ML20005D603

Annual Rept of Oregon State Univ Radiation Ctr & Triga Reactor Jul 1988 - June 1989.

ML20005D603
Person / Time
Site:	Oregon State University
Issue date:	06/30/1989
From:	Anderson T, Dodd B, Higginbotham J Oregon State University, CORVALLIS, OR
To:
Shared Package
ML20005D602	List: ML20005D601 ML20005D603
References
NUDOCS 8909250208
Download: ML20005D603 (63)

Text

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Annual Report 4 of the 3.g Gregon State University C1 Radiation Center and TRIGA Reactor July 1, 1988 - June 30, 1989 To satisfy the requirements of:

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

Technical Specification 6.7(e).

B. Task Order No. 3. under Subcontract No. C84-110499 (DE-AC07-76ER01953) for University Reactor Fuel Assistance-AR-67-88, issued by EGAG Idaho, Inc.

C. Oregon Department of Energy, ODOE Rule No.30-010.

, Written by:

T. V. Anderson, Reactor Supervisor B. Dodd, Reacter Administrator J. F. Higginbotham, Senior Health Physicist D. S. Pratt, Health Physicist A. G. Johnson, Director Submitted by:

A. G. Johnson Director, Radiation Center Radiation Center Oregon State University Corvallis, Oregon 97331-5903 Telephone: (503)737-2341

( September 13, 1989 D $

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109 E. Summary of 0$TR Environmental and Radiation Protection Data L-Year July 1, 1988

1. Liouid Effluents Released (See Table V.B.1) Through June 30, 1989 ,

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Total releasedestimated quantity sewer)(in (to the sanitary of radioactivity curies )(1) 8.53 x 10-4 l

b. Detectable radionuclides in the liquid waste 3H , 60Co I

c. Estimated average concentration of released radioactive material at the point of release i (in microcuries per cubic centimeter) 3.00 x 10-5 I

d. Percent of applicable MPC for released liquid radioactive material at the point 1.01(3) of release (%) s 0.03%(4)

e. Total volume of liquid effluent released, ,

including diluent, which ,

contribution (in gallons)g gtained an OSTR 1 7542 i

O?.P (1) The OSU operational policy is to subtract only detector background from our water analysis data and not background radioactivity in the Corvallis city water.

o (2) Based on values listed in 10 CFR 20, Appendix B, Table 2 Column 2.

(3) Based on values listed in 10 CFR 20 Appendix B. Table 1. Column 2, applicable to sewer disposal.

(4) Total volume of effluent plus diluent does not take into consideration s the additional mixing with the over 7,500,000 gallons per year of liquids and sewage normally discharged by the Radiation Center complex

} into the same sanitary sewer system.

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I fO Year July 1,1986 1 lif # 2. Airborne Effluents Released (See Table V.B.2) Through June 30, 1989

a. Total estimated quantity of radioactivity released (in curies) 6.3

b. Detec;able radionuclides in the gaseous wastell) 41A r (Tg = 1.83 hr)

c. Estimated average atmospheric diluted concentration of argon-41 at the point of release (in microcuries per cubic r centimeter) 4.0 x'10-8 .

d. Percent of applicable MPC for diluted '

concentration of argon-41 at the point '

of release (%) 1.0

e. Total estimated release of radioactivity in particulate form with half-l greater than 8 days (in curies)jygs U1 None Year July 1,1988 ,

3. Solid Waste Released (See Table V.B.3) Through June 30, 1989 '

a. Total amount of solid waste packaged and dieposed of (in cubic feet) 21.0

b. Detectable radionuclides in the solid weste 3H , 46se, 51C r, SaMn, 6BCo,She 60Co 75$ ,, 82Br.

124sb, 132T e, 181C e.

144Ce, !$2gy, 154Eu

c. Total radioactivity in the solid waste (incuries) 5.2 x 10-5 e

(1) Routine gansna spectroscopy analysis of the gaseous radioactivity in the stack discharge indicated that it was virtually all argon-41.

(2) Evaluation of the detectable particulate radioactivity in the stack discharge confirmed its origin as naturally occurring radon daughter

[.. .; products, predominantly lead-214 and bismuth-214, which are not asso-( ,f ciated with reactor operations.

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4. Radiation Exposure Received.by Personne1 . Year July 1. 1988

s. (in mrem) (See Table V.C.1)ul Through June 30, 1989

a. Facility Operating Personnel

1) Average whole body 12

2) Average extremities 72

3) Maximum whole body 80

4) Maximum extremities 500

b. Key Facility Research Personnel

1) Average whole body 0

2) Average extremities 5 3 Maximum whole body 0 4 Maximum extremities 40

c. Physical Plant Maintenance Personnel l 1) Average whole body <1

2) Maximum whole body 8 i

d. 1.aboratory Class Students l

1) Average whole body 0

2) Average extremities 7

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') 3) Maximum whole body 0  ;

4) Maximum extremities 100

e. Campus Police and Security Personnel 1 Average whole body 0 2 Maximum extremities O

f. Visitors l

l 1) Average whole body <1

2) . Maximum whole body 5 l

(1) "0" indicates that each of the beta-gamma dosimeters during the report-7 ing period was less than the vendor's gama dose reporting threshold T. of 10 mrem or that each of the neutron dosimeters was less than the

_,' vendor's threshold of 30 mrem, as applicable.

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S. Number of Routine Onsite and Offsite Year July 1, 1988 Monitoring Measurements and Samples Through June 30. 1989

a. Facility Survey Data i 1) Area Radiation Dosimeters (See Table V.O.1)  !

a) Beta-gansna dosimeter measurements 136 l b) Neutron dosimeter measurements 48  ;

2) Radiation and Contamination Survey Measurements (See Table V.D.3) s6000

b. Environmental Survey Data

. 1) Gamma Radiation Monitoring (See Tables ,

V.E.1andV.E.2)

• i a) Onsite monitoring ,

-- OSU TLD monitors 108 ,

-- Radiation Detection Co. TLD monitors 72

-- Monthly pR/hr measurements 108 b) Offsite monitoring

-- OSU TLD monitors 264

-- Radiation Detection Co. TLD monitors 104 C'j . -- Monthly pR/hr measurements 252 i- 2) Soil, Water and Vegetation Surveys I

(See Table V.E.3) ,

a) Soil samples 16 b) Water samples 16 c) Vegetation samples 56 I

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)) PART IV REACTOR  ;

A. Operatine Statistics l

For the current reporting period, the operating statistics for the [

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OSTR showed modest increases in most of the major categories when compared to the previous period. Operating data by individual category are given in Table IV.A.1 and in Figure IV.A.1. Table IV.A.2 is included for reference and summarizes the operating statistics for the original ,

20% enriched fuel.

The themal energy generated in the reactor during this reporting period was 42.7 megawatt days (MWD). The cumulative themal energy generated by the FLIP core now totals 444.2 MWD from August 1, 1976 ,

through June 30, 1989. Reactor use time averaged approximately 90%

of the normal ninehour, five-day per week schedule. Tables IV.A.3

.- through IV.A.6 detail the operating statistics applicable to this

[. reporting period.

Excess reactivity increased approximately 18t during the current re-porting period. This change was caused by three factors:

1

1. Consumption of the erbium burnable poison in the fuel (increased reactivity).

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2. Fuel element shuffles to even out core burnup (increased reactivity).

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3. Fuel burnup (decreased reactivity).

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< ;f .y Table IV.A.1 OSTR Operating Statistics (Using the FLIP Fuel Core) for the 10-Year Period August 1976 - June 1986 1 JUL 78 1 JUL 79 1 JUL 80 1 JUL 81 1 JUL 82 1 JUL 83 1 JUL 84 1 M 85 Operational Data 1 AUG 76 1 JUL 77 Through Through Through Through Through Threegh Through for Through Through Through 30 JUN 79 30 JUN 80 30 JUN 81 30 JUN 82 30 JUN 83 30 JUN 84 30 m 85 3r M 86 FLIP Core 30 JUN 77(1) 30 JUN 78 1255 1192 1095 1205 1206 1479 Operating Hours 875 819 458 875 (critical) 1005 999 931 943 946 1942 Megawatt Hours 451 4% 255 571 2a.5 10.6 23.8 41.9 41.6 38.8 39.3 39.4 43.4 Megawatt Days 19 25.9 13.4 29.8 52.5 52.4 48.6 49.3 49.5 54.5 Grac 235U Used 24 552 998 973 890 929 904 1924 Hours at Full 401 481 218 Power (1 MW) 0 0 1 0 0 0 0 Number of Fuel 85 0 2 Elements Added cr Removed (-)

372 348 408 396 469 407 403 Number of Irradi- 44 375 329 ction Requests (1) The reactor was shutdown on July 26, 1976 for one month in order to completely refuel the reacter with a new FLIP fuel core.

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, _s Table IV.A.1 (continued)

OSTR Operating Statistics (Using the FLIP Fuel Core) for the Period July 1986 - June 1989 Operational Data 1 JUL 86 1 JUL 87 1 JUL 88 1 JUL 89 1 JUL 90 1 JUL 91 1 JUL 92 1 JUL 93 I JUL 94 1 M 95 for Through Through Through Through Through Throegn Through Threegh Through Through FLIP Core 30 JUN 87 30 JUN 88 30 JUN 89 30 JUN 90 30 JUN 91 30 JUN 92 30 JUN 93 30 JUN M 30 JUN 95 30 M 95 Operating Hours 1172 1352 1170 (critical)

Megawatt Hours 993 1001 1025 Megawatt Days 41.4 41.7 42.7 Grams 235U Used 51.9 52.3 53.6 Hours at Full 980 987 1021 Power (1 MW)

Number of Fuel OIII -2(2) O Elements Added or Removed (-)

Number of Irradi- 387 373 290 stion Requests (1) No fuel elements were added, but one fueled follower control rod was replaced.

(2) Two fuel elements were removed due to cladding deformation. .

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Table IV.A.2 OSTR Operating Statistics with the Original (205 Enriched) Standard TRIGA Fuel Core 1 APR 72 1 APR 73 1 APR 74 1 APR 75 1 APR 76 Oper:tional Data ' 8 MR 67 1 JUL 68 1 JUL 69 1 APR 70 1 APR 71 Through Through Through Through Through Through TetAL: M 67 Through Through Through Through 31 MR 76 26 JUL 76 Through AL 76 far 20% Enriched 31 MR 70 31 MR 71 31 MR 72 31 MR 73 31 MR 74 31 MR 75 Core 30 JUN 68 30 JUN 69 (4)

(1) (2) (3)

Operating Hours 563 794 353 6983 567 855 598 954 705 (critical) 904 610 1 335.94 321.45 408 213 2553 117.24 102.47 138.05 223.77 195.11 497.82 Megawatt Hours 13.99 13.39 17 9 106.4 4.88 4.27 5.75 9.3 8.1 20.74 Megawatt Days 16.81 21.35 10.7 133 7.21 11.7 10.2 26.031 17.57 Grams 235U Used 6.13 5.36 Hours ct Full -- -- -- -- -- -- 096 429 369 58 --

Power (250 kW)

Hours at Full 291 460 205 1790

-- -- 20 23 100 401 200 Power (1 MW)

Number of Fuel 2 2 2 0 94 1 1 Elements Added 70 2 13 1 to Cere (Initial)

Number of Irradt- 396 357 217 4800 391 528 347 550 452 ation Requests 429 433 183 43 39 1940 102 99 249 109 Number of Pulses 202 236 299 (3) heector shut down June 1.1971 for one month for coelfag system (1) Reactor went critical on March 8.1%7 (70 element core; 250 kW). upgradfag.

Note: This period length is 1.33 years as initial criticaf f ty occterred in March of 1%7.

(4) Reector shut down July 26. 1975 for one month for refueltag (2) Reactor shut down August 22.1%9 for one month for upgrading to reactor with a new fell FLIP fuel core. Note: This pertes 1 MW (did not upgrade cooling system). Note: This period length length is 0.33 years.

is only 0.75 years as there was a change in the reporting period from July-June to April-March.

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IV-5 Table IV.A.3 Present OSTR Operating Statistics Annual Values Cumulative Values Operational Data for for for 1 JUL 88 1 AUG 76 FLIP Core Through Through 30 JUN 89 30 JUN 89

1. MWH of energy produced 1,025 10,658

2. MWD of energy produced 42.7 444.2

3. Grams 2350 used 53.6 _ 557.7

4. Number of fuel elements added to or removed from

(-) the core 0 85+3FFCR(1)

5. Number of pulses 19 1,148

6. Hours reactor critical 1,170 14,146

7. Hours at full power (1 MW) 1,021 10,358

8. Number of startup and  !

shutdown checks 252 ,3,265

9. Number of irradiatjgg requests processed W1 290 5,000

10. Number of samples !rradiated 3,177 59,115 (1) Fuel Follower Control Rod, l (2) Each irradiation request could authorize from 1 to 120 samples.

The number of samples per irradiation request averaged 11.0 during 4 the current reporting period.  !

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ms' Table lY.A.4 ,

t OSTR Use Time in Terms of Operational Functions Annual Values Cumulative Values '

for for 1 JUL 88 1 AUG 76 Through . Through 30 JUN 89 30 JUN 89 OSTR Operational Function (hours) (hours)

Checkout, core excess -

and shutdown 377 4,802 Reactorinuse(l) 2,353 20.050 l Total reactor use time 2,730 24.852 s,.-

(1) This function includes preclude time, multiple reactor experiment time, and the time the reactor is in use for teaching but not neces-sarily operating. (Preclude time is the time the reactor is not available for regular use due to performance of surveillance and maintenance items, such as fuel element inspections, transient rod lubrication, control rod calibration, power calibration, as well as sample loading and unloading time.)

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Table IV.A.5 OSTR Use Time in Terms of Specific Use Categorie, Annual Values Cumulative Values l for for l OSTR Use Category 1 JUL 88 1 AUG 76 Through Through 30 JUN 89 30 JUN 89  !

(hours) (hours) -

Teaching (depqrtmental and others)Ui 260 2,994 OUJ research(2) 386 5.782 Off-campus research(2) 791 2.909 Forensic services 11 154(3)

Reactor preclude time 938 8.346 Facilitytime(4) 328 4,472 Visitordemonstration(5) 16 195(6)

Total reactor use time 2,730 24,852 i (1) See Tables Ill.A.2 and Ill.E.1 for teaching statistics.

(2) See Table Ill.A.3 for research statistics.

(3) Prior to the 1981-1982 reporting period, forensic services were grouped under another use category. Since then, this service has been a separate category and the cumulative hours have Seen compiled begin-ning with the 1981-1982 report.

(4) The time OSTR spent operating to meet NRC facility license requirements.

(5) 1his is the time that the reactor was used specific. ally for visitor open-house (demonstration) events. The remainder of the visitors viewed the reactor during times when the reactor was being operated ,

for regularly scheduled research and teaching. .

(6) An error in the preparation of the 1984-1985 report resulted in the reporting of 101 hours0.00117 days <br />0.0281 hours <br />1.669974e-4 weeks <br />3.84305e-5 months <br /> of OSTR operations for visitor demonstration

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while the actual value was 159 hours0.00184 days <br />0.0442 hours <br />2.628968e-4 weeks <br />6.04995e-5 months <br />. The difference of 58 ho0rs is added to this year's report to correct the total cumulative reacter use time value.

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@ Table IV.A.6 OSTR Multiple Use Time (l)

Annual Values Cumulative Values for for Number of Users 1 JUL 88 1 AUG 76 r Through Through  :

30 JUN 89 30 JUN 89 (hours) (hours)

Two 155 1,155 Three 78 278 Four 25 102 Five 6 10 Six '15 23

) Seven 4 4 Total multiple use time 283(2) 1,572(3)

(1) Multiple use time is that time when two or more irradiation requests are being concurrently fulfilled by operation of the reactor.

(2) This represents 24% of the total hours the reactor was critical during this reporting period.

(3) This represents 11% of the total hours the reactor was critical since startup with FLIP fuel in August of 1976.

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1976-77: 19.0 8E8

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1977 78: 20.6 fas O -

1978-79: 10. 6 8888 -

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1979-80: 23.8 181D -

2 20 -- 1980 81: 41.9 sesO -

1981-82: 41.6 ftID .

1982-83: 38.8 fem '

1983-84: 39.3 8410 ~

i 1984-85: 39.4 fem 1985-86: 43.4 see -

. - 1986-87: 41.4 pte -

. 1987-88: 41.7 POS -

. 1988-89: 42.7 ISS .

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76-77 78-79 80-81 82-83 84-85 88-87 88-80 90-91 77-78 79-80 81-82 83-84 85-88 87-88 89-90 Time (Annuoi Reporting Period) i I

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Figure IV.A.1 OSTR Annual Energy Production Vs. Tine (Annual Reporting Period) l. .

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i IV-10 t 8. Experiments Performed During the current reporting period there were 11 approved reactor .

experiments available for use in reactor related programs. The follow-  ;

ing list of reactor experiments identifies the 11 approved experiments. .

Missing numbers signify reactor experiments which are in the inactive ,

file and are not currently being used.

A-1 Normal TRIGA Operation (No Sample Irradiation).

B-3 Irradiation of Materials in the Standard OSTR 1rradiation Facilities. ,

B-11 Irradiation of Materials Involving Specifi: Quantities of Uranium and Thorium in the Standard OSTR 1rradiation Facilities.

B-12 Exploratory Experiments.

l B-21 Beam Port No. 3 Neutron Radiography Facility; Amendment No.1 to B-21; Neutron Holography, l

t l B-23 Studies Using TRIGA Thermal Column.

l General Neutron Radiography.

B-24 B-25 Neutron F1.ux Monitors.

B-29 Reactivity Worth of Feel.

B-30 NAA of Jet. Diesel, and Furnace Fuels.

B-31 TRIGA Flux Happing.

Of the approved experiments on the active list, five were used during the reporting period. A tabulation of information relating to reactor experiment use is given in Table IV.B.1, and includes a listing of the experiments which were used, how often each was used, and the general purpose of the use. Presently, 25 experiments are in the inactive file and could be reapproved for use if needed.

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IV.11 1

Table IV.B.1 Use of OSTR Reactor Experiments (l)

Reactor Experi et Number Zi Research Teaching Forensic Facil(i Time TOTAL A-1 0 33 0 81 114 B-3 132 17 2 0 151 B-11 20 0 0 0 20 B-23 0 2 0 0 2 B-31 2 0 0 0 2 TOTAL 154 52 2 81 289

') (1) This table displays the number of times reactor experiments were used for a particular purpose.

(2) The following tabulation gives the number of each reactor experiment used and its corresponding title:

A-1 Normal TRIGA Operation B-3 Irradiation of Materials in the Standard OSTR 1rradiation Facilities B-11 Irradiation of Materials Involving Specific Quantities of Uranium and Thorium in the Standard OSTR 1rradiation Facilities B-23 Studies Using TRIGA Thermal Column B-31 TRIGA Flux Mapping (3) The time OSTR spent operating to meet NRC facility license requirements.

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C. Unplanned shutdowns j

! There were ten unplanned reactor shutdowns (scrams) during the current i reporting period. Table IV.C.1 contains a sumary of the unplanned shutdowns including a brief description of the cause of each. l l

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IV.13 Table IV.C.) ,

Unplanned Reactor Shutdowns (Scrams) i Number of Type of Scram Occurrences Cause of Shutdown Safety Channel 6 Spurious scram signals. No cause or reason -

could be detemined at the time. (These did not involve actual overpower situations.)

Even though the safety channel ion chamber checked out good, it was decided to change the ion chamber. This action appears to have solved the problem.

% Power Channel 1 Spurious scram signal. Apparent noise in the mode switch caused this scram. Reactor was

• at 100 watts preparing for pulsing operation. ,

When the mode switch was turned, the scram occurred.

Manual 1 The linear channel (blue pen) on the console recorder was behaving abnormally. It was observed that the drive line from the bull wheel to the pen was frayed. The reactor was scrammed and the pen drive repaired.

Manual 1 The reactor top Continuous Air Monitor (CAM) ceased operation (low flow alarm) and the reactor was scrammed manually, it was deter-mined that the CAM blew a fuse. The fuse was replaced and reactor operation was resumed.

Manual 1 This scram was unintentional. The reactor i

operator reached for a writing pen and acci-dentally hit the manual scram button. Reactor operation was resumed.

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I IV-14

^ Changes to the OSTR Facility, to Reactor Procedures, and to Reactor

, t D.

Experiments, and Tests Performed Pursuant to 10 CFR 50.59 The information contained in this section of the report provides a summary of changes and tests performed during the reporting period under the provisions of 10 CFR 50.59. For each item listed, we have included a brief description of the action taken and a summary of the applicable safety evaluation. Although it may not be specifically stated in each of the following safety evaluations, all actions taken under 10 CFR 50.59 were implemented only after it was established by the OSTR Reactor Operations Committee (ROC) that the pruposed activity did not require a change in the facility's Technical Specifications and did not introduce or create an unreviewed safety question as defined in 10 CFR 50.59(a)(2).

1. 10 CFR 50.59 Changes to the Reactor Facility l

l There were eight changes to the reactor facility which were reviewed, approved, and performed under the provisions of 10 CFR 50.59 during l

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the reporting period,

a. INSTALLATION OF A CADMlUM-LINED, lheiME, IRRADIATION TVBE (1) Description The reactor operations staff built and installed a cadmium-lined irradiation tube which can be permanently positioned in the reactor core. As shown in Figure IV.D.1, the facility consists of an air-filled aluminum tube with an offset bend, which is inserted into a convenient B-ring core grid position. The cadmium is approximately 0.025 inches thick and is permanently encased in aluminum inside and out. The tube is positively secured near the top to the center channel which extends across the reactor tank, and has a cap on top to seal it during reactor operation. To eliminate any small pressure increases j due to radiolytic gas production, a suction is drawn

%' on the irradiation tube by connecting a line from the rotating rack vent system to a tee-section on the irradi-ation tube.

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,,) END PLUB AND LOCATING PIN Figure IV.D.1 Cadmium-Lined, In-Core Irradiation Tube

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) NOTE: OSTROP 10 was revised to detail the procedures for using the new irradiation facility, and experiment B-3 was modified to allow irradiations in the cadmium- l lined tube to be performed under this experiment. These  :

changes, which were also made under 10 CFR 50.59, are discussed later in section D.

(2) Safety Evaluation l

The safety considerations for this facility were very i similar to those evaluated for the cadmium-lined pneumatic transfer tube which was previously installed under ROC approval and then recently removed from the reactor core.

The reactivity changes were expected to be about the same, or slightly more negative because a similar amount of cadmium was used and this facility was placed in a

• core location with a slightly higher flux. The cadmium-

,_.., lined tube was therefore estimated to be worth about Y -$2.20 and was expected to reduce the core excess from about $6.50 to about $4.30. As indicated earlier, the tube was secured (bol^ted) to the reactor facility in such a manner that it could not be easily or uninten-

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tionally removed. This prevents any sudden, unplanned addition of reactivity to the reactor.

The tube was constructed in such a manner that the cadmium was completely sealed within an inner and outer aluminum tube. Therefore, there will be no potential for cadmium contamination of the samples or the reactor.

The offset bend in the tube is similar to that of the other in-core facilities and effectively precludes radia-tion streaming from the tube.

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IV-17 The procedures for the facility stipulate that the tube's cap is to be on the top of the tube except when samples are being inserted or removed and whenever the reactor is in operation. Additionally, under normal circumstances, samples will be removed from the tube only after the I reactor has been shut down for a time period adequate I to ensure that .the bulk of the argon-41 activity has decayed (usually overnight). This prevents e.ny unnecessary I release of argon-41 to the reactor bay. The tube could be unloaded after a shorter decay period with specific

$rL health physics approval and appropriate precautions, but this option is no different than that currently employed with the rotating rack.and therefore introduces no new safety considerations.

The procedures and limitations for encapsulation and irradiation of samples using the in-core cadmium-lined facility follow current requirements, particularly those in OSTROP 18. Therefore, no new or untried practices were introduced relative to the actual use of the new facility.

In order to position the cadmium-lined end of the tube into the core's B-ring, a fuel element must first be moved from grid location B1 to G6. The cadmium-lined I

tube can then be inserted into the vacant B-ring position. ,

The estimated reactivity effect at each stage of the move was calculated and is given below:

1. Removal of the element from B1 = -35t

11. Insertion of this element into C5 = +304 111. Insertion of the cadmium-lined tube into B1 = -$2.20 iv. Overall reactivity change = -$2.25

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IV-18 j Core excess measurements were made at each step of the tube insertion procedure outlined above. The control l rods were recalibrated after the tube was inserted and the core excess was remeasured.

b. CHANGES TO THE CADMIUM-LINED, IN-CORE, IRRADIATION TUBE (1) Description Following Reactor Operations Committee approval of the  ;

facility change regarding the cadmium-lined, in-core irradiation tube, the staff reconnended two changes related to the irradiation tube. The first change involved the method to be used for removal and insertion of the sample support seat and the outer container holding the encapsu-lated sample (s) to be irradiated. This change was based ,

on the fact that after the in-core irradiation facility .

was constructed it was discovered that the standard TRIGA

tube handling device used to insert and remove TRIGA tubes for the rotating rack would also easily go down the new irradiation tube provided the handling device's outside borated shield was removed. Therefore, instead of a previously proposed, less desirable method for sample handling it was now proposed that samples be placed in the cadmium-lined facility in aluminum TRIGA tubes modified to have internal threads so that the containers could be inserted and removed using the standard TRIGA tube handling device (fishing pole and grapple) and standard procedures developed for the rotating rack. Similarly, the top of the sample support seat was modified by adding an aluminum TRIGA tube cap so that this too could be easily inserted and removed by this same method.

The second change involved moving the location of the 3 facility's air suction tube to the irradiator tube cap,

) rather than having a "T" in the tube itself, it was determined that this change would allow much more flexi-bility in the angular positioning of the air .9be.

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) (2) Safety Evaluation It was judged that the first change improved the operational safety of the facility. The basis for this was related to the fact that the previously planned use of wire and cord to position and support samples was deleted, and with this deletion went any potential activation of wire, risk of wire breakage, and the risks associated with handling wire or cord when samples are pulled out of the tube. Conversely, the new method for sample insertion and removal employs currently approved procedures, which have been in practice at the OSTR for many years. The new design of the sample support seat also enables the staff to remove the seat easily when not in use, thus preventing any unnecessary activation.

There are no unfavorable safety implications involved with the second change. A slight air suction will still C. .,,a '

be pulled on the tube when it is in the core, in the same manner as before. Repositioning the suction tube on the cap simply allows this tube to be oriented in different directions and thus limite the inconvenience of having the tube in the way of other work which could be going on in the insnediate area.

c. REMOVAL OF AN EXPERIMENTAL WATER RADIDACTIVITY MONITOR (1) Description As part of a graduate research project conducted approxi-mately 8 years ago, an independent water monitoring loop was installed in the demineralizer circuit of the primary water system. The water monitor did not perform satisfac-torily, and once the project was finished, the monitor was never used again. The reactor staff removed the water monitoring loop from the demineralizer circuit.

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Specifically, all of the piping and equipment between

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IV-20 h! ~

valve DV 17 and the tee section where valve DV 22 was attached to the main water line for the demineralizer circuit was removed (see Figure IV.D.2). The tee was plugged and valve DV 17 remained in the system as a sample collection valve. This change resulted in a revision of OSTROP 7 to remove all mention of this water monitoring loop.

(2) Safety Evaluation There were no unfavorable safety implications associated with this facility change as this loop was never used to support reactor operations and the original water monitoring system for the reactor has remained fully functional. Appropriate health physics precautions were taken during the removal of the equipment. No significant contamination was found due in part to the long elapsed

[O ., time since the loop was used, and due to the low radioac-tivity concentration of the reactor primary water,

d. ADDITION OF A PARTICULATE Fil.TER DOWNSTREAM OF THE DEMINERALIZER (1) Description The reactor staff installed a particulate filter downstream of the demineralizer tank in the reactor primary cooling water cleanup system. The new filter prevents resin fines from being introduced into the primary water system.

The filter assembly is mounted next to the east wall of the heat exchanger room adjacent to the demineralizer pump skid. The pipe and valving system (see Figure IV.O.3) is primarily 1" plastic pipe with three valves (two shutoff valves and a bypass valve), two pressure gauges (to measure the pressure drop across the filter) and a drainable housing for the 25 micron filter. The filter housing

's can be shielded by concrete blocks, if needed.

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, f (2) Safety Evaluation There are no' unfavorable safety implications related i to the addition of this new filter. All of the materials used are of good quality, and the system passed all pre-operational-tests. - The new filter is. essentially no I l

different than the existing particulate filter for the  ;

. primary-water which has been in successful operation -

since the OSTR was built. Procedures used to change the new fiiter cartridge are the same as those used for the current filter. The new filter contribuus to increased safety by removing any resin fines which might pass from the demineralizer system..

e. REPLACEMENT OF THE WATER CONDUCTIVITY MONITORING SYSTEM j (1) Description  ;

The reactor staff upgraded the reactor water c'onductivity  !

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@!). monitoring system to a digital readoc .nstrument with automatic temperature compensation.  !

i While the original conductivity monitoring system was still quite reliable, it had one disadvantage related f to the fact that the " cat eye" indicating tube in the  ;

conductivity system had to be changed periodically.

These old-style " cat eye" electronic tubes were becoming increasingly expensive and could possibly become unavail-able in a few years.

. j In order to install the new system the following changes j were mede:

1. Two new signal cables were pulled from the heat exchanger room to the reactor console. The original cable did not have enough wire pairs to accomodate

~T the temperature compensating feature of the new (U..~.

instrument.

Y gy.24 '

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. ., 11. New temperature compensating conciuctivity probes  !

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were ir.'.alled.

iii. Adapters were-installed so that the new probes

-would fit.into the water system piping.

iv. Modifications to the console side cabinet were made to acconnodate the-installation of the new '

digital readout instrument.

(2) Safety Evaulation There is a great deal of historical data documenting the normal range of conductivity for the OSTR primary -

water. Any significant discrepancy observed between established values and'results with the new system could

- and would be immediately investigated using other conduc-I.

tivity instruments.

l- .3- Failure of the new system also does not create any imediate d safety-implications, since the conductivity of the reactor water changes very slowly with time, and thus all'ows-plenty of time for detection and repair cf conductivity l equipment. In addition, the OSTR reactor water is kept et an extremely low level of conductivity, which gives an even greater margin of protection.

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L Because this is new equipment it is expected that the potential for failure will actually decrease and that l-this installation will provide more accurate conductivity

[- readings. Hence, the new device will actually increase reliability and safety, i.

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IV-25

f. MONITORING 0F REACTOR POWER AND FUEL TEMPERATURE WITH A CDAS DURING NON-PULSING OPERATIONS (1)- Description On an as-needed basis, a computer-based data acquisition system (CDAS) can be connected to test terminals on the OSTR console to passively measure reactor power and fuel temperature signals during non-pulsing operations. Termi-nals TP2 and TP3 would be used to measure fuel temperature from the fuel thermocouple amplifier board (card XA16)'

in the left-hand console drawer. These terminals were extended to the rear of the console for easier and safer access. The LINEAR and LOG terminals already on the .

rear of the console would be used to measure the linear and log reactor power, respectively. No active circuitry will exist between the test terminals and the CDAS.

e~- (2) Safety Evaluation

(. )

The referenced terminals are designed to be used as indi-cated above; however, additional measures will be taken to ensure the safe use of the CDAS while measuring reactor power and fuel temperature during non-pulsing operations.

First, a 100-ohm resistor was permanently installed between console terminals TP3 and TP4 for the Nyt circuit, thus ensuring proper connection of the CDAS or other external recording devices to these terminals. Second, the CDAS system will be operationally tested to assure its proper functioning before it is connected to the reactor console.

This will eliminate the possibility that the CDAS will be connected to the console with an improperly functioning or failed component. In addition, cables used to connect the CDAS to console terminals were labeled and color-coded to minimize tN possibility of an incorrect connection.

It is important to note, however, that an incorrect con-nection would create no safety problems and would not

IV [

affect console electronics. The impact would be an incor-rect result on the external recording device. Furthermore, extending terminals.TP2 and TP3 from the thermocouple amplifier board (card XA16) to the rear of the reactor console will increase safety by making these terminals more accassible, by eliminating the need to directly access electronics in the left-hand drawer of the console,

, and by making it easier to confirm that the connections are correct.

An evaluation of the worst-case failure of the CDAS while connected to the reactor console to record reactor power

-~and fuel temperature indicated that the consequences are no worse than those creat(-J by the failure of an existing console component in the same circuit, and such consequences would be immediately obvious to the reactor operator so that' appropriate action could be taken.

(,)

A new OSTROP 26 was written and approved, it details operating procedures for the CDAS when it is being used to measure reactor power level and fuel temperature,

g. MONITORING OF REACTCR PEAK POWER AND- FUEL TEMPERATURE WITH A CDAS DURING PULSING OPERATIONS (1) Description The OSTR staff installed an operational amplifer inside the reactor console cabinet. The amplifier provides a gain of about 100 to amplify the peak power (Nvt) signal taken from console terminals TP3 and TP4. Signal amplifi-cation at these teminals is from about 60 mV to about 6 V. Although the amplifier circuit has been permanently installed inside the reac' tor console cabinet, it will e be connected to Nyt circuit terminals TP3 and TP4 only as needed during pulsing. Also, as needed during pulsing operations, a computer-based data acquisition system

'V IV-27 (CDAS) will be connected to the output from the operational amplifier to measure the peak reactor power during a

. pulse.

To measure fuel temperature during pulsing, the CDAS will also be connected to fuel temperature output terminals' TP2 and TP3 on the back of the OSTR console in order-to passively measure fuel temperature signals from the thermocouple amplifier board (card XA16) in the left-hand console drawer. No active circuitry will exist between the. fuel temperature terminals and the CDAS.-

(2) Safety Evaluation The referenced terminals are designed to be used as indi-cated above; however, additional measures have been taken to ensure the safe use of the CDAS and amplifier during pulsing operations.- First, a 100-ohm resistor has been

\[) pemanently installed between console Nyt teminals TP3 and TP4, thus ensuring proper connection of the CDAS, the amplifier, or other external recording devices to these terminals. Permanently installing a 100-ohm resistor

~

across terminals TP3 and TP4 increases the system's relia-bility by eliminating the need for a jumper cable across TP3 and TP4. Second, the amplifier and the CDAS system will be operationally tested to assure proper functioning before, they are connected to the reactor console. This will eliminate the possibility that these devices will be connected to the console with an improperly functioning or failed component, which will thereby eliminate the chance that a signal from the amplifier or CDAS will affect console electronics. In addition, cables used to connect the amplifier to the console and the CDAS to the amplifier were labeled and color-coded to minimize the possibility of an incorrect connection. It is impor-tant to note, however, that an incorsect connection would 4

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"IV-281

} create no safety problems and would not affect console electronics. The impact would be an incorrect result on the external recording device. Furthermore, only-cables; from input channels of the covered distribution board for the CDAS will be present, and thus no output ,

signal cables will be available for connection between  :

the CDAS and the reactor console. As an added feature, protective diodes were added to the amplifier circuit to isolate the CDAS and the amplifier from the reactor console. This action will prevent the CDAS or amplifier -

L from introducing a measurable charge to the Nyt circuit 1'

capacitor.

• l l

An evaluation of the worst-case failuni of the CDAS and L amplifier while connected to the reactor console to record l- peak pulse power and fuel temperature indicates that

( , the consequences are no worse than those created by the

) failure of an existing console component in the same o circuit, and such consequences would be immediately obvious .

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[

to the reactor operator so that appropriate action could

'. be taken.

p In addition to the above safety considerations, fuel temperature monitoring during pulsing is described and

. fully evaluated by the 10 CFR 50.59 evaluation entitled

" Monitoring of Reactor Power and Fuel Temperature with a CDAS During Non-Pulsing Operations," (see section IV.D.1.f). ,

Operating procedures for the CDAS and amplifier when used during pulsing are included in the new OSTROP 26

" Procedures for the Use of External Monitoring and Record-ing Devices."

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.IV-29.

I h h. INSTALLAT10N'0F AN EXTERIOR LIGHT ON THE EAST WALL OF THE REACTOR BUILDING (1) Description In order to provide additional lightirig for the northern

- half of the Radiation Center's parking lot, and to simulta-neously enhance exterior lighting on the east side of the reacter building, an exterior light was installed t at the not th end of the cast wall of the re&ctor building.

E -ectrical power was supplied by extending a conduit from an interior east wall electrical outlet located acaacer' ta the east fire ex1t door.

(2) Safety Evaluation The conduit paretration 'in the east reactor bay wall, as descrited abose, was filled with an electrical conduit

- which was appropriately cacited ins 1de'and out to prevent (M air leakage. Therefore, this facility change does not.

affect the ability of the reactor building to maintain the originally designed containment integrity, and does not alter the probability of minimal radiological releases as described in the facility SAR. Conseqaently, the change introduces no increase in-the probability or conse-quences of occurrences evaluated in the facility SAR.

Furthermore, no new types of occurrences are introduced and no margin of safety is reduced by the proposed change.

2. 10 CFR 50.59 Changes to Reactor Procedures There were four changes to reactor procedures which were reviewed, approved, and performed under the provisions of 10 CFR 50.59 during the reporting period,

/ -

Y

's?

IV-30

a.- REVISION OF OSTROP 6 (1) Description i OSTROP 6.0. was modified to incorporate revisions necessi- l tated by recent organizational and procedural changes.

In.particular, job descriptions for the Radiation Center Director, the Reactor Administrator, and most of the .,

health physics group were modified. The revised ROC charter was also incorporated into the procedure. In. ,

addition, the. access control procedure for the reactor bay.was changed to incorporate the daytime usc of signs when the security alarms are set during the day.

'(2) Safety Evaluation There are no unfavorable safety implications associated with the. job description changes. All of the necessary responsibilities are well-covered, and the organizational changes have already been approved by the NRC and incor-L porated intu the u5TR Technical Specifications. The changes to the ROC charter were also previously approved ,

by the ROC, and these changes were merely being incorpo-rated into OSTROP 6.0. The slight change to the reactor bay access procedure will not affect the security plan for the reactor. The change only affects those people with reactor bay keys and'these people are the most respon-sible members of the Radiation Center security staff.

Even if the procedure is not followed, security is not compromised, only an alarm is sounded unnecessarily,

b. REVISION OF THE EMERGENCY RESPONSE PLAN (1) Description A number of changes to the OSTR emergency response. plan y were made as a result of the annual review of the plan k.: by the standing subcommittee of the R0C. This review was conducted on October 26, 1988. Most of the changes

IV y , l

[

were required as a result of the revised Radiation Center organization recently approved by the NRC. Remaining changes were merely updates of such things as telephone numbers, first aid qualifications, etc.

(2) Safety Evaluation ,

None of the changes involve revisions of the actual emergency response described in the plan. Instead, the changes simply update the plan to incorporate the current titles and organizational structure of the Radiation Center,'

and involve appropriate modifications to designated lines of. succession in the emergency plan. As a result, none

-of the. changes have any impact on safety. Changes to the plan were also reported to the USNRC under the provi-sions of 10 CFR 50.54(q).

. P, c. REVISION OF OSTROP 10

- n$0 U" (1) Description The reactor staff amended section 10.7 of OSTROP 10 to specify the procedures for moving the cadniium-lined in-core irradiation tube from its storage location in the S-rack to the in-core position, and for returning the tube to the storage location.

In addition to amending section 10.7, the staff changed the title of OSTROP 10 to " Operating Procedures for OSTR Irradiation Facilities."

(2) Safety Evaluation The reactiyity effects of inserting and removing the cadmium-lined in-core irradiation tube have already been addressed in a previous 10 CFR 50.59 safety evaluation.

j The requirement relating to use of a specific set of

"' control rod calibrations corresponding to whether the

es.

IVo32 .

tube is in or out of the core will ensure that accurate measurements of core excess and shutdown margin are made.

A 24 limit for comparing control rod worths is also in-cluded in the revised section 10.7 because this is the estimated error for a rod calibration.

With respect to movement of the tube _in the reactor tank, aluminum TRIGA tubes filled with lead shot will be placed inside the cadmium-lined in-core irradiation tube and the top cap will be sealed to ensure that the tube is neutrally bouyant. Therefore, the impact of accidentally releasing the tube will be minimal. In addition, the tube will not be over the core when it is passed under.

the center channel, and as a result there will be no possibility of dropping the tube on the core.

d. (ADDITIONAL.) REVISIONS OF OSTROP 6.0 p3 4

S (1) Description Three (additional) amendments were made to OSTROP 6,0,

" Administrative and Personnel Procedures." The first amendment states that no external measuring or recording device will be connected to reactor measuring channels or safety channels without ROC approval of a 10 CFR 50.59 safety evaluation and any needed operating procedures.

However, it was not intended that ROC approval be required for normal use of standard diagnostic equipment by the Scientific Instrument Techr.ician or a designated replace-ment. A second amendment to OSTROP 6.0 states that when classes are in the reactor control room, the operator of record will not be the instructor of the class. Finally, a third amendment was added to ensure that any connection of an external system to reactor measuring or safety

~. channels will be checked by the Reactor Supervisor.

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IV-33 ,

[ Again, this tequirement was not' intended to apply to the normal use of standard diagnostic euuipment by the Scientific Instrument Technician or a designated replacement. -

(2) Safety Evaluation All of the three revisions were made to increase safety, and therefore the safety implications are all positive.

The first amendment noted above will prevent the addition of systems which might affect the reactor measuring or safety channels. The second amendment will decrease the possibility of the reactor operator being distracted by also having to instruct a class. The third aaendment

~

!' will help to ensure that any connections made to important console systems are made correctly.

{

l: '3. 10 CFR 50.59 Changes to Reactor Experiments

) There were three changes to reactor experiments which were reviewed, 'i approved, and performed under the provisions of 10 CFR 50.59 during the reporting period.

a. REVISION OF OSTR EXPERIMENTS B-3 AND B-11 (1) Description L

Experiments B-3 and B-11 were revised to add the new cadmium-lined in-core irradiation tube as-one of the standard OSTR irradiation facilities.

l.

l L (2) Safety Evaluation The installaticr. of the cadmium-lined tube in the core provides a new standard irradiation facility. From an operational and health physics standpoint, irradiation of samples in this facility is no different than irradi-

.- ating samples in cadmium cups in the (dummy) sample-holding

-  ! fuel element, in tS rotating rack or in .the pneumatic

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t., , s; y IV-34 hihji transfer facility, in fact, using the new tube will enhance safety by reducing radiation doses associated with the use and handling of cadmium cups, which will not now be needed for many experiments. All irradiations will be controlled following the-usual procedures associated with irradiation requests,

b. REVISION OF-OSTR EXPERIMENT B-31  !

(1)- Description 3 Experiment B-31 was revised to expand the types of materials that can be activated in OSTR facilities for flux-mapping purposes. All OSTR irradiation facilities can be used-including the reactor tank and associated in-core locations, since this was the original intent of Experiment B-31.  ;

a

.(2) Safety Evaluation yg -

I,I No reduction in safety effectiveness results from these x.,

i minor revisions. Reactivity values and radioactivity h- limits have not changed.

L  :

c. REVISION OF OSTR EXPERIMU;TS B-3, B-11 AND'B-1E L (1) Description l

i- Experiments B-3, B-11 and B-12 were revised to prohibit the irradiation of elemental mercury or substances where

! mercury is a major constituent.

l-(2) Safety Evaluation The change increases safety by preventing the introduction j of mercury into the reactor. Mercury reacts with aluminum and could therefore cause undesirable corrosion of reactor components.

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.t c E. Surveillance and Maintenance-1.. Non. Routine Maintenance 18 AUG 88 Replaced a pneumatic damper motor in the reactor building ventilation system.

29 AUG 88 -Replaced the carbon vanes in the stack radioactivity monitor pump.

16 SEP 88 Changed the ion chamber in the reactor safety channel.

12 DEC 88 Installed a 20 micron filter' downstream of the reactor primaty water demineralizer tank.

21 DEC 88 Repaired the low flow alarm on the continucus air monitor.

3 JAN 89 Sent the fuel element handling tool back to the manu-facturer for repair.

6 JAN 89 Installed a i.ew cranking battery f'or the emergency generator.

25 JAN 89 Installed a new test pote7tiometer on the console for the safety channel.

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- 30 JAN 89 Removed an unused water radioactivity monitor from the reactor's primary water purification system.

6 FEB 89 Performed a shuffle of fuel elements to even out burnup.

16 FEB 89 Replaced a diode in the resctor servo system for better servo response.

27 FEB 89 Repaired a frozen pre-heat coil in the reactor building ventilation system. ,

17 MAR 89 Replaced two bearings in the reactor building ventila-tion supply fan.

l 3 APR 89 Replaced bearings in the reactor primary water pump.

4 APR 89 Replaced bearings in the reactor primary water pump motor.

10 APR 89 Replaced bearings (again) in the reactor primary water pump.

10 APR 89 Replaced the GM tube in the stack monitor's particulate

.. channel.

25 APR 89 Replaced the pump motor on the continuous air monitor.

..- ~.

-IV-36

)L 27 APR 89- Replaced the time delay relay in the stack monitor.

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4 MAY 89 Permanently installed a 100 ohm resistor across console terminals TP3 and TP4 for the Nyt circuit.

-12 JUN 89 Installed a new outside light on the northeast side of the reactor building.

13 JUN 89 Replaced the control room access door closed circuit TV monitor.

2. Routine Surveillance and Maintenance

.The OSTR has an extensive ~ routine surveillance and maintenance (S&M) program.. > Examples of typical S&M checklists are presented in Figures IV.E.1 through IV.E.4. Items marked with an asterisk

(*) are required by the OSTR Technical Specifications.

F. Reportable Occurrences In a letter to the USNRC dated April 17, 1989, the OSU Radiation Center

' .j reported an event involving OSTR procedures for hooking up external l~ monitoring equipment to the OSTR console. Based on the nature of

-the event, more stringent procedures were implemented for the connection of any external measuring device to the reactor control console.

No citations were issued as a result of the report and a rcutine onsite inspection by the USNRC.

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e Quarterly Surveillance and Maintenance (3emple. Form) f .

' WILtJEE 8 IRISIEEEE FE LIE N (F / / 19 SSTRIF 14 TARGET DATE sof 70 SRTE REfMARS 8 LIMITS AS FOUND SATE 8E rarsam** CSWLETED INITIALS SURVEILLANCE 8 W !G E = E TO GE PERF0MED

• l REACTOR OPERATIONS CSP'MEE (ROC) AUDIT OF REACTOR gpggyggty CPERATIONS FOR / / OtMRTER

• 2 OtMRTERtf ROC MEETING OUARTERLV

$3 FUEL ELEMENT RADIAT!ON LEVEL E ASUREE NTS IN IIATER f, , ',

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1 PART V PROTECTION I i A. Introduction This section of the report deals with the radiation protection program at the OSU Radiation Center. The purpose of this program is to ensure the safe use of radiation and radioactive materials in the Center's teaching, research, and service activities, and in a similar manner to ensure the fulfillment of all regulatory requirements of the state of Oregon, the U.S. Nuclear Regulatory Comission, and other regulatory agencies. The comprehensive nature of the program is shown in Table V.A.1, which lists the program's major radiation protection requirements and the performance frequency for each item.

The radiation protection program is implemented by a staff consisting-of a Senior Health Physicist, a Health Physicist, a Radiation Protection

. Technologist, and one to five part-time Radiation Protection Technicians

." (see Part II.F). Assistance is also provided by the reactor operations group, the neutron activation analysis group, the Scientific Instrument Technician, and the Radiation Center Director.

The data contained'in the following sections have been prepared to comply with the current requirements of Nuclear Regulatory Comission (NRC) Facility license No. R-106 (Docket No. 50-243) and the Technical Specifications contained in that license. The material has also been prepared in compliance with Oregon Department of Energy Rule No. 345-30-010, which requires an annual report of environmental effects due to research reactor operations. A subunary of required data for the OSTR is provided in Part I.E for quick reference.

Within the scope of Oregon State University's radiation protection program, it is standard operating policy to maintain all releases of radioactivity to the unrestricted environeent and all exposures to radiation and radioactive materials at levels which are consistently "as low as reasonably achievable" (ALARA).

.- '. I V-2

@" Table V.A.1 Radiation Protection Requirements and Frequencies FREQUENCY RADIATI(* PROTECTION REQUIRDENT Daily / Weekly / Monthly Routine area radiation / contamination monitoring.

Weekly Gassna spectroscopy of the (OSTR) continuous air monitor particulate filter.

Monthly Routine response checks of radiation monitoring instruments.

Monitor radiation levels (pR/hr) at the environmental monitoring stations.

Collect and analyze TRIGA primary, secondary, and make up water.

Exchange personnel dosimeters and inside area monitorihg dosimeters and review exposure reports. i Laboratory inspectionc. I Emergency and safety equipment checks. l Neutron generator and tritium assembly contamination survey. l Calculate previous month's gaseous waste' discharge. l l

As Required Process and record solid and liquid waste discharges. i Prepare and record radioactive material shipments. l Survey and record incoming radioactive material receipts. I Monitor and record special radiation surveys. l Perform thyroid and urinalysis bioassays.  !

Conduct orientation and training. l

~

,I Issue radiation work pennits and provide health physics coverage for maintenance operations.

Quarterly Prepare, exchange and process environmental TLD packs.

Collect and process enviror. mental soil, water and vegetation samples.

Orientation for classes using radioactive materials.

Collect and analyze sample from reactor ventilation effluent line.

Exchange personnel dosimeters and inside area monitoring dosimeters and review exposure reports.

Semi-Annual Leak test and inventory sealed sources.

l Floor survey of corridors and the reactor bay.

Calibrate portable radiation monitoring instruments and personnel pocket ion chambers.

Inventory and inspect Radiation Center equipment located at the Student Health Center, Corvallis Fire Department Haz/ Mat van, and Good Samaritan Hospital.

Annual Calibrate reactor stack effluent monitor, continuous air monitors, remote area radiation monitors, water monitor, and air samplers.

Measure face air velocity in laboratory hoods and exchange dust-stop filters and HEPA filters as necessary.

Inventory and inspect Radiation Center emergency equip;nent.

Facility radiation survey of the cobalt-60 irradiator and X-ray machine.

Personnel dosimeter training.

=-

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1

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l Table V.B.1 Monthly Sunenary of Liquid Efflueet Releases to.ttie Sanitary Sewer for the year July 1.1988 through June 30,1989(1)

(OSTR Contribution Shoun in () and Gold Print)

Specific Activity Average For Each Detectable l Concentration Percent cf Radionuclide in the Total Quantity of Released Applicable Total Volume Waste. Where the of Each Detectable Radioactive MPC for of Liquid Date of Total Quantity Release Radionuclide Material at Released Effluent Discharge of Radioactivity Detectable Concentr tion Was Released in the the Point of Radioactive Released.

(Month & Released Radionuclides >l x 10- pCi/cc Waste Release Material Includipa in the Waste (pC1/cc) (Curies) (uC1/cc) (1) 011aent W Year) (Curies)

NOV 88 4.10 x 10-4 g,4g(3)

Radiation H 3.76 x 10-5 4.12 x 10-4 C --- x 3.78 x 10-5 Center _

0.041(4)

Plus OSTR

- - ------ - -------- -------- ------------ ----------- --------- ------- 2872 (3.74 x 10-5) (4.08 x 10 ) (3.75 x y 5) (1.35)

(e.045)k3,I Cont tion (4.08 x 10-4) ( ) (i (3.05 x no- ) )

t3 Ano,e H 5 5.44 x 10-4 APR 89 60Co 3.07 1.86 xx 10 10-7 3.29 x 10-6 Radiation 3.33 x 10-5 1.71(3)

Senter 5.90 x 10-4 65Zn ---

1.14x10-j Plus OSTR g3h 5.M x F 7 1.M x W .

Cs --- 2.76 x 10,7 1

- ------- ---------- -------- ------------ ----------- --------- ------- un 05TR (3N) (2.51 x 10-5) (4.45 x 10-4) (8 88)( }

Con r on (4.45 x 10-4) , ,-g) (2.51 x F5)

) _) (0.035)(4) ,

f Annual Total for ~1.6%(3}

1.00 x 10-3 See Not 1.00 x 10-3 3.50 x 10-5  ;

! Radiation Applicable 0.061 l

Center Above y t

Plus OSTR ,

-- --_--- ---------- -------. .----------- .---------- ----- --- ------- 7542 OSTR Contribution (3N) (8.53 x F4) (3.00 x F5) (1.95)(3)

(8.53 x 10-4) (3 88 x W7 ) (8.838)(4) to Above (44C *)

(1) The OSU operational policy is to subtract only detector background from our water analysis data and not background radioactivity in the Corvallis city water. There were no liquid effluent releases during months not listed.

(2) The total volume of liquid effluent plus diluent does not take into consideration the additional mixing with the over 7.500.000 gallons per year of liquids and sewage normally discharged by the Radiation Center complex into the same sanitary sewer system. ,

(3) Based on values listed in 10 CFR 20. Appendix B. Table 2. Column 2.

(4) Based on values listed in 10 CFR 20. Appendix B. Table 1. Column 2, which are applicable to sewer disposal.  !

7 w

i

=-_. -

V-5 Table V.B.2 Monthly Summary of Gaseous Effluent Releases for the Year July 1, 1988 through June 30,1989(1)

Estimated Average Percent of the.

Atmospheric Applicable Total Diluted MPC for Diluted Total Estimated Concentration of Concentration

. Date of Estimated Quantity of Argon-41 at of Argon-41 at Discharge Radioactivity Argon-4 Point of Release Point of Release (Month & Released Released 2 (1 ) (ReactorStack) (Reactor Stack) year) (Curies)- (Curies) (pCi/ml) (%)

JUL 88 0.42. 0.42 3.1 x 10-8 0.8%

AUG 88 0.42 0.42 3.2 x 10-8 0.8%

'SEP 88 0.63 0.63 5.0 x 10-8 1.2%

OCT 88 0.49 0.49 3.7 x 10-8 o,95 NOV 88 0.41 0.41 3.2 x 10-8 0.8%

DEC 88 0.56 0.56 4.2 x 10-8 1,1g .

JAN 89 0.41 0.41 3.0 x 10-8 0.8%

FEB 89 0.41 0.41 3.4 x 10-8 o,9%

<W.>h MAR 89 0.96 0.96 7.2 x 10-8 1.8%

(W - APR 89 0.30 0.30 2.3 x 10-8 0.6%

MAY 89 0.76 0.76 5.6 x 10-8 1,4%

JUN 89 0.57 0.57 4.4 x 10-8 1.1%

ANNUAL g gE 6.3 6.3 4.0 x 10-8 1,0%

(1) Airborne effluents from the OSTR contained no detectable particulate radioactivity resulting from reactor operations, and there were no releases of any radioisotopes in airborne effluents in concentrations greater than 20% of the applicable MPC value. (20% is a value taken from the OSTR Technical Specifications.)

(2) Routine gamma spectroscopy analysis of the gaseous radioactivity in the OSTR stack discharge indicated the only detectable radionuclide was argon-41.

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Table V.B.3 Annua 1' Summary of Solid Waste Generated and Transferred l for the Year July 1, 1988 through June 30, 1989 ]

Volume of Total Quantity Dates of Solid Waste Detectable of Radioactivity Shipment to Origin of Packaged Radionuclides in Solid Waste U.S.EcogL1gy Solid Waste (CubicFeet) in the Waste (Curies) Company 3-Hydrogen 46-Scandium 51-Chromium 54-Manganese

~~

58-Cobalt 59-Iron TRIGA 60-Cobalt 8/24/88 Reactor 21 75-Selenium 5.2 x 10-5 6/15/89 Facility 82-8romine 124-Antimony 132-Tellurium 141-Cerium Cj 144-Cerium 152-Europium 154-Europium 46-Scandium Radiation 59-Iron 8/24/88 l Center 12 60-Cobal t 6.25 x 10-6 6/15/89 Laboratories 154-Europium l

I 75-Selenium (1) All Radiation Center and OSTR solid radioactive waste is routinely transferred onsite (within the Radiation Center building) to the OSU Radiation Safety Office, where it is held on the University's State of Oregon radioactive materials license, along with other campus waste, prior to shipment to U.S. Ecology by the Radiation Safety Office.

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V-10 Table V.C.1 Annual Summary of Personnel Radiation Doses Received For the Year July 1, 1988 through June 30, 1989 Average Annual Greatest Individual Total Person-mrem Dose (1) Dose (1) For the Group (1)

Whole Whole Whole Personnel Group Body Extremities Body Extremities Body Extremities (mrem) (mrem) (mrem) (mrem) (mrem) (mrem)

Facility Operating Personnel 12 72 80 500 245 1520

. Key Facility Research Personnel 0 5 0 40 0 110 Physical Plant Maintenance Personnel <1 N/A 8 N/A 42 N/A l Laboratory Class Students 0 7 0 100 0 320 Campus Security and Police Personnel 0 N/A 0 N/A 0 N/A Visitors '1

< N/A 5 N/A 135 N/A (1) "0" indicates that each of the beta-gama dosimeters during the reporting period was less than the vendor's gamma dose reporting threshold of 10 mrem or that each of the neutron dosimeters was less than the vendor's threshold of 30 mrem, as applicable. "N/A" indicates that there was no extremity monitoring conducted or required for the group.

,s*%

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. c-V-13 Table V.D.1 Total Dose Equivalent Recorded on Area Dosimeters Located Within the TRIGA Reactor Facility for the Year July 1, 1988 through June 30, 1989 Total Recorded TRIGA Reactor Dose Equivalent (1)(2)

Monitor Facility Location X8(G) Neutron I.D.. (See Figure V.D.1) (arem)(3) (mrem)

MRCTNE D104 North Badge East Wall 65 0 MRCTSE D104 South Badge East Wall 0 0 MRCTSW D104 South Badge West Wall 65 0 MRCTNW D104 North Badge West Wall 90 0 MRCTWN 0104 West Badge North Wall 0 0 MRCTEN 0104 East Badge North Wall 50 0 MRCTES- 0104 East Badge South Wall 250 0 MRCTWS D104 West Badge South Wall '

340 0 MRCTTOP D104 Reactor Top Badge 405 0 MRCTHXS D104A South Badge HX Room 370 0 xMRCTHXW D104A West Badge HX Room 20 0 fiRCD-302 D302 Reactor Control Room 145 0 D332A Reactor Supervisor's Office 15(4) N/A sl.MRCD-302A >

l' (1) The total recorded dose equivalent values do not include natural background contribution and, except as noted, reflect the sumation of the results of 12 monthly beta-gamma dosimeters or four quarterly fast neutron dosimeters for -

each location. A total dose equivalent of "0" indicates that each of the beta-l gamma dosimeters during the reporting period was less than the vendor's gama dose reporting threshold of 10 mrem or that each of the fast neutron dosimeters was less than the vendor's threshold of 50 to 100 mrem, as applicable. "N/A" indicates that there was no neutron monitor at that location.

(2) These dose equivalent values do not represent radiation exposure through an ,

exterior wall directly into an unrestricted area. '

(3) The total recorded dose equivalent values reflect the sumation of eleven monthly I beta-gama dosimeters. September's dosimeters were lost at Radiation Detection  ;

Company during processing. 1 (4) The total dose equivalent reflects the sumation of four quarterly beta-gama j dosimeters.

(v 1 L

l l

l I

1

y .- - - , . == - -

, Le:

V-14 Table V.D.2.

Total Dose Equivalent Recorded on Area Dosimeters g<' Located Within the Radiation Center for the Year July 1. 1988 through June 30, 1989 Total Recorded Radiation Center Dose Eevivalent (1)

Monitor Facility Location XS(G) Neutron 1.D. (See Figure V.D.1) (aren) (mrom)

IRCA-100 Receptionist's Office 0(2) N/A MRCA 120A NAA Temporary Sample Storage. A120A 1780 N/A MACA 126 GAmpus R$0's Radioisotope Receiving Lab 15 (2) N/A MRCCD-60 "Co Irradiator Room 40(2) N/A

.MRCA-130 shielded Exposure Room. A130 0 N/A MRC300ERAY X-Ray Console Room 0I N/A MRCA 134-2 NAA Research 165 s;2ll 2s N/A MRCA-138 Health Physics Laboratory. A138 0 N/A MRCA-146 Gama Analyzer Room (Storage Cave? 15 N/A MRCB-100 Gama A Storage Caves 0 N/A MRCB-114 a Lab (gRa Storage Fact 11ty)er Room l 1530(? 0 MRC8-116-1 R$0's RAM Waste Processing Facility 15 L J N/A' MRCB-116-2 RSO's RAM Weste Facility Compactor Room 15 ( ) N/A .

MRC8-119 Source Storage Room 15 i , J N/A MRCS-119A Sealed Source Storace Room 3740 i , J 20 %

MRC8-120 Instrument Calibrction Facility 0h J N/A MRCB-122-2 Radioisotope Storage Hood 1040( J N/A MRCB-122-3 25 L J N/A MRCB-124 Radioisotope Research Radioisotope Research Laboratory (Hood)

Laboratory 0 M/A MRC8-124-2 Radioisotope Research Laboratory 70 (, ?J I N/A

_ MRCB-124 6 Radioisotope Research Laboratory 0LJ N/A

. cit MRCB 128 Instrument Re ir Shop 0LJ N/A M) -

MRCB-132 MRCC-100 Radioisotope esearch Laboratory Director's Office 200 g J OLJ N/A N/A MRCC-106-H East Loading Dock. C106H 0 N/A MPCC-118 Radio-Chemistry Laboratory Student Counting Laboratory 0(?

0(

N/A N/A.

MRCC-120  ?

MRCC-123N Gama Analyzer Room (Storage Cave). C123 135ll J N/A MRCC-1235 Gama Analyzer Room 0i , J N/A MRCC-124 Student Computer Laboratory 0i , J N/A MRCC-126 . Student Counting Laboratory 0! N/A MRCC-130 Radioisotope Laboratory 0 ll JJ 0(4)

MRCC-134 Gama Analyzer Room (Storage Cave), C134 145 N/A MRCD-102 Pneumatic Transfer Terminal Laboratory 95 ( ? O

• 0i 0 MRCD 102-H 1st Floor Corridor 9 D102 l MRCD-106-H 1st Floor Corridor 9 D106 150I N/A MRCD-200 Senior Health Physicist's Office 115 01 (hh N/A 0

MRCD-204-H 2nd Floor Corridor 9 D204 >

MRCD-300 3rd Floor Conference Room 15 h 0

'MRCBRF Front Personnel Dosimetry Storage Rack Rear Personnel Dosimetry Storage Rack' 0

OLJ (L ) N/A N/A MRCBRR (1) The total recorded dose equivalent values do not include natural background contribution and, except as noted. reflect the sumation of the resJits of 12 monthly beta-gama dosimeters or four quarterly fast neutron dosimeters for 4ach location. A total dose equivalent of "0" indicates that each of the beta-gama dosimeters during the reporting period was less than the vendor's t gasuna dose reporting threshold of 10 mrem or that each of the fast neutron dosimeters was less L

than the vendor's threshold of 50 to 100 mrem, as applicable. "N/A" indicates that there was

! no neutron monitor at that location, l-

! (2) The total dose equivalent reflects the sumation of four quarterly beta-gama dosimeters.

! (3) The total dose equivalent reflects the sumation of eleven monthly beta-gama dosimeters.

l- h September's dosimeters were lost at Radiation Detection Company during processing.

(4) The total dose equivalent reflects the exposure for June 1989 only.

l

. -o o ,

V-16 Table V.D.3 4.g Annual Sumary of Radiation Levels and Contamination Levels Observed Within the Reactor Facility and Radiation Center During Routine Radiation Surveys for the Year July 1,1988 through June 30, 1989 Whole Body Contamination Radiation Levels (mrem /hr). Levels cm (dpm/100 (1)2)

Accessible Location (See Figure V.D.1) Average Maximum Average Maximum TRIGA Reactor Facility:

• ReactorTop(D104) 1 101 <500 <500 Reactor 2nd Deck Area (D104) 6 42 <500 <500 Reactor Bay SW (D104 <1 39 <500 <500 Reactor Bay NW (D104 <1 15 <500 <500 Reactor Bay NE (D104 <1 7 <500 <500 Reactor Bay SW (D104 <1 22 <500 <500 Class Experiments (D104,D302) <1 4 <500 <500 Demineralizer Tank--

Outside Shielding (D104A) <1 7 <500 <500 Particulate Filter--

Outside Shielding (D104A) <1 3 <500 <500 Radiation Center:

- NAA Counting Rooms (A146,B100,C134) <1 3 <500 <500 Health Physics Laboratory

<1 <1 <500 <500 60(A138)

Co Irradiator Room (A128) <1 7 <500 <500 Radiation Research Labs (B114,B122,B124,B132,C130) <1 7 <500 <500 Radioactive Source Storage (B119A) <1 29 <500 <500 l Student Chemistry Laboratory (C118) <1 <1 <500 <500 l

i Student Counting Laboratcries (C120,C126) <1 3 <500 <500 Operations Counting Room (C123) <1 <1 <500 <500 Pneumatic Transfer Laboratory (D102) <1 3 <500 15000(2)

(1) <500 dpm/100 cm2 = Less than the lower limit of detection for the portable survey instrument used.

l l (2) The contamination shown for this location assumes 100% smearing efficiency and

/ was immediately removed. As a result, the average contamination level at this location during the reporting period was, for all practical purposes, <500 dpm l

(- ) per 100 cm2 ,

1 I

I

I .

, Figure V.E.1

- .II Area Radiation Monitor Locations for the TRIGA Reactor, and on the TRIGA Reactor Area Fence h.,l N_

PERIMETER FENCE

\

MRC FE3 - MRC FE2 l OFFICE I BUILDING hI SMRC FE4 MRC FE1G l

C .

l-( MRC TNE e

MRC_TSE e

r R ADI ATION i

MR6 TEN g TRIGA I

I ' REACTOR CENTER

- 1 BAY BUILDIN G ()

MRCTWN $ MRCTSW j

$ 8 i

MRCTNW/ HX $ MRC THXS ,

l GROOM i

MRC THXW D MRC FE5 ' ] [ L10UID .

Juv R. A.WA STE I HOLD UP TANK  ;

L i l

1 MRC FE6 k S _

s TRANSFORMER MRCFE96 i

l STATION MRCFE7 MRC FE8 I

e e a 0 50ft

V-20 Table V.E.1 Total Dose Equivalent at the TRIGA Reactor Facility Fence for the Year July 1, 1988 through June 30, 1989 Total Calculated Dose Equivalent Fence Based on the Annual Average Environmental Total Recorded Total Recorded Monitoring Dose Equivalent Dose Equivalent pR/hr Station Based on R.D. Co. Based on OSU Exposure Rate (See TLDs (1) TLDs(2)(3) (3)

Figure V.E.1) (mrem) (mrem) (mrem)

HRCFE-1 100 81 i 17 77 1 20-MRCFE-2 106 92 1 14 80 1 18 MRCFE-3 106 87 1 3 84 1 19 MRCFE-4 112 85 1 2 92 1 22

,. MRCFE-5 98 78

• 10 76 20 c !.s.1 MRCFE-6 98 85 1 6 81~i 37 O MRCFE-7 101 84 7 75 1 24 MRCFE-8 103 82 2 7 73 1 19

, MRCFE-9 96 79 4 64 1 15 (1) Radiation Detection Company (R.D. Co.) TLD totals include their annual natural background contribution of 80 mrem for the reporting period. Average Corvallis -

area natural background using Radiation Detection Company TLDs totals 88 mrem for the same period.

(2) OSU fence totals include a measured natural background contribution of

. 72 1 2 mrem.

(3) values represent the standard deviation of the total value at the 95%

confidence level.

V-21 F19ure V.E.2 Monitoring Stations for the OSU TRIGA Reactor-For the Year July 1,1988 through June 30, 1989 6 CD C3C3 DOC 3 6QpC"3C3t3 w ~ w ,,u n w a o ,e w m-

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V-22 i

Table V.E.2 Total Dose Equivalent at the Off $1te Gama Radiation Monitoring Stations 1: for the Year July 1.1988 through June 30. 1969  ;

Total Calculated Dose Equivalent  ;

Off-51te Based on the Radiation Total Recorded Total Recorded Annual Average Monitorine Dose Equivalent Dose Equivalent uR/hr Station (l') Based on R.D. Co. Based on OSU Exposure 1 tate (See Fleure TLDs(2) 4 V.E. 7) (mres) TLDs (ares (3))(4) (a(re)n) 15tCTE 2L ---

69210(5? $7

• 21 MRCTE-3 106 69

• 9 (5J , 81 2 17 MRCTE-4 96 85

• 7 69

• 19 MRCTE-5L --- 94 2 13 81 1 15 15tCTE-6 103 95 t 11 81 t 16 MRCTE.7L --- 82 1 9 85 t 10 MRCTE-8 111 72

• 7(5) 90 1 14 MRCTE-9 108 91 2 11 87 t 14 MRCTE-10 91 84

• 15 64

• 15 MRCTE 90 58 1 6(5) 64

• 27 MRCTE-12 103 100 t 9 87 2 20 MRCTE-13L --- 95 2 5 76 1 12

c. MRCTE-14L --- 60 2 6(5) 59 1 16 65210(6) 75 t 14 1O M MRCTE-15 97 94

• 9 80 2 17 MRCTE-16L ---

1. "i MR(,T E-17 95 84 1 7 68

• 18 MRCTE-18L --- 1r2 2 8- 75 2 16

'MRCTE-19 108 101 2 11 87 2 13 MRCTE-20L --- 104 2 14 75 t 16 MRCTE 81 72 2 4 45

• 8 MRCTE-22 86 87

• 9 54 1 13 (1) Monitoring stations coded with an "L" contained one standard OSU TLD pack only.

Stations not coded with an "L" contained. in addition to the OSU TLD pack, one R.D.

Co. TLD monitoring pack.

(2) Radiation Detection Company TLD totals include their annual natural background contribution of 79 mrem for the reporting period. Average Corvallis area natural background using Radiation Detection Company TLDs totals 88 mrem for the same period.

(3) OSU off-site totals include a measured natural background contribution of 81

• 7 mrem.

(4) i values represent the standard deviation of the total value at the 95% confidence level.

(5) The total dose equivalent for three quarterly monitoring periods only. The TLD packet was lost or stolen during one quarter.

(6) The total dose equivalent for three quarterly monitoring periods only. The TLD packet was mistaken for a pipe bomb and was removed and X-rayed by a bomb squad from the state of Oregon.

3

. '?..

V.25

+

e.,-

Table V.E.3 .

t.

Annual Average Concentration of the Total Net Beta Radioactivity (Minus 3) H for Environmental Soil, Water, and Vegetation Samples i for the Year July 1,1988 through June 30, 1989 Sample l L.ocation Annual Average Concentratiog of Sample the Total Net Beta (Minus JH) Reporting (See V.E.2Fig)ure Type Radioactivity (1) Units pC1/cc  ;

1-W Water 2) 2.91 x 10-8 1 3.52 x 10-9 pCi/cc 4-W Water 2) 2.91 x 10-8 3.52 x 10-9 )

11-W Water 2)2.91x10-8 2 3.52 x 10-9 pCi/cc 19-RW Rainwater 2) 2.91 x 10-8 i 3.52 x 10-9 pCi/cc  !

3-5 Soil 9.07 x 10-5 2 1,37 x 10-5 pCi/ gram of dry soil 5-S Soil 1.06 x 10-4 i 1.32 x 10-5 pCi/ gram of dry soil l

20-S Soil 6.23 x 10-5 i 1.95 x 10-5 pCi/ gram of dry soil 21-S Soil 9.74 x 10-51 1.25 x 10-5 pCi/ gram of dry soil 2-G Grass 3.23 x 10-4 1 2.69 x 10-5 pCi/ gram of dry ash 2.95 x 10-4 2.74 x 10-5 pCi/ gram of dry ash

' C.? '1 6-G Grass 7-G Grass 4.04 x 10-4 3.03 x 10-5 pC1/ gram of dry ash 8-G Grass 5.03 x 10-4 1 3.17 x 10-5 pCi/ gram of dry ash 9-G Grass 3.29 x 10-4 2 2.56 x 10-5 pC1/ gram of dry ash 10-G Grass 2.11 x 10-4 2 2.76 x 10-5 pCi/ gram.of dry ash 12-G Grass 2.59 x 10-4 2 2.46 x 10-5 pC1/ gram of dry ash 13-G Grass 4.73 x 10-4 i 3.28 x 10-5 pC1/ gram of dry ash 14-G Grass 2.45 x 10-4 1 2.58 x 10-5 pCi/ gram of dry ash Grass 3.16 x 10-4 1 2.82 x 10-5 pC1/ gram of dry ash 15-G 16-G Grass 3.46 x 10-4 2.55 x 10-5 pCi/ gram of dry ash 17-G Grass 4.08 x 10-4 3.07 x 10-5 pCi/ gram of dry ash 18-6 Grass 4.58 x 10-4 1 3.27 x 10-5 pCi/ gram of dry ash Grass 3.57 x 10-4 2.88 x 10-5 pCf/ gram of dry ash 22-G (1) i values represent the standard deviation of the average value at the 95% confidence level.

3.52 x 10-9 pCi/cc.

(2) Less than lower limit of detection of 2.91 x 10-8

,o*.

9