ML20238A309

Forwards Answers to Questions Posed Following Submittal of May 1987 Formal Answers,Including More Detailed Info Re fire-fighting Equipment in Reactor Room & Nearby Areas & Detailed Analysis of Temp Rise Expected in Fuel Element

ML20238A309
Person / Time
Site:	Dow Chemical Company
Issue date:	08/13/1987
From:	Kocher C DOW CHEMICAL CO.
To:	Alexander Adams NRC
References
NUDOCS 8708200441
Download: ML20238A309 (10)

Text

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DOW CHEMICAL U.S.A. l l

August'13, 1987 MICHIGAN DIVISION

' MIOLAND, MICHIGAN 48640 Alexander Adams Project / Manager Office of Nuclear. Reactor Regulation U. S. Nuclear Regulatory Commission Washington, DC 20555 DOCKET NUMBER 50-264 Sir: ,

1 Enclosed are answers to the questions you posed following submittal of formal answers in May 1987. If you have any further questions please call me.

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i ~C. W. Kocher Reactor Supervisor 1602 Building f

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DOW TRIGA RESEARCH REACTOR i

LICENSE R-108-l DOCKET NO. 50-264 ,

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SUPPLEMENTARY INFORMATION

.10 AUGUST 1987 i i

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1. Please provide an ALARA policy statement signed by a corporate officer or some similar person.

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MICHIGAN OMSloN

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ALARA - AS LOW AS REASONABLY ACHIEVABLE - POLICY STATEMENT l

The principle of ALARA - As Low As Reasonably Achievable - forms the basin of the radiation protection program of Dow Chemical U.S.A.

Close adherence to the principle of ALARA is of paramount importance  !

to the achievement of the Dow goal of minimizing occupational exposures to radiation and releases of radioactive materials.

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I. G. Snyderj ,Jrr  :

l Vice President i gj j Director of Applied Research and Development ,

Dow Chemical U.S.A.  !

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AN OPERATING UNIT OP THE DoW CHEMICAL COMPANY -

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2. Please provide calculations of the possible radiation levels in

'the area of the water purification skid if 250 gallons of water at some known concentration of radioactive materials n re to escape from the reactor primary system and collect ot,the floor (question 3 of the May 1987 submittal).

Consider the release of 250 gallons of water from the reactor into a 10-foot by 50-foot area in the basement of 1602 Building following a Maximum Hypothetical Accident. (MHA) . Assume that the entire release fraction (10-4) of the elements given in table 4 of ref. 1 (Gaseous Fission Product Activity in the TRIGA Element Containing the Greatest Activity Following Operation at 365 MWD), modified by factors of 50/r8 ,

'because of the number of elements in the Dow reactor and by 300/1000 '

because of the power level of the Dow TRIGA reactor. Assume that none of this material is released from the water before the water is leaked into the basement of 1602 Building. We can then calculate an areal concentration of each of the isotopes given in table 4, reference 1, where the total concentration of radioisotopes.is 0.165 pCi/ml and the areal concentration is then 0.36 poi /cm 2. Then assume that we have an infinite plane with each of those concentrations, neglect the absorption of gamma radiation in the water (about 2.16 cm thick) and calculate the dose rate at a position 1 meter above that plane, using the energies and population factors of each of the gamma rays. Use i Equation I-4.1, with h=0 (infinite-plane source), given on page 353 of reference 2, and the appropriate graphs in the same reference for the evaluation of the integrals involved, to find the gamma-ray fluxes for  !

each isotope at the desired point, and use the data of reference 3 to calculate the dose rates due to each gamma-ray of each isotope.

Under these conditions the dose rate at a point one meter above this infinite plane would be 110 mR/hr. The actual dose rate near the i postulated spill would be lower than the calculated dose rate due to l the conservative assumptions.  ;

References:

1. S. C. Hawley and R. L. Kathren, Credible Accident Analyses for TRIGA and TRICA-Fueled Reactors, NURE0/CR-2387, PNL-4028 (1982)  ;

2. T. H. Rockwell III, Editor, Reactor Shielding Calculations, TID- l 7004, Naval Reactor Branch, USAEO, (1956)

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3. H. E. Johns and J. R. Cunningham, Physics of Radiology, 4th Edition, (1983) i i

3. Please provide more detail about the fire-fighting equipment in the reactor room and nearby areas (question 6 of the May 1987 submittal).

One 10-pound C02 fire extinguisher is located about 12 feet from the console in Laboratory 30 'and another is located in contiguous Lab 31. -

Other fire extinguishers are located in adjacent areas.

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4. Please review Reg Guide 8.10 and provide more comments on the ALARA program at the Dow TRIGA Research Reactor f acility (question i 8 of the May 1987 submittal). 1 l

The Dow Chemical Company ALARA policy is expressed in the Policy l Statement.(question 1, above). This statement has been circulated to I the reactor staff and is posted in the facility. Members of the reactor staff receive annual training in radiation protection and the  !

- implications of the ALARA policy. a l

l The Reactor Operations Committee takes a proactive stance with respect  ;

to the evaluation of the facility through the required audits (see the Technical Specifications) under the direction of the Radiation Safety Officer, who is administratively separated from the facility staff. l 1

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E 5. Please provide a copy of the diagram showing the relationship-between the reactor room and areas of unrestricted access, adding scaling information and indicating the points of release (question

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13 of the May 1987 submittal).

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6. Please previde a detailed analysis of the temperature rise expected .n the aluminum-clad fuel element to reactivity insertior s of 0.0216k/k ($3.00) (question 19 of the May 1987 submittal') .

Inadvertent tr e ients have been analyzed in the report of Hawley and l Kathryn (S. C. dawley and R. L. Kathryn, credible Accident Analyses 1 for TRIGA and TRIGA-Fueled Reactors, NUREG/CR-2387, PNL-4028 (1982)), q based upon the detailed analyses reported by West et al (G. B. West et '

al, Kinet3.c Behavior of TRIGA Reactors, GA-7882, General Atomic, San Diego, California (1967)). West et al used a Fuchs-Nordheim point kinetics model, considering both constant and temperature-dependent heat capacities of the fuel elements. Hawley and Kathryn used a simpler equation which assumes a constant heat capacity (pp 18,19; Table 3, Reactivity-Temperature Relationships), since this method predicts a greater mean temperature rise for a given reactivity insertion and is thus more conservative. Hawley and Kathryn calculated (Table 3) a mean temperature rise of 300 0 for a core of aluminum-clad fuel elements with H:Zr ratio of 1.1, a a prompt negative temperature coefficient (a) of 10-4,geff of 0.0073, and insertion of 0.0225 6k/k (S3.08) reactivity. Using their equation AT = ((6k/k)(1-perf)-peff)*2/a for the Dow TRIGA Research Reactor where the value of perf is 0.0070 (due to the stainless-steel-clad fuel elements), the value of the prompt negative temperature coefficient a is 1.2*10-4 , and the maximum available excess reactivity will be .021 6k/k, we calculate the corresponding temperature increase to be AT = 231 0 and if the initial temperature of the fuel is about 30 0 then the i final mean temperature of the fuel is expected to approach 261 0, well below the proposed Safety Limit (500 0) and even more below the temperature associated with phase changes expected in the H:Zr 1:1 aluminum-clad fuel element (535 C).

In the Advanced TRIGA Prototype Reactor (ATPR) used by West et al the measured temperatures were peaked in the B-ring and were found to be considerably lower than calculated (Table VIII, page 33), with a peak t fuel temperature *of 405 0 (measured) for a $3.00 reactivity pulse, i The aluminum-clad fuel element in the Dow TRIGA Reses.rch Reactor will l be positioned in either the E- or the F-ring, far from the peak power  ;

area of the B-ring, in order to be even more conservative with respect i to the operation of this one fuel element. I

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