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Argonaut Reactor Facility Decommissioning Final Rept
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t VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY ARGONAUT REACTOR FACILITY DECOMMISSIONING FINAL REPORT NRC LICENSE NO. R-62 DOCKET NO. 50-124 O'

APRIL, 1987 VIRGINIA POLYTECHNIC INSTITUTE AND STATE UNIVERSITY DEPARTMENT OF HEALTH AND SAFETY BLACKSBURG, VIRGINIA 24061

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SUBJECT:

VIRGINIA TECH REACTOP FIriAL REPORT DEAR l @ Al, SlkS)

This final report describes the dismantlement and disposition (D&D' of the Virginia Tech Argonaut Reactor (VTAR) conducted pursuant to the D&D order issued on October 29, 1986 for Mph 4 R-62.

bi&V The report shows that the D&D order was followed and the facility now meets unrestricted use criteria, specified by the D&D order.

We request that the license R-62 as amended be terminated in its entirety.

Sincere'y, Dr. Keith Furr b

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i TABLE OF CONTENTS NUMBER SECTION PAGE

1.0 BACKGROUND

AND MANAGEMENT 1

1.1

SUMMARY

DESCRIPTION 1

1.1.1 Facility 1.1.2 Reactor 1.1.3 Support Facilities 1.1.4 Decommissioning Option / Man-Rem 1.2 FACILITY OPERATING HISTORY 3

1.2.1 Reactor Operations 1.2.2 Conclusions 1.3 CURRENT RADIOLOGICAL STATUS OF FACILITY 3

1.3.1 Neutron Activated Materials 1.3.2 Contaminated Materials 1.4 DECOMMISSIONING ALTERNATIVE 4

1.5 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES 4

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1. 5.1 Project Organization 1.5.2 Virginia Tech Administrative Controls 1.6 REGULATIONS, REGULATORY GUIDES AND STANDARDS 4

2.0 OCCUPATIONAL & RADIATION PROTECTION PROGRAMS 5

3.0 DISMANTLING AND DECOMMISSIONING TASKS AND SCHEDULES 5

3.1 SCHEDULE 5

3.2 TASK ANALYSIS 5

3.2.1 Baseline Radiological Survey 3.2.2 Core Dismantlement 3.2.3 Experimental Facilities Dismantlement 3.2.4 Decontamination and/or Dismantlement of Fuel Storage Wells 3.2.5 Dismantlement of Process t!ater System 3.2.6 Activated Concrete Removal 3.3 SAFE STORAGE 9

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TABLE OF CONTENTS (CONTINUED)

NUMBER SECTION PAGE 4.0 SAFEGUARDS AND PHYSICAL SECURITY 9

5.0 RADIOLOGICAL ACCIDENT ANALYSIS 10 6.0

'tADI0 ACTIVE MATERI ALS AND WASTE MANAGEMENT 10 7.0 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS 10 8.0 FINAL RADIATION SURVEY 10 8.1 Facility Grid Plan 8.2 Surveys Performed - General 8.3 Special Surveys 8.4 Final Survey Instrainentation FIGURES 13 TABLES 14 APPENDIX A Final Survey Grid Heps 15 APPENDIX B Sample Pages - Final Release Survey 16 APPENDIX C Summary of Release Survey Data 17 APPENDIX D Survey Instrument Data 18

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FIGURES NUMBER TITLE PAGE l

1 FACILITY LOCATION F-1 2

FACILITY FLOOR PLAN F-4 3

VENTILATION SYSTEM (REMAINING COMPONENTS)

F-6 4

ORGANIZATIONAL CHART F-7 5

PROJECT SCHEDULE F-8 6

BACKGROUND RADIATION SURVEY - TABLE F-9 7

FUEL STORACE PITS - DRAWING F-14 8

CONCRETE ACTIVATI0fl PROFILE - DRAWING F-15 9

FACILITY GRID PLAN (FOR FINAL SURVEY)

F-16 O

TABLES NUMBER TITLE PAGE I

AIRBORNE RADI0 ACTIVITY MONITOR DATA T-1 II PROJECT MAN-REM (DOSE)

T-2 III REGULATIONS At!D STANDARDS T-4 IV RELEASE CRITERIA - STANDARD (FROM D&D PLAN)

T-5 V

CONCRETE INVENTORY AND BELOW GRADE S0IL DISPOSED AT T-6 V.P.I. LANDFILL VI RADI0 ACTIVE WASTE SHIPMENT

SUMMARY

T-9 VII DISCHARGES TO SANITARY SEWER T-10 VIII SURVEY INSTRUMENTS USED AT V.P.I.

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1.0 BACKGROUND

AND MANAGEMENT The following sections of this report show that the Virginia Polytechnic Institute and State University Reactor facility now meets unrestricted use limits as specified in the NRC Decommissioning Order and outlined in Section 1.6, Regulations, Regulatory Guides and Standards of this report.

With the exception of removal of the Reactor Fuel, Chem-Nuclear Systems, Inc.

(CHSI) performed all of the major decommissioning operations. The licensee, Virginia Tech, monitored all operations to ensure regulatory compliance.

This report follows the format of the Decommissioning Plan submitted by i

Virginia Tech to the Nuclear Regulatory Commission in July,1986 (Docket No. 50-124).

1.1 SUf1 MARY DESCRIPTION 1.1.1 Facility The Virginia Tech Argonaut Reactor (VTAR) Facility is located on campus in Robeson Hall. Physics Dept. offices, other research laboratories, and classrooms are also located within this building.

Figure 1 shows the Reactor Facility location, Figure 2 shows the floor plan layout of Robeson Hall and the VTAR.

1.1.2 Reactor i

The VTAR utilized a heterogeneous design, graphite reflected / light water moderated core. Operations began in j

December,1959. Licensed for a maximum power level of 10 kW(th), the reactor was later modified and license amended to allow a maximum power level of 100 kW(th).

It was shut down in July 1983, and a possession only license was issued e

in April, 1985. Reactor Fuel was shipped to the DOE in late 1985 and early 1986. Dismantlement and Decommissioning (D&D) operations began in late September, 1986, and the Final Survey was completed in January,1987.

1.1. 3 Support Facilities The remaining facilities that had previously supported VTAR operations and their function /use during D&D operations are listed as follows:

a)

Chemistry Laboratory with Chemical Fume Hood

-used as the Health-Physics counting room.-

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b) 10 Ton Overhead Crane g

-used to handle reactor closure blocks, heavy items, and sectioned concrete pieces.

c)

Ventilation System and Heating Units

-provided negative pressure and thus a constant air flow in the work area (Room 10) during decommission-ing.

-The two facility heating units (located in Rooms 6-and 10) were not used during the decommissioning operations, instead, they were sealed off-tightly to prevent the possibility of contaminating the unit internals.

During the final release survey, they were unsealed and restored to use (Section 8.3).

Operations requiring HEPA. filtered ventilation utilized portable HEPA ventilators which discharged into the facility system.. Monitoring of the ventilation system was accomplished by direct sampling from the discharge stack (using an Eberline RAS-1 pump), and by a G-H detector and TLD chips mounted in the stack. The ventilation system inlet plenum, ducting, and roof fan remain intact and operational (Figure 3).

d)

Reactor Cooling System

-all components of this system (the majority located in the Process Pit) were drained as required, disassembled, surveyed, and disposed of as appropriate. The primary system dump tank was used to 4

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collect water and sludge during concrete sawing operations (Section 3.2.6.)

4 The only remaining compcnent of this system is the secondary system cooling tower located on the roof of Robeson Hall. This component was part of the 500KW upgrade proposal and was never used for reactor operations. The associated system piping located inside Room 10 was surveyed " clean" and dismantled for i

release (see Section 3.2.5).

e)

Below-Grade Reactor Fuel Storage Wells

-the below grade fuel storage areas are located in two, (16) well banks on the NE and NW corners of Room

10. Only the NW bank was previously used t.o store irradiated fuel and other such materials. The NE bank was controlled as a " clean" area. These storage areas were not used during D&D operations.

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1.1.4 Decommissioning Option / Man-Rem Release of the facility for unrestricted use was the decommisioning option chosen. Actual total collective radiation dose to the Project Crew was 7.9 Rem. Higher than expected dose rates inside the core and general work area (s) following removal of " Hot" core components were the main contributors to the collective dose (Table II).

J 1.2 FACILITY OPERATING HISTORY 4

1.2.1 Reactor Operations During the operating history of the VTAR, there were no major radiological accidents /occurences. Pesults of various detailed surveys performed before, during, and after decommissioning operations agree with this statement. Miscellaneous solid, liquid, and gaseous wastes were produced by VTAR operations; however, these did not pose significant decontamination probler..s.

Two minor incidents had occurred during reactor operations.

Both were reported and appropriate cleanup measures were taken. Minor levels of surface (fixed) contamination were found in the bottom of one Fuel Storage well and in the process pit concrete floor. A surface scarifier was used to successfully remove this contamination.

1.2.2 Conclusions Based on actual observations, made during dismantlement, the past operating history data and information was accurate.

Furthermore, all contamination has been successfully removed from the facility to acceptable (release) levels.

1.3 CURRENT RADIOLOGICAL STATUS OF FACILITY The facility had been operated in an extremely clean radiological fashion. This simplified the decommissioning operations.

All of the radioactive portions of the facility have been removed.

This includes the following:

1.3.1 Neutron Activated Materials

-Reactor core closure plugs

-Reactor core, components, and experimental facilities, shield plugs, etc.

-The biological shield structure k

-Reactor core closure plugs 4

-Below grade concrete (floor) and soil from beneath the i

reactor core area...

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1.3.2 Contaminated Materials

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-Reactor coolant / moderator / process system

-Minor contamination found in the bottom of (1) fuel storage well.in the Northwest corner of Room 10, and the process pit floor.

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The final. decommissioning Radiological Survey has been performed and all areas indicate compliance with the criteria for license termination for unrestricted use, d

Surveys by the licensee under the possession only license requirements are no longer' required.

1.4 DECOMMISSIONING ALTERNATIVE Future activities at the reactor facility will not require use of the reactor; an NRC license will not be required. Therefore, no other l

alternative is under consideration.

i 1.5 DECOMMISSIONING ORGANIZATION AND RESPONSIBILITIES 1.5.1 Project Organization Figure 4 shows the organization chart for the dismantling and decommissioning project. CNSI was the prime O

contractor. Personnel experienced in reactor operation and radioactive material handling were responsible for all dismantling operations.

1.5.2 Virginia Tech Administrative Controls The Vice President for Administration and Operations has overall responsiblity for the Reactor Facility.

Administrative responsibility has been assigned to the l'

Department of Health and Safety. Monitoring for regulatory compliance during the decommissioning operations was performed by this department under direction of Dr. Keith i

Furr, Department Head. Health Physics practices and enforcement of Radiological safety was monitored by a University Radiation Safety Officer, Doug smiley.

1.6 REGULATIONS, REGULATOP,Y GUIDES AND STANDARDS The relavant Regulatory guides and standards are listed in Table III. The applied limits for release are those stated in NRC Regulatory Guide 1.86 and the NRC staff position of 5 uR/hr above background at 1 meter (Table IV). Materials near or above these limits were disposed of as Radioactive Waste.

Final survey results indicate that all areas of the facility currently comply wi_th these limits.

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Quality Assurance requirements were followed by the contractor in accordance with:

o 10 CFR 71 Subpart H o

10 CFR 50 Appendix B o

ANSI N45.4 2.0 OCCUPATIONAL AND RADIATION PROTECTION PROGRAMS These standards are outlined in the Decommissioning plan. The contractor used its own Radiological Protection Procedure and Quality Assurance Program with satisfactory results.

3.0 DISMANTLING AND DECOMMISSIONING TASKS AND SCHEDULES The Dismantling and Decommissioning tasks and schedules are outlined in the Decommissioning Plan.

Section 3.2 of this Report gives details of the act4J' project methodology and identifies deviation (s) from, or problems with the planned Project operations.

3.1 SCHEDULE The major task and milestone schedule is shown in Figure 5.

3.2 TASK ANALYSES This section describes the approach which was used to accomplish each significant task of the Virginia Tech Research Reactor decommission-ing Project.

It has been divided into the following main headings:

o Baseline Radiological Survey o

Core Dismantlement o

Experimental Facilities Dismantlement o

Decontamination and/or Dismantlement of Fuel Storage Wells o

Dismantlement of Process Water System l

o Activated Concrete Removal i

3.2.1 Baseline Radiological Survey During Phase I, waste characterization and Baseline Survey data was collected. Several methods were used to collect the required data:

Swipe analysis of the overall facility and of specific suspect areas and components were conducted. Surveys conducted in the ventilation ducting showed no 4

contamination in that system..

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A single 17/8", 6' long concrete core sample was taken i

1 from the north side of the biological shield near core center line, above the beam-tube penetration. This sample was sectioned and analyzed to confirm activation analysis calculations and to identify the radionuclide content of the concrete waste. Other nuclide identification samples included:

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Graphite, aluminum, and steel from the core area o

Corrosion products (rust) from the fuel storage pits o

Activated lead These samples were all analyzed by an off-site Laboratory.

Results were used to complete the waste shipment manifests-in accordance with 10 CFR 61 and address radiological I

controls necessary during operations.

1 Based on the isotopes handled at VTAR and previous contamination surveys, there appeared to be essentially no potential for alpha contamination. Alpha surveys (direct readings and smears) of random points confirmed that alpha i

contamination was at or near background levels and I

substantially below release criteria.

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Background dose rates were measured in buildings and outdoor areas within an approximate quarter-mile radius i

from the reactor using a Pressurized Ion Chamber (PIC) and micro-R meter. These readings were combined to derive an i

average background reading for each instrument (Figure 6A).

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3.2.2 Core Dismantlement t

The dismantlement of core components was executed in a straightforward manner with no attendant mechanical difficulties or unreasonable exoosure. To minimize personnel exposures, the dismantiement of core components was performed early in the project. This removed the 1argest sources of radiation before dismantling the

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experimental facilities and setting up the equipment to

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remove the activated concrete. However, because of

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building accessibility problems (a large building i

construction. site which was located immediately adjacent to l

the Room 10 access doors created several delays and j

complications throughout the project), these packaged i

materials had to be temporarily stored in the building, near the general work area. This caused a measurable increase in the overall collective Nan-Rem (Table II).

For the removal of the core components, crew training in tooling, handling, and task requirements minimized the time 4

required. Long handled cutters and tools were used to maximize the distance to the sources. :

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upon local shielding of the steel items embedded in the concrete, the working background in the core cavity was expected to be in the 25-50 mR/hr range. -However, the actual area dose rate was as high as 300 mR/hr. This also contributed to the overall collective Man-Rem being higher than planned or expected (Section 1.1.4).

4 3.2.3 Experimental Facilities Dismantlement Removal of the thermal column graphite occurred first. A containment tent with HEPA filtered ventilation was erected at the duct opening. Handling and packaging of the stringers took place inside the tent.

Core graphite and components were then packaged per Section 6.0.

The removal of Shield tank duct graphite, gamma j

curtain, and core support plate was delayed until after the non-activated concrete was removed. This allowed for better accessibility and afforded a lower Man-Rem expenditure.

In addition, the diamond wire saw cut at the base of the bio-shield was successful in cutting through the core support plate mounting bolts, making removal quite simple.

The (2) beam tube embedments were removed using a 14" O

coring bit. The 6' long cores were then surveyed, sectioned, and disposed of as appropriate.

3.2.4 Decontamination and/or Dismantlement of Fuel Storage Wells Since the NE bank of storage wells had been controlled as j

" clean" (Section 1.1.3e), confirmatory surveys were performed and the cover was sealed in place until the final survey.

Detailed surveys taken in these wells show that they meet release criteria, The NW bank of storage wells were treated differently.

Baseline survey information indicated measurable levels of j

contamination. This was probably due to various other material (e.g. irradiated items, waste, etc.) which were occasionally stored in these wells. Loose corrosion j

products and fuel element guides (shown in Figure 7) were removed and disposed of as Radioactive Waste.

i Detailed surveys were then taken in each well.

No measurable contamination was found in (15) of the (16) wells. The remaining well was successfully decontaminated to the limits of release criteria.

l All fuel storage well plugs were removed, surveyed and 3

i released for unrestricted use (they remain on site for use i

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f' 3 '. 2. 5 Dismantlement'of Process Water System The old cooling system was disassembled. removed, surveyed and disposed of as metal scrap. The 150 gallon " dump tank", which was used as _a water collection sump during concrete sawing, was later decontaminated, cut open for survey, and released as scrap metal.

l The 500KW upgrade. cooling system components located inside Room 10 (which were never used for reactor operations) were j'

disassembled and surveyed to verify' the absence of j

contamination.The heat exchanger, pump, and piping were surveyed " clean" and retained as salvage. At the point I

where the piping exits Room 10, smears of the interior i

piping were taken to confirm that no contamination was present, and to release the balance of the system. All i

components of this system met criteria for. release. The remaining sections of discharge (sewer) pipe (s) have been surveyed per Section 8.0 and satisfactorily meet release criteria.

l 3.2.6 Activated Concrete Removal Based on baseline survey data, coring of the shield, and a i

surface dose rate and sample survey of the interior surfaces of the shield, a depth profile and contour map of activated concrete was produced (Figure 8).

i Various methods were used to section and " break up" the i

biological shield structure. Radiologically " clean" i

concrete was removed first, leaving an activated concrete section to be demolished and packaged as radioactive waste.

1 Exterior portions of the shield, which were composed of j

releasable materials (below the Table IV limits), were removed, surveyed, sampled, and shipped to the Virginia Tech-owned solid waste disposal site. Samples of these materials were analyzed _on-site and retained with the j

survey data. This releasable concrete will be buried in j

the same manner as other construction debris by the University, after other check surveys performed by NRC.

i Immediately outside the access door from the reactor room, a major facility construction project was underway during the term of the decommissioning work. This resulted in having a narrow, difficult path comprised of dirt and l

gravel for passage of equipment and materials. Due to the j

restricted reactor room access and weight limitations associated with handling materials, the' maximum. practical

'I size for any single piece of concrete to be removed from l

this shield was approximately 10,000 lbs.

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The reactor was first separated into two major pieces using a diamond-impregnated wire saw. The first cut was made e'

through the horizontal beltline of the reactor shield. A second cut was made at the base of the shield. These two horizontal cuts then separated the biological shield into two large slabs; approximately 90 tons each. Water was channeled from the cut area to the process pit where solids were allowed to settle out in the dump tank. Careful monitoring of the sludge and water with the on site MCA System resulted in determinations that most of the cutting fluid met free release criterion (see Section 6.0).

Higher activity ' sludge' which was collected while cutting through the central region of the shield (near the reactor core) was solidified and disposed of as radioactive waste.

The next step after wire sawing was to cut near-surface rebar using a diamond-tipped circular concrete saw.

Vertical cuts on roughly 2-1/2 foot centers were made to define the edges of each five-ton block which would be produced from rock drilling and splitting. This saw-cutting was done to a depth of roughly four inches.

Following these operations, a pneumatically powered rock drill was used to drill 1-7/8 inch holes on approximately 10 inch centers. These holes, drilled to depths up to 6 feet, permitted introduction of hydraulically powered concrete splitters which were used to segment the slab into blocks.

The 10-ton overhead crane was used to rig each concrete block as it was separated from the shield. The block was weighed and moved to a low background area where it was surveyed (Section 8.0).

This involved an assessment of the degree of surface contamination and measurements of the dose rate on contact and at one meter from each side of the block.

Blocks and " chunks" which were below the dose rate and contamination criteria were transported to the University owned landfill (Table V). The radioactive portions of the biological shield amounted to a total volume of approximately 1300 cubic feet.

3.3 SAFE STORAGE AND LICENSE TRANSFER i

All radioactivity has been removed from the VTAR Facility. Sa fe storage of materials within the facility will not be required.

All sources used under License R-02 have been transferred to VPI's NRC Broad License P0. 45-09475-30. These items included a PuBe source, fission chamber and other small sources now listed on the inventory list for the Broad License.

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I The activated lead gamma curtain, which is not currently acceptable for commercial disposal as LLW or hazardous waste has been transferred to the same Broad License above and has been placed in safe storage in a controlled area.

All other radioactive material has been accepted and buried at the Barnwell low-level waste management facility.

4.0 SAFEGUARDS AND PHYSICAL SECURITY The Safeguard and Physical Security plan as described in the D&D Plan remains in effect. All Special Nuclear Material (SNM) listed under NRC License R-62 has been transferred from the reactor facility to the Broad License.

5.0 RADIOLOGICAL ACCIDENT ANALYSIS The fuel was safely transferred off site prior to CNSI D&D operations; there were no accidents related to fuel handling.

6.0 RADIOACTIVE MATERIALS AND WASTE MANAGEMENT During the Decommissioning Project, waste was handled as described in the D&D plan. Approximately 2200 ft3 of materials were packaged as Radioactive waste and shipped for burial to CNSI's Low Level Maste Site at O

Barnwell, S. C.

Five shipments containing a total of approximately 13 O

curies of. activity were made. All packages were classified as Radioactive LSA/ Type A/ Class A waste (Table VI).

No hazardous wastes were generated during the D&D Project. Non-radio-active wastes, including pipe, concrete rubble and blocks, steel, soil, and other materials which met Table IV requirements were disposed of at the University-owned landfill and scrap bin located approximately 3 miles from the reactor (see Table V).

The 1 1/2" X 4' X 4' pb Gamma curtain was sampled, surveyed, and properly packaged for dispositir' at a later time. Chem-Nuclear and U. S. Ecology would not accept this 1e o as LLW since it was shown to be a hazardous material (exceeded limits For the EP toxicity test). Water which met 10 CFR 20. Appendix B relcase limits was discharged to the sanitary sewer system (Table VII). Surveys performed inside these discharge pipe (s) (per Section 8.0) indicate no residual contamination present.

7.0 TECHNICAL AND ENVIRONMENTAL SPECIFICATIONS The dismantling and decommissioning tasks were accomplished with no significant impact on the environment or the health and safety of the i

public.

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~8.0 FINAL RADIATION SURVEY The final release survey was performed in accordance with guidelines found in NUREG-2082, " Monitoring for Compliance with Decommissioning Survey Criteria." The survey used a formalized procedure based on a sampling i

inspection plan which assured adequate coverage and statistical analysis of the results to provide efficient use of the data.

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8.1 FACILITY GRID PLAN The Facility was divided into three types of areas for data 4

l collection and analysis (Figure 9):

o High Potential Contamination Areas l

o Medium Potential Contamination Areas l

0 Low Potential Contamination Areas o

Special Survey Areas Appendix A contains the grid map (s) for each area surveyed.

Areas of low contamination potential were those areas where previous Virginia Tech survey results indicated contamination below release i

limits and the function of that area was not conducive to creating a contamination problem. Areas included in this description were the control room, offices (walls and floors) and all facility ceilings. -

These areas were sectioned into 10' X 10' (3m X 3m) grids. These l

grids were surveyed (as described below) on a statistically random basis. A total of (32) grids were surveyed in these areas to meet statistically satisfactory release criteria.

j Areas of medium contamination potential were those areas where either i

Virginia Tech survey results indicated contamination near or above release limits and/or the function of that area constituted a potential for contamination. Areas-included in this description were f

Room 10 floors and walls and the Chemistry Lab and adjacent rooms.

l These areas were sectioned into 6' X 6' (2m X 2m) grids on the floor and the walls; 90 grids were surveyed on a statistically random

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basis.

Areas of high potential contamination were the reactor shield floor and excavated area, and the pump-process pit. These areas were sectioned into 3' X 3' (lm X 1m) grids; all of these grids were surveyed.

i As part of the total number of grids surveyed on a random. basis for each of the low and medium potential contamination areas, approximately 10 specific grids were surveyed on a stratified basis.

I These grids were in (1) doorways and room corners,-(2) near j

penetrations or obstructions, and (3) specific areas which had j

previously indicated contaminated materials or equipment.

In addition, a detailed survey of the Reactor Pedestal / Room 10 floor expansion joint was conducted to check for deposition of contaminated 1

water / sludge from concrete sawing activities (see Appendix B).

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r%h 8.2 SURVEYS PERFORMED-GENERAL Each designated grid was surveyed in the following manner.

o A gamma (micro-R/hr) point reading was taken at 1 meter above the center of the grid, i

o Beta-gamma and gamma contact readings were taken at five equally spaced points within the grid.

o A G-M beta-gamma surface scan survey of the grid was conducted and the maximum beta-gamma point identified and surveyed, o

A smear survey and gamma dose rate survey was taken at the maximum beta-gamma point. The smear was counted for gross al pha-beta-gamma activity. For rough surfaces where smear surveys were impractical (i.e. " jagged" concrete areas), a direct measurements were taken using a Mini-Scaler with a thin-window GM detector, i

o The data from each five point survey was recorded and the average value was recorded as an unbiased measurement. The measurements taken at the beta-gamma maximum point was recorded as a biased measurement.

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Various random samples were taken (e.g. soil, concrete, etc.)

and analyzed on the ND System 66 MCA as a backup to the routine surveys.

A summary of this data is presented in Appendix C.

I 8.3 SPECIAL SURVEYS i

l The procedure for surveying pipes, drain lines, ventilation system ductwork, and heater was as follows:

o 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

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ends of accessible pipes, the interior surfaces of inlet and outlet vents, and the water traps and exit points of drain lines. No significant radioactivity was found at any of the entry or exit locations.

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A " wipe" was pulled through representative pipes, drain lines, and duct work to ensure no contamination was present. Pipes that contained contamination above Table IV limits were cut up and disposed of as radioactive waste. This included ' portions of the primary coolant system supply and return lines, pool drain pipe, and core cavity drain piping.

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8.4 FINAL SURVEY INSTRUMENTATION All instruments used for these surveys were calibrated in a manner acceptable to the NRC.(as specified in the D&D plan). All calibration sources used are traceable to the National Bureau of Standards. "able VIII lists all the instruments used during the 4

i Project and the Final Survey; Appendix D shows instrumentation data i

for the final survey.

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LOCATION DATE :

24 Jul 86 29/30 Jan 87 RSS-111 PRM-7, RSS-111 PRM-7

============================================================

1 Foyer Area at Derring Hall 8.4 5.0 8.5 3.0

-cement deck & steps Stone sided walls (like Robeson) approx. 20' from 2 sides of detector. (partial walls, 6')

N,o overhead, slightly below grade 2

Grassy open area, (trees, etc.)

13.2 9.5 10.5 5.0 slightly recessed [ below gradel Bldg's all >100 yds away (all sides) approx 75 yds from road (near pond) 3 Drill Field [ open, flat grass) 12.4 9.5 10.5 4.0 Trees @ 75 yds, >100 yds from bldgs.

4 Outside Vawter Bldg. adjacent to 12.4 9.0 9.5 4.0 (2) b1dg. walls [ G100 deg., approx.

50' from ea.1 large tree C 50'.

5 Outside Pritchard Bldg. at base, 7.8 5:5 on contact w/2 walls [in corner)

Stone siding like Robeson, 6 story.

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6 Red Brick Bldg., outer stair well 11.7 8.0 11.8 5.0 approx.

6' below grade, walls e 6' on (2) sides, no overhead.

7 Parking lot [ asphalt). Bldgs. all 8.5 5.0

>100 yds. away. Slightly elevated above general campus layout & Robeson.

8 Outside McBride Bldg. at entrance 7.7 4.0 8.5 3.0 alcove (wedge shaped). (2) 20' walls G12 from detector. Cement steps &

sidewalk, stone [ Robeson) siding.

9 Inside Robeson Hall. Room 224 6.0 3.0 Large open room. 2 levels high approx. 1/2 size of Rx room.

10 Inside Robeson Hall. Room 210 8.7 5.0 (above Rx room) Large classroom w/

sunken speaker's area at base.

11 Outside of Robeson Hall G25 yds.

8.1 4.5 from physics shop door (on pavement) 12 Inside Robeson Hall, Basement level 14.0 10.5

[ cinder block walls 93', on 2 sides).

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F - 17 i

V.P.I. DEC0f!MISSI0 FLING FINAL REPORT TABLES (12 pages)

O

-is-O

l y,

I

/O, :

Table I

(

Airborne Radioactivity Measurement Summary (Continuous monitoring via Regulated Air Sample pump on discharge Stack)

Dates of Total Total Vol.

Mec3ured Activity Operation Hours of Air (j)*

Isotope Concentration 10/17-81.2 hr 4.87E+5 Gross B'/T ND 11/7/86 Gross =

ND K-40 3.65E-9 uCi/ml 11/10-43.2 2.59E+5 Gross E/F ND 11/16/86 Gros s on.

ND K-40 2.72E-11 uCi/ml 11/17-25.3 1.52E+5 Gross B'/t ND 11/19/86 Gross =<

ND K-40 3.59E-ll uCi/ml 12/1-2/86 10.0 6.0E+4 Gross B'/t ND Gross =<

ND K-40 8.76E-ll uCi/ml 12/2-10/86 34.0 2.04E+5 Gross B77 ND Gross d ND K-40 3.97E-10uC1/ml

- Based on an average sample flow rate of 100k/hr w/Eberline RAS-1 pump.

(1)- NOTE:

a.

Gamma Spectranalysis performed with ND66 MCA system b.

Gross B/ as measured with Eberline MS-2 W/HP-210T probe c.

Gross as measured with an Eberline SAC-4 instrument.

Refer to the Project instrument list, Table VII.

ND - None Detected (above background) e i

O T-1

TM LE II A O

V. P. I.

D &

D PROJECT WeeHy Dose (MAN-REM) by SRD.

1.50 1 40 1.30 1'20 M M 1.10 5

a m

M M 5

[l E

E E

E E O.70 0 60 N

O.50 N

0.40 O.30 I

0.20 o,j o 0.00 1

2 3

5 6

7 8

9 10 11 12 j

PROsECT WEEK V.P.I.

DOSE REPORT (END OF PROJECT)

PERSONNEL DOSE EQUIVALENT VS. TASK (BY TLD)

Collective MAN-REM TASK Estimated Actual

_ = = - - - - - - = = - - - - -.

0.25 0.25 Fuel Removal

Miscellaneous Equipment Removal

0.10 0.10 Baseline Survey 0.07 0.05 Core Cavity Material Removal & Packaging 0.75 3.27 Concrete Removal 0.35 2.36 Routine General Area Exposure (s) 0.10 1.90 I

TOTALS:

1.62 7.93 i

l

- These Tasks were performed prior to CNSI's arrival on s.ite.

lO T-2 i

1

TABLE II B i

V.P.I. D & D PROJECT TOTAL MAti-REM SliARE (7.926 REM)

\\

v

,7~M Healt h -P hysics: (24.1 %)

Local Labor: (42.5%)

TRISTATE (3.0%)

d S upervisory: (27.6%)

ENERFAB (2.7%)

a V.P.I.

DOSE REPORT DOSE (R) by TLD NAME COMPANY FUNCTION WHOLE FINGER BODY RING Austin.

G.

TEMP.

Decon. Tech.

O.829 0.036 Becksan.

T.

ENERFAB Vire Saw Sub.

0.028

0.000 Burns.

D.

TRISTATE Saw/ Core Sub.

0.115 0.000 Capp C.

TEMP.

Decon. Tech.

0.868 **

0.010 Ellis.

E.

TEMP.

Health-Physics 0.014 0.000 Evans. G.

TEMP.

Decon. Tech.

0.843-0.017 Hopkins. P.

TEMP.

Decon. Tech.

0.647 0.301

Huff, K.

TEMP.

Health-Physics 0.570 0.000 Huston.

R.

CNSI Ops. Supervisor O.919 0.689 Krasp.

B.

CNSI Health-Physics 0.000 0.000 Lafon.

P.

TEMP.

Decon. Tech.

0.090 0.000 Lester.

B.

TEMP.

Decon. Tech.

0.079 0.000 Manning.

M.

CNSI Proj. Supervisor 0.715 0.394

Martinez, A.

CNSI Ops. Supervisor 0.553 0.000 Matt 11n. S.

TRISTATE Saw/ Core Sub.

0.125 0.000 Montgomery.

R.

VPI Health-Physics 0.384 0.000 Niehaus.

T.

ENERFAB Vire Saw Sub.

0.037 0.000 Riggi, J.

HILBERT Health-Physics 0.960 0.468 Shelor. S.

TEMP.

Decon. Tech.

0.000 0.000 Williams.

V.

ENERFAB Vire Saw Sub.

0.150 ~

0.000 t

TOTALS (REM):

7.926 1.915

Beckman was not issued a TLD: he wore a SRD-this is an assigned dose.

- Lost TLD. Dose assigned by SRD 2

T-3 l

l l

m E

l

{

TABLE III Regulations & Standards Related To Decomissioning Of The Virginia Tech Research Reactor Agency Regulation / Standard Description Federal:

Radiation Protection NRC 10CFR20 Radiation Protection Standards 10CFR50 Domestic Licensing of Production & Utilization Facilities 10CFR71 Shipping of Radioactive Material Reg. Guide 1.86 Decontamination Limits g

00T 49CFP.

Shipping of Radioactive Material EPA NEPA Preparation of an EIS OSHA 29CFR Worker Health & Safety I

Guidelines NRC NUREG-1756 Decomissioning of Research & Test Reactors NUREG-2082 Temination Survey Criteria l

l NUREG-0586 Generic EIS for Decomissioning Nuclear Facilities l

Staff Position

" Guidance and Discussion of

" Guidelines" Requirements for an Application to Terminate a Fon-Power Reactor Facility Operating License",

Revision 1,1984 ANSI ANSI /ANS-15.10 Decomissioning of Research Reactors ANSI /ANS-N13.12 Decontamination Limits b

h k

Y.,,,.

1m..

W N

Table IV ACCEPTABLE SURFACE CONTAMINATION LEVELS b

b NUCLIDEa AVERAGE c MAXIMUMbd REMOVABLE e 2

U-nat, U-235, U-238, and 5,000 dpm a/100 cm2 2

1,000 dpm a/100 cm 15.000 dpm al100 cm associated decay products m

2 2

2 Transuranics, Ra 226, Ra-228, 100 dpm/100 cm 300 dpm/l00 cm 20 dpm/100 cm Th 230,Th 228,Pa 231, Ac-227,1 125,1 129 1

2 2

2 Th-nat, Th-232, Sr-90, 1000 dpm/100 cm 3000 dpm/100 cm 200 dpm/IO0 cm Ra-223, Ra 224 U-232,

~

1126,I131,1133 2

2 1000 dpm #q/100 cm2 Beta-gamma errutters (nuclides.

5000 dpm #9/100 cm 15,000 dpm #9/100 cm with decay modes other than alpha emission or spontaneous fission) except Sr 90 and others noted above.

aWhere surface contamination by both alpha-and beta-gammaemitting nuchdes exists, the limits estabhshed for alpha-and beta-gamma-emitting nuclides should apply independently, bAs used in this table, dpm (disintegrations per minute) means the rate of emission by radioactive material as determined by correcting the counts per minute observed by an appropriate detector for background, cfDciency, and geometric factors associated with the instrumentation.

cMeasurements of average contaminant should not be averaged over more than I square meter. For objects of less surface area, the average should be derived for each such object.

2 dThe maximum contamination level applies to an area of not more than t00 cm,

2

'The amount of removable radioactive material per 100 cm of surface area should be determined by wiping that area with dry filter or soft absorbent paper, applying moderate pressure, and assessing the amount of radioactive material or, the wipe with an appropriate instrument of known efficiency. Men removable contamination on objects of less surface area is determined, the pertinent leveh should be reduced proportionally and the entire surface should be wiped.

~

Maximum Radiati0n Level (at 1 meter from surface) 5 llR/hr (above backoround)

N e

n E

T5 5

l t

n Pa9e 1 of 3 Table V CONCRETE INVENTORY TO VPI LANDFILL Piece No Description Weight (lbs.)

1 Pool Nose 6600 2

SW Corner 4600 3

S Side W - E #1 5800 4

S Side W - E #2 6000 5

W End #1 6000 6

SE Corner 7000 7

S Side W-E #5 (1/2) 4000 8

S Side W-E (1/2) 5200 9

NW Corner 1500 10 SE Corner Partial 1000 11 NE Side E - W #1 6600 12 S Side W - E #4(10F2) 4000 13 S Side W - E #3 7600 14 S Side W - E #4(2 of 2) 2100 15 N Side E - W #2 (Part 1) 4200 16 N Pool Wall 2000 17 N Side E - W #2 (Part 2) 1500 18 NE Corner 7000 19 S Side W - E #5 (1 of 2) 4200 20 S Side W - E #5 (2 of 2) 4100 21 N Side E - W #3 (1 of 2) 3400 22 N Side E - W #3 (2 of 2) 4600 23 N Side E - W #4 (1 of 2) 3300 24 N Side E - W #4 (2 of 2) 3500 25 N Side E - W #3 (1 of 2) 2200 26 NW Corner (1 of 3) 5000 27 NW Side 4800 i

28 Bottom Pool (1 of 2) 5200 29 Bottom Pool (2 of 2) 4300 30 W Side Pool (1 of 2) 5600 31 W Side Pool (2 of 2) 4200 32 No Description 5200 33 No Description 8000 34 Bottom 7000 35 Bottom 3000 36 Bottom 1200 37 Bottom 4000 38 Hear S. W. Corner 1900 39 SE Corner (1 of 4) 900 40 Si Corner ( 2 of 4) 9000 41 NE Corner 8000 42 NE Corner 1500 0.

43 SE Corner (3 of 4) 1200 44 NW Corner (1 of 3) 1100 T-6

l i

Table V (Con't)

Page 2 of 3 CONCRETE INVENTORY TO VPI LANDFILL Piece No Description Weight 45 NW Corner (2 of 3) 1500 46 N Side E - W #1 5500 47 N Side W - E #1 4500 48 S Side (Rod Drive Area) 8000 49 N Side (Rod Drive Area)

(1 of 2) 4000 50 N Side (Rod Drive Area)

(2 of 2) 4000 216,600 lbs.

Total Weight (Large Blocks) 20 Drums concrete 9 650 lbs.............. + 13000 lbs.

35 Blocks of Misc. Concrete (majority of which were 150# or less)......... + 6000 lbs.

Thermal Column door residue.............

4 _ 700 lbs.

226,300 lbs.

Rock drill filings

..................+ 3,450 lbs.

Total Weight (Concrete)................ 239,750 lbs (119.5 Tons)

NOTE:

PER SECTION 3.2.6 EACH PIECE OF CONCRETE WAS SURVEYED IN A LOW BACKGROUND AREA AS IT WAS REMOVED. CONTACT AND IM DOSE RATE READINGS WERE TAKEN WITH A PRM-7 INSTRUMENT. RANDOM SURVEYS WITH THE REUTER-STOKES RSS-lli PIC WERE ALSO TAKEN FOR BACKUP. FIXED / LOOSE CONTAMINATION SURVEYS WERE PERFORMED USING AN OPEN WINDOW GM DETECTOR. MATERIALS THAT DID NOT SATISFACTORILY MEET RELEASE CRITERIA (TABLE IV)

WERE PACKAGED AND DISPOSED OF AS RADI0 ACTIVE WASTE.

O T-7 l

Table V (Con't)

Page 3 of 3 Below Reactor Grade Soil Disposed at V.P.I. Landfill Dates: January 15, 16, 1987 Dose Rate (uR/hr - measured w/PRM-7)

Pallet (1)

Weight (lbs)

CONTACT 0 1 METER 1

900 5-7 5

2 850 5-7 5

3 900 5-7 5

4 900 5-7 5

5 850 5-7 5

6 800 5-7 5

5200 lbs.

( }

(2.36 metric tons)

(1) NOTE:

Each lot of soil removed was placed on a 4' X 6' flat wood

" pallet", spread out : anly to a 4-6" depth, and surveyed / sampled prior to release. Soil that exceeded Table IV limits was packaged and disposed of as Radioactive Waste.

I l

T-8 O

r TABLE ~VI 6

> j\\ '

V.P.I. 01D FROJECT RADIDACTIVE WASTE SHIFFENT SUMMRY JAN '67

\\.

====...........============================.....it._t=============.......============================

SHIPMENT : DATE

VOLUME : NO. of ACTIVITY : WEIGHT :

k0.

! SHIFPED (Ft-3) : FACKAEES !

(Ci) i (Lbs.) :

INVENTORY ~

I

20-Nov-86 !

480.0 '

5l 2.07 : 40,950 : Reactor Graphite, Alue. Fuel Boxes &

i J Control Rod Drives, Concrete Closure -

l-l l

l l

.' blocks,' Misc. Steel Components, Piping, l

Wood, and Laboratory' trash.

2

8-Dec-66 :

359.4 :

5 2.45 : 43,900 Activated Concrete Rubble,' Steel Rebar,

. Core base plate, Concrete Closure Blocks,

-l and Misc. Wood, Lab. trash.

I 3

Il-Dec-86 :

490.0 5:

3.61 : 41,200 i Activated Concrete Rubble, Steel Rebar, i

Reactor Graphite. Misc. Wood and Lab.

trash.

4

-l 19-Dec-86 :

494.0 l 5:

3.90 : 42,800 Activated Concrete, Rubble Steel Rebar.

-l l

Activated Soil, Misc. Wood, piping, and Lab, trash.

5 16-Jan-87 l 210.5 :

16 :

1.05 l 20,250 Activated Concrete Rubble, Steel Rebar, Activated Soil, and Misc. Lab. trash.

TOTALS:

2062.9 :

36 l 13.1 189,100 :

I h0TE(S):

All Packa es were properly classifed as Radioactive LSA, TYPE A, b

ClassA-U!stableWastefordisposalatCNSI'slowLevelWaste i

Site located at Barnwell, SC Fredominant Radionuclide concentrations included :

1 (in decending order) H-3, C-14, Eu-152, K-40, Co-60, & Tc-99 All Shipments were made ' Exclusive Use'.

All Shipments were accepted and buried at the LLW Site.

==============================...;

_s............============22.....========================================

r 9

4

(

l T

e

- END PA6E 1 of 1 4

T-9

.___.7_,_.-._y.

mm...

p

l Q) f Page 1 of 2 TABLE VII Liquid.(H 0) Discharges To Sanitary Sewer System 2

Activity (uC1)

Date Volume (gal)

Source (Eu152)

(Eul54)

(CoS7)

(Co60) 10/28/87 30 14" Coring ops.

0.495 0.146 0.068 10/28 50 Dumptank Background

10/28 30 4" Coring ops.

Background

10/29 5

Sump & Core Drain

Background

10/29 30 Coring ops.

Background

10/29 50 14" Coring ops.

Background

10/29 30 Wire Saw ops.

0.363 0.129 0.060 0.177 10/29 55 Wire Saw ops.

0.577 0.204 0.095 2.03 0.221 i

10/29 30 Wire Saw ops.

10/30 30 Wire Saw ops.

0.231 0.088 0.041 10/30 30 4" Coring ops.

Background

10/30 50 4" Coring ops.

0.958 0.246 0.115 10/30 30 Wire Saw ops.

0.149 0.110 0.051 10/30 30 4" Coring ops.

Background

10/30 55 14" Coring ops.

Background

O 10/30 30 Wire Saw ops.

0.101 0.074 0.035 10/31 50 Wire Saw ops.

Background

10/31 50 Wire Saw ops.

Background

10/31 50 Wire Saw ops.

Background

10/31 50 Wire Saw ops.

Background

10/31 30 Wire Saw ops.

Background

10/31 30 Wire Saw ops.

Background

10/31 30 4" Coring ops.

Background

10/31 50 4" Coring ops.

Background

11/3 539 Wire Saw ops.

Background

11/3 30 Wire Saw ops.

Background

11/3 30 Wire Saw ops.

Background

11/3 30 Wire Saw ops.

Background

11/3 30 Wire Saw ops.

Background

11/3 55 Wire Saw ops.

Background

11/3 50 Wire Saw ops.

Background

11/4 50 Wire Saw ops.

background 11/4 50 Wire Saw ops.

Background

11/4 30 Wire Saw ops.

Background

11/4 30 Wire saw ops

Background

11/6 82 Saw/ Coring ops.

Background

11/7 20 Saw/ Coring ops.

Background

11/7 50 Saw/ Coring ops.

Background

11/8 30 Saw/ Coring ops.

Background

11/10 15 saw/ Coring ops.

Background

11/10 50 Process pit tank

Background

O 11/11 15 Saw/ Coring ops.

Background

T-10 l

Page 2 of 2

~

TABLE VII (Con't) 2 Liquid (H 0) Discharges To Sanitary Sewer System

~

i Activity (uci)

Date Volume (gal)

Source (Eu 54)

(CoS7)

(Co60) 1 (Eul52) 11/11 10 Process pit tank Bkgd*

11/12 2

E. Fuel Pit #15 Bkgd 11/12 15 Saw/ Coring ops.

Bkgd 11/12 1

E. Fuel Pit #13 &l4 Bkgd 11/19 29 Saw/ Coring ops.

Bkgd 0.122 11/21 5

Process pit tank 0.137 11/24 3

HX Core 0.094 0.074 12/1 29 Saw/ Coring ops.

Bkgd*

12/14 60 Saw/ Coring ops.

Bkgd Totals:

2255 gal.(8.54E+6 ml) 3.11 1.00 0.47 2.62 Average Activity: 8.43E-7uCi/ml Total Combined activity: 7.20 uCi i

- No activity above Background as measured on Nuclear Data System-66 MCA calibrated for Marinelli configuration.

l (See calibration data in log book).

l 1

O T-ll f

-..,-.,,,.n

, ~ -., _ -. ~.,.

m..

O TABLE VIII RADIATION DETECTION INSTRUMENTS WINDOW TYPE OF RADIATION SENSITIVITY THICKNgSS INSTRUMENT NUMBER DETECTED RANGE (mg/cm1 USE Eberline E520 2

Beta, Gamma 0-2K mR/hr 30 Survey Eberline F120 2

Beta, Gamma 0-50K cpm 1.4-2.0 Survey Eberline RM 14 2

Beta, Gamma C-50K cpm 1.4-2.0 Survey 4

j Eberline RAS-1 2

N/A N/A N/A Air Sampler Eberline PAC 4G 2

Al pha 0-50K cpm 1.4-2.0 Survey Eberline PRM-7 1

Gamma 0-5 mR/hr N/A Survey Ludlum 19 1

Gamma 0-5 mR/hr 30 Survey Staplex 2

Al pha, Beta,

N/A N/A High Volume Gamma Air Sampler Teletector 1

Beta, Gamma 0-lK R/hr N/A Survey i

Victoreen 495 2

Beta, Gamma 0-50K cpm 1.4-2.0 Survey (Frisker)

Keithley 36100 1

Beta, Gamma 0-20 R/hr Survey Victoreen 856-20 3 Gamma 0.1-10 4 N/A Area mR/hr Monitors Reuter Stokes 1

Gamma 1-999 uR N/A low Level PIC Gamma Surveys Canberra 2402 1

Beta, Gamma 0-999Kcrm N/A Auto Swipe Counter Nuclear Data 1

Gamma N/A N/A Gamma Spectra MCA System 66 Analysis Eberline MS-2 1

Beta, Gamma 0-9999 Counts 1.4-2.0 Scaler 6

o T-12

l

]

A V.P.I. DECOMMISSIONING FINAL REPORT T

l APPENDIX A FINAL SURVEY GRID MAPS (9 PAGES)

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4 APPENDIX P SAf1PLE PAGES V.P.I. FINAL RELEASE SURVEY (5 PAGES)

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GRID SURVEY FORM

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CONTAMINATION /R ADIATION SURVEY REPORT CONTINUATION SHEET U*E AC rivirv e Loc ^ on (pott D.SO L 4 L<f.5 - c~ga m Yl%s&lA AnA4TCC++s)tLTamTVTL s

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V.P.I.

FINAL SURVEY DATA - HIGH POTENTIAL AREAS JAN '87 m

l

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'(_/

GRID SURVEY FORM V.P.I.

FINAL SURVEY JAN '87 AREA SURVEYED:

ROOM 10 FLOOR sRX AREA)

BY: J.'Ri8di/R.. Montgomery GRID GAMMA BETA-BETA-BETA-ALPHA ALPHA ALPHA REMOVABLE

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EXPOSURE GAMMA GAMMA GAMMA SURVEY SURVEY CONT A M.-

CONTAM.

Pt.

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(NET)

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RS-1

1. D. :

a.

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c.

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0 0.00 d.

(NET) 60 10 517 0

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e.

60 10 517 0

0 118.25 f.

100 50 2585 0

0 GAMMA:

AVERAGE:

N/A-N/A 862 N/A N/A O

RS-2 I. D. :

a.

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e.

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AVERAGE:

N/A N/A 862 N/A N/A 0

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N/A N/A 862 N/A N/A O

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i END PAGE 1 OF 19 Page 3 Of 5 J

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Sample Results as Measured on the ND-66 MCA System:

Sample Point Activity (Co60; pCi/9)

A-1 34.0 A-2 82.8 A-3 23.8 B-1 ND C-1 ND D-1 1.7 E-1 2.1 F-1 ND G-1 ND H-1 ND I-1 ND J-1 ND K-1 ND L-1 ND O

C/

Page 5 of 5 n

i.

A

! /

V.P.I. DEC0t9tISSI0t!ING fit!AL REPORT

,. y f

t i

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f b

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4 i

APPENDIX C

SUMMARY

OF RELEASE SURVEY DATA j

(FROM COMPUTER DATABASE) t (9 PAGES) 4

G i

i i

t i

i i

I 1

1

! 1

V. P. I. FINAL RELEASE SURVEY.- OVERALL DATA

SUMMARY

JAN '87 s

(

)

\\_/

============================================================

L(1) --SURVEY OF CEILINGS IN ROOKS 6, 6A, 8,

& 8A j

(1 ORIGINAL DATA PAGE; [5] 3M x 3M GRIDS)

============================================================

SUMMARY

- ROOM 6, 6A, 8

& 8A CEILINGS :

i FIXED CONTAMINATION :

BETA-GAMMA ALPHA Avg. [ net] cpm (pts. [a-e]) -

3.2 0.0 Equivalent dpm/100 cur 2 165.4 0.0 1

Max. [ net] cpm (@ pt. C f) )

10 0

Equivalent dpm/100 cur 2 517.0 0.0 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. [f])

4.3 (dpm/100 cur 2) 0.8 Maximum (taken @ pt. [ f] )

21.5 (dpm/100cm-2 )

4.2 GAMRA DOSE RATE (at 1 Meter) :

Note:

(1) Grid was surveyed for Average (measured @ pt. [al)-

0.4 uR/hr Alpha Contam.

'N Maximum (measured @ pt. Ca))-

2.0 uR/hr

============================================================

L(2) - SURVEY OF ROOM 10 CEILING (2 ORIGINAL DATA PAGES; [10) 3M x 3M GRIDS)

============================================================

SUMMARY

- ROOM 10 CEILING SURVEY :

FIXED CONTAMINATION :

BETA-GAMMA ALPHA

' Avg. [ net) cpm (pts. [a-e]) -

1.0 0

Equivalent dpm/100cmm2 51.7 0.0 Max. [ net) cpm (@ pt. [ f] )

30 0.0 Equivalent dpm/100 cur 2 1550.9 0.0 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. [ f] )

28.0 (dpm/100 cur 2) 0.0 Maximum (taken @ pt. [f3) 96.8 (dpm/100 cur 2) 0.0 GAMMA DOSE RATE (at 1 Meter) :

Note:

(1) Gr1d was surveyed for Average (measured @ pt. Ca])-

0.0 uR/hr Alpha Contam.

h Maximum (measured @s /====================pt.

[a1>-

O.O uR/hr m

========================================

F#;D PAGE 1 OF 9

V.P.I.

FINAL RELEASE SURVEY - OVERALL DATA

SUMMARY

JAN '87 C

5s

============================================================

L (3 ) - SURVEY OF ROOM 108 (ALL ARFf.d)

(4 ORIGINAL DATA PAGES; [17] 3M x 3M GRIDS)

============================================================

SUMMARY

- ROOM 108 SURVEY :

FIXED CONTAMINATION :

BETA-GAMMA ALPHA Avg. [ net) cpm (pts. [a-e]) -

16.5

.0 Equivalent dpm/100 cur 2 851.5 0.1 j

Max. [ net) cpm (@ pt. [ f) )

90 1

Equivalent dpm/100cm-2 4652.6 9.5 4

REMOVABLE CONTAMINATION :

BETA ALPHA I

Average (taken @ pt. [ f] )

58.8 (dpm/100cm-2 )

0.2 j

Maximum (taken @ pt. [f3) 344,0 (dpm/100cm-2) 4.2 GAMMA DOSE RATE (at 1 Meter) :

Note:

(2) Grids were l

surveyed for Average (measured @ pt. [al)-

0.2 uR/hr Alpha Contam.

i Maximum (measured @ pt. [a])-

1.5 uR/hr

============================================================

l L(SUM) - SURVEY OF AREAS HAVING LOW POTENTIAL FOR CONTAMINATION i

(7 ORIGINAL DATA PAGES; [323 3M x 3M GRIDS TOTAL)

============================================================

j

\\

1

SUMMARY

- SURVEY OF LOV POTENTIAL AREAS :

l FIXED CONTAMINATION :

BETA-GAMMA ALPHA l

6.9 0

Avg. [ net) cpm (pts. [a-e])

Equivalent dpm/100cm-2 356.2

.0 Max. [ net) cpm (@ pt. [f1) 90 1

Equivalent dpm/100 cur 2 4652.6 9.5 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. [ f] )

30.3 (dpm/100cm-2) 0.4 344.0 (dpa/100cm-2) 4.2 Maximum (taken @ pt. [ f) )

GAMMA DOSE RATE (at 1 Meter) :

Notes (4) G' rids were surveyed for

[

. Average (measured @ pt. [al)-

0.2 uR/hr Alpha Contam.

i Maximum (measured @ pt. [al)-

2.0 uR/hr

============================================================

END PAGE 2 OF 9

V. P. I. FINAL RELEASE SURVEY - OVERALL DATA

SUMMARY

J AN ' 87 (3

============================================================

j M(1) - SURVEY OF ROOMS 6 & 6A (VALLS & FLOORS)

(4 ORIGINAL DATA PAGES; [ 163 2M x 2M GRIDS)

============================================================

SUMMARY

- ROOM 6 & 6A (VALL & FLOOR) SURVEYS :

===______________________-__________

FIXED CONTAMINATION :

BETA-GAMMA ALPHA Avg. [ net] cpm (pts. La-el) -

3.9

.0 Equivalent dpm/100 car 2 200.3 0.1 Max. [ net] cpm (o pt. C f))

50 1

Equivalent dpn/100cnr2 2584.8 9.5 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. [ f] )

16.3 (dpm/100cm-2) 0.8 Maximum (taken e pt. [f])

86.0 (dpm/100cm-2 )

8.3 GAMMA DOSE RATE (at 1 Meter) :

Note:

(2) Grids were surveyed for Average (measured @ pt. [a))-

0.3 uR/hr Alpha Contam.

O= Maximum (measured @ pt. [al)-

1.5 uR/hr

===========================================================

============================================================

M(2) - SURVEY OF ROOMS 8 & 8A (VALLS & FLOORS)

(3 ORIGINAL DATA PAGES; [ 14 ) 2M x 2M GRIDS)

============================================================

SUMMARY

- ROOM 8 & 8A (VALL & FLOOR) SURVEYS :

FIXED CONTAMINATION :

BETA-GAMMA ALPHA

' Avg. [ net] cpm (pts. [a-e]) -

6.6 0.0 Equivalent dpm/100cm-2 339.7 0.0 Max. [ net) cpm (e pt. [ f) )

30 0

Equivalent dpm/100cm-2 1550.9 0.0 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken G pt. Cf3) 8.9 (dpm/100 cur 2) 0.3 Maximum (taken e pt. [f))

75.3 (dpm/100cm-2 )

4.2 GAMMA DOSE RATE (at 1 Meter) :

Note:

(2) Grids were surveyed for Average (measured @ pt. [al)-

0.4 uR/hr Alpha Contam.

[ )/= Maximum (measured @ pt. [al)-

3.0 uR/hr

\\-

===========================================================

END PAGE 3 OF 9

e V.P.I.

FINAL RELEASE SURVEY - OVERALL DATA

SUMMARY

JAN '87 r"%

============================================================

M(3) - SURVEY OF ROOM 10 VALLS (6 ORIGINAL DATA PAGES; [301 2M x 2M GRIDS)

+

============================================================

SUMMARY

- ROOM 10 VALL-SURVEY :

FIXED CONTAMINATION :

BETA-GAMMA ALPHA Avg. [ net] cpm (pts. [a-el) -

3.4

.0 177.5

.0 Equivalent dpm/100cm-2 Max. Enet] cpm (@ pt. [ f] )

90 1

Equivalent dpm/100cm-2 4652.6 9.5 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. [f])

31.9 (dpm/100 cur 2) 0.6 Maximum (taken @ pt. [f))

161.3 (dpm/100cm-2 )

8.3 GAMMA DOSE RATE (at 1 Meter) :

Note:

(3) Grids were surveyed for Average (measured @ pt. [al)-

.0 uR/hr Alpha Contam.

A(

Maxiraum (measured @ pt.

[a1>-

O.5 uR/hr j========================================================================

============================================================

M(4) - SURVEY OF ROOM 10 FLOOR (7 ORIGINAL DATA PAGES; [30] 2M x 2M GRIDS)

============================================================

SUMMARY

- ROOM 10 FLOOR SURVEY :

FIXED CONTAMINATION :

BETA-GAMMA ALPHA Avg. [ net) cpm (pts. [a-e]) -

6.8

.0 Equivalent dpm/100cm-2 351.5 0.1 Max. [ net] cpm (@ pt. [ f] )

90 1

Equivalent dpm/100cm-2 4652.6 9.5 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. [f3) 20.7 (dpm/100cm-2) 1.1 Maximum (taken @ pt. [ f] )

68.4 (dpm/100cm-2 )

8.3 l

GAMMA DOSE RATE (at 1 Meter) :

Note:

(3) Grids were surveyed for Average (measured @ pt. [al)-

O.1 uR/hr Alpha Contam.

Maximum (measured @

[a3>-

O.5 uR/hr

\\

'====================pt.

========================================

END PAGE 4 OF 9

V.P.I.

FINAL RELEASE SURVEY - OVERALL DATA

SUMMARY

J AN ' 87

.O V

=====================r.======================================

M(SUM) - SURVEY OF AREAS HAVING MEDIUM POTENTIAL FOR CONTAMINATION (20 ORIGINAL DATA PAGES; [90] 2M x 2M GRIDS TOTAL)

============================================================

SUMMARY

- SURVEY OF MEDIUM POTENTI AL AREAS :

FIXED CONTAMINATION :

BETA-GAMMA ALPHA Avg. [ net] cpm (pts. [a-e]) -

4.6

.0 Equivalent dpm/100cm-2 239.2 0.1 Max. [ net] cpm (G pt. [ f])

90 1

4652.6 9.5 Equivalent dpm/100cm-2 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. [ f] )

19.0 (dpm/100 car 2) 0.5 Maximum (taken @ pt. [f])

161.3 (dpm/100cm-2) 8.3 GAMMA DOSE RATE (at 1 Meter) :

Note:

(10) Grids were surveyed for Average (measured @ pt. [al)-

0.2 uR/hr Alpha Contam.

r

Maximum (measured @ pt. [al)-

3.0 uR/hr

============================================================

x_

l

===============n============================================

H(1) - SURVEY OF ROOM 10 FLOOR (Reactor Shield area)

(8 ORIGINAL DATA PAGES; [35] 1M x 1M GRIDS)

============================================================

SUMMARY

- ROOM 10 FLOOR (Reactor Shield Area) :

FIXED CONTAMINATION :

BETA-GAMMA ALPHA

" Avg. [ net) cpm (pts. [a-el) 14.2

.0 Equivalent dpm/100cm-2 732.6 0.1 Max. [ net) cpm (@ pt. [ f1 )

130 1

Equivalent dpm/100cm-2 6720.4 9.5 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. [f])

16.2 (dpm/100 car 2) 0.6 Maximum (taken @ pt. Cf])

172.0 (dpm/100 car 2) 4.2

~

GAMMA DOSE RATE (at 1 Meter) :

Note:

(4) Grids were surveyed for Average (measured @ pt. [al)-

0.1 uR/hr Alpha Contam.

(

Maximum (measured @ pt. Ca))-

1. 0 uR/hr s._

============================================================

END PAGE 5 OF 9

V.P.I.

FINAL RELEASE SURVEY - OVERALL DATA

SUMMARY

J AN ' 87 A

i 4

'%.)

============================================================

H(2) - SURVEY OF ROOM 10 FLOOR (Reactor " Pit" Area)

(4 ORIGINAL DATA PAGES; [20) 1M x 1M GRIDS)

============================================================

SUMMARY

- ROOM 10 FLOOR (Reactor " Pit" Area) SURVEY :

==_______________________________

FIXED CONTAMINATION :

' BETA-GAMMA ALPHA Avg. [ net) cpm (pts. [s-e]) -

25.7 0

Equivalent dpm/100cm-2 1780.8 0.1 Max. Inet) cpm (0 pt. [ f) )

93 1

Equivalent dpm/100cm-2 6451.6 9.5 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. [ f] )

21.2 (dpm/100 cur 2) 1.0 Maximum (taken @ pt. C f])

129.0 (dpn/100cm-2) 8.3 GAMMA DOSE RATE (at 1 Meter) :

Note:

(2) Grids were surveyed for Average (measured G pt. Cal)-

0. 8 uR/hr Alpha Contam.

(measured e (a))-

4.5 uR/hr

("$gMaximum)====================pt.

========================================

g

============================================================

H(3) - SURVEY OF PROCESS PIT AREA (7 ORIGINAL DATA PAGES; [31] 1M x 1M GRIDS)

============================================================

SUMMARY

- PROCESS PIT AREA SURVEY :

FIXED CONTAMINATION :

BETA-GAMMA ALPHA Avg. [ net] cpm (pts. ta-e]) -

13.9

.0 Equivalent dpm/100 car 2 717.1 0.1 Max. [ net) cpm (@ pt. [f3) 150 1

Equivalent dpm/100cm-2 7754.3 9.5 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. (f3) 74.2 (dpm/100cm-2 )

0.0 Maximum (taken @ pt. [f3) 279.5 (dpm/100 car 3)

0. 0 GAMMA DOSE RATE (at 1 Meter) :

Note:

(4) Grids were surveyed for Average (measured e pt. (a))-

0.1 uR/hr Alpha Contam.

s-iMaximum (measured G pt.

[a3>-

1. O uR/hr

============================================================

END PAGE 6 OP 9

I V.P.I.

FINAL RELEASE SURVEY - OVERALL DATA

SUMMARY

JAN '87

(\\_ /

============================================================

H(4) - SURVEY OF PROCESS PIT SUMP (1 CRIGINAL DATA PAGE; [5] < 1M x 1M GRIDS)

=====================================================b======

SUMMARY

- PROCESS PIT SUMP SURVEY :

_______________===-- ___________________________________________________

FIXED CONTAMINATION :

BETA-GAMMA ALPHA Avg. [ net) cpm (pts. [a-e]) -

15.7 0.0 Equivalent dpm/100cm-2 1087.8 0.0 Max. Cnet] cpm (@ pt. C f])

34 0

Equivalent dpm/100cm-2 2358.7 0.0 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken Q pt. [f])

1. 0 (dpm/100cm-2 )

0.8 Maximum (taken @ pt. [ f] )

4.8 (dpm/100cm-2 )

4.2 GAMMA DOSE RATE (at 1 Meter > '

Note:

(1) Grid was surveyed for Average (neasured Q pt. [a))-

0. 0 uR/hr Alpha Contam.

Maximum (measured Q pt. [al)-

0.0 uR/hr

(

============================================================

H(SUM) - SURVEY OF AREAS HAVING HIGH POTENTIAL FOR CONTAMINATION (20 ORIGINAL DATA PAGES; [91) 1M x 1M GRIDS TOTAL)

============================================================

SUMMARY

- SURVEY OF HIGH POTENTIAL AREAS :

FIXED CONTAMINATION :

BETA-GAMMA ALPHA

' Avg. [ net) cpm (pts. ta-e]) -

17.3

.0 Equivalent dpm/100cm-2 1079.6 0.1 Max. [ net] cpm (G pt. [f1) 150 1

Equivalent dpm/100 car 2 7754.3 9.5 REMOVABLE CONTAMINATION :

BETA ALPHA Average (taken @ pt. Cf3) 28.1 (dpm/100 car 2) 0.6 Maximum (taken @ pt. [ f] )

279.5 (dpm/100cm-2) 8.3 GAMMA DOSE RATE (at 1 Meter) :

Note:

(11) Orids were surveyed for Average (measured @ pt. [al)-

0.3 uR/hr Alpha Contam.

Maximum (measured @ pt. [al)-

4.5 uR/hr

,(_,========================================================================

END PAGE 7 OF 9

V. P. I.

FINAL RADIATION SURVEY

SUMMARY

(Measured w/PRM-7 @1M from Surfaces)

,r

~5 V

3Mr3M (L) AREAS S.0 2Mx2M (M) AREAS Y

ff h

e 4.o s

s w

3.0 -

6 3

E.

mo 2.0 l,

i l

i i l

l 1

l l

J l J lu l J dJ a.

s BKG (1) (2) (3)

BKG (1) (2) (3) (4)

BKG (1) (2) (3) (+)

tg gg O AT ON-See Overoll Summ y OV FIXED CONTAMINATION SURVEY

SUMMARY

Contact Measurements w/HP-210T Probe 9

<2 t uvi u (H) ARFA9 g

0 3

.'r 7

h L.

3 E

M)Y 2

2 b

0 4

5

/- 2:e (t) Anta:

cu a uqu;.atAc h

N N

N 4

f, i

1 1

i l l i

i a

i l 1

1

. 1 2

2 s i i

s i l 1 6 1 _i h,m k

b 51_.

M n

G n

v o

(1) (2) (3)

(1) (2) (3) (4)

,gOgON,-See OveroH Summgy tg g,

7_

REMOVABLE CONTAMINATION SURVEY

SUMMARY

100cm-2 Smears Token @ Survey Pt [F]

500

~.

C).

400 3

3Mx3M (L) AREAS 300 iivix iivi (H) AREA 3 8

7 h

N 200 s

8 2Mx2M (M) AREAS E'

O 300 n

7 s.

l 1 l 1

I^

0 T

(1) (2) (3)

(1) (2) (3) (4)

O ON-See Overall Summ y tg gg O

ALPHA CONTAMINATION SURVEYS-

SUMMARY

MAXIMUM VALUES, FIXED / REMOVABLE 14.0 13.0 12.0

, 11.0 10.0 -JWJu--(L)-AREAS 2W2M-(M)-AREAS

? WW (H) * *EAS g

3 3

3 90 N

A k

k k

A m

a m

m oo E

b Y

hh $b b

b b

yo h

d d N d

N d eo 5

b be bk kk b

be k I

3o L

5%

2# B#

L f# 2

.o 2

R2 22 2 22 23 22 M R 2

3.o s

33 Es

$ 33 33 33 33 3 s

2.0 S

2%

R#

R# 2%

R2 22 2 C7 o;o

,o (1) (2) (3)

(1) (2) (3) (4)

DIRECT A

3) h SMEAYt S$JYkY 4

l V.P.I. DEC0i4tISSIONING FINAL REPOP.T l

I APPENDIX D SURVEY INSTRUMENT DATA (2 PAGES)

O 1

t FINAL RELEASE SURVEY SURVEY INSTRUMENT DATA p)

INSTRUMENT MFG./

SERIAL CAL DUE INST.

CORRECT DETECTOR

(

TYPE MODEL NO.

DATE EFF.

FACTOR

BKG.(s)

Micro-R Eberline 476 8-4-87 N/A N/A

.Page#/(uR/hr)

PRM-7 value based on survey location (SEE GRAPH - BELOV)

Beta-Alpha Canberra 786143 N/A B= 19.0%

1.00 B= 4.0 cpm (Scaler) 2402 (Auto)

A= 24.0%

A= 0.0 cpm Beta-Gamma Eberline 1130 2-15-87 0.09 69.37 25 (Scaler)

MS-2 (9.3 %)

cpm Beta-Gamma Eberline 2699 5-5-87 0.12 53.76 50 (Frisker)

RX-14 (12.0 %)

cpm Eberline 5813 4-24-87 0.13 49.63 50 E-120 (13.0 %)

cpm AVG. VALUES ASSIGNED for FRISKERS:

51.70 50 cpm Alpha Eberline 3890 3-25-87 0.21 9.52 0

PAC-4G-3 (21.0 %)

cpm A General Area Radiation Survey w/the PRM-7 was performed in each Survey Location to establish a " Background" dose rate value for that area.

,A These values and calculated averages are plotted below :

V.P.I.

FINAL RADIATION SURVEY PRM-7 Instrurnent (Detector) Bockground 11.0 -

10.0 -

90-x 8.0 -

x Ep 7.0 -

X 3

6.0 -

V V V V V A

X X X c

y 5.0 -- : : : : : : : : A 4.0 -

A 4

a 3.0 -

a x a a a a a

a a a A a a a

x x

a 20-r v x

x x x x a

1.0 -

O V

0.0 Number Of Background Measurements AVG; 1 M AREAS,a

- AREA SURVEYED (INSIDE FACluTY) +

AVG; 2M AREAS,x M AVG; 3M AREAS,9

[

SURVEY INSTRUMENT DATA - FINAL RELEASE SURVEY

................................................ =..................=...

- CORRECTION FACTOR (C.F.) CALCULATIONS

________________L________________

BETA-0AMMA "FRISKER" PROBE (Eberline HP-210T or HP-260):

(Thin-Window " Pancake" GM; Window Size = 15.5 cur 2)

FOR RM-14:

C.F.=

1/ EFFICIENCY-C.F.

x Counts (NET)

Activity (DPM)

=

C.F.=

1/12.0 %

(DPM/ Probe Area)

C.F.=

8.33 or (DPM/15.5cm-2)

C. F.

for 100cm-2 area =

C. F.

x 100/15. 5c m-2

=

8.3 x

6.5 53.76

=

FOR E-120:

C.F.=

1/ EFFICIENCY C.F.

x Counts (NET)

Activity (DPM)

=

C.F.=.

1/13.0 %

(DPM/ Probe Area)

C.F.=

7.69 or (DPM/15.5cm-2)

C.F.

for 100cm-2 area =

C. F.

x 100/15.5cm-2 7.7 x

6.5

=

49.63

=

O

(,j FOR MS-2:

C.F.=

1/EFFICIENCI C.F.

x Counts (NET)

=

Activity (DPM)

C.F.=

1/9.3 %

(DPM/ Probe Area)

O.F.=

10.75 or (DPM/15.5cm-2)

C. F.

for 100cm-2 area =

C. F.

x 100/15.5cm-2 7.7 x

6.5

=

69.37

=

ALPHA SENSITIVE GAS FLOW PROPORTIONAL PROBE (Eberline AC-21)

(Thin-Vindow, Size = 50cm-2)

FOR PAC-4G-3:

C.F.=

1/ EFFICIENCY C.F.

x Counts (NET)

Activity (DPM)

=

C.F.=

1/21.0 %

(DPM/ Probe Area)

C.F.=

4.76 or (DPM/50cm-2)

C.F.

for 100cm-2 area =

C.F.

x 100/15.5ca-2 21.0 x

2.0

=

9.52

=

============================================================

a END PAGE 2 of 2

__

ML20206D610

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