Preclosure Consequence Analyses for License Application Rev 00B, ICN 00

000-00C-MGR0-00900-000-00B

January 2005 

1. PURPOSE 
Radiological consequence analyses are performed for potential releases from normal operations 
in surface and subsurface facilities and from Category 1 and Category 2 event sequences during 
the preclosure period. Surface releases from normal repository operations are primarily from 
radionuclides released from opening a transportation cask during dry transfer operations of spent 
nuclear he1 (SNF) in Dry Transfer Facility 1 (DTF I), Dry Transfer Facility 2 (DTF 2), the 
Canister Handling facility (CHF), or the Fuel Handling Facility (FHF). Subsurface releases from 
normal repository operations are from resuspension of waste package surface contamination and 
neutron activation of ventilated air and silica dust from host rock in the emplacement drifts. 
The purpose of this calculation is to demonstrate that the preclosure performance objectives, 
specified in 10 CFR 63.1 1 1 (a) and 10 CFR 63.1 1 1 (b), have been met for the proposed design 
and operations in the geologic repository operations area. Preclosure performance objectives are 
discussed in Section 6.2.3 and are summarized in Tables 1 and 2. 
Table 1. Performance Objectives for Normal Operations and Category 1 Event Sequences 
Performance Objectives 
Event Sequence Onsite Member Offsite Member 
Type Dose Type Worke? of the Public of the Public 
Normal ope'rations TEDE 5 remlyea? 100 mrem/yearc 15 mremlyeard 
and Category I 
100 mrem~~ea?~~ 
Normal operations Highest TODE 50 remlyear NA NA 
and Category 1 
Normal operations LDE 15 remlyear NA NA 
and Category 1 
Normal operations SDE 50 remlyear NA NA 
and Category 1 
Normal operations External dose: Highest of NA NA 2 mrem in 
and Category 1 DDE, LDE, or SDE for the any 1 hre 
unrestricted area 
NOTES: Radiation exposures and releases from Category I event sequences are aggregated per 
10 CFR 63.1 11(b)(l). 
a10 CFR 20.1201 
b10 CFR 20.1 101 (b) provides for establishment for ALARA goals 
"10 CFR 20.1 301 (a)(l) 
d10 CFR 63.204 
"10 CFR 20.1301 (a)(2). 
ALARA = as low as is reasonably achievable; DDE = deep dose equivalent; LDE = lens dose equivalent; 
mrem = one thousandth of a rem; NA = not applicable; rem = roentgen equivalent man; SDE = shallow dose 
equivalent to skin; TEDE = total effective dose equivalent; TODE = total organ dose equivalent. 
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Table 2. Performance Objectives for Category 2 Event Sequences 
Performance Objectives 
Event Onsite Member of Offsite Member 
SequenceType Dose Type Worker the Public of the Public 
Category 2 TEDE N A N A 5 remlevent 
Category 2 Highest TODE N A N A 50 remlevent 
Category 2 . LDE N A N A 15 remlevent 
Category 2 SDE N A N A 50 remlevent 
LDE = lens dose equivalent; NA = not applicable; rem = roentgen equivalent man; SDE = shallow dose equivalent 
to skin; TEDE = total effective dose equivalent; TODE = total organ dose equivalent. 
Source: 10 CFR 63.1 11 (b)(2) 
Category 1 event sequences are those natural events and human-induced event sequences that are 
expected to occur one or more times before permanent closure of the repository. Category 2 
event sequences are other human-induced event sequences that have at least one chance 
in 10,000-of occurring before permanent closure of the repository. Event sequences that have 
less than one chance in 10,000 of occurring before permanent closure of the repository are 
designated as beyond Category 2 event sequences. 
This calculation performs public dose calculations in Section 6.0. To. support the license 
application, this calculation integrates the results from public dose calculations with worker dose 
results from the following calculations: 
Canister Handling Facility Worker Dose Assessment (BSC 2004a) 
Category 1Eve& Sequences Worker Dose Calculation (BSC 2004b) 
Dry Transfer Facility Worker Dose Assessment (BS C 2004c) 
Fuel Handling Facility Worker Dose Assessment (BSC 2004d) 
Geologic Repository Operations Area Worker Dose Assessment (BSC 2004e) 
GROA Airborne Release Dispersion Factor Calculation (BSC 2004f) 
Normal Operation Airborne Release Calculation (BSC 2004g) 
Remediation Facility Worker Dose Assessment (BSC 2004h) 
Subsurface Facility Worker Dose Assessment (BSC 2004i) 
Transportation Cask Receipt and Return Facility Worker Dose Assessment (BSC 2004j). 
The results of these worker dose calculations (BSC 2004a through 2004j) are used, as required 
by 10 CFR 63.1 1 l(a), to perform worker dose aggregation in Section 6.4.3. Aggregated doses 
are reported in Section 7.0. 
Potential surface and subsurface releases from repository normal operations are discussed in 
Section 6.1.2. Event sequence frequency and categorization are discussed in Section 6.1.3. 
Listings of Category 1 and Category 2 event sequences and the types of waste forms and material 
at risk (MAR) involved in these event sequences are identified in Sections 6.1.4 and 6.1.5. 
This information is used as input to consequence analyses. 
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3.1 METHODOLOGY FOR NORMAL OPERATIONS AND CATEGORY 1 EVENT 
SEQUENCES 
This section discusses the methodology used for evaluating radiological consequences from 
normal operations in the surface and subsurface facilities and Category 1 event sequences. 
3.1.1 Normal Operations and Category 1 Event Sequence Performance Objectives 
Performance objectives for normal operations and Category 1 event sequences are provided in 
10 CFR Part 20 and 10 CFR Part 63, and are summarized in Table 1. They are discussed in 
Section 6.2.3. 
3.1.2 Normal Operations and Category 1 Event Sequence Public Dose Methodology 
This section discusses the method for calculating doses fiom normal operations and Category 1 
event sequences and dose aggregation as required by 10 CFR 63.1 1 1 (a)(2). 
In releases fiom normal operations, radioactive materials are conservatively assumed to be 
released inside a waste transfer cell when a transportation cask is opened. It is assumed that the 
heating, ventilation, and air-conditioning (HVAC) system is operating and the radioactive 
material is vented through the exhaust stack and is released as a radioactive plume that is 
dispersed en route to the site boundary. This results in an acute individual exposure during 
plume passage and a chronic individual exposure to ground contamination and contaminated 
food for 1 year after plume passage. A release duration of 24 hr is assumed for surface and 
subsurface releases (Section 4, Assumption 4.6). 
MACCS2 (ORNL 1998) is designed to simulate accidents with a release duration of 24 hr 
or less. MACCS2 (ORNL 1998), however, is used to simulate normal operations for which the 
release duration is one year. To model normal operations using MACCS2 (ORNL 1998), 
a release duration of 24 hr instead of one year is used. The use of a 24-hr release duration is 
conservative because: (1) a 24-hr release duration has a shorter time for radionuclide decay than 
a 1-year release duration, and (2) a 24-hr release duration results in a longer duration at higher 
ground concentrations of radionuclides and, therefore, higher resuspension inhalation, 
groundshine, and ingestion doses. The inhalation dose is not impacted by the use of a shorter 
release duration because it is not a function of the release duration. This is confirmed by 
sensitivity analysis (Section 6.3.4. I), which shows that modeling normal operations with a 24-hr 
duration instead of a 1-year duration is conservative because the shorter duration MACCS2 
(ORNL 1998) run shows higher doses (Section 6.3.4.1). 
No credit is taken for plume meandering or for plume rise from initial thermal energy to ensure 
that the x/Q value is conservative at the site boundary. Credit is taken for high-efficiency 
particulate air (HEPA) filters to remove airborne radionuclides. For normal operations, credit is 
taken for atmospheric dispersion of radionuclides released via the exhaust stack (Section 4, 
Assumption 4.7). Therefore, the release is modeled as an elevated release with no plume rise 
and the building wake effect on the plume is considered. 
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For Category 1 event sequences involving drops or collisions, or both, of SNF, it is assumed that 
radioactive materials are released inside a waste transfer cell. It is assumed that the HVAC 
system is operating and that radioactive material is vented through the exhaust stack as a 
radioactive plume. The plume is dispersed en route to the site boundary, resulting in an 
acute individual exposure during plume passage and in a chronic individual exposure to 
ground contamination and contaminated food for 1 year after plume passage (Section 4, 
Assumption 4.5). To ensure that the x/Q value is conservative at the site boundary, no credit is 
taken for plume meandering or for plume rise from initial thermal energy. Credit is taken for 
HEPA filters to remove airborne radionuclides. For Category 1 event sequences, no credit is 
taken for the stack height and the release is modeled as a ground-level release (Section 4, 
Assumption 4.17), and the building wake effect is considered. A release duration of 1 hr is 
assumed (Section 4, Assumption 4.5). 
The x/Q values are calculated internally in MACCS2 (OW 1998) using hourly meteorological 
data gathered at the Yucca Mountain site from 1998 through 2002 (Section 6.2.1.1.2). 
MACCS2 (ORNL 1998) uses 8,760 sets of hourly data on wind speed, wind direction, 
atmospheric stability class, and rain intensity collected over a period of 1 year to calculate 
xIQ values. This process generates 8,760 x/Q values, which includes the 16 directional sectors, 
wind speeds, and atmospheric stability classes collected in the annual data for 1999. The x/Q 
value output from MACCS2 (ORNL 1998) is a mean, 5oth percentile, 95th.percentile, or 99Sth 
percentile evaluated at a 11 km distance from the release point for a surface release, or at a 8 km 
distance for a subsurface release (Section 6.2.1.5 and Attachment B, Figure B-1). 
The x/Q values are determined regardless of wind direction or sector for the distance of interest. 
The x/Q values are overall site x/Q values and are not tied to any particular sector, such as the 
sector to the west'site boundary. Because all sector meteorological data are used in x/Q 
calculations and the receptor location (i.e., 11 km from a DTF to the west site boundary) is the 
shortest distance from the release point, the resultant x/Q values are the most conservative and 
bound all site boundary locations. 
The sector-independent MACCS2 (ORNL 1998)-calculated mean x/Q value of 2.69E-06 s/m3 
(Attachment A, Table A-1) is at least one order of magnitude higher than the maximum sector 
chronic X/Q value of 2.OE-07 s/m3 at a distance of 11 km, using Regulatory Guide 1.111, 
Equation 3, as calculated in Calculations of Acute and Chronic "Chi/QV Dispersion Estimates for 
a Surface Release (CRWMS M&O 1999a, Table 6c). 
Category 1 event sequences and normal operational releases use the mean x/Q value. The mean 
xIQ value was found to be more conservative than the 5oth percentile x/Q value. 
Source terms for spent fuel assemblies (SFAs) from pressurized water reactors (PWRs) and 
boiling water reactors (BWRs), with four different combinations of initial enrichment, burnup, 
and decay time, are presented in Table 3. 
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Table 3. Average and Maximum Boiling Water Reactor and 
Pressurized Water Reactor Spent Fuel Assemblies 
Initial 
Enrichment Burnup Decay Time 
Assembly (Percent) (GWdlMTU) (Years) 
Average PWRa 4.0 48 25 
Maximum PWRa 5.0 80 5 
Average BWR~ 3.5 40 25 
Maximum BWR~ 5.0 75 5 
NOTES: BSC 20041 , p. 27 
BSC 2003a , p. 46. 
BWR = boiling water reactor; GWd = gigawatt days; MTU = metric ton uranium; 
PWR = pressurized water reactor; SFAs = spent fuel assemblies. 
Radionuclide inventories in curies per fuel assembly (Ci/FA) for the Average and Maximum 
PWR and the Average and Maximum BWR SFAs are presented in Table 4. Both the Average 
PWR and the Average BWR SFAs are used to calculate the mean doses and the 5othpercentile 
doses for Category 1 event sequences. The calculated mean dose is used because it-is higher 
than the calculated 5othpercentile dose. 
Table 4. Boiling Water Reactor and Pressurized Water Reactor Radionuclide Inventories 
Average PWR Maximum PWR Average BbWR Maximum BWR 
Radionuclide (C~IFA)~ (CilFA)a (CilFA) (C~IFA)~ 
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Table 4. Boiling Water Reactor and Pressurized Water Reactor Radionuclide 
Inventories (Continued) 
Radionuclide Average PWR Maximum PWR Average BbWR Maximum BWR 
(CilFA)= (CilFA)= (CiIFA) (C~IFA)~
. 
Pu-238 2.29E+03 6.80E+03 5.85E+02 2.1 1 E+03 
Pu-239 1.77E+02 1.83E+02 5.35E+01 5.36E+01 
Pu-240 3.18E+02 4.01E+02 1.14E+02 1.48E+02 
Pu-241 2.46E+04 8.00E+04 6.78E+03 2.25E+04 
Pu-242 1.64E+00 3.34E+00 5.08E-01 1.26E+00 
Ru-1 06 1.23E-02 1.33E+04 3.00E-03 3.29E+03 
Sb-125 9.71 E+OO 2.14E+03 2.89E+00 6.21 E+02 
Se-79 4.57E-02 7.35E-02 1.59E-02 2.88E-02 
Sm-151 2.1 1 E+02 3.19E+02 5.39E+01 8.22E+01 
Sn-I 26 3.85E-01 6.83E-01 1.27E-01 2.52E-01 
Sr-90 2.72E+04 6.52E+04 9.54E+03 2.52E+04 
Tc-99 8.98E+00 1.34E+01 3.20E+00 5.35E+00 
Th-230 1.48E-04 3.33E-05 6.09E-05 2.05E-05 
U-232 2.04E-02 5.97E-02 4.63E-03 2.00E-02 
U-233 3.79E-05 2.42E-05 1.06E-05 0.00E+00 
U-234 6.77E-01 5.21 E-01 2.50E-01 2.26E-01 
U-235 7.37E-03 3.28E-03 2.62E-03 9.43E-04 
U-236 1.72E-01 2.23E-01 6.26E-02 9.55E-02 
. U-238 1.48E-01 1.42E-01 6.32E-02 6.07E-02 
Y-90 2.72E+04 6.53E+04 9.54E+03 2.52E+04 
Zr-93 8.94E-01 1.41 E+00 3.39E-01 6.03E-01 
NOTES: a BSC 20041 ,Attachment IX 
BSC 2003a, Attachment XIII. 
BWR = boiling water reactor; CiIFA = curies per fuel assembly; PWR = pressurized water 
reactor; SNF = spent nuclear fuel. 
Airborne release fractions (ARFs) and respirable fractions (FWs) for intact and damaged 
commercial SNF and naval SNF are shown in Table 5. Fuel rods with hairline cracks and 
pinhole leaks are treated as intact he1 rods as stated in ISG-1 (NRC 2002). ARFs and FWs for 
intact commercial SNF assemblies (Table 5) are used for normal operations and Category 1 
event sequences. ARFs and FWs for intact single commercial SNF rods (Table 5) are used for 
Category 2 event sequences because drops or collisions of canistered intact commercial SNF 
rods into transportation casks and waste packages are assumed to be Category 2 event sequences 
(Section 4, Assumption 4.9). 
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Table 5. Spent Nuclear Fuel Airborne Release Fractions and Respirable Fractions 
Airborne Release Respirable Effective Release 
Nuclide Fraction (ARF)a Fraction (RF)a Fraction (ARF x RF)a 
Breach of lntact SNF Assemblies and Rods 
0.3C 1.OC 0.3C
H-3 
Kr-85 0.3C 1.OC 0.3C 
0.3C 1.OC 0.3C
1-129 
CS 2.OE-04' 1.OC 2.OE-04' 
srd 3.0~-05~/5.9~-07' 5.0~-03~/1 1.5~-07~/5.9~-07'
.of 
Ru 2.OE-04' 1.OC 2.OE-04' 
Crud 1.5E-02C' 1.OC 1.5E-02' 
Fuel Fines other than Sr 3.0~-05~/5.9~-07' 5.0~-03~/1 1.5~-07~/5.9~-07'
.of 
Oxidation of Damaged Commercial SNF in ~ir~ 
H-3 0.3 1 .O 0.3 
Kr-85 0.3 1.O 0.3 
1-129 0.3 1.O 0.3 
CS 2.OE-04 1.O 2.OE-04 
srd 1.2E-04 1.O 1.2E-04 
RU 2.OE-04 1.O 2.OE-04 
Crud 1.5E-02g 1.O 1.5E-02 
Fuel Fines other than Sr 1.2E-04 1.O 1.2E-04 
NOTES: aARFs and RFs are for intact assemblies and intact single fuel rods; 
fuel rods with hairline cracks and pinhole leaks are treated as intact fuel rods 
Source: BSC 2004m. Section 6.2.1.3 and Table 11 
lntact commercial SNF assemblies and rods and naval SNF 
Sr is treated as fuel fines 
lntact commercial SNF assemblies and naval SNF 
f .lntact commercial SNF rods 
For crud, the value shown is the "effective ARF," which is the product of a crud 
spallation fraction of 0.15 and an ARF of 0.1 (BSC 2004m, .Section 6.2.1.3) 
Section 4, Assumption 4.16, and Section 6.2.1.3. 
ARF = airborne release fraction; RF = respirable fraction; SNF = spent nuclear fuel. 
Damaged commercial SNF could be oxidized in air during fuel transfer operations inside a waste 
transfer cell. In Table 5, for oxidation of damaged commercial SNF in air, the ARFs and RFs for 
fission product gases, volatile species, and crud are based on Assumption 4.16 (Section 4) and 
the ARFs and RFs for fuel fines are developed in Section 6.2.1.3. 
Potential radiation doses to members of the public that come fi-om inhalation, resuspension 
inhalation, ingestion, air submersion, and groundshine pathways are considered for normal 
operations and for Category 1 event sequences. Whole body and organ doses to a member of the 
public at the site boundary by inhalation, resuspension inhalation, ingestion, air submersion, and 
groundshine pathways are calculated using MACCS2 (OWL 1998, Section 5). 
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To show compliance with 10 CFR 63.1 11(a)(2), the calculated annualized dose, D,, , is 
compared with the regulatory dose limit of 15 mredyear at the site boundary. In addition, the 
calculated TEDE fiom each Category 1 event sequence and any combination of Category 1 event 
sequences that can occur in 1 year are also compared with the regulatory dose limit of 
15 rnredyear at the site boundary. 
3.1.3 Worker Dose Methodology 
As stated in Section 1, the results fiom worker dose calculations (BSC 2004a through 2004j) are 
reported in this calculation and used in Section 6.4.3 to perform worker dose aggregation. 
This section provides a more detailed discussion on the methodologies used in the worker dose 
calculations that are not provided in the referenced calculations (BSC 2004a through 2004j). 
This section describes the methods used to calculate worker dose during normal operations and 
Category 1 event sequences. Regulation 10 CFR 63.1 1 l(a)(l) requires that the geologic 
repository operations area must meet the requirements of 10 CFR Part 20 for radiation 
protection. Shielding must take normal operations and Category 1 event sequences into 
consideration. The primary objective of shielding is to maintain occupational radiation ' 
exposures as low as is reasonably achievable (ALARA), and within the exposure dose limits 
specified in 10 CFR 20.1201. The controlling dose limit for workers (Table 1) is a TEDE 
of 5 rernlyear in 10 CFR 20.1201(a)(l). The general approach used to estimate worker doses is 
to estimate radiation levels in occupied areas, to determine the personnel requirements and 
duration of activities in these areas, and to combine the dose rates, times, and personnel estimates 
to generate the worker dose estimates. 
ALARA design goals, as described in Project Design Criteria ~ocurnent(BSC 2004n, 
Section 4.9.3.3), for occupational workers ensure that both individual and collective annual doses 
are maintained at ALARA levels during normal operations and as a result of Category 1 event 
sequences. Category 1 event sequences are included in ALARA dose assessments. 
Sections 3.1.3.1 and 3.1.3.2 present the methodologies for calculating worker dose and the 
potential dose consequences fiom surface and subsurface releases during normal operations. 
Among the surface facilities, only DTF 1, DTF 2, and the FHF have a potential for airborne 
releases of radionuclides during normal operations or following a Category 1 event sequence. 
Section 3.1.3.5 presents the methodology for calculating the worker dose from direct radiation. 
3.1.3.1 Worker Dose From Airborne Releases 
It is assumed that, for radionuclides released from a waste transfer cell within a surface facility, 
-the HVAC system is operating and the airborne radionuclides are vented through the building 
exhaust stack, then disperse into the atmosphere and reenter the building through the building 
ventilation system air intakes (Section 4, Assumption 4.14). For radionuclides released fiom the 
subsurface facility, it is assumed that the airborne radionuclides are dispersed into the 
atmosphere and reenter the subsurface facility through the subsurface ventilation system air 
intakes (Section 4, Assumption 4.14). 
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The TEDE dose measure, described in ISG-5 (NRC 2003a, p. lo), is expressed as: 
+ 1~y~ ~
TEDE = CEDE + EDE = 1D;:>~~~~~ ; + 1DT&~~,~ ~ (Eq. 3) ~~ 
i i i 
where, 
TEDE = Total effective dose equivalent (rern) 
CEDE = Committed effective dose equivalent (rern) 
EDE = Effective dose equivalent (rem) 
D $h 
j,eftect;ve = Whole body effective inhalation dose from the jthnuclide (rem) 
D ?g
j,eftective = Whole body effective ingestion dose from the jthnuclide (rem) 
-
Whole body effective external dose from thejh nuclide (rem) 
D;&ective -
The TODE (= CDE + DDE) measure is expressed as (NRC 2003a, p. 10): 
TODE, = CDE, + EDE, = 1D: + D;:: + D; where k + effective or skin (Eq. 4) 
i i J 
where, 
CDE, = Committed dose equivalent to the ph organ (rern) 
D;:: = Radiation dose from the jthnuclide to the ph organ from inhalation (rern) 
Ding = Radiation dose from the jthnuclide to the ph organ from ingestion (rem) 
1.k 
D = Radiation dose from the jthnuclide to the ph organ from external exposure 
@em) 
-
k -Organ index, where organs are gonads, breast, lungs, red marrow, bone 
surface, thyroid, and remainder. 
The external dose is the sum of the groundshine dose and the air submersion dose. 
The groundshine dose has been shown (Section 6.3) to be a small contribution to the total dose 
and therefore the external dose is approximated by the air submersion dose. For example, 
MACCS2 (ORNL 1998) Run 4 (Section 6.3) shows that the groundshine dose at 100 m is about 
5.5E-04 mrem, which is less than 0.1 percent of the inhalation dose. 
The ingestion dose is not calculated for onsite workers because no ingestion of contaminated 
food, water, or soil is expected. 
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For airborne releases of radionuclides through building vents, the previous inhalation and 
external doses can be further expressed as (NRC 2003a, pp. 9 and 10): 
DFi = ST, x FA x -X x BR x conv x DCF;;
Q 
where, 
-
-
ST, Release source term per fuel assembly for thejth nuclide (Ci/FA) 
-
FA -Number of SFAs involved in the release per year or per event; FA = 6,316 
PWR SFAs for normal operations per year and FA = 2 PWR SFAs per 
event for Category 1 event sequences 
-X 
Q --Atmospheric dispersion factor (s/m3) 
BR - Breathing rate = 3.33 x m3/s (NRC 2000, p. 9-13) 
DCF,~~ Inhalation DCF of the jth nuclide for the kh organ in Sv/Bq
= 
(Eckerman et al. 1988, Table 2.1) 
DCF;? = Air submersion DCF of the jth nuclide for the khorgan [(sv-~~)/(B~-s)] 
(Eckerman and Ryman 1993, Table 111.1) 
conv = DCF unit conversion factor: 3.7 x 1012 . (rem-Bq)/(Ci-Sv) 
(Eckerman and Ryman 1993). 
The nominal annual capacity and rate of receipt is 3,000 MTHMIyear, as presented in Civilian 
Radioactive Waste Management System Requirements Document (DOE 2004b, Table l), or 
about 6,316 PWR SFAs per year based on 0.475 MTUISFA. It is assumed that 1 percent 
(Section 4, Assumption 4.3) of fuel rods in these SFAs are damaged.and that radionuclides are 
released to the environment. It is assumed that for Category 1 event sequences, an SFA is 
dropped back into a transportation cask and damages another SFA inside the cask during the dry 
transfer of SNF from a transportation cask to a waste package. 
Annual average x/Q values are generated using the NRC-sponsored computer code, Software 
Code: ARCON. V.96 (Atmospheric Relative CONcentrations in Building Wakes [ARCON96]), 
(BSC 2003b), as described in NUREGICR-6331 (Ramsdell and Simonen 1997). ARCON96 
(BSC 2003b) was developed by Pacific Northwest National Laboratory for the NRC to calculate 
xIQ values in plumes for nuclear power plants at control room air intakes in the vicinity of the 
release point. ARCON96 (BSC 2003b) implements a straight-line Gaussian dispersion 
model with dispersion coefficients modified to account for low wind meander and building 
wake effects. ARCON96 (BSC 2003b) also accounts for variations in the location of 
release points. 
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The xIQ value from normal operational releases and Category 1 event sequences are predicted 
using ARCON96 (BSC 2003b). For each receptor of interest, a cumulative probability 
distribution of xIQ is constructed by the code for release time periods of 2 hr, 4 hr, 8 hr, 
and 12 hr. The 8,760-hr probabilistic xIQ distribution is used to determine the annual average 
and the 50 percent weather probability median xIQ values for the receptor of interest. The larger 
of the annual average or median xIQ value is used to calculate the annual average dose to 
workers from airborne radionuclides during normal operations and from Category 1 
event sequences. 
The annual average x/Q value represents the average dilution of an airborne contamination from 
atmospheric mixing and turbulence based on the site-specific atmospheric conditions, the relative 
configuration of the release point and the receptor, and the distance from the release point to the 
receptor of interest. It is the ratio of the average contaminant air concentration at the receptor to 
the contaminant release rate at the release point, and it is used to determine the dose 
consequences for a receptor based on the quantity released, the atmospheric conditions, and the 
distance to the receptor. 
The xIQ values at various onsite receptor locations for airborne releases from surface and 
subsurface facilities were calculated using ARCON96 (BSC 2003b) and the results of the 
calculations are reported in BSC (2004f, Section 6.3). The maximum x/Q values at receptors for 
each of the seven release sources were used to calculate the maximum worker doses for releases 
fiom the surface and subsurface facilities from normal operations and Category 1 event 
sequences, respectively. These xIQ values represent the dispersion factors estimated at the air 
intake point of the surface and subsurface facilities. 
The airborne release source term is the amount of radioactive material in Ci that is released to the 
air per assembly under normal operating conditions or following a Category 1 event sequence. 
The airborne release source term, STj, is calculated by the following, fiom DOE-HDBK-3010-94 
(DOE 1994, p. 1-2): 
Release Source Term (CiIFA) ST, = MARj x DR x ARF x LPF (Eq. 7) 
where, 
MARj = Material at risk; commercial SNF radionuclide inventory per he1 
assembly (CiIFA) 
DR --
Damage ratio; rod breakage fraction (DR = 1.O) 
ARF = Airborne release fraction; the fraction of processed commercial SNF that 
is suspended or released from the SFAs 
LPF = Leak path factor; the fraction of airborne released material that discharges 
into the atmosphere from the HVAC exhaust systems; LPF = 1.OE-04 for 
normal operations and Category 1 event sequences. 
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The SDE is equal to (NRC 2003a, p. 10): 
SDE = D;'$ 
i 
where, 
SDE = Shallow dose equivalent to skin (rem) 
DJY~;~ = Radiation dose from the jthnuclide to the skin from air submersion (rem). 
It is stated in NUREG-1567 (NRC 2000, p. 9-14), that compliance with the LDE limit is 
achieved if the sum of the SDE and the TEDE does not exceed 15 rem; the LDE may be 
expressed as: 
LDE = TEDE + SDE (Eq. 9) 
where, 
LDE = Lens dose equivalent (rem). 
The assumptions in the following paragraph are consistent with the release fractions assumed in 
NUREG-1567 (NRC 2000, p. 9-12) and ISG-5 (NRC 2003a, Attachment, Table 7.1). 
For normal operations, 1 percent (Section 4, Assumption 4.3) of the fuel rods received are 
assumed to have cladding breaches that could cause the release of fission product gases, volatile 
species, and particulates in the gap region. It is assumed that for Category 1 event sequences 
involving drops and collisions, 100 percent of the fuel rods have failed because of the event 
(Section 4, Assumption 4.10). Of the total SFA radioactive inventory, the ARFs given in Table 5 
are used for Category 1 event sequences involving drops and collisions. 
Strontium isotopes including 90~r 
are considered as nonvolatile materials and are, therefore, 
treated as fuel particles (BSC 2004m, Section 6.2.1.2 and Table 11). 
For crud, a crud spallation fraction of 0.15 (BSC 2004m, Section 6.2.1.3) is used, which means 
that 15 percent of the crud becomes loose from the fuel rod surfaces under normal conditions and 
Category 1 event sequences. Of the loose crud, 10 percent (BSC 2004m, Section 6.2.1.3) 
becomes airborne and released during normal operations and Category 1 event sequences. 
An ARF of 10 percent represents the bounding release fraction for the case when venting gases 
pressurize the volume in which there exists loose powdering surface contamination (DOE 1994, 
p. 5-22). An ARF of 10 percent also bounds ARFs for other potential release mechanisms, such 
as venting of pressurized powders or pressurized gases through a powder (DOE 1994, 
Section 4.4.2.3) and suspension of surface contamination from solid material by impact and 
vibration shock (DOE 1994, pp. 5-7 and 5-24). The effective crud ARF is defined as the product 
of the crud spallation fraction (0.15) and the crud ARF (0.1). An effective crud ARF of 0.01 5 is 
given in Table 5. 
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3.1.3.3.6 Axial Power Peaking Factor 
Source terms in Tables 6 and 7 represent the values for uniform fuel burnup; a relative power 
factor of 1 for the entire length of the SFA. Shielding source terms account for the power 
peaking factor to obtain the peak dose rate, because fuel burnup varies axially along the length of 
the SFA. 
Table 6. Maximum and Design Basis Pressurized Water Reactor 
Spent Nuclear Fuel Neutron Source Terms 
NOTES: a BSC 20041, Attachment X; PWR fuel with 5 percent enrichment, 
80 GWdlMTU burnup and 5-year cooling 
BSC 20041, Attachment X; PWR fuel with 4 percent enrichment, 
60 GWdIMTU burnup and 10-year cooling. 
MeV = million electron volt; nls = neutrons per second. 
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Table 7. Maximum and Design Basis Pressurized Water Reactor 
Spent Nuclear Fuel Gamma Source Terms 
Maximum Source Design Basis Source 
(yls per assembly)* (yls per assembly)b 
Gamma Energy 
Range (MeV) Bottom Plenum Top Fuel Bottom Plenum Top
Upper-Lower Region Region Region Region Region Region Region Region 
5.00E-02 1.00E-02 2.33E+15 5.94E+11 5.28E+ll 3.79E+I 1 1.21 E+l5 2.73E+I 1 1.88E+l I 1.75E+l I 
1.00E-01 5.00E-02 6.44E+14 1.16E+ll 6.09E+10 7.43E+10 3.29E+14 5.28E+10 2.77E+10 3.39E+10 
2.00E-01 1.00E-01 5.22E+14 2.83E+10 3.52E+10 1.79E+10 2.45E+14 1.28E+10 1.17E+10 8.19E+09 
3.00E-01 2.00E-01 1.48E+14 1.41 E+O9 1.96E+O9 8.91 E+O8 7.1 3E+l3 6.39E+08 6.33E+08 4.07E+08 
4.00E-01 3.00E-01 9.85E+13 1.90E+09 5.86E+09 1.17E+09 4.55E+13 8.50E+08 1.64E+09 5.33E+08 
6.00E-01 4.00E-01 1.53E+15 1.91 E+09 I.lOE+l I 7.41 E+07 2.26E+14 4.92E+08 2.69E+10 3.37E+07 
8.00E-01 6.00E-01 4.70E+15 4.35E+09 5.95E+10 2.37E+09 2.37E+15 2.91 E+09 1.60E+10 1.86E+09 
1.00E+00 8.00E-01 7.08E+14 1.37E+ll 8.03E+09 7.66E+10 1.22E+14 5.40E+09 2.48E+09 3.41 E+09 
1.33E+00 1.00E+00 4.55E+14 3.38E+13 1.74E+I3 2.17E+13 1.95E+14 1.54E+13 7.97E+12 9.90E+12 
1.66E+00 1.33E+00 1.30E+14 9.53E+12 4.91 E+12 6.12E+12 4.50E+13 4.35E+12 2.25E+12 2.80E+12 
2.00E+00 1.66E+00 1.44E+12 1.87E+03 9.1 9E+O2 1 .I3E+O3 1.52E+11 2.35E+00 1.49E+O2 2.1 5E-02 
2.50E+00 2.00E+00 2.49E+12 2.26E+08 1.16E+08 1.45E+08 5.1 7E+lO 1.03E+08 5.34E+07 6.64E+07 
3.00E+00 2.50E+00 1.10E+11 3.51 E+05 1.81 E+05 2.25E+05 3.79E+09 1.60E+05 8.29E+04 1.03E+05 
NOTES: a BSC 20041, Attachment X; PWR fuel with 5 percent enrichment, 80 GWdlMTU burnup and 
5-year cooling 
BSC 20041, Table 10; PWR fuel with 4 percent enrichment, 60 GWdlMTU burnup and 
10-year cooling. 
MeV = million electron volt; yls = photons per second. 
For PWR fuel, shielding calculations use a conservative peaking factor of 1.25, based on the 
predicted heat profile for a PWR assembly provided in The TN-24P PWR Spent-Fuel Storage 
Cask: Testing and Analyses (Creer et al. 1987, p. 3-29). This factor is directly applied to the 
calculated gamma dose rate as a multiplier; the gamma source strength is reasonably 
proportional to fuel burnup. The multiplier for the neutron dose rate is determined to be the ratio 
of the neutron source strength at the peak burnup to that for uniform burnup, as described in 
Dose Rate Calculation for 21-P WR Waste Package (BSC 2004p, Section 5.2.1). 
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SCALE V4.3, from NUREGICR-0200 (ORNL 1997), contains the modules for performing 
source term and shielding calculations. For Yucca Mountain applications, SCALE can be used 
for both source term and shielding calculations. The ability of updating bumup-dependent cross 
section library in SCALE provides a more accurate determination of the source terms than the 
ORIGEN code provided in RSIC Computer Code Collection, ORIGEN 2.1, Isotope Generation 
and Depletion Code, Matrix Exponential Method (ORNL 1991). 
QAD-CGGP (CRWMS M&O 1995) and PATH (Su et al. 1987) are both point-kernel integration 
codes for gamma shielding problems, capable of explicitly treating three-dimensional source 
shield configurations. These two codes are similar in capabilities and produce results in 
satisfactory agreement. PATH provides additional features to treat multiple sources and various 
source types in a single run. 
Flux-to-DCFs (BSC 2002, Section 5.6.1) used in the shielding analyses are consistent with 
ANSIIANS-6.1.1-1977, pp. 4 and 5. No degradation or loss of shielding materials from an event 
sequence has been identified. 
The computer codes listed in this section are not used to produce results in this calculation. 
3.1.3.5 Worker Dose From Direct Radiation 
During the preclosure period, some surface and subsurface facility workers could be exposed to 
direct radiation when work requires them to be in close proximity to contained sources, such as 
transportation casks, storage casks, or waste packages. If necessary, fixed or portable shielding 
materials are used to reduce the dose rates to which workers are exposed. The dose rate at a 
distance from the contained source was calculated using the MCNP computer program 
(Briesmeister 1997). The dose assessment involves calculations of annual individual doses 
to workers. The dose contributions from contained radiation sources, such as SFAs in a 
transportation cask or waste package, are obtained from the shielding calculations using NRC 
accepted computer codes, such as MCNP. These shielding calculations generate dose maps 
around hctional areas within the facilities for determining dose rates at worker locations. 
Dose assessments are performed by job hction or worker group using time-motion inputs and 
dose rates calculated at worker locations. The time-and-motion inputs define the process step, 
location, number of workers, and duration of worker occupancy. The individual doses are 
calculated for each process step and summed over the process steps to obtain cumulative external 
exposures to workers on an annual basis. Outputs of the assessment consist of a matrix of 
operations, locations, source, frequency, dose rates, stay times, and doses. In addition, to 
demonstrate regulatory compliance, calculated annual individual doses are used for comparison 
with the ALARA design goal to minimize the number of individuals who have the potential of 
receiving more than 500 mremlyear TEDE (BSC 2004n, Section 4.9.3.3). 
The Transportation Cask Receipt and Return Facility, (TCRRF) provides space, layout, and 
structures that support waste handling operations. Waste handling operations, including 
unloading transportation casks from the transportation carrier, integrate with the onsite cask- 
handling system within the TCRRF protective structure to support the throughput rates 
established for waste emplacement. 
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TCRRF worker groups include: transportation personnel, cask operator, health physicist, and 
gantrylcrane operator. Transportation personnel duties include the movement of a full or empty 
casklcarrier into and out of the transfer bay. The cask operator and gantrylcrane operator work 
together and perform duties that involve removal of personnel barrierslimpact limiters, cask 
transfers from a carrier to a site rail transfer cart, and inspection of casks. The health physicist 
monitors radiation and contamination levels during cask receipt and transfer. 
The DTF provides space, layout, structures, and equipment that transfer uncanistered SFAs from 
transportation casks to waste packages and prepare sealed waste packages for subsurface 
transport and emplacement. Waste handling activities include transportation cask and waste 
package receipt and processing and transfer bay operation. DTF worker groups include: the cask 
and waste receipt operator, transfer bay operator, and health physicist. 
The FHF is a surface facility supporting waste handling operations. The facility receives 
transportation casks, unloads and transfers their contents to waste packages or site specific casks, 
transfers the site specific casks to the aging area, prepares waste packages, and transfers waste 
packages to a transporter for subsurface emplacement. FHF worker groups include: the cask and 
waste receipt operator, transfer bay operator, and health physicist. 
The annual external dose from the contained source in casks per individual worker is 
calculated as (BSC 2004j, Equation 1): 
where, 
-
-Annual external dose per individual worker per work group g (mremlyear)
ED, 
DR,, = External dose rate at location i per individual worker per work 
group g @re&)
-
c. -Duration of the exposure at location i, per cask operation (hrlcask) 
-
TC -Number of processed transportation casks per year (caskslyear) 
WS --
Number of work shifts. 
Worker locations for cask preparation operations are categorized in terms of distances from the 
transportation cask surface. The distances are estimated from the most likely worker locations to 
perform the specific tasks for the cask preparation operations. For hands-on activities, such as 
swipes for surface contamination sampling, the worker is assumed to be 1 m from the cask. 
For processing tasks that are not hands-on, but require worker presence in the area, the worker is 
assumed to be standing at a reasonable distance from the cask. For processing transportation 
casks, workers are assumed to be at one of three distances from the exterior surfaces of a 
transportation cask: 1 m, 5 m, or 10 m. Dose rates in the vicinity of a transportation cask are 
estimated using the 2-m dose rates of the Universal Transport Cask and the radial dose distance 
factors derived from the TN-32 cask (BSC 2004c, p. 11). Average dose rates from the cask 
surface are: 9.10 mrem/hr at 1 m, 1.66 mrem/hr at 5 m, and 0.51 mrem/hr at 10 m (BSC 2004c, 
p. 11). 
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where, 
D,,., 
--Annual dose from Category 1 event releases (mredyear) 
0; 
--
Annual dose from normal operational releases from surface facilities 
(mredyear) 
D$ = Annual dose from normal operational releases from the subsurface 
repository (mredyear) 
0;: --Annual dose from direct radiation fiom contained sources (mredyear). 
Worker dose assessment includes individual doses for both normal operations and Category 1 
event sequences. Total dose, including internal and external exposures, is calculated on an 
annual basis by summing the contributions from normal operations and frequency-weighted 
doses from Category 1 event sequences for demonstration of regulatory compliance. Equation 2 
is used to calculate annualized worker dose from Category 1 event sequences. 
Worker dose assessment includes contributions fiom both surface and subsurface facilities on a 
building-by-building basis; the assessment of subsurface facilities also includes contributions 
fiom operations and activities performed underground. 
3.2 METHODOLOGY FOR CATEGORY 2 EVENT SEQUENCES 
This section discusses the methodology used for evaluating radiological consequences from 
Category 2 event sequences. 
3.2.1 Category 2 Event Sequence Performance Objectives 
The four dose measures of LDE, SDE, TEDE, and TODE used' to evaluate normal operations 
and Category 1 event sequences are also used to evaluate Category 2 event sequences. 
Performance objectives (Section 6.2.3) for Category2 event sequences are summarized in 
Table 2. 
3.2.2 Category 2 Event Sequence Public Dose Methodology 
In Category 2 event sequences, radioactive materials are assumed to be released as a 
ground-level radioactive plume (Section 4, Assumption 4.17) and the building wake effect is 
considered. The plume is dispersed en route to the site boundary, resulting in an acute individual 
exposure during plume passage and a chronic individual exposure to ground contamination and 
contaminated food for 1 year after plume passage (Section 4, Assumption 4.5). No credit is 
taken for plume meandering or for plume rise caused by initial thermal energy to further ensure 
that the x/Q value is conservative at the site boundary. Potential radiation doses from inhalation, 
resuspension inhalation, ingestion, air submersion, and groundshine pathways are considered for 
Category 2 event sequences. Inhalation, resuspension inhalation, ingestion, air submersion, and 
groundshine doses are calculated using MACCS2 (ORNL 1998). 
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The xIQ
values are calculated internally by MACCS2 (ORNL 1948) using hourly meteorological 
data gathered at Yucca Mountain from 1998 through 2002. The 1999 meteorological data are 
values (Section 6.2.1.1.2). The xIQ
used because the data generate conservative xIQ value 
output from MACCS2 (ORNL 1998) for surface releases is a mean, 5oth percentile, 95th 
percentile, or 99.5th percentile evaluated at 11 km for surface releases (Section 6.2.1.5 and 
Attachment B, Figure B-1). 
Category 2 event sequences use the 95" percentile with a 5 percent exceedance of the xIQ
value 
evaluated in accordance with Regulatory Guide 1.145, Section 3, at the closest site boundary 
of 11 km for the maximally exposed individual without regard to sectors. 
Commercial SNF Source Term-Source terms for PWR and BWR SFAs with four different 
combinations of initial enrichment, burnup, and decay time are considered. These combinations 
are presented in Table 3. Radionuclide inventories in Ci/FA for Maximum PWR and Maximum 
BWR SFAs are presented in Table 4. For Category 2 event sequences, both Maximum PWR and 
Maximum BWR SFAs are used to calculate maximum doses; the 95" percentile dose in 
accordance with Regulatory Guide 1.145, Section 3. 
The ARFs and RFs for intact commercial SNF rods (Table 5) are used for Category 2 event 
sequences, because the combined ARF x RF values for intact commercial SNF rods are larger or 
equal to the values for intact commercial SNF assemblies. 
Naval SNF Source Term-A bounding radionuclide release source term was developed which 
covers all naval SNF canister inventories and internal hardware designs. Conservative estimates 
of SNF inventory, crud inventory, fuel damage ratio, and release fractions were used to develop a 
bounding radionuclide release source term which will result in doses larger than those expected 
should an actual Category 2 event sequence occurs. 
For naval SNF assemblies, a conservative radionuclide inventory for a canister of naval SNF at 
five years after shutdown is developed using Naval Nuclear Propulsion Program depletion codes 
in conjunction with ORIGEN-S, which have been qualified for use with naval with naval SNF in 
repository applications. The type of SNF assembly that is used in the analysis is the one that 
represents a majority of the naval SNF assemblies that will be emplaced. The inventory assumed 
up to 10 percent more SFAs in a canister that will be loaded. 
The crud contribution to the source term for naval SNF is calculated using the standard naval 
program shielding procedure. This value is increased by a factor of 2.5 to provide a conservative 
crud concentration at core shutdown for use in developing the source term. 
The fuel and crud damage ratios used to developed the source term were conservatively selected 
to ensure that all canister radionuclide inventories and internal hardware designs are bounded. 
The ARFs and RFs for intact commercial SNF assemblies (Table 5) are used for Category 2 
event sequences involving naval SNF. 
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HLW Source Term-Bounding per-canister HLW radionuclide data were obtained from four 
different DOE sites, including the Savannah River Site, Hanford Site, West Valley, and Idaho 
National Engineering and Environmental Laboratory (INEEL). Projected bounding per-canister 
radionuclide inventory data for nuclides were set equal to the maximum values from sludge 
batches of wastes, vitrified between October 10, 1994 and March 17, 2003, as presented 
in Projected Glass Composition and Curie Content of Canisters from the Savannah River 
Site (U) (Fowler 2003, Section 5.0). The methodology used for Category 2 event sequences 
involving HLW is to demonstrate that the Savannah River Site HLW bounds HLW from the 
other DOE sites. 
In summary, the methodology for Category 2 event sequence public dose calculations is to 
maximize the public dose to account for the variability or uncertainty, or both, in MAR, source 
terms, release fractions, and varying weather conditions. Weather conditions relevant to dose 
calculations include wind speed, wind direction, stability class, and precipitation. 
MACCS2 (ORNL 1998) randomly samples the Yucca Mountain site-specific hourly data over 
a 1-year period, producing 8,760 sets of wind speed, wind direction, stability class, and 
precipitation to calculate x/Q values. 
Each set of weather data calculates one value of xIQ. Depending of the number of sampling 
chosen, thousands of x/Q values could be generated in one MACCS2 (ORNL 1998) run. 
Statistical analysis is performed internally in MACCS2 on these xIQ values. Values for the 
mean, 5oth percentile, 95" percentile, and 99.5th percentile are generated and printed on the 
MACCS2 (ORNL 1998) output file. 
4. ASSUMPTIONS 
The following assumptions, except for Assumption 4.14, are used for calculating public dose; 
Assumption 4.14 is used in calculations in other documents for calculating worker dose. 
4.1 Waste forms involved in normal operations and in Category 1 and Category2 
event sequences include PWR or BWR SFAs, HLW, and naval SNF. 
Basis: The only waste form not included is DOE SNF, because a drop and breach of 
aDOE SNF canister, including the 18 in. or 24 in. standardized canister or 
themulticanister overpack is a beyond Category2 event sequence (BSC 2004k, 
Section 4.2.3). 
This assumption is used in Section 6.1 
4.2 At the highest nominal receipt rate, 3000 MTHM (DOE 2004b, Table I), of commercial 
SNF pass through the CHF, DTF 1, DTF 2, and the FHF each year. It is assumed that for 
the purposes of calculating worker and public doses, fuels received are PWR SFAs. 
Using an average PWR assembly weight of 0.475 MTHMISFA (BSC 20041, p. 8), the 
highest nominal throughput is 6,3 16 SFAsIyear. 
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Basis: It is conservative to assume that 3,000 MTHM of commercial SNF is PWR SFAs, 
because PWR SFAs typically have higher enrichment and burnup than BWR SFAs 
(Table 3). SNF with a higher enrichment and a higher burnup, in general, has a hgher 
total radionuclide inventory. 
Ths assumption is used in Section 6.1.2.1. 
4.3 One percent (i.e., DR of 0.01) of the fuel rods received at the repository are modeled as 
having defect sizes equal to pinhole leaks or hairline cracks and the fission product gases, 
volatile species, and fuel fines are released. Releases from one percent of the 6,316 SFAs 
are used as the source term for the calculation of normal operations doses. 
Basis: The 1 percent fuel breakage assumption is from ISG-5 (NRC 2003a, Attachment, 
p. 7) for normal operations. Historical fuel discharge data have shown that the fuel rod 
failure rate is about 0.01 percent, versus the range of 0.02 to 0.07 percent in the first 
20 years of commercial nuclear power, as described in 2002 Waste Stream Projections 
Report (BSC 2003c, p. G-6). These numbers are much smaller than the 1 percent fuel 
failure rate. 
This assumption is used in Sections 3.1.3.1 and 6.1.2.1 
4.4 The HEPA filters of the surface facility HVAC systems are assumed to be unavailable to 
remove radionuclides for Category 2 event sequences. HEPA filters are assumed to be 
functioning for surface facility normal operations and Category 1 event sequence dose 
calculations. 
Basis: It is conservative to not take credit for the HEPA filters to mitigate Category 2 
event sequences. Credit is taken for the HEPA filters to mitigate normal operational 
releases and Category 1 event sequences. 
This assumption is used in Section 6.1.4,6.1.5, and 6.2.1.3. 
4.5 It is assumed that for Category 1 and Category 2 event sequences, radioactive materials 
are released in a 1 -hr duration. 
Basis: The 1-hr release duration is conservative when compared with the 2-hr x/Q values 
specified in Regulatory Guide 1.145, Section 1.3. 
This assumption is used in Sections 3.1.2, 3.2.2, 6.1.4, 6.1.5, and 6.2.1.5. 
4.6 For calculating public doses from normal operations using MACCS2 (ORNL 1998), the 
release duration is assumed to be 24 hr, which is the longest duration allowed by 
MACCS2. The release is assumed to result in an acute individual exposure during plume 
passage and a chronic individual exposure to ground contamination and contaminated 
food after plume passage. The period of long-term exposure to ground contamination 
and intake of contaminated food is 1 year. 
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Basis: The sensitivity run discussed in Section 6.3.4.1 showed that the use of the release 
duration of 24 hr, instead of 1 year, for normal operations is conservative. 
This assumption is used in Sections 3.1.2,6.2.1.2,6.2.1.5, and 6.3.4.1. 
It is assumed that radionuclides are released from surface facilities during normal 
operations via the exhaust stack. 
Basis: The CHF, DTF 1, DTF 2, and the FHF have a vent that is higher than the adjacent 
structures. For surface facility normal operations, credit is taken for the elevated release 
to disperse radionuclides in the air. 
This assumption is used in Sections 3.1.2 and 6.3. 
For normal operations and Category 1 event sequences, a two-stage HEPA filtration 
system with a particulate removal efficiency of 99 percent per stage (i.e., a HEPA leak 
path factor (LPF)HEPA of O.Ol/stage) is assumed. This gives a combined efficiency 
of99.99 percent for two stages; a HEPA LPF of It is further assumed that the 
HVAC system is removing particulates and cesium in air through two stages of HEPA 
filters in series that are protected by pre-filters, sprinklers, and demisters. 
Basis: A HEPA filter LPF of is more conservative than the value of 2 x for two 
HEPA filters, as presented in NUREGICR-6410 (SAIC 1998, Section F.2.1.3). 
NUREGICR-0722 test data (Lorenz et al. 1980, Table 19) shows that two-stage HEPA 
filters capture almost 100 percent of incoming airborne cesium. NUREGICR-0722 test 
data is applicable because the test was conducted at a temperature of 900C (Lorenz et al. 
1980, p. 48), which conservatively bounds SNF temperatures inside a waste transfer cell 
under normal or accident conditions. 
This assumption is used in Sections 3.1.3.1 and 6.2.1.3. 
Drops or collisions of canistered single commercial SNF rods in transportation casks 
and waste packages are assumed to be Category 2 event sequences. 
Basis: Drops or collisions of commercial SNF in transportation casks and waste packages 
are Category 2 event sequences (BSC 2004k, Section 7). Canistered single commercial 
SNF rods are only a subset of commercial SNF. Therefore, the frequency of drops or 
collisions of canistered single commercial SNF fuel rods in transportation casks and 
waste packages is lower than the frequency of drops or collisions of commercial SNF in 
transportation casks and waste packages. Thus, drops or collisions of canistered single 
commercial SNF rods in transportation casks and waste packages are Category 2 
event sequences. 
This assumption is used in Sections 3.1.2 and 6.2.1.3. 
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..lo For Category 1 and Category 2 event sequences, the DR is assumed to be 1.0 for 
commercial SNF. For Category 2 event sequences, the DR is assumed to be 1.0 for HLW 
in a canister. The DR is the fraction of fuel rods that is assumed to fail by cladding 
breach during an event sequence or the fraction of HLW that is damaged by crush or 
impact, or both. Bounding damage ratios are used for Category 2 event sequences 
involving naval SNF, as documented in NNPP Input for Yucca Mountain Project 
Preclosure Safety Analyses (Gisch 2004, p. 1). 
Basis: These are bounding assumptions for commercial SNF, HLW, and naval SNF. 
This assumption is used in Sections 3.1.3.1,6.1.4, and 6.2.1.3. 
4.11 For Category 2 event sequences, the cask leak path factor (LPF),,Sk is assumed to be 0.1 
for SNF in a transportation cask and 0.01 for HLW in a canister in a transportation cask, 
and 1.0 for naval SNF in a transportation cask or in a canister. The (LPF),,,k is the 
fraction of the ARF that reaches the ventilation system after local deposition, consisting 
of plate-out and gravitational settling, within a transportation cask. 
Basis: This is a conservative assumption. For rail casks involved in end, comer, or side 
impacts, at an impact speed of 60 mph, NUREGICR-6672 (Sprung et al. 2000, 
Table 7.19) calculated an (LPF),,Sk of 0.02 for particles and 0.0008 for cesium, which are 
smaller than the conservatively assumed (LPF),,Sk of 0.1. Both the canister and the cask 
are assumed to have an LPF of 0.1 because HLW is shipped inside of a canister 
within a transportation cask. The combined LPF for HLW is equal to 0.01 (i.e., 0.1 x 0.1 
= 0.01). An (LPF),,sk of 1.0 for naval SNF is conservative. 
This assumption is used in Section 6.2.1.3. 
4.12 For normal operations and Category 1 event sequences, the facility leak path factor 
(LPF)faC is conservatively assumed to be 1.0 inside a waste transfer cell. The (LPF)fa, is 
the fraction of the ARF that reaches the ventilation system after local deposition, 
consisting of plate-out and gravitational settling, within a surface facility. 
Basis: This is a bounding assumption because no credit is taken for local deposition of 
particulates within a surface facility. 
This assumption is used in Section 6.2.1.3. 
4.13 It is assumed that the HVAC system is operating and no airborne material released from 
Category 1 event sequences leaks into space occupied by workers who work in rooms 
adjacent to a waste transfer cell in a DTF or the FHF. 
Basis: This assumption is reasonable because the transfer cell confinement and the 
HVAC system are designed to prevent any leakage to rooms adjacent to the transfer cell 
in the event of a Category 1 event sequence. 
This assumption is used in Sections 6.1.2.1 and 6.1.4. 
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4.14 For normal operations and Category 1 event sequences, it is assumed that for 
radionuclides released from a waste transfer cell within a surface facility, the HVAC 
system is operating and airborne radionuclides are vented through the building exhaust 
stack, dispersed into the atmosphere, and then reenter the building through the building 
ventilation system air intakes. It is assumed that for radionuclides released from the 
subsurface facility, airborne radionuclides are dispersed into the atmosphere and reenter 
the subsurface facility through the subsurface ventilation system air intakes. 
Basis: The calculation of worker doses considers that the concentrations within the 
facility are at equilibrium with the concentration outside of the air intake, which is the 
major source of air into the facility. Other pathways, such as through the entrance air 
locks would contribute concentrations at lower levels than at the air intake because the 
dispersion at these locations will be greater than at the air intake. Therefore, these 
pathways were conservatively neglected. The subsurface ventilation system air intakes 
are the only places that airborne radionuclides could enter the subsurface facility. 
This assumption is used in Section 3.1.3.1. 
4.15 The maximally exposed individual is defined as an individual located at a distance that 
corresponds to the approximate distance between the surface facility or the subsurface 
repository and the nearest point of public access on the repository site boundary, which 
lies to the west. The proposed Land Withdrawal Area boundary (Attachment By 
Figure B-1) is assumed to be the site boundary (i.e., preclosure controlled area boundary). 
A site boundary distance of 11 km (Attachment By Figure B-1) is used to calculate x/Q 
values from radiological releases from the surface facility. This distance corresponds to 
the distance from the DTF ventilation exhaust shaft to the nearest point on the site 
boundary that is the closest point where any member of the public could be standing or 
living at the time of a postulated radiological release. A site boundary distance of 8 km 
(Attachment By Figure B-1) is used to calculate x/Q values from radiological releases 
from the subsurface repository. This distance corresponds to the approximate distance 
between the subsurface repository and the nearest point of public access on the site 
boundary, which lies to the west. 
Basis: The use of the shortest distances of approximately 11 km and 8 km from the 
surface and subsurface release points, respectively, to the site boundary to calculate the 
public dose is conservative. 
This assumption is used in Sections 6.2.1.5,6.3.3, and 6.3.5. 
4.16 It is assumed that the fission product gas, volatile species, and crud release fractions for 
breaches of intact commercial SNF assemblies and rods in Table 5 are applicable to 
releases of fission product gases, volatile species, and crud during oxidation of damaged 
commercial SNF in air in Table 5 (BSC 2004m, Section 6.2.1.3 and Table 1 1). 
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Basis: The volatile species release fractions for breaches of intact commercial SNF 
assemblies and rods given in Table 5 bound the data on 137~s release fractions 
and Io6~u 
during oxidation of damaged commercial SNF in air (i.e., 1.4 x for "'cs, and 7.27 x 
for Io6~u) 
from NUREGICR-6672 (Sprung et al. 2000 p. 7-45). The 30 percent 
release fraction for gases for breaches of intact commercial SNF assemblies and rods 
given in Table 5 bounds the fission product gas release fractions during oxidation of 
damaged commercial SNF in air given in "Effects of an Oxidizing Atmosphere in a Spent 
Fuel Packaging Facility" (Einziger 1991, Figure 8). Because the crud is deposited on the 
outer surfaces of commercial SNF assemblies or rods, the release fraction is expected to 
be the same for both intact and damaged commercial SNF. 
This assumption is used in Section 6.2.1.3. 
The releases from the subsurface exhaust shafts during normal operations and the releases 
from the surface facilities from Category 1 and Category 2 event sequences are assumed 
to be at ground level. 
Basis: An occurrence of a ground-level release of radioactive material, instead of an 
elevated release from the subsurface exhaust shafts, is conservative. Releases from 
Category 1 event sequences are assumed to exit surface facilities through a stack and are 
conservatively modeled as ground level releases. Releases from Category 2 event 
sequences are also modeled as ground level releases. 
This assumption is used in Sections 3.1.2, 3.2.2,6.1.2.1,6.1.2.2,6.1.4, and 6.3. 
For normal operations, and Category 1 and Category 2 event sequences, no credit is taken 
for charcoal adsorbers to remove radionuclides. 
Basis: Currently there is no plan to install charcoal adsorbers. If in the future charcoal 
adsorbers are installed, this assumption would result in-a slightly higher dose and, 
therefore, it is a conservative assumption. 
This assumption is used in Section 6.2.1.3. 
It is assumed that the canister handling system is designed so that a drop of an HLW 
canister inside a surface facility will not exceed a drop height of 276 in. (23 ft). 
Basis: A design requirement will limit lift heights for HLW canisters to less than 23 ft 
above the bottom of a cask, waste package, staging rack, or staging pit (BSC 2004k, 
Section 5.1.1 X). 
This assumption is used in Section 6.2.1.3. 
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4.20 The maximally exposed individual at the site boundary is assumed to receive doses from 
the inhalation, resuspension inhalation, air submersion, groundshine, and ingestion 
pathways for a period of 8,760 hr. The onsite individual member of the public, at 100 m 
or 3 km away from a DTF, the FHF, or a subsurface exhaust shaft is assumed to receive 
doses from inhalation, resuspension inhalation, air submersion, and groundshine 
pathways for a period of 2,000 hr. 
Basis: It is expected that onsite ground contamination after event sequences will be 
detected and removed prior to restart of the repository operation. Unlike facility 
workers who work 2,000 hr a year, an individual member of the public is expected 
to visit the Yucca Mountain site for only a few days in a year. Therefore, the assumption 
of a 2,000-hr exposure period is conservative. One hundred meters is a conservative 
distance from the surface facility stack to any location outside the restricted area. 
This assumption is used in Sections 6.2.1.1.3,6.2.1.5,6.3.1,and 6.3.3. 
4.21 It is conservatively assumed that radionuclides are released from a height of 30 m during 
surface facility normal operations. This release height is used to calculate x/Q values for 
use in dose calculations for surface facility normal operations. 
Basis: A release height of 30 m is smaller than the stack heights of 40 m for DTF 1, 
of 40 m for DTF 2, or of 36 m for the FHF (BSC 2004f, Tables I-1,I-2, and Table 1-14). 
A lower release height would result in a higher radiation dose. Hence, a release height 
of 30 m is conservative. 
This assumption is used in Section 6.2.1.1. 
4.22 For members of the public at the site boundary, the period of long-term exposure to 
ground contamination and intake of contaminated food is 1 year. 
Basis: The assumption of 1-year exposure to ground contamination and intake of 
contaminated food is conservative when compared with an exposure period of 30 days 
recommended in ISG-5 (NRC 2003a, p. 9-1 5). 
This assumption is used in Sections 3.1.2, 3.2.2, 6.1.4,6.1.5, and 6.2.1.5. 
4.23 It is assumed that 154 failed fuel rods per year are expected to be vulnerable to oxidation 
and cladding unzipping. 
Basis: The number of failed SFAs expected to be received at the repository is estimated 
to be approximately 4 percent of the total, using data from NUREGICR-3950 (Bailey and 
Wu 1990, Table 30), Spent Nuclear FueI Discharges from US. Reactors 1994 (DOE 
1996, Table 5), and "Meeting the Challenge of Managing Nuclear Fuel in a Competitive 
Environment" (Yang 1997, Table l), supplemented by data for the period 1980 to 2002 
from The Technical Basis for the Classz~cation of Failed FueI in the Back-End of the 
FueI Cycle (EPRI 1997, p. 4-1). The total number of SFAs expected is approximately 
220,000 (BSC 2003c, Table 2). Using a repository operation period of 25 years 
(BSC 2004n, Table l), 350 SFAsIyear are expected to contain failed or damaged rods. 
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The majority of the failures are of a pinhole or hairline crack variety, while a range of 
10 to 20 percent are failures of cladding larger than a pinhole or hairline crack, as 
indicated in EPRI (1997, p. 4-1). Therefore, about 70 SFAsIyear are expected to be 
vulnerable to oxidation and cladding unzipping during fuel handling operations. 
On average, there are 2.2 rods per failed assembly (EPRI 1997, p. 4-I), which gives 154 
failed fuel rods per year (= 70 SFAs x 2.2 rods/assembly). 
This assumption is used in Section 6.2.1.3. 
4.24 An ARF of 1.2 x is assumed for fuel rods expected to be vulnerable to oxidation and 
cladding unzipping. 
Basis: At temperatures of 500C to 700C, bare fuel pellets are fully oxidized to U3O8 in 
1 hr according to Accident Analysis for Continued Storage (Davis et al. 1998, p. 9). 
The particle size distribution of the resultant powder was determined and the fraction of 
particles potentially small enough to become airborne was found to be 12 percent. 
Of this amount, approximately 1 percent is expected to be respirable. Therefore, the total 
ARF is 1.2 x This is consistent with the ARF of 1.0 x determined by DOE 
(1994, p. 4-3) for complete oxidation of uranium metal in a flowing air atmosphere. 
Data from Oxidation of Spent Fuel Between 250 and 360C (EPRI 1986) indicates that 
even after crack propagation (unzipping) begins, the rate of growth of the crack is slow. 
For example, the rate of growth at 360C was determined to be 2.3 x cmlmin for a 
defect size of 760 pm in diameter (EPRI 1986, Table 3-3). At this rate, after 100 hr a 
14 cm crack in the fuel rod would appear. This is approximately 4 percent of the total 
fuel rod length. Therefore, after 100 hr only 4 percent of the total fuel in the rod would 
be oxidized. Conservatively rounding up to 10 percent and using the ARF of 1.2 x 
for bare fuel pellets, an ARF of 1.2 x is assumed for fuel rods expected to be 
vulnerable to oxidation and cladding unzipping. 
This assumption is used in Section 6.2.1.3. 
4.25 Little or no oxidation is expected to occur for intact fuel or fuel with pinhole or hairline 
cracks during fuel handling operations in the repository. The fuel fine release fraction 
of 3 x for a burst rupture is conservatively used for fuel with pinhole leaks or 
hairline cracks where little or no fuel oxidation occurs. 
Basis: EPRI (1986, p. iii) determined that both the size and shape of the defect appear to 
influence the time to cladding splitting. For temperatures above 283"C, the time to 
cladding splitting was longer for the sharp small defect than for the large circular defect. 
This effect diminished as the temperature decreased. The breaches were induced by 
pressurizing the sample rods at elevated temperatures. Breach sizes ranged from 8 to 
52 pm. 
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MACCS2 (ORNL 1998) calculates doses resulting from accidental radionuclide releases from 
nuclear facilities to onsite and offsite members of the public. The code considers: atmospheric 
transport; short- and long-term mitigation actions; and potential exposure pathways, such as 
inhalation, resuspension inhalation, ingestion, air submersion, and groundshine. 
MACCS2 (ORNL 1998) simulates the impact of accidental atmospheric releases of radiological 
materials on the surrounding environment. The principal phenomena considered in MACCS2 
(ORNL 1998) are atmospheric transport, mitigative actions based on dose projection, dose 
accumulation by a number of pathways including food and water ingestion, and acute and 
chronic exposures. MACCS2 (ORNL 1998) contains simple models with analytical solutions. 
A MACCS2 (ORNL 1998) calculation consists of three phases: input processing and validation, 
phenomenological modeling, and output processing. The phenomenological models are based 
on empirical data and the solutions are analytical in nature and computationally straightforward. 
The modeling phase is divided into three modules: ATMOS, EARLY, and CHRONC. 
The ATMOS module treats atmospheric transport and dispersion of material and its deposition 
fkom the air using a Gaussian plume model with Pasquill-Gifford dispersion parameters. 
The EARLY module models consequences of the accident to the surrounding area during an 
emergency action period. The CHRONC module considers the long-term impact in the period 
subsequent to the emergency action period. 
The applicability of MACCS2 (ORNL 1998) to normal operations is discussed in Section 3.1.2. 
The MACCS2 (ORNL 1998) calculated x/Q values (Attachment C) are more conservative than 
the xIQ values calculated based on Regulatory Guide 1.111 for normal operations 
(Section 3.1.2). 
5.2 EXEMPT SOFTWARE 
The Microsofi Excel 97 spreadsheet program is used to perform simple calculations as 
documented in Section 6. User-defined formulas, input, and results are documented in sufficient 
detail in Section 6 to allow for independent duplication of various computations without recourse 
to the originator. This software is considered exempt from the requirements of LP-SI.l lQ, 
Software Management. 
6. ANALYSIS 
6.1 RADIONUCLIDE RELEASES FROM NORMAL OPERATIONS AND EVENT 
SEQUENCES 
This section discusses radionuclide releases from normal operations and Category 1 and 
Category 2 event sequences, as well as event sequence frequency calculation. Waste forms 
involved in normal operations and in Category 1 and Category 2 event sequences include PWR 
or BWR SFAs, HLW, and naval SNF (Section 4, Assumption 4.1). 
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6.1.1 Repository Operations 
NUREG-1804 (NRC 2003b, p. 2.1-29) requires a discussion of modes of geologic repository 
operations area operation. This section discusses facility operations in the repository operations 
area, including the CHF, DTF 1, DTF 2, the FHF, the TCRRF, the subsurface facility, and the 
SNF aging system. Operations relevant to consequence analyses may include waste handling, 
maintenance, waste retrieval, and SNF aging. 
Normal operations associated with waste handling, waste retrieval, and SNF aging are discussed 
in Sections 6.1.2.1 through 6.1.2.3. Doses to facility workers from general facility maintenance 
operations are discussed in Section 6.4.1.2. 
6.1.2 Radionuclide Releases from Normal Operations 
This section discusses normal operations in surface and subsurface facilities and the potential 
release of radionuclides from these facilities. 
6.1.2.1 Potential Releases from Surface Facilities 
The repository surface facilities are located near the North Portal entrance to the repository. 
Waste handling operations are camed out in the geologic repository operationi area. 
The facilities necessary to receive, package, age, and emplace waste in the repository are located 
in this area. 
Waste transfer operations are performed in several waste handling facilities, including the CHF, 
DTF 1, DTF 2, and the FHF, as described in Internal Hazards Analysis for License Application 
(BSC 2004q, Section 6.4). 
The SNF and vitrified HLW are transported to the repository in NRC-certified transportation 
casks in accordance with applicable federal regulations. When an SNF or HLW transportation 
cask is received, personnel verify the shipping manifests and inspect and survey the cask and its 
trailer or railcar. The cask is then moved to a restricted area for staging (BSC 2004q, 
Section 6.4). 
From the staging area, casks are delivered to TCRRF for transfer to a site rail transfer cart and 
the transportation cask buffer area, as applicable. Casks delivered to the FHF are delivered 
directly to the facility on public conveyance (truck trailer or railcar) from the staging area; there 
is no FHF interface with a site rail transfer cart or transportation cask buffer area (BSC 2004q, 
Section 6.4). 
When an empty waste package is scheduled for processing, it is moved into the appropriate 
waste handling facility. The waste handling facility operators prepare and configure the empty 
waste package for SNFMLW transfer. The steps for waste package preparation include 
positioning the waste package for transfer, inserting the inner shell lid and transfemng the waste 
package to the appropriate SNFMLW transfer area (BSC 2004q, Section 6.4). 
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After waste package loading is completed, the inner lid is placed on the waste package. Then the 
waste package is transferred to the waste package closure cell where the inner lid is welded first 
and then the middle and outer lids are placed on the waste package and welded, the welds are 
inspected, and the weld stresses are mitigated. Waste packages may be staged (DTF 1 and 
DTF 2 only) or transferred to the waste package loadout area where the waste package is 
surveyed for contamination, down-ended onto the emplacement pallet, positioned for lifting 
collar removal, and transferred into the waste package transporter. The waste package 
transporter is moved into the subsurface facilities to an emplacement drift, where the waste 
package is remotely transferred to an emplacement gantry for final positioning in the 
emplacement drift (BSC 2004q, Section 6.4). 
It is assumed that 3,000 MTHWyear (Section 4, Assumption 4.2) of commercial SNF pass 
through DTF 1, DTF 2, and the FHF. 
It is further assumed that 1 percent (Section 4, Assumption 4.3) of the fuel rods in the nominal 
annual throughput have cladding breaches and that fission product gases, volatile species, and 
fuel fines are released to the environment. Therefore, releases from 1 percent of fuel rods in the 
6,316 PWR SFAs are used as the source term for the calculation of normal operations doses. 
It is assumed that during normal operations of the surface facilities, there are releases from the 
exhaust stacks of DTF 1, DTF 2, and the FHF (Section 4, Assumption 4.17). There are no 
releases expected from the CHF during normal operations, because the CHF handles only sealed 
canisters and a sealed canister does not leak. 
Failed fuel is defined in 10 CFR Part 961, p. 607, as the fuel that has failed cladding, namely, 
hairline cracks and pinhole leaks or larger. Failed fuel has already released its radionuclide 
inventory in the gap region between the cladding and the fuel pellets, while at a nuclear 
powerplant. Historical fuel discharge data show that the fuel rod failure rate is about 
0.01 percent, versus a much smaller range of 0.02 to 0.07 percent in the first 20 years of 
commercial nuclear power (BSC 2003c, p. G-6) (Section 4, Assumption 4.3). 
Failed commercial SNF will be shipped to the repository in screen-end or closed-end canisters in 
a transportation cask. Failed commercial SNF canisters will be transferred from the cask to a 
waste package inside a waste transfer cell in a DTF or the FHF, during normal operations. 
Exposing the failed fuel in air while in a DTF or the FHF, could cause oxidation of the fuel 
(UOz) to higher oxides, such as U308. When U308 starts to form, the fuel pellet volume expands 
and eventually cladding could unzip. 
It'is assumed that for normal operations, radioactive materials are released in a duration of 24 hr 
(Section 4, Assumption 4.6) via the building exhaust stack (Section 4, Assumption 4.7) as a 
radioactive plume that is dispersed en route to the site boundary. The release is assumed to result 
in an acute individual exposure during plume passage and a chronic individual exposure to 
ground contamination and contaminated food after plume passage (Section 4, Assumption 4.6). 
The period of long-term exposure to ground contamination and intake of contaminated food is 
assumed to be 1 year (Section 4, Assumption 4.6). 
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Table 8. Annual Releases from Subsurface Facility 
Normal Operation (Continued) 
Normal Operations Release (Cilyear) 
Waste Package Surface Contamination (Continued) 
Source: BSC 20049, Table 14 
Annual releases (Table 8) are based on the postulated activation of air and silica dust in the 
subsurface facilities during normal operations as summarized in BSC (2004g, Table 14). 
Subsurface releases include waste package surface contamination and radionuclides generated by 
the activation of air and '6N) and dust (I%, 24~a, 42~
"~1, 31~i, and 55~e). Nitrogen-16 
(I%) is not considered in the dose assessment because of its short half-life and the associated 
long travel time fiom the release point to the site boundary. 
The releases from the subsurface exhaust shafts during normal operations are assumed to be at 
ground level (Section 4, Assumption 4.17). 
6.1.2.3 Potential Releases from Retrieval Operations 
The geologic repository operations area must be designed to preserve the option of waste 
retrieval throughout the period during which wastes are emplaced and thereafter, until NRC 
review of the information obtained from a performance confirmation program is completed 
(1 0 CFR 63.1 1 1 (e)(l)). A permanent retrieval could potentially require the entire underground 
repository to be emptied. For permanent retrieval, a separate license would be required to 
buildnew facilities. No airborne releases of radionuclides are expected for normal 
retrieval operations. 
6.1.3 Event Sequence Frequency Calculation and Categorization 
Categorization of event sequences based on frequency calculations and design bases are 
presented in BSC (2004k, Section 7). Design bases are a set of requirements to be implemented 
by design to prevent releases or to mitigate radiological colisequences. Bounding Category 1 
and Category 2 event sequences are identified in BSC (2004k, Section 7). 
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It is assumed that for this event sequence, radioactive materials are released in a duration of 1 hr 
(Section 4, Assumption 4.5) as a ground-level plume (Section 4, Assumptions 4.5 and 4.17) that 
is dispersed en route to the site boundary, resulting in an acute individual exposure during plume 
passage and a chronic individual exposure to ground contamination and contaminated food after 
plume passage (Section 4, Assumption 4.5). The period of long-term exposure to ground 
contamination and intake of contaminated food is 1 year (Section 4, Assumption 4.5). 
It is assumed that for worker dose calculations, the HVAC system is operating and no airborne 
material released from Category 1 event sequences leaks into space occupied by workers in 
rooms adjacent to the waste transfer cell of a DTF or the FHF (Section 4, Assumption 4.13), 
because the transfer cell confinement and the HVAC system are designed to prevent any leakage 
to rooms adjacent to the transfer cell in the event of a Category 1 event sequence. 
Credit is taken for HEPA filters to remove radionuclides (Section 4, Assumption 4.4). 
6.1.4.2 Event Sequence GET-03B (Table 9) 
During the transfer of individual commercial SFAs from a transportation cask to a waste package 
in a transfer cell, an SFA could collide with heavy equipment or structures. It is assumed that the 
cladding of 100 percent of the fuel rods is damaged and that radionuclides are released to the 
environment (Section 4, Assumption 4.10). 
The release of radionuclides is divided into two phases. The first phase involves a release 
of fission product gases, volatile species, fuel fines, and crud immediately following a 
cladding breach. The second phase involves oxidation of U02 to U3O8and cladding unzipping. 
During the fuel oxidation period, only fission product gases and volatile species are released. 
The release fractions for the first phase are given in Table 5. The release fractions of fission 
product gases and volatile species from fuel oxidation are included in the release fractions given 
in Table 5. 
It is assumed that for this event sequence, radioactive materials are released in a duration of 1 hr 
(Section 4, Assumption 4.5) as a ground-level plume (Section 4, Assumptions 4.5 and 4.17) that 
is dispersed en route to the site boundary, resulting in an acute individual exposure during plume 
passage and a chronic individual exposure to ground contamination and contaminated food after 
plume passage (Section 4, Assumption 4.5). The period of long-term exposure to ground 
contamination and intake of contaminated food is 1 year (Section 4, Assumption 4.5). 
It is assumed that for worker dose calculations, the HVAC system is operating and no airborne 
material released from Category 1 event sequences leaks into space occupied by workers in 
rooms adjacent to the waste transfer cell in a DTF or the FHF (Section 4, Assumption 4.13), 
because the transfer cell confinement and the HVAC system are designed to prevent any leakage 
to rooms adjacent to the transfer cell in the event of a Category 1 event sequence. 
Credit is taken for HEPA filters to remove radionuclides (Section 4, Assumption 4.4). 
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6.1.5 Radionuclide Releases from Category 2 Event Sequences 
Drops or collisions of SFAs, HLW, and naval SNF in canisters, transportation casks, or waste 
packages are determined to be Category 2 event sequences (BSC 2004k, Section 7). It is 
assumed that the cladding of spent fuel rods is damaged and that radionuclides are released in a 
1-hr duration as a ground-level release (Section 4, Assumption 4.5). Two Category 2 event 
sequences, that bound other: Category 2 event sequences (BSC 2004k, Section 7) with respect to 
frequency and MAR, are identified in Table 10. 
Table 10. Bounding Category 2 Event Sequences 
Event Expected Number of 
Sequence Occurrences Over the Bounding Material 
Identifier Description Life of the Repository at Risk 
GET-OIA Drop of a transportation cask, without 5.8E-01 74 BWR or 36 PWR 
impact limiters, in the CHF, DTF, FHF, SFAs, 5 DOE HLW 
or TCRRF canisters, or 1 naval SNF 
canister 
Drop of the inner lid of a transportation 5.8E-01 74 BWR or 36 PWR 
cask, site specific cask, or waste SFAs, 5 DOE HLW 
package into a transportation cask, site canisters or 1 naval SNF 
specific cask, or waste package in the canister 
CHF, DTF, or FHF 
BWR = boiling water reactor; CHF = Canister Handling Facility; DOE = U.S. Department of Energy; DTF = Dry 
Transfer Facility (DTF 1 or DTF 2); FHF = Fuel Handling Facility; HLW = high-level radioactive waste; PWR = 
pressurized water reactor; SNF = spent nuclear fuel; TCRRF = Transportation Cask Receipt and Return Facility. 
Source: BSC 2004k, Table 59 
6.1.5.1 Bounding Category 2 Event Sequence Involving Boiling Water Reactor and 
Pressurized Water Reactor Spent Nuclear Fuels 
Events GET-O1A and GET-02B (Table 10) bound other Category 2 event sequences involving 
PWR and BWR SNFs. Event GET-O1A involves a drop of a transportation cask and breach of 
PWR or BWR SFAs. It is assumed that the event sequence results in breaches of 36 PWR SFAs 
or 74 BWR SFAs and releases of radionuclides to the environment. 
It is assumed that for this event sequence, radioactive materials are released in a duration of 1 hr 
as a ground-level radioactive plume that is dispersed en route to the site boundary, resulting in an 
acute individual exposure during plume passage and a chronic individual exposure to ground 
contamination and contaminated food after plume passage (Section 4, Assumption 4.5). 
The period of long-term exposure to ground contamination and intake of contaminated food 
is 1 year (Section 4, Assumption 4.5). 
No credit is taken for HEPA filters to remove radionuclides (Section 4, Assumption 4.4). 
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6.1.5.2 Bounding Category 2 Event Sequence Involving Naval Spent Nuclear Fuel 
Events GET-O1A and GET-02B (Table 10) bound other event sequences involving naval SNF. 
Event GET-O1A (Table 10) involves a drop of a transportation cask and breach of a naval 
SNF canister. It is assumed that the event sequence results in damage to naval SNF assemblies 
and releases of radionuclides to the environment. 
It is assumed that for this event sequence, radioactive materials are released in a duration of 1 hr 
as a ground-level radioactive plume that is dispersed en route to the site boundary, resulting in an 
acute individual exposure during plume passage and a chronic individual exposure to ground 
contamination and contaminated food after plume passage (Section 4, Assumption 4.5). 
The period of long-term exposure to ground contamination and intake of contaminated food 
is 1 year (Section 4, Assumption 4.5). 
No credit is taken for HEPA filters to remove radionuclides (Section 4, Assumption 4.4). 
6.1.5.3 Bounding Category 2 Event Sequence Involving Vitrified HLW 
Events GET-O1A and GET-02B (Table 10) bound other event sequences involving 
vitrified HLW. Event GET-O1A (Table 10) involves a drop of a transportation cask and breach 
of five vitrified HLW canisters. It is assumed that the event sequence results in breaches of five 
HLW canisters and releases of radionuclides to the environment. 
It is assumed that for this event sequence, radioactive materials are released in a duration of 1 hr 
as a ground-level radioactive plume that is dispersed en route to the site boundary, resulting in an 
acute individual exposure during plume passage and a chronic individual exposure to ground 
contamination and contaminated food after plume passage (Section 4, Assumption 4.5). 
The period of long-term exposure to ground contamination 'and intake of contaminated food 
is 1 year (Section 4, Assumption 4.5). 
No credit is taken for HEPA filters to remove radionuclides (Section 4, Assumption 4.4). 
6.1.6 Interactions of Hazards and Controls 
Much of the equipment used for handling and moving SNF and HLW, such as cranes, gantries, 
and transporters, is remotely operated. To prevent the initiation of a Category 1 or Category 2 
event sequence upon a loss of power, the instrumentation and control systems for remotely 
operated equipment are designed to be failed safe so that a crane or gantry does not drop loads 
(BSC 2004k, Section 5.1.2.1). 
There are no procedural safety controls, such as operator actions, credited to prevent a 
Category 1 or Category 2 event sequence (BSC 2004k, Section 5.1.2.1). 
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The HVAC system is required to mitigate potential releases of radionuclides from surface 
facilities during normal operations and Category 1 event sequences. This system is required 
to be operable when waste handling operations are conducted in a surface waste 
handling facility. The fan and HEPA filters are not required to be operable following a 
Category 2 event sequence. 
Facility design features, such as moderator control in the waste transfer cell, are required to 
prevent criticality. 
6.2 INPUTS 
This section discusses the input parameters used in public and worker dose calculations. 
6.2.1 Public Dose Calculation Design Inputs 
This section discusses input files, output files, and input parameters used in MACCS2 
(ORNL 1998) analyses of radionuclide releases from normal operations and Category 1 and 
Category 2 event sequences. 
6.2.1.1 MACCS2 Input Files 
Seven input files are necessary to run MACCS2 (ORNL 1998) for event sequences. The first 
file, ATMOS, provides radionuclide inventory and release information. The second file, 
EARLY, provides information related to the early phase of the event sequence. The third file, 
CHRONC, provides information on the long-term chronic exposure phase. A fourth file 
provides hourly meteorological data, and a fifth file provides site-specific data including 
population data, land use data, and grid spacing information. The remaining two input files, 
those being the dosimetry file and the food pathway file, are identified in the first five input files. 
The dosimetry file, current applications use YMPD825.W, is defined in the EARLY input file 
YMPEA1.D.JP. The food pathway file, current applications use YUCCALA.BIN, is defined in 
the CHRONC input file YMPCHl .INP. These file names are not changed; if the dosimetry file 
is changed, then it is necessary to rerun the COMIDA2 module with the same dosimetry file 
because MACCS2 checks the two files to ensure that they are based on the same dosimetry file 
used in the current MACCS2 run. 
MACCS2 (ORNL 1998) is run using a batch command file called RUNMAX2. Use of this file 
requires specification of additional file names that identify the files' source of input information 
required by MACCS2. 
The command line for running the PWR base case is: 
RUNMAX2 LAPWROl YMPEAl YMPCHl YMP1999 YMPSIT LAPWROl 
This command line directs MACCS2 (ORNL 1998) to read the ATMOS data set from file 
LAPWROl.INP, the EARLY data set from YMPEA1, the CHRONC data set from YMPCH1, 
the meteorology data set from YMP1999, and the site data set from file YMPSIT. The output 
results are written to file LAPWRO1.OUT. The dosimetry file YMPD825.W and the food 
pathway file YUCCALA.BIN are used as previously described. 
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The command line for running the HLW base case is: 
RUNMAX2 LAHLWOl YMPEA YMPCHl YMP1999 YMPSIT LAHLWOl 
This command line directs MACCS2 (ORNL 1998) to read the ATMOS data set from file 
YMHLWO 1 .INP, the EARLY data set from YMPEA, the CHRONC data set from YMPCH1, the 
meteorology data set from YMP1999, and the site data set fiom file YMPSIT. The output results 
are written to file LAHLWO 1 .OUT. 
The command line for running the naval SNF base case is: 
RUNMAX2 LANAVOl YMPEA YMPCHl YMP1999 YMPSIT LANAVOl 
This command line directs MACCS2 (ORNL 1998) to read the ATMOS data set from file 
LANAVOl.INP, the EARLY data set from YMPEA, the CHRONC data set from YMPCHl, the 
meteorology data set from YMP1999, and the site data set fiom file YMPSIT. The output results 
are written to file LANAVO1 .OUT. 
To modify one of the base cases for a different event sequence, the following parameters are 
defined and entered into the ATMOS file. The new file is prepared by first copying the 
appropriate existing file, for example LAPWROl.INP, to a file with a representative name, such 
as changing the "01" to a "02" to represent a new event sequence and then editing the file to 
represent the new event sequence. 
Six sets of ATMOS input files are prepared based on different fuel types: a Maximum PWR SFA 
of 5 years decay, an Average PWR SFA of 25 years decay, a Maximum BWR SFA of 5 years 
decay, an Average BWR SFA of 25 years decay, a HLW canister, and a naval fuel canister. 
These files are used to describe additional event sequences by modifying specific parameters in 
the input files. 
The parameters likely to need modification for a new event sequence are listed in Table 1 1. 
Each parameter listed in Table 11 is entered in the ATMOS type file using a keyword. Table 11 
also provides the keyword and section in the MACCS2 user manual (ORNL 1998) that describe 
the parameter. Base case files are set up to release the activity over a 1-hr period. If a different 
release period is desired, then that value is entered. If a different value is entered, then the value 
for parameter PMTIMBAS is also changed to the same value. This inhibits special plume 
meander corrections. The maximum value allowed for the release duration is 24 hr. 
The values for the initial sigma-y and sigma-z values should not be changed unless the building 
height or width (not an input) is changed. An initial sigma-y of 10.3 m and an initial sigma-z 
of 12 m (Table 11) are for a 25.9-m high building that is 44.5 m wide (i.e., FHF building 
dimensions) (BSC 2004f, Table 2) and a release height of 0 m (ground level) (Section 4, 
Assumption 4.17) or 30 m (Section 4, Assumption 4.21). These values are used for each 
MACCS2 run except for Run 3 (Section 6.3). An initial sigma-y of 0.1 m and an initial sigma-z 
of 0.1 m (Table 11) are used for MACCS2 Run 3 (Section 6.3) in which no building structures 
are present at the subsurface exhaust shafts. 
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6.2.1.1.3 MACCS2 Output File 
MACCS2 (ORNL 1998) output results are printed to the last named file on the command line; 
extension OUT. The first part of the file is a record of the parameter values used in the analysis 
and provides a record of the analysis. Results are provided as mean values and for 5oth, goth, 
95th, 9gth, and 99.5th percentiles. Three sets of results are provided: x/Q values, whole body 
effective dose, and organ doses. Information on dose contributions fiom specific pathways is 
also contained in the output file. 
A descriptive title is entered for the event sequence in the ATMOS file, near the start; keyword 
name RIATNAMl. The title is enclosed in single quote marks. The title appears on the output 
file as a check that the correct file was used in the analysis. A convention in the input files is that 
lines starting with an asterisk (*) are comment lines and are not read by the program. Only those 
lines recognized as containing data are given a number in the left column. 
In the MACCS2 output file, the values of the parameter "Ground-Level Dilution, x/Q (s/m3)" at 
100 m, 8 km, or 11 km for mean values and for 5oth, goth, 95th, 9gth, and 99.5th percentiles are 
provided. Immediately following the x/Q value outputs, the centerline doses at various distances 
away from the release point are provided for mean values and for 5oth, goth, 95th, 9gth, and 99.5th 
percentiles. The centerline doses include the parameter "L-EFFECTIVE TOT LIF," for whole 
body dose excluding ingestion dose, and the parameter "L-BONE SUR TOT LIF," which is the 
highest TODE excluding ingestion dose. Among the internal organs, bone surface is the organ 
that receives the highest radiation dose. The ingestion dose is printed on the last page of the 
output file under the heading "MAXIMUM ANNUAL FOOD DOSE (EFFECTIVE)." 
The offsite public TEDE, at 11 km or 8 km away from the release point for normal operations 
and Category 1 event sequences, is determined by summing up the mean value of the parameter 
"L-EFFECTIVE TOT LIF" and the ingestion dose. The TEDE at 11 km or 8 km from the 
release point for Category 2 event sequences is determined by summing up the 95th percentile 
value of the parameter "L-EFFECTIVE TOT LIF" and the ingestion dose. The TODE at 11 km 
or 8 km from the release point for normal operations and Category 1 event sequences is 
determined by summing up the mean value of the parameter "L-BONE SUR TOT LIF" and the 
ingestion dose. The TODE at 11 km or 8 km from the release point for Category 2 event 
sequences is determined by summing up the 95th percentile value of the parameter "L-BONE 
SUR TOT LIF" and the ingestion dose. 
The onsite public TEDE at 100 m from the release point for normal operations and Category 1 
event sequences do not include ingestion dose and are calculated based on an exposure period 
of 2,000 hr (Section 4, Assumption 4.20). Therefore, the annual TEDE needs to be adjusted to 
reflect an exposure period of 2,000 hr instead of 8,760 hr. 
The onsite public TEDE, at 100 m from the release point for normal operations, is the mean 
value of the parameter "L-EFFECTIVE TOT LIF" in cohorts 1 and 2 (i.e., combined EARLY 
and CHRONC outputs) multiplied by a factor of 0.228, which is equal to 2,000 hr divided 
by 8,760 hr. 
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The onsite public TEDE, at 100 m from the release point for Category 1 event sequences, is 
determined by summing up the mean value ofthe parameter "L-EFFECTIVE TOT LIF" in 
cohort 1 (the EARLY phase) and the parameter "L-EFFECTIVE TOT LIF" in cohort 2 
(the CHRONC phase) multiplied by a factor of 0.228. 
6.2.1.2 Source Terms 
Source term input parameters are needed in the ATMOS file. Source term is defined as 
concentrations or inventories of radionuclides in SNFs or waste forms to be received and handled 
at the repository. The concentration or inventory of each radionuclide in a waste form is 
expressed as curies per SFA, curies per unit weight of waste form, or curies per canister. A curie 
is a unit quantity of any radioactive nuclide in which 3.7 x 10" disintegrations occur per second. 
Source terms for waste forms to' be received and handled at the repository are needed for 
consequence analyses. Source terms are a function of the initial fuel enrichment, fuel compound, 
cladding type, moderator type, and reactor operating history. Source terms are calculated using a 
computer program that performs a point depletion and decay calculation. 
Commercial Spent Nuclear Fuel-Average and maximum source terms for PWR and BWR 
commercial SNF are from BSC (20041) and BSC (2003a). 
The SAS2H sequence in SCALE V4.3, from NUREGICR-0200 (ORNL 1997), is used to 
calculate the PWR and BWR source terms for selected SFAs as a function of assembly average 
burnup and cooling time. The prime functional module of the SAS2H code sequence used is the 
ORIGEN-S module. This module performs a point depletion and decay calculation of a selected 
fuel type with user-specified irradiation conditions and decay times. The resulting source terms 
are then extracted from the SAS2H output and are used as input to consequence analyses. 
Source terms for PWR and BWR SFAs with four different combinations of initial enrichment, 
bumup, and decay time are considered in this consequence analysis and are presented in Table 3. 
For Category 1 event sequences, both Average PWR and Average BWR SFAs are used to 
calculate mean doses and 5oth percentile doses. The calculated mean dose is used because it is 
higher than the calculated 5oth percentile dose. For Category 2 event sequences, both Maximum 
PWR SFAs and Maximum BWR SFAs are used to calculate doses under 95th percentile weather 
conditions in accordance with Regulatory Guide 1.145, p. 3. Radionuclide inventories in CiEA 
are presented in Table 4 for each nuclide and fuel type evaluated. 
Crud (Commercial SNF')-Crud is activated corrosion products found on the exterior surface of 
SFAs, primarily caused by the irradiation of reactor intemals and imperfect water chemistry 
control in a reactor coolant system. Crud can be released to the environment during an accident 
involving commercial SNF. After decaying for 5 years, the nuclide species that have significant 
and 60~o. The commercial SNF assemblies have initial crud 
activity in the crud are 55~e 
activities at the time of discharge from the reactor as shown in Table 12. 
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These surface areas are bounding estimates based on SFAs with the highest known surface areas, 
which are a South Texas PWR assembly (BSC 20041, p. 27) and an ANF 9 x 9 JP-4 BWR 
assembly (BSC 2003a, Table 45). The crud source term for Category 2 event sequences is based 
on Maximum PWR or Maximum BWR SNF with a 5-year decay time. Using Equations 14 
and 15, the 5-year crud source term is given in Table 13. The crud source term for Category 1 
event sequences is based on Average PWR or Average BWR SNF with a 25-year decay time. 
Using Equations 14 and 15, the 25-year crud source term is given in Table 13. 
Table 13. Five-year and 25-year Crud Source Terms 
Radionuclide 5-year Crud 25-year Crud 
In Crud Source (CilFA) Source (CiIFA) 
Fe-55 PW R 7.45E+02 4.64 
Fe-55 BWR 3.50E+02 2.18 
CO-60 PW R 3.26E+01 2.35 
CO-60 BWR 1.09E+02 7.87 
BWR = boiling water reactor; CVFA = curies per fuel assembly; 
PWR = pressurized water reactor. 
DOE Spent Nuclear Fuel-DOE SNF is shipped in two types of disposable canisters, namely, 
the DOE standardized canister and the multicanister overpack. DOE SNF is received at the 
repository in transportation casks containing the sealed disposable canisters. ISG-5 (NRC 2003a, 
Attachment, p. 3) states that for casks having closure lids that are designed and tested to be leak 
tight as defined in ANSI N14.5-97, detailed consequence analyses are not necessary. 
Because the DOE standardized canister and multicanister overpack are designed and tested to the 
ANSI N14.5-97 leak-tight standard after dropping from a design basis drop height, a drop and 
breach of DOE canisters or multicanister overpacks is a beyond Category 2 event (BSC 2004k, 
Table 16) and detailed consequence analyses are not necessary as stated in ISG-5 (NRC 2003a, 
Attachment, p. 3). 
Naval Spent Nuclear Fuel-Some drops or collisions involving a naval canister may result in a 
breach of the canister but no fuel damage. These events do not correspond to any bounding 
Category 2 event sequences listed in Table 10, because they do not cause fuel damage. For this 
category of events, a release source term based on radionuclides involving only crud was 
developed (Gisch 2004, p. 1). A table of radionuclides, for crud and total activity, for a 
representative naval SNF canister 5 years after shutdown is provided in Gisch (2004, Table 1). 
The crud inventory for naval SNF is calculated using the standard naval program shielding 
procedure. This value was increased by a factor of 2.5 to provide a conservative crud 
concentration for use in developing the source term. Table 14 identifies a radionuclide release 
source term for Category 2 event sequences in which a naval SNF canister is breached and fuel is 
not damaged. 
A conservative damage ratio, along with the crud ARF and RF (Table 5) were applied to the 
naval crud inventory, using a conservative leak path factor of 1.0 (Section 4, Assumption 4.1 1). 
The release source term in Table 14 takes no credit for HEPA filtration, plateout, or deposition 
of radionuclides. 
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Table 14. Crud Release From a Naval Spent Nuclear Fuel Canister 
Radionuclide Activity (Ci) 
Am-241 2.7E-06 
Am-242m 1.5E-08 
Am-243 2.3E-08 
C-14 7.8E-03 
Cm-242 2.1 E-08 
Cm-243 1.7E-08 
Cm-244 2.3E-06 
Cm-245 2.OE-10 
Cm-246 7.8E-11 
Cm-247 2.3E-16 
Cm-248 7.4E-16 
Co-60 4.1E-01 
CS-137 2.8E-04 
Fe-55 4.4E-01 
1-1 29 3.1 E-08 
N b-93m 9.5E-03 
N b-94 1.6E-04 
Ni-59 2.3E-03 
Ni-63 2.3E-01 
Np-237 2.3E-11 
Pu-238 1.9E-06 
Pu-239 3.1 E-07 
Pu-240 2.OE-07 
Pu-241 6.1 E-05 
Pu-242 2.3E-09 
Se-79 1.2E-09 
Sn-126 3.5E-09 
Sr-90 2.8E-04 
Tc-99 7.8E-06 
Th-232 7.4E-13 
U-232 1.IE-08 
Zr-93 1.6E-06 
Source: Gisch 2004. Table 1 
Table 15 identifies a source term, based on radionuclide inventories involving releases from crud 
and fuel-bearing regions for bounding Category 2 event sequences (Table 10) that could result in 
a canister breach with fuel damage. These Category 2 event sequences involve slapdowns onto 
unyielding surfaces resulting in the highest canister strain levels and the highest potential for fuel 
damage, because of contact between the SFAs and the canister basket support plates. 
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Table 15. Fuel and Crud Release From a Naval Spent Nuclear Fuel Canister 
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Table 15. Fuel and Crud Release From a Naval Spent Nuclear 
Fuel Canister (Continued) 
Activity -All Pathways Activity -Inhalation 
Nuclide Except Inhalation (Ci) Pathway Only (Ci) 
Sb-125 9.8E-04 4.9E-06 
Se-79 1 .OE-07 1.7E-09 
U-238 3.6E-10 1.8E-12 
Y-90 1.3E-01 6.7E-04 
Zr-93 3.8E-06 1.6E-06 
Source: Gisch 2004, Table 2 
The fuel and crud damage ratios, used to develop the source term, were conservatively selected 
to ensure that canister radionuclide inventories and internal hardware designs are bounded. 
ARFs and RFs for standard commercial SNF (Table 5) are used for Category 2 event sequences 
involving naval SNF. A conservative leak path factor of 1.0 was used (Section 4, 
Assumption 4.1 1). Separate activity columns are listed: one for dose calculations for all 
pathways' except inhalation, and one that applies the RFs in Table 5 for dose calculations in the 
inhalation pathway. The release source term in Table 15 takes no credit for HEPA filtration, 
plateout, or deposition of radionuclides. 
Vitrified HLW-Vitrified HLW forms from the Savannah River Site (SRS), Hanford Site, 
West Valley, and MEL are received at the repository in sealed canisters inside 
transportation casks. The per canister source terms for vitrified HLW from the four sites are 
provided in Table 16. 
The per canister source term for vitrified HLW shipped from SRS (Table 16) is from 
Fowler (2003, Table 2, column 3); this source term is the projected maximum radionuclide 
inventory per canister based on waste tank sludge batches collected as of March 17,2003. 
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Table 16. Radionuclide Inventory Per Vitrified High-level Radioactive Waste Canister 
Nuclide SRS(Gila Hanford(cilb West Valley (Ci)' INEEL (cild 
Am-24 1 2.28E+02 4.65E+02 5.07E+02 2.61E+00 
Am-2421-11 1.30E-01 0.00E+00 2.74E+00 0.00E+00 
Am-243 3.68E-01 9.99E-02 3.28E+00 0.00E+00 
Ba-I 37m 6.24E+04 6.62E+04e 3.00E+04= 1 .40E+04e 
C-14 0.00E+00 1.06E-07 1.30E+00 0.00E+00 
Cd-I 13m 0.00E+00 2.69E+01 0.00E+00 0.00E+00 
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Table 16. Radionuclide Inventory Per Vitrified High-level Radioactive Waste 
Canister (Continued) 
Nuclide SRS (Ci)= Hanford (cilb West Valley (Ci)' INEEL (cild 
U-238 5.1 7E-02 1 .O1 E-02 0.00E+00 0.00E+00 
Y-90 7.62E+04 7.38E+04 2.80E+04 1.52E+04 
Zr-93 1.88E-01 5.76E+00 2.58E+00 0.00E+00 
NOTES: a Fowler 2003, Table 2, column 3 
DOE 2004c, Table 5, column 4 
WVNS 2001, Table 5, column 4 (high estimate case for 1996) 
CRWMS M&O 1999b, p. 111-1 
Ba-137m = 0.946 x Cs-137 (Eckerman and Ryman 1993, p. 203). 
INEEL = Idaho National Engineering and Environmental Laboratory; SRS = Savan'nah River Site. 
The per canister source tenn for vitrified HLW shipped from the Hanford Site (Table 16) is fiom 
Waste Treatment and Immobilization Plant (WTP) High-Level Waste (HLW Canister 
Production Estimates to Support Analyses by the Yucca Mountain Project (DOE 2004c, Table 5, 
column 4); this source term is the estimated bounding radionuclide inventory per canister based 
on HLW canisters produced fiom the 241-AZ-101 waste tank decayed to January 1, 2010. 
Among the waste tanks in the Hanford tank farm, the 241-AZ-101 waste tank has been 
determined to be the waste tank that will contain the maximum concentration of radionuclides 
for the Hanford HLW canisters (DOE 2004c, p. 6). 
The per canister source tenn for vitrified HLW (Table 16), shipped fiom West Valley, is the 
estimated highest radionuclide inventory per canister, based on a 100 percent fill canister with 
radionuclides decayed to the year 1996, fiom West Valley Nuclear Services Company Waste 
Form Qualzfication Report (WVNS 2001, Table 5, column 4). 
The per canister source term for vitrified HLW shipped fiom INEEL (Table 16) is fiom 
DOE High-Level Vitrzfied Waste Dose Calculation (CRWMS M&O 199923, p. ItI-1). 
Dose calculations show that SRS HLW has the highest dose consequence of HLW from the 
four DOE sites, those being Hanford, INEEL, SRS, and West Valley (CRWMS M&O 199923, 
p. 111-1). 
6.2.1.3 Release Fractions 
Radionuclide release fractions are needed in the ATMOS file. The Category 1 and Category 2 
event sequence total release fiaction is defined as the fraction of total inventory of a given 
radionuclide that is released to the environment fiom a commercial SNF element following an 
event sequence; for example, drop of a fuel element. The release fraction for commercial SNF is 
primarily a measure of the inventory of fuel particulates, gases, and volatile species present in a 
breached fuel element. 
The source term released fiom Category 1 and Category 2 event sequences are a hnction of the 
MAR, DR, ARF, RF, (LPF)cask, (LPF)fac, and (LPF)HEPA: 
Release Source Term STj = MARj x DR x ARF x RF x (LPF),,sk x (LPF)facx (LPF)H~PA
(Eq. 16) 
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The ARFs and FWs for intact commercial SNF assemblies (Table 5) are used for normal 
operations and Category 1 event sequences. The ARFs and FWs for Type 3a (fuel rods with 
intact cladding) are used for Category 2 event sequences because the combined ARF x FW values 
for Type 3a are larger or equal to the values for intact commercial SNF assemblies. 
Because drops or collisions of Type 3a in transportation casks and waste packages are assumed 
to be Category 2 event sequences (Section 4, Assumption 4.9), the ARFs and FWs for Type 3a 
are not applicable to normal operations and Category 1 event sequences. 
Damaged commercial SNF in damaged fuel cans will be shipped to the repository inside a 
transportation cask as stated in ISG-1 (NRC 2002). ISG-1 allows utilities to ship fuel with 
pinhole leaks or hairline cracks as intact fuel (NRC 2002). Intact fuel, damaged fuel in canisters, 
and fuel with pinhole leaks or hairline cracks will be transferred, from a transportation cask to a 
waste package, inside a waste transfer cell during normal operations in a DTF or the FHF. 
Exposing the failed fuel to air, while inside a DTF or the FHF, could cause oxidation of the fuel 
Wo2) to higher oxides, such as U3O8. When U3O8 starts to form, the fuel pellet volume expands 
and eventually cladding could unzip. 
The oxidation of SNF in air, as a two-step process of the form U02+U02.4+U308, is described 
in Hanson (1998, p. iii). The transition from U02+U02.4 does not result in appreciable 
density changes. The transition from U02.4+U308 results in a 20 percent less dense phase. 
The increase in volume as SNF oxidizes to U3O8 places stress on the cladding, which may split 
as a result. 
The oxidation process first progresses by the U02+U02.4 reaction. Once the SNF oxidized to 
UO2.4, a plateau is reached where the fuel resisted oxidation to higher oxides. Following this 
plateau oxidation resumes until the U3O8 phase is reached. Hanson (1998, p. iii) found that the 
U02.4+U308 reaction is strongly dependent on both temperature and bumup. SNF strongly 
resisted oxidation beyond UO2.4 at either low temperature or high bumup. 
Little or no oxidation is expected to occur for ifitact fuel or fuel with pinhole or hairline cracks 
during fuel handling operations in the repository (Section 4, Assumption 4.25). The fuel fine 
release fraction of 3 x for a burst rupture is conservatively used for fuel with pinhole leaks 
or hairline cracks where little or no fuel oxidation occurs (Section 4, Assumption 4.25). 
In addition to 1 percent of the fuel rods received at the repository having pinhole leaks or hairline 
cracks (Section 4, Assumption 4.3), it is assumed that 154 failed fuel rods (Section 4, 
Assumption 4.23) received at the repository per year are expected to be vulnerable to oxidation 
and cladding unzipping. An ARF of 1.2 x is assumed for fuel rods expected to be 
vulnerable to oxidation and cladding unzipping (Section 4, Assumption 4.24). 
In summary, for fuel rods with pinhole or hairline crack failures, a fuel fine release fraction 
of 3 x (Section 4, Assumption 4.25) (Table 5) is used to calculate the normal operation 
doses. For the purposes of calculating doses it is assumed that 1 percent of the total fuel rods 
received at the repository have pinhole or hairline crack failures. 
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In summary, for fuel rods with cladding failures greater than pinhole or hairline cracks, a fuel 
fine release fraction of 1.2 x 1o-~(Section 4, Assumption 4.24) (Table 5) is used to calculate the 
normal operation doses. This release fraction is applicable to 154 fuel rods per year (Section 4, 
Assumption 4.23) that have cladding failures greater than pinhole or hairline cracks. 
It is assumed that the fission product gas, volatile species, and crud release fractions for breaches 
of intact commercial SNF assemblies and rods in Table 5 are applicable to releases of fission 
product gases, volatile species, and crud during oxidation of damaged commercial SNF in air in 
Table 5 (Section 4, Assumption 4.16). 
HLW Release Fractions-The formation of particulates from an impact breach of a HLW 
canister is based on ANSUANS-5.10-1998, p. 15. The results are based on empirical 
measurements of impact tests on U02, ceramic, and glass-simulated waste forms. Small-scale 
laboratory tests establish each correlation for the percentage of respirable size release fractions 
created during impacts. 
This method uses the available results to develop a method to estimate the ARFs of HLW 
canister glass that could be released as airborne particulates. Based on these methods, the release 
fraction of respirable airborne particulates or the release fraction pulverized into respirable sizes 
(< 10 pm) from a drop event (PULF) formed following an impact is estimated as: 
PULF = 2E-4 cm3/joule * EN (Eq. 17) 
where, 
PULF = release fraction pulverized into respirable sizes (< 10 pm) from a drop 
event (units = dimensionless) 
-
EN -impact energy density in impacted HLW 
-1 .OE-07 jo~le-s~/~-cm~ 
-* p * g * h (CRWMS M&O 1999b, Attachment V) 
where, 
-
P -density of the HLW dropped 
-
-2.75 g/cm3 (CRWMS M&O 1999b, Section 5.2.5) 
-
-gravitational constant 
g 
-
-980.7 cds2 (CRWMS M&O 1999b; Attachment V) 
h --
drop height in cm. 
PULF for HLW Canister Drop 
From Equation 17: 
PULF = 2E-4 cm3/joule * 1 .OE-7 jo~le-s~/~-cm~ 
* 2.75 g/cm3 * 980.7 cds2 
* (drop height) cm 
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Calculated PULF for selected drop heights (CRWMS M&O 1999b; Attachment V) are: 
Drop Height PULF 
80 in. (203 cm) 
264 in. (671 cm) 
330 in. (838 cm) 
448 in. (1 138 cm) 
The drop height of 448 in. represents the worst case with a PULF of 6.14E-05, which is used in 
the public dose calculations for HLW canisters because it bounds the other values. It is assumed 
that the canister handling system is designed so that a drop of an HLW canister, inside a surface 
facility, will not exceed a drop height of 276 in. (23 ft) (Section 4, Assumption 4.19) and, 
therefore, the use of a 448-in. drop height is conservative. 
Naval Spent Nuclear Fuel Release Fractions-ARFs and RFs used for Category 2 event 
sequences involving naval SNF are presented in Section 6.2.1.2. 
HEPA Filter Leak Path Factor-The HEPA filter (LPF)HE~A 
refers to the removal of 
particulates provided by HEPA filters present in the surface. facility ventilation system. 
For normal operations and Category 1 event sequences, a two-stage HEPA filtration system with 
a particulate removal efficiency of 99 percent per stage (i.e., a HEPA LPF of O.Ol/stage) 
is assumed (Section 4, Assumption 4.8), which is consistent with NRC-recommended credit for 
accident dose evaluations in Regulatory Guide 1.140, p. 1.140-4, and Regulatory Guide 1.52, 
p. 1.52-5, for a combined efficiency of 99.99 percent for two stages; a HEPA LPF of 
It is further assumed that the HVAC system removes particulates and cesium in air through two 
stages of HEPA filters in series, which are protected by prefilters, sprinklers, and demisters 
(Section 4, Assumption 4.8). 
of is more conservative than the value of 2 x for two stages of HEPA 
An (LPF)HEPA 
filters, as discussed in NUREGICR-64 10 (SAIC 1998, Section F.2.1.3) (Section 4, 
Assumption 4.8). NUREGICR-0722 test data (Lorenz et al. 1980, Table 19) show that the two- 
stage HEPA filters capture almost 100 percent of incoming airborne cesium. 
No credit is taken for HEPA filters in mitigating radiological consequences from Category 2 
event sequences. Credit is taken for HEPA filters in mitigating radiological consequences from 
normal operations and Category 1 event sequences (Section 4, Assumption 4.4). 
For normal operations, and Category 1 and Category 2 event sequences, no credit is taken for 
charcoal adsorbers to remove radionuclides (Section 4, Assumption 4.1 8). 
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6.2.1.4 Dose Conversion Factors 
In MACCS2 (ORNL 1998), DCFs for inhalation, cloudshine, groundshine, and ingestion 
pathways are generated by the FGRDCF module. 
The output filename generated by the FGRDCF module is YMPD825.W. This file is called by 
the EARLY file and the CHRONC file during MACCS2 execution. The YMPD825.W file 
contains DCF data from Federal Guidance Reports Nos. 11 (Eckerman et al. 1988) and 12 
(Eckerman and Ryman 1993) for inhalation, cloudshine, groundshine, and ingestion pathways, in 
accordance with NUREG- 1804 (NRC 2003b, Section 2.1.1 S). 
DCFs for inhalation are dependent on the chemical form of the radionuclide, which is 
represented by the lung clearance class (D = daily, W = weekly, Y = yearly) and the fractional 
uptake from the small intestine to blood (fl). Some nuclides have only one lung clearance class, 
for example, 3~,
whereas others have multiple lung clearance classes, for example, 239~~. 
The inhalation DCFs used by YMPD825.INP are from Federal Guidance Report No. 11 
(Eckerman et al. 1988, Table 2.1). The air submersion DCFs for gonads, breast, lungs, red 
marrow, bone surface, thyroid, remainder, effective-whole body, and skin used by 
YMPD825.W are from Federal Guidance Report No. 12 (Eckerman and Ryman 1993, 
Table 111.1). The remainder is a weighted combination of five remaining organs or tissues 
receiving the highest doses, such as the liver, kidneys, spleen, brain, small intestine, upper large 
intestine, lower large intestine, or other organs, excluding the skin, lens of the eye, and 
extremities (Eckerman et al. 1988, p. 6). 
6.2.1.5 Location of Maximally Exposed Individual 
The maximally exposed individual is defined as an individual located at a distance that 
corresponds to the approximate distance between the surface facility or the subsurface repository 
and the nearest point of public access on the repository site boundary, which lies to the west 
(Section 4, Assumption 4.15). 
This individual is assumed to be present at the site boundary and exposed to ground 
contamination and intake of contaminated food for one year (Section 4, Assumptions 4.5 
and 4.6). 
A site boundary distance of 11 km (Attachment B, Figure B-1) is used to calculate xIQ
values 
from radiological releases from the surface facility. This distance corresponds to the distance 
from the DTF ventilation exhaust shaft to the nearest point on the site boundary that is the closest 
point where any member of the public could be standing, or living, at the time of a postulated 
radiological release (Section 4, Assumption 4.15). 
A site boundary distance of 8 km (Attachment B, Figure B-1) is used to calculate x/Q values 
from radiological releases from the subsurface repository exhaust shafts. This distance 
corresponds to the approximate distance between the subsurface repository and the nearest point 
of public access on the site boundary, which lies to the west (Section 4, Assumption 4.15). 
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Attachment B (Figures B-1 and B-2) shows the locations of the site boundaries, restricted areas, 
and preclosure dose limits. Doses to members of public onsite were calculated using MACCS2 
at various distances from the release points. For releases from the surface or subsurface 
facilities, the maximum dose to members of the onsite public occurs at the shortest distance from 
the release point. Because members of the public can be at any distance outside the restricted 
area and within the site boundary, the dose to members of the public is calculated at 100 m. 
The maximum dose for subsurface releases occurs at the shortest distance from the subsurface 
facility exhaust shafts. Doses to members of the public at 100 m and 3 km for releases from a 
subsurface exhaust shaft are reported in Section 6.3.1. 
The dose to members of the public at 3 km from subsurface facility releases is used in dose 
aggregation with the dose to members of the public at 100 m from surface facility releases in 
Section 6.3.2. 
The maximally exposed individual at the site boundary of 11 km is assumed to receive doses 
from the inhalation, resuspension inhalation, air submersion, groundshine, and ingestion 
pathways for a period of 8,760 hr (Section 4, Assumption 4.20). 
The onsite individual member of the public, at 100 m or 3 km away from a DTF, the FHF, or a 
subsurface exhaust shaft, is assumed to receive doses from inhalation, resuspension inhalation, 
air submersion, and groundshine pathways for a period of 2,000 hr (Section 4, Assumption 4.20). 
6.2.1.6 Food Consumption and Production Rates 
The consequence analysis using MACCS2 (ORNL 1998) requires that any farmland used in the 
50-mi radius from the point of a radioactive release be defined. Annual consumption rates of 
foods are from a 1997 regional survey of the population in the vicinity of Arnargosa Valley, as 
described in Characteristics ofthe Receptorfor the Biosphere Model (BSC 2004r). 
This analysis is conducted using the 1997 survey of food consumption rates for each food type, 
as shown in Table 17. These values are entered into input files for the COMIDA2 module. 
The COMIDA2 module uses the food consumption rates in Table 17 to generate the food 
pathway factor data file YUCCALA.BIN. Because the food production rates are not used in this 
calculation to calculate dose to maximally exposed individual members of the public from foods, 
the values of these input parameters are not discussed. 
This file is believed to provide conservative estimates of the individual dose from food and the 
total population dose from foods. The food pathway file name is specified in the CHRONC 
input file by keyword BIN -FILE. 
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Table 18. MACCS2 Computer Runs 
Run 
No. ATMOS File Out~utFile Remarks 
1 Surface releases from 1 percent of an annual throughput of 
3000 MTHM, average PWR inventory with HEPA filters, 
-elevated release 
2 Oxidation of 154 damaged fuel rods per year, average PWR 
inventory with HEPA filters, elevated release 
Subsurface releases from waste package surface contamination, 
neutron-activated air and dust without HEPA filters, ground- 
level release 
LAPWROI .INP LAPWROI .OUT 1 assembly, average PWR inventory with HEPA filters, ground-level 
release 
1 assembly, maximum PWR inventory without HEPA filters, ground- 
level release 
LABWROI .INP LABWROI .OUT 1 assembly, average BWR inventory with HEPA filters, ground-level 
release 
1 assembly, maximum BWR inventory without HEPA filters, ground- 
level release 
LAHLWOI .INP LAHLWOI .OUT 1 SRS HLW canister inventory without HEPA filters, ground- 
level release 
1 Hanford Site HLW canister inventory without HEPA filters, ground- 
level release 
1 West Valley HLW canister inventory without HEPA filters, ground- 
level release 
1 INEEL HLW canister inventory without HEPA filters, ground-level 
release 
LANAVOI .INP LANAVOI .OUT 1 naval SNF canister inventory without HEPA filters, ground- 
level release 
LASENOI .INP LASENOI .OUT Sensitivity analysis of Run 1 :surface releases from 1 percent of an 
annual throughput of 3000 MTHM, average PWR inventory with 
HEPA filters, elevated release. The release duration is changed 
from 24 hr to 1 hr 
Sensitivity analysis of Run 4: 1 assembly, average PWR inventory 
with HEPA filters, ground-level release. The deposition velocity is 
changed from 0.01 m/s to 0.001 m/s 
Sensitivity analysis of Run 4: 1 assembly, maximum PWR inventory 
with HEPA filters, ground-level release 
Sensitivity analysis of Run 4: 1 assembly, average PWR inventory 
without HEPA filters, ground-level release 
BWR = boiling water reactor; HEPA = high-efficiency particulate air; HLW = high-level radioactive waste; INEEL = 
Idaho National Engineering and Environmental Laboratory; PWR = pressurized water reactor; SNF = spent 
nuclear fuel; SRS = Savannah River Site. 
Source: ORNL 1998 
The ATMOS file in MACCS2 provides the radiological inventory, defines the radial distances of 
16 sector wind rose, establishes the dry or wet deposition factors, provides the release fractions 
for the predefined isotopic groups, and integrates the site-specific weather data with a selected 
frequency distribution. Only the ATMOS file is changed when a different case is run. The input 
parameters used in ATMOS, EARLY, and CHRONC output files for each of the sixteen cases 
are given in Attachment C. 
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Normal operations and Category 1 event sequence dose results are from the MACCS2 
(ORNL 1998) output file at either the mean or 5oth percentile probability level, depending on 
which probability level gives a higher dose. 
Category 2 event sequence dose results are from the MACCS2 (ORNL 1998) output file at the 
95th percentile probability level. Calculated doses are converted from Sv to mrem using a 
conversion factor of 1 Sv equal to 1.0 x lo5mrem. The majority of the dose is received during 
the initial plume passage. 
Public dose includes contributions from inhalation and cloudshine during plume passage plus 
resuspension inhalation, ingestion, and groundshine pathways during the long-term 
exposure period. Internal doses are calculated using a dose commitment period of 50 years. 
The values of DR, ARF, RF, (LPF),aSk, and (LPF)fac used in these MACCS2 (ORNL 1998) 
computer runs are summarized in Table 19. 
Table 19. Values of DR, ARF, RF, (LPF),,k, (LPF)hc, and (LPF)HEPA 
Used in MACCS2 
Run No. I DR ARF 
1.o 
1.o 
1.o 
1.o (g) 
0.1 (c) 
0.1 (P) 
0.1 (crud) 
1.o 
1.0 (g) 
0.1 (c) 
0.1 (PI 
0.1 (cN~) 
0.01 
0.01 
0.01 
0.01 
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Table 19. Values of DR, ARF, RF, (LPF),,k, (LPF)hc, and (LPF)HEP~ 
Used in 
MACCS2 (Continued) 
Run No. DR ARF 
12 0.3 (g) 
See 2.OE-04 (c) See 
3.OE-05 (p) ~ote~ 
1.5E-02 (crudlb 
13 0.3 (g) 
0.0Ia 2.OE-04 (c) 
3.OE-05 (p) 
1.5E-02 (crud)b 
14 0.3 (g) 
1.0 2.OE-04 (c) 
3.OE-05 (p) 
1.5E-02 (crudlb 
15 0.3 (g) 
1.0 2.OE-04(c) 
3.OE-05 (p) 
1.5E-02 (crudlb 
16 0.3 (g) 
1.0 2.OE-04(c) 
3.OE-05 (p) 
1.5E-02 (crud)b 
NOTES: (c) = cesium; (g) = fission product gases; (p) = particulate. 
a DR=O.OI is used in calculating the release fraction input to MACCS2 
(ORNL 1998) 
For crud, the value shown is the "effective ARF," which is the product of a crud 
spallation fraction of 0.15 and an ARF of 0.1 (BSC 2004m, Section 6.2.1.3) 
Radionuclide releases for Category 2 event sequences involving naval SNF were 
developed by the Naval Nuclear Propulsion Program (Gisch 2004) 
RFs for radionuclide releases for Category 2 event sequences involving naval 
SNF are identified in Table 5. 
ARF = airborne release fraction; DR = damage ratio; LPF = leak path factor; RF = respirable 
fraction. 
Run 1 involves a HEPA-filtered release from an annual throughput of 3,000 MTHM of PWR 
SFAs in a DTF, or the FHF, while Run 2 involves a HEPA-filtered release from oxidation of 
154 damaged fuel rods per year in a DTF or the FHF. 
Run 3 simulates an unfiltered subsurface release of waste package surface contamination, 
neutron-activated air, and silica dust during normal operations. 
Run 4 involves a HEPA-filtered release from a drop or collision of an Average PWR assembly 
while Run 5 simulates an unfiltered release from a drop or collision of a Maximum 
PWR assembly. 
Run 6 involves a HEPA-filtered release from a drop or collision of an Average BWR assembly 
while Run 6 simulates an unfiltered release from a drop or collision of a Maximum 
BWR assembly. 
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Runs 8, 9, 10, and 11 involve unfiltered releases from a drop or collision of an SRS HLW 
canister, a Hanford Site HLW canister, a West Valley HLW canister, and an INEEL HLW 
canister. 
Run 12 simulates an unfiltered release from a drop or collision of a naval SNF canister. 
Run 13 is a variation of Run 1; the radionuclide release duration input parameter used in Run 1 
was changed from 24 hr used in Run 1 to 1 hr used in Run 13. 
Run 14 is a variation of Run 4; Run 14 changes the particle deposition velocity input parameter 
used in Run 4 from 0.01 m/s to 0.001 m/s. 
Run 15 is a variation of Run 4; Run 15 changes the source term used in Run 4 from one average 
PWR assembly to one maximum PWR assembly. 
Run 16 is a variation of Run 4; Run 4 takes credit for HEPA filters while Run 16 does not. 
MACCS2 (ORNL 1998) analyses results are summarized in Tables 17, 18, and 22 to 24. 
6.3.1 Doses to Members of the Public from Normal Operations and Category 1 
Event Sequences 
MACCS2 (ORNL 1998) was used to calculate public TEDE to members of the public onsite at 
100 m and at the site boundary of 11 lun from normal operations (Table 20) and Category 1 
event sequences (Table 21). 
Table 20. Public Doses from Normal Operations 
Public Public Public 
Releases from: 
Normal operations: surface releases from 1 percent of 
3,000 MTHM SNF per year and oxidation of 154 
damaged commercial SNF rods per year, Average PWR 
SFAs, HEPA-filtereda 
Normal operations: subsurface releases of waste 
package surface contamination, activated air and silica 
dust. unfilteredb 
NOTES: a Table 18, MACCS2 Run 1 + MACCS2 Run 2 
Table 18, MACCS2 Run 3. 
HEPA = high-efficiency particulate air; mrem = one thousandth of a rem; MTHM = metric tons of heavy metal; 
PWR = pressurized water reactor; rem = roentgen equivalent man; TEDE = total effective dose equivalent. 
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Table 21. HEPA-Filtered Public TEDE for Category 1 Event Sequences 
Event 1 I 
II MAR 
(No. SFAS) 
HEPA-Filtered Public 
TEDE at 100 m 
(mremlevent) 
HEPA-Filtered Public 
TEDE at 11 km 
(mremlevent) 
Sequence Category 1 Avg Avg Avg Avg 
Identifier Events PWR~ BWR~ PWR' BWR~ 
GET-03D Drop of a DTF and FHF 
commercial SNF Waste Transfer 
assembly onto Cells 
another assembly 
in a transportation 
cask 
GET-03B Collision involving DTF and FHF 
a commercial SNF Waste Transfer 1 
assembly Cells 
NOTES: Avg PWR Fuel: Enrichment = 4.0 percent; Bumup = 48 GWdIMTU; Decay Time = 25 years. 
Avg BWR Fuel: Enrichment = 3.5 percent; Bumup = 40 GWdIMTU; Decay Time = 25 years. 
a Table 18, MACCS2 Run 4 
b~able
18, MACCS2 Run 6. 
Avg = average; BWR = boiling water reactor; DTF = Dry Transfer Facility (DTF 1 or DTF 2); FHF = Fuel Handling 
Facility; GWd = gigawatt day; HEPA = high-efficiency particulate air; MAR = material at risk; mrem = one 
thousandth of a rem; MTU = metric tons uranium; PWR = pressurized water reactor; rern = roentgen equivalent 
man; SFA = spent fuel assembly; SNF = spent nuclear fuel; TEDE = total effective dose equivalent. 
The maximally exposed individual at the site boundary of 11 km is assumed to receive doses 
from the inhalation, resuspension inhalation, air submersion, groundshine, and ingestion 
pathways for a period of 8,760 hr (Section 4, Assumption 4.20). The onsite individual member 
of the public, at 100 m or 3 km away from a DTF, the FHF, or a subsurface exhaust shaft, 
is assumed to receive doses from inhalation, resuspension inhalation, air submersion, and 
groundshine pathways for a period of 2,000 hr (Section 4, Assumption 4.20). OnlyHEPA-
filtered public TEDEs are reported in Table 2 1. 
In Table 20, the doses at 11 km are larger than the doses at 3 km, because: (1) the doses at 11 km 
include the ingestion pathway while the doses at 3 km do not, and (2) the doses at 11 km 
assumes an exposure period of 8,760 hr while the doses at 3 km assume an exposure period of 
only 2,000 hr. 
6.3.2 Sum of Doses from Normal Operations and Category 1 Event Sequences 
The dose from an average PWR assembly is higher than the dose from an average BWR 
assembly (Table 21) and, therefore, the average PWR dose per SFA is used in the calculation of 
the sum of frequency-weighted Category 1 event sequence doses using: 
where, 
FAdrop = Maximum number of SFAs affected in a drop event 
-
NdroP -Number of drops per year 
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F&OH = Maximum number of SFAs affected in. a collision event 
-
-
Ncoll Number of collisions per year 
-
-
DP, Average PWR drop or collision dose per SFA. 
The results of the dose aggregation calculation using Equation 18 are shown in Table 22. 
Table 22. Sum of Public Doses From Normal Operations and Category 1 Event Sequences 
Public Public 
Number TEDE at TEDE at 
of 100 m 11 km Public TEDE Public TEDE 
at 11 km
Event events1 (mreml (mreml at 100 m 
Sequence Releases MAR Year event)a eventlb (mrernl~ear)~ (mrern~~ear)~ 
Identifier from: 0(2) (3) 0(5)=(2)x(3) (6)=(2)x(4) 
N A Normal N A N A N A N A Annual Dose Annual Dose 
operations: 
3.02E+OOC 3.51 E-OIC 
surface 
releases from 
64 Average 
PWR SFAs per 
year, HEPA- 
filtered 
Normal N A N A N A Annual Dose Annual Dose 
operations: 
1.I7E-02' 1.67E-02' 
subsurface 
releases of 
waste package 
surface 
contamination, 
activated air 
and silica dust, 
unfiltered 
Drop of a Frequency Frequency 
commercial Weighted Weighted 
SNF assembly Dose Dose 
onto one 9.28E-01 4.32E-03 
assembly in a 
transportation 
cask 
Collision Frequency Frequency 
involving a Weighted Weighted 
commercial Dose Dose 
SNF assemblv 4.64E-01 2.1 6E-03 
Total: 4.43E+00 3.74E-01 
NOTES: a Onsite member of the public is located at 100 m away from a DTF or the FHF and at 3 km away from 
the subsurface exhaust shafts and is exposed to radiation for a period of 2,000 hr (Section 4, 
Assumption 4.20) 
b 
Offsite member of the public is located at 11 km away from a DTF or the FHF, and at 8 km away from 
the subsurface exhaust shafts and is exposed to radiation for a period of 8,760 hr (Section 4, 
Assumption 4.20) 
'Table 20, columns 1, 2, and 3 
d Table 21 
eTable 9, column 3. 
DTF = Dry Transfer Facility (DTF 1 or DTF 2); FHF = Fuel Handling Facility; HEPA = high-efficiency particulate air 
[filter]; NA~= not applicable; PWR = pressurized water reactor; SNF =spent nuclear fuel. 
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Table 23. Bounding Unfiltered Public Doses for Category 2 Event Sequences 
Highest 
TEDE 
Bounding Category 2 Events (mremlevent) 
Event GET-OlA(a): Drop and breach of a 4.76E+01 6.71 E+01 
transportation cask containing 36 maximum 
PWR SFAs without HEPA filtersb 
Event GET-OlA(b): Drop and breach of a 3.71E+01 
BWR transportation cask containing 74 
Maximum BWR SFAs without HEPA filtersC 
NOTES: Maximum PWR Fuel: enrichment = 5.0 percent; burnup = 80 GWdIMTU; decay time = 5 years. 
Maximum BWR Fuel: enrichment = 5.0 percent; burnup = 75 GWdIMTU; decay time = 5 years. 
aThe bone surface dose is the highest TODE 
Table 18, MACCS2 Run 5 
Table 18, MACCS2 Run 7. 
BWR = boiling water reactor; CDE = committed dose equivalent; DDE = deep dose equivalent; HEPA = high-
efficiency particulate air; LDE = lens dose equivalent; mrem = one thousandth of a rem; PWR = pressurized 
water reactor; rem = roentgen equivalent man; SDE = shallow dose equivalent to skin; TODE = total organ dose 
equivalent. 
Table 24. Unfiltered Public Doses From Breached Vitrified High-level Radioactive Waste Canisters 
and Naval Spent Nuclear Fuel Canisters 
TEDE Highest TODEa SDE LDE 
Bounding Category 2 Events (mremlevent) (mremleven t) (mremlevent) (mremleven t) 
Event GET-02B(n): Drop and breach of 1.I3E+00 5.86E+00 4.91 E-02 1.18E+00 
one naval SNF canister without HEPA 
filtersb 
Event GET-02B(a): Drop and breach of 
five SRS HLW canisters without HEPA 
filtersC 
Event GET-02~(b): Drop and breach of 
five Hanford Site HLW canisters 
without HEPA filtersd 
Event GET-02B(c): Drop and breach of 
five West Valley HLW canisters without 
HEPA filterse 
Event GET-02B(d): Drop and breach of 
five INEEL HLW canisters without 
NOTES: a The bone surface dose is the highest TODE for HLW canisters and naval SNF canisters 
Table 18, MACCS2 Run 12 
Table 18, MACCS2 Run 8 
Table 18, MACCS2 Run 9 
Table 18, MACCS2 Run 10 
'Table 18, MACCS2 Run 11. 
HEPA = high-efficiency particulate air; HLW = high-level radioactive waste; INEEL = Idaho National Engineering 
and Environmental Laboratory; LDE = lens dose equivalent; mrem = one thousandth of a rem; rem = roentgen 
equivalent man; SDE = shallow dose equivalent to skin; SRS = Savannah River Site; TEDE = total effective dose 
equivalent; TODE = total organ dose equivalent. 
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6.3.4 Sensitivity Analysis and Treatment of Uncertainty 
This section discusses sensitivity analysis performed for and treatment of uncertainties in public 
doses. A discussion of uncertainties in worker doses is beyond the scope of this report because 
the calculated worker doses presented in this calculation were provided by various references. 
6.3.4.1 Sensitivity Analysis 
A maximum release duration of 24 hr allowed in MACCS2 (ORNL 1998) is used for normal 
operations (Section 4, Assumption 4.6). An initial investigation (Table 18, Run 13) was 
performed whereby doses from a 1-hr and a 24-hr release are calculated; doses fiom the 
shorter 1 -hr release duration are higher than the longer 24-hr release duration. 
The finding of the investigation is considered reasonable because the radionuclides inhaled and 
ingested have a slightly longer time to decay and produce alpha and beta particles. Therefore, 
the use of a release duration of 1 day instead of 1 year with the same amount of radionuclide 
releases for normal operations is conservative. 
For MACCS2 (ORNL 1998) base cases, a particle deposition velocity of 0.01 m/s is used. 
For sensitivity analysis, a particle deposition velocity of 0.001 rn/s is used in MACCS2 Run 14 
(Table 18) with results showing a dose decrease by a factor of about 4. 
MACCS2 Run 15 (Table 18) has shown that the calculated doses increase by approximately a 
factor of 3 by use of a maximum PWR assembly instead of an average PWR assembly. 
MACCS2 Run 16 (Table 18) has shown that without HEPA filters the calculated doses at 100 m 
and 11 km away from the release point for a drop event involving two PWR SFAs could exceed 
the 100 mrem and 15 mrem dose limits for onsite and offsite members of the public, 
respectively. 
6.3.4.2 Treatment of Uncertainty 
Uncertainties in consequence analyses are addressed by using conservative inputs and 
assumptions. For example, for calculating public and worker doses for normal operations, it is 
conservatively assumed that 1 percent of received SFAs contain fuel rods with damaged 
cladding. As stated in Section 6.1.2.1, historical fuel discharge data have shown that the fuel rod 
failure rate is about 0.01 percent, versus 0.02 to 0.07 percent in the first 20 years of commercial 
nuclear power (BSC 2003c, p. G-6). This shows that the true fuel rod failure rate is at least one 
order of magnitude smaller than the 1 percent failure rate used in the consequence analyses. 
For calculating public and worker doses for Category 1 event sequences involving drops and 
collisions, it is conservatively assumed that 100 percent fuel rod cladding is breached. This is at 
least a factor of five larger than the best estimate value. 
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A comparison between MACCS2 Run 4 and Run 15 (Table 18) has shown that using the 
maximum PWR assembly inventory and the 95th percentile xIQ value instead of the average 
PWR assembly inventory and the mean xIQ value, the calculated doses for Category 1 event 
sequences involving drops and collisions increase by a factor of about 9. Even with this 
increase, the sum of normal operational doses and frequency-weighted Category 1 event 
sequence doses is still well within the regulatory dose limits given in Section 7. 
Uncertainties in calculating public doses for Category 2 event sequences are addressed by using 
conservative inputs and bounding assumptions. The bounding assumptions include the use of the 
bounding MAR (i.e., the maximum possible number of SFAs allowed in a certified cask) for a 
transportation cask, bounding source terms, 100 percent fuel rod cladding breach, and 95th 
percentile weather condition (i.e., xIQ values). The bounding MAR (i.e., 36 PWR or 74 BWR 
SFAs in a transportation cask) is at least a factor of 1.5 higher than the average value of the 
MAR. A comparison between MACCS2 Run 4 and Run 15 (Table 18) has shown that the 
bounding PWR and BWR source terms are at least a factor of 3 higher than the average PWR 
and BWR source terms. 
The assumption of a 100 percent he1 rod cladding breach (Section 4, Assumption 4.10) is very 
conservative in view of the fact that historical SNF drop incidents at nuclear power plants 
showed only minimum damage to the cladding. Run 4 (Table 18) has shown that the use of the 
95th percentile weather condition is about a factor of 3 higher than the mean value. 
The information on conservatism provided in this section ensures that uncertainties in inputs are 
treated with a reasonable degree of conservatism. 
6.3.5 Direct Radiation Exposures During Normal Operations, Category 1 Event 
Sequences, and Category 2 Event Sequences 
Offsite direct exposures originate only fiom the surface facilities, because subsurface facilities, 
such as the emplacement drifts, are shielded by massive rock. Surface facilities with potential 
sources of contribution to offsite direct exposures include the transportation cask bufferlstaging 
area and the SNF aging system. 
Other facilities, such as the CHF, DTF 1, DTF 2, and the FHF, provide concrete shielding for the 
exterior walls except for the entrance vestibules. This shielding typically reduces the dose rate 
outside the building to 0.25 mrem/hr (BSC 2004n, Section4.9.1.3) except during cask 
handling operations around the vestibule. These dose rates result in a negligible contribution to 
the public dose at the site boundary located at 11 km (Section 4, Assumption 4.15) fiom the 
surface facilities. 
The transportation cask bufferlstaging area provides space for staging of shielded truck and 
rail casks. The dose rate from an individual cask is regulated by the NRC (e.g., 10 mrem/hr at 
any point 2 m fiom the outer lateral surface of the transport vehicle specified in 10 CFR 
71.47(b)(3) for exclusive use shipment only). With multiple casks in the staging area, the dose 
rate will increase, depending on the cask staging configurations. This radiation level affects the 
worker dose, but poses no impact on the public dose at the site boundary because of the large 
distance factor. 
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In addition, 10 CFR 20.1301(a)(2) requires that each licensee shall conduct operations so that the 
dose in any unrestricted area from external sources does not exceed 0.002 rem (2 mrem) 
in any 1 hr. This requirement applies to normal operations and Category 1 event sequences. 
The dose requirements in 10 CFR Part 20 include contributions from direct exposures. 
Direct exposures from the SNF aging system arise from the effects of neutron and gamma 
skyshine. These effects depend on site-specific fuel characteristics of initial enrichment, burnup, 
and cooling time, along with facility layout, site layout, and site boundary characteristics. 
Skyshine radiation dose calculations are typically performed with the MCNP code 
(Briesmeister 1997). Aging cask vendors have included the skyshine analysis in their safety 
analysis reports submitted to the NRC to demonstrate regulatory compliance, such as Topical 
Safety Analysis Report for the HI-Storm 100 Cask System, HOLTEC Report HI-951312, 
Revision 2, NRC Docket No. 72-1014 (NRC 1997, Section 5.4.3, p. 5.4-4). 
NRC (1997, Section 5.4.3, p. 5.4-4) shows that the skyshine dose at 200 m from an independent 
spent fuel storage facility is in the range of 11.6 to 19.3 mredyear (NRC 1997, Table 5.1.9, 
p. 5.1-15) for array configurations 2 x 3,2 x 4, and 2 x 5. This dose would be further reduced by 
permanent or temporary shielding. With this dose level, the public dose at the site boundary 
from skyshine radiation is negligible, based on the distance factor and attenuation by air. 
6.3.6 Doses to Members of the Public Being ALARA 
Regulation 10 CFR 20.1 101(b) states that the licensee shall use to the extent practical both 
procedures and engineering controls based upon sound radiation protection principles to achieve 
occupational doses and doses to members of the public that are ALARA. 
To implement the ALARA goals of 10 CFR 20.1 101 (b), the calculated offsite public TEDE is 
compared with the 10 mremlyear offsite ALARA dose constraint (Section 7) along with other 
project ALARA goals. 
6.4 WORKER DOSE RESULTS 
The results of worker dose calculations are discussed in this section. 
6.4.1 Normal Operations 
6.4.1.1 Worker Dose From Airborne Releases 
The annual doses to surface facility workers from potential surface facility normal operational 
releases are calculated and summarized in BSC (2004g, Table 13) and BSC (2004e, Table 5) for 
a full time worker who works 2,000 hrlyear (BSC 2004g, Section 4.3.10). 
Results show that workers in a DTF, among the 30 receptor locations, receive the highest TEDE 
(Table 25). The estimated maximum annual surface facility worker dose from airborne releases 
is 1.6E+00 mredyear TEDE (Table 25). 
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Table 25. Annual Surface Facility Worker Doses From Normal Operations Releases 
Highest 
TEDE" TODE~ SDE~ LDE~ 
Released from (mremlyear) (mremlyear) (mremlyear) (mremlyear) 
DTF 1 8.4E-01 2.9E+00 3.5E+00 4.3E+00 
DTF 2 4.6E-01 1.6E+00 1.9E+00 2.4E+00 
FHF 2.7E-01 1.5E+00 1.9E+00 2.3E+00 
Maximum of 17A and 178 2.2E-03 NC NC NC 
Subsurface Exhaust Shafts 1.2E-02 NC NC NC 
TotalC 1.6E+00 6.OE+00 7.3E+00 9.OE+00 
NOTES: a BSC 2004e, Table 5 
BSC 20049, Table 13 
Total is the sum of DTF 1, DTF 2, FHF, Max. of 17A and 178, and subsurface exhaust 
shafts. 
DTF = Dry Transfer Facility; FHF = Fuel Handling Facility; LDE = lens dose equivalent; mrem = one 
thousandth of a rem; NC = not calculated; rem = roentgen equivalent man; SDE = shallow dose 
equivalent to skin; SNF = spent nuclear fuel; TEDE =total effective dose equivalent; TODE = total 
organ dose equivalent. 
Three potential sources of airborne radioactive material releases were calculated for the 
subsurface repository under normal operations, which are (BSC 2004g, Section 6): 
0 Radioactive contamination suspended from the external surfaces of the emplaced waste 
packages 
0 Neutron activation products generated from irradiation of ventilation air inside the 
emplacement drifts 
0 Neutron activation products generated from irradiation of the host rocks inside the 
emplacement drifts. 
Annual TEDEs to subsurface workers from potential normal operational releases are calculated 
and summarized in BSC (2004e, Table 5). The estimated maximum annual subsurface facility 
worker dose from airborne releases is 1.8E-01 mremlyear TEDE (Table 26). 
The annual maximum dose to a repository worker was calculated for a full time worker who 
works 2,000 hrlyear at the access main that is taking air from Intake Shaft 1, about 410 m from 
Exhaust Shaft 1 (BSC 2004g, Section 4.4.6). Airborne radioactive contamination in the access 
main may result from re-entry of the released 'material from the exhaust shafts. 
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Table 26. Annual Subsurface Facility Worker Doses From Normal Operations Releases 
TEDE Highest TODE SDE LDE 
Released from (mremlyear) (mremlyear (mremlyear) (mremlyear) 
DTF 1 2.9E-02 NC NC NC 
DTF 2 2.OE-02 NC NC NC 
FHF 3.5E-02 NC NC NC 
Maximum of 17A and 178 1.8E-03 NC NC NC 
Subsurface Exhaust Shafts 9.1 E-02 NC NC NC 
Total 1.8E-01 NC NC NC 
DTF = Dry Transfer Facility; FHF = Fuel Handling Facility; LDE = lens dose equivalent; rnrern = one thousandth 
of a rern; NC = not calculated; rern = roentgen equivalent man; SDE = shallow dose equivalent to skin; 
TEDE = total effective dose equivalent; TODE = total organ dose equivalent. 
Source: BSC 2004e. Table 5 
6.4.1.2 Worker Dose From Direct Radiation 
Annual worker doses from direct radiation from contained sources, such as transportation casks 
and waste packages, during normal operations were calculated for the CHF, DTF 1, DTF 2, the 
FHF, the TCRRF, the Remediation Facility, the SNF aging system, and the subsurface facility 
(Table 27). 
The highest annual dose to subsurface facility workers from exposure to direct radiation 
is 380 rnredyear for an emplacement/retrieval locomotive operator (Table 27), which does not 
exceed the ALARA goal to minimize the number of individuals that have the potential of 
receiving more than 500 rnredyear TEDE (BSC 2004n, Section 4.9.3.3). 
The highest annual dose to surface facility workers from exposure to direct radiation 
is 2,200 mredyear for a cask handling/emplacement/retrieval worker in the SNF aging system 
(Table 27). This dose rate exceeds the ALARA goal to minimize the number of individuals that 
have the potential of receiving more than 500 rnredyear TEDE (BSC 2004n, Section 4.9.3.3). 
For the final design, ALARA design considerations will be included to achieve the 
ALARA goal. 
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Table 27. Annual Worker Doses From Direct Radiation From Facility Normal Operations 
Worker Dose 
Facility 
Remediation 
SNF Aging System 
r-
Subsurface
r 
TCRRF 
NOTES: a Dose values are rounded to two significant values 
" 
b 
BSC 2004~ 
BSC 2004a, Sections 7.6 and 7.7 
Table 8, columns 3,4, and 5; 
BSC 2004c,Table 6, column 5 reports the maintenance worker dose 
BSC 2004d, Tables 7-1 and 7-2 and Section 7.1.5 
BSC 2004h, Table 12, column 2 
BSC 2004e, Table 4 
BSC 2004i. Table 10, column D. 
CHF = Canister Handling Facility; DTF = Dry Transfer Facility (DTF 1 or DTF 2); FHF = Fuel 
Handling Facility; HVAC = heating, ventilation, and air-conditioning; mrem = one thousandth of a 
rem; rem = roentgen equivalent man; SNF = spent nuclear facility; TCRRF = Transportation Cask 
Receipt and Return Facility. 
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6.4.2 Category 1 Event Sequences 
6.4.2.1 Surface Facility Worker Dose 
Category 1 event sequence scenarios and estimated maximum worker dose consequences are 
shown in Table 28. Dose consequences in mrem received by the maximally exposed worker for 
the Category 1 event sequences are calculated in BSC (2004b, Table 9). 
Table 28. Maximum Worker Doses From Category 1 Event Sequences 
Category 1Event TEDEa TODE~ SDE~ LDE~ 
Facilitf Sequencesa xIQ (s~m~)~ (mrem) (mrem) (mrem) (mrem) 
DTF 1 Drop 1.7E-04 4.6E-01 4.6E+00 5.6E+00 6.1 E+00 
collisionb 1.7E-04 2.3E-01 2.3E+00 2.8E+00 3.OE+00 
DTF 2 Drop 1.8E-05 4.9E-02 4.9E-01 6.OE-01 6.5E-01 
collisionb 1.8E-05 2.5E-02 2.5E-01 3.OE-01 3.2E-01 
FHF Drop 2.4E-04 6.6E-01 6.6E+00 7.9E+00 8.6E+00 
collisionb 2.4E-04 3.3E-01 3.3E+00 4.OE+00 4.3E+00 
NOTES: a BSC 2004b, Table 9 
Collision doses are 112 of the SFA drop doses. 
DTF = Dry Transfer Facility; FHF = Fuel Handling Facility; LDE = lens dose equivalent; mrem = one thousandth of a 
rem; rem = roentgen equivalent man; SDE = shallow dose equivalent to skin; TEDE = total effective dose equivalent; 
TODE = total organ dose equivalent. 
6.4.2.2 Subsurface Facility Worker Dose 
There are no Category 1 event sequences associated with the subsurface facility. Therefore, 
there is no subsurface worker dose from Category 1 event sequences. 
6.4.3 Sum of Worker Doses From Normal Operations and Category 1 Event Sequences 
Calculated worker doses do not exceed the Category 1 dose limits (Table 1) for releases from 
normal operations and Category 1 event sequences. 
Aggregation of worker doses calculated for normal operations and Category 1 event sequences 
(Table 28) is shown in Table 29. 
The combination of Category 1 event sequences is the same as that discussed in Section 6.3.2. 
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Table 29. Sum of Worker Doses From Normal Operations and Category 1 Event Sequences 
Frequency-
Weighted Dose 
Event (4)=(2)43) 
Sequence event) or Annual Dose 
Identifier Releases from: (mremlyear) 
N A Normal operations: surface and 1.6= 
subsurface airborne releases, 
HEPA-filtered 
N A Normal operations: direct radiation 
dose to a cask and waste Receipt 
operator at a DTF 
Drop of a commercial SNF 
assembly onto one assembly in a 
transportation cask 
Collision involving a commercial 
SNF assembly 
Total: I 2,202 
NOTES: a Table 25, column 2 
able 27, column 3 
'Table 28, column 4. 
DTF = Dry Transfer Facility (DTF 1or DTF 2); HEPA = high-efficiency particulate air; mrem = one thousandth of a 
rem; NA = not applicable; rem = roentgen equivalent man; SNF = spent nuclear fuel. 
The case of two commercial SNF assembly drops has the highest MAR (i.e., 4 SFAs). 
The radiation dose to the worker resulting from two commercial SNF assembly drops and normal 
operations is 2.92 redyear. 
7. CONCLUSIONS 
MACCS2 (ORNL 1998) and ARCON96 (BSC 2003b) are used to perform the public dose and 
worker dose calculations. Calculated doses are compared to the performance objectives shown 
in Table 30. 
Calculated public doses and worker doses show that 10 CFR 63.1 1 1 (a) and 10 CFR 63.1 1 1 (b) 
performance objectives are met by taking credit for the HVAC HEPA filters to mitigate the 
consequences of normal operations and Category 1 event sequences. 
The outputs listed in Attachment C have been found reasonable as compared to the inputs, and 
the results have been found suitable for their intended use. 
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Table 30. Summary Preclosure Dose Performance Criteria and Evaluation Results 
Standard Limits Results 
Public Exposuresa 
Preclosure standard: 10 CFR 63.204; 15 rnremlyea? 0.4 mremlyearC 
Preclosure Performance Objective for normal 
operations and Category 1 event sequences 
per 10 CFR 63.11 1(a)(2) 
0.009 mremlyeard 
0.4 mremlyeare 
Constraint specified for air emissions of 10 rnremlyea? 
radioactive material to the environment; 
not a dose limitation per 10 CFR 20.1 101 (d)' 
Dose Limits for Individual Members of the 100 mremlyeaPv 12 mrernlyea? 
Public for normal operations and Cate o 
9 ryl 2 mremlhr in any unrestricted area < < 2 mremlhr 
event sequences per 10 CFR 20.1301 from external sources 
Preclosure Performance Objective for any 5 remb 0.05 remb 
Category 2 event sequence per 50 rem organ or tissue dose other 0.2 rem 
10 CFR 63.1 11 (b)(2) than the lens of the eye 
15 rem lens of the eye dose 0.07 rem 
50 rem shallow dose to skin 0.02 rem 
Workers' Exposures 
Occupational Dose Limits for Adults from normal 5 remlyea? 0.001 rem/yearc 
operational emissions and Category 1 event 0.001 remlyeard 
sequences per 10 CFR 20.1201' 
0.001 rem/yeare 
2.2 remlyeat' 
50 remlyear organ or tissue dose 0.02 remlyealk 
other than the lens of the eye 
15 remlyear lens of the eye dose 0.02 remlyea+ 
50 remlyear shallow dose to skin 0.03 remlyeat" 
NOTES: a Results for public exposures are calculated at the site boundary 
TEDE 
Sum of TEDEs from normal operations and Category 1 event sequences 
d 
EDE from a single Category 1 event sequence in 1 year 
eTEDE from any combination of Category 1 event sequences in 1 year and from normal operations 
'10 CFR 63.1 11 (a)(l)would require operations area to meet the requirements of 10 CFR Part 20 
I0 CFR 20.1301(a)(l); dose limit to the extent applicable 
Onsite public dose at 100 m (Section 6.3.2) 
i 
10 CFR 63.1 11 (b)(l) would require repository design objectives for Category 1 and normal operations 
to meet 10 CFR 63.1 11(a)(l) requirements (10 CFR Part 20) 
The highest annual worker dose from direct radiation from facility normal operation 
k 
TODE from any combination of Category 1 event sequences in 1 year and from normal operations 
'Total LDE from anv combination of Cateaorv Ievent seauences in 1 year and from normal o~erations 
rn 
Total SDE from aily combination of catGgdry 1 event sequences in iyear and from normal operations. 
EDE = effective dose equivalent; LDE = lens dose equivalent; mrem = one thousandth of a rem; rem = roentgen 
equivalent man; SDE = shallow dose equivalent to skin; TEDE = total effective dose equivalent; TODE = total 
organ dose equivalent. 
Sources: Tables 18, 22, 23, 28, and 29, and Sections 6.3.2 and 6.4.3. 
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8. REFERENCES 
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Accident Analysis Handbook. NUREGICR-64 10. Washington, D.C. : U.S. Nuclear Regulatory 
Commission. ACC: MOL.20010726.0069. 
Sprung, J.L.; Ammerman, D.J.; Breivik, N.L.; Dukart. R.J.; Kanipe, F.L.; Koski, J.A.; Mills, 
G.S.; Neuhauser, K.S.; Radloff, H.D.; Weiner, R.F.; and Yoshimura, H.R. 2000. Reexamination 
of Spent Fuel Shipment Risk Estimates. NUREGICR-6672. Two volumes. Washington, D.C.: 
U.S. Nuclear Regulatory Commission. ACC: MOL.20001010.0217. 
Su, S.D.; Baylor, K.J.; and Engholm B.A. 1987. PATH Gamma Shielding Code User's Manual. 
GA-A167721, Rev. 1. San Diego, California: GA Technologies. TIC: 241215. 
Tsai, S-S. 2003. Users Manual: MACCS2 Version 1.12, STN: 11000-1.12-00. 1 1000-UM- 1.12- 
00, Rev. 00. Las Vegas, Nevada: Bechtel SAIC Company. ACC: MOL.20030820.0156. 
WVNS (West Valley Nuclear Services Company) 2001. WVDP Waste Form Qualification 
Report -Canistered Waste Form Specifications, Chemical Specification. Chapter 1 of Waste 
Form Qualification Report (WQR). WVDP-186. West Valley, New York: West Valley 
Demonstration Project. ACC: MOL.20020211 .0184. 
Yang, R.L. 1997. "Meeting the Challenge of Managing Nuclear Fuel in a Competitive 
Environment." Proceedings of the 1997 International Topical Meeting on LWR Fuel 
Performance, Portland, Oregon, March 2-6, 1997. Pages 3-1 0. La Grange Park, Illinois: 
American Nuclear Society. TIC: 232556. \ 
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8.2 CODES, STANDARDS, AND REGULATIONS 
10 CFR 20. Energy: Standards for Protection Against Radiation. Readily available. 
10 CFR 63. Energy: Disposal of High-Level Radioactive Wastes in a Geologic Repository at 
Yucca Mountain, Nevada. Readily available. 
10 CFR 71. Energy: Packaging and Transportation of Radioactive Material. Readily available. 
10 CFR 961.2004. Energy: Standard Contract for Disposal of Spent Nuclear Fuel andlor High- 
Level Radioactive Waste. Readily available. 
66 FR 55732. Disposal of High-level Radioactive Wastes in a Proposed Geologic Repository at 
Yucca Mountain, NV, Final Rule. 10 CFR Parts 2, 19,20,21,30,40, 51,60,61,63,70, 72,73, 
and 75. Readily available. 
ANSI N14.5-97. 1998. American National Standard for Radioactive Materials -Leakage Tests 
on Packages for Shipment. New York, New York: American Nuclear Standards Institute. TIC: 
247029. 
ANSIIANS-5.10-1998. Airborne Release Fractions at Non-Reactor Nuclear Facilities. La 
Grange Park, Illinois: American Nuclear Society. TIC: 235073. 
ANSIIANS-6.1.1- 1977. Neutron and Gamma-Ray Flux-to-Dose-Rate Factors. La Grange Park, 
Illinois: American Nuclear Society. TIC: 239401. 
Regulatory Guide 1.1 1 1, Rev. 1. 1977. Methods for Estimating Atmospheric Transport and 
Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors. 
Washington, D.C.: U.S. Nuclear Regulatory Commission. Readily available. 
Regulatory Guide 1.140, Rev. 1. 1979. Design, Testing, and Maintenance Criteria for Normal 
Ventilation Exhaust System Air Filtration and Adsorption Units of Light- Water-Cooled Nuclear 
Power Plants. Washington, D.C.: U.S. Nuclear Regulatory Commission. Readily available. 
Regulatory Guide 1.145, Rev. 1. 1982. Atmospheric Dispersion Models for Potential Accident 
Consequence Assessments at Nuclear Power Plants. Washington, D.C.: U.S. Nuclear 
Regulatory Commission. Readily available. 
Regulatory Guide 1.52, Rev. 2. 1978. Design, Testing, and Maintenance Criteria for 
Postaccident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and 
Adsorption Units of Light-Water-Cooled Nuclear Power Plants. Washington, D.C.: 
U.S. Nuclear Regulatory Commission. Readily available. 
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8.3 PROCEDURES 
AP-3.12Q, Rev. 2, ICN 2. Design Calculations and Analyses. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. ACC: 
DOC.200403 18.0002. 
AP-3.13Q, Rev. 3, ICN 3. Design Control. Washington, D.C.: U.S. Department of Energy, 
Office of Civilian Radioactive Waste Management. ACC: DOC.20040202.0006. 
AP-3.15Q, Rev. 4, ICN 5. Managing Technical Product Inputs. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. ACC: 
DOC.200408 12.0004. 
LP-SI.llQ-BSC, Rev. 0, ICN 0. Software Management. Washington, D.C.: U.S. Department of 
Energy, Office of Civilian Radioactive Waste Management. ACC: 20040225.0007. 
8.4 SOURCE DATA, LISTED BY DATA TRACKING NUMBER 
MOOOOlYMP00001.000MinimumDistances from Selected Yucca Mountain Project Sites to 
Public Lands. Submittal date: 01/26/2000. 
M00210MXHT8491.000. Mean Seasonal Mixing Height Data (0400 and 1600 PST) for Desert 
Rock, Nevada from 1984 through 199 1. Submittal date: 10/30/02. 
M00306WPMM9802.000. Wind Direction Sector, Wind Speed, Stability Class and Precipitation 
Data for MACCS2 98-02 Model. Submittal date: 06/06/2003. 
M00304MACCS280.000. Wind Direction Sector, Wind Speed, Stability Class and Precipitation 
Data for MACCS2 98-00 Model. Submittal date: 04/30/03. 
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.-.. " ./..-a '.<am. /LA*-
Source: BSC 2004s 
NOTE: Notes 1 through 13 are not applicable to this calculation. 
Figure A-1. Stack and Intake Location of Repository Surface Facilities 
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Intake Shaft #2 
Exhaust Raise #2 
South 
Source: BSC 2003d, Figure 10 
Figure A-2. Subsurface Air intake and Exhaust Locations 
A.3 MACCS2-Generated Atmospheric Dispersion Factors 
The output file for each of the 16 MACCS2 computer runs (Table 18) provides information on 
the calculated xIQ values at various distances from the release point (Table A-1). For each 
distance, the x/Q values are provided as mean, 50th, 90th, 95th, 99th, and 99.5th 
percentile values. 
The xIQ values in Table A-1 are used in Section 6.3 to calculate public doses. 
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Table A-1. xIQ Values for Normal Operations and Event Sequences 
;c/a Values at Various Distances 
From Release Point 
Releases from: 
Normal operations in surface 
facilities: surface releases from 
1 percent of 3,000 MTHM SNF per 
year, Average PWR SFAs, HEPA- 
filtereda 
Normal operations in subsurface 
facility: subsurface releases of 
waste package surface 
contamination, activated air and 
silica dust, unfilteredb 
DTF or FHF: Category 1 event 
sequences (drops and collisions)' 
CHF. DTF, or FHF: Category 2 
event sequencesd 
NOTES: a Table 18, MACCS2 Run 1 
able 18, MACCS2 Run 3 
'Table 18, MACCS2 Run 4 
d~able
18, MACCS2 Run 8. 
CHF = Canister Handling Facility; DTF = Dry Transfer Facility (DTF 1 or DTF 2); FHF = Fuel 
Handling Facility; HEPA = high-efficiency particulate air; MTHM = metric tons of heavy metal; 
NC = not calculated; PWR = pressurized water reactor; SFA = spent fuel assembly; SNF = 
spent nuclear fuel. 
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ATTACHMENT B 
PRECLOSURE RADIATION DOSE LIMITS 
Dose limits, both outside and inside the preclosure restricted area, are illustrated in Figures B-1 
and B-2. The extent of the geologic repository operations area is to be considered coincident 
with the restricted area, as illustrated in Figure B-2. 
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ATTACHMENT C -MACCS2 INPUT AND OUTPUT FILE LISTING 
MACCS2 (ORNL 1998) input and output files are listed in Table C-1 for sixteen cases 
(Table 18, Runs 1 to 16) on a compact disk submitted to Document Control. 
Table C-1. Listing of MACCS2 Computer lnput and Output Files 
MACCSZ lnput and Output File Size 
Run No. ~ileName Date Time (kB) 
1 LASUROI .INP 06/26/2004 1:41 PM 44 
1 LASUROI .OUT 06/26/2004 1.58 PM 720 
2 LASUR02.1NP 1011 312004 4:03 PM 44 
2 LASUR02.0UT 1011 312004 4:21 PM 720 
3 LASUBOI .INP 10/27/2004 3:22 PM 44 
3 LASUB01 .OUT 10/27/2004 3:29 PM 719 
4 LAPWROI .OUT 06/23/2004 6:08 PM 71 9 
5 LAPWR02.1NP 06/23/2004 7:35 PM 44 
5 LAPWR02.0UT 06/23/2004 750 PM 71 3 
6 LABWROI .INP 06/23/2004 8:03 PM 44 
6 LABWROI .OUT 06/23/2004 8:17 PM 719 
7 LABWR02.1NP 06/23/2004 8:25 PM 44 
7 LABWR02.0UT 06/23/2004 8:40 PM 713 
8 LAHLWOI .INP 09/23/2004 1:15 PM 44 
8 LAHLWOI .OUT 09/23/2004 1 :29 PM 71 0 
-
9 LAHLW02.1NP 09/24/2004 11 :32 PM 44 
9 LAHLW02.0UT 09/24/2004 11 :42 PM 71 1 
10 LAHLW03.1NP 10/01/2004 150 PM 44 
10 LAHLW03.0UT 10/01/2004 2:00 PM 71 0 
11 LAHLW04.1NP 09/25/2004 9:24 AM 44 
11 LAHLW04.0UT 09/25/2004 9:34 AM 709 
12 LANAVOl.INP 08/30/2004 2:41 PM 46 
12 LANAVOI .OUT 08/30/2004 257 PM 71 8 
13 LASENOI .OUT 06/26/2004 2:39 PM 719 
14 LASEN02.1NP 06/24/2004 251 PM 44 
14 LASEN02.0UT 06/24/2004 3:06 PM 71 9 
15 LASEN03.1NP 0711 312004 12:27 PM 44 
15 LASEN03.0UT 0711 312004 12:46 PM 71 9 
16 LASEN04.0UT 10/25/2004 3:45 PM 71 9 
All YMP1999.lNP 0713 112003 1 :27 PM 163 
All YMPCHI .INP 0511 712004 9:33 AM 11 
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Table C-I. Listing of MACCS2 Computer lnput and Output Files (Continued) 
MACCS2 Input and Output File Size 
Run No. File Name Date Time (kB) 
5, 7, and 8 to 12 YMPEA.INP 08/22/2003 12:43 PM 12 
1,2,3,4,6, 13, and 14 YMPEA1.INP 08/05/2003 3:05 PM 12 
1 and4 to 14 YMPSIT.INP 0711 712003 7:39 AM 8 
3 YSITE8.1NP 05/07/2001 7:27 AM 8 
All YUCCALABIN 0311 012004 4:11 PM 1,307 
All YMPD825.INP 02/22/2000 7:38 AM 663 
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