ML17335A274

Rev 0 to ER-98-009, Preliminary Waste Characterization of DC Cook SG Lower Assemblies

ML17335A274
Person / Time
Site:	Cook
Issue date:	09/01/1998
From:	Savino A, Whittaker M CHEM-NUCLEAR SYSTEMS, INC.
To:
Shared Package
ML17335A271	List: ML17335A270 ML17335A272 ML17335A273 ML17335A274 ML17335A275
References
ER-98-009, ER-98-009-R00, ER-98-9, ER-98-9-R, NUDOCS 9810140054
Download: ML17335A274 (40)

Text

CHEM-NUCLEARSYSTEMS ONLN

'"."ll'R)l. E MFf IN.

PREPARED BY:

INDEPENDENT REVIEWER:

PRINTED OR TYPED NAME Mark Whittaker Anthony Savino SIGNATURE DATE

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~/i /gs DOCUMENTTITLE:

Preliminary Waste Characterization of D.C. Cook Steam Generator Lower Assemblies DOCUMENT NO.

ER-98-009 REV.

0 PAGE 1 of19 98iOi400M 98i008 PDR ADOCK 050003ib P

PDR

1.

SUMMARY

TABLEOF CONTENTS

~Pa e No.

.3 2.

PHYSICAL DESCRIPTION OF STEAM GENERATORS 3.

RADIOACTIVESOURCE CHARACTERISTICS 4.

CHARACTERIZATIONASSUMPTIONS 5.

SOURCE CHARACTERIZATION.

5.1 Microshield Calculations 5.2 Source Distribution 6.

WASTE CLASSIFICATIONAND DOT SUBTYPING.

7.

REFERENCES APPENDIX A AEP SUPPLIED INFORMATION APPENDIX B SHIPPING PAPERS AND DISPOSAL MANIFESTS APPENDIX C MICROSHIELD MODELS AND OUTPUT.

APPENDIX D SURFACE AREA CALCULATIONS 10 10

...... 1 4

..16 17 18 19 LIST OF TABLES AND FIGURES Table 2-1 SGLA Characteristics..

Table 2-2 Internal Surface Area and Material Density Information..

Table 3.1 - Radionuclide Distribution.

Table 5-1 Curie Content in Straight Tube Section..

Table 5-2 SGLA Total Contamination Content Results Table 6-1a DOT Subtyping of D.C. Cook SGLA1 Table 6-1b DOT Subtyping of D.C. Cook SGLA 2 Table,6-1c DOT Subtyping of D.C. Cook SGLA3....

Table 6-1d DOT Subtyping of D.C. Cook SGLA 4 Table 6-2 Disposal Classification of D.C. Cook SGLA 3.

4

..5

.6 10 11 12 13 13 14 Figure 2-1 D.C. Cook Steam Generator Reference Dimensions 4

Figure 2-2 Steam Generator Channel Head Region Components...............

5 Figure 5-1 Microshield Model Representation of Steam Generator Source Region.......9 ER-98-009 Rev. 0 Page 2

1.

Summary This report presents the preliminary analyses performed in support of the source characterization and classification of the four D.C. Cook Unit 2 Steam Generator Lower Assemblies (SGLAs) for American Electric Power, the owner and operator of the D.C.

Cook plant. The radionuclide content of the SGLAs was determined based on currently available isotopic and dose rate information to demonstrate compliance with applicable criteria for transportation and disposal.

A final characterization will be performed after removal of the SGLAs from the storage facilityand prior to shipment for disposal.

2.

Physical Description of Steam Generators The steam generators at D.C. Cook Unit 2 are Westinghouse Model 51 and were placed in service in 1977. The generators were removed in 1988 and the steam domes removed.

The remaining "lower assemblies" were placed in storage at the reactor site.

Similar Westinghouse Model 51 generators were transported from Salem for disposal in 1996. The basic physical dimensions and design criteria of the SGLAs is included in Table 2-1 [from information provided in Appendix A]. A summary sketch of these basic characteristics is provided in Figure 2-1. This information is utilized to develop the surface areas and effective source region density information listed in Table 2-2.

Additionally, a sketch of the channel head region is provided in Figure 2-2.

ER-98-009 Rev. 0 Page 3

Table 2-1 SGLA Characteristics General Information Weight Length Lower Shell Diameter Lower Shell Thickness 476,000 lbs 533 in 135 in 2.82 in 2.16E+05 kg 1354 cm 343 cm 7.16 cm Tube Bundle Data Number of Tubes Tube Bundle Radius Straight Tube Length Tube OD Tube Wall Thickness Wrapper Data Wrapper Thickness Wrapper OD 3388 59.8 in 357 in 0.875 in 0.05 in 0.38 in 124.9 in 152 cm 907 cm 2.22 cm 0.13 cm rn 0.97 cm 317.2 cm Channel Head Data Channel Head Radius Channel Head Thickness 62.81 in 5.16 in 159.5 cm 13.1 cm Figure 2-1 D.C. Cook Steam Generator Reference Dimensions DC COOK UNIT 2 SGLA BEPEREYCE DATA FOR IVASTE CHARACTERIZATIO'A J.CO 76.75 556.75 r

2.82 21.00 14.00 2.19 R.

Smollest Tube 5

Dlo I 75.7 59.84 R

Lorgest Tube 1

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67 0

155.00 O.OIo.

5.25 129.56

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Cnonnel HeoO OAIder Piete T,ons't:on Sbeo Vl~oooer/Bottle I

~ Lower Snell Tube Plots ER-98-009 Rev. 0 Page 4

Figure 2-2 Steam Generator Channel Head Region Components r

62.81 R.

Primory Rcnle Tube Prole 5.16 OVeer Pc',e onloT Channel Read Region Components Table 2-2 Internal Surface Area and Material Density Information Contaminated Surface Areas Total Tube Bundle Inner Surface Straight Tube Surface Area Tubes in Tube Sheet U-Tube Section Surface Area Channel Head Components Channel Hea Tube Shee DividerPlate (tn')

7.04E+06 5.89E+06 3.47E+05 8.04E+05 4.64E+04 2.48E+04 9.20E+03 1.24E+04 (cm')

4.54E+07 3.80E+07 2.24E+06 5.19E+06 2.99E+05 1.60E+05 5.93E+04 8.00E+04 Densities Tubes (Nickel Alloy)

Shells 8 Wrapper Tube Bundle Data Straight Tube Mass Straight Tube Region Volume (Ib/in')

0.298 0.284 9.33E+4 Ibs 4.0E+6 in'g/cm')

8.25 7.86 4.24E+7 g 6.56E+7 cm'ffective Source Region Density 0.646 g/cm'.

Radioactive Source Characteristics The source scaling factor table, included in Appendix A, was developed based on Part 61 laboratory analyses of D.C. Cook wastes during the time the generators were in use.

The 1988 estimated gamma activity per generator, included in Appendix A, was determined by AEP staff and used in addition to the scaling factors to establish the ER-98-009 Rev. 0 Page 5

normalized source in the SGLAs. The normalized source term and the activity estimates from which it was developed are provided in Table 3-1. The AEP 1988 gamma activity estimate assigned a single curie value to a group of several radionuclides.

In developing the normalized source, the single curie value was evenly distributed among the grouped radionuclides.

Scaling factors were given to predict Pu-239 from both Co-60 and from Ce-144.

The Co-60 factor, which provided the larger Pu-239 amount, was used.

The estimated activity was determined for the date of SGLA removal (1988) using the AEP estimated gamma activity and the scaling factors and then was decayed ten years to 1998.

Radionuctides with an activity of less than 1 mCi after decay were neglected.

The estimated activity was normalized to 1 curie to provide the isotopic distribution used to characterize the SGLA source term based on the 1998 external survey data.

II J "Sl Table 3.1 - Radionuclide Distribution Radionuclide Am-241 C-14 Cm-242 CN-244 Co-57 Co-58 Co-60 Cr-51 Fe-55 Fe-59 Mn-54 Nb-95 Ni-63 1988 Estimated Activity(Ci) 3.98e-002 2.01e+000 3.56e-002 4.01e-002 2.43e+000 1.53e+002 2.10e+002 1.19e+002 3.97e+002 7.0e+000 8.0e+000 4.00e+001 8.80e+001 1998 Estimated Activity(Ci) 1.28e-001 2.01e000 2.73e-002 5.64e+001 3.05e+001 2.44E-03 8.21e+001 Normalized Source Term (Ci) 7.33E-04 1.14E-02 1.56E-04 3.21E-01 1.73E-01 1.39E-05 4.67E-01 ER-98-009 Rev. 0 Page 6

Radionuclide 1988 Estimated Activity(Ci) 1998 Estimated Activity(Ci)

Normalized Source Term (Ci)

P,u-238 PU-239 8.73e-002 6.93e-002 8.12e-002 6.93e-002 4.60E-04 3.94E-04 Pu-241 7.07e+000 4.38e+000 2.48E-02 Ru-106 2.43e+000 2.51e-003 1.43E-05 Sb-125 Sn-113 2.43e+000 2.43e+000 1.99e-001 1.13E-03 Te-125m Zn-65 2.43e+000 4.87e-002 2.77E-04 Contamination samples from inside one SGLA are planned to be taken in September 1998 to update the source isotopic distribution. Analysis on these samples willinclude the typical Part 61 analyses to verify the isotopic content of TRU and other hard to detect radionuclides as well as the expected gamma emitting fission/activation products.

The results of these analyses should be available November 1. Ifthese results are significantly different from the distributionin the normalized source. term, this report will be revised to reflect these values.

As compared to other steam generators previously disposed, the normalized-source term reflects higher.TRU quantities as, compared to the other fission/activation products indicating that the normalized source term may be conservatively predicting TRU content.

External radiation surveys were taken on the SGLAs on 6 July 1998. This survey information is included in Appendix A. Measurements taken radially on the straight tube region of the SGLA are-expected to be uniform due to expected uniform deposition of contaminants in the straight tubes.

Due to the storage arrangement, measurements made on the surfaces of the SGLAs facing each other show higher readings due to the contribution from the adjacent SGLA. Calculation of the expected contribution from the adjacent SGLA assuming a uniform dose field equal to that measured on the opposite side, corrected for distance using a Microshield model, shows that the higher reading can be attributed to this contribution rather than to a non-uniform dose field on the measured SGLA. The 30 cm readings corrected for contribution from the adjacent SGLA, averaged over the straight tube region are 21, 21, 22, and 20 mR/hr respectively ER-98-009 Rev. 0 Page 7

for SGLA 1, 2, 3, and 4. These average values are used in calculating the surface area contamination on the straight tubes.

The final characterization will be Performed based on dose rate profiles taken on the SGLAs on removal from the storage facility. However, these dose rates are not expected to change significantly from those measured in July, 1998 except for eliminating the contribution from an adjacent SGLA.

4.

Characterization Assumptions Several assumptions are made in the course of performing the characterization analyses of the steam generators.

These assumptions are utilized to simplify the analysis, while maintaining accuracy in the overall result.

1)A

1. Secondary-side steam generator surfaces contain no activity.

Since the secondary side of the steam generator is exposed only to secondary side water, it is assumed that the secondary side contains only negligible quantities of radioactive contamination.

This assumption has been used for previous steam generator characterizations.

2.

Residual water in plugged tubes contains no activity.

The plugged tubes in the steam generator could contain relatively small amounts of water that seeps into the'tubes during operation of the generators.

It is assumed that this water contains negligible quantities of radioactive material, and is not considered in this characterization.

3.

Uniformity in distribution of primary-side surface contaminates.

Two EPRI reports [2, 3] address the issue of steam generator primary side surface contamination.

These reports indicate that, while the straight tube sections with the SGLAs exhibit fairly uniform surface contamination, the U-tube and tube sheet sections of the heat exchanger tubes contain higher surface contamination values than that of the straight tube sections.

Additional uncertainty exists concerning the relative surface contamination levels between the tubes and the channel head surfaces, including the tube sheet, divider plate, and bowl itself. The studies indicate that the differing materials used for the tubes versus the channel head components, combined with other factors, could result in higher surface contamination values in the channel head region.

To address these issues, this analysis assumes that all surfaces other than the straight tube sections contain surface contamination levels twice that of the straight tube sections.

This factor of two is addressed specifically in the reference [2] study for the various tube sections.

It is reasonable to apply this assumption to the ER-98-009 Rev. 0 Page 8

channel head sections as well, as they are of a similar geometry and represent only a minimal fraction of the total surface area, and thus only a small portion of the total activity in the SGLAs.

5.

Source Characterization Employing the information from the previous sections, the radionuclide content of the SGLAs can be determined from the measured external SGLA dose rates and the SGLA design parameters.

The straight tube section of the lower barrel of the SGLA is modeled with the Microshield [4] point kernel shielding code, using the normalized source term provided in Table 3-1..The shortest straight tube length is approximately 357 inches, not including the 21 inch length of tube in the tube sheet.

The diameter and thickness of the radial source and shielding regions of the model are taken from the data provided by AEP included in Appendix A. A schematic of the Microshield model is provided in Figure 5-1.

Figure 5-1 Microshield Model Representation ofSteam Generator Source Region 356.75" Effective Straight Tube Length

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Wrapper -.38" Void 2.23" Lower Barrel 2.82" The source region is modeled as nickel alloy, Alloy600, at a density of 0.658 g/cc to represent the fraction of the source region cross-section occupied by the tubes.

The void regions are modeled as air, and the wrapper and lower barrel are modeled 'as A 533 steel.

The densities are provided in Table 2-2.

ER-98-009 Rev. 0 Page 9

5.1 Microshield Calculations Analyses are performed with Microshield using the model previously described with the normalized source term provided in Table 3-1. The short-lived daughter of Ru-1 06, Rh-106, was included in the Microshield source term to accurately depict the dose rate.

Short-lived daughters are not included in the source term for characterization.

The calculation produces an exposure rate 30 cm from the surface resulting from a 1 curie normalized source of 3.80 x 10'R/hr/Ci. The measured exposure rates of 21, 21, 22, and 20 mR/hr, respectively, are then divided by this dose-per-curie factor to determine the number of curies in the straight tube section of the SGLA. The resulting activities are 55.3, 55.3, 57.9, 52.6 Ci. This activity is then divided by the surface area of the straight tubes (3.80E+07 cm') to give the activity per unit area.

The results of these calculations are presented in Table 5-1.

Table 5-1 Curie Contentin Straight Tube Section SGLA 1 SGLA 2 SGLA 3 SGLA 4 Average 30 cm Exposure Rate (mR/hr) 21 21 22 20 Activityin Straight Tube Section (Ci)

Areal Activity(NCi/cm')

55.3 1.45 55.3 57.9 1.45 1.52 52.6 1.39 5.2 Source Distribution The straight tube source contamination calculated in Section 5.1 is utilized to determine the contamination on the U-tube and tube sheet sections of the heat exchanger tubes, as well as the channel head components.

As shown in Table 5-1, the straight tube sections of the steam generator heat exchanger tubes contain 1.45, 1.45, 1.52, and 1.39 pCi/cm', respectively,"of radioactive material of the distribution shown in Table 3-1. Using the factor of two discussed in Section 4, the surface contamination level on the remaining primary side surfaces of the steam generator is 2.91, 2.91, 3.05, and 2.77 NCi/cm', respectively.

These surface contamination levels are used to calculate the total number of curies in each steam generator as shown in Table 5-2.

ER-98-009 Rev. 0 Page 10

Table 5-2 SGLA Total Contamination Content Results.

Contaminated Surface Areas Straight Tube Surface Area Tubes in Tube Sheet U-Tube Section Surface Area Channel Head Tube Sheet Divider Plate Total Surface

-'..'A'rea (cm')

3.80E+07 2.25E+06 5.19E+06 1.60E+05 5.93E+04 8.00E+04 SGLA 1 Activity (Ci) 55.3 6.54 15.1 0.47 0.17 0.23 77.8 SGLA 2 Activity (Ci) 55.3 6.54 15.1 0.47 0.17 0.23 77.8 SGLA 3 Activity (Ci) 57.9 6.86 15.8 0.49 0.18 0.24 81.5 SGLA 4 Activity (Ci) 52.6 6.23 14.4 0.44 0.16 0.22 74.1 6.

Waste Classification and DOT Subtyping The shipping and disposal classiTications can be performed for the SGLAs based on the calculated radionuclide content in accordance with regulatory requirements [5, 6, 7, and 8]. This information is important to demonstrate that the SGLAs meet applicable requirements for transportation and disposal.

The DOT subtyping for the SGLAs are shown in Table 6-1a-6-1d.

As shown, the SGLAs contain a greater-than-Type-A quantity of-radioactive material, with a cumulative A, values of 33.5, 33.5, 35.1, and 31.9, respectively.

While the average surface contamination levels were shown in Table 5-2 to be less than the SCO-II limitof 20 pCi/cm', uncertainty in the distribution of activity over all surfaces in the SGLA results in an uncertainty that all areas are less than the SCO-II limit. As such, an exemption from SCO-II limits and packaging requirements will be requested from the DOT as suggested in Reference 9. The total amount of fissile material is 0.64g which is less than 15g; therefore, the shipment qualifies as fissile excepted.

ER-98-009 Rev. 0 Page 11

Table 6-1a DOT Subfyping ofD.C. Cook SGLA 1 Isotope Cu ries A2 Value A2 Fraction Am-241 5.70E-02 0.00541 10.54 C-14 8.85E-01 54.1 0.02 Cm-244 Co-60 Fe-55 Mn-54 Ni-63 Pu-238 Pu-239 PU-241 Ru-106 1.22E-02 0.0108 2.49E+01 10.8 1.35E+01 1080 1.08E-03 27 3.63E+01 811 1.93E+00 1.11E-03 0.27 5.41 3.58E-02 0.00541 3.06E-02 0.00541 1.13 2.31 0.01 0.00 0.04 6.61 5.66 7.15 0.00 Sb-125 Te-125m TOTALS 8.80E-02 2.15E-02 7.78E+01 24.3 243 0.00 0.00 33.48 Table 6-fb DOT Subtyping ofD.C. Cook SGLA 2 Isotope.

Curies Am-241 5.70E-02 0.00541 10.54 A2 Value A2 Fraction C-14 8.85E-01 54.1 0.02 Cm-244 1.22E-02 0.0108 1.13 Co-60 Fe-55 Mn-54 Ni-63 Pu-238 Pu-239 2.49E+01 1.35E+01 1.08E-03 3.63E+01 3.58E-02 3.06E-02 10.8 1080 27 811 0.00541 0.00541 2.31 0.01 0.00 0.04 6.61 5.66 Pu-241 1.93E+00 0.27 7.15 RU-106 1.11E-03 5.41 0.00 Sb-125 8.80E-02 Te-125m 2.15E-02 TOTALS 7.78E+01 24.3 243 0.00 0.00 33.48 ER-98-009 Rev. 0 Page 12

Table 6-1c DOT Subtyping ofD.C. Cook SGLA 3 Isotop'e'.",Curies A2 Value A2 Fraction Am-241 5.97E-02 0.00541 11.04 C-14 9.28E-01 54.1 0.02 Cm-244 1.27E-02 0.0108 1.18 Co-60 Fe-55 Mn-54 Ni-63 Pu-238 2.61E+01 1.41E+01 1.13E-03 3.80E+01 3.75E-02 10.8 1080 27 811 0.00541 2.42 0.01 0.00 0.05 6.93 PU-239 3.21E-02 Pu-241 2.02E+00 0.00541 0.27 5.93 7.49 RU-106 1.16E-03 5.41 0.00 Sb-125 9.22E-02 Te-125m 2.26E-02 TOTALS 8.15E+01 24.3 243 0.00 0.00 35.07 Table 6-1d DOT Subfyping ofD.C. Cook SGLA 4 Isotope Curies A2 Value A2 Fraction Am-241 5.43E-02 0.00541 10.04 C-14 8.43E-01 54.1 0.02 Cm-244 1.16E-02 0.0108 1.07 Co-60 Fe-55 Mn-54 Ni-63 PU-238 Pu-239 2.37E+01 1.28E+01 1.03E-03 3.46E+01 3.41E-02 2.92E-02 10.8 1080 27 811 0.00541 0.00541 2.20 0.01 0.00 0.04 6.30 5.39 Pu-241 1.84E+00 Ru-106 1.06E-03 0.27 5.41 6.81 0.00 Sb-125 8.38E-02 24.3 0.00 Te-125m 2.05E-02 TOTALS 7.41E+01 243 0.00 31.88 ER-98-009 Rev. 0 Page 13

The disposal classification of SGLA C3, which has the largest total activity, is shown in Table 6-2. The disposal volume is 104.52 m'nd the mass is 1.266E+08g.

This classification lists the required nuclides from 10 CFR 61, and demonstrates that the Table 1 and Table 2 isotopes meet the requirements for classification of the SGLAs as Class A waste.

Table 6-2 Disposal Classification ofD.C. Cook SGLA 3 Table 1 Isotopes C14 TC 99 I129 CM242 PU241 TRU >5 yr Half Life Table 1 Total Total Activity(Ci) 9.28E-01 0.00E+00 0.00E+00 O.OOE+00 2.02E+00 8.23E-02 Specific Activity 8.875E-03 Ci/m'.000E+00 Ci/m'.000E+00 Ci/m'.000E+00Ci/g 1.598E-08 Ci/g 6.501E-10 Ci/g Class A Limit 0.8 Ci/m'.3 Ci/m'.008 CI/01 2.00E-06 Ci/g 3.50E-07 Ci/g 1.00E-08 Ci/g Fraction of Table 1 Limits

,1..1 1E-02 O.OOE+00 O.OOE+00 0.00E+00 4.57E-02 6.50E-02 0.12 Table 2 isotopes CO 60 CS137 H 3 Nl 63 SR 90 Isotopes < 5yr Half Life Table 2 Total Total Specific Activity(ci)Activity (Ci/m')

2.61E+01 2.499E-01 O.OOE+00 O.OOOE+00 0

0.000 E+00 3.80E+01 3.638 E-01 O.OOE+00 O.OOOE+00 1.42E+01 1.363 E-01 Class A Limit (Ci/m')

3.5 0.04 1.04E-01 0.00E+00 700 1.95E-04 0.10 Fraction of Class A Limits 700 3.57E-04 1

O.OOE+00 40 O.OOE+00 7.

References

[1]

CNS Procedure EN-AD-010, "Procedure for Waste Characterization of Non-Irradiated Components or Items."

[2]

EPRI-NP-2968, "Primary-Side Deposits on PWR Steam Generator Tubes,"

Electric Power Research Institute, Palo Alto, CA, March 1983.

ER-98-009 Rev. 0 Page 14

[3]

EPRI-NP-3107, "Gamma-Ray Exposure Rate Distribution in a Steam Generator,"

Electric Power Research Institute, Palo Alto, CA, May 1983.

[4]

Grove Engineering, Inc. "Microshield Computer Code," Version 5.01.

[5]

NRC, "Low-Level Waste Licensing Branch Technical Position on Radioactive Waste Classification," (May 1983).

[6]

Code of Federal Regulations, 10CFR Part 61 and 10CFR Part 71.

[7]

Code of Federal Regulations, 49CFR Parts 100 to 177.

[8]

DHEC License CNSI-SC-097, (Barnwell Site Criteria).

[9]

NRC Generic Letter 96-07, "Interim Guidance on Transportation of Steam

. Generators," U.S. NRC Office of Nuclear Material Safety and Safeguards, December 5, 1996.

[10] NUREG-1608, "Categorizing and Transporting Low Specific ActivityMaterials and Surface Contaminated Objects," U.S. Nuclear Regulatory Commission, July 1998 ER-98-009 Rev. 0 Page 15

APPENDIX A AEP SUPPLIED INFORMATION (12 PAGES) g4 ER-98-009 Rev. 0 Page 16

American Bectri~wer Nuclear Generatio~up.... -,.

500 Circle Orive Buchanan, Ml491071373 Mr. John Bender Chem-Nuclear Systems, Inc.

140 Stoneridge Drive Columbia, SC 29210 ANERlCAN'scmrc POWER August 27, 1998

DearJohn:

This letter serves to document two issues related to the dimensions and center of gravity of the steam generators..

We have reviewed the dimensions of the steam generator as shown in CNS Sketch 46628-01, Rev.

0.

The dimensions'in this drawing accurately reflect the dimensions of the steam generators at the Cook Plant. This was reviewed by comparing the CNS sketch to the drawings and other data from Westinghouse.

The primary purpose of our review was to understand the shell thickness and steam generator wrapper thickness for use in Chem-Nuclear's shielding calculations.

Attached are Figure 1-1 "Outline" and Figure 1-2 "General Arrangement" and the cover page from the Vertical Steam Generators Instructions.

These figures show the dimensions that compare with the CNS Sketch.

Also attached is a letter from Westinghouse that confirmed the wrapper thickness.

The issue related to the center of gravity has been reviewed by both Westinghouse and AEP.

The center of gravity is 16.5 feet above the support pad faces.

This is a calculated value for the Westinghouse drawing and does not include any water, sludge,

closures, or other material.

tfyou have any questions, please contact me.

Sincerety Walter T. MacRae Attachments ER-98-009, Rev.

0 Appendix A, Page 1

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Westingh'ouse Electric Corporatioo Tampa DiviYion,P.O. Box 19218, Tampa, Florida 33616 ER-98-009, Rev.

0 Appendix A, Page 2

Westinghouse Proprietary Class 2C Westinghouse Electric Company, a 6ivision of CSS Corporation Energy Systems Hudear Services Division Box 356 Pitrsbursh, Pennsylvania 1523043S5 Mr. John Jensen American Electric Power Replacement Steam Generator Project Once One Cook Place Bridgman, MI49106 AEP-98-121 NSIKPM-98-126 August 4, 1998

Reference:

AEP RSG Data B/0 Checklist gtem N18)

American Electric Power Service Corporation D.C. Cook Units 1 and 2 S/G Non-LOCAData

Dear John,

Attached is the D. C. Cook, Unit 1, steam generator Non-LOCA geometric data requested by D. C. Cook. This information was originally faxcd to Phil Monk (AEP) on Friday, July 31, 1998, in rough format.

This data reflects input used in current non-LOCA licensing basis analyses for Unit 1. This data is for the current Unit 1, Model 51 steam generator design.

Similar data used in LOCA licensing basis analyses is being gathered and will be forwarded=when it becomes available.

Customer Projects Manager S hould you have any questions, please contact Mr. BillHicks on 412-374C734 or me on 412-374MS L Sincerely,

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0 Appendix A, Page 5

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0 Appendix A, Page 6

FIGUR6'.l - OuIIIIIt

RFP 2/19 Attachment H

Table 7

FSO-RSA-B6/3103

"= 4'

' '. C.

Cook nest Estimated Steam Genera or C

Content

~o Co-63 Co-58 I

Cr-51 sb-9$

Zr-9$

I Mn-5f Fe-59 Others~

t

• Other isotope C0-144 37 R7 21 7.0 2.4 1.4 1

Z 3.0 100 4nc1udo Co-57 7n-6$

Rg-103

~ ~

4m~K Z10 f53 1 19 40 14 8

7 1 7 7

)6, Sn-113.

Sb-125, and

&&I%40~

Table P.

I I

I I

I

p. C. Cook,'Vnkt 2 ttqeted Steam Genera or Le uric Content far Yar)aus Ttwes Act(

0 1 month 1 year 5 years lo years

~sembly

'hutdmnl ER-98-009, Rev.

0 Apperidix A, Page 7

Per t.mer Assembly 568 437 198 110 56 BS38S/)j q'l:t I

\\

1

~"

TABLE D.2 RECOMMENDED SCALING FACTORS FOR D.C.

COOK PASTE TYPE:

DRY ACTIVE HASTE SUBTYPE:

NO SUB"TYPE DATE:

OCTOBER 16, 1987 SCALED/SCALING NVCLIDES 14& /60~

55-Fe/ 60-Co 63-Ni/ 60~

90-Sr/137&s 99-Tc/137&s 129-I /137&s 238-Pu/239-Pu 241-Pu/239"Pu 241-Am/239-Pu 242~m/239-Pu 244-Qn/239"Pu 239-Pu/144 ~e 239-Pu/ 60&o SCALING

'ACTOR 9.55E-03 1.89E+00 4.19E-01 1.23E-02 3.44E-03 8;OOE-05 1.26EE.00 1.02E~02 5.74E-01 5.13E-01 5.78E-01 9.98E-03 3.30E-04

~

~

r e

~

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SELECTION METHOD

~

(NUMBER OF APPLIED RULE)

~

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3 Log~an 3 Sample 11642 8 Reference 7, Table 3-7 8 Reference 7, Table 3-7 "':

5 Log-mean 5 Log-mean 5 Log-mean 5 Log-mean 5 Log-mean 8 Reference 7, Table 3-19 3 Log-mean

~

~

r g4 Refers to the selection rule given in the SAIC "Scaling Factor Selection Method". Revision 4.

I~

ER-98-009, Rev.

0 Appendix A, Page 8

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APPENDIX B SHIPPING PAPERS AND DISPOSAL MANIFESTS (0 PAGES)

P'0 BE INCLUDEDIN FINAL WASTE CHARACTERIZATIONREPORTJ ER-98-009 Rev. 0 Page 17

APPENDIX C MICROSHIELD MODELS AND OUTPUT (2 PAGES)

ER-98-009 Rev. 0 Page 18

'age:

1 3OS File: DCCOOKCH.MS5

> 'ate: August 28, 1998

~ime: 8:18:26 AM

)uration: 00:00:11 WicroShield v5.01 (5.01-01072)

Chem-Nuclear Systems, Inc.

File Ref:

Date:

By:

Checked:

Case

Title:

DC Cook

==

Description:==

DC Cook, characterization model @30cm, normalized source Geometry: 7-Cylinder Volume - Side Shields Height Radius Source Dimensions 906.145 cm 29 ft 8.7 in 151.994 cm 4 ft 11.8 in Dose Points X

Y

¹ 1 201.45 cm 453.07 cm 6 ft 7.3 in 14 ft 10.4 in z

0cm 0.0 in Shield Name Source Shield 1

Shield 2 Shield 3 Shield 4 Transition AirGap Shields Dimension 6.58e+07 cm'.664cm

.965 cm 5.664 cm 7.163 cm Alloy600 Air A 533 Air A 533 Air Air 0.646 0.00122 7.86 0.00122 7.86 0.00122 0.00122 Material

~Deneit Nuclide Am-241 C-14 Cm-244 Co-60 Fe-55 Mn-54 Ni-63 Pu-238 Pu-239 Pu-241 Rh-106 Ru-106 Sb-125 Te-125m Source Input Grouping Method: Linear Energy Number of Groups: 15 Lower Energy Cutoff: 0.015 Photons < 0.015: Excluded Library: Grove becceuerels t/Ci/cm'.7121e+007 1.1146e-005 4.2180e+008 1.7334e-004 5.7720e+006 2.3721 e-006 1.1877e+010 4.881 0e-003 6.401 0e+009 2.6306e-003 5.1430e+ 005 2.1136e-007 1.7279e+010 7.101 0e-003 1.7020e+007 6.9945e-006 1.4578e+007 5.9910e-006 9.1 760 e+ 008 3.7710 e-004 5.2910 e+ 005 2.1744e-007 5.2910e+005 2.1744e-007 4.1810e+007 1.7182e-005 1.0249e+007 4.211 9e-006 cunes 7.3300e-004 1.1400e-002 1.5600e-004 3.2100e-001 1.7300 e-001 1.3900e-005 4.6700e-001 4.6000e-004 3.9400e-004 2.4800 e-002 1.4300 e-005 1.4300 e-005 1.1 300 e-003 2.7700e-004 Buildup The material reference is: Shield 4 Integration Parameters Radial Circumferential Y Direction (axial) 10 10 20 Bct/cm'.1239e-001 6.4137e+000 8.7766 e-002 1.8060e+002 9.7331e+001 7.8202e-003 2.6274e+002 2.5880e-001 2.2167e-001 1.3953e+001 8.0453e-003 8.0453e-003 6.3574e-001 1.5584e-001 ER-98-009, Rev.

0 Appendix C, Page 1

'age:

2

)OS File: DCCOOKCH.MS5

<un Date: August 28, 1998 Run Time: 8:18:26 AM

)uration: 00:00:11 EnercCr.

MeV 0.0356 0.1755 0.2279 0.371 0.4377 0.602 0.6546 0.835 1.0649 1.1732 1.3325 1.357 1.5622 A~ctivit hotons/sec 4.331e+07 3.368e+06 5.494e+04 8.758e+05 1.683e+07 9.586e+06 7.429e+06 5.164e+05 1.125e+04 1.188e+1.0 1.188e+10 3.067e+03 8.284e+02 Fluence Rate Results Fluence Rate

~No Buildu 8.790e-1 51 1.583e-07 3.304e-08 9.615e-06 3.965e-04 8.625e-04 9.334e-04 1.666e-04 9.002e-06 1.351e+01 2.118e+01 5.824e-06 2.519e-06 With Buildu 1.786e-26 1.245e-06 3.437e-07 1.183e-04 4:793 e-03 9.121e-03 9.403 e-03 1.426 e-03 6.412e-05 8.932e+01 1.271 e+02 3.450e-05

'.351e-05 MeV/cm'/sec MeV/cm'/sec Ex osure Rate mR/hr

~No Buildu 5.323e-1 53 2.710e-1 0 5.993e-11 1.864e-08 7.761e-07 1.683e-06 1.811e-06 3.152e-07 1.641 e-08 2.415e-02 3.675e-02 1.006e-08 4.192e-09 Ex osure Rate mR/hr With Buildu 1.082e-28 2.133 e-09 6.234e-1 0 2.293e-07 9.381 e-06 1.780e-05 1.824e-05 2.698e-06 1.169 e-07 1.596e-01 2.206e-01 5.958 e-08 2.248e-08 TOTALS:

2.384e+10 3A70e+01 2.165e+02 6.090e-02 3.802e-01 ER-98-009, Rev.0 Appendix C, Page 2

APPENDIX D SURFACE AREA CALCULATIONS (2 PAGES)

ER-98-009 Rev. 0 Page 19

D.C. Cook SGLA Tube and Channel Head Calculations

- 8/28/98 Channel Head Parameters "tubes'm p

b.=8.25 tu

'm3 "tube openings"

'" tubes Channel Head 8 Tube Dimensions rhead

=62.81

~ in ru max r u min

= 2.19 in

/0.875 ue

.875 rtube od 2

in Ltube sheet

=21. in L straight tube Channel Head Area Equations 2

~ X tube sheet

"'rhead

" tube openings'"'r tube I

A head 2'4 x r head'div~latc (2

x r head j.2 2'.

A tubesheet tubes "tube openings'x'r tube tube sheet Straight Tube Area Calculations straight tubes'tube openings'"'ube'traight tube Astraight tubes =5.886 lo'in'traight tubes 'm 7 ~

2 U-Bend Tube Area Calculation ru max+ru min r ave u bend 'ave'"'rtube'n tubes Au bend =8.037'10 'in Au bend =5.185 10 cm L ave

= x'ravc L ave = 97.436'in ER-98-009, Rev.

0 Appendix D, Page 1

Area Totals tube sheet A head =2.479'10 'in tb h t=

4 ~

2 Ahead =1.599'10

'cm Ad;v~late 1.239 10 'in 4 ~

2 A tubesheet tubes =3.465'10 'in S.

2 Adiv~lat'e 7.996'O

'cm

~

4 ~

Atubesheet tubes 2.235 10 'cm

~

6 ~

2 A channel head tube sheet+

head

~

div~late A channel head = 4.638'10 in

~

4 ~

2 Achannel head =2.992 10 cm S.

2 A tubes straight tubes+

u bend ~ Atubesheet tubes Atub s 7'036'IO 'in2 Atubes =4.539 10 'cm 7 ~

2 Straight Tube Bundle Density Calculation tube i tube od - rtube )'a I

2 2~.

tube metal'" 'traight tube'" tube openings tube metal 5.133 10 cm

~

6 ~

3 M tube tube metal'P tube M tube =4.235'10 gm Vbundle =6.56 10 cm M tube bundle bundle P bundle 0.646 gm cm ER-98-009, Rev.

0 Appendix 0, Page 2

AMERICAN ELECTRIC POWER DONALD C.

COOK NUCLEAR PLANT STEAM GENERATOR DISPOSAL EXEMPTION REQUEST ATTACHMENT 7 EVALUATION OF RESIDUAL WATER

t

'I

't%~

~t I

P'f I

7

$i'

Attachment 7

Page 1

Evaluation of Residual Water An evaluation was performed by AEP to determine the volume of residual water remaining in the steam generators.

This attachment summarizes the results of that evaluation.

The water remaining in the generators would be from the secondary, non-radioactive side of the system.

No radioactivity was contributed from the liquid and a

specific low specific activity (LSA) evaluation was not done.

Any radioactivity in the liquid is a

result of cross-contamination with internally deposited surface contamination.

In 1988, when the generators were taken out of service, they were allowed to drain by gravity.

All the material was.expected to drain except that material that could be trapped by a

plugged tube.

The water in the tubes could only enter a plugged tube from the secondary side of the generator.

Based on past experience, none of the steam generator tube plugs have leaked.

The steam generators were periodically tested throughout their in-service life for cracking and potential leaking tubes.

The method used to test the tubes is the standard industry practice termed eddy current testing.

Tube plugs were also subject to visual observation.

From the historical testing data, tubes with a potential for a through wall crack were selected as possibly containing water, and the volume of the tube was determined.

Tubes potentially susceptible to containing entrapped water were conservatively assumed to include tubes with abnormal indications located at the top of the tubesheet, tubes with support plate or anti-vibration bar wear/thinning indications exceeding 75%

throughwall, and three tubes for which test data was not readily available.

Tubes in a tube pull location were assumed to be only half filled because the tube remains open on one end, allowing a portion of the water to drain out.

The following table summarizes the results:

Steam Generator Identification Steam Generator 21 Steam Generator 22 Steam Generator 23 Steam Generator 24 TOTAL Volume of'ntrapped Water (gallons)

58. 62 147.06 213.63 143.81 563.12

AMERICAN ELECTRIC POWER DONALD C.

COOK NUCLEAR PLANT STEAM GENERATOR DISPOSAL EXEMPTION REQUEST ATTACHMENT 8 EVALUATION OF DOSE RATE

l 1'

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4" 4s

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