ML20140D509
| ML20140D509 | |
| Person / Time | |
|---|---|
| Site: | General Atomics |
| Issue date: | 04/18/1997 |
| From: | Bramblett G, Nicolayeff V, Welch A GENERAL ATOMICS (FORMERLY GA TECHNOLOGIES, INC./GENER |
| To: | |
| Shared Package | |
| ML20140D497 | List: |
| References | |
| PC-000482, PC-482, NUDOCS 9704230280 | |
| Download: ML20140D509 (180) | |
Text
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PC-000482/0 GENERAL ATOMICS TRIGA* REACTOR FACILITY DECOMMISSIONING PLAN APRIL 1997 9
00**SS$$0?{s*o" 9 h GENERAL ATDHCS
C~NERAL ATOMICS Em m ., PROJECT CONTROL ISSUE
SUMMARY
DOC. CODE PROJECT DOCUMENT NO. REV.
I
{LE:
RGN 9009 PC-000482 0 General Atomics TRIGA* Reactor Facility Decommissioning Pian APPROVAL (S)
REVISION CM APPROVAU PREPARED RESOURCE / DESCRIPTION /
DATE REV BY SUPPORT PROJECT W.O.NO.
S Y 'y 1/
, SSED 2 > 0 A. J . Welch V. Nicolayeff C htt Initial Issue W. O. 9009.303.055 APS 171997 g,
K. E. Asmussen 4/7/97
>NTINUE ON GA FORM 2175-1
- See List of Effective Pages
=?
GA PROPRIFTARY INFORMATION THIS DOCUMENT IS THE PROPERTY OF GENERAL ATOMICS. ANY TRANSMITTAL OF TVtS DOCUMENT OUTSIDE GA WILL CONF 10ENCE. EXCEPT WITH THE WRITTEN CONSENT OF GA (1) THtS DOCUMENT MAY NOT F;E COPIED IN WHOLE OR IN PART AND WILL BE RETURNED UPON REQUEST OR WHEN NO LONGER NEEDED BY RECIPIENT AND (2) INFORMATION CONTAINED HE COMMUNICATED TO OTHERS AND MAY BE USED BY RECIPIENT ONLY FOR THE PURPOSE FOR WHICH IT WAS TRANSMITTED NO GA PROPRIETARY INFORMATION lPAGE jj OF
- l PC-000482/0 LIST OF EFFECTIVE PAGES Page Number Page Count Revision l
i through ix 9 0 >
l-1 through 1-11 11 0 2-I through 2-27 27 0 3-I through 3-17 17 0
- 4-1 through 4-13 13 0 l t I 5-1 1 0 .
6-1 1 0 l
7-1 1 0 ,
l 8-1 1 0 9-1 1 0
]
10-1 1 0 )
l A-1 1 0 A-1,1 1 0 -
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A-2 through A-66 65 0 )
B-1 through B-24 .24 0 Total 174
)
fl 1 se i l
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PC-000482/0 TABLE OF CONTENTS 11 P R O J ECT C O NT R O L I S SU E S U M M AR YIv. . . . . . . . . . . . . 111 LI ST O F E F F E CTIV E P A G E S . . . . . . . . . . . . vi. . . . . . . . . . . . .
T A B L E O F C O NT E N T S . . . . . . . . . . . . . . . . . vi. . . . . . . . . . . . .
LISTOFFIGURES................................................................................... vil LiSTOFTABLES..................................................................................... 1-1 LIST O F AC RO N Y M S/ A B B REVI ATIO N S . .
.. .1. -1. . . . . . . ... . .. . .
- 1. S U M M A R Y O F P L A N . . . . . . . . . . . . . . . .. ... ...1 -S. . . . . . . . . . . . .
1.1 Introduction . . . . .. .. . . ... . .. . . . ...... . .. . . .. . . ...
............. . ... . 1 -9 1.2 Baekground., . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......... . . .
... .1 -9 4 1.3.1 Reactor Decommissioning Overview . .. .. . .. . . . . . . . . . . . . . ... . .. . . . . .
.. . . . .1 10 1.2.2 Estimated Cost. . . ........... ............. ...... .
.1-10 1.2.3 Availability of Funds... . . . . .... . . . . . . . . . . .. . . . . . . .
1.2.4 Program Ouality Assurance.. .
. . . . . . . . . 2-1 8 ..2-1
- 2. D E C O M MIS S IO NIN G 2.1 Decommissioning Altematives.. ..
. . . . . . . . . . . . . ACTIVITIE ........ ...
S . . . . . . . ..2-1
... . ..2-1 2.2 Facihty Radiological Status.. . ...... . . ... . . .. ..... .. . ....... ..
. 2-5 2.2.1 Facility Operating History . . . .. ..... . .. ..... . . . . . . . . . . . .
. .2-6 2.2.2 Current P.e liological Status of the Facility.. .. . . . . . . . .
2.2.3 Release Crieria-Residual .............
, , . . . Radiation and Contamination Levels. ... . .
2.3 Decommissioning Tasks.. . .. .............. ..........................................29 . . . . . . . .... . . 2-17 2.3.1 Activities and Tasks. .. .. .. . ....... ... ................
2.3.2 Schedule. . . . . . . . . . . . . . . . . . . . . . . . . . . .
......................2-17 2.4 Decommissioning Organization and Responsibilities.... ..............................,.....2-24 2.5 Training Program.. ... .......................
............ ..........2-26 2.6 Contractor Assistance.. .
2.7 Decontamination and Decommissioning Documents and Guides RS AND . . ... .. ...... . .. .. .
31
- 3. PROTECTION OF THE HEALTH ..3-1AND SAF
. . . . . .. 3- 1 3.1 Radiation Protection.. . .. .. . . ... . . . . . . .. . . . . . . .. . ... . . ... . . . . . . . . . 3-3 3.1.2 He alth Physics P rogram. . .. . .. .... . ... .. . . . . . . . . . . . . . .. ... ... .. . . . . . . . . . . . . . . . . . . . . . . . .
.. .... . 3-12 3.1.3 Dose Estimates......... . ..... .. .......... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .3 . 12 3.2 Radioactive Waste Management . ...... ... ... ......... .. ... . 3 13 3.2.1 Fu el Re moval . . ... . .. . ... . .. . ... . .. . ... .. . . . . . . . . . . . . . . . . . ............ ... ... 3-13
... . 3 15 ,
% 3.2.3 Radioactive Waste Disposal ... .. .. . .... .. .. ....
3.2.4 General Industrial Safety Program....... . . . ..........
3.3 Radiological Accident Analyses.... .... . ........ .. . .......
v 4. PROPOSED FIN AL R ADI ATION i l d SURVEY d in the
. . .. . 4-1 PLAN ...
4.1 Description of Final Radiation Survey Plan .......... . . . . . . .
.....41 ...4-2.... ..... .. .. . ..
S u rv e y Pia n. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .l . . . . . . . . . . . . . . . . . . . . . . . . . . .
4.1.2 Means for Ensuring that Sufficient ..............
Data is included to. Achieve
. . . . 4 -2 Statistical Goa
..4-3 4.2 Background Survey Results .................. .. .. .. . . . .
..4-3 4.34.3.1 Release Criteria-Residual Radiation and Equipment Release Criteria .. .... ..... ...... ... . . . .. . ... ....
Contamination Levels . . . . . .. .
.4-5
.. . .. ..4-6 4.3.2 Facility Release Criteria... .... ... . .. .. .. .. .. . . . . . ... . . . . .
4.3.3 Soil Release Criteria . . . ....... .. ... . .. ..... .... .. ..
iv
PC-000482/0 4.4 Measurements for Demonstrating Compliance with R l e ease Criteria.
4.4.2 Measurement Methodolo4.4.1 ,
n Instrumentation-Type, ..
.. 4 7Specifications 4.4.3 Site Survey Grid.. .... . . .gy for Conduct..of. Surveys..g Conditions.. ... .4-7..
4.4.4 Fixed Contamination Survey Protocol ......... .
.4-8 4.4.5 Removable Contamination Survey . . . . .
Protocol
.. . . . .48 Analyzing a d A . ... .4-12 4.5 Methods to be Employed for Reviewing ,
, n uditing Data..
.4 12 4.5.2 Cross-check of Results..4.5.1 ..
Laboratory / Radiological ssurance . . . .. . .
. . .. 4 12 M?asureme 4.5.3 Supervisory and Management Review of . Results
.4-12
. . 4-13
- 5. TECHNICAL SPECIFICATIONS...... .......
..4-13
- 6. PHYSICAL SECURITY PLAN ........... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
- 7. EM ERGENCY PLA N... .......
- 8. ..........................................................51 ENVIRONMENTAL g REPORT..............................................................................6- ........
- 9. CHANGES TO THE DECOMMISSIONING
- 10. PLAN...................................................7-1 .. 8-1 R E FE R E N C E S. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . g. .
APPENDIX A-
SUMMARY
OF CHARACTERIZATION 10-1 R............
APPENDIX B-ENVIRONMENTAL REPORT E S UL TS. . . . . . . . . . . . . . . . . . . . . .
.....................................................A-1 ......B-1
- a V
r- - - - - - - - - - - - - - - ~ - -- - - - - - ' - - --- -
1 PC-000482/0 i
LIST OF FIGURES Figure 1-1-Regional Location.... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .................1-2 Figure 12-GA Site and Surrounding Uses. ... .. .. ...... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ......1-3 Figure 1-3-TRIGA' Reactor Facility (Building 21) Layout .... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..1 -6 Figure 1-4-TRIGA* Reactor Facility and Associated Yard Area Boundary.. . .... . . ... . ...... .. ... .. ...1 7 Figure 2-1-TRIGA* Mark i Chronology....... ........ ... . . .. ......... .. ....... ... ........... . . . . ...........2-2 - -
Figure 2-2-TRIGA* Mark F Chronology..... .... ..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 -4 -
Figure 2-3-Decommissioning Phases .... . . ... . .... .... . .. .............. ..... ... . .................2-11 Fig u re 2 -Mark l Re actor . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..... 2-12 Fig u te 2-5-Ma rk F Reactor . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . ...............2-13 Figure 2-6-Mark til Reactor..... ..... . .. ..... . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . 2- 14 4 Figure 2-7-Decommissioning Schedule ............. ... ... . . . .... ... ........ . .. .. . .... ... . . . . . . .. 2-18 Figure 2-8-Decommissioning Organization......... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...2-22 Figure 4-1-Background Soll Sample Locations................ .. ........ ...... . . . . . . . . . . . . . . . . . .. .. 4 4
, Figure 4-2-Systemata Soil Sampling Method. ...... .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .............4-9 Figure 4-3-Modified Sy?.tematic Sampling System . .... .. ...... .. ......... .......... .................4-10 Figure 4-4-Final Reles',e Survey Grid Designators ......... . ....... . ...... ..............................4-11 LIST OF TABLES Ttble 1-1-Profile of TRIGA* Reactors at General Atomics....... ........ ................... . .. ...... .... ....... ..1 4 Table 2 1-Expected Radionuclides............ ............ . . . ... ...........................................2-6 Table 2 2-Acceptable Surface Contamination Levels............. ......... .. .... . ........... . ................. .. 2-7 Ttble 2-3-Soil and Concrete / Asphalt Rubble Release Criteria......... . ... ....... .. ... ................. ........ 2-8 Table 2-4-Components to be Removed Under Possession Only/ inactive Licenses... . .... ........ .. . 210 Table 2-5-Components with Potential Surface Contamination-Group 1............ ........ ............. .2 15 Table 2-6-Components with induced Radioactivity-Group 2........... .. . ....................................... 2-15 Table 2 7-Reactor Tank Activated Components --Group 3... ..................................... ............... . 2-15 Table 2-8-Equipment Used in Decommissioning Operations-Group 4.. .................... ...... . ......... 2 15 Table 3-1-Specific Health Physics Equipment and instrumentation Use and Capabilities. ................. 3-5 Table 3-2-Preliminary Collective Dose Estimate ............................................. ....... ..... ... ............ 3-11 Tcble 4-1-Typical Background Media Results ........................................................... ... ... .... ... ... 4-3 Teble 4 2-Acceptable Surf ace Contamination Levels ...................................................... ........... .. 4-5 Table 4-3-Soil and Concrete / Asphalt Rubble Release Criteria....... .... ....... ............................ .. .. . 4-7 vi
PC-000482/0
- lST OF ACRONYMS / ABBREVIATIONS ACPR Annular Core Pulsing Reactor ALARA As Low As Is Reasonably Achievable ALI Annual Limit on Intake (see 10 CFR 20)
AMAD Activity Median Aerodynamic Diameter ANSI American National Standards Institute AP Activation Products APPM GA's Accident Prevention Program Manual ARA Airborne Radioactivity Area (see 10 CFR 20)
ASME American Society *of Mechanical Engineers ATPR Advanced TRIGA Prototype Reactor CA Conditional Authorization 6 CAL-DTSC State of California Department of Toxic Substances Control CAL-DHS Califomia Department of Health Services CAL-EPA Califomia Environmental Protection Agency CAL-OSHA Califomia Occupational Safety and Health Act I CAL-RHB Radiological Health Branch of CAL-DHS CAM Continuous Air Monitor CCR California Code of Regulations CDE Committed Dose Equivalent (see 10 CFR 20)
CE Conditional Exemption CFR Code of Federal Regulations em centimeter cpm counts per minute CPR Cardiopulmonary Resuscitation CTI Cryogenic Technology Inc.
D&D Decontamination and Decommissioning DAC Derived Air Concentration (see 10 CFR 20)
DDE Deep Dose Eguivalent (see 10 CFR 20)
DECON Decontamination DNAA Delayed Neutron Activation Analysis DOE U. S. Department of Energy DUT U. S. Department of Transportation dpm disintegrations per minute (measure of radioactivity)
- EBOR Experimental Beryllium Oxide Reactor EDE Eye Dose Equivalent (see 10 CFR 20)
EH&S Environment, Health, and Safety ENTOMB Entombment EPA U.S. Environmental Protection Agency p FFCRs Fuel-Follower Control Rod (s)
FGR Fission Gas Release FLAIR Flashing Advanced Irradiation Reactor FLIP Fuel Lifetime Improvement Program ""
FP Fission Products g gram, a unit of mass GA General Atomics GCFR Gas Cooled Fast Breeder Reactor GERT General Employee Radiological Training GISO GeneralIndustry Safety Orders
- GM Geiger-Mueller HCF Hot Cell Facility HEPA High Efficiency Particulate Air HEU High Enriched Uranium vii
PC 000482/0 HP Health Physics HPGe High Purity Germanium Detector HTGR High Temperature Gas Cooled Reactor LLW Low-Level Waste LSA Low Specific Activity (see 49 CFR)
LSNC GA's Licensing, Safety, and Nuclear Compliance Division MAP Mixed Activation Products . . .
MDCR Minimum Detectable Count Rate MFP Mixed Fission Products micro-R a measure oflow levels of radiation exposure (I millionth of a R or I thousandth of a mR)
MIWP Metropolitan Industrial Waste Program a MkF TRIGA* Mark F Reactor Mkl TRIGA* Mark I Reactor MkIII TRIGA* Mark III Reactor mR milli-Roentgen (unit of radiation exposure) 8 mrad milli-rad mrem millirem (unit of dose equivalence, see 10 CFR 20)
MSDS Material Safety Data Sheets MSHA U.S. Mine Safety and Health Administration mSv milli-Sievert (unit of dose equivalence, see 10 CFR 20) '
NAA Neutron Activation Analysis NCRP National Council on Radiation Trotection and Measurements NFPA National Fire Protectiou Associition NIOSH National Institute for Occupations! Safety and Health NIOSH/MSHA National Institute for Occupational Safety and Health /Mine Safety and Health Administration NIST U.S. National Institute of Standards and Technology NPR New Production Reactor NQA Nuclear Quality Assurance NTS Nevada Test Site NWPF GA's Nuclear Waste Processing Facility OSHA Federal Occupational Safety and Health Acts PB Peach Bottom (Reactor) pCi pico-curie, a unit of radioactivity (2.22 disintegrations per minute)
PCM Personnel Contamination Monitor POL Possession Only License PTS Pneumatic Transfer System PVC Polyvinyl Chloride .
g QA Quality Assurance QAPD Quality Assurance Program Document R Roentgen RA Restricted Area (see 10 CFR 20)
- rad unit of absorbed radiation dose RCRA Resource Conservation and Recovery Act rem Roentgen Equivalent Man (unit of dose equivalence, see 10 CFR 20)
RESRAD DOE Computer Code for Residual Radioactivity Calculations RM Radiation Monitor RO Reactor Operator RWP Radiological Work Permit RWT Radiological Worker Training SAFSTOR Storage SD-DHS-HMMD County of San Diego Depanment of Health Services Hazardous Materials Management Division 1 1
viii 1
mumu PC-000482/0 SDE Shallow Dose Equivalent (see 10 CFR 20)
SNF Spent Nuclear Fuel SNM Special Nuclear Material SS Stainless Steel Sv Sievert (unit of dose equivalence, see 10 CFR 20)
Sxs Samples TEDE Total Effective Dose Equivalent (see 10 CFR 20)
TFFF TRIGA* Fuel Fabrication Facility .
TKF TRIGA* King Fumace TLD Thermoluminescent dosimeter TRF TRIGA* Reactor Facility (GA Building 21)
TRIGA*
TTSL Training, Research, Isotopes General Atomics (registered trademark)
TRIGA ThermalStabilityLab TTSX TRIGA* Thermal Stability X-Ray Room s UC Uranium Carbide USAEC U. S. Atomic Energy Commission USNRC U. S. Nuclear Regulatory Commission WA Work Authorization
?
IX
PC-000482/0
- 1.
SUMMARY
OF PLAN 1.1 Introduction
! Although General Atomics (GA) continues to offer 1RIGA* (Training, Research, Isotopes.
General Atomics) reactors and related facilities, equipment, materials, fuel and services; -
GA has ceased all TRIGA* reactor operations at the GA main site located in San Diego, CA -
(USNRC Licenses R-38, R-67 and R-100). Figure 1-1 shows the regional location of the facility and Figure 1-2 shows the GA site and surrounding uses. GA has decided to shut down the TRIGA* Reacter Facility (TRF) due to reduced demanc for GA's reactor irradiation services. The objective of this Decommissioning Plan is to conduct decontamination (DECON) including dismantlement of the facility and removal of
, contaminated soilto obtain release to unrestricted use by Nuclear Regulatory Commission (USNRC) and State of Califomia of the TRIGA* Reactor Facility and associated . yard area.
The TRF houses three 1RIGA* reactors which have been variously used since 1958 to a
provide controlled neutron and gamma irradiation for diverse research projects. Table 1-1 presents a profile of the three reactors. This Decommissioning Plan has been prepared using the guidance and format of NUREG -1537 Rev. O, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors (Ref.10.1) by GA's Decontamination and Decommissioning Department, which will also implement the Decommissioning Plan, under the guidance of the Licensing, Safety, and Nuclear Compliance Division (LSNC).The final radiological survey will be conducted by GA or ,,
contract survey organization under GA guidance. A confirmatory independent survey will follow if required.
This Decommissioning Plan provides the following:
- A description of the present radiological status of the TRF and yard area e A description of the planned approach to decommissioning the TRF and yard area e Descriptions of the methods that will be utilized to ensure protection of the health and safety of the workers and to protect the environment and the public from radiological hazards associated with decommissioning activities e A description of physical security and material accountability provisions that will be in place during the decommissioning e A description of the intended final radiation survey a A cost estimate for decommissioning the Facility and the source of funding for these y
activities e A schedule for the decommissioning project
. A description of the Quality Assurance Program
. A description of the Training Program e An Environmental Report 1-1
PC-000482/0 '
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2-10 94 l "J/, .. h .' r; % a Figure 1-2-GA Site and Surrounding Uses 13
PC-000482/0 Table 1-1-Profile of TRIGA* Reactors at General Atomics MkF l Mkill item Description l Mkl l General information:
Research. Water-Cooled, Pool-Type, Thermal Privately-Owned Classification:
General Atomics Owner:
TRIGA* Reactor Facility, Bldg. 21,3550 General Atomics Ct., San Diego, CA Location:
Owner Owner Owner Operator:
Owner Owner Owner Licensee:
Ralph M. Parsons Co. Ferver-Dorland Ferver-Dorland Architect / Engineer:
Owner Owner Owner Nuclear Design:
Owner Owner Owner .
Research & Development:
Owner Owner Owner Core Manufacturer:
Owner Owner Owner Construction:
Training, NAA, DNAA. Transient Radiation, Thermionics Studies, n-Radiography g Pnncipal Uses:
Operation:
5/3/58 7/2/60 1/17/66 Initial Criticality:
Active 3/22/95 9/26/73 Date Secured:
R 38 R-67 R 100 USNRC Utilization Facility Lic. #:
50-89 50-163 50-227 USNRC Facility Docket #:
Specifications:
Max. Power, Steady State, MW(t): 0.25 1.5 2.0 Max. Power, Pulsing, MW(t): 1000. 6400. N/A 1.40E+13 3.30E+13 4.40E+13
$ Steady State, (ny):
$ Pulsing (nv): 5.40E+16 1.40E+17 -
4.50E+12 4.40E+13 5.90E+13
$,,,,,,,,,, Steady State, (nv):
$,,,,,,,,,, Pulsing, (nv): 1.80E+16 1.90E+17 -
l Specific Power (kW/kg '"U): 80 420 560 Power Density (kW/l): 3.5 20 26.7 Fuel Material: UZrH, or UZrH,. UZrH,, UZrH,', i Fuel %U, (wt-%): 8.5 8.5,30 8.5 20% 20 %,70 %,93 % 20%,70 %
'"U Enrichment (%):
Fuel Geometry: Cylindrical rods, 1.42"(3.61 cm) dia. x 15" (38.1 cm) active length Cladding Material: 1100F Al or 304 SS l 304 SS l 304 SS !
Cladding Thickness: Al clad: 0.03" (0.076 cm); SS clad: 0.02" (0.051 cm) .7 Fuel Assembly: Circular array Hexagonal array Hexagonal array Core Active Height: 15" (38.1 cm) 15" (38.1 cm) 15"(38.1 cm) 121 121 "*
No. of Available Fuel Positions: 91 Coolant: Light water Light water Light water Moderator: Light water, ZrH Light water, ZrH Light water, ZrH Reflector: Graphite Water Water NOTE-The profile above relates to the general characteristics of the reactors during the respective periods of opera' ion. In the course of operations, each of the reactors were modified to accommodate utilization of the facility by reactor users, such modifications were carried out by the implementation of appropriate changes to the Technical Specifications, or by application of the provisions of 10 CFR 50.59.
14
PC 000482/0 1.2 Background Site and Facility History -
General Atomics The property, on which is situated the General Atomics 'IRIGA* Reactor Site and Facility, was acquired in 1956 from the City of San Diego, as part of a ~290 acre (~117 hectare) tract, by the General Dynamics Corporation, with the expressed purpose of the establishment of the John J. Hopkins Laboratory for Pure & Applied Science, later named General Atomic Division of the General Dynamics Corporation. One of the first goals of the newly-established General Atomic Division of General Dynamics was the development e of a new family of small nuclear reactors, which could be utilized in both industrial and academic applications for training, research, and isotope production. Between 1957 and 1966 three 'IRIGA* reactors were constructed in the 'IRF. Figure 1-3 shows the 'IRF layout. The rooms are identified by number and are described below.
Offices,21/100,101,103 & 104 Dark Room,21/101 A Mkl Reactor and Control Room,21/102 Tool Shop,21/105 Counting Room,21/106 MkF Reactor,21/107 MkF Control Room,21/108 MkIII Control Room,21/109 MkIII Experimental Area,21/110 MkIII Reactor,21/111 Thermal Stability X-Ray Room (TTSX),21/il2 i
Thermal Stability Lab (TTSL), 21/113 North Entry,21/114 Decontamination Room,21/ll5 Figure 1-4 shows the 'IRIGA* Reactor Facility, associated yard area, and boundary of the decommissioning project.
'IRIG A* Mark I Reactor As part of the reactor development program at GA, it was decided to design, build, and ,
operate a prototype unit on the company's Torrey Pines Mesa site. To this end, in late !
g 1957, GA requested and obtained a Construction Permit and Utilization Facility License ;
from the U. S. Atomic Energy Commission (USAEC) to authorize this activity.
Immediately thereafter, working with the Ralph M. Parsons Company as the o'
Architect / Engineer, General Atomics proceeded with construction of the Isotope Reactor Building, Reactor and supporting systems (e.g., Instrumentation TRIGA ,
and C Forced Cooling System, Water Demineralization System, Vendlation/ Exhaust System, Radiation Monitoring Systems, etc.). Following building constmetion and reactor hardware installation, the reactor was brought to initial criticality on May 3,1958.
Continuously operational since that time, the Prototype 'IRIGA Reactor was later designated as the Torrey Pines 'IRIGA* Reactor, and later yet, as the 'IRIGA* Mark I Reactor.
15
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Figure 1-4-TRIGA* Reactor Facility and Associated Yard Area Boundary 1-7
1 PC-000482/0 TRIGA* Mark F Reactor In March 1960, GA submitted an application to the USAEC requesting a Construction Permit and Utilization Facility License for the Flashing Advanced Irradiation Reactor (FLAIR). These documents were issued to GA by the USAEC and thereafter, working with the Fervor-Dorland Engineering Co., Building 21 was modified by the addition of Rooms 21/107 and 21/108 to house the FLAIR Reactor and Reactor Instrumentation &
Control Systems, respectively. This reactor, which was brought to initial criticality on July 2,1960, was continuously maintained and operated by GA from that time until March 22, ,
1995, when the Utilization Facility License was amended, at the request of GA, to authorize Possession-Only activities. During the operating period, the reactor installation was designated as the Advanced 'IRIGA* Prototype Reactor (ATPR) and also later referred to as the TRIGA* MkF Reactor. *
'IRIG A* Mark III Reactor In July 1964, GA submitted an application to the USAEC requesting a Construction Permit and Utilization Facility License for the Thermionic Research Reactor, Following receipt of USAEC approval, GA' and the Fervor-Dorland Engineering Co. proceeded to further modify Building 21, by the addition of Rooms 21/109, 21/110, 21/111, 21/112, 21/113, 21/114, and 21/115, to house the Thermionic Research Reactor and support facilities. This installation was maintained and operated by GA for over 7 years, from initial criticality on January 17,1966 through September 26,1973, when the Facility Utilization License was ;
terminated at the request of GA. The Thermionic Research Reactor, which early on was j redesignated as the 'IRIGA* MkIII Reactor, was situated in Room 21/111 of the TRF where the below-ground reactor tank and various non-fuel reactor core components remain today.
Onwatine Histarv =ad Status The 'IRIGA* MkI Reactor and associated control systems (situated in Room 21/102) are fully operational under USNRC License No. R-38. The reactor fuel and components in the reactor pool are fully assembled and configured for routine operation. A request to amend the license to Possession-Only was submitted on December 17,1996 (R # 10.2).
The 'IRIGA* MkF Reactor (situated in Room 21/107) has been placed in " Possession-Only" status under USNRC License No. R-67 (Ref.10.3), and is currently inoperable. All reactor fuel elements have been removed from the MkF reactor core / shroud and placed in the Fuel Storage Canal. The non-fuel components of the MkF reactor, including the core support structure, bridge shroud, beam tubes, and associated hardware, remain in the p reactor pool. The Fuel Storage Canal portion of the MkF reactor pool currently houses all of the Spent Nuclear Fuel (SNF) elements previously removed from the MkF and MkIII reactors. All required protection barriers and security systems, including those necessary ,
for HEU storage are maintained in accordance with GA's physical protection plan.
The TRIGA* MkIII Reactor (situated in Room 21/111) is inoperable, the associated USNRC License No. R-100 having been terminated in 1975 (Ref.10.4). All fuel elements have been removed from the MkIII reactor core. The MkIII reactor pool has been partially drained of water, and the remaining liquid in the pool has been treated with anti-corrosion agents for long term storage of the remaining activated reactor components still situated in the pool. ]
All TRF building utility services required for normal operation, (i.e., electrical service, l 1
water supply, natural gas supply), are active.
18 l
PC-000482/0
'1RF building ak ventilation and HEPA-filtered building exhaust systems, air supply compressors, and license-required radiological monitoring instrumentation systems normal continuous operation.
All manually-actuated ar.J automateci fire alarm / suppression systems in the TRF are l
operational.
r :. are normal.
Allinstalied1RF security and radiologi',al alarm s, -
Water cooling and demineralization systems servmg both the Mk1 and MkF Reactors remain fully operational.
- 1.2.1 Reactor Decommissioning Overview Prior to decommissioning, the facility will have undergone a cleanout of all extraneous equipment and items except for the fuel. The fuel will be stored in the MkF Reactor stor canal until approval for off- site shipment is received.
Summary of Activities
- 1. Dismantle, decontaminate or package as LLW, the MkIll Reactor components, tank a pit structures.
- 2. Dismantle, decontaminate or package as LLW, the Mkl Reactor components, tank an pit structures.
- 3. Decontaminate any remaining contaminated areas except the MkF Reactor rooms.
- 4. Reroute services to isolate the MkF Reactor and control rooms.
- 5. Dismantle the facility except the MkF Reactor rooms.
- 6. Obtain approval and ship the fuel from the MkF Reactor storage canal.
- 7. Dismantle, decontaminate or package as LLW, the MkF Reactor components, tank an pit structures.
- 8. Decontaminate any n:maining contaminated areas in the MkF rooms and service yard
- o. 9. Dismantle the MkF Reactor rooms. '
- 10. Remove contaminated soil.
~ l 1. Ship the LLW as appropriate throughout the activities.
- 12. Perform the final survey (s) and submit a request to the USNRC and State of Califo for release to unrestricted use.
1.2.2 Estimated Cost Acost estimate was developed for the tasks shown on the schedule (Section 2.3.2), except for removal and shipment of fuel which is not considered a decommissioning task. The total cost estimate is $6,479K which is broken down as follows: f 1-9
~
~
~
PC 000482/0 ILSNC iMark III D&D 960 Mark I D&D 566L {
Mark FD&D 656!
L Other D&D Tasks 775 Outside Contracts l Waste Management 314 280 l Waste Disposal & Shipping
- 140 IQA 92
' Principal Inv.
339 Project Management 339 ilndependent Confirmatory Survey (if required)
I 638 Subtotal lQQ
{ 5,399 ,
Contingency 20%
Total l.080
$6,479K f The estimate for ll.W disposal is based upon the waste being buried* at the NT 1.2.3 Availability of Funds Califomia) licensed facilities and sites n tatein ay 20,1996 of San Dieg submittal also described the method ac ty.by That which GA for funding its total cost of the subject decommissioning.
USNRC acknowledged acceptance of GA's proposal y ,
assurance (Ref.10 the 6)
By letter 1.2.4 Program Quality Assurance The GAManual.
Assurance Quality Assurance (QA) program is described in the GA Co rporate Quality assurance regulations and standards:The GA wQuality ng quality Assurance pro Code of Federal Regulations Title 10, Part 71 (10 CFR 71)
- Transportation of Radioactive Material, Subpart H, " Quality Assurance " ,
Nuclear Facilities."ASME-NQA-1-1989 (Ref.10.7), " Quality Assurarg:
. Program Requirements for Transportation and Storage Inspection e ce, Nuclear C Section,
-. S Matenals expiration Safety date June and 30,2001 Safeguards, (Ref.10.8). Approval No.,1996, 0030, Revision 6 is described in a Quality Assurance c or Program prepared Facility for QAPD the Docu Decommissioning of the TRIGA* Reactor Facility. The invokes the QA Manual on this project and provides project-specific se of the GA Organization, and the QA measures applied to surveys, material shipments, and waste cenification.
dismantlement, radiological planning, QA requ Consistent with the QA Manual and USNRCendix Regulatory Guide 710 A 10.9), the QA program is applied to the variousapproach, project activities in a gra A (Ref.
i.e.,
l 1-10
PC-000482/0 f
the QA effort extended on an activity is commensurate with its importance to safety and its l impact on project goals. ;
The relationship of the QA function to the dismantlement organization and to facility !
management is shown in Figure 2-8. l t
Audits. Insoections. and Manaoement Review F Formal Quality Assurance audits will be performed annually in accordance with ASME- I NQA-1, to verify compliance with the 'IRIGA* Reactor Facility Decommissioning quality ;
assurance program and to verify its effectiveness. These audits will be performed in accordance with written checklists by personnel who do not have direct responsibility for i 4 performing the activities being audited. Audit reports will be distributed to responsible :
management, up to the Senior Vice President level. Follow-up action will be taken, where !
indicated. .
r
- Project technical assessments and QA surveillances will be performed frequently to assess !
compliance with procedures, including Health Physics (HP) procedures. These assessments will be coordinated by the project QA manager. The assessment team will ,
consist of quality assurance and technical personnel. Assessments will be performed in l accordance with a written plan. Assessment reports will be approved by the project QA 3 manager and distributed to the project manager and other project personnel. Follow-up !
action will be taken, where indicated.
Inspections will be performed on procured and fabricated items to verify compliance with !
controlling documents. Inspections will be conducted by qualified inspectors in accordance j with inspection plans prepared by a quality engineer. Disempancies will be documented in a Nonconformance Report, which will be dispositioned by a quality engineer, or a Material Review Board, as appropriate.
' Additional assessments or management reviews will be performed. when deemed [
appropriate by the Project Manager. Such assessments may include Readiness Reviews i pnor to start of a new task, or Management Assessments. [
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- a. 3 i
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1-11 i
PC-000482/0
- 2. DECOMMISSIONING ACTIVITIES 2.1 Decommissioning Alternatives The objective of the 'IRIGA* Reactor Facility Decommissioning Project is to obtain regulatory release of the facility and associated yard area which is shown on Figure 1-4 to unrestricted use. On this basis safe storage (SAFSTOR) or entombment (ENTOMB) were considered inappropriate to GA's future plans.
SAFSTOR poses essentially the same potential risks and environmental impacts as the proposed project, but potentially for a much greater period of time. This altemative would
- necessitate continued surveillance and maintenance of the TRF over a substantial time period. During this period, the risk of environmental contamination would continue to exist. Moreover, development of the land around the GA site over the next few years may significantly incmase the local employment population density and incmase potential for
-, public exposure.
ENTOMB would necessitate continued surveillance and maintenance of the TRF over a substantial time period. During this period, the risk of environmental contamination would continue to exist. Moreover, development of the land around the GA site over the next few years may significantly increase the local employment population density and increase potential for public exposure.
DECON is the option chosen. To the extent possible, decontamination of facility equipment and structural components will be conducted to minim 'e radioactive waste. Dismantlement of the facility and other site structures will be perkmied and the soil remediated as necessary. This would be followed by USNRC and State mspections and release of the site to unrestricted use.
2.2 Facility Radiological Status 2.2.1 Facility Operating History 1RIGA* Mark 1 Startup: 5/58 Shutdown: N/A; USNRC Utilization Facility License. #R-38 presently Active (Possession Only status requested 12/96, Ref.10.2).
Max. Power: 250 kW(t) Steady State ,
.. The Mark I 1RIGA* Reactor was orig *inally constructed by GA to prove the inherent operational safety of (U,Zr)H TRIGA fuel matrix. Figure 2-1 provides a listing of operations conducted in the GA Mkl Reactor. Through December 1996, the accumulated operations of the TRIGA* Mk1 Reactor have totaled approximately 84 MW-days.
2-1
N PC-000482/0 TRIGA* Mark I:
Startup: 5/58 Shutdown N/A, USNRC license presently active Max. Power: 250 kW(t) Steady State 5/58 Ra6aten Streaming thru Dry Tubes 6/58 Reactor Transient Expenments 7/58 Irra&ation of UC in Graphite for FP Diffuaon 8/58 Transient Reactivity Compensation 12/58 Subentical Assemblies 4/59 erramation of PbTe Thermoelectnc Elements 4/59 Determinaton of Temperature Coefficient 4/59 Transient irradiations by Diamond Ordnance Fuse Lab & Signal Corps 12/59 HTGR Fuel Compact Ra$ation Flash 2/60 Irradation of Thermoelectne Expenment 4/60 trra$aton of Cs Cell for Thermionic Direct Conversion 5/60 initial Expenments with in-Core TRIGA' King Fumace 3/61 Expenments with Hgh Hydnde SS Clad Fuel 3/61 Continued Cs Cell irradiation
- 8/61 Irradiation of Semiconductors 11/61 initial Stearate Run in TKF
- 2/62 Generaton of *'Ar gas for HCF Stack Monitor Calibraton 11/63 Pilot irradation of EBOR 2F18xs 2/64 Epithermal Neutron Absorption 6/64 Continued TKF Stearate and FGR Studies 7/64 i Transport Expenment 10/64 in-Core irradiation /Finng of Explosive Actuators e 1/65 trradiaton of Fassionable Matenals in Sealed Canisters 5/65 in-Core Irramation of 'LI 9/65 Modify TKF for Putsing d
3/66 Irradiation of Elemental Na 1/67 trradation of Aqueous Pu Solution 2/68 Fisson Product Decay Rate Expenments 8/69 Irradiation of Cartndge Pnmer in Pneumatic Rabbit
- d. 3/70 Irradaten of Encapsulated ""U, '"U, "'Np and '"Pu 3/70 Use of Graphite Pouch for TKF trradiations 10/70 Irradiation of F 29 Capsule Rods in TKF 2n1 in-Core Oscillator Measurements 3/71 Use of Shielding Matenals in TKF 5/71 trradiation of UZrH in TKF 6/71 Amtnent Temperature TKF trradations 8/71 High-Capacity TKF runs Out of-Core y 931 Simultaneous Operation of Muttspie TKFs 10/71 Installaten of Auxiliary Cadmium-Shielded Pneumatic Transfer System 192 Installation of New Lincar Channel Component 6/72 installation of New Servo-Controller Linear Channel 692 Removal of Freon Cooling System Hardware 8/72 TKF 1rradaten of High-Bumup Fuel Spectmens 3/73 Wet Helium TKF Purge Test Expenments 12n3 installation of "8 Shielding Pneumatic Transfer System Terminus 4/74 Automated TKF Control System installation -
4 4/74 Fuel Hydrolysis Expenments in TKF 2/75 Use of AmBe Starter Neutron Sources 535 Irradaten of Hghly-Radioactive Specimens in TKF 11/76 Irradaten of Highly-Ramoactive Specimens in TKF 4/76 Use of No Purge Gas for TKF 1rradation 12/76 trradaten of PB Compacts in TKF 2/80 installaton of New Operating Consolo 6/80 installation of Uninterruptabfe Power System 7/80 Use of "B Shielded PTS at Higher Power Levels 4/82 Hgh-Pressure Expenments to Determine Diffusion Rates 7/84 Installation of Microprocessor Based Control Systems -
Digital Control Console Phase il 11/85 Pressure Vessel for Neutron Pulse irradiaton Tests 3/86 Pulse-trradiaton of *"U-Doped Concrete 9/86 Dgital Control Console installation Phase til 4/07 installation of PVC Piping in Cold-Leg of Cooling System p 12/87 Installation of Steppino Motors for Control Rod Drive -
2/88 Ogital Control Consofe installation Phase IV 6/88 Square Wave to 250 kW SS Operation Expenments 6/88 Phase V Digital Controt & instrumentation System 1/89 Installaten of Ground Fault Detector Test Circuitry 5/89 Installation of New Scram Timer Circuit 6/89 ..
installation of New CAM 8/89 Irra$ation of NPR Lithium Target Specimens 6/90 installation of New Reactor Core Top Gnd Plate 2/91 Extension of NPR Uthium Target Irradiatons 3/91 Multiple Pulsing Program 3/91 Mo@fication of NM1000 Wide-Range Power Channel 5/91 installstion of NFC-1000 Flux Controller 8/91 Installation of Loss of Magnet Voltage Relay in Scram Loop 3/92 Installation of Large Diameter in-Core irradiation Dry Tubes 9/92 Modification of Digital Communications Network 12/96 TRIGA' Reactor Demonstratens and Operator Training Figure 2-1-TRIGA' Mark l Chronology 22
l PC-000482/0
'IRIGA* Mark F Startup: 7/60 l
Shutdown: 3/95; USNRC Utilization Facility Lic. #R-67 presently limited to Possession Only Max. Power: 1500 kW(t) Steady State The GA TRIGA* Mark F Reactor was originally constructed by GA as a prototypical testing reactor, to act as a proof test reactor for the TRIGA* Reactor supplied by GA to the l
Defense Atomic Support Agency, Bio-Medical Radiation Research Facility, National Naval Medical Center, Bethesda, MD, under Contract No. 27757, dated 4/4/60. In April,1961, while continuing to operate extensively for the Bethesda testing campaign, GA began to utilize the Mark F Reactor on a multi-user basis for several other irradiation experiments, ,
- the most conspicuous of which were the In-Pile Thermionic and Thermoelectric Power :
Conversion Projects and related experiments. Another primary user of the reactor, j beginning in June,1963, was the Department of Defense Special Weapons Testing Center, Los Alamos, NM, involving the in-pile survivability testing of high-explosive ordnance for i the U.S. Army. Figure 2-2 provides a listing of the experiments performed using the l TRIGA* Mark F Reactor. Integrated operation of the TRIGA* MkF Reactor is estimated to be 4,200 MW-days.
'IRIGA* Mark III:
Startup: 4/66 Shutdown: 9/73; USNRC Utilization Facility Lic. #R-100 presently Inactive. ,
1 Max. Power: 2000 kW(t) Steady State ;
i , Constructed by GA as the Thermionic Research Reactor, the GA Mark III TRIGA* Reactor l essential!v roerated at full licensed steady-state power throughout the over 7 year reactor history to: me specific design purpose ofin-pile irradiation of thermionic power conversion experiments. The reactor core configuration was in fact designed around the thermionic cell
! geometries, to attain maximum neutron fluence levels in central core positions. Initially operated at 1500 kW(t) maximum steady-state power, the Mark III was reconfigured in 1970 for higher power levels, and operated thereafter at 2000 kW(t) maximum steady-state l c, power.
At the same time, other experiments utilized the Mark III Reactor on a multi-user (piggy-back) basis, for a broad range of out-of-core irradiations. The most frequent simultaneous l"
user was the DOE-sponsored HTGR Fuel Development Project efforts, which constructed special in-pile " King" furnaces at the perimeter of the Mark III core to execute advanced fuel performance testing, (i.e., fission gas release measurements). Integrated operation of
, the TRIGA* MkIII Reactor is estimated to be 1,600 MW-days.
l 1
2-3 l
PC-000482/0 TRIGA* Mark F:
Startup: 7Ai0 Shutdown: 3 416 Max. Power: 1500 kW(t) Steady State 660 Construebon Authortzed 720 Reactor Startup Expenmental Program 960 BiologicalTargets Authonzed 441 Therrruonic Direct Power Conversion 561 Support of Harry Diamond Laboratory Forest Glen, MD 861 T6, AL c Device Targets Authorized 1061 Uranium-Zirconium Carbide Thermionic Emitter 11E1 Core Loaded with SS Clad ZrH , Fuel Elements 1121 Support of AFRRI (Armed Forces Radiabon Research Institute), Bethesda, MD 1222 Supoort of Sandia National Laboratory, Albuquerque, NM 243 Pulsed, Fueled Expenments Authorized ,
463 Puisog Program Contnuing, to $4.00 6k/k inserbon 663 in-Core irradebon of Explosives, Harry Diamond Laboratory, Forest Glen, MD 7E3 installation of 2"'Trarment Rod 863 Fuel Fission Product Release Studies 963 trradiabonof Ultrapure Al =
1123 Irradiabon of Xylenes for Radiabon Damage Studies 364 Installation of FFCRs and D-Rog Transient Rod 464 Irradabon of AgZrWOH Battenes 664 Irradabon of Laser Rods, Squibs, & Electrorucs 964 Thermsoruc/ Thermoelectric Device Targets Authorized 366 trradiation of 'LIF and Red Phosphorus, Sanda Nabonal Laboratory, Albuquerque, NM 1146 Expenmental Pulsing Program Authorized 246 Expenmental Pulsing Program Authorized 366 Fueled TKF Expertments Authorized 366 TKF HTGR Development Work Authorized 866 Pulsing Program Conhnuing, to $4.60 Sk/k inserbon 7E7 ACPR Puisog Conflguration Authorized 7E7 in-Core TKF Authortzed 867 Roubne Use of TKF for SS & Pulsing irradiation 868 Uranium Vaponzabon from Seawater Meda 868 600 Ciy trradiabon Expenrrent 1068 installabon of FLIP Elements 120 Irradiabon of "U Doped Concrete 129 Modficabon of Target Thermionic Device Design 2n2 Explosive Targets Authortzed 373 ConhnuedTKF Toshng 4/73 installabon of Mark 111 Fuelin Mark F Core 4/14 Reconfiguration of Mark F to Small Core High Power Density 1/75 InstaAabon of Mark til Console to Mark F 4r75 Roubne Puisog Program Developtrent 5775 Pulse Irradabon Tesbng of New TRIGA* Fuel 12775 Conhnued TKF Testng 1076 Cold Neutron Radiography of Zr Program (CTI Nuclear) 11//6 Cryogenic n Moderator Use Authorized (CTl Nuclear) 2r77 Toshng of 1/2' dia. Fuel at Steady State 12/77 Continued operations for CTI Nuclear and TKF 4/78 Continued Tesbng of 1/2" dia. Fuel for Romania TRIGA* Reactor 678 Teshng of Central Flux Trap in Pulsing Mode 11A10 Small Core Pulse Testing 562 High Pressure TKF Testng 1062 1sotope Produchon of Ar & "Br ..
7E3 Fuel Perfotmance Evaluation and improvement Expenments 964 Reactor Modificatiorts for Thermionic Device Irradiabon 568 Misc. Modifications for Continuing Thermionic Device Irradabon 10R3 Conbnued Therrruaruc Device Irradation
'395 Shutdown Figure 2-2-TRIGA' Mark F Chronology 2-4
PC-000482/0 Current Radiological Status of the Facility 2.2.2 General levels Routine radiological surveys show that the radiation d din levels and contamina l measured at the 1RF have consistently i f the been low. A r residual activity am present. This information indicates that the radioactive p facility are confined to the reactor intemals and biological shield.
Estimates of the radioactive inventory can bet the determined neutron l
by consider elements of the materialin question and calculating the duration of exposure o flux and the energies of the incident neutrons. Direct l t of measurements more reliable and will be used extensively during actual removal and/or dis components. d fully
" A detailed baseline survey will be performed as the ill funher first define the D&D task to verif characterize the radioactive materials inventory. This information w basis for specifying the necessary safety measure exposure to personnel is maintained ALARA.
Principal Radioactive Comnonents This section is based upon process knowiedge
- Mark i Reactorand at theis consistent wit as a result of actual decommissioning experience of a TRIGA University of Texas (UT).
The most highly radioactive component is likely to be the Rotary Spe Mark I Reactor which may have a dose rate of = 16 R/hr at I ft.
Other components which may range up to = 5 R/hr at the surface are:
- Upper and Lower Grid Plates e Core Support Structure e Graphite Reflector Assembly Miscellaneous fasteners, especially stainless steel bolts, nuts, helicoils, etc. ,
g Similar components in the MkF and MkIII Reactors are expected to reduced levels of activation due to decay.
Radionuclides The radionuclides expected to be present in detectable levels due irradiation experiments performed are shown in Table 2-1 below.
2-5
~
PC-000482/0 Source I Table 2-1-Expected isotope ' Half-Ufe (yr) Radionuclides AP "Co Notes 5.27 _
- Nb Predominant AP specie still present.
20.000 "2Ta 0.32 Minor AP constituent, half-life may affect disposition.
FP 85 Kr Specie has nearly decayed away.
10.72 "Sr 29.0 Gaseous specie; any remaining inventory in intact fuel.
- Ru 1.02 Beta emitter only; requires chemical analysis.
'85Sb 2.75 Specie has nearly decayed away.
- Cs 2.06 Specie has nearly decayed away.
"7Cs 30.17 Major constituent of FP inventory.
- Ce Predominant FP specie still present. .
0.78
- Eu 8.59 Specie has nearly decayed away.
"5 Eu ; 4.71 Minor constituent of FP inventory.
i Both the fission-product and activation-product isotope spMinor constitu environment, will primarily be limited to those solid MFP/ MAP speciecies, present a 3
Specific activation-product and fission-product es withspecies half-livese> 200 days. (= 0.55 yr).
Building 21 residual surface contamination, ,
xpected to be present atthe willinclude radi levels in detectable 2.2.3 onuclides hsted above.
Release Criteria-Residual Radiation s and Contamination Level This section provides bases and s Facility structures, and the soils.pecific criteria forrelease of materials and equip for release to unrestricted After use decontamination, the structures will be dismantled ,
soils) itself. The survey approa. The Final Release Survey will be for the pro 5849 (Ref.10.10) and ch will use the guidance is described . in Section provided 4 in Draft NUREG/CR-Materials and Eoyipment Release Criteria All materials leaving the Restricted Ama will be surveyed to e are not inadvertently released from the Facility HP procednsure that licensed performing these evaluations. These evaluations. uresevaluations will be adhered towill o owing types of in include the f l for smearable activity evaluated or and/orwith portable or fixe indimet survey instruments or counters will also be emprocess knowledge ,
could su the criteria specified in HP procedures. ployed. Materials and equipment o will be rel criteria for fixed and smearable Those criteria residual eareradioactivity summarized would in Tableb b 2-2. R concentrations of isotopes on the material for beta-gamma emittased upon t -
category of nuclides for beta-gamma emitters ers from apply.
if moreTable than one2-2 would 2-6
PC 000482/0 Table 2 2-Acceptable Surface Contamination Levels
- 2 (dpm/100 cm pa l
Nuclides*
l Average' l Maximum" l Removable
- 15,000 1,000 5.000 U nat,"U,"U, and associated decay products 300 2D f
100 Transuranics, 82*Ra, 22'Ra, 23"Th, areTh, 22'Pa, r2'Ac, '25 , '2 1 3.000 200 1,000 1,000 Th nat. 2"Th. "Sr,22'Ra, a24Ra. 8 20, i2eg, usi, u'l 5.000 15.000 Beta / gamma entiers (nuchdes mth decay )
modes other than alphaemssion d or spui s sous fissi
- Wnere surface contamnat!on by both alpha-and beta / gams-errutung dnuchdes exists, he hmits by beta / gamma-emtbng nuchdes should apply independently. t e factors correctng the counts per nrkste observed by and less appropnate surface detector f associated mth the instrumentabon.
- Measurements of average whi.cmig should rot be averaged over more than i square meter. For obje area, the average should be derived for each such object. 8 eth dry 8 The maximum contamnadon level apphes to an area of not more than 100 cm . t of The amount of removable radioactve matenal dh per 100 ble contamnabon t e surface should cn cm' of surface a radioar*ve matenal on the wipe mth an appropnate instrument tt of known e
. be sped. 7 milhgrams
' The average and maximum radiaton levels associated with surface contamnaton resulting fr should not exceed 02 mradhr at 1 cm and 1.0 mradhr at 1 cm, respectively, t nation measured through not m per square centmeter of total absorber.
- The Decommissioning Project HP Manager's approvalis required before matenal may be released i levels are en excess of the value stated in the " average" column, even it below the "mammum" column.
- Including contamnaton by induced radioactvity,i.e., actvabon.
In evaluation of equipment and materials for fixed or smearable licensed materials painted with other than original manufacturer's paint will not be released u process knowledge demonstrater that the paint was applied to a clean, n surface prior to use in the Restricted Area, (2) the paint is removed, or (3) He approved paint sampling survey demonstrates that radiation h levels under the release criteria. If the potential exists for contamination on inaccessible surfaces, equipment will be assemed to be intemally contaminated unless (1) the eq dismantled allowing access for surveys, (2) appropriate tool or pipe monitors wi acceptable detection capabilities are utilizedl that would accessible areas are representative of the inaccessible s swabs would be representative of the inaccessible areas).
Bulk Materials or Bulk Liquids-Analysis of representative sample (s) with high res gamma spectrometry system. Bulk materials (not to include concrete or a which is addressed separately below) or bulk liquids will be released if no discem
- - facility-related activity is detected. Minimum Detectabl for principal isotopes in release evaluations.
27
___- - - - - ~_ -
PC-000488/0 6sotope Table 2 3-Soll and Concrete / Asphalt Rubble Release C it r eria' Release Cntena Based upon Extemal Exposure Umitations Release Criteria Based upon internal "Go (pCi/g) Exposure Urnitations i3*Cs 1
8' (pCi/g)
'"C s l 10 I
'6'Eu l 15" l
! l
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u l 10 l
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'26 l Sb 7.5
- Sr l 37
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'*'Pu 27' ,,
l 27*
- 'Pu j l
'*Pu 4326" ~
' l l
[Am l 288 l 2@ ,
Natural Uranium
{ l Depleted Uransum 2 58 E l 10' ]
nnched Urarium (22*U and 2"U) 35' Thonum (2"Th and 83'Th) ( (
s 30' Thr release entena shown in this table witruut annotaton l by footnot 10' 5.18 adhenng b the same assumptone that were provided in the coes 2,3, or4 wer extsmal exposure rate above background rmspo&w hsted in rxxe a
y itselfthat for2. the wouldhave bekm. maximum This corresponds year grve approximately b of exp concrete. asphatt. or similar construchon media matenals .
been g 10 pR/hr 6
USNRC for the Hot Ce!! Decommissioningrourx! protecttobyletter athrough coarse dated rubble.1These May 1996 enteria were app These release enteria are based upon past precedent ,
, Robert C. Pierson USNRC to K. E. Asmussen.
8 See Correspondence 70-734: Plan for Obtaining Release K. E. Asmussen of Certain Areas to W.to UnrestnctedT. Crow, dated U October 1 1985and Sta release enteria are based @cn pastse?pro, dent correspondence identfacaton 696-8023, Sub}ect " Docket These5,1981 October estabhsned b st operatons
- present)frompa. Subsect " Disposal or on-site s1orage of ressdual thonum or uraniu or without daughters
- Numbers were estabhshed using, the most hmiting of lung dos Factors from NUREG/CR-0150. Volume 1.0.
y 2, with an s'pha quakt ta t c ord20, where apphcable, lung (nass of 580 grams. and AMAD d Facility R&ase Criteria The proposed decommissioning attemative that has been p
{ the site and facility mdiological scoping ac y. The results characterization of i
may be directly releasable without need for decontamination Deconta i performed on specific areas or materials in morder to beachieve unrestri nation will reducewhen operation radioactive it can be demonstrated waste that volumes.
the Building service systems will on ase e removed from and to -
safety or efliuent/ exposure control compensated for by other means. functions, y are no longer needed to provide completely evaluated due Some areas and components to inaccessibility, hence all areas of the ..
Facility
/ items will w b
ability to decontaminate impractical or not possible to satisfy release crit and ability to demonstrate e evaluated satisfaction for of relea se criteria. If it is have been met), the location / item will be treated nated.
as radioactively contam As was discussed previously, GA will utilize established release criteria as provided in Removable surface contamination ss oning will y wiping or other Project.be eliminate proven methods. Release criteria for fixed and smearable y for beta-residual radioa 2-6
_._.m.__ _ _ _ _ . _ . _ . . - _ . . . - _ . ~ - _._. _ _ _ _ _ _ _-__-_ _ _. __
l I 3
PC 000482/0 gamma emitters would be based upon the relative concentrations ofisotopes on the material l and their respective release criteda if more than one category of nuclide for beta-gamma emitters applies from Table 2-2. ;
! GA may apply the release criteria in Table 2-3 to evaluations of representative samples of i ~
asphalt, concrete, or other similar construction media that have been reduced to rubble.
Concrete slabs may be released based upon demonstration of conformance to Table 2-2 or l 1-evaluation of representative samples by gamma spectrometry showing results below the criteria in Table 2-3.
$ Soil Release Criteria i A summary of the proposed soil release criteria is provided in Table 2-3.
! In situations where more than one isotope is detected in the soil, determination: of conformance to release criteria would be made according to the following method fo< the j* mixture:
l j
b+b+..+bst G, G G, 2 j 1
t where . '
C,, C 2, ... n is the concentration of nuclide 1,2, ... n in the soil above background values G,, G2, ... n is the release criteria of nuclide 1,2, ... n Table 2-3 shows release criteria based upon the most limiting pathway for nuclides that have been detected in soils or in the Facility during the radiological scoping characterization (or those that conceivably could be encountemd). If additional nuclides are encountered ,
during the remediation or Final Release Survey activities, their respective release criteria !
would be determined in the same manner as the values provided above. !
2.3 Decommissioning Tasks 2.3.1 Activities and Tasks !
l Preparation of the Facility for Decommissioning Upon-receipt of the Possessioa Only License amendment for the Mark I Reactor, (anticipated in April /May 1997),the Mark I fuel will be moved to the Mark F fuel storage l
canal pending approval to ship off-site. A scoping characterization study has been i
.. conducted using surveys, calculations, and process knowledge to determine the extent of contamination in the facility and the surrounding yard, (see Appendix A for a characterization summary report).
The yard and facility will have been cleared of any remaining extraneous equipment and materials before commencing the decommissioning tasks. Reactor components to be removed under a Possession Only/ Inactive Licenses are shown in Table 2-4.
2-9 m - - . _ - . _ _ , _ - - ._ , _- -
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l PC-000482/0 I
Table 2-4-Components to be Removed Under Possession Only/ inactive l Licenses TRIGA* graphite dummy elements Pneumatic core terminal and control system and tubes Unused fuel racks Fission chamber neutron detectors lonization chamber neutron detectors
%nset rod drive mechanism and control rod Motor drive mechanisms and control rods j Heat exchanger and primary piping !
Neutron sources Void tank (Mkill only)
Dismantling and Decontaminatine the Facility The decommissioning plan consists of several activities oriented towards the sequential dismantling of the three reactors, the reactor pits and liners, and any associated systems, in a safe manner and in accordance with ALARA principles, and finaDy dismantling of the .
entire facility. There are three distinct sequences in dismantling the reactors. First the Mark l III Reactor will be dismantled followed by the Mark I Reactor and then the Mark F Rea.ctor )
(after removal of the fuel from the storage canal). Figure 2-3 depicts the three i decommissioning phases. Views of the three reactors are shown in Figure 2-4, Figure 2-5 !
and Figure 2-6.
The initial activity will be the radiation survey of the specific reactor room and the l immediate ad.jacent areas. Components above the pool will be removed including pool deck l plates, bridge structure, cables and conduits, and the rotary rack drive dismantled and decontaminated to the extent feasible. Reactor components with induced activity will be removed using grapples and placed in a shielded container for disposal as LLW. This will be followed by a survey and discharge of the reactor pool water. The dismantling of the reactor tanks and pits will proceed after installation of a confinement barrier in the reactor room and a dedicated ventilation system to prevent the spread of airbome contaminants.
The tank will be sectioned in place, removed from the pit and the contaminated sections packaged for disposal. Surface and coring sunples of the concrete biological shield will bc ;
performed to determine the contaminated areas. The contaminated sections will be cut away and packaged. The remains of the concrete pit will be broken up and removed to permit sampling and analysis of the surrounding soil. Shoring and covering of the pit will provide ,
protection until the building is demolished and the final survey performed. The remaining '
tasks are: dismantlement of the confinement barrier, removal of any remaining surface 8 contamination in the rooms, final survey of the rooms, and dismantlement of the rooms.
The packaged waste is to be shipped to the Nevada Test Site (NTS). A final survey will be performed after complete building dismantlement and a request made for release to unrestricted use.
2-10
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i PC-000482/0 Dismantline Seauengg Dismantling will occur sequentially to the detailed schedule shown in Section 2.3.2. Items j for each of the three reactors will be grouped as follows:
i Group i Equipment which does not have induced radioactivity but which may have surface contamination.
Group 2 Core components and other components which have induced radioactivity
- (excluding the reactor tank).
Group 3 Reactor tank liner, anchors and concrete in the proximity of the former location j of the reactor core and which have been neutron activated.
Group 4 Equipment tools and systems which have been contaminated during decommissioning operations.
1 Components and equipment in the four groups are identified in Table 2-5, Table 2-6, Table 2-7 and Table 2-8.
Table 2-5-Components with Potential Surface Contamination-Group 1 Purification System purification loop and deionizer tank piping demineralizer Other Components cables and conduits pool deck plates rotary rack drive (Mkl only) reactor bridge structure pneumatic transfer system Table 2-6-Components with induced Radioactivity-Group 2 -
Rotary specimen rack (Mkl only)
Control rod guide tubes and detector tubes (Mkl and MkF only)
Top grid plate Bottom grid plate Reflector (Mk1 only)
Core support Fasteners and connectorc Pneumatic transfer system terminus (Mkl only)
Table 2 7-Reactor Tank Activated Components-Group 3 Reactor pit liner o* Concrete Anchors Reinforcement bars Table 2-8-Equipment Used in Decommissioning Operations-Group 4 General ventilation system Localized ventilation system Confined barrier Contaminated tools i Contaminated clothing 2-15 l
l
PC-000482/0 -
I The Rotary Specimen Rack in the Mkl Reactor pool is expected to have the highest induced j radioactivity. This expectation is consistent with the University of Texas experience in a ,
similar 'IRIGA* Reactor decommissioning project. The Rotary Specimen Rack and other Group 2 items will be hoisted from the pool and lowered into a shielded steel box which will have been prepared to accept the items. Additional shielding will be provided for worker protection.
After components, equipment and parts listed in Table 2-5 and Table 2-6 have been removed, a confinement barrier will be installed. The purpose of this barrier is to contain airborne contaminants generated during reactor pit demolition, and to prevent their spread in the Reactor Room and possibly in the surrounding areas.
The confinement barrier will consist of a plastic enclosure on a rigid frame which will ~~
surround the mactor pit. Associated with this enclosure will be an independent localized ventilation system which will ensure a negative pressure with respect to the Reactor Room while providing high efficiency filtration on the exhausted air, and a source of clean air ,
supply within the enclosure. .
The Reactor Room will be maintained at a negative pressure with respect to the surrounding -
areas but less than the pressure diffemntial mamtained between the confinement barrier and the Reactor Room.This will ensure that the air will travel from the non-contaminated area to the increasingly contaminated areas.
The activated liner section will be cut using a rotary saw or other conventional equipment and the activated thickness of the concrete will be broken off with a jackhammer. To minimize dust dispersal, a localized fine water mist will be sprayed over the area being demolished. Activated concmte will be removed a section at a tune and supports will be placed in the cavity formed as needed, before proceeding with the next section.
' At the completion of activated concrete removal, dose rate measurements will be made to !
determine if all necessary portions have been removed. As the demolition of activated material proceeds, the radioactive material will be packaged for shipment and disposal.
There am two potential safety concerns during performance of this task: 1) extemal ;
exposure from the activated components of the tank, and 2) inhalation of airborne material.
To minimize the risk, work areas will be monitored frequently and radiation levels will be monitored continuously, to determine sudden changes in the radiological conditions.
i Upon completion of dismantlement tasks in the reactor pit, the confinement barrier will be dismantled and the plastic sheets compacted and packaged. Surface contamination will be >
removed from contaminated portions of the ventilation system and they will then be packaged for disposal. Contaminated clodiing will also be disposed of by compacting and appropriate 1y packaging. The reactor room will then be cleared and all surface contamination removed.
Surveys Radiation surveys of each of the reactor rooms and other applicable locations will be performed prior to building dismantlement. After all of the building structures have been removed and dispositioned as clean or radioactive waste based on survey results, a final survey of the site (soils) will be performed and a request made for release to unrestricted use.
2 16
PC-000482/0 2.3.2 Schedule The project schedule is presented as Figure 2-7. The dscheduled use time from r approval of the Decommissioning Plan to submittal for release of the site to is 22 months. It should be noted that fuel stored in the MkF Reactor storage c dl I
scheduled for timely removal and shipment off site. If the slippage will occur until such time as the fuel is shipped off the GA site.
2.4 Decommissioning Organization and Responsibilities GAis committed to, and retains ultimate responsibility for full compliance with th USNRC .and State licenses and the applicable regulatory requirements during decommissioning. Company principles, policies, and goals will be followed to e standards of performance in accomplishing the decommissioning tasks.
- GA has an established D&D organization which is currently working on D&D of Hot Cell. A similar organization will be assembled for the 'IRF D&D (Figure 2-8) functions of the organization staff are described below. (Note that while functio maintained, individuals performing the functions l be may v advice during decontamination and decommissioning. Specialized contractors w utilized when required under the direction of the Site Decommissioning Manager.
Key Positions Project Manager-The Project Manager has the overall responsibility for su completion of the Project. The Project Manager functions include:
Controlling and maintaining safety during decommissioning activities and p the environment
- Determining project staffing and organization
- Assuring performance to cost and schedule e Reporting of performance e Approving decommissioning plans and procedures e Approving subcontracts
- Approving budgets and schedules e, assistance of the LSNC Division, that the conduct of a
Ensuring, with thedecommissioning activities complies with all the appropriat accordance with die GA licenses.
e The minimum qualifications for the Project Manager are:
e A four year degree in engineering or science e Ten years of project management experience in the nuclear innstry includ decommissioning projects
- Familiarity with the TRIG A* Reactor Facility
- Appropriate training in radiation protection, nuclear safety, hazardous l
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and industrial safety 2-17
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PC-000482/0 Princinal Investigator-The functions of the Principal Investigator include:
Maintaining the TRF and services in a safe operating condition during decommissioning
- Approval of plans and procedures e
Providing engineering support for the decommissioning activities The minimum qualifications for this position are:
A four year degree in Engineering, or a related field e
Five years of experience in a nuclear facility e
y Substantial knowledge of the TRF equipment and services Health Physics Manager-The HP Manager is responsible for providing radiological safety support in the decommissioning of the Facility. This function ensures that activities
, involving potential radiological exposure are conducted in compliance with the GA licenses, Federal and State regulations, and GA procedures. The position includes responsibility for maintaining a surveillance program and identifying, quantifying, and classifying radioactive waste. The development of HP procedures is also included in this function.
The minimum qualifications for this position are:
- A four year degree in HP or a related field
. Three years' supervisory experience in HP e
Ten years' total experience related to radiation safety Certification as Health Physicist, or equivalent education and professional experience in HP desirable Environment. Health. and Safety Engineer-The Environment, Health, and Safety (EH&S) Engineer is responsible for compliance with Federal Occupational Safety and Health Acts (OSHA) and California Occupational Safety and Health Act (CAL-OSHA), and health and safety aspects of GA procedures. Specific responsibilities during decommissioning include:
Conducting a training program to instruct employees in safe work practices e Performing periodic inspections of work areas to identify and correct any unsafe work practices and conditions O e Providing industrial hygiene services e Advising project management on industrial safety issues
,, The minimum qualifications for this position are:
- A four-year degree in a related field e Three years' experience in Industrial Safety in the nuclear industry e Demonstrated experience in conducting an Industrial Safety training program Site Decommissioning Manager-The Site Decommissioning Manager supervises the daily activities of the on-site labor fome and contractors. This position will be responsible for preparing decommissioning procedures and maintaining the decommissioning activities at the TRF site on schedule.
2-23
PC-000482/0 l
The minimum qualifications for this position are: ;
- A four year degree in engineering
- Ten years of management experience in a nuclear facility to include D&D projects
- Knowledge of the TRF operating equipment and services
- Training in radiation protection, nuclear safety, hazardous materials, and industrial ;
safety '
. Experience in worker training in a nuclear environment Onality Assurance Manager-The QA Manager is responsible for implementing and 1
managing the QA program for this Project in accordance with the applicable requirements of ASME-NQA-1, Quality Assurance Program Requirementsfor Nuclear Facilities and 10 , ,
CFR 71, Subpart H, Quality Assurancefor Packaging and Transportation of Radioactive ,
Material, and for certification of nuclear waste for compliance with the Acceptance Criteria of the Waste Disposal Facility.
The minimum qualifications for this position are:
- A four-year degree in engineering or related field ,
e !
- Five years' experience in nuclear quality assurance
- Two years' experience in nuclear decommissioning and waste processing Waste Manager-The function of the Waste Manager is to address all requirements and actions required in the treatment and removal of waste resulting from project activities. This includes all types of waste. The Waste Manager is responsible for radiological, mixed ;
waste, and hazardous waste which is to be characterized, packaged, and shipped to the appropriate disposal site in compliance with applicable regulations and acceptance criteria, or free released.
i The minimum qualifications for thic position ve: ,
l
- A four-year degree in Nuclear Engineering or a related specialty
- Experience and training in radiological and hazardous waste management !
- Appropriate training in radiation protection, nuclear safety, hazards communication, j and industrial safety l '2.5 Training Program i l Training is conducted and controlled in accordance with GA procedures, license j l commitments, and the Work Authorization for the decommissioning project and focuses on -) l safety, knowledge of applicable regulations, and technical requirements. The training l program shall comply with the training requin:ments specified by the USNRC in 10 CFR 19.12 and 10 CFR 71.105(d); by OSHA in 29 CFR 1910.120(e) and 29 CFR .g ;
1910.1200(h); by the Environmental Protection Agency (EPA) in 40 CFR 265.16; by CAleEPA in CCR 22-66265.16; and by The Department of Transportation (DUT) in 49 CFR 172.704.
General Emolovee Radiological Training (GERT 4 Hour)--Training will be provided to personnel required to enter Restricted Areas (with the exception of visitors and infrequent support personnel), including Radiation Areas and some Radioactive Materials Areas, but I not perform " hands-on work" or who may perform limited work with radioactive material.
I 2-24
PC-000482/0 Radiological Worker Training (RWT 16 Houri-Training will be provided to personnel who require unescorted access to Restricted Areas and who may perform more complex !
radiologicaljob functions.
GERT and RWT are required initially. Both are effective for two years except when a change of visitor status (GERT) to worker status occurs, in which case RWT is required. '
Refresher training is provided as required.
Health Physics Technician Training ;
HP Technicians must successfully complete Radiological Worker Training. In addition, HP Technicians must review and understand procedures according to the HP Technician l Procedure Review Sign-off Forms. HP Technicians will also review applicable procedure revisions in a timely manner. HP Technicians will also be familiarized with the Site and i i
Facility characterization results and the contents of this Plan.
Eauinment Operator Trainine t
All equipment operators will have proper training completed and documented prior to '
unsupervised work with the equipment.
Safetv/ Accident Prevention Training ,
GA has an Accident Prevention Program which is defined in the Accident Prevention Program Manual (APPM). All employees are requimd to abide by the requirements of this ,
i Manual. Additional specific Project requirements are specified in the plans and procedures for this Project. These additional requirements arise because of the nature of the work to be ..
performed.
Ha7nrd Communication Trainine-A hazatd communication training program has been developed for this Project in accordance with OSHA 1910.1200 and the GA APPM. This i program promotes awareness of chemical hazards that am present at this Facility, and provides means to communicate those hazards to employees. A designated person will maintain the hazardous materialinventory and Material Safety Data Sheets (MSDS) for on- i site hazardous materials, and provide all Project personnel with information advising them I of the potential for hazardous constituents in the work place. A list of such materials is maintained at thejob site, and copies of the appropriate MSDS are available to site workers upon request. The MSDS form provides more detailed information about the chemical than a label does. A hazardous chemical inventory is maintained which reflects the current ,
supplies located in the work area. Any chemicals not previously located and identified or new chemicals received on the job site will be added to the inventory list.
t.
Contamination Control Trainine-Personnel will be trained in contamination control together with boundary control, ventilation control, etc. Cross contamination will be limited o- by the use of training and radiological controls. Radiological and hazardous material contamination will be strictly controlled during all decommissioning work. This control will be accomplished using qua:ified workers to perform work identified in approved work procedures. In some instances, special briefings and dry-runs may be used to perfect ,
techniques, demonstrate approaches, and qualify the workers. j Resoirator Training-Each individual who may be expected to need the use of a respirator ;
~
l will be required to receive respiratory protection training, be medically qualified to use respirator protection, and receive a quantitative fit test for each specific device that they are i qualified to use. Training will meet the requirements of the U.S. Department of Health, Education, and Welfare, National Institute for Occupational Safety and Health (NIOSH),
2 25 l t
PC 000482/0 and ANSI ZS8.2-1980, Practices for Respiratory Protection (Ref.10.11). Respirator fit tests will be administered before initial assignments to jobs requiring the use of a respirator, -
and will be conducted annually thereafter. Medical qualification will be assessed annually.
i Confined Soace Entry Training-Employees required to enter confined or enclosed spaces will be trained to the OSHA confined space entry requirements. They will be instructed as to the nature of the hazards involved, the necessary precautions to be taken and the use of required emergency and protective equipment, as prescribed by the Health and Safety i Manager or designated person Aconfined space permit must be issued prior to access into i the confined space.
I Emergency Resoonse Training-GA has a Site General Emergency Plan and a GA Site Radiological Contmgency Plan, as required by the USNRC and the State of California. The 1RF has a specific procedure in support of these plans.
Hazardous Materials Traininy Training for hazardous materials is dependent on the job description for each individual and the types and amounts of hazardous materials or hazardous wastes being handled as :
specified in the position's training plan. In general, the training specified for workers and ;
supervisors directly involved with decommissioning includes some or all of the following i training requirements:
HAZWOPER Training Course-OSHA 1910.120,40 hour4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> classroom and 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> on-the-job training specific to hazardous materials. An annual update is provided.
Hn7ardous Materials Pachging-Reviews the requirements for handling and shipping hazardous materials and wastes as required by 49 CFR, the D0f regulations. A refresher update is required every two years.
Waste Accentance Criteria-Training is provided to the requirements established for the disposal site. An annual training update is provided.
Dangerous Waste Regulations-Training to familiarize hazardous waste technicians and supervisors with appropriate hazardous waste requirements for waste designations. A refresher is provided annually or as regulations are updated.
GA Emergency Response Training-Training to familiarize emergency response personnel with actions to be taken in responding to an unplanned release of hazardous or radioactive material from the TRF. Hands-on training in this area includes conducting drills to evaluate response capabilities.
y RCRA Facility Standards Overview Trainiag-This class covers. the requirements established under 40 CFR 264.16 for personnal who may handle hazardous wastes within the Facility. The class covers the Federal Standards and discusses compliance requirements ., ,
for generators of hazardous and mixed wastes. An annual update is provided.
2.6 Contractor Assistanee Contractors will have undergone the GA Quality Assurance approval process when required. Wherever contracting personnel are used on-site, they will: 1) comply with all provisions of GA licenses and permits, and 2) be trained in .accordance with GA's commitments.
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PC-000482/0 Contractors will be used on an as needed basis during decommissioning. The use of contractors will be complementary to the GA staff and will normally provide specialty support.
Tasks where contractors may be used include but are not limited to:
- Shipment and disposal of radioactive and nonradioactive waste materials
- Laboratory testing and analysis
- Concrete cutting e Building demolition
- Asbestos removal and disposal
'
- Design and fabrication of specialty dismantling tooling and equipment
- Specialty engmecing and design services
- Temporary staff augmentation Potential contractors for each identified task will be required to provide a statement of qualifications as part of their bid submittal. The qualifications required will emphasize the following:
- Experience with similar work in a radioactive environment
- Adequacy of qualified workers
- Ability to meet schedule The Quality Assurance organization at GA maintains an approved supplier list and has an extensive approval process which ensures that contractor qualifications are adequate to the need.
Subcontractors who will work with licensed radioactive materials will be required to:
- Attend and complete applicable Radiological Worker Training
- Provide required exposure history information e Read and sign an applicable RWP and comply with instructions
. Be issued proper dosimetry by cognizant HP personnel e Follow all special instmetions given by HP e Be escorted by a cognizant authorized person listed on the TRF WA, unless specifically listed themselves on the TRF WA g, 2.7 Decontamination and Decommissioning Documents and Guides Health Physics, Industrial Health criteria, and other standards that guided the activities described in this Decommissioning Plan are discussed in Section 3.1.2, Health Physics
- Program, Section 3.2.3, Radioactive Waste Disposal and Section 3.2.4, General Industrial Safety Program. Relevant documents and guides used are noted therein and in Section 10, References.
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PC 000482/0
- 3. PROTECTION OF THE HEALTH AND SAFETY OF RADIATION WORKERS AND THE PUBLIC 3.1 Radiation Protection 3.1.1 Ensuring As Low As is Reasonably Achievable Radiation Exposures Decommissioning activities at the GA 'IRIGA* Facility involving the use and handling of radioactive materials will be conducted such that radiation exposure will be maintained as low as is reasonably achievable, taking into account the current state of technology and economics of improvements in relation to the benefits.
ALARA Program GA's current practice is as follows:
- A documented ALARA evaluation will be required for specific tasks if a Health Physicist determines that 5% of the applicable dose limits for the following may be exceeded:
- Total Effective Dose Equivalent (TEDE)
- The sum of the Deep-Dose Equivalent (DDE) and the Committed Dose Equivalent (CDE) to any individual organ or tissue other than the lens of the eye
- Eye Dose Equivalent (EDE) :
- Shallow-Dose Equivalent (SDE) e A documented ALARA evaluation will be mquimd if a Health Physicist determines that TRIGA* effluent averaged over one year is expected to exceed 20% of applicable concentration in 10 CFR 20, Appendix B, Table 2, Columns I and 2.
The Project management positions responsible for radiation protection and maintaining
Methods for Occunational Exposure Reduction Various methods will be utilized during decommissioning to ensure that occupational '
exposure to radioactive materials is kept ALARA. The methods include WA, RWP, special equipment, techniques, and practices as described in the following subsections.
b Work Authorization Anoroval j Authorization for work to be performed in accordance with Facility licenses and/or the ,
a- . Decommissioning Plan must be obtained through the LSNC Division, by preparation and maintenance of a WA. The WA identifies the proposed work scope and activities, quantity .
and form of radioactive materials involved, individuals authorized to perform the work, and applicable work procedures. An estimate of the isotope (s), physical and chemical form, and quantity of radioactive material generated as waste during a twelve-month period is included in the WA. An assessment of the magnitude and significance of estimated releases !
of radioactivity to the environment is also provided. Implementation of operating procedures is contingent upon approval of the WA. Work is performed in strict accordance with the methods and precautions provided in the approved WA.
3-1
PC-000482/0 Radiation Work Permits l RWPs are used when:
e a work task is not described in the Work Authorization, provided that the margin of safety provided by the RWP is comparable to, or greater than, that specified in the WA, e personnel not listed as authorized users on the Work Authorization must perform work, e or outside contractors or subcontractor personnel must perform limited or routine work
' in the Restricted Area (RA).
The RWP is issued in accordance with existing Health Physics procedural requirements, and is initiated by the Principal Investigator or other responsible individual who has good knowledge of the task to be performed and other work being performed in the area. ~
Resoiratory Protection and TEDE ALARA Evaluations l
J The use of engineering controls to mitigate the airborne radiological hazard at the source :
will be the first choice with respect m controlling tiu e,ncentrations of airborne radioactive i material.- There may be, howeve:, circumstances where engineering controls are not ,
practical, or may not be sufficient to prevent airborne concentrations in excess of those that I constitute an airborne radioactivity area. In such circumstances where worker access is L required, respiratory protective equipment will be utilized to limit intemal exposures. Any situation whereby workers are allowed access to an airborne radioactivi:y area, or allowed to perform work that has a high degree of likelihood to generate airborne radioactivity in -i excess of 0.1 DAC, the decision to allow access will be accompanied by the performance of repmsentative measurements of airbome radioactivity to assess worker intake. The results of DAC-hour tracking and air sample results for any intake will be documented.
Workers will provide nasal smears for HP evaluation following the use of respiratory protective equipment for radiological purposes.
Control and Storage of Radioactive Materials l 1
The GA HP Program establishes radioactive material controls that ensure: '
- Determace of inadvertent release oflicensed radioactive materials to unrestricted areas.
- Confidence that personnel are not inadvertently exposed to licensed radioactive materials.
- Minimization of the volume of radioactive wastes generated during the decommissioning.
All materialleaving the Restricted Area will be surveyed to ensure that radioactive material is not inadvertently released from the Facility. The following methods will be utilized, as !
appropriate to the material being evaluated, to survey for licensed materials:
e Direct scans with a portable detector, e Indirect survey by collection of representative smears for removable contamination with analysis, e Collection of representative samples of bulk liquids or solids for analysis.
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3.1.2 - Health Physics Program f i Project Health Physics Pronram-General I i GA Health Physics has procedures in place which will be implemented during the 'IRIGA*
i Decommissioning Project. If new Health Physics procedures are required at some point in the. work to support the decommissioning, they will be developed and approved in accordance with GA Health Physics policy and procedure.
i 'GA senior management is readily accessible to ensure timely resolution of difficulties that may be encountered. The HP Manager, while organizationally independent of the Project i ' staff, has direct access to the Project Manager on a daily basis, and has full authority to act ,
3- in all aspects of protection of workers and the public from the effects of radiation. Conduct !
. of the TRIGA* Decommissioning Project HP program will be evaluated according to GA l
, policy and procedure by both GA Quality Assurance oversight, and GA site HP audit j
- ,. activities, i 4
l Audits. Insoections. and Mannoement Review j ' During the decommissioning project, aspects of the Project may le ,tssessed by the GA
{ Quality Assurance Depanment, through audits, assessments, and inspections of various j aspects of decommissioning performance, including HP as described in Section 1.2.4'.
- Formal audits of the GA Health Physics program are conducted annually in accordance
j aspects of the 'IRIGA* Decommissioning project. !
i
- Additional assessments or management reviews may .be . performed when deemed
! appropriate by the Project Manager and/or the Principal Investigator.
i l Health Physics Eauipment and Instrumentation 1
! GA has selected HP equipment and instrumentation suitable to permit ready detection and i
- quantification of mdiological hazards to workers and the public, and to ensure the validity
- of measurements taken during mmediation and final miease surveys. The selection of ;
1 equipment and instrumentation to be utilized was based upon detailed knowledge of the j radiological contaminants, concentrations, chemical forms, and chemical behaviors that am
. expected to exist as demonstrated during radiological characterization, and as known from
! process knowledge of the working history of the Facility. Equipment and instrumentation js selection also takes into account the working conditions, contamination levels, and source
['
terms that are reasonably expected to be encountered during the ' performance of decommissioning work as presented in this Plan.
The following sections present details of the equipment and instrumentation presently selected for use during the decommissioning. It is anticipated that through retirement of worn or damaged equipment /mstrumentation or increases in quantities of available components or instruments, that new technology will permit upgrades or, at a minimum,
-like-for-like replacements. GA is committed to maintaining conformance to minimum performance capabilities stated in this Plan whenever new components or instruments are
~
selected.
3-3
PC-000482/0 Criteria for Selecting Equipment and . Instrumentation for Conduct of Radiation and Contamination Surveys and Personnel Monitoring A sufficient inventory and vcriety of instrumentation will be maintained on site to facilitate
' effective measurement of radiok gical conditions and control of worker exposure consistent with ALARA, and to evaluate suitability of materials for release to unrestricted use.
Instrumentation and equipment will be capable of measuring the range of dose rates and radioactivity concentrations expected to be encountered during conduct of remediation and dismantlement of the Facility, as well as for final survey measurements, and to less than the minimum values required for release or ALARA decision-making.
HP staff will select instrumentation that is sensitive to the minimum detection limits for the
. particular task being performed, but also with sufficient range to ensure that the full spectrum of anticipated conditions for a task or survey can be met by the instrumentation in use. Consumable supplies will conform to manufacturer and/or regulatory recommendation to ensure that measurements meet desired sensitivity and are valid for the intended purpose.
- GA will continue review of regulatory information notices and bulletins for applicability to Project HP instrumentation.
. Storane. Calibration. Testine. and Maintenance of Health Physics Eauinment and Instrumentation 9
. Survey instruments will be stored in a common location under the' control of Decommissioning Project HP personnel. A program to clearly identify and remove from service any inoperable or out-of-calibration instruments or equipment as described in HP procedures will be adhered to throughout this Decommissioning Project. Survey mstruments, counting equipment, air samplers, air monitors, and personnel contamination monitors will be calibrated at license-required intervals, manufacturer-prescribed intervals (if shorter frequency) or prior to use against standards that are NIST traceable in accordance with Calibration Laboratory procedures, HP procedures, or vendor technical manuals. Survey instruments will be operationally tested daily when in use. Counting equipment operability will be verified daily when in use. The personnel contamination momtors are operationally tested on a daily basis.
Specific Health Physics Equipment and Instrumentation Use and Canabilities Table 3-1 provides details of the HP equipment and instmmentation that has been selected for use in the TRIGA* Decommissioning Project. As discussed catlier, the selection of instrumentation is subject to change as older equipment and instruments are retired. GA will maintain conformance to minimum performance capabilities or better, whenever new ,e components or instruments are selected.
a 3-4
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PC 000482/0 '
1 l
Table 3-1-Specific Health Physics Equipment and Instrumentation Use and Capabilities i a
in trument Model l Detector Type l Instrument i Mmimum Detection i Application i I l Range l I Eberline-RO-2 and 2A lonization RO-2 RO-2 Beta / gamma exposure rate measurements Eber1:ne RO 20 chamber o.5,000 mRtir 0.2 mRtir
)
RO-2A RO-2A 0-50 Rtir 20 mR/hr RO-20 RO 20 0-50 R/hr 0.2 mR/hr Ebertine Teletgs 6112D/B GM tube 0-1,000R/hr 0.1 mR/hr Telescoping detector with GM probe for high range j Ludium-M239F Floor Monitor Gas 0-500,000 cpm 20 cpm Alpha and beta / gamma floor rnonstor 434 cm2 with 2221 ratemeter/43-37 proporbonal pmbe *Cs efficiency approximately 30% 4n r"Pu eff ciency approximately 17% 4n EbIrhne-RM-14,14SA/HP. GM tube RM-14 RM 14 Beta / gamma surface contaminabon measurements e60 probe or HP100 BGS pancake probe 0-50,000 cpm 20 cpm Can be used with several types of probes-informabon for HP.
pmba RM 14SA RM-14SA 260 probe.
0 -5,000,000 cpm 10 cpm "Sr efficiency - 32% 4x 15.5 cm2
, HP100-BGS probeC"Sr efficiency 36% 4n,100 cm 2 !
Ludium Model12 with 43 68 Gas 0-500,000 cpm <600 dpm/100 cm2 Beta-Gamma surface contaminabon measurements probe Proporbonal C&nbarra Low-Level Gas CPU operated Vanes according to Low-level a/D smear samples Alpha / Beta Counbng System Proporbonal count bme Ludlum 177 ZnS(Ag) 0-500,000 cpm 20 cpm Hand-held alpha fresker (50 cm8 area) scintillabon 2"Pu efficiency 15% 4n 8"Th efficiency 23% 4n Ludlum Model19 R Nai (TI) 0-5,000 gR/hr 1 pR/hr Low gamma exposure rates Scintillator (i.e. 5 mR/hr)
Eb:rtine SAC-4 ZnS(Ag) 6 Decade scalar Background is Alpha laboratory measurement of air samples and smears Scintilator generallyless than 0.3 47 mm diameter cpm, making MDCR -
0.4 cpm "Pu efficiency - 40W4n Eberkne BC-4 Shielded GM 6 Decade scalar MDCR - 20 cpm Beta laboratory rneasurement of air samples ano smear.
parnike tube 47 mm diameter "Sr efficiency - 40%4n REGE Canberra S-100 or HPGe N/A Vanes byphoton Gamma laboratory measurement of water, air, sme5rrkdia Iquivalent Gamma-ray samples (e.g., soil, asphalt, Concrete, tar, vegetabon) energy and sample spectroscopy system rnedia Eberhra Personnel Gas N/A <5.000 dpm/100 cm2 p- Personnel contaminabon rnonstor/ walk-in monitor with Contaminabon Proporbonal y rricroprocessor control and radon reject capability.
Monitor PCM 2 <300 dpm/100 cm' a SAIC RADeCO H809V N/A 130 cfm N/A High Volume air sampling for minimum detecton capabihty ,
"HiVol" l
SAIC RADeCO HD 29A N/A 0.5-3.5 ctm N/A Low volume air samphng for long term air samphng !
"Geose Neck" Tcchrical Assoc. GM 10-105 cpm N/A Local airbome morrtor with alarm capabihty FM-5ABN-2CH ;,,vtJiar air l
$onitors Arr;tek MG-4 Air Sataioler N/A 5 4,000 cc/ min. N/A Lapel air sampler for use in chronic exposure situabons Poliev. Method. Frequency. and Procedures The TRIGA* Decommissioning Project will utilize the existing GA HP Program for the Project. This Program prescribes policy, method, and frequency for efHuent monitoring, conduct of radiological surveys, personnel monitoring, contamination control methods, and prDtective clothing usage.
EfHuent Monitoring-The facility exhaust points will be continuously sampled downstream of each HEPA filtration system by an isokinetic sample collection point (a grab-type sample) which is in continuous operation. The particulate sample change-out frequency is approximately weekly; sample media are analyzed for pasticulates using laboratory counting 3-5
i l PC-000482/0 systems in a timely manner after filter media change-out. The GA HP Depanment also
( operates several continuous environmental air sample stations on the main GA site to I further assess the potential for environmental airborne radiological effluents.
Any potentially contaminated liquids that may be generated or stored at the facility will be sampled and analyzed for radioactivity prior to disposal. If these liquids ce fouud to ,
contain radioactivity, they- will be evaluated for disposal options in accordance with applicable permits and State and Federal regulations.
Radiation Survevs-Radiation, airbome radioactivity, and contamination surveys during I decommissioning will be conducted in accordance with approved HP procedure (s). The i purposes of these surveys will be to (1) protect the health and safety of workers, (2) protect i the health and safety of the general public, and (3) demonstrate compliance with applicable -
license, federal, and state requirements, as well as Decommissioning Plan commitments.
HP personnel will verify the validity of posted radiological warning signs during the conduct of these surveys. Surveys will be conducted in accordance with procedures ,
utilizing survey instrumentation and equipment suitable for the nature and range of hazards anticipated. Equipment and instrumentation will be calibrated and, where applicable, operationally tested prior to use in accordance with procedural requirements. Routine surveys are conducted at a specified frequency to ensure that contamination and radiation levels in unrestricted areas do not exceed license, federal, state, or site limits. HP staff will also perform surveys during decommissioning whenever work activities create a potential to impact radiological conditions.
Personnel Monitoring-Intemal and Extemal-Extemal monitoring will be conducted in -
accordance with the prospective external exposure evaluation for the Facility. Prospective extemal exposure evaluations will be performed, at a minimum, on an annual basis, or whenever changes in worker exposures warrant. Visitors to the Facility will be monitored in accordance with mquirements specified in GA HP procedures, and according to the radiological hazards of areas to be entered.
Internal monitoring will be conducted in accordance with the prospective intemal exposure evaluation for the Facility. This prospective irt.ctnal exposure evaluation will be evaluated )
on an annual basis, at a minimum, or whenever significant changes in planned work :
I evolutions warrant it. A comprehensive air sampling program is conducted at the Facility to evaluate worker exposures regardless of whether mtemal monitoring is specified. The results of this air sampling program will be utilized to ensure validity of specified intemal monitoring requirements for Facility personnel. Facility workers and frequent visitors may also be required to panicipate in a periodic in-vivo counting program at Project 'HP Management direction to further ensure that internal exposure is carefully observed. If at ,e any time during the decommissioning, hazards that may not be readily detected by the preceding measures are encountered, special measures or bioassay, as appropriate, will be instituted to ensure the adequate surveillance of worker internal exposure. ,,
Monitoring will be required if the prospective dose evaluation shows that an individual (s) !
' dose is likely to exceed 10% of the applicable limits and for individuals entering a high or 1 very high radiation area. j Respiratory Protection-The GA respiratory protection program provides direction for use ,
of National Institute for Occupational Safety and Health /Mine Safety and Health !
Administration (NIOSH/MSHA) certified equipment. This program is administered by GA HP in consultation with GA Industrial Hygiene.
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i
, i l PC 000482/0 h NIOSH/MSHA approved equipment are air purifying respirators which includes full face
' 3iece assemblies with air purifying elements to provide respiratory protection against lazardous vapors, gases, and/or particulate matter to individuals in airborne radioactive materials areas. Individuals may be required to use continuous or constant flow full-face airline respirators for work in areas with actual or potential airborne radioactivity. The HP '
Manager will also ensure that the respiratory protection program meets the requirements of 4
10 CFR Part 20, subpart H.
3 Maintenance-When respiratory protection equipment requires cleaning, the cartridges will i be removed. The respirator will be cleaned and sanitized after every use with a cleaner / sanitizer and then rinsed thoroughly in plain warm water in accordance with HP .
, procedures.-
i Storage-Respiratory protective equipment will be kept in proper working order. When i
any respirator shows evidence of excessive wear or has failed inspection, it will be repaired
" or replaced. Respiratory protective equipment that is not in use will be stored in a clean dry
! location.
i i
Contamination Control-Contamination control measures that will be employed include the followmg:
Worker training will incorporate methods and techniques for the control of radioactive j materials, and proper use and donning / doffing of protective clothing Procedures incorporate HP controls to minimize spread of contamination during work e
Radiological surveys will be scheduled and conducted by HP j e
. Containment devices such as designed barriers, containers and plastic bags will be used e
- to prevent the spread of radioactive material -
!- e Physical decontamination of Facility, areas or items :
Physical barriers such as Herculite sheeting, strippable paint, and tacky mat step-off pads to limit contamination spread
- Posting, physical arca boundaries and barricades
- Clean step off pads at the entrance point to contaminated areas 1
Personnel entries into radiological contaminated areas will require the use of protective clothing. This clothing will consist of a suitable combination of the following, dependent upon the conditions outlined in the WA or RWP:
- Heavyweight lab coat
- Heavyweight canvas, cotton, or cotton / polyester coveralls
- Heavyweight hoods
...
- Plastic calf-high booties e
l Rubber shoe covers e Plastic or rubber gloves which may require cloth liners. ]
e Tyvek paper coveralls or plastic rain suit disposable outer clothing
- Face shield or other protective device Access Control-The RA will be properly posted so as to prevent unauthorized access.
Engineered Controls-Personnel exposure to airborne radioactive materials will be minimized by utilizing the following engineering controls:
37
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l PC-000482/0 e: Ventilation devices-in-place or portable HEPA filters or Facility ventilation systems, local exhaust by use of vacuums
- Containment devices--<lesigned containment barriers, containers, plastic bags, tents, )
and glove-bags ,
e Source term reduction-application of fixatives prior to handling, misting of surfaces to l
. minimize dust and resuspension
]
. Airborne' Radioactivity Monitoring-Monitoring for the intake of radioactive material is !
required by 10 CFR 20.1502(b) if the intake is likely to exceed 0.1 ALI (annual limit on 1 intake) during the year for an adult worker or the committed effective dose equivalent is i likely to exceed 0.05 rem (0.5 mSv) for the occupationally exposed minor or declared pregnant woman. Air sampling will be performed in areas where airborne radioactivity is ,
present or likely.
Prospective estimates of worker intakes and air concentrations used to establish monitoring -
requirements will be based on consideration of the following: -
. . The quantity of material (s) handled e The ALI for the material (s) being handled 1
- The release fraction for the radioactive material (s) based upon its physical form and use e' The type of confinement being used for the material (s) being handled
- Other factors that may be applicable HP personnel will use technical judgment in determining the situations that necessitate air sampling regardless of generalized, prospective evaluations done for the Facility.
Prior to identifying the location for an air sampler, the purpose of the radiological air sample wili be identified. Various reasons exist for collecting air samples. The following .
are a few examples: F e Estimation of workerintakes e Verification of confinement of radioactive materials ,
e Early waming of abnormal aliborne concentrations of radioactive materials e Determining the existence of criteria for posting an Airborne Radioacti/ity Area (ARA).
Smoke tubes and buoyant markers may then be used to determine air flow patterns in the j area. Air flow patterns may be reevaluated if there are changes at the Facility that may !
impact the validity of the sampling locations. Such factors might include the following: , !
- Changes in the work process e Changes in the ventilation system o Use of portable ventilation that might alter earlier assessments
. After identifying the purpose for the _ air sample and flow patterns are established, air sample locatmns are established as follows:
- For verification of confinement of radioactive materials:
Locate samplers in the air flow near the potential or actual release point.
- Mcre than one sampling point may be appropriate when there are more than one potential or actual release points.
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PC-000482/0 l
l For estimation of a worker intake:
- Samplerintakes will be located as close to the workers breathing zones as practical without interfering with the work or worker General workplace air sampler intakes will not be placed in or near ventilation exhaust ducts unless their purpose is to detect system leakage during normal operation, and if quantitative measurements of workplace concentrations are not required. Locations or number of air samplers will be changed when dictated by modifications to Facility structure, changes in work processes, or elimination of potential sources.
l A suflicient inventory and variety of operable and calibrated ponable and semi-ponable air 1 sampling equipment will be maintained to allow for effective collection, evaluation, and control of airborne radioactive material and to provide backup capability for inoperable i equipment. Air sampling equipment will be calibrated at prescribed intervals or prior to use i
" against cenified equipment having known valid relationships to nationally recognized standards. Table 3-1 lists anticipated air sampling equipment.
)
l When the work being performed is a continuous process, a continuous sample with a j weekly exchange frequency is appropriate, except for situations where short-lived i radionuclides are expected to represent a significant exposure. For situations where short-
) lived radionuclides are imponant considerations, the exchange frequency will be adjusted ;
l accordingly. Longer sample exchange frequencies may be approved by Decommissioning l l Project HP management for situations where airborne radioactive material and nuisance !
l dust are expected to be relatively low. Grab sampling for continuous processes may also be l l approved by GA HP management based upon consideration of variability of the expected I source term for the facility and process. Grab sampling is the appropriate means of airbome sampling for processes conducted intermittently, and for short duration radiological work i
that involves a potential for airborne release.
1 Potential Sources of Radiation or Contamination Exposure to Workers and Public as a Result of Decommissionine Activities l
l Sources of radiation or contamination exposure may be assessed by process knowledge,
, radiological survey data, surveys performed during characterization, previous and current
! job coverage surveys, or daily, weekly and monthly routine surveys.
Classification of potential sources may also be identified by, radionuclide, physical l properties, volatility, and radioactivity.
- s l Worker exposure to significant extemal deep-dose radiation fields is considered unlikely during this project due to the nature of the contaminar.ts. Worker exposure to airbome radioactivity may occur during decontamination operations / work evolutions which may
- - involve abrasives or methods that volatilize loose and/or fixed contamination.
Exposure of the public to extemal or intemal radiation from this Decommissioning Project is not considered credible.
The type (s) of exposure controls used takes into account the current state of technology and the economics of improvements in relation to the benefits. Control of potential sources of radiation exposure to workers and public as a result of decommissioning activities will be achieved through, but not limited to, the use of administrative, engineering, and physical controls.
3-9 l
PC-000482/0
. Administrative controls consist of but are not limited to:
- Administrative dose limits that are lower than regulatory limits
. Training
- Radiological surveys Physical barriers such as radiological warning rope / ribbon, in combination with radiological warning tape, lockable doors / gates as well as information signs and flashing lights or other applicable barriers may also be used.
Engineering controls may consist of but are not limited to:
. HEPA ventilation / enclosures -
- Protective clothing / equipment e Access restrictions / barriers e Confinement Health Physics Policies for Subcontractor Personnel Subcontractor personnel may be used for certain required work during the 'IRIGA*
Decommissioning Project. Subcontractors who will work with licensed radio. active materials will be required to:
- Attend and complete appropriate radiation safety course e Provide required exposure history information .
. Read and sign an applicable RWP and comply with instructions e Follow all special instmetions given by HP
- Be escorted by a cognizant authorized person listed on the WA, unless specifically listed themselves on the WA 3.1.3 Dose Estimates The total projected occupational exposure to complete the Decommissioning of the GA TRIGA* Reactors Facility is estimated to be 27 person <em. A task-by-task breakdown of this dose estimate is provided herein as Table 3-2. Task-specific dose estimates are based on the nature of the work involved in each task item, the expected number of persons to be assigned to each task, and the individual task duration periods as shown on the overall project Schedule for *IRIGA* Reactor Facility D&D, (see Figure 2-8).
This estimate is provided for planning purposes only. Detailed exposure estimates and exposure controls shall be developed in accordance with the requirements of the GA ALARA program during detailed planning of the decommissioning activities. Ama dose rates used for this estimate are based on process knowledge and current survey maps (where available).
3-10
PC 000482/0 Table 3 2-Preliminary Collective Dose Estimate Task Name Dur:. son Men Avg. Dose ... Totaf Does . Si.elosal Total Tase
+ Hours Hours Rate somer - person-rom person-rem person-rem 1 1 TRIGA* Reactor Facility D&D 2 USNRC/ State approval of Decommissioning Plan 3 Decommission Mkill 4 Radiological survey 120 2 0.0002 0.048 5 Remove reactor components above pool W 4 0.0002 0.064 0.11 4
6 Reactor tank soluton 7 Survey, sample, analyze 80 3 0.001 024 8 Pump Liquids into drums for dicposal 40 2 0.0025 02 0.44 9 Remove reac'or wspen nts in unk 10 Grapple, hoist, survey 40 3 0.005 0.6 11 Disassemble 64 3 0.005 0.96 12 Decontarninate or package as LLW 64 3 0.005 0.96 13 Install confinement bamer around reactor pit 40 3 0.0003 0.(D6 2.56 s 14 Al tank and tangential beam tube removal 15 Cut and remove in sections 80 4 0.0025 0.8 16 Segregate clean sections 64 4 0.0025 0.64 17 Package LLW sectons 40 4 0.002 0.32 1.76 l
l 18 Concrete kner 19 Demolish actvated porbon 120 4 0.002 0.96 20 Remove and package 112 4 0.002 0.896 21 Survey remaining concrete 32 2 0.0005 0.(D2 ;
22 Dernolish remaining porton to expose soil 80 4 0.0006 0.16 i 23 Survey soil presumed to be clean 40 2 0.0002 0.016 24 Shore and cover pit 40 4 0.0002 0.032 l
25 Dismantle bamer and package for LLW disposal 32 4 0.0002 0.0256 26 Remove surface contaminabon from reactor room 80 4 0.0002 0.064 2.19 27 Decommission Mkl t 2B Radiological survey 112 2 0.0002 0.0448 29 Remove reactor components above pool 80 4 0.0002 0.064 0.11 ;
30 Remove reactor s v en.ats in pool !
31 Grapple, hoist, survey 72 3 0.005 1.08 32 Disassemble 72 3 0.005 1.08
- D Dow a....ste or package as LLW 64 3 0.005 0.96 3.12 34 Reactor tank water 3 Survey, sample, analyze 40 3 0.0002 0.024 36 Discharge and filter as necessary 40 3 0.0002 0.024 37 install confinement barrier around reactor pit 40 3 0.00(D 0.(D6 0.00 2 Al tank removal 30 Cut and removein sectms 80 4 0.0025 0.8 40 Segregate clean sectons 64 4 0.0025 0.64 41 Package LLW sections 56 4 0.002 0.448 1.89 42 Concrete hner
, . 43 Demoish actvated porton 120 4 0.002 0.96 44 Remove and package 112 4 0.002 0.896 45 Survey remaintog concrete 32 2 0.0005 0.(D2 46 Demoksh remaining porton to expose soil 80 4 0.0005 .016 47 Survey soil presumed to be clean 40 2 0.0002 0.016 48 Shore and cover pit 40 4 0.0002 0.(D2 40 Survey storage wells, ter.me if contarnnated 160 4 0.00(D 0.192 50 Dismantle barrier and package for LLW disposal 40 4 0.0002 0.032 51 Remove surface contamination from reactor room 80 4 0.0002 0.064 2.38 i
3 11
PC-000482/0 Table 3 2--Preliminary Collective Dose 8 stimate Task Task Name
- +
Duration Man Avg.Dosr . Total Dose SutMotal Total Hours -Hours Rate rerre at persorbrem persorprern person-rem 52 Decviiiiisssion remaining areas (except MkF) 53 Remove hot drain lines 104 4 0.0CTJ3 0.1248 54 Remove contaminated secbons from rooms except MkF 80 4 0.0013 f6 0.096 Reroute services to isolate MkF room 224 4 0 0002 56 0.1792 3 Remove make-up water tarsk (D 4 0.0002 0.064 57 Dismantle and dispose of reunining equipment in yard 160 3 0.0003 58 0.144 Dismantle Mkill, TTSL, & TTSX rooms 100 6 0,0002 fB 0.192 Dismantle Mkl rooms 100 6 0.0002 0.192 00 Demolish concrete floors and wells 160 4 0.00CD 61 Ship LLW to NTS 0.192 80 4 0.0003 0.006 128 62 Decv.. mssion MkF (D DOE approval to ship fueloff-site 64 Remove and ship fuel stored in MkF canal 360 4 0.002 2.88 05 Radiolog. cal survey 120 2 0.002 0.48 3.36 06 Remove reactor cvnvunents '
67 Grapple, hoist, survey 80 3 0.005 12 ER Disassemble 80 3 0.005 1.2 09 Decontarranate or package as LLW 64 3 0.005 0.96 3.36 70 Reactor tank water 71 Survey 40 3 0.0002 0.024 72 Discharge and filter 40 3 0.0002 0.024 73 Install confinement bamer cround reactor pit 40 3 0.00CD 0.CD6 74 0.08 Tank removal 75 Cut and terreve in seccons 72 4 0.0025 0.72 76 Segregate clean sechons 64 4 0.0025 0.64 77 Package LLW sechons 40 4 0.002 0.32 1.68 78 Conerste liner ,
79 Demolish acbvated porbon 112 4 0.002 0.ED6 80 Remove and package 120 4 0.002 0.96 81 Survey re nairung concrete 40 2 0.0005 0.04 f2 Demolish remairung porbon to expose soil 80 4 0.0005 0.16 83 Survey soll presumed to be clean 40 2 0.0002 0.016 84 Shore and cover pit 40 4 0.0002 0.032 85 Dismantle bamer and package for LLW disposal 40 4 0.0002 0.CD2 86 Remove surface contarrunabon from reactor room 80 4 0.0002 0.064 87 Dismantle reactor building and dispose 160 6 0.0002 0.192 BB Demoksh concrete floor and dispose 120 4 0.0002 0.006 89 Package contartunated bois and equipment
' 40 6 0.0002 0.048 90 Ship LLW b NTS 72 4 0.0002 0.0576 91 Survey soit 112 2 0.0002 0.0448 2.64 92 Prepare survey report )
50 Subrruttal for release to unrestncted use 27.0 3.2 Radioactive Waste Management ..
3.2.1 Fuel Removal The DOE has agreed that it has a contractual obligation to accept all of the TRIGA* reactor fuel at a DOE designated fuel storage facility (see Contract DE-CR01-83NE4436). A date for shipment has not yet been determined and is the subject of ongoing negotiation.
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PC-000482/0 l l
The current fuel status is as follows:
Mark III Reactor-All fuel elements have been relocated from the reactor core to the Mark F fuel storage canal.
Mark F Reactor-All reactor fuel elements have been removed from the Mark F reactor core / shroud and placed in the Fuel Storage Canal which also contains the irradiated fuel elements previously removed from the Mark III reactor. J Mark I Reactor-The reactor and associated control systems are fully operational pending approval by the USNRC for a possession only license amendment (POL). Upon USNRC approval for a POL, the fuel elements will be transferred to the Mark F fuel storage canal.
3.2.2 Radioactive Waste Procest,ing The processes of decontamination, remediation, and dismantlement of the TRF will result in solid and liquid low level radioactive waste, mixed waste, and hazardous waste. Limited soil remediation is anticipated which will result ir. solid radioactive waste. This waste will be handled (processed and packaged), stored, and disposed of in accordance with applicable sections of the Code of Federal Regulations (CFR), California Code of Regulations (CCR), San Diego County and City Regulations, disposal site Waste Acceptance Criteria, respective State Administrative Codes, GA Licenses and Permits, and the applicable implementing plans and procedures. Radioactive waste processing includes waste minimization or volume reduction, radioactive and hazardous waste segregation, waste characterization, neutralization, stabilization, and solidification.
3.2.3 Radioactive Waste Disposal Low level radioactive waste will be processed and packaged for disposal at the Nevada Test Site under the terms of an agreement in principle with DOE. Specific permission for shipment of'IRF waste to NTS will be requested. The volume of low level radioactive waste is estimated at 4,000 cu. ft. Mixed low level waste which is not treatable at GA's Mixed Waste Processing Facility will be prepared for shipment to off-site commercial processing and disposal facilities such as Envirocare of Utah. Low level mixed waste is expected to be minimal,less than 50 cu. ft.
10 CFR 61, Licensing Requirementsfor Land Disposal of Radioactive Waste, Subpart D-Technical Requirements for Land Disposal Facilities, e tablishes minimum radioactive waste classification, characterization, and labeling requirements. These requirements will
( be met through the implementation of Project Packaging and Characterization Procedures, Disposal Site Waste Acceptance Criteria, and the Quality Assurance Program Document.
Traming will be provided for Project Waste Certification OfIicials, Waste Packaging personnel, anu Waste Characterization personnel to assure conformance to applicable 10 CFR 61 requirements as stated in the specific implementing procedures and plans Quality Assurance conducts audits and surveillances per the Quality Assurance Program Document i based on ASME-NQA-1-1989, which confirms conformance with Disposal Site Acceptance Criteria and applicable 10 CFR 61 requirements. )
10 CFR 71, Packaging and Transportation of Radioactive Material, establishes requirements for packaging, shipment preparation, and transportation of licensed material. l The radioactive waste ' hat will be packaged and shipped will be LSA material. GA is an )
USNRC and State of California Licensee to receive, possess, use, and transfer licensed by-product and source materials.10 CFR 71 requirements will be met through the implementation of Project and GA's Nuclear Waste Processing Facility (NWPF) Packaging 3-53 !
l
PC 000482/0 -
)
and Shipping Procedures. Traming will be provided for Waste Packaging Personnel and
- Waste Shipping Personnel to assure conformance to applicable 10 CFR 71 requirements.
LSNC's-Nuclear Material Accountability Depanment provides compliance oversight and ' -!
off-site shipment notices. Quality Assurance will confirm conformance to 10 CFR 71 .
Subpan H (Quality Assurance) requirements through the implementation of the GA Quality !
Assurance Manual and Quality Assurance Program Documents.-10 CFR 71 applicable Quality Assurance requirements apply to design, purchase, fabrication, handling, shipping, storing, cleaning, assembly, inspection, testing, operation, maintenance, repair, and ;
modification of components of packaging which are important to safety. ,
10 CFR 20.2006, Transfer for Disposal and. Mamfests, establishes requirements for ;
controlling transfers oflow-level radioactive waste intended for disposal at a land disposal l facility; establishes a manifest tracking system: supplements requirements concerning -
transfers and record keeping; and requires generator cenification that transported materials i are properly classified, described, packaged, marked, and labeled, and am in proper ;
condition for transport. These requirements will be met through the implementation of .,
. Project and NWPF Packagir.g and Shipping Procedures with the oversight of GA's !
LSNC's Nuclear Material Accountability Department. -
I Radiological and mixed wastes will be disposed of at disposal sites per the applicable Disposal Site's Acceptance Criteria. Associated implementmg plans and procedures will :
reflect the characterization, processing, mmoval of prohibited items, packaging and transponation requirements. Appropriate documentation will be submitted to designated :
disposal sites including, as required, cenification plans, qualification statements, assessments, waste stream analysis, evaluations and profiles, transponation plans, and waste stream' volume - forecasts. Waste characterization, waste designation, waste :
traceability, waste segregation, waste packaging, waste minimization...and quality !
assurance and training requirements of the designated disposal sites will be incorporated in i implementing procedures to assure conformance to disposal site requirements.
Generator State (Califomia) and Treatment / Storage / Disposal Facility States (Nevada, Utah, etc.) mquirements for radioactive and mixed waste management will be incorporated into '
plans and procedures to assum conformance with applicable state regulations, hcenses, and )
permits. Applicable state regulations include Califomia Hazardous Waste Management Regulations (Califomia Code of Regulations, Title 22), and Utah Department of Environmental Quality Rules (R313) for the control of ionizing radsatwn reflected in j Envirocare's Utah Radioactive Material License, UT2300249. The Project will conform to .
.GA CAL-DHS, Radiological Health Branch (CAL-RHB) License (0145-80) to possess and use source materials as directed by CCR, Title 17. GA will also conform to the CAL-EPA requirements (EPA ID Number CAD 067 638 957) which permit / authorize GA to ,e ;
operate as a genetator of hazardous waste, .to treat hazardous waste on site under Califomia's Tiered Permit program Conditional Authorization (CA) or Conditional Exemption (CE) tiers, and to manage radioactive mixed w es under Interim Status granted by the State of California Department of Toxic Substm m Control (CAL-DTSC).
- j The Project will also conform to the GA Health Permit to r .ge nazardous materials issued by the County of San Diego Department of Heahh Services Hazardous Materials i Management Division (SD-DHS-HMMD). Project Plans and Procedures will also incorporate Metropolitan Industrial Waste Program (MIWP) requirements for the discharge ofindustrial waste waters into the sanitary sewer system managed by the City of San Diego (San Diego Metropolitan Water District).
1 Radioactive waste will be staged in designated controlled areas in accordance with USNRC 10 CFR 19 and 20 requirements, CAL-DHS 17 CCR requirements, and the requirements of GA Nuclear Material License and State of Califomia Material License. Mixed wastes will 3 14
.__m._ . _ . _ _. ._. _ . _ _ _ __.. _ _. _ . _ _ __ _._._
PC-000482/0 i e
be staged in designated control'.ed areas per EPA 40 CFR requirements, CAL-DHS 22 ;
CCR,10 CFR 19 and 20, and er i local and state permits. Measures will be implemented ;
through plans and procedures to cantrol the spread of contamination, limit radiation levels, ,
i prevent unauthorized access, preved unau6iaci material removal, prevent tampering, and prevent weather damage. The designated contrvied areas will be approved by WAs, Radiological Work Permits, and Hazardous Work Perreits i Radioactive and mixed waste material will be packaged fc r shipment per 10 CFR, 40 CFR, l 49 CFR,17 CCR, 22 CCR, and the designated Disposal Site Criteria and placed in permitted interim storage (staged) until shipped.The quanhty of waste packages staged for i shipment will be a function of waste generation and packaging rate, shipment preparation i rate, shipment rate, and disposal site acceptance rate. The objective is to nunimize the l
' qaantity of waste in interim storage (staged). To meet this objective, regular shipments will '
be scheduled throughout the life of the Project to designated treatment, storage, and disposal facilities. ,
Radioactive material storage areas will be contained inside posted restricted areas according [
to existing procedures and consistent with 10 CFR 20.
l 3.2.4 General industrial Safety Program j Project Industrial Safety personnel, supported by GA Industrial Safety personnel and GA's !
Industrial Hygienist, shall be responsible to ensure that the Project meets occupational [
health and safety requirements of Project personnel and the general public. Primary :
i functional responsibility is to ensure compliance with the OSHA of 1973 as implemented by Califomia Labor Code Section 6400 and the General Industry Safety Orders (GISO ;
3203). Specific responsibilities include conducting an industrial training program to instruct i employees in general safe work practices; reviewing Decommissioning Project procedures ;
to verify adequate coverage of industrial safety and industrial hygiene concems and - -
requirements; performing periodic inspections of work areas and activities to identify and i correct any unsafe conditions and work practices; providing industiial hygiene services as !
required; administering the Hazardous Work Authorization Program; and advising Project s management on industrial safety matters and on the results of periodic safety inspections.
The Project is staffed with an Environmental Health and Safety Engineer and supported by GA Industrial Safety and Industrial Hygiene personnel.
, All personnel working on the 'IRF Decommissioning Project will receive Health and Safety ;
l training in order to recognize and understand the potential risks involving personnel health .
and safety associated with the work at the 'IRF. The Health and Safety training l
( implemented at GA is to ensure compliance with the requirements of the USNRC (10 l CFR), the EPA (40 CFR), and both OSHA and CAL-OSHA (29 CFR and CCR Title 8). :
Workers and regular visitors are familiarized with plans, procedures, and operation of ;
- equipment to conduct themselves safely. In addition, each worker must be familiar with l procedures that provide for good quality control. Section 2.5, Trainir.g Program,' provides additionalinformation.
3.3 Radiological Accident Analyses :
I The potential radiological accidents for the decommissioning of the TRIGA* Reactor Facility will be mainly associated with the approximately 250 'IRIGA* fuel elements stored in an existing storage canal within the former Mark F reactor pool complex. This fuel storage may remain in effect during decommissioning of the buildings and facilities related to the Mark I and Mark III reactors if arrangements to remove the fuel off site are delayed.
Factors considered in assessing potential radiological accidents are: -
l t 3 15
l i
PC-000482/0
- 1) Fuel storage and removal 1 1
- 2) Seismicity I i
- 3) Fire
- 4) Otherconsiderations l
Fuel Storage f
The spent TRIGA* fuel elements from the three reactors [ Mark I (R-38); Mark F (R-67);
i and Mark III (R-100)] will be stored in racks in the storage canal associated with the Mark F reactor pool. The storage canal is 3.5 feet deep with adequate radiation protection '
. provided by the depth of water over the stored fuel. ,
l The Possession Only License amendment (Ref.10.3) requires that fuel be moved and ,
v stored in accordance with the existing technical specifications and GA procedures until >
removed from the site. The proposed decommissioning action does not pose any additional criticality or fission product release risk.
Seismicity l
.i San Diego County has been considered one of the more moderate seismic risk regions in :
Southern California. The historical pattem of seismic activity has generally been )
characterized by a broad scattering of small magnitude earthquakes, whereas the '
surrounding regions are characterized by a high rate of seismicity with many moderate-to-
' large-magnitude canhquakes. -
- Arecent study (see Appendix B Ref. 5.8) estimated the probabilities of large canhquakes occurring in Califorma on the major strands of the San Andreas fault system. In addition to the principal traces of the San Andmas fault, earthquakes occurring on the other' major i faults of the system (San Jacinto, Imperial, etc.) were also considered. The study estimated l
that the probability of a magnitude 7 or greater canhquake occurring in the next 30 years in Southern California (along the Southern San' Andreas, Imperial, or San Jacinto faults) is 0.5 or greater. However, a quake of magnitude close to 7 on these fault lines is not expected to significantly impact the GA site because ofintervening distance.
Current information (see Appendix B, Ref. 5.9) however, indicates the Rose Canyon, Coronado Bank, San Diego Trough, La Nacion, and Elsinore fault zones are capable of i generating strong ground motion in the San Diego area. Possible Richter magnitudes for ,e R canhquakes on these faults can be as high as 7.0, 7.5, 7.5, 6.3 and 7.5, respectively.
- Passing approximately 3 miles (5 km) west of the GA site, the Rose Canyon fault is the nearest active fault.
The presence of three small, local faults was confirmed by the Woodward-Clyde ;
Consultants field mconnaissance of the site (see Appendix B, Ref 5.11). An unnamed j fault in the northern portion of the site trends east to west through ptoposed lots 25, 26, !
31, and 32. The Salk fault is mapped in the southern portion of the site and also trends east .
l, to west. A northerly tmnding fault is located in the southeastern area and crosses the !
Genesee Avenue canyon. All of these faults are mapped as being overlain by early !
Pleistocene formations which have not been displaced. Therefore, the faults on-site are not considemd active.
(
I l 3 16 i__. - - - - _
PC-000482/0 Decommissioning activities will involve i decontammation ill assure that these alterations would not render the building unsafe. Decommission n not pose added risk to workers during a seismic event.
Eire The1RF will not contain combustible material in sufDeient gaantity to support a It is possible for a small fire to occur as a result of an e radiological hazard as a result of fire.
d Other Considerations Radiological accidents could occt.r during removal and packaging i of activated c
, and equipment. However, this risk is verylilizing di low considering the handling of activated /contaninated components and control of job activities ut written and approved procedures, will ensure the safe conduct of the project.
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n PC-00 04 82/0
- 4. PROPOSED FINAL RADIATION SURVEY PLAN The final release of the 1RF construction materials, equipment, and site will be performed in accordance with speciGc release criteria for surface contamination, soil, and other bulk materials speciDed in this Plan. The criteria presented in this plan were developed using established criteria from GA's licenses that affect the TRF. The intended course of action for TRF decommissioning, based upon consideration of site and facility radiological characterization results, is to strive to decontaminate structural materials to the extent practicable in balance with radioactive waste minimization considerations, and dismantle the TRF while releasing materials that are below release criteria, and packaging for burial those materials that cannot reasonably be decontaminated.
As such, the Final Release Survey Plan (and subsequent Final Survey Report) discussed in this section deals exclusively with release of the TRF property to unrestricted use following disnentlement of the structures and concrete and asphalt surfaces. This section will, however, also diccuss the survey methods and release criteria that will apply to release decisions for structural matenals and facility equipment.
4.1 Description of Final Radiation Survey Plan The purpose of the Final Radiation Survey is to demonstrate that the radiological condition of the site are at or below established release criteria in anticipation of State and USNRC approval of license amendments removing the Facility as a location to handle licensed materials and remove restrictions from use of the Facility or property and permit its unrestricted use. The proposed decommissioning approach described in this Plan involves the dismantlement of the Facility in a piecemeal manner until the Facility has been torn down and only the site property remains. The Final Release Survey (and report) will deal with release of the TRF site to unrestricted use.
During dismantlement of the TRIGA* Reactor Facility, GA will be systematically decontaminating and removing equipment wherever reasonable for free release to disposal at land 611s or use elsewhere, to the greatest extent practicable. While the Final Release Survey deals solely with release of the TRF site, a discussion of the criteria that will be applied to release determinations for structural materials and facility equipment during facihty dismantlement is also included. GA has developed its Final Release Survey Plan using the guidance provided in NUREG/CR-5849 (draft) (Ref.10.10).
4.1.1 Means for Ensuring that all Equipment, Systems, Structures, and Site are included in the Survey Plan As discussed above, the Final Release Survey will deal principally with the TRF property itself. The need to ensure that all materials removed from the Facility conform to specine release criteria is not diminished by this approach, however. In the Decommissioning Tasks presented in Section 2.3 to this Plan, the systematic approach to dismantlement of facility equipment is presented. Systems will only be removed from service when it can be
.- demonstrated that they no longer are needed to provide important safety or effluent /
exposure control functions, or when their removal has been adequately compensated for by other means. Every item that is to be removed from the TRF will be evaluated for ability to decontaminate and demonstrate satisfaction of release criteria. When it is impractical or not possible to satisfy release criteria (or conclusively demonstrate that they have been met), the item will be treated as radioactive waste. The systematic approach to facility dismantlement will inherently ensure that each and every component or stmeture in the TRF is speci5cally evaluated for release before beginning the Final Release Survey. The Final Release Survey will treat the TRF Yard Area as "affected" and ensure 1009e scans and systematic soil sampling prior to requesting release of the propeny for unrestricted use.
l 4-1
PC-000482/0 4.1.2 Means for Ensuring that Sufficient Data is included to Achieve Statistical Goals GA has developed the TRF Final Release Survey Plan using the guidance presented in NUREG/CR-5849. By using to this guidance, the TRIGA* D&D Project will satisfy the USNRC recommended statistical goals.
4,2 Baekground Survey Results The Final Release Survey Guideline values for residual activity are taken to be levels above the naturally occurring background radiation. The final release measurements will consist of a combination of general area radiation values and samples of media (principally soils).
In addition, a detailed micro-R survey of the site will be performed and compared to background measurements. An action level of 10 micro-R per hour above background has -
been established.
Background radiation as encountered at any location includes contributions due to both natural radiation sources and manmade sources. Natural radiation sources include terrestrial radioactivity due to naturally occurring radioisotopes in soils and construction media, airborne radioactivity (principally radon and radon progeny) from the radioactive decay of certain of these naturally occurring radioisotopes, and cosmic radiation from high speed panicle interactions in the earth's atmosphere. Manmade background radiation as it would impact the Final Release Survey would primarily consist of atmospheric fall-out of fission products due to weapons testing and reactor accidents and any contribution that might exist as a result of other licensees' activities.
The general area background radiation as would be measured with the micro-R meter is influenced by a number of factors, principally the naturally occurring radioactivity in soils and other nearby materials, radon and radon progeny concentrations in the air, and extent of cosmic radiation (which varies with elevation). Due to the number of influences, the ,
natural background varies appreciably from location to location, day-to-day (even time of i day) and season-to-season as related to changing weather conditions and materials in the surroundings. Under ideal conditions, it would be desirable to have available pre- I construction and pre-operational general area background radiation measurements to use as l a reference during Final Release Survey; for the case of the TRF, pre-operational micro-R l measurements are not available due to the age of the Facility. The Final Release Survey will include the establishment of background area radiation levels using the guidance of I NUREG/CR-5849 in the general time frame immediately preceding Final Release 7urvey.
The site and facility characterization study included measurements to estaban background radioactivity in soils, concrete, and asphalt that were considered representative of those that '
would be encountered in the Final Release Survey. One of the principal constituents of global fallout, *Cs, which is found principally as a result of atmospheric weapons testing and reactor accidents is also the principal fission product contaminant at the TRF. The concentration of this isotope in the upper atmosphere has been declining over the past two decades, and is also resulting in a padual decline of the concentration of these isotopes in soils and new building materials. ' Cs has been seen to be persistent in the upper 15 cm (6 in.) of soil with concentrations decreasing beyond this depth (Ref.10.12).
Release criteria that have been established for this 1RF D&D Project were established as an increment in excess of background values. Hence, it was critical to carefulle establish the natural background values of various isotopes that could be of concern in gamma spectrometry evaluations of soils and construction media as well as those which affect the background extemal radiation values. In order to study these concentrations in soils, a careful review of the surrounding region was done to identify locations that held a good 4-2
PC 000482/0 potential to (1) possess similar soil characteristics so that radionuclide retention behavior could be considered similar, and (2) be undisturbed by recent construction activity that would disrupt the upper 15 cm (6 in.) of soil. The TRIGA* benefited from proximity to state and county parks and preserves that had limited development and hence, limited attendant soil disturbance. The sites that were selected are shown in Figure 4-1. Similarly, the background media study attempted to identify locations where concrete or asphalt was believed to be older construction material that may be other than contemporary. Asphalt and concrete background media were collected on unaffected portions of the GA property.
< NUREG/CR-5849 recommends that background media measurements consist of at least ten samples; this recommendation was followed during the background media study.
4 The background media study was composed of seventeen (17) surface soil samples, ten d
(10) asphalt samples, and ten (10) concrete samples. The results of background measurements are summarized in Table 4-1. These data may be supplemented by additional data or measurements prior to conducting the Final Release Survey.
a Table 4-1-Typical Background Media Results Medse Type 238g 235g 232 g
] 137Cs (pCUg) (pCUg) (pCUg) (pCVg)
Sod 0211 20 1.2610.78 0.0810.04 1.7210.92 Asphalt 02810.30 4.3714.60 0.1210.16 5.4111.02 Concrete 0.0410.04 427 126 0.1210.06 3.6210.48 1
, 4.3 Release Criteria-Residual Radiation and Contamination Levels This section provides bases and spcific criteria for miease of materials and equipment, facility structures, and the 'IRIGA property. The reactors and building structures will be ,
dismantled piecemeal and either disposed of as radioactive waste or released to unrestricted use according to the extent of contamination and cost-effectiveness of decontamination. The i
Final Release Survey will principally be for the propeny (i.e., soils) itself. The approach to the survey uses the guidance provided in NUREG/CR-5849.
4.3.1 Equipment Release Criteria All materials leaving the TRIGA* Restricted Ama will be surveyed to ensure that licensed materials are not inadvertently released from the Facility. Decommissioning Project and GA HP procedures will be used in performing these evaluations. These evaluations will include the following types of evaluations:
, Materials and Eauipment-Direct frisk ( -7) with a portable detector and/or indirect survey for smearable activity evaluated with portable or fixed detector. In any situations where process knowledge would suggest a potential for alpha activity, survey with alpha detection instruments or counters will also be employed. Materials will be released to the criteria
, .- specified in HP procedures. Taose criteria are summarized in Table 4-2.
l 1
4-3
. _ . _ - . _ _ _ _ . _ m PC-000482/0 6 s ,
o=i un mens w 1
l C
- 235-95-187-CH l I-5 - 23S-95-186-CH 1 arrey - 235-55-Ic4-CH ,
- 235-95 -190-CH Pines - 235-95-189-CH State Reserve Los Penesquitos Canyon'Rese g
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-235-94-170/171-CH
- 235-94-168/169-CH ' i
- 23S-94-166/167-CH
- 23S-94-164/165-CH
- 23S-94-162/163-CH
'235-94-160/161-CH S mple Locofions
-23S-94-158/159-CH are Approximate.
Figure 41-Background Soll Bample Locations 4-4
PC-000482/0 Table 4-2-Acceptable Surf ace Contamination Levels *
(dpm/100 cm'f3 Nuclides* l i Average
- l Maximumd4 i Removable
- 5,000 15,000 1 1.000 U nat, "U, "'U, & associated decay pruducts 100 I0 ! @
Transuranics,226Ra,22*Ra, a"Th. 22'Th, 83'Pa,227Ac. i2si, iz'l 1.000 3.000 i Z0 Th nat, 8"Th, "Sr az'Ra,22*Ra, asaU, i2.i, mi, is'l 5,000 15.000 1,000 Beta / gamma emitters (nuchdes with decay modes other than alpha emssion or sous finaeous fission) except "Sr and others noted above. I
- Where surface wntaminanon by both alpha-and beta /gamia-emittng nuchdes exists, the hmits estaDhsned for alona- and beta / gamma-emitting nuclides should apply independently.
- As used in this table dpm (disintegrations per minute) rneans tie rate of emission by radioactve matenal as determined by correctng the counts per mnute observed by an appropnate detector for background, efficiency, an geometnc factors associated with the instrumentation.
- Measuremer*2 of average w dan ma, d should not be averaged over more than 1 square meter. For obtocts of less surface area, he
, average should be denved for each such object. 2
- The maximum contamination level applies to an area of not more than 100 cm .
- The amount of removable radioactve matenal per 1002 cm of surface area should be determined by wiping that area with dry filter (e.g , smear) or soft absorbent paper (e g., masslin), applying moderate pressure, and assessing the amount of radioactve matenal on the wipe with an appropnate instrument of known ethciency. When removable wntaminahon on objects of less surface area is determined, then pertinent levels should be reduced proportonally and the entre surface should be wiped.
' The average and maximum radiationlevels associated with surface contaminaton resultng from beta-gamma emitters should not exceud 02 mrad /hr at 1 cm and 1.0 mrad /hr at 1 cm, respectvely, measured through not more than 7 milkgrams per square centmeter of total absorber.
8 Decommissioning Protect HP Manager's approval is required before matenal may be released from the TRIGA' if fixed contaminabon levels are in excess of the value stated in the " average" column, even if below the " maximum" column.
Including contaminabon by induced radioactvity,i.e., actvaton.
In evaluation of equipment and materials for fixed or smearable licensed materials, items painted with other than original manufacturer's paint will not be released unless clear process knowledge demonstrates that the paint was applied to a clean, nonradioactive surface prior to use in the Restricted Area or approval from the Decommissioning Project's Health Physics Manager, has been obtained and an acceptable survey course for this situation has been approved, if the potential exists for contamination on inaccessible surfaces, the equipment will be assumed to be intemally contaminated unless (1) the equipment is dismantled allowing access for surveys,(2) appropriate tool or pipe monitors with acceptable detection capabilities are utilized that would provide sufficient confidence that no licensed materials were present, or (3) it may readily be con:luded 6at surveys from accessible areas are representative of the inaccessible surfacer. (i.e., surveying the internal surface from both ends of a straight pipe from a nonradioactive proces system with cotton swabs would be representative of the inaccessible areas).
Bulk Materials or Bulk Liauids-Analysis of representative sample (s) with high resolution gamma spectrometry system. Bulk materials (not to include concrete or asphalt rubble which is addressed separately below) or bulk liquids will be released if no discernible facility-related activity is detected. Minimum Detectable Activities for bulk materials of less than or equal to 10% of the established soil release criteria (see Table 4-3) will be met for principal'IRIGA* isotopes in release evaluations.
- 4.3.2 Facility Release Criteria The proposed decommissioning alternative that has been presented in this Decommission-ing Plan involves the piecemeal dismantlement of the Facility. The results of the Radiological Characterization Scoping Survey has indicated that most areas show promise for being directly releasable without need for decontamination. Building service systems will only be removed from operation when it can be demonstrated that they are no longer needed to provide imponant safety or elliuent/ exposure control functions, or when their removal has been adequately compensated for by other means. Some areas of the Facility could not be directly evaluated during radiological characterization due to inaccessibility or on-going work activities, hence every item or piece of the Facility that is to be dismantled 45
PC-000482/0 i
will be evaluated for ability to decontaminate and ability to demonstrate satisfaction release criteria in the same manner discussed in Section 4.3.1. When it is impractica ;
possible to satisfy release criteria (or conclusively demonstrate that iney have beeni; stem will be treated as radioactive waste. '
As was discussed in the previous section (Section 4.3.1), GA 'will utilize previou established release criteria as provided in Table 4-2 as surface contamination release cri for this 'IRF D&D Project. Removable surface contamination will be eliminated where possible by wiping or other method, but will not exceed the limits specified in Table 4-2. it l Release criteria for fixed and smearable residual radioactivity for beta-gamma emitters !
- j. . would be based upon the relative concentrations of isotopes on the material and their -!
respective applies fromrelease criteria if more than one categorf of nuclide for beta-gamma emitters Table 4-2.
1 l ~
GA may apply the release criteria in Table 4-3 to evaluations of representative samples asphalt, concrete, or other similar construction media that have been reduced to rubble.
l Concrete slabs may be released based upon demonstration of conformance to Table 4-2 and, evaluation of representhtive samples by gamma spectrometry showing no discernible l licensed activity.
- i. 4.3.3 Soil Release Criteria A summary of the proposed soil release criteria is provided in Table 4-3.
In situations where more than one isotope is detected in the soil, determination of conformance to release criteria would be made according to the following method for the mixture: :
I
\
t b + b + ... + "- 51 G, G 2 G, l where C,, C2 ,...C, is the concentration of nuclide 1,2, ... n in the soil above background values '
G,, G2 ,...G, is the release criteria of nuclide 1,2, ... n Table 4-3 shows release criteria based upon the most limiting pathway for nuclides that have been detected in soils or in the Facility during the radiological characterization. If ,
additional nuclides are encountered during the remediation or Final Release Survey -
activities, their respective release criteria would be determined in the same manner as the .
values provided above.
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>) Table 4-3-Soll and Concrete / Asphalt Rubble Release Criteria
- isotope Release Criteria Based upon External Release Crrteria Based upon internal Exposure Limitations F* oosure Limitations (pCUg) (pCUg)
Co 8' l
'2'Cs 10 l
'8'Cs 15 6 1
Eu 11
'S'Eu 10
'"Eu 05 j "Nb 7.5 '
'a'Sb 37 "Sr 1800'
, F"Pu 26' SPu 27"
- Pu 27" 8d'Pu 43268
, 2*Pu 28d
'"Pu 28d 3* ' A m 2 58 Natural Uransum 10' Depleted Uranium 35" Ennched Uranium (8*'U and "U) 30' Thorium (a"Th and aaeTh) 10'
- The release entena shown in tfus tab 6e without annotabon by footnotes 2,3, or 4 were calculated by the hcensee using RESRAD version 5.18 amenng e the same assumptons that were provided in me corre.pui&is hsted in note 2, below. This wn.5pvie b conservative calculabon of me iunuvo uJs conceitraban of an isotope in me soll that by itself would give approximately 10 pR/hr Extemal exposure rate above background for the maxrnum year of exposure. It is the hcensee's intent b apply enteria from this table b concrete, asphalt, or sanitar construeban moda rnatorials that have been ground to a coarse rubble. These entena were approved by tie USNRC for the Hot Cell Decommissioning project by letter dated May 1,1996, Robert C. Pierson to K. E. Asmussen.
- These release enteria are based upon past precedent Ihrough USNRC and State of Cahfomia approved release limits for the GA site.
See Cou puiidois K. E. Asmussen to W.T. Crow, dated Oc*cber 1,1985, corrompuiiG.,@ identification 696 8023 Subject " Docket 70-734: Plan for Obtaining Release of Certain Areas to Unrestncted Use."
- Thsse release enteria are based upon past precedent establ.shed by USNRC through USNRC Pohey issue SECY-81-576, dated October 5,1981, Subtect "Desposal or orHiite strage of residual thorium or uranium (erther as natural ores or without daughters present) from past operations."
- Numbers were established ustry me most limiting of lung dose (20 mrem /yr) or bone dose (60 mrem /yr) using Dose Conversion Factors from NUREG/CR-0150, Volume 2, with an alpha quakty factor of 20, where apphcable, lung mass of 580 grams, and AMAD of 1.0.
. 4.4 Measurements for Demonstrating Compliance with Release Criteria
{
4.4.1 Instrumentation-Type, Specifications, and Operating Conditions Instmmentation utilized during the Final Release Survey (and equipment and materials 8-survey) will be selected based upon the need to ensure that site residual radiation will not exceed the release criteria,In order to achieve this goal, instrumentation that is sensitive to the isotopes of concem and capable of measuring levels below 75% of the guideline values l for those isotopes will be selected. Instrumentation selected will be based upon the e- recommendations of NUREG/CR-5849, Instrumentation that is available for the Final Release Survey, and their respective minimum detection capability was presented in Table '
3-1 of this plan, Instrumentation that is used in the surveys will be calibrated against sources and standards that an: NIST traceable and representative of the representative isotopes encountered at the TRF, Instruments will be operationally tested daily, or prior to each use, whichever is less frequent, Instruments will not be used in conditions that am not in conformance with manufacturer recommendations.
s I
4-7
PC-000482/0 4.4.2 Measurernent Methodology for Conduct of Surveys The entire 'IRF site will be treated as an "affected" area in accordance with the definition provided in NUREG/CR-5849. The yard area was characterized during facility radiological characterization scoping survey. This Decommissioning Plan presumes that the Facility has been decontaminated to the extent practicable and dismantled prior to the Final Release Survey, and that all asphalt and concrete in the TRF yard area has also been removed prior to the Final Release Survey. The 1RF and any impacted areas under the " footprint" of the dismantled 1RF will be methodically remediated as necessary prior to conduct of the Final Release Survey. The characterization results and the continuous feedback from remediation surveys will be the basis for soil remediation efforts.
The only radionuclides identified in the 3RF yard during radiological characterization effons were "7Cs (predominant nuclide),6 Co and to a lesser extent (in only one soil sample) *Cs. The:,e isotopes are readily detectable using p-y sensitive instrumentation.
Furthermore, all of these isotopes are readily detectable with gamma spectrometry techniques as well.
Each grid will be surveyed initially with a surface scanning instrument system to ascertain locations of any elevated concentrations. Areas where the ponable measurement exceeds pre-set alert levels will be designated as judgment sample locations. The precise location of these judgment sample locations will be determined by scanning the surface of the soil.
Judgment locations will be marked with clearly visible stakes and annotated on grid m'aps of the area.
In addition, systematic sampling will be performed within each grid at locations equidistant between the center and each of the four grid block corners (see Figure 4-2).
All soil samples collected during the Final Release Survey will be analyzed by gamma spectrometry and compared to release limits. The sensitivity of the measurement will be at least 10% for the nuclides *Cs,
- Co, and igamma spectrometry Cs which were the only contaminants found during characterization. If any location within a grid requires remediation in order to support a decision in favor of release to unrestricted use, the entire affected grid will be scanned again and re-sampled in accordance with the above methodology after completion of remediation efforts.
4.4.3 Site Survey Grid As was discussed in the previous section, this Decommissioning Plan presumes that the Facility has been decontandnated to the extent practicable and dismantled prior to the Final Release Survey, and that all asphalt and concrete in the TRF yard area has also been removed prior to the Final Release Survey. Hence, the Final Release Survey will deal solely with residual radioactivity in the TRF soils. In developing the Final Release Survey -
approach,it was determined that the TRF would be tmated as "affected" and essentially be surveyed for 100% coverage. In order to support this survey, the TRF would be gridded and staked into areas that are 10 meters by 10 meters (i.e.,100 m2 size grid sections). This .
approach matches the approach that was taken during the soils radiological characterization scoping survey. Figure 4-4 shows the intended survey grid for the TRF Final Release Survey.
4-8
PC 000482/0
< 10M y o
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,* g LOCATIONS OF SYSTEM ATIC SOIL SAMPLING ea Figure 4-2-Systematic Soll Sampling Method 49
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- CLOSELY SPACED TRI ANGULAR GRID PATTERNS
- e Figure 4-3-Modified Systematic Sampling System 4 10
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PC 00048210
.4.4.4 Fixed Contamination Survey Protocol The surfaces of equipment and materials will be surveyed in accordance with Project and GA HP procedures for release of equipment and materials to unrestricted use. Direct frisk will be performed with either a portable Geiger-Mueller, or a gas flow proportional detector, as dictated by the minimum detectable activities of the instrument / probe, or beta-scintillator for the contaminants of concem and the associated release criteria. In any !
situations where process knowledge would suggest a potential for alpha activity, survey with alpha detection instruments or counters will also be employed. In evaluation of equipment and materials for fixed or smearable licensed materials, items painted with other than original manufacturer's paint will not be released unless (1) clear process knowledge demonstrates that the paint was applied to a clean non-radioactive surface prior to use in the restricted area, (2) the paint is mmoved or (3) Health Physics approved paint sampling survey demonstrates that radiation levels under the paint are below the release criteria. If the potential exists for contamination on inaccessible surfaces, the equipment will be assumed to be intemally contaminated unless (1) the equipment is dismantled allowing access for ,
-surveys,-(2) appropriate tool or pipe monitors with acceptable detection capabilities are utilized that would provide sufficient confidence that no licensed materials were present, or .
(3)it may readily be concluded that surveys from accessible areas are representative of the !
inaccessible surfaces (i.e., surveying the internal surface from both ends of a straight pipe from a nonradioactive process system with cotton swabs would be repmsentative of the inaccessible areas). The results of contamination surveys will be recorded either on survey maps or special release logs. Results of all surveys will be compared to average and maximum criteria prior to any material being released.
4.4.5 Removable Contamination Survey Protocol Removable contamination will be assayed by collection of 100 cm2 smears from surfaces, or as practicable. The smear samples will be evaluated using suitable hand-held instruments or low level beta counting systems. As discussed in Section 4.4.4, smears will be evaluated for alpha contaminants if process or survey information mcommends this, though ;
'IRF Decommissioning Health Physics personnel routinely evaluates a portion of its e positive smears for. alpha contamination. Survey evaluations are recorded in the same manner described in Section 4.4.4. .
4.5 Methods to be Employed for Reviewing, Analyzing, and Auditing Data 4.5.1 Laboratory / Radiological Measurements Quality Assurance During decommissioning survey activities, many direct and indirect measurements and ,o sample media samples will be collected, measured, and analyzed for radiological contaminants. The results of these surveys will be utilized to evaluate the suitability of the material or item for release to unrestricted use, or whether decontamination of structures, ,* '
components, and the surrounding site have achieved the desired result. Sample collection.
analysis, and the associated documentation will adhere to written procedures and meet the guidance of the USNRC, as well as comply with recognized industry recommendations and good practices. Outside (i.e., non-GA) laboratories selected to analyze 'IRF decommissioning samples shall be approved by the GA Quality Assurance organization and listed on the QA Approved Suppliers List maintained by the GA Quality Assurance Group.
Organizations that perform radiological monitoring measurements recognize the need to establish quality assurance programs to assure that radiological monitoring measurements are valid. These programs are established for the following reasons: (1) to readily identify deficiencies in the sampling and measumment processes to those individuals responsible 4 12
I PC-000482/0 ;
for these activities so that prompt corrective action can be taken, and (2) to routinely l monitor the survey and laboratory measurement results in order to assure that results and !
conclusions are valid.
. Written procedures will be used for sample collection in order to ensure that samples are representative. Written procedures will also be utilized for sample preparation to ensure that media are prepared in accordance with laboratory specifications. Achain of custody will be maintained on all radiological samples to ensure integrity of the sample. Quality control records for laboratory counting systems will include the results of measurements of radioactive check sources, calibration sources, backgrounds, and blanks. Records relating to overalllaboratory performance will include the results of the analysis of quality control samples such as analytical blanks, replicates, inter-laboratory cross-check samples, and other quality control analyses.
4.5.2 Cross-check of Results I Final radiation survey results that are based upon laboratory measurements via gamma !
spectroscopy are subject to a laboratory quality assurance program. An important j complement to this quality assurance program is cross-check of results (Ref.10.13). GA l will cross-check gamma spectroscopy results by a method consisting of three separate i aspects intended to complement normal laboratory quality assurance measures. Replicate l sample counting of approximately 5% of the analytical load will be performed to compare results to previous results for the same sample. The laboratory will not be made aware of the identify of replicate samples. Additionally, periodic blank samples (samples with no licensed materials) will be sent to the laboratory, again without identification, to check for laboratory contamination. Finally, periodic cross-check blind spiked samples of known isotopic constituents will be sent to the laboratory without identification, to allow Project Health Physics management to compare results to stated assay. Wherever- significant questions regarding data validity are identified, affected results will be recounted after ;
system performance is restored to acceptable levels.
4.5.3 Supervisory and Management Review of Results i
Radiological surveys are conducted by Health Physics Technician staff members who are trained and qualified. In addition, radiological surveys and sample results are reviewed by a senior level member of the Project Health Physics staff other than the individual that performed the survey. Final Radiation Survey data is also reviewed by the Project HP Manager and the 'IRF D&D Project Manager.
8 9*
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PC-000482/0
- 5. TECHNICAL SPECIFICATIONS GA has requested amendments to the Technical Specifications for the Mark 1 Reactor in a " Request for License Amendment to Withdraw Authorization to Operate General Atomics 'IRIGA* Mark 1 Non Power Reactor," Facility License R-38; Docket 50-89 (Ref.10.2).
The Technical Specifications for the Mark F Reactor are shown in " Issuance of Amendment 43 to Facility License No. R-67 -- General Atomics (Ref.10.3).
The Mark III is inactive, Terminated License No. R-100 (Ref.10.4).
As decommissioning progresses, further requests for changes to the Technical Specifications will e be submitted in an application for amendment to the license pursuant to 10 CFR 50.90.
. J e
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- 6. PHYSICAL SECURITY PLAN 1 1
All GA radiation restricted areas are secured from unauthorized entry. During non-working hours, l all nuclear facilities are locked. GA maintains 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> security watchmen to patrol the site. !
l l
- Existing physical security and material control and accounting plans approved by the Nuclear
- Regulatory Commission, as may be amended, will continue to be implemented.
These existing plans meet the requirements in 10 CFR 70.38 for decommissioning, and will be
, maintained as required by the MkF Possession Only License amendment (Ref.10.3), and the
- proposed Mk1 Possession Only License amendment (Ref.10.2) when approved. .
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- 7. EMERGENCY PLAN I
GA has a General Emergency Plan and a Radiological Contingency Plan, supplemented by l
4 procedures specific to the 'IRF, as required by the USNRC and State of California. Training on the
- Radiological Contingency Plan is provided to the Emergency Response and Recovery director and alternates. Emergency Response Team members receive training appropriate for responding to
- emergencies.
1 i l a 4
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- 8. ENVIRONMENTAL REPORT The Environmental Report is provided as Appendix B.
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- 9. CHANGES TO THE DECOMMISSIONING PLAN As the decommissioning progresses, changes to the Technical Specifications up to termination of the license will be via a Request for License Amendment pursuant to 10 CFR 50.90.
GA requests that other changes to the Decommissioning Plan be allowed without prior USNRC approval which involve decommissioning activities unless an unreviewed safety questions is involved. An unreviewed safety question is defined as: "When the possibility of occurrence or the consequences of an accident or malfunction of equipment imponant to safety is increased beyond previously planned decommissioning activities."
Reports, records of change, and retention of documents will be in accordance with the applicable
" portions of 10 CFR 50.59.
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PC-000482/0 i
{ 10. REFERENCES 10.1 NUREG-1537 Rev. 0, " Guidelines for Preparing and Reviewing Applications for the
, Licensing of Non-Power Reactors" i
- . 10.2 Facility License R-38; Docket 50-89
- " Request for License Amendment to Withdraw l Authorization to Operate GA TRIGA* Mark I Non-Power Reactor," December 17,1996 10.3 Issuance of Amendment No. 43 to Facility License No. R-67 General Atomics i
j 10.4 USNRC License No. R-100 (terminated) t
- 10.5 Asmussen, Keith E. Letter No. 696-2581 to Document Control Desk, USNRC, ATTN: '
Mr. Alexander Adams, Jr. And Mr. Charles E. Gaskin, " Docket Nos. 70-0734,50-89 and
- j. 50-163; Decommissioning Financial Assurance," dated May 20,1996 t l# 10.6 Weiss, Seymour H. And Robert C. Pierson Letter to Dr. Keith E. Asmussen, " Financial i Assurance for USNRC Licenses SNM-696, R-38, R-67/ Docket Nos. 70-0734, 50-89,50-j 163," dated July 9,1996 -
I j 10.7 ASME-NQA-1-1989," Quality Assurance Program Requirements for Nuclear Facilities" i
l 10.8 General Atomics Quality Assurance Program Approval by the USNRC Transportation & l l Storage Inspection Section, Spent Fuel Project Office, Nuclear Materials Safety and i 1 Safeguards, Approval No. 0030, Rev. 6, July 9,1996 i 10.9 USNRC Regulatory Guide 7.10," Establishing Quality Assurance Programs for Packaging i
- Used in the Transpon of Radioactive Material" J 1
i 10.10 NUREG/CR-5849," Manual for Conducting Radiological Surveys in Support of License
- Termination," Draft for Comment, June 1992 l 10. I 1 ANSI Z88.2-1980, " Practices for Respiratory Protection" i 10.12 National Council on Radiation Protection and Measurements (NCRP), NCRP Report No.
} 50, " Environmental Radiation Measurements," December 27,1976 i
i 10.13 U. S. Nuclear Regulatory Commission Regulatory Guide 4.15, " Quality Assurance for i
Radiological Monitoring Programs (Normal Operations)-Effluent Streams and the !
- _ a, Environment," Revision 1 February 1979 j 4
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APPENDIX A ,
SUMMARY
OF i CHARACTERIZATION RESULTS l l
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A-1 l
PC 000482/0 LIST CF TABLES Ttble A-1-TRF Area Classifications for Characterization....... .. .. .............. .......... ...... .., ............... A-2 Table A-2-Results of Radiochemical Analyses for TRF Soil Media Samples.. ........ . ......... .. .. .. ..... A-3 Table A-3-Results of Radiochemical Analyses for TRF Asphalt / Concrete Media Samples.... . .. ... .. A 4 LIST OF FIGURES Figure A-1-Grid Map of TRIGA* Reactor Facility used for Media Sample Locations for Radiological f ScopingStudy.........................................................................................................A5 Figure A-2-Grid Map of TRIGA* Reactor Facility showing Radiological Measurement Results from
- Radiological Scopin g St udy. . . . .. . . . . . . . . . . . ... . . . . . . ..... . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . .. . . . . A-6 Figure A-3-Schematic of the Room Layout for the TRIGA* Reactor Facility, Building 21........ ... .... .. A 7 a'
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APPENDIX A
SUMMARY
OF CHARACTERIZATION RESULTS Radiological Sconine Studv ;
A radiological scoping study was performed on the 'IRIGA* Reactor Facility ('IRF) to support l decommissioning planning. Due to the operational status of the TRIGA* Mark i Reactor, the radiological scoping characterization was not an invasive study that involved defacing facility structures by collection of building construction media samples. Outside the 'IRF media samples were obtained for surrounding soils, asphalt, and concrete.
1 The 'IRF was fint evaluated from process knowledge and past radiological surveillance. Based upon this evaluation, various portions of the complex and its associated yard are classified as either l "affected" or " unaffected." Affected areas are areas that have potential radioactive contamination 3
(based on process knowledge) or known radioactive contamination (based on past radiological surveillance). Unaffected areas are all areas not classified as affected. Table A-1 shows the classifications of areas within the facility.
a Table A 1-TRF Area Classifications for Characterization f Affected Areas Unaffected Areas )
Room 102--diark I Reactor and Control Room Room 100-Office j Room 105-Tool Shop Room 101-Office l Room 106-Counting Room Room 101 A-Dark Room (
Room 107-Mark F Reactor Room 103-Office Room 10&--Mark F Control Room Room 104-Office Room 109-Mark III Control Room Restrooms Room 110-Mark III Experimental Area Entry and Of0ce Area Corridors Room 1 Il-Mark III Reactor Electrical Pads Room i12-Thermal Stability X-Ray Room I koom ll3-Thermal Stability Lab Room i14-North Entry Room i15-Decontamination Room I All Yard Area / Roof Mark I Shed Cooling Tower Yard Area Storage Shed Machine Shop Sampling protocol, sample preparation, survey and media result documentation, and analytical methods for the scoping study were based upon Refs. A-1 through A-7. Pages A-8 through A-66 are copies of the survey maps from the study.
Surroundiny Soils. Asohalt. and Concrete O , A grid map of the 'IRF, including building surrounding areas (soils, asphalt, concrete) is provided ;
in Figure A-1. !
Systematic media samples of the'IRF were obtained at grid intersections shown in Figure A-2. In addition, one judgment sample location was used to obtain soil / asphalt samples based on process knowledge. Thirty-two soil media samples, 21 asphalt media samples, and one concrete media sample were obtained for radiochemical analysis. The first approximate 6 inch depth of soil in the shape of a 4 inch diameter cylinder, carved out by a coring tool, was obtained for soil media samples. In one case, location (X,1), no soil could be collected due to the depth of the asphalt.
Asphalt and conc ete media samples were obtained similarly using the coring tool to produce approximate 6 inch tall, 4 inch diameter cylindrical samples of asphalt / concrete. Figure A-2 presents the media sample radiological results entered on the grid map, for the 'IRF.
A-2
PC-000482/0 The results of radiological analyses for soil media samples including sample location and identifica-tion number are provided in Table A-2. The results of radiological analyses for asphalt / concrete media samples including sample location and identification number are provided in Table A-3.
Facility Affected Areas The radiological scoping survey showed the facility areas to be well maintained with only minor amounts of residual radioactivity discovered. With the exception of a single grid location, the yard soil results were all below established release criteria. It is recognized that components associated with the 'IRIGA* reactors, and some surrounding structural materials will be activated as a result of reactor operations. Inside of the facility itself, surveys of affected areas showed very little residual radioactivity. Rooms 102 and 107, the Mark I and Mark F Reactor rooms respectively, both showed several locations of fixed radioactivity on the floor surfaces, and Room 107 also had %
one location ofloose surface activity. Figure A-3 is a schematic of the room layout for the TRF.
Table A 2-Results of Radiochemical Analyses for TRF Soll Media Samples Location & Remarks Sample ID Gamma Isotopic Results (pCi/g) ,
Grid (X. 4). soil sample beneath asphalt 21S-97-001 natural activity only Grid (X 5). soil sample beneath asphalt 21S-97-002 0.G4 "'Cs l Grid (Y. 4). soil sample beneath asphalt 21S-97-003 natural activity only G_ rid (L 5). soil sample beneath asphalt 21S-97-0G4 0.04 '"Cs, 0.10 "'Cs Gwi (Y. 6), soil sample beneath asphalt 2 t S47-005 natural activity only ,
,Crid i (Z. 6), soil sample beneath asphalt 21S-97-006 0 06 *"Co. 0.37 "'Cs l Grid (Y+0.5. 6.5). soil sample beneath asphalt 21S-97-007 0.26 "'Cs j Grid /(Y. 7). soil sample beneath asphalt 21S-97-008 0.03 "'Co. 0.53 "'Cs Grid (X,6). soil sample only 21S-97-009 0.45 "'Co. 0.02 '"Cs.1.24 "'Cs Grid (Y. 5), soil sample beneath asphalt 21S-97-010 natural activity only Grid (Z. 7). soil sample only 21S-97-011 9.24 "'Co. 29.59 "'Cs Grid (W. 4). soil sample beneath asphalt 21S-97-012 natural activity only Grid (Z. 2), soit sample only 21S-97-013 0.43 "'Co. 0.59 "'Cs Grid (Z.1), soil sample only , 21S-97-014 0.69 "'Co. 0.43 "'Cs Grid (Z. 0). soil sample only 21S-97-015 0.30 "'Co. 0.30 "'Cs Grid (Y,0), soil sample only 2 t S-97-016 0.16 "'Co. 0.42 "'Cs Grid ( A.1). soil sample only 21S-97-017 0.95 "'Co.1.26 "'Cs Grid (Z 1), soil sample oniv _
21S-97-018 0.11 "'Co. 0.17 "'Cs Grid (Yg ).l soil sample only 21S-97-019 0.06 "'Co. 0.20 "7Cs Grid (X d), soil sample only 21S-97-020 0.22 "'Co. 0.29 "'Cs Grid (U.1) soil sample only 21S-97-021 0.11 "'Co. 0.60 "'Cs Grid (A. 4) soil sample beneath asphalt 21S-97-023 0.10 "'Cs ,3 Grid (A. 3), soil sample beneath asphalt 21S-97-024 natural actisity only Grid (A. 2). soil sample beneath concrete 215-97-025 0.G8 '"Cs Grid (V. 2). soil sample beneath asphalt 21S-97-026 natural activity only ,, g Grid (V.1). soil sample bencath asphalt 21S-97-027 natural activity only Grid (X,0). soil sample beneath asphalt 21S-97-028 natural activity only Grid (X, 3), soil samph beneath asphalt 21S-97-029 0.06 "'Cs Grid (V. 0). soil sample beneath asphalt 21S-97-030 natural activity only Grid (W. 0) soil sample beneath asphalt 21S-97-031 natural activity only Grid (W.1). soil samnie beneath asphalt 21S-97-032 natural activity only Grid (Y.1). soil sample beneath asphalt 21S-97-033 natural activity only A-3
PC-000482/0 Table A 3-Results of Radiochemical Analyses for TRF Asphalt / Concrete Media Samples Location & Remarks Sample ID Gamma Isotopic Results (pCi/g)
Grid (X. 4). asphalt sample 21 B-974X)] natural activity only Grid (X. 5), asphalt sr..nple 21 B-97-002 natural activitv oniv Grid (Y. 4), asphalt sample 21 B-974X)3 natural activity only Grid (Z, 5). asgi: sample 21 B-974XM 0 48 Cs Grid (Y. 7). asphast sample 21 B-97-005 0.38 "'Co.1.38 "'Cs Grid (Y. 6), asphalt sample 21B-974X)6 0.30 "'Cs Grid (Z. 6). asphalt sample 21 B-974)07 1.32 "'Cs Grid (Y+0.5. 6.5), asphalt sample 21 B-974X)8 2.90 "'Cs
- Grid (W. 4). asphalt sample 21 B-974)09 natural activity only Grid (Y. 5). asphalt sample 21 B-97-010 0.40 "'Co. 0.16 "'Cs Grid (X,3). asphalt sample 21 B-97-011 0.22 "'Cs O Grid ( A. 4). asphalt sample 21B-97-012 0.31 "'Co 0.25 "'Cs Grid (X,0). asphalt sample 21 B-974)l3 natural activity oniv Grid (Y I), asphalt sample 21 P-97-014 0.35 "'Cs Grid (V. 2), asphalt sample 21 B-97-015 natural activity only Grid (V I) asphalt sample 21 B-97-016 natural activity oniv Grid (W.1), asphalt sample 21 B-974)l7 natural activity only Grid (V,0), asphalt sample 21 B-97-018 natural activity only Grid (W. 0). asphalt sample 21 B-97-019 natural activity only Grid (X,1), asphalt sample 21 B-97-020 0.73 "'Cs Grid ( A. 3). asphalt sample 21 B-97-021 natural activity only Grid ( A. 2), concrete sample 2]C-974X)1 natural activity only IJnaffected Areas The remaining rooms and areas of the 'IRF are considered unaffected areas. For these unaffected areas, radiological surveys consisted oflarge area masslin smears of the floor and walls to roughly 2 meters above the floor, at I m2 intervals, floor monitor / surface scan surveys, and contamination surveys of floor drains and sinks. No areas of contamination were discovered in surveys of unaffected areas.
References A-1 Health Physics Procedure HP-524, Laboratory Procedure for Gamma Spectroscopy for Hot Cell Samples, Issue B, October 1994.
3 A-2 Health Physics Procedure HP-525 Review and Evaluation of Gamma Ray Spectral Data for the GENIE-ESP System, Issue A, August 1995.
A-3 Health Physics Procedure HP-522, HCF Soil, Sample Media, Water, and Vegetation
,a Sample Preparation and Exceptions List, Issue A. July 1994.
A-4 Health Physics Procedor: rip-526, Performance Testing and Operation of Scalar Counting Systems, Issue B, August 1994.
A-5 Health Physics Procedure HP-502, Control of Health Physics Instrumentation, Issue A.
July 1994.
A-6 Health Physics Procedure HP-500, Radiological Surveys, Issue A, March 1994.
A7 GA Radiochemistry Procedure ACD:RC-041, Issue B, Gamma Counting on the Canberra S-100 Gamma Ray Spectrometer, April 1994.
A-4
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_ - _ _ - - - - - - - - - - - - - - - - - - - - - - - - - - - ~ ' ' ' - - " " - - ' V e APPENDIX B ENVIRONMENTAL REPORT B-1
TABl.E OF CONTENTS
- 1. P U R P O S E A N D N E E D FO R A CTI O N . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B - 4
- 2. DESCRIPTION OF THE FACILITY, PROFOSED ACTION AND ALTERNATIVES.....B-4 2.1 Facility Description . .. . .. .. ... . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . . . . . . . B -4 2.2 Proposed Action.. . . . . . . . . . .... . . . . . . . . . . . . . . . . . . . . .. ... .. . . . . .... . .B-8
- 3. DESCRIPTION OF THE AFFECTED ENVIRONMENT..... . ...... .... .. ......................B-9 3.1 Man-Made Environment. .. . . . . . . .. . . . . . . . . . . . . . . . .. . . . . .. . . . B-9 3.1.1 Radioactive Materials.... . ...... .. . . . . . . . . . . . . . . . . . .. . . . . . .. . .B-9 3.1.2 Hazardous Materials .. .. . . .. . . . . . . . . . . . . . . . . . . . . . . . . .. . . . ... . B-10 3.1.3 Transportation . ....... . . . . . . . . . . . . . . . . . . . ... . . . . . ... B-11
, 3.1.4 Cultural and Historical Resources ... . .. . . .. .. .. .. .. . . . . . . B-11 3.1.5 Population and Land Use. . .... . . . . . . . .. . . . . . . . . . . . . . . . . . . . .. . B-11 3.1.6 Noise . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . B-12 3.1.7 Aesthetics., .... . . . . .. . . . . . . . .. . . ... . ..... . . . . . B-12 = 3.2 Natural Environment... ... . . . ... .. . . . . . . . . .. . . . .. . .. B-12 3.2.1 Topography, Geology and Seismicity.. .. .. . . .. . . . .. . . ... .. . ... . B-12 3.2.2 Climate and Air Quality.. ........ . . . . . . . . . . . . . ..... .. . . . ... .. B-14 3.2.3 Hydrology... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . . . . . B- 16 3.2.4 Biology....... .... ... . . . . . . . . . . . . . . . . . . . . . . . . .... . . . . . . . . . . . . . . .. . . B-16 3.2.5 Socioeconomics and Environmental Justice... . . ... . . . . . . . . . . . . . . . . . ... .. . B-18
- 4. POTENTIAL ENVIRONMENTAL CONSEQUENCES OF PROPOSED ACTION AND ALTE RN ATlVES .......... ...............................................................B-18 4.1 Human Health Effects..... . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . ...........B-18 4.1.1 Hazard identification.. .. ..... ..... . .. . . . . . . .
. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. B-18 4.1.2 Potential Exposures.... . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . ...............B-19 4.1.3 Transportation . .. . . . . ..... ..... . .... ... . . . . . . . . . . . . ...... ... ..... ..... ..........B-19 4.2 Waste Disposal .... .... .. ... . ...... . .. ... . . .. .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . B-20 4.2.1 Hazardous Waste... . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . B -2 0 4.2.2 Low-Level Radioactive and Mixed Waste ... . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . B-20 4.2.3 Non-Hazardous Solid Waste . .. .. . .. .... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B -2 0 4.3 Noise . . . . .. ..... ................................................. ..........................B-20 4.4 Seismicity.. . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . B-20 4.5 Air Quality... ... ..... ..... .... .. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..........B-20 4.6 Regulatory issues.. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. . . . B-21 4.7 Areas Not Affected.. . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... ..... ....B-21 4.8 Cumulative Effects.. ... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . B -22 4.9 Alternatives to Proposed Action....... ...... ..... .... . ..... .. . ... . . . . . . . . . . . . . . . . . . . . . . . . . .. . B 23 # 5. REFERENCES......................................................................................B-23 9
B2
LIST OF FIGURES Figure B-1-Regional Location... .. .. . . . . . . . .. . . . . . . . . . . . . . . . . ... B 5 Figure B-2-GA Site and Surrounding Uses..... .. . ... .. . . . . . . . . . . . . . . . . . . . . . . . . .. . B-6 Figure B 3-TRIGA* Reactor Facility (Building 21) Layout... ...... . . . . . . . ... .. . . . . . . B-7 LIST OF TABLES Table B-1-List of Potential Radionuclides.. ... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . B-10 Table B 2-Applicability of Environmental Statutes and Regulations.. . . . . . . . . . . . .. ... B-22 a
\
1 B-3
APPENDIX B ENVIRONMENTAL REPORT
- 1. PURPOSE AND NEED FOR ACTION As a result of nuclear research training and isotope production, activities conducted since 1958 for the DOE and its predecessor agency, the Atomic Energy Commission (AEC), and commercial customers, the 1RIGA* Reactor Facility (TRF) has become contaminated with varying amounts of radioactive materials and small amounts of hazardous materials. GA decided to shut down the 'IRF due to reduced demand and continuing private industrial development around the site. D&D of the 1RF will eliminate the potential for future inadvenent environmental releases. The purpose of the D&D activities is to obtain from the NRC and the CAL-DHS release of the site for " unrestricted use." The term " unrestricted use" means that there will be no restrictions on the use of the site, other than those imposed by the City of San Diego zoning ordinance.
o 3. DESCRIPTION OF THE FACILITY, PROPOSED ACTION AND ALTERNATIVES 3.1 Facility Descriptior. The 3RF is located within the General Atomics " Main Site." GA occupies approximately 120 acres (48 hectares) on two contiguous areas approximately 13 miles (21 km) north of downtown San Diego, Califomia, just southwest of the convergence of Interstates 5 and 805, and approximately one mile east of the Pacific Ocean. The two locations are referred to as the " Main Site" and the "Sorrento Valley Site', or collectively as the GA site. The locations of Gain relation to San Diego County is shown in Figure B-1 and Figure B-2. As shown in Figure B-3, the TRF occupies Building2 21 and an outdoor service yard. The interior of the Building 21 has approximately 7,600 ft 0f floor space consisting of offices, three reactor 2 rooms, operating rooms and auxiliary areas. Building 21 is surrounded by a 43,800 ft fenced service yard. The TRIGA* Reactor Facility houses three TRIGA* reactors which have been variously used since 1958 to provide controlled neutron and gamma irradiation for diverse research projects. The Mkl1RIGA* Reactor and associated control systems (situated in Room 21/102) are fully operational under USNRC License No. R-38. The reactor fuel and components in the reactor pool are fully assembled and configured for routine operation. A request to amend r the license to Possession-Only was submitted on December 17,1996. The MkF TRIGA* Reactor (situated in Room 21/107) has been placed in " Possession-Only" status under USNRC License No. R-67, and is currently inoperable. All reactor fuel elements have been removed from the MkF reactor core / shroud and placed in the Fuel Storage Canal. The non-fuel components of the MkF reactor, including the core support structure, bridge shroud, beam tubes, and associated hardware, remain in the reactor pool. The Fuel Storage Canal ponion of the MkF reactor pool currently houses all of the Spent Nuclear Fuel (SNF) elements previously removed from the MkF and MkIII reactors. All required protection barriers and security systems, installed due to the presence of HEU in the Storage Canal, are currently operational. l B-4
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AL Figure B-3-TRIGA* Reactor Facility (Building 21) Layout B7
The MkIII 1RIGA* Reactor (situated in Room 21/111) is currently inoperable, the associated USNRC License No. R-100 having been terminated in 1973. All fuel elements have been removed from the MkIII reactor core. The MkIII reactor pool has been panially drained of water, and the n maining liquid in the pool has been treated with anti-corrosion agents forlong term storage of the remaining activated reactor components still situated in the pool. 2.2 Proposed Aetion The proposed action and the alternatives are as follows: Proposed Action (DECON)-Decontamination and Decommissioning followed by 4 release of the site to unrestricted use. Altemative 1 (SAFSTOR)-In safe storage, the facility is placed and maintained in a condition that allows it to be safely stored and subsequently decontaminated to a level
, permitting release of the property by NRC.
- Altemative 2 (ENTOMB)-In entombment, radioactive materials are encased in a structurally long-lived material such as concrete. The entombed structure is appropriately maintained and surveillance is continued until the radioactivity decays to a level permitting release of the property by NRC.
Implementation of the proposed action would be accomplished by the following tasks:
- 1. Dismantle, decontaminate or package as LLW, the MkIII Reactor components, tank and pit.
- 2. Dismantle, decontaminate or package as LLW, the Mk1 Reactor components, tank and pit.
- 3. Decontaminate any remaining contaminated areas except the MkF rooms.
- 4. Reroute services to isolate the MkF rooms.
- 5. Dismantle the facility except the MkF rooms.
- 6. Obtain DOE approval and ship the fuel from the MkF storage canal.
- 7. Dismantie, decontaminate or package as LLW, the MkF components, tank and pit.
p 8. Dismantle the MkF rooms.
- 9. Ship the LLW.
- 10. Perform the final survey and submit a request to the NRC for release to unrestricted use.
To minimize the risks of inadvenent exposure, contamination and/or radioactive releases, all D&D operations would be implemented with appropriate technical and administrative controls, including:
- 1. Performing all work pursuant to approved procedures implementing an NRC and CAL-DHS-approved Decommissioning Plan. GA would continue to be responsible for assuring and demonstrating compliance with NRC and CAL-DHS licenses, as well as other applicable federal, state or local laws, regulations, licenses and/or permits.
l B-8
- 2. Using containment structures, tents, and bags under negative pressure to isolate operation areas and prevent release of contaminants.
- 3. Employing monitored high-efficiency particulate air (HEPA) filtration systems.
- 4. Maintaining emergency ventilation, power, and supplies, as appropriate.
- 5. Applying ALARAprinciples such as emphasizing radiation protection for workers and the general public, employing personnel and area dosimetry, using personal protective equipment and clothing and conducting work through approved Radiological Work Authorization Permits. The term "ALARA" means as low as reasonably achievable, taking into account the state of technology and the economics of improvements in relation to the benefits to public health and safety, and other societal and socioeconomic e considerations. Health physics staff would have the authority to stop any operations which they believe may involve unusual, unnecessary or excessive radiological risk to the worker, the public or the environment.
A
- 6. Maintaining access control to the site and facility to restrict unauthorized individuals from the work area.
- 7. ' Integrating Quality Assurance and Health and Safety requirements into project documents.
- 3. DESCRIPTION OF THE AFFECTED ENVIRONMENT 3.1 Man Made Environment 3.1.1 Radioactive Materials The public is continuously exposed to radiation from natural sources; primarily from cosmic radiation; extemal radiation from natural radioactive material in the earth and global fallout; and intemal radiation from natural radioactive materials taken into the body via air, water, and food. The public receives and accepts the risks associated with radiation exposures from medical X-rays, nuclear medicine procedures, and consumer products. On average, a member of the public in the United States receives approximately 300 mrem /yr from natural soumes of radiation; approximately 50 mrem /yr from medical procedures; and approximately 10 mrem /yr from consumer products, for a total of 360 mrem /yr (Ref. 5.1).
In San Diego, at elevation near sea level, the background radiation from natural sources is - about 240 mrem /yr and the total background radiation is approximately 300 mrem /yr. The primary contarranation from past TRIGA* activities is contained within the TRIGA* Reactor Building and is monitored under an extensive surveillance and maintenance program. Existing monitoring data and historical information indicate that building
- contamination consists of fission products and activation products. Some TRF reactor #
components are contaminated with radionuclides. This is primarily the result of deposition and adherence of airbome and water-soluble contaminants. The radionuclides shown on Table B-1 potentially exist in the TRF. Table B-1 also shows the half-life of the radionuclides. The half-life (Ref. 5.2) is the time required for half the initial number of nuclei to physically decay. B-9 l
Table B 1-List of Potential Radionuclides Chemical Element isotope Half Life (years) Cobalt "Co 5.27 Niobium Nb 20.00 Tantalum '8"Ta 0.32 Krypton 85Kr 10.72 Strontium "Sr 29.00 Ruthenium *Ru 1.02 Antimony '2'Sb 2.75 Cesium "'Cs 2.06 9 Cesium '37Cs 30.17 Cerium '"Ce 0.78 Europium '5'Eu 8.59
, Europium *Eu 4.71 Radioactive atoms undergo spontaneous nuclear transformations and release excess energy in the form ofionizing radiation. Such transformations are referred to as radioactive decay.
As a result of the radioactive decay process, one element is transformed into another; the newly formed element, called a decay product, will possess physical and chemical properties different from those of its parent, and may also be radioactive. A radioactive species of a particular element is referred to as a radionuclide or radioisotope. Radiation emitted by radioactive substances can transfer sufficient localized energy to atoms to remove electrons from the electric field of their nucleus (ionization). In living tissue this energy transfer can destroy cellular constituents and produce electrically charged molecules (i.e., free radicals). Extensive biological damage can lead to adverse health effect (Ref. 5.3). The adverse biological reactions associated with ionizing radiation, and hence with radioactive materials, are skin injury, cancer, genetic mutation and birth defects (Ref. 5.4). Major types of ionizing 3adiation include alpha panicles, beta, and gamma or X-ray radiation. Alpha particles expend their energy in short distances and will not usually penetrate the outer layer of skin. Alpha panicles represent a significant hazard only when taken into the body, where their energy is completely absorbed bj small volumes of tissues. Beta particles constitute external hazards if the radiation is within a iew centimeters of exposed skin surfaces and if the beta energy is greater than 70 kev. Internally, beta particles deposit much less energy to small volumes of tissue and, consequently, inflict much less damage than alpha panicles. Gamma radiation are of the most concern as extemal hazards. / 3.1.2 Hazardous Materials o Hazardous materials of concem in terms of potential exposure to D&D workers, on-site GA employees and off-site neighbors are elemental lead, cadmium and asbestos. Elemental Lead-The predominant hazardous material in the TRF, in terms of mass, is elemental lead (used primarily in various radiation shielding applications). Most lead contained in the facility consists of solid, non-dispersible bricks, fittings, liners and weights. Lead is a cumulative poison. Increasing amounts can build up in the body eventually reaching a point where symptoms and disability occur. The effects of exposure to lead dust through inhalation and ingestion may not develop quickly. Symptoms may include decreased physical fitness, fatigue, sleep disturbance, headache, aching bones and muscles, constipation, abdominal pains and decreased appetite. lead can also cause B-10
irritation to the skin and eyes. These effects are reported to be reversible if exposure ceases. Systemic effects are possible if a long-term exposure occurs and birth defects have been reported. Asbestos-Asbestos is present in 'IRF construction materials (e.g., floor tiles, roofing material). Asbestos is not a hazard unless it is " friable," that is in powder or fiber form. Inhalation of the fibers can cause asbestosis and lung cancer. Gastrointestinal cancer can be. ' caused by ingestion. Asbestos found to be present in the 'IRF will be removed by a licensed asbestos abatement contractor. Cadmium-Cadmium is present in the 'IRF in the form of metal foil. Inhalation or ingestion of cadmium dust or fumes can affect the respiratory system, kidneys, prostate and blood. Symptoms are: pulmonary edema, dyspnea, cough, tight chest, substemal pain, r headache, chills, muscular aches, nausea, diarrhea, anosmia, emphysema. 3.1.3 Transportation The main roadways in the vicinity of the GA site are shown on Figure B-2. They include Genesee Avenue beyond the southern boundary, John Jay Hopkins Drive beyond a portion of the western boundary, North Torrey Pines Road further to the west, and Interstate 5 to the east. Genesee Avenue is a four-lane primary anerial currently undergoing extensive road reconstruction. North Torrey Pines Road north of Genesee Avenue is a six-lane primary arterial. North of Science Park Road, North Torrey Pines Road becomes a four-lane primary arterial. John Jay Hopkins Drive is a four-lane collector street which connects Genesee Avenue with Nonh Torrey Pines Road. The GA site is generally accessed from the Interstate 5 freeway, exiting on Genesee Avenue and traveling west, tuming north on John Jay Hopkins Drive and east on General Atomics Court. The site can be entered through two entrances shown on the map (Figure B-2) from General Atomics Court and from John Hopkins Court. Traffic onto the site is controlled by a guard posted at a guard station and by personnel at an office reception area. OE-hour access is through a keycard gate at the south entrance. The nearest entrance to the GA compound is approximately 1,500 ft. (457 m) from the 'IRF. 3.1.4 Cultural and Historical Resources t No significant archeological or cultural resources have been found in surveys of the GA site. The National Register of Historic Places mentions no historical structures or sites within the boundary of the plant. There is a state park, called Torrey Pines State Park, located one mile to the northwest of the site, which contains a unique species of pine tree. s No historical, archaeological or cultural properties are believed to be under consideration on I or near the TRF. 3.1.5 Population and Land Use ? The site is located within the Torrey Pines Mesa area and is currently zoned SR (Scientific Research). The University Community Plan designates open space and scientific research land uses for the site. Land uses surrounding the GA site include scientific research and development parks to the north and to the east across Interstate 5, undeveloped land associated with Torrey Pines State Park, research and development parks and a hospital to the west and the University of California at San Diego (UCSD) to the south. Surrounding land uses are shown graphically on Figure B-2. B-11
The present population within the University Census Tract Subregion, in which the main site lies, is primarily of an industrial and university campus makeup, with an estimated daytime total of up to 52,000 people (Ref. 5.5) including about 1,200 GA employees. The University Subregion contains six Census Tracts. The immediate vicinity of the Flintkote Avenue facilities is zoned for industrial activity. Estimates of future growth indicate that the University Subregion could have a daytime total of 57,000 people by year 2000, based upon future industrial growth in the Sorrento Valley area and an increased number of students on the university campus. Because of terrain, zoning, and current land use, most future residential development will occur beyond a two mile radius from the site. 9 Nearby sensitive human populations include: GA non-radiological workers; Agouron Pharmaceuticals, located 0.25 miles (0.4 km) to the west; Children at a day care center, located on John Jay Hopkins Drive, approximately 0.45 miles (0.7 km) to the west; Scripps Green Hospital, located 0.5 miles (0.8 km) to the west; UCSD dormitories located about 0.9 miles (1.5 km) to the south; and Aresidence along Torrey Pines Road across from the UCSD campus (about 1.2 miles or 2 km to the southwest). 3.1.6 Noise Within GA site boundaries, the ambient noise environment is generated by vehicular traffic, jet aircraft, general aviation aircraft and building, heating, ventilating and air conditioning equipment. 3.1.7 Aesthetics The 'IRF is located against a backdrop of coastal bluffs interspersed with steeply sloping canyons. It is in the interior of the GA site and is not visible to adjacent neighbors. However, the 'IRF is visible at a 0.5 mile (0.8 km) distance from Interstate 5 to the east and Scripps Green Hospital to the west. The TRF will be visible from future science-f related development to the northeast.
, 3.2 Natural Environment 3.2.1 Topography, Geology and Seismicity Tococrachy Site topography is typical of coastal San Diego County, with bluffs and mesas interspersed with cliffs and ravines. The mesa runs in a northerly direction paralleling the coast and rising to a height of 400 ft. (122 m) above sea level between the site and the ocean. The topography of the site is characterized by steeply sloping canyons and relatively level mesa areas. The main GA site is on Torrey Pines Mesa about one mile east of the ocean at an elevation of 340 ft. (105 m) above sea level.
B-12
Geology The 'IRF has been built on materials that have been mapped as artificial fill (Ref. 5.6). Areas immediately adjacent to the artificial fill are mapped as Ardath Shale, a member of the La Jolla Group of Eocene Deposits, that is predominantly weakly fissile, olive-gray shale. Across section on the Del Mar quadrangle shows subsurface formations approximately 750 ft. (228 m) northeast of the 7RF. Based on this cross section, the Ardath shale deposit in the 7RF area is approximately 300 ft. (91 m) thick, is underlain by approximately 500 ft. (150 m) of Torrey Sandstone over approximately 250 ft. (76 m) of Del Mar Formation. Sath Soils present at the TRF have been mapped as Huerhuero loam,5 to 9 percent slopes and t eroded (Ref. 5.7). The Huerhuero series soils have developed in sandy marine sediments and consist of moderatQ well drained loams that have a clay subsoil. A representative Huerhuero profile has a surface layer that is brown and pale-brown, strongly acid and , medium acid loam about 12 inches (0.3 m) thick, an upper subsurface layer that extends to a depth of about 41 inches (1.0 m) and is brown, moderately alkaline clay and an underlying brown, mildly alkaline clay loam and sandy loam layer that extends to a depth of more than 60 inches (1.5 m). Small areas of Las Flores and Olivenhain soils and alluvium derived from metabasic and metasedimentary rocks are included in the area. Soils immediately downslope of the 'IRF have been mapped as Altamont Clay,15 to 30 percent slopes (AtF) Huerhuero loam,5 to 9 percent slopes and eroded. The Altamont series consists of well-drained clays that formed in material weathered from calcamous shale. A representative Altamont profile has a surface layer that is dark-brown and light olive-brown, moderately alkaline heavy clay loam about 8 inches (0.2 m) thick that hes over soft calcareous shale. Small areas of Linne clay loam and areas where the soil is only 10 inches (0.2 m) over shale are included in the survey area (Ref. 5.7). There may be localized areas of soil contamination. The extent of contamination will be I defined through the site characterization process. Seismicity San Diego County has been considered one of the more moderate seismic risk regions in Southern California. The historical pattern of seismic activity has generally been characterized by a broad scattering of small magnitude earthquakes, whereas the surrounding regions are characterized by a high rate of seismicity with many moderate-to-large-magnitude canhquakes. s A recent study (Ref. 5.8) estimated the probabilities of !arge earthquakes occurring in , California on the major strands of the San Andreas fault system. In addition to the principal i traces of the San Andreas fault, earthquakes occurring on the other major faults of the # l system (San Jacinto, Imperial, etc.) were also considered. The study estimated that the probability of a magnitude 7 or greater earthquake occurring in the next 30 years in Southern California (along the Southern San Andreas, Imperial, or San Jacinto faults) is 0.5 or greater. However, a quake of magnitude close to 7 on these fault lines is not expected to significantly impact the GA site because ofintervening distance. Current information (Ref. 5.9) however, indicates the Rose Canyon, Coronado Bank, San Diego Trough, La Nacion, and Elsinore fault zones are capable of generating strong ground motion in the San Diego area. Possible Richter magnitudes for earthquakes on these faults can be as high as 7.0, 7.5, 7.5, 6.3 and 7.5, respectively. Passing approximately 3 miles B-13
(5 km) west of the GA site, the Rose Canyon fault is the nearest active fault. Recent excavations (Ref. 5.10) showed definite evidence of Halocene (within the last 10,000 years) activity. h is clear that San Diego has experienced major earthquakes in the recent geologic past. The presence of three small, local faults was confirmed by the Woodward-Clyde Consultants field reconnaissance of the site (Ref. 5.11). An unnamed fault in the northem portion of the site trends east to west through proposed lots 25, 26, 31, and 32. The Salk fault is mapped in the southem portion of the site and also trends east to west. A northerly trending fault is located in the southeastem ama and crosses the Genesee Avenue canyon. All of these faults are mapped as being overlain by early Pleistocene formations which have not been displaced. Therefore, the faults on-site are not considered active. 1 3.2.2 Climate and Air Quality
, Climatolo_cv The Torrey Pines Mesa and Sorrento Valley, as with most of San Diego County's coastal areas, has a semi-arid Mediterranean climate characterized by hot, dry summers and mild, wet winters. The mean annual temperature in the project vicinity is 61 F (33.3 C), with summer high temperatures in the low-90s (50 C) and winter lows in the mid-30s (16 C)
(Ref. 5.12). The dondnating meteorologic feature affecting the region is the Pacific High Pressure Zone, a semipermanent high pressure cell located over the Pacific Ocean. This high pressure cell maintains clear skies for much of the year, drives the prevailing westerly to northwesterly winds, and creates two types of temperature inversions (reversals of the normal decrease of temperature with height) that act to degrade local air quality. When a buoyant polluted air rises, it cools by expansion. If the air around the parcelas is inwarm, an parcel of inversion, the parcel sinks back down toward its source and is effectively prohibited from dispersing. In summer, a marine / subsidence inversion is formed when the warm, sinking air mass in the Pacific High Pressure Zone is undercut by a shallow layer of cool marine air flowing onshore. This inversion forms over the entire coastal plain and allows for mixing below the inversion base at 1,100 - 1,500 ft. (457 m), but not any higher. During the
' winter offshore flow regime, cold air pools in low areas and air in contact with the cold ground cools while the air aloft remains warm. A nightly shallow inversion layer [at about 800 ft. (244 m)] forms between the two air masses which can trap pollutants.
In the summer, when the high pressure system is at its most nonherly extent, eastward-f traveling storm and pressure centers are blocked, resulting in little rain due to frontal activity.The migration of this system to its most southerly extent in the winter allows the transient storm and pressure centers to pass through the area, resulting in winter rains in southem California. 4 The predominant pattern is sometimes interrupted by so-called Santa Ana conditions, when high pressure over the Nevada-Utah area overcomes the prevailing westerlies, sending strong, steady, hot, dry winds east over the mountains and out to sea. Strong Santa Anas tend to blow pollutants out over the ocean, producing clear days. However, at the onset or breakdown of these conditions or if the Santa Ana is wed, air quality may be adversely affected. In these cases, emissions from the South Coast Air Basin to the north are blown out over the ocean, and low pressure over Baja California draws this pollutant laden air mass southward. As the high pressure weakens, prevailing northwesterlies reassert themselves and send this cloud of contamination ashore in the San Diego Air Basin. There is a potential for such an occurrence about 45 days of the year, but the region is adversely B-14
impacted on only about five of them. When this impa t does occur, the combination of transponed and locally produced contaminants produces the worst air quality measurements i recorded in the San Diego basin. ' Local Winds and Disnersion Data The prevailing day time wind direction is westerly, although easterly winds are almost as ' common during the winter months. During the day, the westerly winds developing from the Pacific high-pressure system are reinforced by the sea-land breeze caused by the Pacific Ocean resulting m stronger average wind velocities [6 to 9 mph (10 to 15 km/h)] than from the easterly land breeze [1 to 7 mph (1.6 to 11.6 km/h)]. The land breezes are most l common during stable conditions and dominate the flow toward the ocean during the night ) and early morning hours. The airflow in either direction is channeled effectively by r I topographical features of the area. Strong winds are infrequent; the strongest recorded was 51 mph (82 km/h) from the southeast in 1944. Data from an on-site meteorological system were used to provide atmospheric stability and
- wind frequency results. The on-site annual wind data are consistent with the wind rose data l from the Miramar Naval Air Station. i l
Pmcipitation ) The average annual rainfall for the city of San Diego is 10.4 in. (26.4 cm), but relctively l large variations in monthly and seasonal totals occur. The average monthly precipitation l from 1940 through 1970 ranges from 2.15 in. (5.5 cm) in February to 0.01 in. (0.03 cm) in July. Approximately 75% of the annual precipitation occurs from November through March. The maximum annual precipitation during the last 60 years was 24.9 in. (63.3 cm) occurring in 1941. j i Air Ouality 1 Under state regulations, the study area is within the San Diego Air Basin (SDAB). The l concentration of pollutants within the SDAB is measured at eight stations maintained by the County of San Diego Air Pollution Control District (APCD) and the Califomia Air Resources Board (ARB). Air quality at a panicular location is a function of the type and amount of pollutants being emitted into the air locally and throughout the basin and the dispersal rates of pollutants within the region. The air quality monitoring station nearest the project area is located in a school ground at Ninth and Stratford Coun in the City of Del Mar. This is four miles (6.4 km) north of the site. Air quality measurements are expressed as the number of days on which air pollutant levels exceed state and federal clean air s standards. Under federal regulations, the GA facility is located in the southwestem portion of the San Diego Interstate Air Quality Control Region. The Environmental Protection Agency (EPA) ? has designated this region as an " attainment area" for sulfur dioxide and nitric oxides, indicating that the concentrations of these pollutants are below the federal air quality standards. The n gion was classified as a "nonattainment area" with respect to carbon j monoxide, ozone, and small suspended paniculates (PM i o) some years ago, but in recent , ! years only ozone federal standards have been axceeded. j In 1993 at the APCD monitoring station in Del Mar, ozone exceeded the state standard on 19 days and the federal standard on three days. This is characteristic of the entire SDAB. B-15 L l
In 1992 and 1993, the maximum 24-hour measured level of particulates less than 10 microns in' size in the SDAB was found to exceed the state standard on several days. Annual average measured PM,o levels were marginal with state standards. However, neither the 24-hour nor the annual federal standard for PM,o was exceeded. 3.2.3 Hydrology Groundwater The 'IRF is located within the Southwestern ponion of the Sole lad Basin. The Soledad Basin makes up the northwestern part of the Los Penasquitos hydrographic subunit and has not been developed for water supply purposes. No groundwater wells am present at or y immediately adjacent to the 'IRF. Ground water beneath the 'IRF is approximately 300 feet below ground surface. Test borings on the GA site ranging from approximately 6 to 30 ft. (1.8 - 9.1 m) did not encounter groundwater. There is currently no reason to suspect that a any groundwater contamination exists under the TRF. Further studies may be conducted if warranted during D&D activities. Surface Water Based on ground surface elevations and surface drainage patterns, surface run-off from the
'IRF Controlled Yard Area currently flows primarily northerly, across paved and unpaved surfaces in the service yard.
The 'IRF is located within the Los Penasguitos Creek drainage basin. Drainage runs through the Soledad Valley into Los Penasquitos Creek, which flows to the northwest and empties into the Pacific Ocean. Detention basins and silt collection structures have been
, constructed for the development of the Tormy Pines Science Park that surrounds and includes the GA site to ensure that adverse downstream impacts will not occur from stormwater run-off.
Surface water downstream from the site cannot be used domestically because of its intermittent flow and dirty condition during periods following rainstorms or heavy run-offs. No freshwater recreation areas exist within the local vicinity. Agriculture is not prevalent because soils are not well suited for agriculture, pmcipitation is limited, and groundwater quality (primarily in Penasquitos Valley) is considered marginal or inferior for irrigation. Water use in the vicinity of the site is limited by the ephemeral nature of many streams and the high suspended solids content of flow during the winter. f Floods do not represent a danger to the site as it is situated approximately 340 ft. (103 m) above the valley floor on a mesa. Also, drainage downstream from the site to the Pacific
.~ Ocean is unrestricted. The 'IRF is not located within a 100-Year Flood Zone.
9 Wastewater collection services are supplied to the GA site by the San Diego Department of Public Utilities. Wastewater from the site is discharged _through the City's sewer system to the Point Loma treatment plant. Any wastewater released to the city tmatment system must meet the requirements of the San Diego Industrial Waste Discharge Permit. 3.2.4 Biology Vegetation The GA site is professionally landscaped. The open space surrounding the 'IRF and the GA site is a combination of disturbed / developed lands, several eucalyptus groves and three B 16
l i distinct types of native or naturalized plant communities; coastal mixed chaparral, coastal sage scrub, and southern California grassland. No federally-listed endangered plant species are known to exist on or near the GA site (Ref. 5.13). The most significant natural areas in the vicinity of the site are Torrey Pines Park, Torrey Pines State Reserve, and Los Penasquitos Lagoon and associated marsh. These areas are located west and northwest of the site along the coast (Figure B-2). In addition to providing n:latively undisturbed refuge-like habitats, the park and reserve contain a rare species of pine tree, the torrey pine (Pinus torreyana). This species is endemic to California, known to occur only in San Diego County and on Santa Rosa Island. Recional Wetlands
?
Stormwater run-off from the TRF and the GA site flows into the Los Penasquitos Lagoon. The Los Penasquitos Lagoon and associated marsh are designated by the California Department of Fish and Game as a wetland area. The saltwater marsh and lagoon support a diverse fish fauna and a mussel fauna of about 20 species. The Pacific little-neck cochral ' and common little-neck clam are the most common mussel species. A total of approximately 30 species of salt-marsh plants occurs in the Los Penasquitos Lagoon. The predominant vegetation in the marsh and lagoon is pickleweed (Solicormia). Soliconnia subtenninalis occurs in the drier areas; Soliconnia virginica, in the lower-lying areas. Pickleweed filters out most of the suspended material brought in by upstream drainage. Wildlife A 1994 survey of the area adjacent to the 'IRF conducted by Natural Resource Consultants identified several mammal, birds and reptile species, with the majority of these occurring in the brushland habitats (coastal sage scrub and coastal mixed chaparral). Raptors utilize the grassland and to a lesser extent the brushland habitats on the site for foraging. Raptors are protected in Califomia and are considered sensitive due to the general trend of declining populations in many species and their importance in the ecological structure of biological communities. Two species observed in the brushland habitats around the site, black-tailed gnatcatcher (Polioptila melanuria californica) and the orange-throated whip tail (Cnemidophorus hyperythrus heldingi) appear to be experiencing declines in their populations in coastal San Diego County. The black-tailed gnatcatcher is a species of special concem and is listed by the California Department of Fish and Wildlife Service as endangered. The Torrey Pines Park, Torrey Pines State Reserve, and Los Penasquitos Lagoon and associated marsh area provides habitat for several species of shorebirds and waterfowl, as g well as two federally listed endangered species of birds, the light-footed clapper rail (Rallus longirostris levipes) and the Califomia least tern (Sterma albifrons ' browni). These species have been declining because of human disturbance and water *^ pollution that destroyed nesting and feeding habitats. The Belding's Savannah sparrow # (Passerculus sandwichensis beldingi), listed by the state as endangered, is also associated with the pickleweed habitat of the lagoon. It, too, has been declining because of human developments affecting its habitat. None of these unique wildlife species have ever been observed on the site. During the biological survey conducted of.the adjacent area (Natural Resource Consultants, May 10,1994), a total of three bird species were observed on the site. These include the house fm' ch (Carpodacus mexicanus), common raven (Corvus corax), and mourning dove (Zenaida macroura). A single fence lizard (Sceloporus occidentalis) was also B 17
& ._a.e_ . - _.&_.LJ. .- s _ _ .-..a. - a,-.a. _._wa 4 .- ,
4 a observed. There are no wildlife species recognized as twe or endangered by any resource protection agencies known to habitat within the TRF boundary. 4 3.2.5 Socioeconomics and Environmental Justice The socioeconomic environment of the GA facility consists of a well-estrblished, diverse, middle-income community consisting of research institutions, a medium-sized univenity, light industry, tourism, and residences. The setting is attractive physically with the nea-by California coastline, the Torrey Pines Paric, and picturesque La Jolla. The road system is adequate with both interstate highways and secondary roads. GA operations do not constitute a large percentage of the area's economy. 5
- 4. POTENTIAL ENVIRONMENTAL CONSEQUENCES OF PROPOSED ACTION AND ALTERNATIVES l This sectinn discusses the potential direct and cumulative effects of the proposed action on 2 ___.. _ human health and the environment.
4.1 Human Health Effects { Types of exposures that could lead to human health effects considered in this report are worker and off-tite exposures to hazardous chemicals or radioactive materials during
- routine activities or potential accidents on site, or during a transponation accident off-site (involving hazardous or radioactive waste removal). This section identifies and discusses
- potential hazards that may affect workers on-site or people off-site during normal or routine 4
TRF D&D activities. Impacts of the hazards mlative to human health and safety are summarized in Section 4.1.2. J 4.1.1 Hazard identification During the initial site characterization and the final site survey, site workers would be taking readings and measurements of any contamination using direct reading instruments and sampling techniques. Hazards during this work are mostly those involving extemal radiation, inhalation of hazardous or radioactive materials, or dermal contact with these materials. For the D&D activities, the key hazards would involve external radiation, inhalation of hazardous or radioactive materials, or dermal contact with those materials during decontamination, dismantling, packaging and disposal of reactor and ancillary equipment, f the TRF structure, and contaminated soil (if necessary). Generally, the D&D steps described in Section 2 of the Decommissioning Plan could involve the hazards as itemized below: Hazards-Hazards include:
- Extemal radiation for workers working around radioactively contaminated equipment and materials.
. Dermal contact with both radioactive and hazardous materials.
- Inhalation of any hazardous or radioactive materials.
B-18
l i e Possible confined spaces in tents, bags or small rooms with associated oxygen content f~ j and asphyxiant concerns.
- Heavy equipment movement dangers.
l
- . Epic
- No flammables or explosive materials are expected to be present.
' Controls-For workers, project procedures and conformance with GA licenses and regulatory requirements including but not limited to:
l e Radiological Work Permits, Work Authorizations, and Hazardous Work Authorization .' procedures, as required;
- 29 CFR 1910.120 requirements for PPE, air monitoring, work zone controls, medical I
. surveillance and bio-assay program, personnel training, emergency response, and health and safety plan; 1- A e personal dosimetry per 10 CFR 20; i
- .- confined space entry procedures per 29 CFR 1910.146; i
* ' HEPA filter removal of contaminants; j.
e dust filter removal of contaminants. b - 4.1.2 Potential Exposures i The collective dose equivalent estimate to workers for the entire D&D project is 27 person- ] rem. The decommissioning tasks will take approximately 2 years. Total person hours j involving radiological exposure is estimated to be 6,000 hours. ! Potential exposures of the public, including the most exposed person, from radioactive L ' efiluents released during the decommissioning activities and radioactive waste shipments are estimated to be: e ' external dose <10 mrem /yr
. . intemal dose < 1 mrem /yr The anticipated potential exposures to the public after license termination is none. The site will have been released to unrestricted use. .
'4.1.3 _ Transportation %
The primary project impacts to the environment due to tranaportation could occur when - shipments of waste travel from the site; Transportation would be cerheted in accordance p with applicable dor, EPA, and NRC regulations. During such transport, hazardous and radioactive materials would be effectively packaged to prevent significant radiation external to the truck. Thus, the primary impacts are accident risk and emissions / noise from the trucks themselves. The truck route into or from the GA property coming from or gong to San Diego is along Roselle Street west from the Interstate 5 freeway, then along Dum:ill to the gated GA entrance. This entire route from Interstate 5 to the GA gate covers a distance of about 1/2-mile. B-19
Truck shipments of concem consist of hazardous waste and radioactive waste leaving the site. During D&D activities, short-term transportation effects would include employee trips, , which occur under existing conditions, a small number of contractor trips, and less than 12 l truck trips for hazardous and radiological waste transfer. Traffic, circulation and parking l effects are expected to be minor due to the small increase in trips and the short duration of I this action and would not significantly impact the surrounding roadways. 4.2 Waste Disposal ] l 4J.1 Hazardous Waste Small amounts of solid and liquid hazardous waste from D&D activities would be l 9 accumulated in satellite accumulation areas. After accumulation for up to 90 days, the waste would be transferred by a licensed contractor to authorized off-site commercial treatment and disposal facilities or recyclers. l A 4.2.2 Low-Level Radioactive and Mixed Waste Low-level radioactive waste, including any contaminated soil, would be packaged in accordance with the Nevada Test Site Waste Acceptance Criteria. Liquid waste is filtemd or solidified and solid waste is compacted, whenever possible, in accordance with the 1 appropriate regulations. The waste would be shipped to DOE's Nevada's Test Site for disposal. The small amount of mixed waste which may be generated would be temporarily stored at the GA Nuclear Waste Processing Facility for treatment and disposal arrangements. 4.2.3 Non-Hazardous Solid Waste D&D activities will generate uncontaminated construction debris which would be sent to a local sanitary landfill. 4.3 Noise During D&D activities, noise will be generated by equipment such as jackhammers, scabblers and concrete saws. Backhoes and other heavy equipment could also be used for partial dismantling activities. On-site workers will be outfitted with ear protection devices. The closest off-site business g is Agouron Pharmaceuticals,Inc. which is approximately 0.25 miles away. Noise from D&D activities would not impact employees or off-site businesses. 4.4 Seismicity D&D activities would involve the removal of surface contamination and dismantlement activities. Dismantlement plans and specifications would be reviewed by a structural engineer to assure that activities would not render the building structurally unsafe, should an earthquake occur. D&D activities would not increase the risk to 'IRF workers during a seismic event. 4.5 Air Quality Several D&D< elated activities could minimally impact air quality due to both mobile and stationary source emissions. A small amount of mobile source emissions such as carbon B 20
monoxide and nitrogen oxides could be released from contractors' trucks and cars. l However, the San Diego Air Pollution District does not set thresholds for determination of ' significant emissions from mobile source emissions. Due to the temporary nature of the ; truck trips and the small number, mobile source emissions would be low. l Stationary source emissions could be released during decontamination, building i dismantlement and solid remediation but am expected to be negligible. Any releases from j decontamination would occur within Building 21. Hazardous materials would be located , inside the building. Standard asbestos abatement procedures, under the oversight of the ! San Diego County Air Pollution Control District, will be used to remove any asbestos. j During building demolition, particulates could be emitted in the form of fugiuve dust. i Contaminated dust would be controlled by tarping, periodic watering and possibly tenting. j t ! Fugitive dust could be released during removal and packaging of any contaminated soil. I However, dust control measures would be implemented, such as soil compacting, periodic ! watering and tarping. f Site workers would be protected during decontamination, demolition and soil excavation ~ l
. activities through air monitoring and the use of PPE and respirators when required.
The proposed action is only a temporary potential source of air emissions. Negligible ! amounts of mobile source and stationary source, demolition and soil remediation emissions l would be produced and would not affect regional attainment standards. [ 4.6 Regulatory issues .. Table B-2 discusses the applicability of various state and federal regulations 'for the i proposed action, t 4.7 Areas Not Affected ' The proposed action would not affect the following areas: l Popuhtion and I and Use-The proposed action would inemase the compatibility of GA l with other science research activities on-going within the GA site. Future use of the l Building 21 site could result in the addition of employees or tenants at GA. j Cultural Resources-There are no cultural resources on the GA site. '
'b Aesthetics-The proposed action would only be visible from Interstate 5, located i approximately 0.5 mile (0.8 km) to the east and Scripps Green Hospital, located 0.5 mile (0.8 km) to the west. The 'IRF is not currently visible to adjacent neighbors. Temporary
_4 D&D activities will be compati.>le with continuing industrial development of the F surrounding areas. The remaining site would be used for other industrial-related purposes. Biology-There are no known sensitive or endangered species on the 'IRF site. l Hydrologv-The site elevation is approximately 340 feet above mean sea level. It is not in a wetland, nor is it in a 100-year flood plain. i 1 B 21
4 Tab 8e B 2-Applicability of Environmental Statutes and Regulations Statut@e7ation Evaluation Applicabili ty Nabonal Environmental Policy Act (NEPA) The evaluabon for potenbal environmental impacts are contained in this Yes 3 document Endangered Species Act No critical habitats exist in the affected area, and no adverse impacts to No threatened or endangered species are expected to result from the proposed acbon. Floodplain / Wetlands Reculabons The proposed acbon is not located within a wetland or in a floodplain No i Fish and Wildlife Goordinabon Act The proposed action does not modify or impact fish and wildlife in any way or No l modify any bodies of water rnore than 10 acres in surface area. Fs rmland Protectson Policy Act The proposed acton does not affect pnmo or unique farmiands No Nitional Histonc Preservation Act There are no histoncal sites or areas in tha location of the proposed acbon. No AU+n.an Indian Religious Freedom Act The proposed action does not interfers with the nght of Nahve Amencans to No , exercise their tradibonal freedom 4 Wild and Scenic Riveld Act The proposed action does not involve waterways designated as wild and scenic No nyers. I Resource and Gonservabon Recovery Act The proposed acton may include the generabon, packaging and transportabon Yes , (RCRA) of mixed waste J Gomprehensive Environmental Response, Any required release reportng would be performed in compliance with Yes i Compensation and Uability Act (CERCLA) CERCLA requirements. 4 Federal insecticide, Fungicide and RodenDcide The proposed action is not involved in the distnbution, use or disposal of any No 1 Act (FIFRA) insecticides. fungicides or rodenticides. I MbWndes an did Act (T5GA) Asbestos may be encountered dunng D&D operabons which would be property Yes packaged and disposed of in acrordance with TSCA. 1 Gl2an Air Act (GAA) Asbestos may be encountered dunng the project which will be contained in No enclosed spaces, property packaged and disrmM of. GLan Water and Safe Dnnking Water Act The proposed action is not expected to affect surface water bodees or water No . supplies. Air emissions would be below waming levels. I l No;se Gontrol / Ot Noise levels that could adversely affect workers and staff will be mitigated by No I providing ear protection for workers and relocabon of staff to areas away from i the acbvibes. The public is not expected to be impacted from the noise. ' Hazardous Matenais Transportabon Act The proposed acbon will require shipment of radioactive matenals and mixed Yes ! (HMTA) wastes. All waste will be packaged and shipped in appropnate containers and disposed of at heensed facilities National Emissions Standards for Hazardous The EPA has stated that NE5 HAP 5 are applicable to NRG licensed facilibes. Yes Air Pollutants (NESHAPS) Comphance with smission standard would be demonstrated. l Atoriuc Energy Act License required. Compliance with environmental and worker protection Yes standard j G Infomia Environmental Quahty Act 6.rQA) Proposed acbon does not tngger discretionary review by a state agency. No Califorrua Health and Safety Gode, Div. 20, Proposed acbon must comply with worker safety regulations. Yes i Chapter 7.6, Articles 13.14 j G..lifomia Integrated Waste Management Act Transportation of low-level radioactive waste would require Yes nonficatiorVconsultation and manifest. i Galiforrua Gode of Regulations Titie 17 Div.1, License required. Compliance with environmental, worker, and public Yes Chaptor 5. Subchapter 4. Radiaton protecton standard. r 4.8 Cumulative Effects No cumulative effects are expected from the proposed action, as discussed below: lluman Health-The radiological exposures of the public due to expected D&D activities at 4 the TRF are expected to be < 10 mrem /yr(extemal) and < mrem /yr (internal). This amount of inemmental exposure would be an insignificant addition to the normal background
, exposure level.
4 Traffic-The temporary contractor and waste transport trips would contribute an insignificant amount to the average number of daily trips designed for Genesee Avenue and John Jay Hopkins Drive. Waste G neration-The proposed action could generate approximately 4,000 cubic feet of low-level radioactive waste from D&D activities. The Nevada Test Site is designated for the disposal of this waste and has sufficient capacity to receive the waste. Cultural Resources-No cultural resources would be impacted. B-22
Population and Land Use--Only temporary employment for a few contractors would be provided by the proposed action. No increase in population would occur. Land use would not change. l 1 Noise-D&D activities would occur in an industrial area and would largely occur within Building 21. They would not-contribute significantly to off-site background noise levels ; due to the relative isolation of the site. Aesthetics-D&D activities would not be visible to adjacent neighbors. D&D activities would only be visible from Interstate 5 and Scripps Green Hospital, both located approximately 0.5 miles (0.8 km) away. After being released to unrestricted use, the TRF tite would be used in a manner consistent with the existing GA site. Geology. Soils. Seismicity and Hydrology-All D&D activities would be localized and stctmwater runoff would be contained and tested. - Eggpnal Air Oualitv-The San Diego Air Basin is a non-attainment area for carbon f monaxide, ozone, and small suspended particulates (PM io ). The proposed action is ; tcaiporary. Asmall number of vehicle trips would be generated and would convibute only negligible amounts of these pollutants to the basin. Hydrology-No changes to any land forms would occur and no radionuclides or
- hazardous materials would be released to storm water run-off.
Biological Resources-No resources have been identified on the TRF site, nor would D&D activities effect off-site resources. l 4.9 Alternatives to Proposed Action I Alternative 1 to Proposed Action-Safe Storage (SAFSTOR) This alternative poses essentially the same potential risks and environmental impacts as the proposed project, but potentially for a much greater period of time. This altemative would necessitate continued surveillance and maintenance of the TRF over a substantial time period. During this period, the risk of envimamental contamination would continue to exist. Moreover, development of the land around the GA site over the next few years may significantly increase the local employment population density and incmase potential for public exposure. This alternative is not environmentally preferable.- Alternative 2 to Pronosed Action-Entombment (ENTOMB) g This alternative would necessitate continued surveillance and maintenance of the 'IRF over ~' a substantial time period. During this period, the risk of environmental contamination - would continue to exist. Moreover, development of the land around the GA site' over the k ;' next few years may significantly increase the local employment population density and increase potential for public exposure. This alternative is not environmentally preferable.
- 5. REFERENCES 5.1 National Council on Radiation Protection and Measurements (NCRP). loni7ing Radiation Exoosure of the Population of the United States. Report No. 93.1987. ;
5.2 Walker, F. W., Perrington, J. R., and Felner, F. Nuclides and Isotooes. General Electric Company.14th Edition.1989. : B-23
1 4 5.3 U.S. EPA. " Risk Assessment Guidance for Superfund, Volume 1 Human Health ! Evaluation Manual (Part A)." Office of Emergency and Remedial Response, U.S. EPA, 1 Washington D.C.1989. i 5.4 U.S. EPA. " Risk Assessment Methodology Draft Environmental Impact Statement for Proposed NESHAPS for Radionuclides." Vol.1. U.S. Environmentrl Protection Agency, , Office of Radiation Programs. Washington D.C. 5.5 Source Point. "1990 Census Total Population and Housing Units." San Diego Association ] of Governments. April 1991. 5.6 Kennedy. " Geology of the San Diego Metropolitan Area, California." Bulletin 200A. 1975. 7 4 5.7 USDA Soils Conservation Service. " Soil Survey San Diego Area, California." 1973. I 4 5.8 Algermissen, S. T. et al. "Probabilistic Earthquake Acceleration and Velocity for the United States and Puerto Rico." U. S. Geological Survey Map MF-2120. 5.9 Berger, V. And D. L. Schug, "Probabilistic Evaluation of Seismic Hazard in the San Diego-Tijuana Metropolitan Region," Environmental Perils - San Diego Region, P. L. ! Abbott and W. J. Elliott, Editors, San Diego Association of Geologists,1991. 4 5.10 Lindvall, S. C., T. K. Rockwell, and C. E. Lindvall. "The Seismic Hazard of San Diego, i Revised: New evidence for magnitudes 6+ Halocene earthquakes on the Rose Canyon fault zone." Proceedings,4th U. S. Conference of Earthquake Engineering. May 1990. 5.11 Woodward-Clyde Consultants. " Preliminary Geotechnical Reconnaissance of the Torrey Pines Science Park." January,1988. 5.12 San Diego Air Pollution Control District.1993 Annual Report.
- 5.13 City of San Diego Planning Department, Environmental Quality Division. Envirorunental Impact Report: Torrev Pines Science Center Planning Industrial Development. EDQ No.
- 86-0884. 1986.
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