Text

Annual Operating Rept for Jul 1978-June 1979
	ML19290A080
Person / Time
Site:	Oregon State University
Issue date:	06/30/1979
From:	Anderson T, Andrea Johnson, Ringle J; Oregon State University, CORVALLIS, OR
To:	;
Shared Package
ML19290A079	List: ML19290A078; ML19290A080;
References
	NUDOCS 7909100260
	Download: ML19290A080 (110)
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OREGON STATE UNIVERSITY TRIGA REACTOR ANNUAL REPORT To satisfy the requirements of:

A. U.S. Nuclear Regulatory Commission License R-106 (Docket No. 50-234),

Section 6.7(e) of the Technical Specifications, for the period July 1, 1978 through June 30, 1979.

B. U.S. Department of Energy Fuel Fabrication Contract No. EY-76-C-06-1953, for the period July 1, 1978 through June 30, 1979.

C. Oregon Department of Energy, DOE Rule No.30-010, for the period July 1, 1978 through June 30, 1979.

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 Radiation Center Oregon State University Corvallis, Oregon 97331 Telephone: 503-754-2341 August 31, 1979 30,-3

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TABLE OF CONTENTS page I. INTRODUCTION AND

SUMMARY

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

SUMMARY

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

SUMMARY

OF OSTR ENVIRONMENTAL AND RADIATION PROTECTION DATA I-3

1. Liquid Waste Date . . . . . . . . . . . . . . . . . . . I-3

2. Gaseous Waste Data. . . . . . . . . . . . . . . . . . . I-4

3. Solid Waste . . . . . . . . . . . . . . . . . . . . . . I-4

4. Radiation Exposure Received by Facility Personnel and Vis i to rs ( i n mrem) . . . . . . . . . . . . . . . . . . . I-5

5. Number of Area and Offsite Environmental Monitorin Samples Evaluated . . . . . . . . . . . . . . . . g ... 1-5 II, GENERAL INFORMATION ...... ........ 11-1 A. RADIATION CENTER. . . . . . . . . . . . . . . . . . . . . . II-l B. FACULTY MEMBERS HOUSED AT THE RADIATION CENTER. . . . . . . II-2 C. RESEAhJH 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-5 D. CLASSIFIED STAFF AT THE RADIATION CENTER. . . . . . . . . . II-5 E. REACTOR OPERATIONS STAFF. . . . . . . . . . . . . . . . . . II-5 F. REACTOR OPERATIONS COMMITTEE. . . . . . . . . . . . . . . . II-6 G. RADIATION S'FETY COMMITTEE. . . . . . . . . . . . . . . . . II-6 2057 306

Page III. OPERATIONAL IlATA . . . . . . . . . . . . . . . . . . . . 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-12

1. 10 CFR 50.59 Changes. . . . . . . . . . . . . . . . . . III-12

2. Other Changes . . . . . . . . . . . . . . . . . . . . . III-12

3. Planned Changes . . . . . . . . . . . . . . . . . . . . III-13 F. MAINTENANCE AND SURVEILLANCE. . . . . . . . . . . . . . . . III-16

1. Ma i n te n an ce . . . . . . . . . . . . . . . . . . . . . . III-16

2. Tests and Inspections . . . . . . . . . . . . . . . . . III-17 G. REPORTABLE OCCURRENCE . . . . . . . . . . . . . . . . . . . III-17 IV. UTILIZATION DATA . ..................... IV-1 A. TEACHING PROGRAMS . . . . . . . . . . . . . . . . . . . . . IV-1 B. RES E ARCH P ROGRAMS . . . . . . . . . . . . . . . . . . . . . IV-8 C. PUBLICATIONS RESULTING FROM OSTR OPERATIONS THAT WERE REPORTED TO THE RADIATION CENTER. . . . . . . . . . . . . . IV-15

1. Publica tions in Print . . . . . . . . . ........ IV-15

2. Publications in Press . . . . . . . . . . . . . . . . . IV-16

3. Reports and Papers. . . . . . . . . . . . . . . . . . . IV-16 D. COMMERCIAL OR NON-ACADEMIC UTILIZATION. . . . . . . . . . . IV-17 E. PUBLIC RELATIONS. . . . . . . . . . . . . . . . . . . . . . IV-17 F. PLANNED CHANGES IN UTILIZATION. . . . . . . . . . . . . . . IV-17 2057 307

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I. Page V. ENvlR0tiMENTAL AND RADI ATION PROTECTION DATA; JULY 1, 1978 - JUNE 30, 1979 . . . . . . . . . . . . V-1 i

, A. IflTRODUCTION . . . . . . . . . . . . . . . . . . . . . .. V-1 B. A

SUMMARY

OF THE NATURE AND AMOUNT 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 . . . . , , . . ,.. . ,, V-1

1. Liquid Waste (summarked on a, monthly bas.is) . . . . . V-l (a) The radioactivity disctarged durin; the reporting period ... . . . . . . . . . . .. V-l (1) The total estimated jiuantity of radio-activity released isin 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 s

(4) A summary of the total release (in curies) for each radionuclide determined in (2) s above 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 (inclt ding diluent) released d ring each period of release .~ .. . . . . . . . . . .. V-2 Y

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2. Gaseous Waste (summarized on a monthly basis) . . . . V-2 (a) The radioactivity discharged during the reporting period . . . . . . . . . . . . . . . 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/cu~oic centimeter) and the fraction of the applicable MPC value . . . . . . . . . . . . . . . . . . V-3 (5) The total estimated quantity cf radic-activity in particulate form with half-lives greater than e ght days (in curies) released durirg the re-portir.s ceriod,as determined by an appropriate particulate mnitoring system. . .......... ...... V-3 (6) The average concentration ot radioactive particulates with half-1" ves greater than eight days (in microcuries/cubir centimeter) released during the re-porting period ....... ...... V-3 (7) An estinate of the averagn concentration of other significant racionuclides present in the gaseous i aste discharge (in terms of microcurie ;/ cubic centi-meter) and the fraction of the applicable f1PC value for the repercing period,if the estimated release was greater than 207, of the applicable !!PC .......

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3. Solid tlaste (summarized on an annual basis). . . . V-4 (a) The radioactivity discharged during the reporting period ............. 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 shi pped o ff-si te) . . . . . . . . . . . . . V-4 C. AN ANilVAL

SUMMARY

OF THE RADIATION EXPOSURE RECEIVED BY FACILITY PERSONilEL AtlD BY VISITORS Irl TERMS OF THE AVERAGE RADIATION EXPOSURE PER ItiDIVIDUAL AtiD THE GREATEST EXPOSURE PER IrlDIVIDUAL FOR EACH OF THE TWO GROUPS . . . . . . . . . . . . . . . . . . . ..... V-4 D. Att ANilVAL

SUMMARY

OF THE RADIATION LEVELS AtlD THE LEVELS OF C0!!TAMINATIO!! OBSERVED DURING ROUTItiE SURVEYS PER-FORMED AT THE FACILITY IN TERMS OF THE AVERAGE AND THE HIGHEST LEVELS ................... V-5 E. THE LOCATI0tl AtlD 11AGilITUDE OF THE MAXIMUM MEASURED OP, CALCULATED DIRECT RADIATION LEVEL Ill UtlRESTRICTED AREAS DUE TO DIRECT RADIATION FROM THE FACILITY, AND DIRECT RADIATI0fl FROM FACILITY EFFLUEtlTS . ... 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 11ETH03S AND THE RESULTS OF EllVIR0flMENTAL SURVEYS PERFORMED OUTSIDE THE FACILITY ................ ...... V-16

1. The Onsite Environmental Monitoring Systems . .

V-16

2. The Offsite Environmental Monitoring Systems . . V-21 2?057 310

-vi-LIST OF TABLES TABLE TITLE PAGE Table III-l Three 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-18 Table III-9 Quarterly Tests & Inspections . . . . . . . . III-19 Table III-10 Semi-Annual Tests & Inspections . . . . . . . III-20 Table III-11 Annual Tests & Inspections ......... III-21 Table IV-1A OSTR Teaching Hours . . . . . . . . . . . . . IV-3 Table IV-18 OSTR Operator Training Hours ........ IV-3 Table IV-2 Statistics of Students in Nuclear Engineering and Nuclear Science Courses . . . . . . . . . IV-4 Table IV-3 Other Educational Institutions Using OSTR . IV-6 Table IV-4 Graduate Students Doing Thesis Research that Used the OSTR . . . . . . . . . . . . . . . IV-7 Table IV-5 OSTR Research Hours . . . . . . . . . . . . . IV-9 Table IV-6 Sunnary of Oregon State University TRIGA Research Projects and Funding Agencies ... IV-10 Table IV-7 Radiation Center Scheduled Visitors, July 1, 1978 - June 30, 1979 ........ IV-18 2057 31I

-vii-TABLE TITLE PAGE Table V-1 Monthly Summary of Liquid Waste Discharges for the Year July 1, 1978 through June 30, 1979 .................... V-35 Table V-2 Monthly Summary of Gaseous Waste Discharges for the Year July 1,1978 through June 30, 1979 .................... V-36 Table V-3 Annual Summary of Solid Waste Discharges for the Year July 1,1978 through June 30, 1979 . V-37 Table V-4 Annual Summary of Radiation Exposure Received by Facility Personnel and Visitors for the Year July 1, 1978 through June 30, 1979 . . . V-38 Table V-5 Annual Summary of Radiation Levels and Con-tamination Levels Observed During Routine Radiation Surveys for the Year July 1, 1978 through June 30, 1979 . . . . . . . . . . . . V-39 Table V-6 Total Dose Equivalent Recorded on Operating-Area Film Badge Monitors Located Inside the TRIGA Reactor Facility for the Year July 1, 1978 through June 30, 1979. . . . . . . . . . V-42 Table V-7 Total Dose Equivalent at the TRIGA Reactor Area Fence for the Year July 1,1978 through June 3 0, 197 9 . . . . . . . . . . . . V-43 Table V-8 Annual Average Concentrations of Gross Beta Radioactivity for Offsite Environmental Soil, Water, and Vegetation Samples for the Year July 1, 1978 through June 30, 1979 ..... V-45 Table V-9 Annual Totals for Offsite Airborne Gamma Monitoring Stations for the Year July 1, 1978 through June 30, 1979 ......... V-48 2057 312

-viii-LIST OF FIGURES FIGURE TITLE PAGE Fig. III-l OSTR Annual Energy Production Vs. Time (Fiscal Year) . . . . . . . . . . . . . . . III-3 Fig. III 2 Symmetrical Care Before June 1979 . . . . . III-14 Fig. III-3 Assymetrical Care, June 1979 ....... III-15 Fig. V-1 Operating-Area Film Badge Monitor Locations of the TRIGA Reactor ........... V-40 Fig. V-2 Area Radiation Monitor Locations for the TRIGA and AGN Reactors, and the TRIGA Reactor Area Fence ............ V-41 Fig. V-3 Monitoring Stations for the OSU TRIGA Reactor, January 6, 1976 through June 30, 1979 ................... V-44 2057 313

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I. INTRODUCTION AND'

SUMMARY

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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 in-troduced in the 1976-77 report.

2. The reporting period will be for one year: 1 July 78 to 30 June 79.

3. All of the information included in this annual report may not be of interest to all recipients and will reqt ..e 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 year i for the original 20% enriched core. A table showing the important operational data for this period (1967-1976) is included as Table III-2, however. This year's report will review the operating history of the 70% enriched FLIP core (1976-present).

-he 70% enriched FLIP core is established as the historical base for subsequent reports.

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

B.

SUMMARY

OF OSTR USE DURING REPORTING PERIOD During the year July 1,1978 to June 30, 1979 the Oregon State TRIGA Reactor:

1. Generated 10.6 MWD of energy.

2. Consumed 13.3 grams of 235U.

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3. Pulsed 130 times.

4. Two fuel elements were added to the core during the reporting period.

5. Acconmodated 10 courses in nuclear engineering and nuclear engineering technology; six courses in nuclear chemistry; and provided demonstrations for classes in lower division chemistry and general science. Also, 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months of reactor time was used to furnish special training in reactor operation for four Taiwan Power Company students. (Reactor use time for teaching and instruction totaled 202 hours0.00234 days 0.0561 hours 3.339947e-4 weeks 7.6861e-5 months .)

6. Two reactor operators are starting 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 three hours of reactor time has been used in the training during this reporting period.

7. Accommodated 52 research projects. (Reactor use time for research programs totaled 363 hours0.0042 days 0.101 hours 6.001984e-4 weeks 1.381215e-4 months .)

8. During a typical week, operated about 22 hours2.546296e-4 days 0.00611 hours 3.637566e-5 weeks 8.371e-6 months / week.

9. Accommodated 1,095 scheduled visitors and several hundred unscheduled visitors during university open house events.

(Reactor use time for visitor demonstration totaled 31 hours3.587963e-4 days 0.00861 hours 5.125661e-5 weeks 1.17955e-5 months .)

10. Reactor use time averaged 55%, based on a 40-hour week (eight hours a day, five days a week).

Research technicians from "un Ismail Atomic Research Centre.

I-3 C.

SUMMARY

OF OSTR Ef1VIR0f1MEllTAL A!1D RADIATI0fl PROTECTI0fl DATA Year July 1,1978 Through June 30, 1979

1. Liquid Waste Data (See Table V-1):

a. Total estimated quantity of radioactivity released (in curies)* 2.30 x 10-5

b. Detectable radionuclides in liquid waste 51Cr, 54Mn.

58Co, 60Co 3

H

c. Estimated average concentration of released radioactive material at the point of release (in microcuries per cubic centimeter) 2.32 x 10-6

d. Percent of applicable MPC for released liquid radioactive material at the point of release (%) 0.174

e. Total volume of liquid effluent released, including diluent, (in gallons)** 2620

The OSU operational policy is to subtract only detector background from our water analysis data and not background radioactivity in Corvallis city water.

- Total volume of effluent plus diluent does not take into con-sideration the additional mixing with apprc imately 95,000 to 115,000 gallons per year of liquids and sewage normally dis-charged by the Radiation Center complex into the same sanitary sewer system.

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I-4 Year July 1,1978 Through June 30, 1979

2. Gaseous Waste Data (See Table V-2):

a. Total estimated quantity of radioactivity released (incuries) 10.62 4I

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

c. Estimated average atmospheric diluted concentration of Argon-41 at the point of release (in micro-curies per cubic centimeter) 5.97 x 10-8

d. Percent of applicable fiPC for diluted concentration of Argon-41 at the point of release (%) 1.49

e. Total estimated release of radioactivity in particulate form with hal f-lives greater than 8 days (in cucies)** NONE

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

a. Total amount of solid waste packac.J and disposed of (in cubic feet) 9.50

b. Detectable radionuclides in solid waste 60Co, 59Fe, 24Na, 56Mn, 51Cr, 75Se

c. Total radioactivity in solid waste (in curies) 6.51 x 10 -4

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 sr.ack 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-5 Year July 1,1978 Through June 30, 1979

4. Radiation Exposure Received by Facility Personnel and Visitors in mrem) See Table V-4 :

a. Facility operating personnel (mrem)

(1) Average whole body 8.00 (2) Average extremities 208.00 (3) Maximum whole body 30.00 (4) Maximum extremities 680.00

b. Facility research personnel (mrem)

(1) Average whole body 0.00 (2) Average extremities 82.00 (3) Maximum whole body 0.00 (4) Maximum extremities 450.00

c. Visitors (mrem)

(1) Average whole body 1.00 (2) Maximum whole body 25.00

5. Number of Area and Offsite Environmental Monitoring Samples Eva?uated:

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 fance 108

d. Integrating ionization chambers on the reac cor facility fence 468

e. pR/hr measurements around the peri-meter of the reactor facility fencc. 234

f. Offsite environmental soil samples 16

g. Offsite environmental water samples 2057 319 14

h. Offsite environmental vegetation samples 56

I-6 Year July 1,1978 Through June 30, 1979

i. Offsite vendor supplied TLD monitors 44

j. Offsite OSU TLD monitors 228

k. Offsite integrating ionization chambers 572

1. pR/hr measurements at the offsite airborne gama monitoring stations 494 2057 320

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II GENERAL INFORMATION u242alb i

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II-l II. GENERAL INFORMATION A. RADIATI0il CErlTER 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 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 C. ar 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.

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3. Neutron radiography and neutron diffraction.

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

5. Measurement of various types of ionizing radiation.

6. Consultation in the application of radioisotopes and radiation research.

7. Exploratory programs on the novel uses of radio-isotopes and radiation.

B. FACULTY MEMBERS HOUSED AT THE RADIATION CENTER

Wang, Chih H. (Professor)

Director, OSU Radiation Center Reactor Administrator Head, Department of Nuclear Engineering

Bennett, Casey W. (Instructor)

Chemistry (nuclear chemistry)

Binney, Stephen E. (Associate Professor)

Nuclear Engineering (nuclear instrumentation)

Daniels, Malcolm (Professor)

Chemistry (radiation chemistry)

Dodd, Brian (Assistant Professor)

Nuclear Engineeri.'g (health physics)

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

Physics (radiation chemistry)

Hornyil., Karl (Associate Professor)

Nucle . Engineering (safety analysis and reactor kinetics)

Jansen, George (Visiting Professor)

Nuclear Engineering (nuclear fuel cycle)

Johnson, Arthur G. (Associate Professor)

Nuclear Engineering (health ohysics)

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

General Science (radiation biology) 2057 323

Reactor users for research and/or teaching.

II-3

Loveland, Walter D. (Associate Professor)

Chemistry (nuclear chemistry)

Peddicord, K. Lee (Assistant Professor)

Nuclear Engineering (themohydraulics)

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 (neutron radiography and fuel management)

Schmitt, Roman A. (Professor)

Chemistry (neutron activation analysis-Lunar geology)

Spinrad, Bernard I. (Professor)

Nuclear Engineering (reactor design and nuclear fuel cycles)

Thomas, T. Darrah (Professor)

Chemistry (photoelectron spectroscopy)

Woods, W. Kelly ('lisiting Professor)

Nuclear Engineering (energy systems analysis)

Reactor users for research and/or teaching.

C. RESEARCH PERSONNEL HOUSED AT THE RADIATION CENTER

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

Ma, Maw-Suen Chemistry R.A. Schmitt

2. Graduate Students Name Degree Field A_dvisor Ades, Maurice PhD Nuclear Engr K.L. Peddicord Bomben, Ken PhD Chemistry T.D. Thomas

Reactor users for research and/or thesis work.

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II-4 Graduate Students (continued)

Name Degree Field Advisor

Chick, Steve PhD Chemistry W.D. Loveland

Conard, Roberta MS Chemistry R.A. Schmitt

Dzata, Francis K.A. MS Chemistry R.A. Schmitt Fyke, David R. MS Anal Chen S.E. Binney Guidotti, Timothy E. MS Nuclear Engr K.L. Peddicord

Ghannam, Lina M. MS Chemistry W.D. Loveland Hedberg, Thomas MS Rad Biology D.J. Kimeldorf Hindalolla, Suraj PhD Chemistry R.A. Schmitt Huang. K.Y. MS Nuclear Engr A.H. Robinson Jackstn, John T. MS Physics A.H. Robinson Joo, tan Sem MS Nuclear Engr B.I. Spinrad 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

Narccor-Tsey, Winfred MS Nuclear Engr J.C. Ringle

Nelsen, Janet MS Nuclear Engr A.H. Robinson

Nielsen, Larry MS Nuclear Engr K.L. Peddicord

-0ylear, Joan M. MS Nud aar Engr K.L. Peddicord Ozaki, Calvin MS Rad Biology D.J. Kimeldorf

Poeton, Richard MS Gen Science A.G. Johnson Polkinghorne, Steve MS Nuclear Engr S.E. Binney Prici c ad, Andrew MS Nuclear Engr B.I. Spinrad

Priest, George PhD Geology R.A. Schmitt Reardon, Patrick T. MS Nuclear Engr B.I. Spinrad Reid, Bruce MS Nuclear Engr K. Hornyik

Rivera, Ma Rita MS Chemistry R.A. Schnitt Robinson, Cheryl A. MS Nuclear Engr K.L. Peddicord

Scherpelz, Robert I. MS Nuclear Engr S.E. Binney

Schofield, Paul MS Nuclear Engr A.H. Robinson Scott, James D. MS Rad Biology D.J. Kimeldorf Sittner, Valerie J. MS Rad Biol D.J. Kimeldorf

Smith, Monty PhD Nuclear Chem R.A. Schmitt Snowhill, Elaine MS Pharmacy V. Smith

Sterbentz, James MS Nuclear Engr K.L. Peddicord

Taylor, Cynthia MS Geology R.A. Schmitt

Tollefson, Dennis A. MS Nuclear Engr A.H. Robinson Tripathi, Amitabh MS Nuclear Engr B.I. Spinrad

Ungerer, C. Andy MS Chemistry W.D. Loveland Van, Phuong Dong MS Nuclear Engr A.H. Robinson

Wang, Lancelot S.K. MS Nuclear Engr A.H. Robinson Wong, Bright M.K. MS Nuclear Engr B.I. Spinrad Wu, Chi-Hung PhD Nuclear Engr B.I. Spinrad Yoshihara, Grant M. MS Nuclear Engr K. Hornyik Youssefnia, Mohammad H. MS Nuclear Engr J.C. Ringle

Reactor users for research and/or thesis work.

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

Abu, M.P.H. Reactor Operations (Malaysia) J.C. Ringle

Ali, Yusof Reactor Operations (Malaysia) J.C. Ringle

Barton, John P. Neutron Radiography (IRT Corp.) C.H. Wang Besar, Idris Health Physics (Malaysia) A.G. Johnson

Fawa ri s , A.H . Radioecology (Louisiana St. U.) W.D. Loveland

Fukuoka, Takaaki Meteorite Studies (Japan) R.A. Schmitt

Gooding, James Geology (Univ. of flew Mexico) R.A. Schmitt

Hamzah, Razali Reactor Operations (Malaysia) J.C. Ringle

Kamat, Kasbun Reactor Operations (Malaysia) J.C. Ringle

Khair, Nahrul Reactor Operations (Malaysia) J.C. Ringle

Knaus, Ronald M. Nuclear Chemistry (Louisiana St) W.D. Loveland

Laul, J.C. Neutron Act. Anal. (BNWL) R.A. Schmitt

Leeman, William Geology (Rice University) R.A. Schmitt

Nunnelley, Lewis Nuclear Chemistry T.D. Thomas Pilus, A. Rahman Health Physics (Malaysia) A.G. Johnson

Yunus , Yaziz Reactor Operations (Malaysia) J.C. Ringle D. CLASSIFIED STAFF AT THE RADIATI0fl CENTER Name Titl e 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 Flickinger, Evelyn Secretary Keen, Robin Ann Administrative Assistant Moeller, Wanda Clerical Specialist Schneider, Mary K. Clerical Assistant Smith, Vernon N. Chemist Woodrow, Doyle Scientific Instrument Technician Ungier, Leon Research Assistant-Unclassified

Reactor users for research and/or thesis work.

E. REACTOR OPERATIONS STAFF Titl e Name Reactor Administrator C.H. Wang Assistant Reactor Administrator J.C. Ringle Reactor Supervisor T.V. Anderson 2057 326

II-6 REACTOR OPERATIONS STAFF (continued)

Titl e flame Senior Reacter Operators J.C. Ringle S.E. Binney T.V. Anderson J.M. Oylear (student)

Reactor Operators W.T. Carpenter Senior Health Physicist A.G. Johnson Health Physicist B. Dodd Radiation Specialist S.L. Bennett F. REACTOR OPERATIONS COMMITTEE Name Affiliation J.C. Ringle (chairman) Nuclear Engineering T.V. Anderson Radiation Center S.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 (Rad. Biol.)

G. RADIATI0fl SAFETY C0!!MITTEE Name Affiliation S.E. Binney (chairman) Nuclear Engineering A.G. Johnson Radiation Center John Kelley Radiation Safety Office J.E. Nixon Food Science & Technology C.C. Calligan Crmputer Center S.C. Fang Agricultural Chemistry D.J. Reed Biochemistry-Biophysics R.C. Worrest General Science 2057 327

f

.s l.r 4

1

- 111. 0PERATIONAL DATA 908M

-2057L328 1

s i O

III-l Ill. OPERATIONAL DATA A. REVIEW

l. The OSTR has operated for more than 12 years.

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

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 sho-t periods of time, due to the lack of sufficient cooling capacity.

4. In June 1971 the cooling capacity was upgraded to allow continuous operation at I f1W.

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 three year statistics with the FLIP core.

7. See Figure III-l for a graphical review of the OSTR's three 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 12 years, but will only report and review the FLIP core. It will, however, establish the new 70% enriched fuel as the historical base for subsequent reports, fiore detailed information concerning the 20% enriched itandard core can be obtained from the 1976-77 Annual Report dated 31 August 77.

2'057 329

III-2 Table III-l THREE YEAR OSTR STATISTICS 1 Aug 76 1 Jul 77 1 Jul 78 to to to FLIP Core 30 Jun 77* 30 Jun 78 30 Jun 79 Operating Hours (critical) 875 819 458 Megawatt Hours 451 496 255 fiegawatt Days 19 20.6 10.6 Grams 235U Used 24 25.9 13.4 Hours at Full Power (1 MW) 401 481 218 Number of Fuel Elements Added 85 0 2 to Core (initial loading)

Number of Irrad-iation Requests 443 375 329

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

2057 330

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 30 -

25 -

E A

ii!

!20 - O t; e a

e a I g 15 -

5 10 -

5 76I77 0

77-78 78 79 FISCAL YEAR 2057 331

Table III-2 OSTR 51ATISilCS WITH 207. [NR!CHED CORE TNLM7 to 'l 7tftoMJ, l JLC to M T AM76j to 1 ANF717to AIRT~l~AfD to to7T ~T APR to 7(~llW77~TATR to to 7f-10M i 30 JUN 68 30 JUN 69 i 31 MAR 70 31 MAR 71, 31 MAR 72 MR 67 to JR 76

,__ 31 MAR 73 31 MAR 74 31 MAR 75 31 MAR 76,26 JUL 76

! i . ; .. ... t ....

} .

. Operating Hours, I ! l (critical) i 904 ! 610 567 855 l 598 954 705 563 794 35 3 6903 cegawatt Hours 117.24 102.47 138.05 223.77 195.11 497.82 335.94 321.45 408 213 2553

^

W gawatt Days 4.88 4.27 5.75 9.3 8.1 20.74 13.99 13.39 17 9 106.4 Grams ? "U Used ; '

6.13 5 .36 , 7.21 11.7 10.2 26.031 17.57 16.81 21.35 10.7 133 I

. Hours at Full l 429 58 I' 856 Power (250 kW) ! 369 l -- -- -- --

i l Hours at Full '

' l' 20 23 100 l 200 291 460 205 1700 Tower (I HW) -- --

401 j l 1 l l fwitse r o f F uel i I I 6 Elew nts Added 70 2 13 1 1 1 2 2 2 0 , 44 I to Core (initial); ,

Numter of Irrad- i ! l lation Pequests 429 433 391 528 34 7 550 452 346 357 217 4100 Nu*r of Pulses 202 2 36 299 102 98 249 109 183 43 39 1560

Reat. tor became critical on March 8, 1%/ (/0 element core; 250 6W). ***Reactar shan ' wen June 1,1971 for one month for cooling system Note: This period length is 1.33 years as initial critical upgradt.iq.

occurred out of phase with the reporting period.

- - - Reactor shutdown July 26, 1976 for one wonth for refueling

- Reactor shutdown August 22, 1969 for one month for upgrading to reactor with new full FLIP core. Note: This period length 1 m (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 frne July-June tn April-Marc h. ]

E-N '

cD LD N

tr4 tr4 N

III-5 B. OPERATING STATISTICS The utilization of the OSTR for the reporting period declined 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 10.64 MWD. (The cumulative thermal energy generated by the FLIP core now totals 50.2 MWD for Aug. 1, 1976 to June 30, 1979). See Tables III-1 and III-3 through III-5 for this reporting period's statistics.

Reactor use time averaged $55% of our 8-hour day, 5-day week schedule.

Our present rate of excess reactivity decrease with the FLIP core is about 3c/ MWD. Our present core excess (after the addition of two FLIP fuel elements) is approxi-mately $6.70. The initial FLIP core excess was $7.17.

So far the reactivity loss per MWD with the FLIP core is about the same as with the 20% fuel. The fuel manufacturer (General Atomic) reports that with the FLIP fuel we should initially expect a 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 years and is a result of the burnable poison in the fuel .

2057 333

III-6 Table III-3 PRESEflT OSTR OPERATION STATISTICS FLIP 1 July 78 Cumulative to 1 Aug 76 Reactor Operations 30 Jun 79 to date

1. MWH of energy produced 255 1202

2. MWD of energy produced 10.6 50.2

3. Grans 23sU used 13.4 63.3

4. flumber of fuel elements added to core 2 84 + 3 FFCR**

5. Number of pulses 130 31 6

6. Hours reactor critical 458 , 2152 1

7. Hours at full power (1 M,1) 218 1100

8. fiumber of startup and shutdown checks 246 725 l

9. fiumber of irradiation '

requests processed

329 l

1147

10. flumber of samples irradiated 4526 , 16,300 l

Each request authorized from 1 to 120 samples to be irradiated (the number of samples per irradiation request averaged about 14)

- Fuel Follower Control Rod 2057 334

III-7 Table III-4A OSTR USE TIME FLIP 1 July 78 Cumulative Overall Reactor to 1 Aug 76 Operatic,n Time 30 Jun 79 to date Statistics (hours) (hours)

1. Checkout, core excess and shutdown 358 1033

2. Load and unload samples 91 306

3. Reactor in operation

699 2657

4. Total reactor use time 1148 3996

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 78 Cumulative In:pection and to 1 Aug 76 Demonstration Time 30 Jun 79 to date Statistics (hours) (hours)

1. Training (departmental) and others)L2 202 627

2. OSU research >3l 289 1374

3. Off-campus research >3 l 74 257

4. Reactor preclude time 552 1666

5. Visitor demonstration 31 72

6. Total reactor use time 1148 3990 IIncludes sample loading and unloading. -

2 See Tables IV-1 and IV-2 for teaching statistics. 20,37 "

3See Table IV-5 for research statistics.

/

III-8 Yable III-5 OSTR MULTIPLEI USE TIME l FLl?

1 Jul 78 l Cumulative to 1 Aug 76 30 Jun 79 to date Number of Users (hours) (hours)

1. Two users 37 220

2. Three users 10 49 4

3. Four users 0.2 1.7 I'

4. Total multiple 47.22 270.73 use time 1

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

210F, of total hours the reactor was critical this year.

312% of total hours le reactor was critical since startup with FLIP toel August 1976.

2057 336

/

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 number; 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.)

B-29 Reactivity measurements for fuel worth.

B-30 Irradiatior, 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.

2057 337'

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,10 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.)

2057 338

III-ll Table III-6*

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

43 -- 17 50 l

B-3 148 !

15 42 10 i 1 21 6 !

B-8 l 4 l 1 25 --

6 i 36 i !

B-21 28 3 -- -- --

l 31 B-23 i -- i -- 7 -- --

7 B-24 ! 4 -- --

l -- -- 4 i

i i ,

B-25 --

1, -- 4 -- -- 4

, l l i '

i j l

B-29 ,

-- -- 3 -- --

3 I

I ,

i l B-30 -- -- 2 -- --

2 I

B-31 '

1 1 l

TOTAL 184 20 126 , 10 24 364 l

Table displays the number of times a specific experiment was used in a particular area.

2057 339

III-12 D. UNPLANNED SHUTDOWNS There was a total of three unplanned shutdowns that occurred during the reporting period. See Table III-7 for tabulation.

Table III-7 UNPLANNED SCRAMS Type Scram Occurrences Cause Linear Scram 2 Reactor power range switch turned the wrong .iay--usually by students--during, power increases.

External Scram 1 Beam port #1 has a series of inter-locks associated with beam closure and access to target area. One lcad shutter was not closed far enough to trip the micro switch when the access !

! door was starting to roll open caus- i j a scram. ;

E. CHANGES IN FACILITY

1. 10 CFR 50.59 Changes 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 August of 1978, the annunciator panel was moved from its temporary location, on top of the left console side cabinet, to a position in front of and above the operator. This change was made in anticipation of

[

III-13 adding a future instrument rack to the top of the left side cabinet. The annunciator panel position is an improvement over the old location.

b. In June of 1979, the core fuel elements were re-arranged to favor an increased neutron flux to beam port #4, and to increase the transient rod worth. The core configuration was changed from a symmetrical to an asymmetrical arrangement. See Figure III-2 for the symmetrical configuration and Figure III-3 for the asymmetrical arrangement.

The transient. rod worth increased from $2.25 to

$2.78. Mechanical and electrical stops are now used to restrict the transient rod worth to $2.35, as required by our Technical Specifications.

c. In June of 1979, a graphite filler plug was removed from beam port #4. This graphite plug was inserted into the reflector about six years ago to enhance the neutron in the tangential beam port (BP #3).

A water-filled aluminum plug was inserted into BP #4 to replace the graphite plug in hopes the neutron beam would be increased for the neutron radiography program.

The core excess decreased 23c with this change of plugs.

3. Planned Changes

a. He plan, in the very near future, to upgrade about nalf of our existing console electronics. A 2057 341

Symmetrical Core Before June 1979

,igure III-2 E

NUR :Pd%

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g ze 1<~1 eee

@ =<- g m eee 8== g ,,ec<,11 e mee o <.1, g ,,,,1, ,ee au m a ,_ 11 ,,c,< m

Asymmetrical Core June 1979 III-15 Figure III-3 W

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1 x@~w@O

@ _-",,', @ c - < 1,1, E g ze , 1<,1 ee

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REGULATING R0D EMPTY SAFETY R0e 2057 343 g emee11 ,,c,<11,

III-16 wide-range linear, wide-range log, and safety channel would be added, replacing our present linear, log, and startup channels. We have re-quested an amendment to our Technical Specifications in a letter dated April 16, 1979 to the NRC to accommodate this electronics upgrading.

b. We are also planning to increase our pulsed reactivity insertion limit from $2.35 to $2.55.

The letter dated April 16, 1979 to the NRC also requests a change in our Technical Specifications to allow this.

c. He plan to add new radiation surveillance equip-ment to our facility in the form of a new constant air monitor and a new area radiation monitor. This equipment has been ordered, and it should be installed, calibrated, and operational by the end of 1979.

F. MAINTENANCE AND SURVEILLANCE

1. Maintenance _

a. Oct '78: The reactor room exhaust fan motor shorted out. The field windings were replaced and the fan put back in service. The reactor down time was about two days total.

b. Jan '79: The regulating rod drive gear box was re-placed . The bull wheel in the gear train was worn out. We had a replacement gear box in our spare parts inventory; thus reactor down time was only about fivt nours.

2057 344

III-17

c. Jun '79: The transient rod drive piston ring was replaced . The old ring was deteriorating and letting too much air past the piston resulting in inconsistent reactivity in-sertion rates and peak power values during pulsing. The ring that was replaced was the original ring installed over 12 years ago. The reactor down time was about 2.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months .

2. Tests anr'. Inspections The OSTR has a routine test and inspection surveillance program. These T&I lists are presented in Tables III-8 through III-ll. Those items marked with an asterisk (*)

are required by the Technical Specifications.

G. REPORTABLE OCCURRENCE On June 7, 1979, the measured shutdown margin for the OSTR was $0.48, nine cents less than our Technical Specifications (Part 3.2) limitation of $0.57. This was reported to the NRC, Region V, Office of Inspection and Enforcement by telephone on June 8,1979 and by a letter dated June 19, 1979.

This incident occurred when the reactor core was changed from a symmetrical arrangement (See Figure III-2) to a asym-metrical arrangement. This change was made to accommodate a neutron radiography research program.

2057 345

. .;i. .i.-o ... ..

APPENDIX J T & I's FOR THE MONTH OF I DATE i DUE DATE : COMPLETED INITIAll

TEST OR INSPECTION TO BE PERFORMED I .

1.> Functional check of reactor water level alarms.

2. Measure reactor primary water system pH. l 3.l Measure the pH of the bulk shiled tank wah r. l 4.lCheckTRIGAtankwateractivity.

i ,

5. Emergency power systems battery liquid level .

fandterminalchecks. l 6.lEmergencyevacuationalarmsystembattery

! liquid level and terminal checks. I i i i

7. Ir.spect brushes on rabbit system blower motor. i l j 8.l Functional check of evacuation alarm. l l

9. Blow down the t. Ansient rod air accomulator tank.

10.lCalculatetheaveragemonthlyconductivity.

! (Average conductivity must be less than 5 jmicro mhos per centimeter.) l l 11.fChangethelightbulbinthegreenlight. l

, j l !

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

13. ' check filter tape speed on stack monitor i

1"/hr).

j14.!LubricatetheTRIGAtubeloadingtoolasneeded. >

l l I ; i l15.!Checkcamoillevel.

16.fPropanetankliquidlevelcheck(% full) l l 2057 346

III-19 Table III-9 T & I's FOR THE QUARTER OF DATE TEST OR INSPECTION TO BE PERFORMED DUE DATE COMPLETED lINITIAlf

1.i RCC audit reactor operations.

2.fInspectandoil (as needed) solenoid operated l lvalvesinrabbitsystem.

3. Ti=e the sample insertion and retrieve time interval of the rabbit system.

4. Check lazy susan for unknown samples. l l

5 Functional check of emergency lighting (16 units) l 6.[ Westronics Recorder: clean slide wire contacts.

7. Varian Recorders: clean, inspect and lubricate; j replace reference cells.

i .'

8. MAP IB I l

(a) Check HV source.

(b) Check ratemeter test position. l l l l (c) oil drive motors (fas : and slow).

9.. RM I-110 Area Monitors l l (a) Check HIGH & LCW alarms. l

, (b) Check ! S volt supply. j ;

I (c) Check HV (.70 te : use AX-30 test ard l calibrate meter.) l l 10., Arm system alarm checks: l j l I

t CIGN 1 1 2 l 3 ! 4 i 5 I 6 1 7 fl !

i ALT i i ! l !

! LIGHTI I I i 1 f

PANELi i l i I l l l l ,

i , ANN I I I I l l l 2

i i l

II. 'M Sper emluates operators w.% cornments.

2057 347

APethula L SEMI-ANNUAL T & I's FOR TEST OR INSPECTION TO BE PERFORMED DUE DATE INITIAL COMPLETED

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 Withdrawal Insertion l I

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. l t

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.

'0. Lubricate lazy susan drive and indicator assembly bearings.

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

12. Console: Perfcrm check list (Appendix I in GA manual 47615).

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

I

14. Westronics Recorder: Check zero and calibration.

} }kh

,,,, c_. v . . . a nNiUAL 6 a l's FUit DATE TEST OR INSPECTION TO BE PERFORMED DUE DATE INITIAL COMPLETED

1. [BI ANNUAL]

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

2. Annual report (due 30 June + 60 days)

3. Calibrate control rods.

4. Calibrate reactor power. '

5. Calibrate bulk H 2O temp. meter.

l

6. Calibrate the constant air monitor.

7. Stack Monitor (a) Calibrate particulate monitor.

(b) Calibrate gas monitor.

8. Calibrate the Area Radiation Monitor. i t 9. .c aduct evacuation drill.

t10.lCalibratethereactorwateractivitymonitor. !

til. Cou.n rate meter discriminator check.

Draw new curve. j 12.I Inspect standard rod drive mechanisms. l l

13. Change oil if needed in the thermal column door I 8 drive assembly reduction gear casing.

14. Change oil in cam blower and oil motor.

f !

15. Lubricate thermal column door drive assembly as needed.

i .

16.1 Check beam port loadinI tool hydraulic reservoir jlevel and lubricate me"hanisms as needed. ,

All Rx eperater requa lif. cat ion 4tS SNM inventer y t

19 Intrusien AlOrm Rerponse. Drif t !

2e Securi ty Guorci Trco ninq '

2057 349

III-22 The core change had been reviewed and approved by the Reactor Operations Co.nmittee. The change was carried out in sequences, with rod worths being measured and shutdown margins calculated at the end of each sequence. After the final sequence, the core excess reactivity had increased more and the control rod worths had increased less than was pre-dicted from the previous sequence. This resulted in a shut-down margin slightly less than the limit in the Technical Specifications.

As soon as this situation was discovered, two fuel elements were moved and the shutdown margin went up to $0.74.

See Figure III-3 for the final asymmetrical core arrangement.

More details regardinr, this incident can be found in our letter to NRC dated June 19, 1979.

2057 350

n. .

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. 1 W

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IV. UTILIZATION! DATA-pggR 8 MERE 3

2057l35.1- :

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4-4 ..

1 51

' ~ *,

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IV-1 IV, UTILIZATION DATA A. TEACHING PROGRAMS

1. The OSTR was used to accommodate 10 courses in nuclear engi-neering and nuclear engineering technology. These courtes were:

NE 101 Nuclear Engineering Orientation NE 102 Nuclear Engineering Orientation NE 441 Nuclear Reactor Experiments NE 442 Nuclear Reactor Experiments NE 203 Nuclear Radiation Detection and Measurement NT 203 Nuclear Radiation Detection and Measurement NT 311 Radiation Protection NT 312 Radiat')n Protectinn NT 491 Nuclear Technology Experiments NT 492 Nuclear Technology Experiments Six chemistry courses utilized the OSTR. They were:

CH 107 General Chemistry Laboratory CH 419 Radioactive Tracer Methods CH 528 Activation Analysis CH 316 Radiochemistry CH 207 General Chemistry Laboratory CH 515 Experimental Nuclear Chemistry

2. A special class was held for four Taiwan Power Company stu-dents, which was an introduction to nuclear reactor operations.

The training included at least 10 startups and shutdowns for each student, with power changes in manual and automatic modes, rod balancing, temperature effects, scram recovery, and approach to critical. A total of 29 hours3.356481e-4 days 0.00806 hours 4.794974e-5 weeks 1.10345e-5 months of reactor time was required for this class during the reporting period.

See Tables IV-1 A, IV-2, IV-3, and IV-4 for data showing the use of the OSTR to accommodate teaching and academic programs.

2057 352

IV-2

3. Two trainees from Malaysia arrived September,1978. They en-rolled in several academic classes and started their train-ing for a reactor operator license in June,1979. The Malay-sian operator training was integrated into the renular reactor schedule and therefore their hours of on-the-job training are not reflected in any of the teaching statistics. (See Table IV-1B.) The Malaysian operators are Yaziz Bin Yunis and Mohd.

Puad Bin Haji Abu.

2057 353

IV-3 Table IV '. A OSTR TEACHING HOURS ,

j 1 Jul 78 Cumulative to 1 Aug 76 30 Jun 79 to date

Description (hours) (hours)

Departmental 170 567 Chemistry (58)

Nuclear Engineering (83)

Nuclear Engineering Technology (29) i Special Classes i Reactor Operator Trainees l 29 57 :

Urban League Workshop 3 10 Total Teaching Time2 ,3 202r+ 634 1 Special training class was conducted for four Taiwan Poiter Company 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 Accamplishment Yaziz Bin Yunis Reactor Operator Mohd. Puad Bin Haji Abu Reactor Operator Total Training Hours 3 2057 354

IV-4 Table IV-2 STATISTICS OF STUDENTS IN NUCLEAR ENGINEERING AND NUCLEAR SCIENCE COURSES Number of Students l WIN SPR FAL Course Cr. Course Title 1978 1979 1979 Nuclear Engineering Cources NE 10l* 3 Nuclear Engineering Orientation 17 -- --

NE 102* 3 Nuclear Engineering Orientation --

18 --

NE 103x 3 Nuclear Engineering Fundamentals -- --

20 NE 201 3 Nuclear Energy Fundamentals 28 -- --

NE 202 3 Nuclear Radiation & Matter --

23 --

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

26 NE 406 Projects 5 3 6 NE 407 1 Seminar -- 20 --

NE 420 3 Intro. to Nuclear Reactor Analysis -- --

31 NE 421 3 Nuclear Reactor Analysis & Comput. 34 -- --

NE 422 3 Nuclear Reactor Analysis & Comput. -- 32 --

NE 423 3 Nuclear Reactor Analysis & Comput. -- --

30 NE 441** 3 ' Nuclear Reactor Experiments 29 -- --

NE 442** 3 ! Nuclear Reactor Experiments -- 23 --

NE 461 3 ' Reactor Thermohydraulics & Power Gen. 21 -- -- 1 NE 462 3 Reactor Thermohydraulics & Power Gen. 31 -- --

NE 463 3 Reactor Design -- --

17 ;

NE 464 3 Reactor Safety Engineering --

25 --

NE 465 3 Nuclear Fuel Cycle --

38 --

NE 481 3 Nuclear Materials -- -- 34 NE 501 Research 1 1 3 NE 503 1-15 Thesis 9 8 9 NE 505 Reading & Conference 1 6 12 NE 506P Projects --

NE 507 Seminar -- 2 --

NE 511 2 Neutron Trans. Theory 7 -- --

NE 512 2 Neutron Trans. Theory 5 --

NE 513 3 Nuclear Reactions Var. Thry. -- -- 5 NE 521 3 I Reactor Environmental Problems 6 -- --

NE 522 3 I

Reactor Safety Problems -- 4 --

NE 523 3 Advanced Reactor Design -- -- 7 NE 531 3 ; Nuclear Reactor Kinetics -- 7 --

NE 535 2 i Nuclear Reactor Burnup 11 -- --

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

17 --

NE 553 3 ! Computational Methods for Nuc. React. -- -- 10 NE 581 3 j Selected Topics in Reactor Theory -- 10 --

0STR used occasionally for demonstration experiments.

- 0STR heavily used.

~

2057 355

IV-5 Number of Students '

FAL WIN SPR I

Course Cr. Course Title 1978 1979 1979 Nuclear Engincering Technology Couroco NT 10l* 3 Nuclear Engineering Orientation 5 -- --

NT 102x* 2 Nuclear Engineering Orientation -- 3 --

l NT 103x 3 Nuclear Engineering Fundamentals -- -- 1 NT 201 3 Nuclear Energy Fundamentals 5 -- --

j NT 202 3 Nuclear Radiation & Matter -- 4 --

NT 203** 3 Nuclear Radiation Detection & Meas. -- -- 4 NT 311** 4 Radiation Protection -- 13 --

-- i 8 NT 312** 4 Radiation Protection --

Nuclear Engr. Calculations -- -- 7 NT 330x 3 Reading & Conference 1 NT 405 1 NT 405A 1 Reading & Conference (Nuclear Engr. Calc.) 10 -- --

-- 1 7 NT 406 Projects NT 407 Seminar 6 --

NT 410 3 Field Practice 2 -- --

NT 411 3 Nuclear Rules & Regulations 12 -- --

NT 413 3 Nuclear Plant Environmental Impact -- -- 11 NT 431 3 Nuclear Power Plant Technology 16 -- --

NT 432 3 Nuclear Power Plant Technology -- 14 --

NT 433 2 Nuclear Quality Assurance -- -- 13 NT 491 ** 3 Nuclear Technology Experiments 9 -- --

NT 492** 3 Nuclear Technology Experiments -- 9 --

Chemistry Courses CH 105 4 General Chemistry Lecture -- 6 31 94 i 2 67 General Chemistry Labs -- --

2 CH 316** 3 Nuclear Reactor Chemistry 32 -- -- .

CH 419** 4 Radioactive Tracer Methods 10 -- --

Experimental Nuclear Chemistry 4 CH 515** 3 CH 528** 3 Activation Analysis --

11 -

! l OU:cr Coursco i l '

GS 460 3 Radiation Health 13 -- --

4 !

Special Class for Taiwan Power Company -- --

0STR used occasionally for demonstration experiments.

- 0STR heavily used.

2057 356'

s-_

IV-6 Table IV-3 OTHER EDUCATIONAL INSTITUTIONS USING OSTR*

Number of Number of Number of Facul ty Students Visits to Involved Involved OSTR University of Oregon 5** 7** 7 Louisiana State University 1 0 1 i

Rice University 1 0 1 l

Urban League llorkshop 2 7 2 .

(Portland) i !

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."

- Includes researchers and students from other universities working through the University of Oregon. (See Table IV-4.)

2057 357

Table IV-4 GRADUATE STUDENTS DOING THESIS RESEARCH THAT USED THE OSTR Name Degree Department Advisor Thesis !

I Oregon State University A. Ungerer MS Chemistry Loveland ; Air Analysis R. Cobler MS Chemistry Dymond Galapagos Rise Hydrothermal Study R. Poeton MS General Science Johnson Effect of Cadmium Ratio on TLD Flux Calibration M. Smith PhD Chemistry Schmitt Chemical & Petrological Characterization of Individual Rock Casts in a Drecciated Meteorite K. Keasier PhD Chemistry Loveland Stable Activable Tracer for an Estuarine Environment G. Priest PhD Geology Taylor Analysis Little Walker Rocks University of Oregon G. Nixon '

PhD University of British Columbia Through the University of Oregon i Roberts N/A University of Georgia Through the University of Oregon i I

J. Metz PhD Biology Bishop A Study of Photosynthetic Apparatus in Green Algea S. Goldberg lPhD Geology Gol es Anorthosite Genesis C. Bow i PhD Geology Goles Geochemistry of Galapagos Lavas ,

H. Nashland !

PhD Geology McBirney Petrology of the Upper Border Group of the Skaergaard Intrusion C. White MS Geology McBirney Sodium Analysis of Mt. Hood Geologic Samples ;

N C~3 LJ7 2

-J L u

LJ)

CO

IV-8 B. RESEARCH PROJECTS Fifty-two research projects utilized 363 hours0.0042 days 0.101 hours 6.001984e-4 weeks 1.381215e-4 months of reactor time. Thirty-two of these research projects were from Oregon State University,19 were from the University of Oregon, and one was from Battelle florthwest in Richland, Washington.

Three of the Oregon State University projects were con-ducted and correlated with other universities and institutions.

These organizations were:

1. Rice University

2. Louisiana State University

3. Battelle florthwest Labor c'ory*

Several of the University of Oregon projects were also corre-lated with other universities. These institutions were:

1. McGill Univers.'ty (Montreal, Canada)

2. University of British Columbia (Canada)

3. Washington State University

4. University of Georgia

5. University of Tubigen (Germany)

6. University of California, Riverside.

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.

2057 359

IV-9 Table IV-5 OSTR RESEARCH HOURS 1 Jul 78 Cumulative to 1 Aug 76 Reactor Research 30 Jun 79 to date Hours Statistics (hours) (hours)

OSU Research* 289 1374 (About 10% of this time was thesis & research combined)

Off-Campus Research* 74 257 (About 14% of this time was thesis & research combined)

Commercial O 0

' I,

! Total Research* l 363 1631

Includes sample load'.ng and unloading time.

NOTE: Research hours, OSU funded: 114 Research hours, other funded: 249 2057 360

.Ta b_l e I_V-6

SUMMARY

OF OREGON STAIE UNfVERSITY TRIGA RESEARCH PROJECTS AND FUNDING AGENCIES r

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

1. R.A. Schmitt, Ag Chemist ,, Toxicol[of Brominated Oils 1.J. ..s l e y OSU INAA of Bromine in Marine Organisms N!!'IS

2. R.A. Schmitt Ag Chemistry Toxicology of Brominated Fatty Acids 1.J. Tinsley 050 Evaluation of the Distribution and N!!HS Absorption of drominated Fatty Acids

3. R. A. Schmitt Chemistry, OSU Lunar and Meteoritic Activation M. Smith Chemical & Petrological Characteri. NASA Analysts for Thesis (Ph.D.) zation of Rock Casts in a Brecciated Meteorite 4 R.A. Schmitt, Chemistry GU Chemical Studies of Lunar, INAA of Selected Samples NASA M.S. Ma Meteoritic and Terrestrial St.mples S. R.A. Schmitt, Chemistry. OSU Trace Element Studies of Volcanic W.P. Leeman INAA for Selected Trace Elements in Rice University

& Rice Univ. Rocks a variety of Volcanic Rocks From French and Hawaii Polynesia 6.

R.A. Schmitt, Geology, OSU Composition Analysis of Selected INAA for Trace Elements in Little G. Priest OSU Gedogy Dr.pt.

Volcanic Rocks for Thesis (Ph.0) Walker Volcanic Center

7. R.A. Schmitt Chemistry. OSU Carbonate Crys'

AA Crystals Found in Geothermal A. Weibel OSU Radiation Center Investigations to be Analyzed for Uranium & Thorium

8. R.A. Schmitt, Chemistry, OSU INAA of Ar ican Soils r Examine Soils Frou Kenya, Africa A. Weibel OSU '.adiation Cent r

+

to Correlate with Pubitshed Data

9. R.A. Schmitt, Forestry Sci. '

D INAA for Dy Tracer in Forest OSU L. Norris Lab, OSU f y Tracer studies Spraying Forestry Science Lab.

10. R.A. Schmitt, 050 l Western Wheat Quality control INAA for Funge Which is Found in E. Trione OSU Radiation Center I Wheat. To Determine any Of fference 1 in Elemental Content.

11 R . A . Sc hmi t t , Chemistry. OSU High Purity Crystal Production Crystals Analyzed for Purity to Help O' o fhemistry W. Fredericks Esplain Aberations on Light & Other Sources of Radiation

12. lR.A.Schmitt, Chemistry, OSU forensic Investigations INAA for Selected Trace Elements in Oregon Law Enforce-V.N. Smith a Variety of Forensic Samples ment Agencies N 13. R. A. Scinnitt, Chemistry, OSU Elemental Abundances in Meteoritic INAA & RNAA of Selected Chondrules NASA 2

g T. Tukuoka Specimens u Determine REE, Sr. and BG Content 4

' in Meteorites O CD '

O O

Table IV-6 (continuedl Listing Name of Person (s) Department and Number Using Reactor Institution Project Title Description funding Agency

14. R.A. Schmitt Chamistry, OSU Rock Studies of Terrestrial Basalts INAA of Terrestrial Basaltic Rocks NASA

15. l A.H. Robinson Nuclear Engr. I Neutron Radior raphy Studies of Development of High Speed Neutron D00 lLiquidPropeliaats Padiography of Burning Propellants

16. W. Loveland Chemistry. 05U Radiochemical Studies of Low Energy To Measure Mass, Charge, Angular. USDOE

& Relativistic Heavy Ion Reactions Energy, & Momentum Distributions by NAA of Target Materials

17. W. Loveland Chemistry OSU , Willamette River Tracer Experiment Develop a Tracer Method for Tracing USDI for Pollutants fluid Bound Substances in fresh Water

18. W. Loveland Chemistry. OSU Atmo aheric Trace Elements Assoc- [NAA of Atmospheric Air Samples of Water Resources lated With forest fires Forest fires & Slash Burning

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

70. W. Loveland Chemistry OSU Hydrospheric Trace Elements and Development of a Cheap, fast Means U.S. Dept. of Interior Their Uss in Wate Pollutant of Tracing the Behavior of fluid-Tracers Bound Substances in Natural Waters, such as Rivers

21. W. Loveland Chemistry OSU Precipitation Scavenging of Analyze for Trace Eleents in Rain Air Resources Center Tracers Released into frontal Water LOing NAA Storms

22. W. Loveland Chemistry, OSU Hydrospheric Trace Elemerts in Activable Tracers Being Developed USDI K. Keasier Water Pollutant Tracing for to Trace Soluble Materias; in an Thesis (PhD) Estuarine Environment

23. S.E. Binney Nuclear Engr. , i f xtraction of Uranium from Sea Various Muhods and Materials for Exxon Nuclear OSU l Water the Extraction of Uranium from research grant l Seawater are Beino Tested and l Analyzed 24 S.E. Dinney Nuclear Engr., NAA in the Zirconium Metals INAA of Samples f rom the Plant Teledyne Wah Chang OSU ! Industry for Uranium / Thorium Streams & Sludge Ponds in a Albany research grant Zirtonium Separation Plant

25. S.E. Binney Nuclear Engr., I " fluorine" !NAA of Zirronium Salts for Teledyne Wah Chang OSU .

Selected Elements Albany research grant

26 S.E. Binney Nuclear Engr., '

Thorium Assay Thorium Assay in Tissue by OSU Radiation Center [

N A.G. Johnson OSU Delayed Neutron Counting ,

g <_ _ _ _ . _ __ _

w CD CD CD N

__ M_ -.

Table IV.6 h ntinued]

l Listing Nanie of Person (s) Department and

Number Using Reactor Institution Project Title Description Funding Agency

27. A.G. Johnson Osu General E f fect of Cadmium patto on TLD To 5tudy the Variation of TLD Response OSU Radiation Center R. Poeton Science flum Calibrations to Thermal Neutrons as a f unction of Cadmium Ratio

28. J.P. Barton Nuclear Engr., Evaluate heatron Nadiography Development of Neutrcn Radiography of ANL OSU Capabilitaes for Reactor Large Fuel BunJles Inside Substantial Analysis and safety Program Steel Shrouds; The Results will t,e Specifically Relevant to the Fast Breeder Reactor Systems

29. K.S. Krane Physics, OSU Angular Correlation Measurements Nuclear Spectroscupic Study of Excited OSU Physici Department of Nuclear Gaouna Ray 5tates in Odj. Mass spherical Nuclet by Gamma-Gamma Angular Currelation Tech-nique

30. R.M. knaus Radiation Cente Lanthanum & Molybdenum Sorption INAA of Roots Exposed to La and Mo to OSU Radiation Center OSU 8 Louisiana in Willow Roots Determine Sorp to the Root Surface State University'

31. J. Dymond Oc e a no gra phy , i Galapagos Rise Fydrology Study Geochemical Study of Galapagos by NAA NSF R. Cot,ler 050 l For Thesis (Ph.D)

32. J. Coritss Oseanography, Galapagos Project INAA for Trace Metals of Trench Rock NSF 050 and Mn Crust, Sediments & Hydrothermal Geposits

33. J.C. Laul Battelle NW Lunar Chemical Charactert2ation INAA for Cbemical Study of Lunar and ERCA prime contract Meteorite samples EY-76-C-06-1830, sp.

1 agreement B-29210-pF

31. G.G. Goles Geology, Univ. ' Bushveld Granttes: Petrogenesis Study of Granites of the Bushveld Geological Society of MacCaskle of Oregon Compl ex , South Africa America I

35. G.G. Gales Geology, Univ. Rockwell Hanford Analys t s Determination of Compositions of DOE of Oregon Columbia River Basalts. Purpose to j see if Safe Reservoirs for Disposal

! of Radioactive Waste can be Found

36. G.G. Goles University of WACV-1 for Thesis (Ph.0) Dynamic Geochemistry of Iztaccthuati University of Oregon G. Nixon British Columbia (Snowy-Broad) & University of Through Univer- British Coli,mbia N __. _ _ _ _ _

O 37. G.G. Goles Geology, Univ. No Name Trace Element Analp ? s of Ruck University of Oregon 7 g of Oregon Powders for Coastal Mining Co. y

@ 33. G.G. Goles University of E xperimental Petrology of Test or Exp. Glasses for Na University of Oregon Mr. Welli Oregon Na-Glasses & NSF O

tr4

Ta',le IV.6 (t ont inued)

Listing Name of Person (s) Department and using Reactor Institution Project Title Desc ription Funding Agency Number

39. G.G. Goles, MtGill Univ. &

INAA of Mutton Bay 5yenttic Complex Trace & Rare Earth Element Deter- University of Oregon i

i Mr. Martin Univ. of Oregon & Associated Rodingites 'hinat f or of Syenites & Rodingites & McGill University

40. G.G. Coles, Univ , of Cal . Investigit!nn of Rocks of the John INAA of Selected Rock Samples for University of Oregon P. Robinson (Riverside) Day fm., Oregon Pe t rologie Interpreta t io'is & Univ, of Cal .

(Riverside)

41. G.G. Goles, Geology, Univ. Geological and Geochemical Survey INAA of Selected Volcanic specimens 0.0.G.M.I. State of C. White of Oregon of Young Cascade Volcanisms for from the Central Cascades (including Oregon Thesis (Ph.D) both Western Cascades and High Cas-cades) uf Oregon. EsPecially Mt.

Hood N f'

42. G.G. Goles, Geology, Univ. Petrology of the Upper Border Group INAA Trace Element Analysis to NSF H.R. Naslund of Oreg 1 uf the Shaergaard Intrusion Correlate Urper Burder Group 6

i to Major Elen.ent Data

} __ l 43. G.G. Goles, University of A Study of Photosynthetic Apparatus Determination of Manganese Levels NSF I N.I. Bishop, Oregon in Green Algae in Photosynthetic Membranes of Wild J. Metz Type & Mutant Algae l

u j 44. G.G. Goles, Geol ogy, taniv . INAA of Granite Rocks for Thesis INAA of Ore Samples Felated to NSF Throsgh the Univ.

Mr. Stormer, of Oregon Work Mr. Roberts Thesis of Georgia l

Mr. Rat >ert s Through Univ.

l l

of Georgia

+

University of Investigation of Oregen Basalts in IhAA of Selected Geological Basalts University of Oregon

45. G.G. Coles, Mr. toeschke Tubigen delation to Their Tectonic Setting of the Pacific Northwest (Germany)

Through Univ.

of Oregon

46. V. Tromsaor f, Geology, Univ. Regional & Contact Metamorphism, Petrogenesis of Metamorphic Rocks University of Oregon B. Evans, of Oregon Swiss Alps l In the Swiss Alps G.G. Goles Through Univ.

of Washington

~

47. G.G. Goles, Geology, INAA of Metavolcanics From Penn. To Determine Meta Samatic Changes University of Oregon Mr. Crawford University of Between Airphttolite & Granalite l Oregon Facres using Trace & Rare f arths G.G. Goles, Geology, Univ. 1 Anorthosites of the Adirondachs INAA of Se'.ec ted Geological Samples University of Oregon N 43.

5. Goldberg of Oregon i Thesis (Ph.0) Fron Ajiron u hs, New York -

0 .-- - . - - ._- y . - -- -

LTl 49. G.G. Goles, Geology, Univ. I Geochemistry and Petrogenesis INAA of SelK ted Roth Samples for University of Oregon [

y, C. Bow of Oregon '

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IV-15 C. PUBLICATIONS RESULTING FROM OSTR OPERATIONS THAT WERE REPORTED

'TO .HE RADIATION CEI

1. Publications ir4 erint Ma M.-S. and R.A. Schmitt 1979. Genesis of the cumulate eucrites Serra de Mage and Moore County. Meteoritics, 14, 81-90.

Ma M.-S. and R.A. Schmitt 1979. A chemical review of the various types of lunar mare low K basalts. In Lunar and Planetary Science X, Lunar and Planetary Science Institute, Houston, Texas, pp. 753-755.

Wentworth S. , G.J. Taylor, R.D. Warner, K. Keil , M.-S. Ma and R.A. Schnitt 1979. The unique nature of Apollo 17 VLT mare basalts. In Lunar and Planetary Science X, Lunar and Planetary Science Institute, Houston, Texas, pp. 1332-1334.

Ma M.-S. and R.A. Schmitt 1979. Chemistry of a new type of Apollo 11 low-K mare basalt. In Lunar and Planetary Science X_, Lunar and Planetary Science Institute, Houston, Texas, pp. 759-861.

Ma M.-S. , R.A. Schmitt, R.D. Warner, G.J. Taylor and K. Keil 1979. Composition, petrography, and genesis of Apollo 17 high-Ti mare basalts. In Lunar and Planetary Science X, Lunar and Planetary Science Institute, Houston, TE as, pp. 765-767.

Ha M.-S. , R. A. Schmitt, R.L. Niel sen, R.D. Warner, G.J. Taylor, and K. Keil 1979. Luna 16 basalts and breccias: new chemical and petrologic data. In Lunar and Planetary Science X, Lunar and Planetary Science Institute, Houston, Texas, pp. 762-764.

Ma M.-S. and R.A. Schmitt 1979. Chemistry of lithic frag-ments from the Apollo 17 drill core sections 70004 and 70005.

In Lunar and Planetary Science X, Lunar and Planetary Science Institute, Houston, Texas, pp. 756-758.

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

Refractory and siderophile element variations among chon-drules: evidence for primary compositional differences.

In Lunar and Planetary Science X, Lunar and Planetary Science Institute, Houston, Texas, pp. 443-445.

Scherpelz R.I. and S.E. Binney 1978. "A Review of the Delayed Fission Neutron Technique," Nuclear Instruments and Methods, 154, 413 (1978).

2058 006

IV-16

2. Publications in Press Ma M.-S. , R.A. Schmitt, R.L. Nielsen, G.J. Taylor, and R.D. Warner 1979. Petrogenesis of Lunar 16 aluminous mare basalts. J. Geophys. Res. Letts. (submitted).

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 data, magma types, and petrogenesis. Proc. Lunar and Planet. Sci. Conf.10th, Geochim. Cosmochim. Acta.

Suppl .10 (in press).

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 ana Planet. Sci . Conf.10th Geochim. Cosmochim. Acta Suppl .10 (in press).

Fodor R.V. , J.L. Berkley, K. Keil , J.'l. Husler, M.-S . Ma and R.A. Schmitt 1979. Petrology and chemistry of basalt drilled from the Galapagos spreading center, DSDP Leg 54.

Initial Reports of the Deep Sea Drillin Washington (U.S. Geol. Printing Office)g Project V., (in press).

Curtis D.B. and R.A. Schmitt 1979. The petrogenesis of L-6 chondrites-insights from the chemistry of mir.e als.

Geochim. Cosmochim. Acta (in press).

Tinsley I .J. and R.R. Lowry 1979. " Bromine Content of Lipids of Marine Organisms," Journal of American Oil Chemists Society (in press).

3. Reports and Papers J.P. Barton 1978. " Feasibility of Neutron Radiography for Large Bundles of Fast Reactor Fuel," IRT Corporation Final Report #6247-004, IRT Corp., Box 80817, San Diego, California, 92138.

A.G. Johnson and R. Poeton 1979. "The Influence of Cadmium Ratio on Thermal Neutron Measurements with TLD-600 Lithium Fluoride Thermoluminescent Dosimeters," Radiation Center Project Report, Oregon State University.

G.G. Goles (Dept. of Geclogy, University of Oregon) 1978.

Final Report of subcontract SA-911 of Prime Contract No.

EY-77-C-06-1030 for: Rockwell International Corporation, Atomics International Division, Rockwell Hanford Operations, Richland, Washington 99352.

2058 007

IV-17 NOTE: This report by Dr. Goles may ee of interest to those that may be (or are) involved in under-ground high-level waste storage. His findings suggest sub-surface basalts, under and near the Hanford reservation, probably were affected by alternation processes. This condition may strongly suggest that the storage of nuclear wastes in deep chambers, hollowed out within basalts, may not be a simple matter, and may be undesirable as contaminants could conceivably be brought back to or near the surface environ-ment.

D. COMMERCIAL OR N0fi-ACADEMIC UTILIZATI0fl None E. PUBLIC RELATI0flS The continued interest of the general public in the TRIGA reactor is evident in the number of people who have toured the facility. In addition to several hundred visitors during university open house events and interes ed individuals who happened to be in the vicinity, an estimated 1,095 people were given pre-planned and sche iuled tours this fiscal year.

See Table IV-7 for scheduled visitor statistics.

F. PLANNED CHANGES IN UTILIZATION One shift (40 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months per week) operation is the current OSTR utilization mode. This can and will increase, however, dep?nding u;>on the demand for OSTR time.

2058 008

IV-18 Table IV-7 RADIATION CENTER SCHEDULED VISITORS July 1,1978 - June 30,1979 N0. OF DATE NAME VISITORS 10-06-78 Willamina High School 15 10-12-78 American Nuclear Society 38 11-21-78 American Society for Metals 17 11-30-78 Oakridge High School 7 12-02-78 American Nuclear Society Student Chapter 11 12-18-78 NET Workshop 10 1-27-79 Beaver Open Fouse 200 1-29-79 NE 102x 20 2-02-79 Oregon House Committee on Energy & Environment 12 3-05-79 Cub Scouts 15 3-19-79 NET Workshop 10 3-28-79 Japanese Peace Rally Team 20 4-04-79 Youth Energy Awareness Day 180 4-05-79 LBCC 15 4-10-79 ESPREE Day 80 4-12-79 OSU - English Language Institute 10 4-18-79 Salem Academy 25 4-20-79 Vernonia High School 10 4-23-79 LBCC 21 4-24-79 LBCC 20 4-25-79 LBCC 20 4-27-79 Oregon Elementary School Teachers 30 5-04-79 Mom's Weekend 70 5-08-79 OSU - Chemical Engineering 351 82 5-10-79 OSU - Industrial Arts 12 5-11-79 Hidden Valley High School 20 5-11-79 Grants Pass High School 20 5-15-79 OSU - English Language Institute 12 5-18-79 OSU - Chemistry 206 10 5-18-79 Monroe Union High School 15 5-22-79 Chemeketa 6 5-24-79 Gardner Junior High School 30 5-30-79 Western View Junior High School 25 6-26-79 Pacific Power & Light Company 7 T0TAL 1,095 2058 009

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V-i TABLE OF CONTENTS Page V. ENVIRONMENTAL AND RADIATION PROTECTION DATA)

JULY 1, 1978 - JUNE 30, 1979 . . . . . . . . . . . - 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 OF THE LICENSEE AS MEASURED AT OR PRIOR TO THE POINT OF SUCH RELEASE OR DISCMARGE . . . . . , , , ,,,.

V-1

1. Liquid Waste (summarized on a monthly basis) . . . . . V-1 (a) The radioactivity discharged during the reporting period ............... V-1 (1) The total estimated quantity of radio-activity releanei (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) above for the reporting period, based on represatative 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 during each period of release . .............. V-2 2058 011

V-ii Pige.

2. Gaseous Waste (summarized on a monthly basis) . . . . V-2 (a) The radioactivity discharged during the reporting period . . . . . . . . . . . . . . . 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 quant' 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 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 20% of the applicable MPC . . . . . . . . V-4 2058 012

V-iii Page

3. Solid Maste (summarized on an annual basis). . . . V-4 (a) The radioactivity discharged during the reporting period .......... ... 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 shi pped off-site) . . . . . . . . . . . . . V-4 C. Att 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. Att ANNUAL S'JMMARY 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 cat.CULATED 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 Effluents ................... V-10 F. AN ANNUA'

SUMMARY

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

V-16

1. The Onsite Environmental Monitoring Systems . ..

V-16

2. The Offsite Environmental Monitoring Systems ..

V-21 2058 00

V-iv Page LIST OF TABLES AND FIGURES Table or Figure # Title Ta bl e V-1 Monthly Sumary of Liquid Waste Discharges for the Year July 1,1978 through June 30, 1979 . . . . . . . . . . . . ......... V-35 Table V-2 Monthly Sumary of Gaseous Waste Discharges for the Year July 1,1978 through June 30, 1979 . ....... ......... .... V-36 Tabl e V-3 Annual Sumary of Solid Waste Discharges for the Year July 1,1978 through June 30, 1979 V-37 Tabl e V-4 Annual S~mnary of .,adiation Exposure Received by Facility Personnel and Visitors for the Year July 1,1978 through June 30, 1979 ... V-38 Ta bl e V-5 Annual Summary of Radiation Levels and Con-tamination Levels Observed During Routine Radiation Surveys for the Year July 1,1978 through June 30, 1979 .. .. ....... V-39 Fig. V-1 Operating-Ar ea Film Badge Monitor Locations for the TRIGA Reactor . .. .. . V-40 Fig. V-2 Area Radiation Monitor Locations for the TRIGA and AGN Reactors, and the TP, IGA Reactor Area Fence . . . . . . . . . . . . . V-41 Table V-6 Total Dose Equivalent Recorded on Operating-Area Film Badge Monitors Located Inside the TRIGA Reactor Facility for the Yeer July 1, 1978 through June 30, 1979. ......... V-42 Table V-7 Total Dose Equivalent at the TRIGA Reactor Area Fence for the Year July 1,1978 through June 30, 1979 ............ V-43 Fig. V-3 Monitoring Stations for the OSU TRIGA Reactor, January 6,1976 through June 30, 1979 .... V-44 lable V-8 Annual Average Concentrations of Gross Beta Radioactivity for Offsite Environmental Soil, Water, M Vegetation Samples for the Year July 1, s/P through June 30, 1979 . . . . . . V-45 Ta bl e V-9 Annual I for Offsite Airborne Gamma Monitoring Stations for the fear July 1,1978 through June 30, 1979 ............ V-48 2058 014

V-1 V. ENVIRONMENTAL AND RADIATION PROTECTION DATA.; JULY 1, 1978 THROUGH JUNE 30, 1979 A. INTRODUCTI0fl The data contained in this section have been prepared to comply with the requirements of Nuclear Regulatory Coamission (NRC) Facility License No. R-106 (Docket No. 50-243) and the Technical Specifications contained in Appendix A to that license, dated July 21, 1976. The material has also been prepared in com-pliance 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 the unrestricted environment and all occupational exposures to radiation and radioactive materials are consis-tently maintained "as low as reasonably achievable."

B. A SUf1 MARY OF THE NATURE AND A!!0VNT OF RADI0 ACTIVE EFFLUENTS RELEASED OR DISCHARGED TO THE ENVIRONS BEYOND THE EFFECTIVE CONTROL OF THE LICENSEE ..S 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 contained in Table V-1.

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

2'058 015

V-2 (2) The detectable radiencclides pre:

this waste.

(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.

(4) A summary of the total release (in curies) for each radionuclide determined ir. (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 fraction of tle 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 Tabl e V-2.

2058 016

V-3 (1) The total estimated quantity of radioactivity released (in curies) determined by an approp-riate sampling and counting method.

(2) The detectable radionuclides present in this

~

waste.

(3) The total estimated quantity of Argon-41 re-leased (in curies) during the reporting period based on data from an appropriate monitoring system.

(4) The estimated average atmospheric diluted concentration of Argon-41 released during the reporting period (in terms o f microcuries/

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

(5) The total estimated quantity of radioactivity 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) lhe average concentration of radioactive par-ticulates with half-lives greater than eight days (in microcuries/ cubic centimeter) re-leased during the reporting period.

2058 017

V-4 (7) An estimate of the average concentration of other sign" icant radionuclides present in the gaseous waste discharge (in terms of microcuries/ cubic centimeter) and the fraction of the applicable MPC valu' 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 sumarized according to the following items. All solid waste data pertaining to these items are contained in Tabl e V-3.

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

(2) T!.e 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.

C. All AtlNUAL Suf1 MARY OF THE RADIATION EXPOSURE RECEIVED BY FACILITY PERS0flNEL AND BY VISITORS Ill TERMS OF THE AVERAGE RADIATION EX-

' POSURE PER INDIVIOUAL Ai!D THE GREATEST EXPOSURE PER 1NDIVIDUAL FOR EACH OF THE TU0 GROUPS The annual summary of the radiation exposure received by facility personnel and visitors for the applicable reporting period is contained in Table V-4.

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V-5 D. AN ANNUAL

SUMMARY

OF THE RADIATION LEVELS AND THE LEVELS OF CONTAMINATION OBSERVED DURING ROUTINE SURVEYS PERFORMED AT TdE 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 DIRECT RADIATION LEVEL IN UNRESTRICTED AREAS DUE TO DIRECT RADIATION FROM THE FACILITY, AND 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 caiculated) direct radiation level in the 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).

The demineralizer tank was removed from its original position as indicated in Figure V-1 to location "A" in Figure V-1 on January 3,1972, 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) 2058 019

V-6 of the subcritical pile and the utire 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 Lnrestricted areas from the subcritical pile and the demineralizer tank to essentially zero millirem per year.

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 dcmineralizer platform (see Figure V-1), and 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 to the enrestricted area 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 facility as a whole, and the need for relatively frequent access into reactor experimental and irradiation facilities. Both of these latter activities create a small potential for low levels of direct radiation exposure of reactor facility origin in immediately adjacent unrestricted areas.

2058 020

V-7 Direct radiation levels in unrestricted areas (which potentially may arise from the TRIGA facility) are eval-uated on the basis of three different types of radiation measurements. First, direct radiation levels are measured and analyzed as part of our routine radiation survey program.

Second, data from area monitoring film badges installed at strategic locations within the TRIGA reactor operating-area are routinely documented and utilized to indicate locations where direct radiation from the facility might be entering unrestrit;ed areas. The film badge data are corrected, as appropriate, to reflect radiation attenuation in the reactor facility walls. Finally, assessment of the direct radiation levels in unrestricted areas is conducted on the basis of area monitoring data collected through our thermoluminescent dosimetry (TLD) program.

The locations of pertinent vendor supplied (the vendor being Radiation Detection Co. [R.D. Co.], Mt. 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., CaSO4 TLD 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 2058 021

V-8 (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 i972. Figure V-2 also shows the location of thi- - " 0. 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 unrestricted areas. Nevertheless, we believe the data from inside the facility reflect tha 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 position on 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 M_onitor Radiation Center _T_RIGA, S_outh 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).

2058 022

V-9 After the addition of the previously described shielding and the reactor area fence, direct radiation levels in unre-stricted areas due to the TRIGA facility dropped to essentially background levels. Data presented i:. Table V-6 show the ex-tremely low annual doses recorded inside the reactor facility's operating-area. Table V-7 provides verification of the essentially normal annual radiation levels measured in the unrestricted areas, as recorded by the area monitoring TLD's located on the reactor area fence, and as indicated by the direct microroentgen per hour exposures rates taken at each area n:onitoring station. See footnote (7) of Table V-7 for a further explanation of the pR/hr data and its application.

We are continuing our efforts to achieve closer agree-ment between R.D. Co. and OSU TLD data. At the present time, R.D. Co. uses a considerably higher background than OSU, and they may be reporting 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 isolate and eliminate suspected sources of vendor error, should they exist.

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.

2058 023

V-10

2. The Maximum Direct Radiation Level in Unrestrictad 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 reviewed 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 normally present a significant direct radiation potential, particularly when the radionuclide composition of the radioactivity is examined. In our particular operation, the majority of the liqaid radioactive effluent is now normally associated with a single annual domineralizer resin change. When released from the reactor facility, this liquid is mixed with up to 3000 gallons of waste water from the Radiation Center laboratories on a batch basis before final discharge into the unrestricted area (the sanitary sewer system).

The annual average concentration for total reactor facility radioactivity in liquid effluents entering the unrestricted area equaled 2.32 x 10-6 Ci/cc for the year July 1,1978 through June 30, 1979. With respect to this value and the total radioactivity released in the liquid effluent, recall that no city water background radioactivity 9

2058 024

V-ll has been subtracted. Also, note that the main contributor to the microcuries released is tritium (m 22.4 pCi). Even though nearly all of the liquid effluent volume from the reactor facility now originates during the changing of de-mineralizer resins, which occurs about everj 12 months, .

appears that little, if any, of the tritium is of reactor ori gin . Our routine analysis of city water indicates a normal tritium background concentration within a range of 3.42 x 10-5 Ci/cc to 1.53 x 10 -0 pCi/cc for the year July 1, 1978 through June 30, 1979. Our annual average concentration for tritium based on all liquids released to the unrestricted area fron, the reactor facility is within this background range at 2.26 x 10-6 Ci/cc for the year July 1,1978 through June 30, 1979. If the tritium is omitted from the total radioactivity released in the reactor's liquid effluent and a calculation performed using the remaining radioactivity (N 0.55pCi),some of which is also city water background, the annual average concentration for reactor facility radioactivity entering the unrestricted area becomes 5.56 x 10-8 Ci/cc for the year July 1, 1978 through June 30, 1979.

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 (0.174%) of the appropriate unrestricted 2058 025

V-12 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 b" 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, 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- pCi/cc Argon-41 unre-stricted area maximum permissible concentration, or a value of 4 x 10 -6 Ci/cc. On page 4-58 (Table 4.11 of the Safety Analysis Report) it is concluded that under the most un-favorable atmospheric conditions (with the 55 foot stack)

. 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 for one 2058 026

V-13 year under the most unfavorable atmospheric conditions,

-ile 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 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 pCi/cc while operating at 1000 kW, the annual average Argon-41 concentration, as measured by the facility stack monitor, has always been much less than the assumed calculational value of 4 x 10-6 Ci/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 inches above ground level. As a result of the new stack height, new atmospheric dispersion calculations were necessary in order to evaluate the atn:ospheric dilution of gaseous effluents from the 2058 027

V-14 reactor facility. The results of the original 1968 calcu-lations and the first evaluation following the stack change were included in Table 2 of OSU's hay 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 USNRC 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.

Using the same basic assumptions employed for the shorter stack, and in particular a continuous Argon-41 discharge rate of 100 MPC, the 1973-7A 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 cor.-

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 1978 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 2058 028

V-15 the 197? 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 Jose 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 concentratinn is always much lower than the 4 x 10-6 Ci/cc value used for purposes of calculation.

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 pe' 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, (5.97 x 10-8 Ci/cc) as derived 2058 029

V-16 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 0.224 mrem for the year July 1,1978 through June 30, 1979.

F. Afl AfiflVAL SUttMARY OF THE GEf!ERAL METHODS Afl0 THE RESULTS OF EtlVIR0flMEf4TAL 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.

1. 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.

2058 030

V-17 The reactor facility radioactivity stack monitoring system consists of a continuous moving-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, remain. in operation at all times while the reactor is 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 integrating scaler is also attached to the gas channel to increase the accuracy of determining Argon-41 rel ea sed . 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 monitoring is the data derived from its operation which 205B OM

V-18 can be applied to determining potential exposures in unrestricted areas from gaseous radioactive effluents.

Onsite area monitoring film badges consist of standard personnel-type beta-gamma-neutron film packs, 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.

Onsite area monitoring using TLD's now consists of two different types of dosimeters, both located at 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 4 TLD's prepackaged by R.D. Co., and exchanged on 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 metal boxes previously used to house the film packs, and are accompanied at each lo-cation by the second TLD monitoring package which is prepared and interpreted by OSU. Each OSU TLD moni-toring device normally consists of three lithium fluoride chips, presently Harshaw TLD-700's, exchanged on a quarterly basis.

2058 032

V-19 Prior to April 1976, each onsite group of three OSU TLD chips was packaged first in a plastic mount which was then placed inside an outer container consisting of 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 to those presently being used by the Oregon Radiation Control Section in their TLD pecgram. In April 1976, 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 r, . ~ tor area fence monitoring st'tions. 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 ac each of the nine reactor area fence positions identified in Figure V-2.

In addition to the above monitoring devices, each of the nine reactor area fence monitoring positions is presently s. uipped with two 0-200 mrem gamma-sensitive integrating ionization chambers (self-reading pocket dosimeter type). These dosimeters are located inside the thin sheet metal box at each fence monitoring station, which also contains the two different TLD monitoring packets. The ionization chamber dosimeters are read every two weeks ar,d are used as backup monitors for each station.

2058 033

V-20 For the July 1,1978 through June 30, 1979 reporting period, an additional onsite environmental monitoring program was conducted. This program involved the bi-weekly (every 2 weeks) measurement of the direct radiation exposu.'e rate in terms of microroentgens per hour (pR/hr) at each reactor fence monitoring station. Measurements were taken with an Eberline Instrument Co. micro-R per hour rate meter containing a 1" x 1" NaI detector. The 26 annual readings were then averaged and ultimately converted to an expected (calculated) annual mrcm 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.

The routine analysis of gross radioactivity in liquid effluents (with isotopic identification as appropriate) prior to discharge into the unrestricted 4

2058 034

V-21 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 areas due to reactor operations include a soil, water, and vegetation monitoring program, and an airborne gamma monitoring program.

The soil, water, and vegetation mcqitoring 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 reactar 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 ai-borne gamma 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 ofisite airborne gamma monitoring stations were implemented to increase the total number of these stations now in use to nine-teen. See Figure V-3 for the location of the nineteen airborne gamma monitoring stations.

2058 035

V-22 As of January 1,1975, the coding technique used to designate each specific offsite monitoring station was modified slightly to inoicate 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 r ' "L" after the station number. For example, MRCTE-2L is interpreted as Monitor 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 ustJ on the reactor area fence). At offsite stations where only an OSU TLD monitor is used, the copper tube contain;ng 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-sist of a thin weather-tight aluminum box mounted on a post about four feet cff the ground. Each of these stations include one R.D. Co. TLD pack, one standard 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, 2058 036

V-23 the OSU TLD's are not enclosed inside the aluminum box, but instead the copper tube is taped directly onto the box mounting post at the station. All TLD monitors in the offsite airborne gamma environmental monitoring pro-gram are exchanged on a quarterly interval beginning January 1 of each year. The ionization chamber (dosi-meters) are read once every 2 weeks throughout the year.

For the July 1,1978 through June 30, 1979 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 cluivalent for each location based on an annual average pR/hr exposure rate.

A summary of the environmental monitoring results for the year July 1,1978 through June 30, 1979 is given below, and includes, as appropriate for the measurement under consideration:

(a) The number of sampling locations.

(b) The total number of s .aples per year.

2058 037

V-24 (c) The annual average concentration of gross aadioactivity, in some cases, concen-tratiev of soecific 4adionuclides in the mediu... ba' g assayed.

(o, The to.al 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 ar.d offsite programs.

Reactor Facility Stack flonitor, (onsite):

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

(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 monitor is given in Table V-2. As indi-cated in this table, the only gaseous component identified has been Argon-41, 2058 038

V-25 and only naturally occurring particulate 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, arid was within a range of 3.76 x 10-10 Ci/cc to 8.60 x 10-12 Ci/cc.

Reactor Facility Area Monitoring Film Badges, Reactor Fence TLD's. Integrating Ionization 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 CaSO 4

TLD 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 measurc;..ents are made. All of these have application as onsite environmental monitors.

2058 039

^

~

~

s .

V-26 (b) Since each film badge within the TRIGA facility is changed once per month, there is a total of 96 different samples 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) each year. There are a total of 26 pR/hr measurements made at each of the nine fence monitoring stations each year for a total of 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, fjorth 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 2058 040

V-27 Figure V-2. Fence monitoring locations are coded Monitor Radiation uenter 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 integrating 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 Contribt.ted 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 froia the collection point into the unrestricted area.

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

July 1,1978 through June 30,1979 = 1 reactor liquid effluent sample before release to the collection point. July 1,1978 through June 30, 1979 = 1 reactor liquid effluent sample before release from the collection point to the un-restricted area.

2058 00

V-28 (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 level of external radiation from radioactivity in liquid effluent.

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

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

(b) Samples (as available) are taken at each location on a quarterly basis. Samples have been collected as follows:

July 1,1978 through June 30,1979 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 location 'lW was dry on two sampling dates during the year July 1,1978 through June 30,1979.)

2058 042

V-29 I

(c) The annual average concentration of gross beta radioactivity for the offsite environmental soil, water and vegetation samples is given in Table V-8. Identification of specific radionucifdes 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 radus of the reactor building.

In general, samples are collected over a local area having a rddius 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 mrem gamma pocket dosimeters. Eight stations have only a standard OSU TLD moni-toring pack. In addition, each of the nine-teen monitoring stations is included in the ongoing program for measurement of the pR/hr exposure rate. '

20r8 3 v43

V-30 (b) The TLD's at each airborne gamma monitoring station are changed once every calendar quar-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 thips per station x 4 changes per year = 228 samples). The two backup Ponitors (integrating ionization chamber dosimeters) are read every two weeks, which results in approximately 572 individual dosimeter readings each year. There are a total of 26 pR/hr measurements made at each of the nineteen airborne gama monitoring stations each year for a total of 494 in-dividual measurements annually.

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

Like the soil, water, and vegetation sam-pling locations, most of the airborne gamma monitoring stations are within a 1000 foot radius of the reactor building. These lo-cations generally correspond to the atmo-sphere (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, the OSU TLD data, 2058 044

V-31 and results obtained from the pR/hr measurements. See footnote (8) of Table V-9 for a further explanation of the pR/hr data and its application.

This is the second complete year for the vendor supplied TLD monitors, which were substituted for the previous vendor supplied environmental film packs. As already indicated in the next to the last paragraph of section E-1 of this report, OSU is still somewhat re-luctant 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 our program 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.

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 background for the airborne gamma monitoring stations, and still do not believe reported values are always representative of what most stations are experiencing. We increased the number of background stations during 1976, and 2058 045

V-32 between July 1,1978 and June 30, 1379 we continued to make a series of d', rect background measurements with our pR/hr monitoring equipment (started July 1,1977) in order to obtain a better profile of the background variation. The results have increased our faith in our background values, but we plan to continue studying this variable.

From our viewpoint, the major purpose of the airborne gamma monitoring stations is to give an indi-cation of general increases or trends in unrestricted area radiation levels which might be linked to Argon-41 released from the OSU TRIGA. Past experience (over the last nine years) has shown that annual results per lo-cation vary slightly from year to year. Although the data have not been included in this report, by following t.he mrem per year history for a single station and com-paring the annual mrem total to the curies of Argon-41 emitted for the corresponding year, it becomes evident that there is no consistent pattern to the results, and that other factors must be responsible for the minor mrem per year variations. For example, such variations may be the result of small annual differences in cosmic or

- terrastrial background, fallout, etc. In any event, the small amount of Argon-41 released annually does not seem to be a significant factor in determining the total mrem per year reported for any particular monitoring ,

2058 046

V-33 station. A comparison of the data contained in Table V-9 to past results from these monitoring stations and to the values in footnote (8) of Table V-9 leads us to the conclusion that there has been no meaningful increase in the unrestricted area gamma radiation levels due to Argon-41 released by the OSU TRIGA during the defined reporting period.

2058 047

V-34 References

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

2. U.S. Environmental Protection Agency, " Estimates of Ionizing Radiation Doses in the United States, 1960-2000," ORP/CSD 72-1, Office of Radiation Programs, Rockville, Maryland (1972).

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

2058 048

Table V-1 f10NTHLV SumARY OF LIQUID WASTE DISCHARGES FOR THE YEAR JULT 1,1978 THROUGH JUNE 30, 1979 III Date of Total Quantity of Detectable Specific Activity for Olscharge Radioactivity Total Curies of Average Concentra- Percent of Total Radionuclides Each Radioactive Each Detectable l (month) Released (to in the Waste tien of Released Appi lcabl e Volume of ftaterial in Waste Radionucitoe Radioactive Material MPC for E f fluent, sanitary sewer) Discharge Where Released in the (curies) at Point of Release Released including Released Concentra- Waste (to sanitary sever) Radioactive Diluent tion After Dilution (curtes) (uCl/cc) Material Released was > 1.0 x 10'7 uCl/cc (%) (to sanitary (pC1/cc) sewar)

(gallons)

July-78 NONE NONE NOT APPtlCABLE NONE NOT APPLICABLE NOT APPLICABLE NONE August-78 NONE NONE NOT APPLICABLE NONE NOT APPLICABLE NOT APPLICABLE NONE September-78 NOM NONE NOT APPLICABLE NONE N)T APPLICABLE NOT APPLICABLE NONE October-78 NONE l NONE NOT APPLICABLE NONE NOT APPLICABLE NOT APPt! CABLE NONE November-78 NONE NONE NOT APPLICABLE NONE NOT APPLICABLE NOT ADPLICABLE NONE I

December-78 2.30 x 10' Cr -------------- 2.21 x 10'O 68 Mn -------.--.--- 3.58 x 10' 58 Co -------------- 4.87 x 10-8 2.32 x 10 -6 0.174 2620 60 Co - ------------ 4.48 x 10' 2.26 x 10 -6 3

H 2.24 x 10 January-79 NONE NONE NOT APPLICABLE NONE NOT APPLICARLE NOT APPLICABLE NONE February-79 NONE NONE NOT APPLICABLE NONE NOT APPLICABLE NOT APPLICABLE HONE March-79 NONE

{ NONE NOT APPLICABLE NONE NOT APPLICARLE NOT APPLICABLE NONE

{ April-79 NONE NONE NOT APPLICABLE NONE NOT AprLICABLE NOT APPLICABLE NONE

May-79 l NONE NONE l NOT APPLICABLE NONE NOT APPLICABLE NOT APPLICABLE NONE June-79 NONE NONE NOT APPLICABLE NONE NOT APPLICABLE NOT APPLICABLE NONE Annual Value 2.10 x 10 SEE AB0VE NOT APPLICABLE 2.30 x 10'$ 2.32 x 10 -6 0.174 2620 I2I III .

The OSU operational policy is to subtract only detector background from our water analysis dita and not background '

radioactivity in the Corvallis city water.

g N Total volume of ef fluent plus diluent does not take into consideration the additional mixing with arprontmately 95,000 to 115.000 g de gallons per year of ligelds and sewage normally discharged by the Radiatinn Center complex into the same sanitary sewer system.

LD CX)

CD 4

4

Table V-2 MONTHLV

SUMMARY

OF GASEOUS WA5TE DISCHARGES FOR THE VEAR JULY 1.1978 THROUGH JUNE 30. 1979 Total Total istimated Average Percent of the Total Estimated Average Estimated Average Percent of Estimated Estimated Atmospheric Appilcable Quantity of Concentration Concentration of MPC if the Date of Radioactivity Quantity of Diluted MPC for Diluted Radioactivity in of Radioactive Other Significant Estimated Discharge Rel ea sed Argon-4} Concentration of Concentration Particulate Form Particulates Radionuclides in Release was (Month) (Curles) ReleasedLI) Argon-41 at of Argon 41 at with Half- (fe Released With Discharge if 320% of the l

(Curles) Point of Release Point of Release >B DaysI 1 Hal f-L ife >8 Days >20% of the Appilcable (reactor stac6) (reacter stack) (Curles) (Curies) Appitcable MPC MPC l l (uC1/cc) (%) (uC1/cc)

July-78 1.57 l 1.57 1.03 x 10 2.58 NONE NOT APPLICABLE NOT APPt ! CABLE l NOT APPLICABLE August-7R 1.04 l 1.04 6.83 x 10' 1.71 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE ptember-78 1.53 l 1.53 1.04 x 10'I 2.60 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE October-78 0.52 l 0.52 3.44 x 10 0.86 NONE NOT APPLICABLE NOT APPLICARLE lNOTAPPLICABLE November-78 0.17 l 0.17 1.19 x 10- 0.30 NONE NOT APPLICABLE NOT APPLICABLE h07 APPLIJABLE Dec amber-78 0.90 ! 0.90 5.91 x 10~0 1.48 l NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE January-79 0.62 1 0.62 4.11 = 10'O 1.03 NONE NOT APPLICABLE NOT APPLICABLE l NOT APPLICABLE February-79 0.97 0.97 7.08 x 10' l.77 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE 0.29 -0 Ma rc h-79 l 0.29 1.96 x 10 0.49 NONE NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE April-79 1.89 1.89 1.29 x 10' 3.22 NONE NOT APPLICABLE NOT APPLICABLE NOT APPi! CABLE May-79 0.54 0.54 3.57 x IO " 0.89 I NONE l NOT APPLICABLE NOT APPLICABLE NOT APPLICABLE June 79 0.5R 0.58 3.92 x 10- l 0.98 NONE l NOT APPLICABLE NOT APPLICABLE NOT APPtlCABtf Annual

-8 Value 10.62 10.62 5.97 x 10 l 1.49 l NONE lNOTAPPLICABLE NOT APPLICABLE NOT APPLICABLE Routine gaarna 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. I N W m

C

LD CD CD Ln CD

V-37 Table V-3 ANNUAL

SUMMARY

OF SOLID WASTE DISCHARGES FOR THE YEAR JULY 1,1978 THROUGH JUNE 30, 1979 Total Amount Detectable Total Quantity Dates of Shipment of Solid Waste Radionuclides of Radioactivity and Disposition (l)

Packaged in the Waste in Solid Waste (cubic feet) (curies) 9.50 Cobalt-60 6.51 x 10-4 December 12, 1978 l

I ro..- 59 May 22, 1979 Sodium-24 Manganese-56 Chromium-51 Selenium-75 L

(I)All solid waste is transferred to our waste disposal service vendor, Nuclear Engineering Company, for burial at their installation at Richland, Washington. Transfer takes place at Oregon State University, Radiation Center, Corvallis, Oregon.

s 2058 051

V-38 Table V-4 AflNUAL

SUMMARY

OF RADIATION EXPOSURE RECEIVED BY FACILITY PERSONNEL AND VISITORS FOR THE YEAR JULY l, 1978 THROUGH JUNE 30, 1979 Average Annual Exposure Greatest Individual for Eacn Personnel Group ,

Exposure per Personnel Group Personnel Group Whole Body Extremities Whole Body Extremities (mrem) (mrem) (mrem) (mrem) 0perating faer 1 8.00 208.00 30.00 680.00 Re earc sonnel 0.00 82.00 0.00 450.00 Facility Visitors:

Film Badges 0.00 (1) 0.00 (1)

Pocket Dosimeters 1.00 (1) 25.00 (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, extremity dosimeters are not normally necessary and no visitor data are available for extremities.

2058 052

V-39 Table V-5 ANNUAL

SUMMARY

OF RADIATION LEVELS AND CONTAMINATION LEVELS OBSEF'IED DURING ROUTINE RADIATION SURVEYS FOR THE YEAR JULY 1,1978 THROUGH JUNE 30, 1979 Direct Radiation Levels Contamination Levels (mrem /hr)(Sy+ neutrons) (dpm/100 cm2 )(gy)(1)

Average Maximum Average Maximum Reactor Top < 1.00 100.00 < 370 < 370 Sample Handling Area < l.00 100.00 < 370 < 370 Reactor Room Floor < 1.00 50.00 1 370 1 370 Beam Port Facilities < 1.00 50.00 1 370 s 370 Outside Inside Outside Inside Demineralizer Tank Shield Shield Shield Shield Avg. Max. Avg. Max. Avg. Max. Avg. Max.

<1.00 10.00 5.00 75.00 (2) (2) < 370 < 370 Class Experiments < l.00 215.00 < 370 < 370 (1)No contamination was found at the designated locations during the entire reporting period. The 370 dpm/100 cm2 value used in this table is based on the normal beta efficiency and a net count rate equal to the normal background counting rate for the portable survey meters routir.ely used in the field to screer, for radioactive contamination (i.e., field measure-ments would normally have to show a gros'; counting rate equal to twice the normal background counting rate before contamination would be considered present). However, in addition to normal field screening, smears suspected of containing removable radioactive contamination are routinely counted in a more sensitive radiation detection system. Based on usual counting times, a normal instrument counting efficiency and a normal background counting rate, during the current reporting period such a detection system typically pro-vided a lower limit of detection (LLD) at 95% confidence of approximately 11-12 dpm for the radionuclides normally expected to be on the smears.

Smearing efficiency for radioactivity removal is conservatively assumed to be s10%, and positive smear results would usually be multiplied by 10 before final conversion to dpm/100 cm2 ,

(2)Not an applicable measurement.

2058 053

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2058 053

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'l l V-42 Ta bl e V-6 TOTAL DOSE EQUIVALENT RECORDED ON OPERATING-AREA FILn B' ,dE MONITORS LOCATED INSIDE THE TRIGA REACTOR FACILITY FOR ".hE YEAR JULY 1,1978 THROUGH JUNE 30, 1979 Total Recorded mRen for the Year July 1,1978 Location (l) Through June 30,1979(2)(5)

MRCTNE 0 MRCTSE 0(3)

MRCTSW 0 MRCTNW 15 MRCTWN 0 MRCTEN 0 MRCTHXS 0(4)

MRCTHXW 0 (1)These locations do not represent radiation exposure through an exterior wall directly into an unrestricted area.

(2) Totals do not include natural background contribution.

(3) Film badge opposite subcritical oile. Shield for subcritical pile completed February 23, 1972.

(4) Film badge opposite particulate filter for reactor primary water cleanup system. Shield around this filter was completed July 10, 1972.

(5)These area film badge monitors are excharged on a monthly interval, and the total mrem value reflects the summation of results from 12 different film packets.

2058 056

Table V-7 TOTAL DOSE EQilIVALENT AT THE TRIGA REACTOR AREA TENCE FOR THE VEAR JULY 1,1978 THROUGI JUNE 30, 1979 Tctel Calculated mrem for the year July 1,1978 Through June 30 Total Recorded mrem for the Year 1979 Based on the Annual Average July 1,1978 Through June 30 Total Recorded mrem for the Year July 1,1978 pR/hr Exposure Rate Measured LocationJ1] 1979 Based on R.D. Co. TLD's (8) Through _ June 30, 1979 Based on OSU TLD's_(3)(4)_(n at Each location _(4)L7) _

MRCTE-1 124.0 54.0 + 9.7(5) 72.0 + 11.0 71.0 ? 20.0 MRCFE-2 140.0 57.0

13.9 76.0 + 15.4 75.0 + 17.0 MRCFE-3 146.0 49.0 + 15.1 65.0 + 16.3 74.0 + 20.0 MRCFE 4 137.0 4 7.0 + 9.2 63.0 t 10.4 77.0

22.0 MRCFE-5 138.0 44.0 + 10.6 59.0 e 11.8 62.0 + 22.0 HRCTE-6 130.0 51.0 + 16.2 68.0 ! 17.9 63.0

28.0 MRCTE-7 124.0 53.0 + 13.2 71.0 + 14.6 68.0 + 23.0 MRCFE-8 126.0 60.0 < 8.7 80.0 + 9.R 63.0 ' 14.0 MetCIE-9 1 35.0 48.0 - 9.6 64.0 t 10.9 64.0 t 15.0 (II The TRIGA reactor area .ence was instilled September 15, 1972.

Radiation Detection Company (R.D. Co.), Mt. View. Califernia, TLD totals include an annual natural bac ground contribution of 115.0 mRam for the reporting period.

(3)05U fence TLD totals include a measured three calendar quarter natural background contribution of 52.0 + 13.1 mrem for data in the lef t coltunn (see footnote 5), and a calculated four calendar quarter annual natural background contribution of 70.0 + 14.9 pRem for data in the (qht column (see footnote. 5 and 6).

( I+ values represent the standard deviation it the 95% confidence level .

(6I Total mrem values shown in this column ar- based on data from only thret calendar quarters. Ducing the 1976-79 reporting perf od, one quarter's equipment data (the third quarter of 1978, which is the first quarter of the current rerorting period) were completely invalidated due to TLD malfunction.

(6) Total mrem values shown in this column are estimated annual (four calendar qtarter) values for each location. The estimated values were derived by summing the three quarters of measured data and a fourtn value obtained by averaging the three quarters of measured data. The N latter value (the average of the three measured quarters) was used to represent the best estimate of the missing quarter's data for the 1978-79 reporting period.

O W (7)The annual average microroentgen (pR) per hour exposure rate for each location was determined taken at 2 week intervals throughout the year.

by averaging 26 sepa rate uR/hr measurements y 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months per year and then converting microroentgens to millirem.The total mrem for the period was calculated by multiplying this average pR/hr value by range between about 7.0 and 11.0 pR/hr (Ref.1)(excluding areas of unusually high naturalThese radioactivity). Normal exposure rates */hr values for correspond I the U.S. (terrestrial to annual dose equivalent totals of about 59 to 91 mrem per year.

O cosmic dose equivalent for Oregon to be about 110 mrem per year. The U.S. EPA (Ref. 2,3) estimates the total annual terrestrial plus b W TLD monitoring packets are exchangad on a quarterly interval .

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i i V-45 Table V-8 ANNUAL AVERAGE CONCENTRATIONS OF GROSS BETA RADI0 ACTIVITY FOR OFFSITE ENVIRONMENTAL S0IL, WATER, AND VEGETATION SAMPLES FOR THE YEAR JULY l,1978 THROUGH JUNE 30, 1979 Sample Identification Number, Annual Average Concentration of Gross Type & Reporting Units Beta Radioactivity (minus 3H)(1)(2)(3) 1-water (pCi/cc) 5.01 x 10-8 2.28 x 10 -8 N 4-water (uci/cc) 1.12 x 10-7 4.17 x 10-8 ll-water (pCi/cc) -8 7.18 x 10 1.67 x 10-8 19-rainwater (pCi/cc) 6.27 x 10-8 1.75 x 10-8 5.43 x 10 - 1 2.33 x 10 -6 pC 3-soil ( gf ry soil) 5-soil (gram of ry soil) 4.44 x 10-5 2.00 x 10 -6 C -5 20-soil (gram o"f ry soil) 5.26 x 10 2.15 x 10-6

-5 21-soil (gram o"f ry soil) 6.90 x 10 2.87 x 10 -6 pC 2-grass (gy,,gf ry ash) 1.28 x 10 1 2.72 x 10

-5 6.-grass (g m of ry ash) 9.31 x 10 2.98 x 10 -6

-6 7-grass (gram o"f ry ash) 1.67 x 10-4 1 2.86 x 10 2058 059

V-46 Sample Identification Number, Annual Average Concentration of Gross Type & Reporting Units Beta Radioactivity (minus 3H)(1)(2)(3) 8-grass ( "i gf d d) 1.41 x 10- 12.75 x 10-6 "i -4 9-grass (g m of d ash) 1.20 x 10 1 2.34 x 10-6

" -4 10-grass (gram of d ash) 1.48 x 10 13.15 x 10-6

-0 12-grass (gram of d y ash) 1.49 x 10-4 12.76 x 10

" -4 13-grass (gram of d y ash) 1.26 x 10 1 2.92 x 10-6

" 9.59 x 10 -5 2.85 x 10 -6 14-grass (gram of d ash)

" 9.11 x 10 -5 2.38 x 10-6 15-grass (gnm of d ash) 16-grass (g am of d ash) 9.03 x 10-5 2.53 x 10-6

" -4 -6 17-grass (g m of d y ash) 1.46 x 10 1 2.88 x 10

" -6 18-grass (g m of d ash) 1.57 x 10- 12.96 x 10

" -4 -6 22-grass (g m of m) 1.76 x 10 13.29 x 10 (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 < the LLD were averaged in at the corresponding LLD concentration.

2058 060

V-47 (3)For this report, the lower limit of detection (LLD) has been defined as the smallest concentration of radioat.tive material in a sample that has c 95% probability of being detected (4.66 times the standard deviation of the average background value obtained with a blank sample, provided the relative standard deviation of the background [the co-efficient of variation) is less than 25%). For the year July 1,1978 through June 30, 1979, the LLD for gross B in water samples averaged 2.20 x 10-8 pCi/cc and ranged between 2.37 x 10-8 pCi/cc and 2.02 x 10-8 pCi/cc. For gross 8 in vegetation samples the LLD averaged 5.70 x 10-6 uCi/gm and ranged between 1.12 x 10-s pCi/gm and 8.97 x 10-6 pCi/gm.

For gross S in soil samples the LLD averaged 5.77 x 10 6 pCi/gm and ranged between 1.20 x 10-s pCi/gm and 6.13 x 10 6 pCi/gm.

(4)This sample location was dry for two calendar quarters during the reporting period (the third quarter 1978 and the second quarter 1979).

Therefore, no samples were collected for these intervals.

2058 061

Table V-9 ANNUAL TOTALS FOR OFFSITE AIRBORNE GAMMA MONITORING STATIONS FOR THE YEAR JULY 1,1978 THROUGH JUNE 30, 1979 Total Calculated mrem for the Year July 1, 1978 Through June 30, Total Recorded mrem for the Year Total Recorded mrem for the Year 1979 Based on the Annual Average, July 1,1978 Through June 30, 1979 July 1,1978 Through June 30, 1979 uR/hr Exposure Rate Measured at Monitorin1 5tation Based on R.D. Co. TLO's(1)(9) _BasedonStandardOSUTLD's(4)15M9{ Eachlocation_(Sl(8)

M'CTE-2L -----(2) 63.0 + 3.6 R4.0 + 4.I II 62.0 + 22.0 MRCTE-3 138.0(3) 65.0 +10.2 87.0 + 11.5 76.0 t 12.0 MRCTE-4 133.0 6 3.0 + 5.6 R4.0 + 6.5 70.0 + 15.0 MRCTE-5L -----

64.0 t 8.0 85.0 + 9.I 73.0 + 18.0 MRCTE-6 150.0 68.0 + 5.4 91.0 + 6.2 79.0 + 16.0 MRCTE-7L -----

75.0 +11.7 100.0 +13.2 75.0 + 16.0 MRCTE-8 1 38 .0 75.0

6.0 100.0 + 6.7 80.0

20.0 MRCTE-9 152.0 65.0 + 6.2 87.0 + 7.0 86.0 + 29.0 MRCTE-10 156.0 59.G + 6.4 79.0 t 7.3 66.0 + 16.0 MRCTE-11 119.0 52.0 + 6.1 69.0 + 7.1 60.0 + 16.0 MRCTE-12 150.0 60.0 + 4.7 80.0 + 5.4 86.0 t 19.0 MRCTE-13L ----- 60.0 t 8.1 80.0 t 8.9 70.0 + 17.0 MRCTE-14L ----- 63.0 t 4.4 84.0 + 5.0 75.0 + 21.0 MRCTE-15 140.0 52.0 *10.0 69.0 +11.3 82.0 t 15.0 MRCTE-16L ---.- 53.0 + 6.0 71.0 t 6.9 77.0 + 18.0 MRCTE-17 145.0 44.0 t 7.3 59.0 t 8.1 68.0 t 13.0 MRCTE-18L ----- 54.0 8.5 72.0 + 9.6 69.0 + 13.0 MRCTE-19 156.0 58.0 + 8.0 77.0 + 8.9 83.0

14.0 MpCTE-20L ----. 53.0 +11.8 71.0 13.6 68.0 + 20.0

(

Radiation Detection Company (R.D. Co.), Mt. flew. California, TLD totals f rclude an annual natural background contribution of 115.0 mrem for the reporting period.

Monitoring stations coded with an "L" contain one standard OSU TLD monitoring pack only. (No. R.D. Co. TLD pack.)

(

Monitoring stations nqi coded with an "L" contain nna R.D. Co. TLD monitoring pack, two 0-200 mrem garmna pocket dosimeters and one standard OSU TLD monitoring pack.

OSU offsite airborne gamma TLD totals include a measured three calendar quarter natural background contributir, of 45.0 t 11.1 mrem for data in the lef t column (see footnote 6), and a calculated four calendar quarter annual natural background contribution of 60.0 + 12. 5 mrem for data in the right column (see footnotes 6 and 7).

+ values represent the standard deviation at the 95". confidence level .

otal mrem values shown in this column are based on data from only three calendar quarters. During the 1978-79 reporting period, one quarter's data (the third quarter of 1978, which is the first quarter of the current reporting period) were completely invalidated due to TLD equipment malfunction.

III Total mrem values shown in this column are estimated annual (four calendar quarter) values for each location. The estimated values were derived by summing the three quarters of measured data and a fourth value obtained by averaging the three quarters of measured data. The latter value (the average of the three neasured quarters) was used to represent the best estimate of missing quarter's data g for the 1978-79 reporting period.

Q The annual average microroentgen (uR) per hour exposure rate for each location was determined by averaging 26 separate uR/hr measurements C taken at 2 week intervals throughout the year. The total mrem for the period was calculated by multiplying this average uR/hr value by C3 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months per year and then converting microroentgens to millirem. Normal uR/hr values for the U.S. (terrestrial plus cosmic radiation) range betwaen about 7.0 and 11.0 uR/hr (Ref. I)(excluding areas of unusually high natural radioactivity). These exposure rates correspond to annual dose equivalent totals of about 59 to 93 mrem per year. The U.S. EPA (Ref. 2,3) estimates the total annual terrestrial plus cosmic dose equivalent for Oregon to be about 110 mrem per year.

C g (9)TLD monitoring packets are exchanged on a quarter;y interval. 8 g

N

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ML19290A080

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