ML12025A089

Dow Chemical Company, Transmittal of Dtrr Revised Responses to RAI Questions 52, 54, and 56 in Support of the License Renewal

ML12025A089
Person / Time
Site:	Dow Chemical Company
Issue date:	01/20/2012
From:	O'Connor P Dow Chemical Co
To:	Geoffrey Wertz Division of Policy and Rulemaking
References
Download: ML12025A089 (9)

Text

The Dow Chemical Company Midland, Michigan 48667 January 20, 2012 Mr. Geoffrey Wertz Research and Test Reactors Licensing Branch Division of Policy and Rulemaking Office of Nuclear Reactor Regulation

Subject:

The Dow Chemical Company- License No. R-108; Docket No. 50-264 Enclosed are the DTRR Revised responses to RAI questions 52, 54, and 56 in support of the license renewal. They are attachment 1, attachment 2, and attachment 3 respectively.

Attachment 1 is being submitted with designation

'Withhold from public disclosure under 10 CFR 2.390" as it contains security related information.

Should you have any questions or need additional information, please contact the Facility Director, Paul O'Connor, at 989-638-6185.

I declare under penalty of perjury that the foregoing is true and correct.Executed on January 20, 2012 Paul O'Connor, Ph.D.Director Dow TRIGA Research Reactor Subscribed and sworn to before me this ,=-2 __day of January, 2012 No ary ubiMichigan My Commission Expires: NOAYJudy Lynn Raymond ..NOTARY PUBUC, MIDLAND COUNTY, MICHIGAN WMY COMMISSION EXPIRES MAY 13, 2012 cc: Wayde Konze, R&D Director -Analytical Sciences Paul O'Connor, Director Siaka Yusuf, Reactor Supervisor AJo;o Attachment 2 RAI 54 and DTRR Response I

54. NUREG-1537, Part 1, Section 13.1.3, "Loss of Coolant" requests the applicant to provide analysis that assures that doses to the public that could result from a loss of coolant accident do not exceed 10 CFR Part 20 limits. DTRR SAR, Section M. 1.1, Table 7 presents exposures resulting from a loss of coolant accident.

There is no statement regarding occupational or public dose limits and whether they are met. Please explain this accident analysis in further detail and in terms of meeting the regulatory limits.DTRR response: The water level in the surrounding area is above the core height and therefore tank breach will not result in a total loss of coolant. However, a site-specific analysis was completed for an uncovered core after several hours of operation at 300 kW and reported in the SER for DTRR, U.S. Nuclear Regulatory Commission, 1989). The results from this analysis are in the following table: Time after complete Direct radiation

-18 ft Indirect Radiation loss of coolant directly above core shield top edge of the (R/hr) tank (R/hr) (Position 1 or 4 in Figure 54-1)10 seconds 3000 0.78 1 day 360 0.090 1 week 130 0.042 1 month 35 0.012 Exposures inside the Reactor Room The elevated radiation fields generated from this hypothetical accident will be highly collimated above the reactor pool. The core sits inside a 17" diameter opening in the reflector.

The top of the reflector is 16' below the top of the 76" diameter reactor pool. The reactor room roof is 12'above the top of the reactor pool. Based on the maximum scattering angle that direct radiation may be emitted from the reactor (from the far left side of the core to the far right side of the reactor pool, and vice-versa), the direct radiation beam from the core will only have a diameter of 12.3 feet at the roof of the reactor room, which is still much smaller than the entire room.Therefore, workers and members of the public located outside of the reactor room will not be exposed to the direct radiation from the reactor core. Responders to the incident will also avoid the area of the room directly above the core in order to avoid exposure to the direct radiation from the reactor core.Exposures outside the Reactor Building To estimate potential radiation exposure levels from scattered radiation outside of the reactor room (indicated as position 2 in Figure 54-1, which is outside of the reactor building) measured radiation scatter data from dose rates during the operation of the neutron beam tube will be used 2 to estimate potential dose rates generated around the reactor building.

Specifically, during an operation of the central beam tube in 1991, measurement surveys were made and reported to the Radiation Safety Committee, (The Dow Chemical Company, 1991). This report documented measured gamma and neutron doses during the operation of a neutron beam tube that consisted of a 1.5" streaming pathway that allowed collimated neutrons to travel through a helium filled aluminum pipe without being shielded.

Figure 54-1 shows the layout of the reactor pool and the areas surrounding the reactor room.The highest total gamma and neutron dose rate measured during this survey was 4.2 mrem/hr and was located on the east side of the building (marked as location #2 in the drawing).

During this survey, a measurement of 5.9 mrem/hr was made also on the east side of the reactor but outside of the intense direct radiation field (marked as location #1 in the drawing).

This is proportional to the 1991 predicted radiation field of 780 mrem/hr (a factor of 132), immediately after an incident.

Therefore, using this factor, the 4.2 mrem/hr measured outside of the reactor building will be equivalent to 554 mrem/hr with the reactor pool fully empty. This number represents the highest expected dose rate, outside the reactor room, from a totally exposed core. Note that this takes no credit for the decay of the fission products during the time that it would take to drain the reactor pool.Reactor room radiation alarms are monitored by Dow Security.

In the event of an alarm, Dow Security would immediately respond and clear the area around the reactor of personnel.

Area radiation alarms are set at 2 mrem/hr, which would be reached if the reactor pool was drained to approximately 75% of its full volume. This response would occur within 30 minutes. Assuming that the water is pumped out of the pool at a rate of 10 gal/min (capacity of the pump that is stored in the reactor room), the water remaining in the pool (>2500 gallons) will not be able to be completely emptied for over 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. At this rate of removal and conservatively assuming a linear increase in dose rate as water levels decline, the dose rate outside the building will be less than 100 mrem/hr at a time of 30 minutes after the alarms go off. Therefore, the highest potential dose to a member of the public from this incident would be a Total Effective Dose Equivalent of 50 mrem, assuming an individual was located immediately outside the emergency door on the east side of the reactor for the entire duration of the incident until they were cleared from the area'by security.D (mrem) : 100 mrem/hr

• 0.5 hr =50 mrem This is the maximum conservative estimate of the public dose during a hypothetical loss of coolant accident.The Dow Chemical Company operates an on-site fire department, which would respond to this incident and be able to refill the reactor pool within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of the incident occurring.

Assuming that this response requires an individual to be located in the reactor room for 30 minutes and in the control room for the remaining 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />, a total employee Total Effective Dose Equivalent of less than 565 mrem would be received by an employee due to this incident.3 D (mrem) = (780 mrem/hr

• 0.5 hr) + (50 mrem/hr

• 3.5 hr) = 565 mrem This is the maximum conservative estimate of the occupational dose during a hypothetical loss of coolant accident.It is important to note that location 3 shown in figure 54-1 and whose dose rate is addressed above, is the only occupied location just outside of the reactor room.4 0v w Om0 0 N Note: Drawing not to scale Attachment 3 RAI 56 and DTRR Response I

56. NUREG-1537, Part 1, Section 13.1.6, "Experiment Malfunction" requests the applicant to provide analysis of an experiment malfunction event. DTRR SAR, Section M. 1.4, does not include analysis of an experiment failure with release of radioactivity.

Please provide an analysis and consequences of an experiment malfunction for the experiment with the highest potential release of radioactivity.

DTRR response: All experiments are reviewed before insertion and all experiments are separated from the fuel cladding by at least one barrier for example, the pneumatic tube and central thimble. All experiments that could damage components of the reactor are required by technical specification to be double encapsulated.

Samples are typically under 8 grams, with a majority of the samples irradiated consisting of carbon, hydrogen and oxygen (plastic and organics).

Consider a fueled experiment which has the potential to release significant amount of fission product. Such an experiment involving fuel element is the limiting experiment from a radioactivity release scenario at the DTRR. The dose consequences of the release of 10 microCi of 1-131-1-135 from such experiment are calculated for workers assuming that 100%of the material is released into the reactor room, and the ventilation system is shut off, causing the material to be trapped within the rector room. It is assumed that the worker spends 60 minutes within the reactor room to resolve the incident.

The release of the iodine would generate a concentration of 7.69x10-8 Ci/m3 inside the reactor room. Worker doses are calculated using Dose Conversion Factors for effective dose for 1-131 (conservative for iodine radionuclides) from Federal Guidance Report #11 (Eckerman, et al. 1988). This calculation will bound the dose to any member of the public who is located within the building and any exposure estimates to workers located within laboratories adjacent to the reactor room.The total effective dose to the worker is calculated to be 3.04 mrem.Note that the exposure scenario for a ruptured fuel element, in response to RAI 52, which is the analysis for a fueled experiment, is the limiting experiment from a radioactivity release scenario.

From that analysis, the 1-131 contributed 42% of the total dose for the scenario.Since this accident scenario would have a similar mix of radionuclides, it is not anticipated that contributions from additional radionuclides would increase this dose estimate by more than a factor of 3, which would keep exposure estimates from this scenario well below the dose estimates for the fuel rupture accident scenario and 10 CFR Part 20 exposure limits.For exposures to members of the public, it is assumed that the ventilation system is operational and vents the released iodine outside the reactor building.

Based on the ventilation rate and volume of the reactor room, it would take 2.6 minutes to have one full air change of the reactor room and release all of the iodine, which results in a release rate of 6.41 x 10-8 Ci/sec from the facility.

Downwind air concentrations at the plant fence-line located 23 m to the west of the reactor building are determined following guidance in Regulatory Guide 1.145 (U.S. Nuclear Regulatory Commission, 1982) to be 6.80 x 10-9 Ci/m 3.Doses to members of 2 the public are calculated using Dose Conversion Factors for effective dose and dose to the thyroid for 1-131 (conservative for iodine radionuclides) from Federal Guidance Report #11 (Eckerman, et al. 1988).The total effective dose equivalent to the maximally exposed offsite member of the public is calculated to be 1.4 mrem.References U.S. Environmental Protection Agency. 1988. Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion.

Federal Guidance Report No. 11. Washington, D.C.: U.S. Environmental Protection Agency.The Dow Chemical Company. 1991. Radiation Dose and Exposure Rate Evaluations During Neutron Radiographic Operation of the Dow TRIGA* Research Reactor at 100 and 240 Kilowatts, Special Analysis, Michigan Division Analytical Laboratory, 1602 Building, November 19, 1991. HEH RAD14(8).U.S. Nuclear Regulatory Commission.

1983. Atmospheric Dispersion Models for Potential Accident Consequence Assessments at Nuclear Power Plants. Regulatory Guide 1.145.Washington, DC: U. S. Nuclear Regulatory Commission.

U.S. Nuclear Regulatory Commission.

1989. Safety Evaluation Report related to the renewal of the facility license for the research reactor at the Dow Chemical Company. NUREG-1312.

Washington, DC: U.S. Nuclear Regulatory Commission.

U.S. Nuclear Regulatory Commission.

Regulatory Guide 1.111, "Methods for Estimating Atmospheric Transport and Dispersion of Gaseous Effluents in Routine Releases from Light-Water-Cooled Reactors," U.S. Nuclear Regulatory Commission, Washington, DC.3