Text

Supplemental Information for Cladding Fuel Accident - Updated FSAR Section 13 for License Renewal Application for Facility Operating License R-120 for the North Carolina State University Pulstar Research Reactor
	ML22159A237
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Site:	North Carolina State University
Issue date:	06/07/2022
From:	Hawari A; North Carolina State University
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	EPID L-2017-RNW--0026
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Nuclear Reactor Program Campus Box 7909 NC STATE Department of Nuclear Engineering 2500 Stinson Drive Raleigh, NC 27695-7909 UNIVERSITY www.ne.ncsu.edu/nrp 919.515.3347 (voice) 919.513.1276 (fax)

June 7, 2022 U.S. Nuclear Regulatory Commission Document Control Desk Washington, DC SUBJECT SUPPLEMENTAL INFORMATION FOR CLADDING FUEL ACCIDENT-UPDATED FSAR SECTION 13 FOR LICENSE RENEWAL APPLICATION FOR FACILITY OPERATING LICENSE R-120 FOR THE NORTH CAROLINA STATE UNIVERSITY PULSTAR RESEARCH REACTOR (EPID NO. L-2017-RNW-0026)

License No. R-120 Docket No. 50-297 Please find enclosed an Updated Final Safety Analysis Report Section 13.2.6 for the License Renewal Application (EPID NO. L-2017-RNW-0026). The updated FSAR section incorporates discussions during the NRC Renewal Audit Meeting that took place in May 2022.

If you have any questions regarding this submittal or require additional information, please contact Colby Fleming at (919) 515-3347 or ncsorrel@ncsu.edu I declare under penalty of perjury that the forgoing is true and correct. Executed on 7 June 2022.

Sincerely,

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Ph Director, Nuclear Reactor Program North Carolina State University

Enclosures:

001 Supplement - NCSU FSAR Section13 - 7J UN2022

Supplement 1 - NCSU FSAR Section 13 - 7JUN2022 13.2.6 Mishandling or Malfunction of Fuel Events which could cause accidents of this type at the NCSU PULSTAR include:

1. Fuel handling accidents where an element is damaged severely enough to breach the cladding.

2. Failure of the fuel cladding due to manufacturing defect or corrosion.

13.2.6.1 Cladding Failure Accident In the cladding failure accident (CFA), submersion and internal dose to occupants inside and members of the public outside the reactor building are of concern. Credit is taken for the retention of many fission products and other radionuclides present in the primary coolant system. For the CFA, the fission product inventory in the NCSU PULSTAR reactor core is based on 6320 MW*days of operation at 2 MW (equivalent to the fuel burnup limit of 20,000 MWd/MTU).

The activities of the fission products which would be found in the fuel pin annulus gap between the zircaloy-2 cladding and the UO2 pellets are estimated using the data for commercial reactor fuel given in NUREG 1887[13-16] and NUREG/CR2507[13-17] by adjusting for power level and fuel temperature.

NUREG 1887 provides specific core isotopic inventory for commercial reactor fuel in Bq/MW.

Additional radionuclide inventory data given in IAEA Publication 53[13-18] for research reactors were also included in the CFA analysis for the NCSU PULSTAR reactor fuel.

The activity in the fuel gap is dependent on the type of fuel, fuel dimensions, fuel burnup, linear power generation, radionuclide half-life, and centerline fuel temperature. PULSTAR fuel and commercial power reactor fuel have similar dimensions, cladding, chemical form, and 235U enrichment. The major difference is the fuel temperature. Low temperature data on low enriched UO2 sintered pellets of low enrichment is provided in NUREG/CR2507. NUREG/CR2507 gives data for low temperature fuel and burnup similar to that for NCSU PULSTAR fuel. For this analysis, the gap release fraction is based on data and equation 14 given in Section III of NUREG/CR2507 for low temperature UO2 commercial fuel (density of 10.0 to 10.6 grams per cm3 with a surface area to volume value of 6 cm-1 and for a fission energy release of 200 MeV per fission):

= 1 x 107 0.5 + 1.6 x 1012 () Equation 13-5 where, is 2 0.316 = 6.25 / for the NCSU PULSTAR is the radioactive decay constant in s-1 For the CFA, it is assumed that 25 fuel pins at maximum burnup suffer a clad rupture and all the fission products contained in the annulus of the 25 fuel pins are released into the reactor pool with a minimum water depth of 14 feet. The fuel pins are well protected due to the design and construction of the fuel assembly, as described in Section 4. There is no credible scenario that could result in the failure of all 25 fuel pins, nevertheless, the analysis assumes that the damage occurs due to a handling error resulting in mechanical shock to the fuel pins.

All of the fission gases and 3 percent of the iodine and bromine isotopes are assumed to be released from the primary coolant and into the reactor bay air volume at the surface of the reactor pool. Fission gases decay to produce short-lived particulate radionuclides in the reactor air volume. Particulate fission products released to the reactor pool are retained in the reactor pool.

For the purpose of assessing the consequence of released fission products, it is assumed that all of the radioactive fission gases in the fuel gap for 25 fuel pins and 3 percent of the halogens activity in the fuel gap from 25 fuel pins escape into the reactor building air volume. The released activity is uniformly mixed in the free air volume over the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months given that the air exchange rate in confinement is 2.35 hours4.050926e-4 days 0.00972 hours 5.787037e-5 weeks 1.33175e-5 months . Assuming immediate release of the gap inventory as described accounts for all the radioactivity that could be released. Continuing releases of activity into the reactor building is not considered to occur since the fuel temperature is low, which limits diffusion from the fuel, and the reactor pool evaporation rate of approximately 1 gallon per hour is negligible relative to the reactor pool volume of approximately 14,000 gallons.

The anticipated sequence of events which would occur for a postulated 25 fuel pin failure is as follows:

1. Short-lived fission products are assumed to decay to negligible levels prior to reaching the fuel pin gap due to the transit time through the fuel pellet. Even though fuel movements occur after a decay period following reactor operation, no credit is taken for this decay prior to beginning fuel movement as this time is variable. The remaining fission products in the fuel pin gap escape into the reactor primary coolant from the cladding failure.

2. 100 percent of the fission gases and 3 percent of the iodine and bromine isotopes are assumed to be released from the primary coolant to the reactor building air volume at the reactor pool surface. Solubility reduces the iodine and bromine fission products available for escape into the reactor building air volume. Fission gases decay to produce short-lived particulate radionuclides in the reactor building air volume. Particulate fission products released to the reactor pool are retained in the reactor pool.

3. The radiation level increase in the stack gas or stack exhaust radiation monitors would cause the normal ventilation system to automatically place the reactor building into confinement and thus prevent release of unfiltered air. The Radiation Monitoring System alarm automatically initiates evacuation of the Reactor Building as well.

4. In confinement mode the normal ventilation fan is shut down and a 600 cfm confinement fan automatically starts and passes the reactor building air through a filtration system prior to its discharge from the 100-foot stack. The filtration system has a high efficiency particulate air (HEPA) filter with a removal efficiency of 99.97% for particulates and an activated charcoal filter with a removal efficiency of 99% for halogens. Prior to use, and as a routine surveillance after installation, the HEPA and charcoal filters are tested for leakage. Retention factors of 99.97 percent for particulates and 90 percent for halogens are used in the CFA analysis.

5. After passing through the confinement filters, and prior to discharge from the stack, 600 cfm flow would normally be diluted again with exhaust flow from the ventilation system serving the south wing of the Burlington Engineering Laboratories; however, no allowance for this additional dilution is assumed in this analysis.

6. The reactor building free air volume is 2.4x109 cm3 as indicated in Section 11. The fission product gases escaping from the pool are assumed to be uniformly distributed throughout the reactor building. Activity passes through the confinement filter system and is exhausted out the reactor stack at the rate of 600 cfm.

7. The fission product release scenario is calculated to last approximately 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with allowance for decay. Various offsite locations downwind of the stack would be exposed to the plume during this time. Therefore, an average release concentration for the entire 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is used with no allowance for the R-63 ventilation dilution. Since the stack exhaust exit is elevated and not accessible by the public, a dispersion factor for various offsite locations is calculated to quantify radiation dose to the public associated with the 25 fuel pin failure scenario.

The assumption of mixing in the reactor building air volume is conservative since no dilution is assumed for the 600 cfm of fresh air which must leak into the reactor building as makeup. Based on the reactor building volume of 2.4x109 cm3 and a purge rate of 600 cfm, this analysis makes the additional assumption that the entire fission product inventory is removed from the reactor building in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months corresponds to approximately 10 complete air volume exchange of the free air volume. An average concentration for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months was used in the dose calculations. Average concentration for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months is also used as an exposure time as 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allows for evacuation of nearby public areas. The average concentration is given by the following:

[1 ]

() = (0) = (0) 0 Equation 13-6

[1(0)1] 1 ] 2 ]

2() =

[2 1]

{ [1 1

[1 2

}

where, is the removal rate constant given by the sum of the air exchange rate in confinement and the radioactive decay constant, is time, or 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for this analysis (0) is the initial activity released from the reactor pool is the free air volume of 2.4x109 cm3 (0) is the initial concentration in the reactor building = (0)

() is the average concentration over time t 2() is the average concentration of decay product (2) from parent nuclide (1) over time t Table 13-14 provides results from the above discussion for CFA. The following notes apply to Table 13-14 and Table 13-15:

a. Linearly scaled to 20,000 MWd/MTU from 30,000 MWd/MTU as stated in NUREG 1887 [13-16] .

b. Scaled to 2 MW from previous FSAR value at 1 MW.

c. Particulates are produced by decay of parent fission gases in the reactor building air volume.

d. Taken from IAEA Publication 1308[13-18] and adjusted for 2 MW.

e. Gap release fraction calculated using Equation 13-5.

f. Average concentrations were calculated using Equation 13-6. Released average concentrations were adjusted for confinement filter retention.

Table 13 Reactor Core Inventory, Gap Inventory and Airborne Concentrations for a Cladding Failure Accident at 2 MW Reactor Reactor 2 MW Reactor Stack Stack 2 MW 2 MW Gap Bldg. Bldg.

Core Design Building 24h 2h Nuclide Core Inventory 24h 2h FHA C(0) C(ave) C(ave)

C(ave) C(ave)

Ci/MW Ci Gap Ci Fraction(e) µCi/ml µCi/ml(f) µCi/ml(f) µCi/ml(f) µCi/ml(f)

Kr83m(d) 4.49E+03 8.98E+03 3.57E-03 9.93E-06 1.49E-06 7.76E-08 7.76E-08 7.42E-07 7.42E-07 Kr85(a) 3.17E+02 6.34E+02 1.81E-01 7.15E-03 7.55E-05 7.41E-06 7.41E-06 5.09E-05 5.09E-05 Kr85m 8.00E+03 1.60E+04 9.91E-03 1.55E-05 4.13E-06 2.97E-07 2.97E-07 2.44E-06 2.44E-06 Kr87 1.60E+04 3.20E+04 1.05E-02 8.18E-06 4.36E-06 1.87E-07 1.87E-07 1.92E-06 1.92E-06 Kr88 2.30E+04 4.60E+04 2.26E-02 1.23E-05 9.42E-06 5.87E-07 5.87E-07 5.20E-06 5.20E-06 Kr89(d) 3.78E+04 7.56E+04 5.00E-03 1.65E-06 2.08E-06 6.37E-09 6.37E-09 7.65E-08 7.65E-08 Xe131m 3.30E+02 6.60E+02 3.61E-03 1.37E-04 1.50E-06 1.47E-07 1.47E-07 1.01E-06 1.01E-06 Xe133 5.70E+04 1.14E+05 3.99E-01 8.75E-05 1.66E-04 1.61E-05 1.61E-05 1.11E-04 1.11E-04 Xe133m 2.00E+03 4.00E+03 8.81E-03 5.50E-05 3.67E-06 3.49E-07 3.49E-07 2.44E-06 2.44E-06 Xe135m(d) 9.22E+03 1.84E+04 2.69E-03 3.65E-06 1.12E-06 1.49E-08 1.49E-08 1.78E-07 1.78E-07 Xe135 1.10E+04 2.20E+04 1.95E-02 2.22E-05 8.14E-06 6.77E-07 6.77E-07 5.14E-06 5.14E-06 Xe137(d) 5.14E+04 1.03E+05 7.49E-03 1.82E-06 3.12E-06 1.15E-08 1.15E-08 1.38E-07 1.38E-07 Xe138 5.70E+04 1.14E+05 1.60E-02 3.51E-06 6.66E-06 8.23E-08 8.23E-08 9.86E-07 9.86E-07 I131 2.80E+04 5.60E+04 2.47E-01 1.10E-04 3.08E-06 3.00E-07 3.00E-08 2.07E-06 2.07E-07 I132 4.00E+04 8.00E+04 3.53E-02 1.10E-05 4.41E-07 2.53E-08 2.53E-09 2.33E-07 2.33E-08 I133 5.70E+04 1.14E+05 1.55E-01 3.40E-05 1.94E-06 1.76E-07 1.76E-08 1.27E-06 1.27E-07 I134 6.30E+04 1.26E+05 3.42E-02 6.79E-06 4.28E-07 1.46E-08 1.46E-09 1.60E-07 1.60E-08 I135 5.00E+04 1.00E+05 7.53E-02 1.88E-05 9.41E-07 7.39E-08 7.39E-09 5.80E-07 5.80E-08 Br83(b) 4.49E+03 8.98E+03 4.06E-03 1.13E-05 5.07E-08 2.96E-09 2.96E-10 2.70E-08 2.70E-09 Br84m(d) 1.40E+02 2.80E+02 2.56E-05 2.28E-06 3.20E-10 1.81E-12 1.81E-13 2.17E-11 2.17E-12 Br84(b) 8.25E+03 1.65E+04 3.48E-03 5.28E-06 4.35E-08 1.05E-09 1.05E-10 1.22E-08 1.22E-09 Br85(d) 1.07E+04 2.14E+04 1.38E-03 1.61E-06 1.73E-08 4.82E-11 4.82E-12 5.79E-10 5.79E-11 Cs138(b,c) 1.41E-07 4.24E-11 1.59E-06 4.76E-10 Rb89(c,d) 2.70E-08 8.09E-12 3.23E-07 9.69E-11 Rb88 (c) 5.19E-08 1.56E-11 4.08E-07 1.22E-10

Table 13-15 lists the TEDE and thyroid TODE to occupational personnel during evacuation and for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of response with a APR respirator. Total effective dose-equivalent (TEDE) is the sum of the deep dose-equivalent from fission gases and committed effective dose-equivalent from inhalation of halogens and particulates. Thyroid total organ dose-equivalent (TODE) is the sum of the deep dose-equivalent and committed dose-equivalent. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months average concentrations were used for the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months doses for response with a SCBA respirator.

The additional following notes apply to Table 13-15:

g. Correction of 0.1 for finite room size for noble gas as stated in Section 11.

h. Protection factor of 100 for a negative pressure air purifying respirator as stated in 10 CFR Part 20 Appendix A.

i. Dose conversion factors were used from 10 CFR Part 20 Appendix B Table 1 to convert concentration of airborne activity to dose. Corrections for exposure time were made.

For personnel in the reactor building at the time the CFA occurs, an exposure time of 6 minutes is used based detector response time and evacuation time.

The time for the detector response to activate the building evacuation alarm is estimated as 4 minutes. Time for released activity to reach the ventilation system and detection by the radiation monitoring system is less than 2 minutes. The detector response time is based on a ventilation system flow rate of greater than 60 feet per minute and duct length of 100 feet and detector response time of 0.5 minutes. A response time of 4 minutes is assumed for conservatism.

The time for personnel to physically exit the reactor building is estimated as 2 minutes.

Evacuation time were measured from various locations inside the reactor building to the evacuation exit point and ranged from 10 seconds to 1 minute following initiation of the reactor building evacuation signal. Also, evacuation times were calculated for an average walking pace of 3 mph for the furthest distance in the reactor building to the evacuation exit point. Both the measured time and estimated walking time gave a time of 1 minute.

A travel time of 2 minutes is assumed for conservatism.

If an emergency entry after a CFA into the reactor building is necessary, such entries would be made using respirators. Occupational dose to emergency response personnel inside the reactor building was calculated for the first 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months following the CFA release while wearing a negative pressure air purifying respirator (APR) with a protection factor of 100. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is recognized as being an excessive exposure period but is used as 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months accounts more than 10 air changes of the reactor building and therefore reduces the initial concentration of airborne activity to negligible levels.

Self-contained breathing apparatus (SCBA) are available with a protection factor of 10,000, as stated in 10 CFR Part 20 Appendix A, if needed. SCBA respirators have a limited air supply, so personnel would enter and exit for brief times. A total of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months wear time was used for response with a SCBA respirator.

Occupational Dose Results Results from Table 13-15 indicate TEDE and TODE are within 10 CFR Part 20 occupational limits.

Occupational limits are 5 rem TEDE and 50 rem TODE. Shallow dose-equivalent and lens dose-equivalent were not calculated since these doses are not controlling per 10 CFR Part 20 Appendix B.

Table 13 24 Hour Total Effective Dose-Equivalent to Occupational Personnel TEDE for 24 Hours TODE for 24 Hours TEDE TODE from Nuclide With Respirator (APR) With Respirator (APR) from Evacuation Evacuation (rem(h)(i)) (rem(h,i))

(rem(g,i)) (rem(i))

Kr83m(d) 3.71E-09 3.71E-09 4.66E-08 4.66E-08 Kr85(a) 1.89E-05 1.89E-05 4.44E-04 4.44E-04 Kr85m 5.16E-06 5.16E-06 8.91E-05 8.91E-05 Kr87 2.18E-05 2.18E-05 2.24E-04 2.24E-04 Kr88 1.18E-04 1.18E-04 1.76E-03 1.76E-03 Kr89(d) 5.21E-04 5.21E-04 3.82E-04 3.82E-04 Xe131m 9.41E-08 9.41E-08 2.20E-06 2.20E-06 Xe133 4.15E-05 4.15E-05 9.65E-04 9.65E-04 Xe133m 9.17E-07 9.17E-07 2.09E-05 2.09E-05 Xe135m(d) 3.12E-06 3.12E-06 9.92E-06 9.92E-06 Xe135 2.03E-05 2.03E-05 4.06E-04 4.06E-04 Xe137(d) 7.80E-04 7.80E-04 6.90E-04 6.90E-04 Xe138 4.16E-05 4.16E-05 1.23E-04 1.23E-04 I131 9.25E-03 3.85E-01 2.16E-03 9.00E-02 I132 2.65E-05 3.68E-04 3.64E-06 5.06E-05 I133 1.29E-03 4.84E-02 2.82E-04 1.06E-02 I134 5.35E-06 4.39E-07 I135 1.41E-04 3.36E-03 2.66E-05 6.34E-04 Br83(b) 4.22E-07 5.92E-08 Br84m(d) 7.99E-07 1.09E-08 Br84(b) 5.44E-07 3.14E-08 Br85(d) 4.32E-05 2.89E-07 Cs138(b,c) 8.32E-07 4.24E-06 Rb89(c,d) 8.18E-07 2.70E-07 Rb88(c) 1.03E-07 1.04E-06 Total = 1.23E-02 4.39E-01 7.60E-03 1.06E-01 Total with SCBA 2 h SCBA: 3.85E-03 2 h with SCBA: 4.43E-03 Respirator

Public Dose Public TEDE is calculated for occupied and other locations outside the reactor facility. The following exposure pathways and atmospheric conditions were used:

1. Submersion and inhalation doses were calculated for occupied locations listed in Table 11-13 for (i) fumigation conditions, (ii) calm winds, and (iii) normal (GPM) weather conditions for exposure periods of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

2. TEDE was also calculated at distances from 30 m to 5000 m for (i) fumigation conditions and (ii) normal weather conditions (GPM) for elevations up to 30 m for periods of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

3. The maximum [X/Q] value for each weather condition at a given location was used. [X/Q]

equations from Section 11 were used; equation 11-15 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , equation 11-16 and 11-23 for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , equation 11-27 for fumigation, and equation 11-28 for calm winds. Equation 11-15 is on the plume centerline for all stability classes. Equation 11-16 is on the plume centerline for stability classes A, B, and C. Equation 11-23 is used for stability classes D, E, and F. Wind speed used was 1 m/s.

4. Dose conversion factors were used from 10 CFR Part 20 Appendix B Table 2 to convert downwind concentrations of airborne activity calculated to dose.

5. Average concentrations for exposure durations of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months were used:

2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months allows for evacuation of the reactor site and other nearby locations.

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months allows for a minimum of 10 air changes of the reactor building, which reduces the initial concentration of airborne activity to negligible levels.

6. Direct exposure from the reactor stack and overhead plume were calculated as depicted in Figure 11-9 and equations 11-43 through 11-46 for released activity from a CFA.

Fumigation conditions provide the least amount of atmospheric dispersion and therefore result in doses that are conservative. Results using the GPM are more typical since fumigation or calm wind conditions occur infrequently.

Results for locations outside the reactor facility are given in the following tables:

Table 13-16 and Table 13-17 lists TEDE from submersion and inhalation to occupied areas near the reactor facility under fumigation conditions, calm winds, and the Gaussian Plume Model (GPM). The GPM uses a wind speed of 1 m/s.

Table 13-18 lists TEDE from submersion and inhalation to locations outside the reactor facility at distances from 30 m to 5000 m and elevations from 0 m (ground) to 30 m for the GPM.

Table 13-20 lists TEDE from direct exposure at two locations outside the reactor facility within the 100 m from the reactor stack.

Table 13 24 Hour Total Effective Dose-Equivalent to Occupied Areas Outside the Reactor Facility Distance Height [X/Q] s/m3 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TEDE Building or x z Fumigation Calm Wind GPM Fumigation Calm Wind GPM Location meters meters 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months mrem mrem mrem All 30 to 100 up to 12 8.54E-03 3.99E-04 3.66E-04 2.13E-01 9.96E-03 9.14E-03 All 100 to 150 up to 12 2.46E-04 4.73E-05 2.39E-04 6.14E-03 1.18E-03 5.97E-03 All 150 to 5000 up to 30 2.00E-03 4.89E-05 2.15E-03 4.99E-02 1.22E-03 5.37E-02 Withers, 50 12 5.38E-03 1.73E-04 3.65E-04 1.34E-01 4.32E-03 9.11E-03 Mann Broughton, 70 12 3.97E-03 9.36E-05 3.61E-04 9.91E-02 2.34E-03 9.01E-03 Riddick Patterson, 90 12 3.17E-03 5.80E-05 3.63E-04 7.91E-02 1.45E-03 9.06E-03 Ricks DH Hill 150 30 2.00E-03 2.17E-05 2.15E-03 4.99E-02 5.42E-04 5.37E-02 Cox 175 12 1.73E-03 1.58E-05 2.17E-04 4.32E-02 3.94E-04 5.42E-03 Dabney 200 24 1.54E-03 1.22E-05 5.12E-04 3.84E-02 3.05E-04 1.28E-02 Hillsboroug 200 15 1.54E-03 1.22E-05 2.59E-04 3.84E-02 3.05E-04 6.46E-03 h St.

Talley, 200 12 1.54E-03 1.21E-05 2.05E-04 3.84E-02 3.02E-04 5.12E-03 Reynolds Carroll, 325 12 9.93E-04 4.61E-06 1.64E-04 2.48E-02 1.15E-04 4.09E-03 Syme North 350 20 8.23E-04 3.99E-06 1.76E-04 2.32E-02 9.96E-05 4.39E-03

Table 13 2 Hour Total Effective Dose-Equivalent to Occupied Areas Outside the Reactor Facility Distance Height [X/Q] s/m3 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months TEDE Building or x z Fumigation Calm Wind GPM Fumigation Calm Wind GPM Location meters meters 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months < 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months mrem mrem mrem All 30 to 100 up to 12 8.54E-03 3.99E-04 3.66E-04 1.39E-01 6.48E-03 5.95E-03 All 100 to 150 up to 12 2.46E-04 4.73E-05 2.39E-04 4.00E-03 7.68E-04 3.88E-03 All 150 to 5000 up to 30 2.00E-03 4.89E-05 7.57E-03 3.25E-02 7.94E-04 1.23E-01 Withers, 50 12 5.38E-03 1.73E-04 3.65E-04 8.74E-02 2.81E-03 5.83E-03 Mann Broughton, 70 12 3.97E-03 9.36E-05 3.61E-04 6.45E-02 1.52E-03 5.86E-03 Riddick Patterson, 90 12 3.17E-03 5.80E-05 3.63E-04 5.15E-02 9.42E-04 5.90E-03 Ricks DH Hill 150 30 2.00E-03 2.17E-05 7.57E-03 3.25E-02 3.52E-04 1.23E-01 Cox 175 12 1.73E-03 1.58E-05 2.17E-04 2.81E-02 2.57E-04 3.52E-03 Dabney 200 24 1.54E-03 1.22E-05 1.49E-03 2.50E-02 1.98E-04 2.42E-02 Hillsboroug 200 15 1.54E-03 1.22E-05 2.59E-04 2.50E-02 1.98E-04 4.21E-03 h St.

Talley, 200 12 1.54E-03 1.21E-05 2.05E-04 2.50E-02 1.97E-04 3.33E-03 Reynolds Carroll, 325 12 9.93E-04 4.61E-06 1.75E-04 1.61E-02 7.49E-05 2.84E-03 Syme North 350 20 8.23E-04 3.99E-06 4.90E-04 1.51E-02 6.48E-05 7.96E-03

Table 13 24 Hour and 2 Hour Total Effective Dose-Equivalent to Locations Outside the Reactor Facility 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Public TEDE 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Public TEDE Distance Height Height x

Ground 10 meter 20 meter(a) 30 meter(a) Ground 10 meter 20 meter(a) 30 meter(a) meters mrem mrem mrem mrem mrem mrem mrem mrem 30 9.31E-05 2.57E-03 2.86E-02 5.43E-01 6.06E-05 1.68E-03 1.86E-02 1.07E+00 50 2.87E-03 7.23E-03 2.92E-02 2.05E-01 1.87E-03 4.71E-03 1.90E-02 4.24E-01 70 5.78E-03 7.75E-03 2.49E-02 1.08E-01 3.76E-03 5.04E-03 1.90E-02 2.31E-01 90 6.50E-03 7.76E-03 1.92E-02 6.71E-02 4.23E-03 5.05E-03 1.81E-02 1.47E-01 100 2.63E-03 4.70E-03 1.63E-02 1.13E-01 1.71E-03 3.06E-03 1.06E-02 2.49E-01 150 3.90E-03 5.05E-03 1.24E-02 5.36E-02 2.54E-03 3.29E-03 9.33E-03 1.23E-01 200 4.23E-03 4.81E-03 8.79E-03 3.18E-02 2.75E-03 3.13E-03 9.44E-03 7.51E-02 250 4.11E-03 4.37E-03 6.47E-03 2.13E-02 2.68E-03 2.84E-03 9.36E-03 5.13E-02 300 3.67E-03 4.13E-03 5.06E-03 1.54E-02 2.39E-03 2.69E-03 8.53E-03 3.77E-02 350 3.61E-03 3.80E-03 4.39E-03 1.17E-02 2.35E-03 2.47E-03 7.96E-03 2.91E-02 400 3.38E-03 3.44E-03 3.82E-03 9.20E-03 2.20E-03 2.37E-03 8.11E-03 2.32E-02 450 3.09E-03 3.09E-03 3.32E-03 7.46E-03 2.01E-03 2.33E-03 7.93E-03 1.90E-02 500 2.79E-03 2.77E-03 2.94E-03 6.19E-03 1.85E-03 2.26E-03 7.58E-03 1.60E-02 600 2.26E-03 2.23E-03 2.55E-03 4.48E-03 1.89E-03 2.22E-03 6.71E-03 1.18E-02 700 1.84E-03 1.82E-03 2.19E-03 3.41E-03 1.81E-03 2.17E-03 5.84E-03 9.10E-03 800 1.52E-03 1.50E-03 1.88E-03 2.70E-03 1.69E-03 2.07E-03 5.07E-03 7.28E-03 900 1.27E-03 1.26E-03 1.62E-03 2.19E-03 1.72E-03 1.97E-03 4.43E-03 5.99E-03 1000 1.08E-03 1.07E-03 1.41E-03 1.82E-03 1.70E-03 1.87E-03 3.89E-03 5.03E-03 1200 8.31E-04 8.46E-04 1.09E-03 1.32E-03 1.58E-03 1.87E-03 3.07E-03 3.70E-03 1400 7.26E-04 7.34E-04 8.86E-04 1.02E-03 1.43E-03 1.78E-03 2.53E-03 2.90E-03 1600 6.41E-04 6.41E-04 7.43E-04 8.20E-04 1.44E-03 1.67E-03 2.14E-03 2.37E-03 1800 5.67E-04 5.63E-04 6.39E-04 6.83E-04 1.41E-03 1.56E-03 1.87E-03 1.99E-03 2000 5.04E-04 4.99E-04 5.60E-04 5.84E-04 1.36E-03 1.46E-03 1.65E-03 1.72E-03 2500 3.96E-04 4.08E-04 4.28E-04 4.26E-04 1.20E-03 1.23E-03 1.29E-03 1.28E-03 3000 3.39E-04 3.43E-04 3.46E-04 3.35E-04 1.04E-03 1.05E-03 1.06E-03 1.03E-03 3500 2.93E-04 2.93E-04 2.89E-04 2.75E-04 9.13E-04 9.12E-04 9.01E-04 8.57E-04 4000 2.56E-04 2.54E-04 2.48E-04 2.33E-04 8.07E-04 8.02E-04 7.82E-04 7.36E-04 5000 2.01E-04 1.98E-04 1.91E-04 1.78E-04 6.47E-04 6.40E-04 6.17E-04 5.75E-04 Note:

(a) No areas are occupied at 20 m and 30 m elevations at distances (x) from 30 m to 150 m.

For direct exposure from distances within 100 m of the stack, two lines of activity in the overhead plume and upper 20 m of the stack are used as described in Section 11.1.1.1 Figure 11-9.

Table 13-19 list the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months averaged source term and Table 13-20 lists the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months TEDE at 2 locations within 100 m of the stack.

Table 13 Fission Product Activity within 100m of the Stack Overhead Line Stack Line Nuclide 24 h Average 24 h Average Curie Curie Kr83m(d) 2.20E-06 3.05E-07 Kr85(a) 2.10E-04 2.91E-05 Kr85m 8.40E-06 1.17E-06 Kr87 5.29E-06 7.34E-07 Kr88 1.66E-05 2.30E-06 Kr89(d) 1.80E-07 2.50E-08 Xe131m 4.15E-06 5.76E-07 Xe133 4.55E-04 6.32E-05 Xe133m 9.88E-06 1.37E-06 Xe135m(d) 4.21E-07 5.84E-08 Xe135 1.92E-05 2.66E-06 Xe137(d) 3.25E-07 4.51E-08 Xe138 2.33E-06 3.23E-07 I131 8.49E-07 1.18E-07 I132 7.16E-08 9.93E-09 I133 4.98E-07 6.91E-08 I134 4.14E-08 5.75E-09 I135 2.09E-07 2.90E-08 Br83(b) 8.38E-09 1.16E-09 Br84m(d) 5.13E-12 7.11E-13 Br84(b) 2.96E-09 4.11E-10 Br85(d) 1.37E-10 1.89E-11 Cs138(b,c) 1.20E-09 1.66E-10 Rb89(c,d) 2.29E-10 3.18E-11 Rb88(c) 4.41E-10 6.11E-11

Table 13 24 Hour Total Effective Dose-Equivalent from Direct Exposure within 100m of the Stack 24 h TEDE in mrem 24 h TEDE in mrem Source At Location 1 At Location 2 (x,y,z) = (50 m, 0 m, 0 m) (x,y,z) = (50 m, 0 m, 12 m)

Stack Line 4.49x10-5 5.15x10-5 Overhead Line 7.22x10-4 1.54x10-3 Total 7.67x10-4 1.59x10-3 Public Dose from Reactor Building Activity in the reactor building air also represents a source of direct radiation. For this calculation, it is assumed the sources remain in the reactor building and are uniformly distributed within the air volume. The reactor building walls were assumed to be equivalent to 12 inches of ordinary concrete.

Steel doors to the reactor are 0.25 inch thick. Using the above assumptions, the direct radiation level outside the reactor building wall was calculated to be 2x10-3 to 5x10-2 mrem/h, or 0.05 to 1.2 mrem in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The steel door locations have the higher dose rate. Reactor doors are located inside of Burlington Laboratory, i.e. there is no direct access from the reactor building to locations outdoor areas. This calculation was made using computer code Microshield 5.[13-20]

Public Dose Results Results in Tables 13-16 through 13-20 indicate the maximum public TEDE to occupied areas for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is estimated to be less than 0.25 mrem. Therefore, the cladding failure accident (CFA) public dose as analyzed is well within annual TEDE limit of 100 mrem given in 10 CFR Part 20 for members of the public. No emergency protective action recommendation is necessary based on radiation dose.[13-19]

v t e Letter Sequence Supplement
EPID:L-2017-RNW-0026, License Renewal (Open)
Initiation Request, Request, Request, Request, Request, Request, Request, Request, Request, Request Acceptance... Supplement Administration Questions 21 May 2018 18 October 2018 23 August 2021 Answers 17 July 2018 1 November 2018 14 February 2019 2 February 2022 Results Approval... Other: ML18312A305, ML18312A306, ML18312A307, ML18312A309, ML18312A310, ML18312A311, ML19046A033, ML19221B601, ML19221B602, ML19269B707, ML19308C717, ML20084K708, ML20084K713, ML20084K716, ML24040A160
MONTHYEAR2019-02-14 [Table View]

ML22159A237

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