ML19325F187
| ML19325F187 | |
| Person / Time | |
|---|---|
| Site: | 05000294 |
| Issue date: | 11/30/1989 |
| From: | CHEM-NUCLEAR SYSTEMS, INC. |
| To: | |
| Shared Package | |
| ML19325F186 | List: |
| References | |
| NUDOCS 8911160047 | |
| Download: ML19325F187 (74) | |
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- MICHIGAN STATE UNIVERSITY TRIGAJEACTOR DECOMMISSIONING PROJECT
- TERMINATION RADIATION SURVEY
SUMMARY
and FINAL REPORT . 7
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US NRC LICENSE R-114 DOCKET No. 50-294 Michigan State University -- l Engineering liuilding East Lansing, Michigan
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;f DECOMMISSIONING FINAL REPORT and
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TERMINATION RADIATION SURVEY RESULTS l i j c MICHIGAN STATE UNIVERSITY , j TRIGA' REACTOR DECOMMISSIONING PROJECT ; l \ r i e O : i OPERATING LICENSE NO. R-114 i DOCKET NO. 50-294 MICHIGAN STATE UNIVERSITY ! ENGINEERING BUILDING ' EAST LANSING,. MICHIGAN i f Prepared by: CHEM-NUCLEAR SYSTEMS, INC. NOVEMBER,1989
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, '[ ) TABLE OF CONTENTS y PAGE NO. i
1.0 BACKGROUND
AND MANAGEMENT . . . . . . . . . . . . . . . . 1 1.1
SUMMARY
DESCRIPTION . . . . . . . . . . . . . . . . . . . 1 1.1.1 Reactor . . . . . . . . . . . . . . . . . . . . . 1. 1.1.2 Support Facilities . . . . . . . . . . . . . . . . 2 1.2 FACILITY OPERATING HISTORY. . . . . . . . . . . . . . . . 3 , 1.3 CURRENT RADIOLOGICAL STATUS OF FACILITY . . . . . . . . . . 4 l L 1.4 DECOMMISSIONING ALTERNATIVE . . . . . . . . . . . . . . . 5 l 1.5 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES . . . . 5 1.6 REGULATIONS, REGULATORY GUIDES AND STANDARDS. . . . . . . 6 L 1.6.1 Generml. . . . . . . .............. 6
; (~~ 1.6.2 Release Criteria . . . . . . . . . . . . . . . . . 6 l hs-1.7 TRAINING AND QUALIFICATIONS . . . . . . . . . . . . . . . 7 2.0 OCCUPATIONAL & RADIATION PROTECTION PROGRAMS. . . . . . . 8 2.3 CONTRACTOR ASSISTANCE . . . . . . . . . . . . . . . . . .10 F
2.4 COST ESTIMATE AND FUNDING . . . . . . . . . . . . . . . .10 3.0 DISMANTLING AND DECOMMISSIONING TASKS AND SCHEDULES . . .11 3.1 TASKS . . . . . . . . . . . . . . . . . . . . . . . . . .11 3.1.1 Baseline Radiological Survey . . . . . . . . . . .11 3.1.2 Dismantlement Operations . . . . . . . . . . . . .12 3.2 SCHEDULE. . . . . . . . . . . . . . . . . . . . . . . . .12 ( uJ 1
E" . 1 a " L; I i 9 ic TABLE OF CONTENTS (CONTINUED)_ - h c T , t
+ 1
-! PAGE NOx !
3.3 . TASK ANALYSES .. . . . . . . . . . . . . . . . . . . . 12. , 3.3.1 Dismantlement. . . . . . . . . . . .. . . . . . 14 3.3.2 Water Disposal . . . . . . . . . . . . . . . . . 15 1 F 3.3.3 In-tank Operations Set-up. . . . . . . . . . . . 18 I 3.3.4 . Reactor Cavity Liner Decontamination . . . . . . 18 i 3.3.5 Radiological Surveys of Accessible Tank Areas.-. 19 3.3.6 Excavation Requirements . . . . . . . . . . . . 19 ! 3.3.7 Concrete Excavation. . . . . . . . . . . . . . . 20 j-L 3.4- SAFE STORAGE . . . . . . . . . . . . .- . . . . . s . . 21
- t 4.0 SAFEGUARDS AND PHYSICAL SECURITY. . . . . . . . . . . . 22 r '
5.0 RADIOLOGICAL ACCIDENT ANALYSES. . . . . . . . . . . 23 j
; r 6.0 RADIOACTIVE MATERIALS AND WASTE MANAGEMENT. . . . . . . 24 6.1 FUEL DISPOSAL . . . . . . . . . . . . . . . . . . . . . 24 s }f-w(
L) 6.2 RADIOACTIVE WASTE PROCESSING. . . . . . . . . . . . . . 24 '[ 6.2.1- Liquid Radioactive Waste . . . . . . . . . . . . 24 6.2.2 Solid Radioactive Waste. . . . . . . . . . . . . 25 ; 7.0 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS. . . . . . . 26 7.1 -TECHNICAL SPECIFICATIONS DURING DECOMMISSIONING . . . . 26 7.1.1 Administrative Controls . . . . . . . . . . . . 26 7.1.2 Surveillance Requirements . . . . . . . . . . . 27 l. L 8.0 TERMINATION RADIATION SURVEY PLAN . . . . . . . . . . . 28 l i 8.1
SUMMARY
OF TERMINATION SURVEY METHODOLOGY . . . . . . . 28 8.1.1 Low Potential Areas. . . . . . . . . . . . . . 29 l 8.1.2 Medium Potential Areas . . . . . . . . ., . . . . 29 l 8.1.3 High Potential / Contaminated Areas. . . . . . . . 29 . i 8.1.4 Induced Activity Areas . . . . . . . . . . . . . 29 l 8.1.5 Additional Grid Surveyn. . . . . . . . . . . . . 30 l
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w. i I f TABLE OF CONTENTS (CONTINUED) r~Y PAGE NO. V 8.2 TERMINATION SURVEY PROCEDURE. . . . . . . . . . . . . . 30 [ f 8.2.1 Grid Surveys . . . . . .. . . . . . . . . . . . 30 8.2.2 Other Surveys. . . . . . . . . . . . . . . . . . 31 8.2.3 Reactor Cavity Surveys . . . . . . . . . . . . . 31 8.3 TERMINATION RADIATION SURVEY RESULTS (
SUMMARY
) . . . . 31 8.4 INSTRUMENTATION . . . . . . . . . . . . . . . . . . . . 32 L 8.5 FINAL SITE RELEASE. . . . . . . . . . . . . . . . . . . . 33 TABLES 2.0A COLLECTIVE DOSE
SUMMARY
. . . . . . . . . . . . . . . . 9 l 2.0B WHOLE BODY EXPOSURE RANGE . . . . . . . . . . . . . . . 8 l !
l
3.1 BACKGROUND
EXPOSURE RATE SURVEY (" BENCHMARK") . . . . . 13 lrs 3.3A EXPOSURE RATES FOR VARIOUS REACTOR COMPONENTS . . . . . 14
- 3.3B INVENTORY OF RADIOACTIVE COMPONENTS SHIPPED TO ATU. . . 16 l
3.3C
SUMMARY
OF LIQUID DISCHARGES TO THE SANITARY SEWER. . . 17 3.3D AIRBORNE RADIOACTIVITY CONCENTRATIONS . . . . . . . . . 18 ! 3.3E EXPOSURE RATE (S) INSIDE THE REACTOR TANK . . . . . . . 19 APPENDICES ; A FIGURES . . . . . . . . . . . . . . . . . . . . . . . 34 B TERMINATION RADIATION SURVEY DATA . . . . . . . . . . . 36
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1.0 BACKGROUND
AND MANAGEMENT 7
~' The following sections of this report show that the Michigan State University Reactor Facility now meets unrestricted use limits as specified in the NRC Order Authorizing Dismantling of Facility and Disposition of Component Parts dated July 11, 1989 and as outlined in Section 1.6, Regulations, Regulatory Guides and Standards of the Decommissioning Plan.
With the exception of removal of the reactor fuel, Chem-Nuclear' systems, Inc. performed all of the major decommissioning operations. The Licensee, Michigan State University (MSU), monitored all operations to ensure regulatory compliance. This report follows the format of the Decommissioning Plan submitted by MSU to the Nuclear Regulatory Commission in January, 1989 (Docket No. 50-294). 1.1
SUMMARY
DESCRIPTION The 250 KW(th) MSU TRIGA Mark I Reactor Facility is located on (~ the first floor of the Engineering Building on the main campus (_)) of Michigan State University. The Facility consists of the '
) reactor room (184) with adjacent laboratory (184A), the control room (room 188), the reactor supervisor's office (room 190) and a counting laboratory (room 192). Figure 1.1A illustrates this layout within the Engineering Building. A view of the Engineering Buildinc floor plan and relative location of the l
Building on the MSU vampus is provided in Figures 1.1B & 1.1C. l 1.1.1 Reactor The MSU TRIGA (Training Besearch and Isotopes general Atomic) Mark I was a General Atomic heterogenous pool-type reactor. As shown in Figure 1.1D, this reactor was ! extremely simple in physical construction. The reactor core and reflector assembly were located near the bottom ( of a 1/4 in, thick aluminum tank that had an outside l diameter of 6.5 ft and was 24 ft 8 in, deep. The tank was installed below ground and surrounded externally by ordinary concrete (see Figure 1.1E). l tO V 1 l l
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l~ The MSU TRIGA Reactor was designed for non-pulsing
. ('~~)' operation at 3 power level '.sp to and including 250 KW and i
! for ' pulsed operation of up to 250 MW. It contained approximately _ D r.a of D.0% enriched uranium-zirconium-hydride (U-ZrH,; '1RIGA-type fuel'. For reactor start-up, i a 1.9 Ci AmBe neutron source installed in a holder with the same size and shape as a fuel element.was used. To provide for cooling of the Reactor during operations, pool' water was circulated by a primary cooling system thru the tuba' side of a heat exchanger (mounted on the-north l wall of the reactor room) .- Heat was then transferred via the secondary water supply system to a cooling tower { located within a ventilation penthouse on the roof of the reactor building. Experimental Facilities installed at the MSU reactor .j facility included a rotary specimen rack and a central thimble. By design, no provisions were made in the MSU TRIGA Mark I reactor biological shield for any horizontal j beamports or thermalizing columns (the lower tank region j (' ( is underground and inaccessible). Because of this simplistic design, dismantlement and removal of activated component materials and biological shield was relatively straightforward. 1.1.2 Support Facilities
- a. ,qontrol and Instrumentation System 1
1 The MSU reactor facility used a control system similar in design to other TRIGA reactors. The major l l-components of the reactor control and instrumentation l system, including pen recorders, meters, rod controls, and annunciators were located in the control console in the control room.
'It should be noted that all of the reactor fuel was shipped off-site to an approved recipient prior to the commencement of on-site decommissioning operations.
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- b. Ventilation System ,I
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i /- The ventilation system for the reactor facility consists primarily of a roof-mounted ' exhaust fan (connected to a laboratory fume hood in room 184A), pneumatically operated flow dampers (mounted in the - ' walls of the reactor room), and an air- l circulating / heating unit (also located in room 184).
~
These components operate together to provide a slightly ' negative pressure in the reactor facility with respect i to the rest of the Engineering building.- By doing this, direct control of the air flow intc/out-of the facility as well as emergency isolation capabilities ' are maintained locally.
- c. Radioisotone Storaae Wells Located within the concrete bio-shield structure, adjacent to the reactor tank are three (3) 10-in.
diameter, 16-ft deep steel-lined storage wells. The actual location of these wells is shown in Figure 1.1E
/3 (cutaway view of reactor tank and biological shield).
Q NOTE: For additional information regarding the reactor facility, components, and supporting equipment, please refer - to the , Decommissioning Plan (Section 1.1). 1.2 FACILITY OPERATING HISTORY Michigan State University was issued Facility License No. R-114 to operate their TRIGA Mark I reactor on March 21, 1969. The facility served as a source of ionizing radiation and neutrons for various research programs and as a tool for training students in the principles of reactor operation. These activities generally caused activation to the core components and surrounding structures only. Other activities at the facility generally caused activation of solid or encapsulated materials without effects on the reactor room environment. During the operating history of the MSU TRIGA reactor facility, there were no major radiological accidents and/or occurrences. 3
V' . t a Operation of the reactor was terminated in October, 1987. The
; reactor fuel was shipped off-site ~during the months of May and
'}
June, 1989 and the License was modified to a " possession-only" status on June 7, 1989. The NRC Order authorizing dismantling and disposition of ' component parts was issued to MSU on July 11, 1989. 1.3 CURRENT RADIOLOGICAL STATUS OF FACILITY The MSU reactor facility had been operated in an extremely clean radiological fashion. This was confirmed by the results of routine radiological surveys conducted in and around the facility. In most cases, even inside the reactor room, I radiological' surveillance data was statistically indistinguishable from background measurements. . During the dismantlement phase, only minor levels of surface contamination were encountered. Surveys of even the in-core reactor' components generally revealed contamination levels less than the limits for free release. This condition not only
'~
/ simplified dismantlement operations, it ultimately reduced the V] overall decontamination requirements for the Facility.
Maintenance of this clean radiological condition was continued through the concrete demolition phase. Because of the low specific activity found even in the most activated portions of the shield, only minor levels of airborne and/or loose surface contamination were detected during concrete breaking operations.
.Again, this aided in preserving the clean radiological status of .
the Facility. Overall, the dismantlement / decontamination (demolition) operations were successful in removing the radioactive portions from all areas of the Facility. The termination radiation survey of the Facility has been performed and all results' indicate compliance with the release criteria for unrestricted use (see Section 8.0 of this report). O V 4
6 J 7 1.4 DECOMMISSIONING ALTERNATIVE 8 l' MSU selected the DECON alternative for decommissioning. This option required the immediate removal of all radioactive material ; down to residual levels in order to permit release for unrestricted use. To accomplish this, the following tasks were performed: i 1.4.1 Removal of all radj.oactive materials, neutron activated components and structures to meet the acceptance criteria of 5 uRoentgen/hr above background at 1 meter from the - surface. 1.4.2 Decontamination of equipment, components and surfaces ; within the limits set forth in Regulatory Guide 1.86 for acceptable residual surface contamination levels for Unrestricted use, i 1.4.3 Removal of all non-radioactive reactor components and associated reactor systems.
\ 1.4.4 Shipping off-site the radioactive materials resulting C- from decontamination and decommissioning of the facility.
Future activities at the reactor facility will no longer require an NRC license. 1.5 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES
~
The MSU decommissioning organization utilized personnel with experience in reactor operation and radioactive material handling. A contractor, Chem-Nuclear Systems, Inc. , was employed to perform all of the major tasks of the decommissioning. The Organizational Chart for the MSU reactor Decommissioning Project is shown in Figure 1.5. Personnel experienced in reactor operation and radioactive material handling were responsible for all dismantling operations. . lO 'V 5
i 1.6 REGUIATIONS. REGUIATORY GUIDES AND STANDARDS j 1.6.1 General. ; The dismantling operations, radioactive material shipments, and radiological surveys were conducted in > accordance with the relevant Federal and State regulations, regulatory guides and standards associated with nuclear research reactor dismantling. A complete ; listing and discussion of these documents is provided in Section 1.6 of the Decommissioning Plhn. During the on-site operations, the MSU staff acted in an , , oversight and advisory capacity to the subcontractor to ensure full compliance with Federal and State regulations as well as MSU policies and procedures. 1.6.2 Release Criteria The acceptance criteria for unrestricted use (release of the Facility is that published by The Standardization and Special Projects Branch, Division of Licensing (Office of (j Nuclear reactor Regulation), of the United States Nuclear Regulatory Commission. Limits are as follows: o 5 uRoentgen/hr above background at 1 meter from the surface or 10 mrem /yr ebove background, concidering reasonable proximity and occupancy'; and, , o Permissible surface contamination levels given in Table I of NRC Regulatory Guide 1.86. A copy of this information and Table I of Regulatory Guide 1.86 is provided as Figure 1.6. As indicated by the termination radiation survey results,
'The alternate criteria of 10 mrem /yr exposure with reasonable occupancy and proximity considerations was determined to be inapplicable for this Facility.
All areas were required to be in compliance with the 5 uR/hr 91 meter limit. Refer to the Decommissioning Plan and NRC Clarification request / Licensee response documer.tation for further information. O V 6 <
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the dismantlement / decontamination operations were ('] successful in rendering the facility in compliance with ;
'N~J the limits au specified above.
1.7 TRAINING AND OUALIFICATIONS A specific training and qualificatiori program' for on-site personnel involved with the MSU TRIGA reactor decommissioning was developed by the contractor and reviewed / approved by MSU. This program was administered by the project Radiological Control Supervisor (RCS). All work activities were directly supervised by the contractor's on-site Radiation Control Organization and monitored by the MSU office of Chemical, Radiological, and Biological Safety to ensure compliance with NRC and MSU policies. l O l l l 7
2.0 OCCUPATIONAL AND RADIATION PROTECTION PROGRAMS f~ - Y.) , The occupational and radiation protection programs and standards utilized by the Facility during the dismantling operations are outlined in the Decommissioning Plan- (Section 2.0). These - Programs proved to be effective in providing a safe work-place as well as maintaining personnel exposure ALARA throughout all on-site-project operations. There were no personnel injuries, incidents, or unauthorized releases of radioactivity to the environment. Table 2.0A provides a breakdown and summary of the contractor's ' man-rem budget for the decommissioning project. .The overall total of less than 3 man-rem compares favorably with the original estimate of 9.05 man-rem which was presented in the Decommissioning Plan (Section 1.1.7). A statistical summary report of the personnel monitoring information for all project personnel in the format prescribed by 10 CFR 20.408 is shown in the following table (Table 2.0B). f} NJ TABLE 2.0B NUMBER OF l INDIVIDUALS IN WHOLE BODY EXPOSURE RANGE (REMS) EACH RANGE No Measurable Exposure 1 , Measurable Exposure less than 0.1 6* l l 0.1 to 0.25 7 l 0.25 to 0.5 3 0.5 or greater 0 .
- MSU Persomel accounted for (5) of these individuals
--- ---- ---- ----------------------~~~~~~~-~~-
TABLE 2.0 c MSU TRIGA ReactorDemcommissioning Project - COLLECTIVEDOSE
SUMMARY
, (for Contractor Personnel, Expressed in Man-Rem)
/ \
L ;} 0l0READINGSIREnG TLOM REnQ , EXTREntmES(REhQ
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WEEK AfAJOR TASK (S) staN 9tgw TOTAL Bag Gr.gw TOTAL L. HAND' R HAND TOTAL 1 Site Set-up and Baseline Radiological Survey Bkg Bkg Bkg 2 Dismantlement of Reactor Core Components 0.32 0.16 0,47 (including the Rotary Specimen Rack) 3 Completed removing Reactor Core Components " 0.52 0.68 1.20 (Reflector & Stand); removed Neutron Start up Source 4 Removed Aluminum Tank Liner (lower region of tank 0.04 0.06 0.10 only); began Concrete &cavation Operations
. . SUBTOTAL /11.0 REACINGS * - 0 88; 0 90 : 1._77 a712 0.654 '1'366 .
iC.406{ 041]G815 5 Concrete &cavation Operations: Reactor Component 0.30 0.18 0.48 Packaging (for ATU) t 6 Concrete &cavation Operations: Made Shipment of 0.40 0.25 0.64 . Reactor Components to ATU 7 Concrete Scavation; Removed Steel Piling from 0.00 0.01 0.0's Lower Tank Region 8 Completed &cavation In Reactor Pit: Demobilized 0.00 0.00 0.00 3 SUBTOTAL /11.D READfNGS * - G70: 043 ~ :1.12 0602 {b.340 .942 ~ !N/Rl lN/RN 'N/R GRAND TOTAL = 1.57- 1.32 2.89 1.314 0.994 2.300 0.405 0410 G815 NOTES: * - TLD'S WERE PROCESSED MONTHLY
- N/R' - Not Required v * * - the majority of the does received during week (3) was due to the hmited available area to store the high activity reactor componente in the Facility.
WHOLE-l.aODY EXPOSURE TO PERSONNEL 1.5 1.4 1.3 1.2 < i.i 1.0 g i f ::: m e g
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M i W i P E #, iSI M i M i i 1 2 3 4 5 6 7 0 h PROJECT WEEK (ON-SITC) ( Starr Personnel M . Crew 9
2.3 CONTRACTOR ASSISTANCE a 6
# Michigan State University contracted chem-Nuclear Systems, Inc. , j";.
to perform all of the major tasks associated with the reactor
.,- decommissioning project. This included the following:
2.3.1 Preparation and follow-up of documents necessary to amend (R-114) the MSU reactor operating license to a
" possession-only" status; 2.3.2 Preparation of the Decommissioning Plan; follow-up as ,
necessary to secure an NRC Dismantlement Order; , 2.3.3' Controlled dismantlement of the reactor and all associated systems; packaging and shipping of the same to' Arkansas Tech University; 2.3.4 Decontamination / demolition operations necessary to release the Facility for unrestricted use; l 2.3.5 Shipping of the radioactive material; l l p)
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2.3.6 Preparation of the Final Report; and, 2.3.7 Backfilling of the empty reactor pit and basic Refurbishment Activities (yet to be accomplished). 2.4 COST ESTIMATE AND FUNDING i The total program cost of the decommissioning operation was l estimated to be less than $500,000; contracted work was estimated l to be less than $400,000. Actual costs to date are within these l- estimates. i 1 5 - 10 1 1 l'
. - ,~ __ , _ . . _ _ . . - _ _ . _ _ , , . _ _ . _ . . . .- . _ __.. -
l 4 1 3.0 DISMANTLING AND DECONTAMINATION TASKS AND SCHEDULES The MSU dismantling'and decontamination operations included those tasks required to remove all radioactive components and materials from the defueled reactor Facility such that at the time of the termination radiation survey, all areas of the Facility would be within the limits' for release as specified in the decommissioning
. plan (and Section 1.6 of this report).
3.1 TASKS c To prepare the site and facility for the actual dismantling and ' decontamination operations, a number of tasks and activities were performed. Included were the following: o Set up of radiological equipment, instruments, and control systems o Establishment of work zones with barriers and mar'kers as appropriate. I V o Implementation of administrative procedures and training program (s) in accordance with the. occupational and Radiological Protection Program l. ! o Bioassays and physical examinations of all on-site workers - l l 3.1.1 Baseline Radiolocical Survev In addition to the tasks listed above, a Baseline Radiological Survey was performed to fully characterize the initial radiological condition (s) within the reactor facility. These surveys included measurements for both fixed and loose beta-gamma surface contamination along with general area gamma dose-rate surveys. To confirm the absence of alpha contamination, cursory surveys (direct readings and smears) of random points were also conducted. O V 11
In order to properly evaluate gamma exposure-rate survey l ') results with respect to the 5-uR/hr-above-background
'/ release criteria (see Section 1.6.1), it was necessary to establish a benchmark for the background exposure rate for
-the general locale. Measurements were taken in buildings and outdoor areas within an' approximate quarter-mile [
radius.from the' reactor Facility (see Figure 1.lC) using , a Pressurized Ion Chamber (PIC) and "Iaicro-R" meter *. The ! results of this survey are shown in Table 3.1. 3.1.2 Dismantlement Operations The actual dismantlement and decontamination operations were initiated following the completion of. all prerequisite activities described above. Further details regarding these operations are provided in Section 3.3 of this report. 3.2 . SCHEDULE Decommissioning operations began immediately following the q issuance of the Dismantlement Order and Contractor mobilization. Q An activity barchart showing the various tasks and time-scaled sequence is displayed in Figure 3.2. This schedule was originally developed during the planning phase and was used to assess the project development and maintain critical path timing. Bars representing the original " baseline" schedule as well as the actual performance dates are shown. 3.3 TASK ANALYSES L l This section describes the basic approach that was followed to accomplish the most important tasks of the MSU TRIGA reactor decommissioning project. These tasks are described and analyzed below. Where the actual operations differed from that planned, l- a brief discussion concerning the change (s) is also included. l *the "PIC" used was a Reuter Stokes, Model RSS-Ill Environmental Radiation Monitor; the " micro-R" meter was Ludlum Model-19 gamma scintillometer (w/l" x 1" Nal detector). ( 12
TABLE
3.1 BACKGROUND
GAMMA EXPOSURE RATE SURVEY MEASURED VALUE(S)* ; .h. LOCATION Nat" FIC '"
)- QESCRIPTION
(' 1 Engineering bulUing hallway outside room 184 (1st floor). 9 14
' (Detector @ approx. 2M from walls and 1M from floor) 2 Engineering bulMing hallway outsUe room 136(1st floor). 7 10 (Detector @Spprox. 2M from walls and 1M from floor) 3 Engineering building hallway outside room MR21 (Basement). 6 9 (Detector @appror. 2M from walls & 1M from floor) 4 Sidewalk at Abbot entrance and most circle drive. 5 8 (Detector @tMabon theground) 5 Front entrance of museum in staltwell to basement. 5 8 (Taken @lM from floor & approv.1.5M from walls) 6 Auditorium truck bay on north side of complet. Detector 6 9
@lM above the ground & approx.1.5M from bidg wall.
7 Auditorium west entrance on cw1 crete steps. Detector @ IM 8, 12.5 above the ground & approx. 2M from the walls ; 8 Stadium southwest entrance at entrance to sections 12 & 6 9.2
- 13. Detector @ 1M above the ground & approx. 34 from walls.
9 Athletic field, south of the Ice arena. 4 8 Detector @ 1Mabove the ground. 10 Cyclotron building; southwest sMe. Detector @ 1M above 5 9.2 the ground & appror.15M from the building. 11 Fartr; lane road (13M from road west side & 200M south of 5 8.7 p scrvice road (see map). Detector @ 1M above the ground. ~ 12 Engineering research building at ORCBS cntrance. Detector 7 10.8
@ IM above the ground & approx.1M from the building wall.
MINAMUM - 4.0 8.0 AVERAGE - i 6.1 L
- I 9.7 NOTES ,* MAXIMt#4 - 9.0 14.0
* . MEASUFED VALUES are in *microAcontgen per hour * (ufVht)
** = *Nel* values were obtained using a Ludium Modet-t9
- micro-Roentgen
- Instrument which utilises a 1* a 1* Sodium ladde detector.
*** *PIC* values were obtained being a Neuter-Stokee NOS-lit Enviroamanial manhoring instrument wMch utares a Freesurtaed lon Chamber detector.
~ BACKCROUND RADIATION (CAMMA) SURVEY -
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e MODEL-19 (Nol) 5 RSS-111 (PIC) l 13 1
s , I 3.'3.1~ Dismantlement i
. l ')' Upon completion of prerequisite activities outlined in !
Section 3.1, dismantlement of the various ' components, ! equipment, and support systems commenced. Generally, these operations progressed from the least contaminated, lowest activity components to the higher activity, more radiologically significant reactor core components.- With the simplistic design of the TRIGA Mark I reactor and low levels of contamination, these operations were able to be performed in a straightforward manner as anticipated. The-component with the highest exposure rate was the Rotary Specimen Rack which measured up to 14 R/hr -(contact) . To maintain personnel' exposure AIARA, a special shielded handling device was designed to transfer this component from the reactor pool to it's shielded shipping container. , In order to provide a reasonable approximation of the dose rates to be expected from the various reactor components and biological shield, calculations were performed during , the planning phase to estiinate the concentration of
,G radionuclides that had resulted from neutron activation C of these materials. Table 3.3A lists the major components and the exposure rates based on the neutron activation analysis versus the actual measured exposure rates.
l t l l l l - TABLE 3.3A - Exposure Rates for Various Reactor Components l l EXPOSURE RATE (01M) ITEM Calculated Actual Rotary Specimen Rack 11.2 R/hr 3 - 6 R/hr Grid Plate (top) 1.7 mR/hr 1.0 mR/hr Grid Plate (bottom) 2.2 mR/hr 1.5 mR/hr SS Assembly Bolts 0.2 mR/hr* 0.5 mR/hr*
- The exposure rate from the bolts at contact is approximately 200 ns/hr (Calculated);
300 ndt/hr Actual. 14
l q. 1 j As discussed in the Decommissioning Plan (Section 3.3.1), () all of the reactor components and supporting equipment were packaged and sh.ipped to Arkansas Tech University for re-use. A summary. list and activity estimate (s) for the radioactive materials that were transferred to ATU is
.provided in Table 3.3D. ,
3.3.2 Water DisDosal Following the removal of all components from the reactor
- tank with the exception of the Reflector & Stand assembly, the reactor pool water was thorough,1y agitated to ensure uniform suspnnsion of particulate material. A sample was obtained and analyzed -independently by a responsible, qualified laboratory' for gross alpha, g.~oss beta and tritium activity. As uhown in Table 3.3C the results indicated activity levels below the limit (s) for release to the sanitary sewer per 10 CFR 20. The pool water was discharged to the MSU sewer system via a tap (connection) in room 184A. ,
r~T
' Controls for Environmental Pollution, Santa Fe, New Mexico 15
i b ' TABLE 3.3B i
< j t a I
Michigan State University TRIGA Reactor Decommissioning Project iI A i I INVEN'lDRY OF MAJOR REACTOR COMPONENTS AND RADIOACTIVITY ESTIMATES i - %s) ; APPROX. ESTIMATED EXPOSURE RATE l (CONTACT)
~
COMPONENT WEIGHT ACTIVITY * ( ! Aluminum Clad Or.phite Reflector wh 1700 lbs 0.005 Ci 300 500 mR/hr b Aiurninum Bottom GridMare
. Aluminum Top GridMate i
. ContmlRod Guide Turnes (3 en)
Aluminum Rotary Specimen Rack 115 lbs 2.5 Ci 4 9 R/hr Neutron Detectors: 0.0303 Ci 1 15 mR/hr
. Compenanted Jan Chambers (2 en) 12 lbs(en) ,
. Uncompenanted Jon Chamber 12 lbs
. Fission Chamber 12 lbs 0.002 Ci(1) 1.10-180 inR/hr Alum. clad Graphite Refector Elements (18 ea) 4 lbs(en) 0.007 Ci 20-tio mR/hr Aluminum-clad Boron Control Rods (3 ea) 7 lbs (en) 0.001 Ci 12-18 mP1hr l
l TIIE FOLLOWING ITEMS WERE ALSO SHIPPED TO ATU AS RADIOACTIVE MATERIALI HOWEVER, m THEY CONTRIBUTED LESS THAN 15 TO THE OVERALL RADIOACTIVITY INVENTORY: , f . t L [: lon Chamber Support Ring 18 lbs N/A 2 mR/hr 1 Aluminum Central Thimble Assembly 2 lbs N/A < 1 mR/hr i Aluminum Fuel Storage Racks (4 en) 18 lbs (es) N/A < 1 mR/hr Fuel Inspection Tool whdapters 45 lbs N/A < 1 mR/hr l
- j. Fuel Handling Tool 9 lbs N/A < ! mR/hr l
Misc. Reactor Manipulator & Handling Tools 75 lbs N/A < 1 mR/hr l l Primary Water System Components 1750 lbs N/A < l mR/hr (Piping, lleat-exchanger, Dcmin. & Filter, Pump) l' * - TIIE MAJOR ISOTOPE IS Cobalt-60 l 3779 LBS 3 Ci (TOTALS) l l: b bf[$ b b S $$ E E b N A N N $ b b b d$ $ N $ $ fd E i b k N N d b N b b b b s OTilER ITEMS THAT CONTAIN OR ARE M ADE UP OF ISOTOPES OTHER TilAN Cobalt-60: I' (These items were also included in the shipment to ATU) Neutron Startup Source (AmBe) I lb l 88 Ci 3-9 R/hr Beta / Gamma (/ Fission Chamber -NOTE (1) above- (SNM) U 235 1.68 gm 3.5E46 Ci See Above l. l L 16
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MSU TRIGA Reactor Decommissioning - DISCHARGES TO THE SANITARY SEWER SYSTEM (WATER)
- ACTMTYgGQ- ' ; - WOMEDfSCNMSW-
-ESTEMATED ACTMTYMLEASEDesCV-SOWICEOFWATER SAMPLE! - DATE ^ BETA- _
Vn. _ . . - ,_ BETA-- H-3 GAMMA ALPHA
- _DATE GALLONS - UTERS H-3 ~- GAMMA' ALPHA TOTAL TAP WATER (BASEUNE) MSU-02 7/10/89 900 10 < 2 N/A WA N/A NA N/A N/A WA Sec. Water System Water MSU-01 7/10/89 < 500 8 < 2 7/14/89 2500 9463 < 5 0.06 < 0.02 < 48 Reactor Tanic Water rASU-03 7/21/89 36.327 9 < 2 7/24/89 5957 22549 819 0.20 < 0.05 < $19 4 Residual Rx Tank Water MSU-04 8/14/89 87.034 7 < 2 8/31/89 138 520 45 0.00 < 0.00 < 45.3 Ground Water from Tank Annutus MM S/23/89 59.014 361 26 9/1/89 50 189 11 0.07 0.00 11.2 Ground WatM (during D&D Ops.) MSU-07 8/29/89 2.773 69 < 6 9/1!89 250 946 3 0.07 < 0.01 < 2.7 Ground Water (aner D&D Ops.) M SU-08 9/1/89 < 500 11 4 9/4-15/89 576 2180 < 1 0.02 0.01 < 1.1 Ground Water (ater D&D Ops.) MSU-09 9/19/89 905 57 20 9/19-30/89 576 2180 2 0.12 0.04 2.1 TOTALS- 10.047 38.028
-w aea., u-em e w TOTALS (ESTJ: < 886 < 0.6 < 0.1 < as6.7
-a tJouro/M-3 www A2 to stwtw r-en eco o "
$ w g w 200 o O coo o se soo o b
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g_. h 300 o 200 o $
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o.o , _ i _ . , b, . . . M/A 7/14/89 7/24/89 8/31/89 9/1/99 9/1/29 9/a-t s/ss/19-30/e9 oATE G 3 vatte) [2RD [w-3] poA ; ; :- crv(acry g c y- w -
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c , 3.3.3 In-tank Onerations setun
/ Upon completion of tank pumping operations and removal of the Reflector & Stand assembly, specially designed in-tank scaffolding was installed. A containment enclosure .
was then erected over the reactor cavity and immediate surrounding area to preclude the spread of dust ' and - airborne radioactivity during demolition operations. Containment ventilation was established by use of a 2000 cfm HEPA-filtered exhaust blower located external to the enclosure. A program of airborne radioactivity monitoring was conducted inside the containment structure as well as downstream of the HEPA filter. As shown in Table 3.3D (below), results. of this monitoring thrpughout the demolition operations indicated levels well below the established (10 CFR 20] limits for airborne radioactivity in uncontrolled areas. TABLE 3.3D
/~T L_-]
AIRBORNE CONCENTRATION (uCi/ml) LOCATION MONITORED AVERAGE MAXIMUM Inside Containment 8.6E-11 8.6E-10 Downstream of HEPA filter 7.6E-12 4.0E-11 3.3.4 Reactor Cavity Liner Decontamination Interior surfaces of the reactor tank (cavity liner) were surveyed and found to have only minor icvels of curface contamination (i.e. <1000 dpm/100cm') . A general wipe-down was sufficient to effectively decontamiraate these i surfaces to well below release levels. rh V 18
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I 3.3.5 Radioloalcal Surveys of Accessible Tank Areas !
! 'V Surveys conducted in the reactor cavity following ; decontamination of the tank wall surfaces indicated a .
I general area' exposure rate of approximately 1 - 2 mR/hr. , i Table 3.3E compares the estimated exposure rates based on I the analysis performed during the planning phase and the actual survey results.
- TABLE 3.3E -
Exposure Rate (s) Inside the Reactor Tank (General Area, Approx. 1 Meter from Tank Bottom) TANK WALLS: Calculated Actual Concrete 0.9 mR/hr 2 mR/hr , Aluminum 0.001 mR/hr Bkg* TANK FLOOR (Includina Rebar): p)- t ( Concrete 10.0 mR/hr 0.8 mR/hr Aluminum 0.1 mR/hr Bkg*
*
- the aluninum liner materlet removed from the lower tank region was released as non-radioactive waste.
3.3.6 Excavation Recuirements To validate the predicted concrete activation profile, several key sampling locations were selected on the j activated cavity wall surface. Samples were taken at various depths and analyzed by gamma spectroscopy to determine the concentration and profile of the activation products. Results of these surveys indicated that the l walls in the lower tank region (from a depth of 1 foot below the reactor pit floor to an elevation of , approximately 8 foot) would be required to be excavated to a depth of 18 - 24 inches.
- i. ()
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e { 3.3.7 Concrete Excavation In . order - to proceed with concrete excavation it was i necessary to remove the aluminum liner in that area first. To. accomplish this, the liner was first segmented into manageable size pieces which were then pryed/ peeled from the concrete wall. The sections were then thoroughly surveyed for fixed / loose surface contamination as well as gamma exposure rate at 1 meter. All portions of the aluminum liner that were removed were found to be well below the criteria for free release and were disposed of as clean waste (see Section 6.2). . Excavation of activated concrete followed liner removal. The basic approach that was employed for concrete removal involved using a combination of non-explosive demolition compound (BRISTAR") and hydraulically powered rock drill and jackhammer equipment. Using hydraulic-powered equipment reduced the levels of noise and dust generation p- in.the confined tank area. BRISTAR" is essentially a proprietary blend of silicates and organic compounds . which, when loaded into pre-drilled holes (1.25-1.5" in dia.) in the concrete, exerts a powerful expansion force thereby splitting / fracturing the concrete mass. Once fractured, hand-held demolition tools can be used effectively to remove large quantities of concrete. l The first phase of the concrete removal plan required the l excavation of the tank base to a depth of approximately l 1 foot. The tank base concrete was tne only portion of l the lower shield structure that contained reinforcing steel (rebar). Dose-rate in the tank was reduced to about 0.5 mR/hr after retpoving this material. o
\
1 l l: ; i i L ') . In order to proceed with excavation of the tank walls,
'd it was necessary to burrow underneath the walls at the l-tank base in order to create a relief area for subsequent ,
P splitting of the material above it. A horizontal'line of I holes around the inside . circumference of the tank, 1 approximately 18" deep and at a 45* angle was then drilled and loaded'with BRISTAR". This operation was continued-in successive campaigns starting from the bottom and working up the walls, alternating with demolition and waste removal campaigns. Demolished concrete material was packaged into D.O.T.17-H 55 gallon drums and hoisted out of the tank cavity. Excevation proceeded in this fashion i until the lower tank walls (and embedded steel pilings) L we e removed to a depth of 18-24". A total. of approximately 450 cubic feet of activated concrete was removed from the shield. Some of the difficulties encountered during the concrete demolition operations included the following: o Locating and removing slightly activated steel piling [ (/ material embedded in/behind the concrete shield structure (this material was later determined to- , actually be the main contributor to the general area [ exposure rate in the reactor pit) o Continual seepage of ground water into the reactor cavity (the work area was approximately 10 feet below the water table). Mainly, this resulted in having to re-package some of the waste drums due to unacceptable quantities of water being transferred into them during loading operations. 3.4 SAFE STORAGE There is no requirement for Safe Storage since all radioactivn material has been removed from the MSU reactor Facility. O V 21
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t p 4.0 SAFEGUARDS AND PHYSICAL SECURITY {( ) f All' reactor fuel and Special Nuclear. Material has been removed l from the Facility; therefore, no Physical Security Plan is required. MSU was granted a waiver from this requirement when the license was amended to a possession only status. At this time, areas within the Facility are being controlled'in accordance with the requirements of the Technical Specifications discussed in Section 7.0 of this report. , P d f G 1
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,_ s 5.0 RADIOIDGICAL ACCIDENT ANALYSIS I x'~' /'
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No' radiological accident analysis was required since the reactor i' - fuel had been safely transported off-site prior to the' ; dismantling operations. 9 I a t-l'
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, 6.0 RA'oIOACTIVE MATERIALS AND WASTE MANAGEMENT 1,
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6.1 FUEL DISPOSAL' As indicated in the Decommissioning Plan, the. reactor fuel was removed from the MSU Facility prior to performing the dismantlement operations. This was performed under the direction of the Department of Energy and in conformance with applicable NRC and DOT Regulations. 3 6.2 RADIOACTIVE WASTE PROCESSING During the dismantlement operations, radioactive materials in liquid, solid, and particulate forms were generated. Miscellaneous non-radioactive wastes, including pipe, wood, plastic sheeting, metal (including aluminum reactor tank liner material removed from the reactor pit) and other materials which meet the established release criteria will be disposed of at a local sanitary landfill. Currently, these materials are being Q'w/ held in two specially identified / locked metal "dumpsters" located outside the Engineering building; pending final inspection by the , NRC. . 6.2.1 Licuid Radwaste In general, liquid wastes requiring processing (solidification) as radioactive waste were not generated during the decommissioning operations. Other slightly radioactive liquids that were part of the facility inventory and were required to be disposed of included the following: o secondary circulating System Water o Primary Cooling System (including reactor pool) water o Ground Water seepage into the Reactor Fit 24
Y These liquids were sampled, analyzed and released to the (a) sanitary sewage system.since, in all cases, the conditions of 10 CFR 20. 303 were . satisfied. A summary of the liquid i discharges and estimated activity released to the sanitary
. sewer system is shown in Table 3.3C.
It should be noted that currently, due to the continual seepage of groundwater into the empty reactor pit area, l l routine pumping operations to the sanitary sewer are being performed per Section 7.1.2 of this report. It will be , necessary to continue such operations until final release l of the Facility is obtained and the reactor pit can be ! backfilled. Data regarding these discharges (to date) is included in Table 3.3C. As indicated, small amounts of radioactivity are being released along with this water; however, analysis indicates that this activity is j primarily due to natural sources (Radium-i226 and l Potassium-40). A copy of Table 3.3C is being maintained ; to provide record of these discharges which, on ' the I average, amount to approximately 250 gallons per week.
, 6.2'.2 Solid Radwaste i .U]/ l The solid radioactive waste generated during l
decommissioning activities was packaged on-site in D.O.T. ) 17-H 55 gallon steel drums for shipping and disposal. A , total of approximately 450 cubic-feet of waste containing ; 0.23 millicuries and weighing approximately 20 tons was ! shipped to the Low-Level-Waste disposal site in Barnwell, SC. Inventory consisted primarily of concrete rubble and j miscellaneous compactible trash. Only Low Specific Activity, Class A waste was generated during the Decommissioning Project. As mentioned in Section 3.3.1, all of the operational reactor components and equipment (radioactive and non-radioactive) were packaged in appropriate containers and shipped to Arkansas Technical University for re-use. 25
-e#---. - ~ .- . . , _ . , . , .-
h i l j m ; f ) w/- 7.0 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS Technical Specifications for the decommissioning operations were-developed and included in Section 7.0 of the Decommissioning Plan. They replaced those specified in License R-114 under the ,
" possession-only" status and were in effect during the ;
dismantlement / decontamination operations. In the opinion of MSU and it's contractor, the dismantling and > decommissioning tasks were accomplished with no significant
' impact on the environment or. the health and safety of the public.
7.1 INTERIM TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS - Since all radioactive portions of the Facility have been removed and the termination radiation survey indicates compliance with the release criteria, only the following Technical and Environmental Specifications are necessary and are currently in , i g' effect: 7.1.1 Administrative Controls The MSU Reactor Supervisor or qualified designes. is responsible to monitor and control access to the empty reactor Facility. Security consists of locked doors and controlled-key issue system to prevent casual, inadvertent, and/or unauthorized entrance. Although there are no radiological safety concerns at this time, the open, empty reactor pit poses a potential safety hazard in that area. C g 26
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l -- 't) 7.1.2 Surveillance Reauirements l Due to the continual seepage of ground water into the empty reactor pit area, a sump-pump system has been installed. The discharge from this system is directed through a filter-canister and then into the sanitary sewer l system piping in room 184A. Although it is set up to operate automatically on a float switch, periodic ,
< monitoring of the system to ensure proper operation shall s be performed. Periodic sampling and analysis of the water t being pumped will be continued on an approximate monthly l basis to monitor activity levels and verify that these I discharges are well below the allowable limits per 10 CFR !
20.302. j i l L - l 1 i r 1 , ( 1 l 1 :
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F~ i a f i 8.0 TERMINATION _ RADIATION SURVEY PLAN Sinva the goal of decommissioning under the DECON option is to ultimately release the site for unrestricted use, a termination j radiation survey of the Facility is required to be performed in j order to demonstrate that the site satisfies the acceptable l criteria for such release as specified in Section 1.6. This i survey was conducted following the completion of all l dismantlement / decontamination (demolition) operations as described , in Section 3 of this report. 4 i As discussed in the Decommissioning Plan, a detailed methodology for the termination radiation survey was developed during the i planning phase of the project. Specific guidance as documented ! in NUREG 2082, " Monitoring for Compliance with Decommissioning j Survey criteria" was incorporated into this plan. f 8.1 TERMINATION SURVEY MPTHODOLOGY [ The reactor facility was divided into the four following types l of e.reas for data collection and analysis. 7-s ; (s/ !
- 1. Areas of low potential for surface contamination ;
i
- 2. Areas of medium potential for surface contamination
- 3. Areas of high potential for surface contamination or are known to have surface contamination present l
- 4. Areas whera surfaces / materials have become activated from ;
exposure to a neutron flux (e.g. inside the reactor tank / bio-shield). I l Historical survey information/ data obtained from each area along with an analysis of the functional use(s) served by that area were used to determine what category the area was classified as [ for survey purposes, l' k 20 i 1
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l 8.1.1 It2W Potential Areas (.s);
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Areas of low contamination potential were those areas where previous MSU survey results indicated contamination i below release limits and the function of that area was not [ conducive to creating a contamination problem. Since the ! MSU facility is contained in a relatively small area, only the ceilings were considered as low potential areas. Each ; ceiling area was sectioned into 3-meter by 3-meter grids; ; all ceiling grids (14 total) were surveyed. l i I 8.1.2 }iedium Potential Areas Areas of medium contamination potential were.those areas ; where the function of that area constituted a potential l for contamination. These areas were sectioned into 2- ! meter x 2-meter grids. Surfaces included in this description were all of the walls inside the reactor ' Facility. Approximately 30 of these grids were surveyed on a statistically random basis. ; i fm 8.1.3 Hiah Potential / Contaminated Ar_ tag Areas of high contamination potential were the reactor cavity, storage vault and pit area, and all reactor Facility floor surfaces. These areas were sectioned into , 1-meter x l-meter grids; all of these grids were surveyed. { 8.1.4 Induced Activity Areas susceptible areas where materials could contain induced activity included the aluminum reactor tank liner and surrounding concrete bio-shield structure (in the lower l section of the reactor pit, near the reactor core). All remaining surfaces of the aluminum tank liner and the entire lower (excavated) section of the reactor pit were [ sectioned into 1 meter x 1 meter grids. Additional survey measurements were required to be made in these grids (see , Section 8.2.3 - below). All grids in this areas were surveyed. 29
l 1 _ 8.1.5 Additional Grid SyryAy2 i
, i i
\
'v) In addition to the grids surveyed on a random basis in !
the medium contamination potential areas, approximately : 10 specific grids on a stratified basis were surveyed. ; These grids were in (1) doorways and room corners, (2) - near penetrations or obstructions, and (3) specific areas , which have previously indicated contaminated materials or , equipment. ! 8.2 TERMINATION SURVEY PROCEDURE (
SUMMARY
) l I 8.2.1 Grid Surveys (Fixed / Loose Surface Contamination) [ i Each designated grid was surveyed in the following manner. ' (A diagram showing a typical survey grid and the various measurements taken is provided as Figure 8.2A). j I
- a. Beta-gamma and gamma point readings were taken at five (j equally spaced points within the grid (points "a" thru "e"). l i
- b. A G-M beta-gamma survey scan of the grid was conducted !
and the maximum beta-gamma point identified and ! surveyed (point "f"). [
- c. A smear survey was taken at the maximum beta-gamma j point (point "f"). The smear was counted for gross i beta-gamma activity. [
- d. Data from each of the five point survey was then averaged and the value recorded as an unbiased l measurement. The measurements at the beta-gamma maximum point were recorded as a biased measurement. ,
An example of a typical termination radiation survey form l used to document this data is provided as Figure 8.2B. i 30 l
8.2.2 Other Survevc r \ (" / The procodure for surveying pipes, drain lines, and l ductwork was basically the same as described above. The ! nearest accessible locations were surveyed with suitable ; instruments and wiped for removable contamination as far inside as could reasonably be reached. These locations . included the ends of accessible pipes, the interior ! surfaces of inlbt t.nd outlet vents, and the water traps ! and exit points of drain lines. Since only background I levels of radioactivity were found at all entry and exit i locations, further survey (s) were not warranted, 8.2.3 Reactor Cavity Surveys (Induced Activity Areas) l In addition to the measurements performed per Section 8.2.1 for fixed / loose surface contamination in these areas, a gamma-dose rate measurements was taken at 1 meter , away from the center of each grid (perpendicular to point "a"). This survey was performed to ensure compliance with j the release criteria of 5 uR/hr above background at one ,
/G meter as specified in Section 1.6.
i i It should be noted that a " Pressurized Ion Chautber" l instrument as well as a scintillation (event) counter was ! used for the dose rate surveys conducted in the reactor ; cavity area (Section 8.1.3). ; I 8.3 RESULTS OF,,,_THE TERMINATION RADIATION SURVEY (
SUMMARY
) t All radiological surveys required by this plan have been l conducted in all areas of the MSU Retitor Facility. Due to the : large amount of data involved with this task, a computer program i was utilized to assist in compiling tne information. From this program, summary sheets were produced for each of the Facility rooms (areas) surveyed. These summary pages along with a grid ' map are presented in Appendix B to this report. The information displayed on there pages follows the grid survey plan described in the previous Sections. Additional line-item explanations are l as follows:
/^
i
\ 31
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p F ! i i
+ 1 "
SUMMARY
OF GRID SURVEY RESULTS FOR"! Identifies vhat area of I () the facility was surveyed l I
" TOTAL NUMBER OF GRIDS SURVEYED" This is the total number l of grids that were j actually surveyed in that j erea. l
" DIRECT SURVEY OF GR1D POINTS a - e" This section provides summary data from the ;
surveys conducted per ! ' Section 8.2.la.
" DIRECT SURVEY OF MAX. POINT ( f) . . . " This section provides i summary data from the '
surveys conducted per l Section 8.2.1b. , t
" SURVEY FOR LOOSE SURFACE CONTAM.... This section providos i summary data from the i survoys conducted per Section 8.2.lc. l
" SURVEY OF GAMMA EXP. RATE 9 1H..." This section provides '
summary data from the ( surveys conducted per ; ig] Section 8.2.1c. ; f As noted above an example of a raw data sheet used for these suurveys is provided as Figure 8.2A. j l 8.4 INSTRUMENTATION l All instruments used throughout the decommissioning project + l (including the Termination Radiation Survey) were calibrated in [ accordance with procedures recognized as acceptable to the NRC. ; L All calibration sources used are traceable to the National Bureau ; l of Standards. A complete list of these instruments is contained i in Figure 8.4. l An example listing of the actual instruments used during a Termination Radiation Survey along with pertinent calibration, source check, efficiency, and background data is included in l Figure 8.28. l l 32
f . . ., s I U'< ; l' ' # y,;Ui 1 l J;. l t , j i 8.5 FINAL SITE RELEASE h j At the conclusion a ~ successful Termination: Radiation Survey ) inspection conducted by the NRC, HSU plans to backfill the j reactor. tank cavity-stith concreto and refurbish the facility. l I k, ! a I. ) i. 3 I i i a
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y- - iv ( ) r<. , cetu wettan systius twe - i . k , i APPENDIX A ! i t / ; e t l i f I r
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q-f- -- ; i !' CHa u wuctsan systsws suc l 7 APPENDIX A i l' I FIGURES 3 l l .
. NO . TITLE INE to.
1.1A MSU REACTOR FACILITY LAYOUT . . . . . . . . . . . . . . A-1 1.1D MSU ENGINEERING BUILDING FLOOR PIANS. . . . . . . . . . A-2 1.1C MSU REACTOR SITE I4 CATION (CAMPUS MAP) . . . . . . . . . A-3 ! 1.1D TRIGA MARK I REACTOR. . . ............... A-4 i t V) 1.1E CUTAWAY VIEW OF REACTOR TANK AND BIO-SHIELD . . . . . . A-5 I 1.5 PROJECT ORGANIZATIONAL CHART. . . . . . . . . . . . . . A-6 i 1.6 TABLE I, NRC REG. GUIDE 1.86. . . . . . . . . . . . . . A-7 1 3.2 ACTIVITY BARCl! ART (SCl!EDULE). . . . . . . . . . . . . . A-8 . 8.2A SURVEY GRID LAYOUT / MEASUREMENTS . . . . . . . . . . . . A-9 8.2B TERMINATION RADIATION SURVEY DATA FORM . . . . . . . . . A-10
- 8.4 RADIOLOGICAL SURVEY INSTRUMENTATION . . . . . . .. . . .A-11 O as c > .
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I t FIGURE 1.5 t i i-MICHIGAN STATE UNIVERSITY ! TRIGA Reactor DeCorrunissioning Project ' 1[g f04 1AFETY C0 r,tUlf 9RGAM11AT10N l I l PatssetNT l 1
.... Paovost ANo V.r. rom ACAotMIC AFFAIRS 4 i
i V.P. Fon RestAnCH I DEAN of ENGINtgk3NG ANo ** GAAouAft $TUolts t RtActon $ArtTY I t ConMITTtt DFrset or RAo Attons. : 6 CHEM CAL ANo ** i SloLoGICAL $AFtTY ' i DIRECfot
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RActATsoN, i CutMacAL ANo BiotocacAt ! l $AftTY
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j -... NoCLEAR RcACTot LAtokAfotY I [ coNTRACKm3ORGANIZAT10ft ~~~~~~^ O r- "~~ " J f l OfSATDNS RIOULA1 DRY i MANA g AFTAIRS DIY1 SON I J l
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rr t c , , FIGURE.1.6 ; (FROM NRC REG, GUIDE 1.86): r ( ' TABLEI
\
3;I
- . V' ACCEPTABLE SURFACE CONTAMINATION LEVELS NUCLIDEs AVERAGEb c MAXIMUMbd REMOVABLEb e U-nat, U 235.U 238, and 5,000 dpm a/100 cm2 15.000 dpm a/100 cm 2 1,000 dpm a/100 cm2 associated decay products Trensuranles, Ra 226, Ra 228, 100 dpm/100 cm2 300 dpm/100 cm2 20 dpm/100 cm2 Th 230,Th 228, Pa 231, Ac 227,1 125,1 129 Th-nat. Th 232, Sr.90, 1000 dpm/100 cm2 3000 dpm/100 cm 2 200 dpm/100 cm2 Ra 223, Ra 224, U-232, 1126,1131,1133 .
Bete. gamma emitters (nuclides . 5000 dpm M/100 cm2 15,000 dpm h/100 cm2 1000 dpm M/100 cm 2 with decay modes other than alpha emission or spontaneous fission) except Sr.90 and others noted above. I
*Where surfece contamination by both alphe and beta gamma <mitting muchdes exists, the limits established for alpha and telmmna-emitting nuclides should apply independently.
bas used in this table, dpm (di Lategrations per minute) erwant the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, etnciency, and geometric factors associated with the ' instrumentation. ( ' Measurements of everage contaminant should not be averaged over more than I square meter. For objects of less surface area, the r everage should be dertved for each such object. 2
'The mar *. mum contamination level applies eo an eres of not more than 100 cm .
2
*The amount of removable eedionettve snatcrial per 100 cm of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderste pressure, and essessing the amount of radio 6:tive material on the wipe with an appropriate instrument of known efficiency. When removable contamination on objects of less surface area is deterrnined, the pertinent leveh should be reduced proportionally and the entire surface should be wiped.
1 FROM " Guidance and Discussion of Requirements for an application to terminate a Non Power Reactor Facility Operating License" paragraph 2:
~
At this time the acceptance criteria being used by the staff for unrestr.icted use are either 5/. Roentgen /hr above background at I meter from the surface or 10 mrem /yr above background, considering reason-able proximity and occupancy, and the permissible surface contamination levels given in Table.1 of Regulatory Guide 1;86. , 4
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O() A-7
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4 4 4 d 4 SU .. t _j sd U s dL . .. d .. U d srs forseme se9u. pet? ACTIVITY 17 14 12 09 De 06 .3 09 29 26 24 21 39 16 13 11 06 05 95 02 30 OCT NOW CEC m FEB mAft ern day waf Jtes Jtt AUG SEP OCT 90d 3EI JW FEB B448% AdPR APR BR 88 98 89 99 89 99 99 99 99 89 99 99 89' 99 89 90 '90 90 90 90 l I l L f1 ecce= 02 04 - N!P Dece*0*1ssip1pe P3sn *CDMELETEs h2 09 i Suppet /Coci'd. DEC Plannt4 pC04PLETEe o2 09 womere ee >,ees. mu tA's. eCentE E. 04 99 ---- , yU/pstC 1oprTwo e eCOeqILE To 04 16 ._ _ , . . , pre;. Pe), pr<ceepe14Sched eCOB@LETEe 10 4 14 - s' rep <rN CA IDNeecyce eC0ma i b4 12 --- Purch/Stee'/Srate touto.G Sag 3s 'Conectike 2o 20 & s4ECOR FIEL OF5'-SITE l 10 29 e "" Pnositel i!wees.E Bheesseye eCu mptI*Eg 09 10 ers " Staf f peoutf./tep./Trewrl eCOoU tee'. E2 29 en (4sesnt3EDeces Supperi Em 10: *ConsM.ETEe ItO S2 mr). CetP3 jflet r3-Of-Effeet) sCOD@LETEe I 10 49 Deiij. targMtl C Segre {tLOC aCos@LE'Em
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- Setup Ar4rs. Syst ems. Ei Iret. PCOBEPLETEe bg Og A A eiLESToasi: s-osE) PLara ADP8 TOYED 20 e *Trete Staf f oCtp@LETEi l10 l gio le e *Irverte-y G Bese11w tasr+ey tCC*53LE In 4 !22 26 a> DissaantIeCOrceve sta Cpersone* tS eCoup'LETEe 2 42 4 Decen.%teoceen 'imescovreCle *CCiB8t.KTE s 22 30 de Caste /S* oce C:sce9ent14Ecutol *C METEe les 30 h hAnalyaeg01 mese eptef eCDMOLETEe
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15 2 Ed = *swm as-eunit orowsn1s 54 eIILESTopE 5-es*QCT cpiapLET J )N A & f4 S 3 gn a t_tJr e e Legend- M =In prog ess C = Planned ED= Critical + = Baseline " " ' " ' " - - -
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ear chart scey: Each character represents 4 time uni t (s) aw
4
. lt-FIGURE 8.2A O(~;
1 i SURVEY GRID LAYOUT / MEASUREMENTS (TYPICAL.) ORNL-DWG. 80-17790 I - i
!= TYPICALLY ~ 2m
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; l MEASUREMENTS MADE.
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- a. - f. = Survey Points O
A-9
F' FIGURE 8.2B (Examph Fcrm) Michigan State University TRIGA Reactor Decommissioning Project TERMINATION RADIOLOGICAL SURVEY
!I >
Gnio S/ZE(circio one): SURVEYLOCA T/ON(S): REACTOR PIT (Room 184) zu x eM 3M x 3M OTHER _ _ . . TURVEYED BY B. Rigby/A. Bohacheff CAL DUE CORRECllON DETECTOR INSTRUMENT MFG /MODFL SERIAL # DATE EFJJlQlENCY FACTOR BACKGROUND Micro-R Ludlum M-19 S4806 12/30/89 N/A N/A 6 ! BETA-GAMMA (Smears) Ludlum M-2S29 _ 61535 1/8/90' O.25 4.0 43 ALPHA (Smears) - Ludlum M-2929 61535 1/8/90 0.32 3.1 0 BETA-G AMMA (Diroct) Eberline ESP 2 824 12/22/89 0.17 6.1 40 SPECIAL NOTES / REMARKS: GRID SURVEY OF POlNTS (a) - (c) & SCAN SURVEY (((SUEAR RES IiS REMARKS G LAYOUT Bota-Gamma Gamma (uR/hr) (at point F) R b e Pt. Grou cpm Net cpm dpm@/obo Gross Not Beth. Nmma I a. 36 -4 ND 6 Bkg 54 ori.. enm D e b. 48 8 48 6 Bkg 11 Not e..
# c. 76 36 218 6 Bkg 44 dem/roacms ~
e o d. 20 -20 ND 6 Bkg Alpha' _ R camms unair er u- e. 24 -16 ND 6 Bkg_ Grou cem
- a. 9 oson A VG. 41 1 S3 6 0 0 Notepm 28 Bkg not f. 20 -20 ND n/a n/a _
ND dem/immt GRID SURVEY OF PO/NTS (a) - (c) & SCAN SURVEY (f) SMEAR RESULTS REMARKS G LAYOUT Bota-Gamma Gamma (uR/hr) (at po/nt F) R b e Pt. orossepm Not erm dombrobo Gross not Octa-Gamma 1 n. 52 12 73 6 Bkg 64 o,ou cem D e b. 64 24 145 6 Bkg 11 Netcom
# c. 48 8 48 6 Bkg 44 avm/roacmr
. a d. 68 28 170 0 Bkg Alpha' ,
R comma unter et M: e. 48 8 48 6 Bkg oionscom
- a. 9 oron A VG. SG 16 97 6 0 0 Notcom 29 Bkg nor f. 48 8 48 n/a n/a ND dom /roocmr GRID SURVEY OF PO/NTS (a) - (c) & SCAN SURVEY (f) SMEAR RESULTS REMARKS O LAYOUT Beta-Gamma Gamma (uR/hr) (at point F)
R b e Pt. Grosscem Not epm dpm@/obo Gross Not_ Bota-Gamma I a. 20 -20 ND 6 Bkg 50 oroucom i D e b. 52 12 73 6 Bkg 7 Not com l # c. 64 24 145 6 Bkg 28 divn/roacmr l e o d. 36 -4 ND 6 Okg Alpha' Gamms uninv e t u: e. 32 -8 ND 6 Bkg oren com
- a. 9 oross AVG 41 1 44 6 0 0 Noterm g,0 Bkg not f. 36 -4 ND n/a n/a ND dom /rnmi ,
*
- Surwys lar Alpha contamination are required only on a sjmr-checkhurtincaroon tesis Surveyo 's Signaturo: Dato: Pago .l__ of /b
FIGURE 8 4 h i Ed Ul0 LOGICAL SUR VEY INTRUMEINTA TION AND EOUIPMENT 12E1' ,l l
'(',
v Michigan State University TRIGA Reactor DecommissioniPg Project i MFO-MODEl. No. SERIAtJ M ,DATEL APPLICATION ($1 l- INSTRUMl?NTS: l Eberline E 120B w/HP 210T 11944 6/30/89 Beta gamma Surface Contamination Eberline RM 20 w/HP-260 1892 6/22/89 Personnel Monitoring (Beta-samma) I Eberline kM 20 w/ IIP 260 1895 6/22/89 Personnel Monitoring (Beta samma) i i D>erline ESP 2 824 6/22/89 Various (see Di'TECTORS below) Eberline ESP 2 323 6/22/89 Various (see DETECTORS below) l Ludlum )8 2929 w/4310 61535 7/8/89 $mear & Airborne Sample Counting i Eberline MS 2 w/HP 210T 925 7/7/89 Scalor; ut up for Beta-samma counting Low level Gamma Radiation Surveys I Ludlurn M 2220 w/4410 48403 6/30/89 Eberline E 520 w/HP-270 2081 6/30/89 Med. Range Betacamma Radiation Surveys Ludlum MODEL 19 54806 6/30/89 Low Range Gamma Radiation Surveys = Reuteratokes PSS Ill V 4032 7/20/89 Environmental Radiation, Monitor (gamena) Victoreen 450 P 451 5/15/89 Mod High Range Garama Radiation Surveys
]
Eberline PNR 4/NRD 4122 $/8/89 Neutron Radiation Survey lastrunwnt 1 Eberline AMS3 1962 5/12/89 Continuous Beta Osmma Altborne Monitor f
~
I DliTE'CTORS: (UsED WITIn 0 D erline HP 210T 15192 E 120E Thin window OM Detector w/Tungstun Shield Eberline HP 260 707534 RM 20 #1992 Thin-window OM Detector w/o Shield 'l Eberline HP 260 707535 RM 20 #1895 Thin-window OM Deector w/o Shield g Eberline HP 210L 707528 ESP 2 M24 Thin-window OM Detector w/ Lead Shield ( Eber?ine llP 270 707527 ESP 2 M24 Medium Range OM Detector j l Eberline HP 290 707529 ESP 2 M24 High Range OM Detector ; Eberline AC4 7 0057 ESP 2 M24 Eberline ; l Eberline HP 210L 707531 ESP 2 #823 Thin window OM Detector w/L4nd Shield i Eberline HP 270 707530 ESP 2 #823 Mediurn Ranle OM Detector l 70742 E$P 2 H23 High Range GM Detector l Eberline HP 290 Eberline AC 3 7 0040 ESP 2 H23 Dierline Ludlum 4310 PROS $742 M 2929 Simidianeous Alpha,Incta,Oarnma canting Dierline HP 210T 15192 MS2 Thin window OM Detector wrrungstun Shield Decrline SH 4A N/A MS2 sample Holder Drawer (amears, air samples) Ludlum 44 10 5708 M 2220 2' x 2' flat scintillation type detector Dierline HP 270 N/A E 520 Mediurn Range OM Detector l SOURCl?S/ STANDARDS: j Eberline Cs 137 C6405 N/A Check $ource for response checks Eberline Co 137 410 N/A Chuk Source for response checks Dierline Sr/Y 90 352/88 5/4/89 Source Standard for Scaler Checks I;berline Th 230 12066 5/13/89 Source Standard for Scaler ChmLs AflSCl?LI ANI?OUS: Eberline RAS l Pump N/A N/A Air $ ample Pump for low Vol. Air $nmples f s ataplex High Vol 169990 N/A Air $ ampler for high-Volume Air Samples i 0 200mR 4/7/89 self Reading Personal Dosimeters ] ( 11cndia Dosimeters Dosimeter Charger 909 104585 N/A Charger for Scif Reading Dosimeters l l Realistic Soued Level Meter 33-2050 N/A Audiometric Monitoring Device A-11 J
1 CHEM NUCLE AR SYSif WS INC. p --s
%_/
APPENDIX B 1 1 -. i 1 l l 1 1 I l l O x ! c > l
( l cm u weu a 6vsn us we
.l It i l
f APPilNDIX 11 I t I (. ( TliRMINATION R ADigrION SLIRVEY DATA FORMS I i i i-AREA No. of PAGFS ROOh! 184A (Laboratory) 3 l ROOhi 184 (Reactor Room) 3 ,
) REACTOR PIT (in Room 184) 3
, t h1A TERIAL STORAGE / PIT AREA (also in room 184) 3 i ROOh! JSS (CONTROL ROOhf) 3 ; ROOh1190 (ODice) 3 i ROOh1192 (llcalth-Physics Lab.) 3 ; 2) t C 37 P
E : TERMNA TION MADIA TKW SURVEY Aar she MSU TRIGA Ranctor C+x-;:'::':--a^.g Pro}ect > I
SUMMARY
OF GRID SURVEY RESULTS FOR. TOTAL NUMBER OF GRIDS SURVEYED. Ch MOOM 104A (Lab 0fatory) N ; Q. POINTSl'n' 't'): Total Number of Grid Points 'a' 'f' = 156 - 4 USING A THIN-WINDOW BETA-GAMMA DETECTOR (Eberline HP-210T) Awage Value - 97.3 Maximum Value-. ;304 N0t#tr@ robe *(Beta-Gammg Direct Survey of Max. Point ('I') as identified during Scan Survey .. , Antage Value - 100.5 Maximum Value- 304, N0t optr@fobe * (Beta-Gamma) arsutts or ocu-owwa coivr ev=vcv i o.. - c.e - o., - o,e - { o.s -
. 4 c.4 - ,, , , .
o.3 - p ,
'7 Y ' '
o,a - Y ' i, Y ' E e' * [. '
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g , , , , v' r
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,_ ..?" .. im _7 .
?_?' _ I _
es as 47.s o so si sa os o es se av se as so si sa ao si es s. 7s vs a 7 9 (colNvs . ) USING A GAMMA SCINTILLOMETER (Ludlum Modet.19 wf2*x2' Nat detectort Amage Value - 0.5 Maximum Value - 4 NotvMr(gamm$ , u.va. or o== ackwr .v wv , se - 13 - I is - si- t [ io - 3 7-G O- s
.g .-
4- N 7 g- N lM N 3- Ik Imet W N N M
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NOTES:
* - f0f dif0Cl Sutwys, the units dpm/' probe
- were used10 indicate that the actualSurfaC0 stes that was Surveyed is equalto the SurfaC0 ares Of the detsctOf itself(f0f the HP-210T this is about 19 Cm2).
l l TEMMNA TMN MADMTMN SUMWY not Nee MSU TMNEA Meector C+:{ :;'::': -4,2 MJ}ect
~
SUMMARY
OF GRID SURVEY RESULTS FOR: > (j~~'s ( ROOM 194A(LaboratorA (Continued) U SURWY R3R LOCGE SURFACE CQNTAMINA TDN A T MAX POINT (*t*h: TOTAL NUMSER OF SMEARS TAMEN - - as
- (Total No. of Smears Analysed for Alpha)- 6-Average Value - 26.4 bamimum Value - '106.6 dpm/100cm3(Seta-gamma) [
0.0 0.0 dpm/100cm2(alpha') umet suwer cowvam mois em o.e t ce o., o.s o.s o.+ o.3 o.2 - ll ll ir
' II II '!
AVG.(e/g), , ,,
*s e v .s e oo si u n se u w st u n eo si a u si n u ts te e t O DETA. GAMMA M
( W RE M TS
%/
SURMEY OF GAMMA EXPOSURE RA TE A T(1) METER ABOVE GRID CENTER (Point "a')': TOTAL No. OF GRIDS REQUIRING THIS SURVEY - 0 Average Value - N/A Maximum Value - N/A uR/hr @ 1 Meter t 5 f i NOTE (S):
- surv.ye ior Aich. ooni. min.iion .. ui,.d on,y on e egoi <h.co rinc.iion t ..
" - Th. Gemme .mposure rate 91 M surv.y wee e.quir.d only in thoes ar.as with potential for induc.d activity
. . - - ~ . . , . - , - . - - - . . . _ . . . - - - . - - - , . - . _ . - . . . . . - - . - - -
C
^
4 O '
, _ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... , T 4
. .. t l .? . .
- ROOM 184A - -
si i .: i
'61-59 :v. _ : : 60 ; _ .
j . i _
- 4
- 1 . ,
j . .m - i E.ar .a.b = w.4. l l l . i
. 62 -
63 64 !
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a -. . -
; '~l . . .y : j
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.- [
- .75 176 ~ : i "3 ^ '
- 9 :
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66 '
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- : '45 4E. F50l_ ' 65 -- '" ' .. 51 .,
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,e j F53
. ... ....s.
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= F52 ;F55U.[.F56 .- ---T .; .
- 7 I :43 .. ;g 41 c. .
- I . !' -
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77 67 78 .
._.....-...........i........... * ,
. . . A:w.^ k daish:J.
- 6 . .
- Ce.i lina o 69 :
70 : 7 '. :_ s e, - 7 .
- e. . . c e r s . .
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. .s ..- t,
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fggg, ,7 f 8 I..' - l, [ s,z 7 : E- -
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e 8
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i
-------SMx JM Ch7DS . : s, ' ^ = -
e I a 6 . . r .-
- 72 -
73 74 CE >. ; ; I m FDZt7Fhf . 2 i
~: -
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e : ' 'C77
=
} ====m am7sroaMS) ' - -
- - - T , , ..-
- p ;. . ~; ; ;
g _-__ _. .-.. - .... . ' . ,g _. ___ _ m.__- -
.- m.._ _m ..
Amrar zur s/is~ l 5 l i _ _._ _ - - _ . _ . . _ . _ - . . . _ . _ . . . . . _ . . _ . _ _ _ _ _ _ __ .~
t TERMNA TN3N MAOM DON SURVEY Av she MSU TMNRA Mancew C+:-:- .:'::':-;.%g Proniect : i SUMMARf OF GRID BURVEY RFSULTS FOR. TOTAL NUMBER OF GRIDS SURVEYED: ;
^XM IN tReectori??-5) ?$ b 4
DIRECTSURVEYB(CONTACT) ATGRIO P0lNTS(*n* *f*1: > Total Number of Grid Points 'a' 'l' - 427
^
USING A THIN-WINDOW BETA-GAMMA 'M!TECTOR tEberline HP-2107) . j Awage Value- _ _ . 76.4 _ _
. Maximum Value - 479 Not4nr@ robe *(Beta-Gamm4_. !
Direct Survey of Max. Point (*f') as identified during Scan Survey 4 , Awage Value - . N.9 l Maximum Value- Sti Not4wr& robe * (Beta-Gamm4 i ,
- m. ., .r, we .- ,
, f
..e -
y '
^ ...- i i
o ,' ,
, e o i p ,'.' / ', 5 ,
r' "v
[ . i
, ,q T '9;,i r
're
[ ,, vi I '
"i g,
fr , , , , ,, ri,' i Tr'
.. Pi t 9r" "' , ' ,
I' . * ,, , M7 , , , h'r < r'k ' [*4C . ,rev.tus w of .r ; ,_ ra r.-- Z o..,,nnn.,......n............,,,,m..........,. ma, . l v ,. - e4 , USING A GAMMA SCINTILLOMETER (Ludlum Model-19 w/2'a2*. Nal detectori
. Awspe Value . 1.0 ^. Maximum Value - S NetuRhr(gamm$^ ,
ammA1. er e sena oesse .6sserv
,e - [
,. =
,, - j
.. =
E ..
. - A 1B(
a- In M t 3 e. Ig M M 3 M 4m aE as M M Il 3gg am In . 3 se s. n r := rr =- -
= = - - -
- r =-
u . . ,, u ,. n .. = = .,. . u a .. . . - ..... . .. n u ,,.. .. ...
= . . ~- . ui i NOTES:
. For direct sunsys, the units dpm/' probe
- were used to indicate that the actualsurface area that was sun *M is equal to the surface area of the detector itself(for the HP-210TIMs is about 19 cm2).
l l
TEnuwA n0N MActA rtw SURVEY nw on usu TnhaA neeenor C+:-:- - '- ':--:g Protect i
SUMMARY
OF GRID SURVEY RESULTS FOR: { ROOM 144(Reactor rcom) (Continued) BURVEY FOR LCOGE BURFACE CONTAMINADON AT MAX. PONTf*f*1: i
; TOTAL NUMBER OF SMEARS TAKEN - ,
78 (Total No. of Smears Analyzed for Alpha)- ' 14 Avera0e Value - ~ 10.4 Maximum Value - 76.0 dpm/100cm2(Seta-gamma) ;
; 0.0 <
0.0 dpm/100cm2(alpha')
- i. mass ames : - . time assev i
4. t 4.
- 4. '
)
Ei O O O O '
, h _
n .,n e n ., o. m _ - ,, .m 4 e n u a n , n n n n , u = n w ei o n , ..n .m mm . u ii. . i is u , 4
. o == = === '"* " * * " !! - amu =
SURVEY OF GAMMA EXPOSURE RA TE A T(1) METER ABOVE GRID CENTER (Point "a *)'* , TOTAL No. OF GRIDS REQUIRING THIS SURVEY - 0 Average Value - N/A Maximum Value - N/A. ufVhr @ 1 Motor l
)
l \
\
l w NOTE (S):
. - s. y. io, Aiph. coni. min.iion ,. ..a.ir.o ani, on . . poi-ch.exiv.,ine. tion d..i..
l l . . - m .. .. ,. ..,. . m , ... . . . ,, i. is . , . .,in -. i.i ... ...,, l l l l 1 - - - - - - - - , -
i O . , O ' O l: I- I' ' B_QQUA l gu .g ' 3 ; ; i
'. 'C 8 '
C-- 9 l -~ l -
- 79 : 80 : 81 : .
' :83' i.
.- : C 10- - a . .
a- , a
.1
, -.4,._____ , -+.
=. o
..- -, . :gu a i l- s. . ;- - S- g ;
, ., s
- . :c i.
x .;
+;;
i l- ~ i' !
-C 11. - C 12 : -
i
~ i.
; 84 85 e "- s; i ,-
- 88
+ , :
f
~
i
; .$, l. ~ . {. r- - ._ l emmama
<- -e . i e -
. , m -
, e u -
-a ._ ; .
-4 E=i.
~
~ ,; ;
107
; 108 i F63 i
3 t-y .- cl 89 Ih 1. I I F72- - t wj I~- 4 C 13- C 14 : - - . - - - - . - l~~ l ul - E m.,w. f
.. ' l. . .
-- g ' '
ll i i:
! Fat 1 -
- s i* Li
; *! 109 i'110 . ' I '- F I --A ;
j
.c_ _. ______,- w___'
. . e
;; ' -!r . T w
+F95 i - ' 91 J - 92 -
I (Ceiling)
'f97 . . .
,i j
. ._.s. . : i.c.
L.W I
- i ! ~-
95 i 96 ! i : -- 111. ;;, 112
- O- t . .
i
' < a
-(192 l
.! . .i _, M.5 ely_ ;> .;t ,i - <!
r i r. . ._. ss _ i . 1_ .. j SFili , ' 97. <l, . gs L !.
~
^;
- i E
)I __ i-t ,-t , , , _ . -
.{- l j 113 : 114 I I4 - i ; a.L_._41m..ma
" l a I
( , l___ '
- :l. --t 119 [
- LE LTR* -
i
\ -
---/Mr IMGRIDS :
j ;
~&
j ______..a ,ygggay : 99 j1 [ 02 '
. f 'j
-- ; y
- --------Jus Js czar : ; +
- 93 ~E I
i' mfys) :- N EWF6) : J * -
? !
====r MIvRTBEAsh) . : : : ; ;
i csm svarirs 103 - 104 - 105 - 106 -!2 - i 94 : I -
.....x.- n._.....---.n_+w.a=
,,,,,, g, ,p , , .
b (Tei,*W
- M b A
.- . . . . . - - . . . . . ~ . - . . . ~ - . .._ _ ._.__ _. - _..___ _ _ ..- _ _...- -,.-_ ____._- ___ -
c , TERMINATION RADIA TION SURVEY 16r the MSU TRet9A Reactor Decommisekning Pro}ect l J
SUMMARY
OF GRID SURVEY RESULTS FOR: TOTAL NLWBER OF GRIDS SURVEYED. l REACTORPfTtRoom 104) 62 DIRECT SURVEYS (CONTACT) AT GRID POINTS l'a* *f*h TotalNumber of Grid Points'a' 'l' = 266 USING A THIN-WINDOW dETA-GAMMA DETECTOR (Eberline HP-2iDT1 '
' Average Value - 3CL7 l ,
Marimum Value a : SN Netopaprobe*(Beta Gamma); Direct Survey of Max. Point ('l') as identified durin0 Scan Survey , b 1 Average Value - 67,3_ ~ _ Maximum Value- NT, Net c>m4vobe* (Beta-Gamma) RESULTS Olr sETA-eAMwA 'PoodP 96fWEY 1 c.e - 0.8
- 0,5 -
0.4 - 0.5 -
.- g 0.4
- T
- Ds-F P T
- o. -
m ., , , [ pJ *'- (ag.) _u , P .,
,4,. .
P.U.,_.-u,., a.., s so s 34 se se so 4 +4 es 4a no as 04 as as so sa se os as to 72 74 vs te l v (roonSs.N) USING A GAMMA SClNTILLOMETER (Ludlum Model-19 w/2*x2' Nat detectork Averagw Value - 1,3 Maximum Value - $ NetUR/hr(gamma) neau, on eau a w se 14 - 13
- 13 -
i.- f so - e-
- y. .-
h y.
.jl .- =
.- -- a i- -- - --i
.- N N N M MM M
$^% I. m-et so Nna MWemai- Im o
n ao 3: 3. m u .o . .. .. .. o a u u u .o a u a u to n v. v. n
- c~, .".'T ' 2^>""
i
' % NOTES:
* - For direct surveys, the units dpm/* probe
- were used to indicate that the actualsurface area that was surveyed is equalto the surface area of the detector itself(for the HP-210T this is about 19 cm2).
$ .,.p-.--+w,- -
, - - . --.i.._, - p. , w ,g.y. ,, - . , , . - - - . g,m., y ----,,.--,-,,-m - - - - , - - - . -- , ,,.--, .. -,
. 1 l
i TERMINA TION RADIA TION SURVEY hw the MSU TRIGA Reactor Decommissioning Pro}ect j i
) $( CT PfT( 84) (CoretMuod)
SURVEYFOR LOCSESURFACECONTAMINATION ATMAX POINT /*f"): ' TOTAL NUM8ER OF SMEARS TAIGN - l 52 - j
' (Total No. of Smears Analyzed for Alpha)- ' .12 AverageValue .. 88.0 dpm/100cm2(Beta-gamma) d
, 14.0? . Maximum Value -
- 0.0, 0.0 dpm/tooem2(atha')-
moor warace ca c= now sumsy . 5 ; o.e o.e U.7 , oS R o.s o.4 c.s o.2 . o.1 D ii lio n ,11 o o O sio 110 l AWB$ --. _.._' _,, , l 28 30 32 34 34 3e en 42 4,46 48 to 82 64 to 84 to 62 64 e6 88 7o 72 74 76 7e I D at7a-omuA nesEs " " I" " Y 'd.pHa mesut7s \ SJRVEY OF GAMMA EXPOSURE RA TE AT(1) METER ABOVE GRID CENTER (Point 'n'l* *: TOTAL No. OF GRIDS REQUIRING THIS SURVEY - - 52 Averspo Value - 1.1 Maximum Va ue - 3 uR/hr @ 1 Meter l l a=ua n.oianow sumcv .iu amove once L I l l 4 l f k 4 ca9 _.. _.. ._ _.. _. _ l Le 3o 32 34 se Je so 42 44 46 4e to 62 64 Se as so 62 64 ee to 7o 72 74 74 78 9 (see, emove t. ef
'V l NOTE (S):
l ' - Surveye for Alphe contamination were required only on a spot-checkherincation basis.
* * . The Gamme exposure rate $1 M eurvey wee required only in those areas with potential for induced actMty 1-l l... . . _ . . - - ._ _ . - - - . . _, _ _ _ _ _ - . . _ _ _ .
. - - - - - - _ n. .
ri. - % i T
! U - \. '-
1Hx JN GRmT i i CXJB SUEVa72ZP 1, 8
- Reactor Pit Grid Layout (Room 184)'I,gll la
.1 Access ladder location (Reference Point) tm ; l_, _-j - L -
, .-- g 7
, (- ,.
y, a, e- p.. .,,. . . . m.. 7~
. %, s:. .
.,, n.m.y a, . . . . . _. .
- -~
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~. . .n u- :n -
![:
I
<~ Di >[' I ^J
; .a . . , 0:b.:- S'-- [s
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p- -m i s. L e j._-' .
')
.k g
-- g.7
.- * # g,
*'.,,_,y t;; ~ - ~: ,
a ;,, . .v; u. ~-e ;3;
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.. . , m , ..
l'.,u ~~." ;A *_ 2,A!'". '.'.yV *6
-; '4 '"
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>W
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- k .
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m _ ;; *~;
^ "
. cq st-" .- ss l- Q_
;v ,"'7',.
.l_
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w S- 52 '
,;;3 ,. 2 x
] '+ ?i ~' .{
- Q,-
.,s p
.- , w_
..y A .n , n. -
, , n: , -
n. N 1 (CAVITY FLOOR AREA) l-? i -- .' p M .
~
;_ _f. -
-%,t<
p ,.. - 4- .
, g '
.' ..l S6Si<
- a. : - -~
m -- -w s _..--s MO.
; - . . . . ~ ..:,
- ,2 < - - ~ ~ .. .. ~' p: ; t , ' , ' 7, : w pf, -w ::;;';.
s, - _ - - _ . . - x , ~ ,. ,a~ y'
.e a...
2 - --
-?-
x pygg '
. x
.3;f
- 3
.78 \. n +
.--- :; u - ,
79 } w.
~f
~
v 'f ly P6a .
. . x g;;us~f, -
a
- a. - - - . . . . _ -
. .. :w_._-- -_
. .-. = -- - - - -
\1 -
TERMNATION RADIATION SURVEYhw the MSU TRNBA Reeckv C x .^:"':-#.2 Ptc}ect . It
SUMMARY
OF CRID SURVEY RESULTS FOR. TOTAL NUMBER OF GRIDS SURVEYED. MATEMAL STORAGEMTAREA 20 DIRECT SURVEYS (CONTACT 1 AT GRID POINTS ('a' 'f'): Total Number of Grid Points *a' *f* = 140 s - s . -p.r v g g USING ATHIN-WINDOWBETA-GAMMA. DETECTOR (Eberline HP-21(yT) '
? '. ,
N.0i Maximum Value - ;
' Amage Value - . . 206 Netc>m4wobe*(Beta-Gemm$ .
Direct Survey of Max. Point ("f')'as identified during Scan Survey
. Average Value-- ; 14.5 l Maximum Value '
Q
; 123 Net #m4wobe*(Beta-Osmm$ '
HERA.T5 0F WCTA-0AnsedA 90ed' SLNWEY 0.. - 0.a - 0.7 = . 4 0.8 - 0.8 -
.,- 0.4 - e 0.3 -
02 - P O.1 - F5 W 9 9 9 9 T, 9 e , (Aj-), , ,. . . . -- 1 2 3 4 56 7 8 9 1011 121313A1418161718 it 30 2122 23 P ,26 26 27 , 9 (Poens . )
>- j s USING A GAMMA BCINTIL'LOMETER tLudium Model-19 wl2'x2' Nai detectork, ,
- Amrage Vnove- , ,
. 0,0 ~ Maximum Value_-: ,
^ 1 Not UWhr(gam'nN .
ammus or oa A ,m e suwer 14 - 13 - 13 - k to -
*). .-
.]
3-3- 1- . (AV&)-t m i aa4 ea7 e o to it ja is taats to to 17 is to ao at sa as se as as a7
= c~,.f.".' "?)" "
NOTES:
* - For direct surveys, the units dpm/' probe' were used to Indicate that the actual surface area that was surveyed is eQualto the surface area of the detector itself(for the HP-210T this is about 19 cm2).
1 1 I TERMNATION RA.MATION SURswY tur ne MSU TRIGA Rencw Decommissioning Pro}ect '
SUMMARY
OF GRID SURVEY RESULTS POR:
' (] MATERIAL STORAGE PIT AREA (Continued) ,
O l BURVEYFOR LCCGESURFACECONTAMINATION ATMAX POINT (*f'h: O
. TOTAL NUMBER OF SMEARS TAMEN'- ~ ; 38 i (Total No. of Smears Analyzed for Alpha)- . 6 Average Value - - 36.g Maximum Value - ' 1172.0 cipm/100cm2(Beta-gamma) .! )
. . 0.0 : s ;0.0 dpm/100cm2(alpha') , . !
1 Loonc m e er cowr m w w sua e 0.. . 0.5 0.7 o.s o.a 0.4 o.3
.- o Il q) II 0.1 ii ,,
AVG.(t/C) ,, ,, ,, D, 1 2 3 4 6 6 7 8 9 to 11 la 1313A1418 to 17181. So 2122 23 24 28 26 27 a eru-wu asu's "** ("""*" T b =mm , 1 SURVEY OF GAMMA EXPOSURE RA TE AT(1) METER ABOVE GRIO CENTER (Point "s')* *: ~ TOTAL NoiOF GRIDS REQUIRING THIS SURVEY - -0' Average,Value -- -
% N/A' ' Maximum Value - N/A uR/hr @ 1 Meter, ,
J Q G NOTE (S):
* - Surveye for Alpha contamination were required only on a spot <heckhorl6 cation besie.
** - The Gamma exposure rate 91 M eurvoy was required only in thoes areas with potential for induced actMty
.~.
+, * --ee-, -- - - - -,..-,,.,ee- ,-.-~w-e -e--,---+-.,,_,----.-r- ~r~--e-w--- e-==--*--e* =- -- e-+
a 1 MATERIAL STORAGE PIT AREA l l (Storage Pit)- , 3.
- s; y, R-ll!"
l R-l'. R ',- R-3,- , 4; 7q . R-12 l .R-5 ? c o
%vi
.R-4 ,t
~o
._.2..
!d': (Rvu.'I -b u li 4'f' '00r'l !riCF)'
l'
' (doory- '
R-13' 'R-14::
' ' ' ~ '
.R-6? '
si 4 j '(g,7 a
'/ I I '1
.(2,; H--~ --
, ' .' R-15 f
'R-82' ~P-9L 'R-Id' (Storage Vault)
(StorageV ells) R-13A v v v - c = c ,
" ~~
r d*Nk-16 / R-20 / IR 24 rd .R-17 r R-21/ r ' " R-25
+
we . w ,e
-ww
,, i:
" t f-
, ->. R-18 r^ *~ R-22' . r^ 'R-26
~-
- ,- ,pggyp r ^. rA JHx JM GRIDS R-19 r .R-23. R-27
' ' cap sunnw
_ u .a -
-, - 3 . ,<y. . . _ ,m- . - -
. . ~ _
i l TERMINA TION RADIA TION SL)RVEYI6r she MSU TRIGA Reactor Deconomisstatung Pro}ect l l 1 l
SUMMARY
OF GRID SURVEY RESULTS FOR. TOTAL NUMBER OF GRIDS SURVEYED. I
,o l ;q"j CONTRot. ROOM 31 j i
i DIRECT SURVEYS (CONTACT) AT GRID POINTS l'a' 't'l- 4 1 1 l Total Number of Grid Points 'a' *l' = 155 USING A THIN-W'NDOW BETA-0 AMMA DEThCTOR (Eberline HP-210T)
. Averege Value - 88.6 Maximum Value . . 447 Net 40m4)tobe'tBeta-Gamma) >
Direct Survey of Max. Point ('l') as identified during Scan Survey
' Average Value - 106.9 . Maximum Value - : 340 Notdpm4>tobe'(Beta. Gamma) news or octa-omm wwr auwev _
0.9
- o.e -
c.7 - o.s - { o.S - r o.4 - , o.3 - 7 . F , i' , 4 1 1' , 9 l D o.2 - 5 J ' * ,
' 7, o ,
- t , , ,, ,' Wr '" , O i ' y , o ' *
- E_ - -s .
-.) '
'o '
E 28 29 30 3132 33 34 35 34 37 38 3e 40 4142 43 44 36 34 38 42 43 44 48 47 62 65 66 67 4 8 OsuD L.DCATION i T tmms o. - o l USING A GAMMA SCINTILLOMETER (Ludlum Modai-19 w/2*x2' Nal detectork Average Value - 0.6 Maximum Value - . 3 Net un'hr(gamma) acws or o== w sumev 18 14 - 13 - 13 - 11 - 10
- e-y.
y-4 S-
. O*
4*
- g. ]lN 15 NJIM 3 sec ammeunaer Mins
- MtaM k N N 4 1E '88 N 4 M O
38 39 30.1132 33 34 3a as 37 as se 40 4143 43 44 36 36 as 43 43 44 46 47 S3 p M 67 4 8 m (Peinte e - ) N-] NOTES:
* - For direct surveys, the units dpm/' probe" were used to indicate that the actualsurface area that was sutwyed is equal to the surface area of the detector itself(for the HP-210Tthis is about 19 cm2).
1 l
. 1 7ERMINATION RADIA TION SURVEY kw she MSU TRIGA Reactor r==w '^ ^'=,lr.2 Ptc}ect }
l
SUMMARY
OF GRID SURVEY RESULTS FOR: f9 v CONTROL ROOM (Continued) BURVEY PCR LOOSE SURFACE CONTAMINA110N A T MAX. POINTl'f'k l
~ '
. TOTAL NUMSER OF SMEARS TAMEN .~ 81 (Total No. of Smears Anaiyaed for Alpha)- .6 i Avera0e Value . .16.4 Maximum Value - , ' 100.0: dpm/100cm2(Beta'-Damma) .
0.0 -
. 0.0 dom /100cm2 (alpha') ..
Looes numer ca rvam mo aunviv 0.9 0.8 0.7
. O.s R 0.s 0.4 0.3
.- DE s 0.1 i, ll 38 si Il 11 si si
[ 2e 3e 30 3132 33 34 36 36 37 38 3e 404142 43 44 36 3e 34 42 43 44 46 47 62 66 69 67 4 8 k O acta-GAmuA ngsuL nr.eULis m l SURVEY OF GAMMA EXPOSURE RA TE A T/1) METER ABOVE GRID CENTER (Point 'n') * *: l . TOTAL No. OF GRIDS REQUIRING THIS SURVEY - O l AverageValue ~ N/A Maximum Value - N/A uRMr @ 1 Meter t l l l i l l l l l' I J NOTE (S):
* - sureye for Alpha contamination were required only on a spot-checkAmrification basis.
" - The Gamma exposure rete @1 M suray wee required only in those arose with potential for induced actMty l
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t TERMMMToON MADIA110N SURVEYSor the MSU TRsGA Reackv C :c-- :'::':42 Pro}ect SulelARY OF GRID SURVEY RESULTS FOR. TOTAL NUMBER OF GRIDS SURVEYED.
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M 100tOMCO) 14 DIRECTCURVEYG(CONTACT) ATGRID P0lNTS(*n'- *f*); Total Number of Grid Points 'a' 'f' = 90 ;
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USING ATHIN-WINDOW BETA-GAMMA DETECTOR (Eberlirid HPA2107) ' J Aserage Vake - _ 60,4 Maximum Value ' ' '433 NetdiT,,i&(Peta-Gamma) _ Direct Survey of Max. Point (*l') as identified during Scan Survey . , a- ,
- Aserage Value'.: ' 106.9 : Maximum Value - - ; 409 'Not t$wr@ robe * (Beta-Gamma) .
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fulLTS OF SCTA-GAinsA N Suwtv O.. - 0* , 0., - gg .... . 0.8 - c.. - E 9 T O.3 - ,
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USING A' GAMMA SCINTILL'OMETER tLudium Model-19 w/2*x2' Nal detectork -
. Aserage Value- - 0.9_ . Maximum Value -; ~
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" NOTES:
- For direct surnys, the units dpm/' probe
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t, [. TERMNATKW RAOIATKW SURVEYhw dw MSU TRGA ReactorC :-z . '::':-#2 Moinct - ,
SUMMARY
OF GRIO EURVEY RE8JLTS FOR: ({ ROOM 190(Office) (Continued) I RURVEYPCR LQQGE BURFACE CQNTAMINA TKW AY MAX. 'POmtT/*f*k
,T0fAL NUMBER OF SMEARS TAMEN 7 : >- .18
' (Total No of Smears Analyzed for Alpha)- : 4 2 AverageValue . 8.6 : MaximumValue ' '
- 57.1 ' dpm/100cm2 (Seta-Gamma) '-
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, Avera0e Value . N/A ~ Maximum Value - N/A; . UR/hr @ 1 Meter -
NOTE (S):
* - Surveye for Alphe contaminellon were required only on a epot. check /verif6 cation basis.
" - The Gamme exposure rate 91 M eurvey wee required only in those eroes with potential for induced actMty
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( TERMNA110N MADIATION SURVEY nor the MSU TRKEA Reactor C= ^ :'::':2.ai Pro}ect - i 1 1
SUMMARY
OF GRID SURVEY RESULTS FOR. TOTAL NUMBER OF GRIDS SURVEYED. MOOM t89(Henith-Ph," Lab.) 20 l DIRECTSURVEYS(CONTACT) ATGRIO POINTSl'n' 't'k i Total Number of Grki Points 'a' *l' = 100 1 USING'A THIN-WINDOW BETAIGAMMA DETECTOR (Eberline'HP-2107)1 - Awage Value - 93.0 - Maximum Value l419 Not#mproce*(Beta-Gamm$ Direct Survey of Mat Point (*f') as klontified during Scan Survey q. > <
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'OTES:
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f- 1 l l l TERMNA TION RADIA TION SURVEY Sor the MSU TRIGA Reactor C+: T;'::':--E.g Project 1
SUMMARY
OF GRID SUF M RESULTS FOR: l J l-{ POOM 192 (Health-Physics cab.) (Continued) j i SURVEY FOR I 008E SURFACE COMTAMINA TION A T MAX. POINT (*f'k -l 1 TOTAL NUMBER OF SMEARS TAKEN ' 20 : , (Total No. of Smears Analyzed for Alpha)- '6 Average Value - s 16.6- Maximum Value - ' 63.6. dpm/100cm2 (Bota-gamma):
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* - Surveys for Alpha contamination were required only on a spot-check /verincation basis.
" - The Gamma exposure rate 91 M eurwy was required only in thoes areas with potential for induced activity
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