BWR Source Term Generation and Evaluation Rev 00A, ICN 00
 
000-00C-MGRO-00200-000-00A

July 2003

1 PURPOSE 
This calculation is a revision of a previous calculation (Ref. 7.5) that bears the same title and has 
the document identifier BBAC00000-01717-0210-00006 REV 01. The purpose of this revision 
is to remove TBV (to-be-verified) -41 10 associated with the output files of the previous version 
(Ref. 7.30). The purpose of this and the previous calculation is to generate source terms for a 
representative boiling water reactor (BWR) spent nuclear fuel (SNF) assembly for the first one 
million years after the SNF is discharged from the reactors. This calculation includes an 
examination of several ways to represent BWR assemblies and operating conditions in SAS2H in 
order to quantify the effects these representations may have on source terms. These source terms 
provide information characterizing the neutron and gamma spectra in particles per second, the 
decay heat in watts, and radionuclide inventories in curies. Source terms are generated for a 
range of burnups and enrichments (see Table 2) that are representative of the waste stream and 
stainless steel (SS) clad assemblies. During this revision, it was determined that the burnups 
used for the computer runs of the previous revision were actuallv about 1.7% less than the stated, 
or nominal, burnups. See Section 6.6 for a discussion of how to account for this effect before 
using anv source terms from this calculation. The source term due to the activation of corrosion 
products deposited on the surfaces of the assembly from the coolant is also calculated. 
The results of this calculation support many areas of the Monitored Geologic Repository (MGR), 
which include thermal evaluation, radiation dose determination, radiological safety analyses, 
surface and subsurface facility designs, and total system performance assessment. This includes 
MGR items classified as Quality Level 1, for example, the Uncanistered Spent Nuclear Fuel 
Disposal Container (Ref. 7.27, page 7). Therefore, this calculation is subject to the requirements 
of the Quality Assurance Requirements and Description (Ref. 7.28). The performance of the 
calculation and development of this document are carried out in accordance with AP-3.12Q, 
Design Calculation and Analyses (Ref. 7.29). 
2 METHOD 
The SAS2H sequence in SCALE 4.3 is used to calculate the thermal and radiation source terms 
for selected fuel assemblies as a function of assembly average burnup and cooling time. The 
prime functional module of the SAS2H code sequence utilized is the ORIGEN-S code. This 
code performs a point depletion and decay calculation of a selected fuel type with user-specified 
irradiation conditions and decay times. The crud (activated corrosion products) source terms are I 
determined via a simple spreadsheet calculation in Excel. 
.3 ASSUMPTIONS 
3.1. It is assumed that a single assembly can approximate various BWR assembly types, and 
that the source terms generated will not be greatly affected by this approximation in 
geometry. This assumption results in the path A and B representations for all assemblies 
in this calculation being identical to a representative assembly. Information for a General 
Electric (GE) 213 8x8 assembly is assumed to generate conservative source terms. This 
assumption is based on the fact that the assembly has a high initial heavy metal loading 

Title: BWR Source Term Generation and Evaluation 
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(IHML), available operating data, and assembly hardware information. In the cases where 
the data on the hardware is lacking in sufficient detail to model it in SAS2H, data from 
other BWR assemblies are used to fill in a conservative manner. Different IHMLs can be 
accounted for by adjusting the fuel length, which is calculated from the MML and the fuel 
density. The basis for this adjustment is that source terms are known to be sensitive to 
initial uranium loading and burnup (Ref. 7.3), but they are expected to be much less 
sensitive to variations in fuel transverse dimensions. This assumption is to be confirmed 
during a future sensitivity calculation concerning geometric representations in SAS2H. 
This assumption is used throughout Section 5. 
The channel of the stainless steel (SS) clad assemblies is assumed to be Zircaloy-4. This 
assumption is corroborated by Ref. 7.2, which states that only the earliest versions of 
BWR assemblies used SS channels (those for Dresden-1, Humboldt Bay, and Big Rock 
Point), and that Zircaloy-4 became the standard material. Therefore, the SS clad 
assemblies calculated in this report use a Zircaloy-4 channel. This assumption is used 
throughout Section 5. 
It is assumed that the flux scaling factors for the assembly hardware regions (bottom endfitting, 
plenum, and top end-fitting) are 1.5 times the values (Ref. 7.6, Table S.1, p. vi) 
recommended by the U.S. Nuclear Regulatory Commission (NRC). The rationale is that 
this assumption provides about 50% margin for the neutron-activated sources in the 
hardware regions. This assumption is used in Section 5.2. 
The cladwater rod (WR) material for SS assemblies is assumed to be SS348H. This 
assumption is corroborated by information in Ref. 7.15, p. 2A-16. This assumption is 
used in Section 5, Table 4. 
The hardware parts of the BWR assembly described in Ref. 7.15, p. 2A-158, and used in 
Table 5 (see footnote 5), are assumed. The rationale for this assumption is that fuel 
assembly data are proprietary information and that the information in Ref. 7.15 is the best 
available. Since a 50% margin is applied to the neutron flux scaling factors (Assumption 
3.3) to account for the uncertainty in the information, this assumption will result in 
conservative activation in the hardware regions. This assumption is used in Section 5, 
Table 5. 
It is assumed that the ANF 9x9 JP-4 assembly will generate the bounding crud (activated 
corrosion products) source term and that the physical characteristics of the ANF 9x9 JP-4 
assembly can be obtained from Ref. 7.2, p. 2A-24. Again, the rationale is that fuel 
assembly data are proprietary and the information in Ref. 7.15 is the best available 
technical data for this assembly. The rationale for using this assembly is that the crud 
source is proportional to the surface area that is exposed to coolant. Since the ANF 9x9 
JP-4 assembly has a larger surface area than the GE 8x8 assembly, it will generate a more 
conservative (higher) crud source that will serve as the bounding crud source for the 
representative BWR assembly. This assumption is used in Sections 5.2 and 5.4.4, and in 
Attachment VI. 

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4 USE OF COMPUTER SOFTWARE AND MODELS 
4.1 SOFTWARE APPROVED FOR QA WORK 
Program Name: SCALE 
Versioflevision Number: 4.3 Hewlett Packard (HP) 9000 Version, HP-UX V10.20 
Computer Software Configuration Identifier (CSCI) Number: 300 1 1 V4.3 (Ref. 7.1) 
Computer Type: Hewlett Packard 9000 Series 
Computer Processing Unit Name and Civilian Radioactive Waste Management System 
(CRWMS) Management and Operating (M&O) Contractor Tag Number: 'Bloom,' 
CRWMS-M&O Tag 700887 
The SAS2H computer code sequence of the SCALE V4.3 code system was developed to perform 
depletion analyses to obtain radiation and heat generation sources of spent fuel to be used in 
subsequent analyses. It is appropriate for this application, has been validated over the range it is 
used, and was obtained from Software Configuration Management in accordance with the 
appropriate procedures. The echo of the input files is contained in the *.cut files, which are listed 
in Attachment XII and available in the CDs of Attachment VII. The *.cut files, which consist of 
the input echoes and the final ORIGEN-S output, are sufficient to independently repeat the 
calculation. 
4.2 EXEMPT SOFTWARE 
4.2.1 Excel 
The Microsoft Excel 97 spreadsheet program is used to perform simple calculations as 
documented in Section 5 of this calculation. The user-defined formulas, input, and results are 
documented in sufficient detail in Section 5 to allow for independent repetition of the various 
computations without recourse to the originator. This software is considered exempt from the 
requirements of AP-SI.lQ, Software Management (Ref. 7.31, Section 2.1.6). 
4.2.2 Script Files 
Titles: neutrons, gammas, curies, watts 
Versioflevision Number: All are version 00 

Title: BWR Source Term Generation and Evaluation 
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Computer Type: HP 9000 Series 
The specific task of each script is noted in Table 1. The script files are provided in Attachments 
VII and XI. They are executed with the 'awk' command by typing 'awk -f (script file) (input 
*.cut file name) > (output file name). These files are intended for use only with the appropriate 
*.cut SAS2WORIGEN-S output files listed in Attachment XII. The output of the script files has 
been verified by visual inspection. This software is considered exempt from the requirements of 
AP-SI. lQ, Software Management (Ref. 7.3 1, Section 2.1.1). 
Table 1 .Script Files 
5 CALCULATION 
File Name 
neutrons 
gammas 
watts 
curies 
5.1 REVISION HISTORY 
Function 
Extracts the total (alpha-n plus spontaneous fission) neutron source table from a '.cut file 
Extracts the gamma source from the light element, actinide, and fission product contributions from a *.ci 
Extracts the total thermal output from the light element, actinide, and fission product contributions from 6 
Extracts the tables of nuclide curies from a *.cut file for the light element, actinide, and fission product q 
Several calculations have been performed to provide BWR SNF source terms for shielding and 
thermal calculations. Ref. 7.3 was the first in a series of calculations to provide source terms, but 
was limited to a handful of burnups and enrichments that represented the anticipated average and 
bounding waste stream assemblies. The source terms represented only the first twenty-five years 
after discharge from the reactor. In addition, only the ORIGEN-S module was used to bum and 
decay the assembly. Ref. 7.4 replaced Ref. 7.3, and used the SAS2WORIGEN-S sequence to 
calculate the burnup and decay of the assembly to 1 million years. Ref. 7.4 included a wide 
range of burnups and enrichments, and the inclusion of a crud source. Reference 7.5 was a 
revision of Ref. 7.4, and was intended to provide the most complete set of conservative source 
terms for the SS clad and representative BWR assemblies in the expected waste stream. This 
calculation is a revision of Ref. 7.5 to clear TBV-4110. 
The last two revisions incorporate several changes. While it had been assumed in the previous I 
revision that increasing the fuel density slightly would have a negligible impact on the source 
terms, this result was not verified by the calculations in Ref. 7.4, pp. 46 and 47. Therefore, in the 
Iast two revisions, the fuel density for the waste stream is constant, and the fuel length is 
calculated from the heavy metal loading (see Section 5.5.2.1). Secondly, SS clad assemblies are 
included in this calculation. In addition, impurities are accounted for in the fuel, and the material 
definitions are modified slightly to increase the amounts of those elements that contribute to the 
source terms (particularly cobalt in steel). The number of enrichments and bumups (see Table 2) 
were increased, as well as the number of time steps out to 1 million years (see Table 44). The I 
sources provided are expanded to include the inventory in curies of 61 radionuclides of interest 
(see Section 6.5). Calculations are performed to evaluate the effects of impurities in the fuel, and 

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the discussion of the ten axial node calculations is revised. Additional calculations are 
performed to provide radionuclide inventories for screening purposes. I 
5.2 SELECTION OF CONSERVATIVE PARAMETERS 
The inputs for this calculation are chosen to lead to conservative source terms. This section 
discusses the main inputs and the reasons they are used. It covers several different parameters. 
The first of these is the geometry for the waste stream and SS clad assemblies. 
In this calculation, the geometry used to model the assemblies in SAS2H corresponds to a GE 213 
fuel assembly. The GE 213 has a high uranium loading and a large amount of assembly 
hardware, supporting more fission product generation and hardware activation. Therefore, this 
assembly provides a conservative basis for the BWR waste stream. The GE 213 assembly IHML 
is increased to 200 kg for the waste stream and SS clad assemblies to provide slightly higher 
source intensities. 
As mentioned in Section 5.1, additional heavy metal mass is accounted for by increasing the fuel 
length, rather than the fuel density. A longer active fuel length and a lower density rather than a 
shorter fuel length and higher density decreases the fuel self-shielding. This results in a higher 
flux and consequently to a higher source intensity. In this calculation the IHML of the waste 
stream and SS clad assemblies is accounted for by the fuel length rather than the fuel density (see 
Section 5.5.2.1). 
The irradiation of the assembly hardware is calculated by inputting the desired amount of light 
element material into the fuel region in SAS2H. When the sources are calculated, only the light 
elements are included. The SS clad assemblies use waste stream files for the bottom and top 
assembly regions. The hardware sources are calculated by irradiating the material in the fuel 
region; scaling factors (SF) are used to account for the lower flux seen in the non-fuel regions of 
the reactor. The SFs for the neutron flux, provided in Ref. 7.6, Table S.l, have an uncertainty of + 50%. To generate conservative source terms for the non-fuel regions of the BWR assembly, 
the SFs used in this calculation represent 150% of those listed in Ref. 7.6, Table S.l, p.vi 
(Assumption 3.3). 
Consideration has also been given to the material definitions. The compositions of Inconel X- 
750,Zircaloy-2 and -4, SS 304, and SS 348H are representative of materials used in the 
manufacture of nuclear fuel assemblies. These compositions use the maximum amounts of 
cobalt given by the references and a 0.08 wt% cobalt impurity (Ref. 7.7, p. 45) for the steels. The 
composition for SS 348H given by Ref. 7.9, Table 1, includes a 0.2 wt% cobalt content. The 
balance of the remaining elements are representative of the material compositions for each 
material, but are biased towards the maximum amount of Sn, Ni, and Nb. Impurities are also 
included in the fuel itself, and the concentrations are given in Table 13. 
Not all the activation sources can be calculated with SAS2H. The source due to corrosion 
material that accumulates on the surfaces of the assembly from the flow of coolant (crud) is also 
calculated. A bounding crud source term is based on the Exxon ANF 9x9 JP-4 assembly I 

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(Assumption 3.6). This representation of the ANF 9x9 JP-4 does not use GE 213 dimensions. I 
The AN@ 9x9 JP-4 has a greater surface area exposed to coolant, and the crud source is heavily 
dependent on the available surface area. This dependence comes from the source being 
calculated with a radioisotope activity density (ci1cm2). Three estimates for radioisotope 
densities are used, from three different sources (Ref. 7.10,7.11, and 7.12). This calculation is 
discussed in detail in Section 5.5.3. 
The enrichments calculated range from 0.71 1 wt% (natural) to 5.5 wt% (the complete list is 
shown in Table 2). This is done to cover the wide range seen in the waste stream and avoid the 
need to extrapolate for information on assemblies currently being developed. The number of 
time steps is increased to 180, providing more detailed information for the first 100 years after 
SNF is discharged from the reactors. 
5.3 CALCULATION DESCRIPTION 
This calculation uses SAS2WORIGEN-S to generate source terms for the BWR SNF. Due to the 
variations between BWR assembly types, calculations are performed to justify a conservative 
configuration in SAS2H with which to represent all BWR assemblies in the waste stream. 
Sections 5.5.1.1 - 5.5.1.4 develop source terms for several variations of BWR assemblies - 
covering a range of water rods (WRs), moderator density, and gadolinium doped rods (GDRs). 
Section 5.5.1.5 presents the results of this parametric study. The configuration representing a 
conservative combination of the number of WRs, GDRs, and thermal-hydraulic information is 
used in Section 5.5.2 to generate two different representations of the assembly. 
The first representation determined in Section 5.5.2 uses one node to describe the fuel region. 
This is done for the waste stream, SS clad, and the thermal design basis fuel (DBF - the 3.74%, 
48 GWd/MTU case) assemblies. The combinations for which source terms are calculated are 
shown in Table 2. The second representation divides the active fuel region into ten axial nodes. 
The ten-node calculation covers only one assembly with a 5.05% initial enrichment, all the 
burnups shown, and a 200 kg IHML. This is done to justify modeling a BWR assembly with one 
axial region; large variations in moderator density with core height require some representative 
density to be used for the whole core. An axial peaking factor (APF) is calculated from 
information in Ref. 7.13, pp. 290-544, and is intended to be applied to the one-node calculation 
results to compensate for under-predictions. Since the nodal burnup data from pp. 290-544 of 
Ref. 7.19 (a revision of Ref. 7.13) are essentially identical to the data from Ref. 7.13 (differing 
generally only in the 5" significant digit), APFs based on either Ref. 7.13 or Ref. 7.19 will be 
identical for design purposes. Section 6.1 shows a comparison of the ten-node and one-node 
calculation with an APF. This demonstrates the use of the APF, and addresses concerns about 
using a one-node representation of a BWR assembly to develop source terms. 
The thermal DBF is used to illustrate the effect of fuel density on source terms. The waste 
stream and SS clad assembly cases are calculated with a 200 kg IHML. The thermal DBF is 
calculated with both 170 and 200 kg IHML. For the thermal DBF, the extra metal is accounted 
for by increasing the fuel density. As is seen in Section 6.2, the increased fuel density results in a 
less conservative source. As mentioned in Section 5.2, the waste striam and SS clad assemblies 

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account for the higher IHML by increasing the fuel length rather than the density. It should be 
noted that the cases comparing the one- and ten-node calculation, and the thermal DBF 
calculations (see notes in Table 2) are unchanged from Ref. 7.4, and include none of the revisions 1 
noted in Section 5.1. These cases are not to be compared directly to the waste stream and SS 
clad assemblies in this calculation. 
Table 2.BWR Enrichments, Burnups, and IHMLs for Source Terms are ~alculated' 
Initial ""U (wt%) 0.711, 1.0, 1.5, 2.0,2.5, 3.0,3.5*, 3.74, 4.0*, 4.5, 5.0, 5.05, 5.5 
Initial Heavy Metal Mass (kg) 200*, 170 
Final Assembly Average Burnup 
(GWdIMTU) 0.001, 0.01, 0.1, I*, lo*, 20*, 30*, 40*, 49, 50, 60, 70,75 
5.4 CALCULATION INPUTS 
This section outlines the information used in the calculation of the source terms. Several 
parameters are listed as calculated; the calculations of these parameters are shown in the tables or 
in Section 5.5. 
5.4.1 Physical Description of BWR Fuel Assembly 
The physical characteristics of a GE 213 8x8 BWR fuel assembly are provided in Tables 3,4, 
and 5. Figure 1 presents a general drawing of a BWR assembly. This is the information required 
to generate the SAS2H input files needed to create the BWR source terms. The number of digits 
cited herein may be a result of a calculation or may reflect the input from another source; 
consequently, the number of digits should not be interpreted as an indication of accuracy. I 
Table 3.GE 213 8x8 BWR General Assembly Data 
Assembly Parameter Value Units Reference 
Total Number of Rod Positions 64 /assembly Ref. 7.19, p. 5 I 
Number of WRs 1,2,or4 /assembly Ref. 7.19, pp. 5, 11 I 
Number of Fueled Bumable Poison Rods 0-9 Ref. 7.19, p. 13 
Gadolinium Enrichment in the Burnable Poison Rod: 3.0-4.0 wt% 
I 
Channel Material Zircaloy-4 N A ~ Ref. 7.14, p. A-1 I 
1. The parameters shown with an asterix are used for the SS clad assemblies. The entries in bold italics are used 
only for the thermal DBF. The 5.05 wt% enrichment is used only for the comparison between the one- and ten-node 
calculations. The *.cut files for the 5.05 wt% are part of Ref. 7.26. 
2. Not applicable. 
I 

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Table 4.GE 213 8x8 BWR Assembly Dimensions 
Fuel Rod Plenum Length 1 1.24 inches 28.5496 cm 
6.65+0.85+0.84+0.435= inches 22.2885 
Ref. 7.14, p. C-12 
Top End Fitting ~ e n ~ t h ~ 8.775 ' cm 
Total Assembly Length 171.40 inches 435.356 cm 
CladMlR ater ria? SS 348H N A N A NA Assumption 3.4 
Table 5.BWR Assembly Hardware 
Region Reference Part Name Ib/Assembly kg/Assembly Material 
Ref. 7.14, p. A-8 Tie plate 4.409 2.000 SS 304 
TOP Assumption 3.5 Compression sprin~ 1.279 0.580 lnconel X-750 
Channel calculated calculated Zirc-4 
Ref. 7.14, p. A-8 Tie plate 10.516 4.770 SS 304 
Bottom Calculated Channel calculated calculated calculated 
Channel calculated calculated Zirc-4 
Water rod calculated calculated Zirc-2 I SS 348H 
Plenum Getters 1.360 0.617 SS 304 
Ref. 7.14, p. A-8 Plenum springs 3.748 1.700 SS 304 (Inwnel X-750)' 
Channel calculated calculated Zirc-4 
Fuel Ref. 7.14, p. A-8 Spacer gridsY 4.299 10.717 1.950 10.325 Zirc-4 1 lnconel X-750 
Water rod calculated calculated Zirc-2 I SS 3481-1 
3. Channel length for part 2 is used. 
4. Axial dimensions are taken from a drawing in Ref. 7.14. Note that the plenum spring hardware is replaced with a 
more conservative material, Inconel (Table 5, footnote 8). 
5. The extra 0.435 inches is taken from the fact that the dimensions on the drawing do not add up to 171.40 inches; 
to compensate for this, the top end fitting is adjusted to take on the extra length; because of the location, this will 
have negligible impact. 
6. Information is used for the SS clad assemblies. 
7. Information for the compression springs is taken from Ref. 7.15, p. 2A-158, and is originally given in kg. 
8. Since SS304 was not the most stressing material, Inconel X-750 was substituted. 
9. There are seven incore spacers total. Spacer placement is noted on the bottom of Ref. 7.14, p. A-8. 

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Upper Tie Plate 
Plenum Springs 
Central WR 
Spacer Grid 
Fuel Rod 
Channel 
Lower Tie Plate 
Figure 1. BWR Assembly Drawing 

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5.4.2 Operating Parameters 
The parameters in this section, presented in Tables 6,7, and 8 represent operating conditions for 
a Quad Cities, Unit 2 assembly (C3 is used, however, the choice of assembly is arbitrary). Table 
6 provides general information. Information in Table 7 is required to determine the moderator 
temperature in Sections 5.5.1.3 and 5.5.2.2. The data in Table 8 (Ref. 7.13, pp. 333-334) are 
used to determine moderator density profiles, power profiles, and burnups for the parameter 
I 
study, one- and ten-node comparison cases, and the thermal DBF calculations. The fuel 
temperatures shown in Table 9 (Ref. 7.19, pp. 333-334) are an update of those given in Table 8 
(Ref. 7.13, pp. 333-334). The newer fuel temperatures are used to calculate the average fuel 
temperature (as done for the one-node calculations) for the waste stream and SS clad 
calculations. Reference 7.19 is Revision 01 of Ref. 7.13 (Revision 00). It should be noted that 
the ten-node water density data for assembly C3 from Ref. 7.19 are identical to the data from 
7.13, and the ten-node burnup data for assembly C3 differ at most by one in the significant 
digit. Since the temperature data from Ref. 7.13 in Table 8 are not used for the waste stream and 
SS clad source term calculations whose results may be used for other calculations such as waste 
package shielding and thermal studies, the information from Ref. 7.13 can be considered 
"reference only" in accordance with AP-3.15Q, Managing Technical Product Inputs (Ref. 7.33). 
The results of calculations using information available only from Ref. 7.13 are used solely 
internally to this document. 
Table 6.Typical Assembly/Core Operating Parameters for the GE 8x8 Assembly 
Parameter Value Units Reference 
Number of Assemblies In Core 724 N A Ref. 7.19, p. 4 
Total Core Thermal Power 251 1 MWth Ref. 7.19, p. 3 
. I 
Average Assembly Power 251 1/724=3.468 MWIassem bly Calculated 
I 
Cladding Temperature 620 K Ref. 7.16, p. S2.6.12, Table S2.6.4 
Ref. 7.19, p.5 
I 
Operating Pressure 1020 psia 1 
Table 7.Steam Table Information 
[ Temperature ("C) I Temperature (K) I Pressure (bar) I Pressure (psia) I Reference 4 I 
290 ( 563.15 1 74.449 1 1079.7915 
Ref. 7.18, p. 81 1 
300 1 573.15 1 85.917 1 1246.1207 

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Table 8.Moderator Density, Fuel Temperature, and Burnup Profiles 
I 
Pnint 6 1 7 1 8 1 9 1 1 0 1 1 1 1 1 2 1 1 3 1 
Burnup (GWdMTU) I 
Additional information regarding the final nodal burnup for all the assemblies included in Ref. 
7.13 (pp. 290-544) is also used in this calculation. Due to the amount of information, it is 
presented in Attachment III. This information is used to generate Figure 6. 
10. Node 1 is at the bottom of the assembly; 10 is the top. 

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Table 9.Updated Fuel Temperature Profiles 
Also required for this calculation is the height of each node in the active fuel region. This 
information is shown in Table 10. 
Table 10. Node ~eights" 
Node 1 2 3 4 5 6 7 8 9 10 
Height(cm) 15.24 30.48 30.48 45.72 30.48 45.72 45.72 45.72 64.11 15.24 
11. The dimensions of the nodes shown in Table 10 (Ref. 7.13, p. 15, Fuel Type 9) do not represent all of the 
assembly types listed in Ref. 7.13. Fuel assemblies E-M use a slightly different node description, with nodes 9 and 
10 measuring 48.87 and 30.48 cm, respectively. 

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Table 1 1 details information taken from Ref. 7.13, Table 3-8, p. 18 or from Ref. 7.19, Table 3-8, 
p. 18. This table lists the information on the data points (DP) and state points (SP) for which 
thermal hydraulic data are given in Ref. 7.13 and Ref. 7.19. The effective full power days 
(EFPD) are used along with the thermal hydraulic information to determine average values for 
the SAS2H input. 
Table 11. Data Point and State Point Information 
DP I Cycle I Cycle Position 
DPI 19 I BOC'~ 
DP3 1 10 I BOC 
DP4 1 10 I MOC 
10 EOC 
DP5 11 BOC 
DP6 I 11 I MOC 
I 11 I EOC 
-- 
DP7 1 12 I BOC 
DP8 1 12 I MOC 
1 12 I EOC 
SP9 1 13 I BOC 
SPIO 1 13 I MOC 
SP11 13 MOC 
DP12 13 MOC 
SP13 13 MOC 
1 13 I EOC 
SP14 1 14 I BOC 
DP15 1 14 1 MOC 
BU~~UD I EFPD 1 
3554.1 6 MW~MTU") 1 180.3 
9544.1 1 MWdlMTU 1 484.2 I 
0 0 
2820.04 MWdlMTU 142.2 
5229.25 MWdlMTU 263.7 
0 0 
201.61 MWd/MTU 10.1 
2257.20 MWdlMTU 1 112.94 
4484.31 MWdlMTU 1 224.4 
- 
0 l o 
1878.65 MWdlMTU 1 93.56 - - I 1 - I I 
SP16 1 14 I MOC 1 4238.45 MWdMTU 1 211 .09 
12. BOC - Beginning of cycle 
13. MOC - Middle of cycle 
14. STU - Short ton uranium 
15. EOC - End of cycle 
16. MTU - Metric ton uranium 

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5.4.3 Material Specifications 
Table 12 describes the hardware materials used in this analysis. The atomic weights and isotopic 
abundances required to specify the fuel are given in Table 13. Table 14 povides n list of 
impurities that are included in the fuel composition. The compositions of the fuel rods and the 
GDRs are calculated in Sections 5.5.1.2 and 5.5.2.1. 
Table 12. Chemical Compositions of Materials Used in SAS2H Calculations 
Material Element Symbol wt% 
Oxvaen 0 0.12 
Chromium I ~r I 0.1 
Iron I Fe 1 0.2 
Nickel Ni 0.08 
Tin Sn 1.7 
Zirconium Zr 97.8 
1 Density 1 6.56 g/cm3 
I oxwen 1 0 1 0.12 
Chromium I ~r I 0.1 
I Iron I Fe 1 0.2 
Tin I sn 1 1.7 
Zirconium I ~r 1 97.88 I 
Density 1 6.56 g/cm3 
Nickel I Ni ( 70.23 
Chromium I ~r 1 15 
Iron I Fe 18 
Niobium I Nb 11 
Titanium Ti 2.4 
Aluminum A1 0.7 
lnconel X-750 
Cobalt Co 1 
Manganese I Mn 1 0.8 
Silicon I si 1 0.4 
Copper I Cu 1 0.4 
Carbon I c I 0.06 
I sulfur I 0.01 
I Carbon I c I 0.08 
- - 
Manganese I Mn 12 
Phosphorus 1 P 1 0.045 
Sulfur I s 1 0.03 
Silicon Si 0.75 
Chromium Cr 19 
Nickel Ni 10.42 
cobalt" Co 0.08 
Nitrogen I N I 0.1 
Iron I Fe 1 67.495 
Reference 
Ref. 7.8, Table 2 
Ref. 7.8, Table 2 
- 
Ref. 7.20, p. 11 
Ref. 7.32, SEC IIA, SA -240, T 
17. This cobalt impurity is not taken from the reference, but is added for this calculation. 

Title: BWR Source Term Generation and Evaluation 
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Table 12. Chemical Compositions of Materials Used in SAS2H Calculations (Continued) 
-- 
Table 13. Atomic Weights and Abundances Used in Fuel Material ~alculations'~ 
Material 
SS 348H 
The concentrations of element impurities in fresh light water reactor fuel, in parts of element per 
million parts of heavy metal (ppm), are provided in Table 14. These values are provided by Ref. 
7.7, Table 5.4, and reflect actual measured concentrations instead of the maximum allowable I 
concentrations given in purity specifications. 
18. The abundance percents for Gd used in this calculation are taken from Ref. 7.23, pp. 36-37 
Element 
Carbon 
Manganese 
Silicon 
Chromium 
Nickel 
Phosphorus 
Sulfur 
Cobalt 
Niobium 
Tantalum 
Iron 
Symbol 
C 
Mn 
Si 
Cr 
Ni 
P 
S 
Co 
Nb 
Ta 
Fe 
wt% 
0.07 
2 
1 
18 
13 
0.045 
0.03 
0.2 
1 
0.1 
64.555 
Reference 
Ref. 7.9, Table 1 

Title: BWR Source Term Generation and Evaluation 
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Table 14. Nonactinide Composition of UOz 
5.4.4 Crud Source Inputs 
In addition to the source terms for an assembly based on the actinides, fission products, and 
activated light elements, it is also necessary to estimate the source due to the activated corrosion 
products from the coolant deposited on the surfaces of the assembly (crud). These surfaces 
include all the areas of the assembly exposed to the flow of coolant. A bounding estimate of the 
I Element 
Li 
B 
C 
N 
F 
Na 
Mg 
Al 
Si 
P 
CI 
Ca 
Ti 
V 
Cr 
BWR assembly surface area is based on an ANF 9x9 JP-4 assembly (Assumption 3.6), which has I 
79 fuel rods and two WRs. The information for this assembly is used due to the large array size 
and fuel rod size that provides a more conservative estimate of the surface area. The fuel rods for 
this assembly have a 0.424-inch (1.077-cm) outside diameter (OD) and a length of 163.84 inches 
(416.15 cm) (Ref. 7.2, p. 2A-24). Information on the size of the WRs is not included in this 
reference, so they are conservatively estimated to have an OD equivalent to the fuel rod pitch. 
Both the inner and outer surface areas of the WRs are included in the estimation of the surface 
area of the assembly (the WR clad is assumed to have a negligible thickness to ensure 
conservatism). The inner surface area of the assembly channel is also included. The channel 
dimensions used for calculating the surface area for crud buildup are a width of 5.44 inches 
(13.8176 cm) (Ref. 7.2, p. 2A-24) and a 0.08-inch (0.2032-cm) thickness (see Table 4). I 
Three estimates are used for the radioisotope activity density used in calculating crud sources. 
These are summarized in Table 15. The first estimate is from the Nuclear Regulatory 
Commission (NRC) NUREG-1567 (Ref. 7.10, Table 9.2) and is due solely to 6 0 ~ o . The second 
estimate from Ref. 7.1 1, p. 15, is also for 6 0 ~ o and agrees with the NRC value. The third 
estimate is from Ref. 7.12, Table 2. This estimate has activities for eight radionuclides and is 
included in this calculation for information purposes only. It is recommended that the NRC 
value be used for crud activity. The reason is that the 6 0 ~ o activity from the NRC agrees with 
that from Ref. 7.11 and is nearly three times that from Ref. 7.12. Emitting two very energetic 
photons and some beta particles, 'OCO is the most dominant isotope for crud. Since other 
radionuclides in Table 15 either have shorter half-lives or emit much lower intensity radiation, 
they become insignificant by the time the SNF arrives at the potential repository. 
Concentration (ppm) 
1 .O 
1 .O 
89.4 
25.0 
10.7 
15.0 
2.0 
16.7 
12.1 
35.0 
5.3 
2.0 
1 .O 
3.0 
4.0 
Element 
Mn 
Fe 
CO 
Ni 
Cu 
Zn 
Mo 
Ag 
Cd 
In 
Sn 
W 
Pb 
Bi 
Concentration (ppm) 
1.7 
18.0 
1 .O 
24.0 
1 .O 
40.3 
10.0 
0.1 
25.0 
2.0 
4.0 
2.0 
1 .O 
0.4 

Title: BWR Source Term Generation and Evaluation 
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Table 15. Radioisotope Activity Densities and Half Lives Used in Crud Source Calculations 
5.5 CALCULATION 
5.5.1 Parameter Study 
Calculations are done to compare the possible combinations of WRs and GDRs in order to 
quantify the differences between the assembly types. For the purpose of determining a 
conservative configuration, one node of the active fuel region is modeled, and assembly hardware 
other than the WR and the channel is not included. The 44groupndf5 cross section library is used. 
The 'skipcellwt' and 'skipshipdata' shielding calculation options are skipped because the 
shielding sequence of SAS2H is unqualified. The latticecell option is used. 
5.5.1.1 Path A and B Descriptions 
SAS2H requires two geometric descriptions to be input. The path A description represents the 
fuel and moderator cell, and is used to generate cross sections that represent the fuel, clad, and 
moderator. The dimensions used for the path A geometry are shown in Table 16. 
Table 16. SAS2H PATH A Description for Parameter Study 
I Mclad 1 2 
Variable Name 
Lattice type 
Pitch 
Fuel OD 
Mfuel 
Nmod 
Clad OD 
The second geometry description required is known as the path B and represents the entire 
assembly. For this, parts of the assembly are represented with equivalent area concentric circles. 
Because there are several possibilities for the number of WR, GDRs, and moderator densities for 
Value Used 
Square pitch 
1.6256 cm 
1.0643 cm 
1 
3 
1.2268 cm 

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each assembly, comparison cases are examined in order to determine a conservative combination 
to use for the source term generation. The combinations examined in this calculation are taken 
from Ref. 7.17, pp. 50-55. The path B descriptions for these combinations are listed in Table 1.7. 
Since these descriptions are used solely internally to this calculation to determine a conservative 
configuration, the information taken from Ref. 7.17, can be used as "reference only." 
Table 17. SAS2H PATH B Descriptions for Parameter studylg 
Variable 
Name 
Mixes 
Radius, cm 
Value Used 
Mixes 
Radius, cm 
7 
1.06426 
Mixes 
Radius. cm 
0 GDR% 
2 WRs 
4 
1.2268 
Mixes 
Radius, cm 
These path B geometries are also shown in Figures 2 and 3. In addition to the above variations, 
Path B #1 Path B #2 Path B #4 Path B #6 Path B #8 1 7 GDRs, 
Rs 
7 - 
1.22805 
3 
1.29704 
Mixes 
Radius, cm 
Mixes 
Radius, cm 
severai moderator densities are examined. ~ h e s e are summarized in Table 18. 
0 GDRs, 
WRs 
4 
1.30935 
500 
7.3371 6 
Table 18. Moderator Densities Used for Parameter Study 
7 GDRs, 
4 WRs comment 
WR 
3 
1.83429 
- -- 
4 
7.57331 
7 
8.59824 
Density Description 1 
W R clad 
500 
7.3371 6 
10 GDRs, 
4 WRs 
In-channel moder. 
4 
7.57331 
7 
8.59824 
13 I Low density (similar to the outlet) - 0.2 g/cm3 I 
Comment 
6 
0.5321 3 
2 
0.61340 
Homogenized f u ~ 
1 
2 
6 
0.53213 
6 
3 
0.91715 
Channel 
Bypass moderato 
19. Information is taken from Ref. 7.17, pp. 50-55. 
GDR 
2 
0.61340 
500 
2.72951 
Determined from core follow data from Ref. 7.13, for node 5. These cases use multiple cycles that vary the moderator 
see Table 20 for this calculation.) 
Full reference density - 0.7396 g/cm3 (Ref. 7.13, Table 2-2, p. 5 or Ref. 7.19, Table 2-2, p.5) 
4 
0.5321 3 1 
3 
0.91715 
4 
2.82015 
7 
3.24983 
I 
I 
Average density for all nodes and cycles (see Table 37) - 0.4435 g/cm3 
2 
0.613340 
500 
2.6851 3 
Cladding 
3 
0.91715 
4 
2.77721 
7 
3.24983 
In-cell moderator 1 500 
2.24654 
Homogenized fuel 
4 
2.32358 
7 
2.71 900 
Channel material 
Bypass, in-channe 

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The path B descriptions for the assemblies containing GDRs use the scheme shown in Figure 2. 
The assemblies with WRs only are described by the scheme in Figure 3. 
Channel material cell 
Homogenized fuel and 
moderator cell, R, 
/ 
GDR, R1 
Bypass, in-channel, and WR 
a moderator cell, R6 
In-cell moderator, R3 
Cladding, R2 
Figure 2. Path B for SAS2H Calculations, Descriptions 4, 6, and 8 
Channel material cell, R5 
Homogenized fuel and 
moderator cell, R, 
Bypass, in-channel, WR 
moderator cell, R, 
In-cell moderator, Rg 
' WR 
cladding, R2 
Figure 3. Path B for SAS2H Calculations, Descriptions 1 and 2 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Page: 25 of 58 - 
5.5.1.2 Fuel Material Specifications 
This calculation requires the description of six materials: regular fuel (with no gadolinium), 
GDRs, cladding, in-channel moderator, bypass moderator, and channel material. The equations 
used to determine the isotopic wt%s of these materials are shown below. The regular fuel rods 
are specified with the standard composition U02; only the breakdown of the uranium by isotope 
is required as input (see Equations 1 through 3). For the GDRs, the wt%s of all the isotopes are 
specified. This is done by first using the information calculated with Equations 1 through 3 (Ref. 
7.34, p. 20), and then determining the molecular weights of gadolinium oxide (Gd2O3) and U02 
with Equation 6 (by using the values determined with Equations 4 and 5). Equations 7 and 8 
determine the weight fractions of uranium, gadolinium, and oxygen. From these weight 
fractions, the isotopic wt% of the required isotopes of uranium and gadolinium are calculated 
with Equations 9 - 1 1. 
(1 1 molecular weight)urmium = 0.01 * 
atomic weight 
wt% 
(1 1 molecular weight) gadO'ini" 
atomic weight 
(Eq. 5) 
- (molecular weight )u02 or Gd20, - 
(# of U or Gd atoms) * (molecular weight) + (# of 0 atoms) * (atomic weight ),,,,,, 
(Eq. 6) 
molecular weight, ,, 
(weight fraction), or ,, ,,, = (# U or Gd atoms) * 
molecular weightm2 Gd203 
(weight fracti~n)~,,,~ = 1 - (weight fraction), ,, (Eq. 8) 
(wt% ~ d ) = enrichment GDR * (wt% Gd)i * (weight fraction) gdo,inium (Eq. 9) 
(wt% u)? = (1 - enrichmentGDR) * (wt% UIi * (weight fraction), (Eq. 10) 

Title: BWR Source Term Generation and Evaluation 
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These calculations are shown in Attachment I, Worksheet 'parameter study.' The atomic weights 
and isotopic abundances used in this calculation are summarized in Table 13. Node 5 of 
assembly C3 (Ref. 7.13, Tables 3-3,3-4, and 3-6 or Ref. 7.19, Tables 3-3,3-4, and 3-6) is used in 
the study of different combinations of GDRs and WRs. For this node, the fuel is 3.2 wt% 
initially enriched in 2 3 5 ~ . The GDRs are 3.0 wt% enriched with gadolinium. Table 19 shows the 
I 
weight percentages of the isotopes as they are described in the SAS2H files. 
Table 19. Weight Percentages of Isotopes for Materials Defined in SAS2H for the Parameter 
Study 
SAS2H Material 
isotope lsotope 
Fuel GDR In-channel Bypass Identifier 'Iadding Channel Moderator Moderator 
0 8016 11.8536 11.8952 0.12 0.12 0.8888 0.8888 
I Density (g/~rn~)~' 1 9.9695 1 9.9695 1 6.56 1 6.56 1 Varied with cyc 0.7396 I I 
Because SAS2H does not permit the modeling of the gas gap between the fuel and the cladding 
for a central GDR, a smeared fuel density is used (the same approximation is made for the 
regular fuel rods as well). This has a negligible effect on the calculation, and is a common 
approximation. The smeared density is calculated from Equation 12: 
20. See Eq. 12. 

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pellet radiud 
p = (% theoreticd density* pthe ( clad inside radiud 
Yielding: 
5.5.1.3 Fuel Temperatures and Moderator Densities 
The temperature of the cladding and channel is taken to be that of the cladding described in Table 
6. This temperature is in the sample calculations for SAS2H in the SCALE manual. The fuel 
temperature used for all parameter study cases is a cycle weighted average (see Table 20), and is 
determined from the operating information presented in Tables 8 and 11. The EFPD used as a 
weighting factor corresponds to the MOC. For BOC data points, the difference between the 
EFPD given in Table 11 and the EFPD shown in Table 20 is simply the previous MOC value 
subtracted from the BOC value. 
It is also necessary to calculate a cycle weighted moderator density (referred to as 'density 4' in 
Table 18), as well as moderator densities for each cycle (described as the 'density 1' case in Table 
18). For the 'density 1' case, MOC values are used if possible; for cycle 13 an average is taken 
over the MOC values. 
Table 20. Average Fuel Temperature and Moderator Densities for Parameter stud#' 
DP 
4 
5 
6 
7 
8 
9 
10 
11 
12 
13 
(EFPD, * (Temperature or density) , ) 
dota point weighted average = 
Total EFPD 
Total I 11540.11 
Straight average 
Weighted average over all cycles 
21. The fuel temperature and moderator density information in Table 20 correspond to node 5 of the C3 assembly of 
the Quad Cities reactor (Ref. 7.13, Table 4-141, pp. 333-334). 
22. See Table 18. I 
EFPD from 
Table 11 
(days) 
222.47 
467.48 
180.3 
484.2 
142.2 
263.7 
10.1 
112.94 
224.4 - - 
324.73 
1 100.4 
11 29.2 
EFPD 
(days) 
222.47 
245.01 
180.3 
303.9 
142.2 
121.5 
10.1 
102.84 
11 1.46 
100.33 
0.4785 
0.4727 
Fuel Temperature 
(K) 
121 1.5 
1142.8 
1276.3 
1 133.7 
1059.2 
1006.2 
1011.9 
101 8.5 
1078.7 
1065.3- -- 
Moderator Density (g/cm3) 
0.4559 
0.4610 
0.4587 
0.4700 
0.4776 
0.4852 
- 
0.4860 
0.4935 
0.4975 
0.4993 
Moderator Densities for Each 
cycle ( g / ~ m ~ ) ~ ~ 
0.4559 
0.4587 
0.4776 
0.4964 (average over the values for tt 
-I 

Title: BWR Source Term Generation and Evaluation 
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The moderator temperatures are determined from the operating pressure given in Table 6 and the 
steam table information given in Table 7. The calculation is shown in Table 21. 
Table 21. Moderator Temperature Calculation 
Temperature (K) Pressure (psia) 
553.15 931.01 17 
563.15 1079.7915 
ATemperature 
moderator temperture = temperature, + * (operating pressure - pressure, ) 
A Pr essure 
moderator temperature = 553.15 + ( 563*15-553'15 
*(1020-931.0117) 
1079.7915 - 931.01 17 
moderator temperature = 559.13 K 
5.5.1.4 Overall Assembly Input 
As previously stated, for the purposes of this parameter study, one node is examined. The 
selection of this node is arbitrary, and serves only to use uniform values for all the calculations. 
Node 5 information is used. The overall assembly information for each geometry examined is 
given in Table 22. 
Table 22. SAS2H Assembly Input for Parameter Study 
I Variable Name I Path B #I I Path B #I2 I Path B #4 1 Path B #6 I Path B #8 I 
Npinlassm 62 60 62 60 60 
Fuelngth 30.48 30.48 30.48 30.48 30.48 
Ncycles 4 4 4 4 4 
Nliblcyc 3 3 3 3 3 
Printlevel 5 5 5 5 5 
lnplevel 2 2 2 2 2 
Asmpitch 15.24 15.24 15.24 15.24 15.24 
Numxtotal 6 6 6 6 6 
Mxrepeats 1 1 1 1 1 
Mixmod 3 3 3 3 3 
, Facmesh . 1 .O 1 .O . 1.0 . 1.0 . 1.0 
The parameter study cases use a similar irradiation history as node 5 of assembly C3 of the Quad 
Cities reactor. The choice of this assembly is arbitrary, but is chosen because this assembly 
reached a high burnup. The burnup information for this node is given below in Table 23. The 
thermal-hydraulic information for this node is detailed in Tables 18,20, and 21. 

Table 23. Burnup Information for Parameter Study 
Uranium mass in node: 14.144 kg 
BOC 11 BOC 12 BOC 13 EOC 13 
Cumulative Node Burnup (GW~/MTU)~~ 15.657 32.173 38.990 49.301 
Burnup for Cycle (MWdMTU) 15657 16516 681 7 10311 
EFPD for Cycle 467.48 484.2 263.7 387.07 
Nodal Power (MW) 0.4737 0.4825 0.3656 0.3768 
5.5.1.5 Parameter Study Results 
As noted in Ref. 7.21, p. 119, it is important to "obtain a good estimate of the time-averaged 
moderator density at an axial location to best calculate spent fuel composition." Ref. 7.21 makes 
the following conclusions: 
Sensitivity to moderator changes appears to be close to linear for most of the nuclides. 
The actinides are most sensitive to changes in moderator density. Decreased moderator 
density results in increased quantities of actinides, and vice versa. The density decrease 
results in spectral hardening which in turn results in increased Pu production, and decreased 
2 3 5 ~ depletion due to increased Pu fission. 
Of the fission products, only '49~m, 151sm, and ' 3 5 ~ s are sensitive to moderator density, 
increasing quantities with decreasing moderator density, as the spectral hardening increases 
Pu fission, which produces the Sm nuclides more rapidly than U fission. 
This last conclusion is demonstrated in the results of the SAS2H calculations where the grams of 
the mentioned isotopes are printed. This is shown in Table 24. 
Table 24. Parameter Study Comparison of Fission Product Masses 
Grams 
Path B ensi it^*^ Isotope 
1 day 100,000 10 years 100 years 1000 years 10,000 years ears 
Sm14' 0.021 7 0.0339 0.0339 0.0339 0.0339 0.0339 
2 Sml" 6.19 6.1 9 6.19 6.19 6.19 6.19 
C S ' ~ ~ 6.26 6.27 6.27 6.26 6.25 6.08 
23. From Table 4-141, pp. 333-334 of Ref. 7.13 (essentially identical to Table 4-141, pp. 333-334 of Ref. 7.19). 1 
24. As noted in Table 18. Density 2 is the full density, while density 3 refers to an outlet (low) density. 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00~-MGRO-00200-000-OOA Page: 30 of 58 1 
These conclusions are also shown in the results of the sample cases in this calculation, with the 
lower outlet m.oderator density (at the top of the core) generating the highest source. 
Tables 25 - 28 present the total gamma and neutron sources for the five BWR path B descriptions 
with the moderator densities described in Table 18 and at different time steps. A normalization 
of radiation sources to that obtained with density 1 of the moderator is provided in these tables to 
address the influence of moderator density on the source terms provided by the different BWR 
assemblies (or path B descriptions). There is little variation between the geometric models for a 
given moderator density as is seen from the normalized results. The information in these tables 
is calculated in Attachment V, Worksheet 'total.report(density)'. 
Table 25. Comparison of Total Neutron and Gamma Sources for Different BWR Assemblies, 
Moderator Density 1 

Title: BWR Source Term Generation and Evaluation 
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Table 26. Comparison of Total Neutron and Gamma Sources for Different BWR Assemblies, 
Moderator Density 2 
I Total Gamma Sources Over Time 
Path B 
Time (years) 1 2 4 6 8 
I Time (years) I Percentage of Moderator Density 1 Results for Gammas 
Total Neutron Sources Over Time 
1 2 4 6 
2.27E+07 2.41 E+07 2.34E+07 2.46E+07 
1.1 1 E+07 1.22E+07 1 .15E+07 1.25E+07 
6.01 E+05 6.47E+05 6.19E+05 6.63E+05 
I I I 
Percentage of Moderator Density 1 Results for Neutrons 

Title: BWR Source Term Generation and Evaluation 
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Table 27. Comparison of Total Neutron and Gamma Sources for Different BWR Assemblies, 
Moderator Density 3 
Total Gamma Sources Over Time J 
Path B 
1 2 4 6 8 
Time (years) 
Percentage of Moderator Density 1 Results for Gammas 
I Total Neutron Sources Over Time I 
path B 
1 2 4 6 8 
Time (years) 
0.00274 2.97E+07 3.15E+07 3.23E+07 3.29E+07 3.35E+07 
10 1.49E+07 1.61 E+07 1.62E+07 1.69E+07 1.72E+07 
100 8.17E+05 8.76E+05 8.96E+05 9.25E+05 9.42E+05 
Time (years) Percentage of Moderator Density 1 Results for Neutrons 
0.00274 116% 116% 120% 118% 118% 
10 118% 1118% 11 22% 119% 120% 

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Table 28. Comparison of Total Neutron and Gamma Sources for Different BWR Assemblies, 
Moderator Density 4 
I Total Gamma Sources Over Time I 
1 1 2 1 4 1 6 1 8 1 Time bears) 
Time (years) 
100% 
100% 
100% 
1000 102% 
Percentaae of Moderator Density 1 Results for Gammas 
I Total Neutron Sources Over Time I 
Path B 
1 2 4 6 8 
Time (years) - 
0.00274 2.60E+07 2.75E+07 2.74E+07 2.84E+07 2.88E+07 
10 1.29E+07 1.40E+07 1.36E+07 1.45E+07 1.47E+07 
100 6.99E+05 7.50E+05 7.39E+05 7.77E+05 7.87E+05 
1000 2.06E+05 2.23E+05 2.1 8E+05 2.32E+05 2.35E+05 - 
10000 7.93E+04 8.48E+04 8.29E+04 8.75E+04 8.85E+04 - 
100000 2.39E+04 2.51 E+04 2.44E+04 2.55E+04 2.56E+04 
Time hears) Percentage of Moderator Density 1 Results for Neutrons 

Title: BWR Source Term Generation and Evaluation 
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As shown in Table 29, the results from shielding calculations (Ref. 7.22, p. 19) for the dose on 
the outside radial surface of a waste package indicate the neutron dose dominates on the 300 - 
1000+ year time frame. At these times it represents 79 - 98% of the dose. At times greater than 
30,000 years, the neutron dose on the surface of the waste package represents 5 - 6% of the total. 
Table 29. Percentage of Surface Dose Due to Neutrons for the 21 PWR Waste 
5 10 50 100 300 1000 10,000 30,000 Region of the Waste Package Surface years years 
Top section of the active fuel region 5% 8% 1 1 ( 9 % 1 79% 1 96% 1 18% 1 5% 
Middle section of the active fuel reaion I 6% 1 9./0 1 12% 1 10% 1 80% 1 97% 1 20% I 6% - I I I I 
Bottom section of the active fuel region 1 5% 1 8% 111./0 19Y0 1 79% 1 97% 1 18% 1 5% 
Figure 4 and Figure 5 (generated in Attachment V, Worksheet 'Den4.%ofmodl.g') show the 
neutron and gamma sources over time for the five path B descriptions (at density 4). The results 
I 
are shown relative to path B #l. 
Figure 4. Percentage Results for the Gamma Sources for the Various Path B Descriptions 
25. The waste package referred to has an Alloy 22 inner barrier and a 5 16 carbon steel outer barrier. 

Title: BWR Source Term Generation and Evaluation 
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As seen in the Figure 4, all the assemblies are close to each other at early times, with those 
containing the GDRs becoming less conservative in the 300 - 1000+ years time frame. However, 
as time progresses, the GDR bearing assemblies again become more conservative. In all cases, 
the results of all the assemblies are within a few percent of each other. The graph of the neutron 
results (Figure 5; generated in Attachment V, Worksheet 'den4.%modln') is much more straight 
forward, with path B #8 yielding more conservative results, sometimes by as much as 14%. For 
these reasons, path B #8 is used in the subsequent calculations. 
.001 0 01 0 1 1 10 100 1000 10000 100000 
Time (years) 
Figure 5. Percentage Results for the Neutron Sources for the Various Path B Descriptions 
5.5.2 One- and Ten-Node Calculations 
5.5.2.1 Material Specifications 
The thermal DBF calculations use a fuel density determined from the IHML (200 or 170 kg) and 
the total fuel volume. This results in a density of 11.64 g/cm3 for the 200 kg loading and 9.89 
g/cm3 for the 170 kg loading. These calculations are shown in Attachment I, worksheet 'density'. 
The lower density yields a higher per MTU normalized source for both gammas and neutrons, as 
shown in Section 6.2. Based on these results, the source terms for the waste stream calculations 

Title: BWR Source Term Generation and Evaluation 
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of this calculation use a fuel length that has been increased to account for any additional loading 
desired, with the fuel density kept at 9.9695 &m3 for all cases (see Attachment I, Worksheet 
'length calculation'). As done in Section 5.5.1.2, the fuel specifications are determined from 
Equations 1 - 11. The calculations are performed in Attachment I, in Worksheets '5.5' through 
'natural.' The results are summarized in Table 30. 
Table 30. Weight Percents for Fuel Rods and GDRs for All Enrichments 
Weight % of All Isotopes in GDRs I 
The next parameters calculated are the light element masses to be irradiated by SAS2H, 
representing the assembly hardware. The first step is to determine the channel mass using 
Equation 15 and the information from Tables 4 and 5. 
Channel mass = channel height * 
(channel outside width ' - channel inside width ')* channel density (Eq. 15) 
= 38.597 kg 
The calculated channel mass is higher than any of the available reference values, and is therefore 
used. This is done to include as much tin, a gamma source for shorter cooling times, in the 
model as possible. Using this value, and the information in Table 4, the mass of channel in each 
region is calculated, as shown in Table 3 1, along with the mass of WR in the fuel and plenum 
regions. These calculations are shown in Attachment II, Worksheet 'compositions', cells F9, H9, 
19, N9, P9, and J9. 
'57~d 
tssGd 
rsoGd 
Weight % of Uranium Isotopes in Fuel Rods 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 
0.4073 
0.6465 
0.5690 

Title: BWR Source Term Generation and Evaluation 
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Table 31. Calculation of WR and Channel Mass in Different Assembly Regions 
Description (Eq. 16) Result: Mass (k$ 
length in top region 
Mass of channel in top end fittinc mass = * channel mass 0.9949 
channel length - I i length in bottom region 
Mass of channel in bottom end fi mass = I * channel mass 0.6512 
channel length 
length in plenum region 
Mass of channel in plenum regio * channel mass 2.6754 
channel length 
length in fuel region 
Mass of channel in fuel region mass = * channel mass 34.276 
channel length - 
( length in fuel region 
Mass of WR in fuel region26 maSS = * WR mass 3.1 099 1 water rod length - 
length in plenum 
Mass of water rod in plenum reg mass= *WR mass 0.2428 
water rod length 
The masses of the hardware shown in Tables 5 and 31 are multiplied by the weight fractions of 
the constituent elements that compose the material. The calculation for the thermal DBF is 
performed in Attachment 11, Worksheet 'compositions'. The light element calculations for the 
waste stream and the SS clad assemblies are performed in Worksheets '8 1 1 .compositions' and 
'81 1 .ssclad.compositions,' respectively. The resultant elemental mass is then multiplied by the 
appropriate scaling factor for a GE assembly (Ref. 7.6, Table S.1, p. vi) to account for the 
hardware's location in the assembly. These results are shown in Tables 32 and 33. The 
calculations are performed in Attachment 11, Worksheet 'Table 5.3.2.1-2' for the thermal DBF, 
'8 11 .lite.el.tablesl for the waste stream, and '8 11 .ssclad.lite.el.tables' for the stainless steel clad 
assemblies, using results taken from the '*compositions1 Worksheets mentioned earlier. 
26. The fractions of WR in the fuel and plenum regions are also used to describe the fraction of cladding in the fuel 
and plenum regions. 

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Table 32. Light Element Masses Used for the Thermal DBF and One-Node Comparison case2' 
Top End Fitting Region 
Element Masses (kg) 
I DBF I Case 
0 1 0.0002 1 0.0002 
Plenum Region 
Masses (kg) 
SF: 0.300 
0.0036 0.0036 
Fuel Region Masses Bottom End Fitting 
(kg) Region Masses (kg) 
SF: 1,000 
- 
SF: 0.225 
Thermal One-Node Thermal One-Node 
DBF Case DBF Case 
27. These are the light element masses calculated in Ref. 7.4, Table 5.3.2.1-1.2. They do not contain any fuel 
impurities or extra cobalt impurities mentioned in Section 5.1. The one-node case refered to is that is used to 
I 
compare against the ten-node case. This calculation is also unchanged from Ref. 7.4. I 

Title: BWR Source Term Generation and Evaluation 
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Table 33. Light Element Masses Used for the Waste Stream and SS ~ssemblies~~ 
As noted previously, there are seven spacer grids in the core. To evaluate the effect of these 
grids on the peaking factor determined in thb calculation, the spacer grids are accounted for in 
the ten-node representation. This is done by determining which nodes the grids fall into (from 
the locations given in Ref. 7.14, p. A-8 and the node heights given in Table lo), and irradiating 
the material via the light element option in SAS2H, along with the channel and cladding 
material. The node fraction information is calculated in Attachment II, Worksheet 'ten node', 
cells I1 - S6, and is shown in Table 34. The fuel region light elements are shown on this 
worksheet in cells A4 - H25, with the final light element calculations performed in cells A26 - 
S120. The results are summarized in Table 35. 
28. These light elements contain the impurities for the fuel and hardware mentioned in Section 5.1. They are used 
for the final source term calculations. I 

Title: BWR Source Term Generation and Evaluation 
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Table 34. Calculation of Node Fraction of Whole Active Fuel Height 
Node: 1 1 1 2 
Height (inches) l 6 118 
Percentage of Spacer ~ r ( 0 1 69.231 
Percentage of Channel a 4.14 8.28 
Table 35. Light Element Masses for Ten-Node Representation of the Active Fuel Region 
5.5.2.2 Fuel Temperatures and Moderator Densities 
Weighted moderator densities and fuel temperatures are used for all one-node calculations. The 
weighted averages are calculated with Equation 13 as was done in Section 5.5.1.3. These 
calculations are shown in Attachment III, Worksheet 'C3 Average Values', cells Dl5 - W15, and 
A29 - F42. The results are summarized in Tables 36 and 37. 

Title: BWR Source Term Generation and Evaluation 
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Table 36. Average Values of Thermal Hydraulic Parameters for Thermal DBF and One-Node 
Comparison Case 
DP E ~ ~ b ~ f r o m Table 11 EFPD (days) Moderator Density (g/cm3) Fuel Temperature (K) 
(days) 
4 222.47 222.47 0.4403 1059.0 
5 467.48 245.01 0.4351 1073.3 
6 180.3 180.3 0.4333 1061.9 
7 484.2 303.9 0.4402 1071.3 
8 142.2 142.2 0.4450 995.0 
9 263.7 121.5 0.4506 967.0 
10 10.1 10.1 0.4510 910.2 
11 1 12.94 102.84 0.4556 91 6.7 
12 224.4 1'1 1.46 0.4580 949.1 
13 324.73 100.33 0.4593 970.5 
Total EFPD (days) Weighted Average (g/cm3) Weighted Average (K) 
1540.1 1 0.4435 1026.7 
Table 37. Average Values of Thermal Hydraulic yuameters for Waste Stream and SS Clad 
Assemblies 
DP EFPD from Table 11 EFPD (days. ldavc\ I Moderator Density (g/cm3)30 I Fuel Temperature (KIT 
\--I-/ 
4 222.47 222.47 0.4403 91 6.9 
5 467.48 245.01 0.4351 920.3 
6 180.3 180.3 0.4333 91 5.8 
7 484.2 303.9 0.4402 91 8.2 
8 142.2 142.2 0.4450 852.8 
9 263.7 121.5 0.4506 829.9 
10 10.1 10.1 0.451 0 786.9 
11 1 12.94 102.84 0.4556 791.9 
12 224.4 111.46 0.4580 81 8.2 
13 324.73 100.33 0.4593 834.7 
Total EFPD (days) Weighted Average (g/cm3) Weighted Average (K) 
1540.1 1 0.4435 883.1 
For the ten-node representation, cycle averaged values are used for the nodal fuel temperatures, 
and the moderator density is adjusted through the H20 fraction option in SAS2H. The cycle 
average fuel temperatures are determined from Equation 13, except that the formula is applied to 
each node, rather than a weighted average over the nodes. These calculations are shown in 
Attachment III, Worksheet 'C3 Average Values', cells A16 - N27, and are summarized in Table 
38. 
29. These values are based on the updated fuel temperatures from Ref. 7.19. The calculation is shown in Worksheet 
'8 12-c3average values.' 
30. The densities and temperatures shown in this table are weighted averages over node height. 

Title: BWR Source Term Generation and Evaluation 
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Table 38. Values of Thermal Hydraulic Parameters for Ten-Node ~alculation~' 
5.5.2.3 Assembly Parameters 
For each enrichment listed in Table 2, source terms for all burnups listed are calculated (with the 
exceptions previously noted). These combinations provide a sampling of the BWR waste stream 
for later analysis. Each combination of enrichment and burnup is then decayed with the time 
steps shown in Section 5.5.2.4. The average assembly power is determined in Table 39 by 
dividing the total MWth of the reactor by the number of assemblies in the core. The relation for 
determining the EFPD for the selected fuel burnups is shown by Equation 17: 
EFPD * Assembly power (MW 1 assembly) 
Bumup (MWd I MTU) = 
Uranium Loading (MTU 1 assembly) 0%. 17) 
Table 39. SAS2H Input for the Assembly DepletionIDecay Parameters 
Power = 3.468 MW IHML = 200 
Due to the error that can be incurred by using too large of a cycle time, for the one-node 
representations of the assemblies, the EFPD days are broken down into steps of 200 days per 
cycle or less. Therefore, all cycle lengths are 200 EFPD with the exception of the final cycle, 
which is listed in Table 39. 
3 1. These values are based on the same fuel temperatures used for the thermal DBF and the one-node comparison 
case- those from Ref. 7.13, Table 4-141, pp. 333-334. I 

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For the ten-node representation of the assembly, the EFPD and burnup data for assembly C3 are 
used to determine the burn history (Attachment IV, Worksheets 'burnup, passl' and 'burnup, 
pass2', columns G, J, M, P, S, V, Y, AB, AE, and AH, rows 6 - 16). From this data, the power 
generated in each node is determined (columns E, H, K, N, Q, T, W, Z, AC, and AF, rows 6 - 
16). These powers are then used to calculate the desired burnups (columns F, I, L, 0, R, U, X, 
AA, AD, and AG, rows 6 - 16). For burnups greater than 30 GWdJMTU, the life of assembly C3 
is repeated until the desired burnup is reached (the Worksheet 'burnup, passl' is used for the 10 - 
30 GWd/MTU calculations, and Worksheet 'burnup, pass2' is for the higher burnups). The final 
results are shown in Tables 40 and 41. Due to the difficulty in changing the fuel temperatures 
from cycle to cycle, the cycle averaged values are used, and maintained throughout the 
calculations. 
Table 40. Ten-Node Burn Histories for 10.20, and 30 GWdIMTU 
As previously stated, after one full lifetime, the cycles are repeated until the desired bumup is 
reached. For these extra cycles, the power at which the nodes are burned is the same as for the 
previous cycles, but the length of the bum is adjusted to reach the desired assembly averaged 
burnup. The nodal burnup is shown in the following table along with the EFPD for the 
respective cycle. 

Title: BWR Source Term Generation and Evaluation 
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Table 41. Ten-Node Burn Histories for 40 - 75 GWdMTU 
Table 42 summarizes the remaining information required by SAS2H. 
Table 42. SAS2H Assembly Input 
Value for Thermal Value Used for Ten-Node Value used for Waste 
Variable Name DBF and One-Node Calculation Stream and SS Clad 
Comparison Case Calculations 
Npinlassm 60 60 60 
Fuelngth 368.91 Node height 427.1 13' 
Ncycles 1 cycle 10 cycles 1 cycle 
N l i blcyc 16 libraries per cycle 3 libraries per cycle 16 libraries per cycle 
Lightel 17 17 33 
Printlevel 5 5 5 
lnplevel 2 2 2 
Asmpitch 15.24 cm 15.24 cm 15.24 cm 
Numxtotal 6 6 6 
Mxrepeats 1 1 1 
Mixmod 3 3 3 
Facmesh 1 .O 1 .O 1 .O 
5.5.2.4 Time Steps for ORIGEN-S Decay Calculations 
The ten-node representation is decayed to the following times: 5,6,7,8,9, 10, 11, 15, and 20 
years after discharge.33 Table 43 lists the time steps used for the thermal DBF cases. The time 
steps chosen are based on the time the assembly is expected to reach a total thermal output of 400 
watts, or the time when the assembly can be loaded into the waste package and are shown in 
Table 43. I 
32. Due to the results of the density comparison in Ref. 7.4 of this calculation, the fuel length is increased rather I 
then the fuel density to account for an increased IHML. The calculation is performed in Attachment I, Worksheet 
'length calculation'. 
33. Other time steps were included in the ORIGEN-S case for the ten-node representation, but are unusable due to 
an error in the input. 

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Table 43. Time Steps Used for Decay Calculation of Thermal DBF (years) 
1 day (5 16 17 8 
11.01 (1 1.02 11 1.03 (1 1.04 111.05 1 1.06 
11.15 111.2 111.25 111.3 111.35 11.4 
For the waste stream and SS clad assembly calculations, the determination of the assembly 
loading time is left for subsequent analysis. In order to provide adequate data for future use, 180 
time steps are used in the ORIGEN-S input files. These time steps are shown in Table 44. 
Table 44. Time Steps Used for Decay Calculation of Waste Stream and SS Clad Assemblies 
(years) 
5.5.3 Radionuclide Inventories for Performance Assessment 
Several separate SAS2WORIGEN-S cases are provided in this calculation to determine average 
and bounding radionuclide inventories for specific years. The average and bounding BWR 
assemblies are derived from the results of Ref. 7.24 and listed below. The characteristics of the 
average BWR assembly are estimated based on the average BWR assembly of Case A with full 
inventory (83,800 MTU) in Table 5 of Ref. 7.24. For that case, the characteristics of the average 
BWR assembly are 3.02 wt%, 33.6 GWdlMTU, and 25.3 years old with an initial uranium 
loading of 177 kg. A comparison study, based on the data base in Ref. 7.2 (Appendix lC), 

Title: BWR Source Term Generation and Evaluation 
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indicates that per initial MTU loading the average BWR assembly selected in this calculation is 
more conservative than any average BWR assembly for the scenarios in Table 5 of Ref. 7.24 (see 
Attachment XIV). The characteristics of the bounding BWR assembly are also derived from Ref. 
7.24 (Attachment III, bin.dat files). From those files the following bounding BWR SNF 
characteristics are noted: initial uranium loading of 197 kg, burnup of 65.55 GWd/MTU, initial 
uranium enrichment of 4.28 wt%, and cooling time of 5 years. It should be noted that there is no 
single assembly in the waste stream with these combined characteristics. Rather, these are the 
bounding characteristics of each parameter in the entire waste stream. Compared to these SNF 
characteristics, the bounding BWR assembly selected here is more conservative. The 
characteristics of the average and bounding BWR SNF assemblies for this calculation are: 
Average BWR assembly: 
Maximum BWR assembly: 
3.5%, 40 GWdMTU, 25 years old 
5.0%, 75 GWd/MTU, 5 years old 
The source terms for these assemblies are generated for the years 2033,2133,2233,2333,2433, 
2533,3033,4033,5033, 12,033,22,033,32,030,102,033,302,033, and 1,002,033. The age 
mentioned above is the age at 2033. The light elements from the hardware regions are included 
in the fuel region, as it is not necessary to keep the assembly region separate to determine overall 
radionuclide inventories. The *.cut files for these cases are included in the CDs of Attachment 
VII. The radionuclide inventories are provided in Attachment Xm. 
5.5.4 Calculation of Crud Source Terms 
The activity of the crud on the surface of the BWR assemblies at time zero is determined simply 
by multiplying the calculated surface area by the corrosion product activity (given in per unit area 
of surface). The surface area is calculated in Attachment VI, and is shown in Table 45. 
Table 45. BWR Surface Area for Crud Calculation 
I I Value Units I 
The crud activity is then decayed using Equation 18. 
- t * l n 2 
Rod OD 
WR OD (rod pitch) 
# of Rods 
Rod Length 
Channel ID 
Channel OD 
Channel Length 
WR outside area + inside area (estimated from rod pitch) 
rod pitch* PI ' 4; WR ID and OD consewatively approximated as fuel rod pitch 
N ( t ) = N ( 0 ) e t l I 2 ( ~ q . 18) 
where t2 is the half-life and t is the decay time in years. The crud source terms are calculated in 
Attachment VI, and the results are shown in Section 6.4. 
(Data from ANF 9x9 JP-4,5 assembly) 
BWR Assembly Surf. Area 11 681 47.97 lcm2 
(Rod surface + channel inner surface + WR's inside and outside surfaces) 
1.07696 
1.45288 
79 
416.1536 
13.4112 
13.8176 
447.548 
81 71.07563 
cm 
(2 W Rs) 
cm 
cm 
cm 

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6 RESULTS 
This section presents the results of this calculation. The outputs of this calculation are reasonable 
compared to the inputs, and the results are suitable for the intended use. The uncertainties are 
taken into account by consistently using the most conservative approach; the calculations, 
therefore, yield a conservatively bounding set of results. Results of the parameter study are 
presented in Section 5.5.1.5. More information on these results is available in Attachment V. 
6.1 AXIAL PEAKING FACTOR 
One task of this calculation is to examine an APF to be used in future shielding calculations. The 
axial effects can be of great concern, for the reasons mentioned in Section 5.5.1.5. Figure 6 
shows a plot of APF versus burnup for the measured data from Ref. 7.13, pp. 290-544. Since the 
nodal burnup data from pp. 290-544 of Ref. 7.19 (a revision of Ref. 7.13) are essentially identical 
to the data from Ref. 7.13 (differing generally only in the 5~ significant digit), APF versus 
burnup data based on Ref. 7.19 would be identical for design purposes to that in Figure 6. 
Therefore, the conclusions about APFs are still valid. 
0.00 10.00 20.00 30.00 40.00 50.00 
Assembly Average Burnup (GWdIMTU) 1 
Figure 6. Axial Peaking Factor versus Assembly Average Burnup 
This graph (generated in Attachment III, worksheet 'APF.vs.Burnup') indicates the maximum 
value of the nodal burnup divided by the average assembly burnup. The purpose of using an 
APF in shielding calculations is to obtain a peak surface dose without having to model many 
different source regions; rather, the peak can be reached by modeling one region. In order to do 

Title: BWR Source Term Generation and Evaluation 
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this, the source used must encompass all the peak sources in the problem. When the one-node 
source is multiplied by the APF and smeared over the problem volume, it should still bound the 
smaller sources in the multi-node description of the problem. 
The main purpose of the APF calculation is to demonstrate that a one-node model of a BWR can I 
be used to generate source terms such that when used with an APF it will provide conservative or 
comparable results with a more detailed ten-node representation of the assembly. In order to 
compare the one- and ten-node representations of the assembly, total neutron and gamma sources 1 
from both are specified on a particle per second per unit height basis. The one-node value is 
multiplied by the APF, and the difference between the increased one node value and the 
maximum ten-node value is then divided by the one node value. These comparisons can be seen 
in Attachments VIII and M in their entirety. The APF used in this comparison is 1.4 and is based 
on Figure 6. The results favor the one-node representation for the gamma sources, for decay 
times greater than five years (and are within a few percent for the five year decay time), but 
indicate that the ten-node model used in this calculation is more conservative for the total 
neutron source. Attachment IX shows the comparison of the neutron sources. It is apparent from 
these results that a larger APF should be considered for any analysis where the neutron dose is 
expected to dominate. The calculations in Attachment IX indicate that an APF of 2.0 would have 
been sufficient to favor the one-node representation for neutron sources. 
6.2 INITIAL HEAVY METAL LOADING 
The results of using different MMLs are shown below for the thermal DBF. The data shown are 
for a normalized source on a per MTU basis. The calculations that generate Figures 7 and 8 are 
shown in Attachment X. 

Title: BWR Source Term Generation and Evaluation 
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3.00E+00 - - *- . 170 kg loading A 
1s I t 
-200 kg loading I I I I 
0.01 0.1 1 10 
Mean Photon Energy (MeV) 
2.50E40 -- 
f 
Figure 7. Normalized Gamma Source per MTU for Thermal DBF at 11 and 31 1 Years 
0.00 2.00 4.00 6.00 8.00 10.00 12.00 14.00 
Mean Neutron Energy (MeV) 
-6-- 170 kg loading at 31 1 yrs 
*200 kg loading at 31 1 yrs 
Figure 8. Normalized Neutron Source per MTU for Thermal DBF at 11 and 31 1 Years 
i i 
I,I I I 
I I 
I I I 

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6.3 EFFECTS OF FUEL IMPURITIES 
Source terms fbr one of the cases (waste stream assembly with 3.74 wt% 2 3 5 ~ initial enrichment) 
are generated with and without impurities (see Table 13) in the fuel. Negligible differences in 
radiation spectra and thermal powers are observed. However, the results for 3 6 ~ 1 and '"c are 
summarized in this section. Attention is given to 3 6 ~ 1 and '"c, long-lived radionuclides with 
half-lives of 3.01E+05 and 5,715 years (Ref. 7.23), respectively, because of their very high 
solubility-limit in aqueous concentrations (Ref. 7.25, p. 6 - 7). 16cl and '"c activities at 
discharge as a function of burnup are presented in Table 46 and are plotted in Figure 9. 3 6 ~ 1 and 
'"c activities vary with initial fuel enrichment, fuel burnup, and impurity content. For a given 
fuel enrichment and impurity content, the activities of these two radionuclides increase with fuel 
burnup. For a given burnup and impurity content, the activities decrease with increasing fuel 
enrichment. Bounding 3 6 ~ 1 and 14c activities of 1.09E-02 and 6.62E-01 Cilassembly, 
respectively, are obtained for an assembly with natural uranium and 75 GWdJMTU burnup at 
discharge. 
Table 46. 3 6 ~ ~ and 14c Activities for the Waste Stream Assembly with 3.74 wt% 2 3 5 ~ Enrichment 
Bumup (GWdIMTU) 
1 
10 
20 
30 
%I (Cilassembly) 
6.61 E-05 
6.74E-04 
1.44E-03 
2.33E-03 
40 
50 
60 
70 
36C (Ci/ppm/assembly) 
1.25E-05 
1.27E-04 
2.72E-04 
4.40E-04 
75 7.62E-03 1.44E-03 4.14E-01 4.63E-03 
3.38E-03 
4.57E-03 
5.82E-03 
7.03E-03 
14C (Cilassembly) 
3.03E-03 
3.1 1 E-02 
6.70E-02 
1.11E-01 
I4C (Ci/ppm/assembly) 
3.39E-05 
3.48E-04 
7.49E-04 
1.24E-03 
6.38E-04 
8.62604 
1 .lOE-03 
1.33E-03 
1.64E-01 
2.28E-01 
3.00E-01 
3.76E-01 
1.83E-03 
2.55E-03 
3.36E-03 
4.21 E-03 

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0 10 20 30 40 50 60 70 80 
Burnup (GWdIMTU) 
36 Figure 9. CI and 14c Activities per Unit Impurity per Assembly at Discharge as a Function of Burnup 
for Waste Stream Assembly with 3.74 wt% 2 3 5 ~ Enrichment 

Title: BWR Source Term Generation and Evaluation 
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6.4 CRUD SOURCES 
The results of the crud source calculations are shown in Table 47. The NRC values are 
recommended. 
6.5 WASTE STREAM AND SS CLAD SOURCE TERMS 
Due to the large amount of information generated by this calculation, the results are provided as 
electronic files on three compact discs (CDs) (Attachment VII). These results are identical to the 
output files of the previous revision of this calculation (Ref. 7.30). 

Title: BWR Source Term Generation and Evaluation 
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Table 47. Crud Sources 

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6.6 USE OF SOURCE TERMS 
During the revision and checking of this calculation, it was discovered that the actual values of 
burnup, based on the input for the SAS2WORIGEN-S runs, were about 1.7% less than the 
nominal values of burnup stated in Table 2. The SAS2WORIGEN-S runs are valid, but the 
source term information (e.g., radionuclide masses and activities; decay heat generation rates; 
and neutron and gamma sources) contained in the '*.cut' files on the compact discs of Appendix 
VII should be taken to correspond to the actual bumups from the SAS2WORIGEN-S cases, 
rather than the stated, or nominal, burnups. To use the source term information, determine the 
stated, or nominal, burnup from the first part of the '*.cut' file name as explained in Attachment 
XII. Then, associate the information from that '*.cut' file with the actual bumup in Table 48 
before using the source term information. 
Table 48. Stated and Actual Values of Burnup Associated with '*.cut' Files 
Source term information from the '*.cut' and '*.output7 files of the SAS2WORIGEN-S runs in 
the folders "Impurity-study," "parameter-study-output," "ten-node-output," and 
"THERMALRMALDBFoutput"on compact disc one of Attachment VII should not be used for 
assemblies expected to be received at a potential repository. Those cases were run solely to form 
the basis for deciding what kinds of cases to be run for BWR assemblies in an expected waste 
stream. Furthermore, the '*.cut7 files in folder "ten-node-output" and the '*.output' files in 
folder "parameter-study-output" on compact disc one of Attachment VII cannot be considered 
product output in accordance with AP-3.15Q, Managing Technical Product Inputs (Ref. 7.33), 
since some information (Table 8 and Table 17) for those runs was taken from Ref. 7.13, pp. 333- 
334, and Ref. 7.17, pp. 50-55, which are "reference only." Therefore, the information in the 
'*.cut' files of folder "ten node output" and the '*.output' files of folder 
"parameter study output" on compact disc one of Attachment VII should not be used for any 
other calculations or analyses. 

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7 REFERENCES 
CRWMS M&O 1997. SofhYare Code: SCALE. V4.3. HP. 3001 1 V4.3. 
DOE (U.S. Department of Energy) 1992. Characteristics of Potential Repository Wastes. 
DOEIRW-0184-R1. Four volumes. Washington, D.C.: U.S. Department of Energy, Office 
of Civilian Radioactive Waste Management. ACC: HQO. 19920827.0001; 
HQO. 19920827.0002; HQO. 19920827.0003; HQO. 19920827.0004. 
CRWMS M&O 1997. BWR Assembly Source Terms for Waste Package Design. 
BBA000000-01717-0200-00024 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL. 19970710.0078. 
CRWMS M&O 1999. BWR Source Term Generation and Evaluation. BBAC00000- 
01717-0210-00006 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL. 19990309.035 1. 
CRWMS M&O 1999. BWR Source Term Generation and Evaluation. BBAC00000- 
017 l7-O2lO-OOOO6 REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.20000113.0334. 
Luksic, A. 1989. Activation Measurements and Comparison with Calculations for Spent 
Fuel Assembly Hardware. Volume 1 of Spent Fuel Assembly Hardware: 
Characterization and 10 CFR 61 Classification for Waste Disposal. PNL-6906. 
Richland, Washington: Pacific Northwest Laboratory. ACC: NNA. 19890926.0124. 
Ludwig, S. B. and Renier, J. P. 1989. Standard- and Extended-Bumup PWR and BWR 
Reactor Models for the ORZGEN2 Computer Code. ORMITM- 1 101 8. Oak Ridge, 
Tennessee: Oak Ridge National Laboratory. TIC: 203557. 
ASTM B 81 1-90. 1991. Standard Specification for Wrought Zirconium Alloy Seamless 
Tubes for Nuclear Reactor Fuel Cladding. Philadelphia, Pennsylvania: American 
Society for Testing and Materials. TIC: 239780. 
ASTM A 240lA 240M-97a. 1997. Standard Specification for Heat-Resisting Chromium 
and Chromium-Nickel Stainless Steel Plate, Sheet, and Strip for Pressure Vessels. West 
Conshohocken, Pennsylvania: American Society for Testing and Materials. TIC: 
23943 1. 
NRC (U.S. Nuclear Regulatory Commission) 2000. Standard Review Plan for Spent 
Fuel Dry Storage Facilities. NUREG-1 567. Washington, D.C.: NRC. TIC: 247929. 

Title: BWR Source Tern Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Page: 56 of 58 1 
Sandoval, R.P.; Einziger, R.E.; Jordan, H.; Malinauskas, A.P.; and Mings, W.J. 1991. I 
Estimate of CRUD Contribution to Shipping Cask Containment Requirements. SAND88- 
1358. ~lbuquerque, New Mexico: ~andia National Laboratories. ACC: 
MOV. 19960802.01 14. 
Jones, R.H. 1992. Spent Fuel Corrosion Product and Fuel Cleaning Assessment. Los 
Gatos, California: Robert H. Jones, P.E., Consultant. ACC: HQX.19920825.0007. 
CRWMS M&O 1998. Summary Report of Commercial Reactor Criticality Data for 
Quad Cities Unit 2. B00000000-01717-5705-00096 REV 00. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL. 19980730.0509. 
Larsen, N.H.; Parkos, G.R.; and Raza, 0. 1976. Core Design and Operating Data for 
Cycles 1 and 2 of Quad Cities 1. EPRI NP-240. Palo Alto, California: Electric Power 
Research Institute. TIC: 237267. 
DOE (U.S. Department of Energy) 1987. Appendix 2A: Physical Descriptions of LWR 
Fuel Assemblies. Volume 3 of Characteristics of Spent Fuel, High-Level Waste, and 
Other Radioactive Wastes Which May Require Long-Tern Isolation. DOEIRW-0184. 
Washington, D.C.: US. Department of Energy, Office of Civilian Radioactive Waste 
Management. ACC: HQX. 19880405.0024. 
NRC (U.S. Nuclear Regulatory Commission) 1997. SCALE, RSZC Computer Code 
Collection (CCC-545). NUREG/CR-0200 Rev. 5. Washington, D.C.: U.S. Nuclear 
Regulatory Commission. TIC: 235920. 
CRWMS M&O 1999. CRC Depletion Calculations for Quad Cities Unit 2. B00000000- 
017 17-02 10-00009 REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL. 19990929.012 1. 
ASME (American Society of Mechanical Engineers) 1993. Steam Tables, 
Thermodynamic and Transport Properties of Steam. 6th Edition. New York, New York: 
American Society of Mechanical Engineers. TIC: 103243. 
CRWMS M&O 1999. Summary Report of Commercial Reactor Criticality Data for 
Quad Cities Unit 2. B00000000-01717-5705-00096 REV 01. Las Vegas, Nevada: 
CRWMS M&O. ACC:, MOL. 19990917.01 84. 
INCO Alloy International 1988. Product Handbook ZNCO Alloys. Huntington, West 
Virginia: INCO Alloys International, Inc. TIC: 239397. 
Hermann, O.W. and DeHart, M.D. 1998. Validation of SCALE (SAS2H) Isotopic 
Predictions for BWR Spent Fuel. ORNUTM-133 15. Oak Ridge, Tennessee: Oak Ridge 
National Laboratory. TIC: 245042. 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Page: 57 of 58 1 
CRWMS M&O 1998. Time-Dependent Shielding Analysis for Commercial SNF in the 
21 PWR Waste Package. BBAC00000-01717-0210-00003 REV 00. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL. 19981 109.0382. 
Parrington, J.R.; Knox, H.D.; Breneman, S.L.; Baum, E.M.; and Feiner, F. 1996. 
Nuclides and Isotopes, Chart of the Nuclides. 15th Edition. San Jose, California: GE. 
TIC: 233705. 
CRWMS M&O 2000. Waste Packages and Source Terms for the Commercial 1999 
Design Basis Waste Streams. CAL-MGR-MD-000001 REV 00. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL.20000214.0479. 
CRWMS M&O 1995. Total System Peg?omzance Assessment - 1995: An Evaluation of 
the Potential Yucca Mountain Repository. B00000000-0 17 17-2200-00 136 REV 0 1. Las 
Vegas, Nevada: CRWMS M&O. ACC: MOL.19960724.0188. 
CRWMS M&O 1999. BWR Source Term Generation and Evaluation. BBAC00000- 
01717-0210-00006 REV 00. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL.1998llO6.0117. 
CRWMS M&O 1999. Classification of the MGR Uncanistered Spent Nuclear Fuel 
Disposal Container System. ANL-UDC-SE-000001 REV 00. Las Vegas, Nevada: 
CRWMS M&O. ACC: MOL. 19990928.0216. 
DOE (U.S. Department of Energy) 2003. Quality Assurance Requirements and 
Description. DOEIRW-00333P, Rev. 13. Washington, D.C.: U.S. Department of 
Energy, Office of Civilian Radioactive Waste Management. ACC: 
DOC.20030422.0003. 
AP-3.12Q, Rev. 2, ICN 0. Design Calculations and Analyses. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. ACC: 
DOC.20030403.0003. 
CRWMS M&O 1999. Output Files for BWR Source Term Generation and Evaluation. 
BBAC00000-0 17 l7-O2lO-OOOO6 REV 01. Las Vegas, Nevada: CRWMS M&O. ACC: 
MOL. 19991 1 11.0701. 
AP-SI. IQ, Rev. 5, ICN 1. SofhYare Management. Washington, D.C.: U.S. Department 
of Energy, Office of Civilian Radioactive Waste Management. ACC: 
DOC.20030708.0001. 
ASME (American Society of Mechanical Engineers) 2001. 2001 ASME Boiler and 
Pressure Vessel Code (includes 2002 addenda). New York, New York: American 
Society of Mechanical Engineers. TIC: 25 1425. 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Page: 58 of 58 
AP-3.12Q, Rev. 4, ICN 2. Managing Technical Product Inputs. Washington, D.C.: U.S. 
Department of Energy, Office of Civilian Radioactive Waste Management. ACC: 
DOC.20030627.0002. 
Bowman, S.M.; Hermann, O.W.; and Brady, M.C. 1995. Sequoyah Unit2 Cycle 3. 
Volume 2 of SCALE-4 Analysis of Pressurized Water Reactor Critical Configurations. 
ORNLITM-12294N2. Oak Ridge, Tennessee: Oak Ridge National Laboratory. TIC: 
244397. 
8 ATTACHMENTS 
The hardcopy attachments are listed in Table 49. Attachment VII is a CD containing output files, 
'*.cut' files, and script files. The contents of Attachment VII are listed in Attachment XII. 
Table 49. Attachments Supporting Documentation of Source Term Generation and Evaluation 
Description 
Wt% calculations for fuel 
Light element mass calculations 
Operating conditions 
Bumup history calculation 
Parameter study results 
Crud source term calculations 
Electronic copies of files, including '*.cut" files and script files 
10-node gamma source vs. 1 -node gamma source calculations 
10-node neutron source vs. 1 mode neutron source calculations 
Initial heavy metal loading comparison calculations, and jbcl and 14c fuel impurity graph 
Script files 
SAS2HlORIGEN-S *.cut files and output files 
Radionuclide inventories for Performance Assessment 
Comparison of Source Terms per MTU of 4 Average BWR SNF Assemblies 
Attachment Number 
I 
II 
111 
IV 
V 
VI 
VII 
Vlll 
IX 
X 
XI 
XI I 
Xlll 
XIV 
No. of 
es 
16 
14 
11 
10 
10 
3 
3CDs . 
10 
12 
6 
1 
16 
10 
1 

Title: BWR Source Term Generation and Evaluation Worksheet 'length calculation' 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment I; Page 1-1 of 1-16 
I 
(Ref. 7.23, pp. 48- 
49) WT% 
15.9994 100 15.9994 Note for 000-00C-MGRO-00200-000-00A: 
234.0409 0.04423 1.88981 E-06 Corrected formula for U weight fraction in U02; 
235.0439 5 0.00021 2726 No effect on final length estimates. 
236.0456 0.023 9.74388E-07 
238.0508 94.9328 0.003987921 
237.8963629 U weight fraction in U02 
Fuel rod volume (using clad inner diameter) 10.96 theoretical density 
0.88958071 7 smear density 
total fuel cross sectional area in assembly 9.9695 
53.37484304 193.1818182 kg U02 required for 170 kg U loading 
1931 81.81 82 In grams 
19377.28253 cubic centimeters required 
:uel length required = 363.041 4897 cm 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'density' Attachment I; Page 1-2 of 1-16 
I 
Calculations for Fuel Density. Calculation of wt%'s for SAS2H input. 
Elementllsotope 
o w e n 
U-234 
U-235 
U-236 
U-238 
Atomic weight 
(Ref. 7.23, pp. 48- 
49) WT% 
15.9994 100 15.9994 Note for 000-00C-MGRO-00200-000-00A: 
234.0409 0.04423 1.88981 E-06 Corrected formula for U weight fraction in U02; 
235.0439 5 0.00021 2726 No effect on final density estimates. 
236.0456 0.023 9.74388E-07 
238.0508 94.9328 0.003987921 
237.8963629 U weight fraction in U02 
Fuel rod volume (using clad inner diameter) 
325.3730432 
total fuel volume in assembly 
19522.38259 
smear density 
9.9695 
mass in assembly 
194628.3932 
in kg 
194.6283932 
10.96 theoretical density 
I 227.2727273 kg U02 required for 200 kg U loadingpp 
227.2727273 
227272.7273 For regular fuel rods: 
11.641 64908 density required to yield a 200 kg loading or U 
Fuel rod volume (using clad inner diameter) 10.96 theoretical density 
325.3730432 
total fuel volume in assembly 9.9695 
19522.38259 mass in assembly 193.1818182 kg U02 required for 170 kg U loading 
194628.3932 
in kg 
193.1818182 
1931 81.81 82 For regular fuel rods: 
9.895401 71 5 density required to yield a 170 kg loading of U 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Calculation of wtWs for SAS2H input for Parameter Study 
Worksheet 'parameter study' 
Elemenfflsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151 A 9 7 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
159.927 
Eq. 6: (uranium oxide) 269.9507793 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881 46431 7 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 18535683 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight recents of all isotopes in gad doped fuel rods 
IsotopelElement Value Eq. Used 
Eq. 4: 
Eq. 5: 
Attachment I; Page 1-3 of 1-16 
I 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet '5.5' Attachment I; Page 1-4 of 1-16 
I 
Calculation of wt%'s for SAS2H input. 
Element/lsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.91 97 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
159.927 
Eq. 4: 
Eq. 5: 
5.5 initial U-235 enrichment and gad rods with 3.0 wt% initial enrichment of gadolinium 
WITH GADOLINIUM: 
Molecular weight of uranium oxide and gadolinium oxide 
Eq. 6: (uranium oxide) 269.87971 43 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881 4331 04 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 18566896 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight %of all isotopes in gad doped fuel rods 
IsotopelElement Value Eq. Used 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Calculation of wt%'s for SAS2H input. 
Elementllsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.9197 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
159.927 
Worksheet '5.05' Attachment I; Page 1-5 of 1-16 
I 
Eq. 4: 
Eq. 5: 
Eq. 6: (uranium oxide) 269.89361 8 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881 43921 2 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen In uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 18560788 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight %of all isotopes in gad doped fuel rods 
Isotope/Element Value Eq. Used 
Wgen 1 1.8977 11 
U-234 0.0382 10 
U-235 4.31 77 10 
U-236 0.01 99 10 
U-238 81 .I238 10 
Gd-152 0.0052 9 
Gd-154 0.0567 9 
Gd-155 0.3852 9 
Gd-156 0.5328 9 
Gd-157 0.4073 9 
Gd-158 0.6465 9 
Gd-160 0.5690 9 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet '5.0' 
Y 
Attachment I; Page 1-6 of 1-16 
1 
Calculation of wt%'s for SAS2H input. 
Elemenfflsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1 ) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.9197 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
159.927 
Eq. 4: 
Eq. 5: 
Calculations for one- and ten-node comparison cases: 5.05 initial U-235 enrichment and gad rods with 3.0 wt% initial enrichment of gadolinium 
WITH GADOLINIUM: 
Molecular weight of uranium oxide and gadolinium oxide 
Eq. 6: (uranium oxide) 269.8951 629 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium In uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881 439891 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 185601 09 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight %of all isotopes in gad doped fuel rods 
Isotope/Element Value Eq. Used 
Oxygen 11.8976 11 
U-234 0.0378 10 
U-235 4.2750 10 
U-236 0.01 97 10 
U-238 81 .I672 10 
Gd-152 0.0052 9 
Gd-154 0.0567 9 
Gd-155 0.3852 9 
Gd-156 0.5328 9 
Gd-157 0.4073 9 
Gd-158 0.6465 9 
Gd-160 0.5690 9 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet '4.5' 
Calculation of wt%'s for SAS2H Input. 
Attachment I; Page 1-7 of 1-1 6 
I 
Elementhsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.9197 
153.9208 
1 54.9226 
155.9221 
156.9239 
157.9241 
159.927 
Eq. 4: 
Eq. 5: 
4.5 initial U-235 enrichment and gad rods with 3.0 wt% initial enrichment of gadolinium 
WITH GADOLINIUM: 
Molecular weight of uranium oxide and gadolinium oxide 
Eq. 6: (uranium oxide) 269.91 061 18 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881 446677 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 1 8553323 
Eq. 8 (gadolinium oxide) 0.132422605 . .- 
Weight %of ail isotopes in &d doped fuel rods 
IsotopelElement Value Eq. Used 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Calculation of wt%'s for SAS2H input. 
Elemenfflsotope 
oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1 ) 
U-238 (From Eq. 3) 
Gd- 1 52 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd- 1 58 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.91 97 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
159.927 
Worksheet '4.0' Attachment I; Page 1-8 of 1-1 6 
I 
4.0 initial U-235 enrichment and gad rods with 3.0 wt% initial enrichment of gadolinium 
WITH GADOLINIUM: 
Molecular weight of uranlum oxlde and gadolinium oxide 
Eq. 6: (uranium oxide) 269.9260609 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranlum or gadolinium oxlde 
Eq. 7 (uranium) 0.881 453462 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 18546538 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight % of all isotopes in gad doped fuel rods 
IsotopelElement Value Eq. Used 
Eq. 4: 
Eq. 5: 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet '3.74' Attachment I; Page 1-9 of 1-16 
Calculation of wt%k for SAS2H input. 
Elemenfflsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1 ) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd- 155 
Gd-156 
Gd- 157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.91 97 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
1 59.927 
Eq. 4: 
Eq. 5: 
Eq. 6: &raniurn oxide) 269.9340945 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881 45699 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 1854301 
Eq. 8 (gadolinium oxide) 0.1 32422605 
Weight %of all Isotopes in gad doped fuel rods 
lsotopeElement Value Eq. Used 
Oxygen 
U-234 
U-235 
U-236 
U-238 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd- 160 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet '3.5' Attachment I; Page 1-10 of 1-16 
I 
Calculation of wt%'s for SAS2H input. 
Elemenfflsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.91 97 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
159.927 
Eq. 4: 0.004202692 
237.94271 01 
Eq. 5: 0.006360027 
157.2320428 
Eq. 6: (uranium oxide) 269.941 51 01 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881 460247 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 18539753 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight % of all isotopes in gad doped fuel rods 
IsotopelElement Value Eq. Used 
o w e n 11.8956 11 
U-234 0.0257 10 
U-235 2.9926 10 
U-236 0.0138 10 
U-238 82.4696 10 
Gd-152 0.0052 9 
Gd-154 0.0567 9 
Gd-155 0.3852 9 
Gd-156 0.5328 9 
Gd-157 0.4073 9 
Gd-158 0.6465 9 
Gd-160 0.5690 9 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
1,. Calculation of wt%'s for SAS2H input. 
Element/lsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.9197 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
1 59.927 
Worksheet '3.0' 
3.0 initial U-235 enrichment and gad rods with 3.0 wt% initial enrichment of gadolinium 
WITH GADOLINIUM: 
Molecular weight of uranium oxide and nadolinlum oxlde 
Eq. 6: (uranium oxide) 269.9569589 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadollnium in uranium or gadollnium oxlde 
Eq. 7 (uranium) 0.881 46703 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 1 853297 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight %of all isotopes in gad doped fuel rods 
Isotope/Element Value Eq. Used 
Eq. 4: 
Eq. 5: 
Attachment I; Page 1-1 1 of 1-16 
I 

Title: BWR Source Term Generation and Evaluation, 
Document Identifier: 000-00C-MGRO-00200-000-OOA 
Calculation of wt%'s for SAS2H input. 
Elemenfflsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.91 97 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
159.927 
Worksheet '2.5' Attachment I; Page 1-12 of 1-16 
I 
Eq. 6: (uranium oxide) 269.9724072 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881473813 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide. 
Eq. 8 (uranium oxide) 0.1 185261 87 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight % of all isotopes in gad doped fuel rods 
IsotopelElement Value Eq. Used 
Eq. 4: 
Eq. 5: 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet '2.0' Attachment I; Page 1-13 of 1-16 
I 
Calculation of wt%'s for SAS2H input. 
Element/lsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
38 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.91 97 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
159.927 
Eq. 4: 
Eq. 5: 
Eq. 6: (uranium oxide) 269.9878543 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881 480594 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.118519406 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight %of all isotopes in gad doped fuel rods 
lsotopeElement Value Eq. Used 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Calculation of wt%'s for SAS2H input. 
Elemenfflsotope 
o w e n 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151 
153.9208 
154.9226 
155.9221 
156.9239 
157.9241 
159.927 
Worksheet '1 -5' Attachment I; Page 1-14 of 1-16 
I 
1.5 initial U-235 enrichment and gad rods with 3.0 wt% initial enrichment of gadolinium 
WITH GADOLINIUM: 
Molecular weight of uranium oxide and gadolinium oxide 
Eq. 6: (uranium oxide) 270.0032997 
362.4622855 Eq. 6: (gadolinium oxide) 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881 487374 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 1851 2626 
Eq. 8 (gadolinium oxide) 0.1 32422605 
Weight %of all isotopes in gad doped fuel rods 
IsotopelElement Value Eq. Used 
Eq. 4: 
Eq. 5: 

Title: BWR Source Term Generation and Evaluation Worksheet '1.0' Attachment I; Page 1-1 5 of 1-16 
i Document Identifier: 000-00C-MGRO-00200-000-00A 
Calculation of wt%'s for SAS2H input. 
Element/lsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 ' 
Gd-156 
Gd- 157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) 
15.9994 
234.0409 
235.0439 
236.0456 
238.0508 
151.9197 
153.9208 
154.9226 
155.9221 
1 56.9239 
157.9241 
159.927 
Eq. 4: 
Eq. 5: 
1.0 initial U-235 enrichment and gad rods with 3.0 wt% initial enrichment of gadolinium 
WITH GADOLINIUM: 1 
Molecular weight of uranium oxide and gadolinium oxide 
Eq. 6: (uranium oxide) 270.01 8742 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881494152 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen In uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 18505848 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight % of all isotopes in gad doped fuel rods 
lsotopelElement Value Eq. Used 
O~gen . 11.8923 11 
U-234 0.0066 10 
U-235 0.8550 10 
U-236 0.0039 10 
U-238 84.6393 10 
Gd-152 0.0052 9 
Gd-154 0.0567 9 
Gd-155 0.3852 9 
Gd-156 0.5328 9 
Gd-157 0.4073 9 
Gd-158 0.6465 9 
Gd-160 0.5690 9 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'natural' Attachment I; Page 1-16 of 1-16 
I 
Calculation of wt%'s for SAS2H Input. 
Eiementllsotope 
Oxygen 
U-234 (From Eq. 2) 
U-235 
U-236 (From Eq. 1 ) 
U-238 (From Eq. 3) 
Gd-152 
Gd-154 
Gd-155 
Gd-156 
Gd-157 
Gd-158 
Gd-160 
Atomic weight 
(Ref. 7.23, pp. 36- 
37; 48-49) WT% wtWatomic wt% 
15.9994 100 
234.0409 
235.0439 
236.0456 
238.0508 99.2803874 0.41 7055466 
151.9197 0.2 0.001 31 6485 
153.9208 2.18 0.0141 63128 
154.9226 14.8 0.095531 575 
155.9221 20.47 0.131 283506 
156.9239 15.65 0.099729869 
157.9241 24.84 0.157290749 
159.927 21 -86 0.136687364 
Eq. 4: 
Eq. 5: 
0.71 1 initial U-235 enrichment and gad rods with 3.0 wt% initial enrichment of gadolinium 
WITH GADOLINIUM: 1 
Molecular weight of uranium oxide and gadolinium oxide 
Eq. 6: (uranium oxide) 270.0276654 
Eq. 6: (gadolinium oxide) 362.4622855 
Weight fraction of uranium or gadolinium in uranium or gadolinium oxide 
Eq. 7 (uranium) 0.881 498068 
Eq. 7 (gadolinium) 0.867577395 
Weight fraction of oxygen in uranium and gadolinium oxide 
Eq. 8 (uranium oxide) 0.1 18501932 
Eq. 8 (gadolinium oxide) 0.132422605 
Weight % of all isotopes in gad doped fuel rods 
IsotopelElement Value Eq. Used 
o w e n 1 1.8920 11 
U-234 0.0046 10 
U-235 0.6079 10 
U-236 0.0028 10 
U-238 84.8900 10 
Gd-152 0.0052 9 
Gd-154 0.0567 9 
Gd-155 0.3852 9 
Gd-156 0.5328 9 
Gd-157 0.4073 9 
Gd-158 0.6465 9 
Gd-160 0.5690 9 

Tlle: 8WR Source Term Generation and Evaluation Worksheet 'mposrtions' Attachment II; Page 11-1 of 11-14 
Document Identifier: 000-00C-MGRO-00200-000-00A I 
t I 
Channel cladding 
fractions in 
the various 
regions 
fractions in 
he various 
regions 
- Bottom 
channel 
zirc4 
0.6512 
0.0008 
0.0007 
0.0013 
0.0091 
0.6394 
0.0258 
0.0693 
0.8880 
0.0169 - 
water rods 
zirc 4 
0.2428 - 
- 0.0003 
0.0002 
0.0005 
0.0034 
0.2383 
top 
plenum 
fuel 
bottom - Plenum 
channel 
zirc 4 
2.6754 - 
- 0.0032 
0.0027 
0.0054 
0.0375 
2.6267 
Water rod 
volume (2 Cladding 
water rods) 51 1.089 volume 6919.51 
Water rods Cladding 
mass 3.353 mass 45.392 
Channel 
volume 
(nA3) 5883.737 
lannel mass 
(kg) 38.597 
fuel 0.928 
plenum 0.072 
ven channel 
mass - - 
cladding - 
zirc 2 
42.105 - 
0.050! 
0.042' 
0.042' 
0.021' 
0.589! 
41 .%O' 
- TOP 
:ompression 
spring 
Incnnel 
0.58 
- 
channel - 
zirc4 
0.9949 - 
- 0.001 2 
0.0010 
0.0020 
0.0139 
0.9768 - 
- 
- 
plenum 
:!adding 1 spkgs *:el ipacer grids water rods A hardware Tie plate 
ss304 
1.99989 
tie plate - 
ss 304 
4 . m - 
- 
material 
mass 
Wt% 
0.12 
0.1 
0.1 
0.05 
1.4 
98.23 
0.12 
0.1 
0.2 
1.4 
98.18 
Material 
Zircaloy-2 
Zircaby-4 
- 
iconel X-75( 
- ss 304 
Symbol 
0 
Cr 
Fe 
Ni 
Sn 
Zr 
3.56 Mg/m3 
0 
Cr 
Fe 
Sn 
Zr 
6.56 MgIm3 
Ni 
Cr 
Fe 
Nb 
Ti 
Al 
Co 
Mn 
Si 
Cu 
C 
S 
C 
Mn 
P 
S 
Si 
Cr 
Ni 
N 
Fe 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'Table 5.3.2.1 -2' Attachment II; Page 11-2 of 11-14 
I 
top end fitting regions Scale factor = 0.15 
compression I 
channel spring tie plate I raw scaled 
0 0.0012 1 0.0012 0.0002 
cladding channel getters water rods springs I raw scaled 
0 0.003944 0.00321 0.000291 1 0.0074 0.0022 
I fuel region one node 
cladding channel spacers water rods spacers I raw 
0 0.050526 0.041 131 0.00234 0.003732 1 0.0977 

Tlle: BWR Source Term Generation and Evaluation 
Document Identifier: WO-OOC-MGR&00200-000M)A 
Worksheet Yen node' Attachment It; Page 11-3 of 11-14 
I 
Spacer? no yes yes yes yes yes Yes yes yes no 
% d 1 spacer 0.0000h 69.231% 30.769% 100.0000A 100.OW% 61.53836 107.691% 30.769% 200.0000% 0. 
lei region 
ne node 
water 
cladding channel spacers rods 
0 0.050526 0.041 131 0.00234 0.00373; 
water 
Me 1 cladding channel rods - - 0 0.050526 0.041131 0.093732 
NODE: 1 2 3 4 5 6 7 8 9 1 
dimension 6 12 12 18 12 18 18 18 25.24 
fraction of fuel 0.04137931 0.082759 0.082759 0.1241379 0.08275862 0.124138 0.1241379 0.124138 0.174069 0.04137 
light elements are determined by multiplying the hardware by the percent In the fuel region 
takinlng the spacer material, dividing by the number of spacers 
and thenmultlpling by the percentage of one spacer In that node 
clad. wr. I I 

Title: BWR Source Term Generation and Evaluation 
Document Identlfler: 000-00C-MGRO-00200000-00A 
Worksheet 'ten node' Attachment II; Page 11-4 of 11-14 
I 
I water and I 
node 3 cladding channel spacers rods spacers channel spacers 1 total 
0 0.050526 0.041131 0.00234 0.003732 0.0079 0.0001031 0.0080 
Al 0.00227661 0.0000 0.0001 0.0001 
C 0.00019514 0.0000 8.58E-06 0.0000 
Co 0.0032523 0.0000 0.000143 0.0001 
Cr 0.042105 0.034276 0.00195 0.00311 0.0487845 0.0066 0.00223 0.0088 
Cu 0.00130092 0.0000 5.72E-05 0.0001 
Fe 0.042105 0.068551 0.0039 0.00622 0.0260184 0.0097 0.001315 0.0110 
Mn 0.00260184 0.0000 0.000114 0.0001 
Nb 0.0032523 0.0000 0.000143 0.0001 
N 0.0000 0 0.0000 
Ni 0.021053 0.22840903 0.0017 0.01004 0.0118 
P 0.0000 0 0.0000 ' 
S 3.2523E-05 0.0000 1.43E-06 0.0000 
Si 0.00130092 0.0000 5.72E-05 0.0001 
Sn 0.589476 0.47986 0.027299 0.04354 0.0921 0.0012 0.0933 
Ti 0.00780552 0.0000 0.000343 0.0003 
Zr 41.36014 33.6519 1.914412 3.053391 6.4606 0.084149 6.5447 
I clad, wr. I I water and I 
node 5 cladding channel spacers rods spacers channel spacers 1 total 
0.050526 0.041 131 0.00234 0.003732 0.0079 0.0003341 0.0082 
I clad, wr. 
and 1 
node 4 cladding channel spacers water rods spacers channel spacers total 
0 0.050526 0.0411309 0.00233988 0.003732 0.01 18 0.0003343 0.0122 
Al 0.0022766 0.0000 O.wO3252 0.0003 
craa, wr, 
and 
node 6 cladding channel spacers water rods spacers channel spacers 
0 0.050526 0.041 1309 0.00233988 0.003732 0.01 18 0.0002057 
- 
total - 0.0120 
0.0002 
0.0000 
0.0003 
0.0143 
0.0001 
0.0171 
0.0002 
0.0003 
0.0000 
0.0227 
0.0000 
0.0000 
0.0001 
0.1405 
0.0007 
9.8592 - 

Tlle: BWR Source Term Generation and Evaluation 
Document Identifier: 000dOC-MGRO.002OMMO-OOA 
Worksheet 'ten node' Attachment II: Page 11-5 d 11-14 
I 
I . . 
water and I 
total 
0.0122 
0.0004 
0.0000 
0.0005 
0.0177 
0 . m 
0.0191 
0.0004 
0.0005 
0.0000 
0.0378 
0.0000 
0.0000 
0.0002 
0.1423 
0.0012 
9 98% 
water 
clad, wr, 
and I 
node 9 cladding channel spacers rods spacers 
0 0.050526 0.041131 0.00234 0.003732 
total - 0.0173 
0.0007 
0.0001 
0.0009 
0.0283 
0.0004 
0.0289 
0.0007 
0.0009 
0.0000 
0.0689 
0.0000 
0.0000 
0.0004 
0.2015 
I and I 
node 10 - 0 
Al 
C 
Co 
Cr 
Cu 
Fe 
Mn 
Nb 
N 
Ni 
P 
S 
Si 
Sn 
cladding channel spacers water rods spacers - 0.050526 0.0411309 0.00233988 0.003732 
0.0022766 
0.0001951 
0.0032523 
0.042105 0.0342757 0.0019499 0.00311 0.0487845 
0.0013009 
0.042105 0.0685514 0.0038998 0.00622 0.0260184 
0.0026018 
0.0032523 
channel - 0.0039 
0.0000 
0.0000 
0.0000 
0.0033 
0.0000 
0 . m 
0.0000 
0.0000 
0.0000 
0.0009 
0.0000 
0.0000 
0.0000 
0.0461 

TaDIe5.3.2.1. SpedIIUll~n 01 N ~ P ~ I S used in SAS2H and MCNP ulsul.tlons 
- 
Tap 
88Jbn( 
l m e l 
0 58 - 
- 
04073 
0 0870 
o w 
o w 
00139 
0 W41 
O W 
00546 
0 W23 
omw 
00303 
0 M31 - 
me, Cl~ddng 
5883.737 I &)' 511.0892 vdum 6919.507 plenum 0.0893 
0.8880 
bonm 00169 
An2 
0.w: 
0.w: 
om 
ow: 
OE! 
3211 - 
- 
- 
- 
45.392 38.597 Water & mi 3.353 
29.937 kg I 
I FWI 
channel spacergrids water md 
adding re 

Tltle: BWR Source Term Generation and Evaluation 
Document Identifier: ~ - M G R 0 0 0 2 ~ 0 0 A 
Worksheet '8.1 1 .compcsitions' Attachment II; Page 11-7 of 11-14 I 
n the fuel mat itself. 
PWR mass (k 539.77273 
J I I r d 
;sembly vdume (=)I I 54194.05 
isembly vdum (031 19522.38 
I I I 
Impurities in the fuel: 1 I I I I I I I I 
I I DWR I I RWR I I 

Title: BWR Source Ten Generation and Evaluation 
Document Identlfler: OO&OOC-MGR0-00200-00(HXlA 
Worksheet '8.1 1.compositions' Attachment II; Page 11-8 of 11-14 
I 
BWR light elements (kg) I I I I I I I I 
1 
w O.OM)~ at mass (9) wl fr 
pb 0.0002 b10 10.012937 0.16431 1.136E-07 
bi 0.MXX)B b-11 11.009306 0.81569 4.571 E-07 
at mass (g) wt fr 
li-6 6.015122 0.065 6.667E-08 
li-7 7.016004 0.935 8.222E-07 
8.888E-07 

Title: BWR SourceTerm Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Table 5.3.2.1. Spciflcatlon of 
materials used in SAS2H and 
MCNP calculatlons 
Worksheet '81 t .ssciad.compositions' Anachment il; Page 11.9 of 11.14 
I 
Stainless Steel Clad assemblies : 1 
Other Hardware is the same as 
lor the Waste Stream cases. 
That information is taken from 
the 81 1 .compositions worksheet. 
Only the water mds and cladding 
are calculated here 
!I and plenu 
Channel 
fractions in 
the various 
regions 
Channel volume (cmW.3) 5883.737 
channel mass (kg) 38.597 
tilven channel mass 29.937 kg 
Material 
I ss 348H 
I re ion on1 . -7- 
Water rod 
volume (2 waler 
rods) 511.0891957 Cladding volume 6919.507762 
Water rods mass 4.104 Cladding mass 55.564 
PLtNUM FUEL 
cladding 
fractions in 
Symbol 
C 
Ihe various 
regions + 
hardware 
material 
mass 
Wt% 
0.07 
wafer rods cladding 
ss348H 1 ss WH 
0.297149444 4.023031409 
water rods 
ss 3481.1 
3.806896797 
cladding 
ss 348H 
51.54061592 
0.00021 0.00282 0.0027 0.0361 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'new.lite.el.tablesl Attachment 11; Page 11-10 of 11-14 
I 
op end fitting regions Scale factor = 0.15 
compression I 
channel spring tie plate I raw scaled 
0 0.0012 1 0.0012 0.0002 
bottom end fitting Scale factor = 0.225 
channel tie plate I raw scaled 
0 0.000781486 1 0.0008 0.0002 
I plenum Scale factor = 0.3 
plenum I 
cladding channel getters water rods Springs I raw scaled 
0 0.003944 0.00321 0.000291 ( 0.0074 0.0022 
I fuel region one node 
cladding channel spacers water rods spacers raw 
0 0.050526 0.041 131 0.00234 0.003732 0.0977 
A1 0.002277 0.0023 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
p end fitting regions Scale fac 
compression 
channel spring tie plate 
0 0.0012 
Worksheet '81 1 .lite.el.tables ' Attachment II; Page 11-1 1 of 11-14 
I 
r = 0.151 plenum Scale factor = 
I plenum 1 
raw 
0.0012 
0.0041 
0.0019 
0.0074 
0.4680 
0.0023 
1.3982 
0.0446 
0.0058 
0.0020 
0.61 57 
0.0009 
0.0007 
0.0173 
0.0169 
0.0139 
0.9739 - 
scaled cladding channel getters water rods springs raw 
0 0.0002 0 0.0039 0.0032 0.000291 0.0074 
Al 0.0006 Al 0.0119 0.0119 
bottom end fitting Scale factor = 0.2251 
channel tie plate 
0 0.0008 
F raw 
0.0008 
0.0000 
0.0038 
0.0038 
0.9069 
0.0000 
3.2208 
0.0954 
0.0000 
0.0048 
0.4970 
0.0021 
0.0014 
0.0358 
0.01 1 1 
0.0000 
0.6374 - 
scaled =I 
scaled 
0.0022 -1 0.0036 
fuel region one node 
fuel pellets cladding channel spacers water rods spacers 
0 27.2727 0.050526 0.041131 0.00234 0.003732 
impurities raw 
27.3704 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet '81 1 .lite.el.tables ' Attachment II; Page 11-12 of 11-14 
I 
Adding the fuel region hardware and impurities: I I 
I I I I I I 
1 top 1 bottom 
I I 

Title: BWR Source Term Generation and Evaluation Worksheet '81 1 .ssclad.lite.el.tables ' 
Document Identifier: 000-00C-MGRO-00200-000-00A 
I plenum Scale factor = 0.3 
plenum I 
Attachment II; Page 11-13 of 11-14 
I 
'**Heav oxygen content 
water rods actually for 
e oxygen in the fuel 

Title: BWR Source Term Generation and Evaluation Worksheet '81 1 .ssclad.lite.el.tables ' 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment II; Page 11-14 of 11-14 
I 
plenum lfuel hardware Impurities 1 total fuel plenum 
0.00002 1 Ag 0.0000 0.0000 

Title: BWR Source Term Generation and Evaluation 
Document Identlfler: 000-00C-MGRO-W2~?0-00000A 
Worksheet 'C3 AVERAGE VALUES Anachment Ill; Page 111-1 of 111-11 I 

TBe: BWR Source Term Generation and Evaluation 
Document Idenliner: mW-MGROWZOOmWA 
Wodsheet 'APF CALCS Amchment Ill; Page 111.2 d 111-11 
I 
Final Assembly burnup by nodeand assembly 
APFs: A l A Z A 3 M & M A 7 A 8 
1 15.24 7.68 7.99 7.52 7.43 8.10 8.26 7.02 7.73 
2 30.48 29.24 30.37 28.85 28.60 30.88 30.82 27.39 29.42 
3 30.48 36.82 3834 38.83 38.47 39.22 38.31 34.99 37.42 
4 45.72 38.44 40.65 39.91 40.15 41.53 40.37 37.66 40.29 
6 30.48 40.72 41.84 41.W 42.02 42.83 41.46 39.29 41.88 
6 45.72 41.31 42.14 41.48 42.50 43.01 42.14 40.08 42.14 
7 45.72 41.13 41.69 41.20 42.03 42.21 42.24 40.37 41.79 
8 45.72 39.50 40.35 39.37 40.82 40.08 40.67 39.71 40.16 
9 M.11 31.03 30.77 30.25 31.71 30.50 31.44 30.39 30.86 
10 16.24 8.55 8.33 8.32 8.58 8.16 8.66 7.92 8.20 
average 34.88 35.58 34.80 35.53 35.99 35.80 33.55 35.36 
rnaxapf 1.18 1.18 1.19 l.X) 1.20 1.18 1.19 1.19 
9 84.11 32.54 31.89 30.19 30.12 31.53 
10 15.24 918 8.92 8.26 8.19 8.60 
average 38.23 35.57 35.08 35.55 35.61 
rnaxapf 1.19 1.19 1.21 1.19 1.18 
C1 C2 C3 C4 C5 C6 C7 C8 C9 C10 C11 C12 C13 C14 
15.24 9.50 9.08 9.82 8.58 8.39 9.01 8.88 7.63 8.55 8.74 7.18 8.01 8.35 8.53 
30.48 35.28 33.92 35.82 32.28 32.01 33.55 33.66 28.89 32.77 32.95 28.04 30.43 31.44 32.88 
30.48 46.18 43.26 46.39 41.44 41.01 42.73 43.44 36.97 42.08 42.27 36.37 38.77 39.71 42.M 
45.72 48.02 48.32 48.14 44.89 44.00 45.87 48.22 39.95 44.38 45.27 39.32 41.85 42.20 44.37 
30.48 49.28 47.79 49.30 48.54 45.63 47.01 47.36 41.47 45.25 46.79 41.01 43.30 43.68 45.31 
45.72 49.22 47.99 49.23 48.79 48.01 47.34 47.26 42.16 45.11 46.71 41.73 43.85 44.44 46.19 
45.72 47.87 48.96 47.87 46.73 45.41 48.81 48.08 42.75 43.75 48.32 41.77 43.45 44.65 43.87 
45.72 44.15 43.84 44.48 43.69 42.79 43.87 42.82 41.55 40.53 43.W 39.78 41.21 42.55 40.83 
84.11 31.31 31.78 31.82 31.55 31.57 31.89 30.46 30.77 29.15 30.70 29.26 30.80 31.88 29.18 
15.24 8.85 9.01 8.86 8.66 8.78 8.99 8.28 8.44 7.84 8.47 7.87 8.67 9.09 7.55 
avenge 37.28 37.23 
max apl . 1 . M 1.22 

WMsbheel'APF CALCS' Anadunent 116 Page 111-3 d III-11 I 
7 45.72 40.28 40.92 39.92 40.11 38.55 40.10 40.81 40.50 40.11 39.93 39.58 41.92 41.23 39.58 
8 45.72 38.63 37.42 38.63 38.65 35.52 38.78 37.03 38.48 38.32 37.11 36.02 38.19 37.32 38.03 
9 48.97 27.87 28.25 27.73 27.77 27.27 28.M 27.47 27.10 27.29 26.21 28.94 28.48 27.71 27.42 
10 30.48 7.98 8.21 7.94 8.03 7.84 8.18 7.71 7.80 7.75 8.10 7.50 8.11 8.02 7.90 
average 33.60 33.83 33.25 33.35 31.93 33.10 3 3 . 2 5 3 3 . ~ 33.12 33.08 32.38 34.47 33.75 32.45 
rnarepf 1.26 1.24 1.25 1.25 1.24 1.25" - 128 , 1.27 it26 1.25 '1.25 1.25' ' 1.26" 1.26 
7 45.72 40.M 40.88 39.93 37.38 38.48 40.15 37.47 40.87 40.13 38.83 
8 45.72 38.59 38.99 38.83 33.91 38.28 37.41 33.95 38.78 36.71 34.83 
9 48.97 27.89 27.80 28.15 28.76 28.15 28.42 25.74 27.51 27.84 28.20 
10 30.48 8.13 8.18 8.47 7.44 8.45 8.43 7.44 8.11 8.28 7.75 
average 32.38 33.48 32.85 30.68 31.82 33.22 30.78 33.89 32.99 31.85 
lmarapf 1.26 1.25 1.26 1 . 2 8 . 1.23 ,'1.24 1.28 1.27. 124 1 . 2 6 

Anachment Ill; Page 111-4 d 111-11 
I 
~~~ ~~~ ~ ~ ~~~~ ~~~ ~~ ~- 
7 45.72 25.93 27.33 28.38 29.53 29.22 29.58 26.32 25.46 25.27 29.53 26.W 
8 45.72 23.40 24.64 25.66 25.54 26.84 26.37 24.50 23.14 23.28 2612 23.46 
9 48.97 16.56 17.08 18.22 17.53 18.86 18.25 17.85 16.46 16.82 18.41 16.59 
10 30.46 4.38 4.50 4.90 4.62 5 4 5.02 4.83 4.35 4.67 5.11 4.39 
average 21.51 22.66 23.41 23.67 24.51 24.65 22.16 21.12 20.72 24.70 21.53 
maxapf 1.28 1.29 1.27 1.26 1.26 1.30 1.28 1.27 1.27 1.29 1.29 
Jl J2 J3 J4 J6 J6 J7 J8 J9 JlO J11 512 513 J14 J15 J16 
1 15.24 3.70 3.35 3.30 3.17 3.51 3.73 3.03 2.90 3.61 3.M 3.13 3.69 3.31 3.28 3.23 3.03 
7 45.72 20.08 18.66 19.34 16.86 19.58 20.56 17.92 17.85 20.29 20.38 18.71 20.61 18.99 18.83 19.61 18.19 
8 45.72 17.82 16.38 16.86 17.00 17.12 18.14 15.69 15.76 17.67 17.76 16.81 17.90 16.55 16.95 17.56 15.92 
9 46.97 11.61 10.66 10.97 11.38 11.13 11.85 10.50 10.77 1155 11.62 11.03 11.58 10.75 11.30 11.60 10.44 
10 30.48 2.95 2.72 2.80 2.93 2.86 3.07 2.74 2.87 3.03 3.05 2.91 3.01 2.77 2.90 2.91 2.73 
average 16.86 15.51 15.89 15.38 16.23 17.10 14.64 14.59 16.74 18.83 15.34 17.09 15.64 15.67 15.46 13.86 
maxapf 1.31 132 ,,1.34 1.30 1.32 1.31. 1 1.33 ,132 1.32 1.33 1.33 1.31 1.28 1.43 

Titk BWR Source Term Generation and Evaluation 
Dosumen1 Identiller: WDWC-MGROWZOOmWA 
WolkDheet 'WF CALCS' Attachment Ill: Page 111-501 tIi.11 I 
8 45.72 18.21 16.41 17.96 17.70 
9 48.97 12.48 11.05 12.W 12.03 
10 30.48 3.05 2.72 2.93 2.92 
average 17.09 15.34 17.18 15.76 
maxapl 1 . 1.31 1.31 ,' 1.28 
.~~ ~ ~ ~ 
6 45.72 7.87 7.75 7.71 7 . 7.88 7.71 7.02 7.59 7.71 7.91 7.76 7.76 7.84 7.90 7.68 7.66 
7 45.72 7.42 7.17 7.24 6.% 7.15 7.14 6.58 7.05 7.13 7.30 7.14 7.11 7.01 7.21 7.02 6.99 
8 45.72 8.51 6.27 8.34 5.83 6.23 6.23 5.75 6.12 6.20 811 6.27 6.11 8.13 8.18 8.M 8.03 
9 48.97 4.39 4.20 4.27 3.90 4.14 4.15 3.84 4.05 4.12 4.25 4.13 4.W 3.99 4.02 3.94 3.91 
10 30.48 1.10 1.03 1.05 0.94 1.01 1.01 0.93 0.99 1.W 1.W 1.03 0.97 0.98 0.98 0.98 0.94 
average 5.91 8.13 2, 5.86, 8.27 6.11 5.84 6.02 6.07 5.92 5.98 
d x a p f 1.35 1. 1.38 1.33 1.35 1.38 '.1.38?.1.37 1.37 1.36 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: OM)-OOC-MGRO-O0200-000-00A 
Worksheet 'APF CALCS' Attachment Ill; Page 111-6 of 111-1 1 
I 
Thermal Hydraulic Information tor assembly C3 
Data Data Data Data Data Data Data Data Data Data 
point 4 point 5 point 6 point 7 point 8 point 9 point 10 point 11 point 12 point 13 
water Fuel water Fuel water Fuel water Fuel water Fuel water Fuel water Fuel water Fuel water Fuel water Fuel 
Node density temp density temp density temp density temp density temp density temp density temp density temp density temp density temp 
15.24 1 0.7396 672.5 0.7396 680.2 0.7396 654.9 0.7396 678.8 0.7396 674 0.7396 647.9 0.7396 637.4 0.7396 641.4 0.7396 663.9 0.7396 687.1 
30.48 2 0.7396 1050 0.7396 1061 0.7396 968.2 0.7396 996.3 0.7396 934.6 0.7396 852.9 0.7396 813.3 0.7396 821.3 0.7396 890.9 0.7396 949.6 
15.24 10 0.2595 659.9 0.2392 693 0.2383 641.9 0.2421 679.1' 0.2461 660.7 0.2512 668 0.2516 628.3 0.2552 628 0.2576 623.7 0.259 634.1 
Averages: 0.4403 1059.0 0.4351 1073.3 0.4333 1061.9 0.4402 1071.3 0.4450 995.0 0.4506 967.0 0.4510 910.2 0.4556 916.7 0.4580 949.1 0.4593 970.5 
I Cycle I I laveraged H20frac 
DP EFPD Node Fuel temps 4 5 6 7 8 9 10 11 12 13 
4 222.47 1 1669.1 1.0 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 1.0000 
I ,-". . . I I I Averages 
I I water Fuel I 
EFPD density temp - 222.47 0.4403 1059.0 
lode 5 only 5 - 1211.5 1211.5 0.4559 0.4559 
1142.8 0.461 
1276.3 1276.3 0.4587 0.4587 
11 33.7 0.47 
1059.2 1059.2 0.4776 0.4776 
1 006.2 0.4852 
1011.9 1053.42 0.486 0.49643 
1018.5 0.4935 
1078.7 0.4975 
1065.3 0.4993 
1100.4 0.47847 
11 29.2 0.47268 
Data Data Data 
point 11 point 12 point 13 

Title: BWR Source Term Generation and Evaluation Worksheet 'APFvsBURNUP1 Attachment Ill; Page 111-7 of 111-1 1 
Document Identifier: 000-00C-MGRO-00200-000-00A I 
10.00 20.00 30.00 40.00 
Assembly Average Burnup (GWdIMTU) 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment Ill; Page 111-8 of 111-1 1 
I 
Average Burnup 
34.88 
35.58 
34.80 
35.53 
35.99 
35.80 
33.85 
35.36 
36.23 
35.57 
35.06 
35.55 
35.61 
40.37 
39.51 
40.57 
38.67 
38.09 
39.1 9 
38.89 
35.50 
37.20 
38.59 
34.56 
36.42 
37.28 
37.23 
38.51 
39.99 
38.38 
34.44 
38.08 
37.1 6 
39.34 
37.96 
37.88 
37.74 
33.60 
33.83 
33.25 
33.35 
31.93 
33.10 
33.25 
33.09 
33.12 
33.08 
32.38 
34.47 
33.76 
APF 
1.18 
1.18 
1.19 
1.20 
1.20 
1.18 
1.19 
1 .I9 
1.19 
1.19 
1.21 
1.19 
1.18 
1.22 
1.21 
1.22 
1.21 
1.21 
1.21 
1.22 
1.20 
1.22 
1.21 
1.21 
1.20 
1.20 
1.22 
1.20 
1.21 
1.20 
1.19 
1.20 
1.20 
1.20 
1.20 
1.20 
1.20 
1.25 
1.24 
1.25 
1.25 
1.24 
1.25 
1.26 
1.27 
1.26 
1.25 
1.25 
1.25 
1.26 
Assembly 
A1 
A2 
A3 
A4 
A5 
A6 
A7 
A8 
B 1 
B2 
B3 
B4 
65 
C1 
C2 
C3 
C4 
C5 
C6 
C7 
C8 
C9 
C10 
C11 
C12 
C13 
C14 
D 1 
D2 
D3 
D4 
D5 
D6 
D7 
D8 
D9 
Dl 0 
E 1 
E2 
E3 
E4 
E5 
E6 
E7 
E8 
E9 
El 0 
El 1 
El 2 
El 3 
Worksheet 'APF vs Burnup data' 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment Ill; Page 111-9 of 111-1 1 
I 
Worksheet 'APF vs Burnup data' 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment Ill; Page 111-10 of 111-1 1 
Worksheet 'APF vs Burnup data' 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment Ill; Page III-11 Of 111-1 1 I 
FOR ONE NODE MODEL (~verages:l 
1 1 ~oderatorl Fuel 
DP I EFPD I densities I temp 
4 1 222.47 222.47 1 0.4403 1 916.9 
> 1; 1 324.71 100.33 0.4!593 i 834.7 1 sum 2432.52 1540.11 0.4468 858.6 averages 
wei hted avera es 0.4435 883.1 
Worksheet '812-C3 AVERAGE VALUES' 

I Tltle: BWR Source Term Generation and Evaluation Worksheet 'bumups, new lows' Attachment IV; Page IV-1 of IV-10 
Document Identifier: 00000C-MGRO-002WOOO-00A 
Determination of bum histories 
Thermal Hydraulic information for assembiy C3 
for the original forms, see spreadsheet operatinng.condiiond.xls I 368.91 0.5421 37649 kg u per cm height I I Data point 4 1 Data point 5 
I 
node mass- determined by 
heioht'mass U per cm 
MW per node; 
determined 
from C3 data 
and calculated Calculated Burnuo 
Node height in cm height height in the reactor Node I node mass burnup (GWUMTU) I MW per node Calculated burnup Bumup (GWUM1 
15.24 8.262 15.24 1 I 0.0509 1.1401 1.371 1 0.0544 2.2921 2.983 
I 15.24 8.262 368.91 10 1 0.0453 1.0137 1.219 1 0.0601 2.2372 3.001 
Averages: 1 5.4322 5.0 6.04 1 5.5872 10.0 12.89 
from dictionaw I ton 1016 kg 200kg=0.196841 Note for ~00C-MGR~002W~W0-00A: Use of 1016 kg (long ton) instead of 
Assembly power used for 1000 kg (metric tonne) is the main cause of the discrepancy in b~rn;~s 
one node calcs: 3.468 0.001 200kg=0.196841 0.1 96850394 
I I (Average 
EFPD used for calculated I EFPD from Quad Cties Ibumups assemblies 
Data point burnups shown above data need to see EFPD for one node model 
4 185 222.47 10 567.6 
5 140 245.01 20 1135.2 
Burnup calculated 
Cycle from C3 data Desired burnups 
4 6.04 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRMO200-000-WA 
--- - -- - 
Worksheet 'bumups, new lows' Attachment IV; Page IV-2 of IV-10 I 
Data point 6 Data point 7 Data point 8 Data point 9 
Calculated Bumuo Burnup 
MW 'per node Calculated bumup (GWdlMTU) 
0.0516 6.7860 6.786 
0.3372 26.7470 26.747 
0.4018 34.1000 34.1 
0.6409 35.6220 35.622 
0.4490 36.0370 36.037 
0.6995 35.7130 35.713 
0.7173 34.4620 34.462 
0.6878 31.7920 31.792 
0.7066 23.1 950 23.195 
Calculated Bumup 
MW per node bumup (GWd/MTU) 
0.0398 6.8583 7.372 
0.2638 26.9865 28.687 
0.3464 34.4144 36.647 
0.5769 35.971 1 38.45 
0.4016 36.401 6 38.99 
0.6202 36.0883 38.753 
0.6651 34.8645 37.722 
0.7220 32.2289 35.331 
0.7481 23.51 78 25.81 

Tltle: BWR Source Term Generation and Evaluation 
Document Identlfler: 000-00C-MGRO-00200-00000A 
Worksheet 'burnups, new lows' Attachment IV; Page IV-3 of IV-10 
I 
Data point 10 Data point 11 Data point 12 Data point 13 
Calculated Burnup 
MW per node burnup (GWdlMTU) 
0.0352 7.4150 7.415 
0.2274 28.8260 28.826 
0.2978 36.8290 36.829 
0.5375 38.6690 38.669 
0.4057 39.2380 39.238 
0.6258 39.0080 39.008 
0.5988 37.9660 37.966 
0.5448 35.5530 35.553 
Calculated Burnup Calculated Burnup 
MW per node burnup (GWdlMTU) MW per node burnup (GWdIMTU) 
0.0370 7.8750 7.875 0.0471 8.5100 8.51 
0.2356 30.2920 30.292 0.2978 32.3010 32.301 
0.3067 38.7380 38.738 0.3902 41.3700 41.37 
0.5488 40.9460 40.946 0.6671 43.9460 43.946 
0.4125 41.8050 41.805 0.4666 44.9520 44.952 
0.6353 41.6440 41.644 0.6878 44.7370 44.737 
0.6129 40.5090 40.509 0.6347 43.3630 43.363 
0.5531 37.8480 37.848 0.5406 40.2790 40.279 
MW per node Calculated burnup (GW~MTU) 
0.0575 9.2080 9.208 

Tlle: BWR Source Term Generation and Evaluation Worksheet 'burnu~s. oassl' Attachment IV; Page IV-4 of IV-10 . . 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Determination of bum histories 
Thermal Hydraulic information for assembly C3 
for the original forms, see spreadsheet operatinng.wndiiind.xls 
I 368.91 0.542137649 kg u per cm height 1 I Data point 4 1 Data point 5 
I 
node mass- determined bv I MW per node; 
determined 
from C3 data 
height'mass U per cm 
Node height in cm height height in the reactor Node - - 15.24 8.262 15.24 1 
30.48 16.524 45.72 2 
30.48 16.524 76.2 3 
45.72 24.787 121.92 4 
30.48 16.524 152.4 5 
45.72 24.787 198.12 6 
45.72 24.787 243.84 7 
45.72 24.787 289.56 8 
64.11 34.756 353.67 9 
15.24 8.262 368.91 10 
Averages: 
from dictionary= 1 ton 1016 kg 200kgd. 
Assembly power used for 
one node calcs: 3.468 
EFPD used for calculated 
Data point burnups shown above 
4 185 
5 140 
6 180.3 
7 80 
8 142.2 
9 15 
10 10.1 
11 102.84 
12 111.46 
13 100.33 
sum: 
0.4408 4.9346 5.934 0.4507 9.7521 12.616 
0.6189 6.9287 8.332 0.5377 12.8872 16.304 
0.9416 7.0276 8.451 0.7875 12.8988 16.235 
0.5858 6.5586 7.887 0.5240 12.3268 15.857 
0.7949 5.9333 7.135 0.7982 11.8434 15.025 
0.7096 5.2963 6.369 0.7991 10.8825 14.268 
0.6306 4.7067 5.66 0.7696 10.0067 13.267 
0.6138 3.2672 3.929 0.8060 7.1 757 9.61 1 
0.0453 1.01 37 1.219 0.0601 2.2372 3.001 
5.4322 5.0 6.04 5.5872 10.0 12.89 
a1 Note for 000-00C-MGRO-002CU000-00A: Use of 1016 kg (long ton) instead 0 
1000 kg (metric tonne) is the main cause of the discrepancy in burnups 
0.001 200kg=0.196841 0.196850394 
1 Average 
bumups 
EFPD from Quad CIies assemblies 
data need to see EFPD for one node model 
Pass 2 
" see pass 2 worksheet 
scussed in Section 6.6. 
Burnup calculated 
Cycle from C3 data Desired burnups 
4 6.04 
5 12.9 10 
6 17.8 
7 26.3 20 
8 29.6 
9 32.3 30 
10 32.5 
11 34.5 
12 36.9 
13 39.1 
4 45.14 40 
5 52 50 
6 56.9 
7 65.4 60 
8 68.7 
9 71.4 70 
10 71.6 
11 73.6 
12 76 75 
13 78.2 

Title: BWR Source Term Generation and Evaluation 
Document identiller: 000-00C-MGR&00200-ooO.00A 
Worksheet 'burnups, passl' Attachment IV; Page IV-5 of IV-10 
I 
Data point 6 Data point 7 Data point 8 Data point 9 
Bumup Calculated Burnup Burnup Calculated Bumup 
MW per node Calculated burnup (GWdMTU) MW per node bumup (GWdMTU) MW per node Calculated burnup (GWdlMTU) MW per node burnup (GWdMTU) 
0.0430 3.9220 3.922 0.0537 4.4422 5.698 0.051 6 6.7860 6.786 0.0398 6.8583 7.372 
0.3674 16.6250 16.625 0.3926 18.5256 23.845 0.3372 26.7470 26.747 0.2638 26.9865 28.687 
0.5229 22.0090 22.009 0.4694 24.2816 30.642 0.401 8 34.1000 34.1 0.3464 34.4144 36.647 
0.9334 23.0250 23.025 0.7275 25.3731 31.945 0.6409 35.6220 35.622 0.5769 35.971 1 38.45 
0.6440 22.6840 22.6M 0.5160 25.1819 32.1 73 0.4490 36.0370 36.037 0.401 6 36.4016 38.99 
0.901 1 21.5800 21.58 0.8254 24.2440 31.7 0.6995 35.7130 35.713 0.6202 36.0883 38.753 
0.7722 19.8850 19.885 0.8533 22.6391 30.347 0.71 73 34.4620 34.462 0.6651 34.8645 37.722 
0.6261 17.8210 17.821 0.8177 20.4600 27.646 0.6878 31.7920 31.792 0.7220 32.2289 35.331 
0.6169 12.8110 12.811 0.8570 14.7835 20.304 0.7066 23.1950 23.195 0.7481 23.5178 25.81 

Data point 10 1 Data point 11 
I 
Calculated Burnup 
MW per node burnup (GWUMTU) 
0.0471 8.5100 8.51 
0.2978 32.3010 32.301 
0.3902 41.3700 41.37 
0.6671 43.9460 43.946 
0.4666 44.9520 44.952 
0.6878 44.7370 44.737 
0.6347 43.3630 43.363 
0.5406 40.2790 40.279 
0.4771 29.0100 29.01 
0.0290 8.1110 8.111 
4.2380 36.9 36.89 
Data point 12 Data point 13 
Burnup 
MW per node Calculated burnup (GWUMTU) 
0.0575 9.2080 9.208 
0.3506 34.4300 34.43 
0.431 8 43.9920 43.992 
0.6732 46.6710 46.671 
0.4546 47.7120 47.712 
0.6831 47.5020 47.502 
0.661 1 46.0390 46.039 
0.5880 42.6590 42.659 
0.5363 30.5580 30.558 
- 
1 Title: BWR Source Ten Generat~on and Evaluation Worksheet 'burnups, passl' Attachment IV; Page IV-6 of IV-10 
Document Identlfler: 000-00C-MGR(M02~00A I 
I 
i 
L t 
i 
f 
f 
r - 
- 
- - 

Tltle: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-W00A 
Worksheet 'bumups, pass2' Attachment IV; Page IV-7 of IV-10 
I 
Determination of bum histories for long bums 
Thermal Hydraulic information for assembly C3 
BURNUPS FOR 2nd PASS I I Data point 4 I Data point 5 
node mass- determined by 
height'mass U per cm Calculated BU~IUD I height height in the reactor I Node 
15.2400 8.2622 15.24 I 1.0000 
I 30.4800 16.5244 45.72 2.0000 
30.4800 16.5244 76.2 3.0000 
45.7200 24.7865 121.92 4.0000 
30.4800 16.5244 152.4 5.0000 
45.7200 24.7865 198.12 6.0000 
45.7200 24.7865 243.84 7.0000 
45.7200 24.7865 289.56 8.0000 
64.1100 34.7564 353.67 9.0000 
15.2400 8.2622 368.91 10.0000 
Averages: 
from dictionary= 1 ton 1000kp 200kg=0.196841 
MW per node burnup (GW~MTU) I MW per node Calculated bumup Burnup (GWdlMTU 
0.0509 39.3034 1.371 1 0.0544 41.6224 2.983 
Assembly power used for Note for 000-00C-MGRO-00200.OM)-OoA: Use of 1016 kg (long ton) instead of 1000 kg (metric tonne) is the 
one node CACS: 3.468 0.001 the discrepancy in bumups discussed in Section 6.6. 
t t r D used for calculated 
Data point bumup ORIGINALEFPD assemblies need to see node model Cycle Bumup Desired bumups 
4 33 222.47 10 567.6 4 6.04 6.04 
5 175 245.01 20 1135.2 5 12.9 10 6.86 
6 180.3 180.3 30 1702.8 6 17.8 4.9 
7 110 303.9 40 2270.4 7 26.3 20 8.5 
8 142.2 142.2 50 2838.0 8 29.6 3.3 
9 55 121.5 60 3405.6 9 32.3 30 2.7 
10 10.1 10.1 70 3973.1 10 32.5 0.2 
11 102.84 102.84 75 4256.9 11 34.5 2 
12 65 111.46 12 36.9 2.4 
13 100.33 100.33 13 39.1 2.2 
sum: 1540.11 Pass 2 4 45.14 40 0.9 

I Data point 6 1 Data point 7 I Data point 8 I Data point 9 
I BumuD I Calculated BU~UD I Burnup I Calculated Bumup 
MW per node Calculated bumup (GWUMTU) MW per node burnup (GWCVMTU) MW per node Calculated burnup ( G w U ~ U ) MW per node bumup (GWUMTU) 
0.0430 43.0220 3.922 0.0537 43.7372 5.898 0.0516 45.8860 6.786 0.0398 46.1513 7.372 
0.3674 55.7250 16.625 0.3926 58.3384 23.845 0.3372 65.8470 26.747 0.2638 66.7252 28.687 

Tltle: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'burnups, pass2 Attachment IV; Page IV-9 of IV-10 
I 
I Data point 10 I Data point 11 I Data point 12 I Data point 13 I 
Calculated Burnup Calculated Burnup Calculated Burnup Burnup 
MW per node burnup (GWdlMTU) MW per node burnup (GWdlMTU) MW per node burnup (GWdlMTU) MW per node Calculated burnup (GWdIMTU) 
0.0352 46.5150 7.415 0.0370 46.9750 7.875 0.0471 47.3453 8.51 0.0575 9.2080 9.208 


Tills: BWR Source Term Generation and Evaluation 
Document Idenlifler: 00040C-MGRO40200000-OOA 
Attachment V: Page V-1 of V-10 I 
) s w n ~ l tme,wan model' 1 M . 2 mod.14 mod.1. 6 moda(.8 
Gamma 0 002739726 2 31 19E.16 2 3286E.16 2 3088E.16 2 32G9E11G 2 3236E.16 
10 11393E.14 11379E.14 11390E.14 1 1371Et14 11365Et14 
100000 2 3560E104 24940E.04 24540E-01 2 5740E-04 2 5940E-04 
>emw 4 -1 mod.1 2 m 4 rnw 6 model. 8 
Ganma 0 002731726 2 3189E.16 23339E.lG 2 3173E.16 2 3325E.16 2 3297E.16 
10 11410E-14 11396E.14 11411E.14 11390E-14 11385E+14 
100 1244OE-13 12359Etl3 12440E.13 12345E.13 1 2345E+13 
1000 27428~-11 26263Etll 28176E-11 26187E-11 26864E.11 
lo0000 5.7740E+09 5.5953E109 5.9044E+09 5.6482E+09 5.7091E+09 
Neutron 0.002739726 2.6020Et07 2.7520E+07 2.7430E107 2.8440E107 2.8820E+07 
!rtgntagsp ofrodsll values 
100% 101% 100% 101% 101% 
100Y0 100% 100% 100% 100% 

TlUe: BWR Source Term Generation and Evaluation 
Document Identlfler: 000-00C-MGRO-W2W4W4OA 
worksheet 'total.repon (density)' Attachment v; Page V-2 of V-10 I 

Title: BW R Source ~ e r m Generation and Evaluation Worksheet 'Den1 .%of mod1 .gl 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment V; Page V-3 of V-10 
I 
Density 1 : Percent of model 1 values for gamma sources 
-0- Model 1 +Model 2 
--+--Model4 +Model6 
+ Model 8 
0.001 0.01 0.1 1 10 100 1000 10000 100000 
Time, years 

Title: BWR Source Term Generation and Evaluation Worksheet 'den1 .%mod1 n' Attachment V: Page V-4 of V-10 
I Document Identifier: 000-00C-MGRO-00200-000-00A 
Density 1 : Percent of model 1 values for neutron sources 
I I I I I 1 
.001 0.01 0.1 1 10 100 1000 10000 100000 
Time, years 
+ Model I 
+ Model 
8 

Title: BWR Source Term Generation and Evaluation Worksheet 'Den2.%of mod1 .g' 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment V; Page V-5 of V-10 
1 
Density 2: Percent of model 1 values for gamma sources 
--o- Model 1 +Model 2 
---+--Model4 +Model6 
+Model 8 
I 
I 
\. I 
\, - 9 
I 
.001 0.01 0.1 1 10 100 1000 10000 100000 
Time, years 

Title: BWR Source Term Generation and Evaluation Worksheet 'den2.%modl n' Attachment V; Page V-6 of V-10 
Document Identifier: 000-00C-MGRO-00200-000-00A I 
Density 2: Percent of model 1 values for neutron sources 
+ Model 
2 
- - t --Model 
4 
+ Model 
6 
+I+ Model 
8 

Title: BWR Source Term Generation and Evaluation Worksheet 'Den3.%of mod1 .gl 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Density 3: Percent of model 1 values for gamma sources 
Attachment V; Page V-7 of V-10 
I 

Title: BWR Source Term Generation and Evaluation Worksheet 'den3.%modl n' Attachment V; Page V-8 of V-10 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Density 3: Percent of model 1 values for neutron sources 
+ Model 
1 
-o- Model 
2 
- - t --Model 
4 - Model 
6 
-+ Model 
8 
Time, years 

Title: BWR Source Term Generation and Evaluation Worksheet 'Den4.%of mod1 .gl 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment V; Page V-9 of V-10 
I 
Density 4: Percent of model 1 values for gamma sources 
I +Model 1 * Model 2 
1 10 100 1000 10000 100000 
Time, years 

Title: BWR Source Term Generation and Evaluation Worksheet 'den4.%mod I n' Attachment V; Page V-10 of V-10 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Density 4: Percent of model 1 values for neutron sources 
I I 

Title: BWR Source Term Generation and Evaluation Worksheet 'Sheetl' 
Document Identifier: 000-OOC-MGRO-00200-000-00A 
Attachment VI; Page VI-I of VI-3 
BWR assembly parameters to maximize surface are: 
Rod OD 1.07696 cm 
Water rod OD (rod pitch: 1.45288 
#of Rods 79 (2 Water rods) 
Rod Length 416.1536 cm 
Channel ID 13.4112 cm 
Channel OD 13.8176 
Channel Length 447.548 cm 
Water rod outside area+ inside area (estimated from rod pitch 8171.075627 
rod pitch' PI ' 4: WR inside and outside diameters consewatively approximated as fuel rod pitcl 
p a from ANF 9x9 JP-l.5 assembly 
BWR Assembly Surf. Area 168147.97 cmA2 
!,Rod surface + channel inner surface + WR's inside and outside surfaces 
Crud per unit area 
C0-60 1.25E-03 CiicmA2 
(from SAND88-1358, Ref. 7.11. p. 15: 
NRC recommended value: 125E-03 CilcmA2 
From Jones reoort. Table 2. D. 7 
. . 
Zn 65 7.30E-05 CiicmA2 
Zr 95 5.80E-05 CilcmA2 
Half Life information (Ref. 7.23, pp. 24.25, 28 
Cr 51 27.70 davs 
312.10 days 
2.73 years 
70.88 days 
44.51 days 
5.27 years 
100.00 years 
243.80 days 
0.8545 years 
2.7300 years 
0.6675 years 

Title: BWR Source Term Generation and Evaluation Worksheet 'Sheetl' Attachment VI; Page VI-2 of VI-3 I Document Identifier: 000-00C-MGRO-00200-000-00A 
Fuel Age, years since 
reactor discharge 
0 
5 
6 
7 
8 
9 
10 
11 
15 
20 
25 
30 
35 
40 
45 
50 
55 
60 
65 
70 
75 
80 
85 
90 
95 
100 
200 
Crud (Ci) for regular assembly, 
using SAND88-1358 Co 60 value 
(Ref. 7.1 1, p.15) 
210.18 
108.91 
95.49 
83.72 
73.40 
64.36 
56.43 
49.47 
29.24 
15.15 
7.85 
4.07 
2.1 1 
1.09 
0.57 
0.29 
0.15 
0.08 
0.04 
0.02 
0.01 
0.01 
0.00 
0.00 
0.00 
0.00 
0.00 
0.00 
Crud (Ci) for regular assembly, using 
NRC Co 60 values (Ref. 7.10. Table 

Title: BWR Source Term Generation and Evaluation Worksheet 'Sheetl' 
Document Identifier: 000-00C-MGRO-00200-000-OOA 
Attachment VI; Page VI-3 of VI-3 
-uel Age, years since reacto 
discharge 
0 
5 
6 
7 
8 
9 
10 
11 
15 
20 
25 
30 
35 
40 
45 
50 
55 
60 
65 
70 
75 
80 
85 
90 
95 
100 
200 
h d (Ci) for assembl 

Title: BWR Sourn Term Generauon an0 Evaluation 
Documenl Identiller. OWOOC.MGR0O0200.000-OOA 
Wolksheet 'all.bums.lOnode.~am.lolals ' Anachrnent VIII; Page VIII-1 ofVlll-10 I 
Time, years 
10 GWdiMTU 0.002739726 
C. 
olthls burn 
is Incorrect 
I inlormation); 
It Is not used 
for the comparlron 
Time, yean 
20 GWWTU 0.002739726 
5 
Time, yearn 
30 GWWTU 0.002739728 
5 
Raw Data 
6 16 30 48 60 78 96 114 139.24 145.24 
Node 1 Node 2 Node 3 Node 4 Node 5 Node 6 Node 7 Node 6 Node 9 Node10 
Raw Data 
6 18 30 48 60 78 96 114 139.24 145.24 
Node 1 Node 2 Node 3 Node 4 Node 5 Node 6 Node 7 Node 8 Node 9 Node10 
4.32Et12 3.91Et1.3 6.33E+13 7.96Et13 5.27E+13 7.59E+13 7.10E+13 6.41E+13 654E+13 4.18E+12 
Raw Data 
6 18 30 48 60 78 96 114 139.24 145.24 
Node 1 Nods 2 Nods 3 Node 4 Node 5 Node 6 Node 7 Node 8 Node 9 Node10 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'all.burns.lOnode.gam.totals ' Attachment VIII; Page Vlli-2 of Vlll-10 
I 
The, years 
40 GWdItvlTU 0.002739726 
Time, years 
50 GWdItvlTU 0.002739726 
5 
6 
7 
8 
9 
10 
11 
15 
20 
Time, years 
60 GWdItvlTU 0.002739726 
5 
6 
7 
8 
9 
10 
11 
15 
20 
Raw Data 
6 18 30 48 60 78 96 114 139.24 145.24 
Node 1 Node 2 Node 3 Node 4 Node 5 Node 6 Node 7 Node 8 Node 9 Node10 
6 
3.22E+15 
1.80E+13 
1.54E+13 
1.38E+13 
1.28E+13 
1.21E+13 
1.15E+13 
1.11E+13 
9.75E+12 
8.56E+12 
Raw 
6 
Node 1 
12 
2.18E+16 
1.49E+14 
1.30E+14 
1.17E+14 
1.09E+14 
1.02E+14 
9.74E+13 
9.33E+13 
8.14E+13 
7.11E+13 
Data 
18 
Node 2 
12 
3.17E+16 
2.01 E+14 
1.73E+14 
1.54E+14 
1.41E+14 
1.32Et14 
1.25E+14 
1.19E+14 
1.03E+14 
8.96Et13 
30 
Node 3 
18 
4.97E+16 
3.30E+14 
2.81E+14 
2.50E+14 
2.28E+14 
2.12E+14 
2.00E+14 
1.90E+14 
1.64E+14 
1.42E+l4 
48 
Node 4 
6 12 
3.44E+15 2.44E+16 
2.21Et13 1.98E+14 
1.90E+13 1.71E+14 
1.70E+13 1.53E+14 
1.57E+13 1.40E+14 
1.48E+13 1.31E+14 
1.41E+13 1.24E+14 
1.35E+13 1.18E+14 
1.19E+13 1.02E+14 
1.04E+13 8.89E+13 
Raw Data 
6 18 
Node 1 Node 2 
12 
3.13E+16 
2.71 E+14 
2.30E+14 
2.02E+14 
1.83E+14 
1 .7OE+l4 
1.59E+14 
1.51E+14 
1.29E+14 
1.12E+14 
30 
Node 3 
18 
4.68E+16 
4.35E+14 
3.67E+14 
3.22E+14 
2.91E+14 
2.69E+14 
2.51E+14 
2.38E+14 
2.02E+14 
1.74E+14 
48 
Node 4 
60 78 
Node 5 Node 6 
18 
3.97E+16 
3.31 E+14 
2.81 E+14 
2.49E+14 
2.27E+14 
2.10E+14 
1.98E+14 
l.88E+l4 
1.61 E+14 
1.39E+14 
96 
Node 7 
18 
3.50E+16 
3.00E+14 
2.56E+14 
2.27E+14 
2.08E+14 
1.93E+14 
1.82E+14 
1.73E+14 
1.49E+14 
1.29E+14 
1 I 4 
Node 8 
25.24 
3.24E+16 
2.82E+14 
2.45E+14 
2.21E+14 
2.04E+14 
1.92E+14 
1.82E+14 
1.74E+14 
1.51E+14 
1.32E+14 
139.24 
Node 9 
12 
3.12E+16 
2.95E+14 
2.49E+14 
2.18E+14 
1.97E+14 
1.82E+14 
1.70E+14 
1.60E+14 
1.36E+14 
1.17E+14 
60 
Node 5 
18 
4.73E+16 
4.33E+14 
3.65E+14 
3.21E+14 
2.90E+14 
2.67E+14 
2.50E+14 
2.36E+14 
2.01 E+14 
1.73E+14 
78 
Node 6 
18 
4.69E+16 
4.17E+14 
3.52E+14 
3.10E+14 
2.80E+14 
2.58E+14 
2.42E+14 
2.29E+14 
1.94E+14 
1.67E+14 
96 
Node 7 
18 
4.45E+16 
3.78E+14 
3.21E+14 
2.83E+14 
2.57E+14 
2.38E+14 
2.23E+14 
2.11E+14 
1.80E+14 
1.56E+14 
114 
Node 8 
25.24 
4.41 E+16 
3.56E+14 
3.07E+14 
2.76E+14 
2.53E+14 
2.37E+14 
2.24E+14 
2.13E+14 
1.84E+14 
1.60E+14 
139.24 
Node 9 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-OOO-OOA 
Worksheet 'all.burns.lOnode.gam.lo1als ' Attachment VIII; Page Vlll-3 of Vlll-10 
I 
Time, years 
70 GWdlMTU 0.002739726 
The, years 
> 
75 GWMTU 0.002739726 
Raw Data 
6 18 30 48 60 78 96 114 139.24 145.24 
Node 1 Node 2 Node 3 Node 4 Node 5 Node 6 Node 7 Node 8 Node 9 Node10 
6 
2.66E+15 
2.93E+13 
2.53E+13 
2.28E+13 
2.11E+13 
1.98E+13 
1.88E+13 
1.80E+13 
1.58E+13 
1.38E+13 
Raw 
6 
Node 1 
12 
1.67E+16 
2.70E+14 
2.32E+14 
2.06E+14 
1.88E+14 
1.75E+l4 
1.64E+14 
1.56E+14 
1.34E+14 
1.16E+14 
Data 
18 
Node 2 
12 
2.32E+16 
3.62E+14 
3.06E+14 
2.68E+14 
2.41E+14 
2.22E+14 
2.07E+14 
1.95E+14 
1.65E+14 
1.43E+14 
30 
Node 3 
18 
3.84Et16 
5.84Et14 
4.92Et14 
4.30E+14 
3.86Et14 
3.54E+14 
3.30E+14 
3.10E+14 
2.61 E+14 
2.24E+14 
48 
Node 4 
12 
2.67E+16 
4.03E+14 
3.39E+14 
2.95E+14 
2.65E+14 
2.42E+14 
2.25E+14 
2.12E+14 
1.78E+14 
1.52E+14 
60 
Node 5 
18 
4.10E+16 
6.02E+14 
5.05E+14 
4.40E+14 
3.94E+14 
3.61E+14 
3.35E+14 
3.15E+14 
2.65E+14 
2.27E+14 
78 
Node 6 
18 
4.32E+16 
5.88E+14 
4.94E+14 
4.30E+14 
3.86E+14 
3.53E+14 
3.28E+14 
3.09E+14 
2.59E+14 
2.22E+14 
96 
Node 7 
18 
4.53E+16 
5.39E+14 
4.54E+14 
3,97E+14 
3.56E+14 
3.27E+14 
3.04E+14 
2.87E+14 
2.42E+14 
2.08E+14 
114 
Node 8 
25.24 
4.41 E+16 
5.16E+14 
4.42E+14 
3.93E+14 
3.58E+14 
3.32E+14 
3.12E+14 
2.96E+14 
2.53E+14 
2.18E+14 
139.24 
Node 9 
6 
3.17E+15 
2.82E+13 
2.43E+13 
2.19E+13 
2.02E+13 
1.90E+13 
l.8OE+l3 
1.72E+13 
1.50E+13 
1.31 E+13 
145.24 
NodelO 

Tltle: BWR Source Term Generation and Evaluation 
Document Identifier: WOOOC-MGR040200M1040A 
Values p r Inch 
I 
Worksheet 'all.bums.lOnode.gam.totaIs ' Attachment VIII; Page Vllld of Vlll-10 I 
APF 
1.25E+00 
1.41 E+00 
1.41 E+00 
1.41 E+00 
1.41 Et00 
1.41 E+00 
141E+00 
1,41E+00 
1.41 E+00 
1.40E+00 
I 721E+11 3.26E+12 5.27E+12 4.42Et12 4.39Et12 422Et12 394E+12 356Et12 259E+12 697Etll 
Valves per inch 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 00040C-MGRO-002004OO-OOA 
Worksheet 'all.bums.lOnode.garn.totals ' Attachment V111; Page Vlll-5 of Vlll-10 
I 
Time, years 
40 GWdiMTU 0.002739726 
5 
6 
7 
6 
9 
10 
11 
15 
20 
Time, years 
60 GWdiMTU 0.002739726 
lues per Inch 
6 18 30 48 60 78 96 114 139.24 145.24 APF 
5.37E+14 1.82E+15 2.64+15 2.76E+15 2.65+15 2.44E+15 2.20E+15 1.94E+15 1.29E+15 4.69E+14 1.47E+00 
6 18 30 48 60 78 96 114 139.24 145.24 APF 
5.67E+14 1.87+15 2.44+15 2.56+15 2.71E+15 2.85+15 2.89E+15 2.70E+15 1.90E+15 5.74+14 1.37+00 
4.24E+12 1.97E+13 2.71E+13 2.93+13 3.01E+13 2.95E+13 2.81E+13 2.52E+13 1.70E+13 4.03+12 1.40E+00 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Time, years 
70 GWdlMTU 0.002739726 
Time, years 
75 GWdlMTU 0.002739726 
5 
6 
7 
8 
9 
10 
11 
15 
20 
lues per inch 
Worksheet 'all.burns.1Onode.gam.totals ' Attachment VIII; Page Vlll-6 of Vlll-10 
I 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-OOO-OOA 
Worksheet 'all.bums.1Onode.gam.totals ' Attachment VIII; Page Vlll-7 of Vlll-10 
I 
data for node 3 
of this burn 
is incorrect 
(see 
hlghiighted 
information); 
I it is not used 
for the com~arison 
Source ' APF 
\ 
Gamma totals 1 node source per unit 
from 1 node Source'APFl145.24, height- max ten node differencelsource per 
case 1.4 Source per unit height source per unit height unit height for 1 node Time, years 
1- 
1.70E+17 2.37E+17 1.63E+15 -5.21E+14 -32% 10 GWd/MTU 0.002739726 
5.14E+14 7.20E+14 4.96E+12 2.29E+10 0% 5 
4.39E+14 6.14E+14 4.23E+12 1 .O1 E+ll 2% 6 
3.96E+14 5.55E+14 3.82Ec12 1.38E+ll 4% 7 
3.70E+14 5.17E+14 3.56E+12 1.56E+ll 4% 8 
3.51E+14 4.91E+14 3.38E+12 1.64E+ll 5% 9 
3.37E+14 4.71 E+14 3.25E+12 1.68E+ll 5% 10 
3.25E+14 4.55E+14 3.13E+12 1.68E+ll 5% 11 
2.88E+14 4.04E+14 2.78E+12 1.60E+ll 6% 15 
2.54E+14 3.55E+14 2.44E+12 1.44E+11 6% 20 
Source ' APF 
Gamma totals 
from 1 node 
1 node source per unit 
Source'APFl145.24, height- max ten node differencelsource per 
case 1.4 Source per unit height source per unit height unit height for 1 node Time, years . - - 1.75E+17 2.45E+17 1.69E+15 -7.12+14 -42% 20 GWd/MTU 0.002739726 
Source ' APF 
Gamma totals 1 node source per unit 
from 1 node Source'APFl145.24, height- max ten node differencelsource per 
case 1.4 Source per unit height source per unit height unit height for 1 node Time, years 
1.82E+17 2.55E+17 1.75E+15 -3.69E+14 -21% 30 GWdlMTU 0.002739726 
1.50E+15 2.10E+15 1.45E+13 1.86E+ll 1 % 5 
1.31 E+15 1.83E+15 4.94E+11 4% 6 1.26E+13 
1.18E+15 1.66E+15 1.14E+13 6.62+11 6% 7 
l.lOE+15 1.54E+15 1.06E+13 7.56E+11 7% 8 
1.03E+15 1.45E+15 9.97E+12 8.08E+11 8% 9 
9.85E+14 1.38E+15 9.49E+12 8.37E+11 9% 10 
9.44E+14 t.32+15 9.09E+12 8.50E+ll 9% 11 
8.25E+14 1 .16E+15 7.95E+12 8.32E+11 10% 15 
7.21E+14 1.01E+15 6.95E+12 7.61 E+l 1 11% 20 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'all.burns.lOnode.gam.totals ' Attachment ViII; Page Vlll-8 of Vlil-10 
I 
Source APF 
Gamma totals 1 node source per unit 
from 1 node Source'APF1145.24, height- max ten node differencelsource per 
case 1.4 Source per unit height source per unit height unit height for 1 node Time, years 
1.90E+17 2.66E+17 1.83E+15 -9.32E+14 -51% 40 GWdmnTU 0.002739726 
1.99E+15 2.79E+15 1.92E+13 1.14E+ll 1 % 5 
1.74E+15 2.43E+15 1.67E+13 5.1 1 E+11 3% 6 
1.57E+15 2.19E+15 1.51E+13 7.44Et11 5% 7 
1.45E+15 2.03E+15 1.40E+13 6% 8 8.84E+11 
1.36E+15 1.90E+15 1.31 E+13 9.69E+11 7% 9 
1.29E+15 1.81E+15 1.24E+13 1.02E+12 8% 10 
1.23E+15 1.73E+15 1.19E+13 1 .04E+12 9% 11 
1.07Et15 1.50E+15 1.03Et13 1.04E+12 10% 15 
9.33Et14 1.31Et15 8.99Et12 9.58Et11 11% 20 
Source ' APF 
Gamma totais 
from 1 node 
1 node source per unit 
Source'APFl145.24. height- max ten node differencelsource per 
case 1.4 Source per unit height source per unit height unit height for 1 node 
1.98E+17 2.78E+17 1.91E+15 -7.15Et14 -37% 50 GWdmnTU 
Time, years 
0.002739726 
Source ' APF 
Gamma totals 1 node source per unit 
from 1 node Source'APF1145.24, height- max ten node differencelsource per 
case 1.4 Source per unit height source per unit height unit height for 1 node Time, years 
2.06E+17 2.89Et17 1.99E+15 -8.99E+14 -45% 60 GWdmnTU 0.002739726 
2.95Et15 4.13Et15 2.84E+13 -1.68E+12 -6% 5 
2.56E+15 3.58E+15 2.47E+13 -5.44El11 -2% 6 
2.30E+15 3.21E+15 2.21E+13 1.51 E+l 1 1% 7 
2.11E+15 2.95Et15 2.03E+13 5.88E+11 3% 8 
1.96E+15 2.75Et15 1.89E+13 8.67E+11 5% 9 
1.85E+15 2.59Et15 1.78E+13 1.05E+12 6% 10 
1.76E+15 2.47E+15 1.70E+13 1.16E+12 7% 11 
1.51Et15 2.12E+15 1.46E+13 1.30E+12 9% 15 
1.31E+15 1.84Et15 1.26Et13 1.23E+12 10% 20 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'all.burns.1Onode.gam.totals ' Attachment VIII; Page Vlll-9 of Vlll-10 I 
Source ' APF 
1 node source per 
;amma totals unit height- max ten 
from 1 node Source'APFl145.24, node source per unit differencelsource per 
case 1.4 Source per unit height height unit height for 1 node Time, years 
2.17E+17 3.04+17 2.09E+15 -4.25E+14 -20% 70 GWdlMTU 0.002739726 
Source ' APF 
1 node source per 
Gamma totals unit height- max ten 
from 1 node Source'APF1145.24, node source per unit differencelsource per 
case 1.4 Source per unit height height unit height for 1 node Time, years 
2.21+17 3.09E+17 2.13E+15 -4.02+14 -19% 75 GWdMTU 0.002739726 
3.62E+15 5.06+15 3.49E+13 -1% 5 -2.92E+11 
3.13E+15 4.38+15 3.01+13 4.96+11 2% 6 
2.79+15 3.90E+15 2.69E+13 9.86+11 4% 7 
2.55E+15 3.56E+15 2.45E+13 1.30E+12 5% 8 
2.36E+15 3.31E+15 2.28E+13 1.49E+12 7% 9 
2.22E+15 3.11E+15 2.14E+13 1.61E+12 8% 10 
2.11E+15 2.95+15 2.03+13 1.68E+12 8% 11 
1.80+15 2.52+15 1.73E+13 1.72E+12 10% 15 
1.55E+15 2.18E+15 1.50+13 1.59E+12 11% 20 


TRle: BWR Source Term Generation and Evaluation 
Document Identifier: OO(MOC-MGR(M02O(KXXMOA 
Worksheet 'all.burns. 1Onode.neut.totals' Attachment IX; Page IX-1 of IX-12 I 
RAW DATA 
Node 1 Node 2 Node 3 Node 4 Node 5 Node 6 Node 7 Node 8 Node 9 Node10 
8.22E+03 
3.36E+03 
3.39E+03 
3.41 E+03 
3.44+03 
3.46+03 
3.49E+03 
3.51 E+03 
3.58E+03 
3.66E+03 
RAW 
Node 1 
4.79+04 
1.33E+04 
1.34E+04 
1.35E+04 
1.36E+04 
1.37E+04 
1.37Et04 
1.38E+04 
1.41E+04 
1.43E+04 
DATA 
Node 2 
Time, years 6 18 - 
!O GWUMTU 0.002739726 
5 
6 
7 
is incorrect; 8 
B is not used 9 
in the final 10 
comparison 11 
to determine 15 
from one 
7.29E+04 
1.03E+04 
1.03E+04 
1.03E+04 
l.WE+04 
1.04+04 
1.04E+04 
1.04E+04 
1.05E+04 
1.05E+04 
RAW 
Node 1 
5.01 E+05 
8.09E+04 
7.95E+04 
7.82+04 
7.70E+04 
7.58E+04 
7.47Et04 
7.36E+W 
6.96E+04 
6.52E+04 
DATA 
Node 2 
Node 4 
1.29E+05 
2.93E+04 
2.92Et04 
2.92Et04 
2.91E+04 
2.91E+04 
2.90E+04 
2.90E+04 
2.88+04 
2.84E+04 
Node 5 
60 - 
1.72E+05 
4.03E+04 
4.02E+04 
4.02Et04 
4.02E+04 
4.01E+04 
4.01E+04 
4.01E+04 
3.99E+M 
3.96E+04 
Node 6 
78 
2.33+06 
5.82+05 
5.64+05 
5.48Et05 
5.32E+05 
5.1 7+05 
5.02E+05 
4.88+05 
4.36Ec05 
3.81+05 
Node 3 Node 4 
1.65E+06 
4.35E+05 
4.22E+05 
4.09E+05 
3.97Et05 
3.85E+05 
3.74E+05 
3.63+05 
3.24E45 
2.82E+05 
Node 5 
2.36E+06 
6.14E45 
5.95E+05 
5.77E+05 
5.60E+05 
5.44E+05 
5.28Et05 
5.13E+05 
4.58E+05 
4.00E+05 
Node 6 
Node 7 Node 8 Node 9 NodelO 
2.03+06 
4.96E+05 
4.82E+05 
4.68E+05 
4.55+05 
4.42+05 
4.30+05 
4.18E+05 
375E+05 
3.30+05 
Node 7 
1.56E+06 
3.41 E+05 
3.32+05 
3.24E45 
3.1 5E+05 
3.07E+05 
3.00+05 
2.92Et05 
265E+05 
2.36E+05 
Node 8 
8.24E45 
1.38E+05 
1.36+05 
1.34Et05 
1.33E+05 
1.31+05 
1.29E+05 
1.27E+05 
1.21E45 
1.15+05 
Node 9 
8.33+04 
1.22E+04 
1.22E+04 
1.22E+04 
1.22E+04 
1.21+04 
1.21E+W 
1.21 E+04 
1.20E+04 
1.19Et04 
NodelO 
Time ears 6 6 48 60 78 96 114 139.24 145.24 - 
30GWdNTU O.WZ739726 2.71E45 1.84Et06 4.18+06 8.00Et06 6.02+06 9.32E+06 8.81Et06 7.13E+06 3.75E+06 2.97E+05 4.96E47 
5 3.28E+04 3.97E+05 1.33EM 2.87+06 2.30Et06 3.65E+06 3.38+06 2.52E+06 8.97E+05 4.37Et04 1.74E47 
6 3.22E+04 3.85E+05 1.29EM 2.77Ed16 2.22+06 3.52E+06 3.26E+06 2.43E+06 8.69E+05 4.27Et04 1.68+07 
7 3.17E+04 3.74E+05 1.24Et06 2.67Et06 2.14E+06 3.39Et06 3.15E+06 2.34E+06 8.44+05 4.19+04 1.62+07 
8 3.13E+04 3.63E+05 120E+06 2.58E+06 2.07+06 3.27E+06 3.04E+06 2.26+06 8.19E+05 4.11E+04 1.57E+07 
9 3.09E+04 3.52E45 1.16+06 2.49E+06 2.00E+06 3.16E+06 2.93E+06 2.19E+06 7.95+05 4.04E+04 1.51E+07 
10 3.05E+04 3.42E+05 1.12E+06 2.41E+06 1.93Et06 3.05E+06 2.83E+06 2.11E+06 7.72E+05 3.96E+W 1.46E+07 
11 3.01E+04 3.32E+05 1.08E+06 2.32+06 1.86E+06 2.94E+06 2.73E+06 2.04E+06 7.49E+05 3.89E+04 1.41E+07 
15 2.86E+04 2.97E+05 9.47E+05 2.02+06 1.62E+06 2.56E+06 2.38E+06 1.78Et06 6.68E+05 3.63E+04 1.23+07 
20 2.70E+04 2.59E+05 8.03Et05 1.71E+06 1.36E+06 2.15E+06 2.00E+06 1.51+06 5.82E+05 3.35E+04 1.04Et07 
RAW DATA 

Title: BWR Source Term Generation and Evaluation 
Document Identlfler: 000-00C-MGRO-00200-000-WA 
Worksheet 'all.bums.1Onode.neut.totals' Attachment IX; Page IX-2 of IX-12 
I 
Node 1 Node 2 Node 3 Node 4 Node 5 Node 6 Node 7 Node 8 Node 9 Node10 
40 GWUMTU 0.002739726 
5 
6 
7 
8 
9 
10 
11 
15 
20 
5.06E+06 
1.40E+06 
1.35E+06 
1.30E+06 
1.26E+06 
1.21 E+06 
1.17E+06 
1.13E+06 
9.90E45 
8.39E45 
DATA 
Node 2 Node 3 Node 4 Node 5 Node 6 Node 7 Node 8 Node 9 NodelO 
1.25E+06 
2.42E+05 
2.34E+05 
2.26E+05 
2.1 9E+05 
2.13E+05 
2.06E+05 
2.00E+05 
1.77E45 
1.53E45 
RAW 
Node 1 
2.86E+07 
1.57E+07 
1.51E+07 
1.46E47 
1.40E+07 
1.35E47 
1.30E+07 
1.25E+07 
1.08E+07 
8.99E+06 
Node 3 
5.38E+07 
3.15E+07 
3.03E+07 
2.92E+07 
2.81E47 
2.70E+07 
2.60E+07 
2.51E+07 
2.16E+07 
1.79E+07 
Node 4 
3.91E+07 
2.32E+07 
2.23E+07 
2.15E+07 
2.07E47 
1.99E+07 
1.92E+07 
1.85E+07 
1.59E+07 
1.32E+07 
Node 5 
5.84E+07 
3.44E+07 
3.31 E+07 
3.18E+07 
3.06E+07 
2.95E+07 
2.84E+07 
2.73E+07 
2.35E+07 
1.95E+07 
Node 6 
5.35E+07 
3.06E+07 
2.95E+07 
2.83E+07 
2.73E+07 
2.63E+07 
2.53E47 
2.44E+07 
2.10E+07 
1.74E+07 
Node 7 
4.27E+07 
2.29E+07 
2.21E47 
2.13E+07 
2.05E+07 
1.97E+07 
1.90E+07 
1.83E+07 
1.58E+07 
1.31 E+07 
Node 8 
2.13E47 
7.88E+06 
7.60E+06 
7.33E+06 
7.07E+06 
6.82E+06 
6.58E+06 
6.34E+06 
5.50E+06 
4.61 E+06 
Node 9 
1.36E+06 
2.79E+05 
2.70E+05 
2.61 E+05 
2.53E+05 
2.45E+05 
2.37E+05 
2.30E+05 
2.03E+05 
1.74E+05 
NodelO 

Tnle: BWR Source Term Generation and Evaluati~ 
Document Identifier: 00QOOC-MGRW200MM-M)A 
Worksheet 'all.bums.lOnode.neut.tdals' Attachment IX; Page IX-3 Of IX-12 I 
Node 1 Node 2 Node 3 Node 4 Node 5 Node 6 Node 7 Node 8 Node 9 Node10 
Time, years 6 18 30 48 60 78 96 139.24 
L 
70GWd/MTU 0.002739726 2.67E+06 3.36+07 9.32E+07 1.89+08 1.46E+08 2.24E+08 2.01E+08 1.48E+08 6.46+07 2.65E+06 l.llE+09 
5 8.74E+05 2.05E+07 6.36E+07 1.25E+08 9.30E+07 1.40E+08 1.26E+08 9.37E+07 3.56E+07 9.92E+05 6.98E+08 
6 8.43E+05 1.97+07 6.08+07 1.18+08 8.79+07 1.32E+08 1.19+08 8.91E+07 3.42E+07 9.56E45 6.62E+08 
7 8.13E+05 1.90E+07 5.82+07 1.13+08 6.33+07 1.25E+08 1.13+08 6.50E+07 3.29E+07 9.22E45 6.30E+08 
8 7.85E+05 1.83E+07 5.57+07 1.08+08 7.93+07 1.19E+08 1.07+08 8.13E+07 3.17E+07 8.90E+05 6.01E+08 
9 7.57E45 1.76+07 5.35E47 1.03E+08 7.56+07 1.13E+08 1.02+08 7.76E+07 3.05E+07 8.59E+05 5.75E+08 
10 7.31E45 1.69+07 5.13E+07 9.65E+07 7.22+07 1.08E+08 9.76+07 7.46E+07 2.94+07 8.29E+05 5.50E+08 
11 7.06E+05 1.63+07 4.93E+07 9.45+07 6.91+07 1.03+08 9.34+07 7.15E+07 2.83E+07 8.00E+05 5.27E+08 
15 6.13E+05 1.41+07 4.22E+07 8.05E+07 5.87+07 8.76E47 7.93E+07 6.11E47 2.44+07 6.94E+05 4.49+08 
20 5.16E+05 1.17+07 3.50+07 6.66+07 4.65E+07 7.23E+07 6.55+07 5.06E+07 2.03+07 5.83E+05 3.72+08 
RAW DATA 
Node 1 Node 2 Node 3 Node 4 Node 5 Node 6 Node 7 Node 8 Node 9 Node10 
Time, years 6 18 30 48 60 78 96 114 139.24 145.24 

Tills: BWR Swrce Term Generatim and Evaluation 
D o s u m Identuler: CC-OX-MGRDW20DWDWA 
Time, yean 8 12 12 18 12 18 - 10 
GWdiMTU 0.002739726 1.37E.w 399E43 9.87E103 1.12E+Cd 1.07E4 95BE+03 
5 S.BOE+M l . l l E 4 3 2.21E+W 2.51E+03 2.44E+03 224E+03 
6 565E& l . l l E 4 3 220E+03 2.ME+03 2.44E+W 2.23E+W 
7 56QE+02 112E43 220E+03 2.49E+03 243E+03 223E+03 
8 5.73E+02 1.13E43 2.20E.w 2.49EtW 2.43E+03 2.23Et03 
Time, years 
20 
GWUMN O.OUZ3972f 
5 
info 
is inmnm: 
Time, yean 
30 
GWmTU O.OUZ39728 
5 
as p r inch 

Tnle: BWR Source Term Generation and Evaluation Wortsheet 'all.bums.1Onode.neut.totals' Attachment IX; Page IX-5 of IX-12 
I Document Identlfler: 0W-WC-MGRMXnwMXHX)A 
Time, years 
40 
GWdlMTU 0.002739726 
5 
6 
7 
6 
9 
10 
11 
15 
20 
6 
1.18E+05 
1.69E+04 
1.65E4 
1.60E+04 
1.56E+04 
l.WE+04 
1.48E4 
1.44E+04 
1.31E+W 
1.17E+04 
6 
data per inch 

Tnle: BWR Source Term Generation and Evaluation 
Document Identlfler: MM-OOC-MGRO0020000(MOA 
Worksheet 'all.bums.1Onode.neut.totals' Attachment IX; Page IX-6 of IX-12 
I 
Time, years I 6 
70 I 
GWdMTU 
Time, years 
75 
GWdMTU 0.002739726 
5 
6 
7 
8 
9 
10 
11 
15 
20 
6 
data per inch 
6 - 
5.1 1E+05 
1.91 E+05 
1.84E+05 
1.76E+05 
1.72E+05 
1.65E+05 
1.60E+05 
1.54E+05 
1.34E+05 
1.12E+05 
6 APF 
I 

Tltle: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'all.bums.lOnode.neut.totals' Attachment IX; Page IX-7 ot IX-12 I 
6.97E+03 
eutron totals from 1 
node case 1.400 145.24 Diierence % diierence Time, years 
10 
NUtt: 
node 3 
info 
is incorrect; 
R is not used 
in the final 
comparison 
to determine 
% difference 
from me nodl 
7.70E+04 
eutron totals from 1 
node case 1.400 145.24 Difference % diierence 
2.93E45 
eutron totals from 1 
node case 1.400 145.24 Diierence 90 diierence Time, years 
30 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-0OOM)A 
Worksheet 'all.bums.lOnode.neut.totals' Anachment IX; Page IX-8 of IX-12 
I 
7.44E+05 
Neutron totals from 1 
node case 1.400 145.24 Difference % difference Time, years 
40 
GWCUMTU 0.002739726 
5 
6 
7 
8 
9 
10 
11 
15 
20 
1.55E+06 
Neutron totals from 1 
node case 1.400 145.24 Difference % difference Time, years 
50 
2.25+08 3.14E+08 2.16E+06 -1.09E+06 -50% GWCUMTU 0.002739726 
1.07E+08 1.50E+08 1.03E+08 -9.05E+05 -88% 5 
1 .03E+08 1.44E+08 9.91E+05 -8.69E45 -88% 6 
9.91E+07 1.39E+08 9.55E+05 -8.35+05 -87% 7 
9.55E47 1.34E+08 9.20E+05 -8.02E+05 -87% 8 
9.20E+07 1.29E+08 8.87E45 -7.72E45 -87% 9 
8.87E47 1.24E+08 8.55E+05 -7.42E+05 -87% 10 
8.55E47 1.20E+08 8.24E45 -7.14E+05 -87% 11 
7.38E47 1.03E+08 7.1 1E+05 -6.12E+05 -86% 15 
6.15E47 8.61E+07 5.93E+05 -5.07E45 -86% 20 
2.85E+OB 
Neutron totals from 1 
node case 1.400 145.24 Merence %difference Time, years 
60 

Worksheet 'all.burns.lOnode.neut.totals' 
4.88E+06 
Neutron totals from 1 
node case 1.400 145.24 Difference % difference Time. vears 
70 
Attachment IX; Page IX-9 of IX-12 1 
Title: BWR Source Term Generation and Evaluation 
Document Identifier: 00040C-MGRO-00200400-00A 
6.30E+06 
Neutron totals from 1 
node case 1.400 145.24 Difference % difference Time. years 
75 




Tltle: BWR Source Ten Generation and Evaluation 
Document Identltler: WO-00C-MGRO-O02OMXX)-WA 
worksheet 'Source.compare' Anachment X; Page X-1 of X-6 I 
ierrnal design basis fuel after 11 yean of cooling time. 
Neutrons 
normalized normalized 
raw. Normalized. Normalized. swrcdoer sourcdoer 
Min Max Mean 3.74%.11yrs 3.74%.170kg.llyrs 200/170 2Wkg 170 kg mtu, 200 kg mu, 170 kg 
6.43E+W 2.WE+01 1.32E+01 2.71E+06 4.19E+06 65% 1 B5E-02 1.86E-02 9.25E-02 1.09E-01 
O.WE+W 
1.WE-11'. 1.WE-08 5.01E-09 0.00E+W O.WE+W O.WE+W O.WE+W O.WE+W O.WE+W 
1.47E+08 2.26E+08 65% l.WE+W l.WE+W 
Neutrons: Thermal DBF at 31 1 years I 
normalized normalized Percentage of 
raw, swTce!per swrcel~er 170kg loading 
Min Max Mean 3.74%.170kg.311yrs 3.74%.2Wkg.311yrs 200/170 mlu, 170 kg mu. 200 kg vilue 
1.32E+01 7.64E+04 4.59E+04 60% 9.53E-02 7.14E-02 75% 
9.77E+ffi 6.62E+05 68% 1.22E+00 1.03E+00 84% 

Title: BWR Source Term Generation and Evaluation 
Document Idenlifler: 000-00C-MGRMXn~00(MOA 
worksheet 'Source.compare' Attachment X; Page X-2 of X-6 I 
Gamma sources r 
% Of 
170kg 
loading 3.74%.170kg.49 3.74%.49GWd. 
value mean energy GWd.1 l y 1~ 
85% 0.03 3.82E+14 3.93E+14 
Raw 
%of 170kg 
loading value 
103% 
105% 
104% 
104% 
105% 
92% 
99% . 
94% 
95% 
87% 
103% 
97% 
91% 
83% 
65% 
65% 
65% 
65% 
101% 
Normalized per mtu 
% of 170ko loadino 
loading value, 
normalized 
1.74%.170kg.4 3.74%.49GWd.1 sources per 
1.26E-08 6.88E-09 
2.47E-09 1.35E-09 
5.24E-10 2.86E-10 
Raw % of - - 
3.74%. 170ko.49 3.74%.49GWd.3 3.74Ye.170ka.49G value (normalied 170ka loadina - - 
mean energy GWd.316 1 1 ~ ~ d . 3 1 iy 3.74%.49GWd.311 y sources per mu) value 
0.03 3.06+12 3.50+12 l.WE+M) 1.43E+M) 85% 114% 
7.50E-02 5.90E+12 7.09E+12 3.25E+00 2.90E+00 89% 120% 


Title: BWR Source Term Generation and Evaluation Worksheet 'Neutron.sources.graph' 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment X; Page X-4 of X-6 
1 
0.00 2.00 4.00 6.00 8.00 10.00 12.00 14.00 
Mean Neutron Energy (MeV) 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-0OOdOA 
neutron REWHR 
Per source particle 
surface 1 
segment: 
energy 
2.50E-08 
1.00E-07 
1.00E-06 
1.00E-05 
1.00E-04 
1.00E-03 
1.00E-02 
l.OOE-O1 
total 
Gammas for flo 
surface 1 
segment: 
energy 
2.00E-02 
3.00E-02 
5.00E-02 
7.00E-02 
l.OOE-O1 
1.50E-01 
3.00E-01 
4.50E-01 
7.00E-01 
1.00E+00 
1.50E40 
2.00E40 
2.50E40 
3.00E40 
4.00E40 
6.00E40 
8.00E+00 
1.40E+01 
total 
dose 
0.00E+00 
0.00E40 
1.70E-15 
3.82E-14 
2.91E-13 
2.83E-12 
1.65E-11 
1.77E-10 
9.12E-10 
9.08E-10 
6.40E-10 
2.15E-10 
5.63E-11 
5.62E-12 
4.03E-12 
6.52E-12 
2.94E-09 
-19 
3.57E42 
2.29E+01 
3.81 E+01 
9.29E+01 
9.12E42 
7.59E42 
7.44E+03 
1.07E44 
5.33E44 
5.68E44 
2.10E+04 
3.45E43 
2.25E+02 
2.34E4l 
0.00E40 
0.00E+00 
O.OOE+M) 
0.00E40 
1.55E45 
fuel region 
error 
0 
0 
0.2542 
0.0816 
0.0347 
0.0153 
0.0076 
0.0052 
0.0044 
0.0051 
0.006 
0.0093 
0.0201 
0.0527 
0.0811 
0.0776 
0.0034 
surface 3 
segment: 
energy 
2.50E-08 
1.00E-07 
1.00E-06 
1.00E-05 
1.00E-04 
1.00E-03 
1.00E-02 
l.OOE-O1 
5.00E-01 
1.00E+00 
2.50E+00 
5.00E+00 
7.00E+00 
1.00E+01 
1.40E+01 
2.00E41 
total 
surface 3 
segment: 
energy 
2.00E-02 
3.00E-02 
5.WE-02 
7.00E-02 
1.00E-01 
1 SOE-01 
3.00E-01 
4.50E-01 
7.00E-01 
1.00E+00 
1.50E40 
2.00E40 
2.50E40 
3.00E40 
4.00E40 
6.00E40 
8.00E40 
1.40E41 
total 
Altachment X: Page X-5 of X-6 
I 
fuel region 
dose error 
Worksheet 'PW R.shielding.results' 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-OA 
Worksheet 'irnpurity.graphl Attachment X; Page X-6 of X-6 
BWR Fuel Impurity calculations for CI-36 an( 
RAW DATA I 
Burnup C136 (5.3 pprn) C14 
1 6.61 E-05 3.03E-03 
10 6.74E-04 3.1 1 E-02 
20 1.44E-03 6.70E-02 
30 2.33E-03 1.1 1 E-01 
40 3.38E-03 1.64E-01 
50 4.57E-03 2.28E-01 
60 5.82E-03 3.00E-01 
70 7.03E-03 3.76E-01 
75 7.62E-03 4.14E-01 
J C-14 
Burnup 
(GWdMTU) 
1 
10 
20 
30 
40 
50 
60 
70 
75 
curies per pprn 
C136 5.3 pprn C14 89.4 pprn 
CL36 (5.3pprn) valuel5.3 C-14 (Cilassembly) valuel89.4 
6.61 E-05 1.25E-05 3.03E-03 3.39E-05 
6.74E-04 1.27E-04 3.1 1 E-02 3.48E-04 
I .44E-03 2.72E-04 6.70E-02 7.49E-04 
2.33E-03 4.40E-04 1.1 1E-01 1.24E-03 
3.38E-03 6.38E-04 1.64E-01 1.83E-03 
4.57E-03 8.62E-04 2.28E-01 2.55E-03 
5.82E-03 1.1 0E-03 3.00E-01 3.36E-03 
7.03E-03 1.33E-03 3.76E-01 4.21 E-03 
7.62E-03 1.44E-03 4.1 4E-01 4.63E-03 
I Burnup (GWdIMTU) 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A ATTACHMENT XI, Page: XI-1 of XI-1 I 
Script File: gammas 
BEGIN {insas=O && intable=O} 
/halt feature/{insas=l) 
/gamma source spectrum/ {if (insas) intable=l; print $0) 
/ to / {if (insas && intable) print $0) 
/totals/ {intable=O} 
Script File: neutrons 
BEGIN { intable=O} 
/alpha-n plus/ {intable=l; print $0) 
/yr/ {if (intable) print $0) 
/E/ {if (intable) print $0) 
/ gamma sources determined / {incable=Ol 
Script File: curies 
BEGIN {intable=O && insas=O ) 
/halt feature/ {insas=l) 
/nuclide iadioactivity/{if (insas) print $0; intablk=l) 
/initial/ {if (insas && intable) print $0) 
/charge/ {if (insas && intable) print $0) 
/E/ {if (insas && intable) print $01 
/total/ {intable=O}' 
Script File: watts 
BEGIN {intable=O && insas=O 1 
/halt feature/ {insas=l) 
/nuclide thermal power, watts/{if (insas) print $0; intable=l) 
/charge/ {if (insas && intable) print $0) 
/initial/ {if (insas && intable) print $0) 
/E/ {if (insas && intable) print $0) 
/total/ {intable=O} 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Paae: XII-1 of XII-16 I 
List of Files on Compact Discs I 
This attachment lists the all the files on Attachment VII, including the '*.cutt files produced by 
the last revision of this calculation. The '*.cut7 files are the sections of the SAS2WORIGEN-S 
output files that contain the input echoes and the final ORIGEN-S output. Other intermediate 
calculations and information generated by SAS2H and included in the output files have been 
removed. In the '*.cutt file name, the initial enrichment of the fuel appears first, followed by the 
'%' symbol. The files listed as 'BWR.max.2.cutf, 'BWR.min.2.cutq, 'BWR.max.3.cut', and 
'BWR.min.3.cut' are used to generate radionuclide inventories for specific years (see Section 
5.5). The '*.2.cut' files generate the information for the years up to and including 12033. The 
'*.3.cutt files generate the information for the years 12033 to 1002033. The files listed as 
'bwr-imp-BU#.cutV and 'bwr-cl2_BU#.cut' are the SAS2WORIGEN-S output files used to 
demonstrate the effect of fuel impurities on the source terms. The 'BUW indicates the burnup for 
that particular case, with "BU4" being the lowest calculated burnup, and "BU13" being the 
highest. The files 'outbwr-imp-5.5-75.0utt and 'outbwr~imp~bound.out' demonstrate the effect 
of enrichment on activity from fuel impurities. 
Source term information from the '*.cut' and '*.output7 files of the SAS2WORIGEN-S runs in 
the folders "Impurity-study," "parameter~study~output," "ten-node-output," and 
'THERMAL-DBF-output"on compact disc one should not be used for assemblies expected to 
be received at a potential repository. Those cases were run solely to form the basis for decidlng 
what kinds of cases to be run for BWR assemblies in an expected waste stream. Furthermore, 
the '*.cut7 files in folder "ten-node-output" and the '*.output' files in folder 
"parameter~study~output7' on compact disc one cannot be considered product output in 
accordance with AP-3.15Q, Managing Technical Product Inputs (Ref. 7.33), since some 
information (Table 8 and Table 17) for those runs was taken from Ref. 7.13, pp. 333-334 and 
Ref. 7.17, pp. 50-55, which are "reference only." Therefore, the information in the '*.cut7 files 
of folder "ten node output" and the '*.outvut7 files of folder "parameter study output77 on 
compact disc one should not be used for any other calculations or analyses. 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Page: XII-2 of XII-16 
CD 1 
File Size (bytes) Date of Transfer Time of Transfer FILE NAME 
Folder: cut-files-for-0.7%-and-1 % 
2,180,169 1 0131 12002 1 1 :37a 0.7%.1OOMWd.bottom.cut 
2,366,251 912011 999 07:04a 0.7%.100MWd.fuel.cut 
2,220,219 912011 999 07: 1 6a 0.794.1 00MWd.plenum.cut 
2,198,825 912011 999 07: 1 6a 0.7/o.100MWd.top.cut 
2,492,014 912011 999 07:02a 0.7%.1OGWd.bottom.cut 
2,726,172 912011 999 06:21 a 0.7%.1OGWd.fuel.cut 
2,533,942 912011 999 06:59a 0.7%.1OGWd.plenum.cut 
2,509,846 912011 999 07:OOa 0.7%.1OGWd.top.cut 
2,058,921 912011 999 07:24a 0.7%.1OMWd.bottom.cut 
2,235,687 912011 999 07: 1 5a 0.7%.1OMWd.fuel.cut 
2,084,893 912011 999 07:23a 0.7/o.10MWd.plenum.cut 
2,075,493 912011 999 07:23a 0.7%. 1OMWd.top.cut 
2,320,930 912011 999 07: 1 3a 0.7%. 1 GWd.bottom.cut 
2,519,406 912011 999 06:59a 0.7%.1 GWd.fuel.cut 
2,355,206 912011 999 07:04a 0.7%.1 GWd.plenum.cut 
2,342,904 912011 999 07: 1 2a 0.796.1 GWd.top.cut 
1,983,474 912011 999 07:27a 0.7%.1 MWd.bottom.cut 
2,145,862 912011 999 07:20a 0.7%.1 MWd.fuel.cut 
2,013,606 912011 999 07:26a 0.7%.1 MWd.plenum.cut 
2,002,074 912011 999 07:26a 0.7%.1 MWd.top.cut 
2,586,845 912011 999 06:57a 0.7%.20GWd.bottom.cut 
2,826,529 912011 999 06:Ol a 0.7%.20GWd.fuel.cut 
2,635,887 912011 999 06:33a 0.7%.20GWd.plenum.cut 
2,606,349 912011 999 06:55a 0.7%.20GWd.top.cut 
2,627,525 912011 999 06:34a 0.7%.30GWd.bottom.cut 
2,866,187 912011 999 06:OOa 0.7%.30GWd.fuel.cut 
2,672,409 912011 999 06:29a 0.7/o.30GWd.plenum.cut 
2,647,417 912011 999 06:32a 0.7/o.30GWd.top.cut 
2,656,978 912011 999 06:31 a 0.7%.40GWd.bottom.cut 
2,896,946 912011 999 05:59a 0.7%.40GWd.fuel.cut 
2,703,168 91201 1 999 06:25a 0.7%.40GWd.plenum.cut 
2,680,574 912011 999 06:29a 0.7%.40GWd.top.cut 
2,684,838 912011 999 06:28a 0.7%.50GWd.bottom.cut 
2,930,750 912011 999 05:57a 0.7%.50GWd.fuel.cut 
2,730,618 912011999 06:09a 0.7%.50GWd.plenum.cut 
2,707,656 912011 999 06:24a 0.7%.50GWd.top.cut 
2,703,487 912011 999 06:24a 0.7%.60GWd.bottom.cut 
2,953,347 912011 999 05:56a 0.7/o.60GWd.fuel.cut 
2,751,385 912011 999 06:06a 0.7%.60GWd.plenum.cut 
2,728,423 912011 999 06:20a 0.7%.60GWd.top.cut 
2,721,846 9120/1999 06:22a 0.7%.70GWd.bottom.cut 
2,977,790 912011 999 05:55a 0.7%.70GWd.fuel.cut 
2,766,354 912011 999 06:05a 0.7%.70GWd.plenum.cut 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Page: XII-3 of XII-16 
File Size (bytes) Date of Transfer Time of Transfer FILE NAME 
2,743,784 912011 999 06:08a 0.7%.70GWd.top.cut 
2,733,145 912011 999 06:09a 0.7%.75GWd.bottom.cut 
2,986,391 912011 999 05:54a 0.7%.75GWd.fuel.cut 
2,779,831 912011 999 06:03a 0.7%.75GWd.plenum.cut 
2,754,511 3 912011 999 06:06a 0.7%.75GWd.top.cut 
2,173,888 912011 999 07: 1 8a 1.0%.100MWd.bottom.cut 
2,360,922 912011 999 07:04a 1.0%.100MWd.fuel.cut 
2,215,118 912011 999 07:16a 1.0%.100MWd.plenum.cut 
2,191,090 912011 999 07: 1 7a 1.0%.100MWd.top.cut 
2,487,896 912011 999 07:02a 1.0%.1OGWd.bottom.cut 
2,718,438 912011 999 06:22a 1 .O%.lOGWd.fuel.cut 
2,524,210 912011 999 06:59a 1.0%.1OGWd.plenum.cut 
2,505,346 912011 999 07:01 a 1 .O%.lOGWd.top.cut 
2,054,416 912011 999 07:25a 1 .0%.1OMWd.bottom.cut 
2,231,286 912011 999 07: 1 6a 1 .0%.1OMWd.fuel.cut 
2,080,274 912011 999 07:23a 1.0%.1OMWd.plenum.cut 
2,071,122 912011 999 07:24a 1 .O%.lOMWd.top.cut 
2,317,422 912011 999 07: 1 4a 1.0%.1 GWd.bottom.cut 
2,515,020 912011 999 07:OOa 1 .O%.1 GWd.fuel.cut 
2,352,338 912011 999 07:04a 1.0%.1 GWd.plenum.cut 
2,339,511 0 912011 999 07: 1 2a 1 .O%.1 GWd.top.cut 
1,985,211 912011 999 07:27a 1.0%.1 MWd.bottom.cut 
2,145,863 912011 999 07:20a 1.0%.1 MWd.fuel.cut 
2,014,227 912011 999 07:26a 1.0%.1 MWd.plenum.cut 
2,001,951 912011 999 07:26a 1 .O%.1 MWd.top.cut 
Folder: Impurity-study 
232,600 1111 111999 03:02a bwr-imp-BU5.cut 
245,543 1111 111999 03:02a bwr-imp-BU6.cut 
255,849 1111 111999 03:02a bwr-imp-BU7.cut 
266,109 1111 111999 03:02a bwr-imp-BU8.cut 
275,063 1111 111999 03:02a bwr-imp-BU1 O.cut 
285,312 1111 111999 03:02a bwr-imp-Bull .cut 
291,998 1111 111999 -- - 03:02a bwr-imp-BU12.cut - -- 
298,025 1111 111999 03:02a bwr-imp-BU13.cut 
206,934 1111 111999 03:02a bwr-imp-BU4.cut 
48,221,732 1111 111999 03:05a outpbwr~imp~bound.out 
48,082,956 1111 111999 03:09a outpbwr-imp-5.5-75.out 
Folder: parameter-study-output 
10,570,169 a2411 999 02:01 a 1 .den1 .output 
10,589,423 a2411 999 02:01 a 1 .den2.output 
10,464,603 212411 999 02:02a 1 .den3.output 
10,563,627 21241 1 999 02:01 a 1 .den4.output 
10,660,383 212411 999 02:01 a 2.denl .output 
10,684,741 212411 999 02:OOa 2.den2.output 
10,558,555 a2411 999 02:02a 2.den3.output 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII, Page: XI14 of XII- 16 
File Size (bytes) Date of Transfer Time of Transfer FILE NAME 
10,660,514 212411 999 02:01 a 2.den4.output 
11,580,632 212411 999 01 :59a 4.denl .output 
11,604,374 2/24/1999 01 :58a 4.den2.output 
1 1,487,893 212411 999 02:OOa 4.den3.output 
11,581,115 2/24/1999 01 :59a 4.den4.output 
1 1,582,053 a2411 999 01 :59a 6.denl .output 
11,605,046 a2411 999 01 :58a 6.den2.output 
1 1,487,183 212411 999 02:OOa 6.den3.output 
1 1,582,654 212411 999 01 :59a 6.den4.output 
1 1,510,874 212411 999 02:OOa 8.denl .output 
1 1,532,190 212411 999 01 :59a 8.den2.output 
1 1,416,229 212411 999 02:OOa 8.den3.output 
11,511,241 212411 999 02:OOa 8.den4.output 
Folder: Performance~Assessment 
174,898 1 1 1411 999 12:49a BW R.ave.3.cut 
233,861 111411 999 12:49a BWR.ave.2.cut 
199,861 1 1 1411 999 12:49a BW R.max.3.cut 
274,606 1 1 1411 999 12:49a BW R.max.2.cut 
855,883 212411 999 01 :52a node.10.50GWd.cut 
865,551 a2411 999 01 :52a node.lO.6OGWd.cut 
864,541 a2411 999 01 :52a node.lO.75GWd.cut 
875,146 a2411 999 01 :52a node.10.70GWd.cut 
786,499 2/24/1999 01 :52a node.lO.1OGWd.cut 
802,903 212411 999 01 :52a node.lO.20GWd.cut 
81 9,694 a2411 999 0 1 :52a node. 10.30GWd.cut 
844,864 a2411 999 01 :52a node.lO.4OGWd.cut 
1,520,079 212411 999 01 :50a node.1 .l 0GWd.cut 
794,333 2/24/1 999 01 :51 a node.l.20GWd.cut 
81 7,038 - p~ 
a2411 999 01 :51 a node.l.30GWd.cut 
843,162 2/24/1999 01 :51 a node.l.40GWd.cut 
851,435 a2411 999 01 :51 a node.l.5OGWd.cut 
859,800 a2411 999 01 :51 a node.1 .GOGWd.cut 
867,272 a2411 999 01 :51 a node.l.70GWd.cut 
856,667 a2411 999 01 :51 a node.l.75GWd.cut 
847,636 212411 999 01:51a node.2.1OGWd.cut - 
869,935 a2411 999 01 :51 a node.2.20GWd.cut 
889,794 a2411 999 01 :51 a node.2.30GWd.cut 
915,762 2/24/1999 01 :51 a node.2.40GWd.cut 
928,724 a2411 999 01 :51 a node.2.50GWd.cut 
941,769 212411 999 01 :51 a node.2.60GWd.cut 
950,690 212411 999 01 :51 a node.2.70GWd.cut 
942,269 212411 999 01 :51 a node.2.75GWd.cut 
854,033 212411 999 01 :52a node.3.1OGWd.cut 
888,734 a2411 999 01 :52a node.3.20GWd.cut 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-OOA Attachment XII, Page: XII-5 of XII-16 1 
/ ~ i l e Size (bytes) Date of Transfer Time of Transfer FILE NAME I 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-OOA Attachment XII. Page: XII-6 of XII-16 
CD 1 
File Size (bytes) 
963,098 
858,462 
886,150 
907,182 
933,402 
944,360 
959,017 
973,525 
962,434 
Time of Transfer 
01 :55a 
01 :55a 
01 :55a 
01 :55a 
01 :55a 
01 :55a 
01 :55a 
01 :55a 
01 :55a 
Date of Transfer 
212411 999 
212411 999 
212411 999 
212411 999 
a2411 999 
2/24/1 999 
a2411 999 
a2411 999 
a2411 999 
FILE NAME 
node.8.75GWd.cut 
node.9.1OGWd.cut 
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node.9.30GWd.cut 
node.9.40GWd.cut 
node.9.50GWd.cut 
node.9.60GWd.cut 
node.9.70GWd.cut 
node.9.75GWd.cut 
Folder: 
2,328,690 
2,142,897 
2,122,146 
2,121,402 
2,179,387 
2,232,863 
2,294,945 
2,178,921 
2,231,407 
2,287,590 
2,287,590 
2,287,590 
2,316,609 
2,345,584 
2,355,307 
1,986,076 
2,042,577 
2,091,463 
2,122,067 
2,349,235 
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2,294,945 - 
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2,294,945 
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Folder: UNIX-script 
02:08a 
02:09a 
02:07a 
02:08a 
02:08a 
02:08a 
02:08a 
02:07a 
02:07a 
02:07a 
06:30a 
02:07a 
02:07a 
02:07a 
02:07a 
02:09a 
02:09a 
02:09a 
02:09a 
02:08a 
02:08a 
02:09a 
06:30a 
02:09a 
02:09a 
02:08a 
06:31 a 
THERMAL-DBF-output 
21241 1 999 
212411 999 
2/24/1999 
2/24/1999 
212411 999 
212411 999 
212411 999 
212411 999 
212411 999 
212411 999 
2/24/1999 
a2411 999 
212411 999 
2/24/1999 
2/24/1999 
212411 999 
212411 999 
212411 999 
212411 999 
a2411 999 
a2411 999 
a2411 999 
a2411 999 -- 
2/24/1 999- 
212411 999 
212411 999 
212411 999 
3.74%.170kg.60GWd.fuel.cut 
3.74%.no, hardware.50GWd.fuel.cut 
3.74%.1OGWd.fuel.cut 
3.74%.170kg.lOGWd.fuel.cut 
3.74%.170kg.20GWd.fuel.cut 
3.74%.170kg.30GWd.fuel.cut 
3.74%. 170kg.40GWd.fuel.cut 
3.74%.20GWd.fuel.cut 
3.74%.30GWd.fuel.cut 
3.74%.40GWd.fuel.cut 
3.74%.49GWd.fuel.cut 
3.74%.50GWd.fuel.cut 
3.74%.60GWd.fuel.cut 
3.74%.70GWd.fuel.cut 
3.74%.75GWd.fuel.cut 
3.74%.no.hardware.lOGWd.fuel.cut 
3.74%.no.hardware.20GWd.fuel.cut 
3.74%.no.hardware.30GWd.fuel.cut 
3.74%.no.hardware.40GWd.fuel.cut 
3.74%. 170kg.70GWd.fuel.cut 
3.74%.170kg.75GWd.fuel.cut 
3.74%.no. hardware.60GWd.fuel.cut 
3.74%.170kg.49GWd.fuei.cut 
3.74%.no.hardware.70GWd.fuel.cut 
3.74%.no.hardware.75GWd.fuel.cut 
3.74%.170kg.50GWd.fuel.cut 
3.74%.no.hardware.49GWd.fuel.cut 
curies 
gammas 
neutrons 
watts 
261 
177 
158 
268 
101811999 
101811 999 
101811999 
101811999 
04:31 a 
04:07a 
04:31 a 
04:31 a 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Page: XII-7 of XII-16 I 
CD 2 
File Size (bytes) l ~ a t e of Transfer l~ime of Transfer l ~ i l e Name 
Folder: 1 %-through-3.5%-cut-files 
2,582,590 912011 999 05:57a 1.0%.20GWd.bottom.cut 
2,822,016 912011 999 06:01 a 1.0%.20GWd.fuel.cut 
2,631,240 912011 999 06:34a 1.0%.20GWd.plenum.cut 
2,599,460 912011 999 06:56a 1.0%.20GWd.top.cut 
2,625,521 912011 999 06:54a 1.0%.30GWd.bottom.cut 
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2,700,880 912011 999 06:25a 1.0%.40GWd.plenum.cut 
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2,930,002 912011 999 05:57a 1.0%.50GWd.fuel.cut 
2,730,624 912011 999 06:09a 1.0%.50GWd.plenum.cut 
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2,699,644 912011 999 06:25a 1.0%.60GWd.bottom.cut 
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2,770,074 912011 999 06:04a 1.0%.70GWd.plenum.cut 
2,748,992 912011 999 06:06a 1.0%.70GWd.top.cut 
2,733,313 9/20/1999 06:09a 1.0%.75GWd.bottom.cut 
2,988,989 912011999 05:53a 1.0%.75GWd.fuel.cut 
2,782,285 912011 999 06:02a 1.0%.75GWd.plenum.cut 
2,755,251 912011 999 06:05a 1.0%.75GWd.top.cut 
2,168,985 912011 999 07: 1 8a 1.5%.100MWd.bottom.cut 
2,352,299 912011 999 07:04a 1.5%.100MWd.fuel.cut 
2,206,465 912011 999 07: 1 6a 1.5%.100MWd.plenum.cut 
2,186,063 912011 999 07: 1 7a 1.5%.100MWd.top.cut 
2,484,072 912011 999 07:02a 1.5%.1OGWd.bottom.cut 
2,714,614 912011 999 06:23a 1.5%.1OGWd.fuel.cut 
2,522,256 912011 999 06:59a 1.5%.1OGWd.plenum.cut 
2,501,264 912011 999 07:01 a 1.5%.1OGWd.top.cut 
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2,081,760 912011 999 07:23a 1.5%.1OMWd.plenum.cut 
2,072,856 912011 999 07:24a 1.5%.10MWd.top.cut 
2,315,908 912011 999 07: 1 4a 1.5%.1 GWd.bottom.cut 
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2,350,804 912011 999 07:04a 1.5%.1 GWd.plenum.cut 
2,337,614 912011 999 07:12a 1.5%.1 GWd.top.cut 
1,987,442 912011 999 07:27a 1.5%.1 MWd.bottom.cut 
2,146,110 912011 999 07:20a 1.5%.1 MWd.fuel.cut 
2,012,118 912011 999 07:26a 1.5%.1 MWd.plenum.cut 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Page: XII-8 of XII-16 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-OOA Attachment XII. Page: XII-9 of XI-16 
CD 2 
File Size (bytes) Date of Transfer Time of Transfer File Name 
2,566,105 912011 999 06:58a 2.0%.20GWd.bottom.cut 
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2,718,424 912011 999 06:22a 2.0%.50GWd.plenum.cut 
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2,722,069 912011 999 06:21 a 2.0/o.60GWd.top.cut 
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2,205,272 912011 999 07: 1 6a 2.5%.100MWd.plenum.cut 
2,188,114 912011 999 07: 1 7a 2.5%.1OOMWd.top.cut 
2,483,304 912011 999 07:02a 2.5/o.10GWd.bottom.cut 
2,701,892 912011999 06:25a 2.5%.1OGWd.fuel.cut 
2,518,656 912011999 06:59a 2.5%.1OGWd.plenum.cut 
2,500,000 912011 999 07:01 a 2.5%.1OGWd.top.cut 
2,059,089 912011 999 07:24a 2.5%.1OMWd.bottom.cut 
2,236,341 912011 999 07: 1 5a 2.5%.1OMWd.fuel.cut 
2,087,313 912011 999 07:22a 2.596.1 0MWd.plenum.cut 
2,078,037 9/20/1999 07:23a 2.5%.1OMWd.top.cut 
2,303,085 912011 999 07: 1 4a 2.5%.1 GWd.bottom.cut 
2,490,291 912011 999 07:02a - -- 2.5%.1 GWd.fuel.cut 
2,337,207 912011 999 07:12a 2.5%.1 ~~dplenum.cut 
2,321,775 912011999 07: 1 3a 2.5%.1 GWd.top.cut 
1,995,005 912011999 07:27a 2.5%.1 MWd.bottom.cut 
2,152,185 912011999 07: 1 9a 2.5%.1 MWd.fuel.cut 
2,016,953 912011999 07:25a 2.5%.1 MWd.plenum.cut 
2,012,365 912011 999 07:26a 2.5%.1 MWd.top.cut 
2.561.055 912011 999 06:58a 2.5%.20GWd.bottom.cut 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Paee: XII-10 of XII- 16 I 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Page: XII-11 of XII-16 I 
CD 2 
File Size (bytes) Date of Transfer Time of Transfer File Name 
2,592,702 912011 999 06:56a 3.0%.20GWd.plenum.cut 
2,567,050 912011 999 06:57a 3.0%.2OGWd.top.cut 
2,605,177 912011 999 06:55a 3.0%.30GWd.bottom.cut 
2,844,344 912011999 06:01 a 3.0%.30GWd.fuel.cut 
2,652,319 912011 999 06:32a 3.0%.30GWd.plenum.cut 
2,624,725 912011 999 06:54a 3.0%.30GWd.top.cut 
2,638,930 912011 999 06:33a 3.0%.40GWd.bottom.cut 
2,881,942 912011999 05:59a 3.0%.40GWd.fuel.cut 
2,692,172 912011 999 06:27a 3.0%.40GWd.plenum.cut 
2,663,386 912011 999 06:30a 3.0%.40GWd.top.cut 
2,671 ,I 92 912011 999 06:30a 3.0%.50GWd.bottom.cut 
2,907,578 912011999 05:58a 3.0%.50GWd.fuel.cut 
2,717,694 912011 999 06:22a 3.0%.50GWd.plenum.cut 
2,694,406 912011 999 06:26a 3.0%.50GWd.top.cut 
2,698,497 912011 999 06:25a 3.0%.60GWd.bottom.cut 
2,943,515 912011 999 05:57a 3.0%.6OGWd.fuel.cut 
2,744,737 912011 999 06:07a 3.0%.60GWd.plenum.cut 
2,722,331 912011 999 06:21 a 3.0%.60GWd.top.cut 
2,721 ,I 61 912011 999 06:22a 3.0%.70GWd.bottom.cut 
2,970,669 912011 999 05:55a 3.0%.7OGWd.fuel.cut 
2,769,307 912011 999 06:04a 3.0%.70GWd.plenum.cut 
2,743,979 912011 999 06:07a 3.0%.70GWd.top.cut 
2,734,208 912011 999 06:08a 3.0%.75GWd.bottom.cut 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Page: XII- 12 of XII- 16 
File Size (bytes) l ~ a t e of Transfer /Time of Transfer l ~ i l e Name 
Folder: 4.0%-through5.5%_and_SSclad 
2,164,211 912011 999 07: 1 9a 4.0%.1OOMWd.bottom.cut 
2,346,359 912011 999 07:11 a 4.0%.100MWd.fuel.cut 
2,196,775 912011 999 07: 1 7a 4.0%.100MWd.plenum.cut 
2,183,739 912011 999 07: 1 8a 4.0%.1OOMWd.top.cut 
2,479,639 912011 999 07:03a 4.0%.1OGWd.bottom.cut 
2,698,207 912011 999 06:26a 4.0%.1OGWd.fuel.cut 
2,515,011 912011 999 07:OOa 4.0%.1OGWd.plenum.cut 
2,496,261 912011 999 07:01 a 4.0%.1OGWd.top.cut 
2,058,599 912011999 07:24a 4.0%.1OMWd.bottom.cut 
2,235,831 912011 999 07: 1 5a 4.0%.1OMWd.fuel.cut 
2,089,035 912011 999 07:20a 4.0%.1OMWd.plenum.cut 
2,077,671 912011 999 07:23a 4.0%.1OMWd.top.cut 
2,300,650 912011 999 07: 1 4a 4.0h.1 GWd.bottom.cut 
2,480,292 912011 999 07:03a 4.0%.1 GWd.fuel.cut 
2,335,670 912011 999 07:12a 4.0%.1 GWd.plenum.cut 
2,318,740 912011 999 07: 1 3a 4.0%.1 GWd.top.cut 
1,995,145 912011 999 07:27a 4.0%.1 MWd.bottom.cut 
2,154,309 912011 999 07: 1 9a 4.0%.1 MWd.fuel.cut 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Paee: XII- 13 of XII- 16 
File Size (bytes) Date of Transfer Time of Transfer File Name 
2,018,457 912011 999 07:25a 4.0%.1 MWd.plenum.cut 
2,014,737 912011 999 07:25a 4.0%.1 MWd.top.cut 
2,546,626 912011 999 06:59a 4.0%.20GWd.bottom.cut 
2,778,646 912011 999 06:03a 4.0%.20GWd.fuel.cut 
2,588,430 912011 999 06:57a 4.0/~.20GWd.plenum.cut 
2,563,942 912011 999 06:58a 4.0%.20GWd.top.cut 
2,604,783 912011 999 06:55a 4.0%.30GWd.bottom.cut 
2,844,893 912011 999 06:01 a 4.0%.30GWd.fuel.cut 
2,654,251 912011 999 06:32a 4.0%.30GWd.plenum.cut 
2,626,533 912011 999 06:34a 4.0%.30GWd.top.cut 
2,633,254 912011 999 06:33a 4.0%.40GWd.bottom.cut 
2,872,784 912011999 06:OOa 4.0%.40GWd.fuel.cut 
2,684,046 912011 999 06:28a 4.0%.40GWd.plenum.cut 
2,656,630 912011 999 06:31 a 4.0%.40GWd.top.cut 
2,662,834 912011 999 06:30a 4.0%.50GWd.bottom.cut 
2,902,870 912011 999 05:58a 4.0%.5OGWd.fuel.cut 
2,712,842 912011 999 06:23a 4.0%.50GWd.plenum.cut 
2,684,626 912011 999 06:28a 4.0%.50GWd.top.cut 
2,691,347 912011 999 06:27a 4.0%.60GWd.bottom.cut 
2,926,851 912011 999 05:57a 4.0%.60GWd.fuel.cut 
2,734,973 912011 999 06:08a 4.0%.60GWd.plenum.cut 
2,714,809 912011 999 06:23a 4.0/~.60GWd.top.cut 
2,720,493 912011 999 06:22a 4.0%.70GWd.bottom.cut 
2,965,873 912011 999 05:56a 4.0%.70GWd.fuel.cut 
2,769,357 912011 999 06:04a 4.0%.70GWd.plenum.cut 
2.744.451 912011 999 06:07a 4.0%.70GWd.to~.cut 

Title: 
Docu - 
: BWR Source Term Generation and Evaluation 
lment Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Page: XII-14 of XII-16 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Page: XII-15 of XII-16 
File Size (bytes) Date of Transfer Time of Transfer File Name 
2,549,177 912011 999 06:58a 5.0%.20GWd.bottom.cut 
2,775,503 912011 999 06:03a 5.0%.20GWd.fuel.cut 
2,590,733 912011 999 06:56a 5.0%.20GWd.plenum.cut 
2,566,513 912011 999 06:57a 5.0%.20GWd.top.cut 
2,601,534 912011 999 06:56a 5.0%.30GWd.bottom.cut 
2,840,290 912011 999 06:01 a 5.0%.3OGWd.fuel.cut 
2,646,728 912011 999 06:33a 5.0%.30GWd.plenum.cut 
2,620,348 912011 999 06:54a 5.0%.3OGWd.top.cut 
2,633,137 912011 999 06:34a 5.0%.40GWd.bottom.cut 
2,874,155 912011 999 06:OOa 5.0%.40GWd.fuel.cut 
2,680,041 912011999 06:29a 5.0%.40GWd.plenum.cut 
2,656,007 912011 999 06:31 a 5.0%.40GWd.top.cut 
2,656,948 912011 999 06:31 a 5.0%.50GWd. bottom.cut 
2,897,098 912011 999 05:58a 5.0%.50GWd.fuel.cut 
2,706,058 912011 999 06:24a 5.0%.50GWd.plenum.cut 
2,682,232 912011 999 06:28a 5.0%.50GWd.top.cut 
2,683,602 912011 999 06:28a 5.0%.60GWd.bottom.cut 
2,922,200 912011 999 05:57a 5.0%.60GWd.fuel.cut 
2,730,282 912011 999 06:20a 5.0%.60GWd.plenum.cut 
2,709,112 912011 999 06:24a 5.0%.60GWd.top.cut 
2,710,938 912011 999 06:23a 5.0%.70GWd.bottom.cut 
2,951,328 912011 999 05:56a 5.0%.70GWd.fuel.cut 
2,759,678 912011 999 06:05a 5.0%.70GWd.plenum.cut 
2,734,648 912011 999 06:08a 5.0%.70GWd.top.cut 
2,725,401 912011 999 06:21 a 5.0%.75GWd.bottom.cut 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A Attachment XII. Page: XII-16 of XII-16 I 
I 1 I 

Title: BWR Source Term Generation and Evaluation Worksheet 'DBWS REV 01' 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment XIII, Page XIII-I of XIil-10 I 
Averages from Ref. 7.24, Table 5; Maximums from Ref. 7.24, Attachment Ill bin.dat files 
For all cases, the information on the following pages 
represents the data set closest to those defined on this page 
Case A , B, and C Averages for the extended baseline (84k MTU) 
- 
Nominal 
BWR: data points: 3.5%, 40 GWdMTU, 25 years old, 200 kg loading (actual burnup is 39.3 GWdA) 
3.02 33592.2401 6 0.18 Case A Average Age 25.32 I 
I Enrich Burnup (MWdlMTU) MTU Case B Average Age 25.31 1 
1 Case C Average Age 25.43 1 
Averages are calculated by multiplying the enrichment, burnup or age by the number of 
assemblies with that given burnup, enrichment, or age, summing, 
and then dividing by the total number in the waste stream 
I Burnup I Burnup I bur nu^ I 
Enrichment (MW~IMTU) Age (yrs) I ~nrichment (MW~IMTU) Age (yrs) 1 Enrichment (MWCVMTU) Age (yrs) 
4.28 65550 5 1 4.28 65550 5 1 4.28 65550 5 
Nominal data 
points: 5.0%,75 GWdMTU, 5 years old, 200 kg loading (actual burnup is 73.8 GWdA) 
For the maximums: 
These are a combination of all the maximum parameters: 
the maximum burnup, the max enrichment, and minimum age (all independent of each other). 
All source term calculations use dimensions 
for the representative assembly 
The cases refer to the arrival scenarios 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'BWR.ave.2.curie.s' Attachment XIII, Page Xlll-2 of Xlll-10 I 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'BWR.ave.2.curie.s' Attachment XIII, Page Xlll-3 of Xlll-10 I 

Title: BWR Source Term Generation and Evaluation Worksheet 'BWR.ave.3.curies' 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment XIII, Page Xlll-4 of XIII-10 I 

Title: BWR Source Term Generation and Evaluation Worksheet 'BWR.ave.3.curies1 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment XIII, Page XllI-5 of XIII-10 I 
Average BWR for longer decay times 
Decay Time 10025. yr 20025. yr 30025. yr 100025. yr 300025. yr 1000025 yr 
Year: 12033 22033 32033 102033 302033 1002033 . ~0214 3.1 2E-02 6.74E-02 9.97E-02 2.42E-01 2.69E-01 9.71 E-02 
~0215 5.59E-04 1.08E-03 1.56E-03 3.63E-03 4.49E-03 4.51 E-03 
Po21 8 3.1 2E-02 6.75E-02 9.97E-02 2.42E-01 2.69E-01 9.72E-02 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'BWR.max.2.curies' Attachment XIII, Page Xlll-6 of Xlll-10 I 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'BWR.max.2.curies' Attachment XIII, Page Xlll-7 of Xlli-10 1 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Worksheet 'BWR.max.2.curies' Attachment XIII, Page Xlll-8 of Xlll-10 I 

Title: BWR Source Term Generation and Evaluation Worksheet 'BWR.max.3.curie.s' 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment XIII, Page Xlll-9 of Xlll-10 I 
Max BWR for longer decay times 
Decay Time 1 10005. yr (20005. yr 
Year: 1 12033 122033 
30005. yr 
32033 
---- 
100005. yr 
102033 
300005. yr 
302033 
1000005 yr 
1002033 

Title: 
Docu 
BWR Source 
ment ldentifie 
Term Generation and Evaluation Worksheet 'BWR.max.3.curie.s' 
!r: 000-00C-MGRO-00200-000-00A 
Attachment XI Ill, Page Xlll-10 of Xlll-10 I 
1 ---- 
Decay Time 110005. yr (20005. yr 30005. yr 100005. yr 300005. yr 1000005 yr 
Year: 1 12033 122033 32033 102033 302033 1002033 

Title: BWR Source Term Generation and Evaluation 
Document Identifier: 000-00C-MGRO-00200-000-00A 
Attachment XIV; Page XIV-1 of XIV-1 
Comparison of Source Terms per MTU of 4 Average BWR SNF Assemblies 
Curies 
Assembly 
Characteristics 
Watts 
3.02%33.6GWd-25.0 
(0.177 MTU) 
Activation Products 
Actinides and daughters 
Fission products 
Total 
Grams 
Phtonsls 
Total intensity I 6.1 110E+15 I 5.1470E+15 I 4.8900E+15 I 5.51 80E+15 1 
3.5Yv40GWd-25y 
(0.2 MTU) 
3.09Yv34.8GWd-22.0~ 
(0.1 77 MTU) 
6.0790E+02 
5.4690E+04 
2.0070E+05 
2.561 0E+05 
Activation Products 
Actinides and daughters 
Fission products 
Total 
Neutronsls 
Worksheet 'compsrc-MTU' 
3.03%33.6GWd-23.0~ 
(0.1 77 MTU) 
7.8770E+02 
6.4560E+04 
2.2570E+05 
2.91 1 0E+05 
6.8200E+02 
6.7620E+04 
2.4960E+05 
3.1 790E+05 
5.4880E+00 
3.3040E+02 
6.0270E+02 
9.4440E+02 
4.81 50E+00 
5.0520E+02 
7.1 290E+02 
1.2230E+03 
7.2580E+05 
9.61 60E+05 
3.7690E+04 
1.7260E+06 
7.0580E+02 
5.91 70E+04 
2.1 100E+05 
2.7090E+05 
7.2580E+05 
9.6340E+05 
3.5950E+04 
1.7260E+06 
1.0670E+07 
3.3440E+08 
3.4560E+08 
4.2350E+00 
3.2880E+02 
5.7250E+02 
9.1 1 1 0E+02 
7.2580E+05 
9.6340E+05 
3.5950E+04 
1.7260E+06 
Activation Products 
Actinides and daughters 
Fission products 
Total 
9.5330E+06 
2.5300E+08 
2.6290E+08 
6.4490E+00 
3.6590E+02 
6.4540E+02 
1.0250E+03 
7.2580E+05 
9.5460E+05 
4.5480E+04 
1.7260E+06 
9.5950E+06 
2.7050E+O8 
2.8060E+08 
Activation Products 
Alpha, N Neutrons 
Fission products 
Total 
1.481 0E+07 
5.401 OE+08 
5.5500E+08