ML20207J391
| ML20207J391 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 03/10/1999 |
| From: | PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20207J374 | List: |
| References | |
| PGE-1078, NUDOCS 9903160260 | |
| Download: ML20207J391 (300) | |
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Trojan Nuclear Plant 1
License Termination Plan 1
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N h I Portland General Electric l l
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,, TROJAN 12 CENSE TERMINATIONPl.AN i l
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TABLE OF CONTENTS )
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- 1. ' GENERAL INFORMATION.. ... . . .. .... . .. . .... .. ..... . ... -. 1-1 l L 1.1- PURPOSE.....................................................................................................................1-1 i L
i L 'l.2 HISTORICAL BACKGROUND........................ ............................................. ............ . 1 -2 )
i 1.3
SUMMARY
OF MAJOR ACTIVITIES AND SCHEDULE.........................................1-3
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3.1 DESCRIPTION
OF MAJOR ACTIVITIES....................................................... 1 -3
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13.1.1 Transition Period ............ ........ .... .. .................... ................................ 1 -3 i l 1.3.1.2 6 .............,....... ....................... 1-4 l i 1.3.1.3 Site Restoration ................................................................................. 1 -4 I l-L I3.2 FINAL RELEASE CRITERIA ...................... .................................... ............. 1 -5 .
! 1.3.3 SCHEDULE FOR DECOMMISSIONING / SITE RESTORATION
- l. ACTIVITIES . . . . ... ....... . . .. . ... . . . ...... . .. ..... ... .. . .. .. . . ..... ... .. .... .. . ..... ..... .. .... ... ...... . . ........ 1 -5 l 1.4 PLAN S UMMAR Y .... .. .. . .. .. . .. . .. . . .. . .. ...... . .. .. ... . . ... . . . . . . ..... ... .. .. . ... . ... .. .. .... .. . . ........... .. ... .. . 1 -6
~1.4.1
SUMMARY
OF SECTION 1 - GENERAL INFORMATION ......................... 1-6 1.4.2.
SUMMARY
OF SECTION 2 - SITE CHARACTERIZATION.......................1-6 1.4.3
SUMMARY
OF SECTION 3 -IDENTIFICATION OF REMAINING SITE DISMANTLEMENT ACTIVITIES ................................................................... 1 -6 1.4.4
SUMMARY
OF SECTlON 4 - REMEDIATION PLANS ............................... 1-6 1.4.5
SUMMARY
OF SECTION 5 - FINAL SURVEY PLAN.................................1-7 1.4.6 -
SUMMARY
OF SECTION 6 - COMPLIANCE WITH THE RADIOLOGICAL
. CRITERIA FOR LlcENSE TERMINATION .................................................. 1-7 1.4.7
SUMMARY
OF SECTION 7 - UPDATED SITE-SPECIFIC ESTIMATE OF REMAINING DECOMMISSIONING COSTS ................................................. 1-7 1.4.8 -
SUMMARY
OF SECTION 8'-EVALUATION OF ENVIRONMENTAL EFFECTS OF LICENSE TERMINATION........................................................ 1 -7 1.5 , REFERENCES FOR SECTION 1.......... ............ ....................... ...... ............................. 1 -8
- 2. SITE CHARACTERIZATION ... .. .. .... ... ..... ...... .. . ... .. . .... .. 2-1 2.1 INTR OD U CTI ON .. . .... ..... .. .. ..... ... . ... . .... . .. . ... .. . . .. ...... . . . . . . . .. . . ... .... . . . . .. ... . .... .. . . .......... . . .. . . 2- 1 ,
2.1.1 PURPOSE.........................................................................................................2-1 J 2.1.2 DEVELOPMENT OF SITE CHARACTERIZATION METHODOLOGY...... 2-1 l e 1 t
2.2 - . FACILITY RADIOLOGICAL STATUS .................. ................................................... 2-3 !
1 l 2.2.1 FACI LITY HI STORY ........................................................................................ 2-3 I I
2.2.1.1- Onerwi n e Hi storv................... ........................................................... 2 -3 2.2.1.2 Radiological History ............................................................,............ 2-3 2.2.1.2.I' E ffluents .. .. .. . ..... ...... .. ... .. .......... . ... .. .. .. .. .. .... ....... ... .. ..... .. .. . ... 2-3 l 2.2.1.2.2 Operational Events ....... .. ................................................. 2-3 {
2.2.2 : RADIOLOGICAL STATUS OF TNP............................................ ............. ... 2-4 1
- 2.2.2.1 Structure s . ..... ... .. . .. . . . .... . .. .. . . . . . . .. ..... .. . .. . .... . . . . .. . ... .... . . . ... . . . ............ .. . ... 2-6 !
2.2.2.2. Systems............................................................................................2-7 ;
p r .2.2.2.3 Activation ... . .. ....... .. . ........ .. ..... . . . . . . .. . .... ...... ... .... .... . ..... .. ...... . . . .. . . .. .. .... 2-9
- 2.2.2.4 Environm ent . . . . . . . . .. . ... .. . . .. . ... ... . . .... . ... . ...... . .. ... .... . .. .. . . . . .. . ..... . . .... .... . 2- 1 0 .
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TROJANLICENSE TERMINATIONPLAN t 2.2.2.4.1 S urface Soil S urvey ............ ................ .............................. 2- 10 f
2.2.2.4.2 Water S urvey .. .......... ........... . . ........ ............. .. . .. .................. 2- 1 1
! 2.2.2.4.3 Bottom Sediment Survey................................................... 2-12 L 2.2.2.4.4 Pavement S uivey ........................ ............ ....................... ... 2- 13 2.2.2.4.5 Exposure Rate Survey ....................................................... 2- 14 2.2.3 C ON C LU S I O N . . . ... ... . .. .. . . . . ... . . .. . . . . . .. . . . . .. . . ......... .. . . . .. . . . . .. . .. . . . . . . . . .. .. .. .. ... .... .. . . . . . . . 2- 14 2.3 OUALITY ASSURANCE PRACTICES AND PROCEDURES................................. 2-16
2.4 REFERENCES
FOR SECTION 2................................................................................ 2-17 APPENDIX 2-1,
SUMMARY
OF NOTABLE RADIOLOGICAL CONTAMINATION EVENTS APPENDIX 2-2,
SUMMARY
OF STRUCTURAL SURVEY RESULTS
- 3. IDENTIFICATION OF REMAINING SITE DISMANTLEMENT ACTIVITIES.... 3-1 3.1 IN TR O DU CTI ON . . .. . . . ... . ... .. . .. .. ..... . .... . . .. .. . .. . . ... .. ... .. .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. ... . ... .. .. ... .. . 3 - 1 3.2 DECOMMISSIONING ACTIVITIES. TASKS. AND SCHEDULES.......................... 3-2 3.2.1 INTROD U CTI ON .... . . .. . . ... . . . . . .. . . . ... .. . ... .... .. . . .. . .. . . . .... . . .. . . . . . . .. ... . ... . . .. . .. . . . . . . . ... ... .. . 3 -2 3.2.2 TRAN S ITI ON PERI OD .................................................. ... ... ............ .......... ..... 3 -2 3.2.3 DECONTAMINATION AND DISMANTLEMENT PERIOD......................... 3-5 3.2.3.1 Overview...........................................................................................3-5 3.2.3.2 Detailed Planning and Engineerine Activities........................ ......... 3-5
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3.2.3.3 General Decontamination and Dismantlement Considerations........ 3-6 l 3.2.3.4 Decontamination Methods ................................................................ 3-7 ;
3.2.3.5 Dismantlement Method s ................................................................... 3-8 !
3.2.3.6 Removal Seauence and Material Handlinc....................................... 3-9 l 3.2.3.7 Svstem Deacti vati on ...................... ..... . ................ ........................... 3- 10 3.2.3.8 Temoorary Svstems to Suoport Decommissionine......................... 3-11 3.3 REM AINING DISM ANTLEMENT ACTIVITIES ..................................................... 3-12 3.3.1 IDENTIFICATION OF REMMNING SYSTEMS, STRUCTURES, AND i C O M PON ENTS . .. . .. . . .. .. .... . . . .. .. . . . . .. . . . . . . . . . . . . . . . . ... .. .. . . .. . . . . . . . . . . . . . . . . . . . . . . .. . ..... . . . . . .. . 3 - 12 !
3.3.2.1 Remainine Structures. Systems. And Components Not Reauired For Soent Fuel Storace (Phase 1 ) .......................................................... 3-12 l 3.3.1.2 Remainine Systems. Structures. And Components Associated With S oent Fuel Storace (Phase 2) ....... .................................................. 3-13 3.3.1.3 Remaining Structures. Systems. And Components (Phase 3) ........ 3-14 3.3.2 GENERAL DESCRIPTION OF AND REMEDIATION CONSIDERATIONS FOR REMAINING SYSTEMS, STRUCTURES, AND COMPONENTS ..... 3-14 3.3.2.1 Reactor Vessel and Internal s........................................................... 3-15 3.3.2.2 Chemical and Volume Control System ......................................... 3-15 3.3.2.3 Component Coolin e Water System .......................................... ..... 3-15 3.3.2.4 S ervi c e Water S vst e m . .............. . ......... .......... ........ .... ... .. ... .......... . . 3 - 16 3.3.2.5 Spent Fuel Pool and Fuel Handline Eauipment.............................. 3-16 3.3.2.6 Spent Fuel Pool Cooline and Demineralizer System
- (Ori ci nal S vstem) ...... .. ........... ....... . ........... .. .... .. ............. ... ............. 3 - 17 ii March 1999 1
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1 L '3.3.2.7 Modular Soent Fuel Pool Cooline and Cleanuo System ................ 3-17 i
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~3.3.2.9 Condensate Demineralizers .............................................. ............. 3- 17 l Steam Generator Blowdown Svstem ............................ ................. 3-17 ,
3.3.2.10 Primary Makeue Water System and Refueline Water '
S torane Tank ... ..... ... . ..... .. . . ... . . . . .. . .. .. . . . .. . . . . .. . ... . . . . .. .. . .. . . .. . . . . . . .. .... . . . . .. 3 - 1 7 l l 3.3.2.11 Elant Effluent Svstem ..................................................................... 3- 18 3.3.2.12 - Containment Vectilation Systems................................................... 3-18 )
3.3.2.13 ' Fuel Buildine and Agiliary Buildine Ventilation Systems........... 3-19 i 3.3.2.14 Conden=*a Demineralizer Buildine Ventilation System............... 3-19 3.3.2.15 Instrument and Service Air System ................................................ 3-19 I
.3.3.2.16 Gaseous Radioactive Waste Svstem ................ .............................. 3-20 I 3.3.2.17 Solid Radioactive Waste System .................................................... 3-20 )
i 3.3.2.18 Liauid Radioactive Waste System .................................................. 3-20 )
1 3.3.2.19 . Radiation Monitorine System ......................................................... 3-21 3.3.2.20 Process Samoline System ............................................................... 3-21 3.3.2.21 Fire Protection System .................................................................... 3-22 3.3.2.22 Electrical S ystems ................ ....... .............. .................... ................. 3-22 3.3.2.23 Containment Build 1n2.................................................................... 3-22 ;
3.3.2.24 Auxiliary Buildinn (Includine Pine Facade)................................... 3-23 3.3.2.25 Fuel B uil d ine ........................... ................................................ ....... 3 -24 3 3.3.2.26 Other Buil dings ........................................ ............. ........... ............... 3 -25 i i
3.4 RADIOLOGICAL IMPACTS OF DECOMMISSIONING ACTIVITIES.................. 3-26 ;
3.4.1 OCCUPATIONAL EXPOSURE ...................................................................... 3-26 i
, 3.4.2 . - RADIOACTIVE WAS'E PROJECTIONS ..................................................... 3-26 !
3.5 REFERENCES
FOR S ECTION 3 ................................................................................ 3-28 l 4
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- 4. SITE REMEDIATION PLANSm .. ... . . . ..... .-... ..... . . . .. . .. . 4-1 l
! 4.1 - INTR ODU CTI ON .. . .. . . . .. ... . ... . . . ... .... . ... .. .. . .. .. .. ..... .. . .... . . . . ...... . . . .. . .. . . .... ....... . . . .. . . . .. . . . .. . .. . . 4- 1 1
4.2 RE MEDI ATION LEVELS ......... ......................................... . ......... .................... ............. 4-2 4.2.1 REMEDIATION LEVEL CALCULATION...................................................... 4-2 4.2.2 CALC OLATION OF TOTAL COST................................................................. 4-3 4.2.3 DETERMINATION OF REMEDIATION ACTION EFFECTIVENESS......... 4-4 4.3 'AL ARA EVALUATION .......... ........................................................... ........ ..... .............. 4 - 5 l
l 4.4 REMEDI ATION ACTION S ........... ....................... ... .. . .................................................. 4-6 l 4.4.1 STRU CTURES . ... . . . .... ........... . . . .. .. .. ... . .. . . . . . .. . .. . . ... .. . . . .. .. . . . . ... .. ... . . .. .. . . . ... . .. ... .. . . ... . 4-6 4.4.2 LANDAREAS...................................................................................................4-7 4.4.3 SYSTEMS..........................................................................................................4-7
4.5 REFERENCES
..............................................................................................................4-8 1
- 5. FINAL SURVEY PLAN .. . .. ... . . .............. .. ...... ......... . . . . ....... 5-1 l
' 5.1 ' IN TR ODU CTI ON .. .. .. . . . .... .. . . . . ... . .. . . .. .. ... ... ... .. . .. ... . ... . .. . . . .. . . .. . . . .. .. .... . .. ... . .. . . .. . .. . .. .. .. .. ... . 5 - 1 I l 5.1.1 PURPOSE...........................................................................................................5-1 l i
5.1.2 SCOPE................................................................................................................5-1 iii March 1999 .
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SUMMARY
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- 5.1.4 . DEFINITION S . .. . . .. . . .. . . . ... .... . .. . . ... . . . . .. . . . . .. . . .. ... .. . . .. . . . . .. . . . . . . . . .. .. . . . .. . .. . . . . . . . . . . . . . . .. .i S UR V E Y O VE R VIE W . .. . . .. .. . .. .. . . . . . . .. . . . . . . . . . .. . . .. . . . . . .. . . . . . . .. . . .. . . .. . . .. . . .... . . . . . . ... . . .. .. ... ...
- 5.2.1 IDENTITY OF RADIOLOGICAL CONTAMINANTS.................................... 5-8 5.2.2 - SITE RELEASE CRITERIA ..
- ........................................................................... 5-8 5.2.2.1. Radiofonical Criteria For Unrestricted Use ...................................... 5-8 5.2.2.2L Conditions Satisfyinn The Site Release Criteria............................... 5-8 j 5.2.3 ' DEVELOPMENT OF DERIVED CONCENTRATION GUIDELINE 1 LEVELS..............................................................................................................5-9 j 5.23.1 Dose Modeling ... . . ... . .. .. ... . .... .... . . . .. .. ... .. .. . ... . . . . ... . . .. . .. ... ... . .. .. . . . .. . ... ... . . 5 -9 l
5.2.3.1.1 . Building Occupancy Scenario ............................................. 5-9 5.2.3.1.2 Building Renovation or Demolition Scenario................... 5-10 5.2.3.1.3 . Residential Farming on Landfill Scenario......................... 5-10 5.2.3.1.4 . Residential Farming at Plant Site Scenario....................... 5-11 5.2.3.2 Derived Concentration Guideline Levels........................................ 5 5.2.3.2.1 Screening DCO L's ............. . ........ ...... . ................... ......... ... 5- 13 5.2.3.2.2 Site-Specific DCGL's............. .... ..................................... 5- 13 5.2.3.2.3 Surrogate Ratio DCGL's ................................................... 5-14 5.2.3.24 Gross Activity DCG L's..................................................... 5-15 5.2.3.2.5- Elevated Measurement Comparison (EMC) DCGL's....... 5-15 5.2.3.2.6 Unity Rule . . . . . . . . .. . .. . .. . .. . . .. . . . . . . . . . . . . . . . . . . . . . . .. .. .. . . . . . . . . . . . ... . .. . . . . . 5 - 16 5.2.4 FACILITY AND SITE CLASSIFICATION.................................................... 5-16 5.2.4.1 Non-Imoacted Areas . ......... .............. ...... .. .... ... . .......... .. ................... 5- 16
[ 5.2.4.2 Imnacted Areas ................... .. ... ......... ..... .... ... . ...... ........ . .................. 5- 17 5.2.4.2.1- Class 1 A' reas .. .................. ...... ... ..... ... . ..... .. ............ ............ 5- 17 5.2.4.2.2 Class 2 Areas . ....... ........ .... . . . ... ................... ........................ 5- 17 5.2.4.2.3 Class 3 Areas ................. ......... .. .... ...... ....... ........... ............. 5- 17 5.2.4.3- Initial C lassification .. ... ......... ........... ....... .. ... . . .. . . ..... ........... ... .......... 5 - 18 5.2.4.4 Channes In Classificati9D...............................................................5-18 5.2.5 . FINAL SGVEY PROCESS ........................................................................... 5- 18 5.2.5.1 S urvey Prenaration ............ .. . ... ....... . ... ... ...... .. .. . .... ........................... 5- 18 l.
5.2.5.2 S urvey Desi nn .. .... . . . . . . . . . .. .. .. . . . . . ... . . . . .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... .. . . . . . . .. 5 - 1 9
,. 5.2.5.3 Survey Data Collection ...................... . ........................................... 5- 19 L 5.2.5.4 Survey Data Assessment ............. .................................................. 5-20 i f
5.2.5.5 S urvey Results .......................... . .................. .... ..... .... ............. ....... . . 5-20 5.2.6 PROJECT MANA GEMENT........ .. ....... ... ...... ....... ... ........................... ...... ....... 5-2 0 5.2.6.1 Final Survey Ornanization .............................................................. 5-20 1 5.2.6.2 Ouality Assurance And Ouality Control (OA/OCL........................ 5-21 !
5.2.6.3 Survey Records And Documentation.............................................. 5-21 l 5.2.6.3.1- Procedures ......................... .......... .. ....... ............ ....... ........ .. 5-21 i
i 5.2.6.3.2 Technical Basis Documents .............................................. 5-21 5.2.6.3.3 Records .. . . . . .... . . .. .. . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . ..... . . . . ... . . . ... . . . . . . .. . . .. 5 -2 2 5.2.6.4 Audits And Indenendent Reviews .................................................. 5-22 5.2.6.5 Control Of Vendor Sunolied Services ...................................... ..... 5-22 ;
L 5.2.6.6' Traini n n .. .. . . ... ... .. .. . . . . . . . . . . . .... . . ... . . . . .. . .. . . . . .. . . . . . . . . . . .. . . . . . . . . .. . . . . . . . . .. .. .. . . . . 5 -22 5.2.6.7- Sched ul e .. ... .. . . . . . . . . . .. . . . . . .. . .. .. . . .. . .. . . . . . . ... . . . . .. . .. .. . . . .. . ... ... . . .. . . . . . .. . . . . . . . . . . 5 -2 3 i
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1 TROJANLICENSE TERMINATIONPLAN 5.3 S U R VEY PREPARATI ON ... ......... ............................................... ............... ................ 5 -24 5.3.1 REM EDIATION LEVELS ...... .......................................................... ..... ....... ... 5-24 bl 5.3.2 . 'ALARA EVALUATION .................................................................. ................ 5-24 5.3.3 TURNOVER.....................................................................................................5-25 5.3.3.1 Turnover Uni ts .................. ....................................................... ...... 5-25 5.3.3.2 Walkdown .... ...... ..... .. .... ..... ..... ......... .. ... ... . . .... .. . .. . .. .. . . .... . . . ... . . .. . .. ... 5 -2 5
.< 3.3.3 Turnover Criteria .......................... .......... ........................................ 5-2 5
-5.3.3.4 Transfer Of Control ................................ ......................................... 5-26 5.3.3.5 Isolation And Control Measures ..................................................... 5-26 5.4 S URVEY DESIGN . . ... . ... ..... . ...... .. ....... ... .... .......... . .. . ....... ..... .. . .... ...... .... . . . ... . . .... . . .~... . 5 -2 7 5.4.1 SU RVEY UNITS .... ... ........ . ... . ... .. . ... .. ... ....... ... ....... ... ... ............. . . .... . . ................ 5-2 7 5.4.1.1 Survey Unit Size .................................................................. ........... 5-2 7 -
5.4.1.2 Reference Coordinate System ........................................................ 5-28 5.4.1.3 . Background Reference Areas ......................................................... 5-28 5.4.2 SCAN MEAS UREMENTS ........... ...... ................................................. .......... 5-29 5.4.3 STATIC MEASUREMENTS ........... .. ............................................................. 5-30 5.4.3.1 Number Of Measurements .............................................................. 5-30 5.4.3.2 Measurement Locations .................................................................. 5-30 5.4.3.3~ Location Identification .................................................................... 5-31 5.4.4 DATA INVESTIGATION............. ................................................ . ........ ........ 5-31 5.4.4.1 Investi nation Levels ... ........................ .................. ................... ........ 5-3 1 5.4.4.2 Investi nation ...................................................................... .... .......... 5 -3 2 5.4.4.3 Remediation .. .. .. ... ... .. . .... . .. . .... .. . .... .. .. .. . . . . .. . ... . ..... . .. .. . .. . . .. .. ... . . .. . ...... 5 3 3 5.4.4.4 Reclassification ........................................................... ................... 5-3 3 5.4.4.5 Resurvey.........................................................................................5-34 5.4.5 - QUALITY CONTROL (QC) MEASUREMENTS .......................................... 5-34 5.4.5.1 ' Tvoe. N=hn. And SM & c ..................................................... 5-34 5.4.5.1.1 Scan Measurements ................................... ....................... 5-3 5 5.4.5.1.2 Static Surface Contamination Measurements.................... 5-35 5.4.5.1.3 Sail and Bulk Material Measurements .............................. 5-36 5.4.5.2 Measurement Accuraev.......................................... ........................ 5-36 5.5 S URVEY DATA COLL ECTION ........................................... ..................................... 5-3 7 5.5.1 SURVEY PEkFC RMANCE ............................................................................ 5-3 7 5.5.1.1 Turnover Survey ........................................ ..................................... 5-3 7 5.5.1.2 - Final S urvev ....... . . .. ... . . .... .. .. .. .. .. .. ... ..... ... . . ..... ... . ... . . . .. .. . .... . . . . .. .. ... .... 5 -3 8 5.5.1.3 lovestigation Survey ................................................... ................... 5-3 8 e* 5.5.2 INSTRUMENTATI ON .......... ............................................. ............................. 5-3 8 5.5.2.1 Instrument Selection ....................................................................... 5-3 8 5.5.2.2 Cahbration And Maintenance .......................................... ............. 5-3 9 5.5.2.3 Response Checks ..................................................................... ..... 5-3 9 5.5.2.4 Minimum Detectable Corne non (MDCL................................. 5-39 5.5.2.4.1. Beta-Gamma Scan MDC for Structure Surfaces............... 5-39 5.5.2.4.2 Alpha Scan MDC for Structure Surfaces.......................... 5-40 5.5.2.4.3 Gamma Scan MDC for Land Areas .................................. 5-40
.5.5.2.4.4 Static MDC for Structure Surfaces.................................... 5-40 5.5.2.5 Detection Sensitivity ................................................................ ...... 5-41 v March 1999
TROJANLJCENSE TERMINATIONPLAN 5.5.3 SURVEY METH O DS ....... ..... ........ ............................. .. ............ ....................... 5-41 O 5.5.3.1 S can Measurements .... ....................... .. ................................ ....... . ... 5 -41
( 5.5.3.2 Static Surface Contamination Measurements................................. 5-42 5.5.3.3 Soil And Bulk Matcrial Samoles .................................................... 5-42 5.5.3.4 S eeci al Measurement s ....................................... .................... ...... .. . 5-4 3 5.5.3.4.1 Cracks, Crevices, and Small Holes ................................... 5-43 5.5.3.4.2 Paint Covered Surfaces ..................................................... 5-43 5.5.3.4.3 Plant Systems, Floor Drains, and Embedded Piping......... 5-43 5.5.3.4.4 Activated Concrete and Other Materials........................... 5-44 5.5.3.4.5 Paved Parking Lots, Roads, Sidewalks, and Other Paved Areas ....................................................................... 5-44 '
5.5.3.5 Investiaation Measurement _s_ ........................................................... 5-44 5.5.4 SAMPLE HANDLING AND ANALYSIS ...................................................... 5-44 5.5.5 DATA MANAG EMENT. ...........................-..... ............................................... 5-44 5.5.5.1 Scan Measurem ents .............. ............................................... .. ...... 5-4 5 5.5.5.2 S tati c Measurem ents .................... ............ ..... ......... ...... ................ ... 5-4 5 5.5.5.3 Data R ecord in e .................... ... . . ................... ......... .......... ............... 5-4 5 5.6 SURVEY D ATA A S S E S S M ENT . ................. ..... . ................................. ........ ........ ....... 5-4 6 5.6.1 DATA VERIFICATION AND VALIDATION......................... ,.............-..... 5-46 5.6.2 GRAPHICAL DATA REVIEW .................................................................... ... 5-47 5.6.2.1 Po st i n e P l o t . . . . . . . . . . ... .. . . .. . .. .. . .... ... .. . .. .... . .. .. . . ... . .. . . .. . . . . . .. . . . . . . . . . . .. . . . . . . .. 5 -4 7 5.6.2.2 Frea u en c y Pl ot . ........................................................................ .... .. . 5-4 7 5.6.3 BASIC STATISTICAL COMPARISONS .......................... ............................ 5-47 O 5.6.3.1 5.6.3.2 Range...............................................................
Median.............................................................. ..........................5-48
. .. .. . . . .. .. . . . . .. . . .. . 5 -4 8 5.6.3.3 Mean...............................................................................................5-49 5.6.3.4 S tan dard Devi ation.............. ................. ................... ............. ....... ... 5-4 9 5.6.4 STATI STICAL TEST... . ..... ...... ..................... . .................................... .. ...... .. .... 5-5 0 5.6.4.1 Acolication Of Statistical Test........................................................ 5-50 5.6.4.2 SignTest........................................................................................5-51 5.6.4.3 Wilcoxon Rank Sum (WRS) Test ................................................... 5-51 l 5.6.5 D ATA C ONC LU S I ON S ........ ............ ................ .............. ............. ........ . ...... . .. 5-5 3 !
5.7 . S U R V EY RE S U LTS . . . . . . . . . . .. . . . . . .. . . ... .. . . .. .. . .. .. . .. .. . . .. . . . . . . . . . . . ... .. . ... . .. .. . . . . .. . . ... . . .. . .. . .. . . . . . . .. 5-5 4 5.7.1 HI S TO RY FI L E . . . . . ... .. . . .. . . . . . . .. .... . . . . .... . . .. .. .. . . . .. . .. . . . .. . . . .. . . . . . . ... .. .. . . . ... . .. . . . . . . . . . . .... 5- 5 4 5.7.2 SURVEY UNIT RELEAS E RECORD ..................................... ...................... 5-54 l 5.7.3 FINAL S URVEY RE PORT ...................................................... ... ........... ........ . 5-5 4 1
5.7.4 OTH E R RE PO RTS . . ... . . . .. . . . .... .. . . .. . .......... . .. . .... . .. ... .. . . .. .. . . . . . . . .. . . . . . . .. . ... . . . . . .... . . . . . 5 - 5 5 ;
5.8 RE FE REN C E S . .. . .. .. . . .. . . . . . . . . . . . . . . . . . ... .. . ... .. .. . . . .. . ... .... . . . . . .. .. .. .. .. ... . ... ... . . .. . . .. . .. .. . . . ... . .. . . . . . . . 5 -5 6 l APPENDIX 5-1, ELEVATED MEASUREMENT COMPARISON (EMC)
APPENDIX 5-2, NUMBER OF STATIC MEASUREMENTS APPENDIX 5-3, BACKGROUND REFERENCE AREAS
- 6. COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE
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TE RMINATI O N ......................................... ..... . ...... ... .. ... ... ................................ 6- 1
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TRIM 4NIJCENSE TERMINATIONPL4N 6.1 DISCUSSION...............................................................................................................6-1
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i 6.2 RE FE RENC ES FOR S ECTION 6............... ............ .. ........ .. ...... ..... ... ...... .... .......... ......... 6-2
- 7. UPDATE OF SITE-SPECIFIC DECOMMISSIONING COSTS ...... .... .. ....... 7-I 7.1 INTR O DU CTI ON . . . .. ... . .. ... .... .. . ... .. . ... . ...... . . .. .. . ... .. ... . . . ... . . . . .. .. . . .. .... . . . . . .. ... . . .. . .. . . . . .. .. . . .. .. . 7-7.2 DECOMMISSI ONING COST ESTIMATE........................ .......................................... 7-2 7.2.1 COST ESTIMATE R ESULTS .... ............................................. ....... ................. 7-2 7.2.2 COST ESTIMATE DESCRIPTION......................................... ......................... 7-2 7.2.2.1 Radiological Decommissioninn Costs .............................................. 7-3 7.2.2.2 Nonradiolonical Decommissionine Costs......................................... 7-4 7.2.2.3 Spent Fuel Mananement Costs.......................................................... 7-4 7.2.2.4 Financial Activity Costs.................................................................... 7-4 7.3 SPENT FUEL MANAGEMENT FUNDING PLAN ........ ............................................ 7-6 7.4 DECOMMI S SIONING FUNDING PLAN .................................................................... 7-7 7.4.1 CURRENT DECOMMISSIONING FUNDING CAPABILITIES.................... 7-7 7.4.2 TNP CO-OWNERS' DECOMMISSIONING FUNDING PLANS ................... 7-7 7.4.2.1 _P_fi E Fun d i n e . . . . .. ... ...... . .... . ... . . . .. . .... . . . .. . . . . . . . . . . . . . . . . .. . .. . . .. .. . . . . . .. . . ... . . . .
7.4.2.2 E WE B/B P A Fund 1n 2 .................... ......... ..... . .. . .. .............. ................. 7-8 7.4.2.3 P P & L Fund i n n . . .. . . . . .. . .. . .. . . .. .. . . . . . . . .. ... . . .. .. . . . . . .. . . .. . . . . . . . . . . . . . . . . . . .. . .. . . ... . . . 7-9 7.5 REFE REN C E S FOR S ECTION 7............................... ........... .. .... .. ... . ..... .. .... ............ ... 7- 10 s
- 8. EVALUATION OF ENVIRONMENTAL EFFECTS OF LICENSE TERMINATION...
....... . . .. ......... ......... ... . ..... . . ..... .... .. .. . ..... ...... . 8-I 8.1 INTR O D U CTI ON . . .. . . . . . ... . .. . . . . . . .. .. . .. . . .. . . .. . . .. . . . . .. . . . . . . . . . . .. . .. . . .. . . .. . . ... . . . . . . . . . . . . . . . . . . . . . . . .. . . . . ..
8.2 CURRENT ENVIRONMENT IMPACT ....................................................................... 8-2
8.3 CONCLUSION
...............................................................................................................8-3 8.4 REFE REN C ES FOR S ECTION 8 ....................................... .... .......... ......... .................. 8-4 LIST OF ABBREVIATIONS AND ACRONYMS INDEX (A) vii March 1999
I TROJANUCENSE TERMINATIONPLAN l
LIST OF TABLES FOR LICENSE TERMINATION PLAN G
Table Title 2-1 Radioactive Effluent Summary, Noble Gases 2-2 Radioactive Effluent Summary, Iodine and Partien:ates (excluding tritium) 2-3 Radioactive Effluent Summary, Liquids 2-4 Structures Burial Volume and Contamination Activity Projections 2-5 Status of Buildings in the Radiologically Controlled Area 2-6 System Burial Volume and Surface Activity Projections j 2-7 Isotopic Distribution (Decay Corrected to 1994 and 1998) 2-8 10 CFR Past 61 Classification by Component One Year After Shutdown 2-9 10 CFR Part 61 C'r,iFcation by Component Five Years After Shutdown 3-1 Status of Major TNP Systems, Structures, and Components as of January 1999 3-2 . Major Components Removed (By Year) 3-3 Radiation Exposure Projections 3-4 Decommissioning Waste Classification and Volume Projections j 5-1 Screening PCGL's j 5-7 Initial Classification ofTrojan Facility and Site 5-3 Survey Design Summary 5-4 Data Renilts and Investigation Conclusions i 5-5 Typi;al Survey Instrumentation !
1 5-6 Typical Detection Sensitivities !
5-7 Survey Results and Conclusions When A Background Reference Area Is Not Used 5-8 Survey Results and Conclusions When A Eackground Reference Area Is Used 7-1 Estimate of" Trust Fund" Decommissioning Costs (1997 dollars) 7-2 Decommissioning Cost Estimate for Trojan Nuclear Plant Itemized Decommissioning Expenditure Schedule (1997 $ x 1000) 7-3 Status of Decommissioning Trust Funds as of December 31,1998 O
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L . 7-4L Portland General Electric Decommissionmg Trust Fund Cash Flow l l As Of12/31/98 (Nominal $ x 1000) I
- j. 7-5 EWEB/BPA Decommissioning Trust Fund Cash Flow ;
. As Of12/31/98 (Nominal $ x 1000) i 7-6 Pacific Power & Light Decommissioning Trust Fund Cash Flow l'
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TROJANLICENSE TERMINATIONPL.d :
. LIST OF FIGURES FOR LICENSE TERMINATION PLAN :
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Figure Title ;
2-1 Radiological Analysis Samples, Trojan Site, Zones 1 through 16
.2-2 Radiological Analysis Samples, Zone 1 !
2-3 Radiological Analysis Samples, Zone 2 2-4 Radiological Analysis Samples, Zone 3 !
2-5, Radiological Analysis Samples, Zone 5 l
2-6 Radiological Analysis Samples, Zone 6 2-7 Radiological Analysis Samples, Zone 7 l 2-8 Radiological Analysis Samples, Zone 8 2-9 Radiological Analysis Samples, Zone 9 2-10 Radiological Analysis Samples, Zone 10 l 2 Radiological Analysis Samples, Zone 11 2-12 Radiological Analysis Samples, Zone 14 i 2-13 Radiological Survey Data, Auxiliary Building, Elevation 5 ft 2-14 Radiological Survey Data, Auxiliary Building, Elevation 25 ft 2-15 Radiological Survey Data, Turbine Building, Elevation 27 ft 2-16 Radiological Survey Data, Turbine Building, Elevation 45 ft 2-17 Radiological Survey Data, Main Steam Support Structure, Elevs! ion 45 ft 2-18 Radiological Survey Data, Control Building, Elevation 45 ft 2-19 Radiological Survey Data, Auxiliary Building, Elevation 45 ft 2-20' Radiological Survey Data, Fuel Building, Elevation 45 ft ,
2-21 Radiological Survey Data, Containment, Elevation 45 ft L2-22 Radiological Survey Data, Turbine Building, Elevation 63 ft l 2-23 Radiological Survey Data, Control Building, Elevation 61 ft and 65 fi
! 2-24' Radiological Survey Data, Auxiliary Building, Elevation 61 ft
! 2-25. Radiological Survey Data, Fuel Building, Elevation 61 ft 2
2-26 Radiological Survey Data, Main Steam Support Structure, Elevation 69 ft 2-27 Radiological Survey Data, Cc,ntainment, Elevation 61 ft !
2-28 Radiological Survey Data, Control Building, Elevation 77 ft x March 1999 l 1
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TROJANLICENSE TERMINATIONPLAN Figure Title
'f 2-29 Radiological Survey Data, Auxilia;y Building, Elevation 77 A 2-30 Radiological Survey Data, Fuel Building, Elevation 77 ft 2-31 Radiological Survey Data, Containment, Elevation 77 ft -
2-32 Radiological Survey Data, Turbine Building, Elevation 93 ft 2-33 Radiological Survey Data, Control Building, Elevation 93 ft 2-34 Radiological Survey Data, Auxiliary Building, Elevation 93 ft 2-35 Radiological Survey Data, Fuel Building, Elevation 93 ft 2-36 Radiological Survey Data, Containment, Elevation 93 A 3-1 Site Plan 3-2 Containment and Auxiliary Buildings, Plan Below Ground Floor 3-3 Containtnent and Auxiliary Buildings, Elevation 45 ft 3-4 Containment and Auxiliary Buildings, Elevation 61 A 3-5 Containment and Auxiliary Buildings, Elevation 77 ft 3-6 Containment and Auxiliary Buildings, Operating Floor and Above 3-7 Turbine Building, Elevation 27 ft and 45 ft 3-8 Turbine Building, Elevation 61 ft and 63 ft 3-9 Turbine Building, Elevation 93 ft 3-10 Decommissioning / Site Restoration Schedule 1 Data investigation Process 5.-2 Data Assessment Process -
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' TROJANLICENSE TERMINATIONPLAN LIST OF EFFECTIVE PAGES Section/Page Revised Date j Title Page March 1999 Pages i through xii . March 1999 Section 1 Pages 1-1 through 1-8 ' March 1999 i Section 2 i Pages 2-1 through 2-17 March 1999 Appendix 2-1 ;
Pages 1 of 5 through 5 of 5 March 1999 Appendix 2-2 :
Pages_1 of13 through 13 of13 March 1999 i Tables 2-1 through 2-9 March 1999 Figures 2-1 through 2-36
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March 1999 :
Section 3 i Pages 3-1 through 3-29 March 1999 l Tables 3-1 through 3-4 March 1999 Figures 3-1 through 3-10 March 1999 Section 4 Pages 4-1 through 4-8 March 1999 A Section 5 Pages 5-1 through 5-56 March 1999 Appendix 5-1 Page1of1 March 1999 Appendix 5-2 Pages 1 of 6 through 6 of 6 March 1999 Appendix 5-3 Pages 1 of 2 and 2 of 2 March 1999 Tables 5-1 through 5-8 March 1999 Figures 5-1 through 5-2 March 1999 .
Section 6 l
Pages 6-1 and 6-2 March 1999 Section 7 I Pages 7-1 through 7-10 March 1999 l Tables 7-1 through 7-6 March 1999
- Section 8 -
Pages 8-1 through 8-4 March 1999 !
List of Abbreviations and Acronyrns (1 page) March 1999 Index (2 pages) March 1099 p
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TROJANLICENSE TERMINATIONPLAN
, 1. GENERAL INFORMATION ~ !
-1.1 PURPOSE The Trojan Nuclear Plant (TNP) License Termination Plan has been prepared in accordance with the requirements of 10 CFR 50.82, " Termination of License" (Reference 1-1) ard the guidance provided in Regulatory Guide 1.179, " Standard Format and Content of License Termination Plans for Nuclear Power Reactors"(Reference 1-2). The TNP License Termination Plan is -
maintained as a supplement to the TNP Defueled Safety Analysis Report (DSAR) l (Reference 1-3) in accordance with .10 CFR 50.82(a)(9)(i). l l
This plan demonstrates that the remainder of the decommissioning activities at the TNP site will be performed in accordance with the regulations in 10 CFR 50.82; will not be inimical to the common defense and security nor to the health and safety of the publ*c; and will not have a significant effect on the quality of the environment.
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TROJANLJCENSE TERM)NATIONPLAN 1
1.2 HISTORICAL BACKGROUND t TNP is located in Columbia County, Oregon, approximetely 42 miles north of Portland, Oregon.
TNP is jointly owned by Portland General Electric (PGE),67.5 percent; the City of Eugene, 30 '
percent through the Eugene Water and Electric Board (EWEB); and Pacific Power and Light /PacifiCorp (PP&L),2.5 percent. POE is the majority owner and has responsibility for ,
operating and maintaining TNP. The Bonneville Power Administration (BPA), a power marketing agency under the United States Dep.soent of Energy (DOE), is obligated through Net Billing Agreements to pay costs associated with EWEB's share of TNP, including decommissioning and spent fuel management costs.
3 TNP, Docket 50-344, achieved initial criticality in December 1975 and began commercial operation in May 1976. The reactor output was rated at 3411 MWt with an approximate net electrical output rating of 1130 MWe. The nuclear steam supply system was a four-loop pmssurized water mactor designed by Westinghouse Electric Corporation.
TNP was shutdown for the last time on November 9,1992. On January 27,1993, after '
approximately 17 years of operation, PGE notified the Nuclear Regulatory Commission ofits ;
decision to permanently cease power operations. The owners' decision was predicated on both :
fmancial and reliability considerations. The NRC amended the TNP Facility Operating License
_ (NPF-1) to a Possession Only License on May 5,1993. On October 7,1993, PGE transmitted an updated Safety Analysis Report for the Defueled Condition (DSAR).
t PGE submitted a proposed TNP Decommissioning Plan (Reference 1-4) and Supplement to the O- Environmental Report (Reference 1-5) on January 26,1995, which were approved by the NRC on April 15,1996 (Reference 1-6). The TNP Decommissioning Plan was submitted and ;
approved in accordance with the Nuclear Regulatory Commission's rule governing >
decommissioning and termination oflicense,10 CFR 50.82, " Application for Termination of License." This rule has since been revised. The revised rule, specifically 10 CFR 50.82(a)(9),
requires all power reactor licensees to submit a license termination plan, either prior to or with the licensee's application for !icense termination, for NRC approval at least two years before
. termination of the license date. This License Termination Plan satisfies the requirc.ment 10 CFR 50.82(a)(9).
O 1-2 March 1999
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. TROJANLICENSE TERMINATIONPLAN 1.3
SUMMARY
OF MAJOR ACTIVITIES AND SCHEDULE
- INP decommissioning is divided into two broad periods: a Transition Period and a .
Decontamination and Dismantlement Period.- Decommissioning will be followed by site - !
restoration. This section provides a brief description of these activities. Details are provided in l Section 3. ,
.1.3.1- DESCRIPTION OF MAJOR ACTIVITIES j
' De Transition Period began with permanent plant shutdown in January 1993 and will continue until spent fuel is transferred to an ISFSI. The Decontamination and Dismantlement Period will 1
begin once the spent fuel is transferred to the ISFSI. Site restoration will begin following ,
termination of the 10 CFR 50 license and involves the final disposition of structures, systems, ;
and components. i
. Storing fuel at TNP during and after plant decommissioning significantly impacts both the I process and costs associated with decommissioning. The TNP contract with DOE," Standard ;
Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste," provides the i basis for the schedule forecast in DOE's annual acceptance priority ranking for receipt of spent ,
L fuel and/or high level waste. The published schedule specifies the first TNP shipment to be in j
- ~2002, and the final shipment is projected for 2018. Recognizing the uncertainty, but with no '
better formal estimate, the contract dates for fuel shipment are currently being used for planmng purposes. l l.3.1.1 Transition Period
- ne Transition Period of TNP decommissioning is nearing completion. PGE continues to ;
maintain systems and components required to support decommissioning and spent fuel storage in accordance with the Facility Operating (Possession Only) License NPF-1 and administrative
. procedures. Activities include assessing the functional req'tirements for systems, structures, and components; deactivating systems, structures, and components; active decontamination and dismantling of systems / components not necessary for assuring safe spent fuel storage in the j spent fuel pool; and maintaining safe spent fuel storage. This effort includes the conduct of detailed decommissioning project planning, preparation of engineering specifications and procedures, procurement of special equipment needed to support decommissioning, and -
negotiation of service contracts required for decommissioning activities.
The facility currently is maintaining its s,nent fuel in the spent fuel pool and undergoing active decontamination and dismantlement activities in accordance with the approved TNP DSAR and Decommissioning Plan. A concurrent effort is underway to construct an Independent Spent Fuel Storage Installation (ISFSI) at.the Trojan site to. facilitate decommissioning of the TNP. As discussed in Section 3, until the spent fuel is transferred to the ISFSI, the decommissioning effort primarily involves removal or in-place decontamination of contaminated systems and
- components not required for support of fuel storage or subsequent decontamination activities.
Fuel transfer to the ISFSI is planned to begin in early-1999. The completion of fuel transfer to 3 1-3 March 1999 i
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TROJANLICENSE TERhf1NATIONPLAN i
I the ISFSI will allow the removal or decontamination in place of systems and components that ,
support the SFP or wet fuel storage, including the SFP itself. i Included in ongoing efforts is the removal of the reactor vessel with intemals (reactor vessel package) from the 10 CFR 50 licensed area of the TNP site, as approved by the NRC on .
October 29,1998 (Reference 1-7). 'Ihe reactor vessel package is expected to be transported in i 1999 for disposal at the US Ecologi low level radioactive waste facility near Richland, .
l Washington. Removal of the reactor vessel package from the 10 CFR 50 licensed area of the l
TNP site will eliminate approximately 2 million curies of activity from the TNP. Not including the spent nuclear fuel that will be transferred to the ISFSI, removal of the reactor vessel and intemals will result in removal of greater than 99 percent of the remaining activity (curies) at the TNP facility. !
1.3.1.2 Decontamination and Dimmantlement Once the spent fuel is transferred to the ISFSI, the Transition Period ends and the ;
3 Decontamination and Dismantlement Period begins. Major activities planned during the
! Decontamination and Dismantlement Period include removing remaining contaminated systems '
, and components, decontaminating structures, and a final radiation survey to verify radioactivity !
i has been reduced to sufficiently low levels to allow for unrestricted release of the site.
Contaminated systems, components, and structural material will be decontaminated or removed and packaged. The packaged material will either be shipped to an off-site processing facility, )
shipped directly to a low level radioactive waste disposal facility, or otherwise handled in !
- accordance with applicable regulations.
3 Decontaminating plant structures may be completed concurrent with removing equipment and systems and may include the use of a variety of techniques ranging from water washing to ;
surface material removal. Demolishing certain buildings may be necessary based on degraded !
structural integrity as a result of decontamination efforts and/or removal of systems and components, surrounding walis, or other barriers.
A final radiation survey, described in Section 5, will be performed to demonstrate that radiological conditions at TNP satisfy the final site release criteria of 10 CFR 20.1402 to support
- unrestricted release of the TNP site and license termination. Upon completion of the final survey, a final survey report will be submitted to the NRC.
i 1.3.1.3 Site Restoration
- Nonradiological site remediation activities are scheduled to be completed following termination j of the Facility Operating (Possession Only) License NPF-1. The primary nonradiological site remediation effort is scheduled to begin around 2018 and conclude in 2019. Some site restoration activities have been completed, and some may continue to be conducted during the Transition and Decontamination and Dismantlement Periods of decommissioning.
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TROJANLICENSE TERMINATIONPLAN A listing and schedule of remaining major license termination activities is provided in Section 3.
O According to this schedule, PGE anticipates the completion of decommissioning activities, l including the final survey ud license termination, by late 2002.
l 1.3.2 FINAL RELEASE CRITERIA TNP decommissioning will safely reduce radioactivity at the site to levels meeting the unrestricted release criteria of 10 CFR 20.1402 (Reference 1-8). The TNP final survey plan I provided in Section 5 describes the scope and methodology of the fmal survey process, quality I assurance measures, access control procedures, and how implementation of the plan will l demonstrate that the plant and site will meet the 10 CFR 20.1402 criteria for unrestricted release i
' of the site. j 1.3.3 SCHEDULE FOR DECOMMISSIONING / SITE RESTORATION ACTIVITIES j A detailed schedule for decommissioning / site restoration activities is presented in Section 3.2. ;
The following is an overview of the current TNP decommissioning / site restoration project i schedule. !
January 1993 -Early 2000 Transition Period Late 1994 - Late 1995 Large Component Removal Project i Late 1996 - Late 1998 Complete planning / building an ISFSI ,
Early 1997 - Late 1999 Reactor Vessel and Intemals Removal Early 1999 - Early 2000 Transfer spent nuclear fuel to the ISFSI pd Early 2000 - Late 2002 Decontamination and Dismantlement Period l
Mid-2001 Submit application for license termination Late 2002 Complete final radiation survey ;
Late 2002 - Mid 2018 Caretaking Mid 2018 - Late 2019 Demolish buildings l
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1 TROJANLJCENSE TERMINATIONPLAN l
1.4 PLAN
SUMMARY
This TNP License Termination Plan describes the process by which decommissioning will be
' completed and the TNP site released for unrestricted use. The plant activities described in the' ;
TNP License Termination Plan are consistent with the activities that already may be conducted under the approved TNP Decommissioning Plan and the TNP DSAR. The following subsections provide a brief summary of the sections presented in the License Term' m ation Plan.
(- .1.4.1 .
SUMMARY
OF SECTION 1 - GENERAL INFORMATION b This section provides the purpose of and regulatory basis for the TNP License Termination Plan, as well as a brief overview of each section contained in the plan. A brief historical background !
and a summary description and schedule of major activities also are provided.
1.4.2
SUMMARY
OF SECTION 2 - SITE CHARACTERIZATION
, In accordance with 10 CFR 50.82(aX9)(iiXA), this section provides a description of the -
radiological conditions at the TNP site. The TNP site characterization incorporates the results of
. scoping and characterization surveys conducted to quantify the extent and nature of contamination at TNP. The results of the scoping and characterization surveys have been and continue to be used to identify areas of the site that will require remediation, as well as to plan remediation methodologies and costs.
q 1.4.3
SUMMARY
OF SECTION 3 -IDENTIFICATION OF REMAINING SITE
.Q DISMANTLEMENT ACTIVITIES In accordance with 10 CFR 50.82(aX9)(ii)(B), this section identifies the major dismantlement a
and decontamination activities that remain at TNP. This information includes those areas and equipment that need further remediation to allow an estimation of the radiological conditions that may be encountered during remediation of equipment, components, structures, and outdoor areas.
5 The status of decontamination and dismantlement of TNP systems, structures, and components, as ofJanuary 1999, is summarized in Table 3-1. Table 3-2 contains a list of major components removed during each year since 1996. As indicated by Table 3-1 and Table 3-2, the majority of radiologically contaminated systems and components not required to support the storage of spent fuel have been deactivated, dismantled, and disposed ofin accordance with the TNP Decommissioning Plan. Additional details regarding completed and remaining decontamination, dismantlement, and radiological controls for these structures, systems, and components are also o
provided.
l 1.4.4
SUMMARY
OF SECTION 4 - REMEDIATION PLANS In accordance with 10 CFR 50.82(aX9)(iiXC), this section describes how remediation actions j may be applied to various areas on the TNP site, inatifies the remediation methodology to be g used, and demonstrates that the remediation methodology is adequate to ensure that the site Q
1-6 March 1999
TROJANUCENSE TERMNAHONPL4N f . release criteria of 10 CFR 20.1402 are met. Verification of the site release criteria is detailed further in Section 5, Final Survey Plan. '
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! 1.4.5
SUMMARY
OF SECTION 5 - FINAL SURVEY PLAN 1
In accordance with 10 CFR 50.82(aX9XiiXD), the TNP Final Survey Plan describes the methods I_
and criteria that will be used to demonstrate that the TNP site meets the radiological release criteria for unrestricted use specified in 10 CFR 20.1402. This plan includes a description of control measures implemented in accordance with approved plant procedures to preclude the
<. recontamination'of clean areas. The TNP Final Survey Plan also incorporates measures to ensure that final survey activities are planned and discussed with the Nuclear Regulatory
- Commission and the Oregon Office of Energy sufficiently in advance to allow the scheduling of
, inspection activities.
, 1.4.6
SUMMARY
OF SECTION 6 - COMPLIANCE WITH THE RADIOLOGICAL CRITERIA FOR LICENSE TERMINATION As described in the approved TNP Decommissioning Plan, the decommissioning objective at the TNP site is to reduce residual radioactivity to a level that permits release of the portion of the site licensed under 10 CFR 50 for unrestricted use. In accordance with 10 CFR 20 and Regulatory Guide 1.179, this section and Section 5, Final Survey Plan, demonstrate that the radiological t criteria of 10 CFR 20.1402 for unrestricted release will be met.
1.4.7
SUMMARY
OF SECTION 7 - UPDATED SITE-SPECIFIC ESTIMATE OF REMAINING DECOMMISSIONING COSTS In accordance with 10 CFR 50.82(aX9XiiXF), this section provides an updated site-specific estimate of remaining decommissioning costs, a comparison of these estimated costs with the present funds set aside for decommissioning, and a description of the means for ensuring adequate funds to complete decommissioning.
1.4.8
SUMMARY
OF SECTION 8 - EVALUATION OF ENVIRONMENTAL EFFECTS OF LICENSE TERMINATION In accordance with 10 CFR 50.82(aX9XiiXG), this section compares the impacts associated with TNP site-specific license termination activities as described in the TNP License Termination Plan with previously analyzed termination activities described in the approved TNP Decommissioning Plan, Supplement to the Environmental Report, and DSAR. The evaluation in this section finds that the activities described in the TNP License Termination Plan result in no l
. significant environmental changes not bounded by the TNP Decommissioning Plan and the
,- previously approved Supplement to the Environmental Report. This evaluation satisfies the l . requirement of 10 CFR 51.53(d) (Reference 1-9), to reflect any new information or significant environmental change associated with proposed decommissioning or fuel storage activities.
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TROJANLJCENSE TERMINATIONPL.AN
1.5 REFERENCES
FOR SECTION 1
.(%
Q 1-1_ Code of Federal Regulations, Title 10, Part 50.82, " Application for Termination ofLicense."
l-2 Regulatory Guide 1.179. " Standard Format and Content of License Termination Plans for Nuclear Power Reactors," January 1999.
1-3 Portland General Electric, " Trojan Nuclear Plant Defueled Safety Analysis Report," Revision 7.
l-4 Eplt. land General Electric Tonical Report PGE-1061, " Trojan Nuclear Plant Decommissioning Plan," Revision 6.
15 Portland General Electric Tonical Reoort PGE-1063. " Trojan Nuclear Plant Supplement to Applicant's Environmental Report - Post Operating License Stage," Revision 3.
1-6 NRC Letter. M. T. Masnik to S. M. Ouennoz, " Order Approving the Decommissioning Plan and Authorizing Decommissioning of the Trojan Nuclear -
Plant," April 15,1996.
1-7 NRC Letter. W. F. Kane ta S. M. Ouennoz, " Authorization of the Trojan Reactor Vessel Package for Transport," October 29,1998, 1-8 Code of Federal Regulations, Title 10, Part 20, " Standards for Protection Against Radiation."
~1-9 Code of Federal Regulations, Title 10, Part 51.53, " Post-Operating License Stage Environmental Reports."
l-8 March 1999 l
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n TROJANIJCENSE TERMINATIONPLAN
- 2. SITE CHARACTERIZATION.
2.1 INTRODUCTION
2.1.1 ~ PURPOSE. -U In accordance with the requirements of 10 CFR 50.82(a)(9)(ii)(A) (Reference 2-i) and guidance of Regulatory Guide 1.179 (Reference 2-2), this section provides a description of the radiological '
conditions at the TNP site. "Ihis TNP site characterization incorporates the results of surveys conducted to quantify the extent and nature of contamination at TNP. The results of TNP site characterization surveys and analyses, provided previously to the NRC as Section 3.1 of the ~
approved TNP. Dammmissioning Plan (Reference 2-3), have been and continue to be used to identify areas of.ths site that will require remediation, as well as to plan reinediation methodologies and costs.
2.1.2 l DEVELOPMENT OF SITE CHARACTERIZATION METHODOLOGY
~.dP's site characterization plan was developed and implemented following permanent shutdown of the plant in 1993 using the guidance available at the time the characterization surveys were conducted. This guidance included Regulatory Guide 1.86 (Reference 2-4), NUREG/CR-5849 (Reference 2-5) and NUREG/CR-5512 (Reference 2-6). Following the development and -
implementation of the site characterization plan, the revised release criteria of 10 CFR 20, l
. Subpart E (Reference 2-7), and guidance of NUREG-1575 (MARSSIM) (Reference 2-8) were l e - issued.- Although the fNP site characterization was conducted under the previous guidance, PGE has elected to conduct final surveys using the most recent 10 CFR 20.1402 release criteria and the MARSSIM approach of applying derived concentration guideline levels to verify that
- allowable release criteria are met.
..NUREG-1575 incorporates a data quality objectives (DQO) process to ensure that survey results are of sufficient quality and quantity to support the decision to release the area for unrestricted use. The DQO process is introduced as a systematic planning tool to determine the type, quantity, and quality.ofdata needed to support decision-making. Because the DQO process is intended to be used in the planning and development effort, applying this process to the completed TNP site characterization is not practicable. Moreover, for areas of the TNP site classified as impacted, this action would have no significant benefit since: 1) many of the
' contaminated systems and components have been removed from the site, making much of the
- site characterization data historical in nature; and 2) these areas will undergo final survey to .
. verify that the unrestricted release criteria are met. l The portion of the TNP site outside the current industrial area was determined from the results of the site characterization to be non-impacted, thus requiring no additional surveys. This
- conclusion was reached without application of the DQO process since, as stated above, the
- guidance in use at the time did not incorporate the DQO concept. PGE has determined that the
- site characterization results, including classification of the non-impacted areas, remain valid since the TNP site characterization activities were planned and conducted using a rigorous s
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TROJANLICENSE TERMINATIONPLAN I systematic method in accordance with the available regulatory guidance as well as the approved Trojan Nuclear QA Program (Reference 2-9).
In accordance with the available guidance, the TNP Site Characterization Plan incorporated the following objectives: ,
- 1. . Determine the initial (post operation) radiological status of the facility;
- 2. Estimate the site source term and isotopic mixture to support decommissioning cost estimation and decision-making; and
- 3. Determine the location and extent of any contamination outside the radiologically controlled areas.
In the plan, the primary decision-maker and key team members were identified. Available resources were specified and relevant deadlines for the survey established. To assure i i
representative data, the site characterization plan identified the method for selecting the type and number of measurements, locating those measurements, and determining the background contribution.
The site characterization process was divided into four areas: structures, systems, activation, and environment. Quality assurance requirements were imposed on the process, which c included training and qualifications, instrumentation, procedures, records, and audits and surveillances. These measures, along with quality control methods for data collection, were implemented to assure data quality.-
A comprehensive report of the results of the TNP site characterization was prepared and made available for review by the NRC. A description of the TNP site characterization results is incorporated into this section, similar to that previously provided in the approved TNP Decommissioning Plan, Section 3.1.
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TROJANLICENSE TERMINATIONPLAN 2.2 FACILITY RADIOLOGICAL STATUS p
v :2.2.1 FACILITY HISTORY
. 2.2.1.I' - Operatinn Rh ,
l TNP achieved initial criticality in December 1975 and began commercial operation in May 1976.
The reactor output was licensed at 3411 MWt with an approximate net electrical output of 1130 MWe. TNP shut down for the last time in November 1992, because of a steam generator !
tube leak precipitated by a failed sleeve. Commercial operation was formally discontinued in January 1993, after approximately 17 years ofoperation. The plant operated for 14 fuel cycles and approximately 3300 effective full power days. ;
l 2.2.1.2 Radiolonical Historv i l
2.2.1.2.1 Effluents i i
TNP operation resulted in limited release of radioactive material through two main pathways: )
gaseous and liquid effluents. The plant routinely monitored these releases. Monitored gaseous, or airborne, pathways during power operation included the Auxiliary / Fuel Building exhaust, Containment Building exhaust (purge), and condenser offgas system exhaust. Other potential j airbome release pathways, not specifically monitored, included the main steam relief valves, steam packing exhauster blower discharge, and turbine building exhaust. Liquid discharge
' pathways included liquid radwaste discharge, steam generator blowdown, and Turbine Building sumpfoily water separator. Effluents were quantified and reported to the NRC in the Semiannual Radioactive Effluent Release Reports. Tables 2-1 and 2-2 contain information on gaseous releases (noble gases, iodines, and particulates) and Table 2-3 provides information on liquid releases by calendar quarter.
2.2.1.2.2 Operational Events-PGE conducted a review of operational events to determine which could potentially impact decommissioning. Primary sources used to determine TNP radiological history included corrective action system documents (e.g., event reports, radiological event reports, and nonconformance reports) and reports to the Atomic Energy Commission, NRC, and the State of Oregon (e.g., licensee event reports). Interviews were also conducted. A discussion of several notable events follows.
Between 1981 and 1982, fuel assembly damage occurred during plant operation which resulted in loose fuel pellets being released into the reactor coolant system (RCS). The damage was -
attributed to mechanical wearing of the fuel cladding (water-jet-induced vibration or baffle !
. jetting) caused by excessive gaps in the lower internal baffle plate joints.
I Fuel assembly damage resulted in high levels of transuranic radionuclides being found in plant systems during operation. Although this did not result in radioactivity releases greater than TNP
- Technical Specifications limits, consideration of the implications was necessary while 1'
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TROJANLJCENSE TERMINATIONPLAN I-j- developing the site characterization survey plan. Bafflejetting was eliminated with
! .( modifications to selected fuel assemblies, peening of baffle plate gaps, and an upflow
( modification to the lower vessel intemals. No further bafflejetting damage was observed.
The fuel assembly damage necessitated implementing special measures to prevent spreading l ' discrete radioactive particles created by these fuel fragments. However, approximately five years later, high levels ofloose surface and airborne contamination occurred while removing the
! reactor head (during stud hole cleaning). Subsequent investigation revealed a partial fuel pellet on the reactor vessel flange and pellet fragments in the lower refueling cavity area. During stud L hole cleaning, air dispersed the fuel pellet ash, contaminating many surfaces in the Containment, Fuel, and Auxiliary Buildings, i
Since this event occurred, much of the contamination has been remediated. Prior to remediation, ;
many areas of TNP required special radiation protection technician coverage and additional protective clothing when accessing locations where discrete radioactive particles could be found.
As a result, surfaces in the Containment, Auxiliary, and Fuel Buildings are assumed to be potentially contaminated and will require, as a minimum, a wipe / wash down to remove loose ;
surface contamination. PGE recognizes the potential for discovering additional discrete i radioactive particles during decontamination and dismantlement activities, and will implement !
enhanced monitoring techniques (e.g., increased surface area swipes) to ensure detection. l 1
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- Releases of radioactive material from the RCS to secondary systems of the plant occurred through steam generator tube leaks during several operating cycles. Leaking tubes were l
! identified in 1978,1979,1981, and 1992. Radiological monitoring of the condenser offgas also i
indicated continued low level leakage from one or more steam generators from cycle 3 until the final shutdown of TNP. Plant shutdowns to correct the primary-to-secondary leakage were made ,
to ensure the plant complied with TNP Technical Specifications operating and effluent release i limits. Primary-to-secondary leakage resulted in some contamination of secondary systems. ;
- Areas and systems affected by leakage were included in routine surveys and the site radiological l characterization survey. Normal leakage from secondary systems also resulted in the contamination of areas outside the Radiologically Controlled Area (RCA), including the Turbine and Condensate Demineralizer Buildings. Radiological controls were established to prevent I spreading contamination.
Although these events were noted, minimal or no offsite radiological consequences resulted from them. In each instance, POE implemented actions to remove and control contamination and instituted corrective actions to prevent recurrence. These events were used to select additional sampling locations during site characterization. Appendix 2-1 contains a description of several TNP radiological contamination events.
2.2.2 RADIOLOGICAL STATUS OF TNP TNP site characterization is being completed in two phases: 1) Phase I, scoping survey / site i characterization; and 2) Phase II, radiological surveys to support TNP dismantlement and decommissioning.
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TROJANLICENSE TERMINATIONPLAN i Phase I, which is complete and compiled in the " Trojan Nuclear Plant Radiological Site l Characterization Report," Revision 0.1 dated February 8,1995 (Reference 2-10), was used to characterize the radiological status of the facility; estimate the site source term and isotopic
- mixture to support decommissioning cost estimates and decision making; determine the location and extent of contamination outside the RCA; and collect background information to help facilitate release of the site for unrestricted use. Phase II is ongoing and involves routine l radiological surveys in support of PGE's current 10 CFR 50 license. Phase II will be used to help support facility decontamination and dismantlement. Phase II will continue using the existing radiation protection program and procedures. Areas that were not, or could not be, surveyed during Phase I have been or will be surveyed during Phase II.
l Phase I methodology and survey results are summarized in this section. The discussion is divided into four general areas: structures, systems, activation, and environment. An estimate of -
l the disposal volume is also provided for contaminated / activated systems, structures, and components.
i The total waste volume and activity discussed in this section does not hiclude the following l materials since these materials are not considered decommissioning waste:
- 1. Nuclear fuel;
- 2. Control rod elements;
- 3. Incore instrumentation hardware installed in fuel elements; and
- 4. Radioactive fluids, filter media, and resins contained in piping, equipment, sumps, etc.
l Nuclear fuel, control rod alements, and incore instrumentation hardware installed in the fuel elements will be stored in the ISFSI.
A calculated radioactivity inventory, excluding the items listed above, is fumished that includes both contamination and activation. Various mixtures of radionuclides were evident in survey samples throughout the affected areas. Predominant nuclides are shown in Table 2-7. These nuclides are typical of those found in pressurized water reactor plants and are similar to those discussed in NUREG/CR-0130, " Technology, Safety and Costs of Decommissioning a Referenced Pressurized Water Reactor"(Reference 2-11). This was expected since TNP was the i reference plant used for NUREG/CR-0130.
p Site characterization survey maps are included in Figures 2-1 through 2-36. These maps show l radiation protection survey data collected during the first quarter of 1994. This data is a
! " snapshot" of the radiological conditions during the survey period. Normally accessed locations are surveyed periodically for radiation / contamination. Survey instrumentation included the following'.
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- 1. Static and scan alpha surface contamination measurements: Eberline ESP-2 f i
instrument with an Eberline AC-3-8 alpha scintillation probe; L
O- 2. ' Static and scan beta-gamma surface contamination measurements: Eberline ESP-2 instrument with a National Nuclear BP-100 beta-gamma scintillation probe; and
- 3. Gamma exposure ratesi Eberline ESP-2 instrument with a SPA-8 gamma -l scintillation probe and/or a Reuter-Stokes RSS-112 pressurized ion chamber. l To assist the reader in locating specific components or areas while reviewing survey data, equipment locations for major TNP plant equipment are provided in Section 3, Figures 3-2 through 3-9. Reflecting plant configuration at the time ofpermanent shutdown, these figures are provided for information only. The equipment shown on these figures may have been deactivated or removed.
The following table summarizes the scoping and characterization survey results for structures and systems as discussed in the following sections. Environmental survey results indicated no l activity above release criteria.
Section Activity (Ci)
Structures 0.031 i
Systems 1070.5' Activation 4.2x10 6b 6
TOTAL 4.2x10
- Not including -
steam generators, pressurizer, or activation.
b y,,,,,,;y;ty is contained in the vessel internals. Activation curies for reactor vessel, clad, insulation, and concrete are approximately 3.1x10' Cl.-
2.2.2.1 Structures Structures were surveyed to determine contamination levels found in TNP buildings.
Operational radiation protection survey data was supplemented by additional surveys to determine the presence and/or level of contamination. The survey focused on areas outside the ,
present RCA to determine remediation needed to release the areas for unrestricted use. j Structures with known contamination were surveyed to characterize the extent of contam~ mation, ,
l including area and depth of contamination penetration. External system surfaces were considered structures for the purpose of this survey.
Using TNP radiological history, biased surveys were conducted to quantify radioactivity based l l On suspected, or known, contamination at a given location. Contamination in excess of the
! cleanup criteria was identified in a limited number of biased survey locations, including Turbine )
L Building locations where primary-to-secondary leakage caused fixed contamination to build up L in floor concrete.
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Unbiased locations of unaffected areas were selected based on random selection of sampling !
locations within areas of TNP where radioactivity above background was not expected, i Unbiased areas included office buildings, Turbine Building, Maintenance Building, etc. Only i two of the sample points had detectable removable contamination levels. At these locations, the l beta-gamma contamination was below 1 percent of the Regulatory Guide 1.86 contamination l limit and the alpha contamination was below 40 percent of the limit.
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Contamination includes both removable and fixed radioactivity. Removable contamination will be decontaminated through simple means such as wiping or mopping. Fixed contamination appears to be deposited in the upper I cm of the concrete and can be removed using surface l destruction techniques (e.g. scabbling). Table 2-4 contains estimates of volume and radioactivity contained in structures requiring remediation. Fixed contamination levels were not measurable in many biased survey room locations (e.g., Fuel and Auxiliary Buildings) because 1 system radioactivity masked the surface contamination. Additional surveys for fixed contamination will be performed as radiation levels are reduced by decay or removal of the radiation source.
Radioactive waste volume estimates for each building elevation were calculated based on the total area of the elevation, the estimated percentage of contaminated area greater than Regulatory Guide 1.86 limits and an assumed penetration depth of I cm. The estimated area of contamination was based on historical data for spills and industry experience and knowledge.
Additional data will be collected, as necessary, to ensure the assumptions used are accurate and l conservative. l An estimate of containment concrete volume requiring disposal was made through sampling.
Sample results indicate that structural surfaces in the containment may require removal of the upper 1 cm layer. Additional material removal may be required due to neutron activation of concrete constituents. Samples were collected from four locations within the Containment Building, including the primary shield wall, reactor head missile shield, B steam generator wall, and the Containment wall. The concrete core bores taken from four locations in Containment i indicate that removal of at least 3 fl of the primary shield wall surrounding the reactor vessel will )
be required to comply with the site cleanup criteria. In the Containment core bores, detectable j radioactivity was found at various depths. !
As summarized in Table 2-4, the total estimated radioactivity on structural surfaces attributed to contamination was approximately 0.031 Ci. Survey maps showing RCA radiological conditions are also included at the end of this section as Figures 2-13 through 2-36. A summary of radiological conditions for each building, by elevation, is contained in Table 2-5 and a description summarizing the results is contained in Appendix 2-2. Removable contamination, 4 general radiation levels, and maximum contact dose rates are identified. l l
2.2.2.2 Systems Systems were surveyed to determine contaminated systems and estimate the quantity of i contamination. Each plant system was evaluated for its likelihood to be contaminated and 4
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TROJANLICENSE TERMINATIONPLAN 3 sampled by direct survey, loose surface swipe, or metal scrapings. Detected activity that could O not be identified as naturally occurring was attributed to plant operations and the system was classified as contaminated. The approach involved grouping plant systems into four categories:
' Cl or contaminated; C2 or potentially contaminated due to cross contamination; I or indeterminate (need more data); and N or not cantaminated.
Samples of systems from these classifications were collected. The samples were predominantly taken fmm systems classified as C2, I, or N to determine the extent of plant system contamination. Less attention was directed at contaminated systems since data was collected for these systems during plant operation and was still representative.
Systems with potential internal contamination were also sampled as part of the site characterization scoping survey. Some systems which were classified as N, and were expected to be only extemally contaminated, were not sampled (as mentioned previously, the external surfaces of a system were considered structures).
. If a system is sufficiently contaminated, its curie content can be estimated by measuring the dose rate from its piping. For systems classified as C1, plant surveys determined the dose rate levels for each system. Conservative values (i.e., maximum average dose rate) were assumed for each system to provide a bounding estimate of curie content. Dose rate and pipe size were used to calculate the activity deposition (Ci/m'). The result was multiplied by the total contaminated surface area of the system to conservatively estimate the system's curie content.
For potentially contaminated systems (i.e., C2 or I), it was not possible to approximate activity O deposition by field dose rate measurements. Scrapings from the system were used, or an estimate was assumed, to determine activity deposition and curie content.
Where ambient radiation areas and physical configuration allowed, the surface activity of systems was determined by direct surveys with beta-gamma detection equipment. Direct surveys were not performed on systems in high ambient radiation areas or where physical configuration prevented the survey.
Swipes were taken to measure removable surface contamination and scrapings were collected to determine fixed surface contamination. Swipes and scrapings were analyzed in a low background area. Some swipes were limited to one location because of physical restrictions, while others were composite swipes used to check large areas and/or various individuallocations.
Alpha contamination at TNP has been detected only when accompanied by detectable beta-gamma activity. Consequently, only samples that showed detectable beta gamma activity were typically counted for alpha contamination.
Individual system scrapings were counted in the laboratory to determine a qualitative radionuclide spectrum. Fixed and removable contamination was found in the contaminated systems. The total radioactivity is not expected to be substantially reduced through nonaggressive decontamination methods. Operational experience, during activities such as 2-8 March 1999
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. . f steam generator primary bowl hydrolasing, indicates contamination is tightly adhered to surfaces {
and will probably require component disposal. '
O. Potential burial waste volumes and surface activities by system are contained in Table 2-6. The I
total system burial volume is estimated at 215,789 ft'. The total surface activity is l conservatively estimated to be 1070.5 Ci. The volume and activity estimates assume that the <
four steam generators and pressurizer are disposed of as pan of the large component removal !
. project. The volume estimate was based on a model developed by TLG Services, Inc (TLG). I The estimate includes the volume of extemally contaminated systems (e.g., electrical cable, !
conduit, piping) not sampled as part of the scoping survey. I i
2.2.2.3 Activation Plant component activation occurred during normal plant operation due to neutron irradiation. l Estimates ofplant component activation were made using operational data. TLG performed neutron transport calculations using TNP specific data. Measurements were made to verify the accuracy of the calculations.
Calculations for components activated by neutron irradiation consist of one dimensional neutron transpon and point neutron activation analyses. Calculations indicate the reactor vessel, vessel internals, and concrete shielding have levels of radioactivity that will require remediation. These calculations were performed using TLG's FISSPEC and 02 FLUX computer codes and the ANISN and ORIGEN computer codes obtained through the Oak Ridge National Laboratories Radiation Shielding Information Center. Ancillary calculations were performed using TLG's ANISNOUT and 02 READ computer codes and Microsoft's EXCEL computer program.
The one dimensional neutron transport model was normalized with data from a -
Westinghouse Electric Corporation report, " Analysis of Capsule V from the Portland General Electric Company Trojan Nuclear Plant Reactor Vessel Radiation Surveillance Program." The radionuclide inventories evaluated in these analyses were for the four major structural material compositions including Type 304 stainless steel, pressure vessel carbon -
steel, concrete, and plate /rebar carbon steel.
A listing of 10 CFR 61 classification by component, one and five years after shutdown, is l included in Tables 2-8 and 2-9. One year following shutdown, the radioactivity content of i 6
activated components was estimated at 4.2 x 10 Ci. Five years following shutdown, the 6
calculated activity of the activated components was approximately 2 x 10 Ci. Predominant !
radionuclides include 55p,, 6 Co, and Ni. I During 1999, PGE intends to remove the reactor vessel with intemals intact (reactor vessel
. package) from the 10 CFR 50 licensed area of the TNP site. As authorized by NRC letter dated October 29,1998 (Reference 2-12), the reactor vessel package will be transported for disposal at the US Ecology low level radioactive waste facility near Richland, Washington. Removal of the !
reactor vessel package from the 10 CFR 50 licensed area of the TNP site will eliminate approximately 2 million curies of activity from the TNP. Not including the spent nuclear fuel O
2-9 March 1999
TROJANIJCENSE TERMINATIONPLAN thit will be transferred to the ISFSI, removal of the reactor vessel and internals will result in ,
removal of greater than 99 percent of the remaining activity (curies) at the TNP facility.
2.2.2.4 Environment
- l. The environmental survey, which included representative outdoor areas, focused on the impact of TNP operation on the environment due to the release of radioactive material. Operational and preoperational environmental monitoring data were used to measure and evaluate the impact.
l Additional sampling was conducted to augment, or better define, areas requiring biased surveys.
Survey results were compared to background data to determine the overall consequences of TNP-operation.
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- During Phase I of site characterization, soil, sediment, and surface water were sampled.
- l. Exposure rates were measured wherever soil was sampled, except where exposure rates were
! influenced by onsite structures. Paved areas onsite were scanned for beta contamination or sampled and analyzed for gamma emitters. A general site map, Figure 2-1, shows the site divided into zones. Survey maps depicting sample points by grid location are provided in l Figures 2-2 through 2-12. Survey maps were not included for zones where no samples were collected (i.e., Zones 4,12,13,15, and 16).
l Biased sample locations were determined from reviewing plant records that documented radiological events at TNP from 1975 to 1993 (see Section 2.2.1.2.2 and Appendix 2-1).
Corrective action programs were reviewed and interviews conducted with PGE personnel to :
! help determine potential sample locations.
2.2.2.4.1 Surface Soil Suney Surface soil samples were obtained at locations on PGE property contiguous with TNP. The l
locations of the soil samples are shown in Figures 2-2 through 2-12. Each sample contained 2 '
approximately 1 liter of material collected from a 1 ft area. Soil samples were analyzed in-house for gamma emitters by gamma spectrometry and control samples were analyzed bg TMA/Eberline in Albuquerque, New Mexico. Selected samples were also analyzed for Sr.
l Background soil samples were collected from four locations around T1 . Two samples were
! used as quality control checks and were not included with the data. The four background locations were:
- 1. PGE owned property in Prescott, Oregon (approximately 0.75 miles north-northwest of TNP containment);
- 2. Water treatment facility in Rainier, Oregon, near radiological environmental sample location 2 (approximately 3.8 miles northwest of TNP containment);
i 3. ~ PGE owned property west of Highway 30 (approximately 1 mile west of TNP containment); and l
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TROJANLICENSE TERMINATIONPLAN l 4. . Northwest of Kalama, Washington, near radiological environmental sample s >
location 11B (approximately 1.4 miles east-northeast of TNP containment).
For soil background measurements, the mean background '37Cs concentration was 0.49 pCi/g with a standard deviation of 0.4 pCi/g and a range of 0.01 to 1.3 pCi/g. Substantial variation in
, background '37Cs concentrations was obcerved between varying soil types. Sandy soils found
! near the river contained low '37Cs concentrations, while clay soils contained higher concentrations.
For the survey of unaffected soil areas, the mean '37Cs concentration was 0.77 pCi/g with a standarti deviation of 0.86 pCi/g and a range of 0.01 to 2.94 pCi/g. Primarily, the nonnaturally occurring isotopes found in soil samples were '37Cs and "Sr. Fallout from atmospheric weapons tests and the Chemobyl accident are the major sources 337 of Cs and "Srin the environment.
"Sr results averaged 0.2 pCi/g with a standard deviation of 0.16 pCi/g and a range of 0.02 to l .0.32 pCi/g. The Sr levels nrasured during the preoperational period ranged from 0.01 to 1.28 L pCi/g with a mean of 0.30 pCi/g.
- Biased survey soil samples were taken onsite where potential soil contamination may have j occurred. Subsurface soil samples taken in 1991 from the tank farm area were also reviewed as l part of the analysis. Samples were taken at 1,2, and 3 ft depths at 5 locations.
The predominant nonnaturally occurring isotope detected was '37Cs. One surface sample, taken l from the tank farm area, also contained low levels of '3dCs (0.010 pCi/g) and "Co (0.044 pCi/g). l The mean value for '37Cs in affected soil samples was 0.10 pCi/g with a standard deviation of :
0.098 pCi/g. The '37Cs content of the 1991 samples was below the cleanup criteria.
i l . 2.2.2.4.2 Water Survey Ground Water The TNP site is located on an impervious rocky ridge that is bounded on one side and end by the Columbia River and on the other side by an old river channel that has been filled with impervious sediments. The rock on the ridge is moderately fractured, but thejoints have been sealed by impervious materials so that there is no apparent leakage of water from many rain-filled foundation and footing holes over a period of several montle. It is apparent, therefore, that ground water does not have the significance at the TNP site that it might have at others where the plant is less isolated from adjacent ground water supplies and where the soil is permeable.
Of the few domestic wells that exist in the area, water samples are periodically collected from i selected wells as part of the TNP Radiological Environmental Monitoring Program. Levels of l~ tritium and gamma-emitting radionuclides in well water samples consistently have been found to
' be below the minimum sensitivity requirements of the sampling program.
i For the above reasons, ground water sampling was not conducted as an integral part of the TNP site characterization. However, the results of the Radiological Environmental Monitoring ;
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TROJAN 12 CENSE TERMINATIONPLAN Program annual surveys a:e submitted to the NRC annually as required by the provisions of the
' TNP license.
' Surface Water-c Surface water was sampled from indicator sites on PGE property surrounding TNP. A 1 gallon
~
sample was obtained from each site for gamma and "Sr analysis and a 60 mi sample for tritium analysis. The water samples were analyzed for gamma emitters using a gamma spectroscopy .
system located onsite. Water samples were analpd for tritium in the onsite counting .
laboratory. Selected samples were analyzed for Sr.
To determine background, water samples were collected from four locations around TNP. The locations included:
- 1. Fishhawk Lake (approximately 18 miles west of TNP containment);-
- 2. Ponds at the intersection of Goble and Bishop Roads (approximately 3 miles -
southwest of TNP containment);
- 3. Kress Lake (approximately 1 mile east-northeast of TNP containment); and
- 4. Deer Island ponds (approximately 7 miles south of TNP containment). i Analyses for gamma emitters and tritium were completed on the samples. No gamma emitters other than naturally occurring radionuclides were identified in the samples. Tritium values were less than detectable. The four samples analyzed for "Sr were less than detectable. Minimum detectable activity (MDA) for '37Cs, tritium, and "Sr was approximately 4,450, and 0.3 pCi/1, '
respectively. .
For the survey of unaffected water areas, samples were collected from random locations in Whistling Swan and Reflection Lakes located on PGE-owned property surrounding the TNP site. :
No nonnaturally occurring radionuclides were detected in the samples by gamma spectrometry.
Neither tritium nor "Sr was detected in the samples.
For the biased survey, samples were taken from the potentially affected Recreation Lake, also ,
located on PGE-owned property surrounding the TNP site. No nonnaturally occurring ;
radionuclides were detected in the samples. MDA's for the biased and unbiased survey analyses were the same.
2.2.2.4.3 Bottom Sediment Survey Bottom sediment samples were taken from PGE property around TNP. Approximately I liter of sediment was obtained at each sampling site. The sediment samples were dried and analyzed for gamma emitters using a gamma spectroscopy system located onsite. Selected sediment samples were analyzed for "Sr by TMA/Eberline.
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- Specific isotopic background sediment samples were not collected. Instead, soil background O results were used as sediment background. Background soil samples were analyzed as part of the site characterization effort, and the mean 337 Cs concentration was 0.49 pCi/g. A comparison of the '7Cs concentration in preoperational sediment samples to the background soil samples showed a high conelation with the sediment mean equal to 0.51 pCi/g and the soil mean equal to 0.49 pCi/g.
In conducting the survey of unaffected sediment areas, samples were taken from Whistling Swan and Reflection Lakes.~ The mean value for '37Cs was 0.36 pCi/g with a standard deviation of 0.22 837 pCi/g and a range of 0.02 to 0.86 pCig The unaffected area sediment samples contain Cs at levels below the release value for Cs. "Sr content of the two sediment samples sent to TMA/Eberline were 0.05 and 0.03 pCi/g. The lower level of detectability for the "Sr analysis was 0.02 pCi/g.- These results are within theycoperational range of"Sr which was from 0.01 to 0.44 pCi/g with a mean of 0.08 pCi/g. The Sr content of the sediment samples was also below the corresponding screening release level.
For the biased sediment survey sample population, samples were taken from the berm and main areas of Recreation Lake. Results of the analyses indicate a mean of 0.28 pCi/g with a standard deviation of 0.37 pCi/g and a range of 0.04 to 1.12 pCi/g. The affected area samples contain
7 Cs in amounts below the release level. No other gamma emitters wear detected.
2.2.2.4.4 Pavement Survey Pavement scans and sampling were performed. Pavement was scanned for beta contamination.
O 2 In areas where there was interference from the RWST, a 1 ft sample was collected and analyzed using a gamma spectroscopy system located onsite.
No specific background pavement locations were monitored for this survey. Sample locations located in the TNP park and recreational areas were used to estimate background levels. Since these areas were unaffected by TNP operation, the survey data for these locations was determined to be an acceptable estimate of background levels of radioactive material in 2
pavement. The mean gross beta reading was 610 dpm/100 cm with a standard deviation of 94 2 2 dpm/100 cm and a range of 456 to 764 dpm/100 cm ,
2 For the survey of unaffected pavement areas, randomly selected 100 ft sections of pavement in other areas of the TNP site which were unaffected by operations were scanned with an ESP-2 and BP-100 detector. 'Ihe mean value was 657 dpm/100 cm2 with a standard deviation of 2
74 dpm/100 cm The range of measurements was from 542 to 788 dpm/100 cm 2, For the biased pavement survey, the affected areas consisted of pavement around the tank farm and its drainage to the west, pavement around the oily water separator, and the paved equipment laydown area sivund the cooling tower. Pavement samples were taken from affected areas with at least two . samples from each affected area. The onig7 detectable nonnaturally occurring radionuclide found in the pavement samples was Cs in low concentrations. The results of the biased samples exhibited a mean of 0.16 pCi/g with a standard deviation of 0.40 pCi/g and a range of 0.019 to 1.5 pCi/g. '37Cs content of the biased pavement O
2-13 March 1999
F7 TROJANLICENSE TERMINATIONPLAN l:
l samples was similar to that found in background and indicator soil samples obtained for site characterization. One sample, taken from the curb at the southeast corner of the circulating 37
.O water pump pit area, had the3 hipest Cs concentration of 1.5 pCi/g. For comparison, conservatively assuming the Cs was from the top I cm of the concrete and covered a 100 cm2 I
area, then the calculated contamination level wo.ild be 799 dpm/100 cm2, 2.2.2.4.5 Exposure Rate Survey Exposure rates were measured at locations where affected and unaffected site characterization indicator soil samples had been collected. The measurements were made with a Reuter-Stokes pressurized ion chamber instrument positioned 1 meter above the sample site.
Data for exposure rate backgmund was collected during preoperational surveys at TNP using a high pressure ion chamber, the same type ofinstrument used during the site characterization survey. The preoperational mean reading was 7.1 pR/hr with a standard deviation of 1.0 pR/hr and a range of 5.6 to 9.4 pR/hr. The survey locations coincide with the Radiological Environmental Monitoring Program locations.
For the exposure rate survey of unaffected areas, surveys were taken at the unaffected soil sampling locations. Exposure rates ranged from 5.2 to 9.0 R/hr at the unaffected area locations.
The mean exposure rate was 6.4 pR/hr. Data compared favorably with preoperational data, indicating no effect from TNP operation.
For the biased survey, exposure rates were measured at affected area soil san.ple sites where it was determined that radioactive content of surrounding structures would not influence the measurements. Measurements made at two locations were influenced by the RWST and were not included. Exposure rates at four locations were not measured because of radiation levels from the RWST. Exposure rates at two locations were not measured because ofradiation levels frora the Low Level Radioactive Waste Storage and Fuel Buildings. The values at the remaining locations ranged from 6.0 to 8.3 pR/hr with a mean of 6.8 pR/hr. This is consistent with background data.
2.
2.3 CONCLUSION
In summary, several general overall conclusions regarding the site characterization survey can be made about the four sections: structures, systems, activation, and environment.
First, plant structures contain radioactive material that will require removal prior to license termination. The contamination consists of radioactive material incorporated (fixed) into the upper layer of concrete / block and deposited on the surface (loose). Although the levels of radioactivity are genenally low, structures within what was the RCA in 1993, including building surfaces and piping, are considered potentially contaminated and will require, as a minimum, a wipe or wash down.
Second, some plant systems contain deposited radioactive material due to plant or:Mion. The majority of the radioactive meterial is contained in RCS piping and systems directly connected to 2-14 March 1999
2 TROJANUCENSE TERMINATIONPLAN
, the RCS (e.g., chemical and volume control system [CVCS), safety injection system, and residual heat removal [RHR] system)~. Although some systems contain contamination, the
% systems are not expected to be greater than Class A waste.
- . 'Ihird, activated components contain the vast majority of the radioactive material not contained L in fuel. Most activity is primrarily concentrated in the vessel intemals and shield wall. The reactor vessel lower internals contain the highest activity. Although radionuclide distributions are pmvided for the reactor vessel and vessel internals, they will have been removed before final survey data collection begins in the Containment. Neutron activation products have been found
- in samples of containment concrete in various structures, including the reactor vessel shield wall, a
steam generator missile shields, and the containment wall itself. Remediation of the activated components will be required to meet the site release criteria and facilitate license termination.
Fourth, and finally, the environmental survey results indicated that no' radioactive material requiring remediation is present in the various materials sampled, and that no radioactivity
, requiring remediation has been spread to the environment outside the TNP industrial area. The final survey may require additional background data for a number of the sample media.
l Preliminary results indicate no radioactivity at TNP has been spread to the environment inside i the industrial area in quantities requiring remediation.
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TROJANUCENSE TERMNADONPLAN 2.3 OUALITY ASSURANCE PRACTICES AND PROCEDURES TNP site characterization activities are conducted under the auspices of PGE-8010, " Trojan Nuclear Plant Quality Assurance Program," utilizing existing QA procedures.- This QA program complies with 10 CFR 50, Appendix B (Reference 2-13), and is incorporated into the TNP D=m=issioning Plan, Section 7. Quality-related items and activities that are controlled under the QA Program, as well as organization responsibilities for program implementation such as -
review and audit responsibilities, are dermed in PGE-8010.
TNP's Nuclear Quality Assurance Program ensures that survey activities are performed in a manner that assures the results are accurate and that uncertainties have been adequately considered. Surveys are perfonned by trained individuals who follow standard written procedures and are using properly calibrated and source-checked instruments. The custody of samples is tracked from collection to analysis, with every step of the process documented in a way that can be' audited. In addition, QA practices ensure that offsite laboratory analyses are conducted using approved Radiological Environmental Monitoring Program (Reference 2-14) procedures. Finally, characterization data, as well as calibration and source check documentation, are maintained as quality-related decommissioning record.4 and are handled in accordance with approved plant procedures.
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g TROJANLICENSE TERMINATIONPLAN
2.4 REFERENCES
FOR SECTION 2 .
2-1 Code of Federal Regulations. Title 10, Pe.st 50.82, " Application for Termination ofLicense."
2-2 Regulato; Guide 1.179," Standard Format and Content of License Termination Plans for Nuclear Power Reactors," January 1999.
2-3 . Portland General Electric Tonical Report PGE-1061. " Trojan Nuclear Plant Decommissioning Plan," Revision 6.
2-4 Regulatory Guide 1.86," Termination of Operating Licenses for Nuclear Reactors," June 1974.
2-5' NUREG/CR-5849," Manual for Conducting Radiological Surveys in Support of License Termination," June 1992.
2-6 Draft NUREG/CR-5512," Residual Radioactive Contamination from Decommissioning," January 1990.
2-7 Code of Federal Renulations, Title 10, Part 20.1402, " Radiological Criteria for Unrestricted Use."
2-8 NUREG-1575, " Multi-Agency Radiation Survey and Site Investigation Manual
'(MARSSIM)," December 1997.
2-9 Ppr' ed General Electric Tonical Report PGE-8010, " Trojan Nuclear Plant Quality ' Assurance Program," Revision 22.
2-10 . Portland beneral Electric " Trojan Nuclear Plant Radiological Site Characterization Report," Revision 0.1, February 8,1995
.2-11 NUREG/CR-0130," Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station," June 1978.
2-12 NRC Letter. W. F. Kane to S. M. Ouennoz, " Authorization of the Trojan Reactor Vessel Package for Transport," October 29,1998.
2-13 Code of Federal Renulations, Title 10, Part 50, Appendix B," Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants."
2-14 Portland General Electric Tonical Report PGE-1021, "Offsite Dose Calculation Manual," Amendment 17.
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l TROJANLICENSE TERMINATIONPLAN i i
- Appendix 2-1 .~ j Summary of Notable Radiological Contamination Events l l
Datt: Various
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Description:
Spills of radioactive liquids on the 45 ft elevation of the Containment Building occurred on !
several occasions during plant operation. Events include draining reactor coolant through a j steam generator manway during maintenance. Other spills occurred during leaks from l miscellaneous valves. -i Radiolonical Conseeuences: ;
1 The radioactive water may have caused contamination of the concrete surfaces in the area. ]
Fixed contamination requiring remediation may be present in the area. Additional l'
! samples /sury ys, including concrete samples, are planned for this area.
Rats: Various l
Description:
The storm drain systems which discharged to Recreation Lake and the Columbia River were l contaminated from a number of sources. For example, the oily water separator and start-up .
- O boiler, which are addressed later in this appendix, both contributed to past storm drain contamination. Other contamination sources included leaks in the electrical facade, contamination of building roof surfaces from main steam relief and air ejector discharges, and rain " rinse out" of activity releases from the PWST and RWST vents, which are also discussed laterin this appendix.
1 Radiolonical Consecuences:
Elevated levels of tritium on the river side drains, and "7Cs,5'Co, and "Co in sludge from the drains, were identified several years ago. Storm drains are still considered potentially I contaminated from past discharges of slightly radioactive liquids. The drains are too small i to allow access for surveys; however, they may be addressed under Phase II of site l characterization or as part of the embedded pipe survey program. Samples of the water and sediment from the drain discharge at Recreation Lake were collected as part of the scoping ;
j_ survey and the results indicated no activity above background ~ l
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!O Appendix 2-l' 1of5 March 1999 a-
g TROJANllCENSE TERMINATIONPLAN Rait: 1975 to early 1980s
Description:
Steam generator primary-to-secondary leakage resulted in contamination of the secondary system, which was sampled in the secondary chemical laboratory. Secondary chemical laboratory drains were routed to the wastewater treatment plant, hqarag:
Contaminated water was discharged to the waste treatment plant. The water from the treatment plant was sampjed and then discharged to the Columbia River. Discharges were included in the Semiannad Efiluent Release Reports. The system was modified to direct the drains to the oily wate:r seprestor which was a monitored release pathway.
Dats: 1975 - 1979
Description:
Based on oral interviews, it appears, that on several occasions, a radwaste system tank over6wed to the Auxiliary Building floors / sumps prior to 1980. The floors and walls were decontaminated (removable activity) and returned to access without protective clothing.
Radiolonical Consecuences:
These events may have resulted in fixed contamination that will require remediation to meet the final survey release criteria. Operational survey data will be collected to determine the extent of contamination in this area.' The radioactive material was confined to the building.
Date: 1978 - 1980
Description:
The plant start-up boiler was contaminated by primary-to-secondary' leaks The start-up boiler used makeup water from the condensate storage tank. The condensate storage tank was contaminated due to primary-to-secondary leakage.
Start-up boiler blowdown was discharged to the pavement around the start-up boiler. The blowdown flowed to the storm drain and Recreaticn Lake. Contamination was found in the boiler and on the pavement near the boiler blowdown line. A plant modification changed the makeup source to demineralized water and rerouted blowdown to the discharge and dilution structure.
Appendix 2-1 2 of 5_ March 1999
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TROJANLICENSE TERMINATIONPLAN e
Radiological Consecuences:
Blowdown from the start-up boiler caused contamination of the curb going to the storm drain and Recreation Lake. A concrete slab was removed and replaced with new concrete.
The releases did not exceed effluent release limits. Scoping surveys indicated that these ar,m (i.e., pavement, start-up boiler, and Recreation Lake) do not contain radioactivity levels greater than background.
Eg!e: Prior to 1980
Description:
The CVCS was over-pressurized releasing contaminrtc! water to the 77 ft elevation of the Auxiliary Building.
Radiolonical Conseauences:
ne contaminated water sprayed onto untmated concrete and concrete block surfaces. The removable contamination was removed. Levels of fixed contamination remained in the concirte. The area was repainted and marked to indicate that fixed contamination was present under the paint. The levels of f:xed contamination are not measurable due to the presence of radioactive material that mask the contamination. Operational samples / surveys, including concrete samples, are planned for this area.
Ragg: - October 1980
Description:
Water samples from the Recreation Lake berm contained radioactive material.
This was caused by water from the oily water separator, which collected
. potentially contaminated oily waste water from the Turbine and Condensate Demineralizer Buildings, being discharged to Recreation Lake. The design of the system' directed the water to the lake. He system was modified to direct the oily water separator discharge to the discharge and dilution structure which was the normal release path for liquid wastes. De oily water separator overflowed on at least one occasion following the modification. The overflow spilled over ground to the site storm drains and consequently, to Recreation Lake.
Radiolonical Consecuences:
At the time of the event, radioactive iodine ('331), cesium (337Cs), cobalt ("Co and "Co),
and tritium were found in the water samples in the berm area of Recreation Lake. The releases did not exceed plant effluent limits. More recent site characterization scoping survey results indicate that these areas no longer contain activity levels greater than i backgrcund.
O Appendix 2 3 of 5 March 1999
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TROJANLICENSE TERMINATIONPLAN ,
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, Ratt: 1981 .i Description
- r i The Main Steam Support Structure and the gravel area surrounding the plant tank farm I l
(south of containment) were contaminated when a main steam relief valve opened during a I steam generator hydrostatic test. This occurrence was noted due to steam generator tube leakage.
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- Radiological Consecuences:
2 The release resulted in the contamination of a large' area (>1000 ft ) of the Main Steam i Support Stmeture. The contamination was fixed in concrete. The release did not exceed :
effluent release limits. Operational surveys indicate the need for remediation. !
Da.ts: 1986,1987, and 1989
Description:
Tritium contamination was identified in the sewer treatment plant, storm drains, and ,
Recreation Lake. The source of tritium was attributed to minor flange leakage from the !
RWST and seal leakoff from condensate transfer pumps, which was allowed to spill on the ground prior to nerouting the leakoff. 'Another possible source was from the oily water separator which is located above a sewer treatment system manway. The 1989 tritium j
.Os releases appear to have been caused by condensation from the PWST and RWST atmospheric vents which allow tritium in the air space above the tank to be released to l
i atmosphere. Actions taken to stop the releases included sealing the R.WST flanges and rerouting the condensate transfer pump seal leakoff. j Radiolonical Conseauences:
Tritium levels in the waters released from TNP in the storm and sanitary sewer system were I less than the permissible concentration specified in 10 CFR 20, " Standards for Protection Against Radiation."
Ratt: April 1987
Description:
As previously discussed in Section 2.2.1.2, during refueling activities in April 1987 an !
airbome contamination event occurred that resulted in the dispen.at of fuel / fission product activity in the Containment Building. High levels of removable contamination were found from the 93 ft to the 205 ft elevations. An investigation revealed several partial fuel pellets and pellet fragments on the reactor vessel flange and in the lower refueling cavity area.
O Appendix 2 4 of 5 March 1999 l l
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r- 1 TROJANLICENSE TERMINATIONPLAN Radiolonical Conseauences: .
The Containment Building inner surfaces were dxontaminated to remove the material.
Surveys indicated contamination levels following the incident were consistent with pre- )
event levels. Discrete radioactive particles continued to be found in the plant following the incident. The presence of these particles will require detailed surveys prior to free release of l material from the RCA during decommissioning activities. The contamination is expected to increase the difficulty in decontaminating the materials in the Containment Building for ;
release. l 1
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V.
TROJAN 12 CENSE TERMINATIONPLAN Appendix 2-2 ,
Summary of Structural Survey Results O A summary of radiological conditions and components in plant areas is presented below.
- Descriptions of plant areas and component locations are indicative of conditions during early 1994. In some cases, the description below may differ from the survey maps at the end of Section 2. In part, this occurs because the survey maps are intended to be a " snapshot" of the radiological conditions during the first quarter of 1994, whereas, the individual summaries below may contain more current data, professional insight, estimates, and assumptions. Recognizing that radiological conditions change, as may building and structures configuration, PGE has attempted to provide additional insight via the summary below. If current radiological conditions are needed, the most recent survey maps should be used.
Biased Survey ~Results Structures Within the RCA Buildings in the RCA include the Containment, Auxiliary, and Fuel Buildings, Radwaste Annex, Main Steam Support Structure, electrical facade, and the Low Level Radioactive Waste Storage Building. Because of past problems with discrete radioactive particles, surfaces in the Containment, Auxiliary, and Fuel Buildings are assumed to be affected by plant operations and,
. therefore, potentially contaminated. It is anticipated that the surfaces will require a minimal wipe / wash down to remove loose surface contamination.
Containment Buildinn Samples of containment concrete were collected by core boring at four locstions. The bores were collected from the reactor shield wall, reactor vessel missile shield, secondary shield wall, and containment dome. The samples consisted of 3 inch diameter bores. The bores were segmented and counted for radioactivity using gamma isotopic analysis. The cores were used to validate the neutron activation analysis, to determine penetrating depth of activation products in concrete structures (other than the primary reactor vessel shield), and to estimate the area extent and levels of fixed surface contamination.
Containment Buildinn - 205 ft Elevation The major components on this elevation include the containment air coolers and the polar crane.
The highest general area radiation reading is 1.8 mrem /hr. Removable contamination levels vary 2
from less than ik to 150k dpm/100 cm . Fixed contamination levels are expected to be 1-Sk 2
dpm/100 cm Material on this level consists of structural steel and grating.
O Appendix 2-2 1of13 March 1999
S TROJANLJCENSE TERMINATIONPLAN
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Containment Buildine - 136/105 A Elevation .
O
, This elevation has ventilation units and is primarily a storage area for the missile shields and ,
head ventilation duct work during refueling operations. With no installed equipment, dose rates range from 0.3 to 0.5 mrem /hr general area. . Highest removable contamination level is 2
2k dpm/100 cm . Fixed contamination levels are expected to be' 1-5 k dpnt/ 100 cm2 . Untreated
. concrete and structural steel are located on this level.
Containmant Buildino - 93 A Elevation This elevation is the main refueling floor and the normal access path for containment. Access is 1 provided to the refueling cavities, top levels of the steam generators and pressurizer shed, and to I the seal table for the incore detectors. Storage area for th: reactor vessel head is also provided. 1 Installed equipment includes the incore detector drive boxes, ventilation units, and electrical l panels. General area dose rates range from less than 0.2 to 100 mrem /hr. Highest removable 2 ]
1 contamination levels are on the refueling upnder which has up to 200k dpm/100 cm . Fixed contamination levels of 5-25k dpm/100 cm are estimated. Surfaces on this level consist of i
treated and untreated concrete, structural steel, and grating. Neutron activation of the concrete in l the containment wall, secondary shield wall, and reactor vessel missile shield were determined l
. by core bore analysis. The primary radionuclides identified were "Co and u2Eu. ;
Containment Buildine - 77 A Elevation i
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This elevation provides access to feedwater piping. General area dose rates range from less !
2 than 0.2 to 1.5 mrem /hr. Removable contamination levels range up to 2k dpm/100 cm , ;
2 i Fixed contamination levels of 5-25k dpm/100 cm are estimated. Surfaces on this level !
l consist of treated and untreated concrete, structural steel, and grating.
l Containment '43ine - 61 A Elevation This elevatio :. contains the regenerative heat exchanger, excess letdown heat exchanger, letdown piping, and access te 'he bottom of the pressurizer shed. The regenerative heat exchanger room is not normally accessed. However, based on past surveys, the room will need remediation. The l letdown piping west of the regenerative heat exchanger room has an 800 mrem /hr hot spot. The pressurizer shed has contact dose rates up to 280 mrem /hr and general area dose rates up to 120 mrem /hr. Elsewhere, general area dose rates range from less than 0.2 to.8 mrem /hr. Highest 2
removable contamination levels measured on this level are 60k dpm/100 cm in the pressurizer 2
- shed. Fixed contamination levels of 5-50k dpm/100 cm are estimated for this level. Surfaces on this level consist of treated and untreated concrete, structural steel, and grating.
Containment Buildine - 45 A Elevation l
! This elevation contains the emergency airlock for containment, safety injection accumulators, pressurizer relief tank, reactor coolant drain tank, both recirculation sumps, and primary access l
. to the bioshield area. General area dose rates range from 0.2 to 170 mrem /hr outside of the 2
bioshield. Removable contamination levels range up to 30k dpm/100 cm at the safety injection ,
I Appendix 2-2 2 of13 Marah 1999 I
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l TROJANLICENSE TERMINATIONPLAN line aru near accumulators B and C. Fixed contamination levels of 5-100k dpm/100 cm2 are estimated for this level. Surfaces on this level consist of treated and untreated concrete, (e structural steel, and grating.
Containment Buildine - Bioshield Area
~
nis area contains four steam generators, four reactor coolaut pumps, e.d ccess to the under vessel area. Contact dose rates are up to 1000 mrem /hr on the RTD bypass manifolds. General area dose rates range from 10 to 250 mrem /hr. Removable contamination levels up to 110k '
dpm/100'cm 2are found in the bioshield. De under vessel a rea has not been surveyed since shutdown. The incore detectors are partially withdrawn re. King this area a high radiation exclusion area. Fixed contamination levels of 5-500k dpm/100 cm2are estimated for this level.
Treated concrete and structural steel make up the surfaces on this level. A 3 inch diameter concrete core bore was taken from the reactor vessel shield wall at approximately the 50 ft elevation (corresponding to near centerline of the reactor vessel). The core was segmented and 4 analyzed for gamma ray emitters. Predominant radionuclides identified were "Co, is2Eu,8"Eu, l 155 p Eu, and '"Cs. Radioactivity was detected to a depth of approximately 50 inches (total shield
' l thickness is 102 inches). Close agreement was noted between the calculated neutron activation :
results and the measured activation values. The comparison was good for is2 Eu in the first 3 inch l segment. The calculated value was 0.29 pCi/g while the measured value was 0.25 pCi/g. The !
agreement was not as good for segments farther from the inner wall, although the calculated ;
values were conservative.
Auxiliary Buildine - 104 fl Elevation '
O This c'evation contains the supply and exhaust ventilation filters for the Auxiliary and Fuel l
' Buildings. Dose rates are less than 0.2 mrem /hr. Removable contamination levels are below Ik .
dpm/100 cm2(see 93 ft elevation below). )
Auxiliary Buildine - 93 fi Elevation
< This elevation contains supply and exhaust fans for the Auxiliary and Fuel Buildings, containment purge exhaust unit, containment access point, and access to the filter pits. Dose rates are less than or equal to 0.2 mrem /hr. Dose rates in the filter pits vary by system and age of
- filter. These will be surveyed at filter changes or when other access is required. Removable 2 contamination levels outside of the filter pits are less than Ik dpm/100 cm2 . F xed contamination 2
. at levels greater than 5k dpm/100 cm is estimated at 19% of the surface area. An average fixed contamination level of 10k dpm/100 cm2 is estimated.
Auxiliary Buildine '7711 Elevation Major equipment on this elevation includes the demineralizer valve galleries and access to the
. . demineralizer vaults, radioactive waste filter glove boxes, and two boric acid evaporators. The highest dose rates'are in the demineralizer valve galleries and are up to 350 rmm/hr contact and 2
up to 50 mrem /hr general area. Removable contamination levels are up to 9k dpm/100 cm in the filter valve gallery. Fixed contamination at levels greater than 5k dpm/100 cm is estimated to eO Appendix 2-2 3 of13 March 1999
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TROJANLJCENSE TERMINATIONPLAN 2
exist'over 19% of the surface area. An average fixed contamination level of 10k dpm/100 cm ;,
n1- e 9 - Auxiliary Buildine - 61 ft Elevation g l
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This' elevation contains the radioactive waste evaporator, waste gas surge tank, waste l
compressors, waste gas decay tanks, spent resin storage tank and pump, and letdown heat :
exchanger. Dose rates in the letdown heat exchanger valve gallery range from 3 to 40 mrem /hr. !'
Dose rates at the spent resin storage tank pump room range from 1.3 to 18 mrem /hr. Contact
- dose rates up to 400 mrem /hr are found on this elevation. Removable contamination levels are 2
. up to 7k dpm/100 cm in the spent resin storage tank pump room. Fixed contamination at levels 2
greater than 5k dpm/100 cm is Mi==W to exist over 11% of the surface area. An average -
. fixed contamination level of 10k dpm/100 cm2 is estimated l
Auxiliary Buildine - 45 ft Elevation i Major components on this elevation include treated waste monitor tanks and pumps, dirty waste monitoring tank and pumps, spent fuel pool cooling pumps, spent fuel pool purification pump, chemistry hot lab, and hot sample room. Dose rates range from less than 0.2 to 5 mrem /hr.
Removable contamination levels are less than ik dpm/100 cm2 with the exception of the hot 2
sample sinks. Fixed contamination at levels greater than 5k dpm/100 cm is estimated to cover 15% of the surface area. An average fixed contamination level of 10k dpm/100 cm2 is estimated.
Auxiliary Buildine - 25 ft Elevation O This elevation is below grade and contained the positive displacement pump, centrifugal charging pumps, sodium hydroxide tank, primary water makeup pumps, boron injection tank, reactor coolant drain tank pumps, chemical waste tank and pumps, and access to the clean waste receiver tanks. Accessible dose rates are up to 60 mrem /hr contact at the boron injection tank and range from less than 0.2 to 12 mrem /hr at the clean waste receiver tanks. Removable 2
contamination levels are up to 3k dpm/100 cm in the boron injection tank area. Fixed ;
contamination at levels greater than 5k dpm/100 cm2 is estimated to exist over 15% of the surface area. An average fixed contamination level of 25k dpm/100 cm2 is estimated.
Auxiliary Bu;ldine - 5 ft Elevation (
.This elevation contains residual heat removal pumps and heat exchangers, clean waste receivcr ;
tanks and pumps, dirty waste drain tank ar'd pumps, Auxiliary Building drain tank and pumps,
. safety injection pumps, containmer y a > umps, and the Auxiliary Building and passageway sumps. The bottom level of the cler rasu receiver tanks was not routinely accessed, however, ;
the area was posted as a high radiation. ca. Contact dose rates range up to 100 mrem /hr, with general area dose rates ranging fmm less than 0.2 to 45 mrem /hr. Removable contamination 2
levels are up to 35k dpm/100 cm at the residual heat removal pumps. Fixed contaminatica at 2
levels greater than 5k dpm/100 cm is estimated to exist over 28% of the surface area. An '
2 average fixed contamination level of 25k dpm/100 cm is estimated.
O Appendix 2-2 4 of13 March 1999
TROJANIJCENSE TERMINATIONPLAN Pine Facade - 77 ft Elevation -
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This elevation contains the component cooling water (CCW) surge tanks, an emergency escape hatch between the pipe facade and the Auxiliary Building and the CCW penetrations into the i
containment. Dose rates are less than 0.2 to 1 mrem /hr general area. Dose rates have ranged up i to 200 mrem /hrgeneral area at the resin header. Removable contamination levels are less than ,
Ik dpm/100 cm . Fixed contamination estimate for this elevation are included in the Auxiliary
^ Building 77 ft elevation. ,
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Pine Facade - 6111 Elevation !
i Major components include tb volume control tank, letdown system piping and penetrations, l access to the vertical resic 4 heat removal pipe chase, and the fuel transfer tube. Dose rates are l 1 to 40 mrem /hr general Lta, with some hot spots on the letdown lines and volume control tank. !
' 2 Contamination levels are up to 20k dpm/100 cm at the letdown line area. A fixed contamination i estimate is included in the Auxiliary Building 61 ft elevation. :
Pine Facade- 45 ft Elevation l
'i This level is the main access to the pipe facade. Components include the boric acid blender, l
- residual heat removal piping, and containment penetrations. General area dose rates range from ,
2 less than 0.2 to 12 mrem /hr. Contamination levels of up to 2k dpm/100 cm are found in this i area. Fixed contamination estimate for this elevation are included in the Auxiliary Building 45 ft l
' elevation. :
Fuel Buildine - 118 ft Elevation This is the Fuel Building crane elevation. Dose rates depend highly on location of the crane.
General area dose rates are normally less than 0.2 mrem /hr. Contamination levels are less than Ik dpm/100 cm 2. This elevation consists of structural steel. No fixed contamination is estimated.
Fuel Buildinn- 104 ft Elevation This elevation is used primarily for equipment (both contaminated and clean) storage. Dose rates depend on material stored in the area. With the absence of material, dose rates are less than 2
0.2 mrem /hr. Contamination levels are less than Ik dpm/100 cm . Fixed contamination at levels 2
)
greater than 5k dpm/100 cm is estimated to exist over 5% of the surface area. An average contamination level of 10k dpm/100 cm2 is estimated.
I Fuel Buildinn - 93 ft Elevation This is the main operating floor for the Fuel Building and includes the spent fuel pool, decon f ~shop, hot machine shop, radioactive material storage areas, the top of the boric acid storage tanks, and access to the containment equipment hatch. General arm dose rates range from less f than 0.2 to 0.8 mrem /hr. Contact dose rates above this depend on storage of radioactive material
. Appendix 2-2 5 of13 March 1999 4
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l TROJANLICENSE TERMINATIONPLAN in the area. Removable contamination levels range to Ik dpm/100 cm2 . Fixed contamination at levels greater than Sk dpm/100 cm2 is estimated to exist over 19% of the surface area. An O average fixed contamination level of 10k dpm/100 cm2 is estimated.
Fuel Buildine - 77 ft Elevation- e l
- Major components on this level include radioactive waste control panels, evaporator concentrates l bolding tanks, CVCS surge tanks, CVCS concentrates pump, spent fuel pool skimmer pump, l' new fuel storage, and access to the cask washing pit. General area dose rates on the level range L from less than 0.2 to 40 mrem /hr. Items stored in the cask wash pit could be high radiation
- - sources. The evaporator concentrates holding tanks are not routinely accessed. Removable
' 2 contamination levels are up to 3k dp100 cm in the cask wash pit. Fixed contamination at
. levels greater than 5k dpm/100 cm is estimated to exist over 15% of the surface area. Average
~ fixed contamination is epim*A at 10k dpm/100 cm2, t
Fuel Buildinn - 61 ft Elevation Major components on this elevation include the CVCS monitor tanks, boric acid storage tanks and pumps, seal water heat exchanger, spent fuel pool heat exchanger, I&C hot shop, a respirator wash facility, and the radioactive waste solidification mom. The solidification room is designed for use with a urea formaldehyde solidification system. This system was abandoned in place and -
the room was used for storing and decontaminating equipment. Dose rates on this elevation depend highly on material stored in the area. General areas range from less than 0.2 to 18 mrem /hr and contact dose rates on installed equipment are up to 35 mrem /hr. Removable 2
i contamination levels are less than Ik dpm/100 cm . Fixed contamination at levels greater than 2
Sk dpm/100 cm is estimated to exist over 15% of the surface area. An average fixed 2
contamination level of 10k dpm/100 cm is estimated.
Fuel Buildine - 45 ft Elevation 1
- This elevation contains the CVCS hold up tanks, hold up tank recirculation pump, gas stripper pumps, CCW pumps and heat exchangers, a respirator maintenance facility, a crane bay, and the Radwaste Annex, which is used for radioactive material storage and radioactive waste compacting.' General area dose rates on this elevation are primarily less than 0.2 mrem /hr. The highest contact dose rate is 180 mmm/hr. Removable contamination levels are less than 2
Ik dpm/100 cm 2. Fixed contamination at levels greater than 5k dpm/100 cm is estimated to 2 i exist over 8% of the surface area. An average fixed contamination level of 10k dpm/100 cm ;,
estimated.
Electrical Penetration Area Containment electrical penetrations are located in this area. The area is posted as a fixed !
contamination area. General area dose rates range from less than 0.2 to 5 mrem /hr. Removable l 2
contamination levels are less than Ik dpm/100 cm . Fixed contamination is known to exist under l the RCS sample line isolation valves. Fixed contamination levels cannot be measured at this l time due to background radiation levels. I LO L
. Appendix 2-2 6 of 13 March 1999
~
TROJANLICENSE TERMINATIONPLAN I
Main Steam Supoort Structure i
r This area contains main steam and steam generator blowdown penetrations and is a transition ;
area between the Containment and Turbine Buildings. The Main Steam Support Structure and the gravel area surrounding the plant tank farm were contaminated when a main steam relief ;
. valve opened during a steam generator hydrostatic test. The steam generator contained a mixture
~
ofprimary and secondary liquid following a steam generator tube leak. This event caused fixed ,
contamination throughout this structure. General area dose rates are less than 0.2 mrem /hr. !
' Although removable contamination levels are less than Ik dpm/100 cm2, there is fixed i contamination and the area below the floor grating on the 45 ft elevation is posted as l contaminated. Fixed contamination surveys indicate a 1300 ft2area is contaminated to levels !
2 greater than 5k df' 100 cm . The average fixed contamination level is approximately i 50k dpm/100 cm . j 1
Steam Generator Blowdown Buildine l
1 Due to primary-to-secondary leakage, this building has some fixed contamination. General area j dose rates are less than 0.2 to 0.4 mrem /hr and removable contamination levels are less than 1 2
Ik dpm/100 cm ,
Low Level Radioactive Waste Storane Buildinn This building no longer exists. The building was construc' edt in 1991 as a long-term storage area O for waste ready for shipment. General area dose rates were less than 0.2 to 20 mrem /hr and depended highlz on building contents. Removable contamination levels were less than
- ik dpm/100 cm .
Wriaht-Schuchart-Harbor (WSH) Radioactive Material Storane Area l
.l The WSH Radioac:ive Material Storage Area has been free-released in accordance with plant procedures'and regulatory requirements.
f Unbiased Survey Results i Structures selected for characterization were divided into manageable survey areas such as a building elevation or group of buildings depending on size. For each survey area, a minimum of 30 randomly selected sampling locations were chosen or an average of 1 measurement location 2
per 500 ft , whichever produced the greater number oflocations. These locations were restricted .
to the floor and lower 6 ft' of wall surface unless surveys indicated the potential for additional j contaminated surfaces. A total of 840 sample points were established in the area surveyed for Phase I of the site characterization.
Only two of the sample points had detectable removable contamination levels. One point is on the roof of the Maintenance Building and had detectable alpha and beta activity. The other point O l Appendix 2-2 7 of13 March 1999 i
w ._. -- , , , - -. -. , -
6: ' -
)
TROJANLICENSE TERMINATIONPL4N l i
j- is on the roof of the Turbine Building and had detectable alpha activity. Neither location is ;
r expected to require remediation.
u 1 L ' Structures Within the Industrial Area I L The following buildings are outside of the RCA, but inside the Industrial Area, and were
, surveyed specifically for the site characterization scoping survey. i L
)
Turbine Buildino - Roof
{
\.
l l Exposure rates at the southeast corner are higher than 5 net R/hr due to " shine" from the i l RWST. Contamination survey results were less than Ik dpm/100 cm2, ;
Turbine Buildine - 93 ft Elevation L q lr l
This elevation is the main operating floor of the plant. The turbine, generator, moisture i separator / reheaters, feedwater heaters, and process radiation monitor for the condenser off gas I
system are on this level. General area exposure rates are below 5 net pR/hr, except at the south ,
end of the building where exposure rates are slightly higher due to the RWST. Biased contamination survey results identified fixed contamination under the condenser off-gas grab ]
i 2 2 sample rack measured at 15k dpm/100 cm over an area of 25 ft . All other survey location )
measurements were below Ik dpm/100 cm2, j
)
i Turbine Buildine - 63 ft Elevation 1 O
This elevation contains the steam jet air ejector, lube oil reservoir, main steam stop and tivottle valves, feed water heaters, and the switchgear room. Three areas were identified with exposure l rates above 5 net R/hr. The first area is the alum tank. Exposure rates up to 12 pR/hr were
! measured at the tank due to the uranium content of the alum. The second area is a location on
- l. the floor about 30 ft west and 15 ft south of the alum tank near the grating over the main steam L stop valves. The' exposure rate at this floor location is 28 pR/hr, probably due to fixed l' contamination under the paint. The third location is on the floor near the floor drain east of the steam jet air ejector. This area was remediated, but exposure rates can still be measured to 22 pR/hr. The drain is marked by a fixed contamination label. Biased contamination survey results identified fixed contamination in two areas on this level: Sk dpm/100 cm2 os er 2
approximately 10 ft near the drain in the southeast corner and 50k dpm/100 cm2 over 2
approximately 32 ft near the drain on the south end. All other survey location measurements )
j were below Sk dpm/100 cm2, l
l Turbine Buildine - 45 ft Elevation l I
This is the grade-level floor of the Turbine Building. The primary components on this floor include the main condensers, make up water treatment system and the air compressors. One area j is identified with ' exposure rates above cleanup criteria. This area is on floor near the flow l transmitter stand by the southwest corner of condenser A. Exposure rates at this floor location are up to 18 net pR/hr. Biased contamination survey results identified fixed contamination :
Appendix 2-2 8 of13 March 1999 p 3 1:
L - __ _ . - - _ _ __
TROJANUCENSE TERMINATIONPLAN measured in the pipe trough between the electric AFW pump and the condensate puinp pit.
2 2 Contamination levels of 50k dpm/100 cm over approximately 40 ft were found. All other 2
~
survey location measurements were below Ik dpm/100 cm ,
Turbine Buildina - 35 and 27 ft Elevation These elevations contain the condensate pumps, neutralizing tank, and the Turbine Building I
sump and pump. The Turbine Building sump is a contaminated area. Site characterization loose contamination surveys were taken from areas other than the sump and results were below the cleanup crheria. General area exposure rates were below 5 net R/hr. However, exposure rates at the floor on the south end of the 35 A elevation ranged up to 6 net pR/hr. Fixed contamination l levels of 5-50k dpm/100 cm2 were identified over an area of 2181 ft2 . The contamination is a i
result of sump overflows and condensate pump and heater drain pump leaks during periods of primary-to-secondary leakage.
Control Buildina Roof General area ex sure rates were below 5 net pR/hr and contamination levels were less than L Ik dpm/100 cm Control Buildinn - 105 ft Elevation This elevation is for control room ventilation systems and contains the control room viewing gallery.' General area exposure rater were below 5 net R/hr and contamination levels were less than Ik dpm/100 cm2, Control Buildinn- 93 ft Eleva'iQB The main areas include tige main control room and the chemistry cold lab. The lab had numerous systems which were either comaminated or potentially contaminated.
Equipment in the chemistry cold lab contained low levels of radoactivity due to primary-to-secondary leakage. However, the survey results of the room (structure) indicated j 2
levels less than Ik dpm/100 cm . Since the site characterization survey, most of the equipment i from the lab has been removed.
l 1 l
Control Buildine- 77 ft Elevation ;
i i
!- The cable spreading room, computer room, and mechanical room occupy this entire elevation.
General area ex sure rates were below 5 net pR/hr and contamination levels were less than l . Ik dpm/100 cm l
Control Building - 61 ft Elevation ,
Primary equipment on this level includes the electrical auxiliaries, emergency batteries,
- mechanical room, and the telephone equipment. General area exposure rates were below 5 net 2
- pR/hr and contamination le"els were less than ik dpm/100 cm ,
/ Appendix 2-2 9 of13 March 1999 e- 6 e - , ,--s
p TROJANIJCENSE TERMINATIONPLAN Control Buildine - 54 ft Elevation O,
his level contains some ventilation equipment and office areas. General area exposure rates were below 5 net R/hr and contamination levels were less than Ik dpm/100 cm2 ,
Control Buildine -45 A Elevation his elevation is the primary access area for the RCA. Also on this level are the Radiation Protection Department offices, counting rooms, and calibration facility. Site characterization surveys were taken outside of the RCA. General area exposure rates were below 5 net pR/hr and contamination levels were less than Ik dpm/100 cm2 ,
Seculity Building This is the current access and egress point for the Industrial Area. Previous surveys have shown no radiological impact on this building. General area exposure rates were below 5 net R/hr and contamination levels were less than Ik dpm/100 cm2 ,
. Almini*ation Building his building is used only for office space. General area exposure rates were' below 5 net pR/hr 2
and contamination levels were less than ik dpm/100 cm . The roof was not surveyed due to access difficulty.
O Central Building I
l
. . I This building is used for office space. General area exposure rates were below 5 net pR/hr and j 2
contamination levels were less than ik dpm/100 cm , j 1
. Chlorine Buildine This building was used to treat the service water system prior to use in the plant. The 2
contamination survey results were below Ik dpm/100 cm . Exposure rates over most of the building were influenced by the RWST.
Condan=ata Demineralizer Building This building was used to treat condensate prior to return to the feedwater system. Due to primary-to-secondary leakage during plant operations, many of the systems in this building are 2
potentially contaminated. The structure survey results were below Ik dpm/100 cm . Recent survey results of the hopper room on the 19 ft elevation indicate contamination levels below Ik ;
2 dpm/100 cm and dose rates below 0.2 mrem /hr.
O Appendix 2-2 10 of13 March 1999
, . ..~. .. - - - .... -. - . . - - - . . .. - . .- - - - . - . -
TROJANLJCENSE TERMINATIONPL4N l Discharme and Dilution Structure .
This is the main liquid effluent release point for the plant. Exposure rate measurements in this structure were above 5 net R/hr due to proximity of the RWST. Results of cantamination surveys taken on the 45 level are below Ik dpm/100 cm2 ,
Guard House This was the primary access and egress point for the plant until replaced by a new access control facility in 1992. Survey results for this building have shown no historical impact from plant operations. General area exposure rates were below 5 net R/hr and contamination levels were less than ik dpm/100 cm 2,.
Intake Stmeture Exposuie rate measurements in this structure were above 5 net R/hr due to proximity of the RWST and will be resurveyed following RWST remediation. The contamination survey results were below Ik dpm/100 cm2, Maintenance Building l- : This is the primary maintenance support area for the plant. Exposure rates at the north end of the
- building were above 5 net pR/hr due to proximity to the Low Level Radioactive Waste Storage
!. Building. Exposure rates at the south end of the building were above 5 net pR/hr due to
, proximity of the RWST. Exposure rates by a granite slab in the tool room were up to 7 R/hr.
Results of contamination surveys were below Ik dpm/100 cm2, i Materials Building 2
Contamination survey results were below Ik dpm/100 cm . Exposure rates at the radioactive l I material storage area in the southwest comer of the building were close to 5 net pR/hr. Exposure rates were above 5 net pR/hr in the weld rod and radioactive material storage area at the north end of the building because of the radioactive material stored in the building.
I Plant Modification Shon
~ This building was used for craft sugport during refueling outages. The contamination survey results were below Ik dpm/100 cm . Exposure rates were above 5 net pR/hr at the northwest
-- part of the building due to proximity of the RWST.
Startuo Boiler ;
, i I..
! Exposure rates at this' structure were higher than 5 net pR/hr due to proximity of the RWST.
2 Contamination survey results were below Ik dpm/100 cm ,
t i.
O I March 1999 A ppendix 2-2 11 of13
TROJANLJCENSE TERMINATIONPLAN L
l l Technical Support Center i
Q'q This building is adjacent to the Condensate Demineralizer Building, but does not share a wall or have through-wall penetrations. The building has been used as office space and record storage, l as well as for emergency response. The site contamination survey results were below 2
l Ik dpm/100 cm . Exposure rates were below 5 net R/hr except near the check source in radiation monitor PRM-25 in the basement.
Wright-Schuchart-Harbor (WSH) Warehouse l l j l This was one of the original structures built on the plant site. One small corner of this building l was used for storage ofradioactive tools and equipment. The contamination survey results from outside of the radioactive material storage area were below Ik dpm/100 cm2 . Inside the storage area, dose rates were from 2.5 to 5 mrem /hr and removable contamination levels were less than 2
Ik dpm/100 cm . Exposure rates were above 5 net pR/hr in the entire east half of the building due to radiation from the radioactive material storage area. This building has since undergone final survey and has been found to satisfy the site release criteria as documented in the approved PGE-1074, " Trojan Nuclear Plant Final Survey Report for the ISFS1 Site."
l Structures Outside of the Industrial Area Dosimetry Lab This building contains the dosimetry and environmental monitoring groups. Sealed sources are stored in this building which caused exposure rates to be measured above 5 net pR/hr in the surrounding area during gamma surveys. Once the sealed sources were removed, nothing else was identified that would impact radiological conditions in the building.
Park Structures .
These buildings were not used for plant activities and should be below contamination levels of 2
ik dpm/100 cm . A gamma scan performed on the park office did not identify exposure rates above 5 net R/hr.
Pebble Sprines Warehouse A gamma scan of the building found no exposure rates above 5 net pR/hr.
Sewer Treatment Plant ,
i A gamma scan of the building did not identify exposure rates above 5 net pR/hr.
L '
l South Maintenance Buildina l
This building was surveyed at selected sites to provide background data. General area exposure 2
! rates were below 5 net pR/hr and contamination levels were less than ik dpm/100 cm ,
l Appendix 2-2 12 of13 March 1999 l
l
_ . . - . . _ _ _ _ _ _ _ _ . _ _ _ ~ _ . _ _ _ . . _ . . . _ _ _ . . . _ . _ _ _ . _ _ _ _ . _ _ . _ _ .
TROJANLJCENSE TERMINATIONPL4N i Traininn Building 5O This building was used only as a training facility. A gamma scan of the building did not identify 1
exposure rates above 5 net pR/hr. -
Troian North Building This building was used for engineering and administrative office space. A gamma scan of the building found no exposure rates above 5 net R/hr.
Troian Visitors Information Center This building was surveyed at selected sites to provide background data. General area exposure rates were below 5 net pR/hr except near the check source for the radiation monitor.
Contamination levels were below Ik dpm/100 cm2, O
- Appendix 2-2 13 of13 March 1999 L
g L
TROJANLICENSE TERMINATIONPL4N p Table 2-1
' (, '
Radioactive Emuent Summary, Noble Gases l
l Release (Ci)
First Second Third Fourth l
Year - Quarter Quarter Quarter Quarter i 1993 26.5 13.1 7.3 6.5 1992 24.7 33.1 46.8 102.0 1991 75.7 23.9 40.3 26.6 1990 86.4 32.9 43.2 43.9
)
1989 14.3 228.0 179.0 42.4 l
1988 61.3- 145.0 67.2 126.0 1987 85.4 61.3 9.6 93.0 l 1986 475.0 246.0 117.0 75.0
, 1985 340.0 277.0 166.0 278.0 1984 368.0 331.0 0.183 138.0 1983 65.7 0.0 31.9 '131.0 l 1982 347.0 . 162.0 80.4 275.0 1981 ~ 316.0 399.0 153.0 293.0 1980 161.0 90.1 53.2 86.4 1979 33.9 252.0 131.0 510.0 l 1978 290.0 129.0 1.1 1.4 1977 650.0 1960.0 300.0 160.0 1976 170.0 360.0 26.1 111.0 1975 30.0 i
- Operating license granted November 1975. l 4
o March 1999
r i TROJANIJCENSE TERMINATIONPLAN Table 2-2 l G l Radioactive Emuent Summary, l Iodine and Particulates (excluding tritium) l Release (Ci)
First Second Third Fourth i Year Quarter Quarter Quarter Quarter l
- l 1993 1992 0.0000003 0.00008 0.000002 0.00019 1991 0.00009 0.00038 0.00004 0.00008 1990 0.00092 0.00067 0.00005 0.000008 1989 0.00009 0.00389 0.00031 0.00003 1988 0.00008 0.00093 0.00018 0.00194 1987 0.00050 0.00150 0.00004 0.00016 ;
1986 0.00022 0.00333 0.00216 0.00042 I p 1985 0.00264 0.00184 0.00062 0.00039 U 1984 0.00387 0.00396 0.00054 0.00094 1983 0.00365 0.00095 0.00070 0.00082 1982 0.00318 0.00721 0.00132 0.00278 l 1981 0.0294 0.0419 0.00167 0.00369 1980 0.00265 0.02038 0.00089 0.00120 1979 0.00763 0.00789 0.00131 0.01707 1
0.00600 0.00289 0.00064 0.00044 1978 1977 0.0231 0.0244 0.0023 0.00065 1976 0.0097 0.00032 0.00066 0.00037 b b b 1975 0.00016
- Minimum Detectable Activity, b
Operating license granted November 1975.
, March 1999
TROJANLICENSE TERMINATIONPLAN t
Table 2-3
.75 U Radioactive Emment Summary, Liquids Release (Ci) -
First Second- Third Fourth Year Quarter Quarter Quarter Quarter 1993 0.0453 0.0240 0.0490 0.0087 1992 0.0241 0.0305 0.0150 0.0214 1991 0.0118 0.0261- 0.0114 0.0087 1990 0.0084 0.0560 0.0675 0.0123 1989 0.0308 0.0468 0.0684 0.0156 1988 0.0543- 0.0511 0.0570 0.0382 1987 0.0334 0.0657 0.0539 0.0564 1986 0.0460 0.0847 0.0877 0.0459 1985 0.0757 0.154 0.0870 0.148 1984- 0.0618 0.107 0.0696 0.11l' 1983 0.0463 0.0981 0.107 0.0589 1982- 0.298 0.215 0.264 0.0789 1981 0.265 0.318 0.218 0.193 1980 ' O.127 0.381 0.101 0.178 1979 0.141 0.0456 0.0410 0.327 1978 0.279 0.165 0.0782 0.185 1977 0.450 2.65 1.00 0.0920 1976 0.26 0.94 0.89 0.63 1975 0.02
- Operating license granted November 1975.
i r
March 1999 h-
TROJANLICENSE TERMINATIONPLAN
! Table 2-4 Structures Burial Volume and Contamination
- Activity Projections Building Volume (ft') Activity (mci)*
Containment Building 13,458 24 Auxiliary 2,650 2
- Fuel 4,711 1 l MSSS/EP 629 1 ,
Turbine. 1054 2 Total 22,502 31
- Includes removable and fixed contamination, but not activation.
b Column may not total exactly due to rounding activity estimates.
i i
l-O L
l i
O March 1999
- _ - ~ _ .
o o TROIANUCENSE TERMINATIO ALAN ~
' Page 1 of 8
- Table 2-5 Status of Buildings in the Radiologically Controlled Area as of 1994
-y Removable Maximum-Contamination General Area Contact Elevation Level Dose Rate Dose Rate Building ' (ft) Room or Component (dpm/100 cm2) (mrem /hr) (mrem /hr)
Containment 205 Containment air coolers <2k to 150K <0.2 to 1.8 NA Grating <1k to 3k 0.4 to 0.6 - NA I
105 General area <1k to 2k 0.3 to 0.5 NA 93 General area <1k to 3k <0.2 to 8 NA Refueling bridge 4k 3 to 4 NA ,
Upper refueling cavity Sk to 50k 40 to 70 NA Lower refueling cavity Ik to 20k 1.5 to 5 NA -
Refueling upender 6k to 200k 20 to 100 77 General area <1 to 2k <0.2 to 1.5 NA !
63 Seal table 4k to 30k 2 to 6 NA i
61 General area <1k to 4k <0.2 to 8 800 !
Pressurizer shed ik to 60 k 5 to 120 280 i Regenerative heat exchanger NA 1000 to 6000 NA Excess letdown heat exchanger NA 600 NA March 1999 ,
I
c-TROJANUCEN5E TERMNATIO LAN Page 2 of 8 Table 2-5 Status of Buildings in the Radiologically Controlled Area as of 1994 p-7 Removable Maximum Contamination General Area Contact Elevation Level Dose Rate Dose Rate Building (ft) Room or Component (dpm/100 cm2) (mrem /hr) (mrem /hr) 45 General area <lk to 3k 02 to 10 NA Outside bioshield Safety injection line area <1k to 30k 20 to 170 600 Containment 45 Reactor coolant drain tank and 2k to Sk 1.2 to 20 NA Outside recirculation sump bioshield Pressurizer relief tank 8k to 70k 5 to 30 NA 220 dpm alpha Tendon gallery <lk <0.2 NA 45 General area 2k to 110k 10 to 250 1000 Inside bioshield 58 A/D steam generator platform <lk 18 to 250 700 B/C steam generator platform <1k 18 to 100 1000 A/D resistance temperattw detector 10k to 30k 12 to 80 NA (RTD) platform j
B/C RTD platform 3k to Sk 10 to 80 NA March 1999
. .q O O TRWANIJCENSE TERMINA yPLAN
, Page 3 of 8 Table 2-5 Status of Buildings in the Radiologically Controlled Area as of 1994 -
p-y Removable _
Maximum Contamination General Area ~ - Contact Elevation Level Dose Rate Dose Rate Building (ft) Room or Component (dpm/IDO cm') - (mrem /hr) . (mrem /hr) 67 A/D reactor coolant pump (RCP) 2k 6 to 30 34 platform B/C RCP platform 4k 15 to 20 NA Auxiliary 104 General area . <1k ' <0.2 NA 93 General area <1k <0.2 NA 77 General area <1k <0.2 NA Demineralizer valve galleries <1k to 3k <0.2 to 50 NA Filter valve galley <lk to 9k '<0.2 to 30 350 Auxiliary 77 Boric acid evaporators <1k to Ik 0.5 to 8 NA 61 General area <1k <0.2 NA Waste gas decay tanks <lk <0.2 to 0.4 20 Waste gas compressors <1k <0.2 to 0.3 NA-Spent resin storage tank pump room <1k to 7k 1.3 to 18 NA Letdown heat exchanger valve gallery <1k to 6k 3 to 40 400 45 General area <1k,2k at hot <0.2 to 1.0 NA.
sample sinks March 1999
o o ALAN TROJAN 12 CENSE TERAf1NATIO, Page 4 of 8 Table 2-5 Status of Buildings in the Radiologically Controlled Area as of 1994 p-y Removable Maximura Contamination General Area Contact Elevation Level Dose Rate Dose Rate 2
Building (ft) Room or Component (dpm/100 cm ) (mrem /hr) - (mrem /hr) i Dirty waste monitor tank <lk I to 5 NA Treated waste monitor tanks <1k 1 to 1.8 NA -
Spent fuel pool cooling pumps <1k <0.2 to 1.0 NA 25 General area <1k <0.2 to 0.4 NA Centrifugal charging pumps <1k 0.5 to 0.7 10 Boron injection tank <1k to 3k 0.2 to 8 60 Clean waste receiver tanks <lk 4 to 12 35 5 General area <1k <0.2 to 5 NA i Auxiliary 5 Residual heat removal (RHR) pumps <1k to 35k 3 to 70 NA RIIR heat exchanger <1k to 30k I to 60 NA Clean waste pumps <1k <0.2 to 25 100 Auxiliary building drain tank and pumps <1k 0.5 to 45 50 ,
Dirty waste drain tank and pumps <1k ' O.2 to 30 NA i .p . racade 77 General area <1k <0.2 to 1 NA L
L March 1999
A.
N.] TROJANUCENSE TERMLVA PLAN Page 5 of 8 Table 2-5 Status of Buildings in the Radiologically Controlled Area as of 1994 p-y Removable Maximum Contamination General Area Contact Elevation Level Dose Rate Dose Rate 2
Building (ft) Room or Component (dpm/100 cm ) (mrem /hr) (mrem /hr)
Pipe chase, resin header NA 1 to 200 800 ,
61 General area <1k to 4k 1 to 12 NA Letdown line ik to 20k 2 to 12 8 Pipe chase ik 2 to 4 80 Volume centrol tank <1k 2 to 40 200 ,
45 General area <lk <0.2 to 12 NA 40 Pipe chase 2k 0.6 to 3 NA 25 Pipe chase <lk i NA Fuel 118 Fuel building crane <1k <0.2 NA 104 General area <1k <0.2 to 2.5 NA Fuel 93 General area <1k to Ik <0.2 to 0.8 7 Spent fuel pool <1k <0.2 to 0.5 200 77 General area <1k <0.2 to 1 NA Chemical volume control system surge <1k 0.5 NA tank March 1999 .
g_ s
.Q' d s /
TROJAN 12 CENSE TERMINATIO.frLAN Page 6 of 8 Table 2-5 Status of Buildings in the Radiologically Controlled Area as of 1994 p-y Removable Maximum Contamination General Area Contact Elevation Level Dose Rate Dose Rate 2
Building (fi) Room or Component (dpm/100 cm ) (mrem /hr) ~ (mrem /hr)
CVCS pump room <lk to 2k -1 to 6 NA Cask wash pit 3k 10 to 40 NA 61 General area <lk <0.2 to 1.4 8 Spent fuel pool heat exchanger <1k 0.2 to I NA Seal water heat exchanger <lk 4 to 18 35.
CVCS monitorianks <lk- <0.2 NA Boric acid storage tanks <1k 0.5 to 10 NA 45 General area <1k to Ik at CCW <0.2 to I NA heat exchanger A ,
Radwaste annex <lk <0.2 NA i CVCS holdup tank pumps <lk 0.2 to 1 180 Electrical 70 and 77 General area <1k <0.2 NA l penetration 61 General area <1k <0.2 to 1.0 NA Electrical 51 General area <1k <0.2 NA penetration March 1999
_ _ _ _ _ . . _ _ . . _._m.._ ._..__._.______._._______m____,_m._________m._._ - -
O O TRCLIANLICENSETERMINA170 PLAN R Page 7 of 8 Table 2-5 States of Buildings in the Radiologically Controlled Area as of 1994 p-y Removable Maximum Contamination General Area Contact Elevation Level Dose Rate Dose Rate 2
Building (ft) Room or Component (dpm/100 cm ) (mrem /hr) - (mrem /hr)
~
45 General area <lk outside <0.2 to 5 30 contmiled area Main steam 77 General area <lk <0.2 NA support 69 General area <lk <0.2 NA structure 59 General area <lk <0.2 NA 45 General area <lk <0.2 NA Steam 55 <lk <0.2 NA-generator 45 <1k <0.2 to 0.4 NA blowdown -
l building -
Wright- 45 General area <!k .<0.2 to 5 NA Schuchart-Ilarbor RMSA Refueling 45 Exteriorinside fence <1k 2.5 to 5 NA water storage tank Primary water 45 Exterior <lk <0.2 NA storage tank March 1999
. TROJANI2 CENSE TERMINATICseLAN Page 8 of 8 Table 2-5 ,
Status of Buildings in the Radiologically Controlled Area as of 1994 p-y Removable Maximum ;
Contamination General Area - Contact Elevation Level Dose Rate Dose Rate 2
Building (ft) Room or Component (dpm/100 cm ) (mrem /hr) (mrem /hr)
Condensate 33 Hopper room, sump not included <lk NA NA demineralizer building Radwaste 45 General area <lk- <0.2 to 20 NA storage building NA -Not applicable
~
- Recent survey data is not available. .
t
?
i i
March 1999
p; L TROJANLICENSE TERP:: 70NPLAN l
1-
.. Page 1 of 2 l
/ Table 2-6 3 l Q] -
System Burial Volume and Surface Activity Projections ,
- System Volume (ft') Activity"(Ci) l l Reactor coolant piping- 5,894 221 l
' Pressurizer relief tank 625 <1 Reactor coolant pumps and motors 3,044 134 ,
Control rod drive mechanisms /incore 1,726 83 ,
instrumentaticdservice structure l
Reactor vessel and intemals 8,341 357.9 Spent fuel pool and racks 17,305 150+ :
120-V ac preferred instrument ac 1,400 <1 <
125-V de power 175 <1 4.16-kV ac power 726 <1 :
480-V ac auxiliary load center 5,080 <1 ;
480-V ac motor control center 8,426 <1 l Chemical and volume control 10,968 25 l l
Clean raevaste 5,423 14 '
l Containme'-t building penetrations 188 <1 Control rod drive 85 <1 Dirty radwaste 1,613 <1 Electric heat tracing 164 <1 ,
Electrical (Cablerfray/ Conduit) 60,139 <1
)
Fuel handling system 339 b
j i
Fuel pool cooling and demineralizer 4,632 5.6 i'
Fuel and auxiliary building heating, ventilation, 3,661 <1 and air conditioning (HVAC) i Gaseous radwaste 2,529 <1 HVAC 6,635 <1 l
! Hydrogen recombiners 576 <l
(. Integrated leak rate test instrument line 106 <1 Instrument and service air 1,327 <1 March 1999
TROJANLJCENSE TERMINATIONPLAN l
Page 2 of 2 1 O Table 2-6 V
System Burial Volume and Surface Activity Projections System Volume (ft') Activity'(Ci)
Lighting panel supply -997 <1 Miscellaneous components 1,936 <1 Miscellaneous reactor coolant 3,418 <1 Nuclearinstrumentation 193 <1 !
Oily waste and storm drains 1,882 <1 Containment HVAC 18,869 <1 !
Primary makeup water 3,615 <1 i Process sampling 114 4 Radiation monitoring 134 <1 Reactor nonnuclearinstruments 245 <1 I l Reactor vessel system 116
( Residual heat removal 7,649 36 Safety injection system 7,149 7 Solid radwaste 370 <1 l 4 I
I Steam generator system 3,562 <1 )
Turbine building sump pumps and 639 <1 miscellaneous d
Component cooling water 6,115 <1 d
Condensate demineralizers 2,262 <1 d
Discharge and dilution 3,834 <1 d
Containment spray 1,563 <1 Total 215,789 1070.5 Does not include activation.
To be determined.
Activity included with reactor coolant piping.
d Site characterization survey results identified these systems as contaminated #'
b March 1999
. . . . ~ . . - . . - - . . - . -..
- i. [
TROJAN 1.lCENSE TERMINATIONPLAN I
l l 5 Table 2-7 Isotopic Distribution )
l '(Decay Corrected to 1994 and 1998)
% Activity * % Activity * !
Auxiliary Systems Primary Systems i Radionuclide 1994 / 1998 - 1994 / 1998 Magnesium-54 1.0 / <1 <1 / <1 Iron-55 5.9 / 2.6 61 / 43 1 Cobalt-57 <1 / <1 1 Cobalt-58 <1 / <1 <1 / <1 l Cobalt 23 / 18 15 / 18 i
Nickel-63 52 / 66 19 / 38 Strontium-89 <1 / <1 <1 / <1 l Strontium-90 <1 / <1. <1 / <1 Zirconium-95 <1 / <1 <1 / <1 ,
Ruthenium-106 3.7 / <l <1 / <1 Silver-108m <1 / <1 i Silver-110m <1 / <1 Cadmium-109 1.3 / <1 l
l Tin-ll3 <1 / <1
~
Antimony-125 1.2 / <1 <1 / <1 lodine-129 <1 / <1 <1 / <1 Cesium-134 <1 / <1 Cesium-137 <1 / <1 <1 / <1 l Cerium-144 <1 / <1 <1 / <1 l~ Plutonium-238 <1 / <1 <1 / <1 I 6
Plutonium-239/240 <1 / <1 <1 / <1 !
i i Plutonium-241 11 / 11 <1 / <1
' Americium-241 <1 / <1 ,
~
Curium-242 <1 / <1 <1 / <1
- l Curium-243/244 <1 / <1 Based on clean waste filter analysis (10 CFR 61 Waste Stream.1992).
4_
, Based on 1992 steam generator tube analysis.
Not identified.
March 1999 I-i,
F n Tablek l
10 CFR Part 61 Classification by Cod 10 CFR 61 H-3 C-14 Ca-45 Mn-54 Fe-55 Co-60 Ni-59 NiG Component Classification (Ci/m)
Core Baffle >C 2.494E42 1.012E+02 1.650E+00 3.556E+04 9.244E+05 1.046E+06 4.544E+02 7.635E I
Core Formers >C 1.517E+02 9.709E+01 1.626E+00 7.463E+03 9.219E+05 5.652E+05 3.619E+02 6.821Ed B
C 1.951E401 3376E+00 5.473E-02 1.190E+03 3.092E+04 4.129E+04 1.859E+01 2.691 A 7.963E-04 1.378E-04 2.234E-06 4.858E-02 1.262E+00 1.685E+00 7.589E-04 1.098 %
3; p C IJ73E+01 2.385E+00 3.865E-02 8.428E+02 2.185E+04 2.932E+04 IJi9E+01 1.903Ed VesselClad C 1.504E+0i 2.628E+00 4.356E-02 3.603E442 2.493E+04 2.167E+04 1.573F ull 2.186Ed Vessel Wall A 2.293E-01 2.716E-04 2.038E-04 8.962E+00 2.298E+02 2.900E+0! 5.836E-03 8.439E
- ,, A 8.655E-02 1.432E-02 2.360E-04 2.753E+00 IJ47E+02 1.340E402 8.627E-02 1.188E4 i
A 6.825E+00 2.217E-03 2.702E-01 1.098E-02 1386E+00 3.508E-01 1.700E-05 2.286E-ield B osh eld A 2.597E+01 8.305E-03 1.019E400 1397E-02 5.232E+00 1.143E+00 6.500E-05 8.621Es A 3.821E-01 4.795E-03 3.181E-04 9.640E-01 3.609E+02 1.397E+01 1.108E-02 1.468E+
A 1.243E+0! 3.959E 03 4.865E-01 2.759E-03 2.499E+00 5.204E-01 3.ll7E-05 4.ll8Es o eld g,,,
a d 6;, A 2.544E e0 8.08eE-04 . 952E-02 2.387E-04 5.iiiE-Oi i.043E-Oi e387E.06 8.4i9E.
o ield A 3.961E-01 1.259E-04 1.550E-02 3.156E-05 7.960E-02 1.621E-02 9.949E-07 1.311E4 Re A 6.169E-03 7.268E-05 2.285E-05 4.894E-03 7.169E+00 2.258E-01 1.696E-04 2.252E4
[ oshield A 5.954E-02 1.893E-05 2J28E-03 6.215E-06 1.196E-02 2.445E-03 1.495E-07 1.971E-(
> ^ f oshield A 9.559E-03 3.042E-06 3.741E-04 1.406E-06 1.922E-03 3.955E-04 2.400E-08 3.166E-(
l 1
l TROJANLJCENSE TERMINATIONPLAN ;
Page 1 of 2 l I l
.. . . I mest One Year After Shutdown l
_i Nb-94 Tc-99 Sn-119m Sb-125 Te-125m Sum ofFractions Tab.1 - Tab. 2 Col.1 Tab. 2 Col. 2 Tab. 2 Col. 3 l
3 1.897E+00 4.341E-01 1.631E-10 7.702E41 5.999E-03 12.818 5053.209 109.073 10.907
))
3 7.868E41' l.059E41 1.317E-11 1.924E41 1.941E43 6.793 4087.766 97.441 9.744 l
j i
)
) 6.870E-02 1.740E42 6.478E-17 1.225E43 '8.980E46 0.470 182.215 3.844 0384 I l
APEaWRE 2.804E-06 7.104E-07 2.644E-21 5.001E-08 3.666E-10 0.000 0.007 0.000 0.0Mbh
) 4.872E-02 1.237E-02 3.449E-16 4.588E44 3359E-06 0333 129.011 2.719 M Ayn!!ab 0On i Apritno C2rd 2.936E-02 5366E43 1.474E-16 3.752E-04 3.469E-06 0.251 129.929 3.123 0312 7.638E-05 1.939E44 1.142E-03 1.58IE-03 5.869E-06 0.000 0.412 0.001 0.000 1.943E44 4.270E 05 7.968E-21 1.513E48 1307E-10 0.002 0.729 0,'317 0.002 2.915L 06 3358E-08 1.017E-04 1.515E44 5.620E47 0.000 0.174 0.000 0.000 8.182E-06 4.623E-08 1.400E44 2.405E-04 8.922E47 0.000 0.660 0.000 0.000
- 3.509E45 2.465E-09 9.324E-22 9.968E-09 1.084E 10 0.000 0.588 0.002 0.000 l 3.506E-06 1.061E-08 . 3.210E-05 6.772E-05 2.509E47 0.000 0316 0.000 0.000 i 6.840E-07 1.215E-09 3.679E 06 9.929E 06 3.677E-08 0.000 0.065 0.000 0.000 ?
i 1.060E-07 1.728E 10 5.230E-07 1.479E-06 5.476E-09 0.000 0.010 0.000 0.000
)
i 4.544E-07 9.221E 6.141E-26 1.675E-12 1.425E-12 0.000 0.011 0.000 0.000 1.607E-08 3.029E-Il 9.164E-08 2.400E-07 8.888E-10 0.000 0.002 0.000 0.000 j 4
2.624E49 6.060E-12 1.834E-08 4346E48 1.610E-10 0.000 0.000 0.000 0.000 ,
L 9908160260 D March 1999 u % ,
i i
j (n
N !
t.a Table l
10 CFR Part 61 Classification by Comi l
10 CFR 61 H3 C-14 Ca-45 Mn-54 Fe-55 Co-60 Ni 59 Ni-6 Component Classification - (Ci/m)
)
A 1.008E-03 3.212E-07 3.948E45 2.027E-07 2.028E-04 4.210E45 2.530E-09 3342E<
B W 5
A. 4,210645 1342E-08 1.650E-06 1.128E48- 8.471E-06 1.777E-06 1.056E-10 1.3963 B W. q i
g A' 2.0670 46 6.591E-10 8.102E48 6.166F-10 4.161E-07 8.767E 08 5.184E-12 6.854N
>C 9.614E+01 2.474E+01 4.073E 01 5392E+03 23IIE405 2.29IE+05 1.252E402 1.934E<
e-core Support C 3.785E401 6318E+00 1.056E-01 3.896E+02 6.069E+04 4.335E+04 3.921E+01 5349E<
Columns g A 7.425E-03 1.28 t E-03 2.087E45 3.832E-01 1.183E+01 1.452E+01 7.321E 03 1.037Eq S
A 4.101E-05 6.465E-06 1.080E-07 3.751E-04 6.215E-02 4.443E 02 4.148E-05 5.532E Upper Core C 2.078E+0! 3.699E+00 6.067E-02 8.779E+02 3.452E+04 3.736E+04 2.125E+01 3.014E 4
UpperCore Suppost B 2.670E400 4.186E-01 6.989E-03 2.643E+01 4.020E+03 2.913E+03 2.673E+00 3.575E Columns I l
l I
l
.m gM
%u _ - .... ,+. ,, , ,, _ _ . . , _ . . - - . . ,,. - - - - -
I TROJANLICENSE TERMINATION PLAN Page 2 of 2
-8 ponent One Year After Shutdown Nb-94 Tc-99 Sn-119m Strl25 Te-125m Sum of Fractions Tab.1 Tab. 2 Col.1 Tab. 2 Col. 2 Tab. 2 Col. 3 F
37 2.824E-10 7.986E-13 2.418E-09 5.239E-09 1.943E-Il 0.000 0.000 0.000 0.000 APERTURE 38 1.209E Il 4.161E-14 !.260E-10 2.528E-10 9.346E-13 0.000 0.000 0.000 0.000 hkbh JO 6.000E-13 2.227E-15 6.785E-12
~
1.329E-Il 4.644E-14 0.000 0.000 0.000 0.00630 Availtblo 00 fmer!Ur0 CMd p4 3.557E-01 7.724E-02 4.980E-12 3.097E-02 2.605E-04 2.657 1220.184 27.632 2.763 03 5.026E-02 5.982E-03 1.221E-16 6.073E-04 6.239E-06 0.508 302.959 7.642 0.764 00 2.325E-05 5.908E-06 9.228E-23 1.304E-10 1.OllE-12 0.000 0.068 0.001 0.000 4
13 5.056E-08 6.047E-09 0.000E+00 0.000E+00 0.000E400 0.000 0.000 0.000 0.000 03 5.715E-02 1.293E-02 4.549E-16 7.761E-04 6.373E-06 0.429 190.561 4.305 0.431 i 32 3.366E-03 4.ll6E-04 5.553E-19 2.707E-06 2.776E-08 0.034 20.224 0.511 0.051 ,
1 l
l U O316 02 6 0 - 0L March 1999
j ww-i 1
l 1
r i !
l l
Table $
l i
10 CFR Part 61 Classification by Comp i
10 CFR 61 H-3 C-14 Ca-45 Mn-54 Fe-55 Co-60 Ni 59 Ni C y nt ' Classification (Ci/m) l i
Core Shroud >C 1.992E+02 1.0!!E+02 3301E43 1392E+03 3.182E+05 6.180E405 ' 4.544E+
Core Formers >C 1.212E+02 9.705E401 3.253E43 2.921E+02 3.174E+05 3340E+05 3.61SE+02 6.6I8 Cat' C 1.558E+01 3.374E+00 1.095E44 4.658E+01 1.064E+04 2.439E+04 1,859E+01 2.611 A 6362E-04 ' l.378E44 4.470E-09 1.901E 03 4345E-01 9.959E-01 7.589E-04 1.
Shield C 1.097E+01 2384E+00 7.734E45 3.299E+01 7.520E+03 1.732E+04 1.319E+0! 1.
1 i
VesselClad C 1.202E+0! 2.626E400 8.715E45 1.410E+01 8.583E+03 1.280E+04 1.573E+01 2.121[(
' l,
~
Vessel Wall A 1.832E-01 2.715E-04 4.078E-07 3.508E41 7.911E401 1.714E+01 5.836E-03 8.189E l
i
, A 6.914E-02 ' l .431E-02 4.721E-07 1.077E41 4.637E+01 7.918E+01 8.627E-02 1.153 1st 3" Bioshield A 5.453E+00 2.216E43 5.40$E-04 4.298E-04 4.773E-01 2.073E-01 1.700E-05 2.218d gg A 2.074E401 8301E 03 2.038E-03 5.468E44 1.801E+00 6.756E-01 6.500E-05 '8.365 A ' 3.053E41 4.793E43 6365E 07 ' 3.773E-02 1.242E+02 8.254E+00 1.108E42 1.425@
l g A- 9.928E+00 3.957E-03 9.733E-04 1.080E-04 8.604E-01 3.075E-01 3.ll7E-05 3.996 3rd 6" Bioshield A 2.032E+00 8.082E-04 1.9915-04 9343E-06 1.760E-01 6.163E-02 6387E-06 8.169$
l 4th 6" Bioshield 'A 3.164E-01 1.259E-04 3.100E 05 1.235E-06 2.740E-02 9.578E-03 9.948E-07 1.2723 n eld A 6.169E43 7.268E-05 2.285E-05 4.894E-03 7.169Ew 2.258E-01 1.6%E-04 2.252@
l Sth 6" Bioshield A 4.756E-02 1.892E-05 4.659E-06 2.432E-07 4.Il8E-03 1.445E-03 1.494E-07 1.912@
i wV ,
i TROJANIJCENSE TERMINATION PL.AN l
Page 1 of 2 i
l l
ment Five Years After Shutdown Nb-94 Tc-99 Sn-119m Sb-125 Te-125m SUM OF FRACTIONS Tab.1 Tab. 2 Col.1 Tab. 2 Col. 2 Tab. 2 Col. 3 10 l.897E+00 4.341E-01 2.615E-12 2.831E-01 1.566E-10 12.959 3461.137 105.834 0,5 e 1[?]
10 7.867E-01 1.059E-01 2.lllE 13 7.071E-02 5.069E-Il 6.826 2824.912 94.548 9.4 13 6.869E-02 1.740E-02 1.039E-18 4.503E-04 i - (grjhbis 00 2.345E-13 0.476 125.107 3.730 /9. g7gCgd D1 2.804E-06 7.104E-07 4.240E '!3 1.838E-08 9.573E-18 0.000 0.005 0.0M 0.000 13 4.871E-02 1.237E-02 5.530E-18 1.686E-04 8.771E-14 0.337 88.570 2.638 0.264
$3 2.936E-02 5.366E-03 2.364E-18 1.379E-04 9.058E-14 0.253 91.483 3.030 0.30.1 DI 7.637E-05 1.939E-04 1.830E-05 5.811E-04 1.533E-13 0.000 0.166 0.001 0.000
$1 1.943E-04 4.270E-05 1.278E-22 5.560E-09 3.413E-18 0.002 0.511 0.016 0.002 l
)3 2.914E-06 3.358E-08 1.631E-06 5.567E-05 1.468E-14 0.000 0.137 0.000 0.000 h 8.181E-06 4.623E-08 2.245E-06 8.839E-05 2.330E-14 0.000 0.522 0.000 0.000 .
M 3.509E-05 2.465E-09 1.495E-23 3.663E-09 2.831E-18 0.000 0.238 0.002 0.000 l3 3.50$E-06 1.061E-08 5.148E-07 2.489E-05 6.553E-15 0.000 0.250 0.000 0.000 h 6.840E-07 1.215E-09 5.898E-08 3.649E-06 9.602E-16 0.000 0.051 0.000 0.000 b 1.059E-07 1.728E-10 8.387E-09 5.434E 07 1.430E-16 0.000 0.008 0.000 0.000 f2 4.544E-07 9.221E-12 6.141E 26 1.675E-12 1.425E-12 0.000 0.011 0.000 0.000 PS 1.607E-08 3.029E-I l 1.469E-09 8.819E-08 2.32 iE-17 0.000 0.001 0.000 0.000 March 1999 9003160260so3 -
Table 2 10 CFR Part 61 Classification by Comi 10 CFR 61 H-3 C-14 Ca-45 Mn-54 Fe-55 Co-60 Ni-59 Ni-Component Classification (Ci/m) 6th 6" Bioshield A 7.636E-03 3.041E-06 7.485E-07 5.504E-08 6.615E-04 2.337E-04 2.400E-08 3.072 A 8.056E-04 3.210E-07 7.899E-08 7.934E-09 6.980E-05 2.487E-05 2.530E-09 3.243 o ield A 3.363E-05 1341E-08 3301E-09 4.415E-10 2.916E-06 1.050E-06 1.056E-10 13541 shield A 1.651E-06 6..i88E-10 1.621E-10 2.414E-11 1.432E-07 5.181E-08 5.184E-12 6.6511 B oshi ld
>C 7.681E+01 2.472E+01 8.149E-04 2.110E+02 7.955E+04 1354E+0* 1.252E+02 1.8771 Lower Core Support C 3.023E+01 6.31SE+00 2.113E-04 1.525E+01 2.089E+04 2.56IE+04 3.921E+01 5.1901 Columns A 5.932E-03 1.280E-03 4.176E-08 1.500E-02 4.074E+00 8.582E400 7.321E-03 1.0061 g A 3.276E-05 6.462E-06 2.161E-10 1.468E-05 2.139E-07 2.626E-02 4.148E-05 53671 Up r Core C 1.660E+01 3.697E+00 1.214E-04 3.436E+01 1.18G+04 2.208E+04 2.124E+01 2.9241 Upper Core Support L 2.133E+00 4.184E-01 1398E-05 1.035E+00 1.384E+03 i.721E+03 2.673E+00 3.4691 Columns
TROJANllCENSE TERMINATIONPLAN Page 2 of 2 ,
tent kive Years After Shutdown Nb-94 Tc-99 Sn-119m Sb-125 Te-125m SUM OF FRACTIONS Tab.1 Tab. 2 Col.1 Tab. 2 Col.2 Tab. 2 Col. 3 3 2.623E-09 6.060E-12 2.941E-10 1.597E-08 4.205E-18 0.000 0.000 0.000 0.000 f @ E R.c W- q
@ 2.824E-10 7.986E-13 3.877E-11 1.926E-09 5.073E-19 0.000 0.000 0.000 0.0 W hbbD D 1.208E-Il 4.161E-14 2.020E-12 9.289E-Il 2.440E-20 0.000 0.000 0.000 0.00$d 0 ,
@ 5.999E-13 2.227E-15 1.088E-13 4.883E-12 1.213E-21 0.000 0.000 0.000 0.000 ,
D 3.557E-01 7.724E-02 7.986E-14 1.138E-02 6.801E-12 2.682 845.506 26.812 2.681
@ 5.025E-02 5.982E-03 1.957E-18 2.232E-04 1.629E-13 0.510 215.509 7.415 0.741 2 2.324E-05 5.908E-06 1.480E-24 4.791E-11 2.641E-20 0.000 0.047 0.001 0.000 0 5.058E-08 6.047E-09 0.000E+00 0.000E+00 0.000E400 0.000 0.000 0.000 0.000 S 5.715E-02 1.293E-02 7.295E-18 2.852E-04 1.664E-13 0.433 132.522 4.177 0 418 B 3.365E-03 4.ll6E-04 8.903E-21 9.950E-07 7.248E-16 0.034 14.402 0.4 % 0.050 t l
l 9903160260 -O l
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7 TROJANLICENSE TERMINATIONPLAN
- 3. IDENTIFICATION OF REMAINING SITE DISMANTLEMENT ACTIVITIES
3.1 INTRODUCTION
In accordance with 10 CFR 50.82(a)(9)(ii)(B)(Reference 3-1), this section identifies the major dismantlement and decent:mination activities that remain at TNP as of early-1999. This information details those areas and equipment that need further remediation to allow an estimation of the radiological conditions that may be encountered during remediation. Included herein are schedules for implementation of decommissioning and dismantlement activities, estimates of associated occupational radiation dose, and projected volumes of radioactive waste. The activities discussed herein are the same plant activities that already may be conducted under the approved TNP Decommissioning Plan (Reference 3-2) and the TNP DSAR (Reference 3-3), and thus are not anticipated to involve unreviewed safety questions. In any case, appropriate reviews are conducted in accordance with 10 CFR 50.59 (Reference 3-4) and approved plant procedures to ensure that proposed activities do not involve unreviewed safety questions or changes to the TNP Technical Specifications, and to ensure that they do not impact the safe storage of spent fuel. The TNP facility currently is nearing the end of the Transition Period, and PGE is conducting active decontamination and dismantlement activities in accordance with the approved TNP Decommissioning Plan. Decommissioning activities are coordinated with the appropriate Federal and State regulatory agencies in accordance with plant administrative procedures. In ( order to minimize adverse impact by ongoing decommissioning activities, the Spent Fuel Pool ( (SFP) and other systems associated with safe fuel storage have been mechanically and electrically isolated to create a SFP " island." Dismantlement and decontamination activities continue to be performed as described in Section 3.2 while taking into account specific system considerations of Section 3.3. Intended to provide general information and guidance for work package planning. Section 3.3.2 describes the
- major remaining components of contaminated plant systems and, as appropriate, a description of specific equipment remediation considerations. Table 3-2 contains a list of major components removed during each year since 1996.
F #% 4 3-1 March 1999
TROJANUCENSE TERWNAHONPLAN r i 3.2 DECOMMISSIONING ACTIVITIES. TASKS. AND SCHEDULES This section presents a description of activities and tasks associated with TNP decommissioning. l Detailed planning precedes initiation of each specific activity, and includes engineering design, i as low as reasonably achievable (ALARA) planning, and refinement of cost, schedule, and requimd resources estimates. 3.
2.1 INTRODUCTION
TNP decommi.ssioning is divided into two broad periods: a Transition Period and a l Decontamination and Dismantlement Period. Figure 3-10 illustrates how these periods are _j incorporated into the overall decommissioning schedule. This schedule was used in the preparation of the decommissioning cost estimate' and funding plan discussed in Section 7. More detailed scheduling is prepared as part of pre-job planning. j The Transition Period began with permanent plant clopure in January 1993 and will continue ! until the spent fuel is transferred to the ISFSI. Decontamination and dismantlement of the remaining facility radioactive systems, components, and structures are scheduled to be conducted . upon completion of the transfer of spent fuel to the ISFSI. For planning purposes, it was ! assumed that spent fuel will be stored in the spent fuel pool until 1999, at which time it will be transferred to the ISFSI. Some decontamination and dismantlement activities have occurred and , will continue to occur during the Transition Period. Major activities completed or planned during the Transition Period are described in Section 3.2.2. . Following the Transition Period, the remaining decontamination and dismantlemer.t activities are expected to last from early 2000 to the end of 2002. Major activities planned during the t Decontamination and Dismantlement Period are discussed in Section 3.2.3.
)
Nonradiological site restoration activities involving the final disposition of structures, systems, and components are scheduled to be completed following the termination of Facility Operating (Possession Only) License NPF-1. Some site restoration activities have been completed and . others may continue to be conducted during the Transition and Decontamination and ! Dismantlement Periods. 3.2.2 TRANSITION PERIOD Plant closure activities were initiated following the decision to permanently cease TNP power operations in January 1993. These activities culmirnted with the plant in a safe transition state awaiting decontamination and dismantlement. Detailed project planning and engineering activities for the Decontamination and Dismantlement Period, as discussed in Section 3.2.3.2, will begin during the Transition Period. Plant activities will continue to be implemented in compliance with the existing possession-only license and other regulatory requirements. i O 3-2 March 1999
\
TROJANLJCENSE TERMINATIONPLAN l Removal of the four steam generators and p:essurizer has been completed. These components i were disposed of at the US Ecology low level radioactive waste disposal facility near Richland, O Washington. Removal of the steam generators _and pressurizer was accomplished through a new opening in the south face of the Containment Building. That opening is equipped with a door so
)
that the Containment Building can be maintained in a closed condition when the opening is not i in use. Prior to removal of the components from the Containment Building, low-density cellular ; concrete was placed inside each steam generator and the pressurizer. The concrete fixed internal l
, contamination and provided additional shielding.
Each component was moved via an intemal rail system out of the Containment Building, loaded - by a gantry crane onto a multi-wheeled transporter, moved to a preparation area within the TNP l Industrial Area, and prepared for river barge shipment. The component, transport cradle ! assembly, and transporter were then moved as an integral unit onto a barge at the TNP barge slip and shipped up the Columbia River to the Port of Benton, Washington. The multi-wheeled l transporter was used to off-load the barge and move the component to the US Ecology facility i near Richland, Washington. i During 1999, PGE intends to remove the reactor vessel with intemals intact (reactor vessel l package) from the 10 CFR 50 licensed area of the TNP site. As authorized by NRC letter dated l October 29,1998 (Reference 3-5), the reactor vessel package will be transported for disposal at ! the US Ecology low level radioactive waste disposal facility near Richland, Washington.- Removal of the reactor vessel package from the 10 CFR 50 licensed area of the TNP site will r eliminate approximately 2 million curies of activity from the TNP. Not including the spent nuclear fuel that will be transferred to the ISFSI, removal of the reactor vessel and internals will result in elimination of greater than 99 percent of the remaining activity (curies) at the TNP facility. Additional activities that were completed or are in process during the Transition Period include, but are not limited to the following:
- 1. Assessment of the functional requirements for plant systems, structures, and components.
Plant systems, structures, and components needed to support safe storage of the spent fuel, support spent fuel pool cooling, and facilitate ongoing plant activities ! have been identified.
- 2. Deactivation / removal of plant systems, structures, and components.
A comprehensive plant lay-up program was developed and is being implemented as described in Section 3.2.3.7. Systems, structures, and components not required to support decommissioning and spent fuel storage continue to be decontambiated and/or removed in accordance with the possession only license and approved ; plant procedures.
\
O 3-3 March 1999
TROJANIJCENSE TERMINATIONPLAN i
- 3. Redefinition of regulatory basis for the defueled plant. l P
On July 31,1993, PGE submitted a request to revise the TNP Tecimical Specifications to reflect the permanently defueled status of the plant, That request , was supplemented by PGE on March 8,1994. On March 31,1995, the NRC ' issued Amendment #194 to Facility Operating (Possession Only) License NPF-1. This amendment revised the TNP Technical Specifications to reflect the permanently defueled condition of the plant, and regulatory requirements and ' operating restrictions to ensure the safe storage of spent fuel in the spent fuel
- pool.
l The Final Safety Analysis Report was revised to reflect the permanently defueled ! plant condition and was retitled "Defueled Safety Analysis Report." The DSAR was transmitted to the NRC on October 7,1993. Additional licensing basis documents were also revised to reflect the plant's defueled condition. PGE submitted a proposed TNP Decommissioning Plan and Environmental j Report Supplement on January 26,1995 (Reference 3-6), which were approved > by the NRC on April 15,1996 (Reference 3-7). The TNP facility since has been : undergoing active decontamination and dismantlement in accordance with the : approved TNP Decommissioning Plan. i
- 4. Assessment of the plant's radiological status.
Section 2 presents an assessment of the radiological status of TNP. This assessment was used in developing the approved TNP Decommissioning Plan and this License Termination Plan.
- 5. Licensing'and Construction of the TNP ISFSI.
PGE submitted a 10 CFR 72 license application to the NRC for construction and operation of an ISFSI in March 1996 (Reference 3-8). An Environmental
- Assessment related to the construction and operation of the Trojan ISFSI and a Finding of No Significant Impact were completed by the NRC staff and transmitted by letter dated November 25,1996 (Reference 3-9). In addition, the Physical Security Plan for the Proposed Trojan ISFSI was approved by the NRC staff by letter dated November 20,1996 (Reference 3-10). Both of these approvals were contingent on PGE receipt of a 10 CFR 72 license for the ISFSI.
As documented in the approved PGE-1074, " Trojan Nuclear Plant Final Survey i Report for the ISFSI Site" (Reference 3-11), PGE performed a final survey of the ! ISFSI area to support removal from the 10 CFR 50 license concurrent with l issuance of the 10 CFR 72 license. Spent fuel will be transferred from the spent ! fuel pool to the ISFSI in order to facilitate decontamination and dismantlement. i Once the fuel is transferred to the ISFSI, the Transition Period ends and the I Decontamination and Dismantlement Period begins. O 3-4 Marth 1999
*s - * -T- I~t
TROJANLICENSE TERMINATlONPLAN ~ '
, 3.2.3 DECONTAMINATION AND DISMANTLEMENT PERIOD 3.2.3.1 100erview This section presents a general description of the Decontamination and Dismantlement Period activities for TNP decommissioning. These activities involve the reduction of radioactivity to acceptable levels, allowing for release of the site for unrestricted 'use. This information provides the basis for development of programs and procedures for ensuring safe decommissioning and a
- basis for detailed planning and preparation of decontamination and dismantlement activities.
During this period, the remaining contaminated systems and components will be decontaminated or removed, packaged, and either shipped to an offsite processing facility, shipped directly to a -
- low level radioactive waste disposal facility, or handled by other altematives in accordance with applicable regulations.
Decontamination'of plant structures may be completed concurrently with equipment removal. I Decontamination of structures may include a variety of techniques ranging from water washing I to surface material removal. Contaminated. structural material may be packaged and either j shipped to a processing facility, or shipped directly to a low level radioactive waste disposal facility. Alternative disposal methods, in accordance with applicable regulations, may also be used.- Followind the removal or decontamination of contaminated systems, components, and structures, a comprehensive final radiation survey will be completed as described in Section 5. The survey will verify that radioactivity has been reduced to sufficiently low levels, as stipulated in 10 CFR 20.1402 (Reference 3-12), to allow the release of the site for unrestricted use. Upon completion of the final survey, PGE will submit a final survey report to suppon license termination.' 3.2.3.2 Detailed Plannine and Ennineerine Activities
~
Detailed project plans will continue to be developed in accordance with design control j procedures to support the decontamination and dismantlement activities. The plans are used to { develop work packages, support ALARA reviews, aid in estimating labor and resource
~
requirements, and track decommissioning costs and schedule. Work packages are used to implement the detailed plans and provide instructions for actual field implementation. The work packages address discrete units of work and include appropriate hold and inspection points. Administrative procedures control work package format and content, as well as the review and approval process. i O V 3-5 March 1999
TROJANLICENSE TERMINATIONPLAN 3.2.3.3' General Decontamination and Dismantlement Considerations As has been the practice during the Transition Period and in accordance with the approved TNP Decommissioning Plan, the following general decontamination and dismantlement , - considerations, as applicable, will continue to be incorporated into decommissioning work packages during the decontamination and dismantlement period. Specific considerations are presented in Section 3.3.
~ Dismantlement activities are currently reviewed to ensure that they do not impact the safe
- storage of fuel in the SFP. During the decontamination and dismantlement period dismantlement activities will be reviewed to ensure they do not impact the safe storage of fuel in the ISFSI licensed under 10 CFR 72. Work packages are implemented in accordance with j administrative controls that require evaluations in accordance with the requirements of 10 CFR 50.59.
Temporary shielding is used where practical for ALARA purposes during decommissioning activities. Some dismantlement activities may be performed under water for shielding purposes as well as contamination control. I As currently practiced at TNP in accordance with the approved TNP Decommissioning Plan, the capability to isolate or to mitigate the consequences of a radioactive release will continue to be maintained during decontamination and dismantlement activities. Isolaticu is the closure of - penetrations and openings to restrict transport of radioactivity to the environment. This consideration should not preclude the removal of penetrations and attachments to Containment, provided that openings are closed in a timely manner. Airbome radioactive particulate emissions will continue to be filtered, and effluent discharges monitored and quantified. Consideration is given to the following items:
- 1. Operation of the appropriate portions of the containment ventilation and purge system, or an approved altemate system, during decontamination and j dismantlement activities in the Containment Building; !
- 2. Operation of the appropriate portions of the Auxiliary Building and Fuel Building ventilation system, or an approved alternate system, during decontamination and :
dismantlement activities in the Auxiliary and Fuel Buildings;
- 3. Operation of the Condensate Demineralizer Building ventilation exhaust system during decontamination and dismantlement activities in the Condensate Demineralizer Building; and l I
- 4. Use oflocal high efficiency particulate air (HEPA) filtration systems for activities expected to result in the generation of airborne radioactive particulates (e.g.
grinding, chemical decontamination, or thermal cutting of contaminated components). L O 3-6 March 1999
h l TROJANUCENSE TERMINATIONPLAN Work activities are planned to minimize the spread of contamination. Contaminated liquids are Ov contained within existing or supplemental barriers and processed by a liquid waste processing system prior to release. To minimize the potential for spread of contamination, the following
~
considerations will continue to be incorporated into the planning of decommissioning work activities:
- 1. Covering of openings in intemally contaminated components to confme internal ,
contamination; l 2. Decontamination and dismantlement of contaminated systems, structures, and I l components by decontamination in place, removal and decontamination, or removal and disposal;
- 3. Removal of contaminated supports in conjunction with equipment removal or !
decontamination of supports in conjunction with building decontamination, - I
- 4. Removal of contaminated systems and components from areas and buildmgs pnor -
to structural decontamination (Block shield walls, or portions of other walls, ; ceilings, or floors may be removed to permit removal of systems and j components.), ,
- 5. Removal or decontamination of embedded contaminated piping, conduit, ducts, i plates, channels, anchors, sumps, and sleeves during area and building structural !
decontamination activities; i O 6. Use oflocal or centralized processing smd cutting stations to facilitate packaging i i of components removed in large pie ;es; and
- 7. Removal of small or compact plant components and parts intact, where feasible.
(This includes most valves, smaller pumps, some small tanks, and heat ; exchangers. These components could then be decontaminated in whole or part, and reduced to smaller dimensions in preparation for disposal or release.) ; 3.2.3.4 Decontamination Methods Contaminated systems and components typically are removed and sent to an offsite processing facility, sent to a low level radioactive waste disposal facility, or decontaminated onsite and released. Although large scale chemical decontamination is not anticipated as part of the TNP t decommissioning, limited application has been and may be used to reduce radiation dose rates prior to dismantlement or general area decontamination. Other decontamination methods typically include wiping, washing, vacuuming, scabbling, spalling, and abrasive blasting. Selection of the preferred method is based on the specific situation. Other decontamination technologies may be considered and used if appropriate. i Application of coatings and hand wiping may be used to stabilize or remove loose surface i contamination. Airborne contamination control and waste processing systems are used as 3-7 March 1999 y>+- - , , + - -
, , . - -. .--.-. -. -. _ - . - - -.- - -.~... - ..- - -_-.-
,e L 1
TROJANLICENSE TERMINATIONPLtN ' l necessary to' control and monitor releases. If structural surfaces are' washed to remove contamination, controls are implemented in accordance with approved plant procedures to ensure u; that wastewater is collected for processing by liquid waste processing systems. l Tanks and vessels are evaluated and, if required, flushed or cleaned prior to sectioning'and/or removal to reduce contamination levels and remove sludge. The following considerations are incorporated into tank and vessel sludge removal activities:
- 1. Precautions are taken to ensure that liquid inadvertently discharged from the tank is captured se orocessing by a liquid waste processing system;
- 2. Sludge removed from the tank is stabilized prior to shipment in conjunction with the TNP plant process control program; and 1
- 3. ~ Wastewater will be processed and analyzed before being discharged. i
)
. Concrete that is contaminated or activated may be removed and sent to a low level radioactive l
waste disposal facility, allowed to decay below site release criteria, or handled by other methods ! in accordance with applicable regulations. Removal of concrete should be performed using i methods _ that control the removal depth to minimize the waste volume produced. Vacuum j removal of the dust and debris with HEPA filtration of the effluent should be used to minimize l the spread of contamination and reliance on respiratory protection measures. l l 3.2.3.5 Dismantlement Methods O- Dismantlement methods can be divided into two basic types: disassembly, and cutting or other destructive methods. Disassembly generally means removing fasteners and components in an orderly non-destructive manner (the reverse of the original assembly). Cutting methods include flame cutting, abrasive cutting, and cold cutting.
- Flame cutting includes the use of oxyacetylene and other gas torches, carbon are torches, air or oxy are torches, plasma are torches, cutting electrodes, or combinations of these. Most of the torches can either be handheld or operated remotely with the appropriate devices. Abrasive cutting includes the use of grinders, abrasive saw blades, most wire saws, water lasers, grit blast,
- and other techniques that wear away metal. Cold cutting includes the use of bandsaws, bladesaws, drilling, machining, shears, and bolt / pipe / tubing cutters.
Selection of the preferred method depends on the specific situation. Other dismantlement
- technologies may be considered and used if appropriate. Dismantling of systems includes the
- removal of valves and piping for disposal. Most valves can be removed with the piping. Larger ;
- valves and valves with actuators may be removed separately for handling purposes. Valve l actuators that can be decontaminated are removed from the valves prior to pipe removal where
{ practical, rO 3-8 March 1999 r
TROJANLICENSE TERMINATIONPLAN 3.2.3.6 Removal Seauence and Material Handline Removal sequence may be dictated by access and material handling restrictions or by personnel !
. exposure considerations. In some cases, a top-down approach is used; materials and structures at i the highest elevations are removed first to allow access to components in lower levels. In other i cases, different approaches may prove more efficient.'
l In most cases, the first items removed are those that are not contaminated, or are only slightly ! contaminated, to preclude contamination by other equipment. However, personnel exposure i considerations may not always allow this option. Where non-contaminated equipment cannot be removed first, covers or other protection methods may be used. Similarly, non-contaminated piping should be removed from pipe chases and horizontal pipeways before cutting contaminated pipes. If this is not possible, other precautions, such as covers, are used to minimize the spread ofcontamination. ; Where rapid cutting techniques are available, pipes and equipment can be sectioned into pieces f that are manageable using light rigging or by manual lifting. Where slow cutting techniques are l used, the largest manageable pieces will typically be freed and further reduced at a more l convenient location. In the initial stages of decommissioning, most material removed from the Containment Building , was passed through the equipment hatch into the Fuel Building. The Fuel Building operating : floor pmvided a convenient location for handling and processing of materints because of the , availability of a crane and because it is a relatively open area. The areas above the holdup tanks - were used for decontamination, sorting, or packaging. Since mid-1998, when the Containment l Bubding was removed from the security protected area, the equipment hatch was closed and a ! new opening at grade elevation was created in the south Containment wall to support routme i removal of material from the building. ;
?
The plant is equipped with multiple cranes, hoists, and lifting and transpon systems. These systems can be used to lift and transpon components and equipment to suppon plant decommissioning activities. Forklifts, mobile cranes, front-end loaders, and other lifting and transpon devices also can be used for plant decommissioning activities. The major installed plant cranes, hoists, and liftin~g and transpon devices that are available to suppon decommissioning include:
- 1. Containment Building polar crane;
- 2. Fuel Building overhead crane;
- 3. Auxiliary Building electric hoist; "4.. Auxiliary Building elevator;
- 5. Equipment room monorails; O
V- 3-9 March 1999
. , . _ _ _ _,, -- - ,y+-- w-es
1 TROJAN LICENSE TERMINATIONPl.AN ' l
,6. Spent fuel pool bridge crane; and l 7.- Condensate Demineralizer Building bridge crane.
IaMon requirements for the Containment Building Polar Crane, Fuel Building Overhead Crane, Auxiliary Building Electric Holst, Spent Fuel Pool Bridge Crane, and the Condensate Demineralizer Building Bridge Crane are specified in Trojan Plant Maintenance Procedure MP l-20, " Cranes, Hoists and Winches." Chainfalls and other temporary hoists are inspected and verified to be in' good working condition at the time ofissue from the Tool Room. The ,
' Auxiliary Building Elevator is inspected by a State fa==*er in accordance with State of Oregon '
requirements.
"he Containment Building polar crane is capable of reaching most locations inside the . Containment Building and can handle large, heavy loads. The Fuel Building overhead crane has I
access to a hoistway open to plant grade at the 45 ft elevation. The Auxiliary Building elevator has access to upper floors in the building and can carry small loads. j installed cranes and hoists may be used in conjunction with temporary or mobile lifting and transport devices to support decommissioning. The installed plant cranes, hoists, and other j
- lifting devices can be decontaminated and dismantled when they no longer are required to !
support decommissioning activities. j 3.2.3.7 System Deactivation' ! O Systems or components will continue to be deactivated prior to decontamination and dismantlement. In general, deactivation is implemented by mechanical isolation ofinterfaces with operating plant systems, draining piping / components, and de-energizing electrical supplies. Combustible material (e.g., charcoal from filters, lube oil) is removed from'the deactivated comwnents where possible. Chemicals used in, or resulting from, decommissioning activities are controlled in accordance with the plant chemical safety program. P' ant drawings are revised
; to indicate deactivated portions of systems. Plant procedures are modified to reflect the changes.
Deactivation ofplant systems is administratively controlled by approved plant procedures. Deactivation plant are established to implement the desired system valve lineup changes and electrical isolations.' The design change process is used to remove components, lift electrical leads, install elec trical,iumpers, cut and cap piping systems, or install blank flanges. Plant procedures previde controls over the operation of deactivated system boundary valves. As additional' systems are deactivated, existing isolation boundaries are re-evaluated and changed, as .
- necessary, to reflect the new p! ant condition. Boundary valves are tagged for identification.
3-10 March 1999
TROJANIJCENSE TERMINATIONPLAN 3.2.3.8 Iercoorary Systems to Suroort Decommissionine + Decontamination and dismantlement of tyrtems, structures, and components often require the removal ofinterferences. Removal of Scne of these interferences may eliminata power, service air, and other services used for decommi::::bning. It map also become impractic at some point to continue using installed plant systems. 'lempormy 7.mices and systems can be provided to support decommissioning activities. Temporary modifications to plant structures, systems, and components are controlled by plant design control procedures. Portable electric power packs can be powered from motor control centers, load centers, or the yard loop. These portable load centers can supply cutting, hoisting, temporary lighting, or other power needs. Service air can be provided by portable air compressors using hoses or temporary air manifolds. Demineralized water is available from portable demineralizer skids or portable tankers brought from offsite. Portable hydraulic power centers can be used to power hydraulic equipment. Temporary liquid and solid waste processing systems may be used during decommissioning for ; processing plant waste. These systems may include filters and/or demineralizers, and may be ! used at one or more locations in the waste-processing path.
' Portable radiation monitors and air monitoring equipment can provide localized radiation j monitoring. Localized temporary ventilation equipment and HEPA filtration may be used to !
supplement building ventilation and minimize the spread of radioactive particulate j contamination. O t v 1 i l /O ! (.) 3-11 March 1999
~ TROJANllCENSE TERMINATIONPLAN 3.3 REMAINING DISMANTLEMENT ACTIVITIES m- -
y; 3.3.1 IDENTIFICATION OF REMAINING SYSTEMS, STRUCTURES, AND COMPONENTS This section summarizes the structures, systems, and components remaining at TNP and the associated schedule and status of dismantlement and decontamination. Radiological implications, such as occupational exposure estimates, radioactive waste characterization, and radioactive material release estimates, associated with the remaining TNP site dismantlement activities are summarized in Section 3.4. The activities described herein are the same activities
' that already may be conducted under the approved TNP Decommissioning Plan and the DSAR.
The remaining dismantlement and decontamination activities can be classified into several phases, the implementation of which may overlap. The current phase includes removal or in-place decontamination of contaminated systems and components not required for support of fuel
- storage or subsequent decontamination activities. The second phase includes the removal or decontamination in place of systems and components that support the SFP or fuel storage. The third phase involves the remediation (removal or decontamination in place) of remaining structures, systems, and components, including the SFP itself. These phases primarily are implemented on an area-by-area basis. Using this approach, often only a part of the system will -
be removed, with the remaining portion awaiting removal in an adjacent area or being maintained in service to support spent fuel storage and defueled plant operation. The following Sections 3.3.1.1 through 3.3.1.3 list the major structures, systems, and components remaining at TNP as categorized into one of the three phases described above. The status of decontamination and dismantlement of these systems, structures, and components, as of January 1999, is summarized in Table 3-1. Table 3-2 contains a list of major components removed during each year since 1996. As indicated by Table 3-1 and Table 3-2, the majority of radiologically contaminated systems and components not required to support the storage of spent fuel have been deactivated, dismantled, and disposed ofin accordance with the TNP Decommissioning Plan. Additional detail regarding decontamination, dismantlement, and radiological controls for these structures, systems, and components is provided in Section 3.3.2. 3.3.1.1 Remainine Structures. Systems. And Components Not Reauired For Spent Fuel Storane { Phase 1) The remaining systems and components listed below are not necessary to support spent fuel storage. These systems and components will continue to be dismantled and decontaminated in accordance with Sections 3.2 and 3.3.2. These activities are consistent with those that already may be conducted under Sections 2.2.4 and 2.2.5 of the approved TNP Decommissioning Plan. The following systems are classified as Phase 1, with the applicable section of this TNP License Termination Plan referenced in parentheses:
. - Reactor Vessel and Intemals (Section 3.3.2.1) e Component Cooling Water System (Section 3.3.2.3) e Portions of the Service Water System (Section 3.3.2.4) 3-12 March 1999
r : F.- t TROJANJJCENSE TERMINATIONPL,AN ? y 1
- e- Containment Fuel Handling Systems and Refueling Cavity (Section 3.3.2.5) i 1e Original Spent Fuel Cooling and Demineralizer System (Section 3.3.2.6)
O e i Remaining Portions of Condensate Demineralizer System (Section 3.3.2.8)
'e Refueling Water Storage Tank (Section 3.3.2.10) i l
- l. , e ' Portions of the Plant Efiluent System (Section 3.3.2.11) i
, '* - Ponions of Containment Ventilation Systems (Section 3.3.2.12) e ' Portions of the Fuel and Auxiliary Building Ventilation Systems (Section 3.3.2.13) l e ' Portions of the Compressed Air Systems (Section 3.3.2.15) ;
l >e Gaseous Radioactive Waste System (Section 3.3.2.16) ;
'e . Solid Radioactive Waste System (Section 3.3.2.17) ?
- Portions of the Liquid Radioactive Waste System (Sections 3.3.2.18) i e . Portions of the Radiation Monitoring System (Section 3.3.2.19)
- Process Sample System (Section 3.3.2.20) e . Portions of the Fire Protection System (Section 3.3.2.21) l e Portions of the Electrical Systems (Section 3.3.2.22) . !
e Portions of contaminated embedded piping'.
. l 3.3.1.2 Remainine Systems. Structures. And Cow-:.u.a Associated With Snent Fuel Storane j (Phase 2) l l~ After the spent fuel and spent fuel debris have been removed from the Spent Fuel Pool and transferred to the ISFSI, components and systems listed below will be dismantled and/or decontaminated in accordance with Sections 3.2 and 3.3.2. Some components within the Spent l O P > Peei- x 6 -e a Pri 4 -Pi tie er P * < > -e administrative controls will continue to be used to ensure removal' activities are performed in a-
> etiviti : ^PP Pri - t manner that will not adversely impact the safe storage of spent fuel.
The following list of structures, systerns, and components are classified as Phase 2, with the applicable section of this License Termination Plan referenced in parentheses:
- Contaminated embedded piping'
- Boric Acid Batch Tank (portion of CVCS System)(Section 3.3.2.2) e- Portions of the Service Water System (Section 3.3.2.4) 'l e- Spent Fuel Storage and Handling Equipment (Section 3.3.2.5)
- Modular Spent Fuel Pool Cooling and Cleanup System (Section 3.3.2.7) - .
C Portions of the Fuel and Auxiliary Building Ventilation Systems (Section 3.3.2.13)
* ' Portions of the Compressed Air Systems (Section 3.3.2.15)
)
j e Portions of the Liquid Radioactive Waste System (Sections 3.3.2.18) i
' The majority of contaminated embedded piping.'which primarily includes piping from various radioactive waste dra:n 6;; ems, embedded ventilation ductwork, and buried process piping. is expected to be decontaminated to ;
.-9 acceptable levels as adopted in Section 5, Final Survey Plan. to satisfy site release criteria. ' Ponions of the piping j may be removed. '
13 March 1999
-t ,
,, .. - , , - , . - ,.,.p9-
- . ~ - - .. -
TROJANLICENSE TERMINATIONPLAN i
;e' Portions of the Radiation Monitoring System (Section 3.3.2.19) e Portions of the Fire Protection System (Section 3.3.2.21) pd- e ~ Portions of the Electrical Systems (Section 3.3.2.22) i 3.3.1.3 - Remainina Structures. Systems. And Components (Pha=a 3) i Contaminated structural concrete, steel, and other building materials and remaining components ,
of systems listed below will be removed or decontaminated in place in a manner consistent with Sections 3.2 and 3.3.2. The SFP itselfwill be remediated following removal of fuel and draining o f the SFP water. Decontamination of the structures will be performed on an area-by-area basis, ; with the majority of the decontamination activities occurring following completion of equipment ! removal from an area. Demolition of the building structures is not anticipated prior to j termination of the TNP 10 CFR 50 license. The following list of structures, systems, and components are classified as Phase 3, with the applicable section of this License Termination Plan referenced in parentheses. j e Service Water System (Section 3.3.2.4) . e' Spent Fuel Pool (Section 3.3.2.5) e Plant Effluent System (Section 3.3.2.11)
~
- Containment Ventilation System (Section 3.3.2.12) ,
e Fuel and Auxiliary Building Ventilation Systems (Section 3.3.2.13) !
- Condensate Building Ventilation System (Section 3.3.2.14) e Compressed Air Systems (Section 3.3.2.15) -
e Liquid Radioactive Waste System (Sections 3.3.2.18)
- Radiation Monitoring System (Section 3.3.2.19)
*- Fire Protection System (Section 3.3.2.21) e' Electrical Systems (Section 3.3.2.22) e Containment Building (Section 3.3.2.23) !
- Auxiliary Building (Section 3.3.2.24) i
- Fuel Building (Section 3.3.2.25)
)
e Other Buildings (Section 3.3.2.26) j 3.3.2 - GENERAL DESCRIPTION OF AND REMEDIATION CONSIDERATIONS FOR REMAINING SYSTEMS, STRUCTURES, AND COMPONENTS l This section presents a summary description of the remaining TNP systems, components, and I structures that are known or considered to be intemally contaminated or that may be used to support decommissioning activities. This discussion includes general activities and remediation considerations associated with decommissioning these systems, structures, and components. For reference, a site area map current as of January 1999 is provided in Figure 3-1. Plant layout and general arrangement drawings reflecting the TNP facility at the time of plant shutdown are provided in Figures 3-2 through 3-9. ; Because extemal contamination is generally considered to exist on systems, structures and components located in the RCA's of the plant, it is not specifically noted in the following system 3-14 March 1999
TROJANLICENSE TERMINATIONPLAN discussions. However, systems, components, and structures that are extemally contaminated will l p be decontaminated for release or disposed of as radioactive waste. Q This section is intended to provide general information and guidance for work package planning and is not required to be updated to reflect equipment removal. However, Table 3-2 is updated annually to provide a list ofmajor plant systems and system components that were removed the previous year (beginning with 1996). 3.3.2.1 Reactor Vessel and Intemals During 1999, PGE intends to remove the reactor vessel with intemals intact (reactor vessel package) from the 10 CFR 50 licensed area of the TNP site. As authorized by NRC letter dated October 29,1998, the reactor vessel package will be transported for disposal at the US Ecology I low level radioactive waste facility near Richland, Washington. Removal of the reactor vessel l - package from the 10 CFR 50 licensed area of the TNP site will eliminate approximately i 2 million curies of activity from the TNP. Not including the spent nuclear fuel that will be transferred to the ISFSI, removal of the reactor vessel and intemals will result in elimination of greater than 99 percent of the remaining activity (curies) at the TNP facility. PGE letter VPN-012-99, dated February 2,1999 (Reference 3-13), forwarded to the NRC l Revision 0 of the TNP Reactor Vessel and Intemals Removal Project Safety Analysis Report, PGE-1076 (Reference 3-14). This approved safety analysis report served as the basis for PGE's application for a Certificate of Compliance to allow a single shipment of the TNP reactor vessel package as an exclusive use shipping package in accordance with the requirements of O 10 CFR 71," Packaging and Transportation of Radioactive Material"(Reference 3-15). As approved by the NRC, the reactor vessel package will be one-time shipped as an exclusive use Type B (as exempted) shipping package, as defined by 10 CFR 71.4. 3.3.2.2 Chemical and Volume Control Systal Located in the Fuel Building, the remaining portion of the CVCS primarily supports boron ;
~a ddition to the SFP, The only remaining major component is the boric acid batching tank. !
The boric acid batching tank is not internally contaminated. The tank can be removed in one piece or can be segmented. 3.3.2.3 Component Cooline Water System The major components of the Component Cooling Water (CCW) system have been removed. Poitions of the system piping remain. The system piping is not intemally contaminated. No specific considerations apply to the remaining system equipment. O 3-15 March 1999
TROJANLICENSE TERMINATIONPLAN
,3.3.2.4 Service Water System ;
' The service water system supplies raw water from the Columbia River via the intake structure
~
for emergency make-up to the spent fuel pool and for dilution flow for plant liquid discharges. Raadning major system components include three service water pumps and associated valves, piping, fittings, and instmmentation. The active ponion of the system piping is located in outside areas. The system is not intemally contaminated. No specific considerations apply to the remaining system equipment. ! 3 i 3.3.2.5 Soent Fuel Pool and Fuel Handlina Eauinment ne Spent Fuel Pool and fuel storage structures consist of the Spent Fuel Pool, the spent fuel '; storage racks, the fuel transfer canal, the cask loading pit, and the new fuel storage area. The ; spent fuel pool provides for irradiated fuel storage. Additionally, the spent fuel pool provides a t-sp.wnt radiation shield for personnel. Fuel assemblies are stored in stainless steel spent fuel ! storage racks located at the bottom of the spent fuel pool. The fuel transfer canal and cask : loading pit facilitated handling ofirradiated fuel by providing isolable underwater operating i areas for fuel transfer evolutions. The new fuel storage area provided a protected area for dry, !
- suberitical storage of new fuel assemblies, and now houses ponions of the modular SFP cooling and cleanup system (Section 3.3.2.7). He fuel transfer canal and cask loading pit are connected !
to the spent fuel pool by transfer slots which can be closed and sealed by leak-tight gates. The !' spent fuel pool, fuel transfer canal, and the cask loading pit are reinforced concrete structures with seam-welded stainless steel linings. O The spent fuel pool and remaining fuel handling equipment are located in the Fuel Building. The j l spent fuel pool supports spent fuel storage and will be required until the fuel is moved to the ISFSI. The fuel handling equipment in the spent fuel pool may be required to transfer the fuel to the ISFSI. Additionally, the spent fuel pool bridge crane may be required to support the decommissioning of the spent fuel pool. The system is contaminated. The following specific considerations apply. 3 The potential for high levels of contamination exists for components removed from the spent fuel pool. The spent fuel storage racks are acccesible with the Fuel Building crane and could be removed from the spent fuel pool intact for sectioning and packaging at another location. The liner of the spent fuel pool, fuel transfer canal, and cask load pit may be sectioned for removal. The fuel transfer tube and sleeve may also be sectioned for removal. The fuel handling cranes, the fuel transfer cart, its tracks, and the upender frames may be sectioned into manageable pieces ' for removal. It may be possible to decontaminate and release sections of the liners, and the spent ! fuel storage racks. l Dismantlement of the majority of the spent fuel pool and fuel handling radioactive systems, ! components, and structures is scheduled to be conducted upon completion of the transfer of spent fuel to the ISFSI. Spent fuel will be in the ISFSI, and other items stored in the spent fuel pool I will be moved to alternate storage locations or disposed of, prior to the spent fuel pool liner being sectioned for removal or decontaminated in place. 3-16 March 1999
TROJANLJCENSE TERMINATIONPLAN
- 3.3.2.6- Spent Fuel Pool Cooling and Demineralizer System (Orininal System)
The SFP cooling and demineralizer system was replaced by the modular SFP cooling and
- cleanup system in 1998. Remaining major system components include a skimmer pump and associated valves, piping, and instrumentation. Portions of the system piping which formerly entered the SFP have been have been capped on the exterior of the SFP wall, limiting the SFP piping wetted surface area to a minimum. The modular SFP cooling and cleanup system now provides cooling and water purification functions.
No specific considerations apply to the remaining system equipment. The skimmer filter housing and pump are easily accessible and can be removed intact. 33.2.7 Modular Soent Fuel Pool Cooling and Cleanuo System i The Modular SFP Cooling and Cleanup System removes the decay heat from the spent fuel elements stored in the SFP and purifies the system water inventory. Major equipment associated with the system includes two pumps, two water-to-air coolers, a demineralizer, a filter, and associated valves, piping, fittings, and instrumentation. No specific considerations apply to removal of this system or its components. The system is leased and the components are skid mounted and will likely be removed intact for retum shipment to the supplier when their SFP support function is no longer required. , 3.3.2.8 Condensate Demineralizers O The remaining major system components include a backwash system and associated valves, piping, fittings, and instrumentation. The system equipment is located in the Condensate 1 Demineralizer Building. The backwash receiver tank is potentially internally contaminated. i Other portions of the system have detectable levels of contamination. ) i No specific considerations apply to the remaining system components. 3.3.2.9 Steam Generator Blowdown System The major components of the steam generator blowdown system have been removed. Portions of the system piping remain and are included in the embedded piping scope. No specific considerations apply to the remaining system equipment. 3.3.2.10 Primary Makeun Water System and Refueling Water Storane Tank The remaining major system components include the refueling water storage tank (RWST) and portions of the associated system piping, which is included in the embedded piping scope. The RWST is a venical tank constructed of austenitic stainless steel with immersion heaters. The RWST is located in the tank farm (south of the Containment Building). The RWST and system piping is internally contaminated. 3-17 March 1999 l
TROJANUCENSE TERMINATIONPLAN 1 1 i The followmg specific considerations apply.' Due to its size, the RWST will be sectioned to (' .. facilitate packaging and shipping. It may be possible to decontaminate and release sections of l the tanks' ; 3.3.2.11 Plant Emuent System'
' The Turbine Building sump, oily water separator, solids settling basin, and discharge and i dilution stmeture together comprise the plant emuent system. The plant emuent system ,
provides a means of discharging plant liquid wastes while ensuring compliance with the National ! Pollutant Discharge Elimination System Waste Discharge Permit. l The plant emuent system components are located throughout the plant site. Portions of the system are required to support collection and disposal of waste generated by decommissioning i
- activities. The Turbine Building sump is contaminated. Other portions of the system, such as ;
the oily water separator, may also be contaminated; they are still being used to support ongoing plant activities. These components can be sampled for contamination when they are no longer in ; use. { i The following specific considerations apply. Turbine Building sump contamination will be removed using concrete decontamination techniques described in Section 3.2.3.4. Sump input ; and discharge piping will be checked for contamination to determine the proper method of disposal. Buried and embedded piping may be left in place ifit meets site release criteria. The emuent diffusion pipe can be removed by divers if determined to be necessary. 3.3.2.12 Containment Ventilation Systems The remaining Containment ventilation systems consist of the purge supply and purge exhaust systems. Containment purge exhaust is directed to the primary vent stack that is attached to the outside of the Containme'n t Building. Exhaust air is monitored for radiation and is exhausted through a vent at the top of the Containment Building. Portions of the Containment ventilation systems are located inside the Containment Building, the Auxiliary Building, and the Main Steam Support Structure. The primary vent stack is attached to the outside of the Containment Building. Portions of the system will be used to maintain a habitable environment and control contamination during decommissioning. The systems are internally contaminated. The following specific considerations apply. The systems will remain in service until:
- 1. The individual system or component has been evaluated as not required to support further decommissioning activities; or
- 2. An altemate system has been established; or
- 3. Contaminated components for all systems in the building have been removed or 4
remediated, and the building has been decontaminated. 3 18 March 1999
TROJANIJCENSE TERMINATIONPLAN 3.3.2.13- Fuel Building and Auxiliary Building Ventilation Systems The Fuel Building and Auxiliary Building ventilation systems provide for the supply, heating, , cooling, and exhaust of air for the Fuel and Auxiliary Buildings. The remaining systems include several subsystems: Fuel and Auxiliary Building supply system, Fuel and Auxiliary Building
- exhaust system, spent fuel pool exhaust system, space heating system, radioactive waste annex
' supply and retum, and pump cooling units. Air is exhausted through the primary vent stack that I is attached to the outside of the Containment Building. Exhaust air is monitored for radiation.
The other systems provide heating and cooling for specific areas in the buildings.
; The Fuel and Auxiliary Building ventilation systems are located inside the Fuel and Auxiliary Buildings. Partions of the systems will be used to maintain a habitable environment and control contamination during decommissioning. The systems are internally contaminated. The following specific considerations apply.
The systems will remain in service until: l
- 1. The individual system or component has been evaluated as not required to support further decommissioning activities; or 2.' An altemate system has been established; or
- 3. Contaminated components for all systems in the building have been removed or remediated, and the building has been decontaminated.
O 3.3.2.14 Condensate Demineralizer Buildino Ventilation System ; i The Condensate Demineralizer Building ventilation system provides for supply and exhaust air in the building. Supply air is provided through infiltration and, as appropriate, roof supply fans. Exhaust air is monitored using a sample pump, sample probe, and a radioactive airborne particulate monitoring filter. l The system will remain available for service until the Condensate Demineralizer Building is no longer used to process radioactive waste, and the building is decontaminated. 3.3.2.15 Instrument and Service Air Svilfm The instrument and service air system supplies compressed air required for pneumatic
' instruments, valves, and service air outlets throughout the plant. The system has four air compressors, aftercoolers, air receivers, filters, dryers, and associated valves, piping, fittings, and instrumentation.
The instrument 'and service air system is located in _ buildings throughout the plant. The system may be used for operation _of control valves, dampers, tools, and breathing air. As a portion of
' the instrument and service air system is determined not to be required to support further decommissioning, it may be deactivated and removed. The system is not considered to be b
U intemally contaminated. 3-19 March 1999
~. _ _ _ _ _- -
_ ___>- - ~ . _ _._._. . _ _ __ _ __ _ _ _ . TROJANLICENSE TERMINATIONPIAN L I
. 3.3.2.16 e Gaseous Radioactive Waste System I O The majority of gaseous radioactive' waste system components have been removed and the entire system has been removed from service. Remaining major equipment includes the vent collection :
' header exhaust fan and associated piping, fittings, filters, and instrumentation. Gaseous ;
' radioactive waste system is located in the Auxiliary and Fuel Buildmgs. ,
The following specific considerations apply. The gas collection header and vent collection header exhaust filter housings can be removed intact. 3.3.2.17 Solid Radioactive Waste System
. Major remaining equipment associated with the solid radioactive waste system includes a filter :
handling vehicle, and some piping. Solid wastes generated as a result of plant system operation or d~ ammiraioning activities are processed in accordance with the site Process Control l Program. The solid radioactive waste system components are located in the Auxiliary and Fuel Buildings. The system is contaminated. ) No specific considerations apply for the remaining system equipment. 3.3.2.18 Liouid Radioactive Waste System ; 1 I The liquid radioactive waste system collects, stores, processes,'and disposes of contaminated O J liquids, including radioactive wastewater generated during decommissioning activities. System ' components are used to process water and monitor it during discharge. Major remaining system equipm:st includes two treated waste-monitor tanks and pumps, the dirty waste drain tank and pumps, the Auxiliary Building sumps and pumps, Containment sumps, various filters, and associated valves, piping; fittings, and instrumentation. The liquid waste system components and
. piping are located primarily in the Auxiliary and Fuel Buildings and the Containment Building,
. with several floor and equipment drains located within the Main Steam Support Structure and
- Control Building. The system is internally contaminated.
The following specific considerations apply. System tanks can be sectioned to facilitate , removal. System pumps are small and can be removed intact. The filter skids will be dismantled 4 and the bag filter housings removed intact. System sumps are concrete pits that will be decontaminated. Techniques for concrete decontamination and demolition are noted in Section 3.2.3.4. Temporary water cleanup systems may be used to reduce the amount ofinstalled equipment ~ required to remain operational and still maintain the ability to collect, store, process and discharge radioactive liquid waste. Plumbing modifications may be required to use temporary systems. Temporary systems to support decommissioning are discussed in Section 3.2.3.8.
. Radioactive liquid effluents will be monitored and released in accordance with the requirements of topical report PGE-1021, "Offsite Dose Calculation Manual" (ODCM) (Reference 3-16).
- O 3-20 March 1999
TROJANLICENSE TERMINATIONPLAN 3.3.2.19 Eadiation Monitorinn System . The radiation monitoring system consists of the process and effluent radiological monitoring systems (PERMS) and the area radiation monitoring system (ARMS). The PERMS provide monitoring of gaseous and liquid effluent release paths and selected gaseous and liquid plant 1 yocess systems. The ARMS provide remote monitoring of selected areas at the plant. Portions ! of the radiation monitoring system are considered to be contaminated. 1 The following specific considerations apply. The following PERMS channels will be used during decommissioning activities for effluent monitoring. Temporary power or monitonng 1 channels may be used to support decommissioning. l
- 1. PRM-9 (liquid radioactive waste effluent discharge) monitors liquid waste effluent ' i during discharges. The PRM may be moved from its curnmt location due to possible discharge line plumbing modifications.
l
- 2. PRM-2A (auxiliary building ventilah discharge - particulate channel) is designed to monitor the discharge ofradioactive particulates through the Auxiliary Building l exhaust ventilation stack.
J
- 3. PRM-1 A (containment building ventilation discharge - particulate channel) is designed l
. to monitor the discharge of radioactive particulates through the Containment Building !
exhaust ventilation stack. While spent fuel is in the spent fuel pool, PERMS channels PRM-2A and PRM-2C (auxiliary 01 building ventilation exhaust low level gas channel) and ARMS channels ARM-12 (fuel building, elevation 93 ft, machine shop access) and ARM-13 (fuel building, elevation 93 ft, new fuel storage area) will be required. Localized radiation monitoring will be provided in work areas by temporary monitoring instrumentation, when necessary. Due to the mounting configuration of many of the PERMS detectors, they are not likely to be significantly contaminated. The sample tubing and supply piping for PERM-1 (Containment Ventilation) and PERM-2 (Auxiliary Building Ventilation) are potentially contaminated and will be sectioned as necessary for removal. The skid housings should be dismantled to facilitate removal ofvarious detectors, lead detector housings, and sample piping. Sample pumps should be removed intact for disposal. 3.3.2.20 Process Samoline System 1 The major components of the process sampling system have been removed and required
- sampling is performed locally. Portions of the system piping remain. Portions of the system are ,
intemally contaminated. No specific considerations apply to the remaining system equipment. '
)
~
O 3-21 March 1999 I b
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I TROJANIJCENSE TERMINATIONPLAN ! l
'3.3.2.21 Fire Protection System
':- Q The fire protection system provides manual and automatic fire suppression and automatic fire detection for plant areas. The fire protection system includes the following: portable fire
. extinguishers, water supply and distribution systems, fire suppression system, emergency lighting, and the fire detection and alann system. TIhe main fire pumps are located in the intake structure.
4 The fire protection system is located in buildings and areas throughout the plant site. The system 4 is not considered to be intemally contaminated. The following specific considerations apply. Sections of the fire protection system may be deactivated and armoved from service when no , longer required to .:upport further decommissioning activities or after altemate fire detection and I suppression capability has been established. Such changes are changes to the TNP Fire j Protection Program, and will be made in accordance with the provision of License ; Condition C.(8) of Facility Operating (Possession Only) License NPF-1 and 10 CFR 50.59. l 3.3.2.22 Electrical Systems The electrical system includes the main generator, the switchyard, main and auxiliary transformers, and the 230 kV ac,12.47 kV ac,4160 V ac,480 V ac,120 V ac,250 V de, 125 V de, and lighting distribution systems. The electrical systems are located in buildings and areas throughout the plant site. Portions of j the systems (primarily 12.47 kV ac yard loop,480 V ac, and 120 V ac) may be used to support decommissioning activities. The systems are not considered to be intemally contaminated. ' The following specific considerations apply. .; I Temporary electrical services may be used as required during decommissioning to facilitate dismantling and removal of plant components. When a system or component is no longer required, the electrical supply to the component may be isolated and removed. Plant lighting - may also be required until building demolition. Lighting and electrical power may be provided by temporary services as discussed in Section 3.2.3.8. 3.3.2.23 Containment Buildine The Containment Building consists of two structures on a common foundation. One is the containment itself; the other is the internal structure, referred to as the containment internals, whose function is to provide biological shielding. Supports for equipment, operating decks, access stairways, and platforms are included in the containment internals. The two structures are structurally separated above the foundation by a gap based on considerations of maximum relative displacement during an earthquake. Most equipment has been removed from the Containment Bu' %g. Remaining major equipment in the Containment Building includes the reactor ana iefueling cavity and the
. overhead polar crane. The Containment Building is a fully reinfo ced concrete structure in the shape of a cylinder with a hemispherical roof and flat foundation. The approximate dimensions 3-22 March 1999
- y. _ _ _ _ - _ _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ - _ ._ .. _ ____. _ .
TEOJANLICENSE TERMINATIONPLAN of the Containment Building are: 124-ft inside diameter,203-ft inside height,3%-ft wall ! thickness and 2%-ft dome thickness. O The following discussion provides genend information about the construction details of the l l Containment Building. The reactor cavity and the instrumentation tunnel are located below the foundation slab. The cylindrical section has a post-tensioning system consisting of venical and hoop tendons. The dome has a tyco-way post-tensioning system consisting of hoop tendens and continuous vertical tendons. Containment tendons have been detensioned, and some tendons have been removed, The inside of the concrete shell is steel-lined. The liner plate is coated with an epoxy-phenolic finish that is approximately 5 mils thick generally to a height of 6 ft above the floors, and 2 to 3 mils inorganic topcoat above that. Penetrations in the Containment Building include the equipment hatch, two personnel air locks, and numerous smaller electrical and mechanical penetrations. Additionally, since decommissioning activities started, openings with doors have been created in the south wall of containment to support removal oflarge components and for general material removal considerations. The following specific considerations apply. Ponions of exposed surfaces inside the Containment Building are contaminated. Intemal concrete walls can be decontaminated by water or chemical washing. Surfaces that can not be i decontaminated can be scabbl . or surface ground down to non-contaminated depths. Portions I
- of concrete structures inside the Containment Building (e.g., the primary shield wall) are l activated and could require removal oflarge sections of the concrete. This may be accomplished I by chipping, saw cutting or alternate means. Portions of activated concrete walls may be left in l place, provided they meet site release criteria.
l l Plate steel, structural steel, grating, ladders, and platforms may be decontaminated in place or l may be removed by unbolting or cutting, and rigged out for decontamination or disposal. l The liner of the refueling cavity may be sectioned for removal or decontaminated in place. The polar crane can be used to lift the sections out. Potential for high levels of contamination exists for components removed from the refueling cavity. It may be possible to decontaminate and release sections of the liners. 3.3.2.24- Auxiliary Building Uncludine Pine Facade) The Auxiliary Building has two floors below grade, one at grade (elevation 45 ft), and three floors above grade. The portion at or above grade is structurally connected to the Fuel Building on the east and to the Control Building on the west. A number of framing members in the Auxiliary Building are supported by the Containment Building wall. The exterior walls below grade and slabs are constructed of reinforced concrete. Interior framing members below grade and framing members above grade are structural steel. Exterior walls
- above grade are generally constmeted of concrete masonry block with exterior precast concrete j panels (elevation 45 ft) or metal siding at the upper floors. Interior walls are constructed of 1 concrete block masonry. Portions of the Auxiliary Building are coated with an epoxy surface.
3 23 March 1999 l I
TROJANllCENSE TERMINATIONPLAN ' 'Ibe surface is generally applied to the floors and to a height of 12 inches above the floor, but q may extend up to 6 ft in corridors and selected rooms. V Most of the major contaminated equipment has been removed from the Auxiliary Building since the start of decommissioning activities. Major plant equipment remaining in the Auxiliary Building includes the following items: i
- 1. Ventilation equipment supporting the Auxiliary, Fuel, and Containment
- Buildings;
- 2. Filters; and
- 3. Liquid radioactive waste system components.
Portions ofexposed surfaces in the Auxiliary Building are contaminated. The following specific considerations apply. The surfaces of walls and slabs in traffic areas have protective coatings. Concrete can be decontaminated by water or chemical washing. Surfaces that can not be decontaminated can be scabbled or surface ground down to non-contaminated depths. 3.3.2.25 Fuel Buildine - The Fuel Building contains facilities for storage of spent fuel and systems used for processing liquid wastes generated by plant operation and decommissioning activities. Remaining areas of note in the building include four floors above grade, the spent fuel pool, cask loading pit, new fuel storage pit, and cask wash pit. Portions of the Fuel Building are oated with an epoxy - surface. The surface is generally applied to the floor and to a height of 12 inches above the floor, but extends up to 6 ft in conidors and selected rooms. The walls and base slab of the spent fuel pool are constmeted of thick (approximately 5-ft to 6%-ft) reinforced concrete. The spent fuel pool and fuel handling equipment are discussed in Section 3.3.2.5. Most of the major contaminated equipment has been removed from the Fuel Building since the start of decommissioning activities. Major plant equipment remaining in the Fuel Building includes the following items:
- 1. Spent Fuel Pool;
- 2. Modular SFP cooling and cleanup system; and
- 3. Liquid radioactive waste system components.
Portions of exposed surfaces inside the Fuel Building are contaminated. The following specific considerations apply. The surfaces of wails and slabs in traffic areas have protective coatings. Concrete can be decontaminated by water or chemical washing. Surfaces that can not be decontaminated can be scabbled or surface ground down to non-contaminated depths. Removal of entire wall, or portions of walls may require evaluation of the building's structural integrity. 3-24 March 1999 i
TROJANLJCENSE TERMINATIONPL4N 3.3.2.26 Other Buildings - The Main Steam Support Structure (MSSS) consists of two floors, one at grade (Elevation 45 ft)
; and one at Elevation 63 ft. It is located between the Containment Building and Turbine Building and provided protection and support for the main steam isolation, power-operated relief and safety valves, as well as main steam and feedwater piping. The stmeture is constructed of reinforced concreb and stmetural steel. Portions of the MSSS are potentially contaminated.
De Condensate Demineralizer Building is a three-story, partially below grade structure located west of the Turbine Building. He building is used for temporary storage and processing oflow level radioactive waste prior to disposal. Portions of the building will require decontamination from radwaste processing activities. The Steam Generator Blowdown (SGBD) Building is located south of the Main Steam Support Structure, between the Containment Building and Turbine Building. The building has a neinforced concrete slab floor, reinforced masomy block walls, and a reinforced concrete roof supported by steel beams and metal decking. The building has a foundation curb designed to contain liquid spills. Decontamination of the SGBD building has been completed. The Radwaste Annex is a single-story windowless structure adjacera to the north wall of the Fuel Building. It is utilized for laundry sorting, storage, and frisking, as well as solid waste compaction and dmm storage. Portions of the Radwaste Annex are potentially contaminated. De Wright-Schuchart-Harbor (WSH) Warehouse is a Quonset type building that has been partially removed and is currently used as a staging area to support ISFSI construction activities.
; The WSH Warehouse has undergone final survey as documents iin PGE-1074," Trojan Nuclear Plant Final Survey Report for the ISFSI Site." The WSH Warehouse will eventually be dismantled and portions ofits concrete slab demolished or modified as required.
The Turbine Building is located west of the Containment and Control buildings. It houses the plant's secondary side components, inc?uding the turbine generator, condenser, and feedwater equipment. Portions of the Turbine Building are potentially contaminated. O 3-25 March 1999
TROJANIJCENSE TERMINATIONPLAN l l i 3.4 RADIOLOGICAL IMPACTS OF DECOMMISSIONING ACTIVITIES j The decommissioning activities described herein are conducted under the auspices of the approved TNP Radiation Protection Program and Radioactive Waste Management Program. l These programs continue to be implemented as described in the approved TNP l D-munissioning P!an, Sections 3.2 and 3.3, respectively. The TNP Radiation Protection l Progisii implements the regulatory requirements of 10 CFR 20 (Reference 3-17) through l approved plant procedures established to maintain radiation exposures ALARA. The Radioactive Waste Management Program controls generation, characterization, pucessing, handling, shipping, and disposal of radioactive waste per approved TNP Radiation Protection Progr m, Process Control Fiv ieii, and plant procedures. 3.4.1 OCCUPATIONAL EXPOSURE Table 3-3 documents personnel exposure projections for various decommissioning and fuel storage activities. The total radiation exposure impact for decommissioning and spent fuel management is estimated in Table 3-3 to total approximately 551 person-rem. Originally projected in the approved TNP Decommissioning Plaa to be approximately 591 person-rem, this value has been updated to reflect actual exposure savings for specific projects and revised estimates for normal plant opeis+ ions and fuel transfer to the ISFSI. The estimates contained in Table 3-3 incorporate the folicwing assumptions and bases:
- 1. Area dose rates are based on radiological surveys that have been adjusted to account for radioactive decay to the estimated start of decommisaioning activities;
- 2. The projected exposure for decommissioning activities is based on TNP site information;
- 3. The exposure for removal of the steam generators and pressurizer reflects' actual values; l
- 4. Personnel radiation exposure during the Transition Period is estimated to be approximately 4 person-rem per year, excluding some dismantlement activities- i and !
- 5. Estimated personnel exposure due to the transfer of fuel to the ISFSI is approximately 2.9 person-rem for each of the estimated 34 casks, for a total of approximately 99 person-rem. !
As of December 31,1998, the actual total exposure for decommissioning activities was ; approximately 224 peren-rem. Detailed exposure estimates and exposure controls for specific ) activities are developed during detailed planning per Radiation Protection Program procedures. i 3.4.2 RADIOACTIVE WASTE PROJECTIONS The radioactive waste management program is used to control the generation, processing, I handling, shipping, and disposal of radioactive waste during decommissioning. Activated and 3-26 March 1999
TROJANLJCENSE TERMINATIONPLAN 1
- contaminated systems, structures, and components represent the largest volume oflow level .
- ' radioactive waste expected to be generated during decommissioning. Other forms of waste !
. generated during decommissioning include: '
- 1. Contaminated water; 1
- 2. Used disposable protective clothing;
- 3. Expended abrasive and absorbent materials; !
4 i
- 4. Expended resins and filters; !
d 5. Contamination control materials (e.g., strippable coatings, plastic enclosures); and !
- 6. Contaminated equipment used in the decommissioning process.
Table 3-4 provides projections of waste volumes for decommissioning. The waste volume projections are conservative estimates obtained from the decommissioning cost estimate, and l include actual waste volume amounts for removal of the steam generators and pressurizer. As reflected in these tables, approximately 305,719 ft' oflow level radioactive waste will be generated as a result of decommissioning and spent fuel management activities. As of December 31,1998, an approximate volume of I81,076 ft' oflow level radioactive waste had , been shipped in 304 shipments. Decommissioning plancing at TNP incorporates the assumption that cost-effective waste volume reduction methods are limited. It also assumes significantly contaminated or activated materials i are sent directly to a disposal facility. However, alternative processing methods may be evaluated during decommissioning. i i
- O 3-27 March 1999 i
n,,
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l TROJANLJCENSE TERMINATIONPLAN j l l-
3.5 REFERENCES
FOR SECTION 3 3-1 Code of Federal Regulations, Title 10, Part 50.82, " Application for Termination of l
- License." ? .
) i 3-2 Portland General Electric Topical Reoort PGE-1061," Trojan Nuclear Plant Decommissioning Plan," Revision 6. 3-3 Portland General Electric. " Trojan Nuclear Plant Defueled Safety Analysis Report," ! Revision 7. 3-4 Code of Federal Regulations, Title 10, Part 50.59, " Changes, Tests and Experiments." 3-5 NRC Letter. W. F. Kme to S. M. Ouennoz. " Authorization of the Trojan Reactor Vessel , Package for Transport," October 29,1998. l 3-6 PGE Letter VPN-001-95, " Application for Tennination of License; Transmittal of Proposed Decommissioning Plan, Post-Operating License Environmental Repon, and
- Spent Fuel Management Plan for Trojan Nuclear Plant," January 26,1995. l 3-7 NRC Letter. M. T. Masnik to S. M. Ouennoz," Order Approving the Decommissioning ;
Plan and Authorizing Decommissioning of the Trojan Nuclear Plant," April 15,1996. p 3-8 PGE Letter VPN-012-96, "10 CFR 72, Subpart B - Application for License, Trojan F ISFSI," March 26,1996. 3-9 NRC Letter. L. E. Kokaiko to S. M. Ouennoz. " Environmental Assessment and Finding of No Significant Impact Related to the Construction and Operation of the Trojan Independent Spent Fuel Storage Installation (TAC No. L22102)," November 25,1996. 3-10 NRC Letter. L. E. Kokaiko to S. M. Ouennoz. " Safeguards Evaluation of the Physical Security Plan for the Proposed Trojan Independent Spent Fuel Storage Installation (TAC ' No. L22102)," November 20,1996. 3-11 Portland General Electric Topical Reood PGE-1074, " Trojan Nuclear Plant Final Survey Report for the ISFSI Site," Revision 0. [ 3-12 Code of Federal Regulations, Title 10, Part 20.1402, " Radiological Criteria for Unrestricted Use." 3-13 PGE Letter VPN-012-99, " Trojan Nuclear Plant Reactor Vessel and Internals Removal Project Safety Analysis Report (RVAIR SAR), PGE-1076," February 2,1999. 3-14 Portland General Electric Topical Report PGE-1076, " Trojan Nuclear Plant Reactor Vessel and Internals Removal Safety Analysis Report." I l 3-28 March 1999
I TROJANLICENSE TERMINATIONPLAN 3-15 Code of Federal Regulations, Title 10, Part 71, " Packaging and Transportation of ( Radioactive Material." 3-16 Eprtland General Electric Tonical Report PGE-1021. "Offsite Dose Calculation Manual," l Revision 17. 3-17 Code of Federal Regulations, Title 10, Part 20," Standards for Protection Against Radiation." O V s ' (G c l 3-29 March 1999
l TROJANIJCENSE TERMNAHONPL4N l Page1of3 !. 3 Table 3-1 i 1~ ) l Status of Major TNP Systems, Structures, and Components l as of January 1999
" b'"'
System, Structure or . 8mPonent Fue rage i Reactor Coolant Synem -NO Removed. Reactor VesselIntemals NO Preparations for removal in 1999 are underway. Reactor Vessel NO Preparations for removal in 1999 are underway. Steam Generators NO Removed. Reactor Coolant Pumps NO Removed. Pressurizer and Pressurizer Relief Tank NO Removed. Partially removed. Boric Acid Batch Tank remains Chemical and Volume YES in service to support boration of the SFP if Control System required. Safety Injection System NO Removed. Residual Heat Removal System NO Removed. Containment Spray System NO Removed. p Component Cooling NO Partially removed. x) Water System i Partially removed. Portion of system remains in Service Water System YES service to supply dilution flow for liquid discharges and attemate SFF make-up. Spent Fuel Pool and Fuel . . . aHy remWe4 Ponson in senice. Handling Equipment Spent Fuel Pool Cooling and Demineralizer System NO Removed. (Original) Modular SFP Cooling and In service. This system replaced the original YES Cleanup System system. n ensa
; 7 NO Partially removed.
Steam Generator Partially removed. Some piping is included in NO Blowdown System embedded piping scope. Primary Makeup Water Partially removed. Some system piping is included System NO in embedded piping scope. Refueling Water Storage To be removed in 1999. Some related piping is Tank NO included in embedded piping scope. l March 1999
TROJAN 12 CENSE TERMINATIONPLAN ) Page 2 of 3 '. Table 3-1. Status of Major TNP Systems, Structures, and Components as of January 1999 System, Structure or "I '" 8'"' ' Component 7,,[,,,, Partially removed. System decontamination 8ctivities are in Progress in conjunction with Plant Effluent System NO , Turbme Buildmg activities. System ,meludes a portion of piping in embedded piping scope. , Containment Ventilation ' Systems -
O Partially removed;in service.
i Hydrogen Recombiners nd Removed. : Fuel Building Ventilation YES Partially removed;in service. System
^ ' " "8 YES Partially removed;in service.
V tila ion S t Condensate Demineralizer i In service. Suppcrts use of building as radwaste j Building Ventilation NO gy storage and processing facility. Instrument and Service Partially removed; in service. Air currently ,
.YES Air Systems supplies pressure to SFP gates for sealing.
y Gaseous Radioactive Waste System NO Partially removed. Solid Radioactive Waste System NO Panially removed. i
~ '
Partially removed. Portion of remaining system is used for processing liquid radwaste prior to Liquid Radioactive Waste YES discharge from plant. System dram pipmg and Systern some process piping is included in embedded piping scope. Partially removed. Portions of system remains in Radiation Monitoring service for effluent monitoring ofliquid and YES System ventilation exhausts and criticality monitors at SFP area. Process Sampling System NO Partially removed. Partially removed. Portions of system remain in Fire Protection' System YES . service for fire detection and suppression and alternate SFP make-up. Partially removed. Portions of the systems remain Electrical Systems YES in service to support ongoing plant operation and l decommissioning activities. In service. Most equipment has been removed with Containment Building NO e exception of Reactor Vessel and Intemals. i f March 1999
, _ _ .. ___ __ . _ . ~. _ - - . _ _
TROJANllCENSE TERMINATIONPLAN \ l l Page 3 of 3 l Table 3-1 {g \ Status of Major TNP Systems, Structures, and Components l as of January 1999 l System, Structure or '9" ** **'"' Support Component i p ,,,3,,,,,, Auxiliary Building In service. Part of seismically qualified Control-( ng ee Pig ES . p Auxiliary-Fuel Building Structure Fuel Building In service. Part of seismically qualified Control- l YES Auxiliary-Fuel Building Stmeture. Main Steam Support Equipment removed. Decontamination activitMs Structure NO are in progress. Condensate Demineralizer In service. Used for temporary radwaste storage Building NO and pmcessing oflow level radioactive waste. I do i ing NO Equipment removed. Building decontaminated Radwaste Annex NO In service. Wright-Schuchart-Harbor NRC approved final survey to remove from NO 10 CFR 50 licensed area; Currently used for ISFSI (WSH) Warehouse , , prep' cation activities. q Affected equipment has been removed. 1 O Turbine Building NO Decontamination of affected areas is in progress. Contains a portion of dra,n i and process pipmg included in embedded piping scope. l l !v l March 1999
TROJANIJCENSE TERMINATIONPLAN i L - Page 1 of 4 i Table 3-2 O Major Components Removed (By Year) 1996: l l . Reactor Coolant Pumps and Motors - (4) RCS Piping (up to bioshield wall) i RCS and Steam Generator support structural steel ; l Containment Main Steam and Feedwater Piping and Supports (all B/C loops; partial A/D ! loops) - l A Residual Heat Removal Heat Exchanger . ? l B Residual Heat Removal Heat Exchanger ! !- Positive Displacement Charging Pump / Motor j A Centrifugal Charging Pump Motor B Centrifugal Charging Pump Motor A SafetyInjection Pump / Motor .
^
B Safety Injection Pump / Motor A Containment Spray Pump / Motor l B Containment Spray Pump / Motor B Component Cooling Water Pump / Motor (Note: replaced with smaller pump) l Condensate Demineralizer vessels Decontamination Shop Equipment Portions of Outside Buildings (WSH warehouse, Maintenance Shop) l Low Level Radwaste Storage Building (Replaced by Condensate Demin Bldg) { l I i 1 I ; 1 l l L0 f March 1999
TROJANIJCENSE TERMINATIONPLAN i Page 2 of 4 ! s Table 3 2 Major Components Removed (By Year) 1997: l A/B Residual Heat Removal (RHR) Pumps / Motors ' A/C Service Water Booster Pumps / Motors Letdown Heat Exchanger A/B Boric Acid Evaporator skids , Clean Radioactive Waste Evaporator skid i Seal Water Heat Exchanger ! Dirty Waste Monitor Tank Auxiliary Building Drain Tank and Pumps l Waste Concentrate Hold Tanks and Pump i CVCS Concentrates Hold Tank and Pumps . Tiger Lock Storage Tank Steam Generator Blowdown Tank and Pump Boric Acid Storage Tanks and Pumps Boric Acid Blender and Chemical Addition Tank Containment Post-Accident Sampling System (PASS) Containment Hydrogen Analysis System (CHAS) ; Hold-Up Tank Recirculation Pump 1 Gas Stripper Feed Pumps-Spent Fuel Storage Rack (1) New Fuel Storage Racks Decontamination Area Ventilation Supply Cooling System (AB-5) , B/C Safety Injection Accumulators Control Rod Drive Mechanism Patch Panels Reactor Vessel Neutron Water Bag Racks Flux Thimble Drive Units and Detectors Manipulator Crane Portions of Containment Ventilation Systems CS-1, CS-2, CS-3, CS-4, CS-5, CS-6 Portions of Activated Concrete around Reactor Vessel Portions of Containment Wall for new 10'x10' roll-up access door Bioshield Structural Steel and Equipment Supports - RCP Oil Collection System Remaining Containment Main Steam and Feedwater Piping and Supports O March 1999
( l TROJANLICENSE TERMINATIONPLAN i Page 3 of 4 Table 3-2 Major Components Removed (By Year) 1998: Service Water Booster Pumps / Motors (B/D) Sodium Hydroxide Tank Boron Injection Tank l Clean Waste Receiver Tank Pumps / Motors l Clean Waste Receiver Tanks (A/B) j Chentical and Volume Control System (CVCS) Monitor Tanks and Pumps ; Primary Makeup Water (PMW) Pumps Reactor Coolant Drain Tank (RCDT) pumps Chemical Waste Drain Tank and Pumps Waste Gas Decay Tanks (A/B/C/D) Waste Gas Surge Tank Waste Gas Compressors (A/B) CVCS Volume Control Tank Boric Acid Heat Trace Panels CVCS Evaporator Condensate Demineralizers (T-220 A/B) CVCS Evaporator Feed Demineralizers (T-219 A/B/C) CVCS Mixed Bed Demineralizers (T-210 A/B) O CVCS Cation Bed Demineralizer (T-211) SFP Demineralizer(T-224) SGBD Demineralizers (T-316 A/B) l Reactor Coolant Filter (F-204) I SealInjection Filters (F-210A/B) , l Seal Water Return Filter (F-209) Reactor Coolant Drain Filter (F-307) CVCS Evaporator Concentrates Filter (F-208) CVCS Evaporator Condensate Filter (F-207) CVCS lon Exchange Filter (F-206) Resin Backflush Filter (F-305) SFP Demin Pre-Filter (F-201) SFP Demin After-Filter (F-211) Clean Waste Filter (F-304) SGBD Filter (F-306) Automatic Gas Analyzer RCS Post Accident Sampling System Panel S/G Sample Panel Steam Jet Air Ejector (SJAE) associated vessels and piping i o March 1999
TROJANIJCENSE TERMINATIONPLAN Page 4 of 4 Table 3-2 O Major Components Removed (By Year) 1998: (continued) SJAE Effluent Monitor (PERM-6) ' CVCS Holdup Tanks (A/B/C) Primary Mdeup Water Tank Contaminatec Piping b Various Pipe Chases Component Cooling Water (CCW) Heat Exchangers (A/B) SFP Heat Exchangers (A/B) SFP Cooling Pumps / Motors (A/B) SFP Purification Pump / Motor i Reactor Coolant Drain Tank Containment Air Coolers (8) Hydrogen Mixing Fans Reactor Cavity Cooling Far.s Reactor Cavity Exhaust Fans Containment Sump Pumps (2) . Pressurizer Relief Tank l Containment Motor Control Centers (Containment has temporary power only) Reactor Coolant Pump fans for "A" and "D" RCP's O Component Cooling Water and Containment Spray Piping / Supports Equipment Hatch Trolley Rail System Control Rod Drive Mechanisms Electrical Penetration Assemblies (partial) ' Containment Piping Penetrations (partial) i O March 1999 n e- - - , +
.e TROJANL1 CENSE TERMINATIONPLAN Table 3-3 Radiation Exposure Projections Exposure Activity (person-rem) Steam generators and pressurizer removal 54' Reactor vessel and internals removal 67 Dismantlement Nuclear steam supply system 51 Spent fuel racks 19 Balance ofplant systems 165 Structures - 46 Miscellaneous 20 Subtotal 301 Normal plant operations 30 Fuel transfer to ISFSI 99 6 Total 551
- This value reflects actual total project exposures.
b This value represents a reduction in the original projection by approximately 40 person-rem. l l l i i O March 1999
?? TROJANLJCENSE TERM 1NATIONPLAN Page1of3 Table 3-4 O
~ Decommissioning Waste Classification and Volume Projections Class A Class B Class C Burial Burial Burial Volume Volume Volume item (ft') - (ft') (ft')
Reactor coolant piping 5,894 0 0 y Pressurizer relief tank 625 0 0 :
)
Reactor coolant pumps and motors 3,044 0 0 CRDMs/incore instrumentation / service 1,726 0 0 J l structure removal Steam Generators and Pressurizer 57,800' 0' 0 I
.(Large Component Removal) ;
Reactor vessel and intemals 0 0 8,341 ; Spent fuel racks 16,551 0 0 120 V ac preferred instrument ac 1,400 0 0 125 V de power 175 0 -0 4.16 kV ac power 726 0 0' 480 V ac auxiliary load center 5,080 0 0
.480 V ac motor control center 8,426 0 0 Chemical and volume control 10,% 8 0 0 Clean radwaste 5,423 0 0 Containment Building penetrations 188 0 0 Control rod drive 85 0 0 Dirty radwaste- 1,613 0 0 Electric heat tracing 164 0 0 Electrical (Cables / Tray / Conduit) 60,139 0 0
' Fuel handling system 339 0 0 i Fuel pool cooling and demineralizer 4,632 0 0 )
Fuel and Auxiliary Building heating, 3,661 0 0 ventilation, and air conditioning (HVAC) 1 March 1999 !
- y.
- TROJANUCENSE TERMINATIONPLAN ,
, Page 2 of 3 !
- p, Table 3-4 V
Decommissioning Waste Classification and Volume Projections l Class A Class B Class C l Burial Burial Burial l Volume Volume Volume 3 3 3 Item (A ) (A ) (A ) .; i Gaseous radwaste 2,529 0 0 j HVAC 6,635 0 0 Hydrogen recombiners 576 0 0 Integrated leak rate test instrument line 106 0 0 Instrument and service air 1,327 0 0 l Lighting panel supply 997 0 0 l Miscellaneous components 1,936 0 0 Miscellaneous reactor coolant 3,418 0 0 l Nuclear instmmentation 193 0 0 Oily waste and storm drains 1,882 0 0 Containment HVAC 18,869 0 0 Primary makeup water 3,615 0 0 Primary sampling 114 0 0 Radiation monitoring 134 0 0 Reactor nonnuclear instruments 245 0 0 i
- Reactor vessel system 116' 0 0 l 1
Residual heat removal 7,649 0 0 Safety injection system 7,149 0 0 ; Solid radwaste 370 0 0 Spent fuel pool 754 0 0 Steam generator system 3,562 0 0' l Turbine Building sump pumps and 639 0 0 t miscellaneous Component cooling water 6,115 0 0
'C'N March 1999
. = .. .. - ._. _. _ . . . .. .... .- _
TROJANLICENSE TERMINATIONPLAN I Page 3 of 3 r Table 3-4 - l Decommissioning Waste Classification and Volume Projections 1 Class A Class B Class C Burial Burial Burial Volume Volume Volume Item (ft') (ft') (ft') Condensate demineralizer 2,262 0 0 Discharge and dilution 3,834 0 0 . Containment spray 1,563 0 0 Containment Building 13,458 0 0 AuxiliaryBuilding 2,650 0 0 Fuel Building 4,711 0 0 Main steam supply system and electrical 629 0 0 penetration area Turbine Building 1,054 0 0 Process liquid radwaste 0 3,686 0 Disposal of dry active waste generated 5,942 0 0 Total 293,692 3,686 8,341
- This value reflects ictual burial volumes for the large component removal project.
l l lO March 1999
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//
%'.M .avY ' t' Trojan Nuclear Plant Y ~@ rL. iee -S- LI CENSE TERMINATION PLAN t __. ; _ ,=.r._._._. fi _ , _ . J _ _ p: _1 _ f
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O TROJANUCEh5E TERMINATIONPLAN O . Figure 3-10 l Decommissioning / Site Restoration Schedule Major Activity 1993 l 1994 l 199s l t996 l 1997 l 195s l 1999 l 200e l 20e1 l 2002 l 20e3 l 2ee4 l 2ess l 2ee6 l 2ee7 l 2ees l 2ee9 l 2eto l 2et s l 2e12 l 2st3 l 2e 4 l 2ets l 2e 6 l 2st7 l 2sts Transition Period ; j ji; j ii j Large Component l Removal Project g . i i ISFSI Implementation Reactor Vessel & Internals Removal M i j i i Decontamination ' & Dismantlement Period Final Radiation ! Survey l , (M l 3 l' l i j
'I l Non-Contaminated !, I I i i i Building Demolition l
i l f jN F Major Activity M I March 1999 ; i
' TROJANUCENSE TERMINATION PLAN
- 4. SITE REMEDIATION PLANS N
(V
4.1 INTRODUCTION
L 'Ihe purpose of this section is to describe how remediation actions may be applied to various areas on the TNP site, identify the remediation methodology to be used, and demonstrate that the remediation methodology is adequt2 to ensure that the site release criteria of 10 CFR 20.1402 (Reference 4-1) are met. Verification of the site release criteria is detailed further in Section 5,
" Final Survey Plan."
O O 4-1 March 1999
- -- ._~ _ - - .----
TROJANLICENSE TERMINATIONPLAN
- 4.2 REMEDIATION LEVELS The ALARA evaluation uses action levels, referred to as remediation levels, that are established
]
for various types of remediation actions such as chemical decontamination, wiping, washing, i vacuuming, scabbling, spalling, abrasive blasting, and high pressure washing. A remediation I level is the level of residual radioactivity at which the desired beneficial effects due to the i performance of a given remediation action are equal to the undesirable effects or costs of the action. The methodology and equations used here for calculating remediation levels are from ; draft Regulatory Guide DG-4006, " Demonstrating Compliance with the Radiological Criteria for License Termination" (Reference 4-2). Remediation levels are developed using an unbiased analysis of remediation actions which can both aven future dose (a benefit to society) and cost money (a potential detriment to society). In
. order to compare the benefits and costs of a remediation action, the benefits and costs are given a monetary value. The monetary value of the collective avened dose (the benefit) is compared . with the monetary value of the undesirable effec s (the cost). The remediation level is the point at which the benefits of the remediation action equal the costs.
4.2.1 REMEDIATION LEVEL CALCULATION Remediation levels are calculated using Equation 4-1: RL = C8#' x
#+
O (2000)(PD)(0.025) (F)(A) 1 - e"" M ' . (Equation 4-1) where: RL = remediation level fraction of DCOL (dimensionless)
=
Costr total monetary cost of remediation action ($)
-2000 = value of a person-rem averted ($/ person-rem)
PD = population density for the critical group scenario (persons /m2) 0.025 = annual dose to an average member of the critical group from , residual radioactivity at the DCGL concentration (rem /yr) ! F = remediation action effectiveness (dimensionless) :' 2 A- = r = area monetarybeing evaluated discount rate (y( (m ) ') 1 = radiological decay constant for the radionuclide (y(') I Y '= number of years over which collective averted dose is calculated (years) ; Acceptable values for the equation parameters of population density, PD; the monetary discount rate, r; and the niunber of years, Y, are taken from Table 3.1 of draft Regulatory Guide DG-4006 and are given in the table which follows. The development of the equation parameters of total cost, Costr, and remediation action effectiveness, F, are described in Sections 4.2.2 and 4.2.3. A ! justification is provided where values are used other than those given in the following table or
- calculated in accordance with Sections 4.2.2 or 4.2.3. .
(D -O , 4-2 March 1999
I j ; TROJANLICENSE TERAflNATIONPLAN I l Remediation Level Equation Values l l O)
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Equation Acceptable Value ! Parameter Building Land i PD 0.09 0.0004 - r 0.07 0.03 _ Y 70 1000 4.2.2 CALCULATION OF TOTAL COST The total monetary cost for performing a given remediation action, Costi, is calculated using ! Equation 4-2: I Costr = Costa + Costun + Costsa + Costir + Costnw (Equation 4-2) l l where:
= monetary cost of the remediation action, including mobilization costs Costa
=
Costwo monetary cost for transport and disposal of waste generated by the remediation action
= monetary cost of worker accidents during the remediation action Costace '
Costrr = monetary cost of traffic fatalities during transporting of waste k Costwa,, = monetary cost of dose received by workers performing the remediation l action and transporting waste to the disposal facility Other monetary costs may be included as appropriate for the particular situation. The monetary cost of the remediation action, Costa, is the cost of performing the remediation action, including equipment mobilization and demobilization and labor costs. The cost of waste transport and disposal, Costwo, is calculated using Equation 4-3: Costav = (Va)(Coste) (Equation 4-3) where: VA = volume of remediation Waste produced (m ) Costy = cost of waste disposal ($/m') l l i 4-3 March 1999
TROJANLICENSE TERMINATION PLAN ~ The cost of workplace accidents, Costu,is calculated using Equation 4-4: O = (3,000,000)(4.2x 10*)(Ta) (Equation 4-4) Costsa \ j 3,000,000 ; = monetary value of a fatality equivalent to $2,000/ person-rem (_$) ! 4.2 x 104 = . workplace fatality rate (hrs-1) T3 = worker time required for remediation (person-hrs)
] 'Ihe cost of traffic fatalities incurred during the shipment of waste, Cost, is calculated using Equation 4-5 i
i Costre , a) x 10*) (M) (Equation 4-5) 13.6 1 whem: 3,000,000 = monetary value of a fatality equivalent to $2,000/ person-rem ($) ; V: . =- volume of remediation waste produced (m3) 3.k x 10-s = truck fatality rate per kilometer traveled (km-1) 890 = distance traveled by truck (km) ! 13.6_ '= volume of a truck shipment (m3)
- The cost of remediation worker dose, Cost,,%, is calculated using Equation 4-6:
Costma = (2,000)(Da)(Ti) (Equation 4-6) where: 2,000 = monetary cost of dose received ($/ person-rem) D, = EE me to mmedadon woh @*) T3 = worker time required for remediation (person-hrs) 4.2.3 DETERMINATION OF REMEDIATION ACTION EFFECTIVENESS The remediation action effectiveness, F, is expressed in terms of the fraction of the residual radioactivity removed by the remediation action. It is determined by collecting and analyzing pre-remediation and post-remediation measurements in an area in which the remediation action
' is performed. A sufficient number of measurements are made to establish a consistent fraction.
O I 4-4 March 1999 _ - _ - _ I
i I TROJANLICENSE TERMINATION PLAN 4.3 ALARA EVALUATION ' O The ALARA evaluation compares residual radioactivity levels to calculated remediation levels
-- for possible remediation actions. Where the level of residual radioactivity exceeds the remediation level, the remediation action is considered to have a net benefit. Therefore, it is '
considered cost effective and must be taken for the residual radioactivity to be considered ALARA. Conversely, if the concentration is less than the remediation level, the level of residual ' radioactivity is already considered ALARA and the remediation action is not required to be !
. performed.
The ALARA evaluation is needed only to justify not taking a temediation action. If a decision l has already been made to perform a given remediation action, there is no need to evaluate whether the action is necessary to meet the ALARA requirement. For example, if wiping down surfaces with loose radioactive contamination is a good practice that is applied regardless of radioactive contamination levels, then an ALARA evabation is not required. For those remediation actions considered but not taken, the ALARA evaluation includes the levels of- j residual radioactivity above which those remediation actions would have been justified.- , i Remediation levels do not represent concentration limits that cannot be exceeded. Rather, they ! represent the threshold at which the given remediation action is taken. The ALARA requirement . is met by perfonning the appropriate remediation action and not by being below a specified ' concentration after the action is taken. The ALARA evaluation ensures that efforts to remove residual radioactivity are commensurate with the level of risk the residual radioactivity poses. O
* \
l ll l
)
i l O 4-5 March 1999
n..-- - . - - - - - . - - , - - _ . . - - - - - . . . - - . ~ - ..- _ ~ ~. TROJANLICENSE TERMINATION PLAN 4.4 REMEDIATION ACTIONS Remediation actions are performed in accordance with the general and specific decontamination i and dismantlement considerations of Section 3.2. The process ofidentifying areas that need - l remediation is ongoing and will continue throughout decommissioning. For remediation purposes, areas are categorized as one of three types:
- 1. Structures, which include building interiors and exteriors, major freestanding -
exterior structures, exterior surfaces of plant systems, and paved exterior ground surfaces; 2.- Land areas, which include unpaved exterior ground surfaces; and
- 3. Plant systems, which include interior surfaces of process piping and components.
4 .4 .11 STRUCTURES
- As detailed in Section 3.2, remediating structures may include the use of a variety of techniques ;
ranging from water washing to surface material removal. Several factors determine the choice of j the remediation method for a given application, including the extent of the contaminated area, surface material, depth of contamination, and access considerations. Demolishing certain
- structures may be necessary based on degraded structural integrity as a result of remediation -
efforts and/or removal of systems and components, surrounding walls, or other barriers. ! Remidiation actions performed on exposed surfaces vary with respect to the amount ofresidual
. radioactivity previously identified. Remediation activities may include wiping down an area, j vacuuming to collect dirt and contamination from recesses and corners, and low or high pressure '
water washing of an area. Surfaces may also be remediated by scabbling or grinding the surface. , Surface removal is performed uaing methods that control the removal depth to minimize the - I waste volume produced. For concrete surfaces, remediation methods may include core drilling, concrete sawing, and scarifying or scabbling waUs. This last method removes the surface of walls by bush heads, roto-peen devices, flappers, etc., acd is good for removal of material close to the surface. Other methods include abrasive blastmg, which is good for contamination removal from surfaces that i are not necessarily smooth. Also, chipping guns and jackhammers may be used for removal of concrete surfaces as deep as the first mat of rebar. Certain surfaces with exposed cracks may require more aggressive means ofremoving concrete. Demolition equipment such as the ; BROKK and Dynahoe are used to remove concrete walls and to reduce to rubble any concrete blocks which already have been removed. l Strippable coatings can be used to temove contaminants from surfaces on which other methods , don't work. Waterjet, CO blasting, 2 and other means may be used to clean surfaces which tend l
- _to trap contamination using other methods of remediation. Abrasive blasting may also be used to l
remove contaminants from steel and other surfaces besides concrete, i ! l l 4-6 March 1999
-I
_ _ _ .- . . _ . . . . ~ . . . . . - _ _ . _ _ _ . _ . . . . _ . _ . _ . _ _ _ _ _ _ _ . . . - . _ _ _ . _ _ . E. i TROJANLICENSE TERMINAllONPLAN . P _ 4.4.2 LAND AREAS . ;
.As discussed in Section 2, the remediation of surface and subsurface soils, surface water and .
- groundwater, bottom sediment, and pavement is not expected to be necessary. i 4.4.3 SYSTEMS 4
Contaminated plant systems and components are typically removed and sent to an offsite l processing facility or a low level radioactive waste disposal facility, or they may be -
.. decontaminated onsite and released. The majority of embedded piping is expected to be decontaminated in place and released.
Remediation methods typically include chemical decontamination, wiping, washing, vacuuming, scabbling, spalling, and abrasive blasting.- As detailed further in Section 3, selection of the preferred method is based on the specific situation. Other remediation technologies may be considered and are used if appropriate. 1 O 4 I 't d 1 i F i. [O 4-7 March 1999
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, , . . . . . . , .. ..~ -
. - - ~ . . . . . . . - -.=..- .-. . . . . . . - . . --. - . - . . - .
TROJANLICENSE TERMINATIONPLAN
4.5 REFERENCES
4-1 Code of Federal Regulations, Title 10, Part 20.1402, " Radiological Criteria for
- Unrestricted Use."
r 4-2 Draft Regulatory Guide DG-4006," Demonstrating Compliance with the _ Radiological Criteria for License Termination," August 1998. ; i t i P i 4 i 4 i !O l 4-8 March 1999
TROJANI2 CENSE TERMINATIONPLAN : I
- 5. FINAL SURVEY PLAN 5.1 INTROD'UCTION 5.1.1 . PURPOSE I In accordance with 10 CFR 50.82(a)(9)(ii)(D) (Reference 5-1) and Regulatory Guide 1.179
. (Reference 5-2), this TNP Final Survey Plan describes the final survey process that will be used
~
)'
to demonstrate that the TNP facility and site meet the radiological criteria for license terminationi This plan incorporates the site release criteria of 10 CFR 20.1402 (Reference 5-3) for unrestricted use of the TNP site. , 5.1.2 SCOPE ^ l As detailed in Section 5.2.4, the final survey encompasses structures, land areas, and plant systems which, as a result oflicensed activities, are identified as contaminated or potentially contaminated. The majority of these are located within the Trojan Industrial Area, which is illustrated in Section 3, Figure 3 1. At the time of final survey, the structures will be largely
' intact. The majority of the contaminated systems and components will have been removed prior i to the initiation of the survey data collection in those areas.
The final survey does not include the ISFSI. The ISFSI site has been previously surveyed and the results of that survey are documented in PGE-1074, " Trojan Final Survey Report for the ISFSI Site"(Reference 5-4). The final survey also does not include monitored gaseous and O liquid pl nt effluent discharge pathways. As confirmed by characterization results summarized in Sectica 2, the Trojan Radiological Environmental and Effluent Monitoring Program (REMP) (Reference 5-5) documents compliance with the ALARA criterion'of10 CFR 50, Appendix I
' (Reference 5-6), associated with monitored releases.
-5.1.3
SUMMARY
This final survey plan describes the final survey process, as well as the methodology used to develop guideline values against which residual radioactivity levels remaining at TNP at the time of final survey will be compared. The final survey process is described as a series of sequential steps-survey preparation, survey design, data collection, data assessment activities, and final survey report preparation. However, in practice, this process is iterative since the results from one step may prompt repeating one or more previous steps.- Survey preparation activities begin once dismantlement activities are complete in a given area. !
. An ALARA evaluation is performed to determine, from a cost-benefit perspective, which i remediation actions should be taken in addition to those already planned or completed. The area meets the site release ALARA criterion once any additional remediation actions are completed. i The area is divided into survey units that are classified according to their' potential for residual l
. radioactivity. Survey data are collected from the survey unit according to data collection l
' pattems and frequencies established for each classification. Where residual radioactivity is measured above pre-set levels, an investigation is performed. Based on the results of the investigation, the survey unit may be remediated, reclassified, or resurveyed.
5-1 March 1999
TROJANUCENSE TERMINATIONPLAN i l Three principal types of survey data are collected. They are: 1) scan measurements,2) static ! surface contamination measurements, and 3) laboratory analysis of soil and bulk material samples. Data are verified to be of adequate quantity and quality and to support underlying assumptions necessary for a statistical test to be applied. Where necessary, previous survey steps are re-evaluated and additional data are collected prior to statistical analysia. The survey unit meets the site release dose criterion once the survey data pass the statistical test. Where the data j , fail the statistical test, the survey unit does not meet the site release dose criterion. The data are ' , analyzed and additional data are collected or the survey unit is remediated and resurveyed. , Upon completion of final survey activities, a final survey report will be prepared which summarizes the data and documents the conclusion that the TNP facility and site meet the 10 CFR 20.1402 release criteria and can be released for unrestricted use. i 5.1.4 DEFINITIONS l 1 Italicized words and phrases found within the definition of the term are also defined. j Action Level: The numerical value that will cause the decision maker to choose one of the t altemative actions. It may be a dose- or risk-based concentration level (e.g., DCGL), or a reference-based standard. See investigation level. : ALARA (As Low As Reasonably Achievable): A basic concept of radiation protection which specifies that exposure to ionizing radiation and releases of radioactive materials should be , managed to reduce collective doses as far below regulatory limits as is reasonably achievable considering economic, technological, and societal factors, among others. Reducing exposure at a , site to ALARA strikes a balance between what is possible through additional planning and J management, remediation, and the use of additional resources to achieve a lower collective dose i level. I i ALARA Criterion: The residual radioactivity has been reduced to levels that are ALARA. Alpha (a): Also referred to as afalsepositive decision error, it is the probability of passing a survey unit that should fail. Area of elevated residual radioactivity: An area over which residual radioactivity exceeds a specified value DCGL. I Area Factor: A multiple of the DCGL that is permitted in the area ofelevatedresidual l radioactivity without requiring remediation. j Background Reference Area: An area that has similar physical, chemical, radiological, and biological characteristics as the site area being remediated, but which has not been contaminated l by site activities, from which representative reference measurements are performed for ! comparison with measurements performed in specific sun ey units.
..O 5-2 March 1999 j
1 TROJANLICENSE TERMINATIONPLAN Background Radiation: Radiation from cosmic sources, naturally occurring radioactive material, including radon (except as a decay product of source or special nuclear material), and O global fallout as it exists in the environment from the testing of nuclear explosive devices or from nuclear accidents like Chernobyl which contribute to background radiation and are not under the control of the cognizant organization. Background radiation does not include radiation from source, byproduct, or special nuclear materials regulated by the cognizant Federal or State agency. Beta (p): Also referred to as afalse negative decision error, it is the probability of failing a survey unit that should pass. The complement of beta (1- ) is referred to as thepower of the test. Biss: The systematic or perr,i:: tent distortion of a measurement process which causes errors in j one direction (i.e., the expected sample measurement is different from the measurement's true i value). Calikation: Comparison of a measurement standard, instrument, or item with a standard or instrument of higher accuracy to detect and quantify inaccuracies and to report or eliminate those inaccuracies by adjustments. Chain of Custody: An unbroken trail of accountability that ensures the physical security of samples, data, and records. Class 1 Area: An impacted area where, prior to remediation, there are expected to be locations with concentrations of residual radioactivity that exceed the DCGL. O V Class 2 Area: An impactedarea where, prior to remediation, there are expected to be locations with concentrations of residual radioactivity detectable above background levels, but that do not exceed the DCGL. Class 3 Area: An impactedarea where there are not expected to be locations with l concentrations of residual radioactivity detectable above background levels. I Confidence Interval: A range of values for which there is a specified probability (e.g.,80%, 90%,95%) that this set contains the true value of an estimated parameter. Critical Group: The group ofindividuals reasonably expected to receive the greatest exposure to residual radioactivity within the assumptions of the particular scenario. DCGL (derived concentration guideline level): The concentration of residual radioactivity distinguishable from background radiation which, if distributed uniformly throughout a survey unit, would result in a TEDE of 25 mrem /yr to an average member of the criticalgroup. The average member of the criticalgroup is the individual who is assumed to represent the most likely exposure situation based on the assumptions and parameter values used in the dose model calculation. Decision Maker: The person responsible for the fmal decision regarding disposition of the survey unit. This person is the Manager, Personnel / Radiation Protection. 5-3 March 1999
TROJANIJCENSE TERAf1NATIONPLAN Decommissioning: The process of removing a facility or site from operation, followed by remediation, and license termination. Its objective is to reduce the residual radioactivity in p structures, materials, soils, groundwater, and other media at the site so that the concentration of d each radionuclide contaminant that contributes to residualradioactivity is indistinguishable from the bac5groundradiation concentration for that radionuclide. Direct Measurement: Radioactivity measurement obtained by placing the detector near the surface or media being surveyed. An indication of the resulting radioactivity level is read out directly. Elevated Measurement Comparison (EMC): A simple comparison ofmeasured values against a limit. It is described in Appendix 5-1. Exposure Pathway: The route by which radioactivity travels through the environment to eventually cause radiation exposure to a person or group. False Negative Decision Error: The probability of failing a survey unit that should pass. A statistician usually tefers to afalse negative decision error as a Type 11 decision error. False Positive Decision Error: The probability of passing a survey unit that should fail. A statistician usually tefers to thefalsepositive decision error as a Type Idecision error. Final Survey: Measurements and sampling to describe the radiological conditions of a site, following completion ofremediation activities in preparation for release. Final Survey Plan: A formal document describing in comprehensive detail the necessary quality assurance, quality control, and other technical activities that must be implemented to demonstrate that the Trojan facility and site meet the radiological criteria for license termination given in Subpart E of 10 CFR 20. Thefinal surveyplan provides a process for obtaining data of sufficient quality and quantity in such a manner as to ensure that the results of the work performed satisfy the stated performance criteria. Gray Region: A range of values cf the parameter ofinterest for a survey unit where the consequences of making a decision error are relatively minor. The upper bound of the gray region is set equal to the DCGL, and the lower bound ofthe gray region (LBGR) is a site-specific variable. hupacted Area: An area that has reasonable potential for residual radioactivity from licensed activities. Independent Spent Fuel Storage Installation (ISFSI): A complex designed and constructed for the interim storage of spent nuclear fuel and other radioactive materials associated with spent fuel storage. Investigation Level: A level of radioactivity that is based on the site release dose criterion which, if exceeded, initiates an investigation of the survey measurement. /7 U 5-4 March 1999
_~ - - . - - . . - . I TROJANLICENSE TERAflNATIONPLAN License Termination: Discontinuation of a license, the eventual conclusion to decommissioning. A ! Lower Bound of the Gray Region (LBGR): The concentration to which the survey unit must be cleaned in order to have an acceptable probability of passing the statistical test for meeting the i site release dose criterion. It represents the lower bound of the area of uncertainty regarding the concentration of residual radioactivity in the survey unit. The DCGL represents the upper bound. The width of the gray region (DCGL-LBGR) is also referred to as the shift. l Mean (R): The average value obtained when the sum ofindividual values is divided by the number of values. l Median (0): The center of the data set when data points are ranked in order from smallest to largest. , Minimum Detectable Concentration (MDC): The a priori radioactivity level that a specific instrument and technique can be expected to detect 95% of the time. Non-Impacted Area: An area where there is no reasonable potential for residual radioactivity from licensed activities. Nonparametric Statistical Test: A test based on relatively few assumptions about the exact form of the underlying probability distributions of the measurements. As a consequence, nonparametric statistical tests are generally valid for a fairly broad class of distributions. The Wilcoxon Rank Sum test and the Sign test are examples of nonparametric tests.' Outlier: A measurement that is unusually large or small relative to the rest and therefore is suspected of misrepresenting the population from which it was collected. . Power: The probability of accepting a survey unit as meeting the site release dose criterion l' when it actually does. The power is equal to one minus thefalse negative decision error or Type 11 decision error rate (i.e.,1-p). Precision: A measure of mutual agreement among individual measurements of the same propeny, usually under prescribed similar conditions, expressed generally in terms of the standarddeviation. Quality Assurance (QA): An integrated system of management activities involving planning, implementation, assessment, reponing, and quality improvement to ensure that a process, item, or service is of the type and quality needed. Quality Control (QC): The overall system of technical activities that measure the attributes and performance of a process, item, or service against defined standards to verify that they meet the stated requirements, operational techniques and activities that are used to fulfill requirements for quality. O Range: The measure of dispersion between the largest and smallest values in the data set. b 5-5 March 1999 I l
TROJANLICENSE TERhflNATIONPLAN > Reference Coordinate System: A set ofintersecting lines referenced to a fixed site location or benchmark. Typically the lines are arranged in a perpendicular pattem dividing the survey j location into squams or blocks of equal areas. Other pattems include triangular, polar, and three-dimensional coordinate systems. i Relative Shift (Mo): The ratio of the difference between the DCGL and the LBGR divided by the standard deviation in the concentration. I Remediation: The removal of radiological contaminants from, or their neutralization on, an area l to within levels established as acceptable. Remediation is sometimes used interchangeably with decontamination. , Repeat Measurement: A repeated analysis of the same sample or measurement repeated at the , same location. ' Residual Radioactivity: Radioactivity in structures, materials, soils, groundwater, and other i media at a site resulting from activities under the cognizant organization's control. This includes i radioactivity from all sources oflicensed activities, but excludes background radiation as specified by the applicable regulation or standard. It also includes radioactive materials remaining at the site as a result of routine or accidental releases of radioactive material at the site and previous burials at the site, even if those burials were made in accordance with the i provisions of10 CFR 20. 1 Scan Measurement: Radioactivity measurement obtained by moving the detector over a surface at a specified speed and distance above the surface to detect indiation. An indication of the
. resulting radioactivity level is read out directly.
Sign Test: A nonparametric statistical test used to demonstrate compliance with the site release dose criterion when the radionuclide ofinterest is not present in background and the distribution of data is not symmetric. See also Wilcoxon Rank Sum test. Site: Any installation, facility, or discrete, physically separate parcel ofland, or any building or structure or portion thereof, that is being considered for survey and investigation. Site Release Dose Criterion: The residual radioactivity that is distinguishable from background radiation results in a TEDE to an average member of the criticalgroup that does not exceed 25 mrem /yr, including that from groundwater sources. Standard Deviation: The measure of dispersion from the mean of the ' data set. l Surface Contamination: Residual radioactivity found on building or equipment surfaces and expressed in units of activity per surface area (dpm/100 cm2). I Survey: A systematic evaluation and documentation of radiological measurements with a correctly calibrated instrument or instruments that meet the sensitivity required by the objective of the evaluation. 5-6 March 1999 -
TROJANLICENSE TERMINATIONPLAN Survey Unit: A physical area of specified size and shape with similar characteristics and O potential for residual radioactivity for which data evaluation and statistical analysis are l performed. A separate decision is made for each survey unit as to its acceptability for release. Type i Decisioa Error: Secfalsepositive decision error. Type II Decision Error: Seefalse negative decision error. Unity Rule: A rule applied when more than one radionuclide is present at a concentration that is ! distinguishable from background and where a single concentration comparison does not apply. In this case, the mixture of radionuclide concentrations is compared against DCGL 's by applying the unity rule. The sum of the ratios for all radionuclides in the mixture should not exceed 1.0. Unrestricted Release: Release of a site from regulatory control without requirements for future ; radiological restrictions. Also known as unrestricted use. Validation: Confirmation by examination and provision of objective evidence that the particular i requirements for a specific intended use are fulfilled. In design and development, validation concerns the process of examining a product or result to determine conformance to user needs. l Verification: Confirmation by examination and provision of objective evidence that the ! specified requirements have been fulfilled. In design and development, verification concerns the process of examining a result of given activity to determine conformance to the stated requirements for that activity. Wilcoxon Rank Sum (WRS) Test: A nonparametric statistical test used to determine compliance with the site release dose criterion when the radionuclide of concem is present in background. See also Sign test. i l l 5-7 March 1999
_ - . . - - . - - . - - - _ . _ - _ - . . . - - . . . ~ . . - . . , l' TROJANLICENSE TERMINATIONPLAN ' } l 1 5.2 ' SURVEY OVERVIEW : This section describes the scope and methodology of the final survey process, including quality
- assurance measures, access control procedures, and how implementation of the plan will
- l. demonstrate that the plant and site will meet the 10 GR 20.1402 criteria for unrestricted release of the site and license termination. Also described is the methodology used to develop guideline l values against which residual radioactivity levels remaining at TNP at the time of final survey will be compared. l 5.2.1 IDENTITY OF RADIOLOGICAL CONTAMINANTS The gross radionuclide inventory at the Trojan site was estimated during site characterization in l 1993-1994. The data are compiled in the Trojan Radiological Site Characterization Report ;
(Reference 5-7) and summarized in Section 2 of this Trojan License Termination Plan. Additional data continue to be gathered on radionuclide inventory from routine operational and l decommissioning surveys and supplemental site characterization work performed since : 1993-1994. The predominant beta-gamma emitter on structure surfaces and in plant systems is i "Co; in soil, bottom sediment, and pavement it is '"Cs. 5.2.2 _ SITE RELEASE CRITERIA j 5.2.2.1 Radiological Criteria For Unrestricted Use I The site release criteria correspond to the radiological criteria for unrestricted use given in _ .) 10 CFR 20.1402, which are: '
- l. Dose Criterion: The residual radioactivity that is distinguishable from background radiation results in a Total Effective Dose Equivalent (TEDE) to an average member of the critical group that does not exceed 25 mrem /yr, including that from groundwater sources of drinking water; and 1
- 2. ALARA Criterion: The residual radioactivity has been reduced to levels that are ALARA. .
I 5.2.2.2 Conditions Satisfying The Site Release Criteria Levels of residual radioactivity that correspond to the allowable radiation dose and ALARA l levels above are calculated (derived) by analysis of various scenarios and pathways (e.g., direct l radiation, inhalation, ingestion, etc.) through which exposures could occur. These derived levels, referred to as derived concentration guideline levels (DCGL's), form the basis for the following conditions which, when met, satisfy the site release criteria: i
-1. The average residual radioactivity above background is equal to or below the DCGL; i
O . 5-8 March 1999
TROJANLICENSE TERMINATIONPLAN l l L 2. Individual measurements, representing small areas of residual radioactivity which ' exceed the DCOL, do not exceed the elevated measurement comparison DCGL. The elevated measurement comparison DCGL is described in Section 5.2.3.2.5;
- 3. Where one or more individual measurements exceed the DCOL, the average ;
residual radioactivity passes the Sign or Wilcoxon Rank Sum (WRS) statistical . l test. The application of the statistical test is described in Section 5.6.4.1; and l l
- 4. Remediation is performed where it is ALARA to mduce the levels of residual radioactivity to below the concentrations necessary to meet the DCOL's.
5.2.3 DEVELOPMENT OF DERIVED CONCENTRATION GUIDELINE LEVELS ! l 5.2.3.1- Dose Modeling I Dose models based on NUREG/CR-5512, Volume 1 (Reference 5-8), and appropriate to Trojan are used to calculate the DCOL's. The dose model translates residual radioactivity levels into ! potential radiation doses to the public and is defined by three factors: 1) the scenario,2) the ! exposure pathways, and 3) the critical group. The scenarios described in NUREG/CR-5512 " address the major exposure pathways of direct exposure to penetrating radiation and inhalation and ingestion of radioactive materials. The scenarios also identify the critical group. The critical ; group is the group ofindividuals reasonably expected to receive the greatest exposure to residual l radioactivity within the assumptions of the panicular scenario. The scenarios and their modeling : are specifically designed to be " reasonably conservative" by generally overestimating rather than ! underestimating potential dose. Four scenarios were considered for Trojan. They are: 1) building occupancy,2) building renovation or demolition,3) residential farming on landfill, and 4) residential farming at plant site. These scenarios, described below, represent reasonable and plausible human activities and future uses of the Trojan facility and site. They are based on those described in ! NUREG/CR-5512 and use the same modeling assumptions. In the case of the residential farming scenarios, the scenario definition and exposure pathways are changed from those in NUREG/CR-5512 due to site-specific considerations. No conditions exist at Trojan, outside those incorporated in the NUREG/CR-5512 scenarios and modeling assumptions, which would cause the estimated potential dose to the public to increase. . l l 5.2.3.1.1 Building Occupancy Scenario Because surface decontamination operations may not remove all of the surface radioactivity, a scenario describing surface contamination is considered. This scenario accounts for exposure to J I both fixed and removable thin-layer or surface radioactivity within a structure. This scenario assumes that individuals occupy the building in a passive manner without deliberately disturbing the residual radioactivity on building surfaces. Occupancy of the building is assumed to begin ' immediately after license termination. The exposure duration is assumed to be a full work year (2000 hours). The critical group consists of the buiMing occupants, who are the people who ! work in the building following license terminatio.i. 5-9 March 1999 l 1
' TROJANLICENSE TERMINATION PL4N The pathways that apply to the building occupancy scenario include:
- 1. Extemal exposure to penetrating radiation from surface sources;
- 2.- Inhalation of resuspended surface contamination; and
- 3. Inadvertent ingestion of surface contamination.
5.2.3.1.2 Building Renovation or Demolition Scenario At some point in time, the remaining site buildings may require renovation. Ultimately, they will be demolished.- During renovation or demolition, surface and volume sources are disturbed, creating loose contamination. This loose contamination can produce higher concen'. rations of j radionuddes in the air or on surfaces than the levels in an undisturbed building. This scenano assumes that building renovation or demolition activities occur immediately after license l termination. The exposure duration is a short-term, one-time exposure assumed to occur over a i six-month period. The critical group is the workers involved in building renovation or i demolition who work in, on, and nround the building following license termination. The pathways that apply to the building renovation or demolition scenario include: i l
- 1. External exposure to penetrating radiation from volume sources;
- 2. ' Inhalation of airbome radioactive dust; and
- 3. Inadvertent ingestion ofloose surface contamination. i 5.2.3.1.3 Residential Farming on Landfill Scenario Soil away from the site can be contaminated by licensed radioactive material. This scenario accounts for exposure to radioactive material as a result of the intentional disposal of concrete rubble and other debris such as pavement, masonry, and structural steel. The rubble and other debris has both surface and volumetric radioactive contamination and is hauled away from the
, plant site to a landfill. It is assumed the source term for this scenario is reduced by that fraction of the total rubble and debris volume used to backfill below-grade areas at the plant site. The residual radioactivity is assumed to be distributed in a surface soil layer covering the i contaminated landfill on pmperty that is used for residential and light farming activities. The scenario assumes continuous exposure via multiple exposure pathways to the critical group. The l critical group is the resident farming family who lives on the landfill, grows some portion of l their diet on the site, and drinks water from an on-site well.
The pathways that apply to the residential farming on landfill scenario include: l
- 1. External exposure to penetrating radiation from volume soil sources while outdoors; 5-10 March 1999
TROJANLICENSE TERMINATIONPLAN . l
- 2. Extemal exposure to penetrating radiation from volume soil sources while .
indoors;- O 3. Inhalation exposure to resuspended soil while outdoors; ,
- 4. Inhalation exposure to resuspended soil while indoors, j
- 5. ' Inhalation exposure to resuspended surface sources of soil tracked indoors; ,
i 6.' Direct ingestion of soil;
- 7. Inadvertent ingestion of soil tracked indoors; l
- 8. Ingestion of drinking water from a groundwater source; J
- 9. Ingestion of plant products grown in contaminated soil; ,
- 10. Ingestion of plant products irrigated with contaminated groundwater; !
- 11. Ingestion of animal products grown on-site (i.e., after animals ingest contaminated drinking water, plant products, and soil); and 1
- 12. Ingestion of fish from a contaminated surface water source.
5.2.3.1.4 Residential Farming at Plant Site Scenario Soil at the site can be contaminated from licensed operations by accidental spills, long-term accumulation of material in the soil via effluent releases, or intentional disposal or burial of concrete rubble and other debris such as pavement, masonry, and structural steel with both
~
surface and volumetric radioactive contamination. The residual radioactivity is assumed to be distributed in a surface soil layer covering the plant site on property that is used for residential
- and light farming activities. The scenario assumes continuous exposure via multiple exposure I pathways to the critical group. The critical group is the resident farming family who lives on the l plant site following site remediation and grows some portion of their diet on the site, but drinks water from a source away from the site.
The primary difference between this scenario and the landfill scenario is that the exposure pathways associated with contaminated groundwater and surface water are eliminated in this scenario. Physical characteristics constrain plant site use and make water exposure pathways improbable. The plant site proper is situated in a bedrock outcropping parallel to the Columbia River. It is a ridge of volcanic rock bordered on the east by the Columbia River and on the west
. by a lowland marsh, which formed a channel of the river near the year 1900. The site has 1
excellent drainage. The east side of the rocky ridge drains directly into the Columbia River,
.while runoff on the west side flows into the old river channel and thence by slough until it joins
., the Columbia River. < l 5-11 March 1999 y+
. _ . _ . - _ . . . . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ __._._.__m_
TROJANL1 CENSE TERMINATIONPLAN i i
. The pathways that apply to the residential farming at plant site scenario include:
- 1. External exposure to penetrating radiation from volume soil sources while -
outdoors;
- 2. External exposure to penetrating radiation from volume soil sources while indoors,
- 3. : Inhalation exposure to resuspended soil while outdoors; ;
i
- 4. Inhalation exposure to resuspended soil while indoors; ;
- 5. Inhalation exposure to resuspended surface sources of soil tracked indoors; j l
- 6. Direct ingestion of soil;
- 7. Inadvertent ingestion of soil tracked indoors; v
- 8. Ingestion of plant products grown in contaminated soil: and l
- 9. Ingestion of animal products grown on-site (i.e., aRer animals ingest contaminated plant products and soil).
~5.2.3.2 Derived Concentration Guideline Levels b The surface contamination and radionuclide concentration levels of structures, land areas, and plant systems remaining at the time of the final survey are compared to DCGL's calculated using the dose models. A DCGL is defined as the concentration of residual radioactivity distinguishable from background radiation which, if distributed uniformly throughout a survey unit,would result in a TEDE of 25 mrem /yr to an average member of the critical group. The average member of the critical group is the individual who is assumed to represent the most likely exposte situation based on the assumptions and parameter values used in the dose model calculatioc The DCOL's are cakulated based on the peak annual TEDE dose to the average member of the critical group expected within the first 1000 years aRer license termination.
DCGL's are presented in tenns of surface or volumetric radioactivity concentrations and are 2 expressed in units of dpm/100 cm or pCi/g. O 5-12 March 1999
TROJANUCENSE TEMflNATIONPLAN 5.2.3.2.1 Screening DCOL's A- ; Q Screening DCOL's, given in Table 5-1, were calculated using DandD, the NRC's computerized i dose modeling software. DandD uses the conceptual and math matical models developed in NUREG/CR-5512 and the generic input parameter values preoted in draft NUREG-1549 , (Reference 5-9). Screening DCGL's for surface and volumetric residual radioactivity were j developed using the building occupancy and residential farming at plant site scenarios. For l i volumetric residual radioactivity, the residential farming at plant site scenario more closely matches site conditions and provides DCOL's which are more reasonably conservative than either the building renovation m demolition scenario or the residential farming on landfill , scenario. The exposure pathways were modifiedi as described in Section 5.2.3. For both dose j model scenarios, the peak dose values occur the first year following occupation by the critical group. l 1 i 5.2.3.2.2 Site-Specific DCGL's Site-specific DCOL's may be developed by replacing generic input parameter values with site-specific parameters. The generic input parameter values presented in draft NUREG-1549 pnerally lie within the distributions of reported or expected values (i.e., are not at the extremes of the ranges) and explicitly exclude bounding or unrealistic assumptions. Physical parameters represent real conditions and expected variability across the United States. Behavioral and metabolic parameters represent the expected variability between individuals within the defined critical group. Site-specific data mayjustify the use of site-specific parameter values to reduce the conservatism of the screening DCGL and produce a more realistic estimate of site-specific - conditions. Site-specific parameter values r ed in developing site-specific DCGL's are documented, including the justification for their use. A treatment of uncertainty is included as - part of thejustification. The DandD computer code is designed to permit simple modifications to input parameter values, l such as th: exposure durations, intake rates, or concentrnions in various pathway media. Attachment I to draft NUREG-1549 provides informa: ion regarding the valid ranges for site-specific parameter changes that can be made within Dan iD without an additional uncertainty analysis. Where DandD cannot be easily modified to incorporate the use of site-specific data, other computerized dose modeling software, such as the US Department of Energy's RESRAD and RESRAD-BUILD, may be used to generate site-specific DCOL's. i l
'For the residential farming at plant site scenario, a single input parameter value is modified from the ,
l default values given in draft NUREG 1549. It is the infiltration rate,I, which is set to zero. 'Ihis eliminates the exposure pathways associated with contaminated ground and surface water. Also, by setting the infiltration rate to ; zero, the source term is not reduced by water percolating through the soil and carrying away surface and root zone contamination. The source term is fully available to the remaining exposure pathways. 5-13 March 1999 , 4
., - , . , , , , ,n.-- - - - - ,
TROJANLICENSE TERMINATIONPLAN , l 5.2.3.2.3 Surrogate Ratio DCGL's Surrogate ratio DCGL's may be established for areas where fairly constant radionuclide O ' concentration mtios can be demonstrated to exist. The established ratio a concentrations allows the concentration of every radionuclide to be expressed in terms of any one of them. Likewise, a surrogate ratio DCGL allows the DCOL's specific to hard-to-detect radionuclides in a mix to be expressed in terms of a single radionuclide which is more readily measured. The measured radionuclide is called the surrogate radionuclide. A sufficient number of measurements, spatially separated throughout the area ofinterest, are taken to establish a consistent ratio ofrr.dionuclide concentrations. The number of measurements needed to determine the ratio is based on the chemical, physical, and radiological characteristics of the radionuclides and the site. The surrogate ratio is based on the raost conservative value from the measurement data set. Where the standard deviation of the data set is greater than 1/3 of the mean value, the 95% upper confidence bound is used in place of the most conservative value. The 95% confidence level value of the mean, p., is calculated as follows (derived from EPA QA/G-9," Guidance for Data Quality Assessment"(Reference 5-10), Box 3.2-1, Step 3 equation): p, = x + ' fy,y (Equation 5-1)
. where:
= mean, calculated using Equation 5-10 2
O t(95 w
=
t statistic for 95% confidence at n-1 degrees of freedom (values of can k &ained kom M WG-9, Tame A-O o = .t<95%.ohd stand deviation, calculated using Equatica 5-11 n = number ofmeasurements Once an appropriate surrogate ratio is determined, the DCGL of the measured radionuclide is modified to account for the represented radionuclide according to the following equation (NUREG-1575 (Reference 5-11), Equation 4-1): CGL*r (Equation 5-2) DCGLsn = DCGLsa x [(Cup /Csa)(DCGLsa)] + DCGLn, where: DCGLsn = modified DCGL for surrogate ratio DCGL3 , = DCGL for surrogate radionuclide DCOLg = DCGL for represented radionuclide Cg = Concentration ofrepresented radionuclide ! C, 3
= Concentration of surrogate radionuclide When a surrogate ratio is established using data collected prior to remediation, additional post- . remediation measurements are collected to ensure that the data used to establish the ratio are still 5-14 March 1999 i
TROJAN 12 CENSE TERMINATIONPLAN l appropriate and representative of the existing site condition. If additional post-remediation measurements are not consistent with the pre-remediation data, the surrogate ratio DCGL is re-established. Professionaljudgment is used to determine consistency. 5.2.3.2.4 Gross Activity DCOL's
. Where multiple radionuclides are present at concentrations which exceed 10 percent of their respective DCGL's, a gross activity DCGL may be developed. The gross activity DCGL enables field measurement of gross activity, rather than the determination ofindividual radionuclide j activity, for comparison to the radionuclide-specific DCGL. The gross activity DCGL, or DCGLag, for surfaces or volumes with multiple radionuclides is calculated using the following equation (NUREG-1575, Equation 4-4):
DCGLa . (Equation 5-3) S' + S' +... l" DCGL DCGL2 DCGL, where: f, = fraction of the total activity contributed by radionuclide n LDCGL, = DCGL for radionuclide n Different radionuclides or radionuclide combinations may exist on difTerent portions of the site and require the calculation of one or more site-specific gross activity DCOL's. DCGL's are based upon previously determined radionuclide distributions for specific areas. For areas where the radionuclide distribution has not been determined, the most conservative distribution resulting in the lowest DCGL of those specified areas is used. The distributions are based on the radionuclides identified in composite samples from the specific areas collected both during power operation and decommissioning. If new radionuclide distribution data are obtained during decommissioning and detennined to be more appropriate for use, the DCOL may be re-evaluated and altered during the course of the final survey. For soil contamination, specific radionuclides, rather than gross activity, may be measured. For a known mixture of these radionuclides, each having a fixed relative fraction of the total activity, site-specific DCOL's for each radionuclide may be calculated by first determining the gross . activity DCGL and then multiplying that gross activity DCGL by the respective fractional i contribution ofeach radionuclide. 3 5.2.3.2.5 Elevated Measurement Comparison (EMC) DCGL's The EMC DCGL is the DCOL modified to account for small areas whem elevated residual radioactivity levels exceed the DCGL. It is the level of residual radioactivity over a small, but dermed area, at which there is reasonable assurance the site release dose criterion is still satisfied. The EMC DCGL, or DCOLyyc, is dened assuming e.at se msh MoacMy b concentrated in a much smaller area rather than uniformly over the entire survey unit area. It is calculated and applied where the minimum detectable concentration (MDC) for performing sc: a measurements is larger than the DCOL or where one or more static measurement data points , 5-15 March 1999
. _. ~__ . _ . . _ _ _ _ _ _ . . . . _ ._ _ _ _ . . _ _ . . _ -
TROJANLICENSE TERMINATIONPLAN
' exceed the DCGL. The methodology used to calculate and apply the DCGLeuc is given in l Appendix 5-1.
l , 5.2.3.2.6 Unity Rule ,
^
When the concentrations of different radionuclides appear to be unrelated, such as wheie the radionuclides occurring in background have unknown or variable relative concentrations, there is , little alternative to measuring the concentration of each radionuclide and using the unity rule. : The exception would be in applying the most restrictive DCOL to all of the radionuclides. The unit rule is applied using the following' equation (NUREG-1575, Equation 4-5):
)
i C + C2 C" i
+... 51 (Equation 5-4) l DCGLi DCGL2 DCGL, where:
' ) C. = Concentration ofradionuclide n . l DCGL. = DCGL for radionuclide n ; i l l
\
l 5.2.4 FACILITY AND SITE CLASSIFICATION i l The facility and areas of the site do not all have the same potential for residual radioactivity and, ' accordingly, do not all need the same level of survey effort to demonstrate compliance with the
, site release criteria. Different parts of the facility and areas of the site are grouped into impacted and non-impacted areas based on the potential for residual radioactivity using the criteria given L below. Classification is based on professionaljudgment, in consideration of operational history, l site characterization data, operational surveys performed in support of decommissioning, and l routine surveillances.
5.2.4.1 Non-Impacted Areas e ' Areas that have no reasonable potential for remod adiaactivity from licensed activities are designated as non-impacted areas. These areas do not nwd sy level of survey coverage since there was no radiological impact from site operations. No surveys of non-impacted areas are performed. l
- i
- O l 5-16 March 1999 i
TROJANLICENSE TERAflNATIONPLAN 5.2.4.2 Impacted Areas Areas that have reasonable potential for residual radioactivity from licensed activities are designated as impacted areas. Impacted areas are subdivided into three classes described below and illustrated in the following figure. CLASSES OF IMPACTED AREAS i
~ 8 .b
.c s:
bY 5.e
^
83 ' eoj " y Class,2}$g@gf@ 32 j[3Wdgc@k@Expectedpre-remediationM; Q . s . [ dM u bresidual/$diostivilyf<DCOU.! L- _um
.am .
-o Class 3 No expected residual radioactivity 5.2.4.2.1 Class 1 Areas Class 1 areas are impacted areas where, prior to remediation, there are expected to be locations with concentrations of residual radioactivity that exceed the DCGL. Examples of Class 1 areas !
are: 1) site areas previously subjected to remediation,2) locations where leaks or spills are ) known to have occurred, and 3) areas where discrete radioactive particles may be found. ; 5.2.4.2.2 Class 2 Areas Class 2 areas are impacted areas where, prior to remediation, there are expected to be locations with concentrations of residual radioactivity detectable above background levels, but that do not exceed the DCGL. Examples of Class 2 areas are: 1) locations where radioactive materials were present in an unsealed form,2) potentially contarninated in-plant or on-site transport routes,
- 3) upper walls and overhead areas of buildings or rooms subjected to airbome radioactivity,
- 4) areas where low concentrations of radioactive materials were handled, and 5) areas on the perimeter of former contamination control areas.
5.2.4.2.3 Class 3 Areas Class 3 areas are impacted areas where there are not expected to be locations with concentrations of residual radioactivity detectable above background levels. Examples of Class 3 areas are:
- 1) buffer zones around Class 1 or Class 2 areas, and 2) areas with a very low potential for residual contamination but where information is insufficient to justify a non-impacted classification. Previous remediation precludes an area from being classified as a Class 3 area.
5-17 March 1999
_.m_ _ TROJANLICENSE TERMINATIONPLAN 5.2.4.3 Initial Classification The initial classification of the facility and site is given in Table 5-2. It is based on site characterization data, the history of radioactive materials involvement or the known potential for contamination of an area, and the recommendations of Trojan personnel knowledgeable of site conditions. Site characterization data and radiological history are summarized in Section 2. 5.2.4.4 Changes In Classification The classification of an area may be changed in accordance with approved plant procedures prior to the start of final survey data collecdon in the area. Changes in classification are based on i survey data and other available information that indicate that another classification is more j oppropriate.
]l 5.2.5 FINAL SURVEY PROCESS )
Dismantlement activities occur prior to the start of the final survey process. Impacted systems ; and equipment are removed, remaining structures and surfaces are decontaminated, and ; operational radiation protection (RP) surveys of the work area are performed. The final survey i process begins with survey preparation activities, followed by survey design, data collection, and data assessment activities, and concludes with documentation of survey results. j 1 5.2.5.1 Survey Preparation O Survey preparation, described in Section 5.3, is the first step in the final survey process. The site release ALARA criterion, that residual radioactivity has been reduced to ALARA levels, is widressed during survey preparation. Remediation levels are established for various types of remediation actions. A remediation level is the level of residual radioactivity at which the desired beneficial effects due to the performance of a given remediation action are equal to the undesirable effects or costs of the action. Data from operational RP surveys are used to perform an ALARA evaluation. The ALARA evaluation examines the various remediation actions to determine which of them, if any, have a net benefit in further reducing the levels of residual radioactivity. Those remediation actions are taken and other preparations are made for survey i data collection. Work areas are grouped into tumover units and prepared for survey data collection. A turnover unit is an area which can be isolated and controlled to ensure that radioactive material is not introduced into the area from ongoing decommissioning activities in adjacent or nearby areas. Tools, equipment, and materials not needed for survey data collection are removed. Housekeeping and clean-up activities are completed. Scaffolding and other temporary equipment or material needed for survey data collection are put in place. Routine access, equipment removal, material storage, and worker and material transit in or through the turnover unit are no longer allowed. Once a walkdown has been performed and the tumover criteria met, control of activities within the turnove- unit is transferred from the Decommissioning organization to the Final Survey 5 18 March 1999
TROJANLICENSE TERMINATiONPLAN i organization. The turnover unit is isolated and access is controlled. Isolation and control , measures remain in place through survey data collection until license termination. O 5.2.5.2' Survey Design
\
Survey design, described in Section 5.4, identifies relevant components of the final survey ; process and establishes the assumptions, methods, and performance criteria to be used. Turnover units are divided into survey unit and classified as Class 1, Class 2, or Class 3. Systematic scan and static measurements are prescribed according to a pattem and frequency established for each classification. Investigation levels are established which, if exceeded, initiate an investigation of i the survey data. A measurement from the survey unit that exceeds an investigation level may indicate a localized area of elevated residual radioactivity. Such locations are marked and investigated to determine the area and the level of the elevated residual radioactivity. Depending on the results of the investigation, the survey unit may require remediation, reclassification, and/or resurvey. I i Quality control (QC) measurements are prescribed to identify, assess, and control measurement i error and uncertainty attributable to measurement methods or analytical procedures used in the i data collection process.- QC measurements provide qualitative and quantitative information to ] demonstrate that measurement results are sufficiently free of error to accurately represent the radiological condition of the facility and site. 1 5.2.5.3 Survey Data Collection ' O Survey data collection, described in Section 5.5, consists of the collection of survey data for subsequent analysis and preparation of a final survey report. Survey data collection begins with a turnover survey for survey units where areas of elevated residual radioactivity may be found. The turnover survey, performed prior to the final survey, is designed to' verify that residual radioactivity levels are acceptable and that no additional remediation will be necessary. It is performed using the same methodology, techniques, and quality control requirements as the final ; survey. The turnover survey is a biased survey which uses professional judgment to identify l 1
- measurement locations most likely to have elevated levels of residual radioactivity. If an area of elevated contamination is identified and remediation is determined to be ALARA, it is remediated and resurveyed to ensure it meets the final survey requirements.
The data collected during the tumover survey provides a sound basis for interpreting radiological !
- conditions that may be encountered during the final survey. Also, because the tumover survey is conducted using the same methodology, techniques, and quality controls as those required for the )
final survey, certain data collected during the turnover survey may be used as part of the final survey data set. Following the tumover survey, a final survey is performed. The final survey is a confirmatory survey to ensure that any residual radioactivity meets the 25 mrem /yr TEDE site release dose criterion. Measurement results stored as final survey data constitute the final survey of record and are included in the data set used to determine compliance with the site release dose criterion. 5-19 March 1999
TROJAN LICENSE TERMINATIONPLAN 5.2.5.4 Survey Data Assessment
- ' Survey data assessment, described in Section 5.6, is performed to verify that the final survey data are of adequate quantity and quality. Graphical representations and statistical comparisons of the data are made which provide both qualitative and quantitative information about the data. An assessment is performed to verify the data support the underlying assumptions necessary for the statistical tests. If the quantity, quality, or one or more of the assumptions are called into question, previous survey steps are re-evaluated and additional data are collected as necessary prior to further statistical analysis. The statistical tests are applied and conclusions are drawn from the data as to whether the survey unit meets the site release dose criterion.
)
5.2.5.5 Survey Results Survey results are documented in history files, survey unit release records, and in the final survey report. The final survey report is prepared which summarizes the data and states the conclusions. 5.2.6 PROJECT MANAGEMENT The planning and implementation of the fina'. survey process is performed by Trojan personnel supplemented by contracted personnel. Aspects of the finr.1 survey project are outlined below. 5.2.6.1 Final Survey Organization The organization tesponsible for the final survey project is illustrated in the figure below. The Manager, Personnel / Radiation Protection serves as the principal decision maker and is responsible for the overall implementation of the final survey project. The Manager, Personnel / Radiation Protection is also responsible for the overall integration of the final survey . project with decommissioning activities, and for the interface with the Nuclear Oversight l organizatior, on independent assessments and audits of final survey activities. Responsibilities and interfaces with other key plant positions are described in the Trojan Decommissioning Plan. Final Survey Project Organization i Quality Assurance { - - [g'[Pr ctio Decommissioning . i Final Survey Project Manager Survey Data Survey Preparat. ion Techm. cal Support Collection s .l 5-20 March 1999
m l TROJANIJCENSE TERMINATIONPLAN l l
- The Final Survey Project Manager oversees survey design, implementation, and assessment I activities and ensures they are performed in accordance with this plan. The Final Survey Project Manager approves the final survey results and conclusions.
The Survey Preparation staffis responsible for ensuring the facility and site are appropriately remediated and prepared for tumover to final survey. The Survey Data Collection teams are responsible for collecting survey data in accordance with this plan. The Technical Support staff - is responsible for providing technical suppon such as survey design, procedure preparation, technical evaluations, scheduling and coordination of activities, and data review. 5.2.6.2 Quality Assurance And Quality Control (QA/QC) QA/QC is an integral pan of the final survey process. The oblective of QA/QC, as applied to the final survey process, is to ensure the survey data collected are of the type and quality needed to demonstrate with sufficient confidence that the facility and site are suitable for release to unrestricted use. Steps are described in this plan to ensure: 1) the elem:nts of this plan are correctly implemented as prescribed,2) the quality of the data collected is adequate, and
- 3) corrective actions, when needed, are implemented in a timely manner and confirmed to be effective.
In addition to the above, applicable provisions of the Trojan Nuclear Quality Assurance Program, PGE-8010 (Reference 5-12), are applied to final survey activities. The final survey process is performed in a controlled, deliberate manner, providing assurance of accurate results. Surveys are performed by trained individuals with calibrated instruments following approved plant procedures, data are recorded and reviewed, and documentation is auditable. I 5.2.6.3 Survey Records And Documentation Generation, handling, and storage of final survey design information and survey data are ; controlled by approved plant procedures. Survey records and documentation are maintained as ! quality records and decommissioning records in accordance with approved plant procedures. Where possible, they are maintained as electronic media to reduce data transfer errors and to facilitate the use of statistical tools and the eventual reporting of survey results. 5.2.6.3.1 Procedures Survey activities which are essential to survey data quality are implemented and controlled by i approved plant procedures. 5.2.6.3.2 Technical Basis Documents Technical basis documents are prepared and maintained in accordance with approved plant ! procedures. Technical basis documents show what methods were used, how the methods are derived, underlying assumptions, the basis for deviations from this plan, and other information that should be properly documented. They may include position papers, calculations, computer code verification resuits, file memoranda, correspondence, etc. I 5-21 March 1999 L
TROJANLICENSE TERMINATIONPLAN ' DCGL documentation includes reports generated by DandD to verify the version of the software - g used in the analysis. Site-specific data supporting any changes made to input parameters also is : g included. If other computer models are used, sufficient information is documented to allow for i review of the model, scenarios, and pa.ameters. :
; 5.2.6.3.3 Records Records of activities affecting quality are maintained in accordance with approved plant ;
. procedures and the Trojan QA Program. The final survey report and records showing the results l_ of surveys and calibrations are maintained for a period of 3 years following the completion of the final survey. l 5.2.6.4 Audits And Independent Reviews Audits are performed to verify survey cctivities comply with established procedures and applicable aspects of the Trojan QA Program. Randomly selected survey unit release records ! and other final survey documentation are independently reviewed to ensure that the survey j results are documented in accordance with approved procedures. Audits and independent reviews are performed, and the reports of the results are maintained in accordance with approved ; plant procedures. 5.2.6.5 Control Of Vendor Supplied Services i Quality-related services, such as instrument calibration and laboratory analysis, are procured from qualified vendors whose internal QA program is subject to approval in accordance with the Trojan Nuclear Quality Assurance Program. 5.2.6.6 Training Training is conducted to acin' eve initial proficiency and to maintain that proficiency throughout the final survey process. Personnel performing surveys receive training to qualify in the procedures being performed. Training includes:
- 1. Overview and objectives of this plan;
- 2. - Procedures governing the conduct of the final survey; )
- 3. Operation of the appropriate field and laboratory instrumentation;
- 4. Collection of final survey measurements and samples; and i
- 5. Survey data evaluation.
The extent of training and qualifications is commensurate with the education, experience, and proficiency of the individual and the scope, complexity, and nature of the activity. Records of training are maintained in accordance with approved plant procedures. 1 U 5-22 March 1999 e - g-v - r - ww, ~ w --r m
, ..---~___m--__ - . = . _ . _ _ . . . _ _ . .m . . . - . _ _ _ _ _ . _ _ . _ _ _ . - - _ _ _ _ - . . > _ TROJANLICENSE TERMINATIONPLAN
' 5.2.6.7 Schedule Final survey activities are planned, scheduled, and tracked as a part of the overall decommissioning planning process. The schedule is dependent upon the progress and -
completion of several decommissioning activities. Presently, survey data collection is estimated , to be completed in mid- to late-2002. Final survey activities are planned and discussed with the NRC and the Oregon Office of Energy sufficiently in advance to allow the scheduling of
- inspection activities.. -
b b i i 1 1 l I O 5-23 March 1999
TROJANLICENSE TERMINATIONPLAN 5.3 SURVEY PREPARATION Survey preparation is the first step in the final survey process. Remediation levels are established for various types of remediation actions. Using the remediation levels, an ALARA evaluation determines which remediation actions, if any, need to be performed. The necessary remediation actions are taken, and isolation and control measures are instituted in preparation for survey data collection. The site release ALARA criterion, that residual radioactivity has been reduced to ALARA levels, is satisfied once remediation actions determined necessary are completed. 5.3.1 REMEDIATION LEVELS Remediation levels are established for remediation actions such as chemical decontamination, wiping, washing, vacuuming, scabbling, spalling, abrasive blasting, and high pressure washing. A remediation level is the level of residual radioactivity at which the desired beneficial effects due to the performance of a given remediation action are equal to the undesirable effects or costs of the action. The methodology for calculating remediation levels is based on draft Regulatory Guide DG-4006 (Reference 5-13) and is provided in Section 4. 5.3.2 ALARA EVALUATION Residual radioactivity may remain once dismantlement activities are completed. The ALARA evaluation examines various remediation actions to determine which of them, if any, have a net benefit in further reducing the levels of residual radioactivity. The ALARA evaluation is prepared, approved, and maintained in accordance with approved plant procedures. Operational RP survey data are used to perform the evaluation. Residual radioactivity levels are compared to calculated remediation levels of possible remediation actions which could be, but have not been, taken. Where the level of residual radioactivity exceeds the remediation level, the remediation action is considered to have a net benefit. Therefore, it is considered cost effective and must be taken for the residual radioactivity to be considered ALARA. Conversely,if the i concentration is less than the remediation level, the level of residual radioactivity is already considered ALARA and the remediation action is not required to be performed. The ALARA evaluation is needed only tojustify not taking a remediation action. If a decision i has already been made to perform a given remediation action, there is no need to evaluate whether the action is necessary to meet the ALARA requirement. For example, if wiping down ' surfaces with loose radioactive contamination is a good practice that is applied regardless of radioactive contamination levels, then it does not need an ALARA evaluation. For those remediation actions considered but not taken, the ALARA evaluation includes the levels of residual radioactivity above which those remediation actions would have been justified. Remediation levels do not represent concentration limits that cannot be exceeded. Rather, they represent the threshold at which the given remediation action is taken. The ALARA requirement is met by performing the appropriate remediation action and not by being below a specified concentration after the action is taken. The ALARA evaluation ensures that efforts to remove residual radioactivity rre commensurate with the level of risk the residual radioactivity poses. 5-24 March 1999
N
- l. TROJANIJCENSE TERMINATIONPLAN -
5.3.3 TURNOVER , l Due to the large scope of the final survey and the need for some survey activities to be conducted - l in parallel with dismantlement activities, a systematic approach is established for the turnover of l facility and site areas to the Final Survey organization. 5.3.3.1- Turnover Units The facility and site are divided into turnover units. A turnover unit is a logical combination of structures, land areas, and/or plant systems where dismantlement activities are completed and , which can be isolated and controlled. Properly designed, tumover units facilitate the transfer of j control of areas (and their subsequent removal from routine use) and minimize the impact on planned or ongoing dismantlement activities in adjacent areas. , l
. 5.33.2 ' ' Walkdown A walkdown of the turnover unit is performed prior to turnover. The principal objective of the walkdown is to assess the physical state of the turnover unit and the scope of work necessary to l
- prepare it for final survey. During the walkdown, requirements are identified for accessing,-
isolating, and controlling the tumover unit. Support activities necessary to conduct the final l survey, such as scaffolding, interference removal, and electrical tag-outs, are identified. Safety concems such as access to confined spaces, high walls, and ceilings are also identified. For systems, the walkdown includes a review of system flow diagrams and piping drawings. The O walkdown is performed when the final configuration is known, usually near or after the completion of dismantlement activities. 5.3.3.3 Turnover Criteria The following criteria are satisfied prior to acceptance of a turnover unit by the Final Survey organization. The physical aspects of these criteria are verified during the walkdown.
- 1. Planned dismantlement activities within the turnover unit are completed;
- 2. Planned dismantlement activities affecting or adjacent to the tumover unit are j E completed O_R are evaluated and determined to not have a reasonable potential to
~ introduce radioactive material into the turnover unit; i l
. 3. ' An operational RP survey of the tumover unit is completed and outstanding items are addressed; y
- 4. Planned physical work in,' on, or around the tumover unit, other than routine surveillance or maintenance, is completed; I
l S. Tools, non-permanent equipment, and material not needed for survey data ! collection are removed; r 5-25 March 1999
e , : l TROJANLJCENSE TERMINATIONPLAN \ l h
- 6. Housekeeping, clean-up, and remediation of the turnover unit are completed; 4
- 7. Scaffolding, temporary electrical and ventilation equipment and components, and other material or equipment needed for survey data collection are doc mnented radiologically clean and left in place;
- 8. Worker and material transit paths to/through the turnover unit are eliminated or re-routed;
- 9. Appropriate measures are instituted to prevent the introduction of radioactive 4
material into the turnover unit for ventilation, drain lines, system vents, and other potential airborne and liquid contamination pathways; and
- 10. Measures are instituted to control access and egress and otherwise restrict radioactive material from entering the turnover unit.
5.3.3.4 Transfer Of Control Once a walkdown has been performed and the turnover criteria met, control of activities within the tumover unit is transferred from the Decommissioning organization to the Final Survey organization. The need for localized remediation within the turnover unit may be identified after transfer of control. Localized remediation may be performed under the control of the Final Survey f) V organization. Iflarge areas require remediation, the tumover unit may be transferred back to the Decommissioning organization to be reworked. l 5.3.3.5- Isolation And Control Measures l l The. turnover unit is isolated and access is controlled. Routine access, equipment removal, material storage, and worker and material transit through the area are no longer allowed. One or l more of the following administrative and physical controls are established to minimize the ; possibility ofintroducing radioactive material from ongoing decommissioning activities in adjacent of nearby areas. ; i 1
- 1. Personnel training; i
- 2. Installation of barriers to control access to area; j
- 3. Installation ofpostine with access / egress requirements; ;
- 4. Locking or otherwise securing entrances to the area; and
- 5. Installation of tamper-evident seals or labels.
Isolation and control measures are implemented through approved plant procedures and remain in place through survey data collection until license termination. 5-26 March 1999
- . .. - .- - -- . . . . - - . - . - . - - - .- - ~ ~ - .
TROJANLlCENSE TERMINATIONPLAN l 5.4 SURVEY DESIGN 1 The survey design identifies relevant components of the final survey process and establishes the assumptions, methods, and performence criteria to be used. The survey design is summarized in Table 5-3. For survey design purposes, impacted areas are categorized as one of three types:
- l. 1. Structures, which include building interiors and exteriors, major free-standing exterior structures, exterior surfaces of plant systems, and paved exterior ground surfaces;
- 2. Land areas, which include unpaved exterior ground surfaces; and
- 3. Plant systems, which include interior surfaces of process piping and components.
The application of survey design criteria to structures and land areas will vary based on the type of survey media and the relative potential for elevated residual radioactivity. For plant systems, many of the survey design criteria applicable to structures and land areas do not apply or are dictated by the physical system layout and the accessibility to system piping and compcnents. To accommodate these factors, the survey design integrates both probability-based (random) and
. Judgmental (biased) methods to data collection to achieve the overall objective of the final survey process.
5.4.1 SURVEY UNITS O Areas of the Trojan facility and site classified as impacted am divided into survey units to , V facilitate survey design. A survey unit is a physical area of specified size and shape with similar characteristics and potential for residual radioactivity for which data evaluation and statistical analysis are performed. A separate decision is made for each survey unit as to its acceptability for release. 5.4.1.1 Survey Unit Size Professionaljudgment is used to divide the facility and site into appropriately sized survey units. The area of each survey unit is sufficient to assure the total number of data points, based on the measurement frequency, enables a statistical evaluation of the data collected. Considerations for establishing survey units are physical characteristics, concentration levels, and previous remediation efforts, as well as spatial and logistical considerations. Survey units are sized to ensure that survey data points are relatively uniformly distributed l _ among areas of similar potential for residual radioactivity. As an example, a small, separate survey unit is created for an area of known residual radioactivity instead ofincluding it in a much larger survey unit where the probability for one or more measurements to be taken in the area of known residual radioactivity is greatly reduced. Survey units conform to site characteristics to the extent practical. They have relatively compact i~ shapes unless an unusual shape is appropriate for the site operational history or the site topography. Where possible, existing facility or site characteristics such as horizontal and l-5-27 March 1999
TROJANUCENSE TERMINATIONPLAN ; vertical structural support beams, concrete pour seams, or piping runs are used to define the
. boundaries of the survey units. '
O Survey unit sizes are given in Table 5-3. These sizes give a reasonable measurement density.
'Ihey are based on floor or ground surface area only. The survey unit size may need to be adjusted based on existing features of the facility or site. Where the survey unit surface area
'l includes or only consists of walls or ceilings, the survey unit is sized so as to preserve the dose ,
modeling assumptions of the dose receptor within a contaminated room and/or receiving , exposure from an infinite flat plane source. Ajustification is documented for those cases where the survey unit size does not conform to Table 5-3. 5.4.1.2 Reference Coordinate System A reference coordinate system is used to facilitate the selection of measurement locations and to provide a mechanism for referencing a measurement to a specific location ~so that the measurement location can be relocated. A reference coordinate system is a set ofintersecting ' lines referenced to a fixed site location or benchmark.- Typically, the lines are arranged in a perpendicular pattern, dividing the survey unit into squares of equal area; however, other types of ? patterns (e.g., triangular, polar, or three-dimensional) may be used. Scale drawings, maps, or photographs of the survey unit are prepared, along with an overlay of the reference coordinate system or grid system. It should be noted that the reference coordinate system is intended primarily for reference purposes and does not necessarily dictate the actual spacing or location of measurements. Physical gridding is used only where it is useful and cost effective. Where Class 1 and Class 2 survey units are gridded, the basic grid pattems are at I to 2 meter intervals on structure surfaces and at 10 to 20 meter intervals on land areas. For practical purposes, Class 3 areas may typically be gridded at larger intervals, for example,5 to 10 meters for large surfaces and 20 to 50 meters for land areas. The physical grid layout on structure surfaces is marked by chalk line or other appropriate means along the entire grid line or at line intersections. For land areas, the reference coordinate system is marked by wooden or metal stakes driven into the surface at line intersections, or by other appropriate surface markers. The selection of the appropriate marker depends on the characteristics and routine uses of the surface. 5.4.1.3 Background Reference Areas The residual radioactivity of a survey unit m. y be compared directly to the DCGL; however, the residual radioactivity may contain radionuclides which occur in background. To identify and evaluate those contributions attributable to licensed activities, representative background radionuclide concentrations are established using background reference areas. Background reference areas have similar physical, chemical, radiological, and biological characteristics as the survey unit being evaluated. They are usually selected from non-impacted areas, but are not j limited to natural areas undisturbed by human activities. Surveys are conducted of one or more i background reference areas to determine background levels for comparison with the conditions determined in specific survey units. Appendix 5-3 provides additional discussion in selecting I and applying background reference areas. 5-28 March 1999
TROJANLICENSE TERMINATIONPLAN , Background reference areas are not necessary where: 1) the residual radioactivity does not contain radionuclides occurring in background and the detection method is radionuclide-specific, 1 or 2) the background levels are known to be a small fraction of the DCGL. 5.4.2 SCAN MEASUREMENTS Scan measurements are performed to locate radiation anomalies that might indicate elevated areas of residual radioactivity that require further investigation. They are performed according to a preset pattem established for each classification. The level of scanning effort is proportional to
; the potential for finding elevated areas of residual radioactivity based on the history of the survey unit. l i
Scan measurements of Class I survey units are performed over 100 percent of the accessible3 i' surface area. The scan survey is designed to detect small areas of elevated residual radioactivity that are not detected by the static measurements using the systematic pattem. If the sensitivity of the scanning technique is not sufficient to detect levels of residual radioactivity below the DCGL, the number of static measurements may need to be adjusted. Appendix 5-2 describes how this is done. 4 Scan measurements of Class 2 survey units are performed over 10 to 50 percent of the surface area. _ Class 2 survey units have a lower probability of areas of elevated residual radioactivity 2 than Class I survey units. Those areas with the highest potential for elevated residual radioactivity (e.g., comers, ditches, and drains), based on professionaljudgment, are selected for Q D scanning. If the entire survey unit has an equal probability for areas of elevated residual radioactivity, systematic scans are performed along transects of the survey unit or of randomly selected grid blocks. A 10 percent scanning coverage is appropriate ifit is unlikely that any area would exceed the DCGL. Coverage of 25 to 50 percent is appropriate when there might be locations above the DCOL. Where scanning coverage of greater than 50 percent isjudged appropriate, the survey unit is reclassified as a Class I survey unit. 1 Scan measurements of Class 3 survey units are performed over usually less than 10 percent of 1 the surface area. Class 3 survey units have the lowest probability of areas of elevated residual - radioactivity. Those areas with the highest potential for elevated residual radioactivity, based on professional judgment, are selected for scanning. This provides a qualitative level of confidence
~ that no areas of elevated residual radioactivity were missed by the random measurements and 4
that there were no errors made in the classification of the survey unit. Scan measurements of plant systems are performed, where possible, according to the scan coverage for the class of survey unit. The amount of accessible surface area dictates the actual percentage of the surface area to be scanned. i
'3 Personnel health and safety are taken into consideration when determining whether an area is accessible.
5-29 March 1999
TROJAN LICENSE TERMINATIONPLAN
.5.4.3 STATIC MEASUREMENTS i
p Static measurements provide a quantitative measure of the radioactivity present at the location measured. Static measurements are collected at a frequenr.y and at representative locations throughout the survey unit such that a statistically sound conclusion regarding the radiological condition of the survey unit can be developed. Static measurements may also be collected at locations of elevated residual radioactivity identified by scan measurements. The types of static measurements taken are direct surface contamination measurements and soil and bulk material measurements. If the instruments and techniques used for scan measurements are capable of providing data of
~
the same quality as static measurements (e.g., detection limit, location of measurements, ability , to record ard document resul8 then scan measurements may be used in place of static measurements. The results of w measurements are documented for at least the minimum , number of static measurements. The same logic may be applied for using in situ gamma ; spectrometry instead of soil and bulk material measurements. l 5.4.3.1 Number Of Measurements Thirty measurements are collected per survey unit'. This number of measurements is more than sufficient to apply the statistical tests and to protect against the possibility of some of the data ' being unusable. If the estimated mean is greater than 1/2 of the DCGL or the estimated standard deviation is greater than 1/6 of the DCGL, additional measurements may be required. The methodology described in Appendix 5-2 is used to make this determination. 5.4.3.2 Measurement Locations i Measurements in Class 3 survey units and background reference areas are taken in random , locations. Random means that each measurement location in the survey unit has an equal l probability of being selected. The random selection process uses random numbers which correspond to the survey unit reference coordinates to establish the measurement locations. The random numbers are generated using a random number generator or other random selection method. For Class 1 and Class 2 survey units, a random-start systematic pattern is used in place of a random pattern. This is done to meet a survey design objective to locate small areas of elevated ! residual radioactivity that may exist within the survey unit. The starting point is determined by !
' the random selection process. For a square grid, the physical spacing of the measurement .
locations, L, is determined as follows (NUREG-1575, Equation 5-6): ! L-
^
(Equation 5-5) n For plant systems, system size and accessibility to system interior surfaces may not allow the full number ofmeasurements. O. 5-30 March 1999
1 TROJAN LICENSE TERMINATION PLAN ' where n is the number of measurements and A is the total surface area of the survey unit. The calculated value of L is rounded down to the nearest 1/10 meter. Using the reference coordinates, the measurement locations are identified around the starting point in a perpendicular manner at intervals of L. This process is repeated to identify the pattern of measurement locations throughout the survey unit. Where other than a square grid system is used, the physical spacing of the measurement locations is determined such that they are distributed around the starting point in a systematic, equidistant manner across the survey unit area. Measurement locations selected using a random selection process or a systematic pattem that do ) not fall within the survey unit area or that cannot be surveyed due to site conditions, including l health and safety considerations, are replaced with other measurement locations determined using the random selection process. Supplemental measurement locations are also determined ! using the random selection process. j Measurement locations selected based on professional judgment violate the assumption of l unbiased measurements used to develop the statistical tests and are not used in the statistical evaluation. However, special considerations are necessary for survey units with surface areas 2 2 less than 10 m , land areas less than 100 m , and some plant systems. The data generated from i these smaller survey units are obtained based on professional judgment, rather than on systematic < or random design. 1 5.4.3.3 Location Identification i Measurement locations within the survey unit are clearly identified and documented to ensure j that the., can be relocated if necessary. Actual measurement locations are marked with tags, self- 1 adhesive labels, permanent markings, stakes, notations on survey maps, or equivalent methods. j Each measurement location is identified by a unique identificatiou code or number. The number ! convention allows survey data to be referenced to specific measurement locations identified on I the photographs, drawings, or maps of the survey unit. 5.4.4 DATA INVESTIGATION The data investigation process is illustrated in Figure 5-1. Investigation results are documented in accordance with approved plant procedures. 5.4.4.i Investigation Levels Investigation levels, shown in the following table, are radioactivity levels that are based on the site release dose criterion which, if exceeded, initiate an investigation of the survey measurement. Investigation levels are established for each class of survey unit. 1 i O 1 I 5-31 March 1999 i
TROJANLICENSE TERMINATIONPLAN Investigation Levels Survey Unit Classification Sean Measurements Static Measurements Class 1 >DCOL >DCGL Class 2 >DCGL >DCGL Class 3 >DCGL >0.5 x DCGL
' If the MDC (see Section 5.5.2.4) is greater than the DCGL, the investigation level is the MDC .
l The principal purpose ofinvestigation levels is to guard against the possible misclassification of i the survey unit. They also serve as a QC check on the final survey process. A survey I measurement that exceeds an investigation level may indicate that the survey unit has been improperly classified. It may also indicate a failing survey instrument or a localized area of elevated residual radioactivity where there was a failure in the remediation process. For a Class I survey unit, while measurements above the DCOL are not necessarily unexpected, i any measurement exceeding the DCGL is investigated. The site release dose criterion allows ) individual measurements representing small areas of residual radioactivity to exceed the DCGL. However, any measurement that exceeds the DCGL is subject to the elevated measurement comparison (EMC), described in Appendix 5-1. For a Class 2 survey unit, any measurement above the DCGL is unexpected and is investigated. As there is a low expectation for residual radioactivity in a Class 3 survey unit, any static measurement exceeding 0.5 x DCGL is f investigated. If the scanning MDC exceeds the DCGL, any indication of residual radioactivity during the scan is also investigated. If a background reference area is to be applied to the survey unit, the mean of the background reference area measurements may be added to the appropriate investigation level to which the survey measurements are compared. Where an excessive number of measurements exceed the investigation level, the results are reviewed to ensure that the applied background reference area is appropriate. If any background reference area is determined to be inappropriate, it is adjusted as necessary and documented. 5.4.4.2 Investigation Locations, identified by scan or static measurements, with residual radioactivity which exceeds the investigation level are marked and investigated. The elevated survey measurement is verified ' or confirmed to actually exceed the investigation level. The area around the elevated measurement is _ investigated to determine the extent of the elevated residual radioactivity and to provide reasonable assurance that other undiscovered areas of elevated residual radioactivity do not exist. Scan coverage of the area being investigated is increased to 100 percent. Static measurements are also taken if scan measurements are not capable of providing sufficient data to characterize the elevated residual radioactivity. A posting plot, described in Section 5.6.2.1, is generated to document the area investigated and the levels of residual radioactivity found. 5-32 March 1999
{ TROJANIJCENSE TERMINATION PLAN I 1 l Depending on the results of the investigation, the survey unit may require remediation, l reclassification, and/or resurvey. Possible data results and investigation conclusions are shown i in Table 5-4. l Static measurements above the investigation level that should have been but were not identified l by scan measurements may indicate that the scanning meti od i:: inadequate. In that case, the ; scanning method is evaluated and appropriate corrective actions are taken and documented. j Corrective actions may include rescanning affected survey units. !
- 5.4.4.3 . Remediation Areas of elevated residual radioactivity above the DCGL Euc are remediated to reduce elevated residual radioactivity to acceptable levels. - Based on survey data, it may be necessary to remediate an entire survey unit or only a portion thereof. Remediation activities are beyond the scope of this plan and are addressed in Section 4 of this License Term'mation Plan.
5.4.4.4 Reclassification If survey measurements in a Class 2 or Class 3 survey unit exceed the DCGL or suggest that there may be a reasonable potential that contamination is present in excess of the DCGL, the survey unit is reclassified as a Class I survey unit. A Class 2 or Class 3 survey unit which is remediated is reclassified as a Class 1 survey unit. If survey measurements in a Class 3 survey unit exceed 0.5 x DCGL, the survey unit is reclassified as a Class 2 survey unit. If the extent of the elevated residual radioactivity (and corresponding remediation) is limited, O that area of the survey unit containing the elevated measurements may be separated out into a new survey unit and classified. The remainder of the original survey unit retains its original classification. This is illustrated in the following figure. Reclassification of Elevated Area of Survey Unit Survey Unit G28045A Class 2 O 7 rea ofs elevated \ f f~ ; activjty l Survey Unit G28045B ! Class 1 l l 1 O 5-33 March 1999
l TROJANUCENSE TERMNAHONPLAN i. 5.4.4.5 Resurvey if a survey unit is reclassified or if remediation activities are performed, then a resurvey using the methods and frequency appropriate for the new survey unit classification is performed.
- Other than additional scanning, a complete resurvey of a Class 2 survey unit determined to be a l Class 1 survey unit is not necessary provided remediation is not performed. I L
In the case where a new survey unit is separated out from an existing survey unit, Class 3 survey. j i units need only additional randomly located measurements to complete the survey data set. i i Class 1 and Class 2 survey units require a new survey design based on random-start systematic i measurement locations. ; Where only a small fraction of the area of a Class I survey unit is remediated, replacement ! measurements are collected within the remediated area. Their locations are determined using the l random selection process. ! 5.4.5 QUALITY CONTROL (QC) MEASUREMENTS j
. QC measurements are a component of the survey quality assurance process, and include quality !
checking and repeat measurements. Quality checking and repeat measurements are perfonned to i identify, assess, and monitor measurement error and uncertainty attributable to measurement : 5 methods or analytical procedures used in the data collection process . Quality checking includes i direct observations of survey data and sample collections, and sample preparation and analyses. ! Repeat measurements are multiple measurements at the same location. Repeat measurements ! lO provide quantitative information to demonstrate that measurement results are sufficiently free of error to accurately represent the radiological condition of the facility and site. Results of QC measurements are documented in accordance with approved plant procedures. 1 l 5.4.5.1 Type, Number, And Scheduling The type, number and scheduling of QC measurements are determined by a performance-based method, as described in Section 4.9.2 of NUREG-1575, and in accordance with approved plant ; procedures. This method is based on the potential sources of error and uncertainty, the
. likelihood of occurrence, and the consequences in the context of final survey data accuracy. The primary factors considered here are: 1) the number of persons or organizations involved in the das collection,2) the number of measurement types or analytical methods used, and 3) the time interval over which the data are collected. Other factors include:
- 1. Number of survey measurements collected; j 2. Experience ofpersonnelinvolved;
- 3. Types ofmeasurement methods or sampling and analytical procedures used; 5
- _ The error and uncensinty introduced by the instrumentation used to collect the data are assumed to be l controlled by the perfonnance ofinstrument calibration (Section 5.5.2.2) and response checks (Section 5.5.2.3).
5-34 March 1999 l l [ . . , - - , _ , - - ,, ,- -m- - , - , m--
. 1
' TROJANLJCENSE TERMINATIONPLAN i
- 4. -Variability of survey instruments used;
- 5. Level of radioactivity in the survey unit; and
. 6. How close the measurement level is to the detection limit.
,. De collection of QC ' measurements is initiated early in the data collection process to identify i problems and establish estimates of accuracy. QC measurements continue to be collected for the duration of the survey to verify sources of error and uncertainty are minimized and controlled. He factors that influence measurement accuracy are incorporated in the QC measurement L ' collection design. l L Measurement results from survey units may not readily lend themselves to the QC measurement process (e.g., direct measurement results at or near the detection limit). In this and similar cases, test areas may be set up with hidden adioactive sources or spiked media to allow the collection
' of the QC measurements.-
i 5.4.5.1.1 Scan Measurements _ Quality checking of surface scanning surveys are performed to evaluate the effectiveness of scanning methods for identifying areas of elevated residual radioactivity.' Repeat measurements for scan surveys are not meaningful, as scan surveys are qualitative and not quantitative. The frequency of quality checking of scan surveys is dependent on: 1) the number of surveyors,
- 2) the number of scanning methods employed,3) the time interval over which scanning data are collected, and 4) professionaljudgement.- In addition to quality checking, the ability of surveyors to identify areas of elevated residual radioactivity by scanning is periodically tested in accordance with approved plant procedures.
l 5.4.5.1.2 Static Surface Contamination Measurements l Repeat measurements of static surface contamination measurements are performed to assess i error and uncertainty associated with field measurement methods. Measurement locations are selected based on measurement results and represent the entire usable range ofresidual radioactivity found. The usable range of radioactivity includes the highest measurement result
- i. and the lowest measurement result with an acceptable measurement uncertainty compared to the I desired level of accuracy. Repeat measurements with results at or near the detection limit are not
. used because the measurement uncertainty is usually greater than the desired level of accuracy.
The number of repeat measurements is dependent on 1) the number of surveyors,2) the number . I of static surface contamination measurement methods employed, and 3) the time interval over which the measurement data are collected. For these measurements, the survey objective is to l estimate the variance in the accuracy for the specific method between zero and two times the L estimated variance at the 95% confidence level. Fifteen repeat measurements, based on I Table 4.3 ofNUREG-1575, provide this level of confidence. As a planning guideline, 15 repeat measurements are performed using each static surface contamination measurement ; l . method per surveyor per year. ' O 5-35 March 1999
TROJANUCENSE TERMINATIONPLAN , 5.4.5.1.3 Soil and Bulk Material Measurements O
. v
' QC measurements for soil and bulk materials are performed on split samples to assess error and 6
uncertainty associated with sample methodology and analytical procedures . A split sample is a collected sample that has been homogenized and divided into two or more aliquots for subsequent analysis. Selected samples are split into two separate samples. Both samples are 1 analyzed using the same method, but by ditTerent laboratories. An alternative is to submit both I samples to the same laboratory for analysis. i The number of QC measurements for soil and bulk materials is dependent on: 1) the number of laboratories,2) the number of analytical methods used, and 3) the time interval over which lab
' analysis of the samples is performed. As discussed in Section 5.4.5.1.2, the baseline number for discrete QC measurements is 15. As a planning guideline,15 QC measurements are performed using each analytical method per laboratory per year.
5.4.5.2 Measurement Accuracy Measurement accuracy is estimated using the results of QC repeat measurements compared to I the results of original measurements. For laboratory analysis, the results of the split samples are I compared to one another. The accuracy estimates based on two or more surveyors (or i laboratories) refer to the agreement expected when different surveyors or laboratories perform the same measurement using the same method. ) l Acceptance criteria for measurement accuracy are established by approved plant procedures. i Where the acceptance criteria are not met, an investigation of the data collection and/or sample analysis process is initiated to assess and identify the extent of error or uncertainty. The results ; of the investigation and the corrective actions taken are documented. 4 )
'The error introduced by laboratory instrumentation and the laboratory analyst is assumed to be controlled j by the laboratory internal QC program.
5-36 March 1999 l l l
t TROJAN 12 CENSE TERMINATION PLAN l 5.5 SURVEY DATA COLLECTION Survey data collection begins after the survey unit has been isolated and controlled to ensure that radioactive material is not introduced from ongoing decommissioning activities in adjacent or nearby areas. l 5.5.1 SURVEY PERFORMANCE
; Survey data are collected from the tumover survey, the final survey, and any investigation j surveys performed. The final survey uses both random and biased data collection methods and is i performed using the methodology. techniques, and quality control requirements prescribed in I this plan. The turnover and investigation surveys are biased surveys performed using the same l
' methodology, techniques, and quality control requirements as the final survey. A tumover (
survey, when performed, precedes the final survey. Investigation surveys are performed dunng , the tumover or final surveys, as dictated by survey data results.- l l 5.5.1.1 Turnover Survey l l A tumover survey of Class 1 and Class 2 survey units is performed where extensive ) dismantlement activities occurred and operational RP surveys performed do not provide 1 sufficient confidence that the survey unit is ready for the final survey. The turnover survey is designed to verify that residual radioactivity levels are acceptable and that no additional remediation is necessary. It is conducted using the same methodology, techniques, and quality controls as those required for the final survey. Professionaljudgment is used to: 1) identify measurement locations most likely to have elevated levels of residual radioactivity, and 2) establish the scanning coverage and the number of static measurements to be taken. The data collected during the tumover survey provides a sound basis for interpreting radiological conditions that may be encountered during the final survey. Also, because the tumover survey is conducted using the same methodology, techniques, and quality controls as those required for the final survey, certain data collected during the tumover survey may be used as part of the final survey data set. The data from the turnover survey also provide assurance to the Decommissioning organization that dismantlement and remediation activities are complete. The following example illustrates how the turnover survey and the data collected may be used. Dismantlement activities are completed and isolation and control measures are instituted in an ! area classified as a Class I survey unit. A tumover survey, consisting of 100 percent scan i coverage and biased static measurements, is performed. The survey data collected indicate that there are no areas which exceed the DCGL. Later, however, isolation and control measures are removed and the survey unit is used as a temporary lay-down and storage area. Some or all of the tumover survey data may no longer represent the radiological condition of the survey unit ! and cannot be used as final survey data. Still, that data coupled with the knowledge of the use of the area since the tumover survey was performed justifies the survey unit being classified as Class 2 for the final survey. ! ~
/'N V
5-37 March 1999
TROJANIJCENSE TERMINATIONPLAN 5.5.1.2 - Final Survey i A final survey of each survey unit is performed. The objective of the final survey is to collect data 'of a sufficient type, quantity, and quality such that the statistical tests can be applied to the survey unit and conclusions drawn with confidence regarding the radiological condition of the survey unit. 1 The final survey uses both random and biased data collection methods. Scan measurements are taken from biased selection measurement locations. Static measurement results used in the statistical tests are obtained from randomly selected measurement locations. 5.5.1.3 Investigation Survey
' An investigation survey is performed when one or more survey measurements exceed an l l
investigation level as described in Section 5.4.4. The purpose of the investigation survey is to l define the area and level of the elevated residual radioactivity. The data collected during the investigation survey are used.to characterize the area being investigated and to proside the basis for any further actions to be taken. 5.5.2 INSTRUMENTATION , Commercially available portable and laboratory instruments and detectors are used to perform three types of measurements: 1) surface scanning,2) direct surface contamination measurements, and 3) laboratory analysis of soil and bulk materials. Instrumentation is used to perform other types of measurements as dictated by survey data collection needs. The issuance, control, and operation of survey instrumentation and the use of radioanalytical programs are controlled by approved plant procedures. Related quality records are maintained in accordance with approved plant procedures. 5.5.2.1 Instrument Selection Radiation detection and measurement instrumentation is selected based on reliable operation, detection sensitivity, operating characteristics, and expected performance in the field. As a general rule, instruments used for static measurements are capable of detecting the radiations of concem to an MDC less than 50 percent of the DCGL. This allows detectability of residual radioactivity in Class 3 survey units at the investigation level of 0.5 times the DCGL. 1 l As a general rule, instruments used for scan measurements are capable of detecting the radiations of concem to an MDC less than the DCGL. Typical instrumentation that may be used is identified in Table 5-5. The detectors used for direct surface contamination measurements are typically operated with data logging survey meters. L 5-38 March 1999 1 h L
r TROJANIJCENSE TERMINATIONPLAN 5.5.2.2- Calibration And Maintenance I Instruments and detectors are calibrated for the radiation types and energies ofinterest at the site. Instrument calibration and maintenance are performed in accordance with approved plant procedures. If vendor services are used, these services are conducted in accordance with approved procedures and an intemal QA program that is subject to approval in accordance with the Trojan Nuclear Quality Assurance Program. Radioactive sources used for calibration purposes are traceable to the National Institute of Standards and Technology (NIST) for both Trojan and vendor operations. , d 5.5.2.3 . Response Checks Instrument response checks are conducted to assure constancy in instrument response, to verify the detector is operating properly, and to demonstrate that measurement results are not the result of detector contamination. Instrument response is checked before instrument use cach day. Portable instruments are also checked after instrument use each day. A check source is used that emits the same type of radiation (i.e., alpha, beta, and/or gamma) as the radiation being measured and that gives a similar instrument response. The check source does not necessarily use the same radionuclide as the radionuclide being measured. The response check is performed using a specified source-detector alignment that can be easily repeated. If the instmment fails i*s response check,' it is not used until the problem is resolved. Measurements made betweva the last acceptable response check and the failed check are evaluated and discarded, if appropriate. 5.5.2.4 Minimum Detectable Concentration (MDC) The MDC is determined for the instruments and techniques that are used for survey data collection. The MDC is the concentration that a specific instrument and technique can be expected to detect 95 percent of the time under actual conditions of use. 5.5.2.4.1 . Beta-Gamma Scan MDC for Structure Surfaces The scan MDC, or MDCx.n, for scanning structure surfaces for beta and gamma emitters is determined from Equation 5-6 (draft Regulatory Guide DG-4006, Equation 2): (1.38) 5 . (Equation 5-6) MDC,,, = h E: E, (A/100)I where: 2 MDC n = minimum detectable concentration for scanning surfaces (dpm/100 cm ) 1.38 = scan performance criteria, from draft Regulatory Guide DG-4006 B_ = number of backgr'ound counts in time interval t (ents) p = . surveyor efficiency,0.5, from draft Regulatory Guide DG-4006 Ei = instrument efficiency for emitted radiation (cpm /dpm) l E. = source efficiency for emissions / disintegration 2 A = area ofdetector(cm ) t = time interval of observation while detector passes over the source (min) 5-39 March 1999
- . . - . - . . . . - - . - - - - ~ . - - - _ - - . - - - - . - - . - ..
TROJANLICENSE TERMINATIONPLAN i The value of p represents a mean value for normal tield .onditions and is discussed in Section 6.6 of NUREG-1507, " Minimum Detectable Concentrations With Typical Radiation Survey hstruments for Various Contaminants and Field Conditions"(Reference 5-14). The value of E, is determined taking the dose modeling into consideration. For example, for the building occupancy scenario, the residual radioactivity is assumed to be on the surface, not embedded. Thus for embedded material, adjustments in E, are made to account for attenuation by overlying material. When the intemal pathways (inhalation and ingestion) are dominant, a value for E, that does not account for embedding of residual radioactivity or surface roughness is appropriate for the dose modeling and is therefore applied. The value of t is the actual time that the detector can respond to the source of radioactivity. It depends on the scan speed, the detector size in the direction of the scan, and the area of elevated residual radioactivity. 5.5.2.4.2 Alpha Scan MDC for Structure Surfaces Scanning for alpha emitters differs significantly from scanning for beta and gamma emitters in that the expected background response of most alpha detectors is close to zero. Since the time an area of elevated residual radioactivity is under the probe varies and the background count rate may be less than I cpm, it is not practical to determine a fixed MDC for scanning. Instead, another approach described in Section 6.7.2.2 of NUREG-1575 is used. Given the DCGL and a known scan rate, the probability of detecting an area of elevated residual radioactivity is calculated. When one or more counts are registered, the surveyor pauses scanning end waits for CJ interval of the pause corresponds to a 90 percent probability of detecting counts associated with elevated residual radioactivity at the DCGL. 5.5.2.4.3 Gamma Scan MDC for Land Areas Fm scanning land areas with a sodium iodide gamma detector, the MDC.c.,, values given in Table 6.7 of NUREG-1575 are used. 5.5.2.4.4 Static MDC for Structure Surfaces For static measurements of surfaces, the MDC ;c may be calculated using Equation 5-7 (draft Regulatory Guide DG-4006, Equation 3): i MDC = 3 + (4.65) & (Equation 5-7) I (K) (t) where: MDComu = minimum detectable concentration for static counting (dpm/100 cm') B = background counts during measurement time interval t (ents) ; t = measurement counting time interval (min) K = calibration constant (ents/ min per dpm/100 cm2) 5-40 March 1999
.. - -. ~.- --
. TRGIANllCENSE TERMINATIONPLAN The value of K ray include correction factors for the detection efficiency and detector geometry. j i
5.5.2.5 Detection Sensitivity , The detection sensitivity of typical detectors for surface contamination measure:nents is : estimated and the results are summarized in Table 5-6. The results are shown for the principal : instruments that are expected to be used for alpha and beta-gamma direct surface contamination ; measurements. j Count times are selected to ensure that the measurements are sufficiently sensitive with respect to the DCGL. For example, the count times associated with measurements for surface
- contamination and gamma spectral analysis (soil and bulk materials) are normally set to ensure an MDC ie no greater than 50 percent of the DCGL. 'Ihe scan rate associated with surface !
scans is normally set to enrure an MDCw. of no more than 75 percent of the DCOL. If the l
. MDC exceeds the DCGL, additional static measurements may be required, as discussed in l Appendix 5-1. i 5.5.3 SU'RVEY METHODS Survey measurements are performed in accordance with specific instructions contained in !
approved plant procedures. Measurements include surface scans, static surface contamination measurements, and laboratory analysis of soil and bulk materials. Other measurements, such as removable surface contamination and exposure rate measurements, may be obtained as required.
~5.5.3.I' ' Scan Measurements Scanning is performed to locate small areas of residual radioactivity above the investigation level.- If an area of elevated residual radioactivity is identified during the scan of a survey unit, , ' the area is marked for investigation. l 1 - Structure and plant system surfaces are scanned for beta-gamma emitting radionuclides. Beta scintillation or thin window gas-flow proportional detectors are normally used. Typically, the detector is held less than 2 cm from the surface and moved at 5 cm/s. The scan rate is adjusted I such that residual radioactivity can be detected at the investigation level. ;
Scanning for alpha emitters and low energy beta emitters (<100 kev) are limited to interior ; stmeture surfaces which are reletively smooth and impermeable. Porous or exterior structure surfaces and land areas are generally not scanned because of problems with attenuation and ; media interferences. Wre scanning is performed, alpha scintillation or thin window gas-flow ] proportional detectors are typically used. The detector is kept close to the ground, usually less than 1 cm, and moved at a rate such that there is a high probability of detecting elevated residual ! radioactivity. l Land areas are scanned for gamma-emitting radionuclides. Sodium iodide scintillation iodide detectors are normally used. The detector is held close to the ground surface, usually less than 6 cm, and moved in a serpentine (S-shaped) pattern while walking at a speed that allows the surveyor to detect the residual radioactivity at the investigation level. A scan rate of 5-41 March 1999
. TROJANLICENSE TERMINA TION PLAN l approximately 0.5 m/s is typically used; however, this rate is adjusted depending on the expected detector response.
5.5.3.2 Static Surface Contamination Measurements Static measurements are taken to detect surface contamination. Static measurements are I generally performed by placing the detector on or nur the surface to be measured, taking a discrete measurement for a pre-determined time interval, and recording the reading. A one minute integrated count is a practical time interval for most field survey m' strumentation and provides detection sensitivities that are usually below the DCGL. Ilowever, longer time intervals may be used as warranted. 2 2 Static measurements are taken with 100 cm detectors or are corrected to reflect a 100 cm area. When large area detectors are used, the observed contamination is limited to that contamination 2 that would be acceptable when confmed to an area of 100 cm . Since it cannot be discerned that the observed contamination is uniformly distributed, it is assumed that it could be attributed to an area of 100'cm2or less. In the event that contamination in excess of what would be acceptable for an area of 100 cm2is observed when using a large area detector, an evaluation is performed to ascertain compliance with the DCGL. Static measurements are typically restricted to relatively smooth, impermeable surfaces where ' the radioactivity is present as surface contamination. Because the detector is used in close proximity to the potentially contaminated surface, contamination of the detector or damage to the detector caused by irregular surfaces is considered before perfonning direct measurements. O 5.5.3.3 Soil And Bulk Material Samples I Soil and bulk material samples are collected and measured. Soil samples are generally collected down to a depth of 15 cm at static measurement locations. Sampling at greater depths is done in areas where site charact rization or other information indicates potential contamination at depths greater than 15 cm. Sample preparation may ir ,mde, but is not limited to, removing extraneous
- material, homogenizing, splitting, drying, compositing, and final preparation for counting. For
. QC repeat measurements, the sample obtained from the selected measwement lccation is
- homogenized and divided into separate containers. The separate containers are treated as separate samples throughout the remaining sample handling and analytical process.
Samples of paint chips, tank sediment, sewage sludge, roofing material, concrete, pavement, and other bulk materials are collected for laboratory analysis as part of biased static measurements. Such samples may be collected in drain receptacles, sumps, and other catchments in affected areas. Selected storm drain catchments r.re sampled in accessible locations on the site. These l samps are quantitatively analyzed by gamma spectroscopy for principal gamma-emitting radionuclides and the results compared to the DCOL. If residual radioactivity can be measured at DCGL levels by in situ techniques, this method may be used to replace or supplement the sampling and laboratory analysis approach For gamma- i emitting radionuclides, the above data may also be supplemented by several exposure rate and/or in situ gamma spectrometry measurements. 5-42 March 1999 i l
. , _ ~ _ _ _ ___ _ _ . _ . _ . _ . . _ _ . _ _ . _ _ . . . . _ _ _ _ _ _ . _ _
, TROJANLICENSE TERMINATIONPLAN 5.5.3.4 Special Measurements There are situations, described below, that need special consideration. The historical site assessment and site characterization surveys are used to indicate if residual radioactivity may be present and its extent. 5.5.3.4.'1 Cracks, Crevices, and Small Holes 5 Surface contamination on non-planar or irregular structure surfaces, such as cracks, crevices, and
~
small h*s, may be difficult to measure directly using field survey detectors and established techniques. Where no remediation has occurred and residual radioactivity is not expected above background levels, cracks, crevices, and small holes are assumed to have the same level of residual radioactivity as that found on adjacent surfaces. The accessible surfaces are surveyed the same as other structure surfaces and no special meesurement methods are applied. Where remediation has occurred on surrounding surfaces or where residual radioactivity above background levels is suspected, a representative sample of the surface contamination within the crack, crevice, or small hole is obtained. The level of residual radioactivity is measured and detection sensitivities are adjusted such that reasonable appmimations may be made using indirect measurement techniques. The accessible surfaces are surveyed the same as other structure surfaces except that they are included in places receiving judgmental scans when maning is done at less than 100 percent coverage. 5.5.3.4.2 Paint Covered Surfaces Surfaces painted to fix loose contamination in place are remediated prior to survey data collection. For other surfaces painted after plant start-up, representative samples in areas where it is suspected that elevated levels of residual radioactivity could have been covered over are taken and analyzed. Detection sensitivities are adjusted or remediation is performed as dictated by the sample analysis results.' No special measurement methods are applied to surfaces which were painted prior to plant start-up and have not been painted since. 5.5.3.4.3 Plant Systems, Floor Drains, and Embedded Piping i Surface contamination on internal surfaces, such as plant systems, Boor drains, and embedded i piping, may be inaccessible or difficult to measure directly using field survey detectors and i established techniques. Where no remediation has occurred and residual radioactivity is not ; expected above background levels, inaccessible or difficult to measure intemal surfaces are ! assumed to have the same level of residual radioactivity as that found on accessible surfaces. No special measurement methods are applied. i Where remediation has occuned or where residual radioactivity above background levels is !
' suspected, representative samples of the internal surfaces are obtained. The levels of residual radioactivity are measured and detection sensitivities are established such that reasonable ;
approximations may be made using indirect measurement techniques or calibrated detectors
. extended into piping runs in a controlled manner. Accessible intemal surfaces are surveyed the 5-43 March 1999 L.,
r : TROJAN UCENSE TERMINATIONPLAN l
\
l same as other structure surfaces. Scale and sediment samples may be obtained, if appropriate i and as allowed by system size and accessibility to internal surfaces. J l_ 5.5.3.4.4 Activated Concrete and Other Materials
- Residual radioactivity _ within activated concrete and other materials is measured volumetrically.
Representative samples of the activated concrete or other material are collected and analyzed. Detection sensitivities are established such that reasonable approximations may be made using indirect measurement techniques, such as direct surface measurements or in situ gamma spectrometry. - f 5.5.3.4.5 Paved Parking Lots, Roads, Sidewalks, and Other Paved Areas l l Paved parking lots, roads, sidewalks, and other paved areas are treated as structure surfaces. L Scan and static measurements are taken as prescribed by the survey design. Where remediation
- has occurred or where residual radioactivity above background levels is suspected, direct surface contamination measurements are taken and representative subsurface samples are collected and analyzed. Depending on the size of the paved area and the distribution of the residual radioactivity, the paved area may be a separate survey unit or included as part of a larger survey ut.it.
5.5.3.5- Investigation Measurements i Removable activity, exposure rate, and in situ gamma spectrometry measurements may be taken p and used as' diagnostic tools to further characterize the radiological conditions and to evaluate U potential response actions. 5.5.4 SAMPLE HANDLING AND ANALYSIS When sample custody is transferred (e.g., when samples are sent off-site to another lab for l analysis), a sample tracking record accompanies the sample for tracking purposes. The sample tracking (or chain of custody) record documents the custody of samples from the point of measurement or collection until final results are obtained. Sample tracking records are controlled and maintained in accordance with approved plant procedures. On-site laboratory capabilities are used to perform gamma spectroscopy, liquid scintillation, and
. gas proportional counting of soil and bulk materials and other samples. Off-site laborato.y
- i. radionnalysis services are procured as needed. Laboratory analytical methods are generally i
!' capable of measuring levels at 10 to 50 percent of the DCGL. ! 5.5.5 DATA MANAGEMENT l Survey data are collected from the tumover survey, if performed; the final survey; and any investigation surveys. QC measurements are not recorded as final survey data. The following i . table illustrates acceptable sources of final survey data. Measurement results stored as final ! l survey data constitute the final survey of record and are included in the data set for each survey
. unit used for calculations to determine compliance with the site release dose criterion.
5-44 March 1999 E - - . - - . ._ - - .. . - - . l
TROJANLICENSE TERMINATIONPLAN l s' l i
. Sources Of Final Survey Data O Type of Type of Survey Measurement Turnover. . Final Investigation j t
Scan- V V V ; Static ) V V I 5.5.5.1 - l Scan Measurements' ! Scan measurements performed during the turnover survey and investigation surveys may be fi accepted as final survey data provided the following conditions are met: 1) survey data collection requirements prescribed in Section 5.5 are adhered to; and 2) no remediation is performed or, if ! performed, is localized in nature, contamination is controlled, and the area is rescanned; and
- 3) isolation and control measures are applied and maintained. Scan measurements are not recorded as discrete location measurements except where they are used in place of static measurements. -
i 5.5.5.2 Static Measurements : Static measurements performed during the tumover survey and investigation surveys are based . on professionaljudgment. Since they are biased and not random, they may not be used in the ,
- n statistical tests. However, this does not necessarily preclude their acceptance as final survey V data. Static measurements performed during the tumover survey may be accepted as final survey data provided: 1) survey data collection requirements prescribed in Section 5.5 are adhered to,
)
- 2) 30 or more data points are collected within the survey unit, and 3) none of the data points exceeds the DCGL. The data are stored as characterization data when they do not meet these criteria.
Where a survey unit is remediated and/or reclassified subsequent to a final survey, the affected data are stored as characterization data; otherwise, the data are retained as final survey data. 5.5.5.3 Data Recording Measurements are recorded in units appropriate for comparison to the DCGL by correcting for detector area, and measurement size as applicable. The background, recording units are dpm/100 cm f geometry,or surface contamination and pCi/g efficiency, for rad concentration. Measured numerical values are recorded and include values below the MDC and values that are negative (when the measured value is below the average background).- Measurement results from Inboratory analyses that are below the MDC are reported as the MDC value. Records of survey data are maintained in accordance with approved plant procedures. Survey data records include the identification of the surveyor, type of measurement, measurement location, measurement instrumentation used, measurement results, and time and date 4 Q measurement taken. 5-45 March 1999
..w .
. m. ._ _ _ _ __._ _ _ _ . . _ _ _ _ _ _ _ _ _ . . _. _ . _ _ _ _. _ _ .
TROJANLICENSE TERMINATIONPLAN 5.6 SURVEY DATA ASSESSMENT
+
l The survey data assessment process is illustrated in Figure 5-2. Final survey data, described in . l- s Section 5.5, are reviewed to verifythey are of adequate quantity and quality. Graphical i j representations and statistical comparisons of the data are made which provide both quantitative ; ! and qualitative information about the data. An assessment is performed to verify the data support the underlying assumptions necessary for the statistical tests. If the quantity, quality, or : , one or more of the assumptions are called into question, previous survey steps are re-evaluated l and additional data are collected as necessary prior to further statistical analysis. The statistical tests are applied and conclusions are drawn from the data as to whether the survey unit meets the i site release dose criterion. - l 5.6.1 DATA VERIFICATION AND VALIDATION - t , The final survey data are reviewed to verify they are authentic, appropriately documented, and I ! technically defensible. The review criteria for data acceptability are: ;
- 1. The instruments used to collect the data are capable of detecting the radiation of interest at levels less than the investigation level. If not, acceptable compensatory measures have been taken; !
l i
- 2. The calibration of the instruments used to collect the data is current and
! - radioactive sources used for calibration are NIST traceable;
- n 3. Instrument response is checked before and, where required, after instrument use each day data are collected; I
- 4. The MDCs and the assumptions used to develop them are appropriate for the instruments and the survey methods used to collect the data;
- 5. The survey methods used to collect the data are appropriate for the media and types ofradiation being measured; l 6. Special measurement methods used to collect the data are applied as warranted by survey conditions; 7 The custody of samples collected for laboratory analysis are tracked from the point of collection until final results are obtained; and i l
- 8. The final survey data set consists of qualified measurement results that are representative of the current facility status and randomly collected as prescribed by the survey design.
A discrepancy exists where one or more criteria are not met. The discrepancy is reviewed and the reasons for acceptability of the data or the corrective actions taken to restore data acceptability are documented.
- O r
5-46 March 1999 i i
TROJANLICENSE TERMINATIONPLAN
- 5.6.2 GRAPHICAL DATA REVIEW 4
Survey data are graphed to identify pattems, relationships, or potential anomalies in the data that might go unnoticed using purely numerical methods. A posting plot and a frequency plot are used. Other graphical data representation tools may be used, as appropriate, in addition to or in
- lieu of those described here.
l 5.6.2.1- Posting Plot Posting plots, generated during investigation surveys, are used to identify spatial pattems in the data. A posting plot is simply a map of the survey unit with the data values entered at the measurement locations. The posting plot can reveal spatial inhomogeneities in the survey unit such as patches of elevated residual radioactivity or groupings of measurements that exceed the DCOL. Even in a background reference area, a posting plot may reveal spatial trends in background data that might affect the results of the statistical tests. In some cases, the trends
- could be due to residual radioactivity, but may also be due to inhomogeneities in the survey unit background material.
5.6.2.2 Frequency Plot i
' A frequency plot is used to examine the general shape of the data distribution. A frequency plot is a bar chan of the number of data points witlun a certain range of values. The frequency plot may reveal any obvious departures from symmetry, such as skewness or bimodality (two peaks),
-in the data distributions for the survey unit or background reference area. When the data O distribution is highly skewed, it is often because there are a few elevated areas of residual radioactivity. The presence of two peaks in the data may indicate the existence ofisolated areas of residual radioactivity or a mixture of background concentration distributions due to different soil types, construction materials, etc. The greater variability in the data due to the presence of such a mixture will reduce the power of the statistical tests to detect an adequately remediated survey unit. These situations may indicate the need to more carefully match background reference areas to the survey unit or to divide the survey unit into survey units with more homogeneous backgrounds.
1 5.6.3 BASIC STATISTICAL COMPARISONS Statistical quantities (range, median, mean, and standard deviation) are calculated for the final survey data set where one or more data points exceed the DCGL. The calculated quantities are compared to the values shown in the following table. The statistical comparison values represent assumptions underlying the statistical test to be used. Where the statistical quantity fails the comparision, the data set and/or survey design assumptions are examined. l A U 5-47 March 1999
3 TROJANLICENSE TERMINATIONPLAN Basic Statistical Comparisons j O Q Statistical Quantity. Value Failure Response Range (R) Rs5o Examine data for outliers Median (0) l (0 - R)/ ol s 0.5 Examine data for outliers and anomalies l Mean (R) R s DCGL Apply background reference area or remediate i Standard Deviation (a) o s 1/6 DCOL' Determine if additional measurements are necessary ) I
- The survey design assumes an initial standard deviation value less than or equal to I/6 DCGL.
i
. 5.6.3.1 Range The range, R, is a measure of dispersion between the largest and smallest values in the data set. l
-It si calculated simply by subtracting the smallest value from the largest value. When there are i 30 or fewer data points, values of the range larger than 5 standard deviations are unusual. The I range may be wider for larger data sets. .
Where the range is greater than 5 standard deviations, the data are examined for outliers. Outliers are measurements that are extremely large or small relative to the rest of the data and, l therefore, are suspected of misrepresenting the population from which they were collected. i Outliers may result from measurement collection and recording errors. Outliers may also l represent true extreme values of a distribution, such as areas of elevated residual radioactivity, O- and indicate more variability in the population than was expected. Not removing true outliers and removing false outliers both lead to a distortion of estimates of population parameters. Tests developed to detect outliers in a data set may be used to identify data points that require further
. examination. ~A test alone cannot determine whether a statistical outlier should be discarded or corrected; this decision is based on professional judgment. !
b 5.6.3.2 Median j 1
)
The median,0, is the center of the data set when the data points are ranked in order from smallest i to largest. If the number of data points is odd, it is calculated using Equation 5-8 below: ii = xun + i// 2, (Equation 5-8) If the number of data points is even, then it is calcul ted using Equation 5-9: l
;7 . . In./ 2/ + ru /2/+s> (Equation 5-9) 2 1
where the data are ordered from smallest to largest and labeledmm x , x , . . ., x(n). 5-48 March 1999 1 h.. . . - , , .- , . , , . . - .
- . ~ . _ - - . . - - - - - _ - - . - - . - . - - . - . _ . - . . . . ~ . . - . - . - . . - .
i- TROJANLICENSE TERMINATIONPL4N
- l
. Large differences between the mean and the median are an indication of the skewness in the ,
data. ' A simple test for skewness is to subtract the mean from the median and divide by the standard deviation. Where the result is greater than 40.5, the data are examined for outliers or .i anomalies and professional judgment is used to discard or correct suspect data. . 5.6.3.3 Mean l l The mean, R, is the arithmetic average of a data set. It is calculated using Equation 5-10: l l n i - 1 ' l x = - I x, (Equation 5-10) i=1 i I where-n = number ofmeasurements i = measurement value for the ith measurement xi The mean is compared directly to the DCOL to obtain a preliminary indication of the survey unit j status. If the mean exceeds the DCGL, there are two options. Either the survey unit is remediated since it clearly does not meet the site release dose criterion or a background reference area, described in Section 5.4.1.3, may be applied to the survey unit. The mean of the O background reference area is subtracted from the mean of the survey unit and the ditTerence compared to the DCGL. Where the difference exceeds the DCGL, the survey unit is rejected and remediated. 5.6.3.4 Standard Deviation The standard deviation, o,is a measure of the dispersion from the mean of the data set. It is calculated using Equation 5-11: n E (x - x ) g 1 i=1 o= (Equation 5-11) n-I where the variables have been previously defined. Where the value of the calculated standard deviation is larger than that estimated in the survey design (the ini'tial value is 1/6 DCGL - see Appendix 5-2), an insuflicient number of measurements may have been taken and additional measurements may be necessary.
- O l
5-49 March 1999
Ip TROJANllCENSE TERMINATION PLAN 4 i 5.6.4 STATISTICAL TEST i The Sign or the Wilcoxon Rank Sunt (WRS) statistical test is applied to the final survey data set , where one or more measurements exceed the DCGL. The statistical test is based on the hypothesis that the level of residual radioactivity in the survey unit exceeds the DCGL. There must be sufficient survey data with levels of residual radioactivity at or below the DCOL to ;
' reject this statistical hypothesis and to conclude the survey unit meets the site release dose g criterion. .
L The Sign test and the WRS test are nonparametric tests. The basic distinction between , l parametric and nonparametric statistical tests is that the parametric tests use specifL assumptions i
- about the probability distributions of the measurement data. The most commonly made 1 l assumption is that the data fit a normal distribution. A nonparametric test does not assume nonnal data distribution. It uses fewer assumptions than a parametric test, and consequently requires less information to verify these assumptions and is less vulnerable to being found j L incormet when these assumptions are violated. That is, the correct decision is more likely made l about whether or not the survey data mean exceeds the DCOL, even when the data come from a i skewed distribution. !
l l The Sign test and WRS test assume the data are independent random measurements, and that the ;
. data are statistically independent or that there are no trends in the data. The WRS test also e assumes the data are in a symmetric, but not necessarily normal, distribution and that the background reference area and survey unit distributions are the same except for a possible shift in die mean. Both statistical tests are tests of the median. The parameter ofinterest, though, is the mean. If the assumption of symmetry is valid, then the median and the mean are effectively equal, and the tests also are tests of the mean. If the assumption of symmetry is violated, then the nonparametric tests of the median approximately test the mean.
If outliers or anomalies are present in the data set, the statistical test is performed both with and l' without the questionable data to see what efTect they may have on the results. The statistical test l l results are documented in accordance with approved plant procedures. l 5.6.4.1 Application Of Statistical Test ! The statistical test does not need to be performed when the survey data clearly show that the . l survey unit meets the site release dose criterion. The survey unit clearly meets the criterion if: ;
- 1) every measurement in the survey unit is less than or equal to the DCGL, or,2) where a l background reference area is used, the difference between the maximum survey unit measurement and the minimum background reference area measurement is less than or equal to the DCOL. . In these instances, the statistical test is not applied.
L The statistical test is applied where one or more measurements exceed the DCGL. Similarly for
- a curvey unit where a background reference area is used, the statistical test is applied where the difference between any survey unit measurement and any background reference area l measurement is greater than the DCC L.
i 5-50 March 1999
. - - ~ - . - . . _ - .- _~ . - ... .
TROJANLICENSE TERMINATIONPLAN Survey results and the corresponding conclusions, both when a background reference area is not used and when one is used, are shown in Tables 5-7 and 5-8, respectively. 5.6.4.2 Sign Test The one-sample Sign statistical test is used if the radionuclide of concem is not present in background and radionuclide-specific measurements are made. The Sign test may also be used if one or more radionuclides are present in background at such sm". fractions of the DCGL as to be considered insignificant. In this case, background concent ations of the radionuclides are included with the residual radioactivity (in other words, the : care amount is attributed to facility operations). Thus, the total concentration of the radionuclides are compared to the site release dose criterion. This option is only used ifit is expected that ignoring the background concentration does not affect the outcome of the statistical test. The advantage ofignoring a small background concentration is that no background reference area is needed. The Sign test is applied as follows:
- l. List the survey unit measurements, xi, i = 1,2,3, . . ., n; where n = the number of measurements.
- 2. Subtract xi from the DCGL to obtain the difference (DCGL - xi, i = 1,2,3, . . . , n).
- 3. Discard differences where the value is exactly zero and reduce n by the number of
]
such zero measurements.
- 4. Count the number of positive differences. The result is the test statistic S+. Note that a positive difference corresponds to a measurement below the DCGL and contributes evidence that the survey unit meets the site release dose criterion.
- 5. Compare the value of S+ to the critical values in Table I.3 of NUREG-1575. The table columns equate to the false positive decision error rate, n. The value of a is the probability of passing a survey unit which actually fails to meet the site release dose criterion, which is obtained from the survey design (the initial value is 0.05 - see Appendix 5-2). If S+ is greater than the critical value for the false l positive decision error rate given in the table, the survey unit meets the site release dose criterion. If S+ is less than the critical value, the survey unit fails to meet the criterion.
5.6.4.3 Wilcoxon Rank Sum (WRS) Test The two-sample WRS statistical test is used when the radionuclide of concern appears in background or if measurements are used that are not radionuclide-specific. Because gross 4 activity measurements are not radionuclide-specific, they must be performed for both the survey unit being evaluated by the WRS test and for the corresponding background reference area. C d 5-51 March 1999}}