ML20148D746
| ML20148D746 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 05/23/1997 |
| From: | Hosmer J COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| NUDOCS 9706020092 | |
| Download: ML20148D746 (4) | |
Text
Commonwealth Edison Compan) g 1400 Opus Place Downers Grose, IL NK15 5'o1 May 23,1997 United States Nuclear Regulatory Commission Washington. D.C. 20555 Attention: NRC Document Control Desk
Subject:
Byron Nuclear Power Station, Units 1 and 2 Facility Operating Licenses NPF-37 and NPF-66 NRC Docket Numbers: 50-454 and 50-455 i
" Response to Request for Additional Information"
References:
1.
J. Hosmer letter to the Nuclear Regulatory Commission dated January 31,1997, transmitting Technical Specification Amendment l
Request for Specific Activity 2.
D. Lynch letter to 1. Johnson dated May 2,1997, transmitting Request for i
Additional Information Pertaining to the Proposed Reduction in the Maximum Allowable Dose Equivalent Iodine-131 Concentration in the Byron 1 a
Reference I transmitted the Commonwealth Edison Company's (Comed) request to amend the technical specification to allow for the reduction in the allowable dose equivalent iodinc concentration in the reactor coolant for B ron Unit 1. Subsequent to that submittal the Nuclear Regulatory Commission (NRC) issued 3
a Request for Additional Information (RAI). Attached is Comed's response to the RAI.
If 0u have any qtesticus concerning this correspondence, please contact Denise Saccomando at (630) 3 663-7283.
Sincerely, bS John B. Ilosmer Engineering Vice President a
Attachtnent cc:
D. Lynch. Senior Project Manager-NRR G. Dick, Byron /Braidwood Project Manger-NRR
/\\D S. Burgess, Senior Resident Inspector-Byron A.B. Beach Regional Administrator-Rill OITice of Nelcar Safety-IDNS
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ATTACllMENT REQUEST FOR ADDITIONAL INFORMATION RELATED TO THE PROPOSED REDUCTION IN THE BYRON UNIT I DOSE EQUIVALENT IODINE-131 NRC Ouestion # 1:
Table 6.4-la of the 3 ron Station UFSAR provides information regarding the values of the atmospheric 3
dispersion factors, 4/Q, which are used in dose assessments when evaluating the control room habitability for design basis accidents. State what additional parameters and assumptions were used in your calculations to determine the various values of X/Q in Table 6.4-la. These parameters should include the wind speeds at the 5,10,20 and 40 percentile levels, values of sigma y and sigma z, the period over which the meteorological data was acquired, the wind directions used to calculate the X/Q values and, if appropriate, building measurements (i.e. diameter) and the building cross-sectional area State w hich Murphy-Campe equation you used and your basis for the use of the applicable equation. Finally, provide the input parameters and assumptions you used in determining the maximum X/Q value over the time period of 0 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
Byron Response:
All additional parameters and assumptions used in determining Table 6.4-la X/Q are documented below.
The 5th percentile wind speed is 0.18 m/s u hich was obtained through a cumulative probability distribution analysis using FSAR Table 2.3-25 meteorological data. This table was compiled from on-site meteorology data collected over a three-> car period of record (Jan.1,1974-Dec. 31, 1976). The maximum X/Q value was determined to be that for winds blowing from the north i
containment (unit 2) in an casterly direction (98 degrees cast of north) to the control room intake, located on the auxiliary building / turbine building common wall,100 feet distant (in plan view) from the containment surface.
The percentile levels are associated with time periods and windspeed factors in accordance with the Murphy-Campe paper presented at the 13* AEC Air Cleaning Conference in August 1974.
Only the data associated with the 5* percentile is used in the steam line break analysis. The remaining data is applicable to the control room dose assessment for the Loss of Coolant Accident, which is considered to be an upper bound of all accidents postulated to occur per UFSAR Section 6.4.4.1. The wind speeds and associated drua for cach percentile level are as follows:
Percentile Level Time Period (br)
Wind Speed (m/sec)
Wind Speed Factor 5
0-8
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Since Regulatory Guide 1.145," Atmospheric Dispersion Models for Potential Accident consequence Assessments at Nuclear Power Plants." had not been issued at the time of the original SAR submittal X/Q was calculated based on available NRC guidance. Byron calculations were consistent with the guidance provided in Section 2.3.4.2 of Regulatory Guide 1.70, " Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants,"
j Revision 2, issued September 1975. The approach used is specifically detailed in Meteorology and Atomic Energy - 1968, Section 3-3.5.2, and was not taken from Murphy-Campe.
The equation used to calculate X/Q is the following:
X/Q = 2/uA where: u is the wind speed, and A is the cross sectional area of the building causing the downwind turbulent wake.
The input parameters in determining X/Q are:
2 X/Q = 2/[(0.1828 m/s)(2700 m )j i
= 4.05E-3 sec/m'(Table 4.6-la value)
The model assumes that all cfiluents leave the source surface of arca A. are captured and uniformly mixed in a turbulent downwind " tube" of cross sectional area, A/2. The contents of the tube move downwind with speed, u. The arca, A, was conservatively assumed to be equal to 2
? 700 m, the cross sectional area of an assumed free standing single containment. The model is considered to provide a conservative estimate of atmospheric dispersion, given the complexity of multiple structures on site and the location of the control room air intakes (within the turbine building). Equation (1) of US NRC Regulatory Guide 1.145 similarly reduces to this form at positions near a structure (i e., the product of sigma y and sigma z is small compared to A). The model used to calculate the Byron X/Q values did not use sigma y and sigma z.
NRC Ouestion # 2:
Provide the following values for the steam and water in the secondary side of the Byron steam generators:
a.
The volume of the steam (in cubic feet per generator).
b.
The volume of the liquid (in cubic feet per generator).
c.
The temperature of the steam / liquid mix in the secondary side of the steam generators.
d.
The mass of the secondary coolant (in pounds per generator).
Ilvrrm Resnonse:
The following parameters are for the Byron Unit i steam generators during hot full power operation, (1,000 psia):
a.
The volume of steam in cach generator is 3889 cubic feet.
b.
The volume ofliquid in cach generator is 2070 cubic fect.
c.
The temperature of the steam / liquid mix in the secondary side of the steam generators is
$44 degrees Fahrenheit.
d.
The mass of the secondary coolant is 96.000 lbm for cach generator.
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NRC Ouestion #3:
Section 4.8 of WCAP-14046 states that from 0 to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />,416,573 pounds of steam are released to the environment from the three steam generators in the intact loops. State the amount of steam released from these intact steam generators in the time period from 2 to 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Alternatively, state whether you assume that all of the steam is released during the initial two hours following the main stream line break.
Hvrrm Resnonse:
As stated in Section 4.8 of WCAP-14046, the maximum permissible steam generator primary to secondary leak rate during a Main Steam Line Break (MSLB) was evaluated using a limiting case. The limiting case was based on a 30 rem dose at the exclusion area boundary (EAB) for 0-2 hours after the event. This case is more limiting than the low population zone (LPZ) dose during the entire duration of the accident, as evidenced from the doses included in UFSAR Table 15.0-11. The data included in Table 15.0-11 for the design basis MSLB accident demonstrate that LPZ dose for the course of the accident is significantly lower than the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> EAB dose. Sin:c the dose at the EAB is calculated from 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, only the amount of steam release during this time period was considered in WCAP-14046. The maximum permissible steam generator primary to secondary leak rate was calculated for Byron Unit I using the same methodology and assumptions as described in WCAP-14046.
j For the 24% steam generator tube plugging program, the applicable offsite dose cases for a main steam line break, as defined in Standard Resiew Plan Section 15.1.5, Appendix A, were analyzed. These analyscs considered steam releases up to 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br /> from the three intact steam generators and used the same maximum permissible steam generator primary to secondary leak rate for Byron and Braidwood Unit 1. The results are discussed in UFSAR Section 15.1.5.3. The results show that all applicable dose acceptance criteria are met. The amount of steam release from the three intact steam generators from 0 -
2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is 406,716 lbm and from 2-8 hours is 939,604 lbm as shown in UFSAR Table 15.1-3.
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