ML20136H041

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Co Insp Rept 50-264/69-01 on 690402.No Safety,Noncompliance or Unusual Occurrences Noted.Areas Inspected:Reactor Operation
ML20136H041
Person / Time
Site: Dow Chemical Company
Issue date: 04/15/1969
From: Fiorelli G, Gower G, Grier B
US ATOMIC ENERGY COMMISSION (AEC)
To:
Shared Package
ML20136C633 List:
References
FOIA-85-483 50-264-69-01, 50-264-69-1, NUDOCS 8508200137
Download: ML20136H041 (6)
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U. S. ATOMIC ENERGY COMMISSION REGION III DIVISION OF COMPLIANCE Report of Inspection CO Report No. 264/69-l '

as Licensee:

Dow Chemical Company License No. 2-108 Category E Date of Inspection:

April 2, 1969 Date of Previous Inspection: September 26, 1968 kE Inspected By:

G. Fiore111 Responsible Reactor Inspector April 11, 1969 hSAHA G. C. Gower Reactor Inspector April 11, 1969

/%,- ll b',., f Y. Grier Regional Director April 15, 1969 Reviewed By:

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i Proprietary Information:

None i

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SCOPE An announced inspection was made of the 100 kw TRIGA Mark I Research l

Reactor located in Midland, Michigan.

Mr. G. C. Gower participated in the inspection effort and assisted in the preparation of this report.

Mr. Gower assumes principal inspection responsibility following this visit.

SUMMARY

Safety Items - None.

Noncompliance Items - None.

Unusual Occurrences - None.

Status of Previously Reported Problems - No problems involving followup action were reported in the previous report.

Other Sienificent Items - Visual inspections of the fuel elements revealed j

one element which was slightly bowed.

Refer Section 11, i

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2-Reactor scrams from power loss to the reactor safety circuit system This check was prompted by the reported scram sere confirmed by test.

circuit deficiency noted in connection with a Mark III TRIGA reactor.1/

Refer Section G-3.

Manacement Interview - Inspection results were discussed with Drs. W. H.

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Beamer and O U. Anders. The following items were discussed:

Dr. Anders has ordered the, fabrication of a fuel element measuring jig which will be used to determine the extent of bowing of the element in position F-10 (located in the core fringe).

Replacement of the fuel element will be based on conformance with acceptable tolerances used

~in the fabrication of new fuel elements.

Dr. Beamer informed the inspectors that he considers the probability of demonstrations in close proximity of the Dow Industrial Complex to be very low.

He was sppreciative of the concern to reactor damage, however, he stated damage te the huge and complex chlorination process equipment was considered to be of much more significant safety concern.

Plant security as well as other low enforcement agencies are immediately available to the plant.

He stated that he would notify the CO:III office of any disturbance or activities which appeared directed at the reactor facility.

DETAILS A.

Persons Contacted Dr. W. H. Beamer, Laboratory Director Dr. - 0. U. Anders, Reactor Supervisor Mr. L. G. Silverstein, Radiological Of ficer Mr. J. Charms, Assistant Radiological Of ficer B.

Administration and Organization

- The operating staf f of the reactor has not changed since the last visit.

One new member has been added to the Reactor Operations Committee.

The new member, Dr. G. L. Jewett, is a research chemist and holds a senior reactor operator's license. The committee has met four times since the last visit.

The safety committee minutes reviewed' disclosed the topics of discussion were of a safety nature.

C.

Operations The reactor continued to be used solely for small sample irradiations and activation analyses. Operating runs are made at the maximum license power level of 100 kw.

,g 1/ Memo from J. P. O'Reilly dated 2/17/69.

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A total of seven scrams were experienced since the last inspection.

7 All were inadvertent and due to switching errors, instrument noise and maintenance.

I D.

Facility Procedures The reactor operating procedures were reviewed and upgraded during the period. The changes were covered in a Reactor Operatidns Committee meeting as noted in the meeting minutes' These changes were principally concerned with-' delegations'of responsibility and authority, and defined reactor supervisor alternates in the event Dr. Anders is not available i

to the facility.

E.

Primary System A review of operating records disclosed that the reactor pool temper-ature has been maintained below 120 F as required by the technical specifications. The maximum temperature noted in the records was 800F, The inspector was informed by Dr. Beamer that there had been no in-dications of pool leakage.

A review of the records and discussions with Dr. Beamer did not dis-close any problems with drain leaks, pipe breaks, or unaccountable water losses.

A review of activity measurements of evaporated pool water samples disclosed activities essentially at background levels.

G.

Reactor Control and Core Physics 1.

The maximum available cold clean excess reactivity was noted to be within the 1.5% ak/k limit specified in the technical specifications.

t Control rod calibrations were conducted in accordance with the semi'-annual frequency required by the technical spedifica-tions.

The last measurement was conducted on March 4,1969, using the rising period method.

The following measured values were obtained from the operating logs:

Rod Worth Safety

$2.94 Shim 3.18 Regulating

.73 Total

$6.85 luknd

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. The shutdown margin based on the most reactive of the operable i

rods withdrawn was determined as follows from logbook entries:

Total Rod Worth

$6.85 I

Most Reactive Rod 3.18

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Difference

$3.67 anema Excess Reactivity 2,07 Shutdown Margin

$1.60 (1,1% Ak/k)

This is in complia'nce with the minimum technical specification limit of 0.357.4k/k.

i Control rod scram timing was noted to have been performed l

within technical specification frequency requirements and all times were noted to be within the one second technicial specification requirement.

2 Functional Tests A review of interlock functional testing, and power level safety circui'c functional testing, were all noted to have been performed in accordance with technical specification requirements. No unusual conditions or performances were de-

tected, i^

3.

Safety Circuitry A check into the fail safe design of the safety circuit was 4(,,,,,

made by the licensee. This test was prompted by a call from CO:III and was based on the safety circuit deficiency des-i cribed in the Region I report covering the experience of the j

Cornell University TRIGA facility.

I The test confirmed by rod drop that loss of power to the safety cir-"it channels did result in a reactor shutdown.

H.

_ Core and Internals t

The licensee performed a visual examination of each of the fuel ele-

,j ments in accordance with technical specification requirements.

Dr. Anders informed the inspector that one fuel element was noted to have a very alight blow.

(It should be noted that the fuel elements for the Dow TRICA reactor are irradiated elements used in a C. A. reactor core prior l

to initial loading at Dow.) A jig is currently being fabricated which l

will allow physical measurements of the element to be confirmed.

Dr.

i Anders stated that if the fuel is bowed in excess of' fabrication toler-ances it would be replaced. The fuci element is located in core position F-10 There is ao interference with element insertion or removal from i

the core.

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Dr. Beamer informed the inspector that there have been no in-dications of fuel failures since the original defective element was discovered during the initial startup program.2/ Pool water samples are taken periodically and have not indicated activities significantly above background.

mesmes P.

Radiation Detection i

Radiation experience with the licensee was reviewed with Dr. Beamer and Mr. Silverstein. To d' ate, they have stated that there have been no problems with contamination control or irradiated sampling handling procedures. The following significant points of discussions and ob-servation are summarized-r i

1.

Personnel Exposures Reports from the Landauer company show that the exposures received during cilendar year 1968 are as follows:

0-50 mrem Less than 10 mrem Total 3

8 11 2.

Radiation levels With the exception of the area immediately above the pool water, which is measured to be approximately 10 mr/hr at full power, other reactor areas exhibit minimal radiation level intensities.

3.

Several alarm trips on the continuous air monitor noted in the reactor log were the result of instrument malfunction.

Maintenance records reviewed revealed that this problem was corrected.

4 Calibration and testing of radiation detection instrumentation and alaren was noted to have been completed in accordance with technical specification requirements.

l S.

Experiments and Tests I

Principal experimentation continues to be the irradiation of small samples for tracer production. The range of experiment worth has been l

considerably less than the technical specifications maximum of $2.00.

One experiment was conducted to determine the cause of small power level changes noted to be associated with an empty fuel channel. Based on a series of checks which included consideration of fuel element movement, fuel expansion / cont raction etc., it was found that a phenomenon termed " fluidics" by Dr. Anders was occurring.

S CO Report No. 50 264/68-1.

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. According to Dr. Anders, operation with an empty channel sets up a condition of alternating convective thermal driving forces.

This causes a cyclic variation in the amount of cooling water flowing up through the channel. The reactivity changes are due to channel coolant temperature changes which result from the convective driving head changes. The interactions produced the observed power level

-- g changes independent of C. R. motion.

Rotating the coolant reentry nozzle 600 or shutting the recirculation pumps of f eliminates the condition.

T.

Facility Modifications No f acility modifications nave been made.

V.

Reliability Information 1.

Control rod scram times - Refer Section G.

2 Power level channel calibration - Refer Section G.

3.

Safety circuit interlock and functional testing - Refer Section G.

4 Radiation detection alarm testing - Refer Section P-5.

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ogn, THE DOW CHEMICAL COMPANY M I D L A N D. M I C H I G'A N 48840 Y\\

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vu-December 30, 1971

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Mr. Donald J. Skovholt N.i - -j Assistant Director for Z:'-

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Reactor Licensing

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U. S. Atomic Energy Commission d% MQi Washington, D. C.

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Dear Mr. Skovholt:

This letter is in response to your letter dated October 5, 1971, requesting detailed information concerning radio-activity effluents from our reactor.

The Dow TRIGA Reactor, AEC license No. R-108, Docket No. 50264, has been in operation since July 6, 1967.

It is used primarily for activation analysis studies.

The radioactivity released to unrestricted areas on an annual basis results from the disposal of spent activated samples and the release of argon-41 stemming from the air contained in the " sample" rack and the pneumatic sample transfer system.

Most of the activated samples contain only short-lived isotopes.

In 1970 and 1971, a minimum hold-up time of seventeen days was maintained between the time of activation and the time of disposal.

From this procedure less than 0.0006 curies were released to unrestricted areas annually.

The production of argon-41 accounted for an additional release of less than 0.0004 curies to unrestricted areas ennually.

It has been the practice during this period to incinerate the radioactive waste in Dow's main incinerator.

This unit is operated at approximately 1100*C with an air flow of 50,000 cubic feet per minute.

The effluent temperature is approximately 200*C, and the effluent stack height is 200 feet.

The incinerator is equipped with a triple spray wash system.

The effluent water goes to Dow's water waste treatment facility.

The amount of water effluent from this treatment facility is approximately 50 million gallons per day.

We assume that the 0.0006 curies of activated samples, when burned, were all released in the gaseous effluent from the incinerator.

For this case, the concentration at the

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-D. J. Skovholt December 30, 1971 Several monitoring systems are used to determine radiation exposures to both staff and non-staff.

In the reactor room itself, are located two area monitors.

One is a 4

Nuclear Measurement Corporation continuous air monitor which pumps air into a chamber (~2 liter capacity) and thrcugh a filter paper.

Approximately, 1/4 inch from the filter paper is a thin end window Geiger-Mueller tube, shielded by about 2 inches of lead.

The background of this monitor, stemming from radon daughters, is 1.7 x 10-11 p Ci/cc i

of air.

The second area m6nitor is a G-M tube located 13 feet from the top of the reactor.

It is connected to 4

a ratemeter and is sensitive to approximately 0.1 mr/hr.

Located in the water treatment system of the reactor is another G-M tube monitor, which continuously monitors the l

pool water and has a sensitivity of approximately 0.01 u Ci/ml of water.

All water in contact with the reactor core is cleaned up by means of demineralizers and recycled.

The demineraliser bed is periodically changed, with the old one being first monitor.ed and then incinerated.

Its release of isotopes is included in the above calculations.

In addition to these area monitors, several " laboratory I

monitors" (mica-window G-M tubes connected to ratemeters) are continuously operated within the laboratory building housing the reactor.

One of these is operated in the hood to which is connected the exhaust from the pneumatic sample transfer system of tae reactor.

The others are operated i

in adjacent laboratories.

These monitors have a background of approximately 100 counts per minute.

Many other portable, standard, health physics monitoring instruments are used in the building for routine surveys.

These instruments are used for alpha, beta, gamma, and neutron radiation measurements.

Several multi-channel analyzers, liquid scintillation, and proportional counters are also i;

used for evaluating gaseous, liquid, and solid samples.

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The lower detection limit of these instruments are generally h

1 x 10-7 u Ci.

Film badges for-beta, gamma, and neutron dose measurements

'I are used to evaluate radiation exposures of individuals i'

present in the facility.

The film badges are supplied by R. S. Landauer, Jr. and Company and are sensitive to 10 milli-rems of X-and gamma radiation, 40 milli-rems of hard beta, 20 milli-rems fast neutron, or 10 milli-rems thermal h

neutron.

Pocket dosimeters are also used to evaluate radia-tion exposures, and are sensitive to approximately 2 mr of X-or gamma radiation.

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D. J. Skovholt December 30, 1971 point of release to an unrestricted area (the top of the stack) would be approximately 0 x 10-13 C1/cc of air avr. raced over 1 year.

Incincrations have been carried out only during favorabic weather conditions to provide for maximum dispersion.

If we assume, on the other hand, that all the activity were washed down and released in the water effluent from the plant, the averaged concentration would have been 9 x 10-12 u C1/cc of water at the point of discharge.

The isotopes released in this manner consisted of the following:

Isotone Amount (C1)

Br - 82

<0.000010 Na - 24

<0.000001 Cr - 51

<0.000025 Fe - 59

<0.000070 Sb -124

<0.000060 P

- 32

<0.000250 Sc - 46

<0.000050 Zn - 65

<0.000005 l

Co - 66

<0.000003 Hg -203

<0.000010 Ag -110m

<0.000020 Ba -131

<0.000005 Cd -115

<0.000025 S

- 35

<0.000010 Se - 75

<0.000010 Ni - 63

<0.000001 Total Mixed

<0.000555 The argon-41 is released through a vent to the outside of the reactor building.

The air flow through this vent i

is 10C0 cfm.

The concentration of the argon-41 at the point of release is then less than 3 x 10-11 p Ci/cc of l

air.

The direct radiation levels to the unrestricted area from this facility is undetectable (less than 0.01 mr/hr and less than 10 milli-rem / month).

The direct radiation level i

'from the facility effluents is also undetectable (<1 x 10-14 p C1/cc of air of long lived isotopes, < 2 x 10-11 i

p Ci/cc of air of short lived isotopes, and < 1 x 10-3 p C1/ml of water for all isotopes).

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D. J. Skovholt December 30, 1971 1

In addition, air samples are taken on top of nearby buildings on filters which are counted by gamma spectro-scopy and by beta spectroscopy.

There are ten sampling locations in the reactor room, th' i

building housing the reactor, and adjacent buildings.

The personnel film badges, changed once per month, have l

indicated that personnel are exposed to less than 100 milli-rem per year of X, gamma, or beta radiation at our l

facility, and to less than detectable amounts of neutron radiation.

Area film badges have indicated less than detectable amounts of all radiation.

Air samples in the reactor room have indicated that airborne level of the monitor (1.7 x 10concentrationsofradioactivity{obeatthebackground 1

p Ci/cc of air) except during periods immediately following atmospheric testing of nuclear weapons.

This has been confirmed by gamma spectroscopy and other routine measurements around the reactor.

Air samples taken daily on nearby buildings indicated no release of activity directly attributab' to the operation of the reactor.

Some activity (1 x 10-1 Ci/cc) of long lived isotopes, mainly radium, thorium, and their decay p

products, with some fission products following atmospheric b

i testing of nuclear devices were, however, identified in l

these samples.

We hope that the above information will suffice your need.

In case additional discussion of any of the above items appears desirable, please do not hesitate to contact us again.

Very truly yours,

!!uxell &

H#roldHoylelNhairman Raciation Safety Committee 1701 Building HRH:dda 6

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