ML110730761

OL - Chapter 11 and 12 RAI Responses

ML110730761
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Site:	Watts Bar
Issue date:	02/25/2011
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To:	Division of Operating Reactor Licensing
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WBN2Public Resource From: Stockton, Rickey A [rastockton@tva.gov]

Sent: Friday, February 25, 2011 3:50 PM To: Poole, Justin Cc: Crouch, William D; Fickey, Donald G; Woods, Steven E; Clark, Mark Steven; Boyd, Desiree L

Subject:

Chapter 11 and 12 RAI Responses Attachments: 02-25 Chapter 11 and 12 RAI Responses.pdf

Justin, Attached is the submittal containing the Chapter 11 and 12 RAI Responses. Please call me if you should have any questions.

Rickey Stockton Unit 2 Licensing (423) 365-7741 1

Hearing Identifier: Watts_Bar_2_Operating_LA_Public Email Number: 298 Mail Envelope Properties (6B28FBDBF05ED74B8991E9374A9F54D90A027ABB)

Subject:

Chapter 11 and 12 RAI Responses Sent Date: 2/25/2011 3:49:58 PM Received Date: 2/25/2011 3:50:39 PM From: Stockton, Rickey A Created By: rastockton@tva.gov Recipients:

"Crouch, William D" <wdcrouch@tva.gov>

Tracking Status: None "Fickey, Donald G" <dgfickey@tva.gov>

Tracking Status: None "Woods, Steven E" <sewoods@tva.gov>

Tracking Status: None "Clark, Mark Steven" <msclark0@tva.gov>

Tracking Status: None "Boyd, Desiree L" <dlboyd@tva.gov>

Tracking Status: None "Poole, Justin" <Justin.Poole@nrc.gov>

Tracking Status: None Post Office: TVANUCXVS2.main.tva.gov Files Size Date & Time MESSAGE 208 2/25/2011 3:50:39 PM 02-25 Chapter 11 and 12 RAI Responses.pdf 3976641 Options Priority: Standard Return Notification: Yes Reply Requested: Yes Sensitivity: Normal Expiration Date:

Recipients Received:

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 February 25, 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 CHAPTERS 11 AND 12 REQUEST FOR ADDITIONAL INFORMATION

References:

1. TVA letter to NRC dated December 17, 2010, Watts Bar Nuclear Plant (WBN)

- Unit 2 - Final Safety Analysis Report (FSAR), Amendment 102

2. TVA letter to NRC dated February 15, 2008, Watts Bar Nuclear Plant (WBN) -

Unit 2 - Final Supplemental Environmental Impact Statement for the Completion and Operation of Unit 2 The purpose of this letter is to respond to a number of requests for additional information (RAIs) regarding the Unit 2 FSAR Chapters 11 and 12. provides the responses to RAIs received via email on February 9, 2011. The NRC questions and associated numbering is retained herein. provides the responses for the outstanding Chapter 11 RAIs previously received. provides proposed markups to FSAR Chapter 11 (Reference 1) and the Final Supplemental Environmental Impact Statement (Reference 2). These markups correct identified errors found during the preparation of the Chapter 11 RAI responses. TVA has evaluated these errors and determined that NRC notification is not required under 10 CFR 50.9(b) since the errors do not represent a significant implication for public health and safety or common defense and security.

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information Liquid Waste Management System

1. NRC QUESTION:

Columns 4 through 8 of Table 11.2-5 present five different liquid effluent isotopic spectrums, and the total annual radioactivity, released in liquid effluents with, or without, processing of the different waste streams. These total annual releases are compared to the 5 Ci release limit for each reactor in RM 50-2, as annexed to 10 CFR 50, Appendix I. Amendment 95 made minor adjustments to the activities listed in columns 4 and 5 of Table 11.2-5, and added columns 6, 7, and 8 to include releases from unprocessed steam generator blowdown effluent. Amendment 101 revised Section 11.2.6.5 to describe the radwaste process configurations represented by each column of Table 11.2-5. Amendment 102 added column headers and a footnote to Table 11.2-5 explaining each column. All five of the activity columns (columns 4 through 8) of Table 11.2-5 contain liquid waste contributions from the Tritiated Drain Collector Tank, processed by the CVCS Demineralizer and the Mobil Demineralizer; the Reactor Coolant Drain tank, processed by the Mobil Demineralizer; the unprocessed Laundry and Hot Shower Drain Tank; and the unprocessed Turbine Building drains. In addition to these, Column 4 includes Condensate Demineralizer regeneration backwash and steam generator blowdown effluents that have had Condensate Demineralizer decontamination factors [RAI 11-13 & 14, RAI 11-1 is OPEN] applied. Column 5 also applies the decontamination factors for the Mobile Demineralizer to the Condensate Demineralizer backwash and steam generator blowdown process streams. Column 6 represents no processing of, nor release restrictions on, the Condensate Demineralizer and blowdown effluent streams.

Columns 7 and 8 present the annual activity release if the steam generator untreated effluent concentrations are maintained below 5 E-7 uCi/cc and 3.65E-5 uCi/cc, respectively. However, column 7 and column 8 do not include Condensate Demineralizer backwash wastes.

It is unclear how TVA intends to operate WBN Unit 2 without performing this routine maintenance of the Condensate Demineralizer System [RAI 11-10].

TVA RESPONSE:

Column 7 and 8 in Table 11.2-5 are providing information that the 10 CFR 50, Appendix I yearly regulatory limits can be met without use of the Condensate Demineralizers with the specified activity limitations on the Steam Generator Blowdown. Unit 1 currently operates without use of the Condensate Demineralizers. The Condensate Demineralizers will not be used unless significant primary to secondary leakage occurs. Since the demineralizers are not used, the Steam Generator Blowdown is not treated and there is no demineralizer blowdown or backwash waste stream. This method of operation is acceptable so long as the 10 CFR 50, Appendix I and 10 CFR 20 limits are met. TVA plans to operate Unit 2 in the same manner as Unit 1. Note actual plant releases are accomplished and controlled in E1-1

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information accordance with the Offsite Dose Calculation Manual (ODCM) and releases are not allowed to exceed either the 10 CFR 50, Appendix I or the 10 CFR 20 limits.

2. NRC QUESTION:

Amendment 98 made minor revisions to the values in Tables 11.2-5a and 11.2-5b.

These revisions did not affect the final results presented in Tables 11.2-5a and 11.2-5b, e.g., that extended effluent releases without processing the Condensate Demineralizer regeneration waste through the Mobile Demineralizer will not meet the limits of 10 CFR 20 and is not acceptable. To insure that the limits of Part 20 are met, Amendment 98 also revised Section 11.2.6.5 of the FSAR to include the statement that no untreated wastes are released unless they are below the Lower Limit of Detection (LLD=5E-7 uCi/cc gross gamma [sic]). [This closes RAI 11-2]

However, it is unclear how this statement is consistent with the calculational basis for Table 11.2-5, column 8, which assumes the release of untreated Steam Generator Blowdown effluents at concentrations up to 3.65E-5 uCi/cc. [RAI 11-16].

TVA RESPONSE:

Section 11.2.6.5 of the FSAR (see Amendment 102) no longer includes the statement that no untreated wastes are released unless they are below the Lower Limit of Detection (LLD=5E-7 uCi/cc gross gamma. Section 11.2.6.5 now addresses releases when the Steam Generator Blowdown effluents are at concentrations up to 3.65E-5 uCi/cc.

3. NRC QUESTION:

The staff concurs with TVAs conclusion that operating for an extended period of time without processing the Condensate Demineralizer backwash or steam generator blowdown, as represented by column 6 of Table 11.2-5, is not acceptable. However, the staff cannot agree that the total activities represented by columns 7 and 8 of Table 11.2-5, meet the activity limit of RM 50-2, since neither includes the effluent (backwash) from the routine regeneration of the Condensate Demineralizers. [RAI 11-15] Similarly, the staff cannot conclude that Tables 11.2-5c and 11.2-5d demonstrate that 10 CFR 20 can be met with untreated steam generator blowdown effluents, since they do not include Condensate Demineralizer regeneration backwash effluents. [RAI 11-11 &12; Follow-up RAI 11-1 and 11-2 are OPEN pending resolution]

TVA RESPONSE:

Column 7 and 8 of Table 11.2-5 and Tables 11.2-5c and 11.2-5d show that the RM 50-2 and 10 CFR 20 limits are met without use of the Condensate Demineralizers so long as restrictions are placed on the Steam Generator Blowdown activity. As stated in the RAI response for item 1 above, Unit 1 is currently operated without use of the Condensate E1-2

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information Demineralizers, since primary to secondary leakage is not significant. It is expected that Unit 2 will operate in the same manner. Since the demineralizers are not used, the Steam Generator Blowdown is not treated and there is no demineralizer blowdown or backwash waste stream. This method of operation is acceptable so long as the 10 CFR 50, Appendix I and 10 CFR 20 limits are met. Note actual plant releases are accomplished and controlled in accordance with the ODCM and releases are not allowed to exceed either the 10 CFR 50, Appendix I or the 10 CFR 20 limits.

4. NRC QUESTION:

Amendment 95 updated population on usage data listed in Table 11.2-6.

Amendments 95 and 100 update the whole body and organ doses for the maximum exposed individual in each critical age group listed in Table 11.2-7. These updates resulted in minor changes to the calculated doses, which still meet the design criteria for liquid effluents in 10 CFR 50 Appendix I. As discussed below, the staff performed independent dose calculations to verify the acceptability of the applicants dose assessment. The staff determined that there is sufficient agreement between the TVAs and the staffs results to conclude that the WBN Unit 2 design meets the design criteria of 10 CFR 50 Appendix I and is therefore acceptable.

However, it is not clear which source term was used as the basis for these calculations. [RAI 11-9; RAI 11-3 OPEN pending resolution of the source term assumption]

TVA RESPONSE:

See response to question 11.3.c in Enclosure 2 for the source term.

5. NRC QUESTION (9):

Verify that the changes made to Table 11.2-7 are to conform this table with TVAs re-evaluation of the offsite doses, as presented in the February 15, 2008, Environmental Impact Assessment. If not, describe the liquid isotopic release values used to calculate these doses.

TVA RESPONSE:

The values in Table 11.2-7 have been verified to be consistent with those found in the Final Supplemental Environmental Impact Statement (FSEIS). The liquid isotopic release values found in Table 11.2-5 column 8 were used to determine the doses in Table 11.2-7.

E1-3

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information

6. NRC QUESTION (10):

Amendment 101 revised Section 11.2.6.5 and Amendment 102 added a footnote, explaining the radwaste process configurations represented by each column of Table 11.2-5. Columns 7 and 8 do not include effluents from the Condensate Demineralizer regeneration (backwash) operations. Since Table 11.2-5 represents total annual curies released, how does TVA intend to operate WBN Unit 2 for an entire year without backwashing the Condensate Demineralizers? If not then justify the position that annual releases consistent with Column 8 will meet the 5 Ci limit of RM 50-2 Paragraph A.2 or demonstrate WBN meets the alternate criteria in RM 50-2, Paragraph A.3.

TVA RESPONSE:

Column 7 and 8 in Table 11.2-5 are providing information that the 10 CFR 50, Appendix I yearly regulatory limits can be met without use of the Condensate Demineralizers with the specified activity limitations on the Steam Generator Blowdown. Unit 1 currently operates without use of the Condensate Demineralizers. The Condensate Demineralizers will not be used unless significant primary to secondary leakage occurs. Since the demineralizers are not used, the Steam Generator Blowdown is not treated and there is no demineralizer blowdown or backwash waste stream. This method of operation is acceptable so long as the 10 CFR 50, Appendix I and 10 CFR 20 limits are met. TVA plans to operate Unit 2 in the same manner as Unit 1. Note actual plant releases are accomplished and controlled in accordance with the ODCM and releases are not allowed to exceed either the 10 CFR 50, Appendix I or the 10 CFR 20 limits.

7. NRC QUESTION (11):

Similarly, justify the position that Tables 11.2-5b, 11.2-5c, and 11.2-5d demonstrate compliance with 10 CFR 20 when Table 11.2-5b does not include steam generator blowdown effluents, and Tables 11.2-5c and11.2-5d, do not include condensate demineralizer backwash effluents.

TVA RESPONSE:

Tables 11.2-5c and 11.2-5d show that the 10 CFR 20 limits are met without use of the Condensate Demineralizers as long restrictions are placed on the Steam Generator Blowdown activity. As stated in the RAI response to Item 1 above, Unit 1 is currently operated without use of the Condensate Demineralizers since primary to secondary leakage is not significant. It is expected that Unit 2 will operate in the same manner. Since the demineralizers are not used, the Steam Generator Blowdown is not treated and there is no demineralizer blowdown or backwash waste stream. This method of operation is acceptable so long as the 10 CFR 50, Appendix I and 10 CFR 20 limits are met. TVA plans to operate Unit 2 in the same manner as Unit 1. Note actual plant releases are accomplished and E1-4

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information controlled in accordance with the ODCM and releases are not allowed to exceed either the 10 CFR 50, Appendix I or the 10 CFR 20 limits.

8. NRC QUESTION (12):

In addition, Tables 11.2-5b, 11.2-5c, and 11.2-5d, only represent one unit operation.

Provide an analysis that demonstrates that the effluents from WBN will not result in a member of the public exceeding the dose limits in Part 20 with both WBN units in operation.

TVA RESPONSE:

The values in the last column of Tables 11.2-5b, 11.2-5c and 11.2-5d for two unit operation will be the sum of the total tritium production core (TPC) value for Unit 1 and the total (non-TPC) value for Unit 2; e.g., for Table 11.2-5b, 3.201E-01 + 2.680E-01= 5.881E-01 curies per year. All these sums are less than unity and thus meet the dose limits of 10 CFR 20.

9. NRC QUESTION (13):

The footnote added to Table 11.2-5 by Amendment 102 appears to have some typographical errors. Verify that the term F/H1D in the formulation of Column 5 and Mobile in the definition of D should be, F/H/D and Mobile respectively.

TVA RESPONSE:

In the footnote added to Table 11.2-5 by Amendment 102, the term F/H1D in the formulation of Column 5 and Mobile in the definition of D should be, F/H/D and Mobile, respectively. These items will be corrected in FSAR Amendment 103.

10. NRC QUESTION (14):

In addition the definitions of the terms F and H used in columns 4, 5, and 6 are somewhat confusing. A plain reading of the footnote would indicate that the entire condensate flow that is processed by the Condensate Demineralizer is released from WBN as liquid effluent. Reading this in the context paragraph 11.2.6.5, as revised by Amendment 101, would indicate that the term F represents the total annual activity in the effluent waste from Condensate Demineralizer regeneration operations, not the Condensate Demineralizer flow. Verify that this is the case. If it is, identify the demineralizer (whose decontamination factors are represented by H in the terms F/H and F/H/D) that the regeneration waste is processed through prior to E1-5

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information processing with the Mobile Demineralizer. If it is not the case, provide additional clarification of the terms F/H and F/H/D in the footnote.

TVA RESPONSE:

The term F in columns 4, 5, and 6 represents the total annual activity in the effluent waste from Condensate Demineralizer regeneration operations. The demineralizer whose decontamination factors are represented by H in the terms F/H and F/H/D that the regeneration waste is processed through prior to processing with the Mobile Demineralizer is the Condensate Polishing Demineralizer.

11. NRC QUESTION (15):

Provide information that demonstrates that operating WBN Units 1 and 2 will meet the liquid effluent criteria in RM 50-2, Paragraph A.1 (e.g., 5 mrem to the total body or to any organ per site).

TVA RESPONSE:

From the Unit 1 UFSAR, Table 11.2-6, the highest Total Body value is 0.72 mrem for an Adult; the highest organ (Liver) value is 1.0 mrem for a Teen. These values are the same for the corresponding Unit 2 FSAR Table 11.2-7. When added together, Units 1 and 2 will meet the liquid effluent criteria in RM 50-2, Paragraph A.1.

12. NRC QUESTION (16):

Resolve the apparent conflict between the statement in Section 11.2.6.5 that no untreated wastes are released unless they are below the Lower Limit of Detection of 5E-7 uCi/cc, and the calculational basis for Table 11.2-5, Column 8 (and Table 11.2-5d) that concludes that untreated releases up to 3.65E-5 uCi/cc are acceptable.

TVA RESPONSE:

Section 11.2.6.5 contained in Amendment 102 does not indicate that no untreated wastes are released unless they are below the Lower Limit of Detection of 5E-7 uCi/cc. Section 11.2.6.5 now addresses releases when the Steam Generator Blowdown effluents are at concentrations up to 3.65E-5 uCi/cc.

E1-6

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information Gaseous Waste Management System

13. NRC QUESTION:

Amendments 95 and 98 also made several revisions to the gaseous effluent release analysis parameters presented in Table 11.3-6 with resulting minor changes to the resulting radioactive releases in Table 11.3-7. The radioactive releases listed in Tables 11.3-7 are based on the radioactive source term assumptions in NUREG-0017, adjusted for WBN specific parameters. Table 11.3-7 represent operations with containment purge, while Table 11.3-7c assumes that containment is continuously vented through a filtered release. [RAI 11-18] Section 11.3.7.5 of the FSAR indicates that the estimated releases in Table 11.3-7c were used by TVA in calculating the site boundary doses presented in Table 11.3-10 to demonstrate compliance with 10 CFR 50 Appendix I.

a) However it is unclear if the source term used for Table 11.3-7c (i.e., 1/8% failed fuel) is comparable to the NUREG-0017 source term [RAI 11-19].

b) Also, as discussed below, it is unclear if the basis for the doses presented in Table 11.3-10 is the isotopic releases listed in Table 11.3-7c or Table 11.3-7. [RAI 11-17; RAI 11-7 OPEN]

TVA RESPONSE:

a) The source terms used as a basis for Table 11.3-7c are based on ANSI 18.1-1984. The Nominal values in ANSI 18.1-1984 are the same values used in NUREG-0017. To develop the WBN source terms, the ANSI 18.1-1984 nominal values were adjusted based on WBN specific plant conditions. Therefore, the source term values used as a basis for Table 11.3-7c are comparable to those in NUREG-0017.

b) The individual doses listed in Table 11.3-10 were determined using each nuclides total curies/year listed in Table 11.3-7c, Total Releases (1/8% failed fuel in Ci/yr), with Continuous Filtered Containment Vent.

14. NRC QUESTION:

Amendments 95, 98, and 99 revised Table 11.3-11 significantly lowing the calculated doses and presenting them in the table on a per-unit basis instead of on a per-site (2 units operating) basis. [RAI 11-24] It appears that these changes were made to conform Chapter 11 of the WBN Unit 2 FSAR with the re-evaluation of public doses presented in TVAs Watts Bar Nuclear Plant (WBN) - Unit 2-Final Supplemental Environmental Impact Statement, (FSEIS - submitted to the NRC by letter dated February 15, 2008). [RAI 11-16] The revised doses contained in the doses in FSAR Table 11.3-10 (Amendment 98), exactly match the doses presented in Table 3-21 of the E1-7

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information FSEIS. In response to the staffs questions (RAI 11-7 and Follow-up question 11-3),

TVA stated that the revised (lower) doses were the result of several changes TVA made to the calculation input parameters, and presenting the doses on a single-unit, versus a duel-unit, basis. TVA stated they updated the X/Q, D/Q and joint frequency tables used in their calculations to reflect updated meteorology (e.g., data from January 1986 to December 2005, versus previous based on January 1974 to December 1993 data). In addition, the feeding factors used to adjust the fraction of the time cows are grazing on exposed pasture, was significantly lowered for all sectors with a milk cow. Amendment 100 revised the Table 11.3-8 to reflect the revised input parameters. Several compass sectors, distances, and terrain adjustment factors in Table 11.3-8 were also changed to reflect an updated land-use census.

The staff reviewed the changes in Amendments 95, 98, 99, and 100, against the information in the FSEIS and Appendix I of NUREG-0498, Supplement 2, and identified several discrepancies. The FSEIS states that the doses in FSEIS Table 3-21 are based on the FSEIS Table 3-20, which is consistent with Table 11.3-7 of the FSAR. This seems inconsistent with the statement noted above, that the doses in FSAR Table 11.3-10 (identical to FSEIS Table 3-21) are based on the significantly different radioactive quantity values in FSAR Table 11.3-7c. [RAI 11-17 & 18] In addition, although the doses listed in FSEIS Table 3-21 are identical to those in FSAR Table 11.3-10, the former indicates that the maximum thyroid dose was based on a cow feeding factor of 0.65, while the later indicates that the dose was based on a cow feeding factor of 0.33 (also listed as 0.33 in Amendment 100 to FSAR Table 11.3-8).

Neither of these values agrees with the 0.70 feeding factor given in FSAR Section 11.3.10.1. [RAI 11-20] Several of the distances and directions for the locations of the calculated doses given in FSAR Table 11.3-8 (Amendment 100) do not agree with the information in the FSEIS. [RAI 11-23; RAI 11-4, 11-7, and Follow-up question 11-3 OPEN]

The staff performed independent dose calculations to verify TVAs dose results. The details of the staffs calculations and input parameters assumptions can be found in Appendix I of NUREG-0498, Supplement 2. With the exception of the iodine/thyroid doses, the staffs results generally agree with the TVAs calculations. Bases on its conservative assumptions, the staffs calculations determined that the maximum exposed organ expected from radioactive iodine and particulates in gaseous effluents, is 10.78 mrem. Although both TVAs and the staffs calculations indicate that the design criteria in 10 CFR 50 Appendix I are met (15 mrem per year per unit),

they are not sufficient to determine if the criteria in RM 50-2 are met (15 mrem per year from all light-water-cooled nuclear power reactors at a site).

Therefore, the staff cannot confirm that the WBN Unit 2 can be operated within the dose restrictions of RM 50-2. [RAI 11-3 OPEN]

E1-8

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information Verify that the basis for the Amendment 98 changes to Table 11.3-10 is the revised TVA analysis of the offsite radiation doses as presented in the Final Supplemental Environmental Impact Statement (FSEIS), submitted by letter dated February 15, 2008.

If this is not the case, describe the basis for the revised values in Table 11.3-10.

TVA RESPONSE:

TVA has reviewed the FSEIS and found the land use data presented in Table 3-19 to be in error. Table 11.3-10 of the FSAR will be corrected to reflect the 2007 feeding factors and the offsite radiation doses calculated without terrain adjustment factors. These changes to Table 11.3-10 will be reflected in Amendment 103. A mark-up of the FSEIS, Table 3-19 is provided in Enclosure 3 for NRC information to facilitate review.

15. NRC QUESTION (18):

FSAR Section 11.3.7.5 indicates that the site boundary doses presented in Table 11.3-10 are based on the annual radioactive gaseous releases listed in Table 11.3.7c.

However, the FSEIS indicates that these dose values are based on a source term consistent with FSAR Table 11.3.7. Verify the gaseous release values used to calculate the site boundary doses, and/or explain how two significantly different source terms arrive at the exact same calculated doses.

TVA RESPONSE:

TVA has reviewed the FSEIS and found Table 3-20 to be in error. This was caused by the use of values contained in FSAR Table 11.3.7 instead of values contained in FSAR Table 11.3.7c. The correct source term used for calculating the site boundary doses is FSAR Table 11.3.7c. This accounts for the dose values being same between the FSEIS and the FSAR Table 11.3-10. A mark-up of the FSEIS, Table 3-20 is provided in Enclosure 3 for NRC information to facilitate review.

16. NRC QUESTION (19):

The Continuous Filtered Containment Vent case (Table 11.3-7c) has significantly lower activities for all of the Krypton, Xenon, and Iodine isotopes, than those estimated for the containment purge case listed in Tables 11.3-7, while the other particulate activities released from the Containment Building remain the same.

Describe the filter that selectively removes noble gases and iodine species but not other particulates from the Containment Building Vent gaseous effluents. Provide a basis for assuming normal operations with the containment vent continuously open.

Provide, and justify, the Decontamination Factors (by each isotope class) assumed for continuous containment vent filter.

E1-9

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information TVA RESPONSE:

Particulate releases are taken directly from NUREG-0017 with the 99% HEPA filtration efficiency applied. Therefore these values are independent of the case.

The Noble Gas and Iodine values are calculated separately from the particulates. There is a difference between the two cases because of the differences in the amount of air vented/purged. The first case is continuous venting assumed at 100 cfm for an entire year equates to 7.15E11 cc, where the second case is the purge case assumes 26 cfm (12 hr purges from upper and lower containment and the instrument room) for a total volume of 1.22E13 cc purged. Therefore, since the volumes and source terms are the same, less activity is released for the continuous vent case.

The basis for operating with the containment vent continuously open is that it has been shown the 10 CFR 50 Appendix I limits can be met with this path open. This flow path is automatically closed by a containment vent isolation signal in the event of an accident.

The only decontamination factors used are for the HEPA and charcoal filters which use 70%

for halogens and 99% for particulates, as given in NUREG-0017 Table 1-5 and Section 1.5.2.16.2.

17. NRC QUESTION (20):

Verify that the 1/8% failed fuel source term used as the basis for Table 11.3-7c is comparable to the source term specified in NUREG-0017. If not justify the use of this source term for determining nominal effluent release values.

TVA RESPONSE:

The source terms used as a basis for Table 11.3-7c are based on ANSI 18.1-1984. The Nominal values in ANSI 18.1-1984 are the same values used in NUREG-0017. To develop the WBN source terms, the ANSI 18.1-1984 nominal values were adjusted based on WBN specific plant conditions. Therefore, the source term values used as a basis for Table 11.3-7c are comparable to those in NUREG-0017.

18. NRC QUESTION (21):

The response to RAI 11-4, and the revisions to Table 11.3-8 (Amendment 100) are inconsistent with the text in the FSAR and the FSEIS. Section 11.3.10.1 indicates that the doses are based on the 1994 land-use survey and that a cow feeding factor of 70%

was used. In addition, FSEIS Table 3-21 indicates that a cow feeding factor of 0.65 was used to evaluate the iodine/particulate maximum organ dose value. Resolve these conflicts.

E1-10

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information TVA RESPONSE:

TVA has reviewed FSAR Section 11.3.10.1, Assumptions and Calculation Methods, and found that it incorrectly states the dose to the critical organ from radioiodines, tritium, and particulates is calculated for real pathways existing at the site during a land use survey conducted in 1994. The feeding factor of 70% is the feeding factor associated with the 1994 land use survey. The feeding factors should be from the 2007 Land Use Survey, which is 0.33%. The feeding factor of 65% listed in Table 3-21 of the FSEIS is in error. These changes to FSAR Section 11.3.10.1 will be reflected in Amendment 103. A mark-up of the FSEIS, Table 3-21 is provided in Enclosure 3 for NRC information to facilitate review.

19. NRC QUESTION (22):

Provide a justification for each of the cow feeding factors listed in Table 11.3-8.

TVA RESPONSE:

The feeding factors (fraction of time on pasture) are based upon three farms near the WBN site area. The 2007 data for these three farms are provided below:

Percent Substitutional Feeding for Dairy and Goat Herds 2007 Farm Distance Total /

(meters) Jan Feb Mar Apr May Jun Jul Aug Sep Oct Nov Dec TOTAL 1200 FF 6706 ESE 100 100 100 95 95 95 95 95 95 100 100 100 1170 0.975 0.025 2286 SSW 100 100 100 90 90 90 90 90 90 100 100 100 1140 0.95 0.05 3353 SSW WILL NOT PARTICIPATE IN LAND USE SURVEY 0.33*

• This conservative feeding factor assumes a consumption of the milk by an adult.

20. NRC QUESTION (23):

Describe how the revised (Amendment 100) terrain factors in Table 11.3-8 were determined.

TVA RESPONSE:

TVA uses GELC (Gaseous Effluent Licensing Code) to perform routine dose assessments required by NRC Guide 1.111. For WBN, the NRC stated that adjustments to the GELC results were necessary to account for recirculation effects of spatial and temporal variations in airflow in the vicinity of pronounced river valleys.

E1-11

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information TVA developed site-specific adjustment factors for WBN by comparing results from the GELC model with results from the MESOPUFF II model. These adjustment factors are revised each year to reflect changes based on annual surveys.

Studies performed during 2010 for development of an American Nuclear Society (ANS) standard (specifically by the ANS-2.15 recirculation sub-group) determined that the adjustment factor approach is not acceptable for addressing recirculation issues.

Further, comparisons with other models determined that MESOPUFF II is not suitable for calculating /Q values at WBN receptors, and that GELC adequately estimates /Q for WBN receptors, without any need for adjustments. Therefore, WBN can eliminate the use adjustment factors and use GELC results directly.

These changes will be reflected in Table 11.3-8 in FSAR, Amendment 103. A mark-up of the FSEIS, Table 3-19 is provided in Enclosure 3 for NRC information to facilitate review.

21. NRC QUESTION (24):

Footnote 4 to Table 11.3-10 (Amendment 98) indicates that the maximum thyroid dose is for an infant at 3353 meters in the SSW sector. However, the revised (Amendment 100) Table 11.3-8 data indicates that the 0.33 feeding factor is applied to the location at 3353 meters in the SW direction. In addition, Table I-9 of the FSEIS indicates that the max thyroid/iodine dose is for an individual at 1.42 miles (2285 meters) in the SSW direction. a) Resolve these conflicts. b) Provide information describing how two unit operations at WBN will be within all of the dose criteria in RM 50-2 for gaseous releases.

TVA RESPONSE:

a) TVA has reviewed the FSEIS and found the land use data presented in Table 3-19 to be in error. The land use survey used to develop FSAR Table 11.3-10 was from 2007.

Table 11.3-10 of the FSAR will be revised to include 2007 feeding factors and the offsite radiation doses being calculated without terrain adjustment factors. These changes to Table 11.3-10 will be reflected in Amendment 103. A mark-up of the FSEIS, Table 3-19 is provided in Enclosure 3 for NRC information to facilitate review.

b) The corresponding Unit 1 FSAR table is being revised in the same manner as described in response to question 11.3a in Enclosure 2. When the Unit 1 and Unit 2 tables are combined, the results will be evaluated against the criteria of RM 50-2. The Unit 1 values are similar in magnitude to the Unit 2 values and thus the sum of the two units will meet the RM 50-2 criteria.

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Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information

22. NRC QUESTION:

In WBN Unit 2 FSAR Amendment 95, TVA revised Section 12.2.1.3, Sources During Refueling, to include a discussion of the incore instrumentation thimble assemblies (IITAs) as important radioactive sources during refueling operations. The discussion replaced the previous discussion of the incore detector bottom-mounted instrumentation (BMI) thimble tubes in FSAR Section 12.2.1.3 and Table 12.2-3, Chemical and Volume Control System Seal Water Return Filter. In its letter dated June 3, 2010, responding to NRC staff questions (RAI 12-1), TVA stated that the IITAs and BMI thimble tubes would be exposed to the same neutron flux during power operations and therefore would exhibit radiation dose rates of similar magnitude. The radiological hazards posed by this source term change should be no greater than previously described. Therefore, these changes are acceptable to the staff. TVA should provide an update to the FSAR replacing Table 12.2-3 with the expected source strength values of the freshly irradiated IITAs.

TVA RESPONSE:

TVA will provide an update in a future FSAR amendment.

23. NRC QUESTION:

12.4 Radiation Protection Design Features In FSAR Amendment 97, TVA deleted FSAR Figures 12.3-18 and 19. These figures contained the drawings of WBN radiation protection design features, including controlled access areas, decontamination areas, and onsite laboratories and counting rooms. In lieu of providing drawings depicting these radiation protection design features, TVA provided a description of each. In response to a staff question (RAI 12-

7) regarding the FSAR changes, TVA provided clarifying information in its letters dated June 3 and October 4, 2010. In its October 4, 2010, letter, TVA stated that the WBN Unit 2 access controls to radiological areas (including contaminated areas),

personnel and equipment decontamination facilities, onsite laboratories and counting rooms, and Health Physics facilities (including dosimetry issue, respiratory protection bioassay, and Radiation Protection Management and technical staff) are all common to Unit 1. Furthermore, TVA stated that these facilities are sized and situated properly to support two operating units. Based on TVAs response, the staff concluded that the FSAR changes did not impact the staffs previous safety conclusion, as documented in SSER 18, dated October 1995. Therefore, the changes are acceptable. TVA should provide an update to the FSAR reflecting the information provided in its letter dated October 4, 2010.

TVA RESPONSE:

E1-13

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information TVA will provide an update in a future FSAR amendment.

24. NRC QUESTION:

In FSAR Amendment 97, TVA revised the frequency of the radiation monitor channel operability tests from quarterly to periodically. In its letter dated June 3, 2010, TVA responded to a staff question (RAI 12-8) about what frequency was meant by periodically. In its response, TVA provided a WBN Unit 1 FSAR change package as justification for relaxing the interval between monitor channel operability tests from quarterly to 9 months (a calculated 18 months with a margin factor of two). The staff reviewed TVAs response and the change package, but could not conclude that TVA has provided adequate technical justification to relax the quarterly operability tests.

TVA RESPONSE:

TVA reviewed the subject calculation and determined that it was inadequate to support extending the quarterly operability tests. The evaluation determined that the issue was with the calculation methodology and not the data. The evaluation also determined that it was probable that if the calculation was re-performed correctly it would support extending the quarterly operability test interval.

As a result, the calculation was re-performed and the results supported extending the quarterly operability test interval. Attachment 1 to this letter contains TVA calculation WBN-EEB- EDQ1090-99005, Revision 1, Extending Channel Operational Test Frequency for Radiation Monitors.

25. NRC QUESTION:

In FSAR Amendment 97, TVA also revised the description of the airborne monitoring channels in Section 12.3.4.2.4, Component Descriptions, to reflect the replacement of the seven (7) channels of airborne monitors previously indicated for the Auxiliary Building with four (4) portable airborne monitors. TVA stated in the FSAR that the portable airborne monitors will have a sufficient sensitivity to detect a 10 derived air concentration (DAC)-hour change in airborne radioactivity. In response to a staff question (RAI 12-10), TVA provided additional information in its letter to the NRC dated June 3, 2010, regarding the replacement of the airborne monitors. The use of portable airborne monitors reflects the current operational configuration of Unit 1, and is acceptable to the staff. However, the revised FSAR Section 12.3 contains no discussion of the calibration and operability testing of the portable airborne radiation monitors that replace the seven channels of fixed airborne monitors. The staff lacks sufficient information to determine that these monitors meet the acceptance criteria E1-14

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information in the SRP and thus will provide adequate airborne monitoring at WBN Unit 2, consistent with the requirements of Subpart F, Surveys and Monitoring, of 10 CFR Part 20, § 20.1501.

TVA RESPONSE:

The four portable monitors listed in FSAR Table 12.3-5 are calibrated every 6 months in accordance with site Radiological Control Instructions. This meets the requirements of Subpart F, Surveys and Monitoring, of 10 CFR Part 20, § 20.1501, which requires periodic calibration of the monitors. Weekly source checks are performed in accordance with site Radiological Control Instructions. This meets the requirements of Reg. Guide 8.25 Revision 1.

26. NRC QUESTION:

In FSAR Amendment 101, TVA further revised the description in Section 12.3.4.1.3, Area Monitor Calibration and Maintenance, addressing the calibration and operability testing of area radiation monitors. Rather than specifying appropriate testing frequencies, the revision refers to licensing or TVA program requirements.

The staff lacks sufficient information to determine that these licensing or TVA program requirements are sufficient to meet the regulatory requirements of Subpart F of 10 CFR Part 20, § 20.1501.

TVA RESPONSE:

Subpart F of 10 CFR Part 20, § 20.1501 states:

(b) The licensee shall ensure that instruments and equipment used for quantitative radiation measurements (e.g., dose rate and effluent monitoring) are calibrated periodically for the radiation measured.

The statement licensing or TVA program requirements is made to document the source of testing requirement. The first sentence of the paragraph states: With the exception of the Reactor Building upper and lower compartment post accident monitors, periodic testing of each area monitor includes a channel calibration performed at least once per 22.5 months (18 months plus 25%). This statement provides the information required by Subpart F of 10 CFR Part 20, § 20.1501 for all except the upper and lower containment post accident monitors which the final sentence states are calibrated in accordance with technical specifications. Surveillance requirement SR 3.3.3.2 requires that the upper and lower containment post accident monitors are calibrated at 18 month intervals.

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Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information

27. NRC QUESTION:

In FSAR Amendment 97, TVA added a description of two area radiation monitors for the Spent Fuel Pit (0-RE 90-102 and 103) to the list of monitors in Table 12.3-4, Location of Plant Area Radiation Monitors. In response to a question from the staff (RAI 12-9), TVA responded in its letter dated June 3, 2010, that it would provide information to demonstrate compliance with the requirements of 10 CFR 70.24 and 10 CFR 50.68. At this time, the staff lacks sufficient information to determine that these monitors meet the criteria in 10 CFR 70.24, Criticality accident requirements, and 10 CFR 50.68, Criticality accident requirements, for radiation monitoring in areas where fuel is handled or stored.

TVA RESPONSE:

The referenced CFR requirements relate to criticality monitors for areas where reactor fuel is handled or stored. NRC issued an exemption from the requirements of 10 CFR 70.24 as part of the Unit 1 operating licensing. See the following excerpt from section 2.D.(2) of the Unit 1 operating license, which has been incorporated into the Unit 1 Technical Specifications:

2.D.(2) The facility was previously granted an exemption from the criticality monitoring requirements of 10 CFR 70.24 (see Special Nuclear Material License No. SNM-1861 dated September 5, 1979). The technical justification is contained in Section 9.1 of Supplement 5 to the Safety Evaluation Report, and the staff's environmental assessment was published on April 18, 1985 (50 FR 15516). The facility is hereby exempted from the criticality alarm system provisions of 10 CFR 70.24 so far as this section applies to the storage of fuel assemblies held under this license.

Since the new fuel and spent fuel storage areas are common to both units, TVA concluded that criticality monitors are not required for WBN in areas where the fuel is handled or stored. This is also consistent with TVAs application for Special Nuclear Material License dated November 12, 2009.

Compliance with 10 CFR 50.68(b) is documented in FSAR Section 4.3.2.7, Criticality of Fuel Assemblies.

28. NRC QUESTION:

12.5 Dose Assessment Based on the information provided by TVA in its letter to the NRC dated June 3, 2010, and because historical experience has demonstrated that the average annual collective dose to operate WBN Unit 1 was less that 100 person-rem, the staff E1-16

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information concludes that there is reasonable assurance that WBN Unit 2 can be operated at or below 100 person-rem average annual collective dose. Therefore, FSAR Section 12.4 is acceptable. TVA should update the FSAR to reflect the information provided in its letter the NRC dated June 3, 2010.

TVA RESPONSE:

TVA will provide an update in a future FSAR amendment.

29. NRC QUESTION:

12.6 Health Physics Program In FSAR Amendment 95, TVA made several editorial changes to FSAR Section 12.5 resulting from organizational changes at WBN. With the exception of the following two issues, these did not impact the staffs previous safety conclusion, as documented in SSER 14, dated December 1994, and are therefore acceptable. The remaining two issues are related to the Radiation Protection Manager (RPM) qualifications. FSAR Section 12.5.1 states that, The minimum qualification requirements for the Radiation Protection Manager are stated in Section 13.1.3.

FSAR Section 13.1.3 states that, Nuclear Power (NP) personnel at the Watts Bar plant will meet the qualification and training requirements of NRC Regulatory Guide 1.8 with the alternatives as outlined in the Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A. Specifically, TVA modified its commitment to the personnel qualification standards in Regulatory Guide (RG) 1.8, Qualification and Training of Personnel for Nuclear Power Plants, by adding the caveat, with the alternatives as outlined in the Nuclear Quality Assurance Plan. It was unclear to the staff whether or not TVA was committed to (1) the requirement that the RPM have five years of professional experience, and 2) the three month time limit on temporarily assigning an RPM who doesnt meet the RPM qualifications (ANSI/ANS 3.1-1981, as referenced in RG 1.8). In response to staff questions (RAIs 12-13 and 12-14), TVA clarified in its letter to the NRC dated October 4, 2010, that it will meet the requirements of RG 1.8, Revision 2, and ANSI/ANS 3.1-1981, for all new personnel qualifying on positions identified in RG 1.8, Regulatory Position C.1, after January 1, 1990. These changes are consistent with the staffs acceptance criteria 12.5.A of Section 12.5 of the SRP as they pertain to staff qualifications and are, therefore, acceptable. TVA should update the FSAR to reflect the qualification standards of the RPM as provided in its letter to the NRC dated October 4, 2010.

TVA RESPONSE:

TVA will provide an update in a future FSAR amendment.

E1-17

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information

30. NRC QUESTION:

12.7 NUREG-0737 Items In FSAR Amendment 97, TVA revised the list in FSAR Section 12.3.2.2, Design Description, of post accident activities that need to be accomplished, adding three and deleting the activities at the post accident sampling facility. The staff requested information (RAI 12-6) regarding the dose consequences of these vital missions, including plant layout drawings depicting radiation zones during accident conditions and access/egress routes. By letters dated June 3, 2010, and December 10, 2010, TVA provided dose calculations and plant layout drawings depicting the WBN vital area access/egress routes. The staff noted a number of inconsistencies and deficiencies in the information provided by TVA. These include, but are not limited to:

1) There is not a good correlation between the list of vital areas in FSAR Section 12.3.3, the calculations provided, and the layout drawings, e.g.,

a. Not all vital areas listed in Section 12.3.3 have corresponding calculations or maps (i.e., TSC, control room access/egress).

TVA RESPONSE:

Continuous occupancy of the TSC and Main Control Room (MCR) is required during accident conditions (the TSC is within the MCR habitability zone and has the same dose as the MCR). The accident doses for the MCR/TSC include ingress and egress and are reported in FSAR Chapter 15.5. Consequently, dose maps of the MCR/TSR are not necessary.

b. Not all vital areas indicated in the calculations and maps are listed in the FSAR (e.g., OSC, WBNTSR-114, WBNTSR-084).

TVA RESPONSE:

The OSC is an area from which accident missions are dispatched, dose permitting.

If the accident dose in the OSC is prohibitive, missions can be dispatched from the TSC. The mission dose calculations are done from both the OSC and TSC.

Consequently, the OSC is not considered a vital area relative to dispatch of accident missions. FSAR section 12.3.2.2 will be revised to list any applicable additional areas addressed by the mission dose calculations.

c. Not all calculations (i.e., WBNTSR -086) have corresponding maps.

E1-18

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information TVA RESPONSE:

Calculation WBNTSR-086 is for general surveys of four elevations of the auxiliary building during accident conditions to identify piping and component leaks. Since this is a general area, survey specific locations requiring survey within the building areas are not identified. Consequently, survey maps of the areas are not applicable.

The calculation establishes the general area dose rates and estimated time required to complete the surveys.

2) Several calculations and maps included in the response clearly demonstrate that GDC 19 dose criteria will not be met during the proposed vital area missions.

TVA RESPONSE:

Calculation WBNTSR-087 evaluated refill of the Refueling Water Storage Tank from several different sources. All sources except refill from the spent fuel pit could not be accomplished within the GDC 19 dose limitations. However, the mission can be accomplished from the spent fuel pit source. Several other missions exceed the GDC dose limitations for thyroid dose if self contained breathing apparatus (SCBA) are not utilized. However, in this case, use of SCBA is a special requirement of the calculations.

In summary, all missions can be accomplished within the GDC 19 dose limitations utilizing the special requirements of the calculations.

3) The source term used in the evaluation of a steam generator tube rupture (WBNTSR-084) is not consistent with the source term required in the Design Basis Accident analysis in Chapter 15 of the FSAR (e.g., does not consider an iodine spike in the primary coolant).

TVA RESPONSE:

The liquid source term used for the sample in WBNTSR-084 is the normal RCS source term, which is based on ANSI/ANS 18.1, 1984. The airborne activity used for the mission is that of a LOCA. It is expected that use of the LOCA source terms will bound use of the RCS source term with an Iodine spike. However, TVA will perform the calculation using the steam generator tube rupture source term.

4) Several calculations do not address whether the GDC 19 dose criteria are met, but instead calculate a maximum staytime before exceeding a pre-determined limit, with no indication if the identified access/egress vital action can be performed within the calculated results or whether the pre-determined criteria ensures that GDC 19 will be met.

E1-19

Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information TVA RESPONSE:

Calculations WBNTSR-081 and WBNTSR-082 calculated a maximum stay time before exceeding the GDC 19 dose limits. Both these calculations also calculated the mission dose for a 1/2 hour mission. These calculations will be revised to clarify times required to perform the missions.

5) Several calculations identify an alternate, more limiting accident scenario (labeled EGTS PCO Control Loop Single Failure) without identifying what this scenario is, or why it is the limiting case. In at least two of the calculations (WBNAPSR 87 and

94) this limiting case is only calculated for Unit 1, with a note that the Unit 2 impact will have to be evaluated at a later date.

TVA RESPONSE:

The mission dose calculations originally considered a single failure of one train of Emergency Gas Treatment System (EGTS) concurrent with a LOCA. An EGTS Pressure Control Operator (PCO) Control Loop Single Failure was also considered in the calculations due to a corrective actions program requirement. This new failure (scenario) is also described in the calculation revision log. The two different single failures resulted in different exhaust flows out of the Annulus to the outside environment.

The mission dose was separately calculated for each of these single failures and was shown to be either bounded by the original single failure or resulted in doses less than the GDC 19 dose limits. Mission dose calculations that are currently only applicable to Unit 1 are being updated to make them applicable to Unit 2. The conclusions of the calculations are not expected to change with these revisions.

6) Several of the calculations have lists of operational restrictions (i.e., WBNAPS3 -

124 and 125) with no indication of whether the vital action can be completed within these restrictions, nor is there any indication of how TVA will insure these restrictions will be met.

TVA RESPONSE:

Calculations WBNAPS3-124 and WBNAPS3-125 were issued for design change package EDC 56203. The normal design change control process, as described in procedure NPG-SPP-09.3, requires coordination of changes and special requirements with plant organizations. As part of this process the plant organizations are required to identify procedures that must be revised to incorporate the design output, including special requirements. The procedures must be revised prior to closing the design change. Ability to perform the special requirements is confirmed as part of the procedure revision process.

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Enclosure 1 Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information

7) Several of the dose calculation conclusions state, Therefore, the mission can be performed as long as the sum of occupancy, ingress/egress, and mission doses, for the entire duration of the accident, does not exceed the stated limit. It is unclear to the staff whether or not these mission doses comply with GDC 19. If this statement is intended to indicate that each of the mission dose calculations assumes that the operator has no prior accident-related dose, there should be an assurance that sufficient operators are available to complete all of the necessary missions to mitigate the consequences of the accident.

Based on the above, the NRC staff has insufficient information to conclude that TVA has taken appropriate actions to reduce radiation levels and increase the capability of operators to control and mitigate the consequences of an accident at WBN Unit 2, in accordance with the guidance of NUREG-0737, Item II.B.2, or can maintain occupational doses to plant operators within the requirements of GDC

19. Therefore, the staff cannot conclude that the plant shielding for WBN Unit 2 is acceptable.

TVA RESPONSE:

The intent of the mission dose calculations is to show that critical missions can be accomplished during accident conditions and the dose will remain within the GDC 19 dose limitations. In actual practice, overall doses to plant personnel during accident conditions will be monitored and controlled by Site Radcon during accident conditions under the Radiological Emergency Plan. Individuals performing high dose missions can be released from the site prior to exceeding overall dose limits. Similarly, individuals who have accrued a significant dose prior to performing missions will not be tasked with performing the mission if exceeding the dose limitations is possible. This plan ensures that overall doses to plant personnel remain within regulatory limits during accident conditions. In addition to Operations personnel, many of the mission dose actions are performed by plant support personnel such as Chemistry and Radcon.

Consequently, the plant is adequately staffed to perform the necessary missions and perform other necessary functions during accident conditions and remain within the applicable regulatory dose limitations.

E1-21

Enclosure 2 Watts Bar Nuclear Plant Remaining Chapter 11 Request for Additional Information Preliminary RAIs for FSAR 11 (taken from e-mail from NRC dated 03/23/2010)

Section 11 NRC Question:

3.c Table 11.2-7-Identify the specific source term, models, parameters, and assumptions used in calculating these values.

TVA RESPONSE:

Source Term The source term used in calculating Table 11.2-7 was taken from the following design output documents.

The Liquid Radwaste is addressed by Calculation No. TVAN WBNTSR-093 (Liquid Radioactive Waste Release), which is based on NUREG-0017.

The Steam Generator Blowdown is addressed by Calculation No. WBNTSR-100 (Design Releases to Show Compliance with 10 CFR 20).

Single Unit Steam Single Unit Liquid Generator Single UnitTotals Nuclide Radwaste Blowdown Ci/yr Ci/yr Ci/yr Br-84 1.65E-04 5.23E-04 6.88E-04 I-131 2.63E-02 1.14E+00 1.16E+00 I-132 1.32E-02 1.08E-01 1.21E-01 I-133 5.29E-02 8.57E-01 9.10E-01 I-134 6.26E-03 2.65E-02 3.28E-02 I-135 4.75E-02 4.22E-01 4.70E-01 Rb-88 6.89E-03 7.84E-04 7.68E-03 Cs-134 2.93E-02 1.68E-01 1.98E-01 Cs-136 2.55E-03 1.72E-02 1.98E-02 Cs-137 4.03E-02 2.21E-01 2.61E-01 Na-24 1.86E-02 0.0E+00 1.86E-02 Cr-51 7.03E-03 9.27E-02 9.98E-02 Mn-54 4.99E-03 5.10E-02 5.59E-02 Fe-55 8.09E-03 0.0E+00 8.09E-03 Fe-59 2.42E-03 9.05E-03 1.15E-02 Co-58 2.20E-02 1.44E-01 1.66E-01 Co-60 1.44E-02 1.72E-02 3.16E-02 E2-1

Enclosure 2 Watts Bar Nuclear Plant Remaining Chapter 11 Request for Additional Information Single Unit Steam Single Unit Liquid Generator Single UnitTotals Nuclide Radwaste Blowdown Ci/yr Ci/yr Ci/yr Zn-65 3.82E-04 0.0E+00 3.82E-04 Sr-89 1.92E-04 4.33E-03 4.52E-03 Sr-90 2.20E-05 3.88E-04 4.10E-04 Sr-91 2.84E-04 2.18E-03 2.47E-03 Y-91m 1.68E-04 0.0E+00 1.68E-04 Y-91 9.00E-05 3.00E-04 3.90E-04 Y-93 1.27E-03 0.0E+00 1.27E-03 Zr-95 1.39E-03 1.20E-02 1.34E-02 Nb-95 2.10E-03 8.98E-03 1.11E-02 Mo-99 4.20E-03 9.95E-02 1.04E-01 Tc-99m 3.35E-03 0.0E+00 3.35E-03 Ru-103 5.88E-03 0.0E+00 5.88E-03 Ru-106 7.63E-02 0.0E+00 7.63E-02 Te-129m 1.41E-04 0.0E+00 1.41E-04 Te-129 7.30E-04 0.0E+00 7.30E-04 Te-131m 8.05E-04 0.0E+00 8.05E-04 Te-131 2.03E-04 0.0E+00 2.03E-04 Te-132 1.11E-03 2.93E-02 3.05E-02 Ba-140 1.02E-02 3.48E-01 3.58E-01 La-140 1.62E-02 4.98E-01 5.14E-01 Ce-141 3.41E-04 0.0E+00 3.41E-04 Ce-143 1.53E-03 0.0E+00 1.53E-03 Ce-144 6.84E-03 1.26E-01 1.33E-01 Np-239 1.37E-03 0.0E+00 1.37E-03 H-3 1.25E+03 0.0E+00 1.25E+03 Totals w/o H-3 4.38E-01 4.40E+00 4.84E+00 Totals w/ H-3 1.25E+03 4.40E+00 1.26E+03 In order to ensure that the meaning of the column headings is clear, it is noted that the above numbers are for a single unit rather than for Unit 1. Unit 1 utilizes a tritium producing core (TPC) and thus has different values for the corresponding table.

Assumptions

1. Only the mobile demineralizers will be used for processing of liquid radwaste.

2. All sources, except the Laundry and Hot Shower Tank (LHST) and condensate resin regeneration waste, are collected for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (resulting in about 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> average holdup) prior to release, then discharged instantaneously to the mobile demineralizers for decontamination prior to release to the environment. The condensate resin regeneration E2-2

Enclosure 2 Watts Bar Nuclear Plant Remaining Chapter 11 Request for Additional Information waste collects for 6 days, and the LHST is discharged directly to the environment. An exception to this is the case when there is no processing of the condensate by the Condensate Polishing Demineralizers, and the Steam Generator Blowdown is released directly to the river without processing (this will be a continuous release).

3. This calculation assumes a 365 day/yr/unit operation (i.e., 100% capacity factor) since the plant runs with 18 month fuel cycles; therefore, it is conceivable for the plant to run for the entire year.

4. Only one unit operation is addressed.

5. The unplanned release, which is added to the total, is assumed to be 0.16 Curies/yr based on NUREG-0017, section 2.2.23.1 (1).

6. Liquid Tritium release is 90% of 0.4 Ci/yr/MWt = 0.9

• 0.4

• 3480 = 1262.80 Ci/yr based on NUREG-0017, section 2.2.17.1. The MWt is based on 102% of a nominal power of 3411 MWt.

Model The computer code STP (as described in FSAR Section 15.5.3) is used to determine the annual discharge due to the combination of the Auxiliary Building tanks (Reactor Coolant Drain Tank (RCDT), Turbine Drain Collector Tank (TDCT), Floor Drain Collector Tank (FDCT)), Chemical Volume Control System (CVCS) Letdown, the Turbine Building (TB), and the condensate regeneration waste (consisting of 6 day collection of Steam Generator Blowdown [SGB] and condensate flow). The model consists of a continuous source (all isotopes except noble gasses and N-16) of either Reactor Coolant (RC) and/or Secondary Side Coolant (SSC) and/or Secondary Side Steam (SSS) into an arbitrary volume of 1 tank for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or 6 days, as appropriate. The noble gas daughter products are removed from the volume. The RC, SSC and SSS concentrations consist of ANSI/ANS-18.1-1984 expected reactor coolant, secondary side coolant, and secondary side steam adjusted to WBN operating parameters at 105% power.

The ANSI/ANS-18.1-1984 source is essentially the same as NUREG-0017. The continuous source flow is based on NUREG-0017 values. All sources are summed with an appropriate weighting fraction (from NURGEG-0017) to take dilution into account. The weighting fraction is expressed in terms of fraction of Primary Coolant Activity (PCA).

Parameters Below is a compilation of all leaks/effluents. Unless otherwise specified, the values are from NUREG-0017 Table 1-3. The leakage values are for 1 unit. The isotopes used in the analysis are only those listed in NUREG-0017. For the case of no condensate demineralizer processing of condensate, the regeneration waste is deleted from the total release. Also for this alternate case, the SGB component is modified by multiplying the appropriate Condensate Polishing E2-3

Enclosure 2 Watts Bar Nuclear Plant Remaining Chapter 11 Request for Additional Information Demineralizer decontamination factor of each isotope (essentially undoing the credited processing) to the inventory of each isotope in order to establish the release without processing.

a) Reactor Coolant Pump Seal leakage, 20 gal/day @ 0.1 PCA b) Reactor Containment Cooling System, 500 gal/day @ 0.001 PCA c) Other leaks and drains, 10 gal/day @ 1.67 PCA d) Primary Coolant equipment drains, 80 gal/day @ 1.0 PCA e) Reactor Coolant sampling, 200 gal/day @ 0.05 PCA f) Spent Fuel Pit Liner drains, 700 gal/day @ 0.001 PCA g) Auxiliary Building Floor Drains, 200 gal/day @ 0.1 PCA h) Secondary System Sampling, 1400 gal/day @ 1 PCA (of SSC) (note: NUREG-0017 uses 1E-4 PCA [RC], this calculation uses actual SSC activities, therefore PCA=1 SSC) i) CVCS letdown (via Holdup Tanks), 845 lb/hr (2431.654 gal/day) @ 1 PCA j) Condensate Resin Regeneration Waste consisting of:

1) SGB blowdown = 3E4 lb/hr (86330.93 gal/day) @ 1 PCA (of SSC)

2) Condensate flow = 1.5E7 lb/hr (steam flow) *0.55 (flow split) = 8.25E6 lb/hr @ 1PCA (of SSS) k) Turbine Building floor drains, 7200 gal/ day @ 1 PCA (of SSC) (note: no RC in Turbine Building).

l) LHST release taken directly from NUREG-0017 Table 2-27.

For the condensate regeneration waste, the continuous source varies according to element class, as the Condensate Polishing Demineralizers have variable Decontamination Factors (DFs). The DFs are 0.5 for Cs, Rb; 0 for H3; and 0.9 for I, Br, all others.

The decontamination factors are based on NUREG-0017 and/or vendor data. The various decontamination factors for each demineralizer are:

H-3 Cs, Rb Co-58 All Others CVCS* 1 2 50 50 Mobile Demin 1 1000 100 1000 vendor (ref. 29)

Condensate 1 2 10 10 Demins

• The cation bed gives a minimum decontamination factor of 10 for ionic isotopes (including Cesium). The mixed bed also gives an additional factor of 10 (except for Cesium). The effective decontamination factor is then 10 for Cesium, and 100 for others. The use of the above values is therefore conservative.

The total release is determined by the following formula:

RTOT = [RTANKS + (RCVCS/DFCVCS)]/DFMOBDEM + RLHST + RCONDEMINWASTE + RTB where RTOT = total release R = release E2-4

Enclosure 2 Watts Bar Nuclear Plant Remaining Chapter 11 Request for Additional Information DF = decontamination factor (see table above) subscripts refer to source In the event that the releases from the condensate regeneration are excessive, some of the waste can be treated with the mobile demineralizers. Not all of the condensate regenerative waste can be treated by the mobile demineralizers (the Non-Reclaimable and Neutralization Tank fluids cannot be processed); however, this calculation provides a bounding case which assumes none of the condensate regeneration waste is processed. The equation for the condensate regeneration treatment is:

RTOT = [RTANKS + (RCVCS/DFCVCS)]/DFMOBDEM + RLHST + RCONDEMINWASTE/DFMOBDEM + RTB The formula for the case of direct SGB release and no condenser demineralizer processing is:

RTOT = [RTANKS + (RCVCS/DFCVCS)]/DFMOBDEM + RLHST + RSGB + RTB where RSGB = RCONDEMINWASTE* DFCONDEMIN Results Examination of the above indicates that the total release will exceed 5 Ci/unit (10 CFR 50 Appendix I criteria of 5 Ci/unit), therefore another variant is determined. The variant is where the RSGB is maximized so as to reach the total limit of 5 Ci/yr. The gross gamma concentration can then be back calculated to be 4.402 Ci/yr.

The maximum gross gamma concentration in the SGB release to the river without processing and not exceeding 5 Ci/unit is:

= 3.6528E-5 uCicc Table 11.2-7 Values For determining values found in Table 11.2-7, the model used was that specified in Regulatory Guide 1.109 Equations 1, 2, and 3 for potable water, aquatic foods, and shoreline deposits.

FSAR Section 11.2.9.1 contains the Assumptions and Calculational Methods used to generate Table 11.2-7. Receptor and public water supplies data were taken from Tables 3-14 and 3-15 of the WBN FSEIS. For conservatism, a transit time of zero was assumed for releases to reach aquatic recreation areas and public water supplies.

Calculations were performed using TVA code Quarterly Water Dose Computer Code using equations from Sections 6.3 through 6.7 of WBN ODCM.

E2-5

Enclosure 2 Watts Bar Nuclear Plant Remaining Chapter 11 Request for Additional Information NRC Question 11.3.a:

Table 11.3-10 (formerly 11.3-11) provided in Amendment 98 indicates that only change made is the table number. However, it appears that the entire table has been revised.

Provide the basis for the revised dose number in Table 11.3-10.

TVA Response:

TVA has re-verified Table 11.3-10 due to an issue involving terrain adjustment factors identified in 2010, as described below:

In the past, the TVA used Gaseous Effluent Licensing Code (GELC) to perform routine dose assessments required by NRC Regulatory Guide 1.111. For WBN, adjustments to the GELC results were necessary to account for recirculation effects of spatial and temporal variations in airflow in the vicinity of pronounced river valleys. TVA had developed site-specific adjustment factors for WBN by comparing results from the GELC model with results from the MESOPUFF II model. These adjustment factors were revised each year to reflect changes based on annual surveys.

However, studies performed during 2010 for development of an American Nuclear Society (ANS) standard (specifically by the ANS-2.15 recirculation sub-group) determined that the adjustment factor approach is not acceptable for addressing recirculation issues. Further, comparisons with other models determined that MESOPUFF II is not suitable for calculating /Q values at WBN receptors, and that GELC adequately estimates /Q for WBN receptors, without any need for adjustments.

As a result of the above, the FSAR will be revised to eliminate the adjustment factors and use GELC results directly. Specifically, Table 11.3-10 (Unit 2 only) dose values for Noble Gases and Iodines/Particulates will be revised. In addition, due to elimination of the terrain adjustment factors, the highest dose pathway becomes vegetable ingestion instead of the cow milk with feeding factor. Doses reflected in this table will be of one unit (Unit 2) without a Tritium Producing Core. These changes will be submitted as part of Unit 2 FSAR, Amendment 103.

Once Unit 2 is licensed, the plans are to combine this table with the Unit 1 UFSAR table when the Unit 2 FSAR and the Unit 1 UFSAR are merged.

NRC Question 11.3.b:

Table 11.3-10 (formerly 11.3-11) provided in Amendment 98 indicates the only change made is the table number. However, it appears that the entire table has been revised.

It is unclear if this table is demonstrating releases within the design criteria of 10 CFR Part 50 Appendix I (e.g., per unit) or RM 50-2 (e.g., per site), as committed to in response E2-6

Enclosure 2 Watts Bar Nuclear Plant Remaining Chapter 11 Request for Additional Information to Question 8 of Section 11 in letter dated June 3, 2010 (ADAMS Accession Number ML101600477). Please clarification.

TVA Response:

The corresponding Unit 1 table is being revised in the same manner as described in question 11.3a above. When the Unit 1 and Unit 2 tables are combined, the results will be evaluated against the criteria of RM 50-2. The Unit 1 values are similar in magnitude to the Unit 2 values and thus the sum of the two units will meet the RM 50-2 criteria.

NRC Question 11.3.c:

Table 11.3-10 (formerly 11.3-11) provided in Amendment 98 indicates the only change made is the table number. However, it appears that the entire table has been revised.

The revised title indicates that the doses are for Unit 1 without TPC (Tritium Production Core). If that is accurate:

i) provide the estimated doses with Unit 2 operating, and ii) provide the basis for not including Unit 1 tritium production.

TVA Response:

The table provides dose for Unit 2 as explained in the response to NRC question 11.3a. Note that the actual title of Table 11.3-10 is (For 1 Unit without TPC) rather than (For Unit 1 without TPC) verbiage used in the RAI question.

E2-7

Enclosure 3 Watts Bar Nuclear Plant Proposed FSAR Chapter 11 Markups Proposed Final Supplemental Environmental Impact Statement Markups E3-1

WATTS BAR WBNP-102 Table 2.1-12 Watts Bar 2040 Population Distribution Within 50 Miles Of The Site (Sheet 1 of 1)

Direction 0-10 10-20 20-30 30-40 40-50 Total N 2,541 2,218 2,281 4,460 6,373 17,873 NNE 1,687 11,747 18,599 12,607 2,549 47,189 NE 1,524 3,597 16,808 26,935 80,896 129,760 ENE 1,174 4,918 31,814 72,849 244,656 355,411 E 4,811 9,773 17,518 24,692 46,384 103,178 ESE 890 6,151 19,601 4,909 3,336 34,887 SE 961 19,601 17,155 4,359 3,985 46,021 SSE 2,051 8,838 13,196 3,083 38,513 65,681 S 6,157 4,070 42,757 56,934 16,750 126,668 SSW 599 3,215 39,231 42,901 106,346 192,292 SW 1,056 13,605 14,537 60,959 127,447 217,604 WSW 943 12,996 2,714 2,667 3,603 22,923 W 941 3,150 4,984 2,771 5,249 17,095 WNW 721 1,981 3,729 5,400 19,945 31,776 NW 4,018 3,302 13,705 8,129 14,875 44,029 NNW 3,430 1,586 33,560 11,512 6,092 56,180 TOTAL 33,504 110,748 292,149 345,167 726,999 1,508,567 No. 1 - Replace with data from following page GEOGRAPHY AND DEMOGRAPHY 2.1-17

Table 2.1-12 Watts Bar 2040 Population Distribution Within 50 Miles of the Site (Sheet 1 of 1)

Direction 0-10 10-20 20-30 30-40 40-50 Total N 2,619 1,885 2,778 4,768 6,172 18,222 NNE 2,150 11,762 18,766 14,502 2,547 49,727 NE 1,441 3,783 16,734 29,838 78,334 130,130 ENE 1,110 3,553 29,539 63,798 253,831 351,832 E 1,915 11,352 18,647 30,063 44,013 105,990 ESE 135 6,230 20,120 5,068 3,280 34,833 SE 203 19,852 15,185 3,950 4,822 44,012 SSE 782 8,951 12,907 2,918 48,593 74,151 S 5,823 4,586 42,883 56,430 17,985 127,707 SSW 567 5,725 42,517 46,281 106,392 201,482 SW 1,051 12,978 14,499 62,307 111,795 202,630 WSW 938 12,791 2,837 2,840 3,372 22,778 W 937 3,406 5,555 2,944 5,474 18,316 WNW 717 2,091 4,372 5,654 20,511 33,345 NW 3,998 2,889 18,634 10,462 15,956 51,940 NNW 3,413 1,536 33,843 11,609 5,890 56,290 TOTAL 27,799 113,368 299,818 353,432 728,968 1,523,385 Insert this data into Table 2.1-12

WATTS BAR WBNP-102 11.3.7.3 Expected Gaseous Waste Processing System Releases Gaseous wastes consist of nitrogen and hydrogen gases purged from the Chemical Volume and Control System volume control tank when degassing the reactor coolant, and from the closed gas blanketing system. The gas decay tank capacity permits at least 60 days decay for waste gases before discharge during normal operation.

The quantities and isotopic concentration of gases discharged from the GWPS have been estimated. The analysis is based on input sources to the GWPS per NUREG-0017, modified to reflect WBN plant-specific parameters.

The expected gaseous releases in curies per year per reactor unit are given in Table 11.3-5.

11.3.7.4 Releases from Ventilation Systems A detailed review of the entire plant has been made to ascertain those items that could possibly contribute to airborne radioactive releases.

During normal plant operations, airborne noble gases and/or iodines can originate from reactor coolant leakage, equipment drains, venting and sampling, secondary side leakage, condenser air ejector and gland seal condenser exhausts, and GWPS leakage.

The assumptions used to estimate the annual quantity of radioactive gaseous effluents are given in Table 11.3-6. These assumptions are in accordance with NUREG-0017.

The noble gases and iodines discharged from the various sources are entered in Table 11.3-10. No. 2 - Replace with "11.3-7" 11.3.7.5 Estimated Total Releases The estimated releases listed in Table 11.3-7c have been used in calculating the site boundary doses as shown in Table 11.3-10. Table 11.3-7a is the expected gases released for 1% failed fuel with containment purge. Table 11.3-7 is the annual releases with purge air filters. Table 11.3-7b is the expected gases released for 1% failed fuel with continuous filtered containment vent, and Table 11.3-7c for approximately 1/8%

failed fuel with continuous filtered containment vent.

The dose calculations, based on the estimated total plant releases, show that the releases are in accordance with the design objectives in Section 11.3.1 and meet the regulations as outlined in Section 11.3.7.1. Further, the total plant releases are within the ODCM limits.

11.3.8 Release Points Gaseous radioactive wastes are released to the atmosphere through vents located on the Shield Building, Auxiliary Building, Turbine Building, and Service Building. A brief description, including function and location of each type vent, is presented below.

GASEOUS WASTE SYSTEMS 11.3-7

WATTS BAR WBNP-102 No. 3 - Replace with:

Turbine Building Vents Shield Building Vent Gaseous wastes from the condenser are discharged through the condenser vacuum Wastevent.

exhaust gases Thefrom containment vent, which is a purge and 12-inch the waste diameter pipe,gas decay tanks discharges are discharged at approximately theto760-foot the environment level. Underthrough normal a Shield Building operating vent. Each conditions Shield Building the discharge haswill flow rate onetypically vent.

The vent is of be less than 45 cfm. rectangular cross section (dimension - 2 feet by 7 feet 6 inches) and discharges approximately 130 feet above ground level. The location of the Reactor Building ventsventilation Non-radioactive is shown in airthe equipment from is exhausted layoutthe drawings, Turbine Figure Building 1.2-1.

throughThethe location of theBuilding Turbine Shield Building in relation vents. There to the site are eighteen is shown vents at theon the main 755-foot plant level andgeneral twentyplan, vents Figure 2.1-5. All releases from the Shield Building vent except containment purge air at the 824-foot level (roof level). The effluent flow rates vary for each type of vent.

exhaust monitor discharges are passed through HEPA filters and charcoal adsorbers Generally, the normal flow rates through a typical vent at the 755-foot level is 22,888 cfm prior to release. The effluent discharge rate through the vent is variable; occasionally, and the flow rates through typical vent at the 824-foot level is 28,500 cfm. The general during containment purge, the rate may approach the value which is listed in Figure arrangement of vents on the Turbine Building is shown on Figure 1.2-1. The turbine 9.4-28. The flow path for waste gases exhausted through the vent from the waste gas building is shown on the main plant general plan, Figure 2.1-5.

decay tanks is shown in Figure 11.3-1.

Auxiliary Building Vent Waste gases in the Auxiliary Building are discharged through the Auxiliary Building exhaust vent. In addition, containment atmosphere is continuously vented, during normal operation for pressure control, into the annulus after it is filtered through HEPA and charcoal filters, and subsequently, discharged into the Auxiliary Building exhaust vent. The vent is of the chimney type having a rectangular cross section of 10 by 30 feet. The top of the vent is located atop the Auxiliary Building and discharges approximately 106 feet above grade. Under normal operating conditions, gases are continuously discharged through the vent. Effluent flow rates can be near 224,000 cfm when two Auxiliary Building general exhaust fans and one fuel-handling area exhaust fan are operating at full capacity. Under accident conditions, the Auxiliary Building is isolated, and the Auxiliary Building gas treatment system (ABGTS) is used to treat gaseous effluents. When in service, the ABGTS discharges to the Shield Building exhaust vent. The location of the Auxiliary Building exhaust vent is shown in the equipment layout diagram, Figure 1.2-1. The Auxiliary Building is shown on the main plant general plan, Figure 2.1-5.

Turbine Building Vents Ventilation air is exhausted from the Turbine Building through the Turbine Building vents. There are eighteen vents at the 755-foot level and twenty vents at the 824-foot level (roof level). The effluent flow rates vary for each type of vent. Generally, the normal flow rates through a typical vent at the 755-foot level is 22,888 cfm and the flow rates through typical vent at the 824-foot level is 28,500 cfm. The general arrangement of vents on the Turbine Building is shown on Figure 1.2-1. The turbine building is shown on the main plant general plan, Figure 2.1-5.

Condenser Vacuum Exhaust Vent Gaseous wastes from the condenser are discharged through the condenser vacuum exhaust vent. The vent, which is a 12-inch diameter pipe, discharges at approximately the 760-foot level. Under normal operating conditions the discharge flow rate will typically be less than 45 cfm.

11.3-8 GASEOUS WASTE SYSTEMS

WATTS BAR WBNP-102 Service Building Vent Radiologically monitored potentially radioactive waste gases from the radiochemical laboratory and the titration room are exhausted through HEPA filters via a common duct which discharges to the common Service Building roof exhaust plenum. Exhaust air from the general area discharges to the common Service Building roof exhaust plenum. Separate vents from the common roof exhaust plenum discharge to atmosphere approximately 24 feet above grade. The Service Building is shown on the site plot plan, Figure 2.1-5.

11.3.9 Atmospheric Dilution Calculations of atmospheric transport, dispersion, and ground deposition are based on the straight-line airflow model discussed in NRC Regulatory Guide 1.111 (Revision 1, July 1977). Releases are assumed to be continuous. Releases known to be periodic, e.g., those during containment purging and waste gas decay tank venting, are treated as continuous releases. No. 4 - Replace with "batch."

Releases from the Shield Building, Turbine Building (TB), and Auxiliary Building (AB) vents are treated as ground level. The ground level joint frequency distribution (JFD) is given in Section 2.3. Air concentrations and deposition rates were calculated considering radioactive decay and buildup during transit. Plume depletion was calculated using the figures provided in Regulatory Guide 1.111.

No. 5 - Replace with Estimates of normalized concentrations (X/Q) and normalized deposition "therates ODCM."(D/Q) for gaseous releases at points where potential dose pathways exist are listed in Table 11.3-8.

11.3.10 Estimated Doses from Radionuclides in Gaseous Effluents Individuals are exposed to gaseous effluents via the following pathways: (1) external radiation from radioactivity in the air and on the ground; (2) inhalation; and (3) ingestion of beef, vegetables, and milk. No other additional exposure pathway has been identified which would contribute 10% or more to either individual or population doses.

11.3.10.1 Assumptions and Calculational Methods No. 6 - Replace with "2007" External air exposures are evaluated at points of potential maximum exposure (i.e.,

points at the unrestricted area boundary). External skin and total body exposures are evaluated at nearby residences. The dose to the critical organ from radioiodines, tritium (Unit 1 only) and particulates is calculated for real pathways existing at the site during a land use survey conducted in 1994.

No. 6 - Replace with "2007" No. 6 - Delete To evaluate the potential critical organ dose, milk animals and nearest gardens were identified by a detailed survey within five miles of the plant (Table 11.3-8). Information on grazing seasons and feeding regimes are reflected in the feeding factor. The feeding factor is the fraction of the year an animal grazes on pasture. During the 1994 land use survey, there was one milk cow location identified in which information regarding the feeding regime for the animals, and the ages of onsite consumers of the milk could not be established. Because no specific information is known, it is conservatively assumed that the feeding factor for that location is equal to the worst-GASEOUS WASTE SYSTEMS 11.3-9 No. 6 Delete

WATTS BAR WBNP-102 No. 7 - Delete No. 7 - Replace with "0.33" No. 7 - Replace with "past" case feeding factor identified during the 1994 land use census for any real cow location (i.e., 70% pasture feeding) and that all four age groups are present. Since specific data on beef animals were not available, the nearest beef animal was assumed to be at the point of maximum offsite exposure. Milk ingestion is the critical pathway.

TVA assumes that enough fresh vegetables are produced at each residence to supply annual consumption by all members of that household. TVA assumes that enough meat is produced in each sector annulus to supply the needs of that region. Watts Bar projected population distribution for the year 2040 is given in Table 11.3-9.

Doses are calculated using the dose factors and methodology contained in NRC Regulatory Guide 1.109 with certain exceptions as follows:

(1) Inhalation doses are based on the average individuals inhalation rates found in ICRP Publication 23 of 1,400; 5,500; 8,000; and 8,100 m3/year for infant, child, teen, and adult, respectively.

(2) The milk ingestion pathway has been modeled to include specific information on grazing periods for milk animals obtained from a detailed farm survey. A feeding factor (FF) has been defined as that fraction of total feed intake a dairy animal consumes that is from fresh forage. The remaining portion of feed (1-FF) is assumed to be from stored feed. Doses calculated from milk produced by animals consuming fresh forage are multiplied by these factors.

Concentrations of radioactivity in stored feed are adjusted to reflect radioactive decay during the maximum assumed storage period of 180 days by the factor:

180 1 1 - exp - O i 180 180 ³ exp -Oi t dt = ----------------------------------------

180O i 0

This factor replaces the factor exp (-i th) in equation C-10 of Regulatory Guide 1.109.

(3) The stored vegetable and beef ingestion pathways have been modeled to reflect more accurately the actual dietary characteristics of individuals. For stored vegetables the assumption is made that home grown stored vegetables are consumed when fresh vegetables are not available, i.e.,

during the 9 months of fall, winter, and spring. Rather than use a constant 11.3-10 GASEOUS WASTE SYSTEMS

WATTS BAR WBNP-102 Category Ages (A)* Fraction Teen 13<A<19 0.153 Adult 19<A 0.665

• e.g., someone who is 1 year, 11 months is an infant, while someone who is exactly two years old is a child.

Tables 11.3-10 and 11.3-11 provide the doses estimated for individuals and the population within 50 miles of the plant site.

11.3.10.2 Summary of Annual Population Doses TVA has estimated the radiological impact to regional population groups in the year 2040 from the normal operation of the Watts Bar Nuclear Plant. Table 11.3-11 summarizes these population doses. The total body dose from background to individuals within the United States ranges from approximately 100 mrem to 250 mrem per year. The annual total body dose due to background for a population of about 1,100,000 persons expected to live within a 50 mile radius of the Watts Bar Nuclear Plant in the year 2040 is calculated to be approximately 154,000 man-rem assuming 140 mrem/year/individual. By comparison, the same population (excluding onsite radiation workers) will receive a total body dose of approximately 3.85 man-rem from effluents. Based on these results, TVA concludes that the normal operation of the Watts Bar Nuclear Plant will present minimal risk to the health and safety of the public.

No. 8 - Replace with "210,000" REFERENCES None No. 8 - Replace with "6.66" No. 8 - Replace with "1,500,000" 11.3-12 GASEOUS WASTE SYSTEMS

WATTS BAR WBNP-102 Table 11.3-7 Annual Radioactive Releases With Purge Air Filters (Curies/Year/Reactor)

Table based on operation of one unit.

Contain.(1) Aux. Turbine Total Nuclide Building Building Building Kr-85m 2.00E+01 4.53E+00 1.23E+00 2.58E+01 Kr-85 6.90E+02 7.05E+00 1.86E+00 6.99E+02 Kr-87 1.09E+01 4.27E+00 1.09E+00 1.62E+01 Kr-88 2.84E+01 7.95E+00 2.13E+00 3.85E+01 Xe-131m 1.17E+03 1.73E+01 4.53E+00 1.19E+03 Xe-133m 4.63E+01 1.90E+00 5.21E-01 4.88E+01 Xe-133 3.12E+03 6.70E+01 1.77E+01 3.20E+03 Xe-135m 3.86E+00 3.68E+00 9.80E-01 8.52E+00 Xe-135 1.55E+02 2.40E+01 6.46E+00 1.85E+02 Xe-137 3.18E-01 9.67E-01 2.58E-01 1.54E+00 Xe-138 3.33E+00 3.42E+00 9.06E-01 7.66E+00 Ar-41 3.40E+01 0.00E+00 0.00E+00 3.40E+01 Br-84 6.00E-05 5.02E-02 4.81E-04 5.07E-02 I-131 7.29E-03 1.39E-01 7.08E-03 1.53E-01 I-132 1.61E-03 6.56E-01 1.70E-02 6.75E-01 I-133 3.55E-03 4.35E-01 2.03E-02 4.58E-01 I-134 1.66E-03 1.06E+00 1.47E-02 1.08E+00 I-135 3.16E-03 8.10E-01 3.13E-02 8.45E-01 H-3 1.39E+02 0.00E+00 0.00E+00 1.39E+02 H-3 (TPC)(3)

Unit 1 Only 3.70E+02 0.00E+00 0.00E+00 3.70E+02 Cr-51 9.21E-05 5.00E-04 0.00E+00 5.92E-04 Mn-54 5.30E-05 3.78E-04 0.00E+00 4.31E-04 Co-57 8.20E-06 0.00E+00 0.00E+00 8.20E-06 Co-58 2.50E-04 2.29E-02 0.00E+00 2.32E-02 Co-60 2.61E-05 8.71E-03 0.00E+00 8.74E-03 Fe-59 2.70E-05 5.00E-05 0.00E+00 7.70E-05 Sr-89 1.30E-04 2.85E-03 0.00E+00 2.98E-03 Sr-90 5.22E-05 1.09E-03 0.00E+00 1.14E-03 Zr-95 4.80E-08 1.00E-03 0.00E+00 1.00E-03 Nb-95 1.80E-05 2.43E-03 0.00E+00 2.45E-03 Ru-103 1.60E-05 6.10E-05 0.00E+00 7.70E-05 Ru-106 2.70E-08 7.50E-05 0.00E+00 7.50E-05 Sb-125 0.00E+00 6.09E-05 0.00E+00 6.09E-05 Cs-134 2.53E-05 2.24E-03 0.00E+00 2.27E-03 Cs-136 3.21E-05 4.80E-05 0.00E+00 8.01E-05 Cs-137 5.58E-05 3.42E-03 0.00E+00 3.48E-03 Ba-140 2.30E-07 4.00E-04 0.00E+00 4.00E-04 Ce-141 1.30E-05 2.64E-05 0.00E+00 3.95E-05 C-14 2.80E+00 4.50E+00 0.00E+00 7.30E+00 (1) Includes release from GWPS (2) 4.28E+02 = 4.28 X 102 (3) Tritium values for a Tritim Production Core No. 9 - Delete GASEOUS WASTE SYSTEMS 11.3-21

WATTS BAR WBNP-102 Table 11.3-7a Design (For 1% Failed Fuel) Expected Gas Release Concentration/(Effluent Concentration Limit) With Containment Purge (Sheet 1 of 2)

Single Unit Dual Unit Exp. Rel. Design Design 10CFR20 Operation Operation (Ci/yr) Des/Exp (Ci/yr) (Ci/cc) (ECL) C/ECL C/ECL Kr-85m 2.58E+01 12.28 3.17E+02 1.10E-10 1.0E-07 0.0010951 0.0021902 Kr-85 6.99E+02 33.08 2.31E+04 7.99E-09 7.0E-07 0.0114124 0.0228248 Kr-87 1.62E+01 7.45 1.21E+02 4.18E-11 2.0E-08 0.0020906 0.0041812 Kr-88 3.85E+01 12.33 4.75E+02 1.64E-10 9.0E-09 0.0182306 0.0364612 Xe-131m 1.19E+03 2.91 3.45E+03 1.19E-09 2.0E-06 0.0005971 0.0011942 Xe-133m 4.88E+01 43.24 2.11E+03 7.29E-10 6.0E-07 0.0012142 0.0024284 Xe-133 3.20E+03 111.07 3.55E+05 1.23E-07 5.0E-07 0.2456675 0.4913350 Xe-135m 8.52E+00 5.04 4.29E+01 1.48E-11 4.0E-08 0.0003710 0.0007420 Xe-135 1.85E+02 6.97 1.29E+03 4.46E-10 7.0E-08 0.006375 0.012750 Xe-138 7.66E+00 5.43 4.16E+01 1.44E-11 2.0E-08 0.0007188 0.0014376 Br-84 5.07E-02 2.50 1.27E-01 4.38E-14 8.0E-08 5.478E-07 1.096E-06 I-131 1.53E-01 52.41 8.03E+00 2.77E-12 2.0E-10 0.013875 0.027750 I-132 6.75E-01 4.00 2.70E+00 9.33E-13 2.0E-08 4.67E-05 0.0000934 I-133 4.58E-01 26.85 1.23E+01 4.25E-12 1.0E-09 0.0042535 0.0085070 I-134 1.08E+00 1.65 1.78E+00 6.14E-13 6.0E-08 1.023E-05 2.046E-05 I-135 8.45E-01 7.91 6.69E+00 2.31E-12 6.0E-09 0.0003851 0.0007702 Cs-134 2.27E-03 40.60 9.20E-02 3.18E-14 2.0E-10 0.0001589 0.0003178 Cs-136 8.01E-05 165.20 1.32E-02 4.57E-15 9.0E-10 5.079E-06 1.016E-05 Cs-137 3.48E-03 153.22 5.33E-01 1.84E-13 2.0E-10 0.0009203 0.0018406 Cr-51 5.92E-04 0.29 1.73E-04 5.96E-17 3.0E-08 1.988E-09 3.976E-09 Mn-54 4.31E-04 0.47 2.03E-04 7.01E-17 1.0E-09 7.005E-08 1.401E-07 Fe-59 7.70E-05 3.48 2.68E-04 9.27E-17 5.0E-10 1.853E-07 3.706E-07 Co-58 2.32E-02 5.37 1.24E-01 4.30E-14 1.0E-09 4.298E-05 8.596E-05 Co-60 8.74E-03 1.38 1.21E-02 4.17E-15 5.0E-11 8.333E-05 1.667E-04 Sr-89 2.98E-03 22.45 6.69E-02 2.31E-14 1.0E-09 2.313E-05 4.626E-05 Sr-90 1.14E-03 13.49 1.54E-02 5.33E-15 6.0E-12 0.0008877 0.0017754 Zr-95 1.00E-03 1.71 1.71E-03 5.92E-16 4.0E-10 1.481E-06 2.962E-06 Nb-95 2.45E-03 2.34 5.73E-03 1.98E-15 2.0E-09 9.895E-07 1.979E-06 Ba-140 4.00E-04 0.31 1.26E-04 4.34E-17 2.0E-09 2.171E-08 4.342E-08 H-3 1.39E+02 1 1.39E+02 4.80E-11 1.0E-07 0.0004811 0.0009622 H-3 (TPC) 3.70E+02 1 3.70E+02 1.28E-10 1.0E-07 0.0012775 0.0012775 1 rod 1.53E+03 1 1.53E+03 5.29E-10 1.0E-07 0.0052869 0.0052869 2 rod 2.69E+03 1 2.69E+03 9.30E-10 1.0E-07 0.0092962 0.0092962 C-14 7.30E+00 1 7.30E+00 2.52E-12 3.0E-09 0.000841 0.001682 Ar-41 3.40E+01 1 3.40E+01 1.18E-11 1.0E-08 0.0011752 0.0023504 Total 0.3109694 0.6219388 Total (TPC) 0.3117657 0.6227352 1 rod 0.3157751 0.6267446 2 rod 0.3197845 0.6307539 No. 10 - Delete 11.3-22 GASEOUS WASTE SYSTEMS

WATTS BAR WBNP-102 Table 11.3-7a Design (For 1% Failed Fuel) Expected Gas Release Concentration/(Effluent Concentration Limit) With Containment Purge (Sheet 2 of 2)

Note: The Dual Unit Operation column in the above calculation considers dual unit operation.

Based on the evaluation done for Revision 7, the per unit concentrations are the same for both units. Therefore, the last column is twice the preceeding column except in the case of TPC.

Note: Dual unit operation considers only Unit 1 with TPC.

No. 11 - Delete GASEOUS WASTE SYSTEMS 11.3-23

WATTS BAR WBNP-102 Table 11.3-7b Design (For 1% Failed Fuel) Expected Gas Release Concentration/(Effluent Concentration Limit) With Continuous Filtered Containment Vent (Sheet 1 of 2)

Single Unit Dual Unit Exp. Rel. Design Design 10CFR20 Operation Operation (Ci/yr) Des/Exp (Ci/yr) (Ci/cc) (ECL) C/ECL C/ECL Kr-85m 9.48E+00 12.28 1.16E+02 4.02E-11 1.0E-07 0.0004024 0.0008048 Kr-85 6.78E+02 33.08 2.24E+04 7.75E-09 7.0E-07 0.0110743 0.0221486 Kr-87 5.81E+00 7.45 4.33E+01 1.50E-11 2.0E-08 0.0007480 0.0014960 Kr-88 1.32E+01 12.33 1.63E+02 5.63E-11 9.0E-09 0.0062505 0.0125010 Xe-131m 1.09E+03 2.91 3.18E+03 1.10E-09 2.0E-06 0.0005489 0.0010978 Xe-133m 4.31E+01 43.24 1.86E+03 6.44E-10 6.0E-07 0.0010735 0.0021470 Xe-133 2.90E+03 111.07 3.22E+05 1.11E-07 5.0E-07 0.2227110 0.4454220 Xe-135m 4.68E+00 5.04 2.36E+01 8.15E-12 4.0E-08 0.0002038 0.0004076 Xe-135 8.88E+01 6.97 6.19E+02 2.14E-10 7.0E-08 0.0030561 0.0061122 Xe-138 4.34E+00 5.43 2.36E+01 8.15E-12 2.0E-08 0.0004073 0.0008146 Br-84 5.07E-02 2.50 1.27E-01 4.38E-14 8.0E-08 0.0000005 0.0000010 I-131 1.53E-01 52.41 8.00E+00 2.77E-12 2.0E-10 0.0138277 0.0276554 I-132 6.73E-01 4.00 2.69E+00 9.30E-13 2.0E-08 0.0000465 0.0000930 I-133 4.57E-01 26.85 1.23E+01 4.24E-12 1.0E-09 0.0042433 0.0084866 I-134 1.07E+00 1.65 1.77E+00 6.10E-13 6.0E-08 0.0000102 0.0000204 I-135 8.42E-01 7.91 6.66E+00 2.30E-12 6.0E-09 0.0003837 0.0007674 Cs-134 2.27E-03 40.60 9.20E-02 3.18E-14 2.0E-10 0.0001589 0.0003178 Cs-136 8.01E-05 165.20 1.32E-02 4.57E-15 9.0E-10 0.0000051 0.0000102 Cs-137 3.48E-03 153.22 5.33E-01 1.84E-13 2.0E-10 0.0009203 0.0018406 Cr-51 5.92E-04 0.29 1.73E-04 5.96E-17 3.0E-08 0.0000000 0.0000000 Mn-54 4.31E-04 0.47 2.03E-04 7.01E-17 1.0E-09 0.0000001 0.0000002 Fe-59 7.70E-05 3.48 2.68E-04 9.27E-17 5.0E-10 0.0000002 0.0000004 Co-58 2.32E-02 5.37 1.24E-01 4.30E-14 1.0E-09 0.0000430 0.0000860 Co-60 8.74E-03 1.38 1.21E-02 4.17E-15 5.0E-11 0.0000833 0.0001666 Sr-89 2.98E-03 22.45 6.69E-02 2.31E-14 1.0E-09 0.0000231 0.0000462 Sr-90 1.14E-03 13.49 1.54E-02 5.33E-15 6.0E-12 0.0008877 0.0017754 Zr-95 1.00E-03 1.71 1.71E-03 5.92E-16 4.0E-10 0.0000015 0.0000030 Nb-95 2.45E-03 2.34 5.73E-03 1.98E-15 2.0E-09 0.0000010 0.0000020 Ba-140 4.00E-04 0.31 1.26E-04 4.34E-17 2.0E-09 0.0000000 0.0000000 H-3 1.39E+02 1 1.39E+02 4.80E-11 1.0E-07 0.0004811 0.0009622 H-3 (TPC) 3.70E+02 1 3.70E+02 1.28E-10 1.0E-07 0.0012775 0.0012775 1 rod 1.53E+03 1 1.53E+03 5.29E-10 1.0E-07 0.0052869 0.0052869 2 rod 2.69E+03 1 2.69E+03 9.30E-10 1.0E-07 0.0092962 0.0092962 C-14 7.30E+00 1 7.30E+00 2.52E-12 3.0E-09 0.0008410 0.0016820 Ar-41 3.40E+01 1 3.40E+01 1.18E-11 1.0E-08 0.0011752 0.0023504 Total 0.2696131 0.5392262 Total (TPC) 0.2704095 0.5400226 1 rod 0.2744189 0.5440320 2 rod 0.2784283 0.5480413 No. 12 - Delete 11.3-24 GASEOUS WASTE SYSTEMS

WATTS BAR WBNP-102 Table 11.3-7b Design (For 1% Failed Fuel) Expected Gas Release Concentration/(Effluent Concentration Limit) With Continuous Filtered Containment Vent (Sheet 2 of 2)

Note: The Dual Unit Operation column in the above calculation considers dual unit operation.

Based on the evaluation done for Revision 7, the per unit concentrations are the same for both units. Therefore, the last column is twice the preceeding column except in the case of TPC.

Note: Dual unit operation considers only Unit 1 with TPC.

No. 13 - Delete GASEOUS WASTE SYSTEMS 11.3-25

WATTS BAR WBNP-102 Table 11.3-7c Total Releases (y 1/8 failed fuel in Ci/yr), with Continuous Filtered Containment Vent (Sheet 1 of 1)

Table based on operation of one unit Contain.(1) Aux. Turbine Total Nuclide Building Building Building Kr-85m 3.72E+00 4.53E+00 1.23E+00 9.48E+00 Kr-85 6.69E+02 7.05E+00 1.86E+00 6.78E+02 Kr-87 4.48E-01 4.27E+00 1.09E+00 5.81E+00 Kr-88 3.10E+00 7.95E+00 2.13E+00 1.32E+01 Xe-131m 1.07E+03 1.73E+01 4.53E+00 1.09E+03 Xe-133m 4.07E+01 1.90E+00 5.21E-01 4.31E+01 Xe-133 2.82E+03 6.70E+01 1.77E+01 2.90E+03 Xe-135m 2.26E-02 3.68E+00 9.80E-01 4.68E+00 Xe-135 5.83E+01 2.40E+01 6.46E+01 8.88E+01 Xe-137 3.76E-04 9.67E-01 2.58E-01 1.23E+00 Xe-138 1.69E-02 3.42E+00 9.06E-01 4.34E+00 Ar-41 3.40E+01 0.00E+00 0.00E+00 3.40E+01 Br-84 8.16E-07 5.02E-02 4.81E-04 5.07E-02 I-131 6.74E-03 1.39E-01 7.08E-03 1.53E-01 I-132 1.36E-04 6.56E-01 1.70E-02 6.73E-01 I-133 2.36E-03 4.35E-01 2.03E-02 4.57E-01 I-134 4.26E-05 1.06E+00 1.47E-02 1.07E+00 I-135 8.80E-04 8.10E-01 3.13E-02 8.42E-01 H-3 1.39E+02 0.00E+00 0.00E+00 1.39E+02 H-3 (TPC) 3.70E+02 0.00E+00 0.00E+00 3.70E+02 Cr-51 9.21E-05 5.00E-04 0.00E+00 5.92E-04 Mn-54 5.30E-05 3.78E-04 0.00E+00 4.31E-04 Co-57 8.20E-06 0.00E+00 0.00E+00 8.20E-06 Co-58 2.50E-04 2.29E-02 0.00E+00 2.32E-02 Co-60 2.61E-05 8.71E-03 0.00E+00 8.74E-03 Fe-59 2.70E-05 5.00E-05 0.00E+00 7.70E-05 Sr-89 1.30E-04 2.85E-03 0.00E+00 2.98E-03 Sr-90 5.22E-05 1.09E-03 0.00E+00 1.14E-03 Zr-95 4.80E-08 1.00E-03 0.00E+00 1.00E-03 Nb-95 1.80E-05 2.43E-03 0.00E+00 2.45E-03 Ru-103 1.60E-05 6.10E-05 0.00E+00 7.70E-05 Ru-106 2.70E-08 7.50E-05 0.00E+00 7.50E-05 Sb-125 0.00E+00 6.09E-05 0.00E+00 6.09E-05 Cs-134 2.53E-05 2.24E-03 0.00E+00 2.27E-03 Cs-136 3.21E-05 4.80E-05 0.00E+00 8.01E-05 Cs-137 5.58E-05 3.42E-03 0.00E+00 3.48E-03 Ba-140 2.30E-07 4.00E-04 0.00E+00 4.00E-04 Ce-141 1.30E-05 2.64E-05 0.00E+00 3.95E-05 C-14 2.80E+00 4.50E+00 0.00E+00 7.30E+00 (TPC) Tritium values for a Tritium Production Core (Unit 1 only)

No. 14 - Delete 11.3-26 GASEOUS WASTE SYSTEMS

WATTS BAR WBNP-102 Table 11.3-8 Data On Points Of Interest Near Watts Bar Nuclear Plant (Page 1 of 2)

Chi-over-Q D-over-Q Terrain Milk Distance (s/m^3) (1/m^2) Adjustment Feeding Sector (Meters) Factor Factor Unrestricted Area Boundary N 1550 5.12e-06 8.13e-09 1.70 Unrestricted Area Boundary NNE 1980 6.35e-06 1.23e-08 1.80 Unrestricted Area Boundary NE 1580 1.05e-05 1.10e-08 2.10 Unrestricted Area Boundary ENE 1370 1.23e-05 8.77e-09 1.70 Unrestricted Area Boundary E 1280 1.37e-05 9.66e-09 1.60 Unrestricted Area Boundary ESE 1250 1.43e-05 1.16e-08 1.80 Unrestricted Area Boundary SE 1250 1.11e-05 9.49e-09 1.50 Unrestricted Area Boundary SSE 1250 6.04e-06 8.21e-09 1.50 Unrestricted Area Boundary S 1340 5.33e-06 1.17e-08 1.90 Unrestricted Area Boundary SSW 1550 4.14e-06 1.05e-08 2.00 Unrestricted Area Boundary SW 1670 4.46e-06 7.34e-09 2.10 Unrestricted Area Boundary WSW 1430 5.47e-06 6.37e-09 1.80 Unrestricted Area Boundary W 1460 2.11e-06 2.07e-09 1.20 Unrestricted Area Boundary WNW 1400 2.49e-06 2.38e-09 2.50 Unrestricted Area Boundary NW 1400 2.05e-06 2.13e-09 1.70 Unrestricted Area Boundary NNW 1460 2.68e-06 3.08e-09 1.60 Resident N 2134 2.84e-06 4.21e-09 1.50 Resident NNE 3600 2.69e-06 4.41e-09 1.80 Resident NE 3353 3.84e-06 3.22e-09 2.20 Resident ENE 2414 6.26e-06 3.83e-09 1.90 Resident E 3268 3.97e-06 2.14e-09 1.70 Resident ESE 4416 2.64e-06 1.46e-09 1.90 Resident SE 1372 9.66e-06 8.16e-09 1.50 Resident SSE 1524 4.18e-06 5.56e-09 1.40 Resident S 1585 3.91e-06 8.42e-09 1.80 Resident SSW 1979 2.76e-06 6.64e-09 1.90 Resident SW 4230 1.15e-06 1.43e-09 2.00 Resident WSW 1829 3.61e-06 4.03e-09 1.70 Resident W 2896 7.30e-07 6.01e-10 1.10 Resident WNW 1646 2.26e-06 2.12e-09 2.90 Resident NW 2061 1.03e-06 9.95e-10 1.50 Resident NNW 4389 3.50e-07 2.97e-10 1.00 Garden N 7664 3.13e-07 3.00e-10 1.00 Garden NNE 6173 1.06e-06 1.42e-09 1.50 Garden NE 3829 3.06e-06 2.44e-09 2.10 Garden ENE 4927 2.01e-06 9.39e-10 1.60 Garden E 4991 1.99e-06 9.02e-10 1.50 Garden ESE 6096 1.63e-06 7.77e-10 1.80 Garden SE 4633 1.58e-06 8.97e-10 1.30 Garden SSE 7454 4.74e-07 3.57e-10 1.40 Garden S 2254 2.50e-06 4.94e-09 1.90 GASEOUS WASTE SYSTEMS No. 15 - Replace with attached revised table 11.3-27

WATTS BAR WBNP-102 Table 11.3-8 Data On Points Of Interest Near Watts Bar Nuclear Plant (Page 2 of 2)

Chi-over-Q D-over-Q Terrain Milk Distance (s/m^3) (1/m^2) Adjustment Feeding Sector (Meters) Factor Factor Garden SSW 8100 2.79e-07 4.16e-10 1.40 Garden SW 8100 4.28e-07 4.03e-10 1.80 Garden WSW 4667 9.86e-07 8.06e-10 1.70 Garden W 5120 3.33e-07 2.23e-10 1.10 Garden WNW 5909 1.85e-07 1.13e-10 1.40 Garden NW 3170 5.63e-07 4.78e-10 1.50 Garden NNW 4698 3.18e-07 2.64e-10 1.00 Milk Cow ESE 6096 1.63e-06 7.77e-10 1.80 0.25 Milk Cow ESE 6706 1.35e-06 6.18e-10 1.70 0.03 Milk Cow SSW 2286 2.24e-06 5.20e-09 1.90 0.05 Milk Cow SSW 3353 1.36e-06 2.84e-09 2.00 0.33 No. 15 - Replace with attached revised table 11.3-28 GASEOUS WASTE SYSTEMS

No. 15 - New Data for Table 11.3.8 Table 11.3-8 Data On Points of Interest Near Watts Bar Nuclear Plant (Page 1 of 2)

Milk Distance Chi-over-Q D-over-Q Sector Feeding (Meters) (s/m^3) (1/m^3)

Factor Unrestricted Area Boundary N 1550 3.01e06 4.78e-09 1.00 Unrestricted Area Boundary NNE 1980 3.53e-06 6.82e-09 1.00 Unrestricted Area Boundary NE 1580 4.99e-06 5.23e-09 1.00 Unrestricted Area Boundary ENE 1370 7.24e-06 5.16e-09 1.00 Unrestricted Area Boundary E 1280 8.57e-06 6.04e-09 1.00 Unrestricted Area Boundary ESE 1250 7.94e-06 6.46e-09 1.00 Unrestricted Area Boundary SE 1250 7.40e-06 6.32e-09 1.00 Unrestricted Area Boundary SSE 1250 4.03e-06 5.48e-09 1.00 Unrestricted Area Boundary S 1340 2.81e-06 6.14-e09 1.00 Unrestricted Area Boundary SSW 1550 2.07e-06 5.23e-09 1.00 Unrestricted Area Boundary SW 1670 2.13e-06 3.50e-09 1.00 Unrestricted Area Boundary WSW 1430 3.04e-06 3.54e-09 1.00 Unrestricted Area Boundary W 1460 1.76e-06 1.72e-09 1.00 Unrestricted Area Boundary WNW 1400 9.95e-07 9.50e-10 1.00 Unrestricted Area Boundary NW 1400 1.20e-06 1.25e-09 1.00 Unrestricted Area Boundary NNW 1460 1.67e-06 1.93e-09 1.00 Nearest Resident N 2134 1.90e-06 2.81e-09 1.00 Nearest Resident NNE 3600 1.49e-06 2.45e-09 1.00 Nearest Resident NE 3353 1.75e-06 1.46e-09 1.00 Nearest Resident ENE 2414 3.29e-06 2.01e-09 1.00 Nearest Resident E 3268 2.34e-06 1.26e-09 1.00 Nearest Resident ESE 4416 1.39e-06 7.66e-10 1.00 Nearest Resident SE 1372 6.44e-06 5.44e-09 1.00 Nearest Resident SSE 1524 2.99e-06 3.97e-09 1.00 Nearest Resident S 1585 2.17e-06 4.68e-09 1.00 Nearest Resident SSW 1979 1.45e-06 3.50e-09 1.00 Nearest Resident SW 4230 5.76e-07 7.14e-10 1.00 Nearest Resident WSW 1829 2.13e-06 2.37e-09 1.00 Nearest Resident W 2896 6.64e-07 5.47e-10 1.00 Nearest Resident WNW 1646 7.81e-07 7.31e-10 1.00 Nearest Resident NW 2061 6.88e-07 6.64e-10 1.00 Nearest Resident NNW 4389 3.50e-07 2.97e-10 1.00 Nearest Garden N 7664 3.13e-07 3.00e-10 1.00 Nearest Garden NNE 6173 7.04e-07 9.46e-10 1.00 Nearest Garden NE 3353 1.75e-06 1.46e-09 1.00 Nearest Garden ENE 4927 1.26e-06 5.87e-10 1.00 Nearest Garden E 6372 9.63e-07 3.87e-10 1.00 Nearest Garden ESE 4758 1.25e-06 6.73e-10 1.00 Nearest Garden SE 4633 1.21e-06 6.90e-10 1.00 Nearest Garden SSE 7454 3.39e-07 2.55e-10 1.00 Nearest Garden S 2254 1.31e-06 2.60e-09 1.00

No. 15 - New Data for Table 11.3.8 Table 11.3-8 Data On Points of Interest Near Watts Bar Nuclear Plant (Page 2 of 2)

Milk Distance Chi-over-Q D-over-Q Feeding Sector (Meters) (s/m^3) (1/m^3) Factor Nearest Garden SSW 1979 1.45e-06 3.50e-09 1.00 Nearest Garden SW 8100 2.38e-07 2.24e-10 1.00 Nearest Garden WSW 4667 5.80e-07 4.74e-10 1.00 Nearest Garden W 5120 3.03e-07 2.03e-10 1.00 Nearest Garden WNW 5909 1.32e-07 8.07e-11 1.00 Nearest Garden NW 3170 3.75e-07 3.18e-10 1.00 Nearest Garden NNW 4602 3.28e-07 2.74e-10 1.00 Milk Cow ESE 6706 7.97e-07 3.64e-10 0.03 Milk Cow SSW 2286 1.18e-06 2.74e-09 0.05 Milk Cow SSW 3353 6.80e-07 1.42e-09 0.33

Table 11.3-9 Projected 2040 Population Distribution Within 50 Miles Of Watts Bar Nuclear Plant Population Within Each Sector Element Distance From Site (Miles)

N 0-1 1-2 2-3 3-4 4-5 5-10 10-20 20-30 30-40 40-50 NNE 0 111 32 47 135 893 2071 2166 3453 4040 NE 0 25 25 76 43 796 8591 19187 9342 1194 ENE 0 0 130 208 130 861 3381 19210 30623 54111 WATTS BAR E 0 2 55 53 78 252 2445 9497 38457 136395 ESE 0 2 7 53 38 482 9716 8837 10649 17404 GASEOUS WASTE SYSTEMS SE 0 2 4 47 58 591 4514 12085 3420 300 SSE 0 0 16 35 29 505 17835 10818 3969 3756 S 12 23 3 27 24 714 4018 8056 3899 6362 SSW 0 54 14 24 257 1368 1141 34699 40812 11522 SW 0 34 7 19 32 739 5653 17523 25829 117868 WSW 0 0 5 2 0 519 6490 9411 68565 125338 W 0 10 40 38 30 1281 10369 2091 7134 6571 WNW 2 5 19 59 65 837 965 5337 2839 2035 NW 5 30 10 140 121 244 1461 2925 3440 17598 NNW 0 10 111 113 387 2279 314 7266 7004 9802 Total 0 0 62 87 98 2081 874 18279 4784 2983 19 308 540 1028 1525 14442 79838 187387 264219 517279 No. 16 - Replace with attached revised 11.3-29 WBNP-102 table

Table 11.3-9 Projected 2040 Population Distribution Within 50 Miles of Watts Bar Nuclear Plant Population Within Each Sector Element Distance from Site (Miles)

Direction 0-10 10-20 20-30 30-40 40-50 Total N 2,619 1,885 2,778 4,768 6,172 18,222 NNE 2,150 11,762 18,766 14,502 2,547 49,727 NE 1,441 3,783 16,734 29,838 78,334 130,130 ENE 1,110 3,553 29,539 63,798 253,831 351,832 E 1,915 11,352 18,647 30,063 44,013 105,990 ESE 135 6,230 20,120 5,068 3,280 34,833 SE 203 19,852 15,185 3,950 4,822 44,012 SSE 782 8,951 12,907 2,918 48,593 74,151 S 5,823 4,586 42,883 56,430 17,985 127,707 SSW 567 5,725 42,517 46,281 106,392 201,482 SW 1,051 12,978 14,499 62,307 111,795 202,630 WSW 938 12,791 2,837 2,840 3,372 22,778 W 937 3,406 5,555 2,944 5,474 18,316 WNW 717 2,091 4,372 5,654 20,511 33,345 NW 3,998 2,889 18,634 10,462 15,956 51,940 NNW 3,413 1,536 33,843 11,609 5,890 56,290 TOTAL 27,799 113,368 299,818 353,432 728,968 1,523,385 No. 16 - New Data for Table 11.3.9

No. 17 -

WATTS BAR WBNP-102 Replace with 0.479 Table 11.3-10 Watts Bar Nuclear Plant- Individual Doses From Gaseous Effluents 1.62 (For 1 Unit without TPC)

Effluent Pathway Guideline* Location Dose 0.38 Noble Gases  Air dose 10 mrad Maximum Exposed 0.801 mrad/yr Individual1 1.02

 Air dose 20 mrad Maximum Exposed 2.710 mrad/yr Individual1 1.70 Total body 5 mrem Maximum Residence2,3 0.571 mrem/yr Skin 15 mrem Maximum Residence2,3 1.540 mrem/yr Iodines/ Thyroid 15 mrem Maximum Real 2.715 mrem/yr Particulates (critical organ) Pathway4 No. 17 - Replace with: Total Vegetable Ingestion 0.97 No. 17 -

Breakdown of Iodine/Particulate Doses (mrem/yr)

Replace with Cow Milk with Feeding Factor of 0.33 2.44 0.322 Inhalation 0.174 0.0499 Ground Contamination 0.0405 Submersion 0.0603 0.0685 No. 17 -Replace with "5" Beef Ingestion1 0.0 0.285 Total 2.7148

• 1.6954 Guidelines are defined in Appendix I to 10 CFR Part 50.

No. 17 - Replace with "1280" 1

Maximum exposure point is at 1250 meters in the SE sector.

2 No. 17 - Replace with "E" Dose from air submersion.

3 Maximum exposed residence is at 1372 meters in the SE sector. No. 17 - Replace with "1979" No. 17 - Replace with "child" 4

Maximum exposed individual is an infant at 3353 meters in the SSW sector.

No. 17 -Insert "5 Maximum dose location for all receptors is 1280 meters in the E Sector."

11.3-30 GASEOUS WASTE SYSTEMS

WATTS BAR WBNP-102 Table 11.3-11 Summary Of Population Doses THYROID Infant Child Teen Adult Total Submersion 8.28E-02 1.59E-01 1.44E-01 6.28E-01 9.45E-01 Ground 3.11E-03 3.49E-02 3.17E-02 1.38E-01 2.08E-01 Inhalation 7.45E-02 1.39E-00 7.44E-01 2.64E+00 4.85E+00 Cow Milk Ingestion 4.09E-01 1.98E-00 8.42E-01 1.60E-00 4.83E+00 Beef Ingestion 0.00E+00 3.52E-01 1.77E-01 8.93E-01 1.42E-00 Vegetable Ingestion 0.00E+00 1.18E-00 4.76E-01 1.26E-01 2.92E+00 Total man-rem 5.01E-01 5.10E+00 2.42E+00 7.15E+00 1.52E+01 TOTAL BODY Infant Child Teen Adult Total Submersion 1.42E-02 1.59E-01 1.44E-01 6.28E-01 9.45E-01 Ground 3.11E-03 3.49E-02 3.17E-02 1.38E-01 2.08E-01 Inhalation 4.28E-03 1.14E-01 7.23E-02 2.99E-01 4.90E-01 Cow Milk Ingestion 1.14E-01 6.30E-01 2.39E-01 4.25E-01 1.41E-00 Beef Ingestion 0.00E+00 3.36E-01 1.69E-01 8.52E-01 1.36E-00 Vegetable Ingestion 0.00E+00 1.20E-00 5.08E-01 1.42E-00 3.12E+00 Total man-rem 1.36E-01 2.47E+00 1.16E-00 3.76E+00 7.53E+00 No. 18 - Replace with attached revised table GASEOUS WASTE SYSTEMS 11.3-31

Table 11.3-11 Summary of Population Doses THYROID Infant Child Teen Adult Total Submersion 1.26e-02 1.41e-01 1.28e-01 5.57e-01 8.38e-01 Ground 2.31e-03 2.59e-02 2.36e-02 1.03e-01 1.54e-01 Inhalation 6.62e-02 1.24e+00 6.64e-01 2.36e+00 4.33e-00 Cow Milk Ingestion 3.22e-01 1.57e+00 6.63e-01 1.25e+00 3.81e+00 Beef Ingestion 0.00e+00 3.17e-01 1.59e-01 8.04e-01 1.28e+00 Vegetable Ingestion 0.00e+00 1.04e+00 4.16e-01 1.09e+00 2.55e+00 Total man-rem 4.04e-01 4.34e+00 2.05e+00 6.17e+00 1.30e+01 TOTAL BODY Infant Child Teen Adult Total Submersion 1.26e-02 1.41e-01 1.28e-01 5.57e-01 8.38e-01 Ground 2.31e-03 2.59e-02 2.36e-02 1.03e-01 1.54e-01 Inhalation 3.93e-03 1.05e-01 6.65e-02 2.76e-01 4.52e-01 Cow Milk Ingestion 1.04e-01 5.73e-01 2.17e-01 3.85e-01 1.28e+00 Beef Ingestion 0.00e+00 3.06e-01 1.53e-01 7.74e-01 1.23e+00 Vegetable Ingestion 0.00e+00 1.05e+00 4.40e-01 1.21e+00 2.70e+00 Total man-rem 1.23e-01 2.20e+00 1.03e+00 3.31e+00 6.66e+00 No. 18 - New Data for Table 11.3.11

Completion and Operation of Watts Bar Nuclear Plant Unit 2 Table 3-19. Receptors from Actual Land Use Survey Results Used for Potential Gaseous Releases From WBN Unit 2 Receptor Receptor Distance Sector Number Type (meters) 1 Nearest Residence N 2134 2 Nearest Residence NNE 3600 3 Nearest Residence NE 3353 4 Nearest Residence ENE 2414 5 Nearest Residence E 3139 6 Nearest Residence ESE 4416 7 Nearest Residence SE 1372 8 Nearest Residence SSE 1524 9 Nearest Residence S 1585 10 Nearest Residence SSW 1979 11 Nearest Residence SW 4230 12 Nearest Residence WSW 1829 13 Nearest Residence W 2896 14 Nearest Residence WNW 1646 15 Nearest Residence NW 3048 16 Nearest Residence NNW 4389 17 Nearest Garden N 7644 18 Nearest Garden NNE 6173 19 Nearest Garden NE 3829 20 Nearest Garden ENE 4831 21 Nearest Garden E 8005 22 Nearest Garden ESE 4758 23 Nearest Garden SE 4633 24 Nearest Garden SSE 2043 25 Nearest Garden S 4973 26 Nearest Garden SSW 2286 27 Nearest Garden SW 8100 28 Nearest Garden WSW 4667 29 Nearest Garden W 5150 30 Nearest Garden WNW 5793 31 Nearest Garden NW 3170 32 Nearest Garden NNW 4698 33 Milk Cow ESE 6096 34 Milk Cow ESE 6706 35 Milk Cow SSW 2286 36 Milk Cow SSW 3353 37 Milk Cow NW 8100 Replace this data using updated data in the following table 86 Final Supplemental Environmental Impact Statement

Use this updated data in Completion and Operation of Watts Bar Nuclear Plant Unit 2 place of the data in the prior table Receptors from 2007 Actual Land Use Table 3-19 Survey Results Used for Potential Gaseous Releases From WBN Unit 2 Receptor Receptor Distance Sector Number Type (meters)

1. Nearest Resident N 2134

2. Nearest Resident NNE 3600

3. Nearest Resident NE 3353

4. Nearest Resident ENE 2414

5. Nearest Resident E 3268

6. Nearest Resident ESE 4416

7. Nearest Resident SE 1372

8. Nearest Resident SSE 1524

9. Nearest Resident S 1585

10. Nearest Resident SSW 1979

11. Nearest Resident SW 4230

12. Nearest Resident WSW 1829

13. Nearest Resident W 2896

14. Nearest Resident WNW 1646

15. Nearest Resident NW 2061

16. Nearest Resident NNW 4389

17. Nearest Garden N 7664

18. Nearest Garden NNE 6173

19. Nearest Garden NE 3353

20. Nearest Garden ENE 4927

21. Nearest Garden E 6372

22. Nearest Garden ESE 4758

23. Nearest Garden SE 4633

24. Nearest Garden SSE 7454

25. Nearest Garden S 2254

26. Nearest Garden SSW 1979

27. Nearest Garden SW 8100

28. Nearest Garden WSW 4667

29. Nearest Garden W 5120

30. Nearest Garden WNW 5909

31. Nearest Garden NW 3170

32. Nearest Garden NNW 4602

33. Milk Cow ESE 6706

34. Milk Cow SSW 2286

35. Milk Cow SSW 3353 86 Final Supplemental Environmental Impact Statement

Replace this data using Chapter 3 updated data in the following table Table 3-20. WBN Total Annual Gaseous Discharge Per Operating Unit (curies/year/reactor)

Containment Auxiliary Turbine Total per Nuclide Building Building Building Unit Kr-85m 1.99E+01 4.53E+00 1.23E+00 2.57E+01 Kr-85 6.90E+02 7.05E+00 1.86E+00 6.99E+02 Kr-87 1.09E+01 4.27E+00 1.09E+00 1.63E+01 Kr-88 2.83E+01 7.95E+00 2.13E+00 3.84E+01 Xe-131m 1.17E+03 1.73E+01 4.53E+00 1.19E+03 Xe-133m 4.63E+01 1.90E+00 5.21E-01 4.87E+01 Xe-133 3.12E+03 6.70E+01 1.77E+01 3.20E+03 Xe-135m 3.85E+00 3.68E+00 9.80E-01 8.51E+00 xXe-135 1.55E+02 2.40E+01 6.46E+00 1.85E+02 Xe-137 3.18E-01 9.67E-01 2.58E-01 1.54E+00 Xe-138 3.32E+00 3.42E+00 9.06E-01 7.65E+00 Ar-41 3.40E+01 0.00E+00 0.00E+00 3.40E+01 Br-84 6.00E-05 5.01E-02 4.81E-04 5.06E-02 I-131 7.29E-03 1.39E-01 7.08E-03 1.53E-01 I-132 1.60E-03 6.56E-01 1.70E-02 6.75E-01 I-133 3.55E-03 4.35E-01 2.03E-02 4.59E-01 I-134 1.66E-03 1.06E+00 1.47E-02 1.08E+00 I-135 3.16E-03 8.10E-01 3.13E-02 8.44E-01 H-3 1.37E+02 0.00E+00 0.00E+00 1.37E+02 H-3 (TPC) 3.70E+02 0.00E+00 0.00E+00 3.70E+02 Cr-51 9.21E-05 5.00E-04 0.00E+00 5.92E-04 Mn-54 5.30E-05 3.78E-04 0.00E+00 4.31E-04 Co-57 8.20E-06 0.00E+00 0.00E+00 8.20E-06 Co-58 2.50E-04 2.29E-02 0.00E+00 2.32E-02 Co-60 2.61E-05 8.71E-03 0.00E+00 8.74E-03 Fe-59 2.70E-05 5.00E-05 0.00E+00 7.70E-05 Sr-89 1.30E-04 2.85E-03 0.00E+00 2.98E-03 Sr-90 5.22E-05 1.09E-03 0.00E+00 1.14E-03 Zr-95 4.80E-08 1.00E-03 0.00E+00 1.00E-03 Nb-95 1.80E-05 2.43E-03 0.00E+00 2.45E-03 Ru103 1.60E-05 6.10E-05 0.00E+00 7.70E-05 Ru-106 2.70E-08 7.50E-05 0.00E+00 7.50E-05 Sb-125 0.00E+00 6.09E-05 0.00E+00 6.09E-05 Cs-134 2.53E-05 2.24E-03 0.00E+00 2.27E-03 Cs-136 3.21E-05 4.80E-05 0.00E+00 8.01E-05 Cs-137 5.58E-05 3.42E-03 0.00E+00 3.48E-03 Ba-140 2.30E-07 4.00E-04 0.00E+00 4.00E-04 Ce-141 1.30E-05 2.64E-05 0.00E+00 3.94E-05 C-14 2.80E+00 4.50E+00 0.00E+00 7.30E+00 A companion figure, illustrating the release points for radioactive gaseous effluents from WBN is presented in Figure 3-9.

Final Supplemental Environmental Impact Statement 87

Use this updated data in place of the data in the Chapter 3 prior table Table 3-20 WBN Total annual Gaseous discharge Per Operating Unit (curies/year/reactor)

Containment Auxiliary Turbine Nuclide Total Building Building Building Kr-85m 3.72E+00 4.53E+00 1.23E+00 9.48E+00 Kr-85 6.69E+02 7.05E+00 1.86E+00 6.78E+02 Kr-87 4.48E-01 4.27E+00 1.09E+00 5.81E+00 Kr-88 3.10E+00 7.95E+00 2.13E+00 1.32E+01 Xe-131m 1.07E+03 1.73E+01 4.53E+00 1.09E+03 Xe-133m 4.07E+01 1.90E+00 5.21E-01 4.31E+01 Xe-133 2.82E+03 6.70E+01 1.77E+01 2.90E+03 Xe-135m 2.26E-02 3.68E+00 9.80E-01 4.68E+00 Xe-135 5.83E+01 2.40E+01 6.46E+01 8.88E+01 Xe-137 3.76E-04 9.67E-01 2.58E-01 1.23E+00 Xe-138 1.69E-02 3.42E+00 9.06E-01 4.34E+00 Ar-41 3.40E+01 0.00E+00 0.00E+00 3.40E+01 Br-84 8.16E-07 5.02E-02 4.81E-04 5.07E-02 I-131 6.74E-03 1.39E-01 7.08E-03 1.53E-01 I-132 1.36E-04 6.56E-01 1.70E-02 6.73E-01 I-133 2.36E-03 4.35E-01 2.03E-02 4.57E-01 I-134 4.26E-05 1.06E+00 1.47E-02 1.07E+00 I-135 8.80E-04 8.10E-01 3.13E-02 8.42E-01 H-3 1.39E+02 0.00E+00 0.00E+00 1.39E+02 H-3 (TPC) 3.70E+02 0.00E+00 0.00E+00 3.70E+02 Cr-51 9.21E-05 5.00E-04 0.00E+00 5.92E-04 Mn-54 5.30E-05 3.78E-04 0.00E+00 4.31E-04 Co-57 8.20E-06 0.00E+00 0.00E+00 8.20E-06 Co-58 2.50E-04 2.29E-02 0.00E+00 2.32E-02 Co-60 2.61E-05 8.71E-03 0.00E+00 8.74E-03 Fe-59 2.70E-05 5.00E-05 0.00E+00 7.70E-05 Sr-89 1.30E-04 2.85E-03 0.00E+00 2.98E-03 Sr-90 5.22E-05 1.09E-03 0.00E+00 1.14E-03 Zr-95 4.80E-08 1.00E-03 0.00E+00 1.00E-03 Nb-95 1.80E-05 2.43E-03 0.00E+00 2.45E-03 Ru-103 1.60E-05 6.10E-05 0.00E+00 7.70E-05 Ru-106 2.70E-08 7.50E-05 0.00E+00 7.50E-05 Sb-125 0.00E+00 6.09E-05 0.00E+00 6.09E-05 Cs-134 2.53E-05 2.24E-03 0.00E+00 2.27E-03 Cs-136 3.21E-05 4.80E-05 0.00E+00 8.01E-05 Cs-137 5.58E-05 3.42E-03 0.00E+00 3.48E-03 Ba-140 2.30E-07 4.00E-04 0.00E+00 4.00E-04 Ce-141 1.30E-05 2.64E-05 0.00E+00 3.95E-05 C-14 2.80E+00 4.50E+00 0.00E+00 7.30E+00 A companion figure illustrating the release points for radioactive gaseous effluents from WBN is presented in Figure 3-9.

Final Supplemental Environmental Impact Statement 87

Chapter 3 A tabulation of the resulting calculated gaseous doses to individuals per operational unit is given in Table 3-21.

Replace this data using updated data in the following table Table 3-21. WBN Doses From Gaseous Effluent For Unit 2 Without Tritium Production for Year 2040 Effluent Pathway Guideline1 Location Dose Maximum Exposed Noble Gases J Air dose 10 mrad 0.801 mrad/year Individual2 Maximum Exposed E Air dose 20 mrad 2.710 mrad/year Individual2 Total body 5 mrem Maximum Residence3,4 0.571 mrem/year Iodines/ Skin 10 mrem Maximum Residence3,4 1.540 mrem/year Particulate Thyroid (critical organ) 15 mrem Maximum Real Pathway5 2.715 mrem/year Breakdown of Iodine/Particulate Doses (mrem/yr)

Cow Milk with Feeding Factor of 0.65 2.44 Inhalation 0.174 Ground Contamination 0.0405 Submersion 0.0603 Beef Ingestion2 0.00 Total 2.7148 1

2 Guidelines are defined in Appendix I to 10 CFR Part 50.

3 Maximum exposure point is at 1250 meters in the ESE sector.

4 Dose from air submersion.

5 Maximum exposed residence is at 1372 meters in the SE sector.

Maximum exposed individual is an infant at 3353 meters in the SSW sector.

The estimated annual airborne releases and resulting doses as presented by the 1972 FES, the WBN Unit 1 FSAR, Unit 2, Unit 1 and 2 totals, and recent historical data from WBN Unit 1 (as submitted in the Annual Radioactive Effluent Reports to the NRC) with NRC guidelines given in 10 CFR 50 Appendix I are compared in Table 3-22. These guidelines are designed to assure that releases of radioactive material from nuclear power reactors to unrestricted areas during normal conditions, including expected occurrences, are kept as low as practicable.

Final Supplemental Environmental Impact Statement 89

Chapter 3 A tabulation of the resulting calculated gaseous doses to individuals per operational unit is given in Table 3-21.

Use this updated data in place of the data in the prior table Table 3-21 WBN Doses From Gaseous Effluent for Unit 2 Without Tritium Production for Year 2040 Effluent Pathway Guideline* Location Dose Maximum Exposed Noble Gases  Air dose 10 mrad 0.479 mrad/year Individual1 Maximum Exposed

 Air dose 20 mrad 1.62 mrad/year Individual1 Total body 5 mrem Maximum Residence2,3 0.38 mrem/year Iodines/

Skin 10 mrem Maximum Residence2,3 1.02 mrem/year Particulate Thyroid 15 mrem Maximum Real Pathway4 1.70 mrem/year (critical organ)

Breakdown of Iodine/Particulate Doses (mrem/yr)

Total Vegetable Ingestion 0.97 Inhalation 0.322 Ground Contamination 0.0499 Submersion 0.0685 Beef Ingestion5 0.285 Total 1.6954 Guidelines are defined in Appendix I to 10 CFR Part 50.

1 Maximum exposure point is at 1280 meters in the E sector.

2 Dose from air submersion.

3 Maximum exposed residence is at 1372 meters in the SE sector.

4 Maximum exposed individual is a child at 1979 meters in the SSW sector.

5 Maximum dose location for all receptors is 1280 meters in the E Sector.

The estimated annual airborne releases and resulting doses as presented by the 1972 FES, the WBN Unit 1 FSAR, Unit 2, Unit 1 and 2 totals, and recent historical data from WBN Unit 1 (as submitted in the Annual Radioactive Effluent Reports to the NRC) with NRC guidelines given in 10 CFR 50 Appendix I are compared in Table 3-22. These guidelines are designed to assure that releases of radioactive material from nuclear power reactors to unrestricted areas during normal conditions, including expected occurrences, are kept as low as practicable.

Final Supplemental Environmental Impact Statement 89

Enclosure 4 Watts Bar Nuclear Plant List of Commitments

1. In the footnote added to Table 11.2-5 by Amendment 102, the term F/H1D in the formulation of Column 5 and Mobile in the definition of D should be, F/H/D and Mobile, respectively. These items will be corrected in FSAR Amendment 103.

(Question 9)

2. Table 11.3-10 of the FSAR will be corrected to reflect the 2007 feeding factors and the offsite radiation doses calculated without terrain adjustment factors. These changes to Table 11.3-10 will be reflected in Amendment 103. (Question 14)

3. TVA has reviewed the FSEIS and found Table 3-20 to be in error. This was caused by the use of values contained in FSAR Table 11.3.7 instead of values contained in FSAR Table 11.3.7c. The correct source term used for calculating the site boundary doses is FSAR Table 11.3.7c. As a result, this accounts for the dose values being same between the FSEIS and the FSAR Table 11.3-10. (Question 15)

4. FSAR Section 11.3.10.1, Assumptions and Calculation Methods incorrectly states the dose to the critical organ from radioiodines, tritium, and particulates is calculated for real pathways existing at the site during a land use survey conducted in 1994. The feeding factor of 70% is the feeding factor associated with the 1994 land use survey. The feeding factor of 65% listed in Table 3-21 of the FSEIS is in error and should be 0.33%. These changes to FSAR Section 11.3.10.1 will be reflected in Amendment 103. (Question 18)

5. Further, comparisons with other models determined that MESOPUFF II is not suitable for calculating /Q values at WBN receptors, and that GELC adequately estimates /Q for WBN receptors, without any need for adjustments. Therefore, WBN can eliminate the use adjustment factors and use GELC results directly. These changes will be reflected in Table 11.3-8 in FSAR, Amendment 103. (Question 20)

6. TVA has reviewed the FSEIS and found the land use data presented in Table 3-19 to be in error. The land use survey used to develop Table 11.3-10 was from 2007. Table 11.3-10 of the FSAR will be revised to include 2007 feeding factors and the offsite radiation doses being calculated without terrain adjustment factors. These changes to Table 11.3-10 will be reflected in Amendment 103. (Question 21)

7. TVA will provide an update in a future FSAR amendment. (Question 22, 23, 28, and 29)

8. FSAR section 12.3.2.2 will be revised to list any applicable additional areas addressed by the mission dose calculations. (Question 30.1.b)

9. The liquid source term used for the sample in WBNTSR-084 is the normal RCS source term, which is based on ANSI/ANS 18.1, 1984. The airborne activity used for the mission is that of a LOCA. It is expected that use of the LOCA source terms will bound use of the RCS source term with an Iodine spike. However, TVA will perform the calculation using the steam generator tube rupture source term. (Question 30.3)

E4-1

Enclosure 4 Watts Bar Nuclear Plant List of Commitments

10. TVA will revise calculations WBNTSR-081 and WBNTSR-092 to specify mission times.

(Question 30.4)

11. Mission dose calculations that are currently only applicable to Unit 1 are being updated to make them applicable to Unit 2. (Question 30.5)

12. The FSAR will be revised to eliminate the adjustment factors and use GELC results directly.

Specifically, Table 11.3-10 (Unit 2 only) dose values for Noble Gases and Iodines/Particulates will be revised. In addition, due to elimination of the terrain adjustment factors, the highest dose pathway becomes vegetable ingestion instead of the cow milk with feeding factor. Doses reflected in this table will be of one unit (Unit 2) without a Tritium Producing Core. These changes will be submitted as part of Unit 2 FSAR, Amendment 103.

(Enclosure 2 - Question 11.3.a)

E4-2

Attachment 1 Watts Bar Nuclear Plant Calculation WBN EEB EDQ1090-99005 Extending Channel Operational Test Frequency for Radiation Monitors E4-1