ML111570312

Final Safety Analysis Report (FSAR) - Response to Request for Additional Information (RAI) Regarding Accident Dose Analysis Basis

ML111570312
Person / Time
Site:	Watts Bar
Issue date:	06/06/2011
From:	Stinson D Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
Download: ML111570312 (17)

Text

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 June 6,2011 10 CFR 50.4(b)(6) 10 CFR 50.34(b)

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 NRC Docket No. 50-391

Subject:

Watts Bar Nuclear Plant (WBN) Unit 2 - Final Safety Analysis Report (FSAR) - Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis

References:

1. NRC letter to TVA dated June 11, 2010, "Watts Bar Nuclear Plant (WBN) Unit 2 - Request for Additional Information Regarding Licensee's Final Safety Analysis Report Amendment Related to Accident Dose (TAC NO. ME3091)"

2. TVA letter to NRC dated August 30, 2010, ""Watts Bar Nuclear Plant (WBN) Unit 2 - Final Safety Analysis Report (FSAR) - Response to Requests for Additional Information"

3. E-mail from Justin C. Poole, U.S. Nuclear Regulatory Commission to William D. Crouch, TVA, dated March 4,2011 This letter provides a response to a request to provide additional information comparing the Unit 2 accident dose analysis presented in FSAR Chapter 15.5 with the Unit 1 analysis. provides the requested information. Unit 1 values will be the same as the Unit 2 values except as described in Enclosure 1. There are no new commitments made in this submittal.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 6th day of June, 2011.

Respectfully,

-~~

David Stinson Watts Bar Unit 2 Vice President

U.S. Nuclear Regulatory Commission Page 2 June 6, 2011

Enclosure:

1. Response to RAI Concerning WBN Unit 2 Accident Dose Analysis cc (Enclosure):

U. S. Nuclear Regulatory Commission Region II Marquis One Tower 245 Peachtree Center Ave., NE Suite 1200 Atlanta, Georgia 30303-1257 NRC Resident Inspector Unit 2 Watts Bar Nuclear Plant 1260 Nuclear Plant Road Spring City, Tennessee 37381

Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis NRC Question from Reference 1 15.5-1. To ensure a complete and accurate safety assessment of the proposed changes to the Watts Bar FSAR, the NRC staff needs to assess the safety significant of all of the changes to the current licensing basis (CLB) parameter used in the dose consequences described in Chapter 15.5.

Please provide additional information describing, for each design basis accident described in FSAR Section 15.5 all the basic parameters used in the dose consequence analyses. For each parameter, please list the WBN Unit 1 CLB value, the revised value where applicable as will be applied to Unit 1 and 2, and the basis for any changes mode to the WBN Unit 1 CLB values. The NRC staff requests that this information be presented in separate tables for each accident evaluated.

TVA Response:

The WBN Unit 1 UFSAR addresses the dose consequences of seven postulated design basis accidents in Section 15.5. The accidents addressed are:

1. Loss of AC Power to the Plant Auxiliaries

2. Waste Gas Decay Tank Rupture

3. Loss of Coolant Accident (LOCA)

4. Steam Line Break

5. Steam Generator Tube Rupture

6. Fuel Handling Accident

7. Rod Ejection Accident The Rod Ejection Accident is bounded by the LOCA, and is not addressed in detail.

The following table provides the Unit 1 current licensing basis (column 1), the revised value where applicable as will be applied to Unit 2 (column 2), and the basis for any changes made to the Unit 1 values (column 3). Only those values that have been revised are presented with their basis in columns 2 and 3, respectively. The table numbers specified in the comparison table below are from the UFSAR for Unit 1 and from Amendment 103 of the Unit 2 FSAR. The analyses for steam line break, steam generator tube rupture, and loss of AC power for the plant were updated recently. The reactor coolant system dose equivalent iodine for the steam line break and steam generator tube rupture was set at the Technical Specification (TS) value. The value for secondary side dose equivalent iodine for the loss of AC Power was also set at the TS limit. Unit 2 Table 15.5-16 on page E1-14 was updated due to these revisions. The changes to the Unit 2 Accident Analyses were submitted in Amendment 104. A revised Unit 1 Table 15.1-4 is provided on page E1-13. The new values are reflected in the comparison table.

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Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis Unit 1 CLB parameter Revised Unit 2 Parameter Basis for Change 15.5.1 Loss of AC Power

1. A conservative analysis of the potential 1. Unit 2 replaced 1% failed fuel source term 1. To provide a conservative bounding dose, offsite doses resulting from this accident is with the TS Allowable source term. Also, Unit 2 uses the TS Allowable source term presented with steam generator leakage Unit 1 used I-131 dose equivalent factors of 0.1 Ci/g I-131 dose equivalent for the as a parameter. This analysis from ICRP-2, where Unit 2 used Dose secondary coolant in conjunction with the incorporates assumptions of one percent Factors from Regulatory Guide (RG) 1.109, Dose Factors from RG 1.109, Appendix E.

defective fuel and a realistic source term. Appendix E.

Parameters used in both the realistic and conservative analyses are listed in Table 15.5-1.

2. The steam release from four steam 2. The Unit 2 steam release from four steam 2. The volume of steam released in Unit 1 generators is 455,718 lbs (0-2 hr) and generators is 444,875 lbs (0-2 hr) and reflects the use of the replacement steam 962,213 lbs (2-8 hr). 903,530 lbs (2-8 hr). generators (RSGs). The volume of steam release in Unit 2 reflects the use of the original steam generators (OSGs). The volumes are from TVA calculation WBNTSR080, Revision 8.

3. Primary coolant activity is based on 3. 3.

ANSI/ANS 18.1, 1984 and is given in UFSAR Table 11.1-7.

4. The tritium source term is based on two (2) 4. Unit 2 analysis is based on the standard 4. Unit 2 will not produce tritium.

Tritium Producing Burnable Absorber Rods core and did not consider TPBAR failure (TPBAR) failures.

5. The Unit 1 secondary coolant steam 5. The Unit 2 secondary coolant activity 5. The secondary coolant activity released in activity released for the realistic case is in released is in Table 11.1-7. The activity Unit 2 reflects the use of the OSGs. For Table 11.1-7. The activity was ratioed up was ratioed up to the TS limit of 0.1 uCi/gm conservatism, the specific secondary to 1% failed fuel for the conservative case. dose equivalent I-131 for the conservative coolant activity for Unit 1 is used for Unit 2 case. because the Unit 1 specific secondary coolant activity bounds the Unit 2 specific secondary coolant activity. The volume of Unit 2 coolant released is in Item 2 above.

6. The meteorology for the Exclusion Area 6. 6.

Boundary (EAB) and the Low Population Zone are in Table 15A-2.

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Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis

7. The control room parameters are in 7. The atmospheric dispersion coefficients for 7. The Unit 2 atmospheric dispersion Table 15.5-14. One Control Room Unit 2 control room are: coefficients are based on the Unit 2 Emergency Ventilation System (CREVS) is release location at the Unit 2 vent valve in operation. Interval sec/m3 stacks and the Emergency Control Room 0-2h 2.87E-3 Intake. The Unit 2 atmospheric dispersion 2-8h 2.46E-3 coefficients are from TVA calculation 8 - 24 h 1.14E-3 WBNTSR080, Revision 8 and 1-4d 7.55E-4 WBNAPS3104, Revision 2. The accident 4 - 30 d 6.22E-4 releases terminate at 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

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Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis Unit 1 CLB parameter Revised Unit 2 Parameter Basis for Change 15.5.2 Waste Gas Decay Tank Rupture

1. The parameters for the realistic and 1. There are no revised Unit 2 parameters 1.

RG 1.24 analyses of the Waste Gas Decay because the Waste Gas System is Tank (WGDT) rupture are in Table 15.5-3. common to Units 1 and 2.

2. The radiation source terms for the RG 1.24 2. 2.

analysis are in Table 15.5-4. The reactor has been operating at full power with 1%

defective fuel for the RG 1.24 analysis.

3. The tank rupture is assumed to occur 3. 3.

immediately upon completion of the waste gas transfer, releasing the entire contents of the tank through the Auxiliary Building (AB) vent to the outside atmosphere. The assumption of the release of the noble gas inventory from only a single tank is based on the fact that all gas decay tanks will be isolated from each other whenever they are in use

4. The short-term (i.e., 0-2 hour) dilution 4. 4.

factor at the EAB given in Table 15A-2 is used to evaluate the doses from the released activity.

5. Parameters for the control room analysis 5. 5.

are found in Table 15.5-14.

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Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis Unit 1 CLB parameter Revised Unit 2 Parameter Basis for Change 15.5.3 LOCA

1. The analysis is based on RG 1.4. The 1. 1.

parameters used for this analysis are in Table 15.5.6. In addition, an evaluation of the dose to control room operators and an evaluation of the offsite doses resulting from recirculation loop leakage are presented.

2. The Containment Building Ice Condenser 2. 2.

removal efficiency for elemental and particulate iodine is in Table 15.5.7. There is no removal of organic iodine by the Ice Condenser.

3. The time dependent Emergency Gas 3. The time dependent EGTS flow rates for 3. The time dependent EGTS flow rates for Treatment System (EGTS) flow rates for Unit 1 and 2 corresponding to postulated Unit 1 and 2 are revised to reflect 1 and 2 the Unit 1 CLB are in Table 15.5.8. These single failure of one train of EGTS with a hour operator action to correct the EGTS flow rates correspond to postulated single 250 cfm steady state exhaust concurrent operation to a single fan. The EGTS is failure of one train of EGTS with a 250 cfm with a LOCA are in Table 15.5.8. Table modeled as 2 trains with appropriate steady state exhaust concurrent with a 15.5-8A flow rates are as a result of an recirculation/ exhaust based on revised LOCA. Table 15.5-8A flow rates are as a alternate single failure scenario resulting in flow calculations. The EGTS is assumed result of an alternate single failure scenario one pressure control train in full exhaust to to have a PCO control loop single failure resulting in one pressure control train in full the shield building exhaust stack while the at the beginning of the accident such that exhaust to the shield building exhaust other train remains functional. Both EGTS a maximum two train occurs at the stack while the other train remains fans are in service until operator action is beginning of the accident until operator functional. Both EGTS fans are in service taken to place one fan in standby between action is credited in turning off one fan until operator action is taken to place one one and two hours post accident. The Unit between 1 and 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The Unit 1 and fan in standby between one and two hours 2 exhaust flow rate to the outside Unit 2 EGTS flow rates are based on post accident. The Unit 1 exhaust flow rate environment at one hour post accident is separate analyses of ventilation system to the outside environment at one hour 832 cfm. At this time one EGTS fan is calculations reflecting the physical post accident is 957 cfm. At this time one placed in standby. The exhaust flow rate differences between the units.

EGTS fan is placed in standby. The to the outside environment then becomes exhaust flow rate to the outside 604 cfm.

environment then becomes 694 cfm.

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Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis

4. Based on RG 1.4, a total of 100% of the 4. The Unit 2 core inventory of iodines and 4. The Unit 1 source terms are based on a noble gas core inventory and 25% of the noble gases is in Table 15.1-5. tritium core where the Unit 2 source terms core iodine inventory are assumed to be are based on a standard core.

immediately available for leakage from the primary containment. Of the halogen activity available for release, it is further assumed that 91% is in elemental form, 4% in methyl form, and 5% in particulate form. The core inventory of iodines and noble gases is attached (Table 15.1-4 revised).

5. The tritium source term is based on the 5. The Unit 2 analyses did not consider 5. Unit 2 will not be licensed for tritium failure of two (2) TPBARs TPBARs. production.

6. The parameters used in the analysis of 6. 6.

recirculation loop leakage following a LOCA are in Table 15.5-12.

7. The meteorology for the EAB and the Low 7. 7.

Population Zone are in Table 15A-2.

8. The control room parameters are in 8. 8.

Table 15.5-14.

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Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis Unit 1 CLB parameter Revised Unit 2 Parameter Basis for Change 15.5.4 Steam Line Break

1. The parameters for the Main Steam Line 1. Steam releases provided in Unit 2 1. Unit 2 steam release values are different Break (MSLB) are in Table 15.5-16. Table 15.5-16 are different than values for than Unit 1 values since the Unit 1 steam Unit 1. generators have been replaced by different models. See item 7 below.

2. The primary to secondary leakage rate for 2. 2.

the MSLB accident was determined to be 1 gallon per minute (gpm) for the faulted steam generator loop and 150 gallons per day (gpd) for each un-faulted steam generator.

3. The two methods of determining the 3. For Unit 2 the maximum allowable dose 3. A TS change is being processed to resultant dose for MSLB are: equivalent I-131 value is 14 Ci/gm. change the Unit 1 dose equivalent I-131

a. a pre-accident iodine spike where the from 21 Ci/gm to 14 Ci/gm. This change iodine level in the reactor coolant is necessary since the I-131 dose spiked upward to the maximum equivalent conversion factors were allowable limit of 21 Ci/gm I-131 dose changed from ICRP-2 values to RG 1.109 equivalent just prior to the initiation of values. The Unit 1 TS require that the accident, and RG 1.109 be used to determine dose

b. the reactor coolant at the maximum equivalent I-131.

steady state I-131 dose equivalent of 0.265 Ci/gm with an accident initiated iodine spike consisting of a 500 times increase on the rate of iodine release from the fuel

4. The reactor coolant system (RCS) letdown 4. 4.

flow is 124.39 gpm.

5. The RCS letdown demineralizer efficiency 5. 5.

is assumed to be 1.0 for iodines.

6. ANSI/ANS-18.1-1984 spectrum was used 6. For Unit 2 the ANSI/ANS-18.1-1984 6. See discussion in item 3.

and scaled up to 0.265 or 21 Ci/g spectrum was used and scaled up to equivalent iodine. 0.265 or 14 Ci/g equivalent iodine.

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Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis

7. Steam generator secondary inventory 7. Steam generator secondary coolant 7. The volume of steam released in Unit 1 released as steam to the atmosphere is in inventory released as steam to the reflects the use of the RSGs. The volume Table 15.5-16. atmosphere is: of steam release in Unit 2 reflects the use a) total from the faulted steam generator: of the OSGs. The volumes are from TVA b) (0-30 mins) 96,100 lbm calculation WBNAPS3077, Revision 13.

c) total from the non-defective steam generators (0-2 hr), 433,079 lbm d) total from the non-defective steam generators (2-8 hr), 870,754 lbm

8. Iodine partition coefficients from steaming 8. 8.

of steam generator water:

a. non-defective steam generators initial inventory and primary-to-secondary leakage - 0.01; and

b. faulted steam generator initial inventory and primary-to-secondary leakage - 1.0

9. The meteorology for the Exclusion Area 9. 9.

Boundary (EAB) and the Low Population Zone are in Table 15A-2.

10. The control room parameters are in 10. The atmospheric dispersion coefficients 10. The Unit 2 atmospheric dispersion Table 15.5-14. for Unit 2 are: coefficients are based on the Unit 2 release location at the Unit 2 vent valve sec/m3 stacks and the Emergency Control Room 0-2h 2.87E-3 Intake. The Unit 2 atmospheric dispersion 2-8h 2.46E-3 coefficients are from TVA calculation 8 - 24 h 1.14E-3 WBNTSR080, Revision 8, and 1-4d 7.55E-4 WBNAPS3104, Revision 2.

4 - 30 d 6.22E-4 E1-8

Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis Unit 1 CLB parameter Revised Unit 2 Parameter Basis for Change 15.5.5 Steam Generator Tube Rupture

1. Table 15.5-18 summarizes the parameters 1. The Unit 2 secondary side mass releases 1. The Unit 2 secondary side mass releases used in the Steam Generator Tube from the ruptured steam generator are: from the ruptured and intact steam Rupture (SGTR) analysis. The SGTR 0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 103,300 lbm generator and primary coolant are based thermal and hydraulic analysis documents 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 32,800 lbm on an RCS with the OSGs from TVA use WBN specific parameters and actual calculation WBNTSR008, Revision 13.

operator performance data. The Unit 2 secondary side mass releases from the intact steam generator are:

0 - 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 492,100 lbm 2 - 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 900,200 lbm The primary coolant mass released total and flashed are 191,400 and 10,077.2 lbm, respectively.

2. Two cases were analyzed. Case 1: The 2. For Unit 2 the maximum allowable dose 2. A TS change is being processed to change primary side activity release was equivalent I-131 value is 14 Ci/gm the Unit 1 dose equivalent I-131 from 21 determined by using maximum TS limit Ci/gm to 14 Ci/gm. This change is design reactor coolant activities and an necessary since the I-131 dose equivalent iodine spike immediately after the accident conversion factors were changed from that increases the iodine activity in the ICRP-2 values to RG 1.109 values. The reactor coolant by a factor of 500 times the Unit 1 TS require that RG 1.109 be used to iodine production rate necessary to determine dose equivalent I-131.

maintain a steady state concentration of 0.265 Ci/gm of dose equivalent I-131.

Case 2: The initial reactor coolant activity is at 21 Ci/gm of I-131 equivalent due to a pre-accident iodine spike caused by an RCS transient. For both cases, the secondary side releases were determined using expected secondary side activities, based on ANSI/ANS-18.1-l984 as modified for WBN, and on a 150 gpd/steam generator primary-to- secondary-side leakage.

3. The tritium source term was based on the 3. The Unit 2 analyses did not consider 3. Unit 2 will not be licensed for tritium failure of two TPBARs. TPBARs. production.

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Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis

4. Credit was taken for flashing of the primary 4. 4.

coolant, but "scrubbing" of the iodine in the rising steam bubbles by the water in the steam generator was conservatively neglected.

5. A partition factor of 100 was applied to 5. 5.

iodine in the remaining unflashed coolant which will boil.

6. The atmospheric diffusion coefficients 6. 6.

(/Q,) for the EAB and LPZ are in Table 15A-2.

7. The /Q values were based on release 7. The Control Building atmospheric 7. The Unit 2 atmospheric dispersion from the top of the main steam valve vault. dispersion coefficients for Unit 2 are: coefficients are based on the Unit 2 release The control room parameters are in Table location at the Unit 2 vent valve stacks and 15.5-14. sec/m3 the Unit 2 control room air intake on the 0-2h 2.87E-3 east side of the control building. The Unit 2 2-8h 2.46E-3 atmospheric dispersion coefficients are 8 - 24 h 1.14E-3 from TVA calculation WBNTSR008, 1-4d 7.55E-4 Revision 13, and WBNAPS3104, 4 - 30 d 6.22E-4 Revision 2.

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Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis Unit 1 CLB parameter Revised Unit 2 Parameter Basis for Change 15.5.6 Fuel Handling Accident

1. The parameters used for this analysis are 1. 1.

listed in Table 15.5-20. The bases for the RG 1.25 evaluations are:

a. In the Regulatory Guide 1.25 analysis the accident occurs 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> after plant shutdown. Radioactive decay of the fission product inventory during the interval between shutdown and placement of the first spent fuel assembly into the spent fuel pit is taken into account; and

b. In the RG 1.25 analysis damage was assumed for all rods in one assembly.

2. The assembly damaged is the highest 2. 2.

powered assembly in the core region to be discharged. The values for individual fission product inventories in the damaged assembly are calculated assuming full-power operation at the end of core life immediately preceding shutdown. Nuclear core characteristics used in the analysis are given in Table 15.5-21.

3. The radial peaking factor is 1.65. 3. 3.

4. All the gap activity in the damaged rods is 4. 4.

released to the spent fuel pool and consists of 10% of the total noble gases and radioactive iodine inventory in the rods at the time of the accident with the following gap percentage exceptions which are based on NUREG/CR 5009 as appropriate: 14% of the Kr-85, 5% of the Xe-133, 2% of the Xe-135, and 12% of the I-131.

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Enclosure 1 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis

5. Noble gases released to the spent fuel 5. 5.

pool are released through the Shield Building vent to the environment.

6. The iodine gap inventory is composed of 6. 6.

inorganic species (99.75%) and organic species (0.25%).

7. The spent fuel pool decontamination 7. 7.

factors for the inorganic and organic iodine are 133 and 1, respectively.

8. All iodine escaping from the pool is 8. 8.

exhausted to the environment through charcoal filters. A filter efficiency of 99% is used for elemental and organic iodine for the ABGTS filters and 90% for inorganic iodine and 30% for organic iodine for the purge air exhaust filters.

9. No credit is taken for natural decay either 9. 9.

due to holdup in the Auxiliary Building or after the activity has been released to the atmosphere.

10. The short-term (i.e., 0-2 hour) atmospheric 10. 10.

dilution factors at the EAB and low population zone given in Table 15A-2 are used. The Control Building dilution factors are in Table 15.5-14

11. Two TPBARs in the assembly are 11. The Unit 2 analyses did not consider 11. Unit 2 will not be licensed for tritium assumed to break and release the entire TPBARs. production.

contents of tritium. All of the tritium is conservatively assumed to evaporate into the air.

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Enclosure 2 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis Unit 1 TABLE 15.1-4 CORE AND GAP ACTIVITIES BASED ON FULL POWER OPERATION FOR 1,000 DAYS FULL POWER: 3,565 MWt Isotope Curies/Assembly Total Curies in Core Kr-83m 6.37E+04 1.23E+07 Kr-85m 1.39E+05 2.69E+07 Kr-85 4.56E+03 8.81E+05 Kr-87 2.71E+05 5.23E+07 Kr-88 3.82E+05 7.38E+07 Kr-89 4.72E+05 9.10E+07 Xe-131m 4.94E+03 9.54E+05 Xe-133m 3.01E+04 5.8E+06 Xe-133 9.74E+05 1.88E+08 Xe-135m 1.86E+05 3.59E+07 Xe-135 2.57E+05 4.96E+07 Xe-138 8.24E+05 1.59E+08 I-131 4.67E+05 9.01E+07 I-132 6.79E+05 1.31E+08 I-133 9.74E+05 1.88E+08 I-134 1.08E+06 2.08E+08 I-135 9.12E+05 1.76E+08 E1-13

Enclosure 2 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis Unit 2 TABLE 15.5-16 PARAMETERS USED IN STEAM LINE BREAK ANALYSIS Analysis Value Steam Generator Tube Leak Rate Faulted Steam Generator 1 gpm Per Intact Steam Generator 150 gpd Iodine Partition Factor Faulted Steam Generator 1 Intact Steam Generator 100 RCS Letdown Flow Rate 124.39 gpm Steam Releases Faulted Steam Generator (0-30 minutes) 96,100 lbm Three Intact Steam Generators (0-2 hours) 433,079 lbm Three Intact Steam Generators (2-8 hours) 870,754 lbm E1-14

Enclosure 2 Watts Bar Nuclear Plant Response to Request for Additional Information (RAI)

Regarding Accident Dose Analysis Basis REFERENCES Section 15.5.1 - Loss of AC Power to the Plant Auxiliaries - TVA Calculation WBNTRS080 R8 Section 15.5.2 - Waste Gas Decay Tank Rupture - TVA Calculation WBNTRS064 R9 Section 15.5.3 - Loss of Coolant Accident - TVA Calculations TIRPS197 R21 (offsite) and TIRPS198 R23 (control room operator)

Section 15.5.4 - Main Steam Line Break - TVA Calculation WBNTRS077 R13 Section 15.5.5 - Steam Generator Tube Rupture - TVA Calculation WBNTRS008 R13 Section 15.5.6 - Fuel Handling Accident - TVA Calculation WBNTRS009 R12 E1-15