ML20116A110

Facility Reactor Vessel,Internals,Ex-Vessel Lead, Structural Steel & Reactor Compartment Concrete Shield Wall Radionuclide Inventory

ML20116A110
Person / Time
Site:	Saxton File:GPU Nuclear icon.png
Issue date:	12/12/1995
From:	Griffiths G, Levin A, Seymore F External (Affiliation Not Assigned)
To:
Shared Package
ML20115K054	List: ML20115K051 ML20115K082 ML20115K114 ML20115K141 ML20115K145 ML20115K165 ML20116A044 ML20116A058 ML20116A073 ML20116A081 ML20116A110 ML20116A119 ML20116A132
References
G01-1192-003, G01-1192-003-R00, G1-1192-3, G1-1192-3-R, NUDOCS 9607250298
Download: ML20116A110 (90)

Text

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ATTACHMENT 2

( REFERENCE 14 SUPPORTS QUESTION 30 RESPONSE 1'

THE SAXTON FACILITY l REACTOR VESSEL, INTERNALS, EX-VESSEL LEAD, STRUCTURAL STEEL AND REACTOR COMPARTMENT CONCRETE SHIELD WALL RADIONUCLIDE INVENTORY Document No. G01-1192-003 Prepared for:

GPU NUCLEAR December 1995 Prepared by: App ed y-hm NO M//2/9f f //ktkf Adam H. Levin Geoff/dy Griff)(@s, Project Afgr.

Tec ical Revi by: Approved by:

0%%4 w l2fuhr khk /W/ 9J' Caroly/A. Palmer, 'QA higr' Francis W. Se@iore, PE 9607250298 96071e MEEM PDR ADOCK 05000146 P PDR

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. 0

' PROPRIETARY INFORMATION . Section 0, Page 11 ofix FINAL REPORT 1

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l TABLE OF CONTENTS RE VIS I O N LO G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ix I 1

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1. EXE CUTIVE

SUMMARY

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2. I NTR O D U C TI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 - 1

3. THE SAXTON FACILITY OPERATING HISTORY AND CYCLE DATA ..... 3-1

4. MATERIAL COMPOSITIO NS .... .... .... ... ...... .. .. . ...... ........ ..... ..... . ... .. ... . . .. . ... .. . ..... 4- 1 1

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5. COMPONENT DIMENSIONS AND MATERIALS.......................................... 5-1 5.1 Co re B affle . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 1 i 52 The rm al S hie ld . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . .. . . . . . . . . . . 5- 1 5.3 Re actor Vessel Claddin g . ....... ... . .. . ........ ... . ...... .... . .. ........ . . ...... .. .. . ... . ... .... . . 5-2 5.4 Re acto r Ve s s el. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 -2 5.5 Reactor Ve ssel Insulation ..... .... . .... ... . ...... . ..... .......... .... .. . .. .. .... . . .. . .. ....... ... 5 2 5.6 Lower Core G uide B1ocks . ... .... .. . .... ...... . .... .. ............. . .. .. ...... ... .. ... .. .. .... ..... 5 3 l 5.7 Lowe r Co re Plate . . . . . . . . . .. . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3 5.8 Lower S upport Shroud Tubes . . ....... ... .. . ...... .............. .. ... .. .. . .. . . . .. . . . ... ..... .. . 5-3 5.9 Lower S upp ort Tie Ro ds .. . . .. .. . . . . . . . . .. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . . .. . . .. . . . . . . . 5 -3 5.10 Balance of the Lower Support Assembly ............. ...... ........................... 5-4 5.11 Lowe r Core B arre l . . . . . . . .. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4 5.12 Up pe r Core Pla te . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . .. . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 4 5.13 Up p er Core B arre l . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 4 5.14 Balance of the Upper Core Barrel Assembly...................... ................... 5 5 5.15 Ve ssel Insula tion C an . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . .. . . . .. . . . . . . . 5 5 5.16 Ve ssel S up p o rt C an . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . 5 5 5.17 Le a d S hie l din g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 - 5 5.18 S tructural S teel . . . . . . . . . . . .. . . . . .. . . . . . . . . . . ... . . .. . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 6 5.19 Reactor Comp artment Concre te ................. ......... ........................... . ..... . 5 6

6. - ACTIVATION ANALYSIS CALCULATIONS ... .......... ..... ... . ............ .......... 6-1 6.1 One Dimensional Neutron Transport Calculations............................... 6-1 6.2 Point Neutron Activation Calculations ................................................... 6-2 TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. 0

. PROPRIETARYINFORMATION Section 0. Page 111 ofix ,

FINAL REPORT l l

TABLE OF CONTENTS (cont.) l 1

6.3 The Saxton Facility Operating History .................................................. 6-3  !

6.4 The Saxton Facility Radial ANISN Models............................................ 6 3  !

6.4.1 Radial Model Geome tries.............................................................. 6-4 6.4.2 Radial Model Compositions .......................................................... 6-5 l 6.4.3 Radial Model Calculations .................................. .......................... 6-6  !

l 6.5 The Saxton Facility Axial ANISN Models.............................................. 6 7 l 6.5.1 Axial Model Geometries ................................................................ 6-7 6.5.2 . Axial Model Compositions ........................................................... . 6-8 6.5.3 Axial Model Calculations .............................................................. 6 8 6.6 Results......................................................................................................6-9

7. ACTIVATION ANALYSIS RESULTS AND COMPARISON TO SAXTON. 7-1 CHARACTERIZATION DATA 7.1 ~ Curie Contents at Twenty-Four Years After Reactor Shutdown.......... 7-1 7.2 10 CFR Part 61 Classification ................................................................. 7-1 7.3 Discussio n of Re sults . .. . . . . . . . .. . . .. . . . ... . . . .. ... .. . . . .. . . . . . . . . . . . . . . . . . . . . .. .. . .. . . .. . . .. . . .. . . .. 7-2 7.3.1 Induced Radioactivity Twenty-Four Years After Shutdown of.. 7-2 the Saxton Reactor Pressure Vessel, Internals, Vessel Insulation Can and Vessel Support Can 7.3.2 Induced Radioactivity Twenty-Four Years After Shutdown of . 7 3 Components and Materials Outside the Reactor Vessel Support Can 7.4 Comparison of Activation Analysis Results to Saxton................... ....... 7-4 Characterization Data 7.4.1 Comparison to TLD String Data .................................................. 7 5 7.4.2 Comparison to External Exposure Rate Measurements ............ 7-6 7.4.3 Comparison to Internal Exposure Rate Measurements............. 7-6 TLG SERVICES

SAXTON ACTIVATION ANALYSIS Documeae No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 0. Page iv ofix l FINAL REPORT l

i TABLE OF CONTENTS (cont.)

l l 7.4.4 Comparison to Concrete Boring Samples and............................. 7-7 Stainless Steel Samples from Above l the Operating Water Level 7.4.5 Comparison to Structural Stainless Steel Sample...................... 7-8 at Core Midplane

8. RE FE RE N C E S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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- SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 0, Page v ofix FINAL REPORT l

I LIST OF FIGURES '

6.1 The Saxton Facility Radial Component ANISN Core Layout ..................... 6-18 and Model Segmentation 6.2 The Saxton Facility Radial Component ANISN Model ........................... 6-19 Geometry Overview 6.3 The Saxton Facility Radial Component ANISN Model ........................... 6-20 ,

Thermal Neutron Flux 1

6.4 The Saxton Facility Radial Concrete ANISN Model ............................... 6-21 Thermal Neutron Flux 6.5 The Saxton Facility Radial Water ANISN Model .................................... 6-22  ;

Thermal Neutron Flux  ;

6.6 The Saxton Facility Axial Component ANISN Model ............................. 6-23 Thermal Neutron Flux i

6.7 The Saxton Facility Axial Water ANISN Model ...................................... 6-24 )

Thermal Neutron Flux TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 0, Page vi ofix F.!NAL REPORT LIST OF TABLES E.1 The Saxton Facility Neutron-Induced Curie Contents .................................1-2 of the Reactor Pressure Vessel and Internals E.2 The Saxton Facility Neutron-Induced Curie Contents .................................1-3 Ex-Vessel Lead, Structural Steel and Reactor Compartment Concrete Shield Wall 3.1 The Saxton Facility Operating History .............. .......................................... 3-2 1

4.1 Material Composition - Type 304 Stainless Steel ....................................... 4-2 l l

4.2 Material Composition - Carbon Steel .... ....... . .................. ... .......... ... ........... 4-3 4.3 Material Composition - Ordinary Concrete................................................. 4-4 )

4.4 Material Composition - Ordinary Lead ....................................................... 4-5 ,

1 6.1 Cra nbe rg Fission Sp e etrum . . ... . .. .. . . .. . ... . . ... . .. . . . . .. . ...... . .. . . . .. . . .. . .. . . .. .. . . . . . ... . ... . .. 6- 10 6.E The Saxton Regional Composition Summaries for........................................ 6-11 Radial and Axial Component and Water Model Geometries 7.1 The Saxton Facility Component Activation ............................................. 7-9 i Twenty-Four Years After Reactor Shutdown - Core Baffle l 7.2 The Saxton Facility Component Activation ............................................. 7-10 Twenty-Four Years After Reactor Shutdown - Thermal Shield 7.3 The Saxton Facility Component Activation ... ......................................... 7-11 Twenty-Four Years After Reactor Shutdown - Reactor Vessel Clad 7.4 The Saxton Facility Component Activation ............................................. 7-12 Twenty-Four Years After Reactor Shutdown - Reactor Vessel Wall TLG SERVICES

g SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O i PROPRIETARYINFORMATION Section 0, Page vil ofix l FINAL REPORT i  !

l LIST OF TABLES (cont.)

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7.5 The Saxton Facility Component Activation ............................................. 7-13 2

Twenty-Four Years After Reactor Shutdown - Vessel Insulation Can 7.6 The Saxton Facility Component Activation ............................................. 7-14 l l . Twenty-Four Years After Reactor Shutdown -Vessel Support Can 4

i 7.7 The Saxton Facility Component Activation ............................................. 7-15

Twenty-Four Years After Reactor Shutdown -

Lower Core Plate Guide Blocks 7.8 The Saxton Facility Component Activation ............................................. 7-16 Twenty-Four Years After Reactor Shutdown - Lower Core Plate d

7.9 The Saxton Facility Component Activation ........................... ................. 7-17 l

! Twenty-Four Years After Reactor Shutdown -

Lower Support Shroud Tubes 7.10 The Saxton Facility Component Activation ............................................. 7-18 j j Twenty-Four Years After Reactor Shutdown - Lower Support Tie Rods 7.11 The Saxton Facility Component Activation ............................................. 7-19 i t

Twenty-Four Years After Reactor Shutdown -

Balance of Lower Support Assembly -

7.12 ~ The Saxton Facility Component Activation ............................................. 7-20 Twenty-Four Years After Reactor Shutdown - Lower Core Barrel 7.13 The Saxton Facility Component Activation .................. .......................... 7-21 Twenty-Four Years After Reactor Shutdown - Upper Core Plate-7.14 The Saxton Facility Component Activation - .................................. .......... 7-22 c Twenty-Four Years After Reactor Shutdown - Upper Core Barrel 7.15 The Saxton Facility Component Activation ....... ... ................................. 7-23 Twenty-Four Years After Reactor Shutdown -

Balance of Lower Core Barrel Assembly TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARY INFORMATION Section 0, Page vill ofix FINAL REPORT LIST OF TABLES (cont.)

7.16 The Saxton Facility Component Activation ............................................. 7-24 Twenty-Four Years After Reactor Shutdown -

Balance of Upper Core Barrel Assembly 7.17 The Saxton Facility Component Activation ............................................. 7-25 ,

Twenty-Four Years After Reactor Shutdown - Curie Contents for i Reactor Vessel and Internals 7.18 The Saxton Facility Component Activation ...... ............. . ...................... 7-26 Twenty-Four Years After Reactor Shutdown -

Specific Activities for Ex-Vessel Stainless Steel, Lead and Concrete 7.19 The Saxton Facility Component Activation ............................................... 7-27 Twenty-Four Years After Reactor Shutdown - 10 CFR Part 61 Classification 1

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SAXTON ACTIVATION ANALYSIS Document No. G01-1192-008, Rev. O PROPRIETARYINFORMATION Section 0, Page ix ofix FINAL REPORT I

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SAXTON ACTIVATION ANALYSIS Document No. G01-1192-008, Rev. O PROPRIETARYINFORMATION Section 1, Page 1 of 4 l FINAL REPORT )

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1. EXECUTIVE

SUMMARY

TLG Services, Inc. (TLG) has performed a series of one dimensional neutron transport calculations, coupled with point neutron activation analyses, for the Saxton facility (Saxton) reactor vessel, internals, surrounding lead, structural steel and concrete. Additionally provided herein are estimates of the 10 CFR Part 61 waste classification for the reactor vessel, its internals, the vessel insulation can and the vessel support can. All other lead shield, structural steel and concrete components are expected to be Class A material. The j analysis scenario considered was a final shutdown of the facility on 2 April 1972, after three  ;

cycles of both power and experimental operation, with an ensuing decay period of 24 years to 1 July 1996.

This activation analysis has been perfonned in support of the proposed Saxton

. Decommissioning Project. The results are an integral part of the radioactive materials characterization of the Saxton facility. They will assist in the determination of packaging,

- transporting and disposal requirements of the neutron irradiated portions of the facility.

This analysis indicates approximately 1452 curies of neutron activation products, mainly Iron-55, Cobalt-60 and Nickel-63, will be present in the reactor vessel wall and clad, l internals, insulation can and support can 24 years after shutdown (1 July 1996). The

maxunum specific radionuclide activity of any individual component (averaged over the l component) is estimated to be 2.66 millicuries per gram for the lower core guide blocks. The i

curie estimate does not include the inventory of contamination expected on reactor vessel i and internals surfaces.

1 i Table E.1 provides a detailed breakdown of the radioactive inventory distribution for the j vessel, insulation and support cans. Table E.2 furnishes expected ^ specific radionuclide

contents for lead, structural steel and concrete found in close proximity to the reactor vessel, inside the reactor compartment. Two of the four ex-core neutron chambers supported by structural steel around the core centerline - outside the vessel support cylinder - contain

some lead shielding. Activation analyses of the lead shields reveal maxunum specific

activities on the order of 50.0 pCi/kg. All neutron irradiated lead will be 10 CFR Part 61

!~ ' Class A radioactive waste.

i The concrete walls of the reactor compartment below the operating water level show l insignificant neutron activation. Insignificant neutron activation implies this region will

likely be below the proposed NRC release criteria for volumetric release of structural l material. The concerns related to this region are primarily centered upon surface s

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SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 1, Page 2 of 4 FINAL REPORT decontamination and radioactive material located in cracks and crevices in the concrete wall itself.

Above the operating water level, there appears to be significant neutron activation of the l reactor compartment concrete walls. In some regions, TLG estimates as much as 15 inches

! (38.1 Cm.) of wall will require removal to meet release limits. This is highly dependent upon l the azimuthal position of the concrete relative to the orientation of the core (generally the northern side of the compartment wall received the highest neutron exposure).

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TABLE E.1 THE SAXTON FACILITY NEUTRON-INDUCED CURIE CONTENTS OF THE REACTOR PRESSURE VESSEL AND INTERNALS TWEN'IT-FOUR YEARS AFTER REACTOR SHUTDOWN Weight of Weight of Total Curies 24 Years C3trngorant . Component Ubs.) Componentikjk) After Shutdown Core Bafne 1210 549 642.6 Thermal Shield 10491 4759 132.7 ReactorVessel Clad 2043 927 4.185 ReactorVesselWall 87313 39604 4.737 Reactor Vesselinsulation Can (1) 4994 2265 1.078 ReactorVessel Support Can (1) 5442 2468 3.034 Lower Core Guide Blocks 152 69 182.9 Lower Core Plate 500 227 226.7 Lower Support Shroud Tubes (2) 338 153 8.323 Lower Support The Rods (2) 54 24 2.113 Balance Lower Support Assembly (3) 706 320 0.07440 Lower Core Barrel (4) 1067 484 58.46 Upper Core Plate 264 120 165.0 Upper Core Barrel (4) 454 206 19.71 Balance Lower Core Barrel Assembly (5) 1823 827 0.2593 Balance Upper Core Barrel Assembly (6) 832 377 0.1183 Notes: 1. Indudes weight of component between 3 feet above and 3 feet below the active core.

2. Includes weight of component down to 3 feet below the active core.

3. Weight of lower support shroud tubes, lower support tie rods and associated components more than 3 feet below the active core.

4. Includes weight of component up to 3 feet above the active core.

5. Weight of lower core barrel and associated componen+a more than 3 feet above the active core.

6. Weight of upper core barrel and associated components more that 3 feet above the active core, i TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARY INFORMATION Section 1, Page 3 of 4 FINAL REPORT TABLE E.2 THE SAXTON FACILITY NEUTRON-INDUCED CURIE CONTENTS EX-VESSEL LEAD, STRUCTURAL STEEL AND REACTOR COMPARTMENT CONCRETE l SHIELD WALL l l

TWENTY-FOUR YEARS AFTER REACTOR SHUTDOWN l l

Location Curiggtag.rEntm326.QE/ v-112 l C9_rriennant Structural Steel ist 6* Beyond Vessel Support Can - Core Centerline 1.498E-06 / N A Structural Steel 2nd 6* Beyond Vessel Cupport Can - Core Centerline 1.938E-08 / NA Structural Steel 3rd 6* Beyond Vessel Support Can - Core Centerline 1.442E-09 / N A Structural Steel 4th 6* Beyond Vessel Support Can - Core Centerline 1.916E-10 / N A Structural Steel 5th 6* Beyond Vessel Support Can - Core Centerline 3.093E-11/ N A Structural Steel 6th 6* Beyond Vessel Support Can - Core Centertine 5.493E-12 / NA Structural Steel 7th 6* Beyond Vessel Support Can - Core Centerline 1.140E-12 / NA Structural Steel 8th 6* Beyond Vessel Support Can - Core Centerline 2.524E-13 / NA Reactor Compartment Concrete Wall ist 1* Concrete- Core Centerline 1.176E-15 / 6.447E-15 R: actor Compartment Concrete Wall 2nd 1* Concrete-Core Centerline 1.146E-15 / 6.336E-15 R' actor Compartment Concrete Wall 3rd 1* Concrete-Core Centeriine 1.099E-15 / 6.132E-15 R; actor Compartment Concrete Wall 4th 1* Concrete -Core Centertine 1.042E-15 / 5.864E-15 Reactor Compartment ConcreteWall 5th 1* Concrete -Core Centerline 9.777E-16 / 5.545E-15 Reactor Compartment Concrete Wall 6th 1* Concrete -Core Centerline 9.082E-16 / 5.190E-15 RIactor Compartment Concrete Wall 3rd 6* Concrete -Core Centedine 7.609E-16 / 4.401E-15 Reactor Compartment Concrete Wall 4th 6* Concrete-Core Centertine 5.397E-16 / 3.166E-15 Reactor Compartment Concrete Wall 5th 6* Concrete- Core Centerline 3.313E 16 /1.957E-15 Reactor Compartment Concrete Wait 6th 6* Concrete -Core Centerline 1.369E-16 / 8.027E-16 Reacter Compartment Concrete Wall ist 1* Concrete- Above Water Line 4.644E-13 / 2.295E-12 Reactor Compartment Concrete Wall 2nd 1* Concrete - Above Water Line 4.886E-13 / 2.537E-12 Reactor Compartment Concrete Wall 3rd 1* Concrete- Above Water Line 4.090E-13 / 2.649E-12 Reactor Compartment Concrete Wall 4th 1* Concrete - Above Water Line 4.751E 13 / 2.641E-12 Reactor Compartment Concrete Wall 5th 1* Concrete- Above Water Line 4.458E-13 / 2.539E-12 Reactor Compartment Concrete Wall 6th 1* Concrete- Above Water Line 4.080E-13 / 2.369E-12 Reactor Compartment Concrete Wall 3rd 6* Concrete - Above Water Line 3.210E-12 /1.918E-12 Reactor Compartment Concrete Wall 4th 6* Concrete- Above Water Line 1.982E-13 /1.220E-12 Reactor Compartment Concrete Wall 5th 6* Concrete- Above Water Line 1.045E-13 / 6.544E-13 Reactor Compartment Concrete Wall 6th 6* Ccacrete- Above Water Line 3.806E 14 / 2.402E 13 Localion Cyr_teiper._ Gram. Aa-1_Q3/AqiQQ.rn Cam _.ponent Lead 1st 6" Beyond Vessel Support Can - Core Centerline 3.376E-10 / 3.793E-09 2nd 6* Beyond Vassel Support Can - Core Centertine 4.349E-12 / 4.886E-11 Lead Lead 3rd 6* Beyond Vessel Support Can - Core Centertine 3.138E-13 / 3.52SE-12 The Calculations Consist of one-dimensional neutron transport and point neutron activation analyses of the reactor vessel, its internals, surrounding lead, structural steel and Concrete. They have been performed using the FISSPEC [1] and 02 FLUX [2]

Computer Codes, written by TLG; and the ANISN [3] and ORIGEN2 [4] Computer Codes, obtained through the Oak Ridge National Laboratory's Radiation Shielding Information Center (RSIC). Reduction of the output from these programs and ancillary Calculations I

TLG SERVICES

SAXTON ACTIVATION ANALYSIS Do2ument No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 1, Page 4 of 4 FINAL REPORT were performed using the ANISNOUT [5] and 02 READ [6] computer codes, written by TLG; the Microsoft EXCEL [7] spreadsheet application and the MICROSHIELD [8] point- '

kernel radiation shielding code. l The results of these calculations have been compared to characterization data taken by General Public Utilities Nuclear (GPU Nuclear) employees during the second half of 1995.

In general, the analytical results compare favorably with the measured values. There was some disparity between the TLG-calculated curie contents for concrete and structural steel above the water level, and the concrete core and steel samples taken during the characterization effort. As decommissioning work proceeds, additional structural steel and concrete core samples in the reactor compartment above the water line should be analyzed to assist in the determination of concrete removal depths and disposition of structural steel.

Based upon currently existing federal regulations regarding the transportation of radioactive materials, TLG believes the reactor vessel and its internals may be shipped as a greater-than-Type A quantity, LSA package. The 10 CFR Part 61 classification of the reactor pressure vessel and its internals demonstrates that the waste package meets the requirements to be classiSed as Class C waste. However, final classification of the package will also depend upon surface contamination levels and the results of a visual inspection inside the reactor vessel. Under current federal regulations, this package can be disposed of in currently-existing, near-surface, low-level radioactive waste land disposal facility.

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SAXTON ACTIVATION ANALYSIS - Document No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 2, Page 1 of 1 FINAL REPORT i

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2. INTRODUCTION TLG estimated the neutrou-induced radionuclide inventories at the Saxton facility (Saxton)  ;

using a two-step analytical approach. The first step was to determine the magnitude and ,

spretrum of the neutron flux within and beyond the radial and axial boundaries of the reactor core. This was accomplished using the ANISN one-dimensional neutmn transport computer code with eleven radial and axial geometric models.

The second step was to collapse the ANISN neutron transport calculation outputs into two energy group formats'(fast and thermal, with a 0.411 eV thermal energy breakpoint) and I input the data into a series of ORIGEN2 point activation / depletion calculations. Additional

-input to the ORIGEN2 calculations included historical plant performance data described in o

Section 3 and material compositions found in Section 4. A description of the components 1 considered in this analysis is located in Section 5. Wherever possible, plant operations or

- design data, obtained through General Public Utilities Nuclear (GPU Nuclear) has been employed. The balance of the input data was extracted from available literature and is  !

referenced as such, or interpolated from available data. Conservative assumptions were l made and justified whenever actual plant data was not available.

l Details of the calculations performed may be found in Section 6 of this report. Included are j descriptions of the ANISN and ORIGEN modals and how the data was transferred between  ;

them. Due to asymmetric loading of the reactor core, the core geometry was divided into 8 1 azimuthal segments, from which 4 pairs with equal fuel geometries were derived. The model geometry is shown pictorially in Section 6 of this report. Within each 1/8th azimuthal segment, the azimuthal peak-to-average ratio activation of components is assumed to be 1.414. An axial peaking factor of 1.65 was used from the Saxton design basis documentation.

, The operating water level design basis was located at 731'-8" elevation, approximately 36 inches (91.44 cm.) above the top of the active core region. A visual observation of the water 13 vel (by staining on the reactor compartment concrete wall) was made and confirms that the water level was maintained roughly around the design basis. Section 7 provides results of the activation calculations along with a discussion of how they compare to the field characterization data obtained by GPU Nuclear.

i The material contained herein satisfies the requirements of the task authorized under GPU  ;

Nuclear Contract Number PC-0504936, " Task Order for Saxton Decommissioning Plan

! Work", dated 20 October 1994. All work was performed under TLG's Quality ' Assurance l Program. 1 TLG SERVICES

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3. THE SAXTON FACILITY OPERATING HISTORY AND CYCLE DATA The Saxton facility's operating history and performance data was obtained through operating records provided by GPU Nuclear [9]. The data received included megawatt-thermal operations through the life of the plant. The GPU Nuclear data was employed to estimate the effective full power days (EFPD) and capactiy factor for each cycle. The op: rating history, as compiled by TLG from GPU Nuclear data, is presented in Table 3.1.

GPU Nuclear data was used to determine the burnup characteristics (average neutron flux and actinide mix) of the average fuel assembly via an ORIGEN2 point activation / depletion calculation. Core I was all UO2 fuel assemblies. The second and third cores were both mixed oxide (PuO2 and UO2) and UO2 fuel. TLG used archived GPU Nuclear data which indicated that only UO2 fuel was used on the core periphery [10] (having the most significant impact upon the neutron spectrum beyond the radial core boundary) for both cores.

Core II used all new fuel UO2 fuel assemblies (with one exception) on the core periphery.

Therefore, TLG has assumed the periphery of Core II was loaded with all new fuel assemblies of the same design as Core I. Core III utilized UO2 assemblies from Core I for 11 of the 12 peripherallocations. The 12th fuel assembly was a new UO2 assembly. In order to generate in-core total neutron fluxes and actinides, TLG assumed Core I contained all fresh fuel; Core II had fresh fuel; and Core III contained fuel irradiated previously in Core I. The average fuel assembly was assumed to have an initial enrichment of 5.7 w/0 22U and an assembly burnup of 9,675 mwd /MTU,10,235 MWdSITU and 6,553 MWdaiTU in Cores I, II and III, respectively. An average moderator boron concentration of 828 ppm was estimated from available GPU Nuclear data [11], and assumed for this analysis.

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l TABLE 3.1 THE SAXTON FACILITY OPERATING HISTORY average Date Date Gross Energy Cycle Number Cycle Total Beginning .End Number M Wth-br EFPD Capacity ofDays Neutron Flux Sectors eLCxcle ofcycle ofJ2nyn In cvcle lo C.vc!c Factor Refuelina Reactork C:re 1 4/1W62 8/16/65 1222 207360 367.66 30.09 % 113 4.363E+13 Core il 12/6/65 9/20/68 1020 219349 388.92 38.13 % 448 5.567E+13 Core 111 12/11/69 4/2/72 844 140201 248.58 29.45 % 4.595E+13 Overall Plant Capacity: 27.59 %

Overall MWth Gross: 566,910 i

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4. MATERIAL COMPOSITIONS l

In order to determine the neutron induced radionuclide inventories, four material compositions were evaluated in this analysis: Type 304 stainless steel, carbon steel, j _ concrete and ordinary lead. There were other materials subject to neutron irradiation; l however, from the review of plant documents, there does not appear to be any other i

component of sigmficant mass or proximity to the core which would significantly alter the results of this study.

The elemental compositions of the four materials considered in this analysis may be found in Tables 4.1 through 4.4. Information in these tables was extracted from References

[12,13,14,15].

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MATERIAL COMPOSITION- TYPE 304 STAINLESS STEEL j Grams Per Gram Material  ;

H Ge Eu 2.000E-08 He As 1.940E-04 Gd U 1.300E-07 Se 3.500E 05 Tb 4.700E47 Be Br 2.000E 06 Dy 1.000E-06 B Kr Ho 1.000E-06 C 3.000E-04 Rb 1.000E-05 Er N 4.520E-04 Sr 2.000E-07 Tm O Y 5.000E 06 Yb 2.000E 06 F Zr 1.000E 05 Lu 8.000E-07 Ne Nb 8.900E-05 Hf 2.000E46 Na 9.700E 06 Mo 2.600E-03 Ta Mg Tc W 1.860E-04 Al 1.000E-04 Ru Re Si 1.000E 02 Rh Os 1 P 4.500E 04 Pd Ir S 3.000E 04 Ag 2.000E-06 Pt Cl 7.000E 05 Cd Au Ar in Hg K 3.000E-06 Sn TI ,

l Ca 1.900E 05 Sb 1.230E-05 Pb 6.700E-05 Sc 3,000E48 Ta Bi Tl 6.000E-04 i Po V 4.560E-04 Xe At Cr 1.900E-01 Cs 3 000E-07 Rn Mn 2.000E-02 Ba 5 000E-07 Fr Fe 6.790E-01 La 2.000E-07 Ra Co 1.414E-03 Ce 3.710E 04 Ac Ni 1.000E-01 Pr Th 1.000E-06 Cu 3.080E 03 Nd Pa Zn 4.570E 04 Pm U 2.000E-06 Ca 1.290E-04 Sm 1.000E-07 l

Note: Tabulated data may not sum to 1.000 due to the use of several sources for the data contained therein.

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l l l 1 l TABLE 4.2 MATERIAL COMPOSITION- CARBON STEEL Grams Per Gram Material H Ge Eu 3.100E-06 He As 5.320E-04 Gd U 3.000E-07 Se 7.000E-07 Tb 4.500E 07 Be Br 8.500E 07 Dy I i

B Kr Ho 8.000E-07 C 2.900E-03 Rb 4.800E45 Er l

N 8.400E-05 Sr 1.500E-07 Tm O Y 2.000E-05 Yb 1.000E-06 F Zr 1.000E-05 Lu 2.000E-07 Ne Nb 1.880E-05 Hf 2.100E47 Na 2.300E-05 Mo 5.600E-07 Ta 1.300E-07 Mg Tc W 5.500E 06 i Al 3.300E-04 Ru Re Si Rh Os P 4.000E-04 Pd Ir S 5.000E-04 Ag 2.000E-06 Pt C1 4.000E45 Cd Au At in Hg K 1.200E 05 Sn TI Ca 1,400E 05 Sb 1.100E-05 Pb 8.200E 04 i Sc 2.600E-07 Te Bi Ti 2.000E46 i Po V 8.000E-05 Xe At l Cr 1.700E-03 Cs 2.000E 07 Rn Mn 1.000E42 Ba 2.730E-04 Fr Fe 9 842E 1 La 1.000E-07 Ra I

Co 1.220E-04 Ce 1000E-06 Ac Ni 6.600E 03 Pr Th 1.800E-07 Cu 2.000E-03 Nd Pa .!

Zn 1.000E 04 Pm U 2.000E-07 Ga 8.000E-05 Sm 1.700E-08 .

. i Note. Tabulated data may not sum to 1.000 due to the use of several sources for the data contained therein.  !

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TABLE 4.3 MATERIAL COMPOSITION- ORDINARY CONCRETE Grams Per Gram Material H 1.000E-02 Ge Eu 5.500E-07 He As 7.900E-06 Gd U 2.000E45 Se 9.200E-07 Tb 4.100E47 Be Br 2.400E-06 Dy 2.300E-06 4 B 2.000E-05 Kr Ho 9.000E-07 1

C 1.000E-03 Rb 3.500E 05 Er N 1.200E-04 Sr 4.380E-04 Tm O 5.290E 01 Y 1.820E-05 Yb 1.400E46 F Zr 7.100E 05 Lu 2.700E 07 Ne Nb 4.300E-06 Hf 2.200E 06 Na 1.600E-02 Mo 1.030E-05 Ta 4.400E47 Mg 2.000E-03 Tc W 1.400E-06 Al 3.400E-02 Ru Re a

Si 3 370E-01 Rh Os 4

P 5.000E-03 Pd 3.000E-06 tr 1

S 3.100E-03 Ag 2.000E 07 Pt Cl 4 500E-05 Cd 3.000E-07 Au At in Hg a

K 1.300E-02 Sn 7.000E-06 TI Ca 4 400E-02 Sb 1.800E.06 PD 6.100E 05 j Sc 6 500E-06 Te Bi Ti 2.121E-03 i Po V 1.030E 04 Xe At Cr 1.090E-04 Cs 1.300E-06 Rn Mn 3.770E-04 Ba 9.500E44 Fr Fe 1.400E-02 La 1.300E-05 Ra Co 9 800E-06 Ce 2.430E-05 Ac Ni 3.800E-05 Pr Th 3.500E-06 Cu 2.500E-05 Nd Pa Zn 7.500E-05 Pm U 2.700E-06 Ga 8 800E-06 Sm 2.000E 06 Note: Tabulated data may not sum 101.000 due to the use of several sources for the data contained therein.

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SAXTON ACTIVATION ANALYSIS Document Nc. G01-1192-003, Rzv. O PROPRIETARYINFORMATION Section 4, Page 5 of 5 FIN #1 REPORT TABLE 4.4 MATERIAL COMPOSITION- ORDINARY LEAD Grams Per Gram Material H Ge Eu He As 2.000E 05 Gd U Se Tb Be Br Dy B Kr Ho C Rb Er N Sr Tm O Y Yb F Zr Lu Ne Nb Hf Na Mo Ta Mg Tc W Al Ru Re Si Rh Os P Pd Ir S Ag 1.100E 04 Pt Cl Cd Au Ar in Hg K Sn 2.0,00E-05 TI Ca Sb 2.000E-05 Pb 9.900E-01 Sc Te Bi 5.000E-05 Po I Ti i l

V Xe At Cr Cs Rn Mn Ba Fr Fe 2.000E-05 La Ra Co Ce Ac Ni Pr Th Cu 6.000E-04 Nd Pa Zn 1.000E-05 Pm U Ga Sm Note: Tabulated data may not sum to 1.000 due to the use of several sources for the data contained therein.

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5. COMPONENT DIMENSIONS AND MATERIALS Th3 Saxton facility is a small, Westinghouse pressurized water nuclear reactor (PWR),

originally designed for research and development as well as power generation. The reactor pressure vessel was designed by Westinghouse and fabricated by A.O. Smith. This section

. describes each component within or immediately adjacent to the reactor vessel at Saxton considered in this study. In general, the activation analysis model consists of portions of the reactor internals, the reactor vessel, the vessel insulation can, the vessel support can, surrounding lead, structural steel and concrete primary shield wall three feet above and below the active fuel core. Internal components outside the region of the activation analysis model are discussed briefly below. Unless stated otherwise, all components are austenitic stainless steel, i

Many of the components in this study fall within two or more regions of the neutron activation analysis model and as such, Curie-per gram values represent only an average speciSc activity over the volume of the component considered. The weights of components discussed below were generally derived from assembly drawings, by determining the dimensions of the component and converting to component weight. The following describes the individual major components considered in the activation analysis model.

5.1 Core Baffle The core baffle (actually the lower portion of the lower core barrel at Saxton) is a 3/8" thick, I box like structure immediately surrounding the reactor core. [Other pieces of the baffle l assembly include baffle lower and upper plates, eight lengths of 3 inch (7.62 cm.) Schedule 40 pipe runnmg vertically between the plates and outside the box-like structure, and I gussets providing structural strength to the plate / box structure assembly.] TLG calculated an equivalent cylinder inner radius of the core bafHe of 17.328 inches (44.01 cm.) for input j into the ANISN geometry. The estimated weight of the bafHe assembly as used in the '

activation / depletion calculations was 1,210 pounds (549 kg.).

5.2 Thermal Shield -

The thermal shield is a cylindrical barrel supported at the top by pins welded to the vessel inner wall, and at the bottom by lugs, also welded to the vessel inner wall. This component provides gamma ray shielding to reduce gamma ray heating and heat-related stresses in the l

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reactor pressure vessel wall. The barrel is 3.0 inches (7.62 cm.) thick with an inner radius of 24.5 inches (62.23 cm.), weighing 10,491 pounds (4,759 kg.).

i l 5.3 Reactor Vessel Cladding The reactor vessel is lined with a nominal 0.125-inch (0.32 cm.) stainless steel cladding.

l' This clad serves to prevent the carbon steel pressure vessel from directly contacting the i reactor coolant. The inside radius of the cladding is a nominal 28.875 inches (73.34 cm.).

i Approximately 2,043 pounds (927 kg.) of cladding were included in the activation analysis I models.

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i j 5.4 Reactor Vessel )

l The reactor vessel consists of a laminated plate cylindrical shell, a spherically-dished bottom i head and a removable reactor closure head. The reactor closure head is approximately j spherically-dished, and welded to a matchmg ring flange. The pressure vessel and its head f are constructed of carbon steel. The reactor vessel has a nominal inside radius of 29.00

}* inches (73.66 cm.) and a minimum thickness of 5.00 inches (12.7 cm.) at the core midplane.

The activation analysis models included 87,313 pounds (39,604 kg.) of the reactor vessel, i including the closure head.

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l 5.5 Reactor VesselInsulation l Insulation surrounding the reactor pressure vessel is fiberglass. In some regions, the i Eberglass has a metal mesh blanket with 1.0-inch (2.54 cm.) hexagonal poultry netting on

the inside to provide support. The meshed blanket is banded to the reactor shell by 0.75-inch (1.91 cm.) wide by 0.020-inch (0.05 cm.) thick stainless steel bands spaced on 9.0-inch l (22.86 cm.) centers. Other regions contained fiberglass simply packed inside the insulation can. In all regions, the Eberglass is packed to a density of four pounds per cubic foot (0.064 l gm/cm3).

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! TLG conservatively assumed a void region in place of the fiberglass insulation for the neutron transport calculations. Several inches of fiberglass will not significantly perturb the

.,i neutron flux emanating from the reactor vessel.

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5.6 Lower Core Guide Blocks 1 The lower core guide blocks consist generally of two major pieces. The lower block portion l

, contains a large hole into which the lower end fitting of a fuel assembly was located when i inserted in the reactor core. Attached to and above the block portion is a shroud encompassing slightly more than two sides of an assembly and less than three inches (7.62 cm.) tall. This shroud provides additional assembly locating. One lower core block guide assembly is found beneath each active fuel and dummy assembly, attached to the lower core '

plate. The total weight oflower core guide blocks included in the analysis is 152 pounds (69 kg.).

d j 5.7 Lower Core Plate The lower core plate is a 1.5-inch (3.81 cm.) thick plate, 43.5 inches (110.49 cm.) in diameter.

The lower core plate is the lower support pad for the reactor core and provides a grid structure locating the cruciform control rod blades (and fuel followers). A weight of 500 pounds (227 kg.) was calculated for the model.

5.8 Lower Support Shroud Tubes The lower support shroud tubes provide support feet for the lower core plate and enclosed columns up through which the fuel followers (attached to the control rods) move. The tubes down to three feet below the core were included directly in the activation analysis models, having a total weight of 338 pounds (153 kg.). The balance of the lower support shroud tubes were included in the " Balance of the Lower Support Assembly" found below.

5.9 Lower Support Tie Rods The lower support tie rods connect the lower core plate to the spider assembly located near the bottom of the reactor vessel. They are found around the periphery of the lower core plate. As with the lower support shroud tubes, down to three feet below the core were included directly in the activation analysis models, having a total weight of 54 pounds (24 kg.). The balance of the lower support tie rods were also included in the " Balance of the Lower Support Assembly" found below.

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5.10 Balance of the Lower Support Assembly

Some portions of the lower support shroud tubes, the lower support tie rods as well as other lower internals components were found outside of the activation analysis model. These components are assumed to be relatively low in their total curie contents due to their distance from the active fuel region. Accordingly, a total weight of 706 pounds (320 kg.) was

. included for these components, activated to the specific curies per-gram level found in the

] 12th 3-inch (7.62 cm.) region below the reactor core.

5.11 Lower Core Barrel The lower core barrelis a cylindrical assembly attached to the upper end of the core bafDe.

The barrel separates the inlet and outlet plenums for the reactor. A portion of the lower core barrel was directly included in the activation analysis model, extending up to three feet above the active core. The weight of the lower core barrelin this region was calculated to be 1,067 pounds (484 kg.). An additional 1,823 pounds (827 kg.) of lower core barrel were beyond the limits of the activation analysis model and were considered activated to the specific curies per-gram level found in the 12th 3-inch (7.62 cm.) region above the reactor core.

5.12 Upper Core Plate The upper core plate provides a grid structure for insertion and location of fuel assemblies, and serves as a guide structure for the cruciform control rod blades. The plate is 1.0 inch (2.54 cm.) thick and 40.84 inches (103.73 cm.) diameter. The plate's total weight of 264 pounds (120 kg.) was directly included in the activation analysis.

5.13 Upper Core Barrel The upper core barrel is a cylindrical support structure for the upper core plate as well as the balance of the upper core internals assembly. A portion of the upper core barrel was directly included in the activation analysis model, extending up to three feet above the active core. The weight of the upper core barrel in this region was calculated to be 454 pounds (206 kg.).

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l SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARY INFORMATION Section 5, Page 5 of 6 FINAL REPORT 5.14 Balance of the Upper Core Barrel Assembly An additional 832 pounds (377 kg.) of upper core barrel and other upper internals were beyond the limits of the activation analysis model. They were considered activated to the specific curies-per-gram level found in the 12th 3-inch (7.62 cm.) region above the reactor core.

5.15 VesselInsulation Can 1

Surrounding the reactor pressure vessel insulation is a stainless steel can, 0.25-inch (0.64 l

cm.) thick at the core centerline. This can has an inner radius of 38.44 inches (97.64 cm.)

and 4,994 pounds (2,265 kg.) ofit were included in the analysis model- up to three feet above and three feet below the active core. (Note: The excessive weight of the modeled component is due to the inclusion of a heavy flange near the top of the model region.) The balance of the insulation can was assumed to contribute only insignificantly (less than 0.1%)

to the total curie content of the can.

5.16 Vessel Support Can Supporting the reactor vessel, and outside the vessel insulation can, is the vessel support can. This stainless steel can has an inner radius of 40.50 inches (102.87 cm.) and is 0.50 inch (1.27 cm.) thick at the core centerline. As with the vessel insulation can, only three feet above and three feet below the active fuel core have been included in the analysis models.

l The balance of the support can contributes only negligibly to the total curie content of the can (less than 0.1%), but is assumed to be surface contaminated. A can weight of 5,442 pounds (2,468 kg.) was included in the model.

5.17 Lead Shielding TLG has identified the presence of neutron-irradiated lead shielding within the activation analysis model. The shielding is located surrounding two of the four ex-core neutron detectors. TLG was unable to determine how much shielding surrounds each detector.

Therefore, only curie-per-gram results for this material were provided for regions up to 18 inches (45.72 cm.) beyond the outside wall of the vessel support can.

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SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 5, Page 6 of 6 FINAL REPORT 5.18 Structural Steel Stainless steel structural support members located outside the vessel support can and below I the operating water level are expected to be neutron activated. Curie per gram data for th:se components have been calculated out to a distance of 48 inches (121.92 cm.) beyond the reactor vessel support can.

5.19. Reactor Compartment Concrete The reactor vessel is located in the northwest quadrant of the reactor building at Saxton.

The eastern and southern walls are five feet thick. The western and northern sides of the reactor vessel are enclosed by the northwest quadrant of the cylindrical reactor building.

The reactor building has an outside wall of 1.5 foot (45.72 cm.) thick concrete. Curie per gram activation of concrete above and below the operating water level were considered separately.

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6. ACTIVATION ANALYSIS CALCULATIONS 1

i The activation analysis for the Saxton facility implements an innovative approach developed by TLG. A portion of this approach normally uses available and projected facility neutron fluence data to assist in the prediction of point neutron activation in the reactor vessel, its internals and the primary shield wall. Since this source of data was unavailable for the Saxton facility,.TLG resorted to conservative assumptions as to the radial peaking of the ,

l neutron flux in the reactor core. This analysis simply used a flat radial distribution for i

neutron levels throughout the core, which tends to lead to elevated levels of neutron flux by a factor of two or more on the core periphery (particularly for low-leakage core loading i i patterns). TLG did not perform any normalization of the one-dimensional neutron transport '

calculations to obtain a radial flux map beyond the reactor core boundary. However, TLG did compare the results of the activation analysis to characterization data taken by GPU Nuclear personnel during the second half of 1995. This data comparison is discussed in Section 7.

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1 j Gol One-Dimensional Neutron Transport Calculations The ANISN computer code was used to perform the one-dimensional neutron transport- l calculations required for this analysis. The personal computer version received from RSIC l was validated by TLG personnel on in house computers, and the calculations performed for  ;

this study were performed on similar equipment. The ANISN cylindrical geometry source  :

was used in the radial model calculations, and the slab geometry source was employed for )

! the axial model calculations. -

The SAILOR [16] coupled neutron gamma cross section library was the source of neutron cross sections used in this study. This library contains third order Legendre coefficients of the scattering cross sections (P3), in a 47-neutron and 20-gamma group structure. The data in this library were collapsed from a fine-group structure using weighting functions which represent the reactor vessel, its internals and the biological shield wall during reactor l operation. This library was originally constructed expressly for performing pressure vessel l fluence calculations and has been used extensively by utilities and contractors alike l throughout the commercial nuclear power industry. TLG believes this library to be the best j available for this type of activation analysis.

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[ TLG used two company-developed computer codes to assist in the performance of the one-dimensional neutron transport calculations. The first of these codes is FISSPEC, which l generates the. fission spectrum required by this analysis as input to ANISN. FISSPEC TLG SERVICES

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i allows selection of the Watt or Cranberg fission distribution formulations to calculate a i

normalized neutron energy spectrum from the fission of 22U, in either the CASK-81 or BUGLE-80 (SAILOR) energy spectrum formats. The Cranberg formulation in the BUGLE- I 80 energy group structure chosen for this analysis, as output by FISSPEC, is given in Table j 5.1. j i

l A second computer code, ANISNOUT, was employed to read the ANISN output and tabulate

the information in a format capable of being imported into an EXCEL spreadsheet. In this i tabulation process, ANISNOUT performs the following data manipulations:

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1. The 47-group neutron data are collapsed down to a 4-group format for each '

ANISN output zone. The 4-group format fluxes (one fast, two epithermal and j one thermal) were subsequently collapsed to a two-group format in order to l determine fast-to total and thermal-to-total neutron flux ratios.

2. The thermal neutron flux at the midpoint of each radial mesh interval is

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3. The radius, right boundary flux, zone volume, zone thermal neutron flux and zone total neutron flux are extracted and tabulated.

j Eleven one-dimensional models, nine radial and two axial, were constructed for the i calculations. They are discussed in detaillater in this report.

q 6.2 Point Neutron Activation Calculations j The ORIGEN2 computer code was used to calculate the activation and depletion of

radionuclides for components exposed to the reactor neutron flux.' One-group neutron cross I sections for forty eight reactor types are available as input to the code, two of which have been used specifically for this study. The first of these one-group cross section sets is representative of a neutron spectrum collapsed across the reactor core of a PWR (PWRUS)  !

j with a three-cycle reload pattern. The second is a thermal cross section set (THERMAL) containing only thermal neutron cross sections at 68'F. I l

For each location or component of interest and for each material of interest, the volume-averaged total neutron flux (neutrons /cm2.sec) as calculated by ANISN was input into two i ORIGEN2 calculations. One calculation used the PWRUS cross section set as input, the other used the THERMAL cross section set. THERMAL curie-per-gram results generated were adjusted for local moderator and/or environmental temperatures to reflect the decrease of thermal absorption cross sections at elevated temperatures. Based upon the ANISN TLG SERVICES

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spectral results of thermal to fast neutron flux ratios, the two ORIGEN2 calculations were weighted to obtain location or component-specific neutron activation. '

l Curie estimates have been decayed twenty-four years to 1 July 1996. This date represents ths earliest possible date after which work may begin in and around the reactor vessel in 1 preparation for its removal.

l 6,3 The Saxton Facility Operating History i

The operating history used in this analysis was obtained from data provided by GPU Nuclear, as previously discussed in Section 2. The total in-core neutron flux used in this

analysis was the average core flux as generated by ORIGEN2 calculations. An average plant power level (represented by the availability factor) was assumed for each operating cycle, based upoa a design power level of 23.5 MWth.

4 The Saxton core was designed to accommodate 32 fuel assemblies. However, there was no indication from the available literature that the core ever contained more than ' 21 l assemblies. These 21 assemblies were always loaded into the northern side of the core. ,

6.4 The Saxton Facility Radial ANISN Models The radial one-dimensional neutron transport modeling of the Saxton facility for this study included geometric and material considerations. This input data is described in further detail below. Many of the considerations and assumptions used in the radial models also apply to the axial ANISN models discussed later. Since no surveillance capsule / vessel fluence data was available against which normalization could be performed, the radial component model reactor core region assumed ORIGEN2-calculated neutron sources and a constant power distribution.

Nine radial models were constructed for this analysis. Four of these models explicitly

, portray the various components, radially outward from the reactor core, at the core centerline. Each of these models uses a homogeuized core representative of the fuel volume in a pair of the 1/8th sectors. Each model has a different homogenized reactor core radius.

A second set of four radial models depicts the various components, radially outward from the reactor core, with no water surrounding the vessel support can (representing the region above the operating water level). Again, each of these models represents a different i homogenized core radius. One additional radial model consists of the reactor core surrounded by six feet of water.

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SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARY INFORMATION Section 6, Page 4 of 24 FINAL REPORT Figures 6.1 and 6.2 depict the Saxton radial component ANISN model geometry at the core centerline. The core equivalent diameters fbr the four problems may be found in the table beneath Figure 6.2. The other four radial component models are similar to those found in Figure 6.2 with the exception that there is no water surrounding the reactor vessel support can.

1 6.4.1 Radial Model Geometries The radial component model generally consists of the reactor core, a dummy assembly / water region, a water region, the core bafDe, the thermal shield, the reactor vessel cladding and I wall, the vessel insulation can, the vessel support can, the reactor compartment concrete shield wall and an outer vacuum boundary in cylindrical geometry. The vessel insulation and reactor compartment concrete shield wall rebar have not been modeled explicitly.

These components do not significantly impact the total neutron flux and energy spectrum.

Radial distances from the core centerline at the core mid plane elevation for these components were used. Small components which may be found as part of the reactor  ;

internals (such as surveillance capsules, various bolts and fasteners, etc.) were not included in the models. The models do however include all major air and water gaps, as appropriate.

The ninth radial model consisted of a cylinder of core material, similar to the component radial model with the largest core diameter, bounded on the outside- by six feet. of downcomer core coolant and a vacuum boundary. This model was used to' generate ex-core radial neutron flux reduction factors for those components modeled in the axial component models (see Section 6.5.1 below).

Since ANISN allows only one source / shield geometry (either slab, spherical or cylindrical) in a given problem, the reactor core was modeled as a homogenized cylinder. The Saxton facility reactor core, which consisted of twenty-one fuel assemblies, loaded in an asymmetric fashion, was homogenized to solid right cylinders 36.61 inches (92.99 cm.) tall, using the reactor fuel assembly pitch of 5.418 inches (13.76 cm.). The resulting cylinders for the four core diameters had radii of 43.921,38.037,31.057, and 26.896 centimeters. The asymmetric core loading placed 21 assemblies into the 32 space core configuration, with assemblies loaded towards the north side of the reactor (see Figure 6.1).

The core bafDe was homogenized to a cylinder with an inside diameter of 34.66 inches (88.04 cm.), and a thickness of 0.375 inches (0.95 cm.). The thermal shield geometry used in the ANISN model has an inner diameter of 49 inches (124.46 cm.), and from data supplied by GPU Nuclear, a thickness of 3.0 inches (7.62 cm.). The stamless steel cladding on the reactor vessel wall has a nominal thickness of 0.125 inches (0.32 cm.) and an inner diameter of 57.75 inches (146.69 cm.). The carbon steel laminated plate vessel has a nominal thickness TLG SEFIVICES

i SAXTON ACTIVATION ANALYSIS Document No. G01-1192-008, Rev. O PROPRIETARYINFORMATION Section 6, Page 5 of 24 FINAL REPORT of 5 inches (12.70 cm.) and an inner diameter of 58 inches (147.32 cm.). The reactor vessel insulation has not been included in this model-instead its region was replaced by air.

Surrounding the vesselinsulation is a stainless steel vesselinsulation can. This can has an inner diameter of 76.875 inches (195.26 cm.) and is 0.25 inches (0.64 cm.) thick. The reactor vessel and vessel insulation can is held in place by the vessel support can. The stainless steel vessel support can has an inner diameter of 81 inches (205.74 cm.) and is 0.50 inches

-(1.27 cm.) thick. The support can was surrounded by water during reactor operation to a level just above the vessel inlet and outlet nozzles (Elevation 791'-8"). The radial models were surrounded by 18 inches (45.72 cm.) of concrete (representing the thinnest concrete in the reactor compartment, the containment vessel outer wall) and a vacuum outer boundary.

6.4.2 Radial Model Cornpositions TLG assumed the reactor core region of the Saxton radial component and radial water ANISN models to be comprised of one region. Based upon information provided by GPU Nuclear, TLG identined that only UO2 was placed in locations on the core periphery during all three operating cycles. The first core contained all fresh UO2 assemblies. The second core contained all fresh PuO2-UO2 in the nine interior locations, eleven fresh UO2 assemblies on the core periphery, with one once-exposed UO2 assembly on the core periphery. The final core used once-exposed PuO2-UO2 fuel in the core interior, eleven once-exposed fuel from the first core in core periphery locations, and one fresh fuel assembly on the periphery. Since the neutron spectrum beyond the radial core boundary is generally dependent upon the core l peripheral fuel assemblies, assuming the whole core contained UO2 assemblies was reasonable for the radial models.

The core composition was determined in part by running ORIGEN2 with an average fuel enrichment of 5.7 w/o 235U, and an average cycle burnup of 9.730 GWd/MTU for the first core. The core thermal output, normalized to 23.5 MWt was used. The second cycle also contained an enrichment of 5.7 w/0 23sU fresh fuel, and was exposed at 23.5 Mwt for 9.334 GWd/MTU. The final core was a continued burnup of the first core's fuel for an additional 5.966 GWd/MTU. Refueling outages and decay times after final shutdown may be found in Table 3.1 Resulting 235U, 23sU, 238U, 237Np, 239pu, 240pu, 241pu and 242pu Concentrations for each core at the midpoint in each cycle were calculated using ORIGEN2. These data were extracted and used as ANISN input. Inlet, outlet and average core moderator temperatures taken from GPU Nuclear data were used to calculate moderator densities. Table 6.2 contains the core region composition summary for the Saxton facility radial component and water ANISN models.

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l Compositions of the primary coolant, stainless steel, carbon steel, lead, and concrete components were derived or calculated from various sources. The compositions used as input for the radial component and water ANISN models in the current analysis have been summarized in Table 6.2. Compositions for stainless steel, water, carbon steel and concrete outside the reactor pressure, used in the neutron transport calculations only, were those as defined in the SAILOR cross section library.

l 6.4.3 Radial Model Calculations l

Nine radial ANISN calculations were performed in the course of this analysis. Four models were generated for the region below the operating water level, and four for above the operating level. These models were representative of the eight core sector divisions shown in Figure 6.1. The ninth model utilized a homogenized core with a diameter equal to that l fcund in Sectors A&B of Figure 6.1. This model war used to determine flux reduction factors beyond the radial core boundary for thoea components analyzed in the axial model I below. Each of these models assumed an axially-averaged, flat core power distribution, with the total neutron source equal to that calculated by ORIGEN2 for a four-region core - an average core total neutron flux of 1.481E+14 n/cm2-sec. l Outputs from the radial model ANISN runs were transferred to nine EXCEL spreadsheets, one for each model. No normalization was performed on any of these calculations. TLG expects the results to be conservative by a factor of two to four in the radial direction due to the use of a flat core power distribution. Current light water reactors generally display ,

lower power levels on the core periphery - normally, a factor of two to four lower than the )

core average power level. This occurs for several reasons, including a radial cosine l distribution of the neutron flux and low-leakage core loading patterns. The use of a flat l power distribution across the reactor core at Saxton, results in conservatively' high power  !

levels for peripheral fuel assemblies; and peripheral.to-average core neutron flux levels of close to unity.  !

The results of these calculations for the thermal neutron energy group (<0.411 eV, the energy group which most significantly influences activation for most radionuclides) are plotted below. Figures 6.3 and 6.4 show the radial thermal neutron flux at the core midplane averaged over the core height for model sectors A&B; and with a void space outside the reactor vessel support can, instead of water (to mimic the flux profile above the operating water level), respectively. Figure 6.5 depicts the homogenized reactor core surrounded by 6 feet of water (again, the results of which are used for flux reduction factors 4 l for certain components only).

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SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 6, Page 7 of 24 FINAL REPORT 6.5 The Saxton Facility Axial ANISN Models Two axial ANISN models in slab geor2etry were established for performing the axial neutron flux calculations. The Saxton facility c.xial component ANISN uses homogenized material regions above and below the active core, in three-inch (7.62 cm.) increments. Since no data was available (similar to the surveillance capsule / vessel fluence data) against which normalization could be performed, the axial component model reactor core region assumed ORIGEN2.-calculated neutron sources and a constant power distribution.

The second model- the axial water model- consists of the reactor core surrounded on either side by eight feet of water. Similar to the axial component model, ORIGEN2-calculated neutron sources and a constant power distribution were assumed for the reactor core. The calculational differences associated with an all UO2-assembly core versus the actual mixed oxide / uranium oxide core, was considered to have negligible impact upon this type of study.

l 6.5.1 Axial Model Geometries The two axial models constructed for this analysis included one in which the volume fractions of moderator and components were homogenized over three-inch (7.62 cm.)

)

increments axially away from the reactor core, but within the core diameter; and a second model consisting of the reactor core surrounded on top and bottom by water. The components and moderator homogenized in the axial component model includsd the i following: 1

1. Cold leg water for regions below the reactor core.

2. Lower core guide blocks.

3. Lower core plate.

4. Lower core support tubes.

5. Lower core tie rods.

6. Upper core plate.

Two reasons for the conservative nature of the axial component ANISN calculation are given below:

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1. Homogenized regions containing less than 5410% metal by volume were assumed to be pure water since they would not significantly impact flux levels.

j In these regions, the thermal-to-total neutron flux ratio is also conservative

! due to the presence of water only.

I i

2. No definitive inhrmation was available as to the volumes or weights of the fuel

assembly end fitties. Their absence from the model potentially causes the I activation in components above and below the reactor core to be overestimated by 10-20%.

L

{. Similar to the radial modeling, an axial water model consisting of the reactor core bounded l on one side by eight feet of cold leg water, and the other by ei;'ht feet of hot leg water, was constructed. This model was used to generate ex-core axial neutron flux reduction factors for those components modeled in the radial component models.

6.5.2 Axial Model Compositions The reactor core region of the Saxton facility axial ANISN models consisted of a homogenized average of the eight core regions defined for the radial models. The axial component model contained homogenized regions of water and steel above and below the reactor core. The volume of metal in these regions was established from reactor and internals drawings provided to TLG by GPU Nuclear and other available data. Material j composition for the reactor core and eight feet above and below the reactor core for the i component axial models may be found in Table 6.2.

6.5.3 Axial Model Calculations -

Two axial ANISN calculations, corresponding to the two model geometries, were performed for this analysis. Each of these models assumed an axially-averaged, flat core power distribution, with the total neutron source equal to that calculated by ORIGEN2 with an average core total neutron flux of 1.481E+14 n/cm2-sec.

Output from the axial model ANISN runs was transferred to two EXCEL spreadsheets, one for each model. No normalization was performed for either axial model. The results of these calculations for the thermal neutron energy group (the energy group which most significantly influences activation for most radionuclides) are plotted in Figures 6.6 and 6.7.

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The thermal and total neutron fluxes for various components and locations ofinterest in this i analysis were extracted from the ANISN output using ANISNOUT. Total neutron fluxes, as i obtained from the ANISN calculations as discussed above, were input into a pair of l ORIGEN2 problems for each component, location and/or material of interest. The total neutron fluxes input were corrected for the plant availability factors found in Table 2.1.

The fraction of neutron flux which fellinto the thermal energy range was calculated for each component from the ANISN output data, along with the fraction of neutron flux greater than thermal energies. These spectrum fractions were then compared to the epithermal and thermal spectrum fractions found in the reactor core region of the radial and axial ANISN models. ORIGEN2 calculations for the core spectrum activation (PWRUS cross section library) and thermal spectrum activation (THERMAL library) for each component and material of interest were then performed. A composite neutron activation based upon neutron spectrum contributions and results from the two ORIGEN2 runs was subsequently calculated. Resultant specific curie-per-gram radioactivity for various components, materials and locations of interest in this activation analysis have been reproduced in Section 7.

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TABLE 6.1 I CRANBERG FISSION SPECTRUM Group Number Uoner Enerav (Me\A Fission Soectrum 1 1.7333E+01 3.0727E-05 2 1.4191E+01 1.3937E-04 3 1.2214Ev01 8.8602E-04 4 1.0000E+01 2.1758E-03 5 8.6071E+00 5.0640E-03 I

6 7.4082E+00 1.4930E-02 7 6.0653E+00 2.9341E 02 8 4.9659E+00 7.8886E-02 9 3.6788E+00 7.5235E-02 10 3.0119E+00 4.2879E-02 11 2.7253E+00 4.5335E-02 12 2.4660E+00 1.9428E-02 13 2.3653E+00 3.9054E-03 14 2.3457E+00 2.3620E-02 l l

15 2.2313E+00 7.1581E-02 16 1.9205E+00 7.0649E 02 17 1.6530E+00 8.9113E-02 18 1.3534E+00 1.1641E-01 19 1.0026E+00 6.4048E-02 20 8.2085E-01 2.7920E-02 21 7.4274E-01 4.8115E-02 22 6.0810E-01 3.8772E-02 23 4.9787E-01 4.3549E-02 24 3.6883E-01 2.26970E-02 25 2.9720E-01 3.2565E-02 26 1.8316E-01 1.7139E-02 27 1.1109E-01 8.4151E-03 28 6.7379E-02 4.0683E-03 29 4.0868E-02 1.1527E-03 30 3.1828E-02 6.5991E-04 31 2.6058E-02 2.0109E-04 32 2.4176E-02 2.3572E-04 33 2.1875E-02 6.2846E-04 34 1.5034E-02 5.6466E-04 35 7.1017E-03 1.8407E-04 36-47 0.000E+00 l

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SAXTON ACTIVATION ANALYSIS D:cument No. G01-1192-003, Rev. 0 l PROPRIETARY INFORMATION Section 6, Page 11 of 24 FINAL REPORT TABLE 6.2 SAXTON REGIONAL COMPOSITION SUMMARIES FOR RADIAL AND AXIAL COMPONENT AND WATER MODEL GEOMETRIES Core I(Wb-cm) An ANISN Models U-235 3.171E-04 U-236 6.741E-06 U-238 5.749E-03 Np-237 1.140E-07 Pu-239 9.260E-06 Pu-240 5.553E-07 i Pu-241 8.869E-08 Pu 242 1.776E-09 Core II(a/b-cm) All ANISN Models U-235 2.795E-04 U-236 6.30BE-06 U-238 5.100E-03 Np-237 1.104E-07 Pu-239 8.626E-06 Pu-240 5.440E-07

( Pu-241 9.191E-08 l

! Pu-242 1.053E-09 l Core Ill(s%<m) AII ANISN Models l U-235 2.332E-04 l U-236 1.480E-05 i

U-238 5.081 E-03 Np-237 5.031 E-07 Pu-239 1.735E-05 Pu-240 2.388E-06 Pu-241 9.191E-07 Pu-242 5.576E-08 l Homogenized Coms I Through III(MWT Hr Weighted) (a/b cm) All ANISN Models i

U-235 2.818E-04 U-236 8.566E-06 U-238 5.333E-03 Np-237 2.089E-07 Pu-239 1.102E-05 Pu-240 1.004E-06 Pu-241 2.953E-07 Pu-242 1.520E-08 Active Core Other Materials (sh-cm) Radial Water and Component Models. Axial Water Model l H 3.361E-02

B-10 4.612E-06 l O Cn moderator) 1.680E-02 l O (in fuel) 1.222E-02 Cr 9.250E-04 Fe 3.078E-03 Ni 4.312E-04 i Zr 0.000E+00 Downcomer, Bypass and Core inlet Water Regions (s%-cm) Radial Component and Axial Water Models H 5.131 E-02 B-10 7.040E-06 0 2.565E-02 i

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TABLE 6.2 (cont.)

SAXTON REGIONAL COMPOSITION SUMMARIES FOR RADIAL AND AXIAL COMPONENT AND WATER MODEL GEOMETRIES Stainless Steel Regions (a&cm) Radial Component Model Cr 1.761 E-02 Mn 1.754E-03 Fe 5.858E-02 Ni 8.209E-03 Carbon Steel Regions (a/b-cm) Radial Component Model

Mn 8.606E-04 Fe 8.331E-02 Core Extt Water Region (a%cm) Axial Water Model H 4.975E-02 B-10 6.826E-06 0 2.487E-02 Dummy Assembly Region- Sectors A&B Moderator Volume Fraction 1.0000 Stainless Steel Fraction 0.0000 H 5.029E-02 B-10 6.900E-06 O 2.514E-02 Cr 0.000E+00 Mn 0.000E+00 Fe 0.000E+00 Ni 0.000E+00 Dummy Assembly Region- Sectors C&H Moderator Volume Fraction 0.9888 Stainless Steel Fraction 0.0112 H 4.972E-02 B-10 6.822E-06 0 2.486E-02 Cr 1.970E-04 Mn 1.963E-05 Fe 6.555E-04 Ni 9.185E-05 Dummy Assembly Region- Sectors D&G Moderator Volume Fraction 0.9888 Stainless Steel Fraction 0.0112 H 4.972E-02 B-10 6.822E-06 O 2.486E-02 Cr 1.970E-04 Mn 1.963E-05 Fe 6.555E-04 Ni 9.185E-05 Dummy Assembly Regron- Sectors E&F Moderator Volume Fraction 0.9888 Stainless Steel Fraction 0.0112 H 4.972E-02 B-10 6.822E-06 O 2.486E-02 Cr 1.970E-04 Mn 1.963E-05 Fe 6.555E-04 Ni 9.185E-05 1

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TABLE 6.2 (cont.)

1 SAXTON REGIONAL COMPOSITION SUMMARIES FOR RADIAL AND AXIAL COMPONENT AND WATER MODEL GEOMETRIES i l 32nd mrough 13th 3* Bebw Active Core Regson (a/b-cm) Axial Component Model Moderator Volume Fraction 1.0000 Zircaloy-2 Volume Fraction l Stainless Steel Volume Fraction j

Void Volume Fraction '

H 5.131E-02 l

B-10 7.040E-06 0 2.565E-02 Cr 0.000E+00 Mn 0.000E+00 Fe 0.000E+00 Ni 0.000E+00 Zr 0.000E+00 l 12m 3* Below Active Core Region (a/b-cm) Axiel Component Model l ModeratorVolume Fraction 0.9465 l Zircaloy-2 Volume Fraction Stainless Steel Volume Fraction 0.0535 Void Volume Fraction H 4.856E-02 B-10 6.664E-06 l O 2.428E-02 Cr 9.415E-04 Mn 9.380E-05 I Fe 3.133E-03 l Ni 4.390E-04 Zr 0 000E+00 11m 3* Below Active Core Region (a/b<:m) Axsal Component Model Moderator Volume Fraction 0.9465 Zircaloy-2 Volume Fraction Stainless SteelVolume Fraction 0.0535 Void Volume Fraction H 4.856E-02 B-10 6.664E-06 0 2.428E-02 Cr 9.415E-04 Mn 9.380E-05 Fe 3.133E-03 Ni 4.390E-04 Zr 0.000E+00

~10th 3* Below Active Corts Region (abcm) Axsal Component Model Moderator Volume Fraction 0.9465 Zircaloy-2 Vok:me Fraction Stainless Steel Volume Fraction 0.0535 Void Volume Fraction H 4.856E-02 B-10 6.664E-06 O 2.428E-02 Cr 9.415E-04 Mn 9.380E-05 Fe 3.133E-03 Ni 4.390E-04 l Zr 0.000E+00 l

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SAXTON REGIONAL COMPOSITION SUMMARIES FOR RADIAL AND AXIAL COMPONENT AND WATER MODEL GEOMETRIES 9th 3* Below Active Core Regron (alb-cm) Axial Component Model Moderator Volume Fraction 0.9465 Zircaloy-2 Volume Fraction j Stainless Steel Volume Fraction 0.0535 Void Volume Fraction H 4.856E-02 B-10 6.664E-06 O 2.428E-02 Cr 9.415E-04 Mn 9.380E-05 Fe 3.133E-03 Ni 4.390E-04 Zr 0.000E+00 6th 3* Below Active Core Region (a/b-cm) Axial Component Model '

l Moderator Volume Fraction 0.9465 <

Zircaloy-2 Volume Fraction

)

Stainless SteelVolume Fraction 0.0535 . l Void Volume Fraction i H 4.856E-02 l B-10 6.664E-06 O 2.428E-02 Cr 9.415E-04 Mn 9.380E-05 Fe 3.133E-03 l Ni 4.390E-04 l Zr 0.000E+00 7th 3" Below Active Core Region (a/ts:m) Axial Component Model l

ModeratorVolume Fraction 0.9512 l Zircaloy-2 Volume Fraction Stainless Steel Volume Fraction 0.0488 Void Volume Fraction H 4.880E-02 B-10 6.696E-06 0 2.440E-02 Cr 8.600E-04 Mn 8.567E-05 Fe 2.861E-03 Ni 4.009E-04 Zr 0.000E+00 6th 3' Below Active Core Region (ob-cm) Axsal Component Model Moderator Volume Fraction 0.9743 Zirca!oy-2 Volume Fraction Stainless Steel Volume Fraction 0.0257 Void Volume Fraction H 4.999E-02 '

B-10 6.859E-06 0 2.499E-02 Cr 4.520E-04 Mn 4.503E-05 Fe 1.504E-03 Ni 2.107E-04 Zr 0.000E+00 TLG SERVICES

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TABLE 6.2 (cont.)

SAXTON REGIONAL COMPOSITION SUMMARIES FOR RADIAL AND AXIAL COMPONENT AND WATER MODEL GEOMETRIES 5th 3' Below Active Core Regton (a/tH:m) Axial Component Model 1 Moderator Volume Fraction 0.9743 l Zircoloy-2 Volume Fraction 1 Stainless Steel Volume Fraction 0.0257 Void Volume Fraction H 4.999E-02 j B-10 6.859E-06 )

0 2.499E-02 Cr 4.520E-04 Mn 4.503E-05 Fe 1.504E-03 Ni 2.107E-04 Zr 0.000E+00 l 4th 3"Below Active Core Region (a4H:m) Axsal Component Model Moderator Volume Fraction 0.9743 Zirealoy-2 Volume Fraction Stainless Steel Volume Fraction 0.0257 Void Volume Fraction H 4.999E-02 B-10 6.859E-06 0 2.499E-02 Cr 4.520E-04 Mn 4.503E-05 Fe 1.504E43 Ni 2.107E-04 Zr 0.000E+00 3rd 3" Below Active Core Region (ats:m) Axial Component Model Moderator Volume Fraction 0.8712 Zircaloy-2 Volume Fraction Stainless Steel Volume Fraction 0.1288 Void Volume Fraction ,

H 4.470E-02 B-10 6.133E-06 O 2.235E-02 Cr 2.268E-03 Mn 2.260E-04 Fe 7.547E-03 Ni 1.058E-03 Zr 0.000E+00 2nd 3* Be%w Active Conf Region (a4H:m) Axsal Component Model Modera*st Volume Fraction 0.6230 Zircaloy 2 Volume Fraction Stainless Steel Volume Fraction 0.3770 Void Volume Fraction H 3.197E-02 B-10 4.386E-06 O 1.598E-02 Cr 6.638E-03 Mn 6.613E-04 Fe 2.209E-02 Ni 3.095E-03 Zr 0.000E+00 TLG SERVICES I

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SAXTON REGIONAL COMPOSITION SUMMARIES FOR RADIAL AND AXIAL COMPONENT AND WATER MODEL GEOMETRIES 1st 3' Below Active Core Region (artH;m) Axtel Component Mbdel Moderator Volume Fraction 0.9757 Zircaloy-2 Volume Fraction Stainless Steel Volume Fraction 0.0243 Void Volume Fraction H 5.006E-02 B-10 6.869E-06 O 2.5ar,E-02 Cr 4.270E-04 Mn 4.254E-05 Fe 1.421E-03 Ni 1.991E-04 Zr 0.000E+00 Active Core Region (abcm)- Axial Component Model(Using Core I atom densities)

U-235 3.171E-04 )

U-236 6.741E-06 l U-238 5.749E-03 Np-237 1.140E-07 j Pu-239 9.260E-06 Pu-240 5.553E-07 Pu-241 8.869E-08 Pu-242 1.776E-09 H 3.361E-02 B-10 4.612E-06 O (in moderator) 1.680E-02 )

0 (in fuel) 1.222E-02 1 Cr 9.250E-04  ;

Fe 3.078E-03 )

Ni 4.312E-04 Zr 0.000E+00 l 1st 3' Above Active Core Region (abcm) Axial Component Model Moderator Volume Fraction 1.0000 j Zircaloy-2 Volume Fraction Stainless SteelVolume Fraction Void Volume Fraction H 4.975E-02 B-10 6.826E-06 O 2.487E-02 Cr 0.000E+00 Mn 0.000E+00 Fe 0.000E+00 Ni 0.000E+00 Zr 0.000E+00 l

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TABLE 6.2 (cont.)

SAXTON REGIONAL COMPOSITION SUMMARIES FOR RADIAL AND AXIAL COMPONENT AND WATER MODEL GEOMETRIES 2nd 3' Above Active Coro Region (stH;m) Axial Component Model ModeratorVolume Fraction 0.8070 Zircaloy-2 Volume Fraction Stainless SteelVolume Fraction 0.1930 Void Volume Fraction H 4.015E-02 l B-10 5.509E46 1 0 2.007E-02 l Cr 3.398E-03 Mn 3.385E-04 Fe 1.131E42 Ni 1.584E43 Zr 0.000E+00 3rd ttuough 32nd 3' Above Active Com Regen (a/b<m) Axial Component Model 1 Moderator Volume Fraction 1.0000 l Zircaloy-2 Volume Fraction Stainless SteelVolume Fraction Void Volume Fraction H 4.975E-02 B-10 6.826E-06 0 2.487E-02 Cr 0.000E+00 Mn 0.000E+00 Fe 0.000E+00 Ni 0.000E+00 Zr 0.000E+00 TLG SERVICES

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FIGURE 6.1 i THE SAXTON FACILITY i RADIAL COMPONENT ANISN CORE LAYOUT AND MODEL SEGMENTATION l

i

) H I A s -.. e

. r, 5 North I l

w sd m..h:

v

! ,ev j G jfi{ 0 B x '

j tiblA w

j ng; Eh.

1

@e . c.j g,g pa i

1 F

- c m.: ,e .

q wn EE.5 f

t e s.

l E D l

i I

! Fuel Assembly Location 1

! r Dummy Assembly Location l Core sectors are paired by the quantity of fuel in each sector, and a neutron transport j calculation was performed for each pair. The paired sectors are as follows: A&B, C&H,

D&G and E&F.

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SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARY INFORMATION Section 6, Page 19 of 24 FINAL REPORT FIGURE 6.2 THE SAXTON FACILITY RADIAL COMPONENT. ANISN MODEL GEOMETRY OVERVIEW Reactor Water Shield Compartment Concrete Vessel Support Can  % >

cysfM, Vesselinsulation Can s , . . <

}~

Vessel Clad VesselWall O ,1 l [dlh

p, Thermal Shieu Core same

' , . ..  ; 8b Dummy Fuel Assembly / 3- Reactor Core Water Region component QuietRadiusafa n Core Equivalent Radius, Sectors A&B 43.921 Core Equivalent Radius, Sectors C&H ' 38.037 Core Equivalent Radius, Sectors D&G 31.057 Core Equivalent Radius, Sectors E&F 26.896 Dummy Assembly / Water Gap 43.921 Water Gap 44.012 Core Baffle 44.965 Water Gap 62.230 Thermal Shield 69.850 Water Gap 73.343 Vessel Clad 73.660 Vessel Wall 86.360 Vesselinsulation 97.631 VesselInsulation Can 98.266 Air Gap 102.870 Vessel Support Can 104.140 Water Shield 223.520 Reactor Compartment Concrete Wall 269.240 TLG SERVICES

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l FIGURE 6.3 THE SAXTON FACILITY RADIAL COMPONENT ANISN MODEL THERMAL NEUTRON FLUX FOR MODEL SECTORS A&B l

Saxton Neutron Activation Analysis ANISN Radial Component Model- Sectors A&B Thermal Neutron Flux 1.00E +14 1.00E +13 100E +12 1.00E+11 -

1.00E+10 9

1.00E+09 -

1.00E+06 1.00E+07 -

2 100E@ .

1_

100E +04

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TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003. Rev. O PROPRIETARYINFORMATION Section 6, Page 21 of 24 FINAL REPORT FIGURE 6.4 THE SAXTON FACILITY RADIAL CONCRETE ANISN MODEL THERMAL NEUTRON FLUX FOR MODEL SECTORS A&B Saxton Neutron Activation Analysis ANISN Concrete Model-Sectors A&B Thermal Neutron Flux tu.u i

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l FIGURE 6.5 l THE SAXTON FACILITY RADIAL WATER ANISN MODEL THERMAL NEUTRON FLUX  ;

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SAXTON ACTIVATION ANALYSIS Docurnent No. G01-1192 003, Rev. O PROPRIETARYINFORMATION Section 6, Page 24 of 24 FINAL REPORT FIGURE 6.7 THE SAXTON FACILITY AXIAL WATER ANISN MODEL THERMAL NEUTRON FLUX Saxton Neutron Activation Analysis ANISN Axial Water Model . Thermal Neutron Flux 100E+14-100E+13 -

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SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 7, Page 1 of 27 FINAL REPORT l

7. ACTIVATION ANALYSIS RESULTS AND COMPARISON l TO SAXTON CHARACTERIZATION DATA  :

l l

ANISN and'ORIGEN2 calculations were performed following the methodology as outlined in Section 6, and using the input information from Sections 2 through 6. Curie-per-gram values for the various materials of interest at twenty-four years aiter plant shutdown (1 July 1996) were input into an EXCEL spreadsheet. This spreadsheet calculated the radionuclide contents for each component ofinterest and specific curie contents of materials in the region of the reactor vessel. This section discusses the results of those calculations as well as presenting a comparison to characterization data acquired during the site

, characterization campaign during the second half of 1995.

[ 7.1 Curie Contents at Twenty-Four Years After Reactor Shutdown  !

ORIGEN2 outputs provided the curie-per-gram content for the reactor vessel, internals, i

surrounding lead, structural steel ara concrete in the various model regions. This data was subsequently transferred into an EXCEL spreadsheet to allow summation of the curie contents for each component and material ofinterest, over all regions of the ANISN problem in which the component or material was modeled.

Tables 7.1 through 7.19 present relevant portions of this spreadsheet. Since the volumes in each model region for each component are presented in these tables, the Ci/cm3 (or Ci/gm) specific concentrations for'each radionuclide in each component region may be readily 1 calculated. Tables 7.17 and 7.18 summarize the component and material curie content data. l l

7.2 10 CFR Part 61 Classification i Table 7.19 provides results of the 10 CFR Part 61 low level radioactive waste (LLRW) i classification of the reactor pressure vessel and its internals twenty-four years after reactor shutdown. Note that the 10 CFR Part 61 classificat.ms are based upon the sum of the regions considered in the activation analysis models for each component of interest. The drivers for classification generally are 63Ni and 94Nb. Lead, structural steel and concrete beyond the pressure vessel outer wall which became neutron-activated over the operating life of the plant, will be Class A only LLRW. 1 TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 7, Page 2 of 27 FINAL REPORT 7.3 Discussion of Results As may be seen in Tables 7.1 through 7.16, the region of highest activation occurs over the reactor active core height, with induced activity levels falling rapidly axially beyond the active core. There are some anomalies to this general statement, which arc of significant interest to GPU Nuclear. They occur due to the asymmetry of the reactor fuel loading pattern, and the use of a borated water shield surrounding the reactor during operations.

The results of the activation analysis performed herein are discussed in two separate categories.

1. The radioactivity and waste classification of the reactor vessel, internals, vessel insulation can and vessel support can twenty-four years after shutdown.

2. The specific activity of components and materials outside the reactor vessel support can twenty-four years after shutdown.

7.3.1 Induced Radioactivity Twenty-Four Years After Shutdown of the Saxton Reactor Pressure Vessel, Internals, Vessel Insulation Can and Vessel Support Can Table 7.17 summarizes the curie contents of the reactor pressure vessel, internals, vessel insulation can and vessel support can twenty-four years after reactor shutdown. Table 7.19 provides an estimate of the waste classification of these components. A total of 1,452 curies of radioactivity is estimated to remain on 1 July 1996, with approximately 37 Ci, 594 Ci, j and 811 Ci of 55Fe, *Co, and esNi, respectively. Specific radionuclide activities are expected to vary by approximately a factor of ten azimuthally around the reactor vessel at the core centerline. The highest specific activities are expected to be in the northern quadrant, a factor of two higher than the average. The lowest specific activities are expected to be in the southern quadrant, a factor of five lower than the average. Less than 0.3% (approximately 5 Ci) of the induced radioactivity resides in the vesselinsulation can and vessel support can.

The only core sub-component requiring categorization as greater-than-Class C (GTCC) radioactive waste in accordance with 10 CFR Part 61 are the lower core guide blocks. These components are permanently bolted to the lower core plate. When blended with the lower core plate, the lower core plate assembly will meet Class C limits as shown in Table 7.19.

Individuallower core guide blocks have less than 1 mci 84Nb and may therefore use a factor of ten when considering blending scenarios under the NRC's 10 CFR Part 61 Branch Technical Position on Waste Characterization [17]. Therefore, under current regulations, the lower core plate assembly (the lower core plate and the lower core guide blocks) may be blended for shipment, or assumed to be one component, and shipped to a shallow land disposal facility.

TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No G01-1192-003, Rev. O  ;

PROPRIETARYINFORMATION Section 7, Page 3 of 27 i FINAL REPORT The long decay time after shutdown has allowed significant decay of the strong gamma-emitting isotope 60Co. Similarly, a generally large contributor to the total curie content,

1. Fe, has also had the opportunity to decay to low levels. If suitably packaged, the reactor vessel and internals can qualify for shipment under the current 10 CFR Part 71 regulations for greater than Type A quantity, low specific activity (LSA).

7c3.2 Induced Radioactivity Twenty-Four Years After Shutdown of Components and Materials Outside the Reactor Vessel Support Can Lead, structural stainless steel and concrete surrounding the reactor vessel support can and within the reactor vessel compartment will exhibit induced radioactivity due to neutron exposure. A discussion of each follows below.

i Lead Lead blocks were used to reduce the' level of gamma radiation at two of the four ex core i neutron detectors. The radioactivity in the lead is due to the presence of trace amounts of silver, which under neutron irradiation, forms the long-lived radioactive species los=Ag (and '

its short-lived daughter 108Ag). The lead is assumed to surround the ex core neutron ,

detectors at two of the four detector locations azimuthally around the core centerline.

The azimuthal as well as the radial position of the lead will determine its specific curie level.

However, the azimuthal average specific curie contents oflead in the first, second and third six-inch (15.24 cm.) intervals outside the vessel support can may be found in Table 7.18.

Variations in these specific curie contents are expected to range between +100% to -80% due '

to the asymmetric core loading pattern.

Structural Stainless Steel Structural stainless steel below the operating water level (Elevation 791'-8") is expected to show induced radioactivity well beyond the vessel support can. As shown in Table 7.18, structural stainless steel at the core centerline is expected to contain measurable (above 1 pCi/gm) levels of 60Co more than three feet beyond the vessel support can. Similar to the lead above, variations in these average.s are expected to range between +100% to -80% due to the asymmetric core loading pattern.

The axial position of the steel will also highly influence the induced radioactivity levels.

Beyond three feet above and three feet below the active core, radioactivity in structural stainless steelis expected to be at or below measurable levels.

TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 7, Page 4 of 27 FINAL REPORT Structural stainless steel above the operating water level (Elevation 791'-8") will demonstrate some induced radioactivity. Again, the distribution of radioactivity will be azimuthally sensitive, being at or near its maximum in the northern quadrant of the reactor compartment and its minimum near the southern quadrant. According to calculations, "Co concentrations in the structural stainless steel above the operating water level will average around 126 pCi/gm with a range between roughly 25 and 250 pCi/gm. There is little expected change in the specific curie contents as a function of elevation above the operating water level.

Concrete Concrete below the operating water levelis expected'to show unmeasurable levels of #Co and is2Eu, the radionuclides of concern in concrete exposed to neutron irradiation. In this area, surface contamination of the concrete, migration of water into deep fissures and spalling of the concrete surface, will be of greater concern during decommissioning.

Concrete activation above the operating water level produced significantly different results.

Above this level, measurable amounts of"Co and is2Eu in the range of 0.5 to 3 pCi/gm will be found, at depths up to six inches (15.24 cm.)into the concrete surfaces, as determined by the activation calculations. (Note: The measured levels of induced radioactivity in the concrete above the operating water level are roughly a factor of 10 to 18 higher than the calculated values. Please see the discussion below regarding the comparison between measured and calculated values.) These levels are expected to drop by a factor of 10 at 15 inches (38.10 cm.) into the concrete.

7.4 Comparison of Activation Analysis Results to Saxton Characterization Data GPU Nuclear performed a campaign to obtain characterization data during the second half of 1995. This work was in support of the Decommissioning Plan currently being prepared for the NRC's review. TLG recognized that during this task, GPU Nuclear would be able to obtain data which would validate and potentially benchmark the results of this activation analysis. Several different types of samples were requested by TLG, and GPU Nuclear contributed some additional samples which proved to be beneficialin this comparison.

As previously mentioned in the opening of Section 6, TLG expects the results of the activation analysis to be conservative by roughly a factor of two for components radially outward from the reactor core, and reasonably correct for components above and below the active core. Since a number of the radioactive species in this activation analysis have trace element precursors (particularly aCo,94Nb and is2Eu), variations by a factor of ten in tha final results are not uncommon. However, with field data, uncertainties in the activation analysis results may be reduced or eliminated. TLG's goal through this comparison is to TLG SERVICES  !

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rsv. O PROPRIETARY INFORMATION Section 7, Page 5 of 27 FINAL REPORT l

demonstrate that the results obtained for the Saxton activation analysis are reasonably conservative.

7.4.1 Comparison to TLD String Data Thermoluminescent dosimetry (TLD) strings were placed roughly at the northeast, northwest, southeast and southwest locations outside the reactor vessel support can, roughly 74"-76" from the reactor vertical central axis. Integrated gamma exposures and calculated exposure rates were presented to TLG in Reference [18].

TLG used the MICROSHIELD point kernel shielding computer code, with activation analysis generated specific curie contents, to compute exposure rates at the core centerline.

The following assumptions were made regarding the MICROSHIELD calculations:

1. Calculational results representing an average of the four regions over the core height of 36.61 inches (92.99 cm.) were made. Using these results over the core height only, will tend to underestimate the dose by <5% due to activation of individual components above and below the active core.

i

2. At this distance from the vessel axis, axial peaking (1.65) is ignored. This is expected to have a small, non-conservative impact upon the calculated results.

3. Local peaking (within each model sector) was also ignored at this source-detector distance.

Calculation source-detector geometries representing the north side of the vessel will "see" the higher half of the activated circumference (Sectors A, B, C, and H in Figure 6.1).

Calculations for the south side will "see" the lower half (Sectors D, E, F, and G in Figure 6.1).

The MICROSHIELD calculations generated an azimuthally-averaged, core centerline exposure rate of 558.0 mR/hr. Multiplying in the north side and south side factors l l

applicable due to the core. loading asymmetry, TLG calculates maxunum and mimmum exposure rates at the core centerline, 76" from the reactor vertical central axis to be 955 mR/hr and 197 mR/hr, for the north side and south side of the reactor, respectively.

Based upon data presented by GPU Nuclear to TLG, and a discussion with GPU Nuclear personnel [19), the expected maxunum exposure rates on the north side and south side of the reactor are 350 mR/hr and 70 mR/hr, respectively. These estimates are based upon the TLD string data. Using these estimates, the calculated data appears to be factors of 2.7 and 2.8 for the high and low exposure rate calculations, respectively.

TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARYINFORMATION Section 7, Page 6 of 27 FINAL REPORT This is a reasonably good comparison, particularly when it is recogmzed that the radial activation is likely to be overestimated by a factor of two due to core periphery effects alone.

When this effect is accounted for, the calculations are only 40% high, within a factor of two, considered appropriate for this type of calculation.

7.4.2 Comparison to External Exposure Rate Measurements MICROSHIELD calculations, similar to those performed for the TLD strings 76" from the reactor vertical axis, were made for detector locations on contact with the vessel support can.

These calculations were compared with contact exposure rate measurements made during characterization surveys performed by GPU Nuclear personnel.

MICROSHIELD calculations were made using the following input assumptions:

1. The detector is located 2 cm. from the surface of the vessel support can.

2. The proximity of the detector to the radioactive sources is expected to influence the results. At this source-detector distance, axial peakmg and azimuthal peaking factors become important, particularly for the vessel support and insulation cans. As previously mentioned, the axial peaking factor is 1.65. In this case the detector is assumed to "see" only the maxunum (Sectors A&B) and miniuum (Sectors F&G) quadrants.

Core centerline MICROSHIELD calculations yielded " contact" exposure rates of 14.96 and 1.460 R/hr, for the north side and south side of the reactor, respectively. GPU Nuclear survey data showed north side exposure rates of approximately 5 IUhr and south side rates of roughly 0.4 R/hr.

The MICROSHIELD calculations show an overestimate of 3.0 on the north side and 3.7 on the south side. When removing a factor of two to account for core periphery effects, factors of 1.5 and 1.8 remain. Again, TLG believes this to be a reasonably good comparison to measured results, providing a conservative estimate of the curie contents and radiation levels for the irradiated components.  ;

7.4.3 Comparison to Internal Exposure Rate Measurements l GPU Nuclear was able to place inside the reactor pressure vessel, a TLD string used to measure the exposure rates and determine the axial profile of neutron activation.

MICROSHIELD calculations were performed for a detector located on the vessel vertical axis, core centerline. The only components included in the source term (and considered to TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O  ;

PROP.RIETARY INFORMATION Section 7, Page 7 of 27 1 FINAL REPORT l

contribute significantly to the measured exposure) were the core baffle, thermal shield, upper and lower core plates and the lower core blocks. '

The total calculated exposure rate was 1,728 R/hr. Unlike the ex-vessel exposure rate j measurements, there will be no peaking effects to be considered here. The core periphery l efftets are assumed to reduce the calculated exposure rate contributions from the core baffle i and thermal shield by a factor of two. This reduces the calculated exposure rate to 1,098 R/hr, the exposure rate which should be compared directly to the TLD string data.

Data obtained from GPU Nuclear (20] shows 1378 R/hr at the core centerline. However, GPU Nuclear did note that some difficulty was encountered removing the TLD string, which was estimated to increase the exposure time by 20%, from the estimated 10 minute exposure to 12 minutes. Accounti_3 or f this additional exposure time reduces the measured exposure rate to 1148 R/hr. TLG's calculated values are within 5% of the measured results, considered to be satisfactory.

7.4.4 Comparison to Concrete Boring Samples and Stainless Steel Samples from Above the Operating Water Level i

The activation analysis results for concrete and steel above the operating water level were not expected to yield highly accurate results. Using one. dimensional neutron transport calculations radially outward to the outer surface of the reactor pressure vessel support can,

, and then axially up to above the water level, as was done for this calculation, is likely to underestimate the neutron source term above the operating water level.

In reality, there existed a significant source of neutrons, which " streamed" up and down in the annulus between the vessel wall and vessel support can (the fiberglass insulation did not exhibit a significant deterrent to this neutron source). The only way to appropriately calculate the neutron source term in regions where streaming is anticipated to cause a problem is to perform a three-dimensional Monte Carlo calculation. This type of analysis is not recommended for Saxton since the cost anticipated for this type of work outweighs the benefit expected to be derived from the results. Currently existing characterization data will bound activation of concrete and steelin this region.

TLG has reviewed concrete core data from above the operating water level provided by GPU Nuclear. TLG's estimated specific curie contents above the operating water level appear to be underestimated by a factor of 10 to 18.

i TLG SERVICES

SAXTON ACTIVATION ANALYSIS Document No. G01-1192-003, Rev. O PROPRIETARY INFORMATION Section 7, Page 8 of 27 FINAL REPORT Calculated Values Measured Values Location #Co (uCi/em) 152Eu (uCi/em) #Co (uCi/rm) 152Eu (uCi/em)

North Side 0.420 2.265 4.21 37.95 South Side 0.072 0.387 1.12 6.80 The calculated values for steel activation compare somewhat more favorably with the measured values shown below. Generally, the calculated values are within a factor of 10 of the measured values.

Calculated Values (pCi/gm) Measured Values (pCi/gm)

Location I:ast Wall South Wall East Wall South Wall "Co 79 27 143 182  ;

"Ni 129 44 333 447 1 TLG cannot explain the signi6 cant difference between calculated and field results except that from the field results, there appears to be a sigmficant " streaming" neutron source which cannot be calculated by the analysis techniques used herein. TLG strongly suggests additional core bore and structural steel samples be taken during decommissioning to minimize the amount of LLRW generated.

l 7.4.5 Comparison to Structural Stainless Steel Sample at Core Midplane t The activation of structural stainless steel at the core centerline was determined from a l i piece of the ex-core neutron detector support on the northeast side of the vessel support can. l Tabulated below are the calculated and measured results. Based upon GPU Nuclear input, the middle of the sample appeared to reside roughly 15 inches (38.10 cm.) from the wall of the support can. This corresponds to the 3rd 6-inch water shield stainless steel region in the I activation analysis model.  !

, Calculated Values (oCi/em) Measured Values (oCi/em) l 55Fe 79 5 l 2

1. Co 1172 608 aNi 1837 525 l Again, the comparison of calculated results to measured values is reasonable, except for 55Fe. This particular isotope is diflicult to measure, even under laboratory conditions, and

, should be weighed lightly in this comparison. l l

1 TLG SERVICES

n-2 SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT TABLE 7.

THE SAXTON FACILITY COMll

'IWENTY-FOUR YEARS AFTER -

J CORE BAFB l

l Component Volume H3 C-14 Ar-39 Ca-41 Ca 45 Mn-54 Fe-55 Co-0 local!90 fem ^3) (Curies) (Curies) [CLes) (Curies) (Curies) (Cunesi (Gudes) (Gud!l 12th 3" Below Reactor Core 11th 3" Below Reactor Core 10th 3" Below Reactor Core 9th 3* Delow Reactor Core 8th 3" Below Reactor Core 7th 3" Delow Reactor Core l 6th 3" Below Reactor Core i

Sth 3* Delow Reactor Core  !

4th 3" Delow Reactor Core l

3rd 3" Below Reactot Core j 2nd 3* Below Reactor Core 9.915E+03 1.082E 02 8.246E-03 1.418E 05 1.774 E-06 5.426E-20 1.777E-08 2.574 E-01 4.5491 1st 3* Below Reactor Core 3.720E+03 2.160E 02 1.646E-02 2.831 E-05 3.543E-06 1.083E-19 3.547E-08 5.139E-01 9.0831 Reactor Core 4.231E+04 5.886E-01 4.487E-01 7.716E-04 9.655E-05 2.952E-18 9.667E-07 1.400E+01 2.4751 ist 3" Above Reactor Core 1.266E+04 3.446E-02 2.627E-02 4.518E-05 5.653E-06 1.728E 19 5.660E-08 8.199E-01 1.449%

2nd 3" Above Reactor Core 3rd 3" Above Reactor Core 4th 3" Above Reactor Core l 5th 3* Above Reactor Core 6th 3" Above Reactor Core 7th 3" Above Reactor Core 8th 3" Above Reactor Core 9th 3* Above Reactor Core 10th 3" Above Reactor Core 11th 3" Above Reactor Core 12th 3* Above Reactor Core Total 6.860E+04 6.555E-01 4.997E-01 8.593E-04 1.075E-04 3.288E 18 1.077E-06 1.560E+01 2.757E .

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Document No. G01-1192-003, Rev. O Section 7, Page 9 of 27

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O NI-59 Ni-63 Nb-94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 Total (Cudes) (Gudes) (Cudes) (Cunes) (Cunes) l31 (GudeM (Cudes) (Cudes) (CunM 00 4.542E-02 5.734E+00 1.012E-04 1.987E-05 6.239E 26 4.627E-08 1.129E-08 4.832E-04 1.061 E+01

{00 9.068E-02 1.145E+01 2.020E-04 3.968E-05 1.246E-25 9.238E-00 2.254E-08 9.647E-04 2.117E+01 (02 2.471Eo00 3.120E+02 5.505E-03 1.081 E-03 3.395E 24 2.518E-06 6.144E-07 2.629E-02 5.

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SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT TABLE 7e l

THE SAXTON FACILITY COMI TWENTY-FOUR YEARS AFTER THERMALSE Component Voluma H.3 C-14 Ar39 Ca-41 Ca-45 Mn-54 Fe-55 Cc lacM9.0 (cm^3) Nur gM (Curles) MudgM NurigO (Curles) (Curies) (Curles) @

12th 3" Below Reactor Core 1.844 E+03 1.375E-08 7.607E-09 3.121 E-11 1.555E-12 4.916E 26 3.512E-14 2.301 E-07 11th 3" Below Reactor Core 1.844E+03 2.872E-08 1.589E-08 6.520E-11 3.249E-12 1.027E-25 7.338E-14 4.808E-07 6.0 1.2 10th 3" Below Reactor Core 1,844E+03 6.133E-08 3.394E-08 1.393E-10 6.939E-12 2.193E-25 1.567E-13 1.027E-06 2.7 9th 3" Below Reactor Core 1.844E+03 1.344E-07 7 440E-08 3.052E-10 1.521E-11 4.807E-25 3.435E-13 2.251 E-06 5.9:

8th 3" Below Reactor Core 1.844 E+03 3.035E-07 1.679E-07 6.890E-10 3.433E-11 1.085E-24 7.754E-13 5.081 E-06 1.3$

7th 3* Below Reactor Core 1.844E+03 6th 3* Below Reactor Core 7.088 E-07 3.923E-07 1.609E-09 8.019E-11 2.535E-24 1.811E-12 1.187E-05 3.id 1.668E+04 1.559E-05 8.626E-06 3.539E-08 1.763E.99 5.574E-23 3.983E-11 2.610E-04 6.83 Sth 3" Below Reactor Core 2.409E+04 5.760E-05 3.188E-05 1.308E-07 6.517E-09 2.060E-22 1.472E-10 9.644 E-04 2.52 4th 3" Below Reactor Core 2.409E+04 1.583E-04 8.761 E-05 3.595E-07 1.791 E-08 5.661 E-22 4.045E-10 2.650E-03 7,%

3rd 3" Below Reactor Core 2.409E+04 4.894 E-04 2.708E-04 1.111 E-06 5.537E-08 1.750E-21 1.251 E-09 8.193E-03 2.1a 2nd 3" Below Reactor Core 2.409E+04 1.741E 03 9.635E-04 3.953E-06 1.970E-07 6.226E-21 4.449E-09 2.915E-02 7.7d 1st 3" Below Reactor Core 2.409E+ 04 5.015E-03 2.775E-03 1.139E-05 5.674 E-07 1.793E-20 1.281 E-08 8.396E-02 2.29 Reactor Core 2.940E+05 1.418 E-01 7.849E 02 3.220E-04 1.605E-05 5.072E-19 3.624E-07 2.375E+00 6.27 1st 3" Above Reactor Core 2.409E+04 5.076E-03 2.809E-03 1.153E-05 5.743E-07 1.815E-20 1.297E 08 8.498E-02 2.2q 2nd 3* Above Reactor Core 2.409E+04 1.835E-03 1.015E-03 4.166E-06 2.076E-07 6.561 E-21 4.688E-09 3.072E-02 8.1(

3rd 3" Above Reactor Core 2.409E+04 5.354 E-04 2.963E-04 1.216E-06 6.057E-08 1.915E-21 1.368E-09 8.964E-03 2.9I 4th 3" Above Reactor Core 2.409E+04 1.772E-04 9.809E-05 4.024 E-07 2.005E-08 6.338E-22 4.529E-10 2.967E-03 7,9(

5th 3" Above Reactor Core 2.409E+04 6.587E-05 3.646E-05 1.496E-07 7.453E-09 2.356E-22 1.683E-10 1.103E-03 2.90 6th 3" Above Reactor Core 2.409E+04 2.628E-05 1.455E-05 5.968 E-08 2.974 E-09 9.399E-23 6.716E-11 4.400E-04 1.1@

7th 3" Above Reactor Core 8.031 E+03 3.674E-06 2.033E-06 8.341 E-09 4.156E 10 1.314E-23 9.387E-12 6.151 E-05 1.6%

8th 3* Above Reactor Core 9th 3" Above Reactor Core 10th 3" Above Reactor Core 11th 3" Above Reactor Core 12th 3" Above Reactor Core Total 5.948E+05 1.570E 01 8.690E-02 3.565E-04 1.777E-05 5.615E-19 4.013E-07 2.629E+00 6.9%

m. i I

3 Docurnent No. G01-1192-003, Rev. O Section 7, Page 10 of 27 l

ANSTEC APERTURE y CARD

)NENT ACTIVATION EACTOR SHUTDOWN Also Available on Aperture Card ELD 50 NI59 Ni-63 Nb-94 Tc.99 Sn 119m Sb-125 Te 125m Eu 152 Total n) (Curies) (Gutin) (Gurin) ICurles) LGudu) (Gudn) Kulin) (Gudu) (CEin)

E-06 4.329E-08 5.231 E-06 1.565E 10 3.996E-11 2.251 E-15 5.495E-16 5.471 E-09 1.161E-05 E-05 8.836 E-08 1.093E-05 3.270E 10 8.348E 11 4.704E-15 1.148E-15 1.143E-08 2.426E 05 E-05 1.887E 07 2.334E-05 6.984E 10 1.783E 10 1.005E-14 2.452E-15 2.441E 08 5.181E-05 E-05 4.138E-07 5.115E-05 1.531E-09 3.908E 10 2.202E-14 5.373E-15 5.351 E-08 1.136E-04 E-04 9.337E-07 1.155E-04 3.455E-09 8.821E-10 4.970E-14 1.213E-14 1.208E-07 2.563E-04 E 04 2.181 E-06 2.697E-04 8.070E-09 2.060E-09 1.161E-13 2.833E-14 2.821E-07 5.987E-04 E-03 4.796E-05 5.931 E-03 1.775E-07 4.531E 08 2.553E-12 6.231E-13 6.204E 06 1.317E-02 E02 1.772E-04 2.192E-02 6.559E-07 1.674E-07 9.435E.12 2.303E-12 2.293E-05 4.866E-02 E-02 4.871 E-04 6.024 E-02 1.803E-06 4.602E-07 2.593E-11 6.328E-12 6.301 E-05 1.337E-01

-01 1.506E 03 1.882E-01 5.572E-06 1.423E-06 8.016E 11 1.0565-11 1.948E-04 4.134E-01 E 01 5.357E 03 6.625E-01 1.982E-05 5.061E-06 2.852E.10 6.959E 11 6.930E-04 1.471 E+00

?000 1.543E-02 1.908E+00 5.710E-05 1.458E-05 8.214E.10 2.00SE 10 1.996E-03 4,236E+00

001 4.364 E-01 5.397E+01 1.615E-03 4.123E-04 2.323E.08 5.669E-09 5.645E-02 1.198E+02

+00 1.562E-02 1.932E+00 5.780E-05 1.476E-05 8.314E 10 2.029E-10 2.020E-03 4.288E+00 1 01 5.645E-03 6.981 E-01 2.089E-05 5.333E-06 3.005E-10 7.334E-11 7.302E-04 1.550E+00

-01 1.647E-03 2.037E-01 6.096E-06 1.556E-06

-02 8.769E 11 2.140E-11 2.131E-04 4.523E-01 0.453E-04 6.744 E-02 2.018E-06 5.152E 07 2.903E-11 7.085E-12 7.054E-05 1.497E-01

03 3.037E-04 2.507E-02 7.501 E-07 1.915E-07 1.07CE 11 2.633E-12 2.622E-05 5.565E-02

-02 8.007E-05 1.000E-02 2.993E-07 7.640E-08 4.305E-12 1.051E 12 1.046E-05 2.220E-02

-03 9.130E-05 1.398E-03 4.183E-08 1.068E-08 6.017E 13 1.468E 13 1.462E-06 3.103E-03

+01 4.831E 01 5.975E+01 1.788E-03 4.565E-04 2.572E-08 6.277E-09 6.250E-02 1.327E+02 Mo?2@298 ,

i d ' ann ****

SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION <

FINAL REPORT l TABLE 7.;

i THE SAXTON FACILITY COMl!

TWENTY-FOUR YEARS AFTER {

I REACTOR VESS3 Component Volume H-3 C-14 Ar-39 Ca-41 Ca-45 Mn-54 Fe-55 Co-6 Location (cm^ 3) (Cunes) (Curies) (Curies) (Curies) (Cunes) (Curies) (Curies) (C_y1 12th 3* Below Reactor Core 7.688E+ 04 4.570E-07 2.393E-07 5.762E-10 5.086E 11 1.569E-24 6.462E-13 7.418E-06 1.4923 11th 3* Delow Reactor Core 7.449E+02 9.249E-09 4.843E-09 1.166E 11 1.029E-12 3.176E 26 1.308E-14 1.501 E-07 3.0201 10th 3" Below Reactor Core 1.117E+03 2.963E-08 1.552E-08 3.736E-11 3.298E-12 1.017E 25 4.190E-14 4.810E-07 9.6752 9th 3" Below Reactor Core 1.117E+03 6.494 E-08 3.401E-08 8.189E-11 7.228E-12 2.230E-25 9.184E 14 1.054 E-06 2.1201 8th 3* Below Reactor Core 1.117E+03 1.466E-07 7.677E-08 1.848E-10 1.632E-11 5.034 E-25 2.073E 13 2.380E-06 4.787@

7th 3* Delow Reactor Core 1.117E+03 3.424 E-07 1.793E-07 4.317E-10 3.811E 11 1.176E-24 4.842E-13 5.558E-06 1.1181 Sth 3" Below Reactor Core 1.117E+03 8.327E-07 4.361E-07 1.050E-09 9.268E-11 2.859E 24 1.178E 12 1.352E-05 2.7191 5th 3" Below Reactor Core 1.117E+03 2.130E-06 1.115E-06 2.686E-09 2.371E-10 7.314E-24 3.012E-12 3.458E-05 6.955 1 4th 3* Below Reactor Core 1.117E+03 5.854 E-06 3.066E-06 7.381 E-09 6.515E-10 2.010E-23 8.279E-12 9.503E-05 1.911 1 3rd 3* Delow Reactor Core 1.117E+03 1.810E-05 9.477E-06 2.282E-08 2.014E-09 6.214E 23 2.559E 11 2.938E-04 5.90F@

2nd 3* Below Reactor Core 1.117E+03 6.438E-05 3.371 E-05 8.118E-08 7.165E-09 2.211 E-22 9.104 E-11 1.045E-03 2.1021 1st 3* Delow Reactor Core 1.117E+03 1.854E 04 9.711 E-05 2.338E-07 2.064 E-08 6.367E-22 2.622E 10 3.010E-03 6.0551 Reactor Core 1.364 E+04 5.245E-03 2.747E-03 6.613E-06 5.837E-07 1.801 E-20 7.417E-09 8.514E-02 1.713E(

1st 3* Above Reactor Core 1.117E+03 1.877E-04 9.829E-05 2.367E-07 2.089E-08 6.445E-22 2.654 E-10 3.047E-03 6.1295 I 2nd 3* Above Reactor Core 1.117E+03 6.785E-05 3.553E-05 8.554E 08 7.550E-09 2.329E 22 9.594E-11 1.101 E-03 2.215El 3rd 3* Above Reactor Core 1.117E+ 03 1.980E 05 1.037E-05 2.496E-08 2.203E-09 6.798E 23 2.800E-11 3.214 E-04 6.464 E:

4th 3* Above Reactor Core 1.117E+03 6.554E-06 3.432E-06 8.264 E-09 7.294 E-10 2.250E-23 9.268E-12 1.064 E-04 2.140E:

5th 3" Above Reactor Core 1.117E+03 2.436E-06 1.276E-06 3.071 E-09 2.711 E-10 8.364E 24 3.445E-12 3.954 E-05 7.954 E '

6th 3* Above Reactor Core 1.117E+03 9.720E-07 5.089E-07 1.225E-09 1.082E-10 3.337E-24 1.374E-12 1.578E-05 3.173E l 7th 3" Above Reactor Core 1.117E+03 4.076E-07 2.134 E-07 5.139E-10 4.536E-11 1.399E-24 5.763E-13 6.616E-06 1.331 E '

8th 3" Above Reactor Core 1.117E+03 1.778E-07 9.308E 08 2.241E 10 1.978E-11 6.103E-25 2.514E-13 2.885E-06 5.804E<

9th 3" Abovo Reactor Core 1.117E+03 8.014 E-08 4.196E-08 1.010E-10 8.918E-12 2.751 E-25 1.133 E-13 1.301E-06 2.616EJ 10th 3" Above Reactor Core 1.117E+03 3.717E-08 1.947E-08 4.687E-11 4.137E-12 1.276E 25 5.2F7E-14 6.034 E-07 1.214EJ 11th 3* Above Reactor Core 1.117E+03 1.769E-08 9.263 E-09 2.230E-11 1.969E-12 6.074 E-26 2.502E-14 2.871 E-07 5.776EJ 12th 3" Above Reactor Core 1.117E+03 8.601 E-09 4.504E-09 1.084E-11 9.572E-13 2.953E 26 1.216E-14 1.396E-07 2.808Ea Total 1.158E+05 5.809E 03 3.042E 03 7.324E-06 6.465E-07 1.995E-20 8.215E-09 9.429 E-02 1.897Ec

c-Document No. G01-1192-003, Rev. O Section 7, Page 11 of 27 ANSTEC APERTURE ONENT ACTIVATION tEACTOR SHUTDOWN Also Available on Aperture Card L CLAD Ni-59 Ni-63 Nb-94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 Total d (Curies) (CutifuQ (Curles) (Curies) (Curies) (CV.das) (Curies) (Curies) (Cartes}

04 1.431E-06 1.703E 04 3.412E-09 7.442E-10 2.759E-14 6.734 E-15 1.452E-07 3.292E-04 D6 2.896E 08 3.448E-06 6.907E-11 1.506E-11 5.585E-16 1.363E-16 2.938E-09 6.664 E-06 D6 9.279E-08 1.105E-05 2.213 E-10 4.826E-11 1.789E-15 4.366E-16 9.414E-09 2.135E.05 35 2.034 E-07 2.421 E-05 4.850E-10 1.C58E 10 3.921E-15 9.570E-16 2.063E 08 4.679E-05 35 4.591 E-07 5.465E-05 1.095E-09 2.388E-10 8.852E-15 2.160E-15 4.658E-08 1.056E-04 34 1.072E-06 1.276E-04 2.557E-09 5.577E-10 2.068E-14 5.046E-16 1.088 E-07 2.467E-04 34 2.608E-06 3.104E-04 6.218E-09 1.356E-09 5.028E-14 1.227E-14 2.646E-07 6.000E-04 34 6.670E-06 7.941 E-04 1.591 E-08 3.409E-09 1.296E-13 3.139E-14 6.767E-07 1.535E-03 33 1.833E 05 2.182E-03 4.372E-08 9.635E-09 3.535E-13 8.627E-14 1.660E-06 4.218E-03 33 5.667E-05 6.746E-03 1.351 E-07 2.947E-08 1.093E-12 2.667E-13 5.750E-06 1.304E-02 33 2.016E 04 2.400E-02 4.808 E-07 1.049E-07 3.887E-12 9.487E-13 2.045E-05 4.639E-02 32 5.807E-04 6 913E-02 1.385E-06 3.020E-07 1.120E-11 2.733E-12 5.892E-05 1.336E-01 DO 1.642E-02 1.955E+00 3.917E-05 8.542E-06 3.167E-10 7.729E-11 1.666E-03 3.779E+00 32 5.878E-04 6.997E 02 1.402E 06 3.057E-07 1.133E-11 2.766E-12 5.963E-05 1.352E-01 22 2.124E-04 2.529E-02 5.066 E-07 1.105E-07 4.096E-12 9.997E-13 2.155E-05 4.888E-02 13 6200E-05 7.380E-03 1.478E 07 3.225E-08 1.195E-12 2.917E-13 6.290E-06 1.426E-02 33 2.052E-05 2.443E-03 4.894E-08 1.067E 08 3.957E-13 9.658E-14 2.082E-06 4.722E-03 14 7.628E-06 9.081E-04 1.819E-08 3.967E 09 1.471E 13 3.590E-14 7.739E-07 1.755E-03

)4 3.044E 06 3.623E-04 7.258E-09 1.583 E-09 5.869E-14 1.432E-14 3.088E-07 7.003E-04 14 1.276E-06 1.519E-04 3.043E-09 6.638E 10 2.461E 14 6.006E-15 1.295E-07 2.936E.04

)$ 5.566E-07 6.626E-05 1.327E 09 2.895E-10 1.073E-14 2.619E-15 5.647E-08 1.281E-04 15 2.509E-07 2.987E-05 5.984E 10 1.305E-10 4.839E-15 1.181E-15 2.546E-08 5.774 E-05 15 1,164E-07 1.386E-05 2.776E 10 6.054E 11 2.245E-15 5.478E-16 1.181 E-08 2.678E-05 IG 5.539E-08 6.594E-06 1.321E-10 2.881E-11 1.068E-15 2.607E-16 5.620E-09 1.275E-05 16 2.693E-08 3.206E-06 6.423E-11 1.401E-11 5.193E-16 1.267E 16 2.732E-09 6.197E-06

)0 9.819E-02 2.165E+00 4.338E-05 9.461 E-06 3.507E-10 8.500E-11 1.845E-03 4.185E+00 9(p07270298 - 3 -

,c- m l l

l SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT I l

TABLE 7.4 THE SAXTON FACILITY COMB TWENTY-FOUR YEARS AFTER I REACTOR VESS2 Component ,

Volurne H-3 C-14 Ar-39 Cc-41 Ca-45 Mn-54 Fe-55 Co4 Location fem ^3) (Curies) (Curies) (Cunes) (Curies) (Curies) (Curies) (Cunes) (Quds 12th 3" Below Reactor Core 3.349E+06 7.677E-06 3.766E 07 5.478E-08 2.861 E-10 9.007E-24 2.251 E-11 8.384E-05 1.8903 11th 3" Below Reactor Core 3.243E+04 1.553E-07 7.620E-09 1.108E-09 5.789E-12 1.822E-25 4.554E-13 1.696E-06 3.8253 10th 3" Below Reactor Core 4.865E+04 4.977E-07 2.441 E-08 3.551 E-09 1.855E-11 5.839E-25 1.459E-12 5.435E-06 1.2253 9th 3" Below Reactor Core 4.865E+04 1.091E-06 5.351E-08 7.782E-09 4.065E-11 1.280E-24 3.198E-12 1.191 E-05 2.6863 8th 3" Below Reactor Core 4.865E+04 2.462E-06 1.208E-07 1.757E-08 9.176E-11 2.889E-24 7.218E-12 2.689E-05 6.0633 7th 3" Delow Reactor Core 4.865E+04 5.751 E-06 2.821E-07 4.103E-08 2.143E-10 6.747E-24 1.686E-11 6.280E-05 1.4163 6th 3" Below Reactor Core 4.865E+04 1.399E-05 6.861 E-07 9.979E-08 5.212E-10 1.641 E-23 4.100E-11 1.527E-04 3.4443 Sth 3* Below Reactor Core 4.865E+04 3.577E-05 1.755E-06 2.553E-07 1.333E-09 4.197E-23 1.049E-10 3.907E-04 8.8093 4th 3" Delow Reactor Core 4.865E+04 9.832E-05 4.823E-08 7.015E-07 3.664 E-09 1.154E-22 2.882E 10 1.074 E-03 2.4213 3rd 3" Below Reactor Core 4.865E+04 3.039E-04 1.491 E-05 2.169E-06 1.133E-08 3.566E-22 8.910E-10 3.319E-03 7.4841 2nd 3" Delow Reactor Core 4.865E+04 1.081 E-03 5.304 E-05 7.715E-06 4.030E-08 1.269E-21 3.170E-09 1.181 E-02 2.6623 1st 3" Below Reactor Core 4.865E 604 3.115E-03 1.528E-04 2.222E-05 1.161 E-07 3.654 E-21 0.131 E-09 3.401 E-02 7.6693 Reactor Core 5.937E+05 8.809E-02 4.321 E-03 6.285E-04 3.283E-06 1.033E 19 2.582E-07 9.620E-01 2.169E-1st 3" Above Reactor Core 4.865E+04 3.152E-03 1.547E-04 2.249E-05 1.175E-07 3.699E-21 9.242E-09 3.443 E-02 7.7621 2nd 3" Above Reactor Core 4.865E+04 1.139E-03 5.590E-05 8.130E-06 4.246E-08 1.337E-21 3.340E-09 1.244 E-02 2.806]

3rd 3" Above Reactor Core 4.865E+04 3.325E-04 1.631 E-05 2.373E-06 1.239E-08 3.901 E-22 9.748E-10 3.631E-03 8.1873 4th 3" Above Reactor Core 4.865E+04 1.101 E-04 5.400E-06 7.854E-07 4.102E-09 1.291E-22 3.227E-10 1.202E-03 2.710]

5th 3" Above Reactor Core 4.865E+04 4.091 E-05 2.007E-06 2.919E-07 1.525E-09 4.800E-23 1.199E-10 4.468E-04 1.0073 6th 3" Above Reactor Core 4.865E+04 1.632E-05 8.008E-07 1.165E-07 6.083E-10 1.915E 23 4.785E-11 1.783E-04 4.0193 7th 3" Above Reactor Core 4.865E+04 6.845E-06 3.358E-07 4.884 E-08 2.551E 10 8.031 E-24 2.007E-11 7.475E-05 1.6853 8th 3" Above Reactor Core 4.865E+04 2.985E-06 1 A64E-07 2.130E-08 1.113E-10 3.502E-24 8.752E-12 3.260E-05 7.3503 9th 3" Above Reactor Core 4.865E+04 1.346E-06 6.602 E-08 9.603E-09 5.016E 11 1.579E 24 3.945E-12 1.470E-05 3.3143 10th 3" Above Reactor Core 4.865E+04 6.243E-07 3.063E-08 4.455E-09 2.327E-11 7.325E 25 1.830E-12 6.818E-06 1.5373 11th 3" Above Reactor Core 4.865E+04 2.971E-07 1.457E-08 2.120E-09 1.107E-11 3.486E-25 8.710E-13 3.244 E-06 7.3153 12th 3" Above Reactor Core 4.865E+04 1.444 E-07 7.086E-09 1.031 E-09 5.383E-12 1.695E-25 4.235E-13 1.577E-06 3.5573 Total 5.045E+06 9.756E-02 4.786E-03 6.961 E-04 3.636E-06 1.145E-19 2.860E-07 1.065E+00 2.402E :

" ~~e

o ._-

Document No. G01-1192-003, Rev. O Section 7, Page 12 of 27 ANSTEC APERTURE

)NENT ACTIVATION EACTOR SHUTDOWN Al Availagtegn aWALL Ni-59 NI-63 Nb-94 Tc 99 Sn-119m Sb-125 Te-125m Eu-152 Total (Cudes) (Cudes) (Curies) (Cudes) (Cydt31 (Gud63} (CMdes) (Curies) (Cudes) 4 6.541E-07 8.801 E-05 1.310E 08 3.823E-12 1.545E-13 3.773E-14 3.101E-06 3.727E-04 6 1.334E-08 1.781E 06 2.651E-10 7.736E-14 3.127E 15 7.635E-16 6.274E-08 7.542E-06 5 4.240E-08 5.705E-06 8.493E-10 2.478E-13 1.002E-14 2.446E-15 2.010E-07 2.416E-05 5 9.294 E-08 1.250E-05 1.861E-09 5.432E 13 2.196E-14 5.361 E-15 4.405E-07 5.296E-05 6 2.098E-07 2.823E-05 4.202E-09 1.226E-12 4.956E 14 1.210E-14 9.044E-07 1.195E-04 4 4.900E-07 6.593E-05 9.815E-09 2.864 E-12 1.158E-13 2.827E-14 2.323E-06 2.792E-04 4 1.193E-06 1.603E-04 2.387E-08 6.965E-12 2.815E-15 6.874 E-14 5.648E-06 6.790E-04 4 3.048E-06 4.101 E-04 6.105E-08 1.782E-11 7.201E-13 1.758E-13 1A45E-05 1.737E-03 3 8.378E-06 1.127E-03 1.678E-07 4.897E-11 1.979E-12 4.832E-13 3.971 E-05 4.774 E-03 3 2.590E-05 3.484 E-03 5.187E 07 1.514E-10 6.118E-12 1 A94E-12 1.228E-04 1.476E-02 2 9.213E-05 1.240E-02 1.845E-06 5.385E-10 2.177E-11 5.314E-12 4.367E-04 5.250E-02 2 2.654 E-04 3.570E-02 5.315E-06 1.551 E-09 6.270E 11 1.531E-11 1.258E 03 1.512E-01 10 7.506E-03 1.010E+00 1.503E-04 4.387E-08 1.773E-09 4.330E-10 3.558E-02 4.277E+00 2 2.686E-04 3.614 E-02 5.380E 06 1.570E-09 6.346E-11 1.549E-11 1.273E-03 1.531E 01 2 9.709E-05 1.306E-02 1.945E-06 5.675E-10 2.294E-11 5.600E-12 4.602E-04 5.532E-02 3 2.833E-05 3.812E-03 5.675E-07 1.656E 10 6.693E-12 1.634 E-12 1.343E-04 1.614E-02 3 9.379E-06 1.262E-03 1.879E-07 5.482E 11 2.216E-12 5.410E-13 4.446E-05 5.345E-03 3 3.486E-06 4.690E-04 6.982E-08 2.037E-11 8.235E-13 2.011E-13 1.652E-05 1.986E-03 4 1.391 E-06 1.871 E-04 2.786E-08 8.129E-12 3.286E-13 8.023E-14 6.593E-06 7.926E-04 4 5.832E-07 7.847E-05 1.168 E-08 3.409E 12 1.378E-13 3.364E-14 2.764 E-06 3.323E-04 5 2.544E-07 3.422E-05 5.095 E-09 1.487E-12 6.009E 14 1.467E 14 1.206E-06 1.449E-04 5 1.147E-07 1.543E-05 2.297E 09 6.703E 13 2.709E-14 6.615E-15 5.435E-07 6.534 E-05 5 5.320E-08 7.157E-06 1.066E-09 3.109E-13 1.257E-14 3.069E-15 2.521E-07 3.031E-05 S 3.531E-08 3.406E-06 5.070E-10 1.480E-13 5.980E-15 1.460E-15 1.200E-07 1.442E-05 S 1.231 E-08 1.656E-06 2.485E-10 7.194 E-14 2.908E-15 7.100E-16 5.834E-08 7.013E-06 0 8.313E-03 1.118E+00 1.665E-04 4.859E-08 1.964E-09 4.795E-10 3.940E-02 4.737E+00 PlaQ725~C298b _

.-n l

l SAXTON ACTIVATION ANALYSIS l

PROPRIETARY INFORMATION FINAL REPORT TABLE 7.

THE SAXTON FACILITY COMd TWENTY-FOUR YEARS AFTER (  ;

i VESSELINSULAT Cornponent )

Volume H-3 C-14 Ar-39 Ca-41 Ca-45 Mn-54 Fe-55 Cd Lontlen (cm^U (Gurin) (Curles) (Curies) (Curies) (Curles) (Curles) 12th 3* Below Reactor Core (Curles) 2.978E+03 9.591 E-10 6.184E-10 5.123E-12 1.143E-13 3.841 E-27 5.738E-15 1.758 E-08 7.6 %

11th 3* Below Reactor Core 2.978E+03 2.004 E-09 1.292E-09 1.070E-11 2.387E-13 8.025E-27 1.199E-14 3.673E 08 1.58 13 3* Below Rractor Core 2.978 E+03 4.279E-09 2.759E-09 2.286E-11 5.099E-13 1.714E-26 2.561 E-14 7.846E-08 3.3$

Jth b* Below Reactor Core 2.978E+03 9.379E-09 6.048 E-09 5.010E 11 1.118E-12 3.757E 26 5.612E-14 1.720E-07 7.40 8th 3* Below Reactor Core 2.978E+03 2.117E-08 1.365E-08 1.131 E-10 2.523E-12 8.480E-26 1.267E 13 3.882E-07 1.62 7th 3* Below Reactor Core 2.978E+03 4.945E-08 3.189E-08 2.641 E-10 5.892E-12 1.981 E-25 2.959E-13 9.067E-07 3.93 6th 3* Below Reactor Core 2.978E+03 1.203E-07 7.755E-08 6.423E-10 1.433E-11 4.817E-25 7.196E 13 2.205E-06 9.93l but 7'"abw Reactor Core 2.978E+03 3.076E-07 1.984E-07 1.643E-09 3.665E-11 1.232E-24 1.841E-12 5.640E-06 2.44 4th 3* Below Reactor Core 2.978 E+03 8.455E-07 5.452E-07 4.516E-09 1.007E-10 3.386E 24 5.059E-12 1.550E-05 6.7@

3rd 3* Below Reactor Core 2.978E+03 2.614 E-06 1.685 E-06 1.396E-08 3.114E 10 1.047E 23 1.564E-11 4.792E-05 2.0F 2nd 3* Below Reactor Core 2.978E+03 9.298E-06 5.995F-06 4.966E-08 1.108E-09 3.724 E-23 5.563E-11 1.705E-04 7.37 1st 3* Below Reactor Core 2.978E+03 2.678E-05 1.727Ni 1.430E-07 3.191 E-09 1.073E-22 1.602E-10 4.910E-04 2.18 Reactor Core 3.634 E+04 7.575 E-04 4.884 E-04 4.046E-06 9.026E-08 3.034 E-21 4.532E-09 1.389E-02 6.01 ist 3* Above Reactor Core 2.978E+03 2.711 E-05 1.748 E-05 1.448 E-07 3.230E-09 1.086E-22 1.622E-10 4.970E-04 2.19 2nd 3* Above Reactor Core 5.917E+03 1.947E-05 1.255 E-05 1.040E-07 2.320E-09 7.798E-23 1.165E-10 3.570E-04 1.96 3rd 3* Abose Reactor Core 1.180E+04 1.133 E-05 7.303E-00 6.049E-08 1.350E-09 4.536E-23 6.777E-11 2.077E-04 8.99 4th 3" Above Reactor Core 7.085E+04 2.252E-06 1.452E-05 1.203E-07 2.683E-09 9.020E-23 1.347E 10 4.129E-04 1,79 Sth 3* Above Reactor Core 2.767E+04 3.269E 06 2.108E-06 1.746E-08 3.895E-10 1.309 E-23 1.956E-11 5.993E-05 2.55 6th 3" Above Reactor Core 1.313E+ 04 6.188E-07 3.990E-07 3.305E-09 7.373E 11 2.478E 24 3.702E-12 1.135E-05 4.91 7th 3* Above Reactor Core 1.313E+04 2.595E-07 1.673 E-07 1.386E-09 3.092E-11 1.039E-24 1.553E 12 4.757E-06 2.05 8th 3* Above Reactor Core 1.313E+04 1.132E-07 7.297E-08 6.044E-10 ' 1.348E 11 4.533E-25 8.771E 13 2.075E-06 8.99 9th 3* Above Reactor Core 1.313E+04 5.102E-08 3.290E-08 2.725E-10 6.079E-12 2.043E-25 3.053E 13 9.354 E-07 4.04 10th 3" Above Reactor Core 1.313E+04 2.367E-08 1.526 E-08 1.264E-10 2.820E-12 9.479E 26 1.416E-13 4.339E-07 1.87 11th 3* Above Reactor Core 1.313E+04 1.126E-08 7.262E-09 6.015E 11 1.342E-12 4.511E-26 6.739E-14 2.065E-07 8.93 12th 3* Above Reactor Core 1.313E+04 5.476E-09 3.531 E-09 2.925E 11 6.524E-13 2.193E-26 3.276E-14 1.004E-07 4.34 Total 2.832E+05 8.823E-04 5.689E-04 4.713E-06 1.051E-07 3.534E-21 5.279E-09 1.618E-02 7.00

.r~- w

Document No. G01-1192-003, Rev. O Section 7, Page 13 of 27 l 5 ANSTEC I APERTURE ONENT ACTIVATION CARD bEACTOR SHUTDOWN I l

Also Avaltable on  ;

Aperture Card l i

[ON CAN

.-60 Ni59 NI-63 Nb 94 Tc 99 Sn-119m Sb-125 Te-125m Eu 152 Total rin) (Gil_du) (Curles) (Curiest (Curles) (Gmin) (Curle1) (Gmin) (Curies) (Gud.0) l E-07 2.799E-09 3.885E-07 2.225E-11 6.651E 12 2.387E-17 5.829E 18 6.470E-10 1.172E-06 pE-06 5.847E-09 8.118E-07 4.64BE-11 1.389E-11 4.986E 17 1.218E 17 1.352E-09 2.449E-06

$E-06 1.249E-08 1.734E-06 9.927E-11 2.967E 11 1.065E-16 2.601E-17 2.887E-09 5.231E-06

)E-06 2.737E-08 3.799E-06 2.176E-10 6.504E-11 2.334E 16 5.700E-17 6.327E-09 1.146E-05

?E-05 6.179E-08 8.577E-06 4.912E-10 1.468E-10 5 269E-16 1.287E-16 1.428E-08 2.588E-05 E-05 1.443E-07 2.003E 05 1.147E-09 3.429E 10 1.231 E-15 3.006E-16 3.336E-08 6.044 E-05 E 05 3.510E-07 4.872E-05 2.790E-09 8 339E 10 2.993E-15 7.309E-16 8.113E-08 1.470E-04 fE-04 8.97BE-07 1.24BE-04 7.136E-09 2.133E-09 7.656E-15 1.870E-15 2.075E-07 3.760E-04 (E 04 2.467E-06. 3.425E-04 1.961 E-08 5.863E-09 2.104 E-14 5.139E-15 5.704 E-07 1.033E-03 jE-03 7.638E-06 1.059E-03 6.063E-08 1.612E-08 6.504E-14 1.588E-14 1.763E-06 3.195E-03 E 03 2l/14E-05 3.767E-03 2.157E-07 6.447E-08 2.314E 13 5.651E-14 6.272E-06 1.136E-02 (E-02 7.816E-05 1.085E-02 6.213E-07 1.857E-07 6.665E-13 1.628E 13 1.807E-05 3.273E-02 fE-01 2.211E-03 3.069E-01 1.757E-05 5.253E 06 1.885E-11 4.604E 12 5.110E-04 9.25.8E-01 E-02 7.911E 03 1.098E-02 6.288E-07 1.860E-07 6.746E 13 1.648E-13 1.829E-05 3.313E-02 (E-02 5.682E-05 7.887E-03 4.518E-07 1.350E-07 4.845E-13 1.183E-13 1.313E-05 2.380E-02

E-03 j 3.305E-05 4.588E-03 2.627E-07 7.854 E-08 2.819E-13 6.884E-14 7.641E-06 1.384E-02 (E-02 6.572E-05 9.123E-03 5.224 E-07 1.562E-07 5.604E-13 1.369E-13 1.519E-05 2.752E-02 E 03 9.540E-06 1.324E-03 7.583E-08 2.267E-08 8.135E-14 1.987E-14 2.205E-06 3.995E-03 (E-04 1.806E-06 2.507E-04 1.435E-08 4.291 E-09 1.540E-14 3.761E-15 4.174E-07 7.563E-04 fE-04 7.573E-07 1.051 E-04 6.019E-09 1.799E-09 6.457E-15 1.577E 15 1.750E-07 3.171E-04

-05 3.303E-07 4.584E-05 2.625E-09 7.847E 10 2.816E 15 6.878E-16 7.634E-08 1.383E-04

-05 1.489E-07 2.067E-05 1.183E-09 3.538E 10 1.270E-15 3.101E-16 3.442E-08 6.236E-05

-05 0.907E-08 9.587E-06 5.490E-10 1.641E 10 5.890E-16 1.438E-16 1.597E-08 2.893E-05 00 3.287E-00 4.562E-06 2.813E 10 7.810E-11 2.803E 16 6.845E 17 7.597E-09 1.377E-05 06 1.598E-08 2.218E-06 1.270E 10 3.797E 11 1.363E 16 3.328E 17 3.694 E-09 6.693E-06

-01 2.575E-03 3.574E 01 2.047 E-05 6.118E-06 2.196E 11 5.363E-12 5.952E-04 1.078E+ 00 l

%o72SO2W- v

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SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT TABLE 7.)

THE SAXTON FACILITY COM$

TWENTY-FOUR YEARS AFTER [ ,

VESSEL SUPPOT 1

Component l

Volume H-3 C-14 Ar-39 Ca-41 Ca-45 Mn-54 Fe-55 Co-6!

LQO3.1100 (cm^3) (Gud.eM (Curies) (Curies) (Curies) (Curies) (Curies) (Curies) (Curig 12th 3" Below Reactor Core 6.294 E+03 3.512E-09 1.975E-09 9.113E-12 3.992E 13 1.269E-26 1.021E 14 5.936E-08 1.682$

11th 3" Below Roactor Core 6.294E+03 7.336E-09 4.126E-09 1.904E-11 8.339E-13 2.651E-26 2.133E-14 1.240E-07 3.514$

10th 3" Below Reactor Core 6.294 E+03 1.567E-08 8.813E-09 4.066E-11 1.781E-12 5.683E-26 4.555E-14 2.649E-07 7.505X 9th 3" Below Reactor Core 6.294 E+03 3.434 E-08 1.931 E-08 8.912E 11 3.904E-12 1.241 E-25 9.983E-14 5.805E-07 1.645%

8th 3" Below Reactor Core 6.294E+03 7.752E-08 4.360E-08 2.012E-10 8.812E-12 2.802E-25 2.254E-13 1.311E-06 3.713 %

7th 3" Delow Reactor Core 6.294 E+03 1.811 E-07 1.018E-07 4.699E-10 2.058E-11 6.544E-25 5.264E 13 3.061 E-06 8.6731 6th 3" Below Reactor Core 6.294E+03 4.403E-07 2.477E-07 1.143 E-09 5.005E 11 1.591 E-24 1.280E-12 7.444E-06 2.1091 Sth 3" Below Reactor Core 6.294E+03 1.126E-06 6.335E-07 2.923 E-09 1.280E-10 4.071E-24 3.274E-12 1.904E-05 5.395%

4th 3" Below Reactor Core 6.294E+03 3.096E-06 1.741E-06 8.033E-09 3.519E-10 1.119E-23 8.999E 12 5.233E-05 1.4831 3rd 3" Below Reactor Core 6.294 E+03 9.570E-06 5.382E-06 2.483E-08 1.088E-09 3.458E-23 2.782E-11 1.618E-04 4.564 %

2nd 3" Below Reactor Core 6.294 E+03 3.404E-05 1.915E-05 8.835E-08 3.870E-09 1.230E-22 9.897E-11 5.755E-04 1.631 1 i ist 3" Below Reactor Core 6.294E+03 9.806E-05 5.515E-05 2.545E-07 1.115E-08 3.544 E-22 2.851E-10 1.658E-03 4.697X Reactor Core 7.680E+04 2.773E-03 1.660E-03 7.197E-06 3.153E-07 1.002E-20 8.063E-09 4.688E-02 1.328E(

1st 3" Above Reactor Core 6.294 E+03 9.926E-05 5.582E-05 2.576E-07 1.128E-08 3.587E-22 2.885E-10 1.678E-03 4.754 1 2nd 3" Above Reactor Core 6.165E+04 3.514 E-04 1.976E-04 9.119E-07 3.995E-08 1.270E-21 1.022E-09 5.941 E-03 1.683E 3rd 3" Above Reactor Core 3.378E+04 5.619E-05 3.160E-05 1.45BE-07 6.387E-09 2.031 E-22 1.633E-10 9.499E-04 2.691 E .

4th 3" Above Reactor Core 6.060E+03 3.337E-06 1.877E-06 8.659E-09 3.793E-10 1.206E-23 9.700E-12 5.641 E-05 1.598E Sth 3" Above Reactor Core 6.060E+03 1.240E-06 6.975E-07 3.218E-09 1.410E-10 4.482E-24 3.605E-12 2.097E-05 5.940E:

6th 3" Above Reactor Core 6.000E+03 4.948E-07 2.783E-07 1.284 E-09 5.625E-11 1.788E-24 1.438E-12 8.365E-06 2.370E 7th 3" Above Reactor Core 6.060E+03 2.075E-07 1.167E-07 5.385E-10 2.359E-11 7.499E-25 6.032E-13 3.508E-06 9.938E '

8th 3" Above Reactor Core 6.060E+03 9.049E-08 5.090E-08 2.348E-10 1.029E-11 3.270E-25 2.631 E-13 1.530E-06 4.334 E !

9th 3" Above Reactor Core 6.060E+03 4.080E-08 2.295E-08 1.059E-10 4.638E 12 1.474 E-25 1.186E-13 6.897E-07 1.954 E '

10th 3" Above Reactor Core 6.060E+03 1.893E-08 1.064E-08 4.911E-11 2.151E-12 6.840E-26 5.502E-14 3.199E-07 9.065E -

i 11th 3" Above Reactor Core 6.0SOE+03 0.006E-09 5.065E-09 2.337E-11 1.024E-12 3.255E-26 2.618E 14 1.522E-07 4.314 E:

l 12th 3" Above Reactor Core 6.060E+03 4.379E-09 2.463E-09 1.136E-11 4.978E-13 1.582E-26 1.273E-14 7.402E-08 2.097Ec

! Total 3.086E+05 3.432E-03 1.930E-03 8.907E-06 3.902E-07 1.240E-20 9.978 E-09 5.802E-02 1.644 EC I

I l

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Document No. G01-1192-003, Rev. O Section 7, Page 14 of 27 ANSTEC APERTURE

, CARD

)NENT ACTIVATION EACTOR SHUTDOWN Also Available on Aperture card T CAN Ni-59 Ni-83 Nb-94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 Total (Cudes) (Cunes) ICud60 (Cudes) (Cudes) (Cudes) ICudgM UM60 (Curies) 6 1.075E-08 1.344E 06 4.427E-11 1.200E-11 4.949E-17 1.207E-17 1.637E-09 3.104 E-06 6 3.245E-08 2.B09E-06 9.249E-11 2.510E-11 1.034 E-16 2.522E-17 3.420E-09 6.484 E-06 6 4.795E-08 5.999E-06 1.975E-10 5.380E-11 2.208E-10 5.387E-17 7.304 E-09 1.385E-05 5 1.051E-07 1.315E-05 4.329E-10 1.179E-10 4.840E-16 1.181E 16 1.601E-08 3.035E-05 5 2.372E-07 2.968E-05 9.773E-10 2.662E-10 1.092E-15 2.665E-16 3.614E-08 6.852E-05 5 5.541 E-07 6.933E 05 2.283E-09 6.217E-10 2.552E-15 6.225E-16 8.441E-08 1.600E-04 4 1.348E-06 1.686E-04 5.551E-09 1.512E-09 6.205E-15 1.514E-15 2.053E-07 3.892E-04 4 3.447E-06 4.313E-04 1.420E-08 3.867E-09 1.587E-14 3.873E-15 5.251E-07 9.956E-04 3 9.473E-06 1.185E-03 3.903E-08 1.063E-08 4.363E-14 1.064E-14 1.443E-06 2.736E-03 3 2.929E-05 3.664E-03 1.206E-07 3.286E-08 1.349E-13 3.290E-14 4.461E-06 8.45BE-03 2 1.042E-04 1.303E-02 4.292E-07 1.169E-07 4.798E-13 1.170E-13 1.587E-05 3.009E-02 2 3.001 E-04 3.755E-02 1.236E-06 3.367E-07 1.382E-12 3.371 E-13 4.571 F.-05 8.667E-02 0 8.487E-03 1.062E+00 3.496E 05 9.522E-06 3.908E-11 9.535E-12 1.33E-03 2.451E+00 2 3.037E-04 3.800E-02 1.251E-06 3.408E-07 1.399E-12 3.413E-13 4.627E-05 8.773E-02 1 1.075E-03 1.345E-01 4.430E-06 1.207E-06 4.952E-12 1.208E-12 1.638E-04 3.106E-01 2 1.719E-04 2.151 E-02 7.084 E-07 1.929E 07 7.918E 13 1.932E-13 2.619E-05 4.966E-02 3 1.021 E-05 1.278E 03 4.207E-08 1.146E-08 4.702E-14 1.147E-14 1.556E-06 2.949E-03 4 3.795E-06 4.748E-04 1.564 E-08 4.258E-09 1.748E 14 4.264E-15 5.781E-07 1.096E-03 4 1.514 E-06 1.895E-04 0.238E-09 1.699E-09 6.973E-15 1.701E-15 2.307E-07 4.374E-04 5 6.350E-07 7.944E-05 2.616E-09 7.124E-10 2.924E-15 7.134E-16 9.673E-08 1.834E-04 5 2.769E-07 3.465E-05 1.141E-09 3.107E-10 1.275E-15 3.111E-16 4.219E-08 7.998E-05 5 1.24BE-07 1.562E-05 5.143E 10 1.4 01 E-10 5.749E-16 1.403E-16 1.902E-08 3.606E-05 5 5.792E-08 7.24BE-06 2.386E-10 6.498E-11 2.667E 16 6.507E-17 8.823E-09 1.673E-05 3 2.756E-08 3.44BE-06 1.135E-10 3.092E-11 1.269E-10 3.096E-17 4.198E-09 7.960E-06 i 1.340E-08 1.677E-06 5.520E 11 1.503E-11 6.171E-17 1.505E-17 2.041E-09 3.870E-06 p 1.050E-03 1.314 E+00 4.327E-05 1.178E-05 4.837E-11 1.180E-11 1.600E-03 3.034 E+00

%0725'0298'- e

l 1

SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION

• FINAL REPORT l

TABLE 74 THE SAXTON FACILITY COMfj TWENTY-FOUR YEARS AFTER i l

LOWER CORE PLATE d Component Volume H-3 C-14 Ar-39 Ca-41 Ca-45 Mn-54 Local!gn Fe-55 Cod (crif3) ICutiet (Cunes) (Curies) (Curles) (Curies) (Curles) (Curies) (Qudi 1

12th 3" Below Reactor Core l 11th 3" Below Reactor Core 10th 3" Below Reactor Core 9th 3" Below Reactor Core 8th 3" Below Reactor Core 7th 3" Below Reactor Core 6th 3" Below Reactor Core 5th 3" Below Reactor Core 4th 3" Below Reactor Core 3rd 3" Below Reactor Core 2nd 3" Below Reactor Core 7.487E+ 03 1.577E-01 1.122E-01 7.752E-05 2.478E-05 7,533E-19 8.741E-08 3.543E+00 4.855s 1st 3" Below Reactor Core Reactor Core 1.120E+03 3.152E-02 4.345E-02 3.098E-05 0.607E-06 2.920E-10 3.690E-08 1.369E+00 1.7983 1st 3" Above Reactor Core 2nd 3" Above Reactor Core 3rd 3" Above Reactor Core 4th 3" Above Reactor Core 5th 3" Above Reactor Core 6th 3" Above Reactor Core 7th 3" Above Reactor Core 8th 3" Above Reactor Core 9th 3" Above Reactor Core 10th 3" Above Reactor Core 11th 3" Above Reactor Core 12th 3" Above Reactor Core Total 8.607E+03 1.892E-01 1.556E-01 1.085E-04 3.438E-05 1.045E-18 1.243E-07 4.912E+00 6.653}-

l I

i l

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Document No. G01-1192-003, Rev. O Section 7, Page 15 of 27 I

ANSTEC i

[ APERTURE CARD ONENT ACTIVATION ,

Also Available on SEACTOR SHUTDOWN Aperture Card l 1

l UIDE BLOCKS  !

Ni-59 l Ni-63 Nb94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 Total '

) (C.udes) . (Cudes) (Citries) (Cudes) {Cmies) (CuicM (Cudes) (Cynes) (Cudes) l G1 6.701E 01 8.072E+01 8.794E-04 1.025E-04 1.896E-07 4.627E-08 6.086E-03 1.338E+02

@i 2.196E-01 2.953E+01 3.442E-04 4.224E-05 1.413E-25 2.539E-07 6.201 E-08 4.328E-04 4.918E+01 I

@1 8.897E-01 1.102E+02 1.224E-03 1.447E-04 1.413E-25 4.435E-07 1.083E-07 6.519E-03 1.829E+02 i

l t

%C72S02 98-

.~ -

SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT TABLE 7.!

THE SAXTON FACILITY COMI-TWENTY-FOUR YEARS AFTER J LOWER SUPPORT SH]

Compenent Volume H-3 C-14 Ar-39 Ca-41 Ca-45 Mn-54 Fe-55 Co-(

Location (cm^3) (Curles) (Curlest [Cydel) [Curtes) (Curies) (CudB1) (Gud. ell (CMEb 12th 3" Below Reactor Core 2.470E+03 7.550E-06 3.749E-06 2.662E-09 8.267E 10 2.516E-23 3.005E-12 1.185E-04 1.673E 11th 3" Below Reactor Core 2.470E+03 1.598E-05 7.926E-06 5.503E-09 1.749E-09 5.322E-23 8.216E-12 2.498E 04 3.524E 10th 3" Below Reactor Core 2.470E+03 3.474E-05 1.722E-05 1.162E-08 3.801E-09 1.156E-22 1.312E-11 5.430E-04 7.622s 9th 3" Below Reactor Core 2.470E+03 7.756E-05 3.843E-05 2.517E-08 8.484 E-09 2.580E-22 2.844E-11 1.214 E-03 1.693E 8th 3" Below Reactor Core 2.470E+03 1.804 E-04 8.931 E-05 5.624 E-08 1.073E-08 5 998E-22 6.350E-11 2.825E-03 3.911]

7th 3" Below Reactor Core 2.256E+ 03 3.952E-04 1.956E-04 1.189E-07 4.322E 08 1.314E-21 1.343E 10 6.202E-03 8.5183 6th 3" Below Reactor Core 1.186E+03 5.249E-04 2.596E-04 1.515E-07 5.742E 08 1.745E-21 1.710E-10 8.216E-03 1.1243 5th 3" Below Reactor Core 1.186E+03 1.484E-03 7.339E-04 3.921 E-07 1.625E-07 4.932E-21 4.422E-10 2.310E-02 3.139]

4th 3" Below Reactor Core 1.186 E+03 3.849E-03 1.926E-03 1.123E-06 4.259E 07 1.294E 20 1.266E-09 6.080E-02 8.320]

3rd 3" Below Reactor Core 9.880E+02 6.732E-03 3.552E-03 3.007E-06 7.815E-07 2.382E-20 3.396E-09 1.118 E-01 1.625E 2nd 3" Below Reactor Core 1st 3" Below Reactor Core Reactor Core 1st 3" Above Reactor Core 2nd 3" Above Reactor Core 3rd 3" Above Reactor Core 4th 3" Above Reactor Core 5th 3" Above Reactor Core 6th 3" Above Reactor Core 7th 3" Above Reactor Core 8th 3" Above Reactor Core 9th 3" Above Reactor Core 10th 3" Above Reactor Core 11th 3" Above Reactor Core 12th 3" Above Reactor Core Total 1.915 E+04 1.330E-02 6.824E-03 4.894 E-06 1.505E-06 4.580E-20 5.524 E-09 2.152E-01 3.037E-a- . . ,

Document No. G01-1192-003, Rev. O Section 7, Page 17 of 27 ANSTEC APERTURE

>NENT ACTIVATION ,

CARD BACTOR SH'UTDOWN Also Available on Aperture Card DUD TUBES Ni-59 Ni-63 Nb-94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 Total (Curles) (Cudes) (Curies) (Curles) (Cydel) (Gudas) (Cudes) (Cydes) (Cudes) i 2.406E-05 2.761 E-03 2.989E-08 3.683E-09 5.215E-16 1.602E-16 2.410E-06 4.590E-03 i G.092E-05 5.840E-03 6.275E-08 7.624E-09 2.882E-15 7.117E-16 5.087E-06 9.694 E-03 l 1.107E-04 1.269E-02 1.351 E-07 1.012E-08

' 1.342E-14 3.282E-15 1.102E 05 2.103E-02 2.472E-04 2.833E-02 2.987E-07 3.500E-08 6.557E 14 1.601E-14 2.443E-05 4.686E-02 G.701 E-04 6.587E-02 6.857E-07 7.834E-08 3.462E-13 8.448E-14 5.610E-05 1.087E-01 1.200E-03 1.443E-01 1.486E-06 1.660E-07 1.782E-12 4.350E-13 1.196E-04 2.376E-01 1.673E-03 1.916E-01 1.948E-06 2.069E-07 5.840E-12 1.42SE-12 1.487E-04 3.148E-01 4.730E-03 5.420E-01 5.374 E-06 5.319E-07 4.443E-11 1.085E-11 3.496E-04 8.864E-01 9.333E-02 1.419E+00 1.444E-05 1.512E-06 3.208E-10 7.825E-11 6.041 E-04 2.330E+00

) 3.326E-02 2.593E+00 3.007E-05 3.965E-06 1.649E-09 4.022E-10 6.400E-04 4.363E+00 l

4.327E-02 5.005E+00 5.452E-05 6.523E-06 2.022E-09 4.933E-10 1.961 E-03 8.323E+00 i

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SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION '

FINAL REPORT 1 TABLE 7.G THE SAXTON FACILITY COM TWENTY-FOUR YEARS AFTER '

LOWER CORE 1)

Component Volume H-3 C-14 Ar-39 Ca-41 Ca45 Mn-54 Fe-55 Location (cm^3) Co-6j (Curies) (Curles) (Cyde0 (Curies) (CudejQ (Curies) (Curies) (Cudet 12th 3" Below Reactor Core 11th 3" Below Reactor Core 10th 3" Below Reactor Core 9th 3" Below Reactor Core 8th 3" Below Reactor Core 7th 3" Below Reactor Core 6th 3" Below Reactor Core 5th 3" Below Reactor Core 4th 3" Below Reactor Core 3rd 3* Below Reactor Core 9.442E+03 3.847E-02 2.030E-02 1.718E-05 4.465E-06 1.301 E-19 1.941E-08 6.389E-01 9.283E+

2nd 3" Below Reactor Core 1.888E+04 2.379E-01 1.692E-01 1.170E-04 3.738E-05 1.137E 18 1.319E-07 5.346E+00 7.325E+

1st 3" Below Reactor Core Reactor Core 1st 3" Above Reactor Core 2nd 3" Above Reactor Core 3rd 3" Above Reactor Core 4th 3" Above Reactor Core 5th 3" Above Reactor Core 6th 3" Above Reactor Core 7th 3" Above Reactor Core 8th 3" Above Reactor Core 9th 3" Above Reactor Core 10th 3" Above Reactor Core 11th 3" Above Reactor Core 12th 3" Above Reactor Core Total 2.833E+04 2.764 E-01 1.895E-01 1.341 E-04 4.185E-05 1.273E-18 1.513E-07 5.985E+00 0.253E+'

e w

l Document No. G01-1192-003, Rev. 0 Section 7, Page 16 of 27 ANSTEC APERTURE NENT ACTIVATION CARD 3 ACTOR SHUTDOWN Also Availabi. on Aperture Card ATE NI-59 Ni63 Nb-94 Tc 99 Sn-119m Sb-125 Te-125m Eu-152 Total (W (CMdte (Cunes) (CV.ticM (Curies) (Cudes) (Curies) (Cunes) (Curies) i 1273E-01 1.482E+01 1.718E-04 2.266E-05 9.421E-09 2.298E-09 3.657E-03 2.493E+01

' 1.011Eo00 1.218E+02 1.327E-03 1.546E-04 l 2.860E-07 6.981E-08 9.183E-03 2.018E+02 '

l l 1

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1.938Eo00 1,366E+02 1.499E-03 1.772E-04 2.954E-07 7.211 E-08 1.284 E-02 2.267E+02 l

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SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT TABLE 7.1 THE SAXTON FACILITY COMP TWENTY-FOUR YEARS AFTER I LOWER SUPPORT '

Component Volume H-3 C-14 Ar-39 Ca-41 Ca-45 Mn-54 Fe-55 Co-6t Locahon (caf3) (Curi.es) (Qufies) (Curies) (Curies) (Cunes) (Curies) (Curies) [Qude 12th 3" Below Reactor Core 3.089E+02 9.444 E-07 4.689E-07 3.330E-10 1.034 E-10 3.147E-24 3.759E-13 1.482E-05 2.092E-11th 3" Bebw Heactor Core 3.089E+02 1.999E-06 9.914 E-07 6.884E 10 2.188E-10 6.657E-24 7.775E-13 3.12SE-05 4.408 E.

10th 3" Below Reactor Core 3.089E+ 02 4.345E-06 2.154 E-06 1.4 54 E-09 4.755E-10 1.446E-23 1.641 E-12 6.792E-05 9.534 E-9th 3" Below Reactor Core 3.089E+02 9.701 E-06 4.807E-06 3.149 E-09 1.061 E-09 3.228E-23 3.557E-12 1.518 E-04 2.118 E.

8th 3* Below Reactor Core 3.089E+02 2.256E-05 1.117E-05 7.034 E-09 2.468E-09 7.502E-23 7.942E-12 3.534 E-04 4.892E-7th 3* Below Reactor Core 3.089E+02 5.413E-05 2.678E-05 1.629 E-08 5.920E-09 1.799E-22 1.839E-11 8.494 E-04 1.167E-6th 3" Below Reactor Core 3.089E+02 1.368E-04 6.764 E-05 3.946 E-08 1.496E-08 4.545E-22 4.455E-11 2.141 E-03 2.929 E-5th 3" Below Reactor Core 3.089E+ 02 3.865E-04 1.912E-04 1.022E 07 4.234 E-08 1.285E 21 1.152E-10 6.042E-03 9.177E.

4th 3" Below Reactor Core 3.089E+02 1.003E-03 5.019E-04 2.927E-07 1.110E-07 3.372E-21 3.297E-10 1.584E 02 2.168E-3rd 3" Below Reactor Core 2.574 E+02 1.754 E-03 9.255 E-04 7.835E-07 2.036E-07 6.206E-21 8.849E 10 2.913E-02 4.233E 2nd 3" Below Reactor Core 1st 3" Below Reactor Core Reactor Core 1st 3* Above Reactor Core 2nd 3" Above Reactor Core 3rd 3" Above Reactor Core 4th 3" Above Reactor Core 5th 3" Above Reactor Core 6th 3" Above Reactor Core 7th 3" Above Reactor Core 8th 3" Above Reactor Core 9th 3" Above Reactor Core 10th 3" Above Reactor Core 11th 3" Above Reactor Core 12th 3" Above Reactor Core Total 3.037E+03 3.374 E-03 1.733E-03 1.247 E-06 3.821E 07 1.163E-20 1.407E-09 5.462E-02 7.714E :

m -o min

Document No. G01-1192-003, Rev. O Section 7, Page 18 of 27 ANSTEC APERTURE

)NENT ACTIVATION EACTOR SHUTDOWN Also Available on Aperture Card

'IE RODS Na39 Ni-63 Nb-94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 Total (Curles) (Gmief (Cunes) (Curies) (Cuits) (Gmits) 1021e3) [CMdes) (Curies) 4 3.010E-06 3.453E-04 3.739E-09 4.007E-10 6.523E-17 2.003E-17 3.015E-07 5.741E-04 4 6.370E-06 7.304E 04 7.849E-09 9.537E-10 3.604E-16 8.903E-17 6.363E-07 1.213E-03 8 1.385E-05 1.587E-03 1.690E-08 2.017E-09 1.679E-15 4.105E-16 1.378E-06 2.631E-03 3 3.092E-05 3.544E 03 3.736E-08 4.378E-09 8.201E 15 2.003E-15 3.056E-06 5.862E-03 5 7.193E-05 8.239E 03 8.577E-08 9.798E-09 4.331E 14 1.057E-14 7.017E-06 1.360E-02 2 9.720 E-04 1.976E-02 2.035E-07 2.273E-08 2.441E 13 5.957E 14 1.638E-05 3.254 E-02 2 4.360E-04 4.992E-02 5.076E-07 5.390E-08 1.522E-12 3.713E-13 3.874E-05 8.203E.02 2 1.232E-03 1.412E-01 1.400E-06 1.386E-07 1.158E-11 2.826E 12 9.108E-05 2.309E-01 1 3.214 E-03 3.690E-01 3.761 E-06 3.939E 07 8.357E-11 2.039E-11 1.574E-04 6.071E-01 1 5.799E-03 6.750E-01 7.834 E-06 1.033E-06 4.296E-10 1.048E 10 1.667E-04 1.137E+00 1.098E-02 1.271 E+00 1.386E-05 1.660E-06 5.265E 10 1.285E-10 4.827E-04 2.113E+00 9(,07250292 <

m 4 SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT TABLE 7.f THE SAXTON FACILITY COMd TWENTY-FOUR YEARS AFTER BALANCE OF LOWER SU@

Component -

Volume H-3 C-14 Ar-39 Ca-41 Ca45 Mn-54 Fe-55 Co' LQCatt0D (cm^3) (Curies) (Cunes) (Curies) (Curies) (QUIlell (Curies) (Curies) [ Cud-12th 3" Below Reactor Core lith 3" Below Reactor Core 4.003E+04 1.224 E-04 0.076E-05 4.315E-08 1.340E-08 4.078E-22 4.872E 11 1.920E-03 2.712I i

10th 3" Below Reactor Core 9th 3" Below Reactor Core 8th 3" Below Reactor Core 7th 3" Below Reactor Core 6th 3" Below Reactor Core 5th 3" Below Reactor Core 4th 3" Below Reactor Core 3rd 3" Below Reactor Core 2nd 3" Below Reactor Core 1st 3" Below Reactor Core l

Reactor Core 1st 3" Above Reactor Core .

2nd 3" Above Reactor Core 1 3rd 3" Above Reactor Core 4th 3" Above Reactor Core 5th 3" Above Reactor Core j 6th 3" Above Reactor Core 7th 3" Above Reactor Core 8th 3" Above Reactor Core 9th 3" Above Reactor Core 10th 3* Above Reactor Core 11th 3" Above Reactor Core 12th 3* Above Reactor Core 1

Total 4.003 E+04 1.224 E-04 6.076E-05 4.315E-08 1.340E-08 4.078E-22 4.872E-11 1.920E-03 2.7

<n +:,

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Docuntent No. G01-1192-003, Rev. 0 l Section 7, Page 19 of 27 l

ANSTEC APERTURE

$NENT ACTIVATION EACTOR SHUTDOWN Also Available on Aperture Card

? ORT ASSEMBLY I

l Ni-59 Ni-63 Nb 94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 Total j (Curles) (Curies) (Cydet) (Curies) (Curles) (Cudes) (Cy. del) [Cules) (Curica) 8 3.901 E-04 4.475E-02 4.845E-07 5.971 E-08 8.453E 15 2.596E 15 3.907E-05 7.440E-02 h

i t

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, 3.901 E-04 4.475E-02 4.845E-07 5.971 E-08 8.453E-15 2.596E-15 3.907E-05 7.440E-02 l

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SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT TABLE 7.B THE SAXTON FACILITY COMd TWENTY-FOUR YEARS AFTER (

LOWER CORE B  !

1 i

l Component

]

Volume H-3 C-14 Ar-39 Ca-41 Ca-45 Mn 54 Fe-55 Coe LQ.c all0B fem ^3) (Curies) 12th 3" Below Reactor Core 11th 3" Below Reactor Core i 10th 3* Below Reactor Core l Oth 3* Below Reactor Core 1 8th 3* Below Reactor Core 7th 3" Below Reactor Core 6th 3" Below Reactor Core 5th 3" Below Reactor Core 4th 3" Below Reactor Core 3rd 3* Below Reactor Core 2nd 3" Below Reactor Core 1st 3" Below Reactor Core Rescior Core 1st 3" Above Reactor Core 5.077E+03 2.416E-02 3.318E-02 2.436E-05 7.333E 06 2.230E-19 2.914 E-08 1.043E+00 1.381E(

2nd 3* Above Reactor Core 0.467E+03 1.663E-02 1.167E-02 9.514 E-06 2.572E-06 7.830E 20 1.081 E-08 3.681 E-01 5.212E(

3rd 3* Above Reactor Core 4.894E+03 6.575E-03 3.544E-03 2.058E-06 7.842E-07 2.383E-20 2.307E-09 1.120E-01 1.519E(

4th 3" Above Reactor Core 4.894E+03 2.688E-03 1.348E-03 6.736E-07 2.986E-07 9.063E 21 7.574E 10 4.272E-02 5.700E.

5th 3" Above Reactor Core 4.894E+03 9.181 E-04 4.546E-04 2.493E 07 1.006E-07 3.056E-21 2.811E-10 1.436E-02 1.951E 6th 3" Above Reactor Core 4.894E+03 3.413E-04 1.689E-04 1.003E-07 3.734E-08 1.135E-21 1.132E-10 6.330E-03 7.329E:

7th 3" Above Reactor Core 4.894E+03 1.401 E-04 6.935E-05 4.255E-08 1.533E-08 4.658E-22 4.806E-11 2.197E-03 3.025E 8th 3" Above Reactor Core 4.894E+03 5.998E-05 2.971 E-05 1.880E-08 6.560E-09 1.995E-22 2.123E-11 9.380E-04 1.302E 9th 3" Above Reactor Core 4.894E+03 2.635E-05 1.300E-05 8.587E-09 2.883E-09 8.767E-23 9.696E-12 4.127E-04 5.756E 10th 3" Above Reactor Core 4.894E+03 1.204 E-05 5.968E-06 4.037E-09 1.317E-09 4.007E-23 4.557E-12 1.886E-04 2.642E ,

11th 3" Above Reactor Core 4.894E+03 5.631 E-06 2.793E-06 1.943E-09 6.163E 10 1.875E-23 2.195E-12 8.825E-05 1.242E:

12th 3" Above Reactor Core 4.894E+03 2.694E-06 1.337E-06 9.550E-10 2.949E 10 8.976E-24 1.078E-12 4.231 E-05 5.974 E .

Total 6.048E+04 5.if SE-02 5.049E-02 3.703E-05 1.115E 05 3.392E-19 4.349E-08 1.589 E+00 2.143EC u ~.a

l Document No. G01-1192-003, Rev. 0 i Section 7, Page 20 of 27  ;

i l

ANSTEC l APERTURE  !

CARD  !

)NENT ACTIVATION Also Available on I BACTOR SHUTDOWN APerture Card i RREL i i

l Ni-59 Ni-63 Nb-94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 Total  !

l l

l l

1 1

1.674 E-01 2.254 E+01 2.659E-04 3.328E-05 1.128E-25 1.971 E-07 4.809E-08 3.343E-04 3.761E+01 6.957E-03 8.387E+00 9.711 E-05 1.256E-05 1.996E-08 4.872E-09 7.554 E-04 1.407E+01 i 2.231E-03 2.595E+00 2.644 E-05 2.748E-06 1.969E-09 4.807E-10 4.162E-04 4.259E+00 J 8.644E-03 9.936E-01 9.680E-06 9.148E-07 2.637E-10 6.441E-11 3.632E-04 1.610E+00 2.927E-03 3.356E-01 3.352E-06 3.375E-07 )

3.152E-11 7.703E-12 2.088E-04 5.495E-01 1.088E-03 1.246E-01 1.274 E-06 1.367E-07 i 4.551 E-12 1.110E-12 9.484E-05 2.049E-01 1

, 4.466E-04 5.116E-02 5.283E-07 5.936E-08 7.825E-13 1.909E-13 4.201E-05 8.431 E-02 .

1.913E-04 2.190E-02 2.284 E-07 2.618E-08 1.459E-13 3.561 E-14 1.857E-05 3.616E-02 l 8.390E-00 9.624 E-03 1.016E-07 1.193E-08 2.878E-14 7.039E-15 8.284E-06 1.592E-02 4 l

3.037E-05 4.399E-03 4.685E-08 5.600E-09 6.152E-15 1.501E-15 3.812E-06 7.290E-03 i 1.794E-05 2.057E-03 2.213E-08 2.692E-09 1.360E-15 3.363E-16 1.792E-06 3.416E-03

)i 8.586E-06 9.848E-04 1.06BE-08 1.321 E-09 3.220E-16 8.128E 17 8.599E-07 1.638E-03 l 2.727E-01 3.506E+01 4.047E-04 5.008E-05 1.128E-25 2.193E-07 5.352E-08 2.248E-03 5.846E+01 l

)

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SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT TABLE 7.1 THE SAXTON FACILITY COMF TWENTY-FOUR YEARS AFTER 1 UPPER CORE I Component Volume H3 C-14 Ar-39 Ca-41 Ca45 Mn-54 Fe-55 Co-6 Location (cm^31 ICudcM (Curies) (Curies) (Curies) (C.udes ICunes) (Cunes) IC.uda 12th 3" Below Reactor Core 11th 3" Below Reactor Core 10th 3" Below Reactor Core 9th 3" Below Reactor Core 8th 3" Below Reactor Core 7th 3" Below Reactor Core 6th 3" Below Reactor Core 5th 3" Below Reactor Core 4th 3" Below Reactor Core 3rd 3" Below Reactor Core 2nd 3" Below Reactor Core 1st 3" Below Reactor Core Reactor Core 1st 3" Above Reactor Core 2nd 3" Above Reactor Core 1.499 E+ 04 1.951 E-01 1.369E-01 1.116E-04 3.016E-05 9.183E 19 1.267E-07 4.317E+00 6.112E '

3rd 3" Above Reactor Core 4th 3" Above Reactor Core 5013" Above Reactor Core 6th 3" Above Reactor Core 7th 3" Above Reactor Core 8th 3" Above Reactor Core 9th 3" Above Reactor Core 10th 3" Above Reactor Core 11th 3" Above Reactor Core 12th 3" Above Reactor Core Total 1.499E+04 1.951 E-01 1.369E-01 1.116E-04 3.016E-05 9.183E-19 1.267E-07 4.317E+00 6.112E<

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Document No. G01-1192-003, Rev. 0 l Section 7, Page 21 of 27 l

)

ANSTEC 1

APERTURE  ;

CARD ONENT ACTIVATION I LEACTOR SHUTDOWN Ah Avaliable on Aperture Card i LATE l

4 E E E E iEST Sie',53 I!&"' ("C',52 ,, g i

i i

91 0.159E-01 9.836E+01 1.139E-03 1.473E-04 2.341 E-07 5.714E-08 8.860E-03 1.650E+02 f

91 0.159E-01 9.836E+01 1.139E-03 1.473 E-04 2.341E-07 5.714 E-08 8.860E-03 1.650E+02 l

%O 72fo298'-

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j SAXTON ACTIVATION ANALYSIS l.

t PROPRIETARY INFORMATION l FINAL REPORT l

TABLE 7.[

THE SAXTON FACILITY COMd i

'1WENTY-FOUR YEARS AFTER {

l 1

UPPER CORE B Component Volume H3 C-14 Ar-39 Ca-41 Ca-45 Mn-54 Fe-55 Co-6 LocaligD (cm^3) (Cunes) (Curies) (Cuhes) (Cunes) (Curies) (Cunes) [Cgies) (Gudi 12th 3" Below Reactor Core 11th 3" Below Reactor Core 10th 3" Below Reactor Core 9th 3" Below Reactor Core 8th 3" Below Reactor Core 7th 3" Below Reactor Core 6th 3" Below Reactor Core 5th 3" Below Reactor Core 4th 3* Below Reactor Core 3rd 3" Below Reactor Core i 2nd 3" Below Reactor Core 1st 3" Below Reactor Core Reactor Core 1st 3" Above Reactor Core I 2nd 3" Above Reactor Core 6.698E+02 5.341 E-03 3.74 8 E-03 3.055E-06 8.258 E-07 2.514 E-20 3.470E-09 1.182E-01 1,674 E .

3rd 3" Above Reactor Core 4.262E+03 1.775E-02 9.570E-03 5.555E-06 2.117E-06 6.433E-20 6.230E 09 3.023E-01 4.101E:

4th 3" Above Reactor Core 2.309E+03 3 932E-03 1.972E-03 9.855E-07 4.368E-07 1.326E-20 1.108E-09 6.250E-02 8.339s 5th 3* Above Reactor Core 2.309E+03 1.343E-03 6.651 E-04 3.647E-07 1.472E-07 4.471 E-21 4.113E 10 2.101 E-02 2.854 1 6th 3" Above Reactor Core 2.309E+03 1.994 E-04 2.470E-04 1.467E-07 5.463E-08 1.660E-21 1.656E-10 7.807E-03 1.0723-7th 3" Above Reactor Core 2.309E+03 2.050E-04 1.015E-04 6.225E-08 2.242E-08 6.815E-22 7.031 E-11 3.215E-03 4.4261 Sth 3" Above Reactor Core 2.309E+03 8.774E-05 4.346E-05 2.751 E-08 9.598E-09 2.918E-22 3.106E-11 1.372E-03 1.9043 9th 3" Above Reactor Core 2.309E+03 3.855E-05 1.910E-05 1.256E-08 4.218E-09 1.283E-22 1.419E-11 6.037E-04 8.421@.

10th 3" Above Reactor Core 2.309E+03 1.761 E-05 8.731 E-06 5.906E 09 1.927E-09 5.882E-23 6.668E-12 2.75SE-04 3.866E 11th 3" Above Reactor Core 2.309E+03 8.238E-06 4.087E-06 2.843E-09 9.016E-10 2.744 E-23 3.211 E-12 1.291 E-04 1.817E 12th 3" Above Reactor Core 2.309E+03 3.942E-06 1.957E-06 1.397E-09 4.315E-10 1.313E 23 1.578E-12 6.190E-05 8.740E!

Total 2.572E+04 2.923E-02 1.638E-02 1.022E-05 3.621 E-06 1.101 E-19 1.151 E-08 5.174 E-01 7.080E<

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Document No. G01-1192-003, Rev. O Section 7, Page 22 of 27

' ANSTEC APERTURE ONENT ACTIVATION CARD LEACTOR SHUTDOWN Also Airallable a Aperture Card LRREL Ni-59 Ni-63 Nb-94 Tc-99 Sn-119m St>.125 Te-125m Eu 152 Total

) {Cydes) (Cudes) (Qudca) (Cudest (Cudes) (Cudes) (Cydes) (Cunes) (Curies)

I I

10 2.234E-02 2.693E+00 3.118E-05 4.033E-06 6.410E 09 1.565E-09 2.426E-04 4.517E+00 10 6.022E-02 7.007E+00 7,139E-05 7.418E-06 5.31CE 09 1.298E-09 1.124E-03 1.150E+01 t1 1.265E-02 1.454 E+00 1.416E-05 1.338E 06 3.859E 10 9.424E-11 5.314E-04 2.369E+00 t1 4.282E-03 4.909E-Oi 4.904 E-06 4.937E-07 4.612E 11 1.127E-11 3.054E-04 8.039E-01 11 1.591E-03 1.823E-01 1.883E 06 2.000E-07 6.658E-12 1.624E-12 1.388E-04 2.998E-01 12 6.534E 04 7.485E-02 7.720E-07 8.684E-08 1.145E-12 2.793E-13 6.146E-05 1.233E-01 12 2.798E-04 3.204E 02 3.341E-07 3.830E-08 2.134E 13 5.209E-14 2.717E-05 5.290E-02 3 1.229E-04 1.408E-02 1.486E-07 1.746E-08 4.210E-14 1.030E-14 1.212E-05 2.330E-02 3 5.613E-05 6.435E-03 6.854E-08 8.192E 09 9.000E-15 2.195E-15 5.578E-06 1.067E-02 13 2.635E-05 3.010E-03 3.237E-08 3.938E-09 1.990E-15 4.920E-16 2.621E 06 4.998E-03 4 1.256E-05 1.441E-03 1.563E-08 1.933E-09 4.711E-16 1.189E-16 1.258E-06 2.396E-03 lo 9.022E-01 1.196E+01 1.249E-04 1.364E-05 1.216E-08 2.970E-09 2.452E-03 1.971 E+01 1

9(sO22CO298-i J

SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT TABLE 7.J l

THE SAXTON FACILITY COMD l TWENTY-FOUR YEARS AFTER I 1

l BALANCE OF LOWER CORE !

l l l

Component Volume H-3 C-14 Ar-39 Ca-41 Ca-45 Mn-54 Fe-55 C t

Location (cm^j) (Cunes) (Curies) (Curles) (Cur.ies) (Curies) (Curles) (Curies) [ Cud.

12th 3" Below Reactor Core 11th 3" Below Reactor Core 10th 3" Below Reactor Core 9th 3* Below Reactor Core 8th 3" Below Reactor Core 7th 3" Below Reactor Core 6th 3" Below Reactor Core 5th 3" Below Reactor Core 4th 3" Below Reactor Core 3rd 3" Below Reactor Core 2nd 3" Below Reactor Core l 1st 3" Below Reactor Core Reactor Core 1st 3" Above Reactor Core 2nd 3" Above Reactor Core 3rd 3" Above Reactor Core l

4th 3" Above Reactor Core Sth 3" Above Reactor Core 6th 3* Above Reactor Core 7th 3" Above Reactor Core 8th 3" Above Reactor Core 9th 3" Above Reactor Core

10th 3" Above Reactor Core l 11th 3" Ab>ve Reactor Core 12th 3" Above Reactor Core 1.034 E+05 4.265E-04 2.117E-04 1.512E-07 4.669E 08 1.421 E-21 1.707E 10 6.698E-03 9.457@

Total 1.034 E+05 4.265E-04 2.117E-04 1.512E-07 4.669E-08 1.421 E-21 1.707E.10 6.698E-03 9.4573 l

l l

w emursar

Document No. G01-1192-003, Rev. O Section 7, Page 23 of 27 i

5 ANSTEC .

APERTURE ONENT ACTIVATION CARD ,

tEACTOR SHUTDOWN Awo Available on .

Apature Card 1 1ARREL ASSEMBLY I Ni-59 Ni-63 Nb-94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 Total 0 !fdKitM (C!Ldful) (C.uded (Cunes) (Curles) (Curles) (Cgig.s) (Curies)

ICdEits) l 1

l 32 1.359E-03 1.559E-01 1.691 E-06 2.092E-07 5.097E-14 1.287E-14 1.381 E-04 2.593E-01 32 1.359E 03 1.559E-01 1.691 E-06 2.092E-07 5.097E-14 1.287E-14 1.361 E-04 2.593E-01 9607250298-

c l SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL RFsPORT TABLE 7.

THE SAXTON FACILITY COM]

TWENTY-FOUR YEARS AFTER BALANCE OF UPPER CORE Component Volume H-3 C-14 Ar-39 Ca-41 Ca45 Mn-54 Fe-55 Co-1 Location (cma s) (Curies) (Curies) (Gwies) (Curles) (Curies) (Curies) (Curies) (Gud 12th 3" Below Reactor Core lith 3" Below Reactor Core 10th 3" Below Reactor Core 9th 3" Below Reactor Core 8th 3" Below Reactor Core 7th 3" Below Reactor Core 6th 3" Below Reactor Core 5th 3" Below Reactor Core 4th 3" Below Reactor Core 3rd 3" Below Reactor Core 2nd 3" Below Reactor Core 1st 3" Below Reactor Core Reactor Core 1st 3" Above Reactor Core 2nd 3" Above Reactor Core 3rd 3" Above Reactor Core 4th 3" Above Reactor Core 5th 3" Above Reactor Core 6th 3" Above Reactor Core 7th 3" Above Reactor Core 8th 3" Above Reactor Core 9th 3" Above Reactor Core 10th 3" Above Reactor Core 11th 3' Above Reactor Core 12th 3" Above Reactor Core 4.717E+04 1.947E-04 9.663E 05 6.900E-08 2.131 E-08 6.485E 22 7.791 E-11 3.057E-03 4.316E Total 4.717E+04 1.947E-04 9.663E-05 6.900E-08 2.131 E-08 6.485E-22 7.791E-11 3.057E-03 4.316E v ~s

~~.

Document No. G01-1192-003, Rev. O Section 7, Page 24 of 27 I

ANSTEC l APERTURE l r CARD i

)NENT ACTIVATION EACTOR SHUTDOWN Also Available on ,

Aperture Card l l

MtREL ASSEMBLY  !

i Ni-09 Ni-63 Nb-94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 Total l

/fddlud (Cudes) (Cudes) (Cudes) (Cudes) (Cudesi LC.utins) (Cudes) (Cudes) l l

l i l f

1 \l I

I I

6.203E-04 7.115E-02 7,720E-07 9.547E-08 2.326E 14 5.872E-15 C.213E-05 1.183E-01 6.203E-04 7.115E-02 7.720E-07 9.547E-08 2.326E-14 5.872E 15 6.213E-05 1.183E-01 i

i %C7250298-

l l

I SAXTON ACTIVATION ANALYSIS PROPRIETARY INFORMATION FINAL REPORT I i

l l

l l l

l TABLE 7.( j THE SAXTON FACILITY COMD i

TWENTY-FOUR YEARS AFTER l

CURIE CONTENTS FOR REACTOR!

H-3 C-14 Ar39 Ca.41 Ca.45 Mn-54 Fe-55 Co@

Comoonent (Curles) (Cut!ra) (Curles) (Curfes) (Curies 1 (Curles) (Curfes) LC_yrth Core Baffle 6.555E-01 4.997E-01 l

8.593E-04 1.075E-04 3.288E-18 1.077E-06 1.560E+01 2.7573 Thermal Shield 1.570E-01 8.690E-02 3.56SE-04 1.777E-05 5.615E-19 4.013E-07 2.629E+00 6.9483 Vessel Clad 5.809E-03 3.042E-03 7.324 E-06 6.465 E-07 1.995E-20 8.215E-09 9.429E-02 1.8971 Vessel Wall 9.756E 02 4.786E-03 6.961 E-04 3.636E-06 1.145E-19 2.860E-07 1.065E+00 2.4021' Vesselinsulation Can 8.823E-04 5.689E-04 4.713E-06 1.051 E-07 3.534E 21 5.279E-09 1.618E-02 7.0025 Vessel Support Can 3.432E-03 1.930E-03 8.907E-06 3.902E-07 1.240E 20 9.978E-09 5.802E-02 1.6443 Lower Core Guide Blocks 1.892E-01 1.556E-01 1.085E-04 3.438 E-05 1.045E-18 1.243E-07 4.912E+00 0.653E Lower Core Plate 2.764 E-01 1.895E-01 1.341 E-04 4.185E-05 1.273E-18 1.513E-07 5.985E+00 8.2533 Lower Support Shroud Tubes Lower Support Tie Rods 1.330E-02 6.824E-03 4.894E-06 1.505E-06 4.580E-20 5.524E 09 2.152E-01 3.037 @

3.374 E-03 1.733E 03 1.247E-06 3.821 E-07 1.163E-20 1.407E-09 5.462E-02 7.7141 Balance Lower Support Assembly 1.224 E-04 6.0766-05 4.315E-08 1.340E-08 4.078E-22 4.872E-11 1.920E 03 2.7121 Lower Core Barrel 5.156E-02 5.049E-02 3.703E-05 1.115E-05 3.392E 19 4.349E-08 1.589E+00 2.143E:

Upper Core Plate 1.951 E-01 1.369E-01 1.116E-04 3.016E-05 9.183E-19 1.267E-07 4.317E+00 0.112E.

Upper Core Barre! 2.923E-02 1.638E-02 1.022E-05 3.621 E-06 1.101E 19 1.151E-08 5.174E-01 7.080E' Balance Lower Core Barrel Assembly 4.265E-04 2.117E-04 1.512E-07 4.669E-08 1.421 E-21 1.707E 10 6.698 E-03 9.4571 Balance Upper Core Barrel Assembly 1.947E-04 9.663E-05 6.900E-08 2.131 E-08 6.485E-22 7.791E-11 3.057E-03 4.3163-Totals 1.679E+00 1.155E+00 2.341 E-03 2.532E-04 7.745E-18 2.252E-06 3.706E+0i 5.945E '

l l

1

Document No. G01-1192-003, Rev. O Section 7, Page 25 of 27 ANSTEC APERTURE CARD PNENT ACTIVATION

'EACTOR SHUTDOWN ^$,$",$$8fn, rESSEL AND INTERNALS Ni-59 NI-63 Nb-94 Tc-99 Sn 119m Sb-125 Te 125m Eu 152 Total (GUIlM) (GMdul (GMdM) (GMdM} (GMdM) (GUIlul LGEIM} (GDdM} (GMdQ1) f 2.752Eo00 3.474E+02 6.130E-03 1.204 E-03 3.781 E-24 2.804 E-06 6.842E-07 2.928E-02 6.426E+02 1 4.831 E-01 5.975 E+ 01 1.788E-03 4.565E-04 2.572E-C8 6.277E-09 6.250E-02 1.327E+02 0 1.819E-02 2.165E+00 4.338E-05 9.461 E-06 3.507E-10 8.560E-11 1.845E-03 4.185E+00 0 8.313E-03 1.118 E+00 1.665E-04 4.859E-08 1.964 E-09 4.795E-10 3.940E-02 4.737E+00

) 2.570E-03 3.574E-01 2.047E-05 6.118E-06 2.196E-11 5.363E-12 5.952E-04 1.078E+00 0

' 1.050E-02 1.314 E+00 4.327E-05 1.178E-05 4.837E-11 1.180E-11 1.600E-03 3.034 E+00 8.89FE-01 1.102E+02 1.224 E-03 1.447E-04 1.413E-25 4.435E-07 1.083E-07 6.519E-03 1.829E+02 1 9.138Eo00 1.366E+02 1.499E-03 1.772E-04 2.954E-07 7.211 E-08 1.284 E-02 2.267E+02 p 4.327E-02 5.005E+00 5.452E-05 6.523E-06 2.022E-09 4.933E-10 1.961E-03 8.323E+00 I 9.098E-02 1.271 E+00 1.386E-05 1.660E-06 5.265E-10 1.285E-10 4.827E-04 2.113E+00 2 3.901 E-04 4.475E-02 4.645E-07 5.971 E-08 8.453E-15 2.596E-15 3.907E-05 7.440E-02 p 2.727E-01 3.506E+ 01 4.047E-04 5.008E-05 1.128E-25 2.193E-07 5.352E-08 2.248E-03 5.846E+01 1 8.159E-01 9 836E+01 1.139E-03 1.473E-04 2.341 E-07 5.714E-08 8.860E-03 1.650E+02 0 1.022E-01 1.196E+01 1.249E-04 1.364E-05 1.216E-03 2.970E-09 2.452E-03 1.971 E+01

% 1.359E-03 1.559E-01 1.691 E-06 2.092E-07 5.097E-14 1.287E-14 1.361 E-04 2.593E-01

! 6.203E-04 7.115E-02 7.720E-07 9.547E-08 2.326E-14 5.872E-15 6.213E-05 1.183 E-01 2 6.550E000 8.109E+02 1.265E-02 2.230E-03 4.035E-24 4.039E-03 9.8 f ~ 17 1.700E-01 1.452E+03 i

%c250298- -

,~ -

SAXTON ACTIVATION ANALYSIS PItOPIIIETAItY INFOItMATION FINAL REPOItT TABLE 7.

THE SAXTON FACILITY COM]

TWENTY-FOUR YEARS AFTER SPECIFIC ACTIVITIES FOR EX-VESSEL STAINLESS STEEL, LEAD Al H3 C-14 Ar-39 Ca-41 Ca-45 Mn-54 Fe-55 Comoonent . (Cmial) (Curies) (Gmics) (Gwies) (Curies) (Curies) (Curies) ist 6" Water Shield Stainless Steel 6.867E-09 3.392E-09 2.189E-12 7.492E-13 2.258E-26 2.451 E-15 1.075E4 2nd 6" Water Shield Stainless Steel 8.822E-11 4.360E-11 2.920E-14 9.624E-15 2.899E-28 3.265E-17 1.384 E-0 3rd 6" Water Shield Stainless Steel 6.188E-12 3.079E-12 2.759E-15 6.763E-16 2.041 E-29 3.082E-18 9.775E-1 4th 6" Water Shield Stainless Steel 8.034E-13 4.008E-13 3.964E-16 8.786E-17 2.353E-30 4.427E-19 1.271 E-1 Sth 6" Water Shield Stainless Steel 1.281E 13 6.399E 14 6.656E-17 1.40 l E-17 2.564 E-31 7.430E-20 2.027E-1 6th 6" Water Shield Stainless Steel 2.255 E-14 1.128E-14 1.212E-17 2.468 E-18 0.000E+00 1.353 E-20 3.568E-1 7th 6" Water Shield Stainless Steel 4.491 E-15 2.258E 15 2.819E 18 4.923E-19 0.000E+00 3.150E-21 7.131 E 8th 6" Water Shield Stainless Steel 1.042E-15 5.206E-16 5.488E-19 1.140 E-19 0.000E+ 00 6.119E-22 1.647E 1 1st 1" Reactor Compartment Concrete 1.099E-13 1.122E-16 1.496E-15 1.808E-16 5.160E-30 7.993E-24 2.319E-1' 2nd 1" Reactor Compartment Concrete 1.087E 13 1.108E 16 1.387E-15 1.787E-16 5.113E-30 7.411 E-24 2.291 E 3rd 1" Reactor Compartment Concrete 1.058E-13 1.077E-16 1.258 E-15 1.739E-16 4.991 E-30 6.723E-24 2.234 E-ir 4th 1" Reactor Compartment Concrete 1.018E-13 1.034E 16 1.127E 15 1.672E-16 4.811 E-30 0 021E-24 2.147E-1' Sth 1" Reactor Compartment Concrete 9.682E-14 9.823E-17 9.994 E-16 1.590E-16 4.584 E-30 5.339E 24 2.036E-1e 6th 1" Reactor Compartment Concrete 9.105E-14 9.227E-17 8.786E-16 1.494 E-16 4.318E-30 4.696E-24 1.916E 1' 3rd 3" Reactor Compartment Concrete 7.784E 14 7.872E-17 6.671E 16 1.276E-16 3.701 E-30 3.564 E-24 1.635E-if 4th 3" Reactor Compartment Concrete 5.650E-14 5.699E 17 4.159E-16 9.257E-17 2.595E-30 2.223E-24 1.187E-11 Sth 3" Reactor Compartment Concrete 3.508E-14 3.535E-17 2.379E-16 5.748E-17 1.521E-30 1.271E-24 7.359E 1' 6th 3" Reactor Compartment Concrete 1.432E 14 1,445E-17 1.061 E-16 2.347E-17 4.843E-31 5.668E-25 3.011E 1:

1st 1" Concrete Above Operating Water Level 3.950E-11 4.084 E-14 7.663E-13 6.533E-14 1.972E-27 4.112E-21 8.393E-1 2nd 1" Concrete Above Operating Water Level 4.551E-11 4.655E-14 6.306E-13 7.4 98E-14 2.257E-27 3.383E-21 9.609E 3rd 1" Concrete Above Operating Water Level 4.862E-11 4.941E-14 5.044 E-13 7.994E-14 2.402 E-27 2.706E-21 1.023E-1:

4th 1" Concrete Above Operating Water Level 4.912E-11 4.968E-14 3.965E-13 8.063E 14 2.420E-27 2.127E-21 1.030E 1' Sth 1" Concrete Above Operating Water Level 4.755E 11 4.795E 14 3.073E-13 7.798E-14 2.339E-27 1.648E 21 9.950E-1, 6th 1" Concrete Above Operating Water Level 4,456E-11 4.483E-14 2.354 E-13 7.300E-14 2.189 E-27 1.262E-21 9.294E-1 3rd 3" Concrete Above Operating Water Level 3.613E-11 3.624E-14 1.374E-13 5.912E-14 1.771 E-27 7.367E-22 7.549 E-1 4th 3" Concrete Above Operating Water Level 2.296E-11 2.296E-14 5.585E-14 3.754E-14 1.124 E-27 2.995E-22 4.779E-1 5th 3" Concrete Above Operating Water Level 1.229E-11 1.228E-14 2.136E-14 2.009E-14 6.010E-28 1.14 5E-22 2.561E-1 6th 3" Concrete Above Operating Water Level 4.495E-12 4.4 89E-15 6.861E 15 7.345E-15 2.198 E-28 3.678 E-23 9.354E-1!

H-3 C 14' Ar-39 Ca-41 Ca-45 Mn-54 Fe-55 Csmoonent (Cmial) (Curies) (Curies) (Curies) (Curies) (Curies) (Curies) ist 6" Water Shield Lead 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 7.218E-20 3.169E-1:

2nd 6" Water Shield Lead 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 9.618E-22 4.073 E-1 <

3rd 6" Water Shield Lead 0.000E+00 0.000E+00 0.000E+00 0.000E+00 0.000E+00 9.078E-23 2.869E-1!

H-3 C 14 Ar-39 Ca-41 Ca-45 Mn-54 Fe-55 Comoonen_t (Cunes) (Cunes) (Gm!gs) [Cmics) [Cm!al) [Cudg51 (Curies)

Stainless Steel Above Water Level 5 636E 13 2.785E-13 2.017E-16 6.140E-17 1.839E-30 2.260E-19 8.896E-1; em.

Docurnent No. G01-1192-003, Rev. O Section 7, Page 26 of 27 ANSTEC 1

APERTURE CARD NT ACTIVATION Also Available on Aperture Card

. ACTOR SHUTDOWN D CONCRETE (AT CORE CENTERLINE, EXCEPT AS NOTED)

Co-60 Ni-59 NI-63 Nb-94 Tc 99 Sn-119m Sb 125 Te-125m Eu-152 (Cuded (Cudes) (Cunes) (Cunes) Sudes) (Cudes) (CgigO [CMcM (CMcM 1.498E-06 2.183E-08 2.502E-06 2.624E-11 ..J44 E-12 0.000E+00 1.533E 17 3.740E 18 2.132E 09 1.938E-08 2.603E-10 3.214 E-08 3.412E-13 4.050E-14 0.000E+00 4.160E 21 1.026E-21 2.804E-11 1.442E-09 1.962E-11 2.260E-09 2.667E-14 3.774E-15 0.000E+00 0.000E+00 0.000E+00 2.015E-12 1.916E-10 2.545E-12 2.937E-10 3.610E-15 5.403E-16 0.000E+00 0.000E+00 0.000E+00 2.641E-13 3.093E-11 4.055E-13 4.684E-11 5.883E-16 9.052E-17 0.000E+00 0.000E+00 0.000E+00 4.233E 14 5.493E-12 7.138E-14 8.252E-12 1.051E-16 1.647E 17 0.000E+00 0.000E+00 0 000E+00 7.479E-15 1.140E-13 1.419E-14 1.647E-12 2.250E 17 3.810E-18 0.000E+00 0.000E+00 0.000E+00 1.516E-15 2.524E 13 3.297E-15 3.810E-13 4.813E-18 7.458 E-19 0.000E+00 0.000E+00 0.000E+00 3.447E-16 1.176E-15 8.596E-19 9.957E-17 1.540E-19 1.877E-21 4.172E-28 3.970E-20 1.013E-20 6.447E 15 1.146E-15 8.504E-19 9.837E-17 1.483E 19 1.744 E-21 3.876E-28 3.933E-20 9.871E-21 6.336E 15 1.099 515 8.288E 19 9.573E-17 1.404E 19 1.587E 21 3.527E-28 3.59BE-20 9.043E-21 6.132E 15 1.042E-15 7.976E-19 9.200E-17 1.315E-19 1.426E-21 3.170E-28 2.952E-20 8.024 E-21 5.864E 15 9.777E-16 7.588E-19 8.741 E-17 1.219E-19 1.2C9E-21 2.819E-28 2.727E-20 7.340E-21 5.545E-15 9.083E-16 7.139E-19 8.215E-17 1.120E 19 1.119E-21 1485E 28 2.425E-20 6.533E-21 5.190E 15

'7.609E-16 6.107E-19 7.014E-17 9.210E-20 B.550E-22 1.896E-28 1.981E-20 5.471E-21 4.401E-15 5.397E-16 4.436E-19 5.085E-17 6.391E-20 5.382E-22 1.192E-28 1.194E-20 3.486E-21 3.166E-15 3.313E-16 2.755E-19 3.155E-17 3.880E-20 3.095E-22 6.840E-29 7.784E-21 2.355E-21 1.957E 15 1.369E-16 1.124 E-19 1.289E-17 1.623E-20 1.371 E-22 2.982E-29 4.016E 21 1.299E-21 8.027E 16 i 4.644E-13 3.082E-16 3.606E 14 6.530E-17 9.511E-19 2.123E-25 1.952E-17 4.763E-18 2.295E-12 4.883E-13 3.562E-16 4.129E-14 6.431E-17 7.910E-19 1.765E-25 1.658E 17 4.048E 18 2.537E-12 4.909E-93 3.814E-16 4.395E-14 6.139E-17 6.403 E-19 1.429E-25 1.378 E-17 3.363E-18 2.649E-12

-4.751E-13 3.858E-16 4.428E-14 5.712E-17 5.103E 19 1.138E-25 1.128E-17 2.753E-18 2.641E-12 4.458E-13 3.740E-16 4.280E-14 5.199E-17 4.016E 19 8.956E-26 9.134 E-18 2.229E-18 2.539E-12 4.080E-13 3.506E-16 4.005E-14 4.644E-17 3.127E 19 6.975E-26 7.334E-18 1.790E-18 2.369E-12 3.210E-13 2.845E-16 3.242E-14 3.534E-17 1.889E-19 4.214E-26 4.697E-18 1.140E-18 1.918E-12

-1.982E-13 1.809E-16 2.057E-14 2.110E-17 8.204E-20 1.830E-26 2.250E-18 5.495E 19 1.220E-12 1.045E-13 9.692E-17 1.100E 14 1.093E-17 3.359E-20 7.490E-27 1.006 E-18 2.456E-19 6.544E-13 3.806E-14 3.545E-17 4.024E-15 3.955E-18 1.114E 20 2.483E-27 3.458E-19 8.440E 20 2.402E-13 Co-60 Ni-59 Ni-63 Nb-94 Tc 99 Ag-108 Ag-108m Te-125m Eu-152 (CMcM (CRd ed (CMcM ICunes) (Cydc0 (Cudes) (Cyded (CMeO (Cudes) 2.451 E-12 0.000E+00 6.233E-11 0.000E+00 0.000E+00 3.376E-10 3.793E-09 1.838E-13 0.000E+00 3.272E-14 0.000E+00 8.316E-13 0.000E+00 0.000E+00 4.349E 12 4.886E-11 2.438E-15 0.000E+00 3.094E-15 0.000E+ 00 7.858E-14 0.000E+00 0.000E+00 3.138E-13 3.525E-12 2.210E-16 0.000E+00 Co40 Ni-59 Ni 63 Nb-94 Tc-99 Sn-119m Sb-125 Te-125m Eu-152 l{CyfeO (Cyfe$ (Cudes) (Cudes) (Cudes) (Cunes) (Cudes) (Cunes) (CudgM 1.357E-10 1.786E-12 2.052E-10 2.236E 15 2.788E-16 0.000E+00 3.087E-21 7.537E-22 1.728E-13 1

l l

mn f SAXTON ACTIVATION ANALYSIS PILOPItIETAltY INFOItMATION FINAL ItEPOILT TABLE 7.1' THE SAXTON FACILITY COMP  !

TWENTY-FOUR YEARS AFTER ((

l I

10 CFR PART 61 CLAS' l

10 CFR 61 H-3 C 14 Ca-45 Mn-54 Fe 55 Co-60 Ni-59 Ni-63 Comoonent Clais!ncation (Cl/m^3) (Cl/m^3) (Cl/m^3) (Cl/m^3) (Cl/m^3) (Cl/m ^3) (Cl/m^3) (Cl/m^3) 4 Core Baffle C 9.555E+00 7.283E+00 4.792E 17 1.569E-05 2.273E+02 4.018E+03 4.011E+01 5.064E+05 Thermal Shield B 2.640E-01 1.461 E-01 9.441 E-19 6.746E-07 4.420E+00 1.168E+02 8.123E-01 1.005E+0j Vessel Clad A 5.014 E-02 2.626E-02 1.722E-19 7.091 E-08 8.140E-01 1.637E+01 1.570E-01 1.869E+0" Vessel Wall A 1.934 E-02 9.4 86E-04 2.269E-20 5.669E-08 2.112E-01 4.761 E-01 1.648E-03 2.217E-Os VesselInsulation Can A 3.116E-03 2.009E-03 1.248E-20 1.864E-08 5.713E-02 2.473E+00 9.093E-03 1.262E+

Vessel Support Can A 1.112E-02 6.256E-03 4.020E-20 3.234 E-08 1.880E-01 5.328E+00 3.404E-0 04 Lower Core Guide Blocks GTCC 2.198E+ 01 1.808E+01 1.214E-16 1.444E-05 5.707E+02 7.730E+03 1.034E+02 1.281E+a Lower Core Plate C 9.758E+00 6.691E+00 4.493E 17 5.341E-06 2.113E+02 2.914E+03 4.019E+01 4.823E+0g Lower Support Shroud Tubes B 6.947E-01 3.564E-01 2.392E-18 2.885E-07 1.124E+01 1.586E+02 2.260E+00 2.614E+0j Lower Support Tie Rods B 1.111 E+00 5.704 E-01 3.829E-18 4.633E-07 1.798E+01 2.540E+02 3.615E+00 4.183E+0)

Lower Core Barrel B 8.524 E-01 8.347E-01 5.608E-18 7.190E-07 2.028E+01 3.543E+02 4.509E+00 5.797E+0]

Upper Core Plate C 1.301E+01 9.132E+00 6.126E-17 8.456E-06 2.880E+02 4.078E+03 5.443E+01 6.562E+0g Upper Core Barrel B 1.137E+00 6.370E-01 4.280E-18 4.477E-07 2.012E+01 2.753E+02 3.976E+00 4.650E+03 Volumes H3 C 14 Ca-45 Mn-54 Fe-55 Co-60 NI 50 NI-63 I Blcaded Comp.Qncnts fem ^3) [C.1} (CD (CD (Cl) (Cl) (CD (CD (CD  !

Lower Core Guide Blocke 8.607E+03 1.892E-01 1.556 E-01 1.045E-18 1.243E-07 4.912E+00 6.653E+01 8.897E-01 1.102E+0s Lower Core Plate 2.833E+04 2.764 E-01 1.895E-01 1.273E-18 1.513E-07 5.985E+00 p.253E+01 1.138E+00 1.366E+0$

Volumes H-3 C 14 Ca-45 Mn-54 Fe-55 Co-60 NI59 NI63

{cm^3) (Cl/m^3) (Cl/m^3) (Cl/m^3) (Cl/m^3) (Cl/m^3) (Cl/m*3) (Cl/m^3) (Cl/m^3) i Totals 3.693E+04 1.261E+01 9.346E+00 6.276E-17 7.462E-06 2.951E+02 4.036E+03 5.491E+01 6.684E+0%

w  %

e Document No. G01-1192-003, Rev. 0 ,

Section 7, Page 27 of 27  ;

i i

1 ANSTEC  !

APERTURE CARD NENT ACTIVATION l ACTOR SHUTDOWN Als Avaliable on Aperture Card i

(FICATION Nb-94 Tc 99 Sn-119m Sb 125 Te 125m Fractions Fractions Fractions Fractions a

(Cl/m^3) (Cl/m^31 (Cl/m^31 (Cl/m^3) (Cl/m 31 Table i Table 2 Col.1 Table 2 Col. 2 Table 2 Col. 3 936E-02 1.755E-02 5.511E 23 4.087E-05 9.973E-06 0.726 150.995 7.235 0.723

.006E-03 7.674E-04 4.'124 E-08 1.055E-08 0.021 3.050 0.144 0.014

. 745E-04 8.167E-05 3.028E-09 7.389E 10 0.003 0.560 0.027 0.003

.300E-05 9.630E-09 3.893E-10 9.504E-11 0.000 0.008 0.000 0.000

.328E 05 2.161 E-05 7.754E-11 1.894E-11 0.000 0.040 0.002 0.000 402E-04 3.819E-05 1.568E-10 3.824E-11 0.001 0.130 0.006 0.001 422E-01 1.681 E-02 1.642E-23 5.153 E-05 1.258E-05 1.412 378.401 18.300 1.830 291E-02 6.257E-03 1.043E-05 2.546E-06 0.533 142A98 6.889 0.689 848E-03 3.407E-04 1.056E-07 2.576E-08 0.029 7.729 0.373 0.037 562E-03 0.485E-04 1.734 E-07 4.229E-08 0.047 12.368 0.598 0.060 691E-03 8.251 E-04 1.865E-24 3.626E-06 8.849E-07 0.065 17.127 0.828 0.083 598E-02 9.826E-03 1.562E-05. 3.812E-06 0.745 194.050 9.374 0.937 856E 03 5.304 E-04 4.730E-07 1.155E-07 0.050 13.737 0.664 0.066 Nb-94 Tc 99 Sn 119m Sb-125 Te-125m

{G) (Q) (Q) (Q) [Q) Totals

.224E-03 1.447E-04 1.413E-25 4.435E-07 1.083E-07 1.829E+02

.499E-03 1.772E-04 2.954E-07 7.211E-00 2.287E+02 Sum of Sum of Sum of Sum of Nb-94 Tc.99 Sn 119m Sb-125 Te-125m Fractions Fractions Fractions Fractions

G#fn^3) (Cl/m^3) (Cl/m^3) (Cl/m^31 [Q.1/in^3) Table 1 Table 2 Col.1 Table 2 Col. 2 Table 2 Col. 3

.371 E-02 8.717E-03 3.827E-24 2.001E-05 4.884E-06 0.738 197.474 9.549 0.955 9(son 50298-

I SAXTON ACTIVATION ANALYSIS Document No. G01-1192-008, Rev. O PROPRIETARYINFORMATION Section 8, Page 1 of 2 FINAL REPORT

8. REFERENCES

1. Levin, A., "FISSPEC mU Fission Spectrum Generator for the CASK-81 and BUGLE-80 Cross Section Libraries," Version 2, TLG Services, Inc., August 1995. ,

! 2. Levin, A., "O2 FLUX - A Computer Code to Prepare ORIGEN2 Input Files," l l Version 1, TLG Services, Inc., August 1995.

3. CCC-254, "ANISN ORNL - One Dimensional Discrete Ordinates Transport Code System with Anisotropic Scattering," Oak Ridge National Laboratory Radiation l Shielding Information Center, April 1991. ,

l 4. CCC-371, "ORIGEN2 - Isotope Generation and Depletion Code - Matrix l Exponential Method," Version 2.1, Oak Ridge National Laboratory Radiation Shielding Information Center, May 1991.

5. Levin, A., "ANISNOUT - A Computer Code to Collapse and Summarize ANISN-

! ORNL Output," Version 1.0, TLG Services, Inc., August 1995.

6. Levin, A., "O2 READ - A Computer Code System to Collapse ORIGEN2 Output,"

Version 2, TLG Services, Inc., August 1995

7. Microsoft EXCEL, Version 5.0a, Microshield Corporation, Redmond, WA,1994.

8. Worku, G. -and Negin, C.A., "MICROSHIELD," Version 4.21, Grove Engineering, .

Inc., February 1995. l l 9. TMI site visit for archived document review by TLG Services, Inc., August 28,1995.

10. Ibid.

11. Ibid.

12. NUREG/CR-3474, "Long-Lived Activation Products in Reactor Materials," Evans, J.C., et. al., Pacific Northwest Laboratory, August 1984.

13. PNL-6046, " Spent Fuel Disassembly Hardware and Other Non-Fuel Bearing Components" Characterization, Disposal Cost Estimates and Proposed Repository l Acceptance Requirements," Luksic, A.T., et. al., Battelle Pacific Northwest  ;

l' Laboratory, October 1986.

14. ASM Metals Reference Handbook," 2nd Edition, American Society of Metals, March l 1984.

15. " Engineering Compendium on Radiation Shielding," Jaeger, R.G., et. al., Ed.,

Vol.ume 2, Springer Verlag,1975, Pg.10, Table 9.1-6.

DLC-76, " SAILOR - Coupled, Self Shielded, 47-Neutron, 20 Gamma-Ray, P3, 16.

Cross'Section Library for Light Water Reactors," Oak Ridge National Laboratory .

Radiation Shielding Information Center, April 1991. J

17. " Issuance of Final Branch Technical Position on Concentration Averaging and l i Encapsulation, Revision in Part to Waste Classification Technical Position", U.S.

Nuclear Regulatory Commission, January 17,1995.

l TLG SERVICES

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SAXTON ACTIVATION ANALYSIS Documsnt No. G01-1192-003, Rsv. O PROPRIETARY INFORMATION Section 8, Page 2 of 2 FINAL REPORT

18. Letter, GPU Nuclear to TLG Services, Inc., GPU number 5830-95-081, September i 18,1995, with amendments provided on October, 17,1995, and November 16,1995.

19. Telecon, A. Levin (TLG) to B. Brosey (GPU Nuclear), November 2,1995.

20. Facsimile, B. Holmes (GPU Nuclear) to A. Levin (TLG), October 24,1995.

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