ML20198R254
| ML20198R254 | |
| Person / Time | |
|---|---|
| Site: | University of Iowa |
| Issue date: | 01/06/1999 |
| From: | DUKE ENGINEERING & SERVICES |
| To: | |
| Shared Package | |
| ML20198R218 | List: |
| References | |
| 00752.F03.A01, 00752.F03.A01-DRF-R0, 752.F3.A1, 752.F3.A1-DRF-R, NUDOCS 9901080135 | |
| Download: ML20198R254 (59) | |
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B Duke Engineering EdGServiscs. AD* % W IOWA STATE UNIVERSITY UTR-10 REACTOR DECOMMISSIONING PLAN January 1999 Prepared by Duke Engineering & Services Bolton, MA f I Docunent No. 00752.F03.A01 Rev.O
1 l iE \ I E Duke E.neineering 1 eraservices. , I ADerBuyCaymy TABLE OF CONTENTS l.0
SUMMARY
OF PLAN. . . . . . . . . . . . . . . . . . . . . . . . ... . . . . .. .. 1 1.1 Introduction ... .. .. . ... . . . . . . . . . . . . .. . . . . . . ....... 1 1.2 Background Information on Reactor Facility.. . .. .. .. .. . . . . .1 1.3 Reactor Decommissioning Overview . . . . . . . . . . . . . l.4 Estimated Cost., ... . . . ... . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 2 1.5 Availability of Funds. . . .. . . .. . .. . .. . . . . . . . . . . . . . . .......6 1.6 Project Quality Assurance.. .. .... .. . ... . . . . . . . . . . . . ... . .. .6 2.0 DECOMMISSIONING ACTIVITIES..... . . . . ... . . . .. . . . . . . . . . . . . .. 6 l 2.1 Decommissioning Alternative... . . . . . . . . . . . . . . . ...............6 2.2 Facility Radiological and Non-Radiological Status .. ... . . . . . . . . . . ... .7 2.2.1 Facility Operating History.. .... .. . . . . .. . . . . ........7 I 2.2.2 Current Radiological Status of the Facility . . . . .. 2.2.2.1 Average Area Exposure Rates.. . . . . .
. . . .... . I 1
.12 2.2.2.2 Surface Contamination. .. . . ... . . . . . . . . .. 12 I 2.2.2.3 Reactor Core Activation. .
2.2.3 Radiological Release Criteria.. . .
.I2
.12 2.2.4 Hazardous Materials . . . . .. ... .. . . . . . .. . . . .17 I 2.3 Decommissioning Tasks. . .. . . . . . . . .
2.3.1 Activities and Tasks.. . . .
..............18
. . . . . . . . . . . . . . 18 2.3.1.1 Mobilization and Preparation ..... . . . . . . . . . . . . . . . . . .18 I 2.3.1.2 2.3.1.3 Radiological Containment and Dust Control Plan . ..
Demolition and Construction.. ... . . . . . .. . . . 19
. .... I 8 2.3.1.4 Removal of the Stairs . . . . .. .. ...... . . . . . .... . . 19 2.3.1.5 Plug Removal, Prep and Packaging. . . . . . . . . . . . . . . . . ... . 19 2.3.1.6 Water Tank Demolition. .. . . . . ... ..........................20 2.3.1.7 Graphite & Core Component Removal. . .. ... . . . . . . . . . 20 l 2.3.1.8 2.3.1.9 Bio-Shield Removal . . .... ... .. . .. ..... . . . . . . . . . . . . . . .
Wire Saw Cu tting .. . . . .. . . . . . . . . . . . . . . .. . . . . .. . . . . . . . . . . . . 21
...21 2.3.1.10 Block Segregation . . .. . ... . . . . . . . . . . . . . . . . 21 2.3.1. I 1 Pit Excavation .. .... . .. . .. . . . . . . . . . .. ..............22 2.3.1.12 Tank and Sump Equipment Removal ....... .. ... ... .....22 2.3.1.13 Miscellaneous Equipment Removal.. . .. ... ............22 2.3.1.14 Temporary Staging Areas. . . . . . . . . . . . . . . . .. ..22 2.3.1.15 Waste Disposal . .. ..... ... ... .. . . . . . . . . . . . . . .22 2.3.2 Schedules . . . .. . . . .. .... . . . . . . . . . . . .. .. . . . . . . . 23 2.4 Decommissioning Organization and Responsibilities. . . . . . . . . . . . . . . . 23 2.4.1 Project Management Structure... . ....... . . . . . . . . . . . . . . . . ... 23 2.4.2 ISU Project Manager.......
I 2.4.3 ISU EH&S Director.. 2.4.4 ISU Radiation Safety Officer. .. .
...........28
. . . 23
. . .. 28 2.4.5 ISU Reactor Use Committee. . . . . .. ...........28 2.4.6 DE&S Program Director. ... . . . . . . . . . . . . . . . .. .. . . . 28 2.4.7 DE&S Project Manager . . ... . . . . . . . . . . . . . . . . . 29 2.4.8 DE&S Health and Safety Officer. .. . . . . . . . .. . . 29
'I 2.4.9 Characterization / Final Survey (Phase I and III) Manager .. . .. . .29 Ducunent No. 00752R)3.A01 l' Rev. O
k 1 h Duke Engineering F#A Services. ac e r- o c, , TABLE OF CONTENTS (Continued) 2.4.10 Phase II D&D Superintendent.. . . . . . . . . . . . . . . . . . . . . . . . . . 29 2.4.11 DE&S Quality Assurance Project Plan Coordinator . ..............30 2.5 Training Program.. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ...........30 2.6 Contractor Assistance . . . .. . .... .. . . .. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . .. . 30 2.7 Documents and Guides.. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . ..............30 2.8 Quality Assurance.... .. .. .. . . . . . . . . . . . . . . . . . . . .................................31 3.0 PROTECTION OFTHE HEALTH AND SAFETY OF RADIATION WORKERS AND TH E PUB LIC . .. .. . .. . . . .. .. . . . . . . . . . . ..... . . . . . . . . . . . . . . . . . . ...............32 3.1 Radiation Protewon.. . . . . . . . . . . . . . . . . .. .. . . . . . . . . . . . . . . . . .... . 3 2 3.1.1 Ensuring ALARA ..... . . . . . . . . . . . . . . . . . ........................32 3.1.2 Health Physics Program... . . . . . . .. . . . . . . . . . . . .. . ... . 32 3.1.3 Dose Estimates... ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ................33 3.2 Radioactive Waste Management... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .......33 3.2.1 Fuel Removal . . ... ...... .. . . . . . . . . . . . . . . . . . . . . . . ..............33 3.2.2 Radioactive Waste Segregation and Processing... . . . . . . . . . ... . . . . . . 3 3 3.2.2.1 Demolition Rubble Disposal.. . ... ... ..... .. ..... ........ ... ..... 34 3.2.2.2 Gamma Cu rtai n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. .. 34 3.2.3 Radioactive Waste Transportation and Disposal. ....... ... . . .. .. ..... .... 34 3.2.4 General Industrial Safety Program. . . .... ............... ...............35 3.2.4.1 Management Policy............ . . . . . . . . . . . . . . . . .............35 3.2.4.2 Health & Safety Organization . .. .... .. ... . . ... .. . .. ........35 3.2.4.3 ISU Environmental Health and Safety Program ..................... 36 3.2.4.4 Safety Training & Meetings ..... .. . . ..... ......................37 3.2.4.5 Respiratory Protection Progrant..... .... . ... .. . .... . .... . .... . 37 3.3 Radiological Accident Analysis.... ............. ................... .. .......................38 4.0 PROPOSED FINAL RADIATION SURVEY PLAN .. ... ...... . .. ........ ... . . ..... .. . . 38 4.1 Final Status Survey Plan Under MARSSIM Guidelines ... .. ....... .. .. . .......38 4.2 Data Quality Objectives. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ..........................38 4.3 Final Area Classification and Survey Unit Designation.. ... . . . . . . . . . . . . . . . . .. ... 39 4.4 Instrumentation. ... . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. 42 4.5 Background Determination . . . ..... . ..... ... ........ .. .... ...... . ................42 4.6 Demonstrating Compliance with Release Criteria .. .... .. . ..................42 4.7 Documentation of Final Survey Results.... . . ...... .. .... .. ...... ................43 5.0 TECHNICAL SPECIFICATIONS .. . . ... . . . . ...... ... . .............................43 6.0 PHYSICAL SECURITY PLAN. ............. .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... 43 6.1 Physical Site Security Dming Decommissioning... . .... .... . . ................43 6.2 M aterial S afeguard s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43 s 7.0 EM ER G ENC Y PLAN . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .............43 8.0 ENVIRONMENTAL REPORT.. ... ...... ...... .. .... . . . . . . ........................44 9.0 CHANGES TO DECOMMISSIONING PLAN .. ........ . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 10.0 R E FE R EN CES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 Document No. 00752.R)3.A01 ii Rev.O
t D bb EnCldng ruaservices. A Dubr Emsgy t.-r== LIST OF FIGURES FIGURE 1.1 Location of IS U . . . . . . . . . . . . . . . .. . . ... .. . . . . . .. . . . . . . . . . . . . . . . . . .. ....3 l FIGURE 1.2 Location of NEL on ISU Campus. ... .... . .. ..... . . . . . . . . . . . ............4 FIGURE 1.3 Facility Layout.. . . . . . . . . . . . . . . . . . . ........ . . . . . . . . . . . . . . . . . . . . . . . .........5 FIGURE 2.1 Average Exposure Rates - Basement level (NEL 01) .... .. . ..... .......... . ........I3 i FIGURE 2.2 Average Exposure Rates - First Floor (NEL 02). . ...... .. ...... . . . . . . . . . . ... 14 FIGURE 2.3 Average Exposure Rates - Second Floor (NEL 03).. .. ... . . . . . . . . . . . . . . . . . . ..15 I FIGURE 2.4 Reactor Activation Profile - Plan View.. . ....... . . . . . . . . . . . . . . . . . . ............I6 FIGURE 2.5 Project Organization .... . . . . . . . . . . . . . . .......... ...............................24 I FIGURE 2.6 ISU Management Structure..... ... . . .. .. ... . . . . . . . . . . . . . . . . . . . . . . . . . ... .. . 27 LIST OF TABLES TABLE 2.1 Design Data for the UTR-10 Reactor (LEU Fuel)......... . . . . . . . . . . . . . . . . . .......9 I TABLE 2.2 ISU Site-Specific DCGLs. . . . . . . . . . . ....................................17 TABLE 2.3 Schedule - Key Milestones..... .. . .. .... ..... . .. ... .. . . ... ... ... .. . . . . . . . . 25 I TABLE 3.I NRC Occupational Dose Limits .. ..........................................32 TABLE 3.2 Estimated Collective Doses for Decommissioning Activities... .. ...... ........ .. .. 33 TABLE 4.1 Final Status Survey Unit Designation and Classification.. ............ .............41 TABLE 4.2 Nuclear Engineering Laboratory Final Survey Instrumentation.... ........... .... . 42 i APPENDICES I A PP EN D IX A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 6 APPENDIXB.............................................................................................47 [ [ [ I Docunent No. 00752.F03.A01 iii Rev.O e
A Duke Engineering EnFO Services. ADdeausgyCmysy 1.0
SUMMARY
OF PLAN 1.1 Introduction This Decommissioning Plan is submitted to support Iowa State University's (ISU) Request for Decommissioning Order from the Nuclear Regulatory Commission (NRC) for its University Training Reactor (UTR-10) and associated systems. This document describes the decommissioning methods and controls to be used during the dismantling of the reactor systems, the handling, removal, and disposal of radioactive materials, and the radiological health and safety programs. It also describes the planned final radiological survey of the site which is conducted prior to the release of the site and structures by the NRC for unrestweed use. The current radiological condition of the reactor facility is described in Document No. 00752.F02.A01 "Cluracterization Report for lowa State University UTR-10 Reactor" which accompanies this submittal. The reactor is defueled and in a safe shutdown state. 1.2 Background Information on Reactor Facility The UTR-10 is housed in the Nuclear Engineering Laboratory (NEL) building located on the west edge of the main campus of ISU, in Ames, Iowa. The facility is a two-story, three-level building of brick construction built in 1934 by the U.S. Department of Agriculture. It was deeded to the University in 1946. A map indicating the location of the University with respect to Ames and major highways is shown in Figure 1.1. The location of the NEL building within the ISU campus is shown in Figure 1.2. The building floor space is divided into four levels: the basement (west side only), the ground floor (which includes the central bay), the first floor (west side only) and the second floor which surrounds the central bay. The central bay is approximately 34 feet high and has a floor area of 37 feet by 56 feet, of which a space approximately 37 feet by 38 feet is allocated to the reactor room. The reactor room houses the reactor, which is enclosed in a concrete biological shield, the process pit, the fuel storage pit, and a five-ton bridge crane. The facility layout is shown in Figure 1.3. The UTR-10 is a reactor of the Argonaut type which used uranium enriched to 235 19.75% in U in a graphite reflected, water moderated core. The reactor was installed in 1959 on the ground floor level, central bay area, of the N._L. In 1991 the reactor fuel was changed from its original high-enrichment uranium to low-enrichment uranium. The reactor was controll:d with four window-shade type Boral control rods. Heat from fission was ren.ow' from the primary coolant by a 34,000 BTU /hr shell-and-tube heat exchanger that utilized city water as a heat sink. The reactor was designed to be inherently safe. It would automatically s! ut Document No. 00752.F03.A01 Page 1 of 54 Rev.0
In Duke Engineering EUU Services. ADaBmpCam i down if there was a loss of AC power or if parameters important to safety were exceeded. Reactor specifications are described in Reference 1. The enclosure surrounding the reactor facility includes the central section of the building defined by the interior walls of offices, laboratories, and corridors. The reactor was provided with multiple experiment features including: beam ports, thermal column, shield tank, intemal reflector, rabbit tube, and mdiation cavity. 1.3 Reactor Decommissioning Overview The primary objective of this project is to secure approval from the NRC for releasing the site without radiological restrictions. The scope of work for the project is divided into three phases: Phase I
- Develop and implement a comprehensive characterization of the facility and prepare a Characterization Report.
- Develop and submit a Decommissioning Plan, which describes the methods to be employed to safely dismantle, decontaminate and release the facility to applicable criteria.
Phase II
- Implement the decomrnissioning activities in accordance with the NRC approved plan.
Phase III e Conduct a final site survey to verify that residual radioactivity at the site is below the release criteria.
- Prepare a final report.
1.4 Estimated Cost i The total cost of the project is estimated to be 1.0 million dollars. (See ! Reference 2.) I Docunent No. 00752.F03.A01 Page 2 of 54 Rev.O
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FIGURE 1.2 Location of NEL on ISU Campus Document No. 00752.IV3. A01 Page 4 of 54 Rev.0
Duke nineIneertne easervices. a n* a , c,, Nuclear Engineering Laboratory I lowa State University, Ames, Iowa g Basernent . First Floor { iwto21
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l (~h Duke Ernpineering KOUServices. An e a=w c . ,., i 1.5 Availability of Funds ISU is committed to providing funding for the decommissioning of the UTR-10 reactor. A Letter of Intent is provided under separate cover as part of this submittd. 1.6 Project Quality Assurance ISU, as the owner of the facility, retains oversight and has the overall responsibility to ensure that work is conducted by the contractor in concert with ! the relevant quality assurance guidelines from the NRC.
- The decommissioning contractor, Duke Engineering & Services Inc. (DE&S), is responsible for preparing, maintaining, and implementing quality assurance procedures throughout the life cycle of the project. DE&S uses established procedures for radiation control and quality assurance for all decommissioning 1 work that it undertakes. Survey activities during Phase I and Phase III of the project are covered by the Survey Quality Assurance Project Plan (Survey QAPP) and project-specific quality assurance requirements. Activities during Phase II are covered under the quality assurance plan of NSC Energy Services, who is a subcontractor to DE&S. An independent quality assurance function is defined in the project organization to ensure verification of the implementation of the quality assurance program elements. Details of the quality assurance program are l provided in Section 2.8. !
l 2.0 DECOMMISSIONING ACTIVITIES 2.1 Decommissioning Alternative l ISU selected the DECON alternative for decommissioning the UTR-10 reactor after evaluating three decommissioning alternatives described in the NRC guidance document NUREG-0586 (Ref. 3). The three decommissioning i l alternatives are: DECON, SAFSTOR, and ENTOMB. j
\
The non-power reactor decommissioning requirements in 10 CFR 50.82 (b)(1)(ii) l and (b)(1)(iii) state that decommissioning activities should be completed without I significant delay unless prevented by factors beyond the licensee's control. This limits the choice for non-power reactors to the DECON attemative. The use of option SAFSTOR can be acceptable for a limited period, if factors beyond a l licensee's control delay dismantlement and decontamination. Such factors can include unavailability of radioactive waste disposal facilities, DOE fuel shipment delays, continued operation of another reactor at the site, and the need to protect the health and safety of the public. ISU's choice of the DECON attemative is i ' l Documern No. 00752RD.A01 Page 6 of 54 Rev. O
Eh Duke Engineering Edu Servbcs. ADdeSugycmysey consistent with the NRC requirements mentioned above and the guidance provided in NUREG-0586. For the ISU project, DECON is the preferred alternative for the following additional reasons:
- ISU UTR-10 is a small reactor, rated at 10 kW power level.
- Volume of radioactive waste generated from decommissioning activities is estimated to be small.
Commercial radioactive disposal facilities are accessible to the project.
- Radioactivity is limited to the activated materials in the reactor room structure. Radiological surveys of the reactor environs show that radioactive contamination is limited with elevated activity levels confined to a very small 2
location (~200 cm ) on the reactor floor.
- ISU personnel familiar with the reactor are available.
- The University intends to utilize the space for other programs and facilities after the area is approved by the NRC for unrestricted release.
The DECON alternative involves the removal from the site of all spent fuel assemblies, source matedal, radioactive fission products, and radioactively contaminated material and components that exceed the release criteria. The Decommissioning Plan consists of several tasks oriented towards dismantling the reactor vessel, bio-shield, and the reactor systems and stmetures. Dismantlement and removal activities will be undertaken in a safe manner adhering to all applicable regulations and in accordance with the ISU policy of application of ALARA. In addition to the reactor facilities, the NEL building contains offices for faculty and other ISU staff. Precautions will be taken during decommissioning activities to preclude unauthorized access to controlled areas and to minimize radiation exposure. 2.2 Facility Radiological and Non-Radiological Status 2.2.1 Facility Operating History The UTR-10 is an Argonaut type nuclear reactor designed and built in 1959 by the Advanced Technology 12boratories division of the A.merican Radiator and Standard Sanitary Corporation. Initial reactor criticality was on December 31,1959. Final reactor criticality was on May 8,1998 and Docunent No. 00752M3. A01 Page 7 of 54 Rev.0
13 Duke Engineering EUUServices. m r e c,., reactor operations officially ceased on May 15, 1998. Radiological charactedzation surveys and sampling efforts began in July 1998 and finished in September 1998. The resulting Characterization Report formed the basis for the development of this Decommissioning Plan The reactor had a maximum rated thermal power level of 10 kW, was moderated and cooled by light water, and used graphite as a reflector. i Four window-shade type Boral control rods provided reactor control. i During operation, heat from fission was removed from the primary coolant by a 34,000 BTU /hr shell-and-tube heat exchanger. The reactor was designed to be inherently safe. It would automatically shut down if there l was a loss of.AC power or if reactor parameters impc.' ant to safety were , exceeded. Key design parameters of the ISU UTR-10 reactor are listed in Table 2.1. For over three decades, the UTR-10 reactor provided ISU faculty and researchers with teaching, operator training, and experimental capabilities in nuclear science. From 1960 to 1983, several modifications to the l reactor systems were made. There are described in the Safety Analysis Report (SAR). This SAR was submitted by the University in 1983 to the NRC (Docket No. 50-116, License R-59) as part of its application to renew the operating license (Ref. 4). A Safety Evaluation Report by the Commission was issued as NUREG-1016 in September 1983 (Ref. 5). From the initial operation in 1959 to June 30,1983, the reactor generated about 6,417 kW-hrs of energ5 during 6,757 hours of operation. The average bum-up was 13.7 mg U per year. Effective on September 12, 1990, with the issuance of License Amendment No. 8, and in accordance with 10 CFR 50.64, which required the operators of non-power reactors to convert these reactors to low-enriched uranium (LEU) fuel, the ISU reactor was converted from high-enriched uranium (HEU) fuel to LEU. Since the initial criticality of the LEU core in August 1991, the cumulative energy produced was 305 kW-hrs over an operational time pedod of 869.5 hours. The last Annual Operations Report, which covered the period July 1,1997, to June 30,1998, was submitted to the NRC on July 9,1998. During this period, a total of 44.9 kW-hrs of energy production and 73 hours of operation were recorded. During the lifetime of the facility, with both the HEU and LEU cores, a total of 7629.1 kW-hrs of energy was produced for a cumulative operating time of 9540.7 hours. A list of amendments to the reactor license is given in Appendix A, Table Al. A proposed amendment, No.13, for reducing the number of area monitors from five to two, was requested by the University in a letter to the NRC dated September 14,1998. Docunent No. 00752.103. A01 Page 8 of $4 Rev.O
l Etj)u Services.(~' Duke Engineering a m r-,c ,, l TABLE 2.1 Design Data for the UTR 10 Reactor (LEU Fuel) l Maximum Power Ixvel(kW) 10 I Predicted Critical Mass (g U-235) 3300 Geometry Two 5-27/32" by 19-15/16" slabs separated by 18" of graphite and ; reflected by 12" of graphite ' Moderator-Coolant Light Water Excess Reactivity (% Ak/k) 0.5 Delayed Neutron Fraction err 0.00763 Neutron Lifetime,I (ps) 165 Fuel Assemblies 12 Plates per Assembly 24 Fuel Plate Dimensions (in.) 26 x 3 x 0.05 Enrichment (%) 19.75 U-235 (g/ plate) 12.5 Water Gap (in.) 0.17 Aluminum Cladding (in.) 0.015 Fuel-Containing Matrix U3Si2 + A1 Reflector Material Graphite Dimensions (in.) 44 x 56 x 48 high Control Rods and Drives Regulating Rod 1 Shim Safety Rod 1 Safety Rods 2 Rod Worth (% Ak/k) l Regulating Rod 0.26
- Shim Safety Rod 0.45 Safety Rods 0.45 I
Docunent No. 00732.F03. A01 Page 9 of 54 Rev.0
I MN En9WRC Rh Services. ADa % % TABLE 2.1 Design data for the UTR-10 Reactor (LEU Fuel) (Continued) Rod Speed Regulating (inimin.) 27 Shim Safety and Safety Rods: Withdrawal Rate (in/ min.) 6 Scram (complete insertion) 0.5 s or less ! Control Rods Material Boral i Size Shim Safety and Safety (in.) 8 x 7 x 1/8 Regulating (in.) 2 x 2 x 1/8 i Maximum Reactivity Input Rates (% Ak/k/s) Regulating Rod 0.012 Shim Safety Rod 0.0044 Safety Rods 0.0044 Reactivity Effects Fuel Loading (% AL/k/g) +0.0043 Moderator Temperature Coefficient (% AL/k/K) -0.019 Fuel Temperature Coefficient (% Ak/k/K) -0.0013 Void Coefficient (% Ak/k/% void) -0.22 Process Water Conductivity (micromhos/cm) 1 Flow (gpm) 10 Minimum Temperature (#F) 78 Normal inlet Temperature ( F) 80 ! Normal Outlet Temperature l- at 10 kW ( F) 87 l 4 I Docunen: No. 00752.F03.A01 Tage 10 of 54 Rev.0
l FD Duke Engineering l kl.Ju Services. 1 ADeBagCm9my 2.2.2 Current Radiological Status of the Facility Routine radiological survey data was collected periodically during the operationallife of the facility. Of the more recent data, the measurements conducted by ISU in August 1998 indicated that exposure rates in the I central general areas of the NEL (all 3 elevations) and on the reactor floor (including the process pit) were not greater than background. All smears collected at this time indicated that removable contamination was less than 2 200 dpm/100 cm (beta / gamma). The following exposure rate data was collected at the reactor housing area:
- 0.5 mR/hr 1 meter above core surface i
- 2.0 mR/hr on average 4" above top of graphite ;
e 3.0 rnR/hr maximum (between two of the control rod housings) i
- 8.0 mR/hr 3 feet into core tanks (below top of graphite) 6.0 mR/hr at surface of the bottom of one of the shutdown closures.
A review of audits conducted by the Nuclear Regulatory Commission (and its predecessor, the Atomic Energy Commission) indicates that there were no instances of unusual events which caused any area to become i contaminated. Interviews with ISU personnel, and a review of ISU records, indicate there were no known instances of contaminating events, no instances of an area ever posted as a " Contaminated Area," and no instances of airborne contamination, with the exception of short-lived ! noble gases. 4 In preparation for decommissioning, a radiological survey of the ISU reactor facility was recently conducted by DE&S to assess the current radiological conditions at the site. A characterization / sampling plan was prepared and characterization activities were initiated at the fuel storage pit in July 1998. The remaining characterization survey activities for the facility were concluded in September 1998. The characterization activities were conducted in accordance with the Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM) guidance. The results of this survey indicate that the remaining residual radioactivity is limited to the activated materials in the reactor room structure. A small 2 area of slightly elevated and fixed activity (~ 200 cm ) was also detected on the concrete floor of the process pit near a water sample tap. The areas of the NEL outside of the reactor compartment, including NEL01, NELO2, and NELO3, (see Figure 1.3) were not impacted by Document No. 00752.IV3. A01 Page 11 of 54 Rev.0
-Q Duh0 - E.nginOOdng ADaarBespCuyasy reactor operations. This is supported by historical information, routine surveys performed by the ISU Environmental Hecith and Safety (EH&S)
I department and by the characterization data. 2.2.2.1 Aserage Area Exposure Rates The average area exposure rates in R/hr are shown for each level of the NEL on Figures 2.1,2.2, and 2.3. Elevated exposure I rates ranging from 27 to 54 pR/hr were observed in areas immediately adjacent to Rooms 101 and 201. However, these are not related to reactor operation because the areas in question serve as storage areas for radioactive sources and radioactive I materials for the ISU EH&S Department. Exposure rates up to 23 pR/hr were detected in the southwest stairwell and in an area above the east side of the reactor room. The remainder of the facility produced exposure rates consistent with the outdoor levels near the facility, which ranged from 12 to 28 R/hr. 2.2.2.2 Surface Contamination i Surface contamination was limited to a small area on the process pit floor and averaged 9000 dpm/cm2 . No removable surface I contamination was detected on the surface of the reactor core. Elevated measurements that were observed in the reactor core are due to activation. 2.2.2.3 Reactor Core Activation Contact exposure rate readings collected on the east Beam Port Plug and the Central Thermal Column Stringer indicate that the area of activation extends radially to approximately 40" x 34" I from the core centerline (Figure 2.4). A calculation performed using the 10 CFR Part 61 (Ref. 8) analysis results from four materials collected from the core (concrete, graphite, aluminum, I and steel) indicate an estimated total volume of radioactive waste of 1200 ft'. 2.2.3 Radiological Release Criteria L The NRC Final Rule on License Termination,10 CFR 20.1402 (Ref. 9), provides radiological criteria for release of a site for unrestricted use. ( Acceptability criterion for unre icted use is a maximum Total Effective Dose Equivalent (TEDE) of 25 mrem per year from residual radioactivity above background. Application of As Low As Reasonably Achievable [ (ALARA)is also a requirement. The Derived Concentration Guideline E Document No. 00752.R)3.A01 Page 12 of 54 Rev.0 m e-=r
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l 1 i O Duke Engineering i I ktMServices. An*m o e w=r f Limits (DCGLs) are established based on the criteria defined above. The DCGLs were determined based on the radionuclide analyses of samples { and using the USNRC Model D and D Version 1.0 (Ref.10). The resulting values are presented in Table 14 of the Characterization Report (Ref. 6) and reproduced here in Table 2.2. The results of the final survey will be used to demonstrate that the predicted dose to a member of the public from any residual activity does f not exceed the 25 mrem /yr dose limit. The final survey will be conducted in accordance with the guidelines in MARSSIM. TABLE 2.2 ISU Site-Specific DCGLs Material Total Surface contamination { DCGL (dpm/100 cm 2) Structural Surfaces 5.3E03 Reactor Graphite 5.3E03 Reactor Steel 5.3E03 Reactor Aluminum 5.3E03 Reactor Concrete 1.lE04 { (Activated Portions) Alpha 1.0E02 2.2.4 Hazardous Materials As pan of the characterization process for decommissioning, ISU conducted work to identify any potential hazardous materials at the reactor { facility. The results of this characterization identified three areas of I concern: (1) pipe insulation bearing Asbestos Containing Material (ACM); (2) lead contained in paint; and (3) PCB's contained in paint. These results are summarized hen ; further details are available in { Reference 6. The ACM is contained in the drain pipe insulation which dates back to 1959. The asbestos was aetermined to be non-friable. The pipe along with the ACM will be removed as part of the University's l ongoing asbestos abatemerit program during the D&D activities. Of all the samples analyzed from the reactor room, only one exceeded 50 ppm for PCBs. In this area of the facility, no radioactive contamination is identified, thus precluding the possibility of mixed waste. The University plans to strip the paint containing the PCBs prior to the stan of the reactor decommissioning activities. The older paint containing lead is Docunnd No. 00752.R)3.A01 Page 17 of 54 Rev.0
[7 DUh FJspineering ' Edu SERVICES., A D deSmagy Caysy determined to be only on the steel mezzanine and the old stairs to the mezzanine. These components will be disassembled and disposed in accordance with the state and federal regulations in the earlier stages of decommissioning. 23 Decommissioning Tasks 23.1 Activities and Tasks This section provides a description of the activities that will be undertaken during Phase II of the project. 23.1.1 Mobilization and Preparation i To begin the process of decommissioning and decontamination of the radioactive portions of the reactor facility, the contractor will stage the reactor room and organize the safety and access controls to facilitate safe operations. This will prepare the reactor facility for the component removal and waste packaging operations, and modify the building to meet the project specifications. All workers will become familiar with the facility and undergo any necessary training. The five-ton overhead crane will be inspected and tested by a trained and competent person to ensure safe rigging and hoisting operations. When not in use, the crane will be disabled (locked out) to prevent unauthorized use. Circuits and mechanical systems that are to be decommissioned or that would interfere with D&D operations will be identified from drawings and inspections with the facility staff and University physical plant personnel. Once these items are identified, the sequence and method for isolation and deactivation will be established and implemented. 23.1.2 Radiological Containment and Dust Control Plan The sequence of dismantlement will be to remove all reactor components first and the concrete bio-shield will be demolished last. The contamination levels inside the reactor cavities are not expected to warrant the use of contamination control containment during the removal of graphite and core components for packaging. If required, plastic and wood framing containment tents will be set up at the thermal column end and over the top of oocunen: No. 00752,F03. A01 Page 18 of 54 Rev.0
l"; Duke Er9pINGeffNg ELJU Services. ADaBupCam the core to provide an area for remaval and packaging of the graphite and core components. Air samples from the removal of ; the center of the core area will indicate the need for continued ! airbome controls. I The primary means of dust control during concrete demolition will be water which will be epplied in a low volume mist at points where dust is produced. These areas include the demolition of the cinder-block wall, the ipping of conca:te blocks and the cutting of concrete slabs. . vided dust control measure and to protect adjacent offict ..cupied areas and equipment from dust, the reactor room will be prepared to serve as the containment for fugitive dusts. Installation of plastic l barriers on the walls, at the east door, and at the south ramp, will isolate the room. A HEPA blower, operated dunng periods of dust generating operations, will prevent the spread of dast not captured by local mists. Any water on the floor will be vacuumed into a HEPA vacuum water collection drum for re-use. 23.13 Demolition and Construction Facility modifications include: removal and relocation of the double doors (south end), demolition of the cinder-block wall, and removal and sealing of the second floor door openings at the j north end. In addition, the following items will be relocated: steam heater, crane power boxes and the emergency light I conduit. Waste pile management is a critical factor, and contractor plans for movement of waste materials (radioactive and non-radioactive) will take into account the limited available space. 23.1.4 Removal of the Stalrs , l The north stairs, platform and stairs to the reactor top will be removed using flame torches. None of these items are contaminated. As pieces are sectioned, they will be surveyed and removed for disposal as demolition waste / scrap metal for recycling. Access to the reactor top during the balance of the project will be provided using temporary ladders. 23.1.5 Plug Removal, Prep and Packaging Each plug from the reactor will be removed and staged for surface scabbling and survey. The radioactive ends of each plug will be removed using chipping hammers, drilling / splitting, scabbling or scarification to match the unrestricted release Documem No. 00752.m3. A01 Page 19 of 54 Rev.0
r pdO SONl:CS. Duke Ene neerine A herBuy cuymy criteria. Once the hot ends and sides have been removed (if requimd), the plug will be moved or oriented to a low background area for final survey and sampling. Once the fmal survey is complete the clean plug will be loaded for disposal as demolition waste. 2.3.1.6 Water Tank Demolition The water tank will be cut using a diamond blade concrete cut-off saw and removed as clean waste. Each section of the tank will be surveyed prior to placement in the demolition roll-off waste container. 2.3.1.7 Graphite & Core Component Removal The removal of the graphite from the reactor will be performed in two stages. First, graphite outside the activation zone will be removed. Second, material from the activated zone, which is 40 inches by 34 inches, will be removed. ISU may assume that the entire graphite volume is contaminated, but will make every effort to miease the non-activated portion of the graphite as demolition rubble. The activation zone li ats (distance from the core center) have been verified during the characterization phase. ISU may, with disposal site concurrence, dispose of all the graphite as radioactive material rather than sample and survey each block to meet the project release criteria. If this action is taken, it will has an extremely small impact on the total waste volume estimated for disposal. As graphite is remm ed from the shield cavity, each piece will be numbered and segregated for survey. As the removal operations continue, a survey of each piece will be performed in a low background area. Graphite that meets the project release criteria will be treated as demolition wastes. Radioactive graphite will be stacked in metal boxes for Low Level Radioactive Waste (LLRW) disposal. Core mechanical components, activated support structures and internals will be removed from the core, sectioned using shears, and packaged as radioactive waste in metal boxes. Radioactive materials will be packaged in the metal boxes in a sequence that reduces the exterior dose rate from the box by layering, with components of higher activity inside. Docunerd No. 00752.F03. A01 Page 20 of 54 Rev.0
l l ODuke Engineering ' EUU Services. ADMkw ummyc 2.3.1.8 Bio-Shield Removal l 1 The demolition of the bio-shield and appunenances will commence after the preparations discussed above have been completed and the area is ready to accommodate the dismantlement activities. 2.3.1.9 Wire Saw Cutting l l The use of a diamond impregnated wire concrete saw will l provide a clean cut across the l'ottom of the bio-shield and minimize the noise and vibration during concrete demolition operr.tions. The plan is to section the bio-shield using two horizontal cuts and at least two vertical cuts. The wire saw will be placed at the northeast corner of the room to accommodate the drive wheel and water collection system. The pulkvs at the corners will be adjusted to maintain proper wire i tension during cutting. A water spray will be installed to minimize dust and to lubricate the wire. The liquid waste generated during this process is expected to fill about one-half of a 55 gallon drum that may have to be solidified and sent to a ! licensed facility for disposal. Small dams will be located on the floor to minimize the spread of water. During cutting, water will be directed to a collection point and HEPA vacuumed into a drum. The water will be filtered and re-circulated. At the completion of concrete demolition, the sludge will be sampled and stabilized for disposal. During cutting of clean sections, a separate collection drum will be used to minimize the generation of LLRW. 2.3.1.10 Block Segregation The 150-ton reactor assembly will be wire cut into approximately 16 blocks each weighing no more than 10 tons each. These blocks will be funher reduced using ddlling and splitting and will be lifted using the existing overhead crane. Each block will be drilled and split to reduce the gross weight to less than the capacity of the overhead crane - five tons. To do so, the block will be drilled vertically using an air-powered rotary drill. A hydraulic splitter will be used to split the blocks along the drill line. Rebar inside the block will be exposed in the split and cut using a flame torch. The strategy for block segmentation will be to split the activated section from the clean sections and minimize the volume of LLRW for disposal. As each block is Documem No. 00752.m3. A01 Page 21 of 54 Rev.O
Duke E.neIneerine aservices. ADdeSugyCaymy sectioned, it will be surveyed in a low background area for { activation and packaged as either LLRW or demolition rubble. 23.1.11 Pit Excavation [ Directly underneath the reactor centerline is a small pit consisting of soil that may be radioactive via activation. This { area will be sampled at an appropriate time during decommissioning activities, as necessary. If samples indicate that the material has radioactivity greater than the release criteda, it will be disposed as radioactive waste. 23.1.12 Tank and Sump Equipment Removal The fuel pit, tank and associated piping for the reactor water will be removed to the point where it enters the concrete wall. Embedded piping will be abandoned in place or removed as necessary. All reactor water equipment is expected to be clean and will be scrapped as recyclable metals or disposed as demolition waste. The items in the sump and fuel pit will be decommissioned in accordance with the established 4 requirements. 23.1.13 Miscellaneous Equipment Removal Attached conduit, miscellaneous equipment and non-operational I systems attached to the walls will be removed, surveyed and disposed as demolition waste. 23.1.14 Temporary Staging Areas , A staging area for packaged radwaste will be established in a comer of the reactor room. This ama will be sufficient for the l staging of up to 40,000 pounds of debris and LLRW Once this storage capacity has been reached, a truck will be scheduled for pick-up of the LLRW for disposal. The storage area will be isolated with rope barriers. Each container of waste will be inventoried and the contents of each container noted as to the waste origin, curie content and packaging procedures. 23.1.15 Waste Disposal Waste disposal is discussed in section 3.23. Document No. 00752.F03. A01 Page 22 of 54 Rev.O
O Duke Engineering VUQ Services. ADMk=vc =m 2.3.2 Schedults A summary chart of major milestones and dates is shown in Table 2.3. A complete schedule is shown in Appendix B Table Bl. l 2.4 Decommissioning Organization and Responsibilities ISU as the owner of the facility, has the overall n sponsibility for the work conducted during decommissioning of the UTR-10 reactor. The University retained DE&S as the decoramissioning contractor. The DE&S Project Manager reports to the ISU Project Manager and is responsible for preparing the site for l decommissioning and implementing decommissioning and dismantlement activities. 2.4.1 Project Management Structure , l The project management stnicture for the decommissioning of the ISU reactor is consistent with the guidance provided in Appendix 17.1 of l NUREG-1537 (Ref.11). ISU will continue to be responsible for the I overall supervision, compliance with regulations, and the health and safety of the public. The management positions which comprise the key functional areas within the project organization are represented in Figure ! 2.5 and described in sections below. 2.4.2 ISU Project Manager The ISU Decommissioning Project Manager for the project is also the
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Reactor Manager. The Reactor Manager is responsible for the facility i management and the reactor operation. He has been in charge of the ! operations of the reactor for several years. He is familiar with every I aspect of the reactor, associated facilities, and the University policies and l controls related to radiation protection. He has Level 2 authority as shown in Figure 2.6. Level 1 authority rests with the University President because as Chief Executive Officer he is ultimately responsible for the facility license. l The ISU Project Manager (Reactor Manager) interacts with the University's Reactor Use Committee and the Radiation Safety Officer on matters related to radiation safety, ALARA reviews, and regulatory affairs. He also interacts with the Director of Environmental Health & Safety on health and safety issues. Figure 2.6. shows the Reactor Manager's role in the ISU management structure. Decument No,00752.M3.A01 Page 23 of 54 Rev.0
av Adde any Guys, 18 0 W Wr DESS ISU D&D Radiation l Program Director Safety Officer QAPP Coordinator i DE&S Project Manager I I I I Charactertantion D&D Superintendent unnager or Contractor ' i i i s i HP Technicians Field Personnel Design Englw C nator l l FIGURE 2.5 Project Organization i I I I t [ [ [ Docurrent No. 00752.193. A01 Page 24 of 54 Rev.O e
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C]a sofvices.bb ENSIfBOGfffBg 4D*8ae ow=, LEVEL 1 University President V V Vice President Provost Business & Finance Y V V l Dean of Radiation Safety Director of l Engineering Committee _ Environmental Health & Safety if if Facility Reactor Use Director Committee V ! . LEVEL 2 Radiation j - Safety Officer ! Reactor ! Manager V l V Health Physics l LEVEL 3 Staff i Reactor
' Operations Committee membership Staff FIGURE 2.6 ISU .Sh nagement Structure Docunent No. 00752.83. A01 Page 27 of 54 Rev.0
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A o* a=p ca,=, 2.4.3 ISU EH&S Director The EH&S Director oversees the environmental, health, and safety organization of ISU and is ultimately responsible for implementation of the appropriate policies and procedures. The ISU radiation safety office and the health physics staff are under the jurisdiction of the EH&S Director. 2.4.4 ISU Radiation Safety Officer The Radiation Safety Officer (RSO) at ISU reports directly to the EH&S Director and is responsible for health physics activities. He also acts as a member of the Radiation Safety Committee. During decommissioning activities, the RSO will review and appmve all DE&S Radiation Work Permits (RWPs). He will also ensure that the contractor's health physics staff folhw the radiation protection regulations and policies, including the implementation of ALARA. The RSO has the authority to interrupt or suspend any activity if the method used is deemed unsafe or contrary to appropriate regulations. 2.4.5 ISU Reactor Use Committee During operation of the UTR-10 facility, the Reactor Use Committee was responsible for reviewing and approving all experiments before they were conducted. Since the Committee reviews and approves all safety-related projects, it is responsible for the review and approval of the Decommissioning Plan prior to its submittal to the NRC. In addition, the Committee is responsible for the review and approval of any changes to the Plan under the provisions of 10 CFR 50.59. The Reactor Use Committee members are appointed by the ISU Radiation Safety Committee. The Radiation Safety Committee members are representatives from the various colleges at ISU and are appointed by the University Provost. 2.4.6 DE&S Program Director This individual has the overall programmatic responsibility for the decommissioning projects and for ensuring that home office support is available to the DE&S Project Manager. The Program Director reports to the DE&S Vice President of Decommissioning. This manager will coordinate home office support in the an:as of quality assurance, radiological engineering, civil / structural and mechanical engineering, occupational health and safety and licensing. The Program Director is the interface between DE&S management and the Project Manager in the field. He will provide all technical personnel for Phase I and III work Docunent No. 00752 R)3.A01 Page 28 of 54 Rev.0 l
O Duke Engineering klX2 Services. ADderneycaymy activities (site characterization and final survey) and supprt personnel for Phase 11 decommissioning activities. This individual is alsc. responsible for reviewing and approving the project reports includmg the Site Characterization Report, the Decommissioning Plan and the Final Site Survey Report. 2.4.7 DE&S Project Manager The DE&S Project Manager reports directly to the ISU Project Manager. He has overall accountability and responsibility for project execution. This includes quality assurance, project resources, project cost and schedule, and technical requirements. He will also serve as the health physicist for the project and will be responsible for ALARA program implementation i during all phases of the project. He will interface with the RSO for i approvals of RWPs and other matters as necessary. During Phase 11 work, this manager will be on-site, as necessary, controlling and directing decommissioning activities and interfacing with the ISU Project Manager and other technical representatives. Within the DE&S structure, this individual reports to the Program Director. 2.4.8 DE&S Health and Safety Officer This individual reports to the Project Manager and is responsible for occupational and industrial health and safety. 1 2.4.9 Characterization / Final Survey (Phase I and III) Manager This manager was responsible for developing and executing a j comprehensive characterization of the reactor area during Phase I. Tasks ' included radioactive characterization of the site and preparing a i Characterization Report that formed the basis of this Deconunissioning l Plan. The comprehensive characterization program conducted under MARSSIM guidelines included sampling, radiation surveys, collection and analysis of samples identifying surface and subsurface contamination. This individual will also be responsible for the final survey during Phase 111 and will prepare the Final Survey Report. I 2.4.10 Phase II D&D Superintendent This individual is responsible for the decommissioning and dismantlement work to be performed in Phase II. He will also maintain relationships with y the LLRW processing and disposal contractors and assist in obtaining the final arrangements for disposal. During Phase II, the D&D Superintendent ensures that all work is performed in accordance with project plans and procedures. Docurm No.00752.IMA01 Page 29 of 54 Rev.0
l m Duke Engineering EduServices. A DdeEnngyCmyssy lg The D&D Superintendent's responsibilities include: ensuring data lE collection in accordance with project procedures; managing field operations and executing work plans; enforcing site control; conducting daily safety briefings; and supervising craft and other personnel assigned to work at the site. 2.4.11 DE&S Quality Assurance Project Plan Coordinator l The Quality Assurance Project Plan Coordinator reports to the DE&S l Program Director as shown in Figure 2.5. This individual is outside of the j direct project organization in order to maintain independence of the quality assurance checks and balances. His responsibilities include i preparing and maintaining the quality assurance documents, policies and l ) procedures, and ensuring that these are followed. 2.5 Training Program All personnel directly involved with the ISU decommissioning project will ) complete the following training programs: ' e General employee training in compliance with 10 CFR 19.12 (Ref.12) for all l personnel involved with radioactive materials or who work in the vicinity of radioactive materials.
- Respiratory protection training in accordance with the requirements of NRC Reg. Guide 8.15 (Ref.13), NUREG -1400 (Ref.14), and 29 CFR 1910.134 (Ref.15). l 1
- Training consistent with the requirements of 29 CFR 1910.120.
- Hazard communication training consistent with the requirements of 29 CFR 1910.1200.
2.6 Contractor Assistance DE&S was selected by ISU as the decommissioning contractor. DE&S has extensive experience in the nuclear industry including decommissioning of nuclear power reactors. 1 2.7 Documents and Guides The Characterization Report lists the radiological data obtained from the characterization activities conducted at the site. This data forms the basis of this Docunent No. 00752.F03.A01 Page 30 of 54 Rev.0 l
O ouke snelneering [ ava services. ADe s=e cav=, ( Decommissioning Plan. The Characterization Report is included with this submittal. [ The regulatory documents, guides, and other material used in preparation of this Decommissioning Plan are listed in Section 10, References. ( 2.8 Quality Assurance ISU has ALARA and safety review processes in place to ensure that quality assurance is maintained in all work related to reactor operation and management. DE&S, as the decommissioning contractor, will follow these procedures as applicable to decommission the facility in concert with its existing procedures. DE&S has comprehensive quality assurance procedures for decommissioning projects which cover various phases from characterization surveys, through the dismantlement activities, to the final status survey. Field surveys and collection of radiological data in the field require additional quality provisions under MARSSIM. The DE&S Survey Quality Assurance [ Project Plan (Survey QAPP) (Ref.16) applies to these activities. The Survey QAPP specifies the policies, organization, objectives, and Quality ( Assurance / Quality Control (QA/QC) procedures used by DE&S for site L characterization and final survey activities. The primary goal of the QA/QC program is to identify and specify the implementation of surveying, sampling, and analytical methodologies which will limit the introduction of elTors into the analytical data. The specific purpose of the Survey QAPP is to ensure that the samples are collected, analyzed and reported in a consistent manner and that the quality of the resulting data can be independently evaluated. { For the ISU project, DE&S is using its Environmental bboratory (DESEL) for sample analysis. The DESEL has a comprehensive quality assurance program for laboratory analysis. The DESEL Manual 100, bboratory Quality Assurance Plan (Ref.17), provides the main elements of the QA Program. The laboratory participates in a number of quality assurance cross comparison programs. The laboratory issues semi-annual quality assurance status reports. Procedures are also in place covering all aspects of the movement of samples, such as, sample receipt and general chain-of-custody, sample storage and accountability, quality { assurance of laboratory instrumentation, control of records, and quality control audits. l Document No. 00752.IU3. A01 Page 31 of 54 Rev 0 l
D Duke Engineering [ EL)U Services. ADubrSuyCmymy [ 3.0 PROTECTION OF THE HEALTH AND SAFETY OF RADIATION WORKERS AND THE PUBLIC 3.1 Radiation Protection The NRC radiation protection regulations are codMied in 10 CFR 20. All applicable elements from 10 CFR 20 will be applied throughout the duration of the ISU project. Decommissioning operations will be controlled to maintain occupational exposure well below the exposure limits specified in 10 CFR 20.1201. Table 3.1 lists the occupational exposure limits. In addition, it is the, policy of ISU to apply ALARA to all radiation-related activities. The ISU procedures and policies related to radiation protection are defined in the ISU Radiation Safety Manual (Ref.18). This manual is the official guide for radiation protection at ISU and for the use of radioactive materials. In addition to [ the regulatory requirements, the manual defines the administrative controls, authorization process, personnel exposure limits, personnel monitoring, and the roles of the Radiation Safety Officer and the Radiation Safety Committee. TABLE 3.1 NRC Occupational Dose Limits Exposure 10 CFR 20 Annual Dose Limits Whole Body (Internal + Extemal) TEDE 5 rem Lens of the Eye 15 rem Extremities 50 rem Any Organ, Tissue or Skin 50 rem Declared Pregnant Worker 0.5 rem Minors (under age 18) 0.1 rem Members of Public 0.1 rem 3.1.1 Ensuring ALARA It is the policy of both ISU and DE&S to ensure that all exposures to radiation are kept ALARA. For the ISU decommissioning project, policies specified in the ISU Radiation Safety Manual will be followed. Administrative controls and authorization processes detailed in the manual ( will be applied to the project. All workers involved in the project will have radiation training including ALARA.
- 3.1.2 Health Physics Program Radiological controls throughout the dismantlement process will be ensured through the use of Radiation Work Permits (RWP). RWPs will Docunent No. 00752.F03. A01 Page 32 of 54 Rev.O L
D Duke F.ngineering EUA Services. ADernmepGuymy control use of protective clothing and other requirements, as necessary. I RWPs will be approved by the RSO prior to initiation of work activities. 3.1.3 Dose Estimates For estimating the collective dose estimates for occupational exposure, all activities during the decommissioning am organized under three main groupings. These are: (1) site characterization; (2) systems removal (which includes removal and packaging of shield plugs, removal and packaging of graphite, removal and packaging of core components, and miscellaneous equipment removal); and (3) bio-shield removal and demolition activities (which include removal and packaging of the bio-shield, block segregation, water tank demolition, tank and sump i equipment mmoval, pit excavation, and the shipping of radioactive waste). The estimated upper-bound doses for these three groups are listed in Table 3.2, conservatively using the maximum exposure levels rather than the average values. The total estimated collective dose for the project is 2.4 person-rem. TABLE 3.2 Estimated Collective Doses for Decommissioning Activities Decommissioning Activity Estimated Collective Dose (person-rem) Site Characterization 0* Systems Removal 0.7 l Bio-Shield Removal & Demolition 1.7 l
- Actual measurement data from TLDs 3.2 Radioactive Waste Management 3.2.1 Fuel Removal 1
After the UTR-10 reactor ceased operations on ?"cy 15,1998, it was j defueled. The fuel is currently in storage at the site awaiting shipment to l DOE's Savannah River Site. The fuel will be removed from the site by the time decommissioning activities begin. Currently, it is anticipated that the fuel will be shipped in April of 1999. 3.2.2 Radioactive Waste Segregation and Processing , I As mentioned in Section 2.2.4, possibility of mixed waste is precluded. These sections deal with radioactive waste segregation and processing. Document No. 00752 R)3. A01 Page 33 c,f 54 Rev.O
1 D Duke Engineering EduServices. ADaBavCnn 3.2.2.1 Demolition Rubble Disposal Non-radioactive waste originating during the project will be segregated from the radioactive waste. A local demolition waste contractor will be retained to provide 20-yard roll-on/ roll-off containers and disposal services for non-radioactive debris. The 20-yard containers will be scheduled and staged to prevent / minimize interference with traffic outside the south doors. Debris will be removed using a bobcat with a 1.5 yard bucket or a 5-ton forklift. Rubble will be collected and misted to minimize dust during dumping into the roll-off. The blocks and plugs will be lifted and placed into the roll-off to center the load in the bottom of the container. The debris will be disposed of at the Ames Story Environmental Landfill in Ames, Iowa. Each load of material for disposal will be weighed and a disposal confirmation certificate will be maintained in the project documentation file along with the radioactive sample and survey data. 3.2.2.2 Gamma Curtain The Gamma Curtain is expected to be a radioactive mixed waste requiring special treatment for disposal. The 1300 pound shield will be placed in a container and shipped to Envirocare for encapsulation and disposal at their site in Clive, Utah. 3.2.3 Radioactive Waste Transportation and Disposal Waste transportation to the LLRW disposal site will be accomplished using a permitted carrier with experience in radioactive materials shipments and following the Department of Transportation (DOT) regulations (Ref. 21). The truck will undergo a pre-shipment inspection and radiological survey prior to loading. After package loading, the containers will be secured to the trailer r,ad placards / labels applied as required by regulations. The truck will undergo a safety inspection, driver briefing and outgoing radiation survey prior to dispatch. Emergency response instructions will be included with all shipments. Shipments will be exclusive use and one way, with no intermediate stops for loading or unloading. The proposed and preferred LLRW waste disposal site is Envirocare of Utah, Inc. located in Clive, Utah, 80 miles west of Salt Lake City. This facility, in operation since 1987 has the licenses and permits to accept and dispose LLRW from reactor decommissioning projects. The materials for disposal for the project are expected to be less than 1,066 cubic feet with a Document No. 00752.fU3.A01 Page 34 of 54 Rev.0
I I m Duke v.neIneertne LUUServices. ADerSugyOmymmy maximum of 1,200 cubic feet. Materials that are classified as LLRW will be packaged and shipped directly to Envirocare for disposal from ISU. Changes in waste acceptance policies, pricing and competition may make the use of altemative LLRW disposition methods more attractive when D&D operations actually occur. As an altemative to Envirocare, the decommissioning contractor may recommend use of a waste processor for the receipt, survey, volume reduction, packaging and disposal of some or all of the LLRW from the project. Such processors are licensed for the survey for free release and disposal of LLRW at approved facilities. 3.2.4 GeneralIndustrial Safety Program During the ISU decommissioning project, applicable procedures and requirements from the ISU Environment, Safety and Health Manual will be implemented. 3.2.4.1 Management Policy ISU and its decommissioning contractor, DE&S, are committed to the safe decommissioning of the ISU reactor. The primary objective of the EH&S Program is to protect workers and visitors from industrial hazards that have the potential of developing during decommissioning activities. ISU and DE&S will provide sufficient qualified staff, facilities, and equipment to perform decommissioning in a safe and effective manner. ISU is committed to compliance with federal and state requirements and to the guidance provided through industry standards and good work practices. 3.2.4.2 Health & Safety Organization The existing ISU Environmental Health and Safety Program provides the basis for controlling safety during decommissioning activities. It is augmented by the programs and procedures provided by the decommissioning contractor for specific evolutions during the decommissioning process. The purpose of the health and safety organization is to ensure that the standards of safety are maintained through effective implementation of the EH&S Program. ISU will ensure that the contractor's personnel follow all applicable ISU procedures related to health and safety. ISU Project Manager - The ISU Project Manager has the overall responsibility for safe decommissioning of the ISU reactor and has control over on-site actions. Included in this is the responsibility for assuring effective implementation of Docunent No. 00752.R3.A01 Page 35 of 54 Rev.O
1 (3 Duke Engineering \ KUGServices. l
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the EH&S Program. The Project Manager will ensum l 'I coordination of all organizations involved with decommissioning to achieve the goals of providing a safe work place and reduce industrial hazards. i ISU EH&S Suoervisor. The EH&S Supervisor is responsible 4 for the implementation of the EH&S Program policies and standards. The Supervisor has the authority to cease any work activity when worker safety is jeopardized or an unsafe condition occurs. The DE&S Health and Safety Officer i (Section 2.4.8) will work with the EH&S Supervisor to aI ensure that the EH&S Piogram implementation covers the contractor's personnel. Contractor Decommissioning Suoervisors - All supervisory personnel are responsible for the supervision and direction of , safety practices during decommissioning activities. Contrastor Decommissionine Workers - All plant and , 4 decomraissioning workers are responsible for their own safe l work practices as presented in the Safety Manual and OSHA Standards. I 3.2.4.3 ISU Environmental Health and Safety Program The ISU Environmental Health & Safety Program was developed
, to establish and maintain a safe work place for ISU personnel, contractors, and visitors. The program provides guidelines and procedures to be used to reduce industrial hazards and risks.
The ISU EH&S Manual provides guidelines and requirements which will be incorporated into the detailed decommissioning planning process. The following areas are discussed in the manual: Personnel protection and safety equipment
- Prevention of falls Safe use ofladders and scaffolding
. Safe handling of hazardous substances and materials Safe use of hand and portable powered tools and equipment Document No. 00752.M3.A01 Page 36 of 54 Rev.O I
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Electrical safety Confined space safety requirements Heat stress prevention Hazard communication A Site-Specific Health and Safety Plan consistent with the requirements of 29 CFR 1926 will be prepared by DE&S to support personnel safety during decommissioning activities. 3.2.4.4 Safety Training & Meetings Safety training is conducted as part of the initial project training process. Ongoing training will occur during routine safety meetings and daily work briefings. The safety meetings and briefings will focus on current safety issues and events and provide a fomm for workers to ask questions and give feedback. 3.2.4.5 Respiratory Protection Program A respiratory protection program will be implemented in compliance with 10 CFR 20.1701 though 20.1704, U. S. NRC Regulatory Guide 8.15, ANSI Z-88.2 (Ref.19) and OSHA 1920.134 (Ref. 20). The Respiratory Protection Program will contain the following components to ensure worker safety:
- Written procedures which govern selection, use and maintenance
- Training of employees and supervisors
- Fit testing
- Inspection, issuance and storige
- Air monitoring
- Recordkeeping
- Medical surveillance
- Special respirator use problems and limitations All respiratory protective equipment utilized during the decommissioning effort will be NIOSH/MSHA approved.
Dxunent No. 00752.R)3. A01 Page 37 of 54 Rev.O
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f3 Duke Erwineering tv& Services. A%s w cw=, 3.3 Radiological Accident Analysis Since all credible accident scenarios defined in the Safety Analysis Report (Ref.4) involve the presence of nuclear fuel, the probability of a criticality accident to occur that could impact the occupational or public health and safety during the decommissioning process is precluded. The residual radioactive inventory is very small and there is no possibility of a major event that will impact occupational or public health and safety. 4.0 PROPOSED FINAL RADIATION SURVEY PLAN
- g. 4.1 Final Status Survey Plan Under MARSSIM Guidelines The pmpose of the Final Status Survey Plan under tne MARSSIM guidelines, is to demonstrate compliance with regulations and ensure that any survey unit on site does not contain residual radioactive contamination or elevated activity in excess of the release criteria for unrestricted use.
The results and conclusions of the characterization survey report and/or remediation efforts that describe the current radiological characteristics of the ISU Nuclear Engineering 12boratcry (NEL) site will be used to develop the final status survey plan. Following the decontamination and decommissioning process, the final status survey will consist of radiation measurements, field samples and data evaluations of the ISU NEL site. MARSSIM recommends using the graded approach in designing a final status survey. This type of approach to data collection places the greatest survey efforts on areas that have, or had, the highest potential for residual radioactivity and demonstrates that all radiological parameters do not exceed the established release criteria. 4.2 Data Quality Objectives The Data Quality Objectives (DQO) process will be implemented in the final status survey plan using a graded approach as recommended in MARSSIM. The DQOs for the NEL site are:
- Selecting and verifying survey unit classification.
- Collecting sufficient data of high quality to ensure that a comparison can be made with the release criteria for each survey unit to determine if residual radioactivity in each unit has been reduced to a level below the release criteria.
Document No. 00752.N3.A01 Page 38 of 54 Rev.0
13 Duke Engineering kV& Services. ADeraugyCayesy
- Ensuring that the potential risk from the site as a whole is below the dose limit release criteria.
4.3 Final Area Classification and Survey Unit Designation The final area classification of the ISU NEL facility will be subdivided into Class 1, Class 2, and Class 3 areas, according to MARSSIM. These three categories are defined as follows:
- Class 1 Areas: Areas that had, prior to remediation, known contamination based on characterization survey in excess of the DCGL.
- Class 2 Areas: Areas that had, prior to remediation, a potential for radioactive contamination or known contamination, but are not expected to exceed the DCGL.
- Class 3 Areas: Any impacted areas that are not expected to contain any residual radioactivity, or are expected to contain levels of residual radioactivity at a small fraction of the DCGL.
Since the Characterization Repon has shown that only Class I areas had contamination that could be confirmed, these areas will receive the highest degree of survey effons during the final survey. The area classification of the ISU NEL facility for the final smvey will be as shown in Table 4.1. To facilitate survey design, and ensure that the number of survey data points for a specific site are relatively uniformly distributed among areas of similar contamination potential, the facility is divided into survey units that share a common history or other characteristics, or are naturally distinguishable from other portions of the facility. Survey locations will be clearly identified to provide a method of referencing survey results to survey unit locations. Designated Class I survey units will be subdivided uniformly for survey design, measurement control purposes, and to identify survey locations. Normally, Class 3 survey units will not be gridded. Survey locations will normally be indicated on surfaces by the use of self-adhesive labels, temporary markers, or equivalent methods. The survey of surfaces and stmetures located within the ISU NEL will be conducted based on the operating history, results of the characterization survey report, and/or remediation efforts. Document No. 00752.103.A01 Page 39 of 54 Rev.0
ID Du b Engit9000ing kUU Services. Anwc p, 4 Once the oecontamination & decommissioning process has been complete-d, an ' extensive surface and stmeture survey of the restricted area will be conducted. This survey will include the areas shown in Table 4.1. Docunent No. 00752.RB.A01 Page 40 e754 Rev.O
l ouke EneIneertne 1 daservices. AM %c=y=, TABLE 4.1 Final Status Survey Unit Designation and Classification. Nuclear Engineering Laboratory - Basement
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Un Survey Unit Description 3) Classification NEL01 Floors and Lower Walls 550 3 (including the basement sump)
' Nuclear Engineering Laboratory - First Floor .
"" b" Area U Suney Unit Description Classification 3)
NELO2 Floors and Lower Walls 1396 3 Nuclear Engineering Laboratory - Second Floor Survey Unit Description
" Area U Classification 3)
NELO3 Floors and Lower Walls 1467 3 Central Bay Area - Reactor Room
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b"" U [ Survey Unit Description 3) Classification CBA01 Floors and Lower Walls 240 3 CBA02 Upper Walls and Ceiling 770 3 1 Central Bay Area - Reactor Housing (Foundation) b" " ' Area Ui Survey Unit Description Classification 3) CBA04 Reactor Housing Internal Surfaces <100 1 CBA05 Reactor Housing Internal Surfaces <100 3 Central Bay Area - Process Pit-
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U Survey Unit Description Classification 3) PP001 Process Pit 26 i Central Bay Area - Fuel Storage Pit .-
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U Survey Unit Description Classification 3) FP001 Fuel Storage Pit <100m 2 h Docunent No. 00752.R)3. Aot Page 41 of 54 Rev.0 I
1 C p b Engl W ng \ tua services. Ana,cm l TABLE 4.2 Nuclear Engineering Laboratory Final Survey Instrumentation ] l Instrument / Detector Type , Description; , Measurement Types (s); ) 1 Eberline E-600 Data Logger Electronic Storage of Data ! l 2 Eberline SHP-100 100 cm sealed-gas proportional detector Beta scan i 2 ' Ludlum 43-37 435 cm sealed-gas proportional detector Beta surface contamination Alpha surface contamination Eberline SHP-360 15.5 cm2 GM detector Beta-gamma scan Beta-gamma surface contamination Eberline SHP-300 Pressurized GM Gamma exposure rates l Eberline SPA-3 or 2" X 2" Nal or closed-end coaxial Ge Gamma-scans Canberra Ge Detector detector Low Background Gas Flow Alpha and beta activity on smear samples Laboratory analysis l Proporuonal Detector l Liquid Scintillation Low energy beta activity on smear Laboratory analysis samples l Gamma Spectroscopy Gamma-emitting radionuclide Laboratory analysis identification and activity quantification on smear & bulk material samples 4.4 Instrumentation To perform a final status survey both field survey instmmentation and analytical laboratory equipment will be selected based on: (1) the necessary Minimum Detectable Concentrations (MDC), and (2) stability and reliability under environmental conditions. DE&S will utilize the appropriate instrumentation shown in Table 4.2 to perform the final status survey of the NEL site. 4.5 Background Determination Background reference of the NEL facility will be an area located within the survey unit due to material composition and the fact no "non-impacted" areas are located within the facility. However, special instrumentation (e.g. portable gamma-spec) will be used to assure the area selected for a background reference area is free of radioactive material. 4.6 Demonstrating Compliance with Release Criteria l DCGLs will be used to demonstrate compliance with the criteria. s I Docunent No. 00752.F03.A01 Page 42 of 54 Rev.O
(3 Duke Engineering kUUServices. A Der Buy Canya, 4.7 Documentation of Final Survey Results Documentation of the final status survey will provide a complete and unambiguous record of the current radiological status of survey units located within the ISU NEL facility. Three types of documentation will be utilized during the final status survey: (1) field operations records, (2) laboratory records, and (3) data handling records. A field log book will be maintained by DE&S to serve as a reference document for all field activities performed at the ISU NEL facility not covered by the characterization work plan. A reference copy of the field log book will become part of the permanent project file maintained by ISU. 5.0 TECHNICAL SPECIFICATIONS f' A request for Possession Only License (POL) has been submitted to NRC (Ref. 22). C!mnges to the technical specifications of the reactor are described in this request. 6.0 PHYSICAL SECURITY PLAN 6.1 Physical Site Security During Decommissioning The reactor facility part of the building is secured at all times. The reactor manager has control over the keys and these are issued for access only on the basis of a need to enter. Visitors and non-radiation workers must be escorted by the reactor manager or the RSO designee. During decommissioning activities, all access to the reactor facility will be limited to those personnel trained to perform the decommissioning work. During off-hours, access to the reactor facility will be secured. 6.2 Material Safeguards The reactor currently has no fuel; the fuel is in storage in the fuel storage bay in accordance with the applicable safeguards and procedures. It is scheduled for shipment to DOE's Savannah River Site in April 1999. The decommissioning process will not begin until the irradiated fuel has left the site; thus, there are no issues related to materW safeguards that need to be considered. 7.0 EMERGENCY PLAN Since the reactor has been defueled, the major source of radioactive material has been removed. All credible accidents considered in the Safety Analysis Report (Ref. 4) and that will require the implementation of full provisions of the Emergency Plan, involve fuel failure. Since the fuel will be moved off-site prior to the start of decommissioning Docunwnt No. 00752.F03. A01 Page 43 of 54 Rev.O
f3 Duke Engineedng rOuservices. ADubrSugycmysy work, accidents that involve enough radioactive material to reach the conditions of Notification of Unusual Events Alert or the Site Area Emergency are not possible. For any potential Personnel Emergency, applicable provisions of the Emergency Plan will be implemented. 8.0 ENVIRONMENTAL REPORT ISU has evaluated the environmental impact of the decommissioning action in accordance with 10 CFR 51.45 and 10 CFR 51.30. It is concluded that the decommissioning of ISU UTR-10 will have no significant environmental impact (Ref. 23). 9.0 CHANGES TO DECOMMISSIONING PLAN Any changes after the approval of the Decommissioning Plan by the NRC will be handled in accordance with the requirements of 10 CFR 50.59.
10.0 REFERENCES
(1) ISU, Technical Specifications for the UTR-10 Reactor Facility at Iowa State University, Original August 1983, Amendment 1 November,1988. (2) ISU, Letter to USNRC, Letter of Intent, December 21,1998. (3) USNRC, NUREG-0586, Final Generic Environmental Impact Statement of Decommissioning of Nuclear Facilities. (4) ISU, Safety Analysis Report for the Training Reactor UTR-10, ISU-ERI-Ames 82418, August 1981. (5) USNRC, NUREG-1016, Safety Evaluation Report Related to the Renewal of the Operating License for the Research Reactor at Iowa State University, Docket No. 50-116, September 1983. (6) DE&S' Document No. 00752.F02.A01, Characterization Report for the Iowa State University UTR-10 Reactor, December 1998. (7) USNRC. NUREG-1575, Multi-Agency Radiation Survey and Site Investigation Manual (MARSSIM), December 1997. (8) Title 10, Code of Federal Regulations, Part 61, Licensing Requirements for the Land Disposal of Radioactive Waste. Docunent No. 00752.F03.A01 Page 44 of 54 Rev.0
n Duke EneIneerine cuaservices. ADdeBusycmysey (9) Title 10, Code of Federal Regulations, Part 20, Standards for Protection Against Radiation, various subsections; 10 CFR 20.1402, Final Rule on License Termination, Radiological Criteria for Unrestricted Release. (10) USNRC, D and D, Version 1.0,1998. (11) USNRC, NUREG-1537, Guidelines for Preparing and Reviewing Applications for the Licensing of Non-Power Reactors, Appendix 17.1, February 1996. (12) ritte iv, Code of Federal Regulations, Part 19, Instruction to Workers. (13) USNRC, Regulatory Guide 8.15, Applicable Programs for Respiratory Protection, l October 1976. (14) NUREG-1400, Air Sampling in the Workplace, October 1991. (15) Title 29, Code of Federal Regulations, Part 1910, Occupational Safety and Health. (16) DE&S, Quality Assurance Project Plan For Site Characterization and Final Survey Activities, August,1998. (17) DE&S, Laboratory Quality Assurance Plan, DESEL Manual 100, Revision 0, July 1998. (18) lowa State University (ISU), Radiation Safety Manual, May 1996. (19) ANSI Z88.2, Practices for Respiratory Protection,1980. (20) Title 29, Code of Federal Regulations, Part 1926, Safety and Health Regulations for Constructien. (21) Code of Federal Regulations,49 CFR 100-177, Transportation of Hazardous Materials. (22) ISU, Request for POL, submitted to NRC, December 28,1998. (23) ISU, Environmental Report for the Decommissioning of ISU UTR-10 Reactor, November 1998. l l Docunent Na 00752.F03.A01 Page 45 of 54 Rev.0 m
m Duke 9.ngineering , niraservices. Am%o =, w l 4 . I a i i 4 i- l i l l
- APPENDIX A 1
. License Amendments to ISU UTR-10 Technical Specifications 1 f 4 1 Document No. 00752.RD. A01 Page 46 of 54 Rev. o
D Duke E.ngineering stJOSorrices. A D e lh ec= ,= , APPENDIX A TABLE Al License Amendments to ISU UTR-10 Technical Specifications I
"'," D
,1 ate;amediment beca' m e effective?
1 September 18,1961 2 January 18,1963 3 April 22,1963 October 26,1970
~
4 ~ 5 August 7,1975 6 February 23,1982 7 October 21,1983 l 8 September 12,1990 9 February 14,1992 1 10 November 29,1993 11 December 13,1996 1 12 March 30,1998 13 Submitted on December 28,1998 I The following descriptions of the amendments: l Amendment 1 - September 18,1961 - Clarified facility procedures to monitor changes in core reactivity and report changes to the NRC. Amendment 2 - January 18,1963 - Specified facility fuel handling pmcedures. Amendment 3 - April 22,1963 - Addition of shield tank apparatus and water :reatment system, I change in period scram system, and change in control rod interrupts. Amendment 4 - October 26,1970 - Changed Technical Specification reference to amount of I U-235 in fission chambers. Amendment 5 - August 7,1975 - Increased amount of U-235 allowed at facility. Amendment 6 - February 23, 1982 - Revision to Physical Security Plan requiring license amendment. Amendment 7 - October 21,1983 - License renewal amendment. Docunrra No 00752.f03.A01 Page 47 of 54 Rev, O e
1 l l r i l l bb \ ku G Services. & N W DC \ j ADaarannyCuyasy Amendment 8 - September 12, 1990 - Conversion from high enriched fuel (HEU) to low enriched fuel (LEU). Amendment 9 - February 14, 1992 - Updated Administrative Controls section of Technical 1 l Specifications. Amendment 10 - November 29,1993 - Added definition of " confinement secured" to Technical ! Specificatior.s. Amendment i1 - December 13,1996 - Updated Administrative Controls section of Technical Specifications and definition of " confinement secured." Amendment 12 - March 30,1998 - Added definition of " surveillance time intervals" to Technical Specifications. Amendment 13 - Submitted to NRC on December 28,1998 - Request for a Possession Only License. l l l l l d' f i thrunen No. 00752.R)3.A01 Page 48 of 54 Rev.0
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