ML20136H649

Responds to 851011 Request for Addl Info Re L-77 Research Reactor Decommissioning Plan

ML20136H649
Person / Time
Site:	05000433
Issue date:	11/20/1985
From:	Kroes R CALIFORNIA, UNIV. OF, SANTA BARBARA, CA
To:	Thomas C Office of Nuclear Reactor Regulation
References
NUDOCS 8511250160
Download: ML20136H649 (9)

Text

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. UNI ERSITY OF CALIFORNIA. SANTA BARBARA nEnEEs.Ev. Davis any NE. Los ANGELES

• RIVERSIDE

• SAN DIECO

• SAN FRANCISCO SANTA BARBARA

• SANTA CRtJZ DAVID PIERPONT GARDNER Phendent q/ the Unitureity OFFICE OF THE CHANCEL 1DR SANTA BARBARA, CALIFORNIA 98106 RosEnt A. HurrENaACK Chancellor at Santa Barbara November 20, 1985 Cecil 0. TLomas, Chief Standardization and Special Projects Branch Division of Licensing, NRR U.S. Nuc1 car Regulatory Commission Washington, DC 20555 Re:

Docket No. 50-433

Dear Sir:

Enclosed are twenty-one copies of the reply to the Request of Additional Information dated October 11, 1985, relating to the decommissioning of the University of California, Santa Barbara, L-77 Research Reactor Decommissioning Plan. We hope that this information is sufficient for your review and that we can soon proceed with the decommissioning.

Sincerely, Robert J. Kroes Vice Chancellor Administrative Services Attachments cc: Director of Licensing US NRC Region V 1450 Maria Lane, Suite 210 Walnut Creek, CA 94596 Tru2"ioba Mh!!h P

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i Reply to Request. f or Addi tional Information on L-77 Decommissioning Plan, October 11, 1985 t

l Specific Comments:

(1) Page.1--1st' paragraph. The inner shield tank is composed of=three layers, as shown in Fig.1. The innermost layer, which also serves as a. neutron-reflector, is-composed of 60 vol'. % lead and 40 vol. % diphenyl in an aluminum cyl inder.

The second l ayer is composed of paraffin wax loaded with 10 g/l boron, and encased in an aluminum cylinder. The ou termost l ayer. i s 60% l ead, 40% borated paraf f i n, In a stainless steel cylinder 43 in. 0.D.

The cylinder is welded but containment of the fuel and fission products does not depend on the shielding.

The stainless steel core vessel itself i s. sealed and under vacuum.

The fuel solution is

' introduced or removed through a line which emerges from the bottom of.the core vessel, runs up through the inner shielding to a needle valve, and then to a capped port on the side of the outer shleid tank. A vent or gas-line runs from the top of the core vessel, through the shielding to another needle valve, and then to a capped port. The valves are closed except durin'g the semiann'ual loss-of-vacuum test,

.and the negative pressure in the core vessel is maintained

-from one test to the next.

Thus in this sense the core

. vessel is hermetically sealed.

(2) Page 2-1st paragraph. Fast neutrons are indeed emitted from the reactor into the shield surrounding it', and some activation will have occured.

However, the materials of con'struction, including borated paraffin, were designed to minimize activation outside the reflector region.

The reactor =assembi,y refers to~the core vessel plus shleid.

In

. fact, measurements of the shield water indicate no contamination,'and the radiation levels with the reactor shutdown indicate that any activation of the outer carbon steel reactor shield tank is at backgtound.

The inner shield and core vessel will be disposed of as radioactive waste.

The shield water will be drained after confirming it is not radioactive, and the carbon steel tank will be cut up and disposed of as nonradioactive scrap after checking for activity.

(3) Page 3-Section 1.2.

The reactor was used in a L

laboratory course taught once a year, for one afternoon a week for 5 weeks using the reactor, and 5 weeks using other i

facilities.

Experiments were performed on approach to cr i t:i c al and subcri tical mul tiplication, control rod

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-calibration, and intermittent operation at low power in which no experimental facilties were needed. Just the neutron flux sensing instrumentation. One experiment was performed on the reactivity effect of plastic-and cadmium samples in the central irradiation through tube, with the shield-plugs in place after loading the sample. These samples do not activate much and will have decayed to background level. The f i ssi on power was cal i brated annual l y by irradiating two sets of gold foils, one bare and the other-cadmium covered. The irradiation of each set was at 1 watt for 20-30 minutes. The gold foils will have decayed to background level by decommissioning. No other irradiation ports, or beam ports, were used. The reactor is not equipped wi th a rabbi t. As explained in (2) above, the carbon steel outer shield tank is not appreciably activated. A few activation analysis experiments were attempted in the 1970's, but the l ow neu tron fl ux precl uded si gn i f i can t use.

At 10 watts, the average thermal neutron flux in the core is

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only 2 x 108 n/cm2s, decreasing to less than 10*. of this

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.value at the outside of the diphenyl-l ead refl ec. tor (and innermost shi eld layer). The removable shield plug for the exper ime n t al port is not significantly radioactive. The reactor was seldom operated near 10 watts (usually Just for the radiation level surveillance required by the Technical Specifications). Integrated power for operations in the last 10 years or so is only some 360 watt-hours. Thus we expect that no significant amounts of long-lived radioisotopes were produced. But in any case, the core vessel and inner shield assembly will be treated as radioactive waste.

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(4) Page 3-last paragraph.

At 10'W-(th) thi maximum thermal neutron flux is 3 x 108 n/cm25. The fast neutron flux is about the same. The gamma ray dose. rate in the core is 3 x 103 R/hr. At the surface of the outer shleid tank, the thermal neutron flux is about 40 n/cm2s, the fast neutron flux is about 10 n/cm2s, and the gamma-ray dose rate is about 5 mR/hr. These dose' rates decrease rapidly with distance from the tank, so the effective dose rate (neutron plus gamma ray) in the occupied area is less than 2 mrem /hr.

The reactor is operated for less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> a week even during the lab course period. Thus no significant exposures have been received by staff or students, confirmed by TLD badges, which staff and students wear.

Radiation levels at the position of the area radiation monitors are between 0.1 mR/h and 2 mR/h at 10 watts. The question of activation of the reactor system has been discussed already.

(5) Page 4-top of page.

Tritium contamination does not exist in the reactor restricted area itself, but near an accelerator in the south end of Room 1251. The contamination is small, and occured because of slight leakage of coolant l

3 water from a tritiated target assembly used in the accelerator to generate 14 MeV neutrons by the D,T reaction.

The accelerator is being disposed of as radioactive waste, and the area will be decontaminated by usual wash i ng

. techniques and rechecked. This work is expected to be completed in the next few months, and will not affect the decommissioning of the reactor.

(6) Page 4-Section 1.3.

The entire facility (Broida Room 1251) was resurveyed by Environmental Health and Safety on 4 November 1985 for tritium contamination, using swipes and a Beckman LS-233 liquid scintillation counter.

Removable contamination was less than the sensitivity limit of the counter, corresponding to less than 30 pCi per 100 sq. cm.

Some equipment outside the reactor restricted area was found to be slightly contaminated (80 pCi per sq. cm.)

The equipment will be decontaminated or disposed of as low level radwaste.

This work is currently in progress.

The 0.2 mR/h levels are estimated as an upper limit at the inner shield surface from residual activity in the core vessel and perhaps the inner shleid (the fuel solution will have been drai ned already). The estimated activity from cobalt-60 in the stainless steel core vessel is only 20 microcuries, and activity in the lead, diphenyl, paraffin, and aluminum sbleid is probably less. Thus 0.2 mR/h i s probabl y h i gh.

Even immediately after reactor shutdown, with the fuel and its fission products in the core, we measure less than 0.2 mR/h outside the outer shield tank.

(7) Page 13 - Section 3.2.

The defueling is estimated to require 1 to 2 weeks and vill be performed before the dismantling, of course. Disposal cf the shield'ater will w

t ak,e less than 1 day. Drying the core vessel may take 1 or 2 day ~5.

Disconnection of the control d dalves, thimbles, throughtubes, and fuel and vent lines will take about 3 days. Removal of the shield tank may require f or 2 days.

Packaging of the remaining reactor core vessel and inner t

shl ei d assembl y may take 2 days. Cleaning and termination radiation survey should not require more than 5 days. Thus it should be possible to complete the.dismantilng within 3 weeks as scheduled.

(8) Page 13-Section 3.3.

The control rods have been i

removed before, when the reactor was shipped to UCSB from the University of Nevada, Reno. They were not appr e c i abl y radioactive, and are not expected to be significantly radi oac t ive when di sman tl ed now. The source thimble is also not appreciably radioactive or contaminated.

The vent line should not be appreciably contaminated or radioactive. The fuel line is not expected to be radioactive where cut, and will*have been flushed with water and dried after removing l'

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the fuel solution. Thus normal procedures such as surveying for activity at'the point of cut, checking for contamination, wearing gloves and working on the plastic sheeting, will.be sufficient.

(9) Page 14 - Section 3.4.

At present, plans call for storage of the packaged reactor assembly (core vessel and inner shleid. assembly) in the reactor facility, Broida Hall Room 1251, until shipped to a radioactive waste disposal site. If it is necessary to store the package elsewhere (for example, because the radwaste disposal site will not accept California waste, and the facility is needed for other users), it will be stored in the existing hazardous materi al s hol ding f acil i ty, Building 596, under the General License for the campus from the State of California, an agreement state.

(10)'Page 19-Section 8.0.

Drawing for the Termination Radiation Survey is enclosed.

(11) Page 19, middle par.

It is very unlikely that any hot spots requiring chipping of concrete walls or floor will exist. (The reactor does not use a concrete shleid). No contamination above background has resulted from the operations thus far. Contamination of the vinyl tile or concrete floor during defueling will be avoided by use of plastic sheeting.

If concrete chipping must be done, a respirator mask can be used to keep airborne radiation exposure to workers ALARA, and plastic sheeting can be used to prevent spread of contamination.

(12) Page 20 - 2nd Par.

Note that it is 20 microcuries, not, 20 curies, of cobal t-60 activi ty estimated in the reactor core vessel. If the relative response for Cs-137 and Co-60 are not known, then a calibration with Co-60 will be made.

However, the difference in intrinsic efficiency for a 1 x1

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inch NaI(TI) sclntillator is only about 25% for undegraded Cs-137 and Co-60 spectra.

Since this particular meas.rement is made to determine the exposure from distribuited activity, the spectrum is quite degraded by scattering, and so the difference is considerably less than 25%.

General Comments:

(1) The reactor rbom is. equipped wi th a Nuclear Measurements Corp. beta gamma GM airborne radioactivity monitor (fixed filter paper). Radon background count rate varies from 100 cpm to 3000 cpm depending on weather conditions. Values above 5000 cpm are considered to be indicative of airborne activity from non-natural sources such as the reactor. The

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airborne radioactivity monitor will be operating during decontalmination, dismantling and decommissioning. A count rate above 5000 cpm will require cessation of operations until the cause is determined and corrective measures taken.

(2) As stated in the Environmental Report included with the Decommi ssi on i ng Pl an, section 2.0, the collective dose equivalent to workers is estimated at 0.045 person-rem above background.

(3) The fuel solution will have been removed before the d i sman t l i n g, decontamination, and decommissioning. The surveillance specifications in the curren t l i cense (R-124) technical specifications will continue in force until the fuel is removed from the reactor core. Afterward, it is neither possible nor necessary to maintain the surveillance of power, reactivity, reactor instrumentation, or radiation monitoring of radiation levels during reactor operation.

Therefore the only applicable surveillance specifications during the decommissioning procedure are to monitor for accidental criticality or excessive radiation level or contamination from the fuel. It is al so necessary to maintain security while the fuel is on site. After the fuel is shipped, the facility can be treated as any other l aboratory wi th some radi oac t i ve mater i al s, until all radioactive contamination and materials have been cleared.

Therefore the technical specifications that apply areb (a)

Current Technical Specifications under License R-124 until reactor is defueled.

(b) Surveillance of area and airborne radiation monitors, and periodic contamination survey, until fuel is shipped off s i t'e. Security maintained un: 41 fuel is shipped.

(c) Routine operation of area and airborne radioactivity monitors as needed during dismantling. Then removal from facility after reactor is dismantled and core-shield assembl y is packaged. Normal campus procedures for radi oac t i ve mater i al s f oll owed subsequen tl y, until radioactive material is removed from the room.

Reply to Additional Queries from Region V NRC:

(1) The analysis of the shielding water will be performed by EH&S.

The method is to evaporate to dryness and count the activity.

Analysis f or tritium will not be done, as the source of the tritium activity is the accelerator, not the reactor, and the shield water cannot become contaminated from: the accelerator.

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(2) Copies of all records will be available at the UCSB facility.

If any addi tional records are required from the contractor (Rocketdyne Division of Rockwell In terna t i onal ),

they can be obtained quickly as the contractor's place of business is in Canoga Park, California, about 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> drive by car.

(3) The Decommi ssi oni ng Pl an descr i bes what will be done.

It has been reviewed and approved by the UCSB Radiation Safety Committee, which is the committee responsible for compliance with the isotopes license.

The Pl an has al so

- been reviewed and apaproved by the Reactor Operations Committee, which i s the commi t tee responsi bl e for compliance with the reactor facility license.

The UCSB Environmental Heal th and saf ety group will also survey the facility before releasing it for unrestricted use by other UCSB personnel, in accordance with the EH&S establ i shed procedures. The reviews have been documented, and applicable records will be maintained at UCSB.

(4) N/A l

(5) The table in the decommissioning plan is compatible with L

Regulatory' Guide 1.86, for enriched uranium and mixed l

fission products, or low hazard beta and gamma' ray activation products.

Any con taminat ion shoul d f al l in these categories.

The table should have had a note explaining l

this.

The ambient exposure rate of 5 uR/h above background at 1 meter from any surface is acceptable and should be included in the Plan.

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l-(6) The' response of the NaI(TI) scintillator has been discussed under question (12) of Specific Comments, above.

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Accuracy of the measurement of exposure rate is adequate to determine compliance with 5 uR/h at 1 m, above background, espec i al l y as exposure rates are expected to be much lower than the limit.

l (7) N/A.

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