 
BE-SEEPAGE-WP3-L 
BE-SEEPAGE-WP3-NL 
BE-SEEPAGE-WP4-NL 
BE-SEEPAGE-WP4-L 
BE-SEEPAGE-WP5-L 
BE-SEEPAGE-WP5-NL 
BE-SEEPAGE-WP6-L 
BE-SEEPAGE-WP6-NL 
BE-SEEPAGE-WP7-L 
BE-SEEPAGE-WP7-NL 
BE-SEEPAGE-WP8-L 
BE-SEEPAGE-WP8-NL 
BASIC EVENTS FOR THE MS-IC-1 TOP EVENT OF THE FEPS EVENT TREE 
FOR THE SEISMIC SENSITIVITY CASE 
BE-SEEPAGE-WP4-NL 
BE-SEEPAGE-WP4-L 
BASIC EVENTS FOR THE MS-IC-2 TOP EVENT OF THE FEPS EVENT TREE 
BE-DS-CREVICE-10K Drip Shield Degrades Due to Crevice Corrosion 
BE-DS-EMPLACEMENT Drip Shield Failure Due to Improper installation 
BE-DS-FABRICATION 
BE-DS-GENCOR-10K 
BE-DS-PITTING-10K 
BE-DS-ROCK-FALL 
BE-DS-ROCK-FALL1 
BE-DS-SCC-10K 
BE-DS-SEISMIC 
BE-DS-SEISMIC1 
BE-DS-THERM-EXPAN 
BASIC EVENTS FOR THE MS-IC-3 TOP EVENT OF THE FEPS EVENT TREE 
BE-WP-CREVICE-10K 
BE-WP-EARLY-F 
BE-WP-EARLY-F1 
BE-WP-GENCOR-10K 
BE-WP-PITTING-10K 
Description 
Water Reaches Drift at 10,000 years 
minimum water flux at 10K yrs - 21-PWR AP WP - seismic lith 
minimum water flux at 10K yrs - 21-PWR AP WP - seismic nonlith 
minimum water flux at 10K yrs - 21-PWR CR WP - seismic lith 
minimum water flux at 10K yrs - 21-PWR CR WP - seismic nonlith 
minimum water flux at 10K yrs - 12-PWR WP - seismic lith 
minimum water flux at 10K yrs - 12-PWR WP - seismic nonlith 
minimum water flux at 10K yrs - 44-BWR WP - seismic nonlith 
minimum water flux at 10K yrs - 44-BWR WP - seismic lith 
minimum water flux at 10K yrs - 24-BWR WP - seismic lith 
minimum water flux at 10K yrs - 24-BWR WP - seismic nonlith 
minimum water flux at 10K yrs - DOE Long WP - seismic lith 
minimum water flux at 10K yrs - DOE Long WP - seismic nonlith 
minimum water flux at 10K yrs - DOE Short WP - seismic lith 
minimum water flux at 10K yrs - DOE Short WP - seismic nonlith 
minimum water flux at 10K yrs - DOE MCO WP - seismic lith 
minimum water flux at 10K yrs - DOE MCO WP - seismic nonlith 
minimum water flux at 10K yrs - 44-BWR WP - seismic nonlith 
minimum water flux at 10K yrs - 44-BWR WP - seismic lith 
Early Failure of Drip Shield 
Drip Shield Degrades Due to General Corrosion 
Drip Shield Degrades Due to Pitting Corrosion 
Drip Shield Failure Due to Rock Fall - Lithophysal Zone 
Drip Shield Failure Due to Rock Fall - Nonlithophysal Zone 
Drip Shield Degrades Due to Stress Corrosion Cracking 
Drip Shield Failure Due to Seismic Event - Base Case 
Drip Shield Failure Due to Seismic Event - Seismic Event 
Drip Shield Failure Due to Thermal Expansion 
Waste Package Degrades Due to Crevice Corrosion 
Early Failure of Waste Package 
Early Failure of Waste Package - Seismic Event 
Waste Package Degrades Due to General Corrosion 
Waste Package Degrades Due to Pitting Corrosion 
ANL-EBS-NU-000008 REV 00 II-15 of II-18 
Probability 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
9.740E-007 
1.525E-006 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
9.924E-012 
1.242E-010 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
0.000E+000 
3.310E-002 
0.000E+000 
0.000E+000 
1.000E+000 
0.000E+000 
0.000E+000 
2.800E-005 
0.000E+000 
0.000E+000 
0.000E+000 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table II-1. Basic Events For SAPHIRE Criticality FEPs Analysis (Continued) 
Probability Description Basic Event Name 
0.000E+000 
0.000E+000 
1.000E+000 
Waste Package Degrades Due to SCC 
Failure of Waste Package Due to Seismic Event 
Failure of Waste Package Due to Seismic Event - Seismic Event 
BE-WP-SCC-10K 
BE-WP-SEISMIC 
BE-WP-SEISMIC1 
BASIC EVENTS FOR THE MS-IC-4 TOP EVENT OF THE FEPS EVENT TREE 
1.000E+000 Probability of Being in a Bathtub at 10,000 yrs BE-BATHTUB-10K 
1.000E+000 Liquid Accumulates In Waste Package BE-MS-IC-4 
BASIC EVENTS FOR THE DRIFT TOP EVENT OF THE INITEVENT EVENT TREE 
Rock Type Fraction in Drifts DRIFT 8.500E-001 
BASIC EVENTS FOR FEP TOP EVENT OF INITEVENT EVENT TREE 
1.000E+000 
Igneous Case Disruptive Event 
FEP_BASECASE Base Case Disruptive Event 
FEP_SEISMIC Seismic Case Disruptive Event 
FEP_ROCKFALL Rockfall Case Disruptive Event 
FEP_IGNEOUS 
1.000E+000 
1.000E+000 
1.600E-004 
BASIC EVENTS FOR THE WP-SOURCE TOP EVENT OF THE WP_TYPE EVENT TREE 
WP-SOURCE-CSNF 
WP-SOURCE-DOE 
WP-SOURCE-NAVAL 
Commercial SNF (66.8%) 3.320E-001 
DOE SNF (30.5%) 3.050E-001 
Naval SNF (2.7%) 2.700E-002 
BASIC EVENTS FOR THE WP-IND TOP EVENT OF THE WP_TYPE EVENT TREE 
WP-BWR BWR (26.1%) 3.907E-001 
WP-PWR PWR (40.7%) 3.907E-001 
BASIC EVENTS FOR THE WP-TYPE TOP EVENT OF THE WP_TYPE EVENT TREE 
Naval Short (1.3%) 
WP-TYPE-21ABS 21-PWR Absorber Plate (38.4%) 
WP-TYPE-21CR 21-PWR Control Rod (0.8%) 
WP-TYPE-12LONG 12-PWR Long (1.5%) 
WP-TYPE-44BWR 44-BWR (25.3%) 
WP-TYPE-24BWR 24-BWR (0.8%) 
WP-DOE-SHORT DOE Short (18.9%) 
WP-DOE-LONG DOE Long (10.3%) 
WP-DOE-MCO DOE MCO (1.3%) 
WP-NAVAL-LONG Naval Long (1.4%) 
WP-NAVAL-SHORT 
5.651E-002 
1.965E-002 
3.685E-002 
3.065E-002 
3.065E-002 
6.196E-001 
6.622E-001 
4.262E-002 
5.185E-001 
5.185E-001 
INITIATING EVENTS WASTE PACKAGE TYPE EVENT TREES 
21-PWR With Absorber Plates Waste Package 
21-PWR With Control Rods Waste Package 
1.000E+000 
1.000E+000 
1.000E+000 
1.000E+000 
1.000E+000 
1.000E+000 
1.000E+000 
1.000E+000 DOE MCO Waste Package 
WP01-21-PWR-AP 
WP02-21-PWR-CR 
WP03-12-PWR-LONG 12-PWR Long Waste Package 
WP04-44-BWR 44-BWR Waste Package 
WP05-24-BWR 24-BWR Waste Package 
WP06-DOE-LONG DOE Long Waste Package 
WP07-DOE-SHORT DOE Short Waste Package 
WP08-DOE-MCO 
February 2004 II-16 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
II.7 SAPHIRE END STATE RESULTS FOR CRITICALITY FEPS ANALYSIS 
Table II-2 presents the criticality FEPs analysis probability results as calculated by SAPHIRE. 
The 12-character end-state names are encoded to capture the following information: 
1. The criticality FEPs analysis case (i.e., base case, seismic, or rockfall) 
 B  base case 
 S  seismic disruptive event 
 R  rockfall disruptive event 
2. The waste package type (e.g., 21-PWR with Absorber Plates Waste Package) 
 -PWR1  21-PWR with Absorber Plates Waste Package 
 -PWR2  21-PWR with Control Rods Waste Package 
 -PWR3  12-PWR Long Waste Package 
 -BWR1  44-BWR Waste Package 
 -BWR2  24-BWR Waste Package 
 -DOE0  DOE Waste Package (includes results for the DOE-Long, DOE-Short, and 
DOE-MCO waste package types) 
3. The waste package configuration type (i.e., flow or bathtub) 
 -BT  bathtub waste package configuration 
 -FT  flow-through waste package configuration 
4. The repository location (i.e., lithophysal or nonlithophysal zone) 
 -LI  lithophysal zone 
 -NL  nonlithophysal zone. 
The end states are assigned to each event tree sequence based on the project partition rules 
documented in Section II.5. 
February 2004 II-17 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Table II-2. SAPHIRE End State Probabilities 
Sensitivity Case Probabilitya Base Case Probability End State Name 
Base Case Results 
B-BWR1-BT-LI 0.00E+00 NA 
B-BWR1-BT-NL 0.00E+00 NA 
B-BWR2-BT-LI 0.00E+00 NA 
B-BWR2-BT-NL 0.00E+00 NA 
B-DOE0-BT-LI 0.00E+00 NA 
B-DOE0-BT-NL 0.00E+00 NA 
B-PWR1-BT-LI 0.00E+00 NA 
B-PWR1-BT-NL 0.00E+00 NA 
B-PWR2-BT-LI 0.00E+00 NA 
B-PWR2-BT-NL 0.00E+00 NA 
B-PWR3-BT-LI 0.00E+00 NA 
B-PWR3-BT-NL 0.00E+00 NA 
Rock Fall Disruptive Event Results 
R-BWR1-BT-LI 0.00E+00 NA 
R-BWR1-BT-NL 0.00E+00 NA 
R-BWR2-BT-LI 0.00E+00 NA 
R-BWR2-BT-NL 0.00E+00 NA 
R-DOE0-BT-LI 0.00E+00 NA 
R-DOE0-BT-NL 0.00E+00 NA 
R-PWR1-BT-LI 0.00E+00 NA 
R-PWR1-BT-NL 0.00E+00 NA 
R-PWR2-BT-LI 0.00E+00 NA 
R-PWR2-BT-NL 0.00E+00 NA 
R-PWR3-BT-LI 0.00E+00 NA 
R-PWR3-BT-NL 0.00E+00 NA 
Seismic Disruptive Event Results 
S-BWR1-BT-LI 1.30E-06 1.05E-10 
S-BWR1-BT-NL 1.46E-07 1.39E-12 
S-BWR2-BT-LI 0.00E+00 0.00E+00 
S-BWR2-BT-NL 0.00E+00 0.00E+00 
S-DOE0-BT-LI 0.00E+00 0.00E+00 
S-DOE0-BT-NL 0.00E+00 0.00E+00 
S-PWR1-BT-LI 0.00E+00 0.00E+00 
S-PWR1-BT-NL 0.00E+00 0.00E+00 
S-PWR2-BT-LI 0.00E+00 0.00E+00 
S-PWR2-BT-NL 0.00E+00 0.00E+00 
S-PWR3-BT-LI 0.00E+00 0.00E+00 
S-PWR3-BT-NL 0.00E+00 0.00E+00 
TOTALS = 1.44E-06 1.07E-10 
a Neutronit absorber plate thickness of 44-BWR Waste Package increased to 7 mm 
February 2004 II-18 of II-18 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ATTACHMENT III 
SEISMIC CALCULATIONS SPREADSHEETS (OUTPUT FROM MATHCAD FILES) 
III-1 of III-26 ANL-EBS-NU-000008 REV 00 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
INTENTIONALLY LEFT BLANK 
February 2004 III-2 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ATTACHMENT III - SEISMIC CALCULATIONS SPREADSHEETS (OUTPUT FROM 
MATHCAD FILES) 
The following sections present the Mathcad probability analyses for achieving the minimum 
required seepage to remove the neutron absorber material from the waste package as a result of a 
seismic disruptive event. These analyses are performed for a base case scenario (waste package 
baseline design) and a sensitivity case scenario (absorber plate thickness of the 44-BWR Waste 
Package is increase to 7 mm). 
III.1 SEISMIC BASE CASE ANALYSIS 
The following section presents the Mathcad probability analysis for achieving the minimum 
required seepage to remove the neutron absorber material from the waste package as a result of a 
seismic disruptive event. This analysis utilizes seepage information from Abstraction of Drift 
Seepage (BSC 2003 [DIRS 165564]), seismic event information and waste package and drip 
shield damage information from Seismic Consequence Abstraction (BSC 2003 [DIRS 161812]), 
and waste package boron loss information from Boron Loss From CSNF Waste Packages (BSC 
2003 [DIRS 165890]). The seismic base case is performed assuming that: 
 Water cannot enter a failed waste package until at least 700 years when the average 
drift wall temperature falls below the boiling point (Assumption 5.2.2); 
 The neutron absorber material degradation rate is 1.5 times that of stainless steel 
316 (Assumption 5.1.4); and 
 90 percent of the boron from the neutron absorber material can be removed from 
PWR waste packages with absorber plates and 50 percent can be removed from 
BWR waste packages with absorber plates without criticality concerns (Assumption 
5.1.5). 
The information contained in this section was obtained from the Attachment III.mcd Mathcad 
file of Attachment VII. 
February 2004 III-3 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Probability Calculation of Seismic Event Leading to Potential Criticality 
This calculation evaluates the damage to a drip sheild and a waste package due to a seismic 
event that has the potential to allow advective flow into a waste package. The equations and data 
are from Seismic Consequence Abstraction (BSC 2003 [DIRS 161812]) and 
DTN: M0308SPACALSS.002 (BSC 2003 [DIRS 164822]). 
Drip Shield Damage Due to Seismic Event & Rockfall 
Drip shield damage is based on location within the drift. The drip shields within the 
lithophysal area can only be damaged by vibratory ground motion and not rockfall because rocks 
here will crush and during a seismic event (BSC 2003 [DIRS 161812], Section 6.6.2). The drip 
shields located in the nonlithophysal, however, can be damaged by both rockfall and vibratory 
ground motion. Approximately 85% of the drift lies in the lithophysal zones. 
0 0 
0.55 0 0 0 
:= 2.44 2.68 DSUdat 2.44 0 DSLdat 
5.35 50 5.35 10 
20 50 
Drip shield lower bound 
damage from vibratory 
ground motion based on 
look-up table 
DTN:M0308SPACALSS.002 
(BSC 2003 [DIRS 164822]). 
.
 
. 
Drip shield upper bound 
damage from vibratory 
ground motion based on 
look-up table 
DTN:M0308SPACALSS.002 
(BSC 2003 [DIRS 164822]). 
1 . . 1 . . 
:= DSLdat DSld DSUdat PGVdsu DSud PGVdsl := 
Seismic Sampling Routine 
i 0..8 
:= 
DSLdat 
Mean Annual Exceedance Frequency 
....... 
DSUdat 
:= 
.... 
0 . .
MAF 
:= 
The peak ground velocity (PGV) and 
mean annual frequency are from 
DTN: M0308SPACALSS.002 
(BSC 2003 [DIRS 164822]). := 
.

. 
............. 
B PGV := PGV n 
.
.
................. 
0 . . 
0.159 
0.239 
0.398 
0.796 
1.59 
3.98 
5.57 
7.96 
11.9 
B 
:= 
6.26 10 4 . 
2.78 10 4 . 
9.30 10 5 . 
1.84 10 5 . 
3.07 10 6 . 
2.28 10 7 . 
8.15 10 8 . 
2.60 10 8 . 
6.56 10 9 . 
:= 
Peak Ground Velocity (m/s) 
PGVB 
PGV := t 
.
 
. 
MAF ( ) := log MAF n 
reverse( ) MAFn MAF := t 
reverse( ) PGV 
The interpolation between the PGV and mean annual 
exceedance frequency points is based on a linear 
interpolation using the value of log(MAF) at the individual 
sample points as stated in DTN: M0308SPACALSS.002 (BSC 
2003 [DIRS 164822]). 
February 2004 III-4 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Hazard Curve 
1 .10 4 
1 .10 5 
MAF
1 .10 6 
1 .10 7 
1 .10 8
0 2 4 6 8 10 
PGVB 
Horizontal PGV (m/s) 
DATi 1 , := PGV
i 0 , t DATi 0 , := MAFti 0 , 
.8.183 11.9 
.7.585 7.96 
.7.089 5.57 
.6.642 3.98 
DAT = .5.513 1.59 
.4.735 0.796 
.4.032 0.398 
.3.556 0.239 
.3.203 0.159 
DAT 1 . . 
PGV IE := v 
............. 
DAT 0 . . 
1 10 8 . 
.

. 
v := 
Seismic exceedance frequencies and time to first occurrance of seismic event follow 
uniform distributions (BSC 2003 [DIRS 164822]). 
1 10 4 . 
l IEu := . 
T := 10000 u 
IE := . 
T := 1 l 
ANL-EBS-NU-000008 REV 00 
Mean Annual Exceedance Frequency (1/yr) 
February 2004 
0 . . 
Set up mean annual exceedance frequency versus PGV for sampling (based on seismic event). 
The upper and lower bounds are from 
DTN: M0308SPACALSS.002 (BSC 2003 [DIRS 164822]). 
The lower bound is 1 year based on closure of repository and 
the upper bound is 10,000 years based on regulatory period 
(BSC 2003 [DIRS 164822]). 
III-5 of III-26 Screening Analysis for Criticality Features, Events, and Processes for License Application 
runif( ) n 
ns 
s 0 , 1 , runif( ) n 
ns 
s 0 , 1 , + 
RK10 0 < > 
1 - 
runif( ) n 
ns 
+
ns 
Latin Hypercube Sampling Routine for Evaluation of Seismic Events 
The PGV values, damaged areas, and seepage are obtained using Latin Hypercube Sampling. 
s
i 1 := n 
:= := rnd( ) 1.0 RDi 1 - , 6 rnd( ) 1.0 RDi 1 - , 4 
rnd( ) 1.0 RDi 1 - , 7 rnd( ) 1.0 RDi 1 - , 5 
rnd( ) 1.0 := := 
rnd( ) 1.0 := := rnd( ) 1.0 RDi 1 - , 8 
rnd( ) 1.0 rnd( ) 1.0 RDi 1 - , 9 
n := 20000 sample size 
RDi 1 - , 0 RDi 1 - , 2 
s .. 
i := 
RDi 1 - , 1 rnd( ) 1.0 RDi 1 - , 3 
:= 
:= 
RDi 1 - , 10 rnd( ) 1.0 RDi 1 - , 11 
:= 
RKs are matrixes in which the first column contain a permutation on the integers on the interval [1,n 
RK1:= csort := := := ( ) , ( ) , ( ) , ( ) RD 1 , RK2 csort RD 2 RK3 csort RD 3 RK4 csort RD 4 
RK5:= csort := := , ( ) RD 5 RK6 csort , ( ) RD 6 RK7 csort := ( ) , ( ) RD 7 , RK8 csort RD 8 
RK9:= csort := := := ( ) , ( ) , ( ) , ( ) RD 9 , RK10 csort RD 10 RK11 csort RD 11 RK12 csort RD 12 
s 
+ 
Define sets of random values. Each random value is selected within one of the equiprobable n 
intervals that partition [0,1], one set for each random variable. 
RK1 0 < > 
1 - + 
X 0 := 
RK4 0 < > 
1 - 
< > 
X 3 := 
RK7 0 < > 
1 - 
< > 
X 6 := 
< > 
< > 
X 9 := 
RK2 0 < > 
1 - + s 0 , 1 , 
X 1 := 
RK5 0 < > 
1 - 
< > 
X 4 := 
RK8 0 < > 
1 - < > 
< > 
X 7 := 
RK3 0 < > 
1 - + s 0 , 1 , s 0 , 1 , 
X 2 := 
RK6 0 < > 
1 - 
< > 
X 5 := 
RK9 0 < > 
1 - < > 
< > 
X 8 := 
qunif( ) Xi 0 , , IEl, IEu Sample mean annual seismic exceedance frequency 
Calculate a set of sample values for each of the random variables (i.e., seismic 
exceedance frequency and time to first seismic event. 
i 0 := ..n - 1 s 
IE := si 
si 
T := 
runif 
:= 
+ 
:= 
+ ( ) ( ) ( ) ns , 0, 1 
X 10 < > RK11 0 < > 
1 - runif ns , 0, 1 
X
< > 11 RK12
< > 0 - 1 runif ns , 0, 1 
February 2004 
:= 
rnd( ) 1.0 RDi 1 - , 12 
+ 
:= := 
+ 
qunif( ) Xi 9 , ,Tl,Tu Sample time (when first seismic event occurred) 
ANL-EBS-NU-000008 REV 00 III-6 of III-26 
runif( ) n 
ns 
s 0 , 1 , 
+ 
runif( ) n 
ns 
s 0 , 1 , runif( ) n 
ns 
ns 
s]. 
runif( ) n 
ns 
s 0 , 1 , 
+ 
runif( ) n 
ns 
s 0 , 1 , runif( ) n 
ns 
ns
Screening Analysis for Criticality Features, Events, and Processes for License Application 
Calculate peak ground velocity based on seismic mean annual frequency. 
= 
:= 
linterp , ( ) CDF DS , 0.05 = 0 
ds 
PGV:= linterp( ) IE , , log IEs ( ) 
Calculate drip shield upper bound damage state. 
DSli 
:= 
DSui 
DSds
:= 
CDF := i 
ds 
... 
qunif X DS1dsi 
:= 
DS2dsi 
:=
ns 
ns.1
0 
+ 
.. 
cumulative distribution function 
Drip Shield Damage Information 
Drip Shield Damage Due to Seismic and Rockfall (Nonlithophysal Only) 
if DS1ds i 
:= sort DS2ds 
i 1 
if DS1ndi 
Drip Shield Damage Due to Seismic Event Only 
i ui 100
.linterp PGVdsl ,DSld,PGV 
1 
100 
linterp PGVdsu DSud 
if DS1ui 
i 
... 
.. 
0.001 
:= 
:= 
:= ( ) DS1 
1 
0,   1 10 7 . ,DS1ui 
i 1 , li 
,DS ,DSui 
1 1  0 7 . , .. 
Mean damage area (fraction of drip shield damaged)  
5th percentile damage area (fraction of drip shield damaged) 
95th percentile damage area (fraction of drip shield damaged) 
.. 
0 , DS1ds i 
Sort the drip shield fraction of damage area. 
2.234 10 3 . 
Probability of no damage to the drip shield from rockfall 
 
0.601.PGV. 0.735 
1  1 , ,DS1ndi 
 
.. 
.. 
0.00204PGV3.7767 Mode percent failed area of drip shield due to rockfall 
1 10 3 . ,DS1drfmd 
linterp ,  , ( ) CDF DS 0.95 = 6.486 10. 3 
DS1nd 
DSndi 
:= 
DS1drfmd 
DSdrfmin 
DSdrfmd 
(Maximum value is 1.0) 
Min and max damaged percent of drip shield due to rockfall 
, 1.001 10 3 .   
i 
:=
i 
. 
DSdrfmax:= 100 
if DS1drfmd 
... 
Determine which equation to use: 
if Random Number < RNtest then use Equation 1 (DSeq1). 
if Random Number > RNtest then use Equation 2 (DSeq2). 
.. 
ANL-EBS-NU-000008 REV 00 
v PGVv 
, , . ( ) PGVi 
( ) 
DSds i 
 = ds ns 
ds 
:= 
i 
February 2004 III-7 of III-26 Screening Analysis for Criticality Features, Events, and Processes for License Application 
log DSdrfmd 
RNtest := 
i log DSdrfmax 
:= DSeq1i 
. DSeq2i 
:= 
. 
. 
2 
. 
..
. ( ) ( ) log( ) DS log DS log( ) DS 
+ drfmin 
log DSdrfmax 
drfmax 
. . i 2 , 
. i 2 , 
. 
DSeq1 
. drfmin drfmdi 
.. 
. 
. drfmax 
. 
log DSdrfmin 
.. 
.. 
( ) ( ) 1 X ( ). ( ) ( ) ( ) log DSdrfmax log DS log DS 
. log( ) DS X( ) log( ) DS log( ) DS log DS log( ) DS 
if Xi 2 , . RNtest i 
.. 
.. 
, , 
i 
DSeq2i
.. 
:= 
DS1drfi 
.. 
.. 
Fraction of drip shield damaged area due to rockfall 
:= DSdrfi 100
.DS%drfi 
1
.. 
Total fraction of drip shield damaged area due to seismic event and rockfall 
:= + DS1ds DSdrf DS1dnl 
1 . 1 , ,DS1dnli 
if DS1dnli 
DSdnli 
dnl 
:= 
DS := nl 
Percentage of drip shield damaged area due to rockfall 
DS1drfi 10 DS2drfi 
:= if Xi 3 , . DSndi 
:= 
DS%drfi 
, 
i 
. 
0 , DS2drf 
4.735 10 3 . 
sort( ) DS 
Drip Shield damage information 
ns.1 DSnli 
gnlds gnlds ns 0 i 
= nl 
= 
 
.= 
:= 
, , linterp( ) CDF DS 0.05 9.918 10. 9 
nl linterp ,  , ( ) CDF DS 0.95 = 1.06 10. 2 
ANL-EBS-NU-000008 REV 00 
. log DSdrfmin i 
.. 
.
. drfmin .. 
.. 
. 
Mean damage area (fraction of drip shield damaged) 
5th percentile damage area (fraction of drip shield damaged) 
95th percentile damage area (fraction of drip shield damaged) 
February 2004 III-8 of III-26 
drfmin 
( ) 
drfmdi 
.. 
. 
drfmin . 
. drfmdi 
.. 
. 
.
Screening Analysis for Criticality Features, Events, and Processes for License Application 
Rockfall Conversion 
Fraction of drip shields damaged due to rockfall 
n := 17110 rf 
s rf 
frac := rf 
grfds Mean damage area (fraction of drip shield damaged) 
5th percentile damage area (fraction of drip shield damaged) 
95th percentile damage area (fraction of drip shield damaged) ( )  , 0.95 = 1.82 10. 3 
Calculate waste package upper bound percent damaged area. WP := 0 l 
, := WPu WP1ui i 
Sample waste package percent damage between 
lower (0) and upper (WPu). WPu is set to 1x10 -7 if 
calculated value is less than zero in order for the 
i 
WP1di 
. 
( ) n . 1 . n 
ns 
Fraction of drip shields damaged assuming each realization constitutes a rockfall frac = 0.144 rf 
Fraction of Damaged Area to Drip Shield From Rockfall Only 
drf DS := sort DS rf 
i 
:= 
:= 
. 
. 
0 , ,WP1di 
WP2di 
uniform distribution to be sampled. Then all damaged 
areas less than 1x10-7 are set back to zero. The 1x10 -7 
value is determined as being less than the smallest 
sampled value from WP1u. 
Sorted fraction of waste package damaged area. . Mean damage area (fraction of waste package damaged) 2.577 10 4 . gwp 
i 
( ) 
Drip shield fraction of damaged area information from rockfall only 
ns.1 DSdrfi 
g  rfds = 2.501 10. 3 
ns 0 
rf, 
linterp CDF,DSrf 
Waste Package Damage Due to Seismic Event 
WP1 . := 0.436PGV. 0.305 u 
0,  . 1 10 7 . if WP1ui 
, qunif Xi 4 , WP ,WPu l .. 
1 1  0 7 . . if WP1di 
... 
... 
. WP3 := sort d 100 
WP2d 
1 ... 
Waste Package Damage Information 
ns 1 . WP3di 
gwp = 
ns 0 
, linterp( ) CDF DS 0.05 = 0 
.= 
:=
.= 
:= 
linterp( ) CDF WP3 0.05 = 0 5th percentile damage area (fraction of waste package damaged) d, , 
linterp  , ( ) CDF WP3 , 0.95 = 1.469 10. 3 95th percentile damage area (fraction of waste package damaged) d 
February 2004 III-9 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seepage distributions are based on the License Application (LA) seepage abstraction model (BSC 
[DIRS 165564]). The outputs were fit to Weibull distributions based on the sampling process developed 
to obtain the seepage rates for the low-, mean-, and upper-glacial transition climate case. The 
distributions are for both the Tptpll and Tptpmn zones (see Attachment IV). The Tptpll zone uses the 
collapsed drift seepage rates, while the Tptpmn zone increases the nominal seepage rates by 20 percent. 
Tptpll Seepage (Drift Collapse) 
0.52 r 
STlptl := . r ( ) 1 exp 
8.869 10 3 . 
0.468 
.

.
 
.... 
.... 
0.494 

r 
1.46 10 1 . 
r 
3.83 10 1 . 
... 
. 
.... 
STmptl := . r ( ) 1 exp 
STu := . ptl(r) 1 exp 
.... 
.
 
cr 
3.027 ]cr := 
ni 
2.06 106 
n21.bwf 
cr := 
AW := 10.811 
n21 := 
bwf := 
Bi21 := 
AW 
... 
. 
.... 
... 
. 
.... 
Boron loss from waste packages based on Boron Loss from CSNF Waste Packages (BSC 2003 
[DIRS 165890], Section 6). 
Input parameters for 21-PWR Absorber Plate waste package are from different sources as 
listed in Section 5.1 and 6.4. 
Corrosion rate of Neutronit is based on 1.5 times the corrosion rate for stainless steel 316 data 
(see Assumption 5.1.4). The corrosion rate was fit to a Weibull distribution for use in this 
calculation (see Attachment V). The corrosion rate data for stainless steel 316 can be found in 
DTN: MO0303SPAMCRAQ.000 [aqueous-316L.xls] (BSC 2003 [DIRS 162353]). 
\l := 0.314 Alpha parameter of Weibull distribution 
Beta parameter of Weibull distribution 
\lcr.qweibull( ) Xi 7 , , ]cr Corrosion rate data fit to a Weibull distribution ( gm/yr) 
Atomic Weight of boron (g/mole) (see Table 4.1-12) 
Grams of Neutronit in a 21-PWR Absorber Plate (g) (see Table 6.4-3) 
Weight fraction of boron (see Table 6.4-3) 1.245 10 2 . 
Initial bor on in a 21-PWR Absorber Plate waste package (moles) (see Table 6.4-3) 
Surface area of Neutronit in a 21-PWR Absorber Plate (cm 2) (see Table 5.1-1 
[Assumption 5.1.2]) 
Volume of a 21-PWR Absorber Plate (liter) (see Table 5.1-1 [Assumption 5.1.2]) 
... 
. 
.... 
... 
. 
.... 
:= 4.69 103 Vr21 
5.29 105 SA21 := 
ANL-EBS-NU-000008 REV 00 
Tptpmn Seepage (Degraded Drift) 
0.536 r 
STlptn . := r ( ) 1 exp 
4.949 10 3 . 
0.473 r 
... 
. 
.... 
STmptn r ( ) 1 exp 
8.559 10 2 . 
.... 
.... 0.501 r 
. 
. STuptn(r) 1 exp 
:= 
:= 
.

.
 
2.251 10 1 . 
.... .
 
February 2004 III-10 of III-26 Screening Analysis for Criticality Features, Events, and Processes for License Application 
cm/gm conversion factor con := 1  10 4 .
Density of stainless steel 316 g/cm3 (see Table 4.1-12) 8 l 316:= 
con cr . n k21i i 
l 316. Degradation rate of Neutronit in moles for a 21-PWR Absorber Plate (g/cm 2*yr) 
(see Table 6.4-3) 
s 700, 700 Tsi 
, 
i 
T1si 
Assume no water can penetrate the waste package prior to 
700 years due to evaporation from the decay heat generation 
(see Assumption 5.2.2). 
tdeg + T1si Time required to degrade all of the Neutronit from a 
21-PWR Absorber Plate (years) (see Table 6.4-3) 
.... 
Bi21 := D21 Moles of boron released from Neutronit per year (moles/yr) (see Table 6.4-3) 
:= 
:= if T 
21 := 
i 
i 
t1i 
:= t tdeg t1i , 
.

, 21i 
:= 
21i 21i 
. 
<
n21 
k21i
.SA21 
tdeg21i 
Develop the time available for boron to degrade and flush from waste package. This time is 
based on the occurrance of the first seismic event up to the regulatory period. If the time to 
degrade the Neutronit is longer than the delta time (regulatory period minus time of seismic 
event), then it is set to the regulatory period and delta time becomes zero. 
10000 T1si 
if t1 < tdeg i 
fraction of boron remaining in a PWR waste package (10 percent) (see Assumption 5.1.5) 
237.231 B21loss 
.. 
.. 
per := 0.1 B 
:= B21loss 
if tdeg 
.
 
. := A t i 10000, 0, if tdeg21i 21i 
0 , 1 , 
. := 
Parameters required to solve the boron loss 
equation when the time to degrade is longer 
than the regulatory period of 10,000 years t Ci 
.. 21i 
1 , 0 , 
.... 
..
if tdeg 
if tdeg , , . := 10000, 1 t Ei 
. 
perB.Bi21 
21i 
21i 
21i 21i 21i 
. 
i 0 , 
. 
..
10000, 1 if tdeg21i 
10000, 10000 if tdeg 
Boron Loss Equation (BSC 2003 [DIRS 165890], Section 6) 
. 
.. 
.. 
.v21 t21i 0 , 
tdeg21i 0 , v21 .. 
= 
Vr . 21 D21 
v21 
,
A 
. 
. + := . . . . exp 1 exp tdeg i i . B21loss Ei . 21i 0 , 
.. .... 
. 
... 
. 
Vr21 
... 
.. 
..
Vr21 
.... 
.... 
.
 
.
 
... 
.. 
.... 
.... 
.. 
. ... 
,  root NB ( 1 10 .
11 1 1011 
21 , v21 i , ) v21,  .. 
..
.. 
NB ( 21 v21 i , ) v21 C 
V21dri 
:= 
V21dr 
req21dr := 
1000 
. 
Required drip rate (m 3/yr) into a 21-PWR Absorber Plate waste package to 
degrade and flush out the boron based on sampled corrosion rate and 
time to first seismic event 
February 2004 III-11 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The following is the equation used to calculate the probability that the seepage rate can be at or 
greater than the seepage noted as z 21 (for lithophysal) and z1 21 (for nonlithophysal). This seepage 
is based on the seepage rate making it through the damaged areas of the drip shield and waste 
package. The drip rate value is based on sampling the corrosion rate of Neutronit and sampling 
the time to first seismic occurrance, which determines the time to degrade and flush out the 
Neutronit from a 21-PWR Absorber Plate Waste Package for criticality potential. 
req21dri 
z21 := 
i 
100
.WP1di 
.DS1ds i 
1 
req21dri 
100
.WP1di 
DS 
1 
dnli 
... 
z121i 
:= 
Lithophysal Zone 
Tptpll Seepage (Collapsed Drift) 
r 
8.869~ 10 3 . 
r 
r 
... 
..  
STlptl := . r ( ) 1 exp 
:= . 
... 
1.46~ 10 1 . 
3.83~ 10 1 . 
... 
. 
...  
STmptl := . r ( ) 1 exp 
STuptl(r) 1 exp 
. 
. 
...  
.. 
... 
. 
...  
Q21lspli 
.. 
:= 
.. 
... 
1 STlptl z21 
1 . STmptl z21 
.
i 
i 
. := 
:= 
Q21mspli 
Q21uspli 
.. 
.. 
.. 
.. 
.. 
1 STuptl z21i 
The probability of being in the low-, mean- or upper-infiltration scenario for the glacial transition 
climate is 0.24, 0.41, and 0.35, respectively (BSC 2003 [DIRS 165991], Section 7, Table 7-1). 
, := . .0.194 if 0.24 Xi 5 , if Xi 5 , Q21sfl 0.24 Q21lspli i 
:= . . , 0.154 if 0.24 Xi 5 , if Xi 5 , 0.24 Q21lspn Q21sfn i i 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
z21 is amount of seepage rate (m 3/yr) required to 
reach the drift for lithophysal zone 
21 is amount of seepage rate (m 3/yr) required to z1 
reach the drift for nonlithophysal zone 
Nonlithophysal Zone 
Tptpmn Seepage (Degraded Drift) 
0.536 0.52 r 
STlptn . := r ( ) 1 exp 
4.949 10 3 . 
... 
0.473 0.468 r := 
. 
...  
STmptn r ( ) 1 exp 
.... 
.... 
.... 
.... 
.
.
.
.
0.494 
. := 
. 
STuptn(r) 1 exp 
.
.
.
.
0.501 
2.251~ 10 1 . 
.
. 
.
. 
.... 
~ 
8.559~ 10 2 . 
r 
i 
... 
. 
...  
... 
. 
...  
1 . STlptn z121 Q21lspn i 
. := Q21mspn i 
:= Q21uspn 
:= 
i 
i 
.. 
.. 
.. 
.. 
. . , . , , 0.637 0.65 Q21mspli 
.... 
.. 
.. 
1 STmptn z121 
1 . STuptn z121i 
. 
.. 
.. 
0.514 Q21uspli 
. . . , , , 0.672 0.65 Q21mspn i 
.. 
.. 
0.524 Q21uspn i 
.. 
.. 
February 2004 III-12 of III-26 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The probability of the seimic event causing sufficient damage to allow advective flow to penetrate, 
degrade, and flush out the Neutronit from a 21-PWR Absorber Plate waste package is calculated 
using the equations below for both lithophysal and nonlithophysal. The equations are based on 
the sampling of the probability of sufficient advective flow given a seismic event with magnitude, 
v, at time, t. This probability will be fed into top event MS-IC-1 of the SAPHIRE model (Section 
6.4.2) for seismic event. The probability of a seismic event will be set to 1.0 in the SAPHIRE model 
since it is accounted for in this calculation. 
Lithophysal Zone 
ns.1 
u l u := . Pr21sfl i 7 , Q21sfli 
Pr21sfl 
:= ( ) 1 X . 
( ) 1 X . 
Equations based on Equation 6.4-6 of 
main report, DTN:MO0308SPACALSS.002 
(BSC 2003 [DIRS 164822]) and Seismic 
Consequence Abstraction (BSC 2003 [DIRS 
161812], Attachment VIII, Eq. VIII.2.11). 
. i 7 , Q21sfni 
0 = 
Pr21sfnl 
Pr21sfnl 
n12 := 
Bi12 := 
Vr12 
:= 
:= 
SA12 
k12 
l . 
Nonlithophysal Zone 
u 
Boron loss from waste packages based on Boron Loss from CSNF Waste Packages (BSC 2003 
[DIRS 165890], Section 6). 
Input parameters for 12-PWR Absorber Plate Long waste package are from different 
sources as listed in Section 5.1 and 6.4. 
1.15 106 
n12.bwf 
3.32 103 
3.19 105 
i 
 316.con 
i 
l . 
700, 700 Tsi 
, 
:= 
T1 := si 
.. 
n12 + T1s tdeg12 := 
i i 
. ( ) T T ( ) IE . IE 
ns 0 i 
Mean probability 
ns.1 
. l u ..
.. 
0 i 
AW 
cr . n 
if T
0 = 
s 
.. 
.... 
..
= 
. 
= 
. 
Mean probability 
Grams of Neutronit in a 12-PWR Absorber Plate Long (g) (see Table 6.4-4) 
Initial boron in a 12-PWR Absorber Plate Long (moles) (see Table 6.4-4) 
Volume of a 12-PWR Absorber Plate Long (liter) (see Table 5.1-1 [Assumption 5.1.2]) 
Surface area of Neutronit in a 12-PWR Absorber Plate Long (cm2) 
(see Table 5.1-1 [Assumption 5.1.2]) 
Degradation rate of Neutronit in moles for a 12-PWR 
Absorber Plate Long (g/cm2*yr) (see Table 6.4-4) 
Assume no water can penetrate the waste package prior to 
700 years due to evaporation from the decay heat generation 
(see Assumption 5.2.2). 
Time required to degrade all of the Neutronit from a 
12-PWR Absorber Plate Long (year) (see Table 6.4-4) 
. ( ) T T ( ) IE IE 
ns 
< 
i 
.. 
.
. 
k12i
.SA12 
February 2004 III-13 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Bi12 
D12i Moles of boron release from Neutronit per year (moles/yr) 
(see Table 6.4-4) 
tdeg12i 
i 10000. T1si 
:= tdeg t1i , if t1 < tdeg i 12i 12i 
:= 132.435 
. 
B12loss 
:= 
Develop the time available for boron to degrade and flush from waste package. This time is 
based on the occurrance of the first seismic event up to the regulatory period. If the time to 
degrade the Neutronit is longer than the delta time (regulatory period minus time of seismic 
event), then it is set to the regulatory period and delta time becomes zero. 
Fraction of boron remaining in a PWR waste package (10 percent) (see Assumption 5.1.5) 
B12loss 
if tdeg 
.
 
perB.Bi12 
. 
, 
10000, 0 if tdeg := t 12i 12i 
0 , 1 , 
, . := 
Parameters required to solve the boron loss 
equation when the time to degrade is longer 
than the regulatory period of 10,000 years. t 12i 12i 
1 , 0 , 
, . := t 
t1 := 
t12i 
Ai 
Ci 
Ei 
, 
10000, 1 if tdeg12i 
10000, 10000 if tdeg 12i 12i 12i 
. 
. 
.. 
Vr . 12 D12 .v12 t12i 0 , 
tdeg12i 0 , i 0 , v12 ..
:= . . . . + exp B12loss Ei . 
= 
12i
v12 
..
.. 
Ai. 
.
. 
.
.
10000, 1 , 
1 exp .tdeg12i 0 , 
. 
... 
. 
Vr12 
.. 
..
Vr12 
.... 
.
 
.
 
... 
.. 
.... 
.. 
... 
.... 
.... 
.... 
12 , v12 i , ) v12,  root NB ( 1 10 .
11 1 1011 
. ... 
,  
.... 
..
if tdeg 
if tdeg 
Boron Loss Equation (BSC 2003 [DIRS 165890] Section 6) 
NB ( 12 . v12 i , ) v12 Ci 
:= 
req12dr := 
The following is the equation used to calculate the probability that the seepage rate can be at or 
greater than the seepage noted as z 12 (for lithophysal) and z1 12 (for nonlithophysal). This seepage 
is based on the seepage rate making it through the damaged areas of the drip shield and waste 
package. The drip rate value is based on sampling the corrosion rate of Neutronit and sampling 
the time to first seismic occurrance, which determines the time to degrade and flush out the 
Neutronit from a 12-PWR Absorber Plate Long waste package for criticality potential. 
.. 
... 
..
.. 
V12dri 
z12i 
:= 
.. 
V12dr 
1000 
req12dri 
. 
Required drip rate (m 3/yr) into a 12-PWR Absorber Plate Long waste 
package to degrade and flush out the boron based on sampled corrosion 
and time to first seismic event. 
z12 is amount of seepage rate (m 3/yr) required to 
reach the drift for lithophysal zone .DS1ds i 100
.WP1di 
1 . 
February 2004 III-14 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
req12dri 
z112i 
z112 is amount of seepage rate (m 3/yr) required to 
reach the drift for nonlithophysal zone 
100
.WP1di 
.DSdnli 
1 ... 
Nonlithophysal Zone 
. 1 STlptn z112i i 
Q12lspn i 
.. 
.. 
:= 1 . STmptl z12 i i 
... 
... 
Lithophysal Zone 
1 . STlptl z12 
:= 
:= 
Q12mspli 
Q12uspli 
.. 
.. 
.. 
.. 
.. 
1 STuptl z12i 
The probability of being in the low-, mean- or upper-infiltration scenario for the glacial transition 
climate is 0.24, 0.41, and 0.35, respectively (BSC 2003 [DIRS 165991], Section 7, Table 7-1). 
,   , , , 0.637 0.194 if 0.24 Xi 5 , 
.. 
.. 
0.24 Q12lspli 
:= 
0.514 Q12uspli 
:= 
Q12sfli 
:= 0.65 Q12mspli 
:=   , , 
:= 
Q12lspli 
i 
:= 
Q12mspn i 
Q12uspn i
i 
.. 
1 STmptn z112 
1 . STuptn z112i 
. 
0.524 Q12uspn i 
.. 
.. 
0.672 
.
, 
.. 
.. 
. 
 
if Xi 5 , 
. 
0.65 Q12mspn 
. 
0.24 Q12lspn 
.. 
0.154 if 0.24 Xi 5 , Q12sfn i 
.. .. 
.. 
The probability of the seimic event causing sufficient damage to allow advective flow to penetrate, 
degrade, and flush out the Neutronit from a 12-PWR Absorber Plate Long waste package is 
calculated using the equations below for both lithophysal and nonlithophysal. The equations are 
based on the sampling of the probability of sufficient advective flow given a seismic event with 
magnitude, v, at time, t. This probability will be fed into top event MS-IC-1 of the SAPHIRE model 
(Section 6.4.2) for seismic event. The probability of a seismic event will be set to 1.0 in the 
SAPHIRE model since it is accounted for in this calculation. 
.. 
.. 
Lithophysal Zone 
. 
if Xi 5 , 
u 
. 
ns.1 
 
i 7 , 
, 
u l ..
. := 
.. 
Q12sfl Pr12sfl i 
Pr12sfl 
l .
u l . 
:= ( ) 1 X . 
. ( ) T T ( ) IE . IE 
ns 
( ) T T .( ) IE . IE 
ns 0 i 
Mean probability 
ns.1 
l u .. 
0 i 
. ( ) 1 X 
Equations based on Equation 6.4-6 of 
main report, DTN:MO0308SPACALSS.002 
(BSC 2003 [DIRS 164822]) and Seismic 
Consequence Abstraction (BSC 2003 [DIRS 
161812], Attachment VIII, Eq. VIII.2.11). 
. i 7 , Q12sfni 
..
= 
. 
..
= 
. 
Mean probability 0 = 
0 = 
Nonlithophysal Zone 
Pr12sfnl 
Pr12sfnl 
February 2004 III-15 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Boron loss from waste packages based on Boron Loss from CSNF Waste Packages (BSC 2003 
[DIRS 165890], Section 6). 
Input parameters for 44-BWR Absorber Plate waste package are from different sources as 
listed in Section 5.1 and 6.4 . 
Grams of Neutronit in a 44-BWR Absorber Plate (g) (see Table 6.4-5) 2.15 106 n44 := 
n44.bwf Initial boron in a 44-BWR Absorber Plate (moles) (see Table 6.4-5) Bi44 := 
AW 
Volume of a 44-BWR Absorber Plate (liter) (see Table 5.1-1 [Assumption 5.1.2]) 4.85 103 Vr44 
9.55 105 
:= 
SA44 := Surface area of Neutronit in a 44-BWR Absorber Plate (cm2) (see Table 5.1-1 
[Assumption 5.1.2]) 
con cr . n k44i i 
l 316. Degradation rate of Neutronit in moles for a 44-BWR Absorber Plate (g/cm2*yr) 
(see Table 6.4-5) 
if Tsi 
700, 700 Tsi 
, Assume no water can penetrate the waste package prior to 
700 years due to evaporation from the decay heat generation 
(see Assumption 5.2.2). 
:= tdeg + T1si Time required to degrade all of the Neutronit from a 
44-BWR Absorber Plate (years) (see Table 6.4-5) 
.... 
Bi44 
Moles of boron release from Neutronit per year (moles/yr) (see 
Table 6.4-5) 
i 
.

tdeg , t t1i , 44i 44i 
. 
44i 
per 0.5 
:= 
BB Fraction of boron remaining in a BWR waste package (50 percent) 
(see Assumption 5.1.5) 
:= 1.238 103 B44loss 
:= 
T1 := si
44i 
D44i 
:= 
B44loss 
if tdeg 
.
 
perBB.Bi44 
. , := t 10000, 0 if tdeg Ai 
= 
44i 
, . := 
Parameters required to solve the boron loss 
equation when the time to degrade is longer 
than the regulatory period of 10,000 years t Ci 
 
44i 
0 , 1 , 
44i 
1 , 0 , .. 
.
. 
.
.
, 10000, 1 , . := t Ei 
<
n44 
k 
i
.SA44 44 
tdeg44i 
Develop the time available for boron to degrade and flush from waste package. This time is 
based on the occurrance of the first seismic event up to the regulatory period. If the time to 
degrade the Neutronit is longer than the delta time (regulatory period minus time of seismic 
event), then it is set to the regulatory period and delta time becomes zero. 
t1 . := 10000 T1si 
if t1 < tdeg i 
44i 
44i 
44i 
.. 
..
:=
.... 
..
if tdeg 
if tdeg 
. 
..
10000, 1 if tdeg44i 
10000, 10000 if tdeg 
..
.. 
44i 44i 
.. 
. 
. 
February 2004 III-16 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Boron Loss Equation (BSC 2003 [DIRS 165890], Section 6) 
Vr . 44 D44 .v44. t44i 0 , 
tdeg44i 0 , i 0 , v44 .
. . . + exp 1 exp tdeg A . i NB ( 44 . v44 i , ) v44 Ci B44loss Ei . 44i 0 , v44 Vr44 
.. 
... 
.... 
.
 
.... 
.... 
.... 
,  
:= 
root NB ( 1 10 .
11 1 1011 
44 , v44 i , ) v44,  .
V44dri 
:= 
req44dr := 
. 
Vr44 
Required drip rate (m 3/yr) into a 44-BWR Absorber Plate waste package to 
degrade and flush out the boron based on sampled corrosion and time to 
first seismic event 
The following is the equation used to calculate the probability that the seepage rate can be at 
or greater than the seepage noted as z 44 (for lithophysal) and z1 44 (for nonlithophysal). This 
seepage is based on the seepage rate making it through the damaged areas of the drip shield 
and waste package. The drip rate value is based on sampling the corrosion rate of Neutronit 
and sampling the time to first seismic occurrance, which determines the time to degrade and 
flush out the Neutronit from a 44-BWR Absorber Plate waste package for criticality potential. 
z44i 
:= z44 is amount of seepage rate (m 3/yr) required to 
reach the drift for lithophysal zone 
z144i 
z144 is amount of seepage rate (m 3/yr) required to 
reach the drift for nonlithophysal zone ... 
.... 
Nonlithophysal Zone 
... 
Lithophysal Zone 
.... 
.DS1ds i 
req44dri
DSdnli 
.
i 
.

. 1 STlptn z144 . i i 
.. 
V44dr 
1000 
req44dri 
100
.WP1di 
1
100
.WP1di 
1 
1 STlptl z44 
.. 
. 
. := 1 STmptn z144i 
i 
:= 
:=
... 
... 
:= 
:= 
Q44mspli 
Q44uspli 
. 
.. 
1 STuptl z44i 
:= 
i 
i .. 
.. 
.. 
1 . STuptn z144i 
The probability of being in the low-, mean- or upper-infiltration scenario for the glacial transition 
climate is 0.24, 0.41, and 0.35, respectively (BSC 2003 [DIRS 165991], Section 7, Table 7-1). 
. 
Q44lspn 
Q44mspn 
Q44uspn 
0.65 Q44mspl 
. 
0.194 if 0.24 Xi 5 , 
.. 
.. 
.. 
... 
.
. 
. ... 
. 
. 
.. 
1 . STmptl z44 
. 
. 
. 
0.637 := , . , , . 
:= 
Q44lspli 
Q44sfli 
. 
. 
0.24 Q44lspl 
. 
if Xi 5 , 
.. 
. 0.514 Q44uspli i i 
. 
.. 
. 
,
, := . . . , , , . 0.672 0.154 if 0.24 Xi 5 , if Xi 5 , 0.65 Q44mspn 0.24 Q44lspn Q44sfn i i i 
. 
. 
.. 
.. 
0.524 Q44uspn i 
. 
February 2004 III-17 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The probability of the seimic event causing sufficient damage to allow advective flow to penetrate, 
degrade, and flush out the Neutronit from a 44-BWR Absorber Plate waste package is calculated 
using the equations below for both lithophysal and nonlithophysal. The equations are based on 
the sampling of the probability of sufficient advective flow given a seismic event with magnitude, 
v, at time, t. This probability will be fed into top event MS-IC-1 of the SAPHIRE model (Section 
6.4.2) for seismic event. The probability of a seismic event will be set to 1.0 in the SAPHIRE model 
since it is accounted for in this calculation. 
Lithophysal Zone 
ns.1 
T . Tu u IEl := . Pr44sfl i 7 , Q44sfli 
Pr44sfl ~ 
Tu 
Pr44sfnl := 
( ) 1 X . 
Equations based on Equation 6.4-6 of 
main report, DTN:MO0308SPACALSS.002 
(BSC 2003 [DIRS 164822]) and Seismic 
Consequence Abstraction (BSC 2003 [DIRS 
161812], Attachment VIII, Eq. VIII.2.11). 
. i 7 , Q44sfni 
= 
= 
Pr44sfnl 
n24 := 
Bi24 := 
Vr24 
l . 
1.525 10 6 . 
Nonlithophysal Zone 
T . 
9.74 10 7 . 
Boron loss from waste packages based on Boron Loss from CSNF Waste Packages (BSC 2003 
[DIRS 165890], Section 6). 
Input parameters for 24-BWR Absorber Plate waste package are from different sources as 
listed in Section 5.1 and 6.4 . 
2.63 106 
n24.bwf 
2.64~ 103 
6.93~ 105 
:= 
SA24 := 
:= k24i 
l . 
i 
 316.con 
700, 700 Tsi 
, T1si 
tdeg 
. ( )( ) IE 
ns 0 i 
Mean probability 
ns.1 
u IEl 
0 i 
AW 
cr . n 
if T 
.  
.. :=
.  
.... 
..
= 
. 
..
= 
. 
Mean probability 
Grams of Neutronit in a 24-BWR Absorber Plate (g) (see Table 6.4-6) 
Initial boron in a 24-BWR Absorber Plate (moles) (see Table 6.4-6) 
Volume of a 24-BWR Absorber Plate (liter) (see Table 5.1-1 [Assumption 5.1.2]) 
Surface area of Neutronit in a 2 4-BWR Absorber Plate (cm2) (see Table 5.1-1 
[Assumption 5.1.2]) 
Degradation rate of Neutronit in moles for a 24-BWR 
Absorber Plate (g/cm2*yr) (see Table 6.4-6) 
Assume no water can penetrate the waste package prior to 
700 years due to evaporation from the decay heat generation 
(see Assumption 5.2.2) 
i 
:= 
24i 
( ) 1 X . 
Time required to degrade all of the Neutronit from a 
24-BWR Absorber Plate (years) (see Table 6.4-6) 
<
n24 
k24i
.SA24 
Bi24 := D24i 
. ( )( ) IE 
ns 
~ 
~
si 
.. 
T1s 
.
. 
+
Moles of boron release from Neutronit per year (moles/yr) (see 
Table 6.4-6) tdeg24i 
.
. 
.... 
February 2004 III-18 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Develop the time available for boron to degrade and flush from waste package. This time is 
based on the occurrance of the first seismic event up to the regulatory period. If the time to 
degrade the Neutronit is longer than the delta time (regulatory period minus time of seismic 
event), then it is set to the regulatory period and delta time becomes zero. 
t1i 10000. T1si 
t tdeg t1i , if t1 < tdeg i 
:= 
24i 24i 24i 
:= 1.514 103 
:= 
B24loss B24loss 
, i i 
:= 
Parameters required to solve the boron loss 
equation when the time to degrade is longer 
than the regulatory period of 10,000 years 
 
t24i 
0 , 1 , 
t24i 
1 , 0 , 
i 
..
..
,  := 10000, 1 t 
A := 
Ci 
Ei 24i 24i 24i 
.. 
.. 
. Vr . 24 D24i 0 , 
. tdeg24i 0 , v24 .. 
..
:= . . . . . . exp 1 exp tdeg B24loss Ei . 
=
v24 
, 
Ai 
v24. t24i 0 ,
Vr 24i 0 , 
.. 
... 
... 
.. 
...
24 Vr24 
.... 
.
. 
.
. 
... .
.. 
.... 
.... 
, 
Fraction of boron remaining in a BWR waste package (50 percent) (see Assumption 5.1.5) 
perBB.Bi24 
if tdeg24  10000, 0 if tdeg24i 
root NB ( 1 10 .
11 1 1011 
24 , v24 i , ) v24,  
.. 
. ... 
.. 
..
.. 
,  
Required drip rate (m 3/yr) into a 24-BWR Absorber Plate waste package to 
degrade and flush out the boron based on sampled corrosion and time to 
first seismic event. 
.. 
..
if tdeg24  10000, 1 if tdeg24i 
10000, 10000, if tdeg if tdeg 
Boron Loss Equation (BSC 2003 [DIRS 165890], Section 6) 
The following is the equation used to calculate the probability that the seepage rate can be at or 
greater than the seepage noted as z 24 (for lithophysal) and z1 24 (for nonlithophysal). This seepage 
is based on the seepage rate making it through the damaged areas of the drip shield and waste 
package. The drip rate value is based on sampling the corrosion rate of Neutronit and sampling 
the time to first seismic occurrance, which determines the time to degrade and flush out the 
Neutronit from a 24-BWR Absorber Plate waste package for criticality potential. 
..
.. 
NB ( 24 . v24 i , ) v24 Ci 
V24dri 
:= 
req24dr := 
z24i 
:= 
+ 
req24dri
. 
i 
.... 
.. 
.... 
DS1ds 
.. 
V24dr 
1000 
100
.WP1di 
1 ... 
req24dri 
:= z124i 
.. 
... 
.. 
z24 is amount of seepage rate (m 3/yr) required to 
reach the drift for lithophysal zone 
z124 is amount of seepage rate (m 3/yr) required to 
reach the drift for nonlithophysal zone .DSdnli 100
.WP1di 
1 .. 
... 
... 
February 2004 III-19 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Lithophysal Zone 
1 . STlptl z24i 
Q24lspli 
. 
:= 
i 
:= 
Q24mspli 
Q24uspli 
.. 
1 . STmptl z24 
1 STuptl z24i 
. 
. 
, 
.. 
.. 
0.24 Q24lspl 
:= 
Q24sfl := 
i i 
:= . Q24sfn 
The probability of being in the low-, mean- or upper-infiltration scenario for the glacial transition 
climate is 0.24, 0.41, and 0.35, respectively (BSC 2003 [DIRS 165991], Section 7, Table 7-1). 
.. 
1 STmptn z124 
1 . STuptn z124i 
. 
0.524 Q24uspn i 
. 
if Xi 5 , 
. 
0.24 Q24lspn i 
.. 
.. 
The probability of the seimic event causing sufficient damage to allow advective flow to penetrate, 
degrade, and flush out the Neutronit from a 24-BWR Absorber Plate waste package is calculated 
using the equations below for both lithophysal and nonlithophysal. The equations are based on 
the sampling of the probability of sufficient advective flow given a seismic event with magnitude, 
v, at time, t. This probability will be fed into top event MS-IC-1 of the SAPHIRE model (Section 
6.4.2) for seismic event. The probability of a seismic event will be set to 1.0 in the SAPHIRE model 
since it is accounted for in this calculation. 
Lithophysal Zone 
. 
if Xi 5 , 
u 
ns.1 
. , 
u l ..
:= Pr24sfl 
Pr24sfl 
l . 
u l . 
:= 
. ( ) T T ( ) IE . IE 
ns 
( ) T T .( ) IE . IE 
ns 
Mean probability 
l u .. 
0 i 
ns.1
0 i 
..
.= 
. 
...= 
. 
Mean probability 0 =
0 = 
Nonlithophysal Zone 
Pr24sfnl 
Pr24sfnl 
ANL-EBS-NU-000008 REV 00 
Nonlithophysal Zone 
. 1 STlptn z124 Q24lspn i i 
:= Q24mspn i 
Q24uspn 
. 
. 
. . , , , 0.637 0.194 if 0.24 Xi 5 , 0.65 Q24mspl 
:= 
0.514 Q24uspli i 
:= 
i 
i
i 
. 
. 
0.672 
.
, 
.. 
.. 
. . , , 
. 
0.65 Q24mspn 
.. 
0.154 if 0.24 Xi 5 , i 
. 
. 
. 
. 
. 
1 X . i 7 , 
.. 
Q24sfli 
( ) 1 X . 
( ) 
Equations based on Equation 6.4-6 of 
main report, DTN:MO0308SPACALSS.002 
(BSC 2003 [DIRS 164822]) and Seismic 
Consequence Abstraction (BSC 2003 [DIRS 
161812], Attachment VIII, Eq. VIII.2.11). 
. i 7 , Q24sfni 
February 2004 III-20 of III-26 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Probability Calculation using the following assumptions : 
 Corrosion rate of Neutronit assumed to be 1.5 times the corrosion rate of stainless steel 316, 
which was fit to a Weibull distribution (Assumption 5.1.4) 
Flushing out 90 percent of the boron from the PWR waste package types (Assumption 5.1.5) 
Flushing out 50 percent of the boron from the BWR waste package types (Assumption 5.1.5) 
No water ingress into a waste package prior to 700 years (Assumption 5.2.2) 
The following are the final calculated probabilities of having a seepage rate equal to 
or greater than z # (required) along with a corrosion rate of Neutronit equal to or 
greater than cr n given a seismic event for the different waste package types in the 
different repository zones. 
Lithophysal Zone
21-PWR Absorber Plate waste package 
(see p. III-13) 0 = Pr21sfl 
(see p. III-15) Pr12sfl 0 = 
1.525 10 6 - = Pr44sfl 
(see p. III-20) 0 = 
12-PWR Absorber Plate Long waste package 
0 = 
Pr12sfnl 
Pr44sfnl 
Pr24sfnl Pr24sfl 


 
ANL-EBS-NU-000008 REV 00 
Final Results of Seismic Base Case 
Nonlithophysal Zone 
(see p. III-13) Pr21sfnl 
(see p. III-15) 
44-BWR Absorber Plate waste package 
 (see p. III-18) (see p. III-18) 
24-BWR Absorber Plate waste package 
0 =
= 9.74 10 7 - 
0 = (see p. III-20) 
February 2004 III-21 of III-26 Screening Analysis for Criticality Features, Events, and Processes for License Application 
III.2 SEISMIC SENSITIVITY CASE ANALYSIS 
The seismic sensitivity case analysis is identical to the seismic base case analysis of Section III.1 
with the exception that it is assumed that the neutron absorber plate thickness of the 44-BWR 
Waste Package is increased from 5 mm to 7 mm. A 7 mm plate thickness is identical to the 
design of the 21-PWR with Absorber Plates and the 12-PWR Long waste packages types, but is 
less than the absorber plate thickness of the 24-BWR Waste Package design. The information 
contained in this section was obtained from the Attachment III.mcd Mathcad file of 
Attachment VII. 
III-22 of III-26 ANL-EBS-NU-000008 REV 00 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seismic Sensitivity Case Analysis 
(Increase Neutronit plate thickness of the 44-BWR Absorber Plate waste package) 
Boron loss from waste packages is based on Boron Loss from CSNF Waste Packages (BSC 2003 
[DIRS 165890], Section 6). 
Input parameters from DTN: MO0309SPABRNAM.001 [DIRS 165892] for 44-BWR Absorber Plate 
waste package, except the Neutronit plate thickness is increased to 7 mm for a sensitivity analysis 
(see Section 6.4 for discussion). 
Grams of Neutronit in a 44-BWR (g) 3.152 106 nt44 := 
nt44.bwf Initial boron in a 44-BWR Absorber Plate (moles) Bit44 := 
AW 
Volume of a 44-BWR Absorber Plate (liter) 4.85 103 Vr44 
Surface area of neutronit in a 44-BWR Absorber Plate (cm2) 9.837 105 
:= 
SAt44 := 
k44 cr . n i i 
l 316.con Degradation rate of Neutronit in moles for a 44-BWR 
Absorber Plate (g/cm2*yr) 
700 if T 700 si 
,T si 
Assume no water can penetrate the waste package prior to 
700 years due to evaporation from the decay heat generation 
(see Assumption 5.2.2) 
+ := 
. 
T1si 44 i Time required to degrade all of the Neutronit in a 
44-BWR Absorber Plate (years) 
Moles of boron release from Neutronit per year (moles/yr) 
:= 
T1 := si 
tdegt 44i 
Dt44i 
:= 
.... 
Bit44 
tdegt 44i 
i 
tt tdegt 
.. 
.... 
Develop the time available for boron to degrade and flush from waste package. This time is 
based on the occurrance of the first seismic event up to the regulatory period. If the time to 
degrade the Neutronit is longer than the delta time (regulatory period minus time of seismic 
event), then it is set to the regulatory period and delta time becomes zero. 
t1 . := 10000 T1si 
if t1 < tdegt i 
.
 
, t1i , 44i 44i 44i 
per 
1.815 103 
.
 
perBB.Bit44 
:= 
BB:= 
Bt44loss Bt44loss 
, 
nt44 
SAt 
Fraction boron remaining in a BWR waste package (50 percent) 
(see Assumption 5.1.5) 
10000, 0 if tdegt 
..
0.5 
:= 
if tdegt , . := tt Ai 
< 
k44 
44i 
. 
=
44i 

44i 
0 , 1 , ..
, . := 
Parameters required to solve the boron loss 
equation when the time to degrade is longer 
than the regulatory period of 10,000 years tt 10000, 1 if tdegt if tdegt Ci 
. 
.. 
.... 44i 44i 
1 , 0 , 44i 
10000, 10000 if tdegt 
.
. 
.
.
, , . := 10000, 1 tt if tdegt Ei 44i 44i 44i 
.. 
.. 
. 
. 
February 2004 III-23 of III-26 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Boron Loss Equation (BSC 2003 [DIRS 165890], Section 6) 
A . i NBt ( 44 . vt44 i , ) vt44 Ci 
.. 
..  
...  
root NBt ( 1 10 .
11 1 1011 
44 , vt44 i , ) vt44, ~ 
.. 
... 
Vt44dri 
:= 
reqt44dr := 
Required drip rate (m 3/yr) into a 44-BWR Absorber Plate waste package to 
degrade and flush out the boron based on sampled corrosion and time to 
first seismic event. 
The following is the equation used to calculate the probability that the seepage rate can be at or 
greater than the seepage noted as z 44 (for lithophysal) and z1 44 (for nonlithophysal). This seepage 
is based on the seepage rate making it through the damaged areas of the drip shield and waste 
package. The drip rate value is based on sampling the corrosion rate of Neutronit and sampling 
the time to first seismic occurrance, which determines the time to degrade and flush out the 
Neutronit from a 44-BWR Absorber Plate waste package for criticality potential. 
. 
Vr44 
zt44i 
:= 
z1t44i 
+ 
.DS1ds i 
reqt44dri
.DSdnli 
.. 
Vt44dr 
1000 
reqt44dri 
100
.WP1di 
1
100
.WP1di 
1 ... 
... 
Lithophysal Zone 
.
:= 
1 STlptl zt44i 
i 
:=
... 
..  
:= 
:= 
Qt44mspli 
Qt44uspli 
.. 
.. 
1 STuptl zt44i 
.. 
:= , 
.. 
.. The probability of being in the low-, mean- or upper-infiltration scenario for the glacial transition 
climate is 0.24, 0.41, and 0.35, respectively (BSC 2003 [DIRS 165991], Section 7, Table 7-1). 
0.24 Qt44lspl 
. 
if Xi 5 , 
...  
.. 
1 . STmptl zt44 
, := 
.. 
.. 
. 
. 
if Xi 5 , 
:= 
Qt44lspli 
Qt44sfli 
Qt44sfn 0.24 Qt44lspn
. 
i 
.. .. 
ANL-EBS-NU-000008 REV 00 
Vr . 44 Dt44 .vt44. tt44i 0 , 
tdegt 44i 0 , i 0 , vt44 ..
. . . exp 1 exp tdegt Bt44loss Ei . 44i 0 , vt44 Vr44 
.... 
.... 
.... 
z44 is the amount of seepage rate (m 3/yr) required for 
lithophysal zone. 
z144 is the amount of seepage rate (m 3/yr) required for 
nonlithophysal zone. 
. 
1 STlptn z1t44 Qt44lspn i 
:= 
. 
i 
Qt44mspn 
...  
.. 
.
. 
Nonlithophysal Zone 
. 
i 
Qt44uspn 
. 
1 . STuptn z1t44i 
, 
:= 
i 
i 
0.65 Qt44mspl 
.. 
.. 
0.514 Qt44uspli 
.. 
.. 
0.637 . , 
:= 
i i 
.. 
.. 
1 STmptn z1t44 
. 
... 
. 
. ... 
, ~
. 
.
.
, 
.. 
. 0.194 if 0.24 Xi 5 , .. 
0.672 , , , . 
. 
0.524 Qt44uspn i 
. 
0.65 Qt44mspn 
.. 
0.154 if 0.24 Xi 5 , i i 
.. 
.. 
.. 
.. 
February 2004 III-24 of III-26 Screening Analysis for Criticality Features, Events, and Processes for License Application 
The probability of the seimic event causing sufficient damage to allow advective flow to penetrate, 
degrade, and flush out the Neutronit from a 44-BWR Absorber Plate waste package is calculated 
using the equations below for both lithophysal and nonlithophysal. The equations are based on 
the sampling of the probability of sufficient advective flow given a seismic event with magnitude, 
v, at time, t. This probability will be fed into top event MS-IC-1 of the SAPHIRE model for seismic 
event. The probability of a seismic event will be set to 1.0 in the SAPHIRE model since it is 
accounted for in this calculation. 
Lithophysal Zone 
u u := Prt44sfl 
.  
1.242~ 10. 10 Mean probability Prt44sfl 
l . 
T . Tu := 
. ( )( ) IE 
ns 
. ( ) T T ( ) IE . IE 
ns 
.  
9.924~ 10. 12 
= 
0 i 
ns.1 
u IEl .. 
0 i 
= 
Nonlithophysal Zone 
Prt44sfnl 
Prt44sfnl 
.
.
.
.
. 
Final Results of Seismic Sensitivity Case Analysis 
44-BWR Absorber Plate waste package 
..
= 
. 
= 
. 
Mean probability 
l . 
1.242 10. 10 
= 
Sensitivity Probability Calculation using the following assumptions : 
Corrosion rate of Neutronit assumed to be 1.5 times the corrosion rate of stainless steel 316, 
which was fit to a Weibull distribution (Assumption 5.1.4) 
Flushing out 90 percent of the boron from the PWR waste package types (Assumption 5.1.5) 
Flushing out 50 percent of the boron from the BWR waste package types (Assumption 5.1.5) 
No water ingress into a waste package prior to 700 years (Assumption 5.2.2) 
44-BWR Absorber Plate waste package, Neutronit plate thickness increased to 7-mm 
Lithophysal Zone 
Prt44sfl ~ 
Probabilities for the remaining waste package types remained the same and are , therefore, not 
listed. 
ANL-EBS-NU-000008 REV 00 
ns.1 
l . i 7 , Qt44sfli( ) 1 X . 
Equations based on Equation 6.4-6 of 
main report, DTN:MO0308SPACALSS.002 
(BSC 2003 [DIRS 164822]) and Seismic 
Consequence Abstraction (BSC 2003 [DIRS 
161812], Attachment VIII, Eq. VIII.2.11). 
. i 7 , Qt44sfni 
~ 9.924 10. 12 
( ) 1 X . 
Nonlithophysal Zone 
= Prt44sfnl 
February 2004 III-25 of III-26 Screening Analysis for Criticality Features, Events, and Processes for License Application 
INTENTIONALLY LEFT BLANK 
III-26 of III-26 February 2004 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ATTACHMENT IV 
SEEPAGE ANALYSIS SPREADSHEETS (OUTPUT FROM MATHCAD FILES) 
IV-1 of IV-80 ANL-EBS-NU-000008 REV 00 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
INTENTIONALLY LEFT BLANK 
February 2004 IV-2 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ATTACHMENT IV - SEEPAGE ANALYSIS SPREADSHEETS (OUTPUT FROM 
MATHCAD FILES) 
The following sections presents the Mathcad analyses for the lower, mean and upper seepage 
infiltration rates of the glacial transition climate in both the lithophysal and nonlithophysal zones. 
IV.1 SEEPAGE ANALYSIS FOR LOWER INFILTRATION RATE IN THE 
LITHOPHYSAL ZONE 
The following section presents the Mathcad analysis for the lower seepage infiltration rate of the 
glacial transition climate in the lithophysal zone. The seepage information used in this analysis 
was obtained from Abstraction of Drift Seepage (BSC 2003 [DIRS 165564]). The information 
contained in the section has been abstracted from the seepage glac lower Tptpll driftcollapse 
report.mcd Mathcad file of Attachment VII. 
February 2004 IV-3 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
:= := 
:= := := := 
Seepage rate and seepage fraction calculation followed Abstraction of Drift Seepage 
(BSC 2003 [DIRS 165564], Section 6.7). 
rnd( ) 1.0 RDi 1 - , 12 
rnd( ) 1.0 := 
rnd( ) 1.0 
:= 
rnd( ) 1.0 RDi 1 - , 10 
rnd( ) 1.0 RDi 1 - , 11 
Latin Hypercube Sampling Routine to Generate Random Numbers 
n := 20000 
:= := rnd( ) 1.0 RDi 1 - , 9 rnd( ) 1.0 RDi 1 - , 6 RDi 1 - , 3 
rnd( ) 1.0 RDi 1 - , 7 
rnd( ) 1.0 RDi 1 - , 8 
Sample Size: 
i 1 := ..n 
RDi 1 - , 0 
:= RDi 1 - , 1 
RDi 1 - , 2 
i := 
rnd( ) 1.0 RDi 1 - , 4 
rnd( ) 1.0 RDi 1 - , 5 
RKs are matrixes in which the first column contain a permutation on the integers on the 
interval [1,n]. 
RK1:= csort := := := ( ) , ( ) , ( ) , ( ) RD 1 , RK4 csort RD 4 RK7 csort RD 7 RK10 csort RD 10 
RK2:= csort := := := ( ) , ( ) , ( ) , ( ) RD 2 , RK5 csort RD 5 RK8 csort RD 8 RK11 csort RD 11 
RK3:= csort := := := ( ) , ( ) , ( ) , ( ) RD 3 , RK6 csort RD 6 RK9 csort RD 9 RK12 csort RD 12 
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n 
February 2004 
Define sets of random values. Each random value is selected within one of the 
equiprobable n intervals that partition [0,1], one set for each random variable. 
runif , ( ) n 0, 1 < > RK7 0 < > 
1 - 
X 6 := runif , ( ) n 0, 1 < > RK4 0 < > 
1 - 
X 3 := X 0 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK8 0 < > 
1 - 
X 7 := runif , ( ) n 0, 1 < > RK5 0 < > 
1 - 
X 4 := 
< > RK1 0 < > 
1 - 
X 1 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK9 0 < > 
1 - 
X 8 := runif , ( ) n 0, 1 < > RK6 0 < > 
1 - 
X 5 := 
< > RK2 0 < > 
1 - 
X 2 := 
+
n
n 
+
n 
X := 
+ ( ) , 10 < > RK110 < > 
1 - runif n 0, 1 
n 
IV-4 of IV-80 
< > RK3 0 < > 
1 - 
< > ( ) , RK100 < > 
1 - + runif n 0, 1 
X 9 := 
X := 
+ ( ) , 11 < > RK120 < > 
1 - runif n 0, 1 
n 
i 0 - := ..n 1 
ANL-EBS-NU-000008 REV 00 
:= rnd( ) 1.0
Screening Analysis for Criticality Features, Events, and Processes for License Application 
Capillary Strength 1/  in (Pa) 
1lb 
Sampling from spatial variability to obtain the 1/  value 
.1 := 0  .1 
i 
Sample from uncertainty triangular distribution to obtain .1/ 
Determine which equation to use: 
if Random Number < RN . 1 then use Equation 1 ( . 1eq1). 
if Random Number > RN . 1 then use Equation 2 ( . 1eq2). 
qunif( ) Xi 0 , ,  1lb, 1ub 1/ value 1 := 
RN.1 
eq1i 
.1
eq2i 
.1
i .1
i 
T1 := 
Permeability k in Tptpll Unit (in log 10) 
kTl := 
Overall Capillary Strength 1/  + .1/ 
.. 
i 1 
.11.5 
0.47 kTl := 
kTli 
mean kTl 
Stdev kTl 
Permeability .k in Tptpll Unit (in log 10) 
( ) = .11.5 
( ) = 0.47 
:= .k 
ln( ) qlnorm( ) X ,  kTl, 
Uncertainty follows a triangular distribution := .k .kTll 
ANL-EBS-NU-000008 REV 00 
:= := Spatial variability follows a uniform distribution 1 := 591 780  402 1ub 
:= Uncertainty follows a triangular distribution .105 .1 := 105 u l 
:=
:= l 
:= 
:= 
:= 
( ) .1 . .1l 
2 
( ) .1u . .1l .( ) .1 . .1l 
Xi 1 , l u l 
.1 . 
.1 + 
u 
. . ( ) .1 . .1 .( ) 1 .1  . 
u l i 1 , ( ) 1 X . . . ( ) .1u . .1 .( ) 1 . .1 
i i 
.. 
+ .1 
Mean of lognormal distribution 
Standard deviation of lognormal distribution 
if( ) Xi 1 ,  RN.1, . .1eq1 , 1eq2 .1/ value 
1/ value i 
kTl i 2 , 
:= 0.92 .0.92 0 Tlu Tl 
February 2004 IV-5 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Sample from uncertainty triangular distribution to obtain .k 
Determine which equation to use: 
if Random Number < RN .kTl then use Equation 1 ( .kTleq1). 
if Random Number > RN .kTl then use Equation 2 ( .kTleq2). 
RN.kTl 
:= 
. := 
:= 
.kTleq1i 
.kTleq2i 
Tli 
:= 
i 
:= 
TkTl := 
i 
:= ( ) 
.k 
T1kTl 
ffi 
( ) .kTl . .kTll 
2 
( ) .kTlu . .kTll .( ) .kTl . .kTll 
+ Xi 3 , Tlu Tll .k
Tlu .k 
if Xi 3 , 
. . . ( ) .k . .k .( ) kTl .kTll 
i 3 , ( ) 1 X . . . ( ) .kTlu . .k .( ) k . .k 
 RN.kTl eq1i .k value eq2i 
k value i 
Tl Tlu Tll 
.. 
.. 
 .14, .14,T1kTl 
102.3 x . 
.. 
11.434 . 
, . .kTl , kTl ( ) 
Overall Permeability k + .k 
+ kTl Tli i 
Permeability must lie between -14 and -10 (bounds of SMPA simulations) 
if T1kTl i 
. 
.. 
.. 
Flow Focusing Factor (DTN: LB0104AMRU0185.012 [DIRS 163906]) 
f x ( ) := .0.3137 
.k
 .10, .10 if T1kTl i 
x4 
, 
5.4998x3 
.. 
35.66 x2 . + 
.. 
. 
root f x . ( ) X .100 , x, 0, 6 i 5 , 
+ .
IV-6 of IV-80 
Tll 
.. 
ANL-EBS-NU-000008 REV 00 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Percolation Flux (mm/yr) 
The percolation flux used here is for the glacial transition period only. The percolation flux is 
based on sampling from the lower bound TSPA repository location only 
(DTN: LB0310AMRU0120.002 [DIRS 166116]). 
nnn .. := 0 468 
Lower Bound Percolation Flux 
PF1 := l 0
3.68 
2.65 
2.41 
2.13 
2.41 
2.4 
2.12 
2.76 
1.4 
2.21 
0
1
2
3
4
5
6
7
8
9
PF1l PFtl nnn 0 , nnn 
:= 
Z ( ) ( ) , 0 := round runif n 0, 468 
l Zi 0 , 
, := q 
. . 
PF := PFt i 
q1pffi 
pffi 
( ) 
Adjusted Percolation Flux 
Multiply the flow-focusing factor by the percolation flux, which will be used to obtain the 
seepage rate, seepage fraction, and seepage percentage 
PFi ffi . 
1  1 , if q1 
:= 
Percolation Flux must lie between 1 and 1000 mm/yr (bounds of SMPA simulations) 
if q1pffi 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
 1000, 1000, q1pffi pffi 
.. 
.. 
February 2004 IV-7 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seepage Information from SMPA analysis 
(DTN: LB0307SEEPDRCL.002 [DIRS 164337]) 
m:= 2549 data points 
SMPAdata := 
0 1 2 3 4 5 6 
-14 0 
-14 1 
-14 2 
-14 3 
-14 4 
-14 5 
-14 6 
-14 7 
-14 8 
-14 9 
-14 10 
-14 11 
-14 12 
-14 13 
-14 14 
-14 15 
Develop routine to select correct mean seepage, seepage standard deviation, seepage percent, 
and seepage percent standard deviation based on sampled value of 1/ , k, percolation flux. 
nz := 16 ny := 9 nx := 14 
ii .. := 0 nx 
xi := ii ii ii x 
, 
:= SMPAdataii 2 
jj .. := 0 ny 
:= yj jj y . := 100 jj + 100 jj jj 
kk .. := 0 nz 
:= := . . 14 + kk 0.25 kk zkkk zkk 
ANL-EBS-NU-000008 REV 00 
SMPAdata
<0> is permeability value log(k [m^2]) 
SMPAdata
<1> is capillary strength 1/alpha [Pa] 
SMPAdata
<2> is local percolation flux (mm/yr) 
SMPAdata
<3> is Mean Seepage [kg/yr/WP] 
SMPAdata
<4> is Std. Dev. Seepage [kg/yr/WP] 
SMPAdata
<5> is Mean Seepage [%] 
SMPAdata
<6> is Std. Dev. Seepage [%] 
56.44 1 100 
282.63 5 100 
566.16 10 100 
1135.12 20 100 
2849.95 50 100 
5726.78 100 100 
200 11523.63 100 
300 17369.22 100 
400 23241.94 100 
500 29154.54 100 
35097.8 600 100 
700 41099.26 100 
800 47084.03 100 
900 53190.45 100 
1000 59206.88 100 
55.25 1 200 
IV-8 of IV-80 
9.75 100.6 5.47 
9.8 100.76 27.5 
9.83 100.92 55.13 
9.79 101.17 109.85 
9.71 101.6 272.25 
9.55 102.08 535.98 
9.49 102.71 1064.22 
9.41 103.2 1583.08 
9.3 103.57 2086.65 
9.1 103.94 2552.38 
8.89 104.27 2992.46 
8.69 104.66 3411.36 
8.6 104.91 3860.77 
8.21 105.35 4145.2 
8.06 105.54 4520.61 
9.69 98.48 5.44 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
sms2 := 
i 
smsd2 := 
+ 
sms3 := 
i 
smsd3 := 
+ 
loc represents the location within the matrix of which value to pick for the interpolation process. 
loc1i 
loc 
loc 
:= 
sms1 := 
i 
smsd1 := SMPAdataloc i 
spms1 := SMPAdataloc i 
spsd1i 
:= SMPAdataloc1 
dataloc2 
spms2 := SMPAdataloc 
spsd2 := SMPAdataloc i 
dataloc3 
spms3 := SMPAdataloc 
spsd3 := SMPAdataloc i 
loc
2i 
:=
i
i 
3i 
4i 
floor linterp z zk , TkTli 
+
SMPAdataloc1i
1i 
1i 
i 
kTli 
SMPAdataloc2i 
SMPA 
i 
2i 
2i 
SMPAdataloc3i 
SMPA 
i 
3 
:=
i
i i 
, 
4 , 
5 ,
6 , 
4 , 
5 ,
6 , 3i 
, 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
floor linterp z zk , T 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , ,TkTli 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , T 
i
4i 
4i 
, 
3 , 
4 , 
5 ,
6 , 
:=
i 
i i 
. + . 
i 
( ) 
1\ 
, + . . ... 
( ) , + ( ) ny 1 floor linterp y yj,T 
. 
nx 1 + 
. 
nx 1 + 
i 
, ... 1\ 
. . 
nx 1 + 
. 
.( ) , + ( ) ny 1 floor linterp y yj T 
. 
nx 1 + ( ) + , + ( ) ny 1 ceil linterp y yj,T 
. 
nx 1 + 
1\
i 
.
. 
. 
February 2004 
sms4 := 
i 
smsd4 := SMPAdataloc 
spms4 := SMPAdataloc i 
spsd4 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
.. 
..
SMPAdataloc4 
..
. 
. 
.. 
.. 
..
.. 
.. 
.. 
floor linterp x xi , , qpffi 
.. 
.. 
..
.. 
.. 
.. 
. 
. 
. 
. 
. 
. 
. 
. 
kTli 
+ 
.. 
.. 
4 
.. 
.. 
.. 
.. 
. 
( )... 
. 
nx 1 + 
. nx 1 + 
1\ 
i
. 
. 
.. 
+ 
. 
( ) 
( )... 
. 
nx 1 + 
.
. 
. 
( ) , + ( ) ny 1 ceil linterp y yj T . 
.. 
.. 
.. 
.. 
IV-9 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
loc ceil linterp z zk , T 5i kTli 
+ 
dataloc5i
5i 
5i 
s 
:=
i 
psd5i i 
.  
SMPA sms5 := 
i 
smsd5 := SMPAdataloc 
spms5 := SMPAdataloc i
:= SMPAdataloc5 
ceil linterp z zk , T loc6i kTli 
.. 
.. 
.. 
i 
+ 
sms6 := 
i 
smsd6 := 
:=
i
i 
6 
loc7i 
.. 
.. 
.. 
.. 
dataloc6 
dataloc7i 
.. 
+ .. 
.. 
.  
sms7 := 
i 
smsd7 := 
+ 
:=
i
i 
7 
.. 
..
ceil linterp x xi , , qpffi 
SMPA 
SMPAdataloc6 
spms6 := SMPAdataloc 
spsd6 := SMPAdataloc i 
ceil linterp z zk , ,TkTli 
ceil linterp x xi , , qpffi 
SMPA 
SMPAdataloc7 
spms7 := SMPAdataloc 
spsd7 := SMPAdataloc i 
ceil linterp z zk , ,T loc8i kTli 
.. 
.. 
.. 
.. 
.... 
floor linterp x xi , , qpffi 
i 
, 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
4 , 
5 , 
4 , 
5 , 
4 , 
5 , 
.. 
.. 
8 
,
3 ,
i 
6i 
6 , 
i
3 ,
i 
7i 
6 , 
i
3 , 
8i 
8i 
6 , 
i 
:=
i 
i 
.. 
..
SMPAdataloc8 
sms8 := 
i 
smsd8 := SMPAdataloc 
spms8 := SMPAdataloc i 
spsd8 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
. ( )( ) ny 1 + floor linterp y yj , ,T .. 
.. 
+ 
.. 
.. 
+ 
( ). , + ( ) ny 1 floor linterp y yj,T1i 
.. 
nx 1 + 
( ) , + ( ) ny 1 ceil linterp y yj,T + . . 
1
i 
.. 
.. 
.. 
.. 
.. 
.. 
. 
.. 
nx 1 + 
.. 
.. 
nx 1 + 
.. 
. 
. .( ) + , + ( ) ny 1 ceil linterp y yj,T1i 
.. 
.. 
. 
1
.. 
.. 
.. 
nx 1 + 
. . 
. . 
.. 
IV-10 of IV-80 
. .. 
. 
i
.. 
nx 1 + 
( )... 
.. 
nx 1 + 
... 
( )... 
.. 
nx 1 + 
( ) 
.. 
nx 1 + 
.
( )... 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Develop the upper and lower bound of the randomly generated 1/ , k, and adjusted percolation 
flux. 
Develop the upper and lower bound for permeability (k) for Tptpll Unit 
qq := .1.TkTl i i 
mantissa x . := ( ) x floor qq ( ) 
( ) i tt := floor qqi 
rr , := round( ) mantissa qq ( ) 2 
yy1 := i i , i if , < ( ) rr  0.25, 0 if( ) 0.25 rri  0.5, 0.25 rr 
if , < ( ) yy1  0.5, yy1 if( ) 0.5 yy1  0.75, 0.5, 0.75 zz1 := i i i i 
1 . . tt i i TkTl2 := 
i 
yy2 := 
( ) + zz1 
if , < ( ) rr  0.25, 0.25 if( ) 0.25 rri  0.5, 0.5 rri , i i 
zz2 := i i i i 
:= T 
if , < ( ) yy2  0.5, yy2 if( ) 0.5 yy2  0.75, 0.75, 1 
.1.( ) tt + zz2i i kTl1i 
Develop the upper and lower bound for capillary strength (1/ ) 
hh1 := floor i 
... 
.
. 
T1i 
100 
100 ( ) hh1 . i T11 := 
i 
:= ceil hh2i 
.
. 
T1i 
100 
T12 :=( ) hh2 .100 i i 
... 
.. 
i i i i 
Lower Bound value adjusted percolation flux (q pff) 
:= if q aaa1 5  1 , , qpff 1 , <  1 , if 1 qpff pff i i i i 
.. 
.. 
..
,  < , , 10, 5 aaa1 if aaa1 5  aaa1 if 5 aaa1 i bbb1 := i i i i ( ) ( ) 
if , < ( ) bbb1  10, bbb1 if( ) 10 bbb1  20, 10, bbb1 ccc1 := i i i i i 
ddd1 := i i i i i if , < ( ) ccc1  20, ccc1 if( ) 20 ccc1  50, 20, ccc1 
if , < ( ) ddd1  50, ddd1 if( ) 50 ddd1  100, 50, ddd1 eee1 := i 
if , < ( ) eee1  100, eee1 if( ) 100 eee1  200, 100, eee1 fff1 := i i i i i 
February 2004 IV-11 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ggg1 := i i i i i if , < ( ) fff1  200, fff1 if( ) 200 fff1  300, 200, fff1 
if , < ( ) ggg1  300, ggg1 if( ) 300 ggg1  400, 300, ggg1 hhh1 := i i i i i 
iii1 := i i i i i if , < ( ) hhh1  400, hhh1 if( ) 400 hhh1  500, 400, hhh1 
if , < ( ) iii1  500 iii1 if( ) 500 iii1  600, 500 iii1 jjj1 := i i , i , i i 
kkk1 := i i , i , i i if , < ( ) jjj1  600 jjj1 if( ) 600 jjj1  700, 600 jjj1 
if , < ( ) kkk1  700, kkk1 if( ) 700 kkk1  800, 700, kkk1 mmm1:= i i i i i 
:= q if , < ( ) mmm1  800,mmm1 if( ) 800 mmm1  900, 800, 900 i i i pff1i 
i 
i 
i 
Upper Bound value adjusted percolation flux (q pff). 
, 1 <  5 , if 1 q if q aaa2 := i 5  5 , , qpffi 
.. 
.. 
.. 
.. 
pffi pffi 
bbb2 := i i i i i if , < ( ) aaa2  5, aaa2 if( ) 5 aaa2  10, 10, aaa2 
if , < ( ) bbb2  10, bbb2 if( ) 10 bbb2  20, 20, bbb2 ccc2 := i i i i i 
ddd2 := i i i i i if , < ( ) ccc2  20, ccc2 if( ) 20 ccc2  50, 50, ccc2 
if , < ( ) ddd2  50, ddd2 if( ) 50 ddd2  100, 100, ddd2 eee2 := i i i i i 
if , < ( ) eee2  100, eee2 if( ) 100 eee2  200, 200, eee2 fff2 := i i i i i 
ggg2 := i i i i i if , < ( ) fff2  200, fff2 if( ) 200 fff2  300, 300, fff2 
if , < ( ) ggg2  300, ggg2 if( ) 300 ggg2  400, 400, ggg2 hhh2 := i i i i 
iii2 := i i i i if , < ( ) hhh2  400, hhh2 if( ) 400 hhh2  500, 500, hhh2 
if , < ( ) iii2  500 iii2 if( ) 500 iii2  600, 600 iii2 jjj2 := i i , i , i i 
kkk2 := i i , i , i i if , < ( ) jjj2  600 jjj2 if( ) 600 jjj2  700, 700 jjj2 
if , < ( ) kkk2  700, kkk2 if( ) 700 kkk2  800, 800, kkk2 mmm2:= i i i i 
if , < ( ) mmm2  800,mmm2 if( ) 800 mmm2  900, 900, 1000 i i i qpff2 := 
i 
. T T kTl1i kTli 
:= vTkTli . T T 
Interpolate (Solve) for seepage rate (Tptpll Unit) 
. . q T q T1 pff pff1i 11i i i 
:= := t u qpff . . T1i i q T q T pff1i 11i pff2i 12i kTl1i kTl2i 
February 2004 IV-12 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
spfluxTlmi 
1 t . qpffi.
tqpffi 
. 
.. 
tqpffi 
1 t . qpffi 
1 t . qpffi 
...... 
.. 
. + t 
. +
+ 
qpffi 
tqpff 
. 
i 
1 tqpffi 
......
:= 1 t . qpff . 
:= 
spfluxTlsdi i.
+ t 
. 
.. 
. 
.. 
. 
.. 
.... 
.. 
.. 
+
+
+
+ 
qpffi 
+ tqpffi 
........
. tqpff
. 
. 
i 
.
. 
1 t . qpffi 
1 t . qpffi 
1 t 
tqpff 
. 
i
.. 
..
qpffi 
.. 
.... 
+
+
+
+
+ 
...... 
.. 
.. 
Calculate mean seepage for Tptpll Unit 
1.7321spfluxTlsdi 
QTl1stdli 
:= 
QTl1stdu i 
QTlstdli 
, 
i 
QTlstdli 
i 
, 
+ QTlstd i 
0.1, 0,QTl1spmi 
. 
1.7321.spfluxTlsdi 
if QTl1spmi 
QTl2spm 100 
i 
:= 
QT2perc := 
. 
28.05.2 
.
qpffi
. 
:=
:=
if QTl1stdli 
qunif Xi 1 , 0 
if QT2perci 
:= 
QTlstd :=
:= spfluxTlm QTl1spmi 
QTl2spmi 
i 
QT3perci 
0, .0.00001 QTl1stdli 
, 
QTl1stdu i 
. 
, 
.. .. 
ANL-EBS-NU-000008 REV 00 
1 u . T1\ 1 v . TkTl sms1 ... 
i i i 
. . ... s 1 u . T1\ ms2i i i 
. ... .s u 
i 
. . . ... u s 
.
1 vTkTl 
1 v . TkTl 
1 vTkTl 
T1\i
T1\ 
ms3i
ms4i i i 
. . .s v ms5i 
.. 
.. 
. . s 
TkTli
ms6 1 uT1\ i 
. 
. 
.
.. 
. 
. . 
 . 
1 u . T1\i
v 
i
v . ... u .s 
i 
.
.
.
. u 
.. 
.. 
T1\i
T1\ ms8i i 
... 1 vTkTl i i.
s . ... 
. 
. 
i 
.
.. 
.. 
.. 
i 
. 
.. 
..
1 uT1\i. 
.
. ... 
.. 
.
TkTli
ms7 
.s v 
TkTli 
TkTl 
. 
i
1 v 
. v 
. 
. 
. 
... 
i 
.. 
. 
1 v . TkTl 
.s 
1 v .
. 
. 
.
TkTli 
TkTl
msd3i 
msd4 i 
. 
. 
i 
. .
..
1 uT1\ 
u 
. u 
T1\i
T1\ 
1 u . ... 
... 
msd5i 
. . ... s 
... 
. msd1 s
i 
.
v
s 
. 
.
T1\i 
i
v 
. 
. 
u . . 
. 
. 
.. 
. 
.. 
i 
. . . 
. . 
.
msd2i 
s .
.s TkTli
msd6 
... 
msd8i 
.. 
....
1 uT1\ 
T1\i
T1\ 
. 
. 
i 
.
u 
.. 
.
TkTli
msd7 
s v 
.. 
.. 
i 
.. 
..
TkTli 
TkTl 
. 
. 
. 
. 
. 
.. 
.. 
.. 
. 
Equation to calculate seepage percent is based on seepage rate 
(see SMPA data table) (from DTN: LB0310AMRU0120.002 
[DIRS 166116]) 
. 100, 100,QT2perci 
0 . 0 , if QT2perci 
. 
.. 
. 
IV-13 of IV-80 
Check seepage percent; if above 
100 percent, then recalculate 
seepage back to 100 percent. 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
QTlspri 
mean QTlspr 
:= 
Determine the seepage fraction for Tptpll Unit within the repository and then fit the output data 
to distribution. 
Seepage fraction represents the non-zero seepage rates based on the LHS sampling of all of the 
parameters. 
( )= 3.754 
spr 
16122 
16123 
16124 
16125 
16126 
ab .. := 0 n1Tl 
sort( ) QTl 
reverse( ) Q11 
:= 
998
.Q1Tlab 
1 ... 
. 
Tl 
sort( ) QTl = 
nTl := 16123 
spfrcTl := 
n1Tl := 
Q11 := Tl 
Q1Tl 
Q2Tlab 
:= 
QTl := 
CDFTl 
( ) n 1 . .nTl 
n 
( ) n 1 . .( ) n + 1 Tl 
spr
Tl 
Fit the seepage rates to a Weibull distribution. 
The following equations are from What Every Engineer Should Know About Reliability and 
Risk Analysis (Modarres 1993 [DIRS 104667], p. 109]) 
( ) n1 + 1 . 0.25 
n1Tl  = 3.875 103 
ANL-EBS-NU-000008 REV 00 
Mean Seepage Rate (kg/yr per WP) QT3perci
.qpffi
. 
100 
2.28.05 
0 
0
0 
0.101 
0.101 
0.101 
Seepage fraction (i.e., waste package locations that see 
seepage) spfrcTl = 0.194 
sort( ) Q2 
( . + ab 1) 0.375 := 
ab Tl 
February 2004 IV-14 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
 := root 
........  
........  
n1Tl
0 
=
i 
:=  
.....  
CDFw1ji 
...  
1 
0.75 
0.5 
0.25
0 
1 .10 4 
. 
Plot of raw data versus Weibull distribution 
ji 0 n1Tl 
ji 2 , := 
1 exp . ji := 
CDFwji 0 , 
cumulative density function 
:= .. 
ji PFdata := QTlji 
CDFdata 
CDFw1 
:= 
CDFdata 2 . . 
CDFw 
ANL-EBS-NU-000008 REV 00 
CDFTlji
PFdata ji 
. 
 
.... 
Seepage Rate (Tptpll) data and fit 
1 .10 3 0.01 
PFdata 
Seepage Rate (m3/yr) 
n1Tl 
QTl i i
. . 
.
r ln QTl . 
..
n1Tl 
0 i 1 
.
...
. . ln QTli 
1
r 
.. 
..
n1Tl n1Tl 
.
. 
= 
. 
0 i 
.
... 
.. 
. 
QTli 
0 i 
.. 
1
 
.. 
.. 
 
= 
= 
QTli 
...  
.. 
n1Tl 
... 
 8.869 10 3 . ~ = 
.
..... 
 
.. 
r 
.
. 
... 
... 
...  
IV-15 of IV-80 
r , , 0.1, 4  = 0.52 
......... 
......... 
0.1 1 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
IV.2 SEEPAGE ANALYSIS FOR MEAN INFILTRATION RATE IN THE 
LITHOPHYSAL ZONE 
The following section presents the Mathcad analysis for the mean seepage infiltration rate of the 
glacial transition climate in the lithophysal zone. The seepage information used in this analysis 
was obtained from Abstraction of Drift Seepage (BSC 2003 [DIRS 165564]). The information 
contained in the section has been abstracted from the seepage glac mean Tptpll driftcollapse 
report.mcd Mathcad file of Attachment VII. 
February 2004 IV-16 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
:= := 
:= := := := 
Seepage rate and seepage fraction calculation followed Abstraction of Drift Seepage 
(BSC 2003 [DIRS 165564], Section 6.7). 
rnd( ) 1.0 RDi 1 - , 12 
rnd( ) 1.0 := 
rnd( ) 1.0 
:= 
rnd( ) 1.0 RDi 1 - , 10 
rnd( ) 1.0 RDi 1 - , 11 
Latin Hypercube Sampling Routine to Generate Random Numbers 
n := 20000 
:= := rnd( ) 1.0 RDi 1 - , 9 rnd( ) 1.0 RDi 1 - , 6 RDi 1 - , 3 
rnd( ) 1.0 RDi 1 - , 7 
rnd( ) 1.0 RDi 1 - , 8 
Sample Size: 
i 1 := ..n 
RDi 1 - , 0 
:= RDi 1 - , 1 
RDi 1 - , 2 
i := 
rnd( ) 1.0 RDi 1 - , 4 
rnd( ) 1.0 RDi 1 - , 5 
RKs are matrixes in which the first column contain a permutation on the integers on the 
interval [1,n]. 
RK1:= csort := := := ( ) , ( ) , ( ) , ( ) RD 1 , RK4 csort RD 4 RK7 csort RD 7 RK10 csort RD 10 
RK2:= csort := := := ( ) , ( ) , ( ) , ( ) RD 2 , RK5 csort RD 5 RK8 csort RD 8 RK11 csort RD 11 
RK3:= csort := := := ( ) , ( ) , ( ) , ( ) RD 3 , RK6 csort RD 6 RK9 csort RD 9 RK12 csort RD 12 
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n 
February 2004 
Define sets of random values. Each random value is selected within one of the 
equiprobable n intervals that partition [0,1], one set for each random variable. 
runif , ( ) n 0, 1 < > RK7 0 < > 
1 - 
X 6 := runif , ( ) n 0, 1 < > RK4 0 < > 
1 - 
X 3 := X 0 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK8 0 < > 
1 - 
X 7 := runif , ( ) n 0, 1 < > RK5 0 < > 
1 - 
X 4 := 
< > RK1 0 < > 
1 - 
X 1 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK9 0 < > 
1 - 
X 8 := runif , ( ) n 0, 1 < > RK6 0 < > 
1 - 
X 5 := 
< > RK2 0 < > 
1 - 
X 2 := 
+
n
n 
+
n 
X := 
+ ( ) , 10 < > RK110 < > 
1 - runif n 0, 1 
n 
IV-17 of IV-80 
< > RK3 0 < > 
1 - 
< > ( ) , RK100 < > 
1 - + runif n 0, 1 
X 9 := 
X := 
+ ( ) , 11 < > RK120 < > 
1 - runif n 0, 1 
n 
i 0 - := ..n 1 
ANL-EBS-NU-000008 REV 00 
:= rnd( ) 1.0
Screening Analysis for Criticality Features, Events, and Processes for License Application 
Capillary Strength 1/  in (Pa) 
1lb 
Sampling from spatial variability to obtain the 1/  value 
.1 := 0  .1 
i 
Sample from uncertainty triangular distribution to obtain .1/ 
Determine which equation to use: 
if Random Number < RN . 1 then use Equation 1 ( . 1eq1). 
if Random Number > RN . 1 then use Equation 2 ( . 1eq2). 
qunif( ) Xi 0 , ,  1lb, 1ub 1/ value 1 := 
RN.1 
eq1i 
.1
eq2i 
.1
i .1
i 
T1 := 
Permeability k in Tptpll Unit (in log 10) 
kTl := 
Overall Capillary Strength 1/  + .1/ 
.. 
i 1
0.47 kTl := 
kTli 
.k 
mean kTl 
Stdev kTl 
Permeability .k in Tptpll Unit (in log 10) 
( ) = .11.5 
( ) = 0.47 
.0.92 := .k Tll 
ln( ) qlnorm( ) X ,  kTl, 
Uncertainty follows a triangular distribution := 0.92 0 := .kTlu Tl 
ANL-EBS-NU-000008 REV 00 
:= := Spatial variability follows a uniform distribution 1 := 591 780  402 1ub 
:= Uncertainty follows a triangular distribution .105 .1 := 105 u l 
:=
:= l 
:= 
:=
:= 
( ) .1 . .1l 
2 
( ) .1u . .1l .( ) .1 . .1l 
Xi 1 , l u l 
.1 . 
.1 + 
u 
. . ( ) .1 . .1 .( ) 1 .1  . 
u l i 1 , ( ) 1 X . . . ( ) .1u . .1 .( ) 1 . .1 
i i 
.. 
+ .1
Mean of lognormal distribution .11.5 
Standard deviation of lognormal distribution 
if( ) Xi 1 ,  RN.1, . .1eq1 , 1eq2 .1/ value 
1/ value i 
kTl i 2 , 
February 2004 IV-18 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Sample from uncertainty triangular distribution to obtain .k 
Determine which equation to use: 
if Random Number < RN .kTl then use Equation 1 ( .kTleq1). 
if Random Number > RN .kTl then use Equation 2 ( .kTleq2). 
RN.kTl 
:= 
. := 
:= 
.kTleq1i 
.kTleq2i 
Tli 
:= 
i 
:= 
TkTl := 
i 
:= 
.k 
T1kTl 
ffi ( ) 
( ) .kTl . .kTll 
2 
( ) .kTlu . .kTll .( ) .kTl . .kTll 
+ Xi 3 , Tll .k
Tlu .k 
if Xi 3 , 
. . . ( ) .k . .k .( ) kTl .kTll 
( ) 1 X . . . ( ) .kTlu . .k .( ) k . .k 
 RN.kTl .k value eq2i eq1i 
k value i 
11.434 
Tl Tlu Tll i 3 , 
.. 
.. 
.. 
 .14, .14,T1kTl 
102.3 x . 
, . .kTl , kTl ( ) 
Overall Permeability k + .k 
kTl Tli i 
Permeability must lie between -14 and -10 (bounds of SMPA simulations) 
if T1kTl 
+ 
x4 
.k
 .10, .10 if T1kTl i 
5.4998x3 . + . 
.. 
.. 
Flow Focusing Factor (DTN: LB0104AMRU0185.012 [DIRS 163906]) 
f x ( ) := .0.3137 
root f x . ( ) X .100 , x, 0, 6 i 5 , 
, 
i 
35.66 x2 + . 
.. 
.. 
. 
IV-19 of IV-80 
Tll Tlu 
. 
.. 
ANL-EBS-NU-000008 REV 00 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Percolation Flux (mm/yr) 
The percolation flux used here is for the glacial transition period only. The percolation flux is 
based on sampling from the mean bound TSPA repository location only 
(DTN: LB0310AMRU0120.002 [DIRS 166116]) 
nnn .. := 0 468 
Mean Bound Percolation Flux 
PF1 := m 0
15.97 
19.87 
14.2 
7.59 
16.94 
17.76 
10.45 
27.77 
8.95 
16.02 
0
1
2
3
4
5
6
7
8
9
PF1m PFtm nnn 0 , nnn 
:= 
Z ( ) ( ) , 0 := round runif n 0, 468 
m Zi 0 , 
, := q 
. . 
PF := PFt i 
q1pffi 
:= 
pffi 
( ) 
Adjusted Percolation Flux 
Multiply the flow-focusing factor by to the percolation flux, which will be used to obtain the 
seepage rate, seepage fraction, and seepage percentage 
PFi ffi . 
Percolation Flux must lie between 1 and 1000 mm/yr (bounds of SMPA simulations) 
1  1 , if q1 if q1pffi 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
 1000, 1000, q1pffi pffi 
.. 
.. 
February 2004 IV-20 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seepage Information from SMPA analysis 
(DTN: LB0307SEEPDRCL.002 [DIRS 164337]) 
m:= 2549 Data points 
SMPAdata := 
0 1 2 3 4 5 6 
-14 0 
-14 1 
-14 2 
-14 3 
-14 4 
-14 5 
-14 6 
-14 7 
-14 8 
-14 9 
-14 10 
-14 11 
-14 12 
-14 13 
-14 14 
-14 15 
Develop routine to select correct mean seepage, seepage standard deviation, seepage percent, 
and seepage percent standard deviation based on sampled value of 1/ , k, percolation flux. 
nz := 16 ny := 9 nx := 14 
ii .. := 0 nx 
xi := ii ii ii x 
, 
:= SMPAdataii 2 
jj .. := 0 ny 
:= yj jj y . := 100 jj + 100 jj jj 
kk .. := 0 nz 
:= := . . 14 + kk 0.25 kk zkkk zkk 
ANL-EBS-NU-000008 REV 00 
SMPAdata
<0> is permeability value log(k [m^2]) 
SMPAdata
<1> is capillary strength 1/alpha [Pa] 
SMPAdata
<2> is local percolation flux (mm/yr) 
SMPAdata
<3> is Mean Seepage [kg/yr/WP] 
SMPAdata
<4> is Std. Dev. Seepage [kg/yr/WP] 
SMPAdata
<5> is Mean Seepage [%] 
SMPAdata
<6> is Std. Dev. Seepage [%] 
1 100 
5 100 
10 100 
20 100 
50 100 
100 100 
200 100 
300 100 
400 100 
500 100 
600 100 
700 100 
800 100 
900 100 
1000 100 
1 200 
IV-21 of IV-80 
5.47 56.44 
27.5 282.63 
55.13 566.16 
109.85 1135.12 
272.25 2849.95 
535.98 5726.78 
1064.22 11523.63 
1583.08 17369.22 
2086.65 23241.94 
2552.38 29154.54 
2992.46 35097.8 
3411.36 41099.26 
3860.77 47084.03 
4145.2 53190.45 
4520.61 59206.88 
5.44 55.25 
9.75 100.6 
9.8 100.76 
9.83 100.92 
9.79 101.17 
9.71 101.6 
9.55 102.08 
9.49 102.71 
9.41 103.2 
9.3 103.57 
9.1 103.94 
8.89 104.27 
8.69 104.66 
8.6 104.91 
8.21 105.35 
8.06 105.54 
9.69 98.48 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
sms2 := 
i 
smsd2 := 
+ 
sms3 := 
i 
smsd3 := 
+ 
loc represents the location within the matrix of which value to pick for the interpolation process 
loc1i 
loc 
loc 
:= 
sms1 := 
i 
smsd1 := SMPAdataloc i 
spms1 := SMPAdataloc i 
spsd1i 
:= SMPAdataloc1 
dataloc2 
spms2 := SMPAdataloc 
spsd2 := SMPAdataloc i 
dataloc3 
spms3 := SMPAdataloc 
spsd3 := SMPAdataloc i 
loc
2i 
:=
i
i 
3i 
4i 
floor linterp z zk , TkTli 
+
SMPAdataloc1i
1i 
1i 
i 
kTli 
SMPAdataloc2i 
SMPA 
i 
2i 
2i 
SMPAdataloc3i 
SMPA 
i 
3 
:=
i
i i 
, 
4 , 
5 ,
6 , 
4 , 
5 ,
6 , 3i 
, 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
floor linterp z zk , T 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , ,TkTli 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , T 
i
4i 
4i 
, 
3 , 
4 , 
5 ,
6 , 
:=
i 
i i 
. + . 
i 
( ) 
1\ 
, + . . ... 
( ) , + ( ) ny 1 floor linterp y yj,T 
. 
nx 1 + 
. 
nx 1 + 
i 
, ... 1\ 
. . 
nx 1 + 
. 
.( ) , + ( ) ny 1 floor linterp y yj T 
. 
nx 1 + ( ) + , + ( ) ny 1 ceil linterp y yj,T 
. 
nx 1 + 
1\
i 
.
. 
. 
February 2004 
sms4 := 
i 
smsd4 := SMPAdataloc 
spms4 := SMPAdataloc i 
spsd4 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
.. 
..
SMPAdataloc4 
..
. 
. 
.. 
.. 
..
.. 
.. 
.. 
floor linterp x xi , , qpffi 
.. 
.. 
..
.. 
.. 
.. 
. 
. 
. 
. 
. 
. 
. 
. 
kTli 
+ 
.. 
.. 
4 
.. 
.. 
.. 
.. 
. 
( )... 
. 
nx 1 + 
. nx 1 + 
1\ 
i
. 
. 
.. 
+ 
. 
( ) 
( )... 
. 
nx 1 + 
.
. 
. 
( ) , + ( ) ny 1 ceil linterp y yj T . 
.. 
.. 
.. 
.. 
IV-22 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
loc ceil linterp z zk , T 5i kTli 
5i
5i 
5i 
SMPA 
:= 
sms5 := SMPAdataloc i 
smsd5 := SMPAdataloc i 
spms5 := SMPAdataloc i 
spsd5i 
. 
+ 
:= dataloc5i 
loc6i kTli 
.. 
.. 
.. 
i 
+ 
sms6 := 
i 
smsd6 := 
:=
i
i 
6 
loc7i 
.. 
.. 
.. 
.. 
dataloc6 
dataloc7i 
.. 
+ .. 
.. 
.  
sms7 := 
i 
smsd7 := 
+ 
:=
i
i 
7 
.. 
..
ceil linterp x xi , , qpffi 
SMPA 
SMPAdataloc6 
spms6 := SMPAdataloc 
spsd6 := SMPAdataloc i 
ceil linterp z zk , ,TkTli 
ceil linterp x xi , , qpffi 
SMPA 
SMPAdataloc7 
spms7 := SMPAdataloc 
spsd7 := SMPAdataloc i 
ceil linterp z zk , T 
,
3 ,
i 
6i 
6 , 
i
3 ,
i 
7i 
6 , 
i 
loc8i kTli 
.. 
.. 
.. 
.. 
.... 
.. 
nx 1 + 
floor linterp x xi , , qpffi
.. 
.. 
.. 
nx 1 + 
.. 
. 
. 
.. 
nx 1 + 
.. 
nx 1 + ..
.. 
. 
i 
, 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
ceil linterp z zk , T
4 , 
5 , 
4 , 
5 , 
4 , 
5 , 
.. 
.. 
8 
,
3 , 
8i 
8i 
6 , 
i 
:=
i 
i 
.. 
..
SMPAdataloc8 
sms8 := 
i 
smsd8 := SMPAdataloc 
spms8 := SMPAdataloc i 
spsd8 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
. + . ( )( ) ny 1 + floor linterp y yj , ,T .. 
.. 
. 
.. 
. 
i
.. 
nx 1 + 
. + . . 
( )... 
.. 
nx 1 + 
.. 
.. 
.. 
.. 
1 
( ) + ny 1 ( )+ , floor linterp y yj,T1i 
( ) , + ( ) ny 1 ceil linterp y yj,T1i 
i
.. 
.. 
.. 
( ) , + ( ) ny 1 ceil linterp y yj T + 
.. 
nx 1 + 
( ) 
.. 
nx 1 + 
... 
( )
... 
.. 
.. 
. 
... 
( ) , . 1 ..
.. 
.. 
IV-23 of IV-80 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Develop the upper and lower bound of the randomly generated 1/ , k, and adjusted percolation 
flux 
Develop the upper and lower bound for permeability (k) for Tptpll Unit 
qq := .1.TkTl i i 
mantissa x . := ( ) x floor qq ( ) 
( ) i tt := floor qqi 
rr , := round( ) mantissa qq ( ) 2 
yy1 := i i , i if , < ( ) rr  0.25, 0 if( ) 0.25 rri  0.5, 0.25 rr 
if , < ( ) yy1  0.5, yy1 if( ) 0.5 yy1  0.75, 0.5, 0.75 zz1 := i i i i 
1 . . tt i i TkTl2 := 
i 
yy2 := 
( ) + zz1 
if , < ( ) rr  0.25, 0.25 if( ) 0.25 rri  0.5, 0.5 rri , i i 
zz2 := i i i i 
:= T 
if , < ( ) yy2  0.5, yy2 if( ) 0.5 yy2  0.75, 0.75, 1 
.1.( ) tt + zz2i i kTl1i 
Develop the upper and lower bound for capillary strength (1/ ) 
hh1 := floor i 
... 
.
. 
T1i 
100 
100 ( ) hh1 . i T11 := 
i 
:= ceil hh2i 
.
. 
T1i 
100 
T12 :=( ) hh2 .100 i i 
... 
.. 
i i i i 
Lower Bound value adjusted percolation flux (q pff) 
:= if q aaa1 5  1 , , qpff 1 , <  1 , if 1 qpff pff i i i i 
.. 
.. 
..
,  < , , 10, 5 aaa1 if aaa1 5  aaa1 if 5 aaa1 i bbb1 := i i i i ( ) ( ) 
if , < ( ) bbb1  10, bbb1 if( ) 10 bbb1  20, 10, bbb1 ccc1 := i i i i i 
ddd1 := i i i i i if , < ( ) ccc1  20, ccc1 if( ) 20 ccc1  50, 20, ccc1 
if , < ( ) ddd1  50, ddd1 if( ) 50 ddd1  100, 50, ddd1 eee1 := i 
if , < ( ) eee1  100, eee1 if( ) 100 eee1  200, 100, eee1 fff1 := i i i i i 
February 2004 IV-24 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ggg1 := i i i i i if , < ( ) fff1  200, fff1 if( ) 200 fff1  300, 200, fff1 
if , < ( ) ggg1  300, ggg1 if( ) 300 ggg1  400, 300, ggg1 hhh1 := i i i i i 
iii1 := i i i i i if , < ( ) hhh1  400, hhh1 if( ) 400 hhh1  500, 400, hhh1 
if , < ( ) iii1  500 iii1 if( ) 500 iii1  600, 500 iii1 jjj1 := i i , i , i i 
kkk1 := i i , i , i i if , < ( ) jjj1  600 jjj1 if( ) 600 jjj1  700, 600 jjj1 
if , < ( ) kkk1  700, kkk1 if( ) 700 kkk1  800, 700, kkk1 mmm1:= i i i i i 
:= q if , < ( ) mmm1  800,mmm1 if( ) 800 mmm1  900, 800, 900 i i i pff1i 
i 
i 
i 
i 
. T T kTl1i kTli 
:= vTkTli . T T 
Upper Bound value adjusted percolation flux (q pff) 
, 1 <  5 , if 1 q if q aaa2 := i 5  5 , , qpffi 
.. 
.. 
.. 
.. 
pffi pffi 
bbb2 := i i i i i if , < ( ) aaa2  5, aaa2 if( ) 5 aaa2  10, 10, aaa2 
if , < ( ) bbb2  10, bbb2 if( ) 10 bbb2  20, 20, bbb2 ccc2 := i i i i i 
ddd2 := i i i i i if , < ( ) ccc2  20, ccc2 if( ) 20 ccc2  50, 50, ccc2 
if , < ( ) ddd2  50, ddd2 if( ) 50 ddd2  100, 100, ddd2 eee2 := i i i i 
if , < ( ) eee2  100, eee2 if( ) 100 eee2  200, 200, eee2 fff2 := i i i i i 
ggg2 := i i i i i if , < ( ) fff2  200, fff2 if( ) 200 fff2  300, 300, fff2 
if , < ( ) ggg2  300, ggg2 if( ) 300 ggg2  400, 400, ggg2 hhh2 := i i i i 
iii2 := i i i i if , < ( ) hhh2  400, hhh2 if( ) 400 hhh2  500, 500, hhh2 
if , < ( ) iii2  500 iii2 if( ) 500 iii2  600, 600 iii2 jjj2 := i i , i , i i 
kkk2 := i i , i , i i if , < ( ) jjj2  600 jjj2 if( ) 600 jjj2  700, 700 jjj2 
if , < ( ) kkk2  700, kkk2 if( ) 700 kkk2  800, 800, kkk2 mmm2:= i i i i 
if , < ( ) mmm2  800,mmm2 if( ) 800 mmm2  900, 900, 1000 i i i qpff2 := 
i 
. T T1 11i i 
. T T 11i 12i 
Solve for seepage rate (Tptpll Unit) 
. q qpff pff1i i 
:= := t u qpff . T1i i q q pff1i pff2i kTl1i kTl2i 
February 2004 IV-25 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
spfluxTlm 1 u . T1\ 1 v . TkTl sms1 ... 
i i i i 
1 t . qpffi.
. ... t 1 u . T1\ i i i qpffi 
. . ... .s u 
.. 
tqpffi i i 
. . . ... u 
.
1 vTkTl 
1 v . TkTl 
1 vTkTl 
T1\i
T1\ i i i 
. . .s v 
i i 
1 t . qpffi 
1 t . qpffi 
...... 
.. 
.. 
.. 
. + . . s t 1 uT1\ 
. 
. 
sms2 
ms3
sms4 
TkTl ms5 
ms6i qpffi 
.
.. 
. 
. . 
 . 
1 u . T1\i
v 
i
v . + . ... .s u 
i 
.
.
.
. + u 
.. 
.. 
T1\i
T1\ 
...... 
ms8i i 
1 t 
tqpff 
. 
i 
qpffi 
:= ... 
:= 
spfluxTlsdi 
.. 
.. 
+
+
+
+ 
... 
. msd1 s 1 vTkTl 1 t . qpff . 
i i i.
s . + . ... t 
i 
. 
.. 
. 
.. 
. 
.. 
.... 
.. 
. 
.. 
.. 
i 
. .
.. 
.. 
1 uT1\i.
i
1 v
1 v
i 
. ... 
.. 
. 
TkTl 
.
. 
. 
. 
i 
.. 
.
TkTli
ms7 
.s v 
TkTli 
TkTl 
1 v . TkTl 
.s 
.
i 
. v 
. 
. 
. 
... 
i 
. . 
..
1 uT1\ 
u 
. u 
T1\i
T1\ 
. ... 
... 
msd5i i 
........ 
TkTli 
TkTl 
s . . . ... tqpff
. 
. 
qpffi 
tqpffi
i 
.
v
s 
. 
.
1 uT1\ 
i
v 
. 
. 
u . . . 
.. 
.... 
. 
. 
.. 
. 
. . 
.
msd2i 
s 
msd3i 
msd4 .
.s
... 
msd8i 
. . 
. 
. 
. 
. 
. 
msd6i 
i 
TkTli
msd7 
s . . .
.. 
....
1 uT1\ 
T1\i
T1\ 
. 
. 
i 
.
u 
.. 
. 
v 
. 
1 t . qpffi 
1 t . qpffi 
tqpff 
. 
i 
1 t 
.. 
..
qpffi 
.. 
.. 
i 
.. 
..
TkTli 
TkTl 
+
+
+
+
+
+ 
. 
. 
. 
. 
...... 
. 
.. 
.. 
... 
. 
. 
.. 
Calculate mean seepage for Tptpll Unit 
1.7321spfluxTlsdi 
QTl1stdli 
:= 
QTl1stdu i 
0, .0.00001 QTl1stdli 
QTlstdli 
, 
i 
QTlstdli 
, 
QTl1stdu i 
i 
, 
+ QTlstd i 
0.1, 0, QTl1spmi 
. 
1.7321.spfluxTlsdi 
if QTl1spmi 
QTl2spm 
. 
100 
i 
:= 
QT2perc := 
. 
28.05.2 
.
qpffi
. 
, :=
:=
if QTl1stdli 
qunif Xi 1 , 0 
if QT2perci 
:= 
QTlstd :=
:= spfluxTlm QTl1spmi 
QTl2spmi 
i 
QT3perci 
.. .. 
ANL-EBS-NU-000008 REV 00 
Equation to calculate seepage percent based on seepage rate (see 
SMPA data table) from DTN: LB0310AMRU0120.002 [DIRS 166116] 
. 100, 100,QT2perci 
0 . 0 , if QT2perci 
. 
. 
.. 
IV-26 of IV-80 
Check seepage percent; if above 
100 percent, then recalculate 
seepage back to 100 percent. 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
QT3perc QTlspri i
.qpffi
. 
100 
2.28.05 := 
mean QTlspr 
0 
0
0 
0.1 
0.1 
( )= 175.204 
Determine the seepage fraction for Tptpll Unit within the repository and then fit the output data 
to distribution 
Seepage fraction represents the non-zero seepage rates based on the LHS sampling of all of the 
parameters 
spr 
9713 
9714 
9715 
9716 
9717 0.1 
spfrcTl = 0.514 
sort( ) QTl = 
nTl := 9714
( ) n 1 . .nTl 
spfrcTl := 
n 
( ) n 1 . .( ) n + 1 Tl n1Tl := 
spr Q11 := sort QTl Tl 
Tl Q1Tl 
Q2Tl 
:= 
ab 
QTl 
( ) 
reverse( ) Q11 
ab .. := 0 n1Tl 
:= 
998
.Q1Tlab 
1 . 
Tl 
... 
mean Q . Tl 998 
:= sort( ) Q2 
( )= 340.699 
( . + ab 1) 0.375 := CDFTlab Tl 
( ) n1 + 1 . 0.25 
Fit the seepage rates to a Weibull distribution. 
The following equations are from What Every Engineer Should Know About Reliability and Risk 
Analysis (Modarres 1993 [DIRS 104667], p. 109). 
n1  Tl = 1.028 104 
ANL-EBS-NU-000008 REV 00 
Mean Seepage Rate (kg/yr per WP) 
Seepage fraction (i.e., waste package locations that 
see seepage) 
February 2004 IV-27 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
 := root 
........  
........  
n1Tl
0 
=
i 
:=  
.....  
CDFw1ji 
...  
1 
0.75 
0.5 
0.25
0 
1 .10 4 
. 
Plot of raw data versus Weibull distribution 
ji 0 n1Tl 
ji 
ji 2 , := 
1 exp . ji := 
CDFwji 0 , 
cumulative density function 
:= .. 
PFdata := QTlji 
CDFdata 
CDFw1 
:= 
CDFdata 2 . . 
CDFw 
ANL-EBS-NU-000008 REV 00 
CDFTlji
PFdata ji 
. 
 
Seepage Rate (Tptpll) data and fit 
0.01 
PFdata 
Seepage Rate (m3/yr) 
n1Tl 
QTl i i
. . 
.
r ln QTl . 
..
0 i 
n1Tl 
. 
QTli 
0 i 
.. 
1
 
.. 
.. 
 
= 
= 
QTli 
...  
.. 
n1Tl 
 1.464 10 1 . 
... 
= 
.
..... 
 ... 
.... 
.. 
r 
.
.
1 .10 3 
n1Tl 
1 
.
...
. . ln QTli 
1
r 
.. 
..
n1Tl 
.
. 
= 
. 
0 i 
.
... 
.. 
~ 
0.1 
... 
...  
IV-28 of IV-80 
r , , 0.1, 4  = 0.468 
......... 
......... 
10 1 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
IV.3 SEEPAGE ANALYSIS FOR UPPER INFILTRATION RATE IN THE 
LITHOPHYSAL ZONE 
The following section presents the Mathcad analysis for the upper seepage infiltration rate of the 
glacial transition climate in the lithophysal zone. The seepage information used in this analysis 
was obtained from Abstraction of Drift Seepage (BSC 2003 [DIRS 165564]). The information 
contained in the section has been abstracted from the seepage glac upper Tptpll driftcollapse 
report.mcd Mathcad file of Attachment VII. 
February 2004 IV-29 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
:= := 
:= := := := 
Seepage rate and seepage fraction calculation followed Abstraction of Drift Seepage 
(BSC 2003 [DIRS 165564], Section 6.7). 
rnd( ) 1.0 RDi 1 - , 12 
rnd( ) 1.0 := 
rnd( ) 1.0 
:= 
rnd( ) 1.0 RDi 1 - , 10 
rnd( ) 1.0 RDi 1 - , 11 
Latin Hypercube Sampling Routine to Generate Random Numbers 
n := 20000 
:= := rnd( ) 1.0 RDi 1 - , 9 rnd( ) 1.0 RDi 1 - , 6 RDi 1 - , 3 
rnd( ) 1.0 RDi 1 - , 7 
rnd( ) 1.0 RDi 1 - , 8 
Sample Size: 
i 1 := ..n 
RDi 1 - , 0 
:= RDi 1 - , 1 
RDi 1 - , 2 
i := 
rnd( ) 1.0 RDi 1 - , 4 
rnd( ) 1.0 RDi 1 - , 5 
RKs are matrixes in which the first column contain a permutation on the integers on the 
interval [1,n]. 
RK1:= csort := := := ( ) , ( ) , ( ) , ( ) RD 1 , RK4 csort RD 4 RK7 csort RD 7 RK10 csort RD 10 
RK2:= csort := := := ( ) , ( ) , ( ) , ( ) RD 2 , RK5 csort RD 5 RK8 csort RD 8 RK11 csort RD 11 
RK3:= csort := := := ( ) , ( ) , ( ) , ( ) RD 3 , RK6 csort RD 6 RK9 csort RD 9 RK12 csort RD 12 
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n 
February 2004 
Define sets of random values. Each random value is selected within one of the 
equiprobable n intervals that partition [0,1], one set for each random variable. 
runif , ( ) n 0, 1 < > RK7 0 < > 
1 - 
X 6 := runif , ( ) n 0, 1 < > RK4 0 < > 
1 - 
X 3 := X 0 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK8 0 < > 
1 - 
X 7 := runif , ( ) n 0, 1 < > RK5 0 < > 
1 - 
X 4 := 
< > RK1 0 < > 
1 - 
X 1 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK9 0 < > 
1 - 
X 8 := runif , ( ) n 0, 1 < > RK6 0 < > 
1 - 
X 5 := 
< > RK2 0 < > 
1 - 
X 2 := 
+
n
n 
+
n 
X := 
+ ( ) , 10 < > RK110 < > 
1 - runif n 0, 1 
n 
IV-30 of IV-80 
< > RK3 0 < > 
1 - 
< > ( ) , RK100 < > 
1 - + runif n 0, 1 
X 9 := 
X := 
+ ( ) , 11 < > RK120 < > 
1 - runif n 0, 1 
n 
i 0 - := ..n 1 
ANL-EBS-NU-000008 REV 00 
:= rnd( ) 1.0
Screening Analysis for Criticality Features, Events, and Processes for License Application 
Capillary Strength 1/  in (Pa) 
1lb 
Sampling from spatial variability to obtain the 1/  value 
.1 := 0  .1 
i 
Sample from uncertainty triangular distribution to obtain .1/ 
Determine which equation to use: 
if Random Number < RN . 1 then use Equation 1 ( . 1eq1). 
if Random Number > RN . 1 then use Equation 2 ( . 1eq2). 
qunif( ) Xi 0 , ,  1lb, 1ub 1/ value 1 := 
RN.1 
eq1i 
.1
eq2i 
.1
i .1
i 
T1 := 
Permeability k in Tptpll Unit (in log 10) 
kTl := 
Overall Capillary Strength 1/  + .1/ 
.. 
i 1
0.47 kTl := 
kTli 
mean kTl 
Stdev kTl 
Permeability .k in Tptpll Unit (in log 10) 
( ) = .11.5 
( ) = 0.47 
:= .k 
ln( ) qlnorm( ) X ,  kTl, 
Uncertainty follows a triangular distribution := .k .k 
ANL-EBS-NU-000008 REV 00 
:= := Spatial variability follows a uniform distribution 1 := 591 780  402 1ub 
:= Uncertainty follows a triangular distribution .105 .1 := 105 u l 
:=
:= l 
:= 
:=
:= 
( ) .1 . .1l 
2 
( ) .1u . .1l .( ) .1 . .1l 
Xi 1 , l u l 
.1 . 
.1 + 
u 
. . ( ) .1 . .1 .( ) 1 .1  . 
u l i 1 , ( ) 1 X . . . ( ) .1u . .1 .( ) 1 . .1 
i i 
.. 
+ .1
Mean of lognormal distribution .11.5 
Standard deviation of lognormal distribution 
if( ) Xi 1 ,  RN.1, . .1eq1 , 1eq2 .1/ value 
1/ value i 
kTl i 2 , 
:= 0.92 .0.92 0 Tlu Tl Tll 
February 2004 IV-31 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Sample from uncertainty triangular distribution to obtain .k 
Determine which equation to use: 
if Random Number < RN .kTl then use Equation 1 ( .kTleq1). 
if Random Number > RN .kTl then use Equation 2 ( .kTleq2). 
RN.kTl 
:= 
. := 
:= 
.kTleq1i 
.kTleq2i 
Tli 
:= 
i 
:= 
TkTl := 
i 
:= ( ) .. 
.k 
T1kTl 
ffi 
( ) .kTl . .kTll 
2 
( ) .kTlu . .kTll .( ) .kTl . .kTll 
+ Xi 3 , Tlu Tll .k
Tlu .k 
if Xi 3 , 
. 
. . . ( ) .k . .k .( ) kTl .kTll 
i 3 , ( ) 1 X . . . ( ) .kTlu . .k .( ) k . .k 
 RN.kTl .k value eq2i eq1i 
k value i 
11.434 
Tl Tlu Tll 
.. 
.. 
 .14, .14,T1kTl 
102.3 x . 
.. 
IV-32 of IV-80 
, . .kTl , kTl ( ) 
Overall Permeability k + .k 
kTl Tli i 
Permeability must lie between -14 and -10 (bounds of SMPA simulations) 
, if T1kTl 
35.66 x2 
+ 
x4 
.k
 .10, .10 if T1kTl i 
5.4998x3 . + . 
.. 
.. 
Flow Focusing Factor (DTN: LB0104AMRU0185.012 [DIRS 163906]) 
f x ( ) := .0.3137 
root f x . ( ) X .100 , x, 0, 6 i 5 , 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
i 
. + . 
Tll 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Percolation Flux (mm/yr) 
The percolation flux used here is for the glacial transition period only. The percolation flux is 
based on sampling from the upper bound TSPA repository location only 
(DTN: LB0310AMRU0120.002 [DIRS 166116]) 
nnn .. := 0 468 
Upper Bound Percolation Flux 
PF1 := u 0 
40 
36.19 
35.73 
27.83 
30.74 
40.03 
31.86 
57.08 
18.33 
27.91 
0
1
2
3
4
5
6
7
8
9
PF1u , nnn 0 
PFtunnn 
:= 
Z ( ) ( ) , 0 := round runif n 0, 468 
Zi 0 , 
, := q 
. . 
PF := PFt i 
q1pffi 
:= 
pffi 
u( ) 
Adjusted Percolation Flux 
Multiply the flow-focusing factor by the percolation flux, which will be used to obtain the 
seepage rate, seepage fraction, and seepage percentage 
PFi ffi . 
Percolation Flux must lie between 1 and 1000 mm/yr (bounds of SMPA simulations) 
if q1 1  1 , if q1pffi pffi 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
 1000, 1000, q1pffi
.. 
.. 
February 2004 IV-33 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seepage Information from SMPA analysis 
(DTN: LB0307SEEPDRCL.002 [DIRS 164337]) 
m:= 2549 Data points 
SMPAdata := 
0 1 2 3 4 5 6 
-14 0 
-14 1 
-14 2 
-14 3 
-14 4 
-14 5 
-14 6 
-14 7 
-14 8 
-14 9 
-14 10 
-14 11 
-14 12 
-14 13 
-14 14 
-14 15 
Develop routine to select correct mean seepage, seepage standard deviation, seepage percent, 
and seepage percent standard deviation based on sampled value of 1/ , k, percolation flux 
nz := 16 ny := 9 nx := 14 
ii .. := 0 nx 
xi := ii ii ii x 
, 
:= SMPAdataii 2 
jj .. := 0 ny 
:= yj y . := 100 jj + 100 jj jj 
kk .. := 0 nz 
:= := . . 14 + kk 0.25 kk zkkk zkk 
ANL-EBS-NU-000008 REV 00 
1 100 
5 100 
10 100 
20 100 
50 100 
100 100 
200 100 
300 100 
400 100 
500 100 
600 100 
700 100 
800 100 
900 100 
1000 100 
1 200 
jj 
IV-34 of IV-80 
SMPAdata
<0> is permeability value log(k [m^2]) 
SMPAdata
<1> is capillary strength 1/alpha [Pa] 
SMPAdata
<2> is local percolation flux (mm/yr) 
SMPAdata
<3> is Mean Seepage [kg/yr/WP] 
SMPAdata
<4> is Std. Dev. Seepage [kg/yr/WP] 
SMPAdata
<5> is Mean Seepage [%] 
SMPAdata
<6> is Std. Dev. Seepage [%] 
5.47 56.44 
27.5 282.63 
55.13 566.16 
109.85 1135.12 
272.25 2849.95 
535.98 5726.78 
1064.22 11523.63 
1583.08 17369.22 
2086.65 23241.94 
2552.38 29154.54 
2992.46 35097.8 
3411.36 41099.26 
3860.77 47084.03 
4145.2 53190.45 
4520.61 59206.88 
5.44 55.25 
9.75 100.6 
9.8 100.76 
9.83 100.92 
9.79 101.17 
9.71 101.6 
9.55 102.08 
9.49 102.71 
9.41 103.2 
9.3 103.57 
9.1 103.94 
8.89 104.27 
8.69 104.66 
8.6 104.91 
8.21 105.35 
8.06 105.54 
9.69 98.48 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
sms2 := 
i 
smsd2 := 
+ 
sms3 := 
i 
smsd3 := 
+ 
loc represents the location within the matrix of which value to pick for the interpolation process 
loc1i 
loc 
loc 
:= 
sms1 := 
i 
smsd1 := SMPAdataloc i 
spms1 := SMPAdataloc i 
spsd1i 
:= SMPAdataloc1 
dataloc2 
spms2 := SMPAdataloc 
spsd2 := SMPAdataloc i 
dataloc3 
spms3 := SMPAdataloc 
spsd3 := SMPAdataloc i 
loc
2i 
:=
i
i 
3i 
4i 
floor linterp z zk , TkTli 
+
SMPAdataloc1i
1i 
1i 
i 
kTli 
SMPAdataloc2i 
SMPA 
i 
2i 
2i 
SMPAdataloc3i 
SMPA 
i 
3 
:=
i
i i 
, 
4 , 
5 ,
6 , 
4 , 
5 ,
6 , 3i 
, 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
floor linterp z zk , T 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , ,TkTli 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , T 
i
4i 
4i 
, 
3 , 
4 , 
5 ,
6 , 
:=
i 
i i 
+ . ( ) , + ( ) ny 1 floor linterp y yj,T 
. 
nx 1 + 
1\ 
, + . 
, 1\ 
. . 
nx 1 + 
. 
.( ) , + ( ) ny 1 floor linterp y yj T 
. 
nx 1 + ( ) + , + ( ) ny 1 ceil linterp y yj,T 
. 
nx 1 + 
1\
i 
.
. 
. 
. 
i 
( ) . ... 
. 
nx 1 + 
i 
... 
February 2004 
sms4 := 
i 
smsd4 := SMPAdataloc 
spms4 := SMPAdataloc i 
spsd4 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
.. 
..
SMPAdataloc4 
..
. 
. 
.. 
.. 
..
.. 
.. 
.. 
floor linterp x xi , , qpffi 
.. 
.. 
..
.. 
.. 
.. 
. 
. 
. 
. 
. 
. 
. 
. 
kTli 
+ 
.. 
.. 
4 
.. 
.. 
.. 
.. 
1\ 
i
. 
. 
. 
( )... 
. 
nx 1 + 
. 
nx 1 + 
.. 
+ 
. 
( ) , + ( ) ny 1 ceil linterp y yj T . 
.. 
.. 
.. 
.. 
IV-35 of IV-80 
. 
( ) 
( )... 
. 
nx 1 + 
.
.
Screening Analysis for Criticality Features, Events, and Processes for License Application 
loc ceil linterp z zk , T 5i kTli 
5i
5i 
5i 
SMPA 
:= 
sms5 := SMPAdataloc i 
smsd5 := SMPAdataloc i 
spms5 := SMPAdataloc i 
spsd5i 
. 
+ 
:= dataloc5i 
loc6i kTli 
.. 
.. 
.. 
i 
+ 
sms6 := 
i 
smsd6 := 
:=
i
i 
6 
loc7i 
.. 
.. 
.. 
.. 
dataloc6 
dataloc7i 
.. 
+ .. 
.. 
.  
sms7 := 
i 
smsd7 := 
+ 
:=
i
i 
7 
.. 
..
ceil linterp x xi , , qpffi 
SMPA 
SMPAdataloc6 
spms6 := SMPAdataloc 
spsd6 := SMPAdataloc i 
ceil linterp z zk , ,TkTli 
ceil linterp x xi , , qpffi 
SMPA 
SMPAdataloc7 
spms7 := SMPAdataloc 
spsd7 := SMPAdataloc i 
ceil linterp z zk , T 
,
3 ,
i 
6i 
6 , 
i
3 ,
i 
7i 
6 , 
i 
loc8i kTli 
.. 
.. 
.. 
.. 
.... 
.. 
nx 1 + 
floor linterp x xi , , qpffi
.. 
.. 
.. 
nx 1 + 
.. 
. 
. 
.. 
nx 1 + 
.. 
nx 1 + ..
.. 
. 
i 
, 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
ceil linterp z zk , T
4 , 
5 , 
4 , 
5 , 
4 , 
5 , 
.. 
.. 
8 
:=
i 
i 
,
3 , 
8i 
8i 
6 , 
i 
.. 
..
SMPAdataloc8 
sms8 := 
i 
smsd8 := SMPAdataloc 
spms8 := SMPAdataloc i 
spsd8 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
. + . ( )( ) ny 1 + floor linterp y yj , ,T .. 
.. 
. 
.. 
. 
i
.. 
nx 1 + 
. + . . 
( )... 
.. 
nx 1 + 
... 
( ) 
.. 
.. 
.. 
.. 
1 
( ) + ny 1 ( )+ , floor linterp y yj,T1i 
( ) , + ( ) ny 1 ceil linterp y yj,T1i 
i
.. 
.. 
.. 
( ) , + ( ) ny 1 ceil linterp y yj T + 
.. 
nx 1 + 
( ) 
.. 
nx 1 + 
.. 
.. 
. 
... 
( )... , . 1 ..
.. 
.. 
IV-36 of IV-80 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Develop the upper and lower bound for permeability (k) for Tptpll Unit 
qq := .1.TkTl i i 
mantissa x . := ( ) x floor qq ( ) 
( ) i tt := floor qqi 
rr , := round( ) mantissa qq ( ) 2 
yy1 := i i , i if , < ( ) rr  0.25, 0 if( ) 0.25 rri  0.5, 0.25 rr 
if , < ( ) yy1  0.5, yy1 if( ) 0.5 yy1  0.75, 0.5, 0.75 zz1 := i i i i 
1 . . tt i i TkTl2 := 
i 
yy2 := 
( ) + zz1 
if , < ( ) rr  0.25, 0.25 if( ) 0.25 rri  0.5, 0.5 rri , i i 
zz2 := i i i i 
:= T 
if , < ( ) yy2  0.5, yy2 if( ) 0.5 yy2  0.75, 0.75, 1 
.1.( ) tt + zz2i i kTl1i 
Develop the upper and lower bound for capillary strength (1/ ). 
hh1 := floor i 
T1i 
100 
... 
.
. 
100 ( ) hh1 . i T11 := 
i 
:= ceil hh2i 
T1i 
100 
.
. 
( ) hh2 .100 i T12 := 
i 
... 
.. 
Develop the upper and lower bound of the randomly generated 1/ , k, and adjusted percolation 
flux 
February 2004 
Lower Bound value adjusted percolation flux (q pff) 
:= if q aaa1 5  1 , , qpff 1 , <  1 , if 1 qpff pff i i i i 
.. 
.. 
..
,  < , , 10, 5 aaa1 if aaa1 5  aaa1 if 5 aaa1 i bbb1 := i i i i ( ) ( ) 
if , < ( ) bbb1  10, bbb1 if( ) 10 bbb1  20, 10, bbb1 ccc1 := i i i i i 
ddd1 := i i i i i if , < ( ) ccc1  20, ccc1 if( ) 20 ccc1  50, 20, ccc1 
if , < ( ) ddd1  50, ddd1 if( ) 50 ddd1  100, 50, ddd1 eee1 := i i i i i 
fff1 := i if , < ( ) eee1  100, eee1 if( ) 100 eee1  200, 100, eee1i i i i 
IV-37 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ggg1 := i i i i i if , < ( ) fff1  200, fff1 if( ) 200 fff1  300, 200, fff1 
if , < ( ) ggg1  300, ggg1 if( ) 300 ggg1  400, 300, ggg1 hhh1 := i i i i i 
iii1 := i i i i i if , < ( ) hhh1  400, hhh1 if( ) 400 hhh1  500, 400, hhh1 
if , < ( ) iii1  500 iii1 if( ) 500 iii1  600, 500 iii1 jjj1 := i i , i , i i 
kkk1 := i i , i , i i if , < ( ) jjj1  600 jjj1 if( ) 600 jjj1  700, 600 jjj1 
if , < ( ) kkk1  700, kkk1 if( ) 700 kkk1  800, 700, kkk1 mmm1:= i i i i i 
:= q if , < ( ) mmm1  800,mmm1 if( ) 800 mmm1  900, 800, 900 i i i pff1i 
i 
i 
i 
i 
. T T kTl1i kTli 
:= vTkTli . T T 
Upper Bound value adjusted percolation flux (q pff) 
, 1 <  5 , if 1 q if q aaa2 := i 5  5 , , qpffi 
.. 
.. 
.. 
.. 
pffi pffi 
bbb2 := i i i i i if , < ( ) aaa2  5, aaa2 if( ) 5 aaa2  10, 10, aaa2 
if , < ( ) bbb2  10, bbb2 if( ) 10 bbb2  20, 20, bbb2 ccc2 := i i i i i 
ddd2 := i i i i i if , < ( ) ccc2  20, ccc2 if( ) 20 ccc2  50, 50, ccc2 
if , < ( ) ddd2  50, ddd2 if( ) 50 ddd2  100, 100, ddd2 eee2 := i i i i 
if , < ( ) eee2  100, eee2 if( ) 100 eee2  200, 200, eee2 fff2 := i i i i i 
ggg2 := i i i i i if , < ( ) fff2  200, fff2 if( ) 200 fff2  300, 300, fff2 
if , < ( ) ggg2  300, ggg2 if( ) 300 ggg2  400, 400, ggg2 hhh2 := i i i i 
iii2 := i i i i if , < ( ) hhh2  400, hhh2 if( ) 400 hhh2  500, 500, hhh2 
if , < ( ) iii2  500 iii2 if( ) 500 iii2  600, 600 iii2 jjj2 := i i , i , i i 
kkk2 := i i , i , i i if , < ( ) jjj2  600 jjj2 if( ) 600 jjj2  700, 700 jjj2 
if , < ( ) kkk2  700, kkk2 if( ) 700 kkk2  800, 800, kkk2 mmm2:= i i i i 
if , < ( ) mmm2  800,mmm2 if( ) 800 mmm2  900, 900, 1000 i i i qpff2 := 
i 
. T T1 11i i 
. T T 11i 12i 
Solve for seepage rate (Tptpll Unit) 
. q qpff pff1i i 
:= := t u qpff . T1i i q q pff1i pff2i kTl1i kTl2i 
February 2004 IV-38 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
spfluxTlm 1 u . T1\ 1 v . TkTl i i 
1 t . qpffi.
. t 1 u . T1\i qpffi 
. . u 
i 
.. 
tqpffi 
. . u 
T1\i
T1\i 
. . v 
1 t . qpffi 
1 t . qpffi 
...... 
.. 
.. 
.. 
.
1 vTkTl 
1 v . TkTl 
1 vTkTl 
i
v . + . t 
TkTli
ms6 1 uT1\ 
. 
. 
.. 
.
. 
. . 
 . 
1 u . T1\ 
i
v . + . u .s 
. 
. 
.. 
.. 
T1\i
T1\
.. 
.. 
i 
.
.
.
. + u 
qpffi 
tqpff 
. 
i 
1 tqpffi i 
. 
. 
...... 
:= . 1 vTkTl 1 t . qpff . 
:= 
spfluxTlsdi i 
.. 
.. 
+
+
+
+ 
. + . t 
. 
.. 
. 
.. 
. 
.. 
.... 
.
.. 
. 
TkTli 
TkTl 
. . 
. 
. 
. 
msd3i 
msd4 i 
. . 
..
1 uT1\ 
uT1\i 
. uT1\ 
.. 
. 
1 v .
. 
.. 
.. 
1 uT1\i
.
. 
i 
1 u . 
........ 
.. 
.
TkTli
ms7 
.s v 
TkTli 
TkTl 
1 v . TkTl 
.s 
. 
i
1 v 
. v . tqpff
. 
. 
qpffi 
tqpffi
i 
.. 
....
1 uT1\ 
.
v
s 
. 
.
T1\i 
i
v 
. 
. 
u . . 
.. 
.... 
.. 
. 
. 
T1\i 
. . 
TkTli
msd6 
.
.s
. . 
.
msd2i 
s
... 
msd8i 
. 
. 
i 
.. 
.
TkTli
msd7 
s v 
... 
. 
. 
. 
uT1\ 
. 
1 t . qpffi 
1 t . qpffi 
1 t 
tqpff 
. 
i 
.. 
..
TkTli 
TkTl 
+
+
+
+
+
+ 
...... 
.. 
i 
.. 
. 
.. 
. 
. 
0, .0.00001 QTl1stdli 
, .. 
.. 
..
qpffi 
+ QTlstd i 
. 
. 
.. 
. 
. 
QTl1stdu i 
0.1, 0,QTl1spmi 
0 . 0 , if QT2perci 
. 
.. 
Calculate mean seepage for Tptpll Unit 
1.7321spfluxTlsdi 
QTl1stdli 
:= 
QTl1stdu i 
, QTlstdli 
, 
i 
QTlstdli 
i 
. 
. 
1.7321.spfluxTlsdi 
if QTl1spmi 
QTl2spm 100 
i 
:= 
QT2perc := 
. 
28.05.2 
.
qpffi
. 
, :=
:=
if QTl1stdli 
qunif Xi 1 , 0 
if QT2perci 
:= 
QTlstd :=
:= spfluxTlm QTl1spmi 
QTl2spmi 
i 
QT3perci 
Equation to calculate seepage percent based on seepage rate (see 
SMPA data table) from DTN: LB0310AMRU0120.002 [DIRS 166116] 
. 100, 100,QT2perci 
. 
.. .. 
ANL-EBS-NU-000008 REV 00 
sms1 ... 
i i 
... sms2i i 
... .s 
. ... s 
ms3i
ms4i i 
.sms5i 
.s 
i 
... 
i
ms8i 
... 
i i.
s ... 
i 
... 
... 
i 
... 
... 
msd5i 
. ... s 
... 
. msd1 s
i 
.. 
i 
. 
. 
Check seepage percent; if above 
100 percent, then recalculate 
seepage back to 100 percent. 
February 2004 
. 
IV-39 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Mean Seepage Rate (kg/yr per WP) QT3perc QTlspri i
.qpffi
. 
100 
2.28.05 := 
mean QTlspr 
0 
0
0 
0.101 
0.101 
( )= 486.156 
Determine the seepage fraction for Tptpll Unit within the repository and then fit the output data 
to distribution 
Seepage fraction represents the non-zero seepage rates based on the LHS sampling of all of the 
parameters 
spr 
7263 
7264 
7265 
7266 
7267 0.102 
spfrcTl = 0.637 Seepage fraction (i.e., waste package locations that 
can see seepage) 
:= 
ab 
:= 
sort( ) QTl = 
nTl := 7264
( ) n 1 . .nTl 
spfrcTl := 
n 
( ) n 1 . .( ) n + 1 Tl n1Tl 
:= sort QTlspr 
:= 
Q11Tl 
Tl Q1Tl 
Q2Tl 
QTl 
( ) 
reverse( ) Q11 
ab .. := 0 n1Tl 
:= 
998 
Q1Tlab 
1 ... 
. 
. 
Tl sort( ) Q2 
mean Q . Tl 998 ( )= 763.497 
( . + ab 1) 0.375 := CDFTlab Tl 
( ) n1 + 1 . 0.25 
Fit the seepage rates to a Weibull distribution. 
The following equations are from What Every Engineer Should Know About Reliability and Risk 
Analysis (Modarres 1993 [DIRS 104667], p. 109). 
n1Tl  = 1.273 104 
ANL-EBS-NU-000008 REV 00 February 2004 IV-40 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
 := root 
........  
........  
n1Tl
0 
=
i 
:=  
.....  
CDFw1ji 
...  
1 
0.75 
0.5 
0.25
0 
1 .10 4 
Plot of raw data versus Weibull distribution 
ji 0 n1Tl 
ji 
ji 2 , := 
1 exp . ji := 
CDFwji 0 , 
.
cumulative density function 
:= .. 
PFdata := QTlji 
CDFdata 
CDFw1 
:= 
CDFdata 2 . . 
CDFw 
ANL-EBS-NU-000008 REV 00 
CDFTlji
PFdata ji 
. 
 
Seepage Rate (Tptpll) data and fit 
0.01 
PFdata 
Seepage Rate (m3/yr) 
n1Tl 
QTl i i
. . 
.
r ln QTl . 
..
0 i 
n1Tl 
. 
QTli 
0 i 
.. 
1
 
i 
.. 
.. 
 
= 
= 
QTl 
...  
.. 
n1Tl 
 ~ = 3.83 10 1 . 
... 
.
..... 
 ... 
.... 
.. 
r 
.
.
1 .10 3 
n1Tl 
1 
.
...
. . ln QTli 
1
r 
.. 
..
n1Tl 
.
. 
= 
. 
0 i 
.
... 
.. 
0.1 
... 
...  
IV-41 of IV-80 
r , , 0.1, 4  = 0.494 
......... 
......... 
10 1 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
IV.4 SEEPAGE ANALYSIS FOR LOWER INFILTRATION RATE IN THE 
NONLITHOPHYSAL ZONE 
The following section presents the Mathcad analysis for the lower seepage infiltration rate of the 
glacial transition climate in the nonlithophysal zone. The seepage information used in this 
analysis was obtained from Abstraction of Drift Seepage (BSC 2003 [DIRS 165564]). The 
information contained in the section has been abstracted from the seepage glac lower Tptpmn 
x1.2 report.mcd Mathcad file of Attachment VII. 
February 2004 IV-42 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
:= := 
:= := := := 
Seepage rate and seepage fraction calculation followed Abstraction of Drift Seepage 
(BSC 2003 [DIRS 165564], Section 6.7). 
rnd( ) 1.0 RDi 1 - , 12 
rnd( ) 1.0 := 
rnd( ) 1.0 
:= 
rnd( ) 1.0 RDi 1 - , 10 
rnd( ) 1.0 RDi 1 - , 11 
Latin Hypercube Sampling Routine to Generate Random Numbers 
n := 20000 
:= := rnd( ) 1.0 RDi 1 - , 9 rnd( ) 1.0 RDi 1 - , 6 RDi 1 - , 3 
rnd( ) 1.0 RDi 1 - , 7 
rnd( ) 1.0 RDi 1 - , 8 
Sample Size: 
i 1 := ..n 
RDi 1 - , 0 
:= RDi 1 - , 1 
RDi 1 - , 2 
i := 
rnd( ) 1.0 RDi 1 - , 4 
rnd( ) 1.0 RDi 1 - , 5 
RKs are matrixes in which the first column contain a permutation on the integers on the 
interval [1,n]. 
RK1:= csort := := := ( ) , ( ) , ( ) , ( ) RD 1 , RK4 csort RD 4 RK7 csort RD 7 RK10 csort RD 10 
RK2:= csort := := := ( ) , ( ) , ( ) , ( ) RD 2 , RK5 csort RD 5 RK8 csort RD 8 RK11 csort RD 11 
RK3:= csort := := := ( ) , ( ) , ( ) , ( ) RD 3 , RK6 csort RD 6 RK9 csort RD 9 RK12 csort RD 12 
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n 
February 2004 
Define sets of random values. Each random value is selected within one of the 
equiprobable n intervals that partition [0,1], one set for each random variable. 
runif , ( ) n 0, 1 < > RK7 0 < > 
1 - 
X 6 := runif , ( ) n 0, 1 < > RK4 0 < > 
1 - 
X 3 := X 0 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK8 0 < > 
1 - 
X 7 := runif , ( ) n 0, 1 < > RK5 0 < > 
1 - 
X 4 := 
< > RK1 0 < > 
1 - 
X 1 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK9 0 < > 
1 - 
X 8 := runif , ( ) n 0, 1 < > RK6 0 < > 
1 - 
X 5 := 
< > RK2 0 < > 
1 - 
X 2 := 
+
n
n 
+
n 
X := 
+ ( ) , 10 < > RK110 < > 
1 - runif n 0, 1 
n 
IV-43 of IV-80 
< > RK3 0 < > 
1 - 
< > ( ) , RK100 < > 
1 - + runif n 0, 1 
X 9 := 
X := 
+ ( ) , 11 < > RK120 < > 
1 - runif n 0, 1 
n 
i 0 - := ..n 1 
ANL-EBS-NU-000008 REV 00 
:= rnd( ) 1.0
Screening Analysis for Criticality Features, Events, and Processes for License Application 
Capillary Strength 1/  in (Pa) 
:= := Spatial variability follows a uniform distribution 1 := 591 780 1  402 1ub lb 
:= Uncertainty follows a triangular distribution .105 .1 := 105 u l .1 := 0  
Sampling from spatial variability to obtain the 1/  value 
.1 
i qunif( ) Xi 0 , ,  1lb, 1ub 1/ value 
Sample from uncertainty triangular distribution to obtain .1/ 
Determine which equation to use: 
if Random Number < RN . 1 then use Equation 1 ( . 1eq1). 
if Random Number > RN . 1 then use Equation 2 ( . 1eq2). 
1 := 
RN :=
:= l 
.1 
eq1i 
:= 
.1
eq2i 
:= 
.1
i .1
i 
.. 
Overall Capillary Strength 1/  + .1/ 
i 1 
0.34 
:= 
( ) .1 . .1l 
2 
( ) .1u . .1l .( ) .1 . .1l 
Xi 1 , l u l 
.1 . 
.1 + 
u 
. . ( ) .1 . .1 .( ) 1 .1  . 
u l i 1 , ( ) 1 X . . . ( ) .1u . .1 .( ) 1 . .1 
i i 
.. 
+ .1 
.12.2 Mean of lognormal distribution 
Standard deviation of lognormal distribution 
if( ) Xi 1 ,  RN.1, . .1eq1 , 1eq2 .1/ value 
1/ value i 
kTn i 2 , 
T1 := 
Permeability k in Tptpmn Unit (in log 10) 
kTn := 
kTn := 
kTni 
.k 
ln( ) qlnorm( ) X ,  kTn, 
mean kTn 
Stdev kTn 
Uncertainty follows a triangular distribution := 0.68 := 0 .kTnu 
( ) = .12.2 
( ) = 0.34 
Permeability .k in Tptpmn Unit (in log 10) 
:= Tnl .0.68 .kTn 
February 2004 IV-44 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Sample from uncertainty triangular distribution to obtain .k 
Determine which equation to use: 
if Random Number < RN .kTn then use Equation 1 ( .kTneq1). 
if Random Number > RN .kTn then use Equation 2 ( .kTneq2). 
RN.kTn 
:= .kTn 
. := .kTn 
:= 
eq1i 
eq2i 
:= 
( ) .kTn . .kTnl 
2 
( ) .kTnu . .kTnl .( ) .kTn . .kTnl 
+ Xi 3 , Tnu Tnl . . . ( ) .k . .k .( ) kTn .kTnl 
i 3 , Tn Tnu Tnl 
.k
Tnu .k 
if Xi 3 , 
Tnl 
Tni 
:= 
TkTn := 
i 
:= .. 
.k 
T1kTni 
ffi 
( ) 1 X . . . ( ) .kTnu . .k .( ) k . .k 
 RN.kTn 
.
, . .kTn , kTn ( ) 
Overall Permeability k + .k 
Tni 
kTni 
Permeability must lie between -14 and -10 (bounds of SMPA simulations) 
, if T1kTn 
35.66 x2 
+ 
x4 
.k
 .10, .10 if T1kTn i 
5.4998x3 . + . 
.. 
.. 
Flow Focusing Factor (DTN: LB0104AMRU0185.012 [DIRS 163906]) 
f x ( ) := .0.3137 
root f x . ( ) X .100 , x, 0, 6 i 5 , 
i 
11.434 102.3 x . + . 
.. 
.. 
( ) 
ANL-EBS-NU-000008 REV 00 
.k value eq2i eq1i 
k value .. 
.. 
.. 
 .14, .14,T1kTni 
.
IV-45 of IV-80 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Percolation Flux (mm/yr) 
The percolation flux used here is for the glacial transition period only. The percolation flux is 
based on sampling from the lower bound TSPA repository location only 
(DTN: LB0310AMRU0120.002 [DIRS 166116]) 
nnn .. := 0 468 
Lower Bound Percolation Flux 
PF1 := l 0
3.68 
2.65 
2.41 
2.13 
2.41 
2.4 
2.12 
2.76 
1.4 
2.21 
0
1
2
3
4
5
6
7
8
9
PF1l PFtl nnn 0 , nnn 
:= 
Z ( ) ( ) , 0 := round runif n 0, 468 
l Zi 0 , 
, := q 
. . 
PF := PFt i 
q1pffi 
:= 
pffi 
( ) 
Adjusted Percolation Flux 
Multiply the flow-focusing factor by the percolation flux, which will be used to obtain the 
seepage rate, seepage fraction, and seepage percentage. 
PFi ffi . 
Percolation Flux must lie between 1 and 1000 mm/yr (bounds of SMPA simulations) 
1  1 , if q1 if q1pffi 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
 1000, 1000, q1pffi pffi 
.. 
.. 
February 2004 IV-46 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seepage Information from SMPA analysis 
(DTN: LB0304SMDCREV2.002 [DIRS 163687]) 
m:= 2549 Data points 
SMPAdata := 
0 1 2 3 4 5 6 
-14 0 
-14 1 
-14 2 
-14 3 
-14 4 
-14 5 
-14 6 
-14 7 
-14 8 
-14 9 
-14 10 
-14 11 
-14 12 
-14 13 
-14 14 
-14 15 
Develop routine to select correct mean seepage, seepage standard deviation, seepage percent, 
and seepage percent standard deviation based on sampled value of 1/ , k, percolation flux. 
nz := 16 ny := 9 nx := 14 
ii .. := 0 nx 
xi := ii ii ii x 
, 
:= SMPAdataii 2 
jj .. := 0 ny 
:= yj jj y . := 100 jj + 100 jj jj 
kk .. := 0 nz 
:= := . . 14 + kk 0.25 kk zkkk zkk 
ANL-EBS-NU-000008 REV 00 
SMPAdata
<0> is permeability value log(k [m^2]) 
SMPAdata
<1> is capillary strength 1/alpha [Pa] 
SMPAdata
<2> is local percolation flux (mm/yr) 
SMPAdata
<3> is Mean Seepage [kg/yr/WP] 
SMPAdata
<4> is Std. Dev. Seepage [kg/yr/WP] 
SMPAdata
<5> is Mean Seepage [%] 
SMPAdata
<6> is Std. Dev. Seepage [%] 
27.73 1 100 
138.92 5 100 
277.9 10 100 
555.87 20 100 
1391.67 50 100 
2793.55 100 100 
5610 200 100 
8415 300 100 
11220 400 100 
14025 500 100 
16830 600 100 
19635 700 100 
22440 800 100 
25245 900 100 
28050 1000 100 
26.14 1 200 
IV-47 of IV-80 
14.59 98.86 4.09 
14.65 99.05 20.55 
14.68 99.07 41.19 
14.71 99.09 82.54 
14.66 99.23 205.57 
14.5 99.59 406.7 
14 100 785 
14 100 1178 
14 100 1570 
14 100 1963 
14 100 2356 
14 100 2748 
14 100 3141 
14 100 3590 
14 100 3989 
15 93.21 4.21 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
sms2 := 
i 
smsd2 := 
+ 
sms3 := 
i 
smsd3 := 
+ 
loc represents the location within the matrix of which value to pick for the interpolation process 
loc1i 
loc 
loc 
:= 
sms1 := 
i 
smsd1 := SMPAdataloc i 
spms1 := SMPAdataloc i 
spsd1i 
:= SMPAdataloc1 
dataloc2 
spms2 := SMPAdataloc 
spsd2 := SMPAdataloc i 
dataloc3 
spms3 := SMPAdataloc 
spsd3 := SMPAdataloc i 
loc
2i 
:=
i
i 
3i 
4i 
floor linterp z zk , TkTn 
+
SMPAdataloc1i
1i 
1i 
i 
SMPAdataloc2i 
SMPA 
i 
2i 
2i 
SMPAdataloc3i 
SMPA 
i 
3 
:=
i
i i 
, 
4 , 
5 ,
6 , 
4 , 
5 ,
6 , 3i 
, 
i 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
floor linterp z zk , TkTni 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , ,TkTni 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , TkTni 
floor linterp x xi , , qpffi 
i
4i 
4i 
, 
3 , 
4 , 
5 ,
6 , 
:=
i 
i i 
+ . ( ) , + ( ) ny 1 floor linterp y yj,T 
. 
nx 1 + 
1\ 
, + . 
, 
.
. 1\ 
1\
i 
. . 
nx 1 + 
. 
.( ) , + ( ) ny 1 floor linterp y yj T 
. 
nx 1 + ( ) + , + ( ) ny 1 ceil linterp y yj,T 
. 
nx 1 + 
. 
. 
i 
( ) . ... 
. 
nx 1 + 
i 
... 
February 2004 
sms4 := 
i 
smsd4 := SMPAdataloc 
spms4 := SMPAdataloc i 
spsd4 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
.. 
..
SMPAdataloc4 
..
. 
. 
.. 
.. 
..
.. 
.. 
.. 
.. 
.. 
..
.. 
.. 
.. 
+ 
. 
. 
.
. 
. 
. 
. 
. 
.. 
.. 
4 
.. 
.. 
.. 
.. 
1\ 
i
. 
. 
. 
( )... 
. 
nx 1 + 
. 
nx 1 + 
.. 
+ 
. 
( ) , + ( ) ny 1 ceil linterp y yj T . 
.. 
.. 
.. 
.. 
IV-48 of IV-80 
. 
( ) 
( )... 
. 
nx 1 + 
.
.
Screening Analysis for Criticality Features, Events, and Processes for License Application 
loc ceil linterp z zk , TkTn 5i 
5i
5i 
5i 
SMPA 
:= 
sms5 := SMPAdataloc i 
smsd5 := SMPAdataloc i 
spms5 := SMPAdataloc i 
spsd5i 
. 
+ 
:= dataloc5i 
loc6i 
.. 
.. 
.. 
..
.. 
i 
+ 
sms6 := 
i 
smsd6 := 
:=
i
i 
6 
loc7i 
.. 
.. 
.. 
.. 
dataloc6 
dataloc7i 
.. 
+ .. 
.. 
.  
sms7 := 
i 
smsd7 := 
+ 
:=
i
i 
7 
. 
.. 
nx 1 + 
.. 
nx 1 + ..
.. 
.
( ) , + ( ) ny 1 ceil linterp y yj,T1i 
. . 
,
3 ,
i 
6i 
6 , 
i
3 ,
i 
7i 
6 , 
i 
loc 
.. 
.. 
8i 
.. 
.. 
.. 
.. 
SMPA 
SMPAdataloc6 
spms6 := SMPAdataloc 
spsd6 := SMPAdataloc i 
ceil linterp z zk , ,TkTni 
ceil linterp x xi , , qpffi 
SMPA 
SMPAdataloc7 
spms7 := SMPAdataloc 
spsd7 := SMPAdataloc i 
ceil linterp z zk , TkTni 
floor linterp x xi , , qpffi 
.. 
.. 
.. 
nx 1 + 
.. i 
, 
i 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
ceil linterp z zk , TkTni 
ceil linterp x xi , , qpffi 
4 , 
5 , 
4 , 
5 , 
4 , 
5 , 
.. 
.. 
8 
,
3 , 
8i 
8i 
6 , 
i 
:=
i 
i 
.. 
..
SMPAdataloc8 
sms8 := 
i 
smsd8 := SMPAdataloc 
spms8 := SMPAdataloc i 
spsd8 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
. ( )( ) ny 1 + floor linterp y yj , ,T .. 
.. 
. 
+ 
( ) + ny 1 ( )+ , floor linterp y yj,T1 
+ . 
.. 
.. 
.. 
.. 
1 
i 
i
.. 
.. 
.. 
.. 
nx 1 + 
( ) , + ( ) ny 1 ceil linterp y yj T + 
.. 
.. 
. , 1 
.. 
nx 1 + 
( ) 
.. 
nx 1 + .. 
. 
.. 
.. 
IV-49 of IV-80 
. .. 
. 
i
.. 
nx 1 + 
.
( )... 
.. 
nx 1 + 
... 
( )
...
... 
( ) 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Develop the upper and lower bound of the randomly generated 1/ , k, and adjusted percolation 
flux 
Develop the upper and lower bound for permeability (k) for Tptpmn Unit 
i qq := .1.TkTni 
mantissa x . := ( ) x floor qq ( ) 
( ) i tt := floor qqi 
rr , := round( ) mantissa qq ( ) 2 
yy1 := i i , i if , < ( ) rr  0.25, 0 if( ) 0.25 rri  0.5, 0.25 rr 
if , < ( ) yy1  0.5, yy1 if( ) 0.5 yy1  0.75, 0.5, 0.75 zz1 := i i i i 
1 . . tt i i TkTn2 := 
i 
yy2 := 
( ) + zz1 
if , < ( ) rr  0.25, 0.25 if( ) 0.25 rri  0.5, 0.5 rri , i i 
zz2 := i i i i 
:= T 
if , < ( ) yy2  0.5, yy2 if( ) 0.5 yy2  0.75, 0.75, 1 
.1.( ) tt + zz2i i kTn1i 
Develop the upper and lower bound for capillary strength (1/ ) 
hh1 := floor i 
... 
.
. 
T1i 
100 
100 ( ) hh1 . i T11 := 
i 
:= ceil hh2i 
.
. 
T1i 
100 
T12 :=( ) hh2 .100 i i 
... 
.. 
i i i i 
Lower Bound value adjusted percolation flux (q pff) 
:= if q aaa1 5  1 , , qpff 1 , <  1 , if 1 qpff pff i i i i 
.. 
.. 
..
,  < , , 10, 5 aaa1 if aaa1 5  aaa1 if 5 aaa1 i bbb1 := i i i i ( ) ( ) 
if , < ( ) bbb1  10, bbb1 if( ) 10 bbb1  20, 10, bbb1 ccc1 := i i i i i 
ddd1 := i i i i i if , < ( ) ccc1  20, ccc1 if( ) 20 ccc1  50, 20, ccc1 
if , < ( ) ddd1  50, ddd1 if( ) 50 ddd1  100, 50, ddd1 eee1 := i 
if , < ( ) eee1  100, eee1 if( ) 100 eee1  200, 100, eee1 fff1 := i i i i i 
February 2004 IV-50 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ggg1 := i i i i i if , < ( ) fff1  200, fff1 if( ) 200 fff1  300, 200, fff1 
if , < ( ) ggg1  300, ggg1 if( ) 300 ggg1  400, 300, ggg1 hhh1 := i i i i i 
iii1 := i i i i i if , < ( ) hhh1  400, hhh1 if( ) 400 hhh1  500, 400, hhh1 
if , < ( ) iii1  500 iii1 if( ) 500 iii1  600, 500 iii1 jjj1 := i i , i , i i 
kkk1 := i i , i , i i if , < ( ) jjj1  600 jjj1 if( ) 600 jjj1  700, 600 jjj1 
if , < ( ) kkk1  700, kkk1 if( ) 700 kkk1  800, 700, kkk1 mmm1:= i i i i i 
:= q if , < ( ) mmm1  800,mmm1 if( ) 800 mmm1  900, 800, 900 i i i pff1i 
i 
i 
i 
i 
. T TkTn kTn1i i 
:= vTkTn . i T T 
Upper Bound value adjusted percolation flux (q pff) 
, 1 <  5 , if 1 q if q aaa2 := i 5  5 , , qpffi 
.. 
.. 
.. 
.. 
pffi pffi 
bbb2 := i i i i i if , < ( ) aaa2  5, aaa2 if( ) 5 aaa2  10, 10, aaa2 
if , < ( ) bbb2  10, bbb2 if( ) 10 bbb2  20, 20, bbb2 ccc2 := i i i i i 
ddd2 := i i i i i if , < ( ) ccc2  20, ccc2 if( ) 20 ccc2  50, 50, ccc2 
if , < ( ) ddd2  50, ddd2 if( ) 50 ddd2  100, 100, ddd2 eee2 := i i i i 
if , < ( ) eee2  100, eee2 if( ) 100 eee2  200, 200, eee2 fff2 := i i i i i 
ggg2 := i i i i i if , < ( ) fff2  200, fff2 if( ) 200 fff2  300, 300, fff2 
if , < ( ) ggg2  300, ggg2 if( ) 300 ggg2  400, 400, ggg2 hhh2 := i i i i 
iii2 := i i i i if , < ( ) hhh2  400, hhh2 if( ) 400 hhh2  500, 500, hhh2 
if , < ( ) iii2  500 iii2 if( ) 500 iii2  600, 600 iii2 jjj2 := i i , i , i i 
kkk2 := i i , i , i i if , < ( ) jjj2  600 jjj2 if( ) 600 jjj2  700, 700 jjj2 
if , < ( ) kkk2  700, kkk2 if( ) 700 kkk2  800, 800, kkk2 mmm2:= i i i i 
if , < ( ) mmm2  800,mmm2 if( ) 800 mmm2  900, 900, 1000 i i i qpff2 := 
i 
. T 1 11i i 
. T11i 12i 
Solve for seepage rate (Tptpmn Unit) 
. q q T pff pff1i i 
:= := t u qpff . T1i i q q T pff1i pff2i kTn1i kTn2i 
February 2004 IV-51 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
spfluxTnmi 
1 t . qpffi.
t 1 u . T1\ qpffi 
. u tqpffi 
1 t . qpffi 
1 t . qpffi 
...... 
.. 
+ t 1 uT1\ 
. + u 
+ 
.. 
.. 
T1\i
T1\ 
qpffi 
tqpff 
. 
i 
1 tqpffi 
...... 
:= 1 t . qpff . 
:= 
spfluxTnsd i i.
t 
. 
. 
.. 
. 
.. 
. 
.. 
.... 
qpffi 
1 t . qpffi 
t 
. . 
..
1 uT1\ 
u
u
T1\i 
. u
1 u
i T1\ 
. 
.. 
.. 
.. 
+
+
+
+ 
........ 
+
+ t 
+
+
+
+
+ 
.. 
qpffi 
. 
. 
. 
. 
. 
1 t . qpff 
. 
i 
qpffi 
tqpff 
. 
i 
1 uT1\ 
.. 
.... 
T1\i
T1\ 
.. 
....
1 tqpffi 
...... 
0, .0.00001 QTn1stdl 
.. 
. 
..
.. 
:=
:= 1.2 spfluxTnm 
:= if QTn1spm .. 
Calculate mean seepage for Tptpmn Unit 
.1.7321spfluxTnsd i 
QTn1stdli 
:= 
QTn1stdu i 
QTnstdli 
QTn QTnstd i 
. 
i 
, 
100 
. 
1.7321.spfluxTnsd i 
qpff .28.05 
i 
, 
, stdli 
i 
, :=
:=
if QTn1stdli 
qunif Xi 1 , 0 
QTn2spmi
. 
if QT2perci 
:= 
QTn1spmi 
QTn2spmi 
QT2perc := 
i 
QT3perci 
0 . 0 , if QT2perci 
Equation to calculate seepage percent based on seepage rate (see 
SMPA data table) from DTN: LB0310AMRU0120.002 [DIRS 166116] 
. , 100, 100 QT2perci 
. 
.. 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
1 u . T1\ sms1 ... 
i i 
1 v . TkTni 
. . ... sms2i i 
... . s 
i 
. ... u s 
T1\i
T1\ 
ms3i
ms4 
.
1 vTkTni 
1 v . TkTn 
1 vTkTni i i 
. 
.
. .s v 
. 
TkTni ms5i 
sms6i 
.
.. 
. 
. . 
 . 
1 u . T1\i
v 
i
v . ... .s 
.
i 
.
.s 
.
.
.
. u 
ms7i
ms8i 
. 
. 
. ... 
... 
. msd1 s 
i 
... .s 
1 vTkTni 
1 v . 
. 
. 
TkTni 
.
.. 
.. 
.. 
i 
. 
.. 
.. 
1 uT1\i. 
. .s 
. ... 
... 
i 
.. 
.
TkTni 
TkTni 
1 v .
. . ... 
. 
.
TkTni 
vTkTn 
... 
msd5i 
.. 
.
TkTni 
TkTn 
v
. 
i
1 v 
. v 
. 
. 
. 
.. 
.. 
.
.
.
.
. 
. 
.
msd3i 
msd4 
i 
... s 
. 
.
T1\i 
i
v . 
. 
. 
.. 
..
TkTni 
..
msd6i 
i 
. . 
.
msd2i 
s .
.s
... 
msd8i 
.
u 
.. 
.
TkTni 
. msd7 s 
v 
.
.s . . 
.. 
.. 
i 
.. 
.
TkTni 
.. 
. 
. i
. 
. 
. 
QTn1stdu i 
+ QTnstd i 
Increase the seepage rate by 20 percent to 
account for drift degradation. 
.. . 
0.1, 0,QTn1spmi
. 
. 
. 
IV-52 of IV-80 
Check seepage percent; if above 
100 percent, then recalculate 
seepage back to 100 percent. 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
QTlspri 
mean QTlspr 
:= 
Determine the seepage fraction for Tptpmn Unit within the repository and then fit the output 
data to distribution 
Seepage fraction represents the non-zero seepage rates based on the LHS sampling of all of the 
parameters 
( )= 1.613 
spr 
16913 
16914 
16915 
16916 
16917 
:= 
ab 
:= 
sort( ) QTl = 
nTl := 16914 
spfrcTl := 
n1Tl := 
Q11Tl 
Q1Tl 
Q2Tl 
QTl 
( ) n 1 . .nTl 
n 
( ) n 1 . .( ) n + 1 Tl 
:= sort QTlspr
Tl 
ab .. := 0 n1Tl
( ) 
reverse( ) Q11 
:= 
998 
Q1Tlab 
1 ... 
. 
. 
Tl 
mean Q . Tl 998 
sort( ) Q2 
( )= 10.457 
CDFTl 
( . + ab 1) 0.375 := 
ab Tl 
Fit the seepage rates to a Weibull distribution. 
The following equations are from What Every Engineer Should Know About Reliability and 
Risk Analysis (Modarres 1993 [DIRS 104667], p. 109). 
( ) n1 + 1 . 0.25 
n1Tl  = 3.084 103 
ANL-EBS-NU-000008 REV 00 
Mean Seepage Rate (kg/yr per WP) QT3perci
.qpffi
. 
100 
28.05 
0 
0
0 
0.1 
0.1 
0.1 
spfrcTl = 0.154 Seepage fraction (i.e., waste package locations that can 
see seepage) 
February 2004 IV-53 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
 := root 
........  
........  
n1Tl 
=
i 
 := 
.....  
Plot of raw data versus Weibull distribution 
ji 0 n1Tl 
ji 
ji 2 , := 
1 exp ji := 
CDFwji 0 , 
cumulative density function 
:= .. 
PFdata := QTlji 
CDFdata 
CDFw1 
. 
:= 
CDFdata 2 . . 
CDFw 
ANL-EBS-NU-000008 REV 00 
CDFTlji 
 
. . 
 
.... 
Seepage Rate (Tptpmn) data and fit 
1 .10 3 0.01 
PFdata 
Seepage Rate (m3/yr) 
n1Tl 
QTl i i 
..
0 i 
.
...
. .. 
1
r n1Tl 
. 
QTli 
0 i 
. . 
.
r ln QTl . 
1 
QTli
.. 
.. 
 
= 
...  
0 
~ = 
= 
.. 
n1Tl 
... 
 4.949 10 3 . 
.
..... 
.. 
 
PFdata ji 
.. 
r
.
. 
... 
...  
CDFw1ji 
1 
0.75 
0.5 
0.25
0 
1 .10 4 
n1Tl 
1 . ln QTli 
.. 
.
. n1Tl = 
. 
0 i 
.
... 
.. 
... 
...  
IV-54 of IV-80 
r , , 0.1, 4  = 0.536 
......... 
......... 
0.1 1 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
IV.5 SEEPAGE ANALYSIS FOR MEAN INFILTRATION RATE IN THE 
NONLITHOPHYSAL ZONE 
The following section presents the Mathcad analysis for the mean seepage infiltration rate of the 
glacial transition climate in the nonlithophysal zone. The seepage information used in this 
analysis was obtained from Abstraction of Drift Seepage (BSC 2003 [DIRS 165564]). The 
information contained in the section has been abstracted from the seepage glac mean Tptpmn 
x1.2 report.mcd Mathcad file of Attachment VII. 
February 2004 IV-55 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
:= := := 
:= := := := 
Seepage Rate and Seepage Fraction Calculation using Abstraction of Drift Seepage 
(BSC 2003 [DIRS 165564], Section 6.7) 
rnd( ) 1.0 RDi 1 - , 12 
rnd( ) 1.0 := 
rnd( ) 1.0 
:= 
rnd( ) 1.0 RDi 1 - , 10 
rnd( ) 1.0 RDi 1 - , 11 
Latin Hypercube Sampling Routine to Generate Random Numbers 
n := 20000 
:= := rnd( ) 1.0 RDi 1 - , 9 rnd( ) 1.0 RDi 1 - , 6 RDi 1 - , 3 
rnd( ) 1.0 RDi 1 - , 7 
rnd( ) 1.0 RDi 1 - , 8 
Sample Size: 
i 1 := ..n 
RDi 1 - , 0 
RDi 1 - , 1 
RDi 1 - , 2 
i := 
rnd( ) 1.0 RDi 1 - , 4 
rnd( ) 1.0 RDi 1 - , 5 
RKs are matrixes in which the first column contain a permutation on the integers on 
the interval [1,n] 
RK1:= csort := := := ( ) , ( ) , ( ) , ( ) RD 1 , RK4 csort RD 4 RK7 csort RD 7 RK10 csort RD 10 
RK2:= csort := := := ( ) , ( ) , ( ) , ( ) RD 2 , RK5 csort RD 5 RK8 csort RD 8 RK11 csort RD 11 
RK3:= csort := := := ( ) , ( ) , ( ) , ( ) RD 3 , RK6 csort RD 6 RK9 csort RD 9 RK12 csort RD 12 
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n 
Define sets of random values. Each random value is selected within one of the 
equiprobable n intervals that partition [0,1], one set for each random variable 
runif , ( ) n 0, 1 < > RK7 0 < > 
1 - 
X 6 := runif , ( ) n 0, 1 < > RK4 0 < > 
1 - 
X 3 := X 0 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK8 0 < > 
1 - 
X 7 := runif , ( ) n 0, 1 < > RK5 0 < > 
1 - 
X 4 := 
< > RK1 0 < > 
1 - 
X 1 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK9 0 < > 
1 - 
X 8 := runif , ( ) n 0, 1 < > RK6 0 < > 
1 - 
X 5 := 
< > RK2 0 < > 
1 - 
X 2 := 
+
n
n 
+
n 
X := 
+ ( ) , 10 < > RK110 < > 
1 - runif n 0, 1 
n 
IV-56 of IV-80 
< > RK3 0 < > 
1 - 
< > ( ) , RK100 < > 
1 - + runif n 0, 1 
X 9 := 
X := 
+ ( ) , 11 < > RK120 < > 
1 - runif n 0, 1 
n 
i 0 - := ..n 1 
ANL-EBS-NU-000008 REV 00 
rnd( ) 1.0 := 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Capillary Strength 1/  in (Pa) 
1lb 
Sampling from spatial variability to obtain the 1/  value 
.1 := 0  .1 
i 
Sample from uncertainty triangular distribution to obtain .1/ 
Determine which equation to use: 
if Random Number < RN . 1 then use Equation 1 ( . 1eq1). 
if Random Number > RN . 1 then use Equation 2 ( . 1eq2). 
qunif( ) Xi 0 , ,  1lb, 1ub 1/ value 1 := 
RN.1 
eq1i 
.1 
.1
eq2i 
.1i := 
i 
Overall Capillary Strength 1/  + .1/ 
.. 
i 1 + .1 
.12.2 Mean of lognormal distribution 
Standard deviation of lognormal distribution 0.34 
T1 := 
Permeability k in Tptpmn Unit (in log 10) 
kTn := 
kTn := 
kTni 
mean kTn 
Stdev kTn 
Permeability .k in Tptpmn Unit (in log 10) 
( ) = .12.2 
( ) = 0.34 
:=
ln( ) qlnorm( ) X ,  kTn, 
Uncertainty follows a triangular distribution := .kTnl 
ANL-EBS-NU-000008 REV 00 
:= := Spatial variability follows a uniform distribution 1 := 591 780  402 1ub 
:= Uncertainty follows a triangular distribution .105 .1 := 105 u l 
:=
:= l 
:= 
:= 
( ) .1 . .1l 
2 
( ) .1u . .1l .( ) .1 . .1l 
Xi 1 , l u l 
.1 . 
.1 + 
u 
. . ( ) .1 . .1 .( ) 1 .1  . 
u l i 1 , ( ) 1 X . . . ( ) .1u . .1 .( ) 1 . .1 
.1/ value i i 
.. 
if( ) Xi 1 ,  RN.1, . .1eq1 , 1eq2 
1/ value i 
kTn i 2 , 
:= 0.68 0 .kTnu .0.68 .kTn 
February 2004 IV-57 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Sample from uncertainty triangular distribution to obtain .k 
Determine which equation to use: 
if Random Number < RN .kTn then use Equation 1 ( .kTneq1). 
if Random Number > RN .kTn then use Equation 2 ( .kTneq2). 
RN.kTn 
:= .kTn 
. := .kTn 
:= 
eq1i 
eq2i 
:= 
( ) .kTn . .kTnl 
2 
( ) .kTnu . .kTnl .( ) .kTn . .kTnl 
+ Xi 3 , Tnu Tnl . . . ( ) .k . .k .( ) kTn .kTnl 
i 3 , Tn Tnu Tnl 
.k
Tnu .k 
if Xi 3 , 
Tnl 
Tni 
:= 
TkTn := 
i 
:= ( ) .. 
.k 
T1kTni 
ffi 
( ) 1 X . . . ( ) .kTnu . .k .( ) k . .k 
 RN.kTn 
.
, . .kTn , kTn ( ) 
Overall Permeability k + .k 
Tni 
kTni 
Permeability must lie between -14 and -10 (bounds of SMPA simulations) 
, if T1kTn 
35.66 x2 
+ 
x4 
.k
 .10, .10 if T1kTn i 
5.4998x3 . + . 
.. 
.. 
Flow Focusing Factor (DTN: LB0104AMRU0185.012 [DIRS 163906]) 
f x ( ) := .0.3137 
root f x . ( ) X .100 , x, 0, 6 i 5 , 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
i 
11.434 102.3 x . + . 
.k value eq2i eq1i 
k value .. 
.. 
.. 
 .14, .14,T1kTni 
.
IV-58 of IV-80 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Percolation Flux (mm/yr) 
The percolation flux used here is for the glacial transition period only. The percolation flux is 
based on sampling from the mean bound TSPA repository location only 
(DTN: LB0310AMRU0120.002 [DIRS 166116]) 
nnn .. := 0 468 
Mean Bound Percolation Flux 
PF1 := m 0
15.97 
19.87 
14.2 
7.59 
16.94 
17.76 
10.45 
27.77 
8.95 
16.02 
0
1
2
3
4
5
6
7
8
9
PF1m PFtm nnn 0 , nnn 
:= 
Z ( ) ( ) , 0 := round runif n 0, 468 
m Zi 0 , 
, := q 
. . 
PF := PFt i 
q1pffi 
:= 
pffi 
( ) 
Adjusted Percolation Flux 
Multiply the flow-focusing factor by the percolation flux, which will be used to obtain the 
seepage rate, seepage fraction, and seepage percentage 
PFi ffi . 
Percolation Flux must lie between 1 and 1000 mm/yr (bounds of SMPA simulations) 
1  1 , if q1 if q1pffi 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
 1000, 1000, q1pffi pffi 
.. 
.. 
February 2004 IV-59 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seepage Information from SMPA analysis 
(DTN: LB0304SMDCREV2.002 [DIRS 163687]) 
Data points m:= 2549 
SMPAdata := 
0 1 2 3 4 5 6 
-14 0 
-14 1 
-14 2 
-14 3 
-14 4 
-14 5 
-14 6 
-14 7 
-14 8 
-14 9 
-14 10 
-14 11 
-14 12 
-14 13 
-14 14 
-14 15 
Develop routine to select correct mean seepage, seepage standard deviation, seepage percent, 
and seepage percent standard deviation based on sampled value of 1/ , k, percolation flux. 
nz := 16 ny := 9 nx := 14 
ii .. := 0 nx 
xi := ii ii ii x 
, 
:= SMPAdataii 2 
jj .. := 0 ny 
:= yj jj y . := 100 jj + 100 jj jj 
kk .. := 0 nz 
:= := . . 14 + kk 0.25 kk zkkk zkk 
ANL-EBS-NU-000008 REV 00 
SMPAdata
<0> is permeability value log(k [m^2]) 
SMPAdata
<1> is capillary strength 1/alpha [Pa] 
SMPAdata
<2> is local percolation flux (mm/yr) 
SMPAdata
<3> is Mean Seepage [kg/yr/WP] 
SMPAdata
<4> is Std. Dev. Seepage [kg/yr/WP] 
SMPAdata
<5> is Mean Seepage [%] 
SMPAdata
<6> is Std. Dev. Seepage [%] 
27.73 1 100 
138.92 5 100 
277.9 10 100 
555.87 20 100 
1391.67 50 100 
2793.55 100 100 
5610 200 100 
8415 300 100 
11220 400 100 
14025 500 100 
16830 600 100 
19635 700 100 
22440 800 100 
25245 900 100 
28050 1000 100 
26.14 1 200 
IV-60 of IV-80 
14.59 98.86 4.09 
14.65 99.05 20.55 
14.68 99.07 41.19 
14.71 99.09 82.54 
14.66 99.23 205.57 
14.5 99.59 406.7 
14 100 785 
14 100 1178 
14 100 1570 
14 100 1963 
14 100 2356 
14 100 2748 
14 100 3141 
14 100 3590 
14 100 3989 
15 93.21 4.21 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
sms2 := 
i 
smsd2 := 
+ 
sms3 := 
i 
smsd3 := 
+ 
loc represents the location within the matrix of which value to pick for the interpolation process 
loc1i 
loc 
loc 
:= 
sms1 := 
i 
smsd1 := SMPAdataloc i 
spms1 := SMPAdataloc i 
spsd1i 
:= SMPAdataloc1 
dataloc2 
spms2 := SMPAdataloc 
spsd2 := SMPAdataloc i 
dataloc3 
spms3 := SMPAdataloc 
spsd3 := SMPAdataloc i 
loc
2i 
:=
i
i 
3i 
4i 
floor linterp z zk , TkTn 
+
SMPAdataloc1i
1i 
1i 
i 
SMPAdataloc2i 
SMPA 
i 
2i 
2i 
SMPAdataloc3i 
SMPA 
i 
3 
:=
i
i i 
, 
4 , 
5 ,
6 , 
4 , 
5 ,
6 , 3i 
, 
i 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
floor linterp z zk , TkTni 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , ,TkTni 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , TkTni 
floor linterp x xi , , qpffi 
i
4i 
4i 
, 
3 , 
4 , 
5 ,
6 , 
:=
i 
i i 
+ . ( ) , + ( ) ny 1 floor linterp y yj,T 
. 
nx 1 + 
1\ 
, + . 
, 
.
. 1\ 
1\
i 
. . 
nx 1 + 
. 
.( ) , + ( ) ny 1 floor linterp y yj T 
. 
nx 1 + ( ) + , + ( ) ny 1 ceil linterp y yj,T 
. 
nx 1 + 
. 
. 
i 
( ) . ... 
. 
nx 1 + 
i 
... 
February 2004 
sms4 := 
i 
smsd4 := SMPAdataloc 
spms4 := SMPAdataloc i 
spsd4 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
.. 
..
SMPAdataloc4 
..
. 
. 
.. 
.. 
..
.. 
.. 
.. 
.. 
.. 
..
.. 
.. 
.. 
+ 
. 
. 
.
. 
. 
. 
. 
. 
.. 
.. 
4 
.. 
.. 
.. 
.. 
1\ 
i
. 
. 
. 
( )... 
. 
nx 1 + 
. 
nx 1 + 
.. 
+ 
. 
( ) , + ( ) ny 1 ceil linterp y yj T . 
.. 
.. 
.. 
.. 
IV-61 of IV-80 
. 
( ) 
( )... 
. 
nx 1 + 
.
.
Screening Analysis for Criticality Features, Events, and Processes for License Application 
loc ceil linterp z zk , TkTn 5i 
5i
5i 
5i 
SMPA 
:= 
sms5 := SMPAdataloc i 
smsd5 := SMPAdataloc i 
spms5 := SMPAdataloc i 
spsd5i 
. 
+ 
:= dataloc5i 
loc6i 
.. 
.. 
.. 
..
.. 
i 
+ 
sms6 := 
i 
smsd6 := 
:=
i
i 
6 
loc7i 
.. 
.. 
.. 
.. 
dataloc6 
dataloc7i 
.. 
+ .. 
.. 
.  
sms7 := 
i 
smsd7 := 
+ 
:=
i
i 
7 
. 
.. 
nx 1 + 
.. 
nx 1 + ..
.. 
.
( ) , + ( ) ny 1 ceil linterp y yj,T1i 
. . 
,
3 ,
i 
6i 
6 , 
i
3 ,
i 
7i 
6 , 
i 
loc 
.. 
.. 
8i 
.. 
.. 
.. 
.. 
SMPA 
SMPAdataloc6 
spms6 := SMPAdataloc 
spsd6 := SMPAdataloc i 
ceil linterp z zk , ,TkTni 
ceil linterp x xi , , qpffi 
SMPA 
SMPAdataloc7 
spms7 := SMPAdataloc 
spsd7 := SMPAdataloc i 
ceil linterp z zk , TkTni 
floor linterp x xi , , qpffi 
.. 
.. 
.. 
nx 1 + 
.. 
i 
, 
i 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
ceil linterp z zk , TkTni 
ceil linterp x xi , , qpffi 
4 , 
5 , 
4 , 
5 , 
4 , 
5 , 
.. 
.. 
8 
,
3 , 
8i 
8i 
6 , 
i 
:=
i 
i 
.. 
..
SMPAdataloc8 
sms8 := 
i 
smsd8 := SMPAdataloc 
spms8 := SMPAdataloc i 
spsd8 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
. ( )( ) ny 1 + floor linterp y yj , ,T .. 
.. 
. 
+ 
( ) + ny 1 ( )+ , floor linterp y yj,T1 
+ . 
.. 
.. 
.. 
.. 
1 
i 
i
.. 
.. 
.. 
.. 
nx 1 + 
( ) , + ( ) ny 1 ceil linterp y yj T + 
.. 
.. 
. , 1 
.. 
nx 1 + 
( ) 
.. 
nx 1 + .. 
. 
.. 
.. 
IV-62 of IV-80 
. .. 
. 
i
.. 
nx 1 + 
.
( )... 
.. 
nx 1 + 
... 
( )
...
... 
( ) 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Develop the upper and lower bound of the randomly generated 1/ , k, and adjusted percolation 
flux 
Develop the upper and lower bound for permeability (k) for Tptpmn Unit 
i qq := .1.TkTni 
mantissa x . := ( ) x floor qq ( ) 
( ) i tt := floor qqi 
rr , := round( ) mantissa qq ( ) 2 
yy1 := i i , i if , < ( ) rr  0.25, 0 if( ) 0.25 rri  0.5, 0.25 rr 
if , < ( ) yy1  0.5, yy1 if( ) 0.5 yy1  0.75, 0.5, 0.75 zz1 := i i i i 
1 . . tt i i TkTn2 := 
i 
yy2 := 
( ) + zz1 
if , < ( ) rr  0.25, 0.25 if( ) 0.25 rri  0.5, 0.5 rri , i i 
zz2 := i i i i 
:= T 
if , < ( ) yy2  0.5, yy2 if( ) 0.5 yy2  0.75, 0.75, 1 
.1.( ) tt + zz2i i kTn1i 
Develop the upper and lower bound for capillary strength (1/ ) 
hh1 := floor i 
... 
.
. 
T1i 
100 
100 ( ) hh1 . i T11 := 
i 
:= ceil hh2i 
.
. 
T1i 
100 
T12 :=( ) hh2 .100 i i 
... 
.. 
i i i i 
Lower Bound value adjusted percolation flux (q pff) 
:= if q aaa1 5  1 , , qpff 1 , <  1 , if 1 qpff pff i i i i 
.. 
.. 
..
,  < , , 10, 5 aaa1 if aaa1 5  aaa1 if 5 aaa1 i bbb1 := i i i i ( ) ( ) 
if , < ( ) bbb1  10, bbb1 if( ) 10 bbb1  20, 10, bbb1 ccc1 := i i i i i 
ddd1 := i i i i i if , < ( ) ccc1  20, ccc1 if( ) 20 ccc1  50, 20, ccc1 
if , < ( ) ddd1  50, ddd1 if( ) 50 ddd1  100, 50, ddd1 eee1 := i 
if , < ( ) eee1  100, eee1 if( ) 100 eee1  200, 100, eee1 fff1 := i i i i i 
February 2004 IV-63 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ggg1 := i i i i i if , < ( ) fff1  200, fff1 if( ) 200 fff1  300, 200, fff1 
if , < ( ) ggg1  300, ggg1 if( ) 300 ggg1  400, 300, ggg1 hhh1 := i i i i i 
iii1 := i i i i i if , < ( ) hhh1  400, hhh1 if( ) 400 hhh1  500, 400, hhh1 
if , < ( ) iii1  500 iii1 if( ) 500 iii1  600, 500 iii1 jjj1 := i i , i , i i 
kkk1 := i i , i , i i if , < ( ) jjj1  600 jjj1 if( ) 600 jjj1  700, 600 jjj1 
if , < ( ) kkk1  700, kkk1 if( ) 700 kkk1  800, 700, kkk1 mmm1:= i i i i i 
:= q if , < ( ) mmm1  800,mmm1 if( ) 800 mmm1  900, 800, 900 i i i pff1i 
i 
i 
i 
i 
. T TkTn kTn1i i 
:= vTkTn . i T T 
Upper Bound value adjusted percolation flux (q pff) 
, 1 <  5 , if 1 q if q aaa2 := i 5  5 , , qpffi 
.. 
.. 
.. 
.. 
pffi pffi 
bbb2 := i i i i i if , < ( ) aaa2  5, aaa2 if( ) 5 aaa2  10, 10, aaa2 
if , < ( ) bbb2  10, bbb2 if( ) 10 bbb2  20, 20, bbb2 ccc2 := i i i i i 
ddd2 := i i i i i if , < ( ) ccc2  20, ccc2 if( ) 20 ccc2  50, 50, ccc2 
if , < ( ) ddd2  50, ddd2 if( ) 50 ddd2  100, 100, ddd2 eee2 := i i i i 
if , < ( ) eee2  100, eee2 if( ) 100 eee2  200, 200, eee2 fff2 := i i i i i 
ggg2 := i i i i i if , < ( ) fff2  200, fff2 if( ) 200 fff2  300, 300, fff2 
if , < ( ) ggg2  300, ggg2 if( ) 300 ggg2  400, 400, ggg2 hhh2 := i i i i 
iii2 := i i i i if , < ( ) hhh2  400, hhh2 if( ) 400 hhh2  500, 500, hhh2 
if , < ( ) iii2  500 iii2 if( ) 500 iii2  600, 600 iii2 jjj2 := i i , i , i i 
kkk2 := i i , i , i i if , < ( ) jjj2  600 jjj2 if( ) 600 jjj2  700, 700 jjj2 
if , < ( ) kkk2  700, kkk2 if( ) 700 kkk2  800, 800, kkk2 mmm2:= i i i i 
if , < ( ) mmm2  800,mmm2 if( ) 800 mmm2  900, 900, 1000 i i i qpff2 := 
i 
. T 1 11i i 
. T11i 12i 
Solve for seepage rate (Tptpmn Unit) 
. q q T pff pff1i i 
:= := t u qpff . T1i i q q T pff1i pff2i kTn1i kTn2i 
February 2004 IV-64 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
spfluxTnmi 
1 t . qpffi.
t 1 u . T1\ qpffi 
. u tqpffi 
1 t . qpffi 
1 t . qpffi 
...... 
.. 
+ t 1 uT1\ 
. + u 
+ 
.. 
.. 
T1\i
T1\ 
qpffi 
tqpff 
. 
i 
1 tqpffi 
...... 
:= 1 t . qpff . 
:= 
spfluxTnsd i i.
t 
. 
. 
.. 
. 
.. 
. 
.. 
.... 
qpffi 
1 t . qpffi 
t 
. . 
..
1 uT1\ 
u
u
T1\i 
. u
1 u
i T1\ 
. 
.. 
.. 
.. 
+
+
+
+ 
........ 
+
+ t 
+
+
+
+
+ 
.. 
qpffi 
. 
. 
. 
. 
. 
1 t . qpff 
. 
i 
qpffi 
tqpff 
. 
i 
1 uT1\ 
.. 
.... 
T1\i
T1\ 
.. 
....
1 tqpffi 
...... 
0, .0.00001 QTn1stdl 
.. 
. 
..
.. 
:=
:= 1.2 spfluxTnm 
:= if QTn1spm .. 
Calculate mean seepage for Tptpmn Unit 
.1.7321spfluxTnsd i 
QTn1stdli 
:= 
QTn1stdu i 
QTnstdli 
QTn QTnstd i 
. 
i 
, 
100 
. 
1.7321.spfluxTnsd i 
qpff .28.05 
i 
, 
, stdli 
i 
, :=
:=
if QTn1stdli 
qunif Xi 1 , 0 
QTn2spmi
. 
if QT2perci 
:= 
QTn1spmi 
QTn2spmi 
QT2perc := 
i 
QT3perci 
0 . 0 , if QT2perci 
Equation to calculate seepage percent based on seepage rate (see 
SMPA data table) from DTN: LB0310AMRU0120.002 [DIRS 166116] 
. , 100, 100 QT2perci 
. 
.. 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
1 u . T1\ sms1 ... 
i i 
1 v . TkTni 
. . ... sms2i i 
... . s 
i 
. ... u s 
T1\i
T1\ 
ms3i
ms4 
.
1 vTkTni 
1 v . TkTn 
1 vTkTni i i 
. 
.
. .s v 
. 
TkTni ms5i 
sms6i 
.
.. 
. 
. . 
 . 
1 u . T1\i
v 
i
v . ... .s 
.
i 
.
.s 
.
.
.
. u 
ms7i
ms8i 
. 
. 
. ... 
... 
. msd1 s 
i 
... .s 
1 vTkTni 
1 v . 
. 
. 
TkTni 
.
.. 
.. 
.. 
i 
. 
.. 
.. 
1 uT1\i. 
. .s 
. ... 
... 
i 
.. 
.
TkTni 
TkTni 
1 v .
. . ... 
. 
.
TkTni 
vTkTn 
... 
msd5i 
.. 
.
TkTni 
TkTn 
v
. 
i
1 v 
. v 
. 
. 
. 
.. 
.. 
.
.
.
.
. 
. 
.
msd3i 
msd4 
i 
... s 
. 
.
T1\i 
i
v . 
. 
. 
.. 
..
TkTni 
..
msd6i 
i 
. . 
.
msd2i 
s .
.s
... 
msd8i 
.
u 
.. 
.
TkTni 
. msd7 s 
v 
.
.s . . 
.. 
.. 
i 
.. 
.
TkTni 
.. 
. 
. i
. 
. 
. 
QTn1stdu i 
+ QTnstd i 
Increase the seepage rate by 20 percent to 
account for drift degradation 
.. . 
0.1, 0,QTn1spmi
. 
. 
. 
IV-65 of IV-80 
Check seepage percent; if above 
100 percent, then recalculate 
seepage back to 100 percent 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
QTlspri 
mean QTlspr 
:= 
Determine the seepage fraction for Tptpmn Unit within the repository and then fit the output 
data to distribution 
Seepage fraction represents the non-zero seepage rates based on the LHS sampling of all of the 
parameters 
( )= 102.895 
spr 
9520 
9521 
9522 
9523 
9524 
sort( ) QTl = 
nTl := 9521 
spfrcTl := 
n1Tl := 
Q11 := sort QTl Tl 
Q1Tl 
Q2Tlab 
:= 
:= QTl 
( ) n 1 . .nTl 
n 
( ) n 1 . .( ) n + 1 Tl 
spr
Tl 
ab .. := 0 n1Tl
( ) 
reverse( ) Q11 
:= 
998 
Q1Tlab 
1 ... 
. 
. 
Tl 
mean Q . Tl 998 
sort( ) Q2 
CDFTl
( )= 196.403 
( . + ab 1) 0.375 := 
ab Tl 
Fit the seepage rates to a Weibull distribution. 
The following equations are from What Every Engineer Should Know About Reliability and 
Risk Analysis (Modarres 1993 [DIRS 104667] p. 109). 
( ) n1 + 1 . 0.25 
n1Tl  = 1.048 104 
ANL-EBS-NU-000008 REV 00 
Mean Seepage Rate (kg/yr per WP) QT3perci
.qpffi
. 
100 
28.05 
0 
0
0 
0.1 
0.1 
0.1 
spfrcTl = 0.524 Seepage fraction (i.e., waste package locations that can 
see seepage) 
February 2004 IV-66 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
:=  root 
...  
........  
........  
n1Tl 
i
=
:=  
.....  
. 
Plot of raw data versus Weibull distribution 
ji 0 n1Tl 
ji 
ji 2 , 
ji := 
CDFwji 0 , 
cumulative density function 
:= .. 
PFdata := QTlji 
CDFdata 
CDFw1 
:= 
CDFdata 2 . . 
CDFw 
ANL-EBS-NU-000008 REV 00 
:= 
ji 
 PFdata ji 
. 
 
Seepage Rate (Tptpmn) data and fit 
0.01 
PFdata 
Seepage Rate (m3/yr) 
IV-67 of IV-80 
n1Tl 
QTl i i 
0 i 
.
...
. 
n1Tl 
. 
QTli
.. 
r =
0 i 
. . 
.
r ln QTl . 
1
 
QTli 
= 
..
0 
.. 
.. 
 
=  8.559 10 2 . 
n1Tl 
.
..... 
CDFTl 
. 1 exp 
... 
.
. 
.. 
...  
CDFw1ji 
1 
0.75 
0.5 
0.25
0 
1 .10 4 
.. 
.... 
1 .10 3 
n1Tl 
1 . r , , 0.1, 4 ln QTli 
1
r 
..
n1Tl 
.
. 
= 
. 
0 i 
... 
...  
.
... 
.. 
......... 
... 
~ 
.. 
0.1 
......... 
1 
 = 0.473 
10 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
IV.6 SEEPAGE ANALYSIS FOR UPPER INFILTRATION RATE IN THE 
NONLITHOPHYSAL ZONE 
The following section presents the Mathcad analysis for the upper seepage infiltration rate of the 
glacial transition climate in the nonlithophysal zone. The seepage information used in this 
analysis was obtained from Abstraction of Drift Seepage (BSC 2003 [DIRS 165564]). The 
information contained in the section has been abstracted from the seepage glac upper Tptpmn 
x1.2 report.mcd Mathcad file of Attachment VII. 
February 2004 IV-68 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
:= := := 
:= := := := 
Seepage Rate and Seepage Fraction Calculation using Abstraction of Drift Seepage 
(BSC 2003 [DIRS 165564], Section 6.7) 
rnd( ) 1.0 RDi 1 - , 12 
rnd( ) 1.0 := 
rnd( ) 1.0 
:= 
rnd( ) 1.0 RDi 1 - , 10 
rnd( ) 1.0 RDi 1 - , 11 
Latin Hypercube Sampling Routine to Generate Random Numbers 
n := 20000 
:= := rnd( ) 1.0 RDi 1 - , 9 rnd( ) 1.0 RDi 1 - , 6 RDi 1 - , 3 
rnd( ) 1.0 RDi 1 - , 7 
rnd( ) 1.0 RDi 1 - , 8 
Sample Size: 
i 1 := ..n 
RDi 1 - , 0 
RDi 1 - , 1 
RDi 1 - , 2 
i := 
rnd( ) 1.0 RDi 1 - , 4 
rnd( ) 1.0 RDi 1 - , 5 
RKs are matrixes in which the first column contain a permutation on the integers on 
the interval [1,n] 
RK1:= csort := := := ( ) , ( ) , ( ) , ( ) RD 1 , RK4 csort RD 4 RK7 csort RD 7 RK10 csort RD 10 
RK2:= csort := := := ( ) , ( ) , ( ) , ( ) RD 2 , RK5 csort RD 5 RK8 csort RD 8 RK11 csort RD 11 
RK3:= csort := := := ( ) , ( ) , ( ) , ( ) RD 3 , RK6 csort RD 6 RK9 csort RD 9 RK12 csort RD 12 
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n
runif( ) n 0 , 1 , +
n 
Define sets of random values. Each random value is selected within one of the 
equiprobable n intervals that partition [0,1], one set for each random variable 
runif , ( ) n 0, 1 < > RK7 0 < > 
1 - 
X 6 := runif , ( ) n 0, 1 < > RK4 0 < > 
1 - 
X 3 := X 0 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK8 0 < > 
1 - 
X 7 := runif , ( ) n 0, 1 < > RK5 0 < > 
1 - 
X 4 := 
< > RK1 0 < > 
1 - 
X 1 := 
+
n 
+
n 
runif , ( ) n 0, 1 < > RK9 0 < > 
1 - 
X 8 := runif , ( ) n 0, 1 < > RK6 0 < > 
1 - 
X 5 := 
< > RK2 0 < > 
1 - 
X 2 := 
+
n
n 
+
n 
X := 
+ ( ) , 10 < > RK110 < > 
1 - runif n 0, 1 
n 
IV-69 of IV-80 
< > RK3 0 < > 
1 - 
< > ( ) , RK100 < > 
1 - + runif n 0, 1 
X 9 := 
X := 
+ ( ) , 11 < > RK120 < > 
1 - runif n 0, 1 
n 
i 0 - := ..n 1 
ANL-EBS-NU-000008 REV 00 
rnd( ) 1.0 := 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Capillary Strength 1/  in (Pa) 
1lb 
Sampling from spatial variability to obtain the 1/  value 
.1 := 0  .1 
i 
Sample from uncertainty triangular distribution to obtain .1/ 
Determine which equation to use: 
if Random Number < RN . 1 then use Equation 1 ( . 1eq1). 
if Random Number > RN . 1 then use Equation 2 ( . 1eq2). 
qunif( ) Xi 0 , ,  1lb, 1ub 1/ value 1 := 
RN.1 
eq1i 
.1
eq2i 
.1
i .1
i 
Overall Capillary Strength 1/  + .1/ 
.. 
i 1 
0.34 
T1 := 
Permeability k in Tptpmn Unit (in log 10) 
kTn := 
kTn := 
kTni 
mean kTn 
Stdev kTn 
Permeability .k in Tptpmn Unit (in log 10) 
( ) = .12.2 
( ) = 0.34 
:=
ln( ) qlnorm( ) X ,  kTn, 
Uncertainty follows a triangular distribution := .kTnl 
ANL-EBS-NU-000008 REV 00 
:= := Spatial variability follows a uniform distribution 1 := 591 780  402 1ub 
:= Uncertainty follows a triangular distribution .105 .1 := 105 u l 
:=
:= l 
:= 
:= 
:= 
( ) .1 . .1l 
2 
( ) .1u . .1l .( ) .1 . .1l 
Xi 1 , l u l 
.1 . 
.1 + 
u 
. . ( ) .1 . .1 .( ) 1 .1  . 
u l i 1 , ( ) 1 X . . . ( ) .1u . .1 .( ) 1 . .1 
i i 
.. 
+ .1 
.12.2 Mean of lognormal distribution 
Standard deviation of lognormal distribution 
if( ) Xi 1 ,  RN.1, . .1eq1 , 1eq2 .1/ value 
1/ value i 
kTn i 2 , 
:= 0.68 0 .kTnu .0.68 .kTn 
February 2004 IV-70 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Sample from uncertainty triangular distribution to obtain .k 
Determine which equation to use: 
if Random Number < RN .kTn then use Equation 1 ( .kTneq1). 
if Random Number > RN .kTn then use Equation 2 ( .kTneq2). 
RN.kTn 
:= .kTn 
. := .kTn 
:= 
eq1i 
eq2i 
:= 
( ) .kTn . .kTnl 
2 
( ) .kTnu . .kTnl .( ) .kTn . .kTnl 
+ Xi 3 , Tnu Tnl . . . ( ) .k . .k .( ) kTn .kTnl 
i 3 , Tn Tnu Tnl 
.k
Tnu .k 
if Xi 3 , 
Tnl 
Tni 
:= 
TkTn := 
i 
:= ( ) .. 
.k 
T1kTni 
ffi 
( ) 1 X . . . ( ) .kTnu . .k .( ) k . .k 
 RN.kTn 
.
, . .kTn , kTn ( ) 
Overall Permeability k + .k 
Tni 
kTni 
Permeability must lie between -14 and -10 (bounds of SMPA simulations) 
, if T1kTn 
35.66 x2 
+ 
x4 
.k
 .10, .10 if T1kTn i 
5.4998x3 . + . 
.. 
.. 
Flow Focusing Factor (DTN: LB0104AMRU0185.012 [DIRS 163906]) 
f x ( ) := .0.3137 
root f x . ( ) X .100 , x, 0, 6 i 5 , 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
i 
11.434 102.3 x . + . 
.k value eq2i eq1i 
k value .. 
.. 
.. 
 .14, .14,T1kTni 
.
IV-71 of IV-80 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Percolation Flux (mm/yr) 
The percolation flux used here is for the glacial transition period only. The percolation flux is 
based on sampling from the upper bound TSPA repository location only 
(DTN: LB0310AMRU0120.002 [DIRS 166116]) 
nnn .. := 0 468 
Upper Bound Percolation Flux 
PF1 := u 0 
40 
36.19 
35.73 
27.83 
30.74 
40.03 
31.86 
57.08 
18.33 
27.91 
0
1
2
3
4
5
6
7
8
9
PF1u , nnn 0 
PFtunnn 
:= 
Z ( ) ( ) , 0 := round runif n 0, 468 
Zi 0 , 
, := q 
. . 
PF := PFt i 
q1pffi 
:= 
pffi 
u( ) 
Adjusted Percolation Flux 
Multiply the flow-focusing factor by the percolation flux, which will be used to obtain the 
seepage rate, seepage fraction, and seepage percentage. 
PFi ffi . 
Percolation Flux must lie between 1 and 1000 mm/yr (bounds of SMPA simulations) 
if q1 1  1 , if q1pffi pffi 
.. 
.. 
ANL-EBS-NU-000008 REV 00 
 1000, 1000, q1pffi
.. 
.. 
February 2004 IV-72 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Seepage Information from SMPA analysis 
(DTN: LB0304SMDCREV2.002 [DIRS 163687]) 
m:= 2549 Data points 
SMPAdata := 
0 1 2 3 4 5 6 
-14 0 
-14 1 
-14 2 
-14 3 
-14 4 
-14 5 
-14 6 
-14 7 
-14 8 
-14 9 
-14 10 
-14 11 
-14 12 
-14 13 
-14 14 
-14 15 
Develop routine to select correct mean seepage, seepage standard deviation, seepage percent, 
and seepage percent standard deviation based on sampled value of 1/ , k, percolation flux. 
nz := 16 ny := 9 nx := 14 
ii .. := 0 nx 
xi := ii ii ii x 
, 
:= SMPAdataii 2 
jj .. := 0 ny 
:= yj jj y . := 100 jj + 100 jj jj 
kk .. := 0 nz 
:= := . . 14 + kk 0.25 kk zkkk zkk 
ANL-EBS-NU-000008 REV 00 
SMPAdata
<0> is permeability value log(k [m^2]) 
SMPAdata
<1> is capillary strength 1/alpha [Pa] 
SMPAdata
<2> is local percolation flux (mm/yr) 
SMPAdata
<3> is Mean Seepage [kg/yr/WP] 
SMPAdata
<4> is Std. Dev. Seepage [kg/yr/WP] 
SMPAdata
<5> is Mean Seepage [%] 
SMPAdata
<6> is Std. Dev. Seepage [%] 
27.73 1 100 
138.92 5 100 
277.9 10 100 
555.87 20 100 
1391.67 50 100 
2793.55 100 100 
5610 200 100 
8415 300 100 
11220 400 100 
14025 500 100 
16830 600 100 
19635 700 100 
22440 800 100 
25245 900 100 
28050 1000 100 
26.14 1 200 
IV-73 of IV-80 
14.59 98.86 4.09 
14.65 99.05 20.55 
14.68 99.07 41.19 
14.71 99.09 82.54 
14.66 99.23 205.57 
14.5 99.59 406.7 
14 100 785 
14 100 1178 
14 100 1570 
14 100 1963 
14 100 2356 
14 100 2748 
14 100 3141 
14 100 3590 
14 100 3989 
15 93.21 4.21 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
sms2 := 
i 
smsd2 := 
+ 
sms3 := 
i 
smsd3 := 
+ 
loc represents the location within the matrix of which value to pick for the interpolation process. 
loc1i 
loc 
loc 
:= 
sms1 := 
i 
smsd1 := SMPAdataloc i 
spms1 := SMPAdataloc i 
spsd1i 
:= SMPAdataloc1 
dataloc2 
spms2 := SMPAdataloc 
spsd2 := SMPAdataloc i 
dataloc3 
spms3 := SMPAdataloc 
spsd3 := SMPAdataloc i 
loc
2i 
:=
i
i 
3i 
4i 
floor linterp z zk , TkTn 
+
SMPAdataloc1i
1i 
1i 
i 
SMPAdataloc2i 
SMPA 
i 
2i 
2i 
SMPAdataloc3i 
SMPA 
i 
3 
:=
i
i i 
, 
4 , 
5 ,
6 , 
4 , 
5 ,
6 , 3i 
, 
i 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
floor linterp z zk , TkTni 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , ,TkTni 
ceil linterp x xi , , qpffi 
3 , 
floor linterp z zk , TkTni 
floor linterp x xi , , qpffi 
i
4i 
4i 
, 
3 , 
4 , 
5 ,
6 , 
:=
i 
i i 
. + . 
i 
( ) 
1\ 
, + . . ... 
( ) , + ( ) ny 1 floor linterp y yj,T 
. 
nx 1 + 
. 
nx 1 + 
i 
, ... 
.
. 1\ 
1\
i 
. . 
nx 1 + 
. 
.( ) , + ( ) ny 1 floor linterp y yj T 
. 
nx 1 + ( ) + , + ( ) ny 1 ceil linterp y yj,T 
. 
nx 1 + 
. 
February 2004 
sms4 := 
i 
smsd4 := SMPAdataloc 
spms4 := SMPAdataloc i 
spsd4 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
.. 
..
SMPAdataloc4 
..
. 
. 
.. 
.. 
..
.. 
.. 
.. 
.. 
.. 
..
.. 
.. 
.. 
+ 
. 
. 
.
. 
. 
. 
. 
. 
.. 
.. 
4 
.. 
.. 
.. 
.. 
. 
( )... 
. 
nx 1 + 
. nx 1 + 
1\ 
i
. 
. 
.. 
+ 
. 
( ) 
( )... 
. 
nx 1 + 
.
. 
. 
( ) , + ( ) ny 1 ceil linterp y yj T . 
.. 
.. 
.. 
.. 
IV-74 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
loc ceil linterp z zk , TkTn 5i 
5i
5i 
5i 
SMPA 
:= 
sms5 := SMPAdataloc i 
smsd5 := SMPAdataloc i 
spms5 := SMPAdataloc i 
spsd5i 
. 
+ 
:= dataloc5i 
loc6i 
.. 
.. 
.. 
..
.. 
i 
+ 
sms6 := 
i 
smsd6 := 
:=
i
i 
6 
loc7i 
.. 
.. 
.. 
.. 
dataloc6 
dataloc7i 
.. 
+ .. 
.. 
.  
sms7 := 
i 
smsd7 := 
+ 
:=
i
i 
7 
. 
.. 
nx 1 + 
.. 
nx 1 + ..
.. 
.
( ) , + ( ) ny 1 ceil linterp y yj,T1i 
. . 
,
3 ,
i 
6i 
6 , 
i
3 ,
i 
7i 
6 , 
i 
loc 
.. 
.. 
8i 
.. 
.. 
.. 
.. 
SMPA 
SMPAdataloc6 
spms6 := SMPAdataloc 
spsd6 := SMPAdataloc i 
ceil linterp z zk , ,TkTni 
ceil linterp x xi , , qpffi 
SMPA 
SMPAdataloc7 
spms7 := SMPAdataloc 
spsd7 := SMPAdataloc i 
ceil linterp z zk , TkTni 
floor linterp x xi , , qpffi 
.. 
.. 
.. 
nx 1 + 
.. i 
, 
i 
floor linterp x xi , , qpffi 
3 , 
4 , 
5 ,
6 , 
ceil linterp z zk , TkTni 
ceil linterp x xi , , qpffi 
4 , 
5 , 
4 , 
5 , 
4 , 
5 , 
.. 
.. 
8 
,
3 , 
8i 
8i 
6 , 
i 
:=
i 
i 
.. 
..
SMPAdataloc8 
sms8 := 
i 
smsd8 := SMPAdataloc 
spms8 := SMPAdataloc i 
spsd8 := SMPAdataloc 
ANL-EBS-NU-000008 REV 00 
. ( )( ) ny 1 + floor linterp y yj , ,T .. 
.. 
. 
+ 
( ) + ny 1 ( )+ , floor linterp y yj,T1 
+ . 
.. 
.. 
.. 
.. 
1 
i 
i
.. 
.. 
.. 
.. 
nx 1 + 
( ) , + ( ) ny 1 ceil linterp y yj T + 
.. 
.. 
. , 1 
.. 
nx 1 + 
( ) 
.. 
nx 1 + .. 
. 
.. 
.. 
IV-75 of IV-80 
. .. 
. 
i
.. 
nx 1 + 
.
( )... 
.. 
nx 1 + 
... 
( )
...
... 
( ) 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Develop the upper and lower bound of the randomly generated 1/ , k, and adjusted percolation 
flux 
Develop the upper and lower bound for permeability (k) for Tptpmn Unit 
i qq := .1.TkTni 
mantissa x . := ( ) x floor qq ( ) 
( ) i tt := floor qqi 
rr , := round( ) mantissa qq ( ) 2 
yy1 := i i , i if , < ( ) rr  0.25, 0 if( ) 0.25 rri  0.5, 0.25 rr 
if , < ( ) yy1  0.5, yy1 if( ) 0.5 yy1  0.75, 0.5, 0.75 zz1 := i i i i 
1 . . tt i i TkTn2 := 
i 
yy2 := 
( ) + zz1 
if , < ( ) rr  0.25, 0.25 if( ) 0.25 rri  0.5, 0.5 rri , i i 
zz2 := i i i i 
:= T 
if , < ( ) yy2  0.5, yy2 if( ) 0.5 yy2  0.75, 0.75, 1 
.1.( ) tt + zz2i i kTn1i 
Develop the upper and lower bound for capillary strength (1/ ) 
hh1 := floor i 
... 
.
. 
T1i 
100 
100 ( ) hh1 . i T11 := 
i 
:= ceil hh2i 
.
. 
T1i 
100 
T12 :=( ) hh2 .100 i i 
... 
.. i i i i 
Lower Bound value adjusted percolation flux (q pff) 
:= if q aaa1 5  1 , , qpff 1 , <  1 , if 1 qpff pff i i i i 
.. 
.. 
..
,  < , , 10, 5 aaa1 if aaa1 5  aaa1 if 5 aaa1 i bbb1 := i i i i ( ) ( ) 
if , < ( ) bbb1  10, bbb1 if( ) 10 bbb1  20, 10, bbb1 ccc1 := i i i i i 
ddd1 := i i i i i if , < ( ) ccc1  20, ccc1 if( ) 20 ccc1  50, 20, ccc1 
if , < ( ) ddd1  50, ddd1 if( ) 50 ddd1  100, 50, ddd1 eee1 := i 
if , < ( ) eee1  100, eee1 if( ) 100 eee1  200, 100, eee1 fff1 := i i i i i 
February 2004 IV-76 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ggg1 := i i i i i if , < ( ) fff1  200, fff1 if( ) 200 fff1  300, 200, fff1 
if , < ( ) ggg1  300, ggg1 if( ) 300 ggg1  400, 300, ggg1 hhh1 := i i i i i 
iii1 := i i i i i if , < ( ) hhh1  400, hhh1 if( ) 400 hhh1  500, 400, hhh1 
if , < ( ) iii1  500 iii1 if( ) 500 iii1  600, 500 iii1 jjj1 := i i , i , i i 
kkk1 := i i , i , i i if , < ( ) jjj1  600 jjj1 if( ) 600 jjj1  700, 600 jjj1 
if , < ( ) kkk1  700, kkk1 if( ) 700 kkk1  800, 700, kkk1 mmm1:= i i i i i 
:= q if , < ( ) mmm1  800,mmm1 if( ) 800 mmm1  900, 800, 900 i i i pff1i 
i 
i 
i 
i 
. T TkTn kTn1i i 
:= vTkTn . i T T 
Upper Bound value adjusted percolation flux (q pff) 
, 1 <  5 , if 1 q if q aaa2 := i 5  5 , , qpffi 
.. 
.. 
.. 
.. 
pffi pffi 
bbb2 := i i i i i if , < ( ) aaa2  5, aaa2 if( ) 5 aaa2  10, 10, aaa2 
if , < ( ) bbb2  10, bbb2 if( ) 10 bbb2  20, 20, bbb2 ccc2 := i i i i i 
ddd2 := i i i i i if , < ( ) ccc2  20, ccc2 if( ) 20 ccc2  50, 50, ccc2 
if , < ( ) ddd2  50, ddd2 if( ) 50 ddd2  100, 100, ddd2 eee2 := i i i i 
if , < ( ) eee2  100, eee2 if( ) 100 eee2  200, 200, eee2 fff2 := i i i i i 
ggg2 := i i i i i if , < ( ) fff2  200, fff2 if( ) 200 fff2  300, 300, fff2 
if , < ( ) ggg2  300, ggg2 if( ) 300 ggg2  400, 400, ggg2 hhh2 := i i i i 
iii2 := i i i i if , < ( ) hhh2  400, hhh2 if( ) 400 hhh2  500, 500, hhh2 
if , < ( ) iii2  500 iii2 if( ) 500 iii2  600, 600 iii2 jjj2 := i i , i , i i 
kkk2 := i i , i , i i if , < ( ) jjj2  600 jjj2 if( ) 600 jjj2  700, 700 jjj2 
if , < ( ) kkk2  700, kkk2 if( ) 700 kkk2  800, 800, kkk2 mmm2:= i i i i 
if , < ( ) mmm2  800,mmm2 if( ) 800 mmm2  900, 900, 1000 i i i qpff2 := 
i 
. T 1 11i i 
. T11i 12i 
Solve for seepage rate (Tptpmn Unit) 
. q q T pff pff1i i 
:= := t u qpff . T1i i q q T pff1i pff2i kTn1i kTn2i 
February 2004 IV-77 of IV-80 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
spfluxTnmi 
1 t . qpffi.
t 1 u . T1\ qpffi 
. u tqpffi 
1 t . qpffi 
1 t . qpffi 
...... 
.. 
+ t 1 uT1\ 
. + u 
+ 
.. 
.. 
T1\i
T1\ 
qpffi 
tqpff 
. 
i 
1 tqpffi 
...... 
:= 1 t . qpff . 
:= 
spfluxTnsd i i.
t 
. 
. 
.. 
. 
.. 
. 
.. 
.... 
qpffi 
1 t . qpffi 
t 
. . 
..
1 uT1\ 
u
u
T1\i 
. u
1 u
i T1\ 
. 
.. 
.. 
.. 
+
+
+
+ 
........ 
+
+ t 
+
+
+
+
+ 
.. 
qpffi 
. 
. 
. 
. 
. 
1 t . qpff 
. 
i 
qpffi 
tqpff 
. 
i 
1 uT1\ 
.. 
.... 
T1\i
T1\ 
.. 
....
1 tqpffi 
...... 
0, .0.00001 QTn1stdl 
.. 
. 
..
.. 
:=
:= 1.2 spfluxTnm 
:= if QTn1spm .. 
Calculate mean seepage for Tptpmn Unit 
.1.7321spfluxTnsd i 
QTn1stdli 
:= 
QTn1stdu i 
QTnstdli 
QTn QTnstd i 
. 
i 
, 
100 
. 
1.7321.spfluxTnsd i 
qpff .28.05 
i 
, 
, stdli 
i 
, :=
:=
if QTn1stdli 
qunif Xi 1 , 0 
QTn2spmi
. 
if QT2perci 
:= 
QTn1spmi 
QTn2spmi 
QT2perc := 
i 
QT3perci 
0 . 0 , if QT2perci 
Equation to calculate seepage percent based on seepage rate (see 
SMPA data table) from DTN: LB0310AMRU0120.002 [DIRS 166116] 
. , 100, 100 QT2perci 
. 
.. 
.. .. 
ANL-EBS-NU-000008 REV 00 
1 u . T1\ sms1 ... 
i i 
1 v . TkTni 
. . ... sms2i i 
... . s 
i 
. ... u s 
T1\i
T1\ 
ms3i
ms4 
.
1 vTkTni 
1 v . TkTn 
1 vTkTni i i 
. 
.
. .s v 
. 
TkTni ms5i 
sms6i 
.
.. 
. 
. . 
 . 
1 u . T1\i
v 
i
v . ... .s 
.
i 
.
.s 
.
.
.
. u 
ms7i
ms8i 
. 
. 
. ... 
... 
. msd1 s 
i 
... .s 
1 vTkTni 
1 v . 
. 
. 
TkTni 
.
.. 
.. 
.. 
i 
. 
.. 
.. 
1 uT1\i. 
. .s 
. ... 
... 
i 
.. 
.
TkTni 
TkTni 
1 v .
. . ... 
. 
.
TkTni 
vTkTn 
... 
msd5i 
.. 
.
TkTni 
TkTn 
v
. 
i
1 v 
. v 
. 
. 
. 
.. 
.. 
.
.
.
.
. 
. 
.
msd3i 
msd4 
i 
... s 
. 
.
T1\i 
i
v . 
. 
. 
.. 
..
TkTni 
..
msd6i 
i 
. . 
.
msd2i 
s .
.s
... 
msd8i 
.
u 
.. 
.
TkTni 
. msd7 s 
v 
.
.s . . 
.. 
.. 
i 
.. 
.
TkTni 
.. 
. 
. 
i
. 
. 
. 
QTn1stdu i 
+ QTnstd i 
Increase the seepage rate by 20 percent to account 
for drift degradation 
.. 
. 
0.1, 0,QTn1spmi
. 
. 
. 
IV-78 of IV-80 
Check seepage percent; if above 
100 percent, then recalculate 
seepage back to 100 percent 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
QTlspri 
mean QTlspr 
:= 
Determine the seepage fraction for Tptpmn Unit within the repository and then fit the output 
data to distribution 
Seepage fraction represents the non-zero seepage rates based on the LHS sampling of all of the 
parameters 
( )= 296.006 
spr 
6561 
6562 
6563 
6564 
6565 
:= 
ab 
:= 
sort( ) QTl = 
nTl := 6562 
spfrcTl := 
n1Tl := 
Q11Tl 
Q1Tl 
Q2Tl 
QTl 
( ) n 1 . .nTl 
n 
( ) n 1 . .( ) n + 1 Tl 
:= sort QTlspr
Tl 
ab .. := 0 n1Tl
( ) 
reverse( ) Q11 
:= 
998 
Q1Tlab 
1 ... 
. 
. 
Tl 
mean Q . Tl 998 
sort( ) Q2 
( )= 440.583 
CDFTl 
( . + ab 1) 0.375 := 
ab Tl 
Fit the seepage rates to a Weibull distribution. 
The following equations are from What Every Engineer Should Know About Reliability and 
Risk Analysis (Modarres 1993 [DIRS 104667], p. 109). 
( ) n1 + 1 . 0.25 
n1Tl  = 1.344 104 
ANL-EBS-NU-000008 REV 00 
Mean Seepage Rate (kg/yr per WP) QT3perci
.qpffi
. 
100 
28.05 
0 
0
0 
0.101 
0.101 
0.101 
spfrcTl = 0.672 Seepage fraction (i.e., waste package locations that 
can see seepage) 
February 2004 IV-79 of IV-80 Screening Analysis for Criticality Features, Events, and Processes for License Application 
:=  root 
...  
........  
........  
n1Tl 
i
=
:=  
.....  
. 
Plot of raw data versus Weibull distribution 
ji 0 n1Tl 
ji 
ji 2 , 
ji := 
CDFwji 0 , 
cumulative density function 
:= .. 
PFdata := QTlji 
CDFdata 
CDFw1 
:= 
CDFdata 2 . . 
CDFw 
ANL-EBS-NU-000008 REV 00 
:= 
ji 
 PFdata ji 
. 
 
Seepage Rate (Tptpmn) data and fit 
0.01 
PFdata 
Seepage Rate (m3/yr) 
IV-80 of IV-80 
n1Tl 
QTl i i 
0 i 
.
...
. 
n1Tl 
. 
QTli
.. 
r =
0 i 
. . 
.
r ln QTl . 
1
 
QTli 
= 
..
0 
.. 
.. 
 
=  2.251 10 1 . 
n1Tl 
.
..... 
CDFTl 
. 1 exp 
... 
.
. 
.. 
...  
CDFw1ji 
1 
0.75 
0.5 
0.25
0 
1 .10 4 
.. 
.... 
1 .10 3 
n1Tl 
1 . r , , 0.1, 4 ln QTli 
1
r 
..
n1Tl 
.
. 
= 
. 
0 i 
... 
...  
.
... 
.. 
......... 
... 
~ 
.. 
0.1 
......... 
1 
 = 0.501
10 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ATTACHMENT V 
NEUTRONIT CORROSION SPREADSHEET (OUTPUT FROM MATHCAD FILE) 
V-1 of V-8 ANL-EBS-NU-000008 REV 00 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
INTENTIONALLY LEFT BLANK 
February 2004 V-2 of V-8 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ATTACHMENT V - NEUTRONIT CORROSION SPREADSHEET (MATHCAD FILE) 
The following presents the Mathcad analysis for the sampling of the Neutronit corrosion rate for 
input to the minimum required seepage analysis of Attachment III. The stainless steel type 316 
corrosion information used in this analysis was obtained from DTN: MO0303SPAMCRAQ.000 
(BSC 2003 [DIRS 162353]). The information contained in the section has been abstracted from 
the Attachment III.mcd Mathcad file of Attachment VII. 
February 2004 V-3 of V-8 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Stainless steel type 316 corrosion rate i nformation from experiments listed in 
DTN: MO0303SPAMCRAQ.000 (BSC 2003 [DIRS 162353]) is fit to a Weibull Distribution . 
SS316 is the stainless steel type 316 corrosion rate data ( gm/yr), which is obtatined from 
DTN: MO0303SPAMCRAQ.000 (aqueous-316L.xls), only the J-13 well water data is used (BSC 2003 
[DIRS 162353]). 
nn := 13 Total data points 
mm:= 12 
jj .. := 0 mm 
:= SS1316 
0.037 
0.102 
0.109 
0.152 
0.154 
0.178 
0.203 
0.229 
0.229 
0.254 
0.254 
0.254 
0.279 
.................. 
c := 1.5 
SS316jj 
mm 
SS316 jj jj 
.
mm 
.
.............. 
 
c is used to increase the corrosion rate for Neutronit versus stainless steel 316. 
SS1316jj 
c . 
0 jj 
.
..., . . r , 0.1, 5 ln SS316 ]316 jj 
1 
nn 
1
r 
. 
... 
.. 
. .  
= 
. 
0 jj 
... 
.... 
.
... 
. 
. 
SS316jj 
.. 
. 
r 
0 
.
....... 
.
....... 
. . 

r ln SS316 . 
1 
]316 
:= 
The following equations are from What Every Engineer Should Know About Reliability and 
Risk Analysis, (Modarres 1993 [DIRS 104667] p. 109). 
root 
jj 
.. 
mm 
jj 
SS316
.= 
]316 
jj
. 
.= 
.... 
.. 
nn 
:= 
\316:= 
........ 
........ 
]316 = 3.027 
mm
0 
.= 
..... 
.
.... 
= \316 0.314 
February 2004 V-4 of V-8 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
pdf316( ) x := 
....  
pdf316( ) x
:= cdf316(x) 1 exp . 
0.75 
cdf316 x ( ) 
0.5 
0.25
. 
1
0 
0 
ANL-EBS-NU-000008 REV 00 
...  
probability density function cumulative distribution function 
. 
 
(316.1) 
316.x 
.exp . 
316 
316 
4
3
2
1
0 
0 
...  
316 x 
... 
.
... 
.
.... 
0.11 
316 
... 
.
.
0.12 
316 x 
316 
.
... 
pdf316(x) is the probability density function of 
Neutronit using a Weibull distribution. 
0.43 0.33 0.65 0.54 
.
. 
1.5 x SS 316 Weibull Fit (pdf) 
0.22 
x 
Corrosion Rate (um/yr) 1.5xSS 316 
cdf316(x) is the cumulative distribution function of the 
Neutronit data fit to a Weibull distribution. 
1.5 x SS 316 Weibull Fit (cdf) 
0.36 0.24 
x 
Corrosion Rate (um/yr) 1.5xSS 316 
V-5 of V-8 
0.6 0.48 
February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Plot of raw data versus Weibull distribution 
SS316jj 
1 exp w316jj 316 
....  
:= 
c316 
w316 
. 
1 
0.8 
0.6 
0.4 
0.2
0 
0 
The distribution was checked to see if it can be used to represent the corrosion rate ( m/yr) for 
Neutronit. The Kolmogorov-Smirnov (K-S) Test was chosen to determine the goodness-of-fit of the 
data to the Weibull distribution. The K-S Test is used according to Goodness-Of-Fit Techniques 
(D'Agostino and Stephens 1986 [DIRS 160320], Chapter 4). 

0.13 
10 0.819 
KSw2:= 0.874 
.
. 
sigw := 20 0.843 
50 0.856
.
.... 
.
. 
numw sigw 0 . . 
testw sigw
. . 1 
KSw1:= linterp , ( ) numw testw , nn 
KSw1 = 0.826 
KSw:= if >( ) nn 50,KSw2,KSw1 
.... 
:= 
KSw = 0.826 
:= 
KSw is the Weibull test value at the 0.05 significance shown by 
D'Agostino and Stephens (1986 [DIRS 160320], p. 148). 
ANL-EBS-NU-000008 REV 00 
cumulative distribution function 
316 
( . + jj 1) 0.375 
c316 := 
jj nn + 0.25 
0.5 0.38 0.25 
..... 
Corrosion Rate Data (1.5xSS 316) 
SS316 
Corrosion Rate (um/yr) 1.5xSS316 
This is used to interprete the 0.05 significance level for the 
Weibull distribution as shown by D'Agostino and Stephens 
(1986 [DIRS 160320], p. 148). 
February 2004 V-6 of V-8 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Set up the K-S test by determining the data functional form and the distribution fit of the data. 
316 SS316jj 
:= 1 exp . we316jj 
we316 is the Weibull cumulative 
distribution function of the fitted 
corrosion rate. 316 
.( ) jj + := := DM 
.
nn 
....  
( ) + jj 1 
we316 DP316 we316 
.
. 
jj 316jj jj jj nn 
.... 
. + 
max DM316,DP316 
..... 
KS316 = 0.194 KS316 
:= 
KSw = 0.826 
:= 
KS316mw 
pass := 
( ) 
= nn . mw 0.698 KS316( ) KS316 
If KSm is less than KSn, then distribution cannot be rejcted at 0.05 significance level. 
mw KSw , 0, 1.. 
0 equals reject hypothesis of correct distribution type at the 0.05 significance level. 
1 equals cannot reject hypothesis of correct distribution type at the 0.05 significance level. 
. 
if( ) KS316 > 
pass 1 = 
February 2004 V-7 of V-8 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
INTENTIONALLY LEFT BLANK 
February 2004 V-8 of V-8 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
ANL-EBS-NU-000008 REV 00 
ATTACHMENT VI 
LISTING OF FILES ON CD-ROM 
VI-1 of VI-4 February 2004 Screening Analysis for Criticality Features, Events, and Processes for License Application 
INTENTIONALLY LEFT BLANK 
February 2004 VI-2 of VI-4 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
Time Size (bytes) Date File Name 
EXCEL Directory 
Igneous Files 
FepIgn1.xls 07/30/2003 08:23a 29,184 
FEP Results Binomial Distribution Files 
Binom Dist.xls 11/09/2003 05:20p 20,992 
Time Size (bytes) Date 
MCNP (Igneous) Directory 
File Name 
cpinf01 7/23/2003 12:39p 1,369 
cpinf01o 7/23/2003 12:39p 189,249 
cpinf02 7/23/2003 12:39p 1,369 
cpinf02o 7/23/2003 12:39p 186,863 
cpinf03 7/23/2003 12:39p 1,369 
cpinf03o 7/23/2003 12:39p 188,937 
cpinf04 7/23/2003 12:39p 1,369 
cpinf04o 7/23/2003 12:39p 189,249 
cpinf05 7/23/2003 12:39p 1,369 
cpinf05o 7/23/2003 12:39p 187,773 
cpinf06 7/23/2003 12:39p 1,369 
cpinf06o 7/23/2003 12:39p 189,873 
cpinf07 7/23/2003 12:39p 1,369 
cpinf07o 7/23/2003 12:39p 188,069 
cpinf08 7/23/2003 12:39p 1,369 
cpinf08o 7/23/2003 12:39p 190,020 
cpinf09 7/23/2003 12:39p 1,369 
cpinf09o 7/23/2003 12:39p 189,873 
cpinf10 7/23/2003 12:39p 1,369 
cpinf10o 7/23/2003 12:39p 188,457 
Time Size (bytes) Date 
MATHCAD (Seismic) Directory 
File Name 
Attachment III.mcd 02/10/2004 10:59a 221,219 
seepage glac lower Tptpll driftcollapse report.mcd 02/02/2004 04:03p 257,153 
seepage glac lower Tptpmn x1.2 report.mcd 02/02/2004 04:18p 257,683 
seepage glac mean Tptpll driftcollapse report.mcd 02/02/2004 04:07p 257,057 
seepage glac mean Tptpmn x1.2 report.mcd 02/02/2004 04:21p 257,299 
seepage glac upper Tptpll driftcollapse report.mcd 02/02/2004 04:14p 257,028 
seepage glac upper Tptpmn x1.2 report.mcd 02/02/2004 04:25p 257,363 
File Name Time Size (bytes) Date 
SAPHIRE Directory 
SAPHIRE.zip 01/14/2004 03:07p 231,569 
February 2004 VI-3 of VI-4 ANL-EBS-NU-000008 REV 00 Screening Analysis for Criticality Features, Events, and Processes for License Application 
INTENTIONALLY LEFT BLANK 
February 2004 VI-4 of VI-4 ANL-EBS-NU-000008 REV 00