ML19337A782

or State Univ Triga Reactor Annual Rept for Jul 1979-June 1980

ML19337A782
Person / Time
Site:	Oregon State University
Issue date:	08/31/1980
From:	Anderson T, Sharon Bennett, Andrea Johnson OHIO STATE UNIV., COLUMBUS, OH, Oregon State University, CORVALLIS, OR
To:
Shared Package
ML19337A781	List: ML19337A780 ML19337A782
References
NUDOCS 8009300197
Download: ML19337A782 (121)

Text

. .

. OREGON STATE UNIVERSITY TRIGA REACTOR ANNUAL REPORT To caticfy the requirements of:

A. U.S. Nuclear Regulatory Commission License R-106 (Docket No. 50-243),

Section 6.7(e) of the Technical Specifications, for the period July 1, 1979 through June 30, 1980.

B. U.S. Department of Energy Fuel Fabrication Contract No. EY-76-C-06-1953, for the period July 1, 1979 through June 30, 1980.

C. Oregon Department of Energy, DOE Rule No.30-010, for the period July 1, 1979 through June 30, 1980.

\ Written by:

T.V. Anderson, Reactor Supervisor

~

A.G. Johnson, Health Physicist S.L. Bennett, Radiation Specialist J.C. Ringle, Assistant Reactor Administrator Submitted by:

C.H. Wang, Reactor Administrator

g. Radiation Center

\. Oregon State University Corvallis, Oregon 97331 Telephone: 503-754-2341 August 31, 1980 8009300bi

1 TABLE OF CONTENTS l

-, page I. INTRODUCTION AND

SUMMARY

................. I-1 e A. INTRODUCTION TO OREGON STATE TRIGA REACTOR ANNUAL REPORT. . I-l l

B.

SUMMARY

OF OSTR USE DURING REPORTING PERIOD . . . . . . . . I-l C.

SUMMARY

OF OSTR ENVIRONMENTAL AND RADIATION PROTECTION DATA I-4

1. Liquid Waste Data . . . . . . . . . . . . . . . . . . . I-4

2. Gaseous Waste Data. . . . . . . . . . . . . . . . . . . I-5  ;

l

3. Solid Waste . . . . . . . . . . . . . . . . . . . . . . 1-5

4. . Radiation Exposure Received by Facility Personnel and Vis i tors ( i n mrem) . . . . . . . . . . . . . . . . . . . I-6 l

5. Number of Area and Offsite Environmental Monitoring j Samples Evaluated . . . . . . . . . . . . . . . . . . . I-6  :

II. GENERAL INFORMATION ................... II-l A. RADIATION CENTER. . . . . . . . . . . . . . . . . . . . . . II-l i B. FACU' TY MEMBERS HOUSED AT THE RADI ATION CENTER. . . . . . . II-2 C. RESEARCH PERSONNEL HOUSED AT THE RADIATION CENTER . . . . . II-3

1. Post-Doctorate Research Associates. . . . . . . . . . . II-3

2. Graduate Students . . . . . . . . . . . . . . . . . . . II-3

3. Visiting Scientists and Trainees. . . . . . . . . . . . II-4 D. CLASSIFIED STAFF AT THE RADIATION CENTER. . . . . . . . . . II-5 E. REACTOR OPERATIONS STAFF. . . . . . . . . . . . . . . . . . II-5

. F. REACTOR OPERATIONS COMMITTEE.,. . . . . . . . . . . . . . . II-6 G. RADIATION SAFETY COMMITTEE. . . . . . . . . . . . . . . . . II-6

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page III. OPERATIONAL DATA . . . . . . . . . . . . . . . . . . . . . . III-l

. A. REVIEW. . . . . . . . . . . . . . . . . . . . . . . . . . . III-l B. OPERATING STATISTICS. . ................. III-5 C. EXPERIMENTS PERFORMED . . . . . . . . . . . . . . . . . . . III-9 D. UN PL AN N E D S H UT DOWN S . . . . . . . . . . . . . . . . . . . . III-12 E. CHANGES IN FACILITY . . . . . . . . . . . . . . . . . . . . III-13

1. 10 CFR 50.59 Changes. . . . . . . . . . . . . . . . . . III-13

2. Other Changes . . . . . . . . . . . . . . . . . . . . . III-13

3. Planned Changes . . . . . . . . . . . . . . . . . . . . III-15 F. MAINTENANCE AND SURVEILLANCE. . . . . . . . . . . . . . . . III-17

1. Maintenance . . . . . . . . . . . . . . . . . . . . . . III-17

2. Tests and Inspections . . . . . . . . . . . . . . . . . III-18 G. REPORTABLE OCCURRENCES. . . . . . . . . . . . . . . . . . . III-18 IV, UTILIZATION DATA . . . . . . . . . . . . . . . . . . . . IV-l A. TEACHING PROGRAMS . . . . . . . . . . . . . . . . . . . . . IV-1 B. RESEARCH PROJECTS . . . . . . . . . . . . . . . . . . . . . IV-7 C. PUBLICATIONS RESULTING FROM OSTR OPERATIONS THAT WERE -

REPORTED TO THE RADIATION CENTER. . . . . . . . . . . . . . IV-13

1. Publi ca tions in Print . . . . . . . . . . . . . . . . . IV-13 1

2. Publications in Press . . . . . . . . . . . . . . . . . IV-14 l

3. Reports and Papers. . . . . . . . . . . . . . . . . . . IV-15

4. Papers Submitted to the S venth C TRIGA Owners Conference, March 2-5,1980, San " Jgos California ........ IV-15

.- D. COMMERCIAL OR NON 'cC/?[3:' TILIZATION. . . . . . . . . . . IV-16 E. PUBLIC RELATIONS. . . . . . . . . . . . . . . . . . . . . . IV-16 F. PLANNED CHANGES IN UTILIZATION. . . . . . . . . . . . . . . IV-16 l

-iii-Page V. ENVIRONMENTAL AND RADIATION PROTECTION DATA; JULY 1, 1979 - JUNE 30, 1980 . . . . . . . . . . . . . . V-1 A. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . . . V-1 B. A

SUMMARY

OF THE NATURE AND AMOUNT OF RADI0 ACTIVE EFFLUENTS RELEASED OR DISCHARGED TO THE ENVIRONS BEYOND THE EFFECTIVE CONTROL 0F THE LICENSEE, AS MEASURED AT OR PRIOR TO THE POINT OF SUCH RELEASE OR DISCHARGE . . . . . . . . . . . . V-1

1. Liquid Waste (summarized on a monthly basis) . . . . . V-1 (a) The radioactivity discharged during the reporting period based on the following: ... V-1 (1) The total estimated quantity of radio-activity released (in curies) ...... V-1 (2) The detectable radionuclides present in this waste .............. V-2 (3) An estimate of the specific activity for each detectable radionuclide present,

• ~ if the specific activity of the released material after dilution was greater than 1 x 10 7 microcuries/ cubic centimeter. . . V-2 (4) A summary of the total release (in curies) for each radionuclide determined in (2) abova for the reporting period, based on representative isotopic analysis . . . . . V-2 (5) The estimated average concentration of the released radioactive material at the point of release for the reporting period (in terms of microcuries/ cubic centimeter) and the fraction of the applicable MPC value . . . . . . . . . . . V-2 (b) The total volume (in gallons) of effluent water (including diluent) released durin period of release . . . . . . . . g each

....... V-2

-iv-Page

2. Gaseous Watte (summarized on a monthly basis) . . . . . V-2 (a) The radioactivity discharged during the reporting period based on the following: .... V-2 (1) The total estimated quantity of radio-activity released (in curies) determined by an appropriate sampling and counting method .................. V-3 (2) The detectable radionuclides present in this waste . . . . . . . . . . . . . . . V-3 (3) The total estimated quantity of argon-41 released (in curies) during the reporting period, based on data from an appropriate monitoring system ..... V-3

, (4) The estimated average atmospheric diluted concentration of argon-41 raleased during the reporting period (in terms of microcuries/ cubic centimeter) and the fraction of the applicable MPC value . . . . . . . . . . . . . . . . . . . V-3 (5) The total estimated quantity of radio-

. activity in particulate fonn with half-lives greater than eight days (in curies) released during the re-porting period, as determined by an appropriate particulate monitoring system .................. V-3 (6) The average concentration of radioactive particulates with half-lives greater than eight days (in microcuries/ cubic centimeter) released during the re-porting period .............. V-3

, (7) An estimate.of the average concentration of other significant radionuclides

!'* .present in the gaseous waste discharge i (in terms of microcuries/ cubic centi-meter) ~and the fraction of the applicable

. MPC value for the reporting period, if  ;

the estimated release was greater than 1 20% of the applicable MPC . . . . . . . . . V-4 l j

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3. Solid Waste (summarized on an annual basis) . . . . . . . V-4 (a) The radioactivity discharged during the reporting period based on the following: ..... V-4 (1). The total amount of solid waste packaged (in cubic feet) .......... V-4 (2) The detectable radionuclides present in this waste . . . . . . . . . . . . . . . . V-4 (3) The total radioactivity in the solid waste (in curies) . . . . . . . . . . . . . . V-4 (b) The dates of shipment and disposition (if shipped off-site) ................ V-4 C. AN ANNUAL

SUMMARY

OF THE RADIATION EXPOSURE RECEIVED BY FACILITY PERSONNEL AND BY VISITORS, IN TERMS OF THE AVERAGE RADIATION EXPOSURE PER INDIVIDUAL AND THE GREATEST EXPOSURE PER INDIVIDUAL FOR EACH OF THE TWO GROUPS ............,.............. V-4 D. AN ANNUAL

SUMMARY

OF THE RADIATION LEVELS AND THE LEVELS OF CONTAMINATION OBSERVED DURING ROUTINE SURVEYS PER-FORMED AT THE FACILITY, IN TERMS OF THE AVERAGE AND THE HIGHEST LEVELS ....................... V-5 E. THE LOCATION AND MAGNITUDE OF THE MAXIMUM MEASURED OR CALCULATED DIRECT RADIATION LEVEL IN UNRESTRICTED AREAS DUE TO DIRECT RADIATION FROM THE FACILITY, AND DIRECT

. RADIATION FROM FACILITY EFFLUENTS . . . . . . . . . . . . . . V-5

1. The Maximum Direct Radiation Level in Unrestricted Areas Due to Direct Rcuiation From the Facility . . . . V-5

2. The Maximum Direct Radiation Level in Unrestricted Areas Due to Direct' Radiation From the Facility E f fl uen ts . . . . . . . . . . . . . . . . . . . . . . . V-10

• F. AN ANNUAL

SUMMARY

OF THE GENERAL METHODS AND THE RESULTS OF ENVIRONMENTAL SURVEYS PERFORMED OUTSIDE THE FACILITY . . . V-17

1. The On-Site Environmental Monitoring Systems ...... V-17

2. The Off-Site Environmental Monitoring Systems . . . . . . V-22 N

-vi-LIST OF TABLES a,

TABLE TITLE PAGE Table III-l Four Year OSTR Statistics (using FLIP core) . III-2 Table III-2 OSTR Statistics with 20% Enriched Core ... III-4 Table III-3 Present OSTR Operation Statistics . . . . . . III-6 Table III-4A OSTR Use Time . . . . . . . . . . . . . . . . III-7 Table III-4B OSTR Use Time . . . . . . . . . . . . . . . . III-7 Table III-5 OSTR Multiple Use Time ........... III-8 Table III-6 Experiment Usage Vs. Project ........ III-ll Table III-7 Unplanned Scrams .............. III-12 Table III-8 Monthly Tests & Inspections . . . . . . . . . III 19

. Table III-9 Quarterly Tests & Inspections . . . . . . . . III 20 Table III-10 Semi-Annual Tests & Inspections . . . . . . . III-21 Table III-11 Annual Tests & Inspections ......... III-22 Table IV-1A OSTR Teaching Hours . . . . . . . . . . . . . IV- 2 Table IV-18 OSTR Operator Training Hours ........ IV- 2 Table IV-2 Statistics of Students in Nuclear Engineering i

and Nuclear Science Courses . . . . . . . . . IV- 3 i

Table IV-3 Other Educational Institutions Using OSTR . IV- 5 Table IV-4 Graduate Students Doing Thesis Research that Used the OSTR . . . . . . . . . . . . . . . . IV- 6 Table IV-5 OSTR Research Hours . . . . . . . . . . . . . IV- 8 Table IV-6 Summary of Oregon State University TRIGA

. Research Projects and Funding Agencies ... IV- 9 Table IV-7 Radiation Center Scheduled Visitors, July 1, 1979 - June 30, 1980 ..,. ,., IV- 17

-vii-TABLE TITLE PAGE

, Table V-1 Monthly Summary of Liquid Waste Discharges for the Year July 1, 1979 through June 30, 1980 . . . . . . . . . . . . . . . . . . . . . V 36 Table V-2 Monthly Summary of Gaseous Waste Discharges for the Year July 1,1979 through June 30, 1980 . . . . . . . . . . . . . . . . . . . . . V-37 Table V-3 Annual Summary of Solid Waste Discharges for -

the Year July 1, 1979 through June 30, 1980 . V-38 Table V-4 Annual Summary of Radiation Exposure Received by Facility Personnel and Visitors for the Year July 1,1979 through June 30, 1980 ... V-39 Table V-5 Annual Sunnary of Radiation Levels and Con-tamination Levels Observed During Routine Radiation Surveys for the Year July 1,1979 through June 30, 1980 ............ V-40 Table V-6 Total Dose Equivalent Recorded on Operating-Area Film Badge Monitors Located Inside the

- TRIGA Reactor Facility for the Year July 1, 1979 through June 30, 1980 . . . . . . . . . . V-43

' Table V-7 Total Dose Equivalent at the TRIGA Reactor Area Fence for the Year July 1,1979 through June 30, 1980 ............ V-44 Table V-8 Annual Average Concentrations of Gross Beta Radioactivity for Offsite Environmental Soil, ,

Water, and Vegetation Samples for the ' Year  !

l July 1,1979 through June 30, 1980 . . . . . . V-46 Table V-9 Annual Totals for Offsite Airborne Gamma Monitoring Stations for the Year July 1, 4

1979 through June 30, 1980 . . . . . . . . . . V-49 6 -

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< -LIST-0F FIGURES FIGURE TITLE PAGE Fig. III-l OSTR Annual Energy Production Vs. Time

-(Fiscal Year) . . ... . . . . . . . . . . . . III-3 1

- Fig ~. III-2 Nuclear Instrument Schematic Comparison . . . III-16 Fig. V 0perating-Area Film Badge' Monitw Locations

. for the TRIGA Reactor . . . . . . . . . . . . V 41

. Fig.~V-2 Area Radiation Monitor Locations for the TRIGA and AGN Reactors, and the TRIGA Reactor Area Fence ............. V-42 Fig. V-3 Monitoring Stations-for the OSU TRI^A Reactor, January 6,1976 through June .30, 1980 . -. . ................. V-45 i

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I-l I. INTRODUCTION AND

SUMMARY

A. INTRODUCTION TO OREGON STATE TRIGA REACTOR ANNUAL REPORT

, 1. This year's annual report will use the format that was introduced in the 1976-77 report.

- 2. The reporting period will be for one year: 1 July 79 to 30 June 80.

3. All of the information included in this annual report may not be of interest to all recipients and will require selected perusal. A comprehensive Table of Contents has been included to aid in such a selection.

4. This year's report will not attempt to review in detail the past operating years for the original 20% enriched core. A table showing the important operational data for this period (1967-1976) is included as Table III-2.

(The 1976-77 report is a good source of detailed infor-mation for readers interested in the OSTR's 20% enriched core history.)

5. This year':: report will review the operating history of the 70% enriched FLIP core (1976-present). The 70%

enriched FLIP core has been established as the historical base for subsequent reports.

B. SUPNARY OF OSTR USE DURING REPORTING PERIOD l

e During the year July 1, 1979 to June 30, 1980 the Oregon State TRIGA~ reactor: i

l. Generated 23.8 MWD of energy.

2. Consumed 29.8 grams of 23sU.

S I-2

3. Pulsed 313 times.

4. No fuel elements were added during the reporting period.

5. Accommodated 6 courses in nuclear engineering; 6 courses in chemistry; and provided demon-strations for classes in Civil Engineering, Urban League Workshop and General Science. (Reactoruse time for teaching and instruction totaled 136 houbs.)

6. One senior reactor operator trainee started his training and preparation for an NRC license exami-nation during the reporting period. He is a Radiation Center staff member. A total of 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> of reactor time has been used for this purpose. These

, hours are listed under regular and preclude operating

- time due to the nature of his training.

- 7. Two reactor operators are continuing their training and preparing for reactor operator licensing. These trainees are from Malaysia

• participating in a special research reactor operator training program. A total of 113 hours0.00131 days <br />0.0314 hours <br />1.868386e-4 weeks <br />4.29965e-5 months <br /> of reactor time has been used in their training during this reporting period. These hours are not listed with the teaching hours, but are under regular and preclude hours due to the nature of their training.

.. 8. Accanmodated 53 research projects. (Reactor use time for research programs totaled 650 hours0.00752 days <br />0.181 hours <br />0.00107 weeks <br />2.47325e-4 months <br />.) -

9. During a typical week, the reactor was in use about 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> / week. j

• Research technicians from Tun Ismail Atomic Research Centre.

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I-3 10 Accommodated 1,595 scheduled visitors and several hundred unschedule'd visitors during university open house events.

f . (Reactor use' time for visitor demonstration totaled 27 hours3.125e-4 days <br />0.0075 hours <br />4.464286e-5 weeks <br />1.02735e-5 months <br />'. )

11. Reactor use. time averaged 76%, based on a 40-hour week

-(eight' hours a day, five days ~a week).

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I-4 C.

SUMMARY

OF OSTR ENVIRONMENTAL AND RADIATION PROTECTION DATA Year July 1, 1979

• Through June 30, 1980 l 1. Liquid Waste Data (See Table V-1):

a. Total estimated quantity l of radioactivity released (in curies)* 1.85 x 10~4

b. Detectable radionuclides 58Co, 60Co, SlCr, in liquid waste 75Se, 463f,54Mn, 8

and 3H

c. Estimated average concentration of released radioactive material at the. point of release (in l microcuries per cubic centimeter) -9.97 x 10-6

u. Percent of applicable MPC for released liquid radioactive material at the point of

. release (%) 0.37

e. Total volume of liquid effluent i released, including diluent,.

!- (in gallons)** 4898 i *The OSU operational policy is to subtract only detector background

! from our water analysis data and not background radioactivity in the Corvallis city water.

l l ** Total volume of effluent plus diluent does not take into

! consideration the additional mixing with approximately 95,000 to 115,000_ gallons per year of liquids and' sewage normally discharged by the Radiation Center complex into the same sanitary sewer system.

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I-5 Year July 1, 1979 Through June 30, 1980 l-( 2. Gaseous Waste Data (See Table V-2):

! .a. Total estimated quantity of radioactivity released (in curies)* 23.81 4I Ar

b. Detectable radionuclides in gaseous waste ** (t g=1.83hr)

c. Estimated average atmospheric diluted concentration of argon-41 at the point of release (in microcuries per -7 cubic centimeter) 1.26 x 10 4

+

d. Percent of applicable MPC for diluted concentration of argon-41 at the point of release

(%) 3.15 i

e. Total estimated release of radioactivity in particulate form with half-lives greater than 8 days (in curies)*** NONE 1

- 3. Solid Waste (See Table V-3):

! a. Total amount of solid waste packaged.and disposed of 4 (in cubic feet) 21 .00 1

I b. Detectable radionuclides in 60Co, 59Fe, 24Na, 54Mn, solid waste 65Zn, 51Cr, 58Co i c. Total radioactivity in solid i waste (in curies) 5.96 x 10~4

• The increase in total argon-41 released during the current reporting period parallels very closely the increase in

. reactor megawatt-days (MWD) operating time for~the same period.-

• • Routine gamma' spectroscopy evaluation of the gaseous radioactivity in the stack discharge indicated that it was virtually all. argon-41. .

• Evaluation of the particulate radioactivity in the stack discharge confirmed its origin as naturally occurring radon daughter' products, predominantly lead-214 and bismuth-214, not associated with reactor operations.

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I-6 Year July 1, 1979 Through June 30, 1980 s 4. Radiation Exposure Received by Facility Personnel and Visitors (in mrem)(see Table V-4):

a. Facility operating personnel (mrem)

(1) Average whole body 34.00 (2) Average extremities 490.00

3) Maximum whole body 115.00

4) Maximum extremities- 2110.00

b. Facility research personnel

.(mrem)

(1) Average whole body 5.00

-(2) Average extremities 67.00 (3) Maximum whole body 20.00 (4) Maximum extremities 180.00

c. Visitors (mrem)

~

(1) Average whole body - 1.00 (2) Maximum whole body , 38.00

. 5. Number of Area and Off-Site Environmental Monitoring Samples Evaluated:

a. Area film badges inside the TRIGA facility 96

b. Vendor supplied TLD monitors on the reactor facility fence 36

c. OSU TLD monitors on the reactor facility fence 108

d. Integrating ionization chambers on the reactor facility fence 468

e. pR/hr measurements around the peri-meter of the reactor facility fence 207

. f. Off-Site environmental soil samples 16

g. Off-Site environmental water samples 14

h. Off-Site environmental vegetation -

samples 56

J I-7 1

Year July 1, 1979 Through June 30,-1980

• 1. Off-Site. vendor supplied TLD monitors 44

! a j. Off-Site OSU TLD monitors 228

k. Off-Site integrating-ionization chambers 572

1. pR/hr measurements at the off-site airborne gamma

. monitoring stations 437 4

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II-l II. GENEP'L INFORMATION A. L IATION CEtlTER.

O The Oregon State TRIGA Reactor (OSTR) is housed in the

., Radiation Center at Oregon State University. The Radiation Center was designed and established to: (1) accommodate in-ternal and off-campus instructional programs; (2) support research and development programs involving nuclear science and engineering; (3) provide a place for the use of radio-isotopes and ionizing radiation; and (4) provide fast and thermal neutrons for applicable programs. Construction of a

the Radiation Center was divided into two phases. The first phase was completed in June 1964 and consisted of 32,397 square

, feet of office and laboratory space. The second phase was com-pleted in March 1967, and consisted of a nuclear research reactor housed in a 9,956 square foot building adjacent to the existing Radiation Center. In 1975, temporary space of 1,600 square feet was added for interim accommodation of the fast expanding nuclear engineering program. In 1977, addi-tional temporary space of 1,600 square feet was added. The Radiation Center complex at present totals 45,553 square feet.

Housed in the Center are various types of laboratories and equipment designed to furnish:

1. Instruction programs in nuclear engineering, radiation biology, and nuclear and radiation chemistry.

2. Instrumental and radiochemical neutron activation analysis.

II-2

3. Neutron radiography and neutron diffraction.

4. Irradiation experiments involving x-ray, gamma-ray, or neutrons.

5. Measurement of various types of ionizing radiation.

Consultation in the application of radioisotopes and 6.

radiation research.

7. Exploratory programs on the novel uses of radioisotopes and radiation.

B. FACULTY MEMBERS HOUSED AT THE RADIATION CENTER

<

• Wang, Chih H. (Professor)-

Director, OSU Radiation Center Reactor Administrator Head, Department of Nuclear Engir.eering Badiei, Sue (Research Assistant Unclassified)

. Chemistry (nuclear chemistry)

• Bennett, Casey W. (Instructor)

Chemistry (nuclear chemistry) 9

• Binney, Stephen E. (Associate Professor)

Nuclear Engineering (nuclear instrumentation)

Daniels, Malcolm (Professor)

Chemistry (radiation chemistry)

• Dodd, Brian (Assistant Professor)

Nuclear Engineering (health physics)

Health. Physicist, OSU P. diation Center Fairchild, Clifford E. (Professor)

Physics (radiation chemistry)

• Hornyik. varl-(Associate Professor)

Nuclear Engineering (safety analysis and reactor kinetics)

• Johnson, Arthur G. (Associate Professor)

Nuclear Engineering (health physics)

Senior Health Physicist, OSU Radiation Center Kimeldorf, Donald J. - (Professor)

General Science (radiation biology)

-* Reactor users for research and/or teaching.

II-3

• Loveland, Walter D. (Associate Professor)

Chemistry (nuclear chemistry)

• Peddicord, K. Lee (Assistant Professor)

Nuclear Engineering (thermohydraulics)

Podowski, Michael (Visiting Assistant Professor)

, Nuclear Engineering (reactor kinetics)

- Popovich, Milosh (Vice President Emeritus)

• Ringle, John C. (Associate Professor)

Nuclear Engineering (shielding and safety analysis)

Assistant Reactor Administrator, OSU Radiation Center

• Robinson, Alan H. (Professor)

Nuclear Engineering (ncitron radiography and fuel management) r

• Schmitt, Roman A. (Professor)

Chemistry (neutron activation analysis, lunar geology)

Spinrad, Bernard I. (Professor)

Nuclear Engineering (reactor design and nuclear fuel cycles)

Stratton, Richard (Visiting Professor)

, Nuclear Engineering (nuclear fuels)

Thomas, T. Darrah (Professor)

Chemistry (photoelectron spectroscopy)

Woods, W. Kelley (Visiting Professor)

Nuclear Engineering (energy systems analysis)

C. RESEARCH PERSONNEL HOUSED AT THE RADIATION CENTER

1. Post-Doctorate Research Associates Name Field Advisor Gimzewski, James Chemistry T.D. Thomas

• Ma, Maw-Suen Chemistry R.A. Schmitt Ungier, Leon Chemistry T.D. Thomas

2. Graduate Students Name Degree Field Advisor Azafar, M. A. PhD Nuclear Engr B.I. Spinrad

. - Bomben, Ken PhD Chemistry T.D. Thomas Coomes, Edmund P. MS Nuclear Engr A.H. Robinson

• Dzata,. Francis K.A. MS Chemistry R.A. Schmitt

• Reactor users for research and/or thesis work.

II-4 Graduate Students (continued)

Name Degree Field Advisor Guidotti, Timothy E. MS Nuclear Engr K.L. Peddicord

• Ghannam, Lina M. MS Chemistry W.D. Loveland Ghannam, Musa M. MS Pharmacy J.W. Ayres Henke, Robert K. MS Nuclear Engr K.L. Peddicord

• Hughes, Scott S. PhD Geology R.A. Schmitt Johnson,.bn D. PhD Forest Sci B. Farrell Kazerouni, Mohd. MS Chemistry W.D. Loveland

• Keasler, Ken PhD Chemistry W.D. Loveland

• Kraus, Robert H. PhD Chemistry W.D. Loveland

• LaTouche, Y. David MS Biol Science D.J. Kimeldorf Lopez, Ricardo MS Nuclear Engr K. Hornyik Michaelis, Barbara MS General Sci D.J. Kimeldorf

. Montgomery, Scott MS Nuclear Engr K.L. Peddicord Nassersharif, Bahram MS Nuclear Engr K.L. Peddicord Nelson, Janet MS Nuclear Engr A.H. Robinson

• Nielson, Larry MS Nuclear Engr K.L. Peddicord Nolan, Gary PhD Chemistry T.D. Thomas l Nyone, William MS Chemistry W.D. Loveland l Oertel, Chris P. -

PhD Chemistry W.D. Loveland Oylear, Joan M. MS Nuclear Engr K.L. Peddicord )

• 0zaki, Calvin MS Rad Biology D.J. Kimeldorf !

• Poeton, Richard MS Gen Science A.G. Johnson ,

Polkinghorne, Steve MS Nuclear Engr S.E. Binney i Prichard, Andrew MS Nuclear Engr B.I. Spinrad Reardon, Patrick T. MS Nuclear Engr B.I. Spinrad

. Reid, Bruce MS Nuclear Engr K. Hornyik

• Rivera, Ma Rita MS Chemsitry R.A. Schmitt Robinson, Cheryl A. MS Nuclear Engr K.L. Peddicord

• Schofield, Paul MS Nuclear Engr A.H. Robinson Scott, James D. MS - Rad Biology D.J. Kimeldorf Sietner, Valerie J. MS Rad Biology D.J. Kimeldorf

• Smith, Monty PhD Nuclear Chem R.A. Schmitt Sterbentz, James MS Nuclear Engr K.L. Peddicord Ting, Yine-Ping PhD Nuclear Enge K.L. Peddicord

• Tollefson, Dennis A. MS Nuclear Engr A.H. Robinson

• Ungerer, C. Andy MS Chemistry W.D. Loveland Van, Phuong Dong MS Nuclear Engr A.H. Robinson Vincent III, Andrew MS Nuclear Engr K.L. Peddicord

• Wang, Lancelot S.K. MS Nuclear Engr A.H. Robinson Yen, Yu-Liang PhD Nuciear Engr B.I. SpinrM Youssefnia, Mohammad H. MS Nuclear Engr J.C. Ringle -

3. Visiting Scientists and Trainees Name Field (Affiliation) Advisor

• Abu, M.P.H. Reactor Operations (Malaysia)

J.C. Ringle Besar, Idris Health Physics (Malaysia) A.G. Johnson

• Diazengwe, Mpaka Reactor Operations (Zaire) J.C. Ringle

• Gao , Y . C. Neutron Radiography (China) A.H. Robinson

• Reactor users for research and/or teaching

II-5 Visiting Scientists and Trainees (continued)

Name Field'(Affiliation) Advisor Graslund, Astrid DNA photochemistry (Sweden) M. Daniels

• Liu , Y.G. Neutron activation anal (China) R.A. Schmitt Liu, D.Q. Nuclear fuel (China) B.I. Spinrad

. Luo, C. Nuclear chemistry (China) W.D. Loveland

• Nunnelley, Lewis _ Nuclear cl+mistry (Chemeketa Col) T.D. Thomas Pilus, A. R. Health physics (Malaysia) A.G. Johnson Xu, B. Radiotracer methodology (China) C.H. Wang

• Yunus, Yaziz Reactor operations (Malaysia) J.C. Ringle D. CLASSIFIED STAFF AT THE RADIATION CENTER Name Title Anderson, Terrance V. Reactor Supervisor Bauman, Mary L. Clerical Specialist Bennett, Stephen L. Radiation Specialist Busby, Harold Scientific Instrument Technician Campbell, Ken Custodian Carpenter, William T. Reactor Operator Clark, Judith A. Business Manager Doak, Sandra Clerical Assistant c Flickinger, Evelyn Secretary Keen, Robin A. Administrative Assistant Lovett, Jody Clerical Assistant Moeller, Wanda Clerical Specialist Schenider, Mary K. Clerical Assistant Smith, Vernon N. Chemist Woodrow, Doyle Scientific Instrument Technician E. REACTOR OPERATIONS STAFF Title Name Reactor Administrator C. H. Wang Asst. Reactor Administrator J. C. Ringle Reactor Supervisor T. V. Anderson Senior Reactor Operators J. C. Ringle S. E. Binney T. V. Anderson Reactor Operators W. T. Carpenter Senior Health Physicist A. G. Johnson Health Physicist B. Dodd Radiation Specialist S. L. Bennett

• React.or users for research and/or teaching

II-6 F. REACTOR OPERATIONS COMMITTEE Name Affiliation

.J.1:. Ringle (chairman) Nuclear Engineering T.ll. Anderson Radiation Center Se~E. Binney Nuclear Engineering

-. A. G. Johnson. Radiation Center G. M. Reistad Mechanical Engineering A. H. Robinson Nuclear Engineering R. A. Schmitt Nuclear Chemistry-D. L. Willis. General Science (radiation biology)

G. RADIATION SAFETY COMMITTEE Name Affiliation D. J. Reed (chairman) Biochemistry-Biophysics S. E. Binney Nuclear Engineering A. G.~ Johnson Radiation Center.

i J. P. Kelley Radiation Safety Office J. E. Nixon Food Science and Technology C. C. Calligan Computer Center S. C. Fang . _ Agricultural: Chemistry R. C. Worrest General Science f

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k III-l Ill. OPERATIONAL DATA A. REVIEW t 'l . The OSTR has operated for more than 13 years.

i . 2. From March 1967 to August 1969, the maximum reactor power level was restricted to 250 kW.

3. In August 1969 the. reactor was licensed to operate at a maximum reactor power level of 1 MW. From then until June 1971 the OSTR could operate at 1 MW for only short periods of time, due to the lack of sufficient cooling capacity.

In June 1971 the cooling capacity was upgraded to allow I

.4.

continuous operation at 1 MW.

. 5. In July 1976 the reactor was shut down (for a month) and a new FLIP core (70% enriched fuel) installed.

6. See Table III-l for a tabular -review of the OSTR's four year statistics with the FLIP core.

7. See Figure -III-l for a graphical review of the OSTR's

• four year. energy production with the FLIP core.

8. See Table III-2 for a summary of the OSTR nine year statistics with a standard (20% enriched) core.

9. This year's Annual Report will not attempt to review the

. past 13 years, but will only report and review the FLIP core. We have established the 70% enriched

• fuel .as the historical base for subsequent reports. More detailed information concerning the 20% enriched standard core can be obtained from the-1976-77 Annual Report dated

~

F l 31 August 77.

1

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

.- III-2 Table III-l FOUR YEAR OSTR STATISTICS 1 Aug 76 1 Jul 77 1 Jul 78 1 Jul 79 to to to to FLIP Core 30 Jun 77* 30 Jun 78 30 Jun 79 30 Jun 80 l i

Operating Hours j 458 875 (critical) 875 81 9 i i

- Megawatt Hours 451 496 255 571 l tiegawatt Days 19 20.6 10.6 23.8 Grams 2350 Used 24 25.9 13.4 29.8 Hours at Full 4 01 4 81 218 552 Power (1 MW)

Number of Fuel Elements Added 85 0 2 0 l to Core (initial loading) t i

Number of Irrad-iation Requests 443 375 329 372

^

• Reactor shutdown July 26, 1976 for one month for refueling reactor with new full FLIP core.

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III-3

, Figure III-l OSTR ANNUAL ENERGY PRODUCTION VS. TIME (FISCAL YEAR) 35 -

1976-77: 19 MWD 1977-78: 20.6 MWD 1978-79: 10.6 MWD-1979-80: 23.8 MWD 30 -

25 -

~

5 e A

E E 20 - O ti e B

i g

a-

. g 15 -

5 5

10 -

5 -

0 ' ' ' '

~

76-77 77-78 78-79 79-80 FISCAL YEAR

- - , - . - -y- - - , . <, ,, -- . -.-y

Table III-2 OSTR STATISTICS WITH 201 ENRICEED CORE 8 MR 67 .1 Jtt 68 1 JLA. 69 1 APR 70 1 APR 71 1 APR 72 1 APR 73 1 APR 74 1 APR 75 1 APR 76 to to TOTAL to to to to to to to to 30 JUN 68 30 JUN 69 31 M R 70 31 MR 71 31 M R 72 31 MR 73 31 MAR 74 31 M R 75 31 MR 76 26 JUL 76 MR 67 to JE 76 Operating Hours (critical) 904 610 567 855 598 954 705 563 794 353 6903 I

Megawatt Hours 117.24 102.47 138.05 223.77 195.11 497.82 335.94 321.45 408 213 2553 Megawatt Days 4.88 4.27 5.75 9.3 8.1 20.74 13.99 13.39 17 9 106.4 Grams assU Used 6.13 5.36 7.21. 11.7 10.2 26.031 17.57 - 16.81 21.35 10.7 133 -

Hours at Full Power (250 KW) 429 369 58 -- -- -- -- -- -- -- 856 Hours at full Power (1 Md) -- -- 20 23 100 401 200 291 460 205 1700 e

i Number of Fuel Elements Added 70 2 13 1 1 1 2 2 2 0 94 to Core (initial)

Number of Irrad-1ation Requests 429 433 391 528 347 550 452 396 3'J 217 4100 Number of Pulses 202 236 299 102 98 249 109 183 43 39 1560 l

• Reactor became critical on March 8, 1967 (70 element core; 250 igd). *** Reactor shui %wn June 1,1971 for one month for cooling system Note: This period length is 1.33 years as initial critical upgrading.

occurred out of phase with t e reporting period.

• • • • Reactor shutdown July 26, 1976 for one month for refueling

• • Reactor shutdown August 22, 1969 for one month for upgrading to reactor with new full FLIP core. Note: This period length

, 1 Md (did not upgrade cooling system). Note: This period length is .33. years.

is only .75 years as there was a change in the reporting period from July-June to April-March. C 5

III-5 B. OPERATING STATISTICS The utilization of the OSTR for the reporting period increased to some extent compared to that of the previous

.. year (see Table III-1).

The thermal energy generated in the reactor during the reporting period was 23.78 MWD. (The cumulative thermal energy generated by the FLIP core now totals 74.0 MWD for Aug.1,1976 to June 30, 1980). See Tables III-1 and III-3 through III-5 for this reporting period's statistics.

Reactor use time averaged 476% of our 8-hour day, 5-day week schedule.

Our present rate of excess reactivity decrease with

, the FLIP core is about .5c/ MWD. Our present core excess is approximately $6.67. (The initial FLIP core excess was $7.17.)

Lately the reactivity loss per MWD, with the FLIP core, has been much less than the 3c/ MUD with the 20% fuel. The fuel manufacturer (General Atomic) reports that the FLIP fuel shouli initially decrease in reactivity, and then eventually (at about 120 MWD)see a net gain in reactivity. This net gain should peak in about 4.5 MW L , years and would be a result of the burnable poison in the fuel.

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III-6 l Table III-3 PRESENT OSTR OPERATION STATISTICS FLIP 1 July 79 Cumulative to 1 Aug 76 Reactor Operations 30 Jun 80 to date

1. MWH of energy prodi o.j., 571 1773

2. MWD of energy r coduced 23.8 74.0

3. Grams 23st' used 29.8 93.1

4. Number of fuel elements added to core 0 84 + 3 FFCR**

5. Number of pulses 313 629

6. Hours reactor critical 875 3027,

7. Hours at full power (1 MW) 552 1652

8. Number of startup and shutdown checks 253 978

9. Number of irradiation requests processed

• 1 372 1 1519

10. Number of samples irradiated 3656 19,956

• Each request authorized from 1 to 120 samples to be irradiated (the number of samples per irradiation request averaged about 9)

• • Fuel Follower Control Rod O

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III-7 Table III-4A OSTR USE TIME FLIP 1 July 79 Cumulative Overall Reactor to 1 Aug 76 Operation Time 30 Jun 80 to date Statistics (hours) (hours)

1. Checkout, core excess and shutdown 363 1396

2. Load and unload samples 78 384

3. Reactor in operation

• 1135 3792

4. Total reactor use time 1576 5572

• Includes preclude time. (Preclude is the time the reactor is not available for use due to inspections and maintenance, such as fuel element inspections, transient rod lubrication, control rod calibration, power calibration, etc.)

. Table III-4B OSTR USE TIME FLIP Teaching, Research, 1 July 79 Cumulative Inspection and to 1 Aug 76 Demonstration Time 30 Jun 80 to date Statistics (hours) (hours)

1. Training (departmental) and others)h2 136 763

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_2. OSU researchb3 586 1960

3. D#f-campus researchb3 64 321

4. Reactor prec'ude time 763 2429

5. Visitor demonstration 27 99

6. Total reactor use time 1576 5572 1

2 Includes sample loading and unloading.

3 See Tables IV-1 and IV-2 for teaching statistics.

See Table IV-5 for research statistics.

III-8 i

Q Table III-5 8

OSTR MULTIPLEl USE TIME FLIP 1 Jul 79 Cumulative to 1 Aug 76 30 Jun 80 to date Number of Users (hours) (hours)

1. Two users 112 332

2. Three users 4 53

3. Four users 2 4 l 4. Total multiple 118 2 389 8

. use time l IMultiple use time is that time when more than one ex-i perimenter had samples in the reactor during critical operations.

214% of total hours the reactor was critical this year.

313% of total hours the reactor was critical since startup with FLIP fuel August 1976.

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III-9 C. EXPERIMENTS PERFORMED There are at the present time 12 approved experiments on the active list that can be utilized in reactor-related pro-

. grams. These experiments are listed below:

NOTE: Missing numbers identify those experiments that are in the inactive file and are not being used.

A-1 Reactor operation in any of its modes with no sample irradiation.

B-3 Irradiation of materials in assorted matrices for elements H to Bi inclusive plus natural Th and U for neutron activation analysis.

B-8 Isotope production for elements 1 thru 83 (H to Bi) excluding Cd.

B-ll Nuclear reaction studies by irradiating stable elements to produce any nuclide formed during the neutron irrad-

. iation of natural uranium.

B-12 Exploratory experiments to investigate the TRIGA capability to achieve certain experimental goals. If the TRIGA can achieve the desired goals, a regular ex-periment is drafted.

B-21 Advanced Neutron Radiography using beam port #3.

(Radiography of all conventional items plus ordinance materials.)

B-23 Measure y decay via y detector in thermal column for nuclear engineering labs.

B-24 General neutron radiography using beam port #1.

(Ordinance items excluded from radiography in this experiment.)

. B-25 Neutron flux monitors to be used to measure relative fluxes at various locations in the reactor core and other irradiation facilities. (Fission probes and self-powered neutron detectors.)

4 B-29 Reactivity measurements for fuel worth.

B-30 Irradiation of jet, diesel, and furnace fuels. Irrad-iation of various fuel oils for NAA required a new ex-periment to satisfy the needs of various environmental agencies.

III-10 B-31 TRIGA flux mapping using all irradiation facilities and foils for determining neutron fluxes.

There are 25 experiments in the inactive file that would

, require re-approval of the Reactor Operations Committee before using.

Of the 12 approved experiments, 9 were used during the reporting period.

See Table III-6 for a tabulation of the experiments performed during the reporting period. (This table shows the experiments used, how often each was used, and in which par-ticular area the use occurred.)

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III-ll Table III-6*

EXPERIMENT USAGE VS. PROJECT Research Lab Special Experiment Research Thesis Classes Forensic Projects Total A-1 --

11 33 --

35 79 B-3 87 40 40 1 4 172 B-8 18 12 21 -- 10 61 B-21 33 20 -- -- --

53 B-23 -- -- 1 -- -- 1

. B-24 4 -- 2 -- --

6 B-25 -- -- 1 -- --

1 0-29 -- -- 1 -- -- 1 B-ll' -- --

2 --- --

2 TOTAL 142 83 101 1 49 376

• Table displays the number of times a particular experiment was used for a particular application.

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III-12 D. UNPLANNED SHUTDOWNS There were 14 unplanned shutdowns that occurred during ,

the reporting period. See Table III-7 for tabulation.

Table III-7 s

UNPLANNED SCRAMS Type Scram Occurrences Cause Voltage loss 4 Commercial or inverter power failure.

Linear Scram 2 Linear power range switch turned wrong way during power increases.

Linear Scram 1 Servo system did not track properly. A step change in

servo output occurred momen-tarily.

External Scram 1 Stack monitor blew a fuse High Voltage Scram 3 New high voltage supply unit had the scram point set too high.

Safety Channel ~

Scram 2 Electronic noise from mode switch, when switching from square-wave to automatic, triggered the safety bistable.

Period Scram 1 Scrammed in square-wave.

Period is supposed to-be grounded out during square-wave. This condition corrected.

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III-13 E. _ CHANGES IN FACILITY

1. 10 CFR 50.59 Changes O

There were no 10 CFR 50.59 changes in our facility and no new experiments added to the approved list during the reporting period.

2. Other Changes

a. In December of 1979, the Physical Plant Department finished installing a new air compresr.or for the reactor building. The new compressor will supply air to the transient rod, instead of the old Center general purpose compressor. The long air line from the Center compressor was fabricated from galvanized iron and generated too much particulate rust. The new compressor, with its short run of aluminum pipe, should be relatively free of corrosion products in the air line.

b. It was decided in the summer of 1977 to replace and upgrade part of the electronics in the OSTR console.

Maintenance was becoming more frequent, and locating replacement parts was becoming more difficult. The upgrading replaced the majority of the nuclear system l with new, state-of-the-art electronics.

The new upgrading package replaced the left-hand console electronics drawer which specifically accom-plished the following:

III-14

1) The original multirange 1inear channel, which used an ion chamber, was replaced by a new 9.5-decade linear channal driven by a fission chamber. (No scram features are on the new linear channel .)

2) The original log channel with a period circuit, which used an ion chamber, was replaced by a new 10-decode wide-range log channel with a period circuit, driven by a fission chamber. (The same fission chamber drives the new linear and wide-range log channels.)

3) The original count-rate (startup) channel which used a fission chamber was removed. (It's function was taken over by the new wide-range log and linear channels.)

4) A new safety channel with scram capability, driven by an ion chamber, was added.

5) An additional fuel element temperature circuit was added.

The upgrading package was ordered in February 1978 and installation was completed in January 1980. One of the biggest time delays in the process was the NRC review

. of the Technical Specifications amendment that was re-quested for this change. NRC approval was given 18 December 1979 in the form of Amendment #4 to our facility license. The actual installation of the new package required five weeks, including functional testing.

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III-15 The linearity of the new instrument systems is excellent, and the wide-range capability of the new log and linear channels provides increased operational flexibility and accuracy, especially when a low power run immediately follows a high power run.

See Figure III-2 for a comparison of the "before" and "after" nuclear instrumentation.

c. In late June of 1980, our old reactor top CAM was retired and a new CAM (same manufacturer) was installed to take its place. The new CAM was checked out and calibrated several weeks before it was used as our official monitor.

The new CAM has two channels, a gaseous and a par-ticulate, where the old CAM only had one, a particu-late channel. The new monitor should provide better

. information than the original.

d. On 18 December 1979 the NRC granted our request to increase our pulse reactivity insertion from

$2.35 to $2.55. This change was included in our License Amendment #4.

3. Planned Changes

a. We plan in the 'near future to replace our resistance

- bulb temperature sensor with thermocouples. We feel our temperature indications will then be more accurate with fewer periodic calibration adjustments.

b. We plan to add new radiation surveillance equipment to our facility in the fonn of a new area radiation ,

1 monitor. This equipment has been ordered, and it should be installed, ca .brated, and operational by the end of 1980.

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> FLUX REGULATOR , > FLUX REGULATOR LINEAR LINEAR > RECORDER CIC > RECORDER CHANNEL CHAtlNEL i & SCRAM LOGIC

"(100% OF EACH RANGE) FC -

>STARTUP INTERLOCK LOGIC LOG E II *

> RECORDER LOG CHANNEL > RECORDER CHANNEL & PULSE INTERLOCK

" (1 xW-LOGIC)

IOD S PERIOD & SCRAM LOGIC CIRCUIT p (CRAM a 3 SEC.)LOGIC CIRCUIT " (a 3 SEC.)

LINEAR PERCENT LINEAR UIC y(SCRAM LOGIC POWER 110% OF FULL POWER)

PERCENT SCRAM LOGIC UIC CHANNEL POWER (110% OF FULL POWER)

CHANNEL LINEAR SAFETY UIC ' y(SCRAM LOGIC STARTUP y STARTUP INTERLOCK POWER 110% OF FULL POWER)

FC CHANNEL LOGIC CHANNEL E

BEFORE AFTER k Figure III-2 Nuclear Instrument Schematic Comparison

III-17 F. ftAINTENANCE AND SURVEILLANCE

1. Maintenance

-a. Aug '79: Replaced the transient rod solenoid valve.

1 The old valve was not consistent timewise from full a

' close to full open. That parameter becomes sig-nificant when conducting high speed motion neutron radiography.

, b. Aug '79: Replaced several circuit boards in our inverter. The inverter burned out components in several circuit boards. (The inverter is used to preclude false security alarms during a commercial oowerdip.) A charging transformer was also replaced

in the inverter to replace a defective unit.

c. Dec '79: Have a need to modify our underwater vacuum cleaner. While circulating and filtering the bulk- '

shield tank, the filter discharge connector separated causing water to be discharged-onto the reactor floor.

The filter. discharge connector will be modified before it is used again.

d. Jan '80: Installed a new gear box on the reactor servo control. The gear box was a new design to-replace our existing gear train that had a design deficiency. The deficiency prevented square-wave operation.

e '. Apr '80: Replaced the shim rod LED position indi-cator. The original burned out.

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III-18

f. May '80: Replaced the reactor servo motor with our "off-the-shel f" spare. The motor did not respond to small input signals properly. It was sent back to the vendor for inspection and adjustment.

g. Jun '80: . Replaced the graphite seal in the demin-eralizer pump.

h. Jun '80: Replaced the motor-starter contactor for the reactor room ventilation fan. The original burned out.

2. Tests and Inspections The OSTR has a routine test and inspection surveillance program. These T&I lists are presented in Tables III-8 through III-11. Those items marked with an asterisk (*)

- are required by the Technical Specifications.

G. REPORTABLE OCCURRENCES There were two reportable occurrences during the re-porting period:

1. On May 20, 1980 we apparently inserted about $2.75 in a reactivity pulse, exceeding the limit of $2.55 in part 3.3 of our technical specifications. This was reported to the USNRC, Region V office by telephone on 21 May 80, followed by a written report on 3 June 80.

. 2. On June 26, 1980 an inadequacy occurred in one of our administrative procedures. This was_ discovered on July 2, 1980 and reported to the USNRC, Region V office by telephone on July 3,1980, followed by a written report on July ll,1980.

Table III-8 y y y,) g APPENDIX J T & I's FOR THE MONTH OF

. TEST OR INSPECTION TO BE PERFORMED 00E DATE i INITIR.

CO LETED

• 1. . Functional check of reactor water level alarms.  !

, 2.lMeasurereactorprimarywatersystempH. f ,

3.. Measure the pH of the bulk shiled tank water.

4.{CheckTRIGAtankwateractivity.

5. :iEmergency power systems battery liquid level .

!and terminal checks. j I

6. l Emergency evacuation alanu system battery

{ liquid level and terminal checks.  !

7.IInspect brushes on rabbit system blower motor. i i l l l 8.fFunctionalcheckofevacuationalarm. '

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9. ;' Blow down the transient rod air accumulator tank.

f 10.lCalculatetheaveragemonthlyconductivity.

I(Average conductivity must be less than 5

, jmicromhospercentimeter.) ,

i 11.1 Change the light bulb in the green light. j i '

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. 12. Change the light above side entrance to reactor I building.

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13. l Check filter tape speed on stack monitor j(l"/hr).

t 14.' Lubricate the TRIGA tube loading tool as needed. l l l 1

l15.!Checkcamoillevel. l i 16.fPropanetankliquidlevelcheck(% full) f I

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III-20 Table III-9 T & I's FOR THE QUARTER OF TEST OR INSPECTION TO BE PERFORMED DUE DATE COMPT.ETED l INITIAQ 1.fROCauditreactoroperations.

2. Inspect and oil (as needed) solenoid operated valves in rabbit system. '

3.!Timethesampleinsertionandretrievetime

' interval of the rabbit system.

4. Check lazy susan for unknown samples.

5J Functio.nal check of emergency lighting (16 units)

6. Westronics Racorder: clean slide wire contacts.

7.! Varian Recorders: clean, inspect and lubricate; I

replace reference cells.

8.! MAP 18

{ (a) Check HV source.

: (b) Check ratemeter test position. ,

[ l (c) oil drive motors (fast and slow). j i 9.! RM I-110 Area Monitors l I

(a) Check HIGH & I4W alarms. '

(b) Check 225 volt supply.

(c) Check uv (.7cce: use AX-30 tes: and  :

cciib.~~cte meter. ) g I

10.j Arm system alarm checks: I CHAN 1 2 l 3 1 4 1 5 6 7 AUD 1 1 I

{

LIGHT  ! l i l PANEL 1 I  ! l ANN  ! I i  ! l l 4 11. b S per emluates operator.s - % Cerntnents.

Table III-10 III-21 APPENDIX L SEMI-ANNUAL T & I's FOR TEST OR INSPECTION TO BE PERFORMED DUE DATE INITIAL C0 PLETED

• 1. Functional check of the following interlocks:

(a) Source interlock.

(b) Simultaneous withdrawal of 2 rods.

, (c) Pulse initiation above 1 kw.

(d) Pulse interlock on range switch position 1 MW.

(e) Transient rod cylinder air interlock.

, (f) Pulse mode rod movement interlock.

(g) Prevents pulsing above $2.35 reactivity insertion.

• 2. Test safety circuits below:

(a) Linear channel.

(b) Safety channel.

(c) Manual scram.

(d) Preset times on pulse (less than 15 seconds).

• 3. Check (1) rod drop time (time must be less than two seconds) and (2) rod withdrawal and insertion time.

TRANS SAFE SHIM REG Rod Drop Withdrawall Insertioni e

• 4. Pulse reactor and compare fuel temperature and peak power with previous pulses of the same reactivity insertion.

• 5. Functional check of reactor room ventilation shutdown system.

• 6. Calibrate FE temp. meter.

• 7. ROC MTG at least semiannual.

• 8. Clean and lubricate transient rod internal barrel and piston (check for excessive air leakage).

9. Lube ball nut drive and threaded cylinder on transient rod.

• 10. Lubricate lazy susan drive and indicator assembly bearings.

11. MAP-1B: Disassemble and clean orifice plate for flow indicator.

12. Console: Perform check list (Appendix I in GA manual #7615).

13. AM-2A Air Monitor: Inspect and clean recorder, lightly lubricate recorder bearings.

14. Westronics Recorder: Check cero and calibration.

l APPENDIX M ANNUAL T & I's FOR III-22 l

l TEST OR INSPECTION TO BE PERFORMED. 00E DATE I"I I CO LETED l '*l'.![BI~. ANNUAL].

Remove and inspect all control rods for signs of corrosion and wear.

Annual report (due 30 June + 60 days)

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• 3. Calibrate control rods, i ..

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• 4. Calibrate reactor power. I

• 5. Calibrate bulk H O 2 temp. meter.

• 6. Calibrate the constant air monitor. I 1

• 7. Stack Monitor *

(a) Calibrate particulate monitor. l (b)-Calibrate gas monitor.

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• 8. Calibrate the Area-Radiation Monitor.

{

t 9. Conduct evacuation drill.

t10. Calibrate the reactor water activity monitor. '

, til. Count rate meter discriminator check.

Draw new curve.

12. Inspect standard rod drive mechanisms.

13. Change oil if needed in the thermal column door t

drive assembly reduction gear casing.

l l 14. Change oil in cam blower and oil motor, l

! 15. Lubricate thermal column door drive assembly l .as needed.

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16. Check beam port loading tool hydraulic reservoir level and lubricate mechanisms as needed.

! 411 Rx eperatcv regualifi catien 4tB SNM inventory

_19 Intrusion : Alorm Response. Drif t 2Q Security Gucircl Trcsi ning .

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IV-1 IV. UTILIZATION DATA A. TEACHING PROGRAMS

1. The OSTR was used to accommodate six courses in nuclear

~

engireering. These courses were:

• NE 101 Nuclear Engineering Orientation NE 102 Nuclear Engineering Orientation NE 203 Nuclear Radiation Detection and Measurement NE 441 Nuclear Reactor Experiments NE 442 Nuclear Reactor Experiments NE 531 Nuclear Reactor Kinetics Six chemistry courses utilized the OSTR. They were:

CH 107 General Chemistry Laboratory CH 207 General Chemistry Laboratory CH 316 Radiochemistry CH 419 Radioactive Tracer Methods CH S15 Experimental Nuclear Chemistry CH 528 Activation Analysis See Tables IV-1A, IV-2, IV-3, and IV-4 for data showing the use of the OSTR to accommodate teaching and academic programs.

2. Two trainees from Malaysia arrived in September of 1978.

t They enrolled in several academic classes in addition to starting their training for a reactor operator license in June,1979. The Malaysian operator training has been

, integrated into the regular reactor schedule and therefore these hours of on-the-job training are not reflected in any of the teaching statistics. (See Table IV-18.) The Malaysian operators are Yaziz Bin Yunis and Mohd. Puad 1

Bin Haji Abu.

3. An Assistant Professor in' Nuclear Engineering has been 1

assigned to obtain a senior reactor operator license. l His training commenced just prior to the end of this fiscal year. His training has been integrated into the regular reactor schedule and the hours are shown in Table IV-18.

IV-2 Table IV-1 A OSTR TEACHING HOURS i

1 Jul 79 Cumulative to 1 Aug 76 30 Jun80 to date Description (hours) (hours)

Departmental 130 697 Chemistry 37 Nuclear Engineering 87 Nuclear Engineering Technology 6 Special Classes 6 73 Hidden Valley High Schooll 2 Urban League Workshop 1 Civil Engineering Special Lab. 3 Total Teaching Time2 ,3 134" 770 1

Special training class was conducted for several Hidden Valley High School students.

2 Includes sample loading and unloading.

3See Table IV-2 for class and student statistics "See Table III-4B.

Table IV-1B OSTR OPERATOR TRAINING HOURS Name Training Accomplishment Training Hours Yaziz Bin Yunis Reactor Operator 113 Mohd. Puad Bin Haji Abu Reactor Operator Professor B. Dodd Senior, Reactor Operator 15 i

IV-3 Table IV-2 STATISTICS OF STUDENTS IN 4 NUCLEAR ENGINEERING AND NUCLEAR SCIENCE COURSES Number of Students FAL WIN SPR Course Cr. Course Title - 1979 1980 1980 Nuclear Engineering Courace t NE 10l* 2 Nuclear Engineering Orientation 27 -- --

l NE 102* 2 Nuclear Engineering Orientation --

20 --

NE 103 3 Intro Nucl . Engr. & Comp. -- --

19 NE 201 3 Nuclear. Energy Fundamentals 18 -- --

NE 202 3 Nuclear Radiation & Matter --

17 --

NE 203** 3 Nuclear Radiation Detection & Meas. -- --

14 NE 405 1 Reading & Conference -- --

3 NE 405A 1 Reading & Conference (Therm. Hydraul.) 24 -- --

NE 405A- 1 Reading & Conference (Comp. Analy. Nuc. Eng.) -- -- 10 NE 406 Projects 3 4 10 NE 407 1 Seminar --

32 --

NE 411x 4 Intro Nuclear Reactor Engineering 25 --

NE 412x 4 Intro Nuclear Reactor Engineering --

27 --

NE 413x 4 Intro Nuclear Reactor Engineering -- --

23 NE 421 3 Nuclear Reactor Analysis & Comput. 29 -- --

NE 422 3 Nuclear Reactor Analysis & Comput. --

28 --

NE 423 3 Nuclear Reactor Analysis & Comput. --

--_ 25 NE 430x 3 Nuclear Fuel Cycle --

25 --

NE-431x 3 Reactor Thermal Hydraulics - 27 -- --

NE 432x _2 Reactor Design --

24 --

NE 433x 3 Reactor Design -- -- 24 l NE 435x 2 Nuclear Materials -- --

27 l NE 441** 3 Nuclear Reactor Experiments 32 -- --

NE 442** 3 Nuclear Reactor Experiments --

5 --

NE 461x 3 Radiation Protection Engineering --

34 --

NE.465x 3 Nuclear Rules & Regulations -- -- 32 i NE 503 1-15 Thesis 15 16 14

, NE 505 Reading & Conference l 1 --

NE 506P Projects 3 -- 1 NE 513 '3 Nuclear Reac. Vary. Thry. -- -- 4 NE 514 2 Reactor Neutron Spectra 9 -- --

NE 515 2 Reactor Neutron Spectra --

10 --

NE 521 3 Reactor Environmental Problems 8 -- --

NE 523x 2 Advanced Reactor Design --

3 --

• 0STR used occasionally for denonstration experiments.

• • 0STR heavily used.  !

i

IV-4 Number of Students FAL WIN SPR Course Cr. Course Title 1979 1980 1980 Nuclear Engineering Couroaa (Continued) n NE 524x 3 Advanced Reactor Design -- -- 3 NE 531 3 Nuclear Reactor Kinetics --

6 --

NE 532 2 Reactor Economics 12 -- --

NE 534 3 Power Reactor Dynamics -- --

6 NE 552 3 Computational Methods for Nuc. React. --

21 --

NE 553 3 Computational Methods for Nuc. React. -- --

8 NE 581 1 Sel Top /Adv Nuc Fuels 17 -- --

NE 583 2 Sel Top /Natn'l Energy Futures -- --

8 Nuclear Engincering Technology Couroco NT 406 Projects --

1 --

NT 407 1 Seminar --

5 --

NT 410 3 Field .'ractice 1 1 --

NT 413 3 Nuclear Plant Environmental Impact -- --

8 NT 431 3 Nuclear Power Plant Technology 7 -- --

NT 432 3 Nuclear Power Plant Technology --

8 --

Chemistry Coursec CH 105 4 General Chemistry Lecture --

752 98 2

$ General Chemistry Labs 91 fHf07 CH 316** 3 Nuclear Reactor Chemistry 22 -- --

CH 419** 4 Radioactive Tracer Methods 16 -- --

CH 515** 3 Experimental Nuclear Chemistry -- --

4 CH 528** 3 Activation Analysis --

8 --

Other Cources GS 405A Reading & Conference -- --

2 GS 460 3 Radiation Health -- --

14 GS 501 Research -- --

3

, *0STR used occasionally for demonstration experiments.

• • 0STR heavily used.

IV-5 Table IV-3 OTHER EDUCATIONAL INSTITUTIONS USING OSTR*

4 M

~

Number of Number of NumberofI

Faculty Students Visits to' Involved Involved OSTR-

'U in versity of Oregon 5** 9 ** 10 I

j 1

i t i .

{

j Rice University 1 0  ! 1 i ,

1 .

l l

Orban League Workshop.. 2 7 '

1

(Portland) I 1  ;  ;

}

- *Does not count community college, high school and grade school classes that come through for special tours. These are listed

^under the section on "Public Relations."

i

' ** Includes researchers and students from ot'ler universities working I- through the University of Oregon. (See Table IV-4.)

i 1

l l

e- - - - - - w -

w r--- -xs--w,, - -r .ew,.* ----m- ----,,-r,- x - v~ -w --

, , . . - 4 Table IV-4 GRADUATE STUDENTS DOING THESIS RESEARCH THAT USED THE OSTR Name. Degree Department Advisor Thesis Oregon State University D. Tollefson MS Huclear Engr. Robinson Neutron Radiography u. Liquid Propellants F. Dzata MS Chemistry Loveland Radiotracer Study to Determine Stability of Tracers  !

. in River Water.

L. Ghannam MS Chemistry Loveland Stable Activable Tracers for Toxic Substances T.' Murphy MS Chemistry Schmitt- INAA of Terrestrial Basalts M. Smith PhD Chemistry Schmitt Chemical & Petrological Characterization of Individual Rock Clasts in a Brecciated Meteorite L. Wang MS Nuclear Engr. Robinson Neutron Radiography of Two Phase Flow ,

K. Keasier PhD Chemistry Loveland Stable Activable Tracer for an Estuarine Environment B. Pickett MS General Sci. Johnson Independent Verifics- 'on of OSTR Stack Monitor Accuracy Y.D. LaTouche .PhD General Sci. M.C. Mix INAA of Mytilus Edulis .

R. Gill PhD Forest Res. Lab D. Lavender Study Hemlock Seedling Root Uptake From Soils Un iversity of Oregon S. Goldberg PhD Geology Goles Anorthosite Genesis Mr. MacCaskie PhD Geology Goles Study of Granites of the Bushveld Complex, South Africa Mr. Barnes MS Geology Goles Geology of Cascade Head E. Stimson MS Geology Kays Analysis of Meta-Volcanic Rocks from Baker County W. Avramenko MS Geology Goles Volcanism & Structure of the Echo Mt. Quad. 52 t

A. Rite PhD Yale University Through the University of Oregon G. Nixon PhD University of British Columbia Through the University of Oregon Mr. Roberts PhD University of Georgia Through the University of Oregon J. Bradley PhD Arizona State University Through the University of Oregon

IV-7 B. RESEARCH PROJECTS Fifty-three research projects utilized 650 hours0.00752 days <br />0.181 hours <br />0.00107 weeks <br />2.47325e-4 months <br /> of reactor

'- time. Thirty-six of these research projects were from Oregon

. State University,16 were from the University of Oregon, and one was from Battelle Northwest in Richland, Washington.

Two of the Oregon State University projects were con-ducted and correlated with another university and an institution.

These organizations were:

1. Rice University

2. Battelle Northwest Laboratory

• Several of the University of Oregon projects were also corre-

. lated with other universities. These institutions were:

1. Yale University

. 2. University of British Columbia (Canada)

3. Arizona State University

4. University of Georgia See Table IV-5 for statistics regarding research hours and Table IV-6 for a summary of the research projects.

• Under ERDA prime contract EY-76-6-06-1830.

IV-8 Table IV-5

a. OSTR RESEARCH HOURS 1 Jul 79 Cumulative

, to 1 Aug 76 Reactor Research 30 Jun 80 to date Hours Statistics (hours) (hours)

OSU Research

• 586 1960 4

! Off-Campus Research a 64 321

. Commercial 0 0 Total Research l'2 650 2281 2

Includes sample loading and unloading time.

2 20% of total hours was thesis and research combined.

NOTE: Research hours, OSU funded: 494.

Resear::h hours, other funded: 156 4

% +

- w w p - - -

,.e- _m- --y , ,- ,- - --w-- --

Table TV-6 SupplARY OF OREGON STATE UNIVERSITY TRIGA RESEARCH PROJECTS AND FUNDING ASENCIES 1 Rea ton Project Title Description Funding Agency

1. R.A. Schnitt Ag Chemistry. Toxicology of Brominated Oils INAA of Bromine in Marine Organisms NIIHS 1.J. Tinsley OSU

2. R.A. Schmitt Chemistry. OSU Terrestrial Basalts Analysis of Trace Elements T. N rohy Private

3. R.A. Schmitt Chemistry. OSU Lunar and Meteoritic Activation Chemical & Petrological Characteri- NASA M. Smith Analysis for Thesis (Ph.D.) zation of Rock Clasts in a Brecciated Peteorite

4. R.A. Schmitt Chemistry, OSU Chemical Studies of Lunar. INAA of Selected Samples NASA M.S. Ma Meteoritic and Terrestrial Samples

5. R.A. Schmitt Chemistry. OSU Trace Element Studies of Volcanic INAA for Selected Trace Elements in Rice University W.P. Leeman & Rice Univ. Rocks a Variety of Volcanic Rocks From French and Hawall Polynesia

6. R.A. Schmitt Chemistry. OSU Ore Investigation Ore Samples from Local Mines Checked

• OSU Radiation Center V.N. Smith With INAA for Au, Ag

7. R.A. Schmitt Cheatstry. OSU INAA of Selected Paint Chips Determine Presence of Cu in Selected V.N. Smith OSU Radiation Center Paint samples

8. R.A. Schmitt Chemistry. OSU Cd Ratios of Gold Activation of Very Dilute Gold OSU Radiation Center Solution to Determine Cd Ratios in L.S. & Rabbit

9. R.A. Schmitt Chemistry. OSU Lactose Analysis INAA of Lactose for Mg, Mn, Na OSU Radiation Center

10. R.A. Schmitt- OSU Western Wheat Quality Control INAA for Funge Which is Found in OSU Radiation Center E. Trione Wheat. To Determine any Difference in Elemental Content

11. R.A. Schmitt Geology Dept., High Cascade Volcanics Major & Trace Elements of Volcanic OSU Geology Department E.M. Taylor OSU Rocks From Three Sisters Area S. Hughes

12. R.A. Schmitt Chemistry. OSU Forensic Investigations INAA for Selected Trace Elements in Oregon Law Enforcement V.M. Smith a Variety of Forensic Samples Agencies

13. R.A. Schmitt Chemistry OSU SiO Analysis for Na Determine Na and K in 510: Matrices OSU Radiation Center V.N. Smith

14. D.D. Church Animal Science Nutritional Study of Single Determination of Sludge Value in R. Kellums Animal Science Dept.,

OSU Cell Frotein From Pulp Mills Animal Nutrition OSU

15. A.H. Robinson Nuclear Engr. Neutron Radiography Studies Development of High Speed Neutron D.A. Tollefson OSU of Liquid Propellants Radiography of Burning Propellants D00 $

. - - - ~ . _ _ - _ . . - _ - __ ._- .-. -_. , _ . . . . - c ~_ -

Table IV-6 (continued) 1 Reac i Project Title Description Funding Agency

16. A.H.-Robinson Nuclear Engr., Neutron Radiography of Two-Phase Flow L. Wang ~ Investigation of High Speed Motion Nuclear Engr.. OSU OSU Neutron Radiography of Two-Phase Flow

17. W. Loveland Chemistry. 050 River Trace Experiment F. Dzata Develop a Tracer Method for Tracing USDI Fluid Bound Substances in Fresh .

Water

18. W. Loveland Chemistry. OSU Sediment Transport Studies Activate Sand Samples to Determine Sea Grant Coatic Ratio & Stability

19. W. Loveland Chemistry. OSU herbicide Tracing Use of Tracers to Monitor Herbicide USDI Dispersal

20. W. Loveland Chemistry. OSU Stable Activable Tracers for Toxic Activate Organomebe11fc L. Ghannam , Sea Grant Substance Detection Dy Labeled Tracers

21. W. Loveland Chemistry. OSU Precipitation Scavenging of Tracers Analyze for Trace Elements in Rain Air Resources Center Released into Frontal Storms Water Using NAA C. W. Loveland~ Chemistry. OSU 'Hydrospheric Trace Elements in Activable Tracers Being Developed K. Keasier USDI Water Pollutant Tracing for to Trace Soluble Materials in an Thesis (Ph.D.) Estuarine Environment

23. W. Loveland Chemistry. OSU Indoor-Outdoor Ratios Determine Trace Elements in USDI Cigarette Smoke

24. .B. Dodd Nuclear Eagr., Argon Reduction Froject Study Methods of Reducing Ar OSU Radiation Center OSU Emissions From OSTR

25. S.E. Binney Nuclear Engr., Project Minnesota Uranium Assay in Rock Samples for OSU Radiatic,a Center A.G. Johnson OSU Oregon Department of Energy

26. Ranjit Gill Forest Research Short Term '8P Uptake Study Study Western Hemlock Seedling Forest Research Lab.,

Lab., OSU Root Uptake From Soils OSU

27. Y.D. LaTouche General Science. Activation of Mytilus Edulis Comparison of Elements in OSU OSU Radiation Center Mussels & the Creosote Pflings on Which they are Found a

28. A.G. Johnson General Science. Independent Verification of Study Gas Flows and Ar Concen-i B. Pickett OSU Radiation Center OSU Stack Monitor Accuracy trations in Ventilation System

29. A.G. Johnson OSU Radiation Stack Monitor Calibration Cer.ter

Ar Production for Stack Monitor OSU Radiation Center Standard

30. J.C. Ringle Nuclear Engr., Fission Product Buildup Study Investigate the Fission Product T.V. Anderson OSU OSU Radiation Center Buildup and Decay in TRIGA-FLIP Fuel Elements, and Compare with  ?

l Calculated Results 5 4

• n

= . . e Trbl+ IV-6 (continued)

Listing Name of Person (s) Department and Number Using Reactor Institution Project Title Description Funding Agency 31 K.S. Krane Physics. 05U Angular Correlation Measurements Detemine Nitigte Orders of 8 & y OSU Physics Department

'of **Rb Radiations in ' Rb

32. K.S. Krane Physics. OSU Angular Correlation Measurements - Verify Nuclear Spin-Parity Assign- OSU Physics Department of "Ru ment in the Beta Decay of 3 "RJ

33. K.S. Krane Physics. OSU Angular Correlation Measurements Study Nuclear Spectroscopy of OSU Physics Department of *"Ba Levels of 8"Cs by Decay of 5"Ba 34 K.S. Krane Physics. OSU Angular Correlation Measurements Improve Knowledge of the Electro- OSU Physics Department of 2"Os magnetic Transition Probabilities of the 8"Ir Levels

35. P. Van Nuclear Engr.. OSTR Thermodynamics Measure OSTR Flux and Temperature OSU Nuclear Engineering OSU Profiles

36. J. Corliss Oceanography. Galapagos Project INAA for Trace Metals of Trench NSF OSU Rock and Mn Crist. Sediments

& Hydrothemal Deposits *

37. J.C. Laul Battelle N.W. Lunar Chemical Characterization INAA for Chemical Study of Lunar ERDA prime contract Laboratory and Meteorite Samples EY-76-C-06-1830. sp.

agreement B-29210-KF

38. G.G. Goles Geology. Univ. Bushteld Granites: Petrogenesis Study of Granites of the Bushveld Geological Society of MacCaskie of Oregot Complex. South Africa America

39. G.G. Goles Geology, Univ. Rockwell Hanford Analysis Detemination of Compositions of DOE of Oregon Columbia River Pasalts. Purpose to see if Safe Reservoirs for Reservoirs for Disposal of Radio-active Waste can be Found. ,

40. G.G. Goles l'niversity of WACK-1 for Thesis (Ph.D.) Dynamic Geochemistry of Iztacctbuati University of Oregon G. Nixon British Columbia (Snowy-Broad) & University of British Through Univer- Colun6fa sity of Oregon

41. G.G. Goles Geology. Univ. NA Analysis for In-House Develop Standard for Use in I-Ray University of Oregon Peterson of Oregon Standards Fluorescence Laboratory

42. G.G. Goles University of Experimental Petrology of Test of Exp. Glasses for NA University of Oregon &

Prof. Weill Oregon Na-Glasses NSF M. Shaffer

43. G.G. Goles Geology, Univ. INAA of Granite Rocks for Thesis INAA of Ore Samples Related to NSF Through the Univ.

Prof. Stormer of Oregon Work Mr. Roberts Thesis of Gecrgia Mr. Roberts Through Univ.  ?

of Georgia j O

. . . . e o Tible IV-6 (continued)

Listing Name of Person (s) Department and Project Title Description Funding Agency Number Using Reactor Institution

44. G.G. Goles Geology, Univ. Anorthosites of the Adirondachs INAA of Selected Geological Samples University of Oregon S. Goldberg of Oregon Thesis (Ph.D.) From Adirondachs, New York

45. G.G. Goles, Geo13gy, Univ. Pet ology Studies of Paricutin Study of Mechanisms by Which Magnetic University of Oregon A.R. McBirney of Gregon Volcano Bodies Become Compositional 1y Zoned

46. G.G. Goles Geology, Univ. SE Oregon Volcanic Study NAA of Rock Saaples to Determine University of Oregon C. Hering of,0regon Petrologic History of the Volcanics

47. G.G. Goles Arizona State Not Known Geochemical Studies of Monazites NSF J. Bradley Univ. Through Univ of Oregon

48. G.G. Goles Geology, Univ. Effects of Metamorphism on Study of Metamorphased Rocks in University of Oregon J.M. Rice of Oregon Trace Element Contents the Presence of Fluids of 4

Reasonably Well-Known Composition

49. G.G. Goles Geology, Univ. Geology of Cascade Head Trace F.lement Analysis of Rocks U.S. Geological Survey '

Mr. Barnes of Oregon. Fror. an Focene Volcanic Center as .

an alt in a Petrologic Investigation

50. Dr. Kays Geology, Univ. M.S. Thesis Analysis to Complete Major Elements University of Oregon E. Stimson of Oregon in Meta-Volcanic Rocks From Baker County, Oregon

51. G.G. Goles Geology, Univ. Volcanism & Structure in the Major Element Analysis to Evaluate University of Oregon W. Avramenko of Oregon Central Portion of Echo Mt. Quad. the Geochemistry of Volcanic Rocks on Echo Mt.

52. G.G. Goles Vale Univ. Iron in Melts Data for Magnetic Susceptibility NSF A. Rite Through Univ. Measurements of Oregon i 53 G. Goles Arizona State Uranium in Rhyolites Determine Uranium Content of U.S. Department of M. Sheridan Univ. Through Rhyolites Energy 7 Univ, of Oregon 4

IV-13 C. PUBLICATIONS RESULTING FROM OSTR OPERAr!GNS THAT WERE REPORTED TO THE RADIATION CENTER

1. Publications in Print

, Ma M.-S. , R. A. Schmitt, R.L. Nielsen, G.J. Taylor, R.D. Warner and K. Keil .1979. Petrogenesis of Luna 16 aluminous mare basalts. J. Geophys. Res. Letts., 6_, 909-912.

Warner, R.D., G.J. Taylor, G.H. Conrad, H.R. Northrup, S. Barker, K. Keil, M.-S. Ma and R.A. Schmitt.1979. Apollo 17 high-Ti mare basalts: new bulk compositional dati, magma types, and petrogensis. Proc. Lunar and Planet. Sci. Conf.

,10, Geochim. Cosmochim. Acta. Suppl .10, p. 225-247.

Beaty, D.W., S.M.R. Hill, A.L. Albee, M.-S. Ma and R. A Schmitt.

1979. The petrology and chemistry of basaltic fragments from the Apollo 11 soil, Part I. Proc. Lunar and Planet. Sci,. Conf.

10th, Geochim. Cosmochim. Acta. Suppl .10, p. 41-75.

Wentwortn, S., G.J. Taylor, R.D. Warner, K. Keil, M.-S. Ma and R.A. Schmitt. 1979. The unique nature of Apollo 11 VLT mare basalts. Proc. Lunar and Planet. Sci. Conf.10, Geochim. Cosmochim. Acta. Suppl .10, pp. 207-223.

Taylor, G.J., R. Warner, K. Keil, J. Geiss, K. Marti, E. Roedder, R.A. Schmitt and P. Weiblen. 1979. The 67915 consort 4*c.. searching for pieces of the ancient lunar crust.

In r eference on the Lunar Highland Crust, Lunar and Plane-tafInstitute, Houston,TX,pp. 169-171.

Ma, M.-S. and R.A. Schmitt. 1980. Luna 24 VLT microgabbro and recrystallized basalt-new chemical data. In Lunar and Planetary Science XI, Lunar and Planetary Institute, Houston, TX, pp. 646-648.

Ma, M.-S. and R.A. Schmitt. 1980. Chemistry of lithic fragments from the Apollo 17 drill core sections 70003, 70005, and 70007-II-KREEP and ANT. In Lunar and Planetary Science XI, Lunar and Planetary Institute, Houston, TX, pp. 643-645.

'I Ma, M.-S. and R.A. Schmitt. 1980. Petrogenesis of Apollo 11 mare basalts: new chemical data of 30 basaltic fragments from cores 10004 and 10005. In Lunar and Planetary Science XI, Lunar and Planetary Institute, Houston, TX, pp. 649-651.

Ma, M.-S. and R.A. Schmitt. 1980. Chemistries of lithic

, fragments from the Apollo 17 drill core sections 70003, 70005, and 70007-III-anorthosites and chemically unique fragments. In Lunar and Planetary Science XI, Lunar and Planetary Institute, Houston, TX, pp. 640-642.

IV-14 Ma, M.-S., R.A. Schmitt, R.D. Warner, G.J. Taylor, S. Baker and K. Keil. 1980. Aluminous mare basalts and basaltic-textured KREEPy rocks from Apollo 14 coarse fines. In

. Lunar and Planetary Science XI, Lunar and Planetary Institute, Houston, TX, pp. 655-657.

• Taylor, G.J. , R.D. Warner, K. Keil, M.-S. Ma and R. A. Schmitt.

1980. Major-element compositional variations in KREEP. In

• Lunar and Planetary Science XI, Lunar and Planetary Institute, Houston, TX, pp. 1131-1133.

Gooding, J.L., K. Keil, T. Fukuoka and R.A. Schmitt. 1980.

The origin of chondrules as secondary objects: evidence from chemical-petrological heterogeneities. In Lunar and Planetary Science XI. Lunar and Planetary Institute, Houston, TX, pp. 345-347:

Taylor, G.J., R.D. Warner, K. Keil, M.-S. Ma and R.A. Schmitt.

1980. Silicate liquid imiscibility, evolved lunar rocks and the formation of KREEP. Proc. The Conference on the Lunar Highland Crust., Lunar and Planetary Institute, Houston, TX.

Gooding, J.L. , K. Keil, T. Fukuoka and R. A. Schmitt.1979.

The taetal components of chondrules. Meteoritics, 1_4_,

4 404-407.

Fodor R.V., K. Keil, M. Prinz, M.-S. Ma, A.V. Murali and

- R.A. Schmitt. 1980. Clast-laden melt-rock fragment in the Adams County, Colorado, H5 chondrite. Meteori tic, _1_5_,

5 41 -62.

W. Loveland, The use of stable activable tracers in environ-l

- mental science. 1980. In Short-lived Isotopes in Biology and Medicine, K.A. Krohn, J. Root, Ed. (ACS).

W. Loveland, K. Keasler, L. Ghannam and A. Borovik, Recent developments in the use of stable activable tracers in foren*,1c and marine science. Nuclear Methods in Energy and Environmental Research, J. A. Vogt, Ed. (USD0E,1980).

I Ian J. Tinsley, Robert R. Lowry, Bromine content of lipids of marine organisms. Journal of the American Oil Chemists, Society, Jan.1980, Vol . 57 (31).

2. Publications in Press a

. Warner, R.D., G.T. Taylor, K. Keil, M.-S. Ma and R.A. Schmitt.

1980. Aluminous mare basalts: new data from Apollo 14 coarse fines. Proc. Lunar and Planet. Sci. Conf. lith (in press).

Ma, M.-S., R.A. Schmitt, D.W. Beaty and A.L. Albee. 1980.

The petrology and chemistry of basaltic fragments from the Apollo 11 soil: drive tubes 10004 and 10005. Proc. Lunar and Planet. Sci. Conf. lith, (in press).

IV-15 Gooding, J.L. , K. Keil, T. Fukuoka and R. A. Schmitt.1980.

Elemental abundances in chondrules from unequilibrated chondrites: evidence for chondrule origin by melting of pre-existing materials. Earth and Planet. Sci. Letts.

(in press).

8 3. Reports and Papers s

Ozaki, Calvin and A.G. Johnson. The effects of elevated temperature during irradiation on the response of lithium fluoride TLD-600's and TLD-700's to thermal neutrons _and gamma radiation. (M.S. thesis, C. Ozaki, for submission to Health Physics Journal).

Pickett, B. and A.G. Johnson Independent verification of ti.- calibration of the argon-41 stack gas monitor at the OSU Tk.CA Mark II Reactor. (Research report (thesis equivalent) M.S., B. Pickett).

Preser.tations to the American Oil Chemists' Society September 26-29, 1976, Chicago, Illinois, " Metabolism of Brominated Fatty Acids," lan J. Tinsley, Barbara A. Jones, Robert R. Lowry, Dept. of Agricultural Chemistry, OSU, Corvallis, OR.

May 14-18,1978, St. Louis, Missouri, " Toxic Effects of Brominated Fatty Acids," Ian J. Tinsley, Barbara A. Jones, Robert R. Lowry, Dept. of Agricultural Chemistry, OSU, Corvallis, OR.

April 29-May 3, 1979, San Francisco, California, " Lipid-Bromine Concentrations in Tissues of Rats Fed Brominated Fatty Acids," Barbara A. Jones, Ian J. Tinsley, Robert R. Lowry, Dept. of Agricultural Chemistry, OSU, Corvallis, OR.

4. Papers Submitted to The Seventh TRIGA Owners Conference, March 2-5, 1980, San Diego, California J.C. Ringle, A.G. Johnson, T.V. Anderson. Self-Protection of FLIP Fuel: Calculational Techniques.

A.G. Johnson, B. Dodd, S. Bennett, J.C. Ringle, T.V. Anderson,

. W.T. Carpenter. Self-Protection of FLIP Fuel: Experimental Measurements.

T.V. Anderson, J.C. Ringle, A.G. Johnson, W.T. Carpenter.

Oregon State TRIGA Reactor Console Upgrading.

IV-16 D. COMMERCIAL OR NON-ACADEMIC UTILIZATION None E. PUBLIC RELATIONS t'

a The continued interest of the general public in the 0

TRIGA reactor is evidert. ir the number of people who have toured the facility. In addition to several hundred visitors during university open house events and interested individuals who happened to be in the vicinity, a total of 1,595 people were-given pre-planned and scheduled tours this fiscal year.

s See Table IV-7 for scheduled visitor statistics.

F. PLANNED CHANGES IN UTILIZATION One shift (40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> per week) operation is the current OSTR ntilization mode. At the present time there are no planned changes in utilization.

4 9

r 2

c - - s

IV-17 Table IV-7 RADIATION CENTER SCHEDULED VISITORS July 1, 1979 - June 30, 1980 N0. OF DATE NAME VISITORS 8

7-16-79 OSU English Language Institute 20

, 7-18-79 U of 0 Elementary School Teachers 40 8-02-79 Youth Conservation Corp 40 8-13-79 LBCC 8 9-04-79 OSU Army ROTC 8 10-05-79 CVHS General Science 40 10-09-79 OSU NE 101 10 10-10-79 LBCC ._

20 10-30-79 Gervais High School 50 10-06-79 OSU Chem 107 14 11-09-79 OSU English Language Institute 15 11-01-79 -0SU Chem 107/207 20 11-06-79 OSU Chem 107/207 20 11-07-79 OSU Chem 107/207 20 11-08-79 OSU Chem 107/207 20 11-15-79 OSU Chem 107/207 20 11-12-79 OSU English Language Institute 20 11-13-79 OSU Chem 206 20 11-15 OSU Chem 206 20

, 11-07-79 OSU Chem 107 13 11-08-79 OSU Chem 107 8 11-08-79 OSU Chem 107 15 11-15-79 OSU Chem 107 14 11-15-79 OSU Chem 107 9 12-05-79 Willamette High School 10 2-02-80 Beaver Open House )

100  ;

2-06-80 Carvallis Fire Department 6 2-06-80 OSU English Language Institute 16 2-16-80 OSU Alumni Association 190 2-21-80 OSU NE 102 18 1

2-22-80 LBCC 14 2-22-80 LBCC 10 2-25-80 OSU Equal Opportunities Program 30 2-28-80 OSU Energy Awareness Day 360 3-01-80 . Int'l 4-H Youth Exchange 5 4-03-80 LBCC- 7 4-25-80 Hidden Valley High School 30

, 4-30-80 OSU Civil _ Engineering 50 4-30-80 OSU Chemical Engineering 20 4-30-80 OSU Chemical-Engineering 25 5-01-80 OSU Civil Engineering 50

. 5-01-80 OSU Chemical Engineering 25 5-01-80 OSU Chemical Engineering 25 5-14-80 Department 'of Health 10 5-14-80 Bandon High School 10 5-16-80 0SU Chem 206 40 5-20-80 LBCC 24 5-28-80 LBCC 24 5-29-80 Chemeketa 12 1,595

V-1 TABLE OF CONTENTS Page

. V. ENVIRONMENTAL AND RADIATION PROTECTION DATAJ JULY 1, 1979 - JUNE 30, 1980 . . . . . . . . . . . - V-1 A. INTRODUCTION . . . . . . . . . . . . . . . . ....... V-1

, B. A

SUMMARY

OF THE NATURE AND AMOUNT OF RADI0 ACTIVE EFFLUENTS RELEASED OR DISCHARGED TO THE ENVIRONS BEYOND THE EFFECTIVE CONTROL 0F THE LICENSEE , AS MEASURED AT OR PRIOR TO THE POINT OF SUCH RELEASE OR DISCHARGE . . . . . . . . . . . . V-1

1. Liquid Waste (summarized on a monthly basis) . . . . . V-l (a) The radioactivity discharged during the reporting period based on the following: ...

V-l (1) The total estimated quantity of radio-activity released (in curies) ...... V-l (2) The detectable radionuclides present in this waste . . . . . . . . . . . . . . . . V-2 (3) An estimate of the specific activity for

, each detectable radionuclide present, if the specific activity of the released material after dilution was greater than 1 x 10-7 microcuries/ cubic-centimeter ..

V-2 (4) A sur,ary of the total release (in curies) for each radionuclide determined in (2) above for the reporting period, based on representative isotopic analysis . . . . . V-2

, (5) The estimated average concentration of the released' radioactive material at the noint of release for the reporting period (in terms of microcuries/ cubic centimeter) and the fraction of the applicable MPC value . . . . . . . . . . . V-2 (b) The total volume (in gallons) of effluent water (including diluent) released during each period of rel ease . . . . . . . . . . . . . . . V-2

V-ii Page

2. Gaseous Waste (summarized on a monthly basis) . . . . V-2 (a) The radioactivity discharged during the reporting period based on the following: ... V-2 (1) The total estimated quantity of radioactivity released (in curies) determined by an appropriate sampling and counting method ................. V-3

-(2) The detectable radionuclides present in this waste ............... V-3 (3) The total estimated quantity of argon-41 released (in curies) during the reporting period, based on data from an appropriate monitoring system . . . . . . V-3

. (4) The estimated average atmospheric diluted concentration of argon-41 released during the reporting period (in terms of microcuries/ cubic centimeter) and

, the fraction of the applicable MPC value . . . . . . . . . . . . . . . . . . V-3 (5) The total estimated quantity of radio-activity in particulate form with half-lives greater than eight days (in curies) released during the re-porting period,as determined by an appropriate particulate monitoring system. . . . . . . . . . . . . ..... V-3 (6) Theaverage concentration of radioactive 4 particulates with half-lives greater

^

than eight days (in microcuries/ cubic centimeter) released during the re-porting period ............. V-3 (7) An estimate of the average concentration of other significant radionuclides present in the gaseous waste discharge (in terms of microcuries/ cubic centi-meter) and the fraction of the applicable

- MPC value for the reporting period,if the estimated release was greater than i 20% of the applicable MPC . . . . . . . . V-4 I

V-iii Page,

3. Solid Waste (summarized on an annual basis). . . . y-4 (a) The radioactivity discharged during the reporting period based on the following: . y-4 (1) The total amount of solid waste packaged (in cubic feet) . . . . . . . V-4 (2) The detectable radionuclides present in this waste ............ V-4 (3) The total radioactivity in the solid waste (in curies). . . . . . . . . . . V-4 (b) The dates of shipment and disposition (if shi pped off-si te) . . . . . . . . . . . . . V-4 C. AN ANNUAL

SUMMARY

OF THE RADIATION EXPOSURE RECEIVED BY FACILITY PERSONNEL AND BY VISITORS , IN TERMS OF THE AVERAGE RADIATION EXPOSURE PER INDIVIDUAL AND THE GREATEST EXPOSURE PER INDIVIDUAL FOR EACH OF THE TWO GROUPS . . . . . . . . . . . . . . . . . . . . . . . . V-4 D. AN ANNUAL

SUMMARY

OF THE RADIATION LEVELS AND THE LEVELS OF CONTAMINATION OBSERVED DURING ROUTINE SURVEYS PER-FORMED AT THE FACILITY , IN TERMS OF THE AVERAGE AND THE '

HIGHEST LEVELS ................... V-5 E. THE LOCATION AND MAGNITUDE OF THE MAXIMUM MEASURED OR CALCOLATED DIRECT RADIATION LEVEL IN UNRESTRICTED AREAS DUE TO DIRECT RADIATION FROM THE' FACILITY, AND DIRECT RADIATION FROM FACILITY EFFLUENTS . .... V-5

1. The Maximum Direct Radiation Level in Unrestricted Areas Due to Direct Radiation From the Facility . . . . . . . . . . . . . . . . . . . . V-5

2. The Maximum Direct Radiation Level' in Unrestricted Areas Due to Direct Radiation From Facility Ef fluents ................... V-10 F. AN ANNUAL

SUMMARY

OF THE GENERAL METHODS AND THE RESULTS OF ENVIRONMENTAL SURVEYS PERFORMED OUTSIDE THE FACILITY ....................... V-17

~

1. The On-Site Environmental Monitoring Syster..s . . . V-17

2. The Off-Site Environmental Monitoring Systems .. V- 22

V-iv Page LIST OF TABLES AND FIGURES Table or Figure # Title Table V-1 Monthly Sumary of Liquid Waste Discharges for tFe Year July 1,1979 - through June 30, 1980 .................. V-36 Table V-2 Monthly Summary of Gaseous Waste Discharges for the Year July 1,1979 through June 30, 1980 . . . . . . . . . . . . . . . . . . . . . V-37 Table V-3 Anaual Sumary of Solid Waste Discharges for the Year July 1,1979 through June 30, 1980 . V-38 Table V-4 Annual Summary of Radiation Exposure Received  :

by Facility Personnel and Visitors for the Year July 1,1979 through June 30, 1980 ... V-39 Table V-5 Annual Summary of Radiation Levels and Con-tamination Levels Observed During Routine Radiation Surveys for the Year July 1,1979 through. June 30, 1980 ............ V-40 Fig. V-1 Operating-Area Film r,adge Monitor Locations for the TRIGA Reactcr ............ V-41 9

Fig. V-2 Area Radiation Monitor Locations for the TRIGA and AGN Reactors, and the TRIGA Reactor Area Fence . . . . . . . . . . . . . . V- 42 Tabl e V-6_ Total Dose Equivalent Recorded on Operating-Area Film Badge Monitors Located Inside the TRIGA Reactor Facility for the Year July 1, 1979 through June 30, 1980. ......... V-43 Table V-7 Total Dose Equivalent at the TRIGA Reactor

' Area Fence for the Year July 1,1979 '

through June 30, 1980 ............ V-44 Fig. V-3 Monitoring Stations for the OSU TRIGA Reactor, January 6,1976 through June 30, 1980 .... V- 45

. Table V-8 Annual Average Concentrations'of Gross Beta

. _ Radioactivity for Offsite Environmental Soil, Water, and Vegetation Samples for the Year July 1,1979 through June 30, 1980 . . . ._. . V- 46

- Table V-9 Annual Totals for Offsite Airborne Gamma Monitoring Stations for the Year July 1,1979 through June 30, 1980 ............

V- 49

V-1 V. ENVIRONMENTAL AND RADIATION PROTECTION DATA; JULY 1,1979

.THROUGH JUNE 30, 1980 A. INTRODUCTION The data contained in this section have been prepared to comply with the current requirements of Nuclear Regulatory Commission (NRC) Facility License No. R-106 (Docket No. 50-243) and the Technical Specifications contained in Appendix A to that license.

The material has also been prepared in compliance with Oregon Department of Energy Rule No. 345-30-010, which requires an annual report of environmental effects due to research reactor operations.

Within the scope of this program, all releases of radio-activity to thb unrestricted environment and all occupational exposures to radiation and radioactive materials are consis-tently maintained "as low as reasonably achievable."

B. A SU:11ARY OF THE NATURE AND Af10VNT OF RADI0 ACTIVE EFFLUENTS RELEASED OR DISCHARGED TO THE ENVIRONS BEYOND THE EFFECTIVE CONTROL OF THE LICENSEE, AS MEASURED AT OR PRIOR TO THE POINT OF SUCH RELEASE OR DISCHARGE

1. Liquid Waste (summarized on a monthly basis)

(a) The radioactivity in liquid waste discharged during the applicable reporting period has been summarized according to the following items. All liquid waste data pertaining to these items are containea in Table V-1.

(1) The total estimated quantity of radioactivity released (in curies).

V-2 (2) The detectable radionuclides present in this waste.

(3) An estimate of the specific activity for each detectable radionuclide present, if the specific activity of the released 6 material after dilution was greater than 1 x 10-7 microcuries/ cubic centimeter.

(4) A summary of the total release (in curies) for each radionuclide determined in (2) above for the reporting period, based on repre-sentative isotopic analysis.

(5) The estimated average concentration of the released radioactive material at the point of release for the reporting period (in terms of microcuries/ cubic centimeter) and the 0

fraction of the applicable MPC value.

(b) The total volume (in gallons) of effluent water (including diluent) released during each period when liquid waste was released is also summarized in Table V-1.

2. Gaseous Waste (summarized on a monthly basis)

(a') The radioactivity in gaseous waste discharged during the applicable reporting period has been summarized

. according to the following items. All gaseous waste data pertaining to these items are contained in Table V-2.

V-3

! (1) The total estimated quantity of radioactivity released (in curies) determined by an approp-riate sampling and counting method.

i j , (2) The detectable radionuclides present in this waste.

1 q (3) The total estimated quantity of a rgon-41 re-1 leased (in curies) during the reporting period ,

based on data from an appropriate monitoring system.

1 (4) The estimated average atmospheric diluted concentration of argon-41 released during the 1

reporting period (in _ terms of microcuries/

, cubic centimeter) and the fraction of the applicable MPC value.

(5) The total estimated quantity of radioactivity j in particulate form with half-lives greater than eight days (in curies) released during the reporting period,as determined by an appropriate particulate monitoring system.

(6) The average concentration of radioactive par-i i

ticulates with half-lives greater than eight days (in microcuries/ cubic centimeter) re-leased during the reporting period.

l

]

V-4 (7) An estimate of the average concentration of other significant radionuclides present in the gaseous waste discharge (in terms of

, microcuries/ cubic centimeter) and the fraction-of +he applicable HPC value for the report-ing period,1f the estimated release was greater than 20% of the applicable MPC.

3. Solid Waste (summarized on an annual basis)

(a) The radioactivity in solid waste discharged during the applicable reporting period has been summarized according to the following items. All solid waste data pertaining to these items are contained in

,_ Table V-3.

(1) The total amount of solid waste packaged (in cubic feet).

I (2) The detectable radionuclides present in this waste.

(3) The total radioactivity in the solid waste (in curies) .

(b) The dates of shipment and disposition of solid wastes (if shipped off-site) are also contained in Table V-3.

1 C. AN ANNUAL

SUMMARY

OF THE RADIATION EXPOSURE RECEIVED BY FACILITY PERSONNEL AND BY VISITORS , IN TERMS OF THE AVERAGE RADIATION EX-POSURE PER INDIVIDUAL AND THE GREATEST EXPOSURE PER INDIVIDUAL

. FOR EACH OF THE TWO GROUPS The annual summary of the radiation exposure received by facility personnel and visitors for the applicable reporting period is co'ntained in Table V-4.

V-5 D. AN ANNUAL

SUMMARY

OF THE RADIATION LEVELS AND THE LEVELS OF CONTAMINATION OBSERVED DURING ROUTINE SURVEYS PERFORMED AT DJE FACILITY , IN TERMS OF THE AVERAGE AND THE HIGHEST LEVELS The annual summary of radiation and contamination levels

. observed during routine facility surveys for the applicable reporting period is presented in Table V-5.

E. THE LOCATION AND MAGNITUDE OF THE MAXIMUM MEASURED OR CALCU-LATED CIRECT RADIATION LEVEL IN UNRESTRICTED AREAS DUE TO DIRECT RADIATION FROM THE FACILITY, /ND DIRECT RADIATION FROM FACILITY EFFLUENTS

1. The Maximum Direct Radiation Level in Unrestricted Areas Due to Direct Radiation From the Facility The location and magnitude of the maximum (measured and calculated) direct radiation level in an unrestricted area due to direct radiation from the facility can best be under-

, stood by referencing Figures V-1 and V-2, and Tables V-6 and V-7.

Early in the operating history of the OSU TRIGA reactor, two potential sources of direct radiation from the TRIGA facility were identified. These were the demineralizer tank for the reactor primary water system, and the graphite-natural uranium subcritical pile located in the main reactor room (see Figure V-1).

On January 3, 1972, the demineralizer tank was removed from its original position, shown in Figure V-1, to locat. ion "A" in Figure V-1, and henceforth ceased to be a major

, contributor to the direct radiation from the facility. On February 23, 1972, the east' side (the exterior wall side)

V-6 of the subcritical pile'and the entire demineralizer tank were conservatively shielded with concrete and lead, further limiting any small direct radiation contribution from the demineralizer tank, and effectively reducing the direct radiation to unrestricted areas from both the subcritical pile

and the demineralizer tank to essentially zero millirem per year.

1 With the elimination of the preceding two sources of direct radiation from the facility, two additional sources of lesser magnitude became apparent. One of these was the particulate filter for the reactor primary water system, which is located on the demineralizer platform (see Figure V-1), while the second is.best collectively termed " normal use of reactor experimental facilities and operating areas for research and teaching."

The particulate filter was completely shielded by July 10,1972, and the new shield eliminated any further radiation contribution in unrestricted areas from this source. The second source, relating to normal use of the OSU research reactor, takes into consideration the routine handling of radioactive materials within the entire facility, and the need for relatively freouent access into reactor experimental and irradiation facilities. Both of these latter activities create a small potential for very low level direct radiation exposure (of reactor facility origin) in immediately adjacent unrestricted areas.

V-7 Direct radiation levels in unrestricted areas (which po-tentially may arise from the TRIGA facility) are evaluated on the basis of three different types of radiation measurements.

First, direct radiation levels are measured and analyzed as part of our routine radiation monitoring program. These measure-ments include data from continuously operating area radiation monitoring stations located throughout the TRIGA facility operating-area, plus results obtained by numerous on-the-spot direct radiation measurements made by members of the radiation protection staff both routinely and during special TRIGA facility operations. Second, data from area monitoring film badges installed at strategic locations within ,the TRIGA reactor operating-aret are routinely documented and utilized to indicate locations where direct radiation from the facility might be entering unrestricted areas. The film badge data

. are corrected, as appropriate, to reflect radiation attenuation in the reactor facility walls. Finally, assessment of direct radiation levels in unrestricted areas is conducted on the basis of area monitoring data collected through our thermo-luminescent dosimetry (TLD) program. Most of the TLD's in this program are actually located in unrestricted areas or on the TRIGA reactor area fence surrounding the accessible sides of the TRIGA reactor building (see Figure V-2). There-fore, these monitors are an excellent source of information.

The specific locations of pertinent vendor supplied (the vendor being Radiation Detection Company [R.D. Co.], Mt.

_. ~

V-8 View, California) beta-gamma-neutron area monitoring film badges inside the restricted operating-area at the OSU TRIGA facility are shown in detail in Figure V-1, and are again shown as part of an overall area diagram in Figure V-2.

Figure V-2 also shows the locations of R.D. Co., CaSO 4TLD area monitors (started during the 1977-78 reporting period, replacing the beta-gamma-neutron film packs used previously on the reactor area fence) plus OSU supplied and processed TLD area monitors (normally 3 Harshaw LiF TLD-700 chips per monitor). Both types of TLD area monitors are positioned on the fence surrounding the TRIGA reactor facility. This fence was originally installed in September 1972. Figure V-2 also shows the location of three R.D. Co. beta-gamma-neutron area monitoring film badges used in conjunction with the University's AGN reactor. The AGN reactor and its monitors are not part of this report.

With the addition of the fence around the reactor area, area monitoring film badge-data from inside the TRIGA facility (contained in Table V-6) no longer have a high degree of correlation to direct' radiation levels in surrounding un-restricted areas. Nevertheless, we believe the data from inside the facility reflect the general character of our operation and therefore plan to continue including it in all reports of this type.

In Figure V-1 and V-2, film badge locations within the TRIGA~ reactor facility.are abbreviated to indicate their

V-9 position .cn1 a north, south, east, or west wall of the main reactor bay, or their location in the reactor's adjacent heat exchanger room. For example, MRCTSE is interpreted as Monitor Radiation Center TRIGA,' South badge, E,ast wall of the main reactor bay building. Similarly, MRCTHXS is the badge for the adjacent Heat Exchanger room, South wall. Monitoring locations on the fence are simply designated MRCFE-1 through MRCFE-9, and imply Monitor Radiation Center Fence Environmental (TLD position number).

After the addition of the previously described shielding and the reactor area fence, direct radiation levels in unrestricted areas due to the TRIGA facility dropped to esientially background levels. Data presented in Table V-6 show the generally low annual doses recorded inside the reactor facility's operating-area (the dose for location MRCTNW is' explained in the following paragraph). Likewise, Table V-7 provides verification that the total annual radiation levels present in unrestricted areas adjacent to the reactor area fence were within the range typically expected for natural background in Oregon. This table (V-7) presents

- results from the area monitoring TLD's located on the reactor area fence,'and direct microroentgen per hour exposure rate measurements collected at each fence area monitoring station. See footnote (5) of Table V for a further-explanation of the pR/hr data and its application.

i V-9a TRtGA operating-area film badge monitor MRCTNW recorded an annual dose of 735 mrem for the reporting period. While this value is obviously higher than those normally reported, it was mainly due to the planned -interim storage of a new 3 curie cobalt-60 sealed source purchased for the OSU instrument cali-bration facility. The source was located in its D0T-approved shipping container and the storage area in the reactor room was additionally roped off and posted. Direct radiation levels in that portion of the TRIGA room operating-area monitored by the MRCTNW film were quite low, and ranged between 0.3 and 0.5 mrem /hr inside the reactor building wall. Radiation attenuation in the wall reduced these levels by at least a factor of 4 to 4.5 at corresponding points directly in con-tact with the outside of the wall, and specified radiation levels inside the room were down by a factor of 11 to 12 at points of maximum intensity 3 feet from the outside of the reactor building wall. The designated locations outside the reactor buiiding are, of course, well inside the reactor area fence. and appropriate monitoring stations on the fence did not record abnormally high results during the reporting period. We therefore have concluded that the dose recorded by monitor MRCTNW resulted in no measurable . increase in the -

direct radiation levels in unrestricted areas surrounding the.

TRIGA facility. During August'1980, the cobalt-60 source storage container was additionally shielded to reduce the external radiation levels even further.

V-10 As a final note on the fence monitoring stations, it should be reported that there is little or no occupancy of any specific point on the perimeter of the fence throughout the entire year, and no continuous occupancy of any specific area insid'e the reactor fence.

OSU is also continuing its efforts to achieve closer agreement between R.D. Co. and OSU TLD data. At the present time, R.D. Co. continues to use a somewhat higher annual background than OSU, and we still feel that they may be re-porting dose accumulated during periods when their TLD's are not in service at OSU (e.g., during transportation, etc.).

Present control and QA procedures used by OSU for its outside dosimetry vendor will continue to be carefully scrutinized, and will be modified as deemed necessary during the next year in order to improve the agreement between the two TLD moni-toring systems.

2. The Maximum Direct .~.adiation level in Unrestricted Areas Due to Direct Radiation From the Facility Effluents The location and magnitude of the maximum (measured and calculated) direct radiation level in unrestricted areas due to direct radiation from facility effluents will be re-viewed in light of both liquid and gaseous releases.

As reported in Table V-1, the total annual quantity of radioactivity released in liquid effluents has been quite

~

small. The microcurie quantity for the reporting period in even a few hundred cubic centimeters of solution would not

V-loa normally present a significant direct radiation potential, particularly when the radionuclide composition of the radio-l . activity is examined. In~ our particular operation, the  ;

majority of the liquid radioactive effluent is now normally associated 1

k 9

5 i.

A I

i.

t I

l 4

6 9

9 j

i e

1 4

. , - - - u

-v - , - _ , . -

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

V-ll with a single annual demineralizer resin change. However, during the 1979-80 reporting period liquid radioactive effluent was also released from the reactor bulk shielding tank adjacent to the reactor pool, but not part of the reactor pool water volume or the reactor primary water system. When released from the reactor facility, potentially radioactive liquid is mixed in a tank on a batch basis with up to 3000 gallons of waste water from the Radiation Center laboratories before-final discharge into the unrestricted area (the sanitary sewer system)'. .

The annual average concentration for total reactor facility radice:tivity in liquid effluents entering the unrestricted area equaled 9.97 x 10 -6 Ci/cc for the year July 1,1979 through June 30, 1980. With respect to this value and the total radioactivity released in the liquid effluent, recall that no city water background radioactivity has been subtracted.

Also, note that the main' contributor to the microcuries released is tritium (s 184.6 pCi of tritium out of a total of 185 pCi released). Even though nearly all of the liquid effluent volume from the reactor facility originated during the annual changing of.the demineral Ter resins, or due to the release from the reactor's bulk shielding tank , it appears that little of the tritium is of reactor origin. Our routine analysis of Corvallis city water indicates a normal tritium background concentration within a range of 2.84 x 10-5 pCi/cc to l 3.76 -x 10 -6 Ci/cc for the year July 1,1979 through June 30, 1980.

i I

I t

V-12 Our annual average concentration for tritium based on all liquids released to the unrestricted area from the reactor facility is within this background range at 9.95 x 10-6 pCi/cc for the year July 1,1979 through June 30, 1980. If the tritium is omitted from the total radioactivity released in the reactor's liquid effluent and a calculation performed using the remaining. radioactivity (s 0.4 pCi), some of which is also city water background, the annual average concentration for reactor facility radioactivity entering the unrestricted r ea becomes 2.15 x 10-8 pCi/cc for the year July 1, 1979 through June 30, 1980.

In view of the radionuclides present, and the relative abundance of each, it can be easily determined (as shown in Table V-1) that the annual average concentration of total reactor facility radioactivity in liquid effluents represents but a small fraction (0.37%) of the appropriate unrestricted area maximum permissible concentration. In addition, the average concentration DOES NOT take into consideration the additional mixing with approximately 95,000 to 115,000 gallons per year of liquids and sewage normally discharged by the Radiation Center complex into the sanitary sewer system.

For these reasons, we have concluded that the direct radiation to unrestricted areas due to radioactivity in reactor liquid effluents has been negligible.

On pages 4-53 through 4-58 of the Safety Analysis Report (SAR) for the OSU TRIGA Research Reactor, dated August 1968,

V-13 consideration is given to routine discharge and atmospheric dilution of gaseous effluents from the reactor facility.

This particular analysis in the 1968 SAR was conducted using the original TRIGA facility stack height of 55 feet above ground level. On page 4-57 of this report, it is specified that the activity discharge rate assumed for the purpose of calculation was 100 MPC, meaning 100 times the normal 4 x 10-8 uCi/cc argon-41 unrestricted area maximum permissible

-6 concentration, or a value of 4 x 10 uCi/cc. On page 4-58 i (Table 4.11 of the Safety Analysis Report) it is concluded that under the most unfavorable atmospheric conditions (with the 55 foot stack) a person standing for a full year at the point of maximum concentration would be exposed to less than 9% (8.53%) of the normal unrestricted area MPC for argon-41.

As a result, a person could stand at that point continuously i

for one year under the most unfavorable atmospheric conditions, while the reactor operated continuously at 1000 kW, (and continuously discharged an assumed worst case concentration of 4 x 10-6 Ci/cc of argon-41) and receive a whole body I gamma dose from argon-41 of less than 45 mrem (42.6 mrem) integrated over an entire year's occupancy.

Since the OSU TRIGA does not operate on a 24-hour per day basis, does not operate continuously at 1000 kW while it is: operating, and does not discharge argon-41 at 4 x 10-6 Ci/cc while operating at 1000 kil, the annual average argon-41 concentration, as measured by the facility I l

stack monitor, has always been much less than the assumed I

k

~

V-14 )

l calculational value of 4 x 10-6 pCi/cc. Consequently, the maximum annual dose to the unrestricted area due to direct radiation from gaseous effluents has also been significantly

- less than the nominal 45 mrem per year value projected in the 1968 Safety Analysis Report.

As indicated in OSU's May 16, 1973 report of 10 CFR 50.59 items to the former USAEC Division of Reactor Licen-sing, (a copy of which also went to the former Oregon Nuclear and Thermal Energy Council) on February 23,1972 the TRIGA facility stack height was increased from its original 55 feet above the ground to 65 feet, 10 inthes above ground l evel . As a result of the new stack height, new atmospheric dispersion calculations were necessary in order to evaluate the atmospheric dilution of gaseous effluents from the reactor facility. The results of the original 1968 calcu-1 lations and the first evaluation following the stack change were included in Table 2 of OSU's May 16, 1973 report to the USAEC, and indicated a slightly lower con-centration at the point of maximum concentration using the higher stack. Additional plume studies during 1973 and 1974, and again during 1978 using USNRL Regulatory Guide 1.111, evaluated the influence of the new stack height on gaseous effluent dispersion, and essentially confirmed earlier. data. Only a slight change is introduced if the most unfavorable values from the expanded 1973-74 and newer 1978 study are used. ,

V-15 Using the same basic assumptions employed for the shorter stack, and in particular a continuous argon-41 discharge rate of 100 MPC, the 1973-74 results indicate

~

that for atmospheric conditions giving the highest ground concentration (i.e., the worst atmospheric con-ditions) a person standing at the point of maximum con-centration would encounter approximately 3.018% (as

, opposed to 2.85% in the 1972 report) of the unrestricted area MPC for argon-41. Furthermore, the 1973 study pro-duced a nearly identical value of 3.005% of the unre-stricted area argon-41 MPC. As a result, a person could stand at this point of maximum concentration (currently projected to be 130 meters from the stack as opposed to 150 meters in the 1973-74 calculation, and 135 meters in the 1972 report) continuously for one year under the worst atmospheric conditions, while the reactor contin-uously discharged 100 times the argon-41 MPC, (4 x 10-6 Ci/cc) and receive a whole body gamma dose from argon-41 of 15 (15.03) mrem integrated over an entire year's occupancy.

As we have indicated, the OSU TRIGA does not operate on a 24-hour per day basis, nor does it operate continu-ously at 1000 kW. Also, the facility's annual average argon-41 concentration is always much lower than the 4 4 x'10-6 Ci/cc value used for purposes of calculation.

-e

V-16 As a result, the maximum annual dose in the unrestricted area due to direct radiation from gaseous effluents consistently remains much less than the nominal 15 mrem per year. projected using the new stack height and the 1978 plume dispersion data.

In order to evaluate the maximum dose in the unre-stricted area from gaseous effluents during the reporting period, one should assume continuous annual occupancy at the point of maximum concentration. Furthermore, it will

, be necessary to assume the existence of the most unfavorable meteorological conditions for a full year in order to achieve the maximum concentration at the specified point for one entire year. If these conservative assumptions are applied in conjunction with the reported annual average argon-41 concentration, (1.26 x 10-7 Ci/cc) as derived from actual measurements at the point of release with the facility's continuous stack monitor (see Table V-2),

then the maximum annual dose in the unrestricted area (at 130 meters from the stack under the most unfavorable atmospheric conditions) would be approximately 0.473 mrem for the year July 1,1979 through June 30, 1980.

a m

V-17 F. All AfillVAL SUf4 MARY OF THE GEf!ERAL METHODS Afl0 THE RESULTS OF EllVIR0flMEf4TAL SURVEYS PERFORMED OUTSIDE THE FACILITY The environmental radiation monitoring program will

. be categorized according to onsite and offsite environ-mental monitoring systems. A description of the two

! categories follows.

l. The Onsite Environmental Monitoring Systems _

Onsite radiation monitoring programs which we believe qualify as environmental radiation monitoring systems include the facility radioactivity stack monitor, onsite area moni-toring film badges, TLD's and 0-200 mrem gamma-sensitive integrating ionization chambers (self-reading pocket dosi-

. meter type), and the monitoring procedures associated with the analysis of radioactivity in liquid effluents from the reactor facility. Also, routine (daily, weekly, bi-weekly and monthly) direct radiation surveys conducted by the OSU TRIGA radiation protection staff provide a wealth of essential information on existing radiation conditions throughout the various onsite areas.

The reactor facility radioactivity stack monitoring system consists of a continuously-movino-filter-paper par-

. ticulate monitor, followed by a separate chamber which functions as a gas monitor. The system is consistently .

)

.- placed in operation before the reactor is started up, remains'in operation at all times while the reactor is

V-18 in use, and is kept operable after reactor shutdown until both detection channels reach normal background. The system is equipped with an isokinetic sampling head

, which draws its sample near the point of discharge in the reactor building stack. The system is calibrated at least annually with standardized particulate samples of appropriate types and energies, and with known quan-tities of argon-41 gas. The system reads out continu-ously in both the particulate and gaseous channels, with each channel having its own count rate meter and recorder.

A count lategrating scaler is also attached to the gas channel to increase the accuracy of determining argon-41

, released. The system is equipped with alarm circuits which will automatically shut off the facility ventilation system and close dampers on the intake and exhaust lines in the event preset airborne radioactivity concentration limits are reached. One of the most valuable applications of this system from the standpoint of environmental 1

monitoring is the data derived 'from its operation which '

can-be applied to determining potential exposures in unrestricted areas from gaseous radioactive effluents.

. Ons'ite area' monitoring film badges consist of standard personnel-typa beta-gamma-neutron film packs, e . located at strategic positions inside the reactor facility operating-area (see Figures V-1 and V-2).

The films within the facility are changed once per month.

.V-19 4

Onsite area monitoring using .TLD's now consists of two' different types of dosimeters, both located at  ;

f . identical positions on the reactor area fence (see

~

_ Figure V-2). One type of TLD' monitor is supplied and interpreted by our vendor, Radiation' Detection Company (R.D. Co.),

$ Mt. View, California. The vendor supplied -system utilizes CaSO TLD's prepackaged by R.D. Co., and exchanged on 4

I a quarterly basis. These dosimeters replace the R.D.

Co. beta-gamma-neutron film packs previously used on

- the reactor area fence. The R.D. Co. TLD's 'are located in'the same thin sheet netal boxes previously used to  ;

E ,

house the film packs, and are accompanied at each lo-f cation by the second TLD monitoring package which is l

prepared and interpreted by OSU. Each OSU TLD moni-

! toring device.normally consists of'three lithium i

fluoride chips, presently Harshaw TLD-700's, exchanged

on a quarterly basis.

Prior to April 1976,' each onsite group of three OSU .

TLD ' chips was packaged first'in a plastic mount which l was then placed inside an outer container cor.sisting'of - I a thin walled copper -tube. - The copper tube was subse-

_quently taped to the' reactor area fence. The plastic mount and copper container were essentially identical i

a e

i i

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V-20 i

l-to those presently being used by the Oregon Radiation Control Section in their TLD program. In April 1976, i

the copper tube outer containers were discontinued for the OSU supplied TLD's on the reactor area fence, and the remaining inner plastic mounts were placed inside thin sheet metal boxes located at each of the reactor area fence monitoring stations. This was done to reduce data loss due to increasing theft of the small copper tube TLD packs. OSU and R.D. Co. TLD packs are currently located at each of the nine reactor area fence positions identified in Figure V-2.

l In addition to the above monitoring devices, each of the nine reactor area fence monitoring positions

( is presently equipped with two 0-200 mrem gamma-sensitive i

i .

= integrating ionization chambers (self-reading pocket dosimetertype). These dosimeters are located inside i

the thin sheet ~ metal box at each fence monitoring l station, which also contains the two different TLD l

I monitoring packets. The. ionization chamber dosimeters are read every two weeks and are used as backup monitors for each station.

9 s

i j

V-21 For the July 1,1979 through June 30, 1980 reporting

' period, an additional onsite environmental monitoring program was conducted. This program involves the bi-weekly (every 2 weeks) measurement of the direct radiation exposure rate in terms of microroentgens per hour (pR/hr) at each reactor fence monitoring station. Measurements are taken with an Eberline Instrument Co. micro-R per hour rate meter containing a 1" x 1" Na! detector. The bi-weekly readings (normally 26 annually) are then averaged and ultimately converted to an expected (calculated) annual mrem dose equivalent for each location.

In terms of environmental monitoring, onsite area monitoring films, TLD's, integrating ionization chambers, and direct radiation exposure rate measurements at appropriate locations may be used to estimate maximum potential doses in nearby unrestricted areas due to direct radiation from the reactor facility. Normally, these estimates are made to reflect the annual dose equivalent which could be delivered in the unrestricted area assuming continuous occupancy, although occupancy of unrestricted areas adjacent to the reactor facility is virtually zero throughout the year.

1 The routine analysis of gross radioactivity in l liquid effluents (with isotopic identification as appropriate) prior to discharge into the unrestricted

V-22 area allows evaluation of the reactor facility contri-bution to potential radiation exposures to the general public from this source.

2. The Offsite Environmental Monitoring Systems Offsite environmental monitoring systems useful as indicators of potential radiation dose in unrestricted ,

area due to reactor operations include a soil, water, and ves*tation monitoring program, a,nd an airborne gamma monitoring program.

The soil, water, and vegetation monitoring program centers around the collection of a limited number of samples in each category on a quarterly basis. It is operated in conjunction with the reactor facility and the OSU Radiation Center, and considered useful for in-dicating general trends in gross radioactivity concen-trations for the substances sampled. See Figure V-3 for the location of sampling positions for G-Grass, S-Soil, W-Water, and RW-Rain Water.

The airborne gama monitoring program is generally

. described on pages 4-59 and 4-60 of the August 1968 Safety Analysis Report for the OSU TRIGA Reactor. As of January 1,1975, nine additional offsite airborne gamma monitoring stations were implemented to increase the total rumber of these stations now in use to nine-O teen. See Figure V-3 for the location of the nineteen airborne gama monitoring stations.

V-23 As of January 1,1975, the coding technique used to designate each specific offsite monitoring station was modified slightly to indicate the radiation monitoring devices present at a particular station. Under the new coding system, stations which contain only a standard OSU TLD monitoring pack (described previously in this report) will have an "L" after the station number. For example, MRCTE-2L is interpreted as M_onitor Radiation C_ enter ,TRIGA E_nvironmental Station number 2 with a standard OSU TLD pack in a copper tube being the only monitoring device at this station. (NOTE: The copper tube outer container is still used for all OSU TLD packs employed in the offsite environmental monitoring program.

They were discontinued only for the OSU TLD's used on the reactor area fence). At offsite stations where only dn OSU TLD monitor is used, the copper tube containing the TLD's is taped directly onto a mounting post or other permanent object used to identify the monitoring station.

Stations which have no "L" after the station number con-(

l sist of a thin weather-tight aluminum box mounted on a post about four feet off the ground. Each of these l

l stations includes one R.D. Co. TLD pack, one standard

?

l OSU copper tube TLD monitoring pack identical to those previously described, and two 0-200 mrem gamma-sensitive integrating ionization chambers (self-reading pocket dosimeter type) as backup monitors. At these stations, i

1

i

[

V-24 i

the OSU TLD's are not enclosed inside the aluminum box, but instead the copper tube is taped directly onto the l box mounting post at the station.- All TLD monitors in l, the offsite airborne gamma environmental monitoring pro-l j gram are exchanged on a quarterly interval beginning l

l January 1 of each year. The ionization chamber (dosi-l meters) are read once every 2 weeks throughout the year.

t For the July 1,1979 through June 30, 1980 reporting period, the previously described program for biweekly measurements of the direct radiation exposure rate in pR/hr at each reactor fence monitoring station was ex-tended to include each of the nineteen airborne gamma

, monitoring stations. The data was handled in the same manner as already mentioned and the objective was to derive an expected (calculated) annual mrem dose equivalent for each location based on an annual average pR/hr-exposure rate.

A summary of the environmental monitoring results for the year ~ July 1,1979 through June 30, 1980 is given below, and includes, as appropriate for the measurement l under consideration:

(a) The number of sampling _ locations.

l (b) The total number of samples per year.

O i

- - - , _ _ . , _ , - m , ,- , _ -. . -

V-25 (c) The annual average concentration of gross radioactivity, and in some cases, concen-trations of specific radionuclides in the medium b'eing assayed.

(d) The total annual millirem of external rad- -

iation dose for a particular location as well as a general description of that location.

The data from the environmental monitoring systems will be arranged to correspond to the specific individual systems identified previously in conjunction with onsite and offsite programs.

l j Reactor Facility Stack tionitor, (onsite):

(L) The system has one sampling location as indi-cated previously.

l (b) Samples are continuous; (i.e., prior to, during, and after reactor operation). It is normal for the stack monitor to begin operation as one of the first systems in the morning and to cease operation as one

! of the last systems at the end of a normal operating day.

(c) The annual average concentration of gross radioactivity based on the facility stack i monitor is given in Table V-2. As indi-cated in this table, the only gaseous component identified has been argon-41,

V-26 and only naturally occurring particulate radio-activity (radon daughter products) has been detected by the particulate channel.

The normal concentration for the naturally occurring particulate daughters during the reporting period remained about the same as in previous years, and was within a range of 4.19 x 10-10 pCi/cc to 4.30 x 10-12 pCi/cc.

Reactor Facility Area Monitoring Film Badges, Reactor Fence TLD's, Integrating Inization Chambers and Direct Radiation fleasurements (onsite):

(a) There are presently eight applicable area monitoring film badges within the TRIGA reactor facility operating-area. There are also nine vendor (R.D. Co.) supplied CaSO4 Tl.D monitors plus nine standard OSU TLD monitoring packs and eighteen (2 per station) 0-200 mrem gamma-sensitive inte-grating ionization chambers on the reactor area fence. There are also nine specific locations (the fence monitoring stations) where routine biweekly pR/hr measurements are made. All of these have application as onsite environmental monitors.

V-27 (b) Since each film badge within the TRIGA facility is changed once per month, there is a total of 96 different saaples of this type each year.

Quarterly changes of the fence TLD monitors result in another 36 vendor supplied TLD samples and 108 OSU TLD samples (9 stations x 3 TLD chips per station x 4 changes per year = 108 samples) for these locations each year. The eighteen integrating ionization chambers are read once every two weeks and thus result in approximately 468 samples (readings)eachyear. There are normally a total of 26 pR/hr measurements made at each of the nine fence monitoring stations each

~

year for a total of approximately 234 such measurements annually.

(c) TRIGA internal sampling locations are iden-tified in Figure V-1, with film badges being located on the inside of the indicated walls at approximately head height above the floor.

Locations of the film badges are coded Monitor Radiation C, enter TRIGA, N, orth badge, E,ast wall (MRCTNE) and so on. Locations for the TRIGA internal film badges plus the lo-

$ cations of the fence monitors are shown in

V-28 Figure V-2. Fence monitoring locations are coded Monitor Radiation Center Fence Environ-mental-1 (MRCFE-1) and so on through MRCFE-9.

, TLD monitors on the fence are in sealed moisture-resistant packages inside thin sheet metal mailboxes about four feet off the ground. The inte0 rating ionization chambers are also contained in the metal boxes. Total annual levels of radiation exposure recorded at the area monitoring

, locations are given in Tables V-6 and V-7.

Analysis of Reactor Contributed Radioactivity in Liquid Effluents, (onsite):

(a) TRIGA liquid effluent is analyzed before release to a collection point, and is analyzed again in conjunction with other radioactivity prior to discharge from the collection point into the unrestricted area.

! (b) The total number of samples were as follows:

July 1,1979 through June 30,1980 = 2 reactor liquid effluent samples before release to the collection point. July 1,1979 through June 30, 1980 = 2 reactor liquid effluent samples before release from the collection point to the un-i restricted area.

~

V-29 (c) The liquid effluent data for environmental assessment have been summarized for the re-porting period in Table V-1. Section V-E-2 of this report also addresses the estimated 9

level of external radiation from radioactivity

. in the liquid effluent.

Soil, Water and Vegetation Monitoring Program, (offsite):

(a) For this program there are now a total of 22 sampling locations: 4 s' oil locations, 4 water locations (when water is available), and 14 vegetation locations.

(b) Samples (as available) are caken at each location on a quarterly basis. Samples have been a

collected as follows:

1 July 1,1979 through  ;

June 30, 1980 '

Total number of samples = 86

- Total number of soil samples = 16 Total number of water samples * = 14 Total number ~ of vegetation samples = 56

• (Water sampling locations lW and 4W were each dry on one sampling date during the year July 1,1979 through June 30,1980.)

1 a - - - . -e- p - - w ----p' m y--, , y a w w

V-30 (c) The annual average concentration of gross beta radioactivity for the offsite environmental 6

soil, water and vegetation samples is given in Table V-8. Identification of specific radionuclides is not routinely carried out as part of this program, but would be conducted if unusual radioactivity levels above natural background were evident. Locations of sampling points relative to the reactor facility are given in Figure V-3, and as shown in this figure, most locations are within a 1000 foot radius of the reactor building.

In general, samples are collected over a local area having a radius of about 10 feet at the positions indicated in Figure V-3.

Airborne Gamma Monitoring Program. (offsite):

(a) The offsite airborne gamma monitoring program currently utilizes nineteen stations, and each station is considered a sampling location.

Presently, eleven stations have a vendor (R.D. Co.) supplied CaSO 4 TLD monitor, plus a standard OSU TLD monitoring pack, and two 0-200 mre.n gamma pocket dosimeters. Eight stations have only a standard OSU TLD moni-toring pack. In addition, each of the nine-Lteen monitoring stations is included in the ongoing program for measurement of the pR/hr exposure rate.

l V-31

~

_( b) The TLD's at each airborne gamma monitoring '

station are chenged once every calendar quar-i 1

+

ter for. a total of 44 vendor TLD samples per year, and a total of.228 OSU TLD samples per year (19 stations x 3 TLD chips per station x 4 changes per year = 228 samples). The two backup' monitors (integrating ionization chamber dosimeters) are read every two weeks, which results in approximately 572 individual dosimeter readings each year. There are normally a total of 26 pR/hr measurements made at each of the nineteen airborria gamma monitoring  !

l

stations each year for a total of approximately 1

494 individual measurements annually.

4 (c) Locations of the nineteen airborne gamma monitoring stations are shown in Table V-3.

i Like the soil, water, and vegetation sam- l 1

l pling locations, most of the airborne gamma i

! monitoring stations are within a 1000 foot '

radius of the reactor building. These lo-

cations generally correspond to the atmo-spheric (plume) dispersion' results mentioned

. earlier in this report.

~ The results reported for the airborne gamma moni-toring stations are summarized _ in . Table V-9, and are based on:the vendor supplied TLD data, tiie OSU TLD data, I

. , - - . , _ . - , . . _ . . . . _ , _ . . , _ , _ , _ _ . . _ _ . , .- __..__.._m._ . . . -

V-32 and results obtained from the pR/hr measurements. sea footnote (6) of Table V-9 for a further explanation of the pR/hr data and its application.

This is the third complete year for the vendor 9

supplied TLD monitors, which were substituted for the previous vendor supplied environmental film packs.

As already indicated in the last paragraph of section E-1 of this report, OSU is still somewhat reluctant to accept the R.D. Co. TLD data without greater confidence in their QA and control procedures.

As mentioned, we plan to continue a careful assessment of both our program and their's in this area to ascertain whether our suspicions are real or not. Future reports will hopefully show closer agreement between the OSU and R.D. Co. TLD results, or will provide data allowing one to state more clearly the reasons for the differences.

For this reason, we would favor a national quality assurance program for dosimetry suppliers.

Our in-house OSU TLD program was started in 1974, and we believe a number of improvements have been made in the program since that time. There are, however, a few aspects which we continue to improve and some which may still require added refinement. In particular, we are still continuing to study our reported TLD back-ground-for the airborne gamma monitoring stations, and still do not believe reported values are always i

l

V-33

representative of what most stations are experiencing.

We increased the number of background stations during 1976, and between July 1, 1979 and June 30, 1980, we

! continued to make a series of direct background measure-ments with our pR/hr monitoring equipment (started July 1,1977) in order to obtain a better profile of the background variation. The results continue to increase our faith in our background values, but we plan to extend our study of this variable.

From our viewpoint, the major purpose of the air-borne gamma monitoring stations is to give an indication of general increases or trends in unrestricted area radiation levels which might be linked to argon-41 l

released from the OSU TRIGA. Past experience (over 1

! the last ten years) has shown that annual results per location vary slightly from' year to year. Al though

the data have not been included in this report, by following the mrem per year history for a single station I and comparing the annual mrem total for that station to the curies of argon-41 emitted for the corresponding l

year., it becomes evident that there is no consistent pattern to the results, and that other factors must i

l be responsible for the :ainor mrem per year variations.

l l For example, such variations may be the result of small l

annual differences in cosmic or terrestrial background,

fallout, etc. In any event, the small amount of

V ;

argon-41 released annually does not seem to be a signifi-cant factor which effects the total mrem per year re-ported (or predicted) for any particular monitoring  !

station. A comparison of the data contained in Table V-9  ;

to past results from these monitoring stations, and a comparison to the values in footnote (6) of Table V-9, leads us to the conclusion that there has been no meaning-ful increase in the unrestricted area gamma radiation levels due to argon-41 released by the OSU TRIGA during

the defined reporting period.

D h

7 l

l t

s j

e

.m -,-* - .y e -

, - ,..~, - , -..- - ,y-- - - .- =

V-35 i

1 - References i

! r

1. Eisenbud, Merril, Environmental Radioactivity, Second Edition, p.190, Academic Press, New York, NY (1973),

t

2. U.S. Environmental Protection Agency, " Estimates of Ionizing

! Radiation Doses in the United States, 1960-2000," ORP/CSD j

72-1, Office of Radiation Programs, Rockville, Maryland l (1972).

t i

! 3. U.S. Environmental Protection Agency, " Radiological Quality of i the Environment in the United States,1977," EPA 520/1-77-009, Office of Radiation Programs; Washington, D.C. 20460 (1977).

i s'

l 1.

l

)

i 1

• s

)

I J

i' ,

7

• i-I

.i

, - - - ~ , . - - - , - ,,m, , ,..-. ,- r , , . -

Tabl6 V-1 MONTHLY StM4ARY OF LIQUID WASTE DISCHARGES FOR THE YEAR JULY 1,1979 THROUGH JUNE 30, 1980 III Date of Total Quantity of Detectable Specific Activity for Total Curies of Average Concentra- Percent of Total Discharge Radioactivity Radionuclides Each Radioactive Each Detectable tion of Released Appilcable Volume of (Month & Year) Released (To in the Waste N terial in Waste Radionuclide Radioactive Material MPC for Liq. Effluent, Sanitary Sewer) Discharge Where Released in aa at Point of Release Released Including (Curies) Released Concentra- Waste (To Sanitary Sewer) Radioactive Diluent ,

tion After Dilution (Curies) (pC1/cc) Material Released was > 1.0 x 10-7 pCf/cc (1) (To Sanitary (pCi/cc) Sewer)

(Gallons)

July- 79 NONE NONE NOT APPLICABLE NONE NOT APPLICABLE NOT APPLICARE MONE August- 79 NONE NONE NOT APPLICABLE NONE NOT APPLICABLE NOT APPLICABLE NOME September- 79 NONE NONE NOT APPLICABLE NONE NOT APPLICABLE NOT APPLICABLE NONE October- 79 NONE NONE NOT APPLICABUc NONE NOT APPLICABLE NOT APPLICABLE NONE November- 79 NONE NONE NOT APPLICABLE NONE NOT APPLICABLE NOT APPLICABLE MONE December- 79 Co -= ------- 1.62 x 10- 2.24 x 10-5 -0.79 1898 1.61 x 10-4 H 2.24 x 10 1.61 x 10, Janua y-80 NONE NONE NOT APPLICABLE NOME NOT APPLICAR E NOT APPLICARE MONE

' February- 80 NONE NONE NOT APPLICABLE NONE NOT APPLICABLE NOT APPLICARE NONE March- 80 NONE NONE NOT APPLICABLE NONE NOT APPLICABLE NOT APPLICABLE NONE April-80 NONE NONE NOT APPLICABLE NOME NOT APPLICABLE NOT APPLICARE NONE 46 2.54 x 10-8

_May-80 Sc --------------

51 Cr ====--- - ---

8.12 x 10-8 54 Mn -------------- 4.55 x 10-8 2.40 x 10-5 58 2.11 x 10 0.10 3000 C0 -------------- 5.55 x 10-8 60 Co -------------- 9.48 x 10 -8 75 Se -------------- 6.19 x 10-8 3

H 2.08 x 10-0 2.36 x 10-5 June-80 NONE NONE NOT APPLICABLE NONE NOT APPLICARE NOT APPLICARE MONE ,

-4 4898(2) h"("*,I 1.85 x 10-4 SEE ABOVE NOT APPLICABLE 1.85 x 10 9.97 x 10-6 0.37 II OSU operational policy is to subtract only detector background from our water analysis data and I not background radioactivity in the Corva11 h city water. M (2) Total volume of effluent plus diluent does not take into consideration the additional mixing with approximately 95,000 to 115,000 gallons per year of liquids and sewage normally discharged by the Radiation Center complex into the same sanitary sewer system.  ;

T

- . O Table V-2 MONTHLY SupmARY OF GASEOUS WASTE DISCHARGES FOR THE YEAR JULY 1.1979 THROUGH JUNE 30, 1980 Total Total Estlested Average Percent of the Total Estimated Average Estimated Average Percent of Estimated Estimated Atmospheric Applicable Quantity of Concentration Concentration of MPC if the Date of Radioactivity Quantity of Diluted MPC for Diluted Radioactivity in of Radioact6ve Other Significant Estimated Discharge Released Argon-41 Concentration of Concentration Particulate Form Particulates Radionuclides in Release was (Month & Year l (Curtes) ReleasedLI) Argon-41 at of Argon-41 at fe Released With Discharge if >20% of the (Curles) Point of Release Point of Release with Half-( i

>B Days Half-Life >8 Days >20% of the Appi tcable (Reactor Stack) (Reactor Stack) (Curies) (Curles) Applicable MPC MPC (uC1/cc) (%) (uti/cc)

July- 79 0.47 0.47 3.10 x 10-8 0.78 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE August- 79 1.28 1.28 8.40 x 10 -8 2.10 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE September-75 0.% 0.56 3.79 x 10 -8 0.95 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE October-79 1.73 0.73 1.40 x 10-8 0.35 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE' November- 79 3.05 3.05 2.08 x 10'I 5.20 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE December-79 1.43 1.43 9.40 x 10-8 2.35 MONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE January-80 0.67 0.67 4.41 x 10-8 1.10 NOME NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE February-80 3.48 3.48 2.45 x 10'I 6.13 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE March-80 1.76 1.76 1.16 x 10'# 2.90 NOME NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE April-80 3'.57 3.57 2.43 x 10'I 6.08 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE May- 80 2.76 2.76 1.82 x 10'I 4.55 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICAPM June-80 3.05 l 3.05 2.10 x 10'I 5.25 MONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE 23.81 I3) a1 23.81 53) 1.26 x 10'I 3.15 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE U} Routine gansna spectroscopy evaluation of the gaseous radioacWity in the stack discharge indicated that it was virtually all argon-41.

h (2) Evaluation of the particulate radioactivity in the stack discharge confirmed its origin as naturally occurring radon daughter products, predominantly lead-214 and bismuth-214. not associated with reactor operations. .

II The increase in total argon 41 released during the current reporting period parallels very closely the increase in reactor megawatt-days (MWD) operating time for the same period.

V-38 Table V-3

, ANNUAL

SUMMARY

OF SOLID WASTE DISCHARGES F0". THE YEAR JULY ~1, 1979 THROUGH JUNE 30, 1980

}

}

Total Amount Detectable Total Quantity Dates of Shipment of Solid Waste . Radionuclides of Radioactivity and Disposition (l) 1 Packaged in the Waste in Solid Waste (Cubic Feet) (Curies) 21.00 Sodium-24 5.96 x 10-4 September 25, 15/3 Chromium-51 May 27,1980

Manganese-54 Cobal t-58 Iron-59 Cobalt-60 Zinc-65 II)All solid radioactive waste is transferred to our radioactive waste disposal service vendor, Nuclear Engineering Company, for burial at their installation at Richland, Washington.

r a

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, . .7, _ - . - . , , - , -

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V-39 Table V-4 3 ANNUAL

SUMMARY

OF RADIATION EXPOSURE RECEIVED BY FACILITY PERSONNEL AND VISITORS FOR THE YEAR JULY 1,1979 THROUGH JUNE 30, 1980 Average Annual Exposure Greatest Individual for Each Personnel Group Exposure per Personnel Group p

Whole Body Extremities Whole Boc'y Extremities (mrem) (mrem) (mrem) (mrem) a perating 34.00 490.00 115.00 2110.00-per 1 5.00 67.00 20.00 180.00 e ear sonnel Facility Visitors:

Film Badges- 0.00 (1) 0,00 (1)

Pociat Dosimeters 1.00- (1) 38.00 (1)

(I)0SU TRIGA reactor policy does not normally allow people in the visitor category to become actively involved in the use or handling of radiation or radioactive materials. Therefore, visitor extremity dosimeters are not normally necessary and no visitor data are available for the extremities.

V-40 Table V-5 ANNUAL

SUMMARY

OF RADIATION LEVELS AND CONTAMINATION LEVELS OBSERVED DURING ROUTINE RADIATION SURVEYS FOR THE YEAR JULY l, 1979 THROUGH JUNE 30, 1980 s

a i

Direct Radiation Levels Contamination Levels (mrem /hr)(Sy+ neutrons) (dpm/iOO cm 2

)(gy)(3)

Average Maximum Average Maximum Reactor Top <l.00 143.00 <370 <370 Sample Handling Area <l.00 143.00 <370 <370 Reactor Room Floor <l.00 120.00 5370 1370 Beam Port Facilities <l.00 96.00 1370 1370

. Outside Inside Outside Inside Demineralizer Tank Shield Shield Shield Shield .

Avg. Max. Avg. Max. Avg. Max Avg. Max.

<l.00 3.00 39.00 150.00 (2) (2) (2) (2)

Class Experiments <l.00 200.00 <370 <370 (I)No contamination was found at the designated locations during the entire reporting period. The 370 dpm/100 cm 2 value used in this table is based on the normal beta counting efficiency and a net count rate equal to the normal back-ground counting rate for the portable survey meters routinely used in the field to screen for radioactive contamination.(i.e., field measurements would normally have to show a gross counting rate equal to twice the normal background counting rate before co camination would be considered present). However, in addition to normal field r reening for contamination by direct surveys and smear samples, i those smears suspected of containing removable radioactive contamination are i routinely counted in a more sensitive radiation detection system. Based on usual l counting times, a normal instrument counting efficiency, and a typical background counting rate, during the current reporting period such a detection system typically provided a lower limit of detection (LLD) at 95% confidence of approxi-mately 11-12 dpm for the radionuclides-normally expected to be ca the smears.

Smearing efficiency for radioactivity removal is conservatively assumed to be 410%, and positive smear results would usually be multiplied by 10 before final conversion to dpm/100 cm2 ,

(2)Not an applicable measurement.

V-41

+ -- l o '- A t_

MRC THXW Figure V-1

\

DEMause Auz ER EXCH AN GE.R Ptav ro a m 20'/ LOCATIONS FOR THE TRIGA REACTOR S*

N -

4.(, MRcTHxS TRENCH

( NEAT ExcwAusca )

Damit Ta =

r1 A c- I t r1 n r

$$* 6y FAN Room MRCT5w MR.CTNW l

- l MRC TWA! %

N .

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l TRIGA l RE ACTOR LA15 GO' w

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T SUB CRnTtCAL plLE sma  ;

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MRC TSE % MRC TNE l m , i'

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V-42 Figure V-2

\

AREA RADIATION MONITOR LOCATIONS FOR THE  :

TRIGA AND AGN REACTORS, AND THE TRIGA REACTOR AREA FENCE

• 138'  ;

pq -- - - -- - g - - - _-- _

1 mRC FES mRC FE2 s

I

\ s 8

fo' I

8 mRC FE4 M4 mRC FE1 di ymRC AGNJ V i e 1 msc rNt mac isc LR/hr measurements taken at two week intervals throughout the year. The total mrem for the period is calculated by multiplying <

this average pR/hr value by 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> per year and then by converting microroentgens to millires. Normal pR/hr values for the I-U.S. (terrestrial plus cosmic radiation) range between about 7.0 and 11.0 pR/hr (Ref.1)(excluding areas of unusually high natural radioactivity). These exposure rates correspond to annual dose equivalent totals of about 59 to 93 rRem per year.

The U.S. EPA (Ref. 2.3) estimates the total annual terrestrial plus cosmic cose equivalent for Oregon to be about 110 mBem per year.

(6)TLD monitoring packets are exchanged on a quarterly interval.

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V-46 Table V-8 ANNUAL AVERAGE CONCENTRATIONS OF GROSS BETA RADI0 ACTIVITY FOR OFFSITE ENVIRONMENTAL S0IL, WATER, AND VEGETATION SAMPLES FOR THE YEAR JULY l, 1979 THROUGH JUNE 30, 1980 0

Sample Identification Number, Annual Average Concentration of Gross Type & Reporting Units Beta Radioactivity (minus 3H)(1)(2)(3) 1-water (pC1/cc) 4.39 x 10-0 + 1.15 x 10-0I4) 4-water (uCi/cc) 9.11 x 10-8 1 2.19 x 10-8(4) ll-water (uCi/cc) 9.87 x 10-8 1.06 x 10-8 19-rainwater (pCi/cc) 4.80 x 10-8 1 1.00 x 10 -0 3-soil ( "

am of ry soil) 7.24 x 10-5 1 2.88 x 10-6 5-soil (gram o"f ry soil) 3.79 x 10-5 2.37 x 10-6 20-soil (gram of ry soil) 6.14 x 10-5 2.72 x 10-6 21-soil (gg,,g"f ry soil) 6.25 x 10-5 2.69 x 10-6 2-grass (gram of ry ash) 1.53 x 10-4 4.15 x 10-6 l

1 3 ' 6-grass (gram of ryash) 1.27 x 10 4 13.62 x 10 7-grass (gramof ryash) 1.84 x 10-4 4.31 x 10-6 1

7

. V-47 Sample Identification Number, Annual Average Concentration of Gross Type & Reporting Units Beta Radioactivity (minus 3H)(1)(2)(3)

-4 8-grass ( " 1.05 x 10 3.69 x 10-0 gf d a2) a

" 8.90 x 10-5 2.98 x 10-6 9-grass (gram of d ad)

" 7.51 x 10-5

• 3.79 x 10-6 10-grass (gram of a2) 12-grass (gram of d y ash) 1.70 x 10-4 4.14 x 10-6

" 1.28 x 10-4 i 4.11 x 10-6 13-grass (gp,, gf d y ash) 14-grass (gram of d ash) 9.29 x 10-5 3.91 x 10-6 15-grass (g of y ash) 1.18 x 10-4 3.55 x 10-6 16-grass .(gra of d ag) 8.81 x 10-5 i 3.85 x 10-6 4

1.09 x 10-4 3.47 x 10-6 17-grass (gram of d y ash) 18-grass (gram of d y ash) 1.08 x 10-4 1 3.54 x 10-6 3

22-grass (gram of d y ash) 1.59 x 10-4 3.90 x 10-6 (1)1 values represent the standard deviation at the 95% confidence level .

(2) Annual average concentrations were calculated using sample results .

which exceeded the lower limit of detection (LLD), except that sample results which were s the LLD were averaged in at the corresponding LLD concentration.

V-48 (3)For this report, the lower limit of detection (LLD) has been defined as the smallest amount or concentration of radioactive material in a sample that has a 95%' probability of being detected. It is equivalent to 4.66 times the standard deviation of the detection system's background counting rate -obtained with a-blank sample, provided the relative standard deviation of the background rate (the coefficient of variation) is less than 25%.

For the year July 1,1979 through June 30, 1980, the LLD for gross S in water samples averaged 2.34 x 10-8 pCi/cc and ranged between 7.64 x 10 8 pCi/cc and 1.10 x 10-8 pCi/cc. For gross S in vegetation samples, the LLD

, averaged 8.93 x 10-5 pCi/gm and ranged between 1.29 x 10-5 pCi/gm and 8 3.55 x 10-' pCi/gm. For gross 8 in soil samples, the LLD averaged 7.11 'x 10-5 pCi/gm and ranged between 7.91 x 10-8 pCi/gm and 6.53 x 10-5 pCi/gm.

(4)This sample location was dry for one calendar quarter;during the reporting period (the third quarter of 1979). -Therefore, no sample was collected for this interval.

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Table V-9 ANNUAL TOTALS FOR OFFSITE AIRBORNE GAP 91A MONITORING STATIONS FOR THE YEAR JULY 1, 1979 THROUGH JUNE 30, 1980 Total Calculated mrem for the Year Total Recorded mrem,for the Year July 1, 1979 Through June 30 Total Recorded mrem for the Year 1980 Based on the Annual Average July 1, 1979 Through June 30, 1980 July 1,1979 Through June 30, 1980 Monitoring Station pR/hr Exposure Rate Measured Based on R.D. Co. TLD's (1)(7) Based on Standard OSU TLD't (4)(5)(7) at Each Location (5)(6)

MRCTE-2L MRCTE-3

---

(2) 82.0 t 11.0 65.0 t 26.0 106.0(3) 115.0 1 17.0 76.0 t 21.0 MRCTE 101.0 MRCTE-5L ==-

81.0 i 13.0 70.0 i 15.0 84.0

• 28.0

' MRCTE-6 111.0 75.0

• 25.0

,MRCTE-7L --------

77.0 t 16.0 82.0 t 22.0 MRCTE-8 76.0 i 11.0 76.0 23.0 110.0 89.0 t 12.0 MRCTE-9 120.0 86.0 i 28.0 -

MRCTE-10 102.0 t 36.0 86.0 t 20.0 104.0 82.0 1 11.0 MRCTE-11 105.0 67.0 i 19.0 MRCTE-12 63.0 t 12.0 60.0 t 23.0 112.0 MtCTE-13L 68.0 ti 12.0(8) 85.0 t 21.0 MRCTE-14L 66.0 73.0 t 24.0 MRCTE-15 8.0(8) 75.0 t 8.0 79.0 t 31.0 -

108.0 79.0 t 9.0 MRCTE-16L 81.0 1 25.0 MRCTE-17 77.0 t 10.0 75.0 1 20.0 105.0 66.0 1 12.0 MRCTE-18L 72.0 1 19.0 MRCTE-19 89.0

• 15.0 72.0

• 15.0 120.0 90.0 t 9.0

, MRCTE-20L 81.0 t 21.0 93.0 i 12.0 71.0122.0 UI Radiation Detection Cor,sany (R.D. Co.), Mt. View, California, TLD totals include their annual natural background contribution of 85.0 mrem.

Corvallis area natural background using R.D. Co. TLD's totals s93 mrem for the same period.

(2) Monitoring stations coded with an "L" contain one standard OSU TLD monitoring packCo.

(No. R.D. only.

TLD pack.)

III Monitoring stations and one standard notTED OSu coded with an pack.

monitoring "L" contain one R.D. Co. TLD monitoring pack, two 0-200 mrem gama pocket dosimeters, I4I OSU offsite airbo reporting period. gansna TLD totals include a measured annual natural background contribution of 77.0 t 8.0 mrem for the (5)* values represent the standard deviation at the 95% confidence level.

(6)The annual average microroentgen (pR) per hour exposure rate for each location is normally detemined by averagt 26 separate pq/hr measurements taken at two week intervals throughout the year. The total mrem for the w riod is calculated by multiplying this average pR/hr value by 8760 hours0.101 days <br />2.433 hours <br />0.0145 weeks <br />0.00333 months <br /> per year and then by converting microroentgens to millirem.

(terrestrial Nomal uR/hr values for the U.S.

radioactivity . lusThese cosmic radiation) range between about 7.0 and 11.0 pR/hr (Ref.1)(excluding areas of unusually high natural exposure rates correspond to annual dose equivalent totals of about 59 to 93 mrem per year. The U.S.

EPA (Ref. 2,3 III TLD monitoringestimates the total annual terrestrial plus cosmic dose equivalent for Oregon to be about 110 mrem per year.

packets are exchanged on a quarterly interval, <

(8) Totals for reporting period are based on three calendar quarter's data.The TLD monitors for one quarter were stolen. h 4