ML20149K812
| ML20149K812 | |
| Person / Time | |
|---|---|
| Site: | Saxton File:GPU Nuclear icon.png |
| Issue date: | 02/29/1996 |
| From: | GENERAL PUBLIC UTILITIES CORP. |
| To: | |
| Shared Package | |
| ML20149K805 | List: |
| References | |
| NUDOCS 9602220205 | |
| Download: ML20149K812 (206) | |
Text
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M CHOICE OF DEC0FDlISSIONING ALTERNATIVE AND DESCRIPTION OF ACTIVITIES INVOLVED
- l. -
M PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND SAFETY I EJ PROPOSED FIFAL RADIATION SURVEY PLAN UPDATED COST ESTIt! ATE FOR DEC0!DtISSIONING F!i:THOD CHOSEN AND PLAN FOR ASSURING AVAILABILITY OF FUNDING FOR C05fPLETION OF DEC050!ISSIONING E TECHNICAL AND ENVIRON >! ENTAL SPECIFICATIONS IN PLACE DURING DECOS21ISSIONING M QUALITY ASSURANCE PROVISIONS IN PLACE DURING DECO >D!ISSIONING M PHYSICAL SECURITY PLAN PROVISIONS IN PLACE DURING DECO:D!ISSIONING REFERENCES IIM b AVERY' RE ADY INDEX " INDEXING SYSTE u i
SAXTON NUCMAR EXPERIDENTAL C(MtPORATION k DECOMMISSIONING FLAN TABLE OF CONTENTS 1.0 S UMMARY OP PLAN . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-1 l.1 DESCRIPTION OF DECOMMISSIONING PLAN AND DECOMMISSIONING ALTERNATIVE . . . . . . . . . . . . . . . . . . . . 1-1 1.1.1 Introduction .... ........................... 1-1 1.1.2 Contents of the Decommissioning Plan . . . . . . . . . . . . . . . . 1-3 1.1.3 Administration of the Decommissioning Plan . . . . . . . . . . . . 1-6 1.2 MAJOR TASKS, SCHEDULES AND ACTIVITIES . . . . . . . . . . . . . 1-7 1.2.1 Description of Major Activities . . . . . . . . . . . . . . . . . . . . . 1-7 l.2.2 Schedule for Decommissioning Activities . . . . . . . . . . . . . . . 1-8 1.3 DECOMMISSIONING COST ESTIMATE AND AVAILABILIT( OF FUNDS ........................................ 1-9 1.4 FINAL SITE RFI FASE PLAN ..... ................... 1-10 1 2.0 CHOICE OF DECOMMISSIONING ALTERNATIVE AND DESCRIP'I1ON OF l 1 ACTIVITIES INVOLVED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2-1 ! 1 ~ 1 2.1 DECOMMISSIONING ALTERNATIVE . . . . . . . . . . . . . . . . . . . . 2-1 I 2.1.1 No Action ................................. 2-1
; 2.1.2 Further Deferral of Dismantlement .................. 2-1 2.1.3 Immediate Dismantlement . . . . . . . . . . . . . . . . . . . . . . . . 2-3 2.2 DECOMMISSIONING ACTIVITIES, TASKS AND SCHEDULE .... 2-6 2.2.1 Activities and Tasks ........................... 2-6 2
2.2.2 Schedule ................................. 2-26 2.3 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES . 2-30 2.3.1 Vice Pmsident Nuclear Services Division ............. 2-30 2.3.2 Pmgram Director SNEC Facility . . . . . . . . . . . . . . . . . . . 2-30 2.3.3 Saxton Radiation Safety Officer (RSO) . . . . . . . . . . . . . . . 2-31 f% l i Rev.O
i SAXTON NUCLEAR EXPERIM(NFAL CORPORATION i gj DECOMMISSIONING PLAN 2.3.4 Radiation Safety Committee . . . . . . . . . . . . . . . . . . . . 2-32 2.3.5 Other Support .... .... ... ............... 2-32 2.3.6 Deconunissioning Organization . . . . . . . . ..... ... 2-34 2.4 TRAINING PROGRAMS . . . . ......... ............. 2-36 s 2.5 CONTRACTOR ASSISTANCE ........ . . . ... . .. 2-40 2.5.1 Contractor Scope of Work . . . .. .... ........ . 2-40 2.5.2 Contractor Qualifications and Experience . . . . . ...... 2-41 1 '2.5.3 Contractor Administrative Controls . .......... 2-41 3.0 PROTECTION OF OCCUPATIONAL AND PUBLIC HEALTH AND I SAFETY..... ...... ...................... . . . . . 3-1 3.1 FACILfTY RADIOLOGICAL STATUS ............... ... 3-1 3.1.1 Facility Operating History . . . . . .......... ....... 3-1
)
() 3.1.2 Current Radiological Status of Saxton facility ... ... .. 3-6 4 3.2 RADIATION PROTECTION ......................... 3-37 3.2.1 Radiological Controls Program . . . . . . . . . . . ... .... 3-37 3.3 RADIOACTIVE WASTE MANAGEMENT . . . ... ...... . 3-45 3.3.1 Fuel Disposal ......., ................ ... 3-46 3.3.2 Radioactive Waste Processing . .... . . .... .... 3-46 3.3.3 Radioactive Waste Disposal ...... ............ 3-57 3.3.4 Decommissioning Support Building . . . . ......... . 3-63 3.4 ACCIDENT ANALYSES . . . .. ....... .... ...... 3-68 3.4.1 Introduction ..... ........ ... ..... 3-68 4.0 PROPOSED FINAL RADIATION SURVEY PLAN . . . . . . ......... 4-1 4.1 FINAL RELEASE CRITERIA ... .......... ......... 4-1
% ii Rev.O
I h SAXMN NUCLEAR EXPERIMENTAL CORFOILATION (/ DECOMMISSIONING PLAN l 4.1.1 Site Release Criteria ........................... 4-1 4.1.2 Material Release Criteria .......... ............. 4-2 4.2 FINAL SURVEY METHODOLOGY . . . . . . . . . . . . . . . . . . . . . . 4-3 4.2.1 Overview . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-3 4.2.2 Instmmentation .............................. 4-4 4.2.3 Documentation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5 4.2.4 Quality Assurance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4-5
- 4.2.5 Independent Verification . . . . . . . . . . . . . . . . . . . . . . . . . 4-6 5.0 UPDATED COST ESTIMATE FOR DECOMMISSIONING METHOD CHOSEN AND PLAN FOR ASSURING AVAILABILITY OF FUNDING FOR I l
COMPLETION OF DECOMMISSIONING . . . . . . . . . . . . . . . . . . . . . . . 5-1 l Q 5.1 DECOMMISSIONING COST ESTIMATE . . . . . . . . . . . . . . . . . . . 5-1 . \d
5.2 DECOMMISSIONING FUNDING
....................... 5-2 6.0 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS IN PLACE
- DURING DECOMMISSIONING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 6.1 Technical Specifications .............................. 6-1 6.2 Environmental Specifications ........................... 6-2 7.0 QUALITY ASSURANCE PROVISIONS IN PLACE DURING l
DECOMMISSIONING . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.1 POLICY STATEMENT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.2 ORGANIZATION AND FUNCTIONAL QA RESPONSIBILITIES ... 7-2 7.2.1 General . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 7-2 7.2.2 Organization . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-2 7.2.3 Functional QA Responsibilities . . . . . . . . . . . . . . . . . . . . 7-3 [~h , V iii Rev.O 1
1 I O SAXTON NUCLEAR EXPERIMENTAL CORPORA'nON
) DECOMMISSIONING PLAN i
7.3 SAXTON FACILITY QUALITY ASSURANCE PROGRAM . . . . . . . 7-5 7.3.1 General Requirements .......................... 7-5 7.3.2 General Description . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-6 7.3.3 Saxton Facility Decommissioning Project Quality Assurance 1 i 2 Plan ......................... ........... 7-6 ! l 7.3.4 Procedures and Drawings . . . . . . . . . . . . . . . . . . . . . . . . 7-7 7.3.5 Quality Assurance Training . . . . . . . . . . . . . . . . . . . . . . . 7-8 7.3.6 Design Control ............. . . . . . . . . . . . . . . . 7-9 7.3.7 Procurement Document Control . . . . . . . . . . . . . . . . . . . . 7-9 , 1 1 i 7.3.8 Document Control ........................... 7-10 7.3.9 Contml of Purchased Material, Equipment, and Services . . . . 7-11 7.3.10 Identification and Control or AIaterials, Pans and Components 7-12 4 / 7.3.11 Control of Special Processes ..................... 7-12 ( 7.3.12 Inspection . . . . . . . . . . . . . ................... 7-13
- 7. 3.13 Test Control . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-14 7.3.14 Control of Measuring and Test Equipment . . . . . . . . .... 7-14 7.3.15 Handling, Storage, Shipping and Housekeeping . . . . . . . . . . 7-15
. 7.3.16 Inspection, Test, and Operating Status ............... 7-15 7.3.17 Nonconforming Materials, Parts or Components . . . . . . . . . 7-16 7.3.18 Corrective Action . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-16 1 7.3.19 Quality Assurance Records . . . . . . . . . . . . . . . . . . . . . . 7-17
- 7. 3.20 Audits . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ... 7-18 8.0 PHYSICAL SECURITY PLAN PROVISIONS IN PLACE DURING DECOMMISSIONING ..................................8-1
8.1 DESCRIPTION
.................................. 8-1 l
4 i x iv Rev.0 l l l h
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SAXTON NUCLEAR EXPERIMpnAL CORPORATION i DECOMMISSIONING PLAN i l,
9.0 REFERENCES
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9- 1 4
e 4 4 4 1 1 't l l 1 l ) 1 i i } d a i a P n i i i i i i 3 I t l I f i 1 ) t i 1 a f 4 1 ( v Rev.0 4 ( k E i l
f) i / SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN V List of Tables Table Title Pace 2.1-1 Occupational Dose Comparison Between Decommissioning Alternatives . 2-5 2.2-1 Occupational Exposure for immediate Dismantlement . . . . . . .. .. 2-28 2.2-2 Summary of Applications for Metal Cutting ..... ...... . . 2-29 2.5-1 List of Potential Contractors . ............. . ....... 2-42 3.1-1 Facility Operational History . . . . . . . . . . . . . . . . . . . . . . . . . .. 3-16 3.1-2 Area 1 Radiation Survey . ........ .... ....... ... .. 3-18 3.1-3 Area 2 Radiation Sun'ey . . . . . . . .............. . .. 3-19 3.1-4 Area 3 Radiation Survey . . . . . . . . ............ ..... 3-20 3.1-5 Area 4 Radiation Survey . 3-21
] ............ .... . . . ....
(,,/ 3.1-6 Area 5 Radiation Survey . . .. ............ ... ....... 3-23 3.1-7 Area 6 Radiation Survey . . . . . . . . ........... ......... 3-24 3.1-8 Area 7 Radiation Survey . . ..... .... ... .... . .. 3-26 3.1-9 Area 8 Radiation Survey . . . . . . . . . . . . . ............... 3-27 3.1-10 Alpha Contamination Area 1 . . . . . . . . . . . . . . . . . . . . ...... 3-28 3.1-11 Alpha Contamination Area 2 . . . . . . . ............. . . . 3-29 s 3.1-12 Alpha Contamination Area 3 . . . . . . . ......... .......... 300 3.1-13 Alpha Contamination Area 4 ..... ..... ... . .. .. . 3-31 3.1-14 Alpha Contamination Area 6 . .... .... ........ .... . 3-32 4 3.1-15 Composited General Area Smear Results ............. . . . 3-33 3.1-16 Reactor Vessel / Internals Curie Determination ............... . 3-34 3.1-17 Environmental Sample Analysis . . . . . . . . . . . . . . . . . . . ...... 3-35 3.1-18 Concrete Waste Volume Determination ............. . . . 3-36
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vi Rev.0
1 l l I SAXTUN NUCLEAR EXPERIMENTAL CORPORATION l l s DECOMMISSIONING FLAN 4 i 2 ) List of Tables (continued) i i
- Table Title Eagc 1
, 3.3-1 Examples of Plans and Procedures for Radioactive Waste I' Shipping and Disposal Compliance . . . . . . . . . . . . . . . . . . . . . . . . 3-64 i 3.3-2 Area / Component Waste Classification . . . . . . . . . . . . . . . . . . . . . . 3-65 i k 1 1 i d .I ) v s l 2 1 1 J d vii Rev.0 v
l SAITON NUCLEAR EIFERUMNTAL CORPORATION I ( DECOMMISSIONING PLAN i I List of Figures 1 Figure litic l 1.1-1 Saxton Site Layout 1.1-2 Rotary Bridge Crane 1.1-3 Containment Vessel, Sectional View j 1.1-4 Containment Vessel, Sectional View-West 1.1-5 Containment Vessel, Sectional View-North 2.2-1 Steam Generator 2.2-2 Pressurizer j 2.2-3 Reactor Vessel - Cross Section l 2.2-4 Project Schedule f] 2.3 1 Saxton Organization
/ 2.3-2 Saxton Organization - Operational Phase 3.1-1 Area Topographic Map 3.1-2 Location of Environmental Sample Points 3.1-3 Property Map Saxton Site 3.1-4 Saxton Site 12yout 3.1-5 Bridge Crane, Sectional View - West a
I 3.1-6 Containment Vessel, Sectional - West a 3.1-7 Containment Vessel, Plan above Op Floor 3.1-8 Containment Vessel Above 795'-2" 3.1-9 Containment Vessel Above 781'-4" 3.1-10 Containment Vessel Above 765'-8" 3.1-11 SNEC Monitoring Well I4 cations 3.1-12 SNEC Environmental Monitoring Stations 3.3-1 Preliminary Layout Decommissioning Support Building O viii Rev.O
/ SAXTON NUCMAR EXPEHt1 MENTAL CORFORATION Q] DECOMMISSIONING FLAN 1
SECTION 1.0
SUMMARY
OF PLAN 1 l 1.1 DESCRIFFION OF DECOMMISSIONING PLAN AND DECOMMISSIONING ALTERNATIVE 1.1.1 Introduction The Saxton Nuclear Experimental Corporation (SNEC) plans to decommission the Saxton facility Containment Vessel (CV), the concrete shield wall located around the NW and NE quadrant of , 4 the CV, the tunnel sections that are immediately adjacent to the outer circumference of the CV j and remaining portions of the septic system, weirs, and associated underground piping in preparation for release of the site for unrestricted use. The Saxton facility is a deactivated, pressurized water reactor (PWR), which was licensed to ! operate at 23.5 megawatt thermal (23.5 MWT). The Saxton facility is maintained under a Title 10 Part 50 License and associated Technical Specifications. 'Ihe license was amended to possess but not operate the Saxton facility reactor in 1972. The license expires on February 11,2000 or upon expiration of the SNEC corporate charter, whichever occurs first. , The facility was built from 1960 to 1962 and operated from 1962 to 1972 primarily as a research I and training reactor. The facility was placed in a condition equivalent to a status later defined J by the NRC as SAFSTOR after it was shutdown in 1972. Since then, it has been maintained in a monitored condition. All fuel was removed from the CV in 1972 and shipped to the Atomic Energy Commission (AEC) facility at Savannah River, S.C., who remained owner of the fuel. As a result, neither SNEC or GPU Nuclear have any responsibility relative to the spent fuel from the Saxton facility.
' l-1 Rev. O
I j h SAXTON NUCIEAR EXPERIMENTAL CORPORATION j DECOMMMMONING PLAN ; 1 In addition, the control rod blades and the superheated steam tes*. loop were shipped offsite. Following fuel removal, equipment, tanks, and piping located outside the CV were removed. The buildings and structures that supported reactor operations were partially decontaminated in 1972 through 1974. After the formation of the GPU Nuclear Corporation in 1980, SNEC formed an agreement with GPU Nuclear to use GPU Nuclear and its resources to maintain, repair, modify, or dismantle SNEC facilities as may be required. Both SNEC and GPU Nuclear are subsidiaries of the same parent company, General Public Utilities Corporation, (GPU). While SNEC remains the owner of the facility, a license amendment has been submitted to designate GPU Nuclear as a co-license l holder. GPU Nuclear will then have responsibility to comply with the license and technical - i specifications. GPU Nuclear will carry out the Saxton facility decommissioning on behalf of , p the site owner, SNEC.
- O !
1 Decontamination / removal of reactor suppon structures / buildings was performed in 1987, 1988, and 1989, in preparation for demolition of these stmetures. This included the decontamination of the Control and Auxiliary Building, the Radioactive Waste Disposal Facility, Yard Pipe Tunnel, and the Filled Dmm Storage Bunker, and the removal of the Refueling Water Storage Tank. Upon acceptance of the final release survey by the US Nuclear Regulatory Commission (USNRC), these buildings were demolished in 1992. 4 In November 1994, the Saxton Soil Remediation Project was completed. This was a comprehensive project involving soil monitoring, sampling, excavation, packaging and shipment of contaminated site soil. This program successfully reduced radioactive contamination levels i below the NRC current and presently proposed levels required to meet site cleanup criteria for unrestricted use. A U 1-2 Rev.O
SAITUN NUCMAR EIFBtIMDffAL CORFORATION DECOMMISSIONING FLAN Site-specific radiological and environmental data were obtained in 1995 as pan of the Saxton Site Characterization Plan (6575-PLN-4520.M)(Reference 1) to support the development of the Saxton facility Decommissioning Plan. The scope of the characterization plan extended over areas of the facility that may have become internally or externally contaminated or activated during the facility's operating history. Results of the characterization plan have been used to determine the current radiological status of the facility and are presented in Section 3.1.2. Facility arrangement drawings are provided as Figures 1.1-1 through 1.1-5. (Some components are shown out of plane for clarity.) 1.1.2 Contaata of the Decommiccianing Plan The Decommissioning Plan for the Saxton facility has been prepared and submitted in response to the requirements of 10 CFR 50. Specifically, the Decommissioning Plan has been developed to address the requirements of 10 CFR 50.75(f) and 10 CFR 50.82(a). The guidance of Draft Regulatory Guide Task DG-1005 (Reference 2) has been used in developing the Plan. In addition, the following experience was utilized in development of the plan: I
- other decommissioning plans submitted to the NRC;
- GPU Nuclear experience during Saxton facility reactor support buildings demolition;
- GPU Nuclear experience developed during TMI-2 Post Defueling Monitored Storage efforts;
- consultation with other Part 50 licensees involved with decommissioning activities.
O 1-3 Rev.0 4
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h SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING FLAN
'vl 7he following provides a summary of the requirements of 10 CFR 50 and the corresponding section(s) of the Decommissioning Plan where the information is presented.
10 CFR Requirement Descriotion/ Summary Plan Section 50.75(O Plan containing:
- decommissioning cost estimate 5.0
- major technical factors 2.2 50.75(0(1) Decommissioning alt,rnative 2.1 50.75(0(2) Major technical actions 2.2 x 50.75(0(3) current situation regarding radioactive 3.1.2 waste 50.75(0(4) Residual radioactivity criteria 4.0 1 1
50.75(0(5) Site specific factors which could affect 2.2,3.1.2 j decommissioning 50.82(b)(1) Decommissioning alternative and 2.1,2.2 description of activities 50.82(b)(2) Description of controls and limits on 3.2, 3.3, procedures and equipment to protect 3.4,6.0 occupational and public health and safety 50.82(b)(3) Description of final radiation survey 4.0 50.82(b)(4) Updated cost estimate and funding plan 5.1, 5.2 1-4 Rev.O
d( SAXTON NUCLEAR EXPERIMDfrAL CORFOItATION asComanoNmo ruN 50.82(b)(5) Description of:
- Technical Specifications 6.0
- Quality Assurance provisions 7.0
- Physical Security Plan provisions 8.0 The following is a brief description of each of the Sections presented in the Decommissioning
! Plan.
- Section 1: Summary of the Plan - This section provides a summary of the information presented in each of the sections of the decommissioning plan.
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- Section 2: Choice of Decommissionina Alternatives and Descriotion of Activities -
'(
n This section provides detailed information regarding choice of decommissioning alternatives, major decommissioning tasks and activities, schedule for completion, organizational responsibilities and training program details.
- Section 3: Protection of Occupational and Public Health and Safety - This section provides a description of the site, a summary of the operational history of the facility, the current radiological status of the facility, the Radiation Protection Program, and the radioactive waste management program. This section also presents the accident analysis that bounds events that could occur during decommissioning.
- Section 4: Final Radiation Survey Plan - This section provides the site release criteria and a description of the guidelines that will be used to develop the final .
' I radiation survey plan for the site. t V 1-5 Rev.O I i 4
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SAXTON NUCIEut EXPDti DffAL COBtPORATION (d onco ==ismoN o ruN
- Section 5: Decommissioning Cost Estimate and Fundine Plan - This section pmvides a summary of the decommissioning cost estimate and the funding plan.
- Section 6: Decommissioning Technical Specifications - This section provides a description of the Technical Specifications that will be in effect during the decommissioning effort.
- Section 7: Decommissionine Ouality Assurance Plan - This section provides a description of the quality assurance plan that will be in effect during the decommissioning effort.
- Section 8: Decommissioning Security Plan - This section provides a summary of the site access control and other security plan requirements that will be in effect during 5
\ the decommissioning period.
l
- Section 9: References - This section provides references cited tiuoughout the
- Decommissioning Plan.
1 1.1.3 Administration of the Decommissionine Plan i The Saxton facility Decommissioning Plan is a controlled document and will be reviewed and updated, as required by and in accordance with procedures. Revised pages will be clearly marked by side-bar and Revision number. l i 1-6 Rev.0 1
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l l O SAXTON NUCLEAR EXPERIMENTAL CORPOKATION C) DECOMMISSIONING PLAN 1.2 MAJOR TASKS. SCHRDUIR9 AND ACTIVITIR9 1.2.1 Description of Maior Activities Decommissioning activities have bwn divided into four distinct phases. These phases have been selected based upon the sequence of activities required to decommission the site in an orderly 4 manner.
- 1.2.1.1 Preparation Phase )
l The Preparation Phase provides for the necessary project management and support and includes those activities necessary to organize the Saxton facility work areas and project staff in order to A proceed in an orderly fashion. This phase includes tasks such as: preparation of program plans and implementing procedures, establishing radioanalytical facilities, procuring equipment and i 1 consumables, establishing a waste management area, and installing radiological monitoring I stations. In addition, facilities such as the decommissioning support building and utilities will be added or constructed. l 1.2.1.2 Operations Phase i The decommissioning Operations Phase principally includes removal of equipment and the decontamination of concrete and steel in the CV. The shipping and disposal of equipment and contaminated or activated building material is included in this phase. 1-7 Rev. 0 1
l A SAXMM NUCutAR EXMtIMWTAL CORPORADON Q DECOMMISSIONING FLAN 1.2.1.3 Final Survey Phase The Final Survey Phase includes detailed, comprehensive, and formal radiological surveys of the CV and the surrounding areas. As a result of this survey, evidence will be provided to confirm that radiological and hazardous contaminants have been reduced to acceptable levels in
, the facility and adjacent areas. Further details of the Final Release Criteria are provided in , Section 4.0.
1.2.1.4 Site Restoration Phase i The Site Restoration Phase is the final phase and is started after the facility has been released l by the NRC from the requirements of the NRC license. This phase includes the removal and I scrapping of the CV steel shell to three feet below grade, demolition of all remaining concrete l ,' % to three feet below grade, backfilling with additional structural fill, placement of topsoil and landscaping the site. i 1.2.2 Schedule for Decommissionine Activities A detailed schedule for decommissioning activities of the Saxton facility is presented in Section 2.2.2. . The present Saxton facility Decommissioning Project covers a duration of three and one half years, fmm 1996 through 1999, to complete the four phases of the pmject. 1 i i i l l l-8 Rev.O l l
O SAITON NUCLEAR EXIMEsttMENTAL COItPORATION DECOMMISSIONING PLAN I 4 l.3 DECOMMISSIONING COST ESTIMATE AND AVAILABILITY OF FUNDS l 2 l j A site-specific updated cost estimate was prepared in 1995 for decommissioning the Saxton j
; facility to account for the unique features of the facility and the dismantling of the Containment 1 Vessel. The total estimated cost for decommissioning the Saxton facility is estimated to be
$22.2 Million (1995 dollars). Further details of the site-specific updated cost analysis am i
pmvided in Section 5.0. A a j A Saxton facility decommissioning trust fund was established to fully fund the cost of ; decommissioning the facility for unrestricted use. Further details are presented in Section 5.0. i 1 4 i a i e t T
- s y ..
1-9 Rev.O
,I SAXTON NUCLEAR EXPERIMENTAL CORPORATION
( DECOMMISSIONING FLAN 2 1.4 FINAL SITE RRI RASE PLAN GPU Nuclear plans to meet the proposed site release criteria of 10 CFR 20 for release of the site for unrestricted use. The dose to an average member of the critical public will not exceed 15 millirem in any year for the following 1000 years due to any residual radioactive material
; of plant origin. Additional details of the Final Site Release Plan are provided in Section 4.0.
a C 1 4 1 1-10 Rev.0
i
/ SAXTON NUCLEAR EXPERIMENTAL CORPORATION h DECOMMISSIONING FLAN SECTION 2.0 CHOICE OF DECOMMISSIONING ALTERNATIVE AND DESCRIPTION OF ACTIVITIES INVOLVED 2.1 DECOMMISSIONING ALTERNATIVE The Saxton facility was placed in a condition equivalent to a status later defined by the NRC as SAFSTOR when it was shutdown in 1972. Since then, it has been maintained in a monitored condition and the plant structures, external to the containment vessel, have been dismantled.
The present NRC possessionenly license for the facility expires on February 11, 2000. In recognition of this the Saxton Nuclear Experimental Corporation (SNEC) has evaluated several options for the decommissioning of the Saxton facility. 1 2.1.1 No Action Q./ The No Action Alternative, as described in NUREG-0586 " Final Generic Environmental Impact l Statement on Decommissioning of Nuclear Facilities" (Reference 3), implies that a licensee ~ would abandon or leave a facility as is. This is not a viable decommissioning attemative and, I therefore, is not considered. 2.1.2 Further Deferral of Dismantlement The Saxton facility has been shut down since 1972, therefore, dismantlement has already been ! deferred for greater than 20 years. The option of deferral of dismantlement for an additional 30 years has been evaluated. Thirty (30) year additional deferral has the advantage of further radioactive decay thus reducing overall radiation exposure during dismantlement. Table 2.1-1 provides a comparison of 2-1 Rev.O
(q w/
)
SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN radiation exposure for the various alternatives. In spite of this advantage, deferral for 30 years has sevemi overriding disadvantages. The first 4 is the loss of an experience base currently available. SNEC's parent company, General Public Utilities (GPin, currently employs individuals who worked at the Saxton facility while it operated. Their knowledge of the plant from that era has proven and will continue to be invaluable. In addition, GPU Nuclear has recently remediated and demolished the contml, auxiliary and radioactive waste buildings and structures at the Saxton facility and placed Three Mile Island Unit 2 (TMI-2) in Post-Defueling Monitored Storage (PDMS). The skills of the people who worked on these projects are directly applicable to the remaining work at the Saxton facility and those same people will not be available in 30 years. In addition, a high ground water condition could lead to loss of containment which could either cause an unmonitored miease path or ground water flooding of the lower elevations of the containment vessel. As shown on Figures 1.1-2 through 1.1-5, the Saxton facility reactor vessel and other associated contaminated systems are located below ground level and ground water flooding would create an extremely difficult dismantlement scenario, increase the quantity of resulting radwaste, thus increasing the overall cost. Further, since the inside of the steel liner below gmde is covered by concrete on the inside, degradation of the liner could go undetected. Additionally, the high moisture content of the atmosphere inside the facility would hasten degradation of containment vessel systems and stmetural components (e.g. polar crane and related equipment) which will be needed to suppon dismantlement activities. This would result in making decommissioning activities less safe for workers as the components contiaae to deteriorate. l There is also the disadvantage of the continuing maintenance mquirements for the Saxton facility including an escalating effort to manage the deterioration of the facility over the next 30 years. 2-2 Rev.O
[] SAXTON NUCLEAR EXPERIMEIVTAL CORPORATION DECOMMISSIONING PLAN Q It makes no economic sense to spend money to monitor and maintain a facility that will never be used again. Finally, the cost of radioactive waste disposal in 30 years is likely to be much greater than the cost of disposal at the presently available facilities. The cost of the radioactive waste disposal has been rising at a much higher rate than that of inflation and therefore, it would be more expensive to wait until later to decommission the facility. Sites for the disposal of low level radioactive waste generated in Pennsylvania are currently available at the Barnwell, South Carolina Waste Management Facility and/or Envirocare of Utah; therefore the waste can be sent directly to burial. Future waste disposal choices are le:;s cenain, introducing the possibility of long term radioactive waste storage at the site. This is clearly undesirable due to the location of the site in a flood plain. The facility was never intended to be a long-term radioactive waste e storage site. ( For these reasons. the 30 year additional deferral of dismantlement was not selected. 2.1.3 Immediate Dismantlement The major advantages ofimmediate dismantlement of the Saxton facility are that it most quickly removes components from below ground level, stabilizes the radiological conditions at the site and allows the site to be released for unrestricted use. Immediate dismantlement also allows GPU Nuclear to make use of GPU's remaining Saxton facility and TMI-2 expenise for planning
- and implementing dismantlement activities. In addition, sites for the disposal of low level radioactive waste generated in Pennsylvania are currently available at the Barnwell, South !
l Carolina Waste Management Facility and Envirocare of Utah under present contracts, therefore the waste can be sent directly to burial, thus further minimizing decommissioning costs. O Q 2-3 Rev.0 1 1
l l
/ SAXTON NUCLEAR EXPERIMENTAL CORPORATION Q DECOMMISSIONING FLAN l
The major disadvantage to proceedir.g with immediate dismantlement is that radiation exposure l l to dismantlement personnel is higher for this option as compared to additional deferral. Since l the Saxton facility has been shutdown for over 20 years, the majority of personnel exposure savings to be gained from deferring dismantlement has aheady been achieved. The person-rem determination for the immediate dismantlement option is reasonable and in-line with current industry experience. The 17.2 person-rem difference is small and provides no overall benefit compared with removing the site as a source of radioactive material. l l Radiological conditions at the facility now are at a level that allows workers to safely remove ) components from the facility without threat to the safety of workers or local residents. i Additionally, the technology exists to safely and efficiently decommission the site now. I p t' Immediate dismantlement places the Saxton facility in a stable and secure condition in the shortest amount of time. It has been chosen as the preferred option. 4 4 I l l l l i 2-4 Rev. 0
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SAXTON NUCIEAR EXPEILIMENTAL COItFORATION I ( DECOMMISSIONING FLAN ~ TABLE 2.1-1
- Occupational Dose Comparison Between Decommissioning Alternatives !
, l 1
. Task 30 Year Deferral Immediate Person Person Rem Rem 0 l 9 Asbestos Remediation 2.3 4.9 1 System Dismantlement 8.1 17.7
- Reactor Vessel and Steam Generator 2.9 6.3
! Removal i ", Structure Decontamination and 0.2 0.35
- Dismantlement
'. Waste Management 1.1 2.5 1
Total 14.6 31.8 a 4 e 4 h 2-5 Rev.O 4
SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING FLAN 2.2 DECOMMISSIONING AC'ITVITIFR. TASKS AND SCHEDULE The scope of work for completing DECON decommissiordng at the Sntaa facility includes the removal of activated and contaminated systems and stmetures. Removal or dismantlement and removal will be performed for those systems and structures that do not meet the site release
- criteria discussed in Section 4.0.
The major decommissioning activities and tasks for the selected DECON method are discussed in Section 2.2.1. The schedule for the Saxton facility's decommissioning activities is presented and discussed in Section 2.2.2. It is expected that the tasks and activities described herein will i be further developed as additional detailed engineering and plannmg are performed. Table 2.2-1 provides a determination of the person-rem exposure associated with major dismantlement p activities.
- V 2.2.1 Activities and Tasks l
GPU Nuclear's decommissioning objective is to safely and efficiently remove activated and contaminated systems and stmetums to meet the site miease criteria identified in Section 4.0. I A discussion of the major decommissioning activities and tasks is provided in the following sections. I l 2.2.1.1 System Dismantlement l 1 GPU Nuclear has performed a site characterization study which shows the radioactive contamination and activation levels at the Saxton facility. A chameterization plan was developed and survey methods selected to record the site data in a concise manner. The results of the site t O 2-6 Rev. 0 f
O SAITON NUCLEAR EIMGLBGNTAL CORPORATION DECOMMISSIONING FLAN characterization study and subsequent additional radiological surveys indicate that the following systems are contaminated or activated:
- Main Coolant System
- Pressure Relief System l
- Charging and Volume Control System
- Component Cooling System i
- River Water System ,
- Steam /Feedwater System
- Cooling, Heating and Ventilating System
- Purification System i
- Resin Sluice System e Shutdown Cooling System o Safety Injection System
- Sample System
- Storage Well Cooling System l
- Vents & Drains System
- Waste Liquid System I
- Septic Tanks i Based on the results of the site characterization study, conceptual engineering and planning have been performed to determine the most advantageous approach to decommissioning the activated and contaminated systems and structures. Both conceptual and detailed engineering and planning have and will incorporate such considerations as: regulatory guidance, maintenance of occupational radiation exposure as low as reasonably achievable (ALARA), management of low i
] level radioactive waste (LLRW), industrial safety, environmental impacts, costs, and schedule. ) Another aspect considered is the use of field-proven and state-of-the-art dismantlement techniques. Saxton facility decommissioning activities will be performed under a QA program 4 O 2-7 Rev.0 9
4 i [ SAXTON NUCIZAR EXPERSEENTAL 00RFORA'Il0N DECOMMISSIONING FLAN as described in Section 7.0 of this plan. Similar consideration will be given for the ) dismantlement of non-contaminated / activated systems. Upon NRC approval of the Saxton facility Decommissioning Plan, those systems or structures that do not meet the release criteria will be dismantled and removed. Pipe and metal dismantlement and removal will be performed using shears, portable band saws, diamond wire e saws, abrasive wheel cutting, OD milling machine, or other suitable techniques. Scabblers, and CO2 blasters are options for removal of fixed contamination from concrete. Evaluations of the best alternatives are continuing as part of the further detailed engineering and planning. 'Ihe use J of water will be minimized due to the cost and schedule impact of disposing of the water. ! Table 2.2-2 (Summary of Applications for Metal Cutting) provides a listing of the types of ! dismantlement technologies available for use to dismantle systems and structures.
- A
< These technologies were selected based on a review of their effectiveness as discussed in the l j U.S. Department of Energy Decommissioning Handbook (Reference 4) and experience derived i from cleanup and non-radiological dismantlement activities at Three Mile Island Unit 2. l Radiological surveys, after dismantlement of systems and structures, will be performed to ensure that all contamination levels are at or below the release criteria. If contamination levels are discovered above the release criteria, remedial measures will be evaluated and implemented. All work performed as part of the Saxton facility decommissioning will be performed under the j controls described throughout this plan, and will be consistent with current industry standards
- and practices. These include appropriate radiological controls, radiological monitoring, contamination contml envelopes, local ventilation control with High-Efficiency Particulate Air (HEPA) filters, etc., as required to prevent the spread of contamination. A discussion of the l
t V 2-8 Rev.0 l 2
1 O SAXTON NUCLEAR EXPERIMENTAL CORPORATION V DECOMMISSIONING PLAN ALARA program is provided in Section 3.2 of this plan. l I l The major tasks associated with the Saxton facility decommissioning of contaminated systems are l
- 1) Construction of a Decommissioning Suppon Building adjacent and connected to the 1
containment vessel for segregating and packaging of waste.
- 2) Removal and packaging of any remaining asbestos insulation from affected systems and components.
- 3) Dismantlement and removal of all piping and other ponions of the systems that are
) not at or below the release criteria. This includes the dismantlement and removal d of all three (3) inch and smaller diameter piping that cannot be characterized properly or in a practical manner. It is GPU Nuclear's intention to ship such small bore piping to a licensed vendor for funher decontamination, mdiological surveys, I segregation, and ultimate disposal. l
- 4) Radiological characterization of the ponions of the systems currently remaining at the site to ensure that all contamination exceeding the release criteria described in Section 4.1 has been removed. These systems will then be removed to allow dismantlement and removal of the containment vessel stmeture.
U 2-9 Rev.0
SAX'IDN NUCIRAR EXPERinG!NTAL CORFORATION DECOMMISSIONING FLAN
- i j General l
l In general, the contaminated systems will be dismantled as follows: A. Premquisites ^
- 1. Pmcedures and work instructions are pmpamd and appmved in accordance with >
l approved station procedures. l
- 2. The system is deenergized, electrically disconnected, drained, and tagged out of
- service in accordance with appmved station procedures.
I Waste containers are located in an appmpriate area of the containment vessel for l\ [ 3. transpon to the decommissioning support building. 1
- 4. Radiological Controls preparation of the area to control and monitor contamination in accordance with appmved procedures. Contmis, such as contamination barriers, l catch basins and continuous airbome monitors, may be employed.
t l B. Activities
- 1. Removal of 3 inch and smaller diameter pipe and associated valves and instruments, i
l Although systems have been drained as part of decommissioning activities following i plant shutdown in 1972, there is the potential for residual amounts of potentially contaminated water to remain in the piping. Each piping segment will be walked down , 2-10 Rev.0 f
SAXTON NUCIEut EIFIELIMENTAL COItPOItATION ( DECOMMISSIONING FLAN l
\
l l and low points that have the potential to contain water will be identified. Present vents and drains will be utilized for those sections of piping, as available. A collection assembly will be installed at these locations to catch potentially contaminated water. To access piping located in the overheads appropriate temporary work surfaces will be pmvided. These surfaces may include scaffolding, platforms, and for lower heights, ; step ladders. Rigging will also be incorporated where appropriate to handle either heavy or awkward sections of small-diameter piping. Rigging will be field installed and will take advantage of existing stmetural features or hangers. Overhead hoists suspended by these stmetumi features or hangers, will be attached as appmpriate to piping sections ; p prior to cutting. Slings will be used to connect the hoists to the pipe. l
\ !
To remove the piping, the pipe will be severed at two locations using an appropriate l technique as discussed in Table 2.2-2. The distance between cuts will be selected so that the pipe section can be lifted and transported easily. Also, the section must be l sized to fit inside the selected waste container. After covering the open ends, pipe j sections will be moved to an appropriate location for transport to the decommissioning . support building where it will be packaged for loading onto a transport vehicle. i i
- 2. Removal of larger than 3 inch diameter piping. j l
- I Piping will be accessed as discussed for piping 3 inch and smaller. l To remove piping greater than 3 inch in diameter, the pipe will be severed at two j locations using a split OD track mounted milling machine or other appropriate technique i
,v 4 2-11 Rev.0
'/ SAXTON NUCLEAR EXPERIMFRTAL CORPORATION
\ DECOMMISSIONING PLAN as discussed in Table 2.2-2. Once the pipe is sectioned it will be handled and moved as discussed for piping 3 inch and smaller.
- 3. Dismantlement of System Components System components, pumps, valves, instrumentation, motors, heat exchangers, etc., will be n moved using mechanical techniques. They will be unbolted, disconnected, and removed. These components will be staged to an appropriate location and then transferred to the decommissioning support building where they will be appropriately packaged for shipment to a volume reduction facility. Two system components, the steam generator and the reactor pressure vessel which exceed the lift capacity of the Saxton facility polar crane, and the pressurizer will require specialized handling as
(-~'N discussed in Sections 2.2.1.3 and 2.2.1.4 of this plan.
- 4. Embedded and Inaccessible Piping Embedded and inaccessible piping consists of the piping, equipment and floor drains encased in concrete. The embedded and inaccessible piping that is contaminated will be removed as part of the dismantlement activities and shipped in a manner similar to that discussed above.
Based on past industrial experience, it is expected that cutting with band sa'vs and OD milling machines will not genemte significant levels of airbome contamination as both techniques produce heavier filings as opposed to dust. These filings will be collected in catch basins and vacuumed, as needed. However, GPU Nuclear will verify the mdiological conditions by performing a qualification program on appropriate sections of piping and appropriate contamination control will be provided, as necessary, at each b
\
2-12 Rev.0
i-I SAX 1DN NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN e system breach especially for systems with higher levels of TRU. Controls such as contamination barriers and taking airborne grab samples will be instituted, as necessary, as well as having continuous airborne monitoring. If the qualification program or other data indicate that airborne contamination would be generated, it will be controlled by use of temporary contamination control containments (tents, steel enclosures, glove boxes, and glove bags) with flEPA filters on the exhaust. The HEPA filtered ventilation will exhaust to the building atmosphere. The piping and components from non-contaminated systems will be surveyed following removal. Any piping and components that do not meet the release criteria will be disposed of in the same manner as above.
! 2.2.1.2 Structure Dismantlement As discussed in Section 3.1 the only remaining original buildings and stmetures at the Saxton ;
facility are the Containment Vessel (CV), the concrete shield wall located around the NW and l NE quadrant of the CV, the tunnel sections that are immediately adjacent to the outer circumference of the CV and portions of the septic system, weirs, and associated underground piping. In addition, in order to facilitate decommissioning activities, a decommissioning support building and several personnel trailers will be provided. The temporary structures provided to assist decommissioning will be removed from the site and their location surveyed for release as described in Section 4.0. Containment Vessel: The Saxton facility CV (Figure 1.1-2 through 1.1-5) is a circular steel stmeture approximately 109 ft. tall by 50 ft. in diameter with approximately 50 percent of the structure below grade. The CV is subdivided into a reactor compartment / storage well, primary 2-13 Rev.0 l
,Q t
SAXTON NUCLEAR EXPERIMENTAL CORPORATION Df'X'OMMISSIONING PLAN companment, auxiliary compartment, and an operating Door. These areas are separated from each other by concrete walls, Doors, and ceilings. Additionally the below grade portion of the . CV is lined with concrete. The major tasks associated with dismantlement of the CV are: ,
- 1) Radiological characterization to detennine the depth of penetration of activation and contamination into the concrete stmetun:s, and the extent of contamination on the remainder of building surfaces.
- 2) Removal of systems and components from the containment vessel.
1 g 3) Removal of contaminated concrete and other internal building structures, e.g., ( support steel, grating, etc.
- 4) Decontamination of the CV shell as needed.
- 5) Final site release survey as discussed in Section 4.0 to ensure all contamination above the release criteria has been removed.
- 6) NRC confinnatory survey. ,
1
- 7) Removal of remaining internal building stmeture exclusive of the concrete.
Demolition of the remaining building concrete as necessary to pennit backfill of
- the containment vessel.
- 8) Removal of the containment vessel to 3 feet below grade and backfill the void.
f3 t 2-14 Rev.0
O SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN This is based upon current industry practice and previous experience at the Saxton i site, as described in Section 3.0. Concrete Shield Wall: The concrete shield wall is a small exterior wall built along the NW and 3 NE quadrant of the containment vessel. This wall will be demolished and the uncontaminated concrete used as backfill during site restoration. If applicable, contaminated mbble will be processed as radwaste. ,
- i Tunnel
- The tunnel section immediately adjacent to the CV (Figure 1.1-1) originally carried ,
l system piping between the CV and other buildings on site. This piping was removed as part of ) the Saxton facility decommissicning activities that occurred following plant shutdown in 1972. j
) The major tasks associated with dismantlement of the tunnel are:
- 1) Radiological characterization of the tunnel to detennine the extent of contamination i
penetration into the concrete walls, floors, and ceilings and the extent of 4 contamination on the external CV wall. 1
- 2) Removal of groundwater from the tunnel.
- 3) Removal of contaminated concrete.
- 4) Decontamination or removal of contaminated CV steel.
- 5) Final release survey as discussed in Section 4.0 to ensure all contamination above the release criteria has been removed.
(~
- 2-15 Rev.O
SAX' ION NUCIZAR EXPERDENTAL COItPORATION DECOMMISSIONING FLAN i
- 6) NRC confirmatory survey.
- 7) Demolition of the tunnel to 3 feet below grade and backfill of the exposed tunnel.
l i Other Plant Structures: Ponions of the septic system, weils, and associated undergn>und piping still exist at the Saxton site. These areas will be characterized and ponions not meeting ] i the release criteria of Section 4.0 will be removed. j Teenporary Support Facilities: Trailers brought on-site to suppon the Saxton facility decommissioning and the decommissioning suppon building will be removed following completion of dismantlement activities. Areas they occupied will be characterized as discussed , in Section 4.0 to ensure all contamination above the release criteria has been removed. ! Saxton site: In addition to dismantlement and release of all Saxton facility structures, the i Saxton site will be characterized as discussed in Section 4.0. All areas not meeting the site release criteria will be remediated.
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2-16 Rev.O
l [3 SAI'IDN NUCIEAR EXPEILIMDfrAL CORPOILATION Q DECOMMISSIONING PLAN 2.2.1.3 Steam Generator and Pmssurizer Removal 2.2.1.3.1 Steam Generator Description ] 1 The Saxton facility steam generator, as shown in Figure 2.2-1, is a vertical shell and U-tube type I with an integral moisture separator and steam drier. The vessel shell (secondary side) is a barrel about 4 feet in diameter fabricated of 3 inch thick welded carbon steel plate. It is welded to an upper hemispherical head fabricated of the same material. The upper head is provided with a 15 inch ID man-way opening for access to the steam drum internals. The upper head contains a single 9 inch ID steam outlet nozzle at the top. The lower end is closed with the 9 f 1/2 inch thick tabe plate fabricated of carbon steel with a 3/8 inch thick cladding of 304 stainless l steel. It is drilled to accept the ends of the 736 U-tubes which are made of 5/8 inch OD 304 i (x stainless steel tubing. The lower hemispherical head is a carbon steel forging with a 304 stainless steel cladding. It is divided in half by a welded stainless steel partition plate that fonns the inlet and outlet water channels for the primary coolant side of the steam generator. Each I side of the lower head contains a 15 inch ID man-way opening. The inlet side contains a 12 inch ID coolant inlet nozzle and the outlet side contains a 14 inch ID coolant outlet nozzle. The , complete unit has an overall length of about 20 feet, a diameter of about 4 feet, and a dry weight of about 52,000 pounds. It is supported in place by four equally spaced lugs that are welded f to the outer shell at a location about 32 inches above the unit's center of gravity (i.e., the center
- of gravity for a dry steam generator). Dese four lugs are the primary rigging points.
l 2.2.1.3.2 Pressurizer Description ne Saxton facility pressurizer, as shown in Figure 2.2-2, is a vertical cylindrical vessel with an upper hemispherical head. It is fabricated of 3 inch thick ca1 bon steel with a cladding of 304 stainless steel on its internal surface. The lower head is welded to the bottom of the shell and
'D O
2-17 Rev.0
l t SAX 10N NUCEEut EXPEILIBOWTAL CORPORATION DECOMMISSIONING PLAN I i contains a flanged opening for installation of the heater bundle. The pressurizer contains several penetrations for instrumentation and process connections. It has an overall length of about 18 feet, a diameter of about 4 feet, and a dry weight of 25,000 pounds. It is supported in place by four equally spaced support lugs that are welded to the external surface of the shell and will serve as the primary rigging points for lifting the unit. 2.2.1.3.3 Steam Genemtor/ Pressurizer Curie Contents I
.Both the pressurizer and the steam generator are imernally contaminated from contact with the radioactive primary coolant. Both vessels am likely to require filling with grout for burial, but will qualify as an LSA shipment of a less than A2 quantity, as defined by 49 CFR 173.435. The activity contained in each of the two vessels is shown in Table 3.3-2. Disposal of the vessels p will be by burial. However, consideration will be given to optional disposal techniques such as k/ volume reduction through metal melt or compaction, and pocible decontamination.
2.2.1.3.4 Steam Generator / Pressurizer Removal Sequence
'Ihe approach to removal of the two vessels will first requite that all process piping attachments I
to the vessels be cut using an application from Table 2.2-2 or other suitable techniques and any attached instrumentation removed. Openings created by cutting the attached piping will be temporarily sealed to prevent release of contamination to the surrounding areas during handling. Actual removal of the vessels can be performed by installation of tempomry supports, cutting j of the four hanger rods, then moving the vessels horizontally for lifting up through the 1 removable hatch, or by cutting a new opening in the concrete floor at the 818'-4" elevation I directly above the vessels for removal. Removal of the steam generator and pressurizer vessels 2-18 Rev.0
SArf0N NUCULAR EXHGtEWINTAL CORMMLATION DEconseSSIONING PLAN from the containment can be accomplished through several different pathways. Both vessels will fit through the equipment hatch iflaid horizontally on a trolley. They may be moved into the attached support building for preparation for shipping, or, if coordinated with reactor vessel removal, they may be taken out thmugh the same containment dome opening as the reactor i vessel (Refer to Section 2.2.1.4). The activities involved in preparing the vessels for shipment will include removal, " fixing", or covering any external contamination. Vessel openings will be covered and sealed in a manner that will permit the vessels to be certified as "stmng, tight containers" in accordance with 49 CFR 173.24. A detailed survey will be performed to confirm that the vessels meet the requirements for an LSA shipment of a less than an A2 quantity of radioactive material. and then they will be loaded onto an " exclusive use" vehicle. Grouting of the vessels, if required, will p pmbably be performed at the site prior to shipment. The vessels will be shipped via pubhc
- roads and/or rail to the burial site in accordance with the applicable DOT and NRC regulations 4
in effect at the time of transport. 1 Safety during removal operations will be assured by the use of experienced and qualified l personnel to design and operate standard cranes and rigging equipment. Design specifications l for equipment and operations will invoke the applicable requirements of the Occupational Safety l and Health Administration (OSHA). i I 2.2.1.4 Reactor Pressure Vessel Removal i 2.2.1.4.1 Reactor Vessel Description The reactor vessel, as shown on Figure 2.2-3, is a vertical, right cylindrical vessel, fabricated J fmm carbon steel plates. 'Ihese plates were wrapped and welded to achieve a 5 inch wall . O V 2-19 Rev. 0
SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING LW thickness. Welded to the reactor vessel wall is the hemispherical shaped bottom head. The bottom head is fabricated from carbon steel and has a nominal thickness of 4.50 inches. The top head is flanged and gasketed, bolted to the vessel wall by 36 closure studs, nuts and washers. The top head has a nominal thickness of 5.25 inches, and is also fabricated from carbon steel. All internal surfaces of the reactor vessel are clad with stainless steel. The overall height of the reactor vessel is 18 feet with an external diameter of 68 inches. 'Ihe internal volume of the vessel is appmximately 268 ft'. The reactor vessel has the following openings (nozzles and ports): 4- 12 inch nozzles in the vessel wall (main coolant and safety injection systems); 6 - 3 inch ports in the vessel top head; l 5 - 2 inch ports in the vessel top head; 9 - 3 inch control md ports in the bottom head. The weight of the reactor vessel is supported by the reactor vessel support skirt. The support skirt is welded to the reactor vessel and bolted to the stainless steel reactor vessel support e assembly. 1 The weights of the principal reactor vessel components are as follows: Vessel (without internals or top head) 63,790 lbs. Thennal shield 10,491 lbs. ! , Top head 19,194 lbs. 4 2-20 Rev.O
) SAXTON NUCIEAll EXPERIMENTAL COItPOILATION 4
j DECOMMISSIONING FLAN Internals 7,300 lbs. Studs, nuts and washers 6,372 lbs.
; Steel shot assembly 16.828 lbs. L Total Weight 123,975 lbs.
The reactor vessel is surrounded by a nominal 4.0 inch thick insulation, contained within a 0.25 l inch thick stainless steel can. i j i 2.2.1.4.2 Meactor Vessel Curie Contents i i It has been determined that the reactor vessel contains a total of 1463.8 Curies, distributed as i follows as follows:
\
1,452 Curies (based on Preliminary Activation Analysis results) of neutron activated i internals and reactor vessel walls; and
- 11.8 Curies of radioactivity deposited on the internal surfaces of the reactor vessel ,
i l I , Based on the radioisotopes present, and the total curie content, the reactor vessel and internals can be transported as a Type B quantity of radioactive material, with packaging desiga ; exemptions due to the low Specific Activity of the contents. 'Ihe contents of the package (the I l reactor vessel and internals) meets the requirements of Class C waste, as defined in 10 CFR 61, based on the activation analysis (Reference 5). 2.2.1.4.3 Reactor Vessel Removal Sequence The reactor vessel will be removed from the containment vessel and placed into a sheltered 2-21 Rev.0 4
f SAXmN NUCtEAR EXPERIMENTAL CORmRATION Co M 0N o ruN laydown area to package the vessel for transportation to a licensed disposal facility. The internals will not be removed from the reactor vessel. The following is the sequence of activities associated with the mactor vessel removal operations (Note: This sequence may be modified based on ALARA considerations): A temporary reactor vessel laydown facility will be installed within the site boundary. The facility will provide weather protection for the reactor vessel, and will permit the decommissioning work fome to conduct packaging operations; The piping and instrumentation lines attached to the cmctor vessel will be cut using appropriate cutting technologies. Openings created by the cutting operations will be i temporarily sealed to preclude the release of surface radioactive contamination; Interferences such as steel plates, concrete shield blocks, etc. located directly above l the reactor vessel will be removed to permit a vertical lift of the reactor vessel; An opening in the steel containment vessel dome will be made above the reactor vessel. The opening will be of sufficient diameter such that the reactor vessel can be removed in one piece. When not transferring material through this opening, it will be covered to ensure the weather-tight integrity of the containment vessel dome; The reactor vessel (support skirt) will be separated from the reactor vessel support assembly; Appropriate radiological contamination and airborne control measures will be implemented to prevent the spread of such material prior to removal of the reactor i vessel from the CV; O 2-22 Rev.0 l 1 f e
SAI1ON NUCLEAR EXPEEtIMDffAL COItMMLATION DECOMMISSIONING FLAN i Using a crane, the reactor vessel, including the attached " insulation can", will be ! l lifted from the reactor vessel support assembly. As the vessel is lifted, and the ; exterior surfaces of the reactor vessel or insulation can become accessible, loose ) surface contamination present on the exterior surfaces will be " fixed" or contained within a plastic barrier; i I
- The crane will lift the vessel from the containment vessel and place it on a transporter ;
for conveyance to the sheltered laydown area; and
'Ihe crane will lift the removed section of the containment vessel dome back into place and the weather-tight integrity of the opening will be restored.
2.2.1.4.4 Reactor Vessel Packaging Activities It is intended to utilize the reactor vessel containing the activated internals as its own transportation package. The reactor vessel will be modified to satisfy the requirements of a Type B package (10 CFR 71.51), with an " Exemption for low specific activity" in accordance with 10 CFR 71.52. The following are the sequence of activities (Note: This sequence may be modified based on ALARA considerations): I Design packaging modifications to the reactor vessel and submit an application for package approval in accortlance with the requirements of 10 CFR 71; Subpart D - Application for Package Approval; I Remove (unbolt) the insulation can and remove the insulation surrounding the reactor vessel. Should the reactor vessel package design allow the retention of the insulation can, this activity may not be required; 2-23 Rev.0
I SAXTON NUCIEAR EXPE3tBGINTAL CORPORATION O Co ossioniwo r u n
- Remove the steel shot and steel shot support can fmm the bottom head of the reactor vessel. Should the reactor vessel package design allow the retention of the steel shot l and support can, this activity may not be required; Note: The following activities are mpresentative of the types of activities which may be required in accordance with the application for package appmval.
Inject concrete / grout into the reactor vessel through one or more of the existing l openings. The concrete / grout shall be injected in such a manner to ensure that essentially all interior void volumes have been filled with concrete / grout; Install welded closure caps on all openings in the reactor vessel;
- Install any structural members to the reactor vessel package to meet 10 CFR 71.45,
" Lifting and tie-down standards for all packages";
i Remove or "fix" any externalloose surface contamination to meet 49 CFR 173.443,
" Contamination Control", as necessary;
- - Install (weld /or otherwise attach) shielding around the reactor vessel to meet
- 49 CFR 173.441, " Radiation level limitations", as necessary; )
i
- Label the package in accordance with the requirements of 49 CFR 172.400, and 49 CFR 173.471; lead the reactor vessel package onto a transporter suitable for use on public 2-24 Rev. 0 4
1 l l l SAXTON NUCMAR EXPEILIMENTAL CORPOItATION < DECOMMISSIONING FLAN l highways and/or mil; and
)
Placard the transporter in accordance with the requirements of 49 CFR 172.500. l 2.2.1.4.5 Reactor Vessel Package Transponation The packaged reactor vessel will be transported to the disposal facility in Barnwell, SC on public highv/ays and railways using exclusive use carriers. The following are the activities which will ) take place prior to transponing the mactor vessel: I Roads (public highways) used on the transponation route will be inspected to determine the suitability for handling the weight and clearance requirements of the reactor vessel package. This may include an engineering analysis of the roadway structure, bridges, etc. whem necessary; Overweight permits for public highways will be obtained from the applicable states; Arrangements with the owner of a suitable railhead will be made to permit the transfer of the reactor vessel from the transponer to the railcar; i The reactor vessel package, including the tiedown assembly, will be inspected to l ensure it meets the NRC Certificate of Compliance, and applicable US Depanment of Transportation mie:; local govemmer t authorities will be consulted regarding the " time of transport", to minimize the dismption to community activities. A communications plan will be , developed and implemented to ensure such authorities have been notified. m l 2-25 Rev.O
[] SAXTON NUCILUt EXPE3LIAEENTAL CORPORATION v oscOmHSSIONING FLAN The following are the activities which will take place while transponing the reactor vessel: The transponer, loaded with the reactor vessel package, will travel on public highways to a railhead; The reactor vessel package will be transferred from the transporter onto a railcar equipped with a suitable tiedown assembly. t The reactor vessel package, including the tiedown assembly, will be re-inspected to ensure it meets the NRC Certificate of Compliance, and US Depanment of Transportation, 49 CFR 174, " Carriage by rail", rules; (3 - The railcar will be placarded in accordance with the requirements of 49 CFR 172.500; 4 The railcar, loaded with the reactor vessel package, will transit to a railhead local to the disposal location. Using methods similar to the ones used at the departure j railhead, the reactor vessel package will be transferred to a transporter; The transporter, loaded with the reactor vessel package, will travel on public highways to the final disposal location; and The reactor vessel package will be unloaded from the transporter at the final disposal location using a portable crane. 2.2.2 Schedule v 2-26 Rev.0
SAXTON NUCIEut EXM!3tthE!NTAL CORPOItATMN DECOMAHSSIONING PLAN The schedule for the decommissioning of the Saxton facility is presented in Figure 2.2-4. The schedule divides the decommissioning activities into four distinct phases as discussed in Section 2.2.1 i The schedule for the decommissioning of the Saxton facility follows the concepts and general sequence presented in Atomic Industrial Forum / National Environmental Studies Project repon ~ AIF/NESP-036, " Guidelines for Producing Commercial Nuclear Power Plant Decommissioning j Cost Estimates" (Reference 6). The schedule identifies the key activities, sequences these key 4 activities, and lists the schedule duration for each activity. The schedule is fully integrated with the updated cost estimate presented in Section 5.0. I The following assumptions were made in development of the schedule for the Saxton facility.
- 1. All work is performed during a 10 hour workday,4 days per week with no overtime.
! 2. Where appmpriate, the person-hour estimates reflect adjustments for working in a radiological controlled environment. i
- 3. Multiple crews work parallel activities to the maximum extent possible.
4 i The overall ' duration of scheduled activities for all four phases of the Saxton facility decommissioning pmject is projected to cover a period of three and one half years. l 1 O\ 1 ' Q 2-27 Rev.0
- a. _ - 4 - A =_ m 4._a- # ___ mA...t .e,m -
J1 - A,J3e SAXTON NUC12AR EXPERIMDrrAL CORPORATION 1 d DECOMMISSIONING FLAN
)
TABLE 2.2-1 ! t Occupational Exposure for Immediate Dismantlement j Task Person 1 Re Asbestos Remediation 4.9 i System Dismantlement 17.7 l Reactor Vessel and Steam Generator 6.3 4
- Removal i
l I Structure Decontamination and 0.35 Dismantlement ) Waste Management 2.5 Total 31.8 ? 1 i O 2-28 Rev.0
% (~N
{V (. 1 SAXTON NUCLEAR EXPERIMENTAL FACILITY DECOMMISSIOMNG PLAN TABLE 2.2-2 Summary of Applications for Metal Cutting TECHNIQUE _ ." --a
,m<, T . , , , ",,,
~
r" REN NIBBLERS and SHEARS % 2 in Possible L L Y Small diameter piping MECHANICAL SAWS - s24 in Can be adapted for bar, M L Y Slower than other (Hacksaw, Reciprocating, angle or channel methods Guillotine) CIRCULAR CUTTERS 53 26 in Net applicable M L Y Contammation control by vacuuming chips DIAMOND WIRE Unlimited All Not applicable L M Y Access to both sides of diameters the material is necessary heat exchanger tube bundles PLASMA ARC CUTTING 67 All Any shape, if cutting is H H Y Required relief space made to follow shape of behind workpiece. component OXYGEN BURNER 6 48 All Any shape H L Y Radioactive fumes (1) L= Low, M = Medium, H=High 2-29 Rev.0
Figuer 2.21
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.. -- . - . _~ . . - - . . _ - _ . - - . - . - - - - - . _ . . - - _ . - - . - - . - - . - - . . - - . _ - . . . - . - - - . . . - . - . . y _ _ _ _ _ _ _ _ _ _ _ _ _ SAXTON DECOMMISSIONING PROJECT SCHEDULE Figure 2.2-4 Desenption Duratm ' - - ~ ~ ~~ -199 ( - F 1998 1999 - 1 Saxton CV Decorrmssioneng 693 - [ 61~~[ -o T-~63 .1 . o4 .at_ . I c2 1997-o3T64 1 ~ oi 1 o2 T os I~ o4 l oiT~o2_]W - 1.1 Preparahon Phase 193 7 1.1.1 GPU Nuclear StaM 193 - T _ NRO Review hm-=9 Plan 140 1.2 Operahons Phase 325 1.21 GPU Nuclear StaM MS 122 Operations Actmties 325 t221 Start up Time 18 M 1.22.2 Asbestos Removal 3 M 12.2.3 Lower Levet (SW, SE. & NW Quadrant) 3 M 1224 Meddle Level (SW Quadrant) 69 W 1.225 Middle Level (SE Quadrant) 79 W 1.2.26 Operahng Ficor 23 M W 1.2.2.7 Reactor Cwiwitia..; and Storage Well 5 1.2.28 CV Structures (Concrete and Steet) 68 W
- ~
1.229 RPV and SG Removal 32 M 13 Surw.y Phase 66 M 1.31 GPU Nuclear Stan 66 W 1.32 Final Survey 66 M 1.4 Site Restoration Phase 109 1.41 GPU Nuclear Stan 109 1.42 Restoration ActM!es 109 142.1 NRC Review 52 W 1.42.2 Restoration 57 W i _ ._ _ . _ _
Qj SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN 2.3 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES Overall control and responsibility for conducting all activities safely and effectively rests with the GPU Nuclear President. Reponing directly to the GPU Nuclear President is the Vice i President Nuclear Services Division, the Vice President Technical Functions, the Director i Nuclear Safety Assessment and other support departments as required. The organization is shown in Figures 2.3-1. During the decommissioning operations phase, additional staff will be added to the existing the Saxton facility organization to manage the various dismantlement and decommissioning activities. The organization for this phase is shown in Figure 2,3-2. 2.3.1 Vice President Nuclear Services Division
,n The Vice President Nuclear Services Division assures that all division and corponte activities are perfonned in accordance with corporate policies, applicable laws, regulations, and the Saxton j facility Technical Specifications. Reporting directly to the Vice President Nuclear Services Division is the Program Director SNEC Facility who is responsible for all Saxton facility ]
activities. l 2.3.2 Progmm Director SNEC Facility l The Program Director SNEC Facility is responsible for administration of all Saxton facility functions and for assuring that the requirements of the License and Technical Specifications are implemented. During decommissioning, the Program Director SNEC Facility is responsible for 1 the safe and effective conduct of all decommissioning operations and activities. Reporting l directly to the Program Director SNEC Facility is the SNEC Facility Site Supervisor. The l Program Director SNEC Facility also is the main liaison for the SNEC Project Engineer for i 1 coordinating decommissioning technical support. 4
\ 2-30 Rev.O
/ SAITUN NUCl2AR EXPE3tthENTAL COItFOItA110N
( seconeUSSIONING FLAN 2.3.2.1 SNEC Facility Site Supervisor
'Ihe SNEC Facility Site Supervisor reports directly to the Program Director SNEC Facility and
' shall meet or exceed the qualifications of ANSI N18.1-1971, paragraph 4.3.2. This position provides on-site management and continuing oversight of production activities. The position supervises the day-to-day performance of decommissioning activities by company and contractor employees and ensures the activities are accomplished in a radiologically and industrially safe manner. The position also recommends changes to procedures and schedules to improve the safety and efficiency of work carried out at the Saxton facility. 2.3.3 Saxton Radiation Safety Officer (RSO) The Saxton RSO is responsible for the conduct and oversight of all Saxton facility Radiation 1 1 Safety activities through implementation of the Saxton Radiation Protection Plan. The Saxton RSO will meet or exceed the qualifications of ANSI N18.1-1971, paragmph 4.3.2. All ! Radiological Controls personnel report to the Saxton RSO. The Saxton RSO or a qualified designee shall be present on-site whenever decommissioning activities requiring radiological , controls are in progress. The Saxton RSO and through the RSO all Radiological Contmis personnel have stop work authority in matters relating to or impacting radiation safety. 2.3.3.1 Group Radiological Controls Supervisor (GRCS) Bach Group Radiological Controls Supervisor (GRCS) shall meet or exceed the qualifications of ANSI N18.1-1971, paragraph 4.3.2 or shall be formally qualified through an NRC approved l Radiological Controls Training Program. The GRCS reports to the Saxton RSO, directly , I supervises radiation safety activities at the Saxton facility and is responsible for ensuring compliance with radiation work procedures. ( 2-31 Rev.0 i
- _ _ - _ _ - _ _ = _ _ _ .
4 h} (/ SAXMN NUCIRAR EXPE3tIMlpfrAL CORPORATION DECOMMISSIONING PLAN 2.3.3.2 Radiological Controls Technician 4 Each Radiological Controls Technician shall meet or exceed the qualifications of ANSI N18.1-1971, paragraph 4.5.2. Radiological Controls Technicians will monitor decommissioning activities, measure and record on-the-job radiation exposure information, and operate the Saxton Site Laboratory Facilities, including sampling and analysis. 2.3.4 Radiation Safety Committee I The Radiation Safety Committee mports to the Vice President Nuclear Services Division. The committee consists of at least four members appointed by the Vice President Nuclear Services Division. Three members shall constitute a quorum. The committee is responsible for myiewing all matters with radiological safety implications mlating to activities at the Saxton facility. Meetings shall be held as required by Technical Specifications to review and discuss the events of the preceding period. The committee will review License and Technical Specification changes, maintenance actions, decommissioning operations, audits, and NRC Inspection Reports and corrective actions for deficiencies identified. Written minutes of all meetings shall be prepared and distributed to the Vice President Nuclear Services Division within 30 days of the meeting date. 2.3.5 Other Sunoort Technical Support and Quality Assurance oversight for the Saxton facility project is provided by GPU Nuclear Corporation. O 2-32 Rev.O
l s SAX 1DN NUCLEAR EXPERBEENTAL CORFORATION (/ DECOMMISSIONING FLAN l i 2.3.5.1 Technical Support l GPU Nuclear has assigned a full time Project Engineer to support the Saxton facility decommissioning. The Project Engineer maintains a liaison with the Program Director SNEC Pacility and the SNEC Facility Site Supervisor in order to assure appropriate resources are supplied to the Saxton facility decommissioning. These resources include but are not limited to Engineering, Licensing, Industrial Safety, and Environmental Controls. Individuals in each of these organizations have been designated as having accountabilities for support of the Saxton ; facility decommissioning. 2.3.5.2 Quality Assurance Oversight Section 7.0 of this Decommissioning Plan describes the functional Quality Assurance N responsibilities of the Saxton facility organization. In addition, the GPU Nuclear, Nuclear Safety Assessment Department provides the Saxton facility audit function and is independent of Saxton facility management. Section 7.2 further describes the Nuclear Safety Assessment ! Department role at the Saxton facility. i ! 2.3.5.3 Communications GPU Nuclear has developed a Communications Plan to increase the public's understanding and awareness of the Saxton facility decommissioning project. GPU Nuclear will conduct an on-going communications progmm with the community around the Saxton facility tiuuughout the decommissioning project. This program will be designed to keep the public informed of the activities and events associated with the decommissioning work. A U 2-33 Rev.0
i l l
> SAXTON NUCMAR EXPERIMENTAL CORPORATION
- DECOMMISSIONING PLAN The program will be the responsibility of the GPU Nuclear Communications Division. Key elements of the communications program may include conducting open houses and tours of the plant, issuing a community update newsletter, providing speakers for local civic and community meetings, and working with a citizens task force.
i l The Citizens Task Force (CTF) has been established to ensure there is a vehicle for hearing and addressing public concerns about decommissioning the Saxton Nuclear Experimental facility and to ensure public involvement, open communications and education on decommissioning issues. l l l An independent Inspector has been retained to provide the CTF and the Bedford County j Commissioners with regular reports on inspection activities and serve as a technical consultant to the CTF. b 2.3.6 Decommissionine Oreanization l i During the decommissioning operations phase, additional staff will be added to the existing Saxton facility organization to manage the various activities. The organization before this phase begins is depicted in Figure 2.3-1, the operations phase organization is depicted in Figure 2.3-2. Key additional positions staffed during the operations phase are discussed below. 2.3.6.1 D & D Supervisor The D & D Supervisor reports to the Site Supervisor and is responsible for the dismantling and decommissioning activities. The D & D Work Supervisory Staff reports to the D & D Supervisor. This group is responsible for the perfonnance of the work packages for dismantlement of systems and structures. O V 2-34 Rev.O
4 h SAXTON NUCMAR EXPERIMENTAL CORPORATION O oscomummmma nun 4 i 2.3.6.2 Waste Supervisor 1 2 The Waste Supervisor reports to the Site Supervisor and is responsible for general maintenance activities and decontamination support of decommissioning. The Radwaste Supervisory Staff , reports to the Waste Supervisor. This group is responsible for decontamination activities, radwaste packaging, manifests, and radwaste shipping. 2 I k
- J T l l
i i l l J 2-35 Rev.0 s 1
s s _ v ) Figure 2.3-1 SAXTON ORGANIZATION GPU Nuclear President
, _____________p I
, i E
I
' GPU Nuclear Radiation Safety Other GPUN e Vice President Nuclear Services Committee " '##
- i Division - -l. Division Support Assessment 1
I l ' Communications I l Radiation Safety ---
---- > Program Director SNEC Facility Officer Engineering SNEC Facility Financial Group Radiological Controls Site Supervisor Supervisor (as needed)
Information Management Radiological Controls Other Nuclear Services Legal l Technicians (as needed) Division Departments l l . Key: Direct Reporting Relationship
--------- Indirect Reporting Relationship
~~ x J \
G Figure 2.3-2 SAXTON ORGANIZATION - OPERATIONAL PIIASE GPU Nucicar President
,_____________h I i
l GPU Nuclear Radiation Sarcty Other GPU Nuclear
- , --, guege,,. Safety Vice President Nuclear Services Committee i Division Support
, Division . bess ment s Communications l
l Radiation Safety ---
---- k Program Director SNEC Facility M' Officer En gineering Radiolegical SNEC Facility Financist Engineer Site Supervisor Information Management Group Hadiological Controls Supervisor (s) l Other Nuclear Services IR"I Division Departments Radiological Controls Technician s i
l D & D Super.isor Clerical Security Sched1 Planner / Buyer Waste Supeniser Key: Direct Reporting Relationship
--------- Indirect Reporting RcIntionship
O SAXTON NUCLEAR EXPERIMENTAL CORIORATION DECOMMISSIONING PLAN 2.4 TRAINING PROGRAMS GPU Nuclear will maintain a training program for all personnel performing work functions at the Saxton facility. The program shall be comparable to that presented for personnel working at other GPU Nuclear facilities (Oyster Creek or TMI). However, training shall be specific and relevant to the Saxton site area, existing facilities and work site conditions. All decommissioning employees, i.e., contractor, GPU Nuclear or other, shall receive appropriate training commensurate with the potential hazards to which they may be exposed. The Saxton Radiation Safety Officer (RSO) shall identify the required radiological and safety training for site personnel. All training records shall be maintained for a period of time as required by procedure. Records
\ shall include the decommissioning employee names, training date, type of tmining received, protective equipment authorizations, expiration dates, instructor name(s) and certifications (if
. applicable).
Periodic requalification and continuing training for individuals assigned on a long-tenn basis will be provided as required by procedures. . General Employee Training i
- General Employee Training (GET), shall be provided to all decommissioning employees assigned to conduct unesconed work at the Saxton site. GET training shall be completed in accordance with procedures and shall comply with appropriate Institute of Nuclear Power Operations (INPO) recommendations, Saxton facility GET training shall implement
- t 2-36 Rev.O
O SAXTON NUCLEAR EXPERIMENTAL CORPORATE)N
\ DECOMMISSIONING FLAN
- ANSI N18.1-1971, Selection and Training of Nuclear Power Plant Personnel, section 5.4.
- Radiological controls and radiation protection (applicable portions), from: 10 CFR 19
" Notices, Instructions, and Reports to Workers. 10 CFR 20 " Standards for Protection . Against Radiation."
- Respiratory protection training (as appropriate) from: Regulatory Guide 8.15, October
- 1976, " Acceptable Programs for Respiratory Protection" and NUREG 0041,
" Respiratory Protection for Airborne Radioactive Materials."
l
- Prenatal exposure training from Regulatory Guide 8.13, Rev 1, "Instniction Concerning Prenatal Radiation Exposure."
l , 4 GET training shall also include but not be limited to:
- Plant Security i
- Fire Protection
!
- Quality Assurance
- Industrial Safety
- Fitness for Duty Program and Dnig and Alcohol Policy
- Health and safety infonnation (Hazardous Materials including asbestos, etc.).
f A written exam shall be administered demonstrating successful completion of all GET training pmgrams. In addition, those persons designated as radiation workers shall receive training in the f allowing
,Og
(,' 1 2-37 Rev.O
l SAXTON NUCLEAR EXPERIMENTAL CORFORATION Q,O DECOMMISSIONING PLAN areas as pan of General Employee Training: l
- fundamentals of radiation i
- radiation and contamination measurements and control
- ALARA
- radioactive waste minimization
- Radiation Work Permits j i This course includes practical activities training which includes the following elements:
- protective clothing dressout l e dosimetry
- entering / exiting contaminated areas j
- preparation / implementation of RWPs Q)
< The Training Depanment shall obtain the concurrence of the Radiation Safety Officer when setting training objectives for radiation protection portions of the GET training program. i 4 Job Soecific Training l f I Radiological Controls personnel qualifications shall be as per ANSI 18.1-1971 as described in l Section 2.3.3.2. I 1 4 q Decommissioning personnel shall be provided training or show competency v/ hen necessary, for ) tasks outlined below as required. ,
- lifting and rigging practices
- decontamination techniques a
- use of power / hand tools d 2-38 Rev.O t
. .. .- .- - . _ - - - . - -- _ . - . . - . . ~ . .- - - _- - . - . _ _ - - . - ._.
SAXTON NUCLEAR EXPERIMENTAL CORM) RATION DFf0MMISSIONING PLAN
- crane / heavy equipment operation
- implementation of engineering controls for personnel protection f
- implementation of procedures and work instructions i
! The decontamination and decommissioning work scope shall dictate the level of training necessary to accomplish a specific task or project goal. The scope of the work at the Saxton facility is comprehensive and will require training or knowledge in dismantlement and j decontamination activities. At the discretion of the Program Director SNEC Facility previous experience in a job skill or work evolution may be considered equivalent to comprehensive training.
\
l 2-39 Rev.O
E
/ SAXTON NUCLEAR EXPERIMENTAL CORPORATION
( DECOMMISSIONING PLAN . 2.5 CONTRACTOR ASSISTANCE GPU Nuclear intends to manage and oversee all phases of decommissioning work using - dedicated licensee employees. The use of contractors will be restricted to that needed to augment GPU Nuclear staff or for specialized services which are routine!y employed for such tasks by the industry. 2.5.1 Contractor Scone of Work Tasks where specialty contractors may be utilized in support of decommissioning activities include, but are not limited to the following: T a. Packaging, transportation and disposal of radioactive material and mixed wastes,
- b. Radiological controls staff augmentation.
- c. Contaminated laundry service.
- d. Site security.
3
- e. Building fabrication / trailer rentals.
.: f. Administrative / clerical tasks.
- g. Safety and health / industrial hygiene.
- h. Specialty engineering and design services such as heavy loads management and transportation. ;
- i. Decontamination / volume reduction of contaminated components and water processing.
- j. Dismantlement and demolition of stmetures, I
- k. Excavation.
- 1. Specialized radioanalytical laboratory services.
A list of potential contractors, with corresponding scope of work, is provided in Table 2.5-1. A . (V
)
2-40 Rev. 0 4 i
t f [ 3 SAXTON NtJCLEAR EXPERIMEPTTAL CORPORATM)N DECOMMISSN)NING PLAN 2.5.2 Contractor Qualifications and Experience Each contractor will be evaluated to ensure they meet appropriate qualifications for the tasks that they will perfonn. Contnictors will be evaluated based upon the following criteria: technical and operational capability, cost and schedule compliance, demonstrated experience in providing quality services on similar projects and ability to meet regulatory requirements. l 2.5.3 Contractor Administrative Controls GPU Nuclear management will retain overall responsibility for the perfonnance of all i contractors and provide necessary management oversight to assure that the tasks perfonned by > 4 the contractors are in full compliance with the Quality Assurance Program, the purchase agreement, all GPU Nuclear Corp. radiation protection procedures and standards, and applicable j ,/ regulatory requirements. 4 i l' i 1 l l l i 9 2-41 Rev.0
i 'O SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN P TABLE 2.5-1 List of Potential Contractors P Contractor Name Scope of Work Scientific Ecology Group, Inc Radwaste disposal, decontamination and volume 1560 Bear Creek Rd. reduction services, water processing P.O. Box 2530 Oak Ridge, TN 37830 TLG Engineering Decommissioning consultation / planning, engineering 148 New Milford Rd. services liridgewater, CT Bartlett Nuclear, Inc. Radiological controls support , 60 Industrial Park Rd. Plymouth Industrial Park Plymouth. M A 02360 Frank W. Hake Assoc. Heavy loads management and transportation, 1790 Dock Street decontamination and volume reduction services
} P.O. Box 13464 d Memphis, TN 38113 American Ecology Recycle Center, Decontamination and volume reduction services inc.
109 Flint St. Oak Ridge TN 37830 Envirocare of Utah Radioactive waste disposal 46 W. Broadway Suite 240 Salt Lake City, UT 84101 Chem-Nuclear Systems Radioactive waste disposal 140 Stoneridge Drive Columbia, SC 29210 MDS Labs Industrial hygiene lab support 4418 Pottsville Pike Reading, PA 19605 Interstate Nuclear Services Contaminated laundry services 295 Parker Street P.O. Box 51957 Springfield, M A 01151 s 2-42 Rev.O
O SAXTON NUCLEAR EXPERIMENTAL CORPORATKW DECOMMISSIONING PLAN TABLE 2.5-1 (Continued) List of Potential Contractors ll&W Nuclear Environmental Services, Inc. Radioanalytical lab support Lynchburg Technology Center Lynchburg, VA 24506-1165 Raytheon Nuclear Services Craft labor 30 S.17 Street Philadelphia, PA 19101 liryce Saylor & Sons Excavation, rigging, demolition services 6th Street Altoona, PA 16602 i t \ 2-43 Rev.O
[' SATIDN NUCIEAR EXPERIMDrfAL CORPORA'IlON DECOMMISSIONING FLAN SECTION 3.0 PROTECTION OF OCCUPATIONAL AND PUBLIC IIEALTII AND SAFETY 3.1 EACILITY RADIOLOGICAL STATUS 3.1.1 Facility Operating Historv 3.1.1.1 Facility Description The Saxton facility is a deactivated pressurized water reactor (PWR) which was licensed to opente at 23.5 megawatt thermal (23.5 MWT). It is owned by the Saxton Nuclear Experimental Corporation (SNEC) and maintained by GPU Nuclear Corporation (GPUN). The Saxton reactor A facility is maintained under a 10 CFR 50 License and associated Technical Specifications. The license was amended to possess radioactive material but not operate the Saxton facility reactor i in 1972. The license expires on Febmary 11,2000 or upon expiration of the SNEC corporate l charter, whichever occurs first. The facility was built from 1960 to 1962 and operated from 1962 to 1972 primarily as a research and training reactor. The fuel was removed from the Containment Vessel (CV) in 1972 and shipped to the Atomic Energy Commission (AEC - now The Department of Energy) facility at Savannah River, S.C., who retained ownership of the fuel. As a result, neither SNEC or GPU Nuclear have any responsibility relative to the spent fuel fmm the Saxton facility. The control rod blades and the superheated steam test loop were shipped offsite. Following fuel removal, equipment, tanks, and piping located outside the CV were removed. The buildings and stmetures that supported reactor operations were partially decontaminated in 1972 through 1974. The radiological condition of the facility following shutdown was documented in a report titled
" Decommissioned Status of the Saxton Reactor Facility" forwarded to the United States Nuclear 3-1 Rev.O
I I I SAIMN NUCLEAR EIFERIMDrFAL CORPORATION
;a _ ,uN i
l l Regulatory Commission (USNRC) on February 20,1975 (Reference 7). ]
; After the formation of the GPU Nuclear Corporation in 1980, SNEC formed an agreement with l GPU Nuclear to use GPU Nuclear and its resources to maintain, repair, modify, or dismantle SNBC facilities as may be required. Both SNEC and GPU Nuclear am subsidiaries of the same l
parent company, General Public Utilities Corporation, (GPU). l. 1 While SNEC remains the owner of the facility, a license amendment has been submitted to designate GPU Nuclear as a co-licensee. GPU Nuclear will carry out the Saxton facility decommissioning on behalf of the site owner, SNEC.
- 1 l
l Decontamination and removal of reactor support structures and buildings were performed in j ! 1987,1988, and 1989, in preparation for demolition of these structures. 'Diis included the
'p removal and discharge of appmximately 210,000 gallons of gmundwater fmm various structures, the decontamination of the Control and Auxiliary Building, the Radioactive Waste Disposal j
Facility, Yard Pipe Tunnel, and the Filled Dmm Storage Bunker, and the removal of the Refueling Water Storage Tank. A comprehensive final release survey was conducted from f October,1988 to June,1989, to verify that residual comamination was within USNRC guidelines l for unrestricted use. Details of the decontamination activities and final survey are provided in the " Final Release Survey Report of the Reactor Support Buildings" (Reference 8). Upon acceptance of the final release survey by the USNRC, the buildings were demolished in 1992. Details of the demolition are available in a report titled, "SNEC Reactor Support Buildings Demolition Report." (Reference 9) In November 1994, the Saxton facility Soil Remediation Project was completed. This was a comprehensive plan involving soil monitoring, sampling, excavation, packaging and shipment j of contaminated site soil. This program successfully reduced radioactive contamination levels below the NRC current and presently proposed levels required to meet site cleanup criteria. 8 3-2 Rev. 0
I 1
/7 SAXTON NUCIEAR EXPERIMEPrTAL CORPORATKW Q! DECOMMISSMMING FLAN l
Site-specific radiological and environmental data was obtained in 1995 as pan of the Saxton Site Characterization Plan (6575-PLN-4520.%)in order to support the development of the Saxton facility Decommissioning Plan. The scope of the characterization plan extends over treas of the facility that may have become internally or externally contaminated or activated during the facility's operating history. Results of the characterization plan have been used to determine the curmnt radiological status of the facility and are presented in Section 3.1.2. The operational history of plant systems and suppon buildings and facilities is an important component of determining the radiological status of the facility. Table 3.1-1 provides chronology of major operational events at the Saxton facility since the stan of operation in 1962. 3.1.1.2 Site Layout I l
\ The site is located about 100 miles east of Pittsburgh and 90 miles west of Harrisburg in the Allegheny Mountains, three founhs of a mile nonh of the Borough of Saxton in Libeny Township, Bedford County, Pennsylvania. The site is on the nonh side of Pennsylvania Route 913,17 miles south of U.S. Route 22, and about 15 miles nonh of the Breezewood Interchange of the Pennsylvania Turnpike. Figures 3.1-1 through 3.1-4 identify the location of the site relative to the landmarks of the local ama.
The Saxton facility was built on the east side of and adjacent to the Saxton Steam Generating Station (SSGS) of Pennsylvania Electric Company. This station was located on the east bank of the Raystown Branch of the Juniata River as shown on Figure 3.1-3. The propeny comprises approximately 150 acres. The SSGS was demolished in 1975. The Saxton site comprises 1.1 acms which contained the yard and buildings from the nuclear plant. An additional 9.6 acre area is fenced in around electrical switchyard and buildings still 3-3 Rev.O
' O, SAXMN NUCLEAR EXPERIMDfrAL CORPORATION Q MCOMMISSMN PLAN in use by Penelec (Pennsylvania Electric Company, a subsidiary of GPU). This area adjoins the Raystown Branch of the Juniata River. The Saxton site as well, as a portion of the Penelec area
; and the surrounding uncontrolled lands, are in the 100-year floodplain of the Raystown Branch ,
and Shoup Run. Additional large areas of low lying field and scrub tree growth surround the site. 1 3.1.1.3 Station Stmetures The only remaining permanent Saxton facility buildings and structures are the Containment Vessel (CV), the concrete shield wall located around the NW and NE quadrant of the CV, the tunnel sections that are immediately adjacent to the outer circumference of the CV and portions of the septic system, weirs, and associated underground discharge piping. Concrete barrier walls p have been installed to isolate the open ends of the tunnel that were connected to the Control & Auxiliary Building, the Radioactive Waste Disposal Facility, and the Steam Plant. Portions of I the Steam Plant Tunnel still exist downstream of where it is blocked off. This area will also be verified to meet unrestricted release criteria following dismantlement. 3.1.1.3.1 Area Designations The remaining Saxton Containment Vessel facilities (see Figures 3.1-5 through 3.1-10), consist of eight principle stmetures/ locations designated as the following: Area 1 - Southwest, southeast, and northwest quadrants of CV between elevations 765'-8" (concrete floor) and el. 777'-8" (concrete ceiling) in southwest quadrant, el. 779'-0" (steel platform) in southeast quadrant, and el. 775'-2" (concrete ceiling of the " Rod Room") in northwest quadrant. The area also includes the 3'-6" deep sump located in the floor, and a 4 I foot wide concrete ledge (el. 768'-3") extending around the circumference of the area. (^) 3-4 Rev.0
, SAXTON NUCLEAR EXPERIMENTAL CORPORATION \ DECOMMISSIONING FLAN Area 2 - The southwest quadrant of the CV between elevations 779'-8" (Concrete Floor) and 814'-6" (Concrete Ceiling). Three (3) steel platforms are installed. One platform, el. 789'-4", is located beneath the Steam Generator, Pressurizer and Primary Coolant Pump. The second, el. 795'-2", essentially extends over the entire area of the quadrant and the third, el. 807'-0" is located around the upper heads of the Steam Generator and Pressurizer. Area 3 - The Southeast quadrant of the CV between elevations 781'-4" (steel platform) and 810"-0" (concrete ceiling). One additional steel platform is installed at elevation 795'-2". Both steel platforms extend over the entire area of the quadrant, each containing an 8 foot by 8 foot open hatch. There is a similar opening in the concrete ceiling. Area 4 - The concrete operating floor of the CV (el. 812'-0"), and surfaces up to the top of the CV dome. The area includes: 1) three access hatches (equipment, personnel, and escape); 2) V the concrete walls and platform located in the southwest quadrant containing the steam generator, pressurizer and reactor coolant pump; 3) the polar bridge crane; 4) a steel platform (el. 818'- 0"); and 5) the movable bridge over the reactor vessel and fuel storage area. Area 5 - The below grade concrete lined turmel which wraps around the outer circumference of the CV between azimuths 35 degrees and 270 degrees.
&ra_6 - This area is comprised of the Northwest and Northeast quadrants (reactor compartment and storage well respectively) of the CV between elevations 765'-8" (concrete floor) and 807'-0" (concrete ceiling). All concrete surfaces of this area, including the surmunding concrete walls are lined with a Series 300, four-coat, catalyzed phenolic protective lining.
The reactor compartment contains the reactor vessel, associated piping, and the reactor vessel "can", which was used as a barrier to keep the storage well water off the piping and reactor C (' 3-5 Rev.O
O
\
SAXTON NUCLEAR EXPERIMDffAL CORPORA 110N DECOMMLTRONING PLAN vessel surfaces. There are two steel platforms installed in the reactor compartment. One removable platform at elevation 793'-2", and a second platfomi at elevation 800' 6". The storage well contains the movable fuel storage rack (up/down direction), three demineralizer vessels (Storage well, Boric Acid, and Purif'ication), which are installed at elevation 776', and a concrete pad for storage of the fuel shipping cask (above el. 781'-4"). 4 Area 7 - The outside of the steel liner of the CV extending from Grade level (el. 811') to the top of the CV Dome. This area also includes the concrete shield stmeture located around the circumference of the CV between azimuths 270* and 35' (el. 804' to 814'). l Area 8 - This area includes the tanks and enclosures which received the sanitary discharge from the C&A Building, processed the sewage, and discharged the effluent to the Raystown Branch of the Juniata River. The system consisted of: 1) two concrete septic tanks installed in series, , V Tank "A" and Tank "B"; 2) a concrete pump well contained within a concrete block pump house (no above grade structures remaining); and 3) a concrete Chlorine Contact Tank / Weir unit. 3.1.2 Current Radiological Status of the Saxton facility 4 Site-specific radiological and environmental data was obtained in 1995 as part of the Saxton Site Characterization Plan (6575-PLN-4520.06)in order to support the development of the Saxton facility Decommissioning Plan. The scope of the Plan extends over areas of the facility that may have become internally or externally contaminated or activated during the facility's operating history. The data obtained in this program will be used to determine effective and appmpriate decontamination and dismantling techniques and activity sequencing to support decommissioning. This data was also used for planning radioactive material disposal, assessing potential hazards O N 3-6 Rev.0
SAXTON NUC12AR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN during decommissioning and decontamination work, determining ALARA controls, and accurately scheduling and estimating the cost of the overall pmgmm. 4 Radiological samples and information acquired during the Characterization Plan included locations, areas, and activity levels of structural surfaces, depth and activity levels of contaminant penetration levels into porous or cracked surfaces, location, volume, and activity levels of contaminated soil, location, surface areas and volumes, and activity levels in piping systems and contaminated equipment, calculation and confirmation of activity levels induced by activation in the reactor vessel and associated components, waste classification of contaminated materials, and general area and hot spot radiation levels. Environmental characterization carried out under the Characterization Plan has determined the A radiological characteristics of potential contamination in the soil on the site. In addition, the characterization has determined the location and type of asbestos which was used as thermal system insulation, a Total Metals Analysis to determine the presence and concentration ratio of lead, chmmium, and cadmium metal contaminants in painted surfaces, and a complete Toxicity Characteristic Leaching Procedure (TCLP), inorganic analysis on all sediment and sludge samples. These survey and sampling programs are presented in detail within the Saxton Site Characterization Plan. The following sections present a summary of the results of the Site Characterization Plan program. The radiological status of the following areas is derived from the best available information. The Saxton Site Characterization Report will be updated as deemed appropriate. G 3-7 Rev.O
SAXTON NUCMAR EXPERIMENTAL C05tPORATION V DECOMMISSIONING FLAN 3.1.2.1 Area Dose Rates and Contamination Levels Tables 3.1-2 thmugh 3.1-9 present the Saxton facility Area by Area radiological evaluations for surface contamination on components and stmetures; average general area (GA) exposure rate values, and contact measurements using a thin window G-M detector and other standard , measurement equipment. These are average external beta / gamma contamination levels for 2 locations and components within the Saxton facility and are listed as DPM/100 cm for smears and Net CPM (background subtracted average) for direct contamination measurements. Component average contact mR/hr readings are reported when available. The average general area (GA), exposure rate measurement results are presented for all Saxton facility Area designations in each Area title heading. The direct frisk CPM measurements were made using an Eberline B-140N with HP-210T probe.
.O Tables 3.1-10 through 3.1-14 present typical Area by Area alpha surface contamination smear results (in DPM/100 cm 2), for Saxton facility Containment Vessel structural components (Areas 1, 2, 3, 4 & 6). l l
Regulatory Guide 1.86 limits were used to determine radionuclide surface contamir.ation limits. It is the intent to replace Regulatory Guide 1.86 limits with more recently proposed criteria as they are promulgated. These newer limits will most likely be based on Dmft NUREG-1500 (Reference 19) , NUREG/CR-5512 (Reference 18) and other similar guidance. l 3.1.2.2 Concrete Penetration and Activation Fifty-six (56) concrete core bore samples were collected from around and within the Saxton facility Containment Vessel concrete structure. Eleven additional core bore samples weni taken from unaffectal areas. Core bore samples were approximately 7 centimeters (2.74lin.) in x.J 3-8 Rev.0
/~'g SAX 1DN NUCIEAR EXPERLIMENTAL CORFOItATION Q DECOMMISSIONING FLAN diameter and ran;;ed from appmximately eight centimeters to 152 centimeters (3 in. to 60 in.)
in length. All core bore samples were sliced into 1.3 cm (1/2 inch) sections and analyzed with a HpGe Gamma Spectroscopy detector system. Fmm the counting results, a determination of the depth of penetration of the gamma emitting radionuclides was made. Background core bore samples fmm unaffected locations were taken fmm concrete pours of the same time period as those concrete pours used at the Saxton facility during construction. Selected core bore samples were sent to an off-site analytical laboratory for analysis. The results of this distinctive analysis are presented in the Saxton Site Characterization Report. To determine the release criteria for contaminated concrete, typical hard-to-detect nuclides were scaled in from surface contamination composite smear analysis results providM in Table 3.1-15.
'Ihe resulting concentations were input into the RESRAD computer code (Reference 10).
Concrete contaminated above the unrestricted release criteria defined by RESRAD output is d assumed to require remediation. The volume of concrete to be removed from the Saxton facility is presented in Table 3.1-18. In addition to the actual counting results, neutron activation analysis of the Saxton facility reactor vessel and surrounding support stmetures has also indicated that significant concrete removal will be required. Table 3.1-18 presents the maximum determined depths of affected concrete at the Saxton facility. This table is an Area by Area presentation showing the total concrete volume to be removed by location. These preliminary results indicate that approximately 10,841 cubic feet should be removed. This determincd volume is thought to be conservative and will be adjusted when the results of the off-site concrete sample analysis results become available. m 3-9 Rev.O _____J
l l l [) SAITON NUCLEut EIFEltinEDfrAL COItFORATION V DECOMMISSIONING FLAN l 3.1.2.3 Radionuclide Identification Smears from Areas 1,2,3,4, & 6 were collected and composited by Area. Each group of composited smears were sent to an off-site analytical facility for a complete 10 CFR 61 type analysis. The results of these analysis are shown in Table 3.1-15. Areas 3 and 4 were combined because of previously verified similarities in nuclide composition. Note that "less Than" values are not included in the Table. 3.1.2.4 , Internal Contamination Systems samples were collected by removing sediment, scrapings or cutting out sections of piping systems. The irsults of the off-site analysis of these samples are reported in detail within the Saxton Site Characterization Report. The following information is provide as a summary. Based on off-site sample analysis results, some systems samples have demonstrated high TRU (transuranic) content (as a percentage of total sample activity.) As an example, TRU content l in samples from the purification system are about 86% of that sample's total activity (in pCi/g), and is largely due to the Pu-241 content. System results with signifi-ently different nuclide fractions than those present in surface contamination activities (composite smear results), are shown below: SYSTEM PRINCIPAL NUCIlDES
- 1) Purification System TRU, Co-60
- 2) Pressurizer Spray Line TRU, Ni-63, Co-60
- 3) Shutdown Cooling System TRU, Ni-63 i 4) Parssurizer Surge Line Ni-63, TRU
- - 3-10 Rev.O
SAITON NUC12AR EXPEIURIENTAL CORMMLATION DECOhSGSSIONING FLAN 1
- 5) Component Cooling System TRU, Ni-63
- 6) Stomge Well Piping TRU, Cs-137
- 7) Vents & Drains System Ni-63, Cs-137
- 8) Septic Tank B Ni-63, Cs-137, Pu-238
- 9) Charging and Volume Control System TRU, Cs-137 While the external exposure rate values for the system piping and most components are relatively low (a few Mr/hr - contact), the TRU content within some systems will require the use of containments and engineering controls during the operations / dismantlement phase. Some examples of such controls include glove bags and other contamination enclosures, local HEPA ventiiation, fixative agents, etc.
3.1.2.5 Reactor Pressum Vessel and Internal Activation A neutrun activation analysis of the Saxton facility reactor pressure vessel, internals, surrounding lead, structural steel and concrete was performed by TLG Services of Bridgewater, Cormecticut. The analysis scenario considered was a final shutdown of the facility on April 2nd,1972, after three (3) cycles of both power and experimental operation, with an ensuing decay period of 24 years to July 1st,1996. The analysis indicates that approximately 1452 curies of neutron activation products, mainly Fe-55 (37 Ci), Co-60 (595 Ci) and Ni-63 (811 Ci), will be present in the reactor vessel wall and
- clad, internals, insulation can and support can 24 years after shutdown (July 1st,1996). This does not include the activity from internal and external surface contamination, which is normally a small percentage of the total activity for these components. Table 3.1-16 presents the curie determination for the Saxton facility reactor vessel and internals.
4 l 3-11 Rev.O ! i 1
l ! L SATIDh NUCIEut EIFI!3URENTAL CORFOILATION DECOMMISSIONING FLAN The completed Saxton facility reactor activation analysis is pmsented as pan of the Saxton Site Characterization Repon, (Reference !1). , i 3.1.2.6 Soils and Environmental Results , In November 1993, extensive soil characterization was performed at the Saxton site to determine the contaminant activity levels in the soil. Subsequent analyses were performed during soil remediation work between July and October 1994. Survey procedures and methods to assess i soil radioactivity and hazardous material content were based upon NUREG/CR-5849 and 40 CFR 261. Soil characterization and disposal work am described respectively in procedures 6575-PLN-4742.05, "SNBC Soil Characterization Plan" and 6575-PLN-4542.07, "SNEC Soil Disposal Plan". Results of all the analytical data are specifically described in the "1994 Saxton Soil Remediation Project Repon" dated May 11,1995.
- On average, as a result of the soil mmediation work, the predominant nuclides remaining in the soil are Cs-137 and Co-60 at concentrations less than 1.0 pCi/g and 0.1 pCi/g respectively.
i Although 56,266 ft), containing 11 millicuries, of the soil was disposed at Envirocare and Barnwell, some residual Cs-137 still remains in areas described in the soil remediation repon. Appendix 1 and Figure 3 of the Soil Remediation Report (Reference 12) provide the pre and post Cs-137 concentrations in the soil. Air, sediment and ground water sampling have been and continue to be conducted on the site. 1 Hazardous material analyses were performed by a qualified laboratory using U.S. EPA i guidelines on soil and the soil was found to be nonhazardous. Although most of the affected soil has been removed from the site these analyses are also mpresentative of the soil remairdng. 3-12 Rev.0
~w--m m n wa- ,+wn
SAX 10N NUCLEAR EXPERBEENTAL CORPORATION DELtNSEISSIONING FLAN The total number of samples obtained and the type of analyses that were reported are shown in Table 3.1-17. An Environmental and Hazardous Materials Characterization was performed for the Saxton Nuclear Experirnental facility, as part of the Characterization Plan. Relevant points from this study and the ongoing environmental monitoring program are presented below:
- 1) Ten environmental monitoring well stations plus two potable ground water stations are sampled at least quarterly. One station has shown positive tritium activity. Periodically, low levels of tritium (200-600 pCi/1) have been detected in one of the environmeatal ground water monitoring wells, SX-GW-GEO-5, refer to figure 3.1-11 for well locations.
Buried beneath this area is the RWDS (Rad Waste Disposal System) tunnel. It is likely that this positive result is due to residual tritium from activity in this area arising from plant d operation. Nevertheless, tritium concentrations from this station are well below the effluent concentration specified in the USEPA Primary Drinking Water Standard of 20,000 pCi/1. Gamma scans from this station, as well as all other environmental well stations, have always resulted in less than detectable limits.
- 2) low levels of Cs-137, and at times Co-60, have been detected in aquatic sedirnent samples collected near the Saxton facility at the discharge to the river (SX-SD-Al-4), refer to figure 3.1-12 for sample locations. Cs-137 was also detected in the control station (SX-SD-Ql-4) which is collected upstream from the discharge. These results peaked following the outbuilding demolition when site soil washed to the river from the site storm drain discharge. Following the outbuilding demolition project the storm drain discharge line was plugged and the soil remediation project was completed. Since that time, radionuclide concentrations in aquatic sediment samples have fallen. In addition to Co-60 and Cs-137, KSD, Ra-226, and Th-232 have been detected, however these are naturally occurring 3-13 Rev.O
l i i ifj \ SAXTON NUCIRAR EIFERinGDfTAL CORPORATION DECOMMISSIONING PLAN isotopes and present in all sediment.
- 3) Environmental air particulate sampling stations are put into service whenever any site operation is being performed. The particulate filter is analyzed for gross alpha, gross beta,
; and gamma and only the positive activities detected are listed in this report. Three J
indicating air stations are located amund the site and one control station is located 10 miles . from the site. Generally, the weekly trends of gross alpha and gmss beta activity at all ] stations are similar. Gamma-emitting radionuclides related to the Saxton facility were not , detected on any of the samples. As expected, all of the samples contained naturally-occurring beryllium-7 (Be-7). Concentrations detected are similar on both indicator and 1 control filters. l i The results of the gmund water, aquatic sediment, and air particulate samples, tabulated in this report verify that the doses obtained were well below the applicable dose limits, and concludes that the Saxton facility has not had any adverse effect on the envimament or on the health of the public. i 3.1.2.7 Hazardous Materials 4 i Since the Saxton facility was built in the early 1960's, it is presumed that a majority, if not all, insulation materials used during the construction contain some percentage of asbestos. This same assumption has been made with respect to lead in paint, in that all painted surfaces will be presumed to contain some of lead within the paint materials. I A comprehensive asbestos sampling pmgram was conducted as part of the site characterization. .j I Samples collected were analyzed on-site by polarized light microscopy (PIM). The analysis was
- conducted by a labomtory technician fmm the GPU Nuclear System laboratory in Reading, i
3-14 Rev.0
,~. - . _ . _ . - . , - , ~ . _ _ . _ _ . . . _ . _ _ - . . _ _ - , -
[
; \
SAXTON NUCutAIL EIFERDWINFAL COItPOItATION peCOMneBSIONING FLAN Pennsylvania. The System bboratory is certified by the National Voluntary bboratories l Accmditation Program (NVLAP) for bulk insulation identification of asbestos. i A total of 61 bulk insulation samples were collected, with an additional nine samples taken at - an earlier time. Based upon the results of the samples taken from various locations within the Saxton facility Containment Vessel, it appears that the only insulation that does not contain any asbestos is on the straight pipe runs for the HVAC system. 'Ihese straight pipe runs contam j fiberglass. However, elbows, valves and tee connections for the HVAC system consist of trowled-on insulation containing asbestos. All other bulk insulation ramples contain either chrysotile or amosite asbestos. Any piping system that needs to have the insulation removed, and has not been sampled, will be " presumed C asbestos containing material" (PACM), and treated accordingly until determined otherwise.
- During the asbestos sampling period, paint samples wem collected for lead analysis. Most of the paint samples taken were radioactive and could not be sent to a routine lead analysis laboratory. The one sample that could be analyzed routinely had a lead content of about 1.7%.
This sample was collected from the paint covering the containment building external steel shell. l Two additional samples of paint from the internal surfaces of the containment building were collected. Each of these samples were composite samples from many surfaces throughout the facility. Analysis results from these composited samples are reported within the Saxton Site Characterization Report. i O 3-15 Rev.0 I
i l t l SAXTON NUCIEUt EXPERIMENTAL CORPORATION DECOMMISSIONING FLAN TABLE 3.1-1 Facility Operational History EYcal DAlt Construction authorization February 11,1960 Initial License issued November 15, 1961 Initial criticality April 12,1962 First Electricity Generated November 16,1962
\
Spill of approximately one gallon of November 26,1968 i radioactive water (less than 10 microcuries) I from door of safety injection system pump l house. Storage well leaks possibly 1%8 through layup resulting in extensive contamination of storage well of the internal concrete structures (n\ Experiments with mixed oxides fuel, with fuel cladding intentionally Last fuel cycle December 11, 1969 - l
" failed" resulting in extensive May 1,1972 contamination of piping connected with the main coolant system Unplanned gas release of 7.32 curies May 14,1970 due to rupture of diaphragm in gas compressor regulating valve.
Unplanned gas release of 0.034 curies from August 26,1970
#3 gas decay tank. 1 Unplanned gas release of 80.2 curies from November 29,1971 purification system into charging pump room of the Control & Auxiliary Buildings (previously dismantled and released).
l l l O LJ 3-16 Rev.0 1
SAXTON NUCIEAR EXPERIMDEAL CORPORATION (/ DECOMMISSIONING PLAN TABLE 3.11 Facility Operational IIistory (Continued) Event Date Unplanned gas release of 19.7 curies December 15, 1971 comes from a packing leak of PC-97V, pressurizer vent line. Final shutdown May 1,1972 Nuclear fuel and other removable July-November,1972 "special nuclear materials" shipped off site. By-product material removed from site November,1972 - (with exception of material in Early 1974 exclusion areas) p Facility is in a " mothball" condition February 1975 Groundwater removed from RWDF and Late 1986 - Yard Pipe Tunnel Early 1987 l l l Decontamination of C&A building 1987 & 1988 l RWDF, RWST, and yard pipe tunnel Pennsylvania State University Soil December 1988 & Characterization January 1989 Final release survey of C&A building, October 1988 - RWDF, RWST, and yard pipe tunnel June 1989 EG&G Measurements aerial survey July 1989 Comprehensive Radiological Survey 1991 CV (Scoping Survey) Demolition of C&A building, RWDF, 1992 RWST and yard pipe tunnel Soil Remediation Project June -November 1994 Saxton Site Characterization Study May 1995 3-17 Rev.O
SAX 1DN NUCIEAR EXPERIMENTAL CORPORATION
; DECOMMISSIONING PLAN Table 3.1-2 AREA 1
(~5 mR/hr General Area) Components & Locations Mean Smear Range Smear
- Contact Direct DPM/100cm2 DPM/100cm2 mR/br Frisk CPM 765' 8* El. Concrete Floor ~ 33,100 ~2,100 to -1.2E5 NTT - 24,000 768' 3' El. Concrete Ledge - 15,000 -2,300 to -4.5E4 N/T - 10,000 765' 2" El. Sump - 1.07E5 -5.0E4 to -1.6E5 2.5 Inside N/A 775' 2* El. Con. Ceilmg - 1,100 < 1,000 to ~2,000 Nrr ~770 Concrete Walls - 6,700 -250 to -67,000 N/T - 2.300 781' 4" El. Steel Pitfm. -1,400 < 1,000 to -2,000 N/T ~5,400 1 Other Surfaces - 41,000 -360 to ~4.8E5 N/T - 12,000 ;
l Instrument Racks 1 & 2 - 64,000 - 5.0E4 to ~ 7.5 E4 - 16 N/T I Shutdown Cooling Hx N/T N/T 18.4 Nrr l Shutdown Cooling Pumps - 6.54E5 -5.0E5 to -7.96ES 12 N/T O - 500 < 234 to -880 < 0.2 Nrr Condensate Return Pump [ Condensate Return Tank ~ 2,600 < 234 to - 11,600 < 0.2 Nrr PZR Discharge Tank -7,000 One N/T 63.4 N/T Smear Discharge Tank Pumps - 10,600 N/T 5,9 - 10,800 One Smear Sump Pumps Nfr NTT 2.5 In Sump by N/T I Pumps s SW Demin. Suction Filter Nrr N/T 3.2 N/T Purification Filter - 1.3 E5 N/T 2.6 N/T One Smear SW Demin. Disc. Filter Nfr N/T 6.5 Nfr Rod Room Air Handler -880 - 360 to - 1400 Nrr - 750 Rod Room Vent Fan ~2100 -750 to ~3500 Nrr -700 CRDM'S ~ 6,620 N/T Nrr ~4,000 One Smear Fuel Storage Rack - 8,300 -8,100 to -8,400 N/T N/T Note 1: " Contact Measurements unless otherwise noted. Note 2: *N/A" Not Applicable. 'Nrra - Not Taken.
] .,j 3-18 Rev.0
1 l l l SAXTON NUCLEAR EXPERIMFNTAL CORIGtATION DECOMMISSIONING PLAN
%/
Table 3.1-3 AREA 2 i (~11 mR/hr General Area) l Components & Mean Smear Range Smear ' Contact Direct Frisk Imations DPM/100cm2 DPM/100cm2 mR/hr CPM 779' 8" El. Concrete ~ 55,000 - 1,800 to ~ 1,100,000 N/T ~ 21,000 Floor i 814' 6" El. Concrete ~ 800 ~240 to ~ 1,000 N/T ~800 Ceiling Concrete Walls ~1,800 - 110 to ~24,000 Nfr - 10,600 789' 4" El. Steel Pitfm. ~ 107,000 ~ 1,500 to ~ 300,000 N/T N/T l 795' 2" El. Steel Pitfm. ~ 8,700 ~ 6,200 to ~ l1,000 N/T ~ 7,700 l O t J 807' E!. Steel Pitfm. ~ 3,500 ~ 2,000 to ~ 6,000 Nfr N/T !
's Other Surfaces ~ 75,000 ~ l,150 to ~570,, X) N/T 7,800 Steam Generator ~ 5,300 ~ 2,000 to ~ l3,500 23.2 N/T Pressurizer ~ 9,100 ~ 980 to ~ 33,400 2.7 N/T Main Coolant Pump ~ 55,210 -4,500 to ~ l80,000 2.6 N/T Regen Heat Exchanger ~ 22,200 ~ 1,100 to ~ 100,000 101 N/T Non-Regen Heat - 4,000 - 1,300 to ~ 9,100 6.3 ~ 12,000 Exchanger Instrument Rack ~ 571,000 ~ 32,000 to ~ 750,000 Nfr N/T Over Head Air Handler N/T Nfr N/T N/T Primary Air Handler ~ 15,100 ~ 2,700 to ~ 27,000 4 - 35,000 Note 1: ' Contact Measurements unless otherwise noted.
Note 2: "N/A" - Not Applicable. "Nff" - Not Taken. A V 3-19 Rev.0
SAIlON NUCIEAR EXPERIMDfrAL CORPORATION t n i DECOMMISSIONING PLAN i l Table 3.1-4 4 AREA 3 (~3 mR/hr J General Area) Components & Mean Smear Range Smear
- Contact Direct Frisk l I4 cations DPM/100cm' DPM/100cm2 mR/hr CPM i
] 781' 4" El. Steel Pitfm ~ 8,100 ~ 1,200 to ~47,500 N/T ~ 7,800
- 1 795' 2" El. Steel Pitfm ~ 2,000 - 860 to ~ 4,400 N/T ~ 3,000 )
- i
! Concrete Ceiling ~ 790 ~ 230 to - 1,000 N/T N/T Concrete Walls ~ 610 ~ 230 to ~ 4,900 N/T ~ 6,600 l l
> Other Surfaces . ~ 6,100 ~460 to ~ 59,000 N/T ~ 5,900 Storage Well Heat N/T N/T 1.4 N/T Exchanger
/*
Storage Well Pumps N/T N/r 0.9 N/T { Component Cooling HX ~ 3,800 ~ 1,700 to ~ 8,600 N/T N/T Component Cooling N/T N/T N/T N/T
- Storage Tank
) Component Cooling ~ 16,500 ~6,900 to -38,400 < 0.2 ~ 2,100 Pumps Auxiliary Air Handler - 7,000 - 1,900 to - 12,000 N/T ~ 12,000 Operating Area Air ~ 13,500 - 1,000 to ~ 26,000 N/T ~ 21,000 Handler 4 Note 1: ' Contact Measurements unless otherwise noted, i Note 2: "N/A" - Not Applicable. "N/T" - Not Taken.
;V )
3-20 Rev.O i i
SAXTON NUCLEAR EXPERBGNTAL CORPORATION 7- DECOMMISSIONING MAN Table 3.1-5 AREA 4 (~0.2 mR/hr General Area) Components & Mean Smear Range Smear
- Contact Direct Frisk lacations DPM/100cm2 DPM/100cm2 mR/hr CPM 812' El. Ops Floor -360 ~ 250 to - 1,500 Nfr - 1,700 Concrete 818' El. Concrete ~1,030 - 1,000 to ~ 1,500 Nfr ~ 450 Pitfm. !
l 818' El. Steel Pitfm. ~1,000 s 1,000 Nfr ~ 3,500 j Concrete Walls ~270 ~ 270 to - 1,000 Nfr ~ 380 812' to 818'El. Steel ~270 ~ 250 to ~ 370 Nfr N/T Walls 818' to 845' El. Steel ~280 ~260 to ~520 Nfr Nfr I Walls Movable Bridge -870 ~ 270 to ~ 2,000 N/T ~800 Polar Crane ~1,300 ~280 to -3,500 Nfr ~ 2,900 CV Upper Dome < 1,000 s 1,000 Nfr ~ 60 Equip. Access Hatch ~ 290 ~ 250 to ~ 370 Nfr ~ 1,200 Personnel Access ~530 ~ 260 to ~ 9,100 Nfr N/T Hatch Escape Hatch ~1,700 ~ 790 to - 8,900 Nfr Nfr Other Surfaces ~1,200 -250 to ~5,400 Nfr ~ 1,400 Ventilation Cavity -870 ~ 270 to ~ 2,200 Nfr Nfr Contaminated Pumps ~ 294,000 ~ 1000 to -4,800,000 -0.3 Nfr (3ea) Electrical Distribution ~ 3,500 - 1,040 to ~ 5,300 Nfr Nfr Box 3-21 Rev.0
SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN Table 3.1-5 (Continued) AREA 4 Components & Mean Smear Range Smear
- Contact Direct Frisk Locations DPM/100cm2 DPM/100cm2 mR/hr CPM Tool Rack - 1.500 N/T N/T ~ 2,800 One Smear Fuel Handling Tools N/T N/T N/T Nfr Teleflex Shield ~ 330 ~ 260 to ~ 1,100 N/T Nfr Fuel Trans Cask Skid ~ 400 ~ 250 to ~ 520 N/T N/T -
1 Head Stand ~ 290 ~ 270 to - 350 N/T ~ 600 l l Lights NTT NfT NTT NTT 1 Air Circ. Fan ~ 6,400 - 1,000 to ~ 12,000 N/T ~ 16,800 Note 1:
- Contact Measurements unless otherwise noted.
Note 2: "N/A" - Not Applicable. "N/T" - Not Taken 1 i 4 4 4 f O 3-22 Rev.O
SAXTON NUCLEAR EXPERIMENTAL CORPORATION
\ DECOMMISSM)NING PLAN O
Table 3.1-6 AREA 5 (~0.01 mR/hr General Area) Components & Mean Smear Range Smear
- Contact Direct Frisk Imcations DPM/100cm 3 DPM/100cm2 mR/hr CPM 803' El. Concrete Floor Water Covered Water Covered Water Covered Water Covered 803' to 8Il' El. ~220 s 220 N/T - 30 Concrete walls Concrete Ceiling ~ 220 s 220 N/T - 10 Exposed Steel N/T N/T N/T N/T
/
G 4 Note 1: " Contact Measurements unless otherwise noted. Note 2: "N/A" - Not Applicable. "N/T" - Not Taken. 4 4 v 3-23 Rev.O
SAX'IDN NUCIEAR EXPERIMENTAL CORPORATION b DECOMMISSIONING FLAN LJ Table 3.1-7 AREA 6 (~200 mR/hr General Area) Components & Mean Smear Range Smear
- Contact Direct Frisk Imations DPM/100cm' DPM1100cm2 mR/hr CPM 765' 8" El. Concrete Floor ~ 2.7E6 ~2.4E6 to 2: 3.0E6 N/T N/A 807' El. Concrete Ceiling N/T N/T N/T N/T Concrete Walls RS ~ 31,600 ~ 2,C')0 to - 100,000 N/T N/A Concrete Walls SWS ~ 470,000 ~ 20,000 to ~ 4.8E6 N/T N/A 793' 2" El. Aluminum Platform - 138,000 ~55,000 to -323,000 N/T N/A 768' 3" El. Concrete Ledge ~1.74E6 ~ 420,000 to ~ 2.4E6 N/T N/A 779' 8" El. Concrete Floor ~ 200,000 N/T N/T N/A One Smear 800' 6" El. Aluminum Platform - 240,000 -20,000 to ~ 1.lE6 NTT N/A Reactor Vessel 8.5E5 ~ 60,000 to ~ 2.4E6 5,000 N/A Reactor Vessel Upper Support ~ 29,600 ~2,900 to ~ 140,000 N/T N/A Casing Reactor Vessel Piping N/T N/T 52 N/A RV Support Structure 8.5E5 ~ 60,000 to ~ 2.4E6 5,000 N/A From RV RV Head Lift Rigging N/T N/T Nfr N/A
> Internals Rigging N/T Nfr N/T N/A Support Stand Nfr N/T N/T N/A i Spent Fuel Pool Elevator - 150,000 - 100,000 to - 200,000 Nfr N/A i Super Heated Test Equipment Nfr Nfr 12 N/A 3-24 Rev.O
l l l i SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN U 1 Table 3.1-7 (Continued) l l AREA 6 l Components & Mean Smear Range Smear
- Contact Direct Frisk Locations DPM/100cm2 DPM/100cm2 mR/hr CPM Tripod N/T N/T N/T N/A Spent Fuel Storage Racks ~ 134,000 ~39,000 to ~424,000 N/T N/A Boric Acid Demineralizer ~1.8E6 N/T 8000 N/A One Smear Storage Well Demineralizer ~ 4.8E6 Nfr 280 N/A One Smear Purification Demineralizer ~ 60,000 N/T 240 N/A One Smear Other Surfaces ~ 150,000 - 100,000 to -200,000 N/T N/A i Note 1:
- Contact Measurements unless otherwise noted.
Note 2: "N/A" - Not Applicable. "Nff" - Not Taken. i l O 3-25 Rev.0
l SAXTON NUCLEAR EXPERIMENTAL CORPORATM)N 3
\ DECOMMISSIONING PLAN ,}
s. l Table 3.1-8 AREA 7 (~0.017 mR/hr General Area around CV outside) ) 1 Components & Mean Smear Range Smear
- Contact Direct Frisk tecations DPM/100cm 2
DPM/100cm 2 mR/hr CPM 811' to 817 El. Steel < 1,000 s 1,000 N/T - BKGND Walls 817' to 845' El. Steel < 224 s 224 N/T - 70 Walls 845' to 870' < 224 s 224 N/T -230 l l El. Steel Dome Concrete Wall < 1,000 s 1,000 N/T -BKGND (^g (Horizontal)
-BKGND \j Concrete Wall < 1,000 s 1,000 N/T (Vertical) l Note 1: ' Contact Measurements unless otherwise noted.
l Note 2: "N/A" - Not Applicable. "N/T" - Not Taken. 1 J O G Rev. O 3-26 1
SAXTON NUCLEAR EXPERIMENTAL CORFORATION
/ \ DECOMMISSIONING FLAN
\s.-) Table 3.1-9 AREA 8 (-0.014 mR/hr General Area average) Components & locathms Meun Smear Range Smeur
- Contact Direct Fri.sk CPM DPM/100cm 8 DPM/100cm 2 mR/hr Srptic Tank " A" 0.00023 G.A.
in tank Concrete Flixir Sediment Covered Sediment Covered Sediment N/A Covered Concrete Walls N/T N/T N/T N/T Concrete Ceiling N/T N/T N/T N/T Concrete Manway N/T N/T 0.00011 -BKGND Scotic Tank "B" - 0.00012 G.A. in tank Concrete Fhuir Sedunent Covered Sediment Covered Sediment N/A [,m} Covered \ /
'U Concrete Walls N/T N/T Nrr N/T Concrete Ceihng N/T N/T N/T N/T Concrete Manway N/T N/T 0.00065 -BKGND Pump Well --- 0.0001 G.A.
in well Concrete Floor Sediment Sediment Sediment N/A Covered Covered covered Concrete Walls N/T N/T N/T N/T Concrete Ceihng N/T N/T N/T N/T Concrete Manwmy N/T N/T 0.00035 -BKGND Concrete Bhick N/T N/T Nrr N/T lituseFkwir Concrete Chlorinator / Weir - Concrete Floor < 239 s 239 N/T -BKGND Conerete Walla < 239 s 239 0.0001 -BKGND i Note 1: " Contact Measurements unien otherwix noted. O Note 2: "N/A" - Not Applicable. "N/T" - Not Taken. (v) 3-27 Rev.O i l
SAXTON NUCLEAR EXPERIMENTAL CORFORATION
, DECOMMISSIONING PLAN l
Table 3.1-10 ALPflA CONTAMINATION AREA 1 l Structural Mean Smear Range-Smear Surface DPM/100cm' DPM/100cm2 765' 8" El. Concrete Floor ~ 30 < 8 to ~ 90 768' 3" El. Concrete Ledge - 20 < 8 to -30 765' 2" El. Sump ~ 70 ~20 to - 130 775' 2" El. Concrete ~5 N/T Ceiling One Smear Concrete Walls ~ 10 < 8 to ~ 80 f 781' 4" El. ~5 N/T Steel Platform One smear Other Surfaces ~ 40 < 8 to ~ 570 Note: "N/T" - Not Taken f .I i O 3-28 Rev.0
p SAX' ION NUCLEAR EXPERIMENTAL CORFORATION DECOMMISSIONING PLAN Table 3.111 AIEHA CONTAMINATION AREA 2 Structural Surface Mean Smear Range-Smear DPM/100cm2 DPM/100cm2 779' 8" El. Concrete Floor ~ 30 < 8 to - 120 814' 6" El. Concrete - 10 < 5 to - 10 Ceiling Concrete Walls ~8 ~ 8 to - 10 789' El. Steam Gen. ~ 670 < 8 to 4700 Platform e 795' El. Steam Gen. ~ 24 < 24 to ~ 24 Platform ( 870' El. Steam Gen. ~ 10 < 5 to ~ 25 Platform Other Surfaces - 10 <MDA to -350 1 ( k 3-29 Rev.0
SAX 10N NUCIEAR EXPE3tlMEMfAL CORPORATION DECOMMISSIONING FLAN Os Table 3.1-12 ALPHA CONTAMINATION AREA 3 Structural Sunface Mean Smear Range-Smear 2 2 DPM/100cm DPM/100cm 78l' 4" El. Steel Platform - 10 < 8 to ~ 50 795' 2" El. Steel Platform ~ 10 < 8 to ~ 20 Concrete Ceiling <5 N/T One Smear l Concrete walls -8 < 8 to - 10 Other Surfaces - 10 < 8 to ~ 60 s e Note: "N/T" - Not Taken i 3-30 Rev.O
SAXTON NUCLEAR EXPERIMI!NTAL CORPORATION Og DECOMMISSIONING FLAN V Table 3.1-13 ALPHA CONTAMINATION AREA 4 Structural Surface Mean Smear Range-Smear DPM/100cm' DPM/100cm2 812' El. Ops Concrete Floor <8 <8 818' El. Concrete Platform <5 <5 818' El. Steel Platform <5 <5 l Concrete walls <5 <5 Steel Walls <8 <8 C Polar Crane -5 <8 Other Surfaces -8 < 8 to ~ 10 .. a l 3 m 3-31 Rev.O
( l I SAX 1DN NUCLEAR EXPERIMDfrAL COItPORATION DECOMMISSIONING FLAN Table 3.1-14 ALPHA CONTAMINATION ARFA 6 i Structural Surface Mean Smear Range-Smear DPM/100cm2 DPM/100cm2 765* 8" El. SW Concrete ~ 560 ~ 280 to ~ 840 Floor ! l Concrete Walls Rx Side ~130 ~ 6 to ~ 870 Concrete Walls SW Side ~ 460 ~ 30 to ~ 3,500 , 793* 2" El. Steel Platform ~1,400 - 80 to ~ 4,800 1 768' 3" El. SW Concrete ~ 880 ~ 230 to ~ 1,500 Ledge [D Rx Vessel Support Can -380 - 10 to ~ 2,700 Fuel Storage Rack ~1,100 ~700 to ~ 1,900 800' 6" El. Rx Side Steel ~ 110 ~ 60 to ~ 200 Platform i \ C 's V 3-32 Rev. O
=_ . -- . _ . - _ _ . . . ._
SAXTUN NUCIEAR EXPERIMENTAL CORPOItATION
, [ i DECOMMISSIONING FLAN Table 3.1-15 COMPOSITED GENERAL AREA SMEAR RESULTS J
AREA 1 AREA 2 AREAS 3 & 4 AREA 6 765' 8" Ausiliary Primary Compt. 781'-4",795'2", 812', Spent Fuel Pool and Compt. 818'4" Aus Compt, Op. Floor Reactor Storage Well isotope pCi's As % pCi's As % pC1's As % pCi's As %
<MDA 1.10E-5 0.0001 l C.14 <MDA - 1.10E-5 0.002 -
Ni-59 <MDA - <MDA - <MDA - 1.70E-3 0.010
$r-90 6.8 E-4 0.168 1.20E-2 1.952 <MDA -
2.30E-2 0.140 Fe-55 <MDA - <MDA - <MDA - 2.50E-2 0.152 Tc-99 <MDA - 3.20E-5 0.005 <MDA - 6.20E-4 0.004 Co-60 2.87E-3 0.794 1.49E-2 2.424 2.59E-4 3.905 7.18E +0 43.661 Co-137 3.568-1 98.428 5.63E-1 91.58 6.26E-3 94.382 9.03E+0 54.910 l H.3 <MDA - <MDA - <MDA - 2.00E-3 0.012 Ni-63 1.20E-3 0.332 1.90E-2 3.091 8.90E-5 1.342 1.30E-1 0.791
- Pu-238 4.60E-5 0.013 3.50E-4 0.057 4.00E-6 .060 3.10E-3 0.019 Am-241 1.80E-4 0.050 8.50E-4 0.138 1.20E-5 .181 1.30E-2 0.079 Cm-242 <MDA - 7.40E-6 0.001 <MDA - 1.00E-4 0.001 l Cm-244 <MDA - 1.90E-5 0.003 <MDA - 3.00E-4 0.002 Pu-239 1.00E-4 0.028 8.90E-4 0.145 8.60E-6 .130 8.20E-3 0.050 Pu-241 6.10E-4 0.169 3.70E-3 0.602 <MDA - 2.80E-2 0.170 i Pu-242 <MDA - 4.80E-6 0.001 <MDA - 2.40E-5 0.0001 Note: *<MDA* Less Dan Minimum Detectable Activity N/
3-33 Rev.O
l SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING FLAN
%.)
4 Table 3.1-16 l REACTOR VESSEL / INTERNALS CURIE DETERhENATION l Components Weight in Ibs. Curies 24 Years After Shutdown (July 1,1996) Core Baffle 1,210 642.6 Thermal Shield 10,491 132.7 i Reactor Vessel Clad 2,043 4.2 Reactor Vessel Wall 87,313 4.7 Reactor Insulation Can (1) 5,012 1.078 Reactor Support Can (1) 5,442 3.034 Lower Core Guide Blocks 140 182.9 Lower Core Plate 500 226.7 Lower Support Shroud Tubes (2) 338 8.323 [m)
%/
Lower Support Tie Rods (2) 54 2.113 Balance of Lower Support Assembly 718 0.0744 (3) Lower Core Barrel (4) 1,%7 58.46 Upper Core Plate 264 165 Upper Core Barrel (4) 454 19.71 Balance of Lower Core Barrel 1,823 0.2593 Assembly (5) Balance of Upper Core Barrel 832 0.1183 Assembly (6) Votes:
- 1) Includes weight of component between 3 feet above and 3 feet below the active core.
- 2) includes weight of component down to 3 feet below the active core.
- 3) Weight of lower support shroud tubes, lower support tie rods and associated components more that 3 feet below the active core.
- 4) Includes weight of component up to 3 feet above the active core.
- 5) Weight oflower core barrel and associated components more than 3 feet above the active core.
- 6) Weight of upper core barrel and associated components more than 3 feet above the active core.
O 1 3-34 Rev.O I
SAXTON NUCLEAR EXPERIMENTAL CORPORATION V' DECOMMISSIONING PLAN Table 3.1-17 ENVIRONMENTAL SAMPLE ANALYSIS i i SAMPLE TYPE NUMBER OF ANALYSIS TYPE SAMPLES l Site Characterization (soil) 538 garmna spec i 1 22 transuranics/ l strontium / gamma spec Soil Remediation (soil) 381 gamma spec Hazardous Material Profile 15 TCLP
' []}
Soil Density Profile I compactability/ moisture content Air Particulates 96 gamma spec / Gross Alpha / Gross , Beta Potable Water 19 gamma spec / tritium / Gross Beta l Groundwater 87 gamma spec / tritium Sediment 27 gamma spec l l Total Samples 1186 l O. U 3-35 Rev.O
-- - -- -=. . . - . - -~
f'N g 'x TABLE 3.1-18 ' SNEC FACILITY CONCRETE WASTE VOLUMES AREAS 1-8, SNECSite CV And Surrounding Facility 11/8/95 Depth Of NOTE: Volumes are based on the best currentiy available information. Concrete To Number of Be Removed Volume in Weight Weight AREA ITEM DESCRIPTION Square Feet Cubic Feet (laches) (Ibs) (tons) 1 765* 8" El. Floor 705 6.6 399.5 56862 28.4 1 768' 8" El. Ledge 350 1.3 37.9 5397 2.7 1 765' 8" El. Walls 2543 4.3 911.2 129700 64.9 1 775' 2" El. Ceiling
- 733 0.25 15.3 2174 1.1 2 779' 8" El. Floor 475 5.1 201.9 28734 14.4 2 5' Thick Primary Compartment Wall (See Note 1) 968 59.9 4833.9 688029 344.0 ,
2 Primary Compartment Walls, Other Than 5' Section 1875 5.3 827.9 117845 58.9 2 814' 6" El. Concrete Ceiling
- 475 0.25 9.9 1409 0.7 3 Auxiliary Compartment Walls (All) 2405 2.7 541.1 77020 38.5 3 810' El. Ceiling
- 381 0.25 7.9 1130 0.6 4 818' El Operating Floor 669 0.6 33.4 4761 2.4 4 812' El Operating Floor, includes Top Of Shid Plugs 1202 1.8 180.3 25663 12.8 4 812' El Operating Fir., Interior Walls 341 0.6 17.1 2427 1.2
_5_ CV Tunnel Walls Below Grade 2014 0.4 67.1 9555 4.8 5 CV Tunnel Ceiling Below Grade 953 0 0.0 0 0.0 s S CV Tunnel Floor, Below Grade (Assume 812' Ops. Floor D epth) 953 1.8 143.0 20347 10.2 5 CV Tunnel Ceiling, Exterior (Grade level) 953 0.6 47.7 6782 3.4 6 765* 8" Storage Well Floor 1035 5.6 483.0 68747 34.4 6 Storage Well Walls- SW Side 2465 2.8 575.1 81855 40.9 6 Rx Cavity Side Walls, < Operating Level Water Line (NO S' WALL) 387 1.6 51.5 7337 3.7 6 Rx Cavity Side Walls, > Operating Level Water Line (NO 5' WALL) 545 18 817.5 116358 58.2 6 Rx Cavity Side Ledge At 779' 8" El 113 1.9 17.9 2547 1.3 6 807' El. Ceiling Of Storage Well (Shield Blocks etc.) 574 12.8 612.3 87146 43.6 7 Concrete Walls And Pads Outside Of CV 500 0.2 8.3 1186 0.6 8 Sanitary Sewage Treatment Facilities. Weir, Tanks etc. 993 0 0.0 0 0.0 TOTALS = => 24606 N/A 10841 1543009 772
- Volume Based On A Minimium RemovalDeph Of 0.25 inches.
Note 1: The 5' thick wall between the Rx and Primary Compartments may not be completely contaminated. However, lower sections are contaminated to a depth which may require complete removalof his wall. 3-36 Rev3
l Figure 3.1-1 l Area topographic map i l nuurmacon n w / 133 ,78 'I b' '
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Figure 3.1-2 [
- SAXTON NUCLEAR (
N - EXPERIMENTAL STATION / \ l S .. Location of Environmental i ) j - . Sampling Points 4
- % s (
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'v SAXTON AREA LOCATIONS Average Contamination Levels and General Area Exposure Rates
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ty Plan Above Operating Floor Elev 812'-0" Containment Vessel (CV) up scAte
1 1 Figure 3.1-8 l f' G SAXTON AREA LOCATIONS Average Contamination Levels and General Area Exposure Rates .%- g -3.500 dpm/100 sq cm h;$ MlEN! b AREA 2 gdb ~
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Figure 3.1-10 (~'\ SAXTON AREA LOCATIONS ('y Average Contamination Levels and General Area Exposure Rates h j AREA 1 , A cr--agggg, j *% %; AREA 1
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Figure 3.1-12 SNEC Environmental Monitoring Points n v -
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4 i l l l l sArmN NUCutAR EIFR3tatDirAL CORFORAMON 1 i 3.2 RADIATION PROTECTION l 3.2.1 Radiolorical Contmls Pmgram l The objectives of the Radiological Control program at the Saxton facility are to control radiation hazards to avoid accidental radiation dose, to maintain doses within the regulatory requirements, and also maintain doses to the workers and the general population as low as reasonably achievable (ALARA). These philosophies, policies, and objectives are based on, and implement, the regulations of the Nuclear Regulatory Commission (NRC), as contained in Title 10 of the Code of the Federal Regulations (10 CFR) Parts 19, 20, 50, and 71, and the appmpriate l Regulatory Guides. In addition, the Radiological Contmls pmgram will support the decommissioning effort by providing radiological data information and the documentation of the site release surveys. All decommissioning work will be accomplished in accordance with the Saxton Radiation Pmtection Plan as presented in Saxton facility pmcedure No. 6575-PLN-4542.01 "SNEC Radiation Pmtection Plan", or equivalent. Ensurine that Ocennational Radiation Exposures Are As Low As Is Reasonably Achievable (ALARA) The Saxton facility will be dismantled under the administrative oversight of GPU Nuclear. GPU Nuclear continues to operate with, and supports the ALARA concept. The Saxton facility ALARA procedure has been written by GPU Nuclear personnel and incorporates the current ALARA policy and concepts used by GPU Nuclear in meeting ALARA obligations at TMI and Oyster Creek. Saxton facility decommissioning administrative personnel will continue to enforce high standards with respect to controlling personnel exposure and exposure rate, as well as other radiological monitoring requirements. O 3-37 Rev.O
P SAXTON NUCIEAR EXPERIMENTAL CORPORATM)N DECOMMISSK)NING FLAN (G l Personnel exposure will be the highest during the removal and handling of the activated and/or contaminated components of the containment vessel. 'Iherefore, each major CV dismantling phase will be performed under a detailed plan that will include a work package or work instructions which shall consider methods for reducing exposure to personnel. These work instmetions will be reviewed by Radiological Contmls personnel and concurmd with by the Radiation Safety Officer (RSO) prior to issuance of Radiation Work Permits (RWP's). On-going reviews of work packages, RWP's, and radiological survey data during the work effort will ensure that changing radiological conditions are considered and accounted for. ALARA concepts developed and practiced during the TMI-2 cleanup effort as well as " lessons learned" will be implemented during the Saxton facility CV dismantling operation. ALARA dose considerations will be reviewed and considered for each RWP and work instruction l 1 prepared. Issues such as radiation shielding, worker exposure times and respiratory protection 1 O requirements will be discussed with workers to ensure their full understanding and cooperation l with good radiological controls work practices. When appropriate, worker input will be solicited to efficiently apply the best combinations of ALARA control features during work evolutions. Past TMI-2 experiences should significantly increase the efficiency of the Saxton facility decommissioning effort. Decontamination techniques, radiological measurement methodology and analytical tools developed during the cleanup phase of TMI-2 may be employed to reduce the number of or complexity of work evolutions performed during the Saxton facility decommissioning operation. Organization and Personnel The organizational stmeture of the Radiological Controls program is presented within the overall Saxton facility decommissioning organization in Figure 2.3-2. The administrative organization of the Radiological Controls pmgram is structured to provide excellent communication and 3-38 Rev.0
k 4
; SAXTON NUCIAut EXFERIMMFAL CORPOILATION DECOMMISSIONING PLAN administrative control during the decommissioning effort. . The Radiological Controls staff reposting to the Radiation Safety Officer (RSO) will typically be composed of radiological engineers, radiological controls supervisors and technicians. The exact number will vary according to the number of tasks performed simultaneously and the needs l
of specific job functions. Support from GPU Nuclear radiological engineering personnel will be requested on an as needed basis. i Training The training requirements for Radiological Controls personnel will be similar to those presented for Radiological Controls personnel routinely supporting TMI and Oyster Creek activities. Radiological Controls personnel will meet or exceed the qualifications outlined in ANSI N18.1- j ( 1971 for their respective positions and/or be approved through an NRC approved training j program. Special job functions may be performed after sufficient on site training at the discretion of the RSO. However, all personnel shall have sufficiently documented training l records and/or an extensive experience history demonstrating past competency in the skill area. k Exposure Limits f The Saxton facility administrative exposure limits are controlled by procedure. Administrative limits are approximately eighty percent of those presented in 10 CFR 20 " Standards for Protection Against Radiation". Several of these administrative dose limits are listed below. G 3-39 Rev. 0
p SAX'IDN NUCIEAR EXPERIMENTAL CORPORATION DECOMMISSIONING FLAN ( ( Dose Categorv Yearly Limit Total Whole Body (TEDE)...... .. ......... .. 4 rem Eye (LDE) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 rem Skin (SDE-WB) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 rem Extremity (SDE) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 rem Organ (TODE) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 40 rem Declared pregnant workers shall normally be limited to 0.4 rem total whole body dose for the term of pregnancy using a limit of 0.05 rem / month. Mom restrictive individual limits may be established on a case by case basis, b U Access Control and Radiological Postings The minimum criteria used to establish radiologically controlled amas (RCA) is controlled by procedure. This procedure describes the requirements for radiological postings at the entrances and boundaries of the RCAs. These posting requirements comply with 10 CFR Part 20 and will be used to advise and warn personnel of the radiological conditions that may be encountered in each area. Radiation work will be controlled by use of the Radiation Work Permit (RWP). Radiation Work Permits The Radiation Work Permit (RWP) is a tracking and controlling device. The RWP is used to 4 administer protective measures, clothing and equipment, and tracks exposure incurred on a specific job thus allowing the Radiological Controls group to modify (as necessary) protective requirements and review current conditions. The control of radiation work will be based on the
.O V
3-40 Rev.O
I SAX 10N NUCLEAR EXPERIhENTAL CORPORATION ! 1 tO DECOMMISSIONING FLAN
- J J
best available survey information. Radiological Controls personnel determine the requirements l' for an RWP based on job site and radiological conditions. All personnel assigned to a work site must understand and acknowledge their compliance with the specific mquirements intended. I I Radiation Pmtection Support Facilities d The Saxton radiological controls support facilities include extensive GPU Nuclear resources l , already in place. GPU Nuclear has comprehensive personnel dosimetry, gamma spectroscopy, i liquid scintillation counting and air sampling instmmentation, environmental monitoring, 1 bioassay and whole body counting equipment, and radiological instrument calibration and service at both their TMi and Oyster Creek facilities. GPU Nuclear maintains industrial safety and medical staff at both TMI and Oyster Creek. All of these support elements will be available to j support the decommissioning effort as required. In addition, GPU Nuclear maintains contracts with outside organizations for sample analysis and other support functions. At the Saxton site, routine radiological survey equipment are in use and are maintained and calibrated by GPU Nuclear. GPU Nuclear maintains on the Saxton facility site, a gamma spectroscopy system for quantification and identification of radionuclides, and has access to other portable systems as needed. The GPU Nuclear Environmental Radiological Laboratory (ERL) also maintains considerable on-site envimnmental sampling systems. l l , InstrumentaliQD l Typical portable instrumentation used at the Saxton facility include but are not limited to: i 3-41 Rev.0
SAX 11)N NUCMAR EXPERIMENTAL CORPORATION DECOMMISSIONING FLAN O 1 Instmmentation Radiation Type Range / Scale Application ! EIC RO-2 Beta / Gamma 0 - 5 R/hr Dose Rate Surveys EIC RO-2A 0 - 50 R/hr ASP-1 Multi-purpose Various Portable Survey dependant upon Meter probe used EIC E-520 Gamma 0 - 2 R/hr Dose Rate Surveys EIC E-530N Gamma 0 - 20 R/hr Dose Rate Surveys i l Ludlum Model 19 Gamma 0 - 5 mR/hr Low-level Dose Rate Surveys EIC 140N Beta / gamma 0 - 50000 CPM Personnel and area EIC RM14 contamination monitoring Ludlum 2000 w/ Beta / gamma Digital readout Count Room HP-210 probe instmmentation
! Air Samplers Particulate N/A Airbome Sampling (Various) I 1
1 EIC SAC 4 Alpha Digital readout Count Room j Instrumentation All radiation monitoring equipment used in monitoring personnel exposures shall be calibrated and maintained by a qualified service organization. GPU Nuclear will provide these services 3-42 Rev.0
SAITON NUCt&Ut EIFEREWINTAL CORPORATION i for most of the equipment used at the Saxton facility. In all cases, equipment calibration and maintenance shall conform to the requirements of the GPU Nuclear Quality Assurance Pmgram for Radiological Instruments (6610-QAP-4220.01) or the Saxton Nuclear Quality Assurance Program for Radiological Instmments (6675-QAP-4220.01). In some instances special non-standard radiation measurement equipment may be used to characterize radiation fields or quantify selected radionuclide content. Special non-standard instrumentation shall be calibated against certified standards and be used and maintained by qualified and experienced personnel. When necessary, training and training documentation on specialized equipment or instrumentation shall follow the same format as outlined above for other training programs. Special non-standard measurement instmmentation shall not be used to release materials or personnel for unrestricted use from radiologically controlled areas of the Saxton facility unless the measurement methodology has been appropriately verified, documented ( r and approved by the RSO. l Radiological Surveys
- The minimum radiological survey requirements for radiation, contamination and airborm activity l l
are controlled by procedure. This procedure also provides the requirements for survey i documentation and reviews. Radiological surveys will be performed on a routine and non-q routine basis for verification and documentation of radiological conditions for use in the control l of personnel dose while engaged in Saxton facility decommissioning activities. Radiological j surveys will be performed by qualified and experienced personnel. All radiological survey l instrumentation will be handled in accordance with the aforementioned requirements.
! Control of Radioactive Contamination
, Radioactive surface contamination shall be controlled to the extent possible to minimize the i l 3-43 Rev.0 G
SAXTON NUCLEAR EXPERIMFKTAL CORPORATM)N O I!ECOMMISSK)NING PLAN i O
, possible inhalation or ingestion of radioactivity. The limits used to verify the absence of . contamination are stated in the SNEC Radiation Protection Plan 6575-ADM-4542.01 or equivalent. In addition, the RWP will set the appropriate contamination controls requirements when working in highly contaminated areas. Personnel contamination monitoring requirements are controlled by procedure. This procedure outlines requirements and techniques for personnel frisking. An automated whole body frisker (e.g. PCM-1) can be used in lieu of a whole body ; frisk.
Control of Radioactive Materials A radioactive material control system will be established to ensure radioactive materials are not lost or misplaced in a location where personnel could unknowingly be exposed to radiation and to control the spread of radioactivity to areas where the public might be affected. The D) g radioactive material control program is outlined in section 3.3. V Radiological Effluent Monitorin.g l The Offsite Dose Calculation Manual (ODCM) is a support document of the Technical Specifications and implements SNEC mdiological effluent controls. The ODCM contains the controls, bases, and surveillance requirements for liquid and gaseous radiological effluent. In addition, the ODCM describes the methodology and parameters to be used in the calculation of the off-site doses due to radioactive liquid effluent. This document also describes the methodology used for calculations of the liquid and gaseous effluent monitoring instmmentation alarm and trip set points. The ODCM follows the methodology and models suggested by NUREG-0133 and Regulatory Guide 1.109 Revision i for calculation of offsite doses due to 4 plant effluent releases. Simplifying assumptions have been applied in this manual where applicable to provide a more workable document for implementation of the Radiological Effluent Controls requirements.
/
3-44 Rev.O
SAXTON NUCLEAR EXPERIMEPfrAL CORM) RATION DECOMMISSIONENG PLAN
\' -
1 3.3 RADIOACTIVE WASTE MANAGEMENT The Saxton facility decommissioning will require the handling of a relativel 3 arge l volume of radioactive materials to reduce the residual levels of radioactivity to a level pennitting release of the site for unrestricted use and tennination of the license. Materials that are not decontaminated and released will be processed as radioactive waste. This section of the decommissioning plan presents the programs used to manage and control the processing of solid, liquid and gaseous radioactive waste. GPU Nuclear will continue to ensure appropriate processing, packaging and monitoring of solid, liquid and gaseous wastes during decommissioning by implementing the Radiation Protection Plan, Process Control Procedures and the Radiological Environmental Monitoring Program. These programs will be maintained in compliance with federal and state regulations, disposal site requirements, and any other applicable requirements. The radioactive waste program is (
- implemented through the Radiation Protection Program (Section 3.2). Implementing procedures will be used to control the classification, treatment, packaging and shipment of radioactive ,
1 material. At the time of shutdown, the Saxton facility radiological exclusion area (all surrounded by a j fence) contained the Containment Vessel, Control and Auxiliary Building, and the Radioactive I Waste Disposal Facility. Access to these structures was controlled by lock and key as well as administrative procedures. The condition of the plant was documented in a report entitled
" Decommissioned Status of the Saxton Reactor Facility" forwarded to the NRC on Febmary 20, 1975 (Reference 7). In essence all contaminated and non-contaminated equipment, tanks and piping located outside of the containment vessel were removed. All equipment located inside I the containment vessel, (reactor vessel, vessel internals, steam generator, pressurizer, piping etc.) remained assembled. Reactor control rod blades and some components of the superheated 1
steam test kop were removed from the reactor and shipped offsite for disposal in 1972. l A 3-45 Rev.O
SAXTON NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN During a dismantling effon from 1986 through 1992 all stmetures with the exception of the containment vessel were removed to 3 feet below grade. The foundations were backfilled with clean fill, The site was subsequently surveyed, characterized and contaminated soil was properly removed and disposed of at the Envirocare Disposal Facility in Clive, Utah. The current exclusion area consists of the Containment Vessel and the immediate surrounding area. During decommissioning, significant resources will be expended in processing and disposing of liquid and solid radioactive waste. Radioactive wastes include neutron activated materials, contaminated materials remaining in the containment building and tools and equipment which became contaminated during dismantling activities. 3.3.1 Fuel Disposal Shonly after shutdown during the period from July 1972 through November 1972, the nuclear V fuel, and all other removable special nuclear materials and sources were shipped off site. From November 1972 through early 1974, by-product material, with the exception of incidental quantities contained within the reactor coolant system and surface contamination in exclusion I areas, was removed from the site. 3.3.2 Radioactive Waste Processing Generally, system components will not be decontaminated on-site. If it is detennined that a component or portion of a component is to be decontaminated, an experienced off-site vendor facility will be utilized. Currently the intent is to dismantle the contaminated piping systems and dispose of the material or to decontaminate and free release those materials. The option that is most economically advantageous will be chosen. During decommissioning, liquid and solid mdwaste will be generated. These will be processed O \ 3-46 Rev.O
l l l l ! l SAXTON NUCLEAR EXPERIMI!NTAL COItPORATION i
- I O DECOMMISSIONING FLAN l
and disposed of as necessary. Radioactive wastes include neutron activated materials, contaminated materials remaining in the containment vessel and those items necessarily contaminated on-site during the dismantling activities. The following sections address ! radioactive waste management. l 3.3.2.1 Gaseous Wastes i Since termination of Saxton facility operation in 1972 and prior to dismantlement of all ; radioactive waste systems, radioactive gas had been decayed and released. Therefore, processing of gaseous waste will not be necessary. Temporary HEPA filtration systems may be requimd to contain airborne particulate radionuclides that may be generated during the performance of various decommissioning t activities. The Decommissioning Support Building will be vented through HEPA filters and j ( exhausted to the Containment Vessel (CV) atmosphere. The CV atmosphere will be monitored by portable air samplers and, if necessary, by Continuous Air Monitors (CAMS). The CV ventilation will exhaust via a HEPA filtemd ventilation system. If other activities require control of aistome contamination, portable HEPA filtration units, including those built into vacuum cleaners, will be used. l The effluent monitoring instrumentation will be used to monitor discharget of airbome effluent as mquired, and to demonstrate compliance with the Saxton Offsite Dose Calculation Manual (ODCM) limits as promulgated by applicable regulations. 3.3.2.2 Liquid Wastes Radioactive liquid wastes will be generated during the decontamination and dismantlement of the Saxton facility systems and structums. O 3-47 Rev.O
SAXTON NUCLEAR EXPERIMENTAL CORPORATION h DECOMMISSIONING PLAN V Liquid radioactive wastes generated during decommissioning will be processed as necessary using temporary systems supplied by GPU Nuclear or by experienced vendors and contractors where appropriate. The temporary waste treatment system will be connected to tanks for storage of processed water prior to discharge. Once it has been verified that the stored processed water meets the allowable discharge limits specified in the Off-site Dose Calculation Manual, the water will be released. These systems may include temporary ventilation with filtration for airtwrne contamination control. The liquid waste stream will be processed using techniques which are cost effective and mee, ALARA goals. During earlier demolition activities, installed plant equipment used to process liquid radwaste had been removed. Therefore, temporary filtmtion units or demineralizers will be used as the primary means of treatment for all planned evolutions. Any processed liquids may then be discharged after they have been monitored and approved for release. The effluent monitoring instrumentation will be used to monitor discharges ofliquid effluent as required, and ( to demonstrate compliance with the Saxton Offsite Dose Calculation Manual (ODCM) limits as promulgated by applicable regulations. Filters will be changed-out as appropriate to keep exposures ALARA. Solid waste resulting from the processing will be dewatered or solidified in appmpriate containers for disposal. Containers will be shipped to licensed burial facilities. Makeup water used for flushing will generally be from the existing plant supply. However, water that has been processed through a temporary system may be used if quantities and economics suggest a savings without a reduction in safety. The effluent stream (s) from such activities will be processed as above, by Hitration and demineralization. Temporary systems will allow for processing of water in such a manner that as few hoses as O 3-48 Rev.O
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l l /- SAXTON NUCutAgt EXPEltBENTAL COILFORATION ); n COnansmoNmo ruN reasonable will be used. System components will be positioned or shielded as equinxi to maintain dose rates to workers ALARA. Dere are no current plans to utilize chelating agents in any chemical decontamination activities for Saxton facility systems or structures. Radioactive wastes containing chelating agents will be generated only if essential, and in that case will be mmimized to the full extent possible. Current procedures ensure that burial site criteria for chelates will be met. GPU Nuclear will continue to monitor requirements for packaging and disposal of radioactive waste containing chelating agents. No radioactive waste containing chelating agents will be generated during Saxton facility decommissioning operations that result in packaged radioactive waste that is not consistent with waste form, packaging transportation and disposal requirements existing at the time when the operations are performed. 3.3.2.3 Solid Waste Processing and Volume Reduction Solid radioactive wastes will result from the processing of liquid and airbome particulate waste streams as described above. The majority of solid waste, however, will result from the decontamination and dismantlement of activated and contaminated plant systems and stmetures. Table 3.3-2 provides a conservative determination of the volume of solid radwaste resulting from the Saxton facility decommissioning. The table lists components and structural materials by area and includes all those containing known or suspected radioactivity. This includes those items with surface deposited activity and those which may be remediated to meet free release criteria. He estimated curie content, volume and waste classification as found is shown. In some cases additional information is required to determine the waste classification. When this information is known, a revision to the decommissioning plan will be made to reflect this. I 3-49 Rev.O ! I
. SAITON NUCIEut EIFERIMENTAL COItMNLATION 2
, DECOMMISSIONING FLAN
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i l I This determination is conservative because it does not take credit for any volume reduction ) techniques and, further, because it assumes only direct burial rather than allowing for i j_ decontamination and possible free release, d As indicated in Table 3.3-2, the waste contains approximately 1569 Ci, most of which is due to activation of the reactor pressure vessel intemals. Radioactive waste is expected to be primarily Class A waste. Little mixed waste is expected to be produced during decommissioning. All asbestos will be treated as contaminated unless othenvise indicated. l GPU Nuclear is planning to employ a number of measures with the overall objective of reducing 4 the volume of solid radwaste that will ultimately require disposal at a licensed burial facility, j The data provided in Table 3.3-2 does not reflect the benefits offemd by various volume-reducing techniques which art being contemplated. b
\ l i
The primary components of the solid waste to be generated by the decommissioning of the [ Saxton facility are expected to be disposed of as follows:
)
(1) Reactor Vessel and Internals ! GPU Nuclear plans to remove the reactor vessel with the internals in place, fill the ! intemal void space with concrete / grout, seal the unit, and ship as one piece. Improvements may be needed on some roads and/or bridges. ! (2) Steam Generator l The steam generator will be removed in one piece and all openings sealed. External ! loose contamination will be removed or fixed. The steam generator may be volume reduced at a vendor facility. If not feasible, it will be filled with concrete /gmut and
- shipped as a package for disposal.
3-50 Rev 0
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SAXTON NUCLEAR FIFIGtIMENTAL CORPORATION DECOMMISSIONING FLAN b (3) Pressurizer The pressurizer will be removed in one piece and all openings sealed. External loose contamination will be removed or fixed. The internal volume may be filled with other , 4 l 4 contaminated or activated debris in order to maximize packaging efficiency. If needed, j the internal volume will be filled with concrete / grout to fill void spaces. Options such : 1 as volume reduction or decontamination by qualified vendors will be pursued pnor to 1 1 committing to the disposal option. ! l l 1 (4) Piping and Equipment The contaminated systems piping will be segmented. As cuts are made, a suitable cover will be placed on open ends to preclude the spread of contamination. The small bore piping (i.e., 3 inches diameter and under) will be removed from the system, packaged and shipped off-site to a licensed vendor offering decontamination and volume mduction services. Large bore piping will be moved directly from its removal location to the packaging area. As the containers are filled, they will be moved to a staging area awaiting final preparation and loading for shipment off site to a volume reduction facility. Components and instruments will be bagged for contamination control and handled in a manner designed to minimize contamination spread. Containments will be selected with consideration of proper size, contamination levels and ability to process. That material which can economically be dismantled and decontaminated will be appropriately handled on-site or may be sent to a vendor facility. Material that cannot economically be decontaminated will be placed in proper disposal containers (i.e. LSA containers) and sent to an appropriate processor or burial facility. The accumulation of contaminated piping will not represent an ALARA concern because of the generally low dose rates of the Saxton facility piping. N 3-51 Rev.O
; SA31DN NUCI&ut EIHmtIMENTAL COILFORATION DECOMMISSIONING PLAN l
l (5) Concrete Rubble and Dust Activated or contaminated concrete rubble and dust may be packaged as low specific activity (LSA) material in approved shipping containers. When feasible, this material will be used to fill void space in other radwaste shipping containers. 1 l (6) Dry Active Waste (DAW) DAW consisting of contaminated paper, plastic, coveralls, etc. will be packaged as IJSA ! material in approved shipping containers. DAW will be shipped non-compacted to an f j off-site vendor for volume reduction and processing if supported by ALARA and cost considerations. When feasible, DAW will be used to fill void space in other radwaste I shipping containers. If the preceding is not reasonable, the DAW may be shipped for direct burial. (7) HEPA Filters f-l Engineering controls such as High-Efficiency Particulate Air (HEPA) filtemd ventilation will be required to capture potential aisorne contaminants. Spent .HEPA fdters will be changed out and treated as dry active waste (DAW) radioactive waste, l I l
- (8) Demineralizer Resins and Filters 1
i Radioactive waste treatment systems will be required to process the liquid waste stream i resulting from various decommissioning activities as described above. 1 Filtration and ion exchange processing will be used to remove residual radioactivity in the water. Temporary deminerahzation and filtration systems may be supplied by a 3-52 Rev.O l
i l l SAXTON NUCLEAR EXPERIMDfTAL CORFORATION b DECOMMISSIONING FLAN V , l vendor or by GPU Nuclear. The volume of spent resins and filters required to process the water is determined to be less than 400 cubic feet. Resins generated by water processing systems will be in, or placed in, approved disposal l cc,ntainers dewatered and transported for disposal. 3.3.2.4 Methods for Monitoring Releases of Radioactive Materials The me.thods to be used to monitor releases of radioactive materials (effluents) during the Saxton facility decommissioning will consist of those methods used by GPU Nuclear in the present defueled plant condition, supplemented with additional provisions as necessary. The instrumentation, setpoints, and calculation methodologies specified in the Offsite Dose Calculation Manual (ODCM) are intended to ensure that the limits of 10 CFR 20 are not exceeded. l l l The measures described above will be supplemented through the use of portable radiation l monitoring instrumentation where necessary. For example, continuous air monitors (CAMS) and j portable alarming area radiation detectors will be used when required by facility operations or conditions. i 3.3.2.5 Method of Estimating Types, Amounts and Radionuclide Concentrations of Radioactive Waste Generated During Decommissioning The following provides a description of aspects of the decommissioning as they relate to radioactive waste disposal. Included in this section are descriptions of the method that was used to derive the types and amounts of waste generated and radionuclide concentrations. O) i v 3-53 Rev.0
l l sAxmN NUCutAR EIMtBENTAL CORMtAMON DECOROMBIONING FLAN l The determination of total radioactivity present in the Saxton facility were derived from field radiological measurements, supplemented by analytical data. The activity present within internally contaminated piping and on plant structures was determined by radiological surveys performed by the Saxton facility staff in accordance with station procedures. The radionuclide inventory present within activated hardware internal to the reactor pressure vessel was determined by TLG Services, Inc.(Refemnce 5). This analysis was performed with the ORIGEN2 (Reference 27) and ANISN (Reference 28) computer codes obtained from Oak Ridge Nationallaboratory and two codes written by TLG Services. The activation analysis as well as each of the surveys performed are described in detail in the Site Characterization Report (Reference 10). The characterization report indicates the level of activity and whether the activity is present as surface contamination or within activated hardware. This information will be used to determine O the field processes that will be used to remove the radioactive material. Given these two parameters - the total radioactivity present and the means of mdioactive material removal - an determination of expected radioactive waste volume and its associated radionuclide concentration was developed. I Conservative determination of total activity (Curies), radioactive waste volumes, average activity l concentration and 10 CFR 61 waste classifications are included in Table 3.3-2. For the purposes of establishing the conservative radioactive waste volume presented in that table, no credit was taken for volume reduction techniquer or for packaging efficiency. l It is not anticipated that any of the radioactive material will be contained in a sealed capsule which would meet the definition of "special form" radioactive material. For this reason all radioactive material discussed is considered to be "nonnal form."
- 3.3.2.6 Radioactive Waste Handling a
3-54 Rev.0 .
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i f SAX 1DN NUCIRAR EXHGtthD!NTAL COItMMLATION DECOMMIIBIONING FLAN
; The methodology which will be used to handle radioactive waste generated during the ,
decommissioning project and for the packaging of waste is described below. i j The waste classification and the volume of each waste class expected to be generated during
. decommissioning is provided in Table 3.3-2. Specific measures used to classify waste are addressed in GPU Nuclear's mdwaste procedures which have been developed to ensure compliance with 10 CFR 20,10 CFR 61 and 49 CFR Parts 170-178.
4 i In general, the large components will be removed and prepared for shipment as soon as ! practicable. Surface contaminated objects will be handled in accordance with Saxton facility procedures which have been developed to ensure compliance with 10 CFR 20. The Saxton facility Radiological Control Programs are described in Section 3.2. I. All waste generated will meet the mquirements set forth in Department of Transponation ( l Regulations 49 CFR 173 Subpart I and 10 CFR 71 for the packaging and transport of radioactive , materials. Asbestos waste material is present. Contaminated asbestos waste will be generated by decommissioning or preparatory activities. Contaminated asbestos will be properly labeled and packaged for disposal per GPU Nuclear procedures. In some areas the asbestos may have been previously removed for personnel hazard reduction and to provide access for radiological characterization. Asbestos removal will be accomplished as early in the project as possible. Industry proven methodologies vill be employed to ensure the separation of contaminated and l non-contaminated materials during the decommissioning process. These methodologies will include the establishment of radiological controls consistent with the Radiological Controls Program and the implementation of Institute for Nuclear Power Operations (INPO) good practices as incorporated into GPU Nuclear procedures. The guidance to be provided for station !
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. 3-55 Rev.O
- SAITUN NUCLEAlt EIFEstBII!NTAL CORFOstATION DECOMMISSIONING FLAN monitoring of radioactive materials is provided within Saxton facility procedums. l The waste stmam(s) resulting from the Saxton facility's decommissioning are similar to that resulting from nuclear power plant operations and maintenance. There am no regulatory transportation issues specifically related to the decommissioning of the Saxton facility which are not provided for in existing procedures.
Section 3.3.4 provides a description of the Decommissioning Support Building to be erected adjacent to the Containment Vessel. 3.3.2.7 Preparatory Conditions . The following conditions are assumed to be in place to suppon the stan of dismantlement. However, any missing factor does not preclude the start of work. l l A. All accessible asbestos insulation has been removed to conduct the decommissioning j asbestos free with some minor exceptions where insulation is inaccessible. l l B. The following temporary services have been installed: I
- 1. A source of water to support decontamination needs.
- 2. Sufficient electrical capacity in the Containment. (" Sufficient" may vary during the decommission effort.)
- 3. A Heating, Ventilating and Air Conditioning (HVAC) system with HEPA filtration. The Containment will be cooled to the extent reasonable with respect to human factors and the tasks to be performed.
l l 3-56 Rev.0
i SAXTON NUCLEAR EIFE3tIMENTAL 00RFORATION
\ DECOMh0SSIONING FLAN l
- 4. Temporary air systems to support tool use and breathing air. These may 1 operate fmm the same compressor and distribution header and will be i designed to provide " breathing quality air" as required by applicable standards and regulations.
1
- 5. A temporary water processing system.
1
- 6. Temporary / Portable decontamination facilities / equipment 4
C. The Decommissioning Support Building (DSB) is in place. This building: i 1. Supports the packaging and staging of radwaste awaiting shipment . s
- 2. Provides storage for contaminated tools, equipment, laundry, etc.
$ 3. Is physically connected to the Containment Vessel (including ventilation) and can accommodate handling of all components except the reactor vessel and steam generator. l D. Temporary facilities to support a work force of 40-45 people. l 1 3.3.3 Radioactive Waste Disoosal 3.3.3.1 Waste Characterization i Existing GPU Nuclear pmcedures used for waste handling pmcessing and characterization will be adapted for use as required, utilizing appmval contmis thmughout decommissioning. In addition, isotopic analyses, waste characterization computer codes and activation analyses are t 3-57 Rev.O
1 I I i SArmN NUCurAR EEFERIRENTAL CORFORATION
\ DECOMMllEIONING FLAN 4
some of the methods which have been and will continue to be used to characterize the waste j streams nesulting from the Saxton facility's decommissioning. The procedures follow 10 CFR 20,10 CFR 61, disposal site criteria, and other Federal and State regulations. l 3.3.3.2 Mixed Waste i The only known mixed waste at the Saxton facility is in the form of lead shielding and lead paint. All lead with the excotion of that in two (2) activated stainless steel shields should be able to be decontaminated to free release by an offsite pmcessor. There are approximately f 1,100 pounds of lead in the two activated shields. These materials will be encapsulated or otherwise treated by a vendor for ultimate disposal or recycle. I2ad paint will be acceptable for i burial as part of the component that was painted since the fraction of lead available for leaching i appears to be less than the maximum leached fraction allowed.
- O GPU Nuclear's objective is to generate no new mixed waste during decommissioning activities.
i Procedures currently in place for hazardous and radiological waste management are sufficient to provide the assurance that waste will not be generated arbitrarily and that generated wastes l will be pmperly dispositioned. At this time, no processes are planned to be used during ; decommissioning that will create a non-treatable mixed waste. l l l 1 However, in the event that a mixed waste is identified or inadvertently generated, the programs l I and procedures in place for hazardous waste and for radioactive waste establish the i
- responsibilities, controls and practices necessary to appmpriately handle the waste. These GPU Nuclear pmcedures pmvide the means to minimize adverse conditions as a result of the mixed waste and manage the containment of the waste.
i 3-58 Rev. 0
n SAXTON NUCIEUt EXPEltihGNTAL CORPORATION
/ i DECOMMISSIONING FLAN U
1 3.3.3.3 Radwaste Processing l Radwaste processing will be performed in accordance with approved procedures. These procedures will be patterned after those used at the GPU Nuclear operating facilities. I 1 I l 3.3.3.4 Radwaste Shipping / Transportation Transportation of radwaste will be in accordance with applicable NRC and DOT regulations and plant procedures. Radioactive waste and material will be shipped either by truck including open and closed transport, trailer mounted shipping cask or by a combination of truck and rail. j
- Shipments will be planned in a practical and efficient manner. Facility procedures will be used with appropriate quality oversight to ensure the shipments are in compliance with company !
procedures, regulations and the receiving site license as described in Section 7.0. Table 3.3-1 provides a list of procedures that give the necessary guidance to ensure handling, packaging, shipment, and disposal of radioactive material are performed within compliance and safely. l Packages, packaging, and labeling for radioactive materials and waste will meet all applicable ] regulations and requirements. I I Due to the size of some of the components in both volume and weight, it will be necessary to
- acquire over dimension permits to travel on public highways. GPU Nuclear will work with the l responsible carrier to ensure permits are obtained. It is anticipated that the reactor vessel will be shipped by truck to the closest appropriate rail siding for shipment to the disposal facility.
i GPU Nuclear will also work with both the tmck and rail carrier to ensure a safe shipment. It is possible that due to the heavy load of shipping the reactor vessel that some of the roads and bridges will have to be improved. GPU Nuclear will work with the appropriate organizations and authorities to make the transportation route safe for passage to the destination. 3-59 Rev.O l 1
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SAIMN NUCLEAR EXPE3tBENTAL CORMMLATION [~ s DECOMMISSIONING PLAN Approved computer codes will be used to pmpam necessary shipping and disposal documentation. 3.3.3.5 Radwaste Storage Extended storage of radwaste is not under consideration. GPU Nuclear expects to store surge volumes of radwaste only for shipment consolidation. 3.3.3.6 Pmvious Onsite Disposal l The Saxton facility was not pmviously used for onsite disposal of low-level radioactive waste by land burial, t 3.3.3.7 LLRW Volume Minimization The Saxton facility Decommissioning Plan conservatively includes determinations of radioactive
- waste volumes which take no credit for volume reduction. However, GPU Nuclear management intends to utilize good industry practices mgarding waste minimization in an effort to be cost effective.
4 Specific measums to be used to ensure the minimization of LLRW volume will include, the use j of the following technologies as appmpriate and available. Both in-house and vendor supplied efforts will be utilized on and off site. f Sorting /Segmgation Super-Compaction
- Incineration
- Metal melting
- i O b
3-60 Rev. O i
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- SAXTON NUCIE.AR EXPERIMENTAL CORPORATION DeCOMRGSSIONING FLAN Decontamination Control of work practices 4
l Void minimization in packaging I
- Off-site only Qualified vendor (s) will be utilized to decontaminate / volume reduce contaminated components.
his work will be performed primarily at the vendor facility, with some notable exceptions i mentioned below. Onsite decontamination of removed components will be primarily limited to that required to support packaging and shipment. Some onsite decontamination capability will be provided for tools, support equipment, etc. This will be accomplished through the use of vendor supplied station (s). Decontamination of the upper dome of the Containment stmeture and the polar crane will be performed onsite to the extent possible. Some crane components may require offsite processing. It is the intent to free release as much material as possible. That which is not feasible to decontaminate will be treated as mdwaste. Debris will be recycled / reused to the maximum economic extent reasonably achievable. 3.3.3.8 Offsite Disposal Options GPU Nuclear will continue to monitor and evaluate off-site low level radioactive waste (LLRW) disposal and storage options prior to and during the decommissioning process. The Saxton facility currently has access to two operating LLRW disposal sites - Barnwell, South t a 3-61 Rev.O
s SAX 1DN NUCULAR EXPERIMENTAL CORPORATION DECOMMISSIONING FLAN Carolina and Envirocare, Clive, Utah. nese two facilities are assumed to remain available to the Saxton facility decommissioning project. The Appalachian Compact disposal site is assumed not to be in operation during any phase of the decommissioning. Based on these premises, radioactive wastes which meet the Envirocare facility criteria will be disposed of there, whenever possible. De remainder, including the reactor vessel, will be disposed of at Barnwell. l The following represents the current assumptions for radwaste disposal: 1 A. The Envirocare facility will accept all soil and most (approximately 80%) of the contaminated concrete volume. 1 B. Voids in components will be filled to at least 85 % for disposal. O C. The steam generator and pressurizer will be certified as packages for shipment and i disposal if they are not otherwise processed (melted, decontaminated, etc). ! D. The reactor vessel will be disposed of as described in section 2.2.1.4. i f 3-62 Rev.O
I SAXTON NUCIEAR EXPERIMENTAL CORPORATION DECOMMISSIONING PLAN V) 3.3.4 Decommissionine Sunoort Building A Decommissioning Support Building (DSB) will be provided as the main facility for processing and packaging so!id radioactive components and waste generated during the dismantlement of I the Containment Vessel. The DSB is designed to support the packaging and staging of radioactive waste awaiting shipment and to provide for storage for contaminated tools, equipment, laundry, etc. The conceptual arrangement of the DSB is shown in Figure 3.3-1. A separate section of the DSB will serve as the personnel access facility and will be the major egress / ingress to the Containment Vessel. This area will additionally serve as a dress out area where personnel will be able to dress out in protective clothing and perform frisking after exit in a low dose, contamination free zone. f% i 6 The DSB will be located adjacent to and connected to the Containment Vessel. It will V accommodate the handling of all components except the reactor vessel and steam generator. Routine personnel access to the Containment Vessel will be provided through the present CV access door. An additional opening will be provided for at the DSB connection to allow for the transfer of material. A ventilation system will be designed to ensure air flow is directed from clean to potentially contaminated areas, out through a HEPA filtered unit, exhaust fan and an effluent radiation monitor. O 3-63 Rev.O I
l l \ SAXTON NUCIEAR EXPE3tIMENTAL CORPORATION DECOMMISSIONING PLAN TABLE 3.3-1 EXAMPLES OF PLANS AND PROCEDURES FOR RADIOACTIVE WASTE SHIPPING AND DISPOSAL COMPLIANCE Number Title 1000-PLN-7200.01 GPUN Operational Quality Assurance plan for Three Mile Island Unit I and Oyster Creek 6575-QAP-7200.01 SNEC Quality Assurance Program 1072 Hazardous Material Transportation Incidents 1077 Material Non-conformance Reports and Receipt Deficiency Notices 6575-ADM-4500.29 Radioactive Material Identification and Handling 6610-ADM-4400.03 Surveying of radioactive waste packages t 6610-ADM-4450.01 Curie estimates for radioactive material shipments 6610-PMI-4220.01 Conduct of calibmtion, maintenance and repair of radiological instruments 1104-28F Packaging non-routine radioactive waste l Il04-28J Off-site shipment of Non-waste radioactive material l 1104-280 Radwaste nuclide distribution and conversion factors 1104-28Q Mixed low-level radioactive waste control program 1104-28S Packaging of radioactive material for shipment off-site to a volume reduction vender 1104-28T Waste oil transfer, recirculation, and storage system 1104-63 Control of radioactive materials 6213-PGD-2640 Radioactive material transportation training program unit 3-64 Rev.O
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s TABLE 3.3-2 AVERAGE AVERAGE ACTIVITY uCi's PER CONCENTRATION 100 cm ^2 ESTIMATED VOLUME IN WASTE COMPONENT (uCi/cc) FOR SCO CURIES CUBIC METERS CLASSIFICATION AMEA 1 !+ E919 e6 +++W+ we++m-se Reiss ee sedasEi SHUT DOWN COOUR6TIEAT EXCHANGER 4.26E +00 N/A 9.00E-01 0.21 TBD SHUT DOWN COOUNG PMPS 5.93E-01 N/A 2.52E-01 0.42 Class C CONDENSATE RETURN PUMP 3.00E-03 N/A 1.00E-04 0.03 Class A CONDENSATE RETURN TANK 1.13E-03 N/A 1.00E-04 0.09 Class A DISCHARGE TANK 1.35E-01 N/A 6.27E-01 4.63 Class A DISCHARGE TANK PMPS 4.36E-02 N/A 9.00E-03 0.21 Class A SUMP PUMPS _ _ __ __ 2.68E-01 N/A 6.70E-03 0.03 Class A STORAGE WELL DEMIN. SUCTION FILTER 3.82E-02 N/A 3.26E-03 0.09 Class A PURIFICATION FILTER 2.21 E-01 N/A 1.00E-03 0.005 Class A STORAGE WELL DEMIN. DISCHARGE FILTER 8.06E-02 N/A 2.54E-04 0.003 Class A CCW PUMP FILTER 1.40E-01 N/A 3.00E-03 0.02 Class C ROD ROOM AIR HANDLER N/A 2.3E-02 TBD 2.48 Class A ROD ROOM VENT FAN N/A 2.1 E-02 TBD 0.03 Class A CONTROL ROD DRIVES N/A 1.2E-01 TBD 0.05 Class A INSTRUMENT RACK 1 2.65E-02 N/A 3.59E-02 1.35 Class A INSTRUMENT RACK 2 5.71 E-03 N/A 7.00E-03 1.23 Class A NEW FUEL STORAGE RACKS N/A 3.7E-02 TBD 5.05 Class A REACTOR HEAD INSULATION RING N/A 3.6E-01 TBD 5.45 Class A PUMP LEAKOFF DRAIN TANK 2.78E-02 N/A 6.60E-03 0.24 Class A CONTAINMENT VESSEL SUMP VOLUME _ 7.38E -02 N/A 3.90E-02 0.53 Class A AREA 2
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~
~ '
b E t 'Y STKK~M 'GERnHkTOR 6.~DTET00 M/A 4.liOE + 01 7.12- Ciass C PRESSURIZER 3.86E +00 N/A 2.47E + 01 6.41 Class C MAIN COOLANT PUMP 9.44 E-02 N/A 5.00E-02 0.53 Class A REGEN HEAT EXCHANGER 3.83E + 00 N/A 3.10E + 00 0.81 TBD NON-REGEN HEAT EXCHANGER 4.30E-01 N/A 1.00E-01 0.23 Class C INSTRUMENT RACK 4.71 E-03 N/A 1.09E-02 2.32 Class A l AIR HANDLER IN OVERHEAD _{s_ee Primary) N/A 1.1 E + 00 TBD 0.68 Class A PRIMARY AIR HANDLER N/A 1.1 E + 00 TBD 3.60 Class A TBD - To Be Determined At A Later Time. SCO - Surface Contaminated Object. l l 3-65 Rev 0
m x TABLE 3.3-2 AVERAGE AVERAGE ACTIVITY uCi's PER CONCENTRATION 100 cm^2 ESTIMATED WASTE COMPONENT (uCi/cc) (SCO) CURIES CUBIC METERS CLASSIFICATION p--- (
-AREA 3 !c - - Ma + 6 ' - ntli STORAGE WELL HEAT EXCHANGER 3.98E-02 N/A 4.00E-03 0.10 Class A STORAGE WELL PUMPS 4.01 E-03 N/A 2.00E-03 0.50 Class A COMPONENT COOUNG HEAT EXCHANGF9 9.48E -03 N/A 9.50E-03 1.00 Class A COMPONENT COOUNG STORAGE tat..s 1.32E-02 N/A 1.30E-03 0.10 Class A COMPONENT COOUNG PUMPS 8.71 E -03 N/A 3.70E -03 0.42 Class A INCORE INSTRUMENT DRIVES N/A 1.8E-01 TBD O.93 Class A AUXIUARY AIR HANDLER N/A 3.6E-01 TBD 0.74 Class A OPERATING FLOOR AIR HANDLER N/A 6.3E-01 TBD 5.62 Class A AVERAGE AVERAGE ACTIVITY uCi's PER CONCENTRATION 100 cm ^ 2 ESTIMATED WASTE CO MPON_ENT (u_C_i/cc) (SCO) CURIES CUBIC METERS CLASSIFICATION l AREA 4 il r .
se ~
- 1. M POLAR CRANE N/A l 8.7E-02 TBD 113.27 Class A REFUEUNG BRIDGE N/A j 2.4E-02 TBD 0.68 Class A SMALL C_ONTAMINATED PUMP 2.56E-03 N/A 1.00E-04 0.04 Class A LARGE CONTAMINATED PUMPS 6.23E-04 N/A 2.00E -04 0.32 Class A '
ELECTRIC AL DISTRIBUTION BOX N/A 1.6E-02 TBD 0.91 Class A EQUIPMENT HATCH TROLLEY N/A 4.2E-02 TBD 4.58 Class A I TOOLRACK N/A 8.4E-02 TBD 0.19 Class A FUEL HANDLING TOOLS N/A 8.4E-02 TBD 0.10 Class A TELEFLEX SHIELD (Steel Section) N/A 1.5E-03 TBD 4.25 Class A FUEL TRANSFER CASK SKID N/A 1.8E-03 TBD 2.98 Class A HEAD STAND N/A 1.8E -02 TBD 1.16 Class A UGHTS N/A 8.7E-02 TBD 0.70 Class A AIR CIRCULATING FAN N/A 5.0E-01 __TBD 0.51 Class A TBD - To Be Determined At A Later Time. SCO - Surface Contaminated Object. 3-66 Rev 0
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[N ( )
'V v' s/
TABLE 3.3-2 AVERAGE AVERAGE ACTIVITY uCi's PER CONCENTRATION 100 cm ^2 ESTIMATED WASTE COMPONENT (uCi/ce) (SCO) CURIES CUBIC METERS CLASSIFICATION , i! AREA 6 _ __._m ~~
. m - ^ ^s r9 Ee%%A m, s 91l
" 1.4sn+63 FiKA~C'T UR - Ve$$EL 1.T3E + 02 N/A 12!55 Cias~ s C REACTOR VESSELINSULATION CAN 1.21 E-01 N/A 1.08E + 00 8.95 Class A REACTOR VESSEL SUPPORT CAN 1.66E-01 N/A 3.03E + 00 18.26 Class A SPENT FUEL RACK 6.22E-04 6.0E-01 7.00E-03 11.25 Class A BORIC ACID DEMINERAUZER VESSEL 4.15E +01 N/A 1.74E +01 0.42 TBD STORAGE WELL DEMINERAUZER VESSEL 2.74E + 00 N/A 4.30E-01 0.16 TBD PURIFICATION DEMINERAUZER VESSEL 1.15E + 00 N/A 4.80E-01 0.42 TBD SPENT FUEL POOL ELEVATOR N/A 6.8E-01 TBD 2.72 Class A SUPPORT STAND N/A 6.8E-01 TBD 0.45 Class A INTERNALS RIGGING FIXTURE N/A 6.8E-01 TBD 0.51 Class A REACTOR VESSEL HEAD UFT RIGGING N/A 6.8E-01 TBD 0.51 Class A SUPER HEATED TEST EQUIPMENT TANK 2.05E-02 N/A 7.28E-03 0.36 Class A lilllllllliimmilmmIlmiilllBillil!!!!!ImitilllllillIIIMIllilillillIllIMIIm111l1111111111111111115111!!!!l111111111111111111111111111111111111llll111llll111II[1IM11E111111 GRATING (all areas)
.. : 0.004 . ,
(Steel Only) ... WTl M ^ 3/ft ^2 - aWA*y4fC wwwww w, emmyymmewww . t..-.- %" ' 2.9E-01 1.46E-02 5.74 l Class A
+
- Average Total Total M ^ 3 lik - =p:me w GRATING (Rx Compartment) ' - '
W , T 0.004 % _ - Q? (Aluminum Only) ~ .. W~ %s -
-9 . O _. i hl M ^ 3/ft ^2 #(M ^'z m' 1 ymmrweg;
%. ~ , m--
gem-mweree*xmmweg . p.g?[ L.s 4 8:5E-01' 1!i6E-02 1.56 Class A
, c.
STEP GRATING (all areas) m, gga. .g;gg a ggg 0.002) M.g (Steel Only) JM4r C' ~ , X L J-
+ M ~' M ^ 3/Ir ^2 Fe@";395N ggy@g. W emm . .m .. ... . . emp . m=mp~er * '
519 6 0i ~ 7.99E-04 07f7 Class A p: ~, m:- <rx - a ~.;pg A verage___ Total Total M ^3 >
. : %y SITE PIPING 1.92E + 00 +M4 m 1.34E +01 7.0 Class A & C INSULATION -
TBD _ . M NM TBD 17.2 Class A PIf06E5~S WASTE GESERdTED TBD M - Included TBD TBD RV INTERNAL SURFACE CONTAMINATION N/A /# 1.18E + 01 N/A Included With RV [ TOTALS ^ P: 4 s1 1569 579lp%hdfEnnS4ll TBD - To Be Determined At A Later Time. SCO - Surface Contaminated Object. Note: The Volume Estimate Takes No Credit For Decontamination, Volume Reduction Or Packaging Arrangements. 3 -67 Rev 0
3 no. 3.>.i E REACTOR EQUIP. HATCH
\
I CONT AI NMENT VESSEL
$ ""32" J $'
i [ l P.A.F. s,% / ,_ mQ HATCH
" x
-I ,
- O f
A DECOMMISSIONING s i SUPPORT FACILITIES $ .g a
- (CONCEPTUAL DRAWING) k unwoon 8 l
1 f 10 20 30 , , h SCALE (FEET) h ROLL-UP DOOR TRUCK LOADING AREA
_- . . - . _~ n SAXTON NUCLEAR EXPERIMENTAL CORPORATION (b DECOMMISSIONING FLAN i < 3.4 ACCIDENT ANALYSES Various radiological accident scenarios during the dxommissioning of the Saxton facility have been postulated and examined. The analyses discussed herein used very conservative approaches in treating the source terms, as well as in the methods of calculation. To the extent applicable, these analyses are consistent with approaches used in the NRC's examination of postulated accidents during the decommissioning of the Reference PWR (Reference 14). 3.4.1 Introduction 4 The EPA has established pmtective action guidelines (Refemnce 16) that specify the potential l off-site dose levels at which actions should be taken to protect the health and safety of the public. The EPA pmtective action guidelines (PAGs) are limiting values based on the sum of the effective dose equivalent msulting fmm exposure to external sources and the committed effective dose equivalent incurred fmm the significant inhalation pathways during the early phase of an event. The EPA PAG limits are: EPA PAGs (millirem) k Total Whole Body (TEDE) 1000 3 j Thyroid Committed 5000 Dose Equivalent (CEDE) Skin (CDE) 50,000 Since there is no irradiated fuel stored at the Saxton site, there are no radioactive noble gases or radiciodines available for mlease from the site. This preempts the possibility of accidental 3-68 Rev.O
SAITON NUCLEAll EXEE3tIMENTAL CORPORATION
- DECOMMISSIONING FLAN 1
off-site radiological releases that could approach the PAGs for the skin and thyroid. As a result, the PAG for TEDE is the limiting criteria for decommissioning activities at the Saxton facility. 4 GPU Nuclear has analyzed the decommissioning activities described in this plan to ensure that they will not create the potential for accidental releases that could cause doses at the site boundary to be more than a small fraction of the EPA PAGs. Performing decommissioning
- activities in a manner that keeps off-site doses from even the most unlikely events at a small l; fraction of the EPA PAGs provides for the protection of the health and safety of the public
- without the need for protective actions.
The accident analyses demonstrate that no adverse public health and safety or environmental i l-impacts are expected from accidents that might occur during the Saxton facility's
- decommissioning operations. The highest calculated dose to an individual located at the site ~
f boundary was less than 1.5 miem to the whole body during a postulated materials handling ! accident . This highly conservative, unrealistic scenario is further described in Section 3.4.1.1. The results of other on-site accidents are below this value. The limiting accident case represents ] less than 0.14% of the EPA lower whole body dose limit. As a result, it is concluded that there are no significant radiological consequences to the general public from postulated credible l accidents during the planned decommissioning operations at the Saxton facility. i 3.4.1.1 Material Handling Accident - Dropped Resin Vessel f ( This accident scenario assumes that the steel demineralizer vessel containing the remaining used resins is dmpped during removal from the containment building. This was considered to be the worst case material handling accident, since analysis of a drop of the steam generator and the pressurizer using similar assumptions resulted in less offsite dose. Dropping of the reactor pressure vessel was not analyzed since it would be highly unlikely to rupture during a materials handling accident due to the nature of its construction. The residual activity in the resin vessel 3-69 Rev.0
SAXTON NUCLEAR EXPEILIMDfrAL CORPORATION
-( DECOnOGSSIONING FLAN <v i
is determined to be 17 curies. The nuclide mixture is primarily composed of Co-60 (5.4%), Ni-
] ; 63 (29.9%), Sr-90 (1.8%), Cs-137 (9.5 %), Pu-238 (1.1 %), Pu-239 (3.1 %), Pu-241 (43.8 %),
and Am-241 (3.5 %). When the vessel is dropped, it is assumed to split open, releasing 1.7X104 of the activity in the vessel to the atmosphere. De release fraction of 1.7X10' is considered
, to be conservative based on the following:
J : 4 i Reference 14 describes a release fraction of 1.7X10 for a fue or explosion in ion exchange resins. Dropping the resin vessel would
- provV; far less motive force for releasing activity than a fue or 1
4 explosion. I l Prior to movement, the resin vessel will be filled with grout. As ! a result, the residual activity in the vessel will be more immobilized than would be the case in a normal vessel of resin. v i j No credit is taken for filtration by the HEPA ventilation since it is hypothetically possible that such an event could occur outside the containment building. A total of 28.9 Ciis released j from this accident over an assumed two hour period. I An atmospheric dispersion factor (X/Q) of 4.14X103 sec/m8 is used to calculate the airborne 1 l activity concentration at the site boundary (200 meters) in accordance with Reference 15. This ! conservative value is calculated for a 1 m/s wind speed and a G stability category in accordance ! with Reference 15. Off-site doses are calculated using the parameters and methodology of ] Reference 16. The whole body dose to an individual standing at the site boundary for the , duration of the release is calculated to be less than 1.5 mrem. This is a small fraction of the EPA Protective Action Guide of 1000 mrem for the whole body. Derefore, the container drop accident poses no serious risk to the general public and has no significant environmental impact. i O 3-70 Rev. 0
SAX 10N NUCIEAR EXPERInG!NTAL CORFORATION DECOAUdISSIONING FLAN l \l 3.4.1.2 Fire - Combustible Waste Stored in the Yard l This accident scenario assumes that a SeaLand Van of combustible waste materials is completely consumed by a fire while stored in the yard area of the Saxton facility. This was considered to i be the worst case fire, since the waste is stored outside the containment building and releases would not be contained by building confines or HEPA ventilation systems. The activity in the van is assumed to be 1.79 curies. This amount of activity in the van is 99.8% of the Type A LSA limit for this type of container, which is the maximum shipping class to which such
. containers can be loaded. The construction of higher level shipping containers would prevent the release of significant quantities of activity during a fire. They are also far less likely to be involved in a fire. The nuclide mixture is primarily composed of Co-60 (43.7 %), Ni-63 (0.8 %),
Sr-90 (0,1 %), Cs-137 (54.9 %), Pu-238 (0.02 %), Pu-239 (0.05 %), Pu-241 (0.2 %), and Am-241 (0.08%). At the IJSA limit for a van, this type of contamination pmduces the highest off-site doses of all loose surface contamination characterized in other areas of the building. The maximum fractional airborne release measured during burning of contaminated wastes under similar conditions was 1.5X10", in accordance with Reference 14. No credit is taken for filtration by the HEPA ventilation since it is assumed that the fire occurs in the yard area. A total of 269 Ci is released fmm this accident over an assumed two hour period. Using the same mem:.ological assumptions and dose calculation methodologies as the analysis in Section 3.4.1.1, the whole body dose to an individual standing at the site boundasy for the duration of the release is calculated to be less than 0.3 mrem. This is a small fraction of the EPA Protective Action Guide of 1000 mrem for the whole body. 'Ihe fire accident poses no serious risk to the general public and has no significant environmental impact. l 3.4.1.3 Vacuum Filter-Bag Rupture Sharp objects, such as metal shards, could rupture a filter-bag during surface decontamination
\
3-71 Rev.O
SAITVN NUCIEut EIFERRENTAL CORNMtATION DECOMhGSSIONING FLAN j operations involving the use of a vacuum cleaner. To maximize the calculation of the atmospheric release, the bag mptum is assumed to occur at the time just prior to the bag change 2 (i.e., when the filter bag is full). It is assumed that the vacuum is used to vacuum 2600 m of floor area prior to the bag being changed out, per Reference 14. It is assumed that the average loose surface contamination level on the floor being vacuumed is 3X10 6 dpm/100 cm2 . This is the highest loose surface contamination level identified in the containment building, on the floor of the spent fuel pool, elevation 765'-8". Loose surface contamination levels in the majority of the containment building are orders of magnitude less than this area, so it is believed that this I assumption provides a highly conservative estimate of the airborne activity gen'e rated during this , scenario During the vacuuming process, it is assumed that 50% of the loose surface activity
- on the t rea being vacuumed is removed by the vacuum and collected in the bag per Reference i 14. As a mesult, a total of 0.176 Ci of activity is assumed to present in the bag when the mpture occurs. The nuclide mixture is primarily composed of Co-60 (43.7%), Ni-63 (0.8%), Sr-90 C (0.1 %), Cs-137 (54.9%), Pu-238 (0.02%), Pu-239 (0.05%), Pu-241 (0.2%), and Am-241 i (0.08%). When the filter bag is mptured, all of the collected activity in the bag (0.176 Ci) is assumed to become airborne in the building because of the mechanical and aerodynamic forces of the vacuum cleaner air flow. Since decontamination activities at the Saxton facility will only
- be performed while the building ventilation system is operable, it is assumed that the airborne activity will be collected by the building ventilation system and discharged to the environment through HEPA filters (99.95% efficient per Reference 14). No credit is taken for plateout of particulates on building surfaces or ductwork. A total of 87.8 Ci is assumed to be discharged to the environment, Using the same meteorological assumptions and dose calculation methodologies as the analysis in Section 3.4.1.1, the whole body dose to an individual standing at the site boundary for the duration of the release is calculated to be less than 0.09 miem. 'Ihis is a small fraction of the EPA Protective Action Guide of 1000 mrem for the whole body. The vacuum filter-bag mpture accident poses no serious risk to the general public and has no significant environmental impact.
3-72 Rev.0 4
4 i SAITVN NUCIEAR EIFI!3tIMDfrAL CORFOItATION DECOMMISSIONING PLAN 3.4.1.4 Segmentation of Components or Structures Without or During loss of Local l Engineering Controls
- Segmentation of components or structures can be accomplished by disassembly , cutting, or othe destructive methods. Disassembly of components or structures does not result in destruction of material. 'Ihe potential for radioactive material release is limited to dislodging contamination. Disassembly events are therefore considered bounded by the material handling
) event discussed in Section 3.4.1.1. i The dismantlement of RCS piping is considered to provide the bounding analysis for generation of ai6orne activity, since it is anticipated that the reactor vessel will not require segmentation i for removal. While activated components like the teactor vessel contain the greatest activity 1 levels, the transuranic content of surface contamination in RCS components at the Saxton facility O is the dominant factor in producing offsite doses. As a result, surface contamination of piping in the Safety Injection Piping was used to represent the maximum activity available for release , during segmentation of RCS components. This piping was chosen since it had the highest transuranic content of all piping samples collected during the Saxton facility radiological characterization.
- 'Ihe guidance provided in Reference 14 was used to determine the amount of activity that could be generated during a segmentation cut. To determine the to+al activity generated from a segmentation cut, the following equation was used
i, Total Activity Generated = (Surface Contamination Level)(Kerf Width)(Pi X length of Pipe ID) To determine the maximum generated activity the following values were used: Surface contamination samples from the safety injection piping showed an activity of 208 3-73 Rev.0
l l i SAX'IDN NUCIEut EXPEtDWWTAL CORPORATMM
- f. ~
( DECOMMISSIONING PLAN U Ci per gram. It was assumed that this activity was imbedded in the first 1/16" layer of the piping. The density of stainless steel is assumed to be 8 g/cc (Reference 14). The mass of a 1 cm2 area of piping,0.159 cm thick is 1.27 g. As a result, the activity per 2 unit area in the pipe is assumed to be 208 Ci/g X 1.27 g/cm2 , or 264 pCi/cm , The kerf width used was 0.95 cm. This is conservative since it is the largest kerf width of possible cutting methods that may be employed (Reference 14). 1 De diameter of the pipe assumed to be cut is 78.7 cm (31.7 inches) per Reference 14. I nis assumption represents the longest segmentation cut that would be performed before the release is detected and segmentation secured to terminate the generation of airborne 1 activity. This is a highly conservative assumption, since continuous air monitors located in the area would alert personnel to the release long before the full length of pipe was p cut. Using the above equation, the maximum release to the containment atmosphere is 0.062 Ci. No credit is taken for local engineering controls sir.ce they are assumed to have failed or not be present. The nuclide mixture is primarily composed of Co-60 (17.3 %), Ni-63 (44.0%), Fe-55 (2.5 %), Cs-137 (0.4 %), Pu-238 (1.1 %), Pu-239 (2.4 %), Pu-241 (27.6%), and Am-241 (3.7 %). Since cutting activities at the Saxton facility will only be performed while the building ventilation system is operable, it is assumed that the airborne activity will be collected by the building ventilation system and discharged to the environment through HEPA filters (99.95% efficient 2 per Reference 14). No cmdit is taken for plateout of particulates on building surfaces or ductwork. A total of 30.9 Ci is assumed to be discharged to the environment. 4 Using the same meteorological assumptions and dose calculation methodologies as the analysis in Section 3.4.1.1, the whole body dose to an individual standing at the site boundary for the duration of the release is calculated to be less than 1.5 mrem. This is a small fraction of the O G/ 3-74 Rev.O
gs SAITDN NUCIEAR EXPERIMErfrAL CORPORATION DECOMMISSIONING FLAN ( EPA Protective Action Guide of 1000 mrem for the whole body. The wation accident poses no serious risk to the general public and has no significant en" ict. 3.4.1.5 Oxyacetylene Explosion It is anticipated that segmentation of the reactor pressure vessel will not be required. However oxyacetylene torches may be used to segment RCS piping systems and other piping systems l within the containment building. For the purposes of this accident evaluation, it is assumed that reactor coolant system pipe cutting will be performed using oxyacetylene torches. It is assumed I that the acetylene is stored in an area that does not contain radioactivity, so there is no 1 mdioactive release potential from a postulated storage accident. I Violent explosions can occur when acetylene and oxygen are incorrectly mixed. The degree of
/
explosive violence depends on how closely the gas mixture approximates the ratio for complete combustion. Oxyacetylene explosions can occur from such causes as flow reversal, nozzle obstmetions, or flashbacks. This accident is postulated to occur during cutting of the RCS : piping. It is conservatively assumed that all the RCS piping has the same radiological characteristics as the safety injection piping. This piping was chosen since it had the highest transuranic content of all piping samples collected during the Saxton facility radiological characterization. In addition, it is anticipated that such piping would be one of the more highly ; ! activated piping sections due to it's proximity to the reactor. It is assumed that cutting of this piping system would be performed within a portable ventilated enclosure. It is assumed that all i the filters contained within the portable enclosure are damaged and release all of their contents i 1 to the containment building atmosphere. It is further assumed that there are ten filters and the l 1 I accident occurs when the filters are fully loaded. The mass of material that can be deposited on enclosure HEPA filters without causing serious operational problems, such as excessive pressure drop, varies considerably with the filter O 3-75 Rev.O
~s SAX'tDN NUCLEAR EXPERIMENTAL CORMMtATION DECOMMISSIONING PLAN
[ construction and particle size of the deposited material. In this accident, it is assumed that 2.3 kg of material is deposited per filter (Reference 14), and all of this material is released into the containment building during the explosion. To maximize the results, it is also assumed that about the same amount of material on the walls and floor of the enclosure is also released due to the explosion. As a result, a total of 46 kg of material with a specific activity of 0.038 pCi/g goes airborne in the containment building during the explosion. 1 J l Using the assumptions above, the maximum release to the containment atmosphere is 0.0018 Ci. The nuclide mixture is primarily composed of Co-60 (17.5%), N-63 (44.6%), Fe-55 (2.5%), Cs-137 (0.4%), Pu-238 (1.1 %), Pu-239 (2.4%), Pu-241 (27.4%), and Am-241 (3.7%). Since j 4 cutting activities at the Saxton facility will only be performed while the building ventilation system is operable, it is assumed that the airborne activity will be collected by the building ventilation system and discharged to the environment through HEPA filters (99.95% efficient ( per Reference 14). No credit is taken for plateout of particulates on building surfaces or ductwork. A total of 0.88 Ci is assumed to be discharged to the environ, ment. Using the same meteorological assumptions and dose calculation methodologies as the analysis in Section 3.4.1.1, the whole body dose to an individual standing at the site boundary for the ! duration of the release is calculated to be less than 0.05 mrem. This is a small fraction of the : EPA Protective Action Guide of 1000 mrem for the whole body. The oxyacetylene explosion accident poses no serious risk to the genemi public and has no significant environmental impact. I 3.4.1.6 Explosion of Liquid Propane Gas (L.PG) Leaked From a Front-End leader l l An LPG powered front-end loader for loading concrete rubble and moving equipment is assumed i to be used to support dismantling operations. An accidental leak of LPG is postulated to occur during the loading of concrete rubble in the containment building. During this accident, it is assumed that the pre-filters and filters in both exhaust filter banks are mptured simultaneously O V 3-76 Rev.O
4
. SArmN NUQAut EIMGLEMNTAL CORPORATION i DECONGSSMNNG FLAN
' (two banks with 50 filters per bank per Reference 14). It is further assumed that the filters are
! fully loaded with contaminated concrete material.
i
- l. The mass of material that can be deposited on HEPA filters without causing serious operational problems, such as excessive pressure drop, varies considembly with the filter construction and particle size of the deposited material. In this accident, it is assumed that 2.3 kg of material is deposited per filter (Reference 14), and all of this material is released to the environment during l
l
'the explosion. To maximize the results, it is also assumed that about the same amount of j material on the ductwork is also released due to the explosion (Reference 14). As a result, a j total of 460 kg of material with a specific activity of 0.014 pCi/g goes airborne to the j environment during the explosion. l l !
i Using the assumptions above, the maximum release to the environment is 6500 Ci. The lr ( a nuclide mixture is primarily composed of Co-60 (0.82%), N-63 (0.01 %) and Cs-137 (99.2%), along with small fractions of Sr-90, Pu-238, Pu-239, Pu-241, and Am-241. i Using the same meteorological assumptions and dose calculation methodologies as the analysis in Section 3.4.1.1, the whole body dose to an individual standing at the site boundary for the duration of the release is calculated to be less than 0.4 mrem. This is a small fraction of the EPA Protective Action Guide of 1000 mrem for the whole body. The explosion of LPG ) accident poses no serious risk to the general public and has no significant environmental impact. i 3.4.1.7 In Situ Decontamination of Systems Iarge scale chemical decontamination of systems is not anticipated as part of the Saxton facility decommissioning. However, limited application may be used on systems or tanks to reduce radiation dose rates prior to dismantlement or general area decontamination. This type of decontamination employs the use of liquid decontamination agents that do not readily become
- \
3-77 Rev.O i
l l l 4 SAXTON NUC12AR EXPEREG!NTAL CORFO JION 1 DECOMMIESIONING FLAN l 4 airborne. Even during a spray release, droplets tend to readily plateout on building surfaces and equipment. Those droplets that remain airborne are readily captured by ventilation filtration systems prior to release to the environment. In addition, they are not instantaneous releases as would be the case with the dropped HEPA vacuum or explos:on events. The nature of this type 6 of event allows for mitigation of the release upon detection by airborne radioactivity monitors, whereas the explosion events previously analyzed do not permit mitigating actions until after the , release has already occurred. As a result, radiological releases from accidents involving in situ decontamination of systems are considered bounded by the explosion and dropped vacuum events analyzed in Sections 3.4.1.3, 3.4.1.5, and 3.4.1.6. 3.4.1.8 loss of Support Systems Electric power, cooling water, and compressed air systems provide support to decommissioning ( activities. Ioss of these systems could potentially affect many other systems and plant areas simultaneously. Each of these events is evaluated below. A. Loss of Offsite Power Offsite power is used to energize tools, cranes, lighting and air filtering equipment used during decommissioning operations. A loss of power to tools and lighting being used for decommissioning will result in an intermption of work activities, but does not result in the release of mdioactivity. A loss of power to plant ventilation and filtering systems could result in the disruption of airflow paths and effective utilization of HEPA filters. In the event of loss of offsite power, work activities with the potential for airborne contamination will be suspended.
- A loss of offsite power could result in loss of power to material handling equipment.
l Occupational Safety and Health Administration (OSHA) regulations require that crane hoisting C l d 3-78 Rev.0
i 4 n SAXMN NUCLEAR EXPERIMENTAL CORPORATION DECOMMISSIONING FLAN units be equipped with a holding brake. A holding brake is a brake that automatically prevents 4 motion when the power is off. Although loss of power is not expected to result in crane or hoist failure, this event would be bounded by the material handling event analysis provided in Section 3.4.1.1. 2 1 B. less of Cooling Water l Cooling water may be supplied to air compressors and the decommissioning cutting equipment i and tools. Cutting operations that use cooling water will stop. This does not adversely affect contamination control. Compressed air will be lost if alternate cooling water is not established in a shon period of time. The consequences of a loss of compressed air are analyzed in Section 3.4.1.7.C. l j A loss of cooling water being used for decommissioning will result in an interruption of work activities, but does not result in the release of radioactivity. Therefore, public health and safety are not adversely affected by a loss of cooling water event. C. Ioss of Compressed Air Compressed air will be supplied by air compressors to power pneumatic tools. Upon a loss of compressed air, decommissioning pneumatic tools shut down. This terminates potential releases < from activities using these tools. A loss of compressed air being used for decommissioning will result in an intermption of work activities, but does not result in the release of radioactivity. Therefore, public health and safety are not adversely affected by a loss of compressed air event. 3.4.1.9 External Events i V 3-79 Rev. O
em SAXTON NUCLEAR EXPEILIMENTAL CORPOILATION
/ \ DECOMMISSIONING PLAN V
A review of external events was done to evaluate the effects of natural and manmade events on the radiological consequences of decommissioning activities. The hazards associated with these events are assumed to be consistent with those that could have occurred with the Saxton facility in operation, which wem evaluated in the Saxton facility FSAR. Such events are of extremely low probability. A discussion for each of the analyzed events follows. A. Earthquake Per the Saxton facility FSAR, them has been only one minor earthquake in the area in the past
- 200 years. In the unlikely event that a seismic event would occur during decommissioning, it could initiate a materials handling accident and/or loss of offsite power. These events have been analyzed in Section 3.4.1.1 and Section 3.4.1.8A and found to pose no serious risk to the general public and no significant environmental impact.
B. Flooding As discussed in the Saxton facility FSAR, the highes: flood level on mcord is 809.5 feet, whereas the site grade level is 811 feet. A flooding event at the Saxton facility would typically be preceded by a sufficient warning period to prepare the site for the event by securing decommissioning activities. Most of the potentially removable radioactivity at the Saxton facility is located in the containment building, above the potential flood height. Most of the balance of contamination would be packaged for shipment. Containers that hold high radioactivity materials l are designed for greater levels of stmetural integrity, providing additional protection. In the l unlikely event that a lower radioactivity container is exposed to flood waters and radioactive material is dispersed, the flooding dilution effect results in a radiological consequence significantly less than an airborne release of a similar amount of radioactive material. I l Flooding could initiate a loss of off-site power event. The analysis in Section 3.4.1.8.A concludes that public health and safety are not adversely affected fmm a loss of off-site power n 3-80 Rev.0 '
1 l l SAX 1DN NUCLEAR EIFERIhDNTAL CORFORATION p osComaSMONING FLAN event. C. Tornadoes and Extreme Winds The annual strike probability of a tornada that could cause a significant release of radioactivity from a container or component is very low. In addition, most components and containers that would be vulnerable to a tornado will be packaged awaiting shipment. The integrity of these containers would limit the probability and consequences of a significant release of radioactive materials. Further consideration of the interaction between a tornado and decommissioning is not warranted. An extreme wind event at the Saxton facility would be preceded by a sufficient warning period to prepare the site for the event by securing decommissioning activities. Most of the potentially airborne radioactivity at the Saxton facility is located in the containment building, which protects the components from the effects of extreme winds as discussed in the Saxton facility FSAR. l Components and containers that would be outside the containment building and vulnerable to extreme winds will be packaged awaiting shipment. Containers that hold high radioactivity materials are designed for greater levels of structural integrity, providing additional protection. In the unlikely event that a lower radioactivity container is unprotected and exposed to extreme winds and radioactive material is dispersed, the combination of low radioactivity content and significant dispersion by wind would result in an offsite dose that is bounded by the limiting release of the material handling event analyzed in l l Section 3.4.1.1. D. Lightning The lightning strike annual probability for a decommissioning activity is very low. Although (\ 3 81 Rev.O
i SAXTON NUCIEAR EXPERIMENTAL CORFORATION [ DECOMMISSIONING FLAN d the effects of lightning are localized, a lightning strike could initiate a loss of off-site power event or a fire. The analyses in Sections 3.4.1.2 and 3.4.1.8.A conclude that public health and safety are not adversely affected by these events. Further consideration of the interaction between decommissioning and a lightning event is not warranted. 1 E. Toxic Chemical Event Toxic chemicals air a personnel safety concern. In the event of a toxic chemical event affecting plant personnel, decommissioning activities would be suspended and personnel evacuated as necessary. A toxic chemical event has the potential to initiate a radiological event. The most severe radiological event that could be initiated would be if a personnel injury resulted in an event involving a loaded crane or hoist. A toxic chemical event is therefore considered as an initiating event for a material handling event which is analyzed in Section 3.4.1.1. . (~ 1 ( F. Intruder Event The cause of this type of event could be an individual fmm the general public breaching the security fence and entering a radiological controlled area. The consequences due to radiation exposure of a member of the public from an unauthorized entrance to a radiologically controlled area are not expected to be significant because of the low levels of radiation and contamination thmughout the plant. Areas with high dose rates (> 1R/hr at 30 cm), will be locked in accordance with plant procedures. Radiation exposures are therefore expected to be low and should not pose a significant risk. A less likely accident scenario was also assumed to involve sabotage by a plant employee or a member of the public, resulting in a fire in a radiologically controlled area. The consequence of an accident involving sabotage such as a fire was analyzed in Section 3.4.1.2. The analysis in Section 3.4.1.2 concludes that public health and safety are not adversely affected from a fire t 3-82 Rev.0 l
sArmN NU(LEAR EgyuansefrAL COItPORATION N FLAN , event. G. Forest or Bmsh Fire The Saxton facility site is located in a relatively wooded section in the Allegheny Mountains, three fourths of a mile north of the Borough of Saxton in Liberty Township, Bedford County, Pennsylvania. The area surrounding the containment building and areas where radioactive materials are stored is maintained and kept free of any significant quantities of combustible vegetation. The local fire company in the Borough of Saxton is close by and could respond quickly to fires outside the plant area that could pose a threat of spreading to the plant site. In addition, a forest fire event at the Saxton facility would typically be preceded by a sufficient warning period to prepare the site for the event. A forest fire could initiate a loss of off-site power event which was analyzed in Section 3.4.1.8.A and concluded that public health and safety were not adversely affected from a loss of off-site power event. 3.4.1.10 Offsite Radiological Events ! Offsite radiological events related to decommissioning activities are limited to those associated with the shipment of radioactive materials. Radioactive shipments will be made in accordance ! with applicable regulatory requirements. The radioactive waste management program and the I Quality Assurance Program assure compliance with these requirements. Compliance with these ! requirements ensures that both the probability of occurrence and the consequences of an offsite i event do not significantly atYect the public health and safety.
- 3.4.1.11 Containment Vessel Breach 1
During decommissioning operations it is possible that the containment vessel steel liner could be accidently breached. Such an incident occurred on May 25,1995 during the characterization 3-83 Rev.0 il
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SArtDN NUCLEAR EXPERIMENTAL CORPORATION (g DECOMMISSIONING PLAN V) study when a three inch diameter core bore inadvertently penetrated the steel liner in the rod room sump, 765'-8" elevation. As a result of this liner breach, ground water entered the containment at a rate of approximately one gallon per minute until a plug was installed. The principal concems with any liner breach would be the possibility of radiological contaminants migrating to the surrounding ground water and the ability to contain any in-leakage of ground water into containment. The incident which occurred May 25, 1995 is considered to be the bounding case as this breach occurred at the lowest point in containment resulting in the maximum ground water intrusion. The leak was plugged and no adverse effects have been noted. Increased ground water monitoring has verified no impact on the ground water. Since ground water at the site is typically only a few feet below ground level (81l' elev) a penetration of the steel liner would result in in-leakage. This would preclude the possibility of ground water contamination due to liner penetration. The remaining risk is that of in-leakage. The breach which occurred previously took place at a time of high gmund water levels and at the point of maximum head pressure, these conditions resulted in the highest in-leakage likely to be encountered. The in-leakage was contained and stopped using readily available materials. Such materials and procedures to use them will be in place during the operations phase. Additionally, precautions will be including in pmcedures to minimize the chance that the liner integrity could be challenged. For these reasons, containment vessel liner penetration is a low probability event which also carries a minimal consequence. b V 3-84 Rev.O
SAXTON NUCLEAR EXPERIMENTAL CORPORATION b DECOMMISSIONING FLAN l' O SECTION 4.0 PROPOSED FINAL RADIATION SURVEY PLAN 4.1 FINAL Pm RASE CRITERIA 4.1.1 Site Release Criteria The release for unmstricted use of the Saxton facility site, including structures and any material mmaining on site or areas affected by residual radioactivity as a result of Saxton facility operations, will be based upon the calculated dose consequence to an average individual of the critical population group. GPU Nuclear intends to release the site for unrestricted use upon completion of decommissioning. As the Saxton facility site is wholly contained within Pennsylvania Electric Co. (PENELEC), pmperty, it is anticipated that the ownership of the Saxton facility propeny will revert back to v PENELEC after the NRC license is terminated. l GPU Nuclear intends to meet the criteria of the proposed change to 10 CFR 20 (Reference 22) for site release. The currently pmposed change requires that residual radioactive contamination at the site attributed to licensed operations contribute not greater than 15 mrem per year Total Effective Dose Equivalent (TEDE) to an average individual of the critical population group during the period of 1000 years following site miease. 1 Rubble, debris, soil and structums remaining on the site will be analyzed using the RESRAD code (Reference 10) or equivalent methodology to calculate the dose. Residual contamination
]
types not applicable to RESRAD methodology will be analyzed by guidance deemed appropriate at the time of use. At present such guidance is given in NUREG/CR-5512 (Refemnce 18) and NUREG-1500 (Reference 19). g < i Rev.0 4-1
1 SAXTON NUCLEAR EXPERIMENTAL CORFORATION I DECOMMISSIONING FLAN Contamination and/or migration of radioactive contamination into ground and surface waters 1 with a potential to be used as a source of drinking shall not exceed the National Primary Drinking Water Standards contained in 40 CFR 141. 4.1.2 Material Release CrilCIia Materials released from the restricted area at the Saxton facility will be appropriately surveyed in accordance with existing procedures which incorporate the guidance presented in NRC Circular 81-07 (Reference 20) and NRC Information Notice No. 85-92 (Reference 21). This ;
)
process will ensure that radioactive materials above the specified release criteria are not ; inadvertently released from the site. The limits for free release of materials from the restricted area are as follows: m
* <100 cpm above background for total (fixed and loose) beta-gamma contamination as measured with a pancake G-M detector (HP-210 or equivalent); and,
* < 300 dpm/100 cm 2above background for total (fixed and loose) alpha contamination; and,
* < 1000 dpm/100cm 2above background for loose beta-gamma contamination; and, 2
* <20 dpm/100cm above background for loose alpha contamination.
- Buli liquids, soils, and other free flowing materials used in contaminated or potentially contaminated areas / systems shall be surveyed / analyzed, as appropriate, to verify that isotopic concentrations are less than those specified by the RSO, as promulgated by current NRC guidance.
1 i O , k.) l l 4-2 Rev.0
I I e SAXTON NUCLEAR EXFERInG!NTAL CORPORATION [ DECOMMISSIONING FLAN v 4.2 FINAL SURVEY METHODOLOGY 4.2.1 Overview l The purpose of the final radiation survey is tc riemonstrate that the Saxton facility meets the final site release criteria for unrestricted use. These criteria are based on the dose to an average < member of the critical population group of less than 15 mmm per year, Total Effective Dose Equivalent (TEDE), for a period of 1000 years following site release. In order to demonstrate compliance with these criteria, the contribution to dose from direct exposure, ingestion and inhalation pathways must be determined. 4 The following radiological data will be obtained to calculate the dose in this manner: O
- Surface sctivity levels (removable and fixed),
- Deposited activity in volumetrically contaminated materials, I
- Direct exposum rates,
- Radionuclide concentrations in soil and ground / surface water, )
I 4 l 4
- Total site inventory of residual radioactivity.
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'Ihe data must be accurate and represent actual site conditions while accounting for the 1
considerable variability of natural and enhanced background radiation. This will be achieved through the use of a well planned, documented survey process. The final radiation survey will be carried out under the control of a survey plan which will incorporate sound statistical methods i of data collection and analysis. f n V : 4-3 Rev. 0
I SAXTON NUCLEAR EXPERIMEPfTAL CORPORATION DECOMMISSIONING FLAN The final radiation survey plan will specify:
- The equipment, methods and procedures for the determination of residual radioactivity and external dose rates, l
- Methodology for data review and analysis including the application of appropriate statistical modeling, l l
- Physical control of samples,
- Training and qualification of the survey personnel, i
- Data quality objectives,
- Quality control / verification.
The final survey plan and related implementing procedures will follow the guidance in the applicable standard at the time of the final survey. At present that guidance is contained in NUREG/CR-5849 (Reference 23), however, guidance needed to implement the currently proposed release criteria will be used if available. These are now available in preliminary draft form as NUREG-1505 (Reference 24), NUREG-1506 (Reference 25) and NUREG-1507 (Reference 26). i . 4.2.2 Instrumentation Selection and use of instrumentation for the final radiation survey will be based upon the need I to ensure that the residual radioactivity remaining on site n'eets the release criteria. Instrumentation will be selected which will have a detection capability as used in the field which (m 4-4 Rev.O
c SAXTON NUCI2AR EXPERIMENTAL CORPORATION l l DECOMMISSIONING FLAN ( is at least 75% of the guideline value established for the type of radiation and media being measured. In all cases, such equipment will be selected using the guidance of the applicable standard as referenced in section 4.2.1. 1 Instrumentation used for the final radiation survey and analysis of samples will be calibrated using standards and methods which are traceable to the National Institute of Standards and Technology (NIST). Calibrations will be performed using site specific representative radionuclides. In all cases, equipment calibration and maintenance sludl conform to the a requirements of the GPU Nuclear Quality Assurance Program for Radiological Instmments (6610-QAP-4220.01) or the Saxton Nuclear Quality Assurance Program for Radiological l Instmments (6675-QAP-4220.01). l l 4.2.3 Documentation All aspects of the final radiation survey will be documented and retained in accordance with Saxton facility procedures. The survey report will pmvide a complete record of the facility . radiological status and a comparison to the site release criteria. The report will be sufficient to enable independent or third party re-creation and evaluation of the survey results and a determination as to whether the site release criteria have been met. 4.2.4 Ouality Assurance Quality assurance requirements for all aspects of the project are covered in section 7.0 of this plan, the final radiation survey will be subject to those requirements. Periodic audits and assessments will be conducted to verify compliance with the applicable procedures and requirements. In addition, an independent re-survey of selected areas, stmetures, components and materials will be made by the GPU Nuclear Sr.fety Assessment Department. 4-5 Rev.O
1 1 SAXTON NUCIEAR EXFERSENTAL COItPORATION
.a DECOMMISSIONING FLAN
! 4.2.5 Indanendant Verification It is anticipated that the NRC will perform or have performed an independent confirmation of ] the final radiation survey. This confirmation will verify the adequacy and scope of the final mdiation survey and will serve as confirmation of the achievement of the site release criteria. 2 i The confirmation of the final radiation survey and the report detailing the site radiological J j conditions will serve as the basis for a decision by the NRC to terminate the Saxton facility j license and permit the unrestricted release of the site. i i i 1 i ? s i i i Y A 4 i 4-6 Rev.O i 4
, 1 c SAXTON NUCMAR EXPERIME!NTAL CORPORATION ' DECOMMISSIONING FLAN O' h I SECTION 5.0 UPDATED COST ESTIMATE FOR DECOMMISSIONING METHOD CHOSEN AND PLAN FOR ASSURING AVAILABILITY OF FUNDING FOR COMPLETION OF DECOMhDSSIONING 5.1 DECOMMISSIONING COST ESTIMATE An updated site-specific cost estimate for the Saxton facility decommissioning was prepared by TLG Services (Reference 13). The estimates provided in the study provide for removing all radiological and hazardous contaminants to levels which will allow the facility and adjacent areas to be released for unrestricted use. The cost estimate is based upon remaining inventories of equipment, building structures, and site characterization data. The estimate was performed in accordance with the published study from ~ the Atomic Industrial Fomm/ National Environmental Studies Project report AIF/NESP-036,
" Guidelines for Producing Commercial Nuclear Power Plant Decornmissioning Cost Estimates" (Reference 6) and the U.S. Department of Energy " Decommission:.ng Handbook" (Reference 4).
These references utilize a unit cost factor method for estimating decommissioning activity costs I to standardize the estimating calculations. l The site-specific cost estimate was developed using available Saxton facility drawings, including structural, mechanical, and electrical, piping and component inventories, radiological survey data, and labor rates and radwaste processing / disposal fees. This information was used to l develop the general arrangement of the facility and determine the estimates of building steel and ! concrete volumes, numbers and size of components, and restoration requirements. The total estimated cost to decommission the Saxton facility in 1995 dollars is $22,200,000. The major elements of the cost estimate are (rounded to the nearest $100,000): b d
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l l
l sAxum NuCu.Am ExnmumrAL COmFORADON g DECOMMISSIONING FLAN k l l
- Labor Related Costs $10,600,000
- Radwaste Disposal / Processing $2,400,000
- Transportation Costs $300,000
- Specialty Contactor Services $1,800,000
- Purchased Materials and Equipment $2,800,000
- Installation of Support Facilities $500.000
- Subtotal $18,400,000
- Contingency $3.800.000 TOTAL ESTIMATED COST $22,200,000 l
5.2 DECOMMISSIONING FUNDING
l A Saxton facility decommissioning tmst fund was established to fully fund the cost of t O decommissioning the site for unrestricted release. As of December 31,1995 the fund had a . positive balance of approximately $8.1 Million. The three owners of the Saxton facility, the Metropolitan Edison Co. (Meted), the Jersey Central Power and Light (JCP&L), and the Pennsylvania Electric Co. (Penelec), all subsidiaries of GPU, are continuing tmst fund collections at an annual rate of $2.5 Million. Any cash flow shortfall prior to full collection will be made up from GPU operating funds. By letter SNEC-90-0041 to the NRC, dated July 26,1990 (Reference 17), Saxton Nuclear Experimental Corporation requested an exemption from the requirements of 10 CFR 50.75(c). The Saxton facility was a small demonstration reactor and the facilities have already been substantially decontaminated and dismantled. 'Ihus, the estimated funding necessary to complete decommissioning is far less than the nominal formula calculation of 10 CFR 50.75 (c). Adequate decommissioning funding, based upon the site specific cost estimate described in Section 5.1, is assured given the circumstances of the Sar. ton facility site. l (
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SAITON NUCLEAR EXPERIMFRFAL CORPORATION DECOMMISSIONING FLAN SECTION 6.0 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS IN PLACE DURING DECOMMISSIONING 6.1 Technical Soecifications The reactor at the Saxton facility was permanently shutdown May 1,1972. Amendment No.8 to the Saxton Facility License which was issued on August 15,1972 converted the Saxton Facility License to a Possession Only License and extended the expiration date to February 11,2000. During the next few years limited decommissioning activities at the Saxton facility were carried out. Amendment No. 9 which was issued on .iamtary 10,1974 recognized the completion of fuel shipping from the Saxton facility and eliminated the license to store V) Special Nuclear Material pursuant to 10 CFR Pan 70. Finally, Amendment No.10 recognized the completion of near term site activities and licensed SNEC to store /midntain the Saxton facility but not to perform decommissioning type activitics. The Saxton facility was placed in storage and maintained until 1986 when decontamination activities were begun with the exception of the containment vessel. These activities were completed in 1989. Following performance of a verification survey, License Amendment No. I1 was issued on May 28,1992 permitting demolition of the Saxton Suppon Facilities. Finally on Febmary 22,1995 License Amendment No.12 was issued which permitted characterization activities to be performed in the Saxton facility containment vessel. These activities were performed in April to October 1995 and the results of the Characterization Survey are contained in Section 3.1.2 of this plan.
)1 6-1 Rev.O
p) v SAXTON NUCILut EXPEItIMEPfrAL COltPOItATION DECOMMISSIONING FLAN A Technical Specification Change Request will be submitted under separate cover to pmvide the Technical Specifications necessary to control decommissioning activities at the Saxton facility. 4 The significant changes requested to the Technical Specifications to support decommissioning activities include:
- 1) Permission to perform decommissioning activities at the Saxton facility as described in the Decommissioning Plan.
- 2) Specific authorization to use 10 CFR 50.59 to permit changes in j decommissioning activities without receiving NRC approval if a change to the Technical Specification or an unreviewed safety question is not involved.
/
- 3) Extending exclusion area controls to include the Saxton facility Decommissioning Support Building.
- 4) Establishing specific Technical Specification controls over decommissioning activities.
- 5) Establishing Technical Specification requirements for a Radiological Environmental Monitoring Program, an Off-Site Dose Calculation Manual, and a Process Control Program.
6.2 Environmental Specifications , l As described in NRC Generic Letter 89-01, Environmental Specifications no longer need to l be maintained as part of a Facility's Technical Specifications. However, a Radiological b q 6-2 Rev.0 l
i l SAX 1DN NUCIEAR EXPERIMENTAL CORFORATION DECOMMISSIONING FLAN } Environmental Monitoring Program (REMP) is mquired. A formal REMP program is in place at the Saxton facility. The scope of the Saxton facility REMP program, including specific sampling locations, sampling frequencies, and media, is detailed in Saxton facility procedums. 1 1 ( 1 2 l } 1 i i i 1 3 M 6-3 Rev.0
SAXMN NUCIAut EXPERIMENTAL CORPORATION
!rb DECOMMISSIONING FLAN V
SECTION 7.0 QUALITY ASSURANCE PROVISIONS IN PIACE DURING DECOMMISSIONING 1 l POllCY STATEMENT 7.1 GPU Nuclear will establish and implement this quality assurance pmgram for the Saxton facility Decommissioning Project. The quality assurance pmgram, as applied to activities shall comply with and be responsive to applicable regulatory requirements and applicable industry codes and standards. These activities are for the protection of the health and safety of the public ~ and project personnel, and for adherence to regulations and commitments made to the Nuclear Regulatory Commission, including the control of personnel exposure to 1 ( radiation, control of radioactive material and contamination, and radioactive waste I
% 1 shipments.
Project procedures shall provide for compliance with appropriate regulatory, statutory, I l license, and industry requirements. Specific quality assurance requirements and ! organizational responsibilities for implementation of these requirements shall be specified. Compliance with this pmgram and provisions of project procedures is mandatory for personnel with respect to decommissioning activities which may affect quality or the 1 health and safety of project personnel or the general public. Personnel shall, therefore, . be familiar with the requirements and responsibilities of the program that are applicable l to their individual activities and interfaces. j This project quality assurance program is structured to comply with appropriate regulatory requirements and is funher implemented to assure that dismantling, packaging, l 7-1 Rev.O
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l sixwm Nucuum sx-tanMAL CORF0hDON DECOhGaSSIOfGNG FLAN
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4 j i and shipping activities are conducted in a controlled manner designed to assure quality and to protect the health and safety of both project workers and the general public. ) 7.2 ORGANIZATION AND FUNCHONAL OA RESPONSTRillTIR9 ii j 7.2.1 Geneml 1 1 It is the policy of GPU Nuclear to conduct decommissioning activities in such a manner ! as to ensure protection of the health and safety of the public and the personnel on site. 4 To implement this policy, GPU Nuclear will adhere to the QA requirements established in this Decommissioning Plan and other applicable federal, state, and local requirements. l i GPU Nuclear is responsible for the Quality Assurance Plan implementation. Verification jg of effective plan implementation is the responsibility of the GPU Nuclear, Nuclear Safety Assessment (NSA) organization. NSA personnel shall have the authority and organizational freedom to identify quality problems; to take action to stop unsatisfactory work and control funher processing, delivery, installation or use of nonconforming i items; to initiate, to recommend, or to provide solutions; and to verify implementation of solutions. The persons and organizations performing quality assumnce functions report to a management level that assures the required authority and organizational j freedom are provided, including sufficient independence from cost and schedule. The j individuals assigned the responsibility for assuring effective execution of any portion of the Quality Assurance Plan have direct access to the levels of management necessary to perform quality assurance functions. { ! 7.2.2 Ornnintion The Saxton facility organization is structured on the basis that the attainment of the i \ 7-2 Rev.O
l SAXMN NUCLEAR EXPERIMENTAL CORPORATION
. DECOMMISSIONING FLAN \
objectives of the QA Program relies on those who manage, perform, and suppon the performance of activities within the scope of this plan; and assumnce of this attainment miles on those who have no direct responsibility for managing or performing the activity. t 4 The organizations responsible for the conduct, suppon, and assurance of decommissioning am described in section 2.3 of this Decommissioning Plan. Specific QA requirements and organizational msponsibilities for implementation of these mquirements shall be specified in various implementing procedures and other documents.
- The requirements stipulated in the QA Program shall be imposed on all personnel and organizations, including contractors, who perform the decommissioning activities 4
referenced in Section 7.1 above. O 7.2.3 Functional OA Resoonsibilities 7.2.3.1 GPU Nuclear President The GPU Nuclear Pmsident has the overall responsibility for the establishment, implementation, and effectiveness of the Saxton facility Decommissioning QA Program. This responsibility is administered through management staff as described below. ,I 7.2.3.2 GPU Nuclear Vice President, Nuclear Services Division The GPU Nuclear Vice President, Nuclear Services Division repons directly to the GPU Nuclear President. This position is responsible for the overall Saxton facility decommissioning pmgram and the various suppon elements required to safely decommission the facility. O b 7-3 Rev.O
l l SAXTON NUCIAut EXPERIMENTAL COItMMtAT10N DECOMMISSIONING FLAN v 7.2.3.3 P t> gram Director, SNEC facility The Program Director, SNEC feri)ity reports to the GPU Nuclear Vice President, Nuclear Services Division. The Program Director is responsible to decommission the Saxton facility in a safe and efficient manner in accordance with this QA Program and with all applicable laws, regulations, licenses, and technical requimments. 7.2.3.4 Saxton Radiation Safety Officer (RSO) The Saxton Radiation Safety Officer reports to the GPU Nuclear Director, Nuclear Services and is responsible for the conduct and oversight of all radiation safety activities through implementation of this QA Program and the Saxton Radiation Protection Plan. V 7.2.3.5 SNEC Facility Site Supervisor The SNEC Facility Site Supervisor reports to the Program Director SNEC facility and is responsible for on-site management and continuing oversight of production activities. The Supervisor is responsible for the safe conduct of production activities and effective implementation of QA Progmm requirements. 7.2.3.6 Radiation Safety Committee The Radiation Safety Committee reports to the GPU Nuclear Director Nuclear Services and is responsible for the review of inspection and audit results and corrective actions for deficiencies. The Committee has the organizational freedom, authority and capability to identify safety problems; to initiate, recommend, or provide solutions; and to verify implementation of solutions. G G l l 74 Rev.O j
SAX 10N NUCIRAEt EIFEILInGDfrAL COItPOItATION
.O DECOMMISSIONING PLAN V
7.2.3.7 GPU Nuclear, Nuclear Safety Assessment (NSA) Staff
'Ihe audit function is provided by GPU Nuclear and is independent of Saxton facility management. GPU Nuclear's NSA Department is responsible to provide auditors certified to meet the requirements of ANSI N45.2.23 and to establish an ,
audit program for the activities listed in Section 7.1. The audit reports are forwarded to the GPU Nuclear President, the Vice President, Nuclear Services and the Radiation Safety Committee within 60 days of the completion of the audit. 7.3 SAXTON FACILITY OUALITY ASSURANCE PROGRAM 7.3.1 General Requirements O V A) The project quality assurance program shall:
- 1. Documented by written procedures.
- 2. Carried out throughout the decommissioning project in accordance with those procedures. l l
4 B) The program shall provide control over activities affecting quality or the health and safety of project personnel and the general public. i C) Activities affecting quality shall be accomplished under suitable controlled conditions. Controlled conditions include the use of appropriate equipment; suitable environmental conditions for accomplishing the activity, such as adequate cleanliness; and assurance that all prerequisites for the given activity have been satisfied. A 7-5 Rev.O
~ m SAXTON NUCLEAR EXPERIMENTAL CORPORATION O " - - D) The program shall take into account the need for special controls, processes, test equipment, tools, and skills to attain the required quality, and the need for verification of satisfactory implementation. E) The program shall provide for indoctrination and training of personnel performing activities affecting quality as necessary to assure that suitable proficiency is achieved and maintained. F) The adequacy and status of the program shall be regularly reviewed. G) Management of other organizations participating in the program shall mgularly mview the status and adequacy of that part of the program which they are , implementing. 0 V 7.3.2 General Descriotion l A) The Saxton Facility Decommissioning Project Quality Assurance Program has been established to govern those activities that may affect the quality of the pmject, including the health and safety of the general public as well as project l personnel, i B) The project quality assurance program shall utilize the Saxton Facility Decommissioning Project Quality Assurance Plan (plan) and appropriate implementing procedums to meet its objectives. The plan shall be considered an J overall document which governs the implementing documents, i.e., procedures. 7.3.3 Sall'>n Facility Decommissioning Project Ouality Assurance Plan C t
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7-6 Rev. 0 l 1
l i p SAXTON NUCIEAR EKFERIMENTAL CORPORATION c DECOMMISSIONING PLAN ( l i l A) The plan shall describe in general how compliance with appropriate quality and safety requirements is accomplished. B) The plan shall be issued under the authority of the GPU Nuclear Pmsident and shall be reviewed by the Manager NSA. C) All changes to the project quality assurance plan shall be approved by the GPU Nuclear President and reviewed by the Manager NSA. 7.3.4 Procedures and Drawings A) Procedures, and drawings of a type appropriate to the circumstances, shall be provided for the control and performance of activities which affect quality, health ( and safety of the public or pmject personnel, or regulatory requirements. l B) Procedures and drawings shall include appropriate quantitative or qualitative 1 acceptance criteria for determining that imponant activities have been satisfactorily accomplished. I i ., C) GPU Nuclear departmental level and specialized procedures, plans, policies and documents may be utilized as deemed appmpriate for use at or in support of Saxton facility activiti.:s with the approval of the Program Director, SNEC facility and with the concurrence of a Responsible Technical Reviewer. D) The following typical procedures shall be provided as appropriate: Calibration procedures, Radiation protection pmcedures, Special process procedures, 7-7 Rev.0
l SAXTON NUC1JCAR EXPEILIMENTAL CORPORATION t DECOMMISSIONING PLAN O Transport procedures, Radioactive material packaging and shipment procedures, Demolition procedure (s), Audit procedures, , Administrative control procedures, Emergency procedures, Inspection procedures. E) Drawings and technical manuals of a type appropriate to the circumstances may be used as procedural documents. 7.3.5 Ouality Assurance Training , T N) Training programs shall be established for those personnel performing quality affecting activities such that they are knowledgeable in the quality assurance documents and their requirements and proficient in implementing these requirements. These training programs shall assure the following. A) Personnel responsible for performing these activities are instructed as to the purpose, scope, and implementation of applicable controlling procedures. B) Personnel performing such activities are trained and qualified, as appropriate, in principles and techniques of the activity being performed. C) The scope, the objective, and the method of implementing the training programs are documented. 7-8 Rev.0 1
q SAITON NLFrNAR EIFERBIENTAL CORPORATION [V t DECORGEIBSIONING FLAN D) Methods are provided for documenting training sessions describing content,
! attendance, date of attendance, and the results of the training session, as appropriate.
7.3.6 Desien Contml ;
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A) When dismantling or shipping activities require design, or modification of existing design, controls shall be applied commensurate with the potential impact on j quality or health and safety of project personnel and the general public, t B) Appropriate provisions of design control shall include the specifying of design i input, the correct translation of input in design documents, the verification of design by persons other than the originator, and the assurance that changes to the f design beyond design tolerance are properly reviewed and contmiled. . 7.3.7 Pmcurement Document Contml i f l A) Measures shall be established to assure that applicable regulatory requirements, design bases, and other requirements which are necessary to assure adequate [ quality are suitably included or referenced in the documents for procurement of j material, equipment, and services, whether purchased by GPU Nuclear or by its contractors or subcontractors. To the extent necessary, procurement documents shall require contractors or subcontractors to provide a quality assurance program ,
- consistent with the contractor's potential impact on quality or the health and i safety of project personnel and the general public.
B) Procurement documents shall contain specific technical and quality requirements, as appropriate. 1 A 7-9 Rev.0
i SAXTON NUCMAR EXPERIMDrTAL CORPORATION DECOMnOSSIONING PLAN v C) Procurement documents shall contain provisions which establish the right of access to vendor facilities and records for source inspection and audits as appropriate. D) Procurement documents for contracting packages for transport of radioactive materials shall when appropriate require a copy of the package license, certificate, or other NRC approval authorizing use of the package. The procurement documents shall also require copies of all documents refened to in the license, certificates, or other NRC appmval as applicable which relate to the use and maintenance of the packaging and to the actions to be taken prior to shipment. E) Documents, and changes thereto, initiating procurement of equipment, components, or services shall be approved by appropriate management personnel , and shall be subject to a quality review to ensure applicable regulatory requirements, design bases, quality assurance, and other requirements am adequately satisfied prior to release. 7.3.8 Document Control A) Measures shall be established to control the issuance of documents, such as procedures and drawings, including changes thereto, which prescribe activities affecting quality. B) These measures shall assure that documents, including changes, are reviewed for adequacy and approved for release by authorized personnel, and are distributed to and used at the location where the prescribed activity is performed. C) Changes to documents shall be reviewed and approved by the same organization O V 7-10 Rev. 0
SAXTON NUCLEAR EXPERIMENTAL CORPORATION [- v DECOMMISSIONING FLAN that performed the original review and appmval or another designated msponsible organization.
, D) Required procedures shall be controlled to assure that current copies are made available to personnel performing the prescribed activities. Required pro 2dures shall be reviewed by a technically competent person other than the preparer, and shall be appmved by a management member of the orgamzation responsible for the prescribed activity.
E) Significant changes to mquired procedures shall be reviewed and approved in the I same manner as the original. 7.3.9 Control of Purchased Material. Eauipment. and Services . O) q v A) Measures shall be established to assure that purchased material, equipment and services conform to the procurement documents. These measures shall include provisions, as appropriate, for vendor evaluation and selection, objective evidence of quality furnished by the vendor, inspection at the vendor so irce, and inspection of products upon delivery. B) The effectiveness of the control of contractor services shall be assessed at intervals consistent with the importance of the service. C) The adequacy of vendor's quality assurance program specified in procurement documentation shall be verified prior to use when appropriate. Vendors' adherence to their quality assurance program shall also be verified as appropriate. D) Commensurate with potential adverse impacts on quality or health and safety, O V 7-11 Rev.0 l I
SAXTON NUCLEAR EXFIGtIMENTAL CORPORATION O DECOMMISSIONING FLAN material and equipment shall be inspected upon receipt at the plant site or other GPU Nuclear sites prior to use or storage to determine that procurement requirements are satisfied. 4 B) Material, pans, and components that are to be utilized to fulfill a 10 CFR 71 related function or used for shipment of radioactive materials shall be inspected upon receipt to assure that associated procurement document provisions have been satisfied. Measures shall be established for identifying nonconforming material, parts and components. 7.3.10 Identification and Control of Materials. Parts and Comoonents A) Measures shall be established for the control of critical materials, parts, and components. gx) B) These identification and control measures shall be designed to prevent the use of incorrect or defective material, parts and components. 7.3.11 Control of Special Processes s A) Measures shall be established to assure that special processes, including welding, and nondestmetive examination are controlled and accomplished by qualified I personnel using qualified procedures in accordance with applicable codes, standards, specifications, criteria, and other special requirements. B) Welding of critical lifting and rigging equipment shall be performed in accordance with qualified procedures. Such procedures shall be qualified in accordance with applicable codes and standards and shall be reviewed to assure their technical C ( 7-12 Rev.0
n) SAITON NUCI#1.R EXPERilENTAL CORPORATION DECOleGGSIONING FLAN (V adequacy. C) Measures shall be established that assure welding of critical lifting and rigging equipment is performed by qualified personnel. D) Nondestructive examinations (NDE) of critical lifting equipment shall be performed in accordance with procedures formulated in accordance with applicable codes and standards and shall be reviewed to assure their technical adequacy. E) Measures shall be established that assure nondestructive examination (NDE) are performed by personnel qualified in accordance with applicable codes and standards. l l 7.3.12 Insoection A) Measures shall be established for inspection of appropriate activities to verify conformance with the documented procedures and drawings for accomplishing the activity. l l B) If mandatory inspection hold points, which require witnessing or inspection and beyond which work shall not proceed without prior consent are required, the l specific hold poir.ts shall be indicated in appropriate documents. C) Measures shall be established which assure that acthities associated with technical j services (such as surveillance testing, instrument calibration, laboratory services, l etc.) are inspected by qualified personnel when determined appropriate by quality or other qualified personnel. A d 7-13 Rev.0 l
SAX 11)N NUCMAR FJPERIMENTAL CORPORATION m} [V DECOMMISSIONING FLAN D) Measures shall be established which assure that packages utilized to ship licensed radioactive material offsite are inspected in accordance with the applicable provisions of 10 CFR 71. E) Required inspections shall be performed in accordance with appropriate procedures. Such procedures shall contain a description of objectives, acceptance criteria and prerequisites for performing the inspections. These procedures shall also specify any special equipment or calibrations required to conduct the inspection. F) Personnel performing required inspections shall be qualified. Required inspections shall not be performed by individuals who performed the inspected activity or directly supervised the inspected activity. G) Personnel performing required inspections shall be qualified based upon experience and training in inspection methods. 7.3.13 Test Control
- Measures shall be established to assure that tests necessary to assure quality or health and
- safety are controlled and accomplished in accordance with approved procedures. Such
- tests shall include verification of lifting capacity of cranes prior to use in performing l
critical lifts. 7.3.14 Control of Measuring and Test Equipment i Measures shall be established to assure that tools, gauges, instruments and other measuring and testing devices used in activities important to health and safety are o 7-14 Rev.O
fs SAXTON NUCLEAR EXPERIMENTAL CORFORATION ( DECOMMISSIONING FLAN v properly contmiled, calibrated and adjusted at specified periods to maintain accuracy within necessary limits. 7.3.15 Handline. Storare. Shipoing and Housekeeping Measures shall be established to control the handling, storage, and shipping of radioactive materials and to maintr.in appmpriate levels of housekeeping. A) Amas shall be pmvided for storage of radioactive material which assum physical protection, as low as reasonably achievable radiation exposure to personnel, control of the stored material, and containment of radioactive material as appmpriate. I B) Handling, storage, and shipment of radioactive material shall be controlled by wJ pmcedures based upon the following criteria. i
- 1. Established safety restrictions conceming the handling, storage, and shipping of packages for radioactive material shall be followed.
- 2. Shipments shall not be made unless all tests, certifications, acceptances, and final inspections have been completed.
- 3. Shipping and packaging documents for radioactive material shall be consistent with pertinent mquirement of 10 CFR 71.
7.3.16 Insoection. Test. and Operatine Status A) Appropriate controls consistent with 10 CFR 20 shall be established for the A b 7-15 Rev. O
SAXTON NUCIEAR EXPERIMENTAL CORPORATION g 4 DECOMMISSIONING PLAN V control of radioactive material as well as personnel exposure. B) Inspection, test, and operating status of equipment and components associated with radioactive material shall be established based upon the following criteria.
- 1. Inspection, test, and operating status for radioactive material shall be indicated and controlled by established procedures.
- 2. Status shall be indicated by tag, label, marking or log entry.
- 3. Status of nonconforming puts or packages shall be positively maintained by established procedures.
( ( 7.3.17 Nonconformine_ Materials. Parts or Com_ponents A) Measures shall be established to control materials, parts, or components which l do not conform to requirements in order to prevent their inadvertent use or release for shipment. These measures shall include, as appropriate, procedures for identification, documentation, segregation, disposition and notification to affected organizations. B) Nonconformance items shall be reviewed and accepted, rejected, repaired, or reworked in accordance with documented procedures. 7.3.18 Corrective ACliDD A) Measures shall be established to assure that conditions adverse to quality, such as failures, malfunctions, discrepancies, deviations, defective material and m V) I 7-16 Rev. O
l l 1 I ; I ! SAITON NUCLEAR EXFIDtMDfrAL COBtMHtATION b DECOMMISSIOfGNG FLAN '( equipment, and nonconformances are promptly identified and corrected. B) The identification of the condition adverse to quality, the cause of the condition, and the cortective action taken shall be documented and reported to appropriate levels of management. 7.3.19 Onnlity Assurance Records A) Sufficient records shall be maintained to furnish evidence of activities affecting quality. B) Records shall be identifiable and retrievable. b C) Requirements shall be established concerning record retention, such as duration, ( location, and assigned responsibility. Such requirements shall be consistent with the potential impact on quality, health and safety of the public, safety of project personnel, and applicable regulations. D) Measures shall be established which assure that qualification records of personnel performing special process activities, such as welding, NDE, inspection, etc., are retained. E) Measures shall be established which assure that quality related procurement documents are retained. F) Measures shall be established which assure that appropriate records pertaining to audits are retained. 7-17 Rev.O
SAXTON NUCLEAR EXPERIMEPRAL CORPORATION DECOMMISSIONING FLAN l G) Measures shall be established which assure that records associated with radioactive material and personnel exposure contml are retained. ) 7.3.20 Audits l I l A system of planned audits shall be carried out by the GPU Nuclear NSA to verify compliance with appropriate requirements of the Project Quality Assurance Program and l to determine the effectiveness of the program. The audits shall be performed in accordance with written pmcedures or checklists by appropriately trained personnel not having direct responsibility in the areas being audited. Audit results shall be documented and reviewed by management having responsibility in the area audited. Follow-up action, including re-audit of discrepant areas, shall be taken where indicated. A) Reports of the results of each audit shall be prepared. These reports shall include a a description of the area audited, identification of individuals responsible for implementation of the audited pmvisions and for performance of the audit, identification of discrepant areas, and recommended corrective action as i appropriate, B) Audit reports shall be distributed to the GPU Nuclear President, the Vice President Nuclear Services Division, the Radiation Safety Committee, and to those individuals responsible for implementation of audited provisions. 7-18 Rev.O
l l SAXTON NUCIZAll EIFE3URENTAL CORFOILATION
- ' DECOMMISWONING FLAN Y
C) Measures shall be established which assure that discrepancies identified by audits l or other means are resolved. These measures shall include notification of the l manager responsible for the discrepancy, recommended corrective action, and verification of satisfactory resolution. Discrepancies shall be resolved by the manager responsible for the discrepancy. Line management shall resolve disputed discrepancies. , l l s 7-19 Rev.0 l l
i
, SAI1DN NUCutAR EXPERIMENTAL CORPORATION i DECOMMISSIONING FLAN i
SECTION 8.0 PHYSICAL SECURITY PLAN PROVISIONS IN PLACE DURING DECOMMISSIONING 8.1 DESCRIP'IlON 1
! Physical Security requirements for the Saxton facility decommissioning project are as specified in the Saxton Facility Technical Specifications and are as follows:
1 j 1. Except for authorized entry the following access points shall be maintained locked:
- a. the gate (s) to the exclusion area fence surrounding the CV, i
~
- b. the CV access door (s),
- c. the grating covering the auxiliary compartment stairwell in the CV.
These same requirements will remain in place during the Saxton facility dismantlement activities except for the auxiliary compartment stairwell grating which will no longer require ! locking after the approval of TSCR 57. Additionally, temporary structures built at the 4 Saxton facility in support of dismantlement for handling radioactive material arising from dismantlement activities shall also be maintained locked except for authorized entries, when 3 they are occupied by dismantlement personnel or when activities requiring access are planned. Personnel authorized entry to these areas will be designated in writing by the ! Program Director, SNEC facility. 1 Access control measures consisting of a uniformed " watchman" level security officer (unarmed) will be in place to verify that only authorized personnel are permitted entrance during normal hours of operation. 'Ihe security officer will be provided with means of 1 communication to make all necessary contacts. After normal hours of operation the site
- security fence will be maintained locked and selected facilities will have security alarms in operation. Security checks performed by Pennsylvania Electric Co. personnel will continue to 1
O 8-1 Rev.O i
SAITON NUCLEAR EIFERIMM mm DECOMMISSIONING PLAN be made during non-workdays. Selected Saxton facility management personnel will be provided with the appropriate keys in the event site access is requimd outside of normal hours of operation. Response to unauthorized entry at the Saxton facility has been coordinated with Pennsylvania Electric Co. personnel and is contmiled by plant pmcedum. 8-2 Rev. 0
p SAXMN NUCMAR EXPE3tinEDfTAL CORPORATION f DECOhEWISSIONING FLAN SECTION
9.0 REFERENCES
1.) Saxton Site Characterization Plan , Procedure No. 6575-PLN-4520.06 2.) Draft Regulatory Guide, " Standard Format and Content for Decommissioning Plans for Nuclear Power Plants", Task DG-1005, U.S. NRC, Washington, D.C., September 1989 3.) NUREG-0586 " Final Generic Environmental Impact Statement on Decommissioning of Nuclear Facilities", US NRC, Washington, D.C., August 1988. 4.) Decommissioning Handbook, U.S. Depanment of Energy, DOFJEM-0142P, March 1994, Wasinngton D.C. l 5.) "Saxton Nuclear Experimental facility Reactor Vessel, Internals, Ex-vessel lead, Structural Steel and Reactor Compartment Concrete Shield Wall Radionuclide Inventory" (Proprietary), Document No. G01-1192-003, Rev. O, December 1995, prepared by TLG Services, Inc. 6.) Atomic Industrial Forum / National Environmental Studies Project report AIF/NESP-036, " Guidelines for Producing Commercial Nuclear Power Plant Decommissioning Cost Estimates". 7.) " Decommissioned Status of the Saxton Reactor Facility" forwarded to the United States Nuclear Regulatory Commission (USNRC) on Febmary 20,1975. 8.) " Final Release Survey Repon of the Reactor Support Buildings", GPU Nuclear Repon, Revision 3, dated March 1992. 9.) "SNEC Reactor Support Buildings Demolition Report", GPU Nuclear Report, dated March 1994. 10.) RESRAD, "A Computer Code for Evaluating Radioactively Contaminated Sites," Argonne National Laboratory. 11.) Saxton Site Characterization Report, Draft, GPU Repon, dated January 1996. 12.) 1994 Saxton Soil Remediation Pmject Report, GPU Nuclear, May 11,1995. 13.) " Cost Estimate for the Saxton facility", prepared by TLG Services, Document No. 9-1 Rev.O l t .
SAITUN NUCMAR EINWLIMMTAL COItPORATMIN y - n.iN G01-1192-002), Draft, dated September 1995. 14.) NUREG/CR-0130, " Technology, Safety, and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station", Volumes 1 and 2. 15.) Regulatory Guide 1.145, " Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants", US NRC,1983. 16.) EPA 400-R-92-001, " Manual of Protective Actions Guides and Protective Actions for Nuclear Incidents", US EPA,1991. 17.) SNEC letter No. SNEC-90-0041 to NRC, dated July 26,1990, Docket No. 50-146 18.) NUREG/CR-5512, Volume 1, " Residual Radioactive Contamination from Decommissioning: Technical Basis for Translating Contamination levels to Annual Total Effective Dose Equivalent," U.S. Nuclear Regulatory Commission, October ! 1992. 19.) NUREG-1500, " Working Dmft Regulatory Guide on Release Criteria for Decommissioning: NRC Staff's Draft for Comment," U.S. Nuclear Regulatory Commission, August 1994 20.) NRC IE Circular No. 81-07, " Control of Radioactively Contaminated Material," l May 14,1981. 21.) NRC IE Information Notice No. 85-92, " Surveys of Wastes Before Disposal from Nuclear Reactor Facilities," December 2,1985. 22.) Proposed Rule " Radiological Criteria for Decommissioning," Federal Register Vol. 59, No.161, pp. 43200-43232, August 22,1994. 23.) NUREG/CR-5849, " Manual for Conducting Radiological Surveys in Support of License Termination," draft dated June,1992. 24.) NUREG-1505, "A Nonparametric Statistical Methodology for the Design and Analysis of Final Status Decommissioning Surveys, U.S. NRC, August 1995. 25.) NUREG-1506, " Measurement Methods for Radiological Surveys in Support of Decommissioning Criteria, U.S. NRC, August 1995. 26.) NUREG-1507, " Minimum Detectable Concentrations with Typical Radiation Survey O 9-2 Rev.0
SAXTON NUCLEAR EXPERDENTAL CORPORATION DECOMMISSIONING FLAN Instmments for Various Contaminants and Field Conditions," U.S. NRC, August 1995. 27.) CCC-371, "ORIGEN2 - Isotope Generation and Depletion Code - Matrix Exponention Method," Oak Ridge National Laboratory Radiation Shielding Information Center, May 1991. 28.) CCC-254 "ANISN - ORNL - One Dimensional Discrete Ordinates Transport Code System with Anisotmpic Scattering," Oak Ridge National Laboratory Radiation Shielding Information Center, April 1991. n
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