ML110620687

Final Safety Analysis Report (FSAR) - Response to Chapters 11 and 12 Request for Additional Information

ML110620687
Person / Time
Site:	Watts Bar
Issue date:	02/25/2011
From:	Stinson D Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML110620687 (113)

Text

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.

U.S. Nuclear Regulatory Commission Page 2 February 25, 2011 provides the list of commitments made in this letter. provides TVA calculation WBN EEB EDQ1090-99005, "Extending Channel Operational Test Frequency for Radiation Monitors," which is referenced as part of RAI response of Question 24 provided in Enclosure 1.

If you have any questions, please contact Bill Crouch at (423) 365-2004.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 2 5t h day of February, 2011.

Respectfully, David Stinson Watts Bar Unit 2 Vice President

Enclosures:

1. Response to Chapters 11 and 12 RAIs 2

Outstanding Chapter 11 RAIs

3. Proposed FSAR Chapter 11 Markups and Final Supplemental Environmental Impact Statement Markups

4. List of Commitments Attachment

1. Calculation WBN EEB EDQ1090-99005, "Extending Channel Operational Test Frequency for Radiation Monitors" cc (Enclosures):

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

Enclosure I 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 L 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 uCilcc and 3.65E-5 uCilcc, respectively. However, column 7 and column 8 do not include Condensate Demineralizer backwash wastes.

I 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 I 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 uCilcc 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 uCilcc. [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 TVA 's 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 I 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 applicant's dose assessment. The staff determined that there is sufficient agreement between the TVA's and the staff's 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 TVA's 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.

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Enclosure I 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 and1l.2-Sd, 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 I 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/HID" in the formulation of Column 5 and "Mobi"le" in the definition of "D" should be, "F/HID" 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 "Mobi"le" in the definition of "D" should be, "F/HID" 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/HID") that the regeneration waste is processed through prior to E1-5 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 "FIHID" 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/HID" 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 I 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 uCilcc, 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 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 L 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 TVA's '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 I Watts Bar Nuclear Plant Response to Chapters 11 and 12 Requests for Additional Information FSEIS. In response to the staff's 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 TVA's dose results. The details of the staff's 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 staff's results generally agree with the TVA's calculations. Bases on its conservative assumptions, the staff's calculations determined that the maximum exposed organ expected from radioactive iodine and particulates in gaseous effluents, is 10.78 mrem. Although both TVA's and the staff's 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 I 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:

IVA 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.

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Enclosure I 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 lodinevalues 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 7.15E1 1 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 1.22E1 3 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-O017. 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.

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Enclosure I 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 site area. The 2007 data for these three farms are provided below:

near the WBN Percent Substitutional Feedin for Dairy and Goat Herds 2007 Farm Distance Total I (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 1 100 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.

El-11

Enclosure I 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 x/Q values at WBN receptors, and that GELC adequately estimates x/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 1-9 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 I 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 TVA's response, the staff concluded that the FSAR changes did not impact the staff's 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:

El-13

Enclosure I 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 I 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 TVA's 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 El-14

Enclosure I 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.

El-15

Enclosure I 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:

"21D.(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-1 861 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 TVA's 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 I was less that 100 person-rem, the staff El-16 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 staff's 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 doesn't meet the RPM qualifications (ANSIIANS 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 ANSIIANS 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 staff's 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.

El-17 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 MCRITSR 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 I 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.

El-19

Enclosure I 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.

E1-20

Enclosure I 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 ll.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.

El-21 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 Liquid Single Unit Steam Single Generator Single Unit Totals Nuclide Radwaste Bodw iy CyrBlowdown Ci/yr Ci/yr Ci/yr Br-84 1.65E-04 5.23E-04 6.88E-04 1-131 2.63E-02 1.14E+00 1.16E+00 1-132 1.32E-02 1'.08E-01 1.21 E-01 1-133 5.29E-02 8.57E-01 9.1OE-01 1-134 6.26E-03 2.65E-02 3.28E-02 1-135 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 2.21 E-01 2.61 E-01 Na-24 1.86E-02 0.OE+00 1.86E-02 Cr-51 7.03E-03 9.27E-02 9.98E-02 Mn-54 4.99E-03 5.1OE-02 5.59E-02 Fe-55 8.09E-03 0.OE+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 Watts Bar Nuclear Plant Remaining Chapter 11 Request for Additional Information Single Unit Liquid Single Unit Steam Radwaste Generator Single Unit Totals Nucid Blowdown Ci/yr Ci/yr Ci/yr Zn-65 3.82E-04 0.OE+00 3.82E-04 Sr-89 1.92E-04 4.33E-03 4.52E-03 Sr-90 2.20E-05 3.88E-04 4.1OE-04 Sr-91 2.84E-04 2.18E-03 2.47E-03 Y-91 m 1.68E-04 0.OE+00 1.68E-04 Y-91 9.OOE-05 3.OOE-04 3.90E-04 Y-93 1.27E-03 0.OE+00 1.27E-03 Zr-95 1.39E-03 1.20E-02 1.34E-02 Nb-95 2.1OE-03 8.98E-03 1.11E-02 Mo-99 4.20E-03 9.95E-02 1.04E-01 Tc-99m 3.35E-03 0.OE+00 3.35E-03 Ru-103 5.88E-03 0.OE+00 5.88E-03 Ru-106 7.63E-02 0.OE+00 7.63E-02 Te-129m 1.41 E-04 0.OE+00 1.41E-04 Te-129 7.30E-04 0.OE+00 7.30E-04 Te-131m 8.05E-04 0.OE+00 8.05E-04 Te-131 2.03E-04 0.OE+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 3.41 E-04 0.OE+00 3.41 E-04 Ce-143 1.53E-03 0.OE+00 1.53E-03 Ce-144 6.84E-03 1.26E-01 1.33E-01 Np-239 1.37E-03 0.OE+00 1.37E-03 H-3 1.25E+03 O.OE+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 sin-gle 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 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-001 7, 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-1 8.1-1984 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 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 1 E-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 @ 1 PCA (of SSS) k)

Turbine Building floor drains, 7200 gal/ day @ 1 PCA (of SSC) (note: no RC in Turbine Building).

I)

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-001 7 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 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)]IDFMOBDEM + 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:

(4.4o2Ci/yrl*E6uCi/Ci)

= 3.6528E-5 uCYcc (24hr/day*3E41b/hr*453.59g/lb lcc/g*36Sday/hr)

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 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 x/Q values at WBN receptors, and that GELC adequately estimates x/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 lodines/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 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 I 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 I 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 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 I of 1)

Direction 0-10 10-20 20-30 30-40 40-50 Total N

NNE NE ENE E

ESE SE SSE S

SSW SW WSW W

WNW NW NNW TOTAL 2,541 2,218 2,281 4,460 6,373 17,873 1,687 11,747 18,599 12,607 2,549 47,189 1,524 3,597 16,808 26,935 80,896 129,760 1,174 4,918 31,814 72,849 244,656 355,411 4,811 9,773 17,518 24,692 46,384 103,178 890 6,151 19,601 4,909 3,336 34,887 961 19,601 17,155 4,359 3,985 46,021 2,051 8,838 13,196 3,083 38,513 65,681 6,157 4,070 42,757 56,934 16,750 126,668 599 3,215 39,231 42,901 106,346 192,292 1,056 13,605 14,537 60,959 127,447 217,604 943 12,996 2,714 2,667 3,603 22,923 941 3,150 4,984 2,771 5,249 17,095 721 1,981 3,729 5,400 19,945 31,776 4,018 3,302 13,705 8,129 14,875 44,029 3,430 1,586 33,560 11,512 6,092 56,180 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 WATTS BAR WBNP-1 02 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-001 7.

The noble gases and iodines discharged from the various sources are entered in Table 11'310.-

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 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.

Non-radioactive 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.

Ny j

U 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."I Releases from the Shield Building, Turbine Building (TB), and Auxiliary Building (AB) vents are frefid as nrund_

level. The ground level joint frequency distribution (JFD) is given inlSection 2.3. Air concentrations and deposition rates were calculated considering radioactive and buildup during transit. Plume depletion was calculated using the figures pro d in Regulatory Guide 1.111.

No5 - Replace with Estimates of normalized concentrations (X/Q) and normalized deposr"the ODCM."

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 potent* maximum exposure (i.e.,

points at the unrestricted area boundary). External s and total body exposures are evaluated at nearb residences. The dose to the tical organ from radioiodines, tritium (Unit 1 only)land particulates is calcul t for real pathways existing at the site No. 6 - Dele~te durin land use survey conducted in 994 No. 6 - Replace with "2007" STo 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 orst-GASEOUS WASTE SYSTEMS No 6 Delete I 3-9

WATTS BAR WBNP-102 No. 7 - Delete No. 7 - Replace with "0.33" INO'7- '

_/,.J'

,*./"No.

7 - Replace with "'past"'

case feedi or identified during the ý 9941land 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 1f' exp(-)Xit)dt 1 -exp(-Xi180) 180 J 180ki 0

This factor replaces the factor exp (-Ai th) in equation C-1 0 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

[1,100,0001persons expected to live within a 50 mile radius of the Watts Bar Nuclear Plant in the year 2040 is calculated to be approximately140 an-rem assuming 140 mrem/year/individual. By comparison, the same popul n excluding onsite radiation workers) will receive a total body dose of approxi ately 3.85 man-rem from effluents. Based on these results, TVA concludes that thýnorr$ operation of the Watts Bar Nuclear Plant will present minimal risk to the h alth a d safety of the public.

REFERENCES No. 8-Replace with "210,000"-

None No. 8 - Replace with "6.66" INo.

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.

Nuclide Kr-85m Kr-85 Kr-87 Kr-88 Xe-131m Xe-133m Xe-133 Xe-135m Xe-135 Xe-1 37 Xe-1 38 Ar-41 Br-84 1-131 1-132 1-133 1-134 1-135 H-3 Contain.(1)

Building 2.OOE+01 6.90E+02 1.09E+01 2.84E+01 1.17E+03 4.63E+01 3.12E+03 3.86E+00 1.55E+02 3.18E-01 3.33E+00 3.40E+01 6.OOE-05 7.29E-03 1.61 E-03 3.55E-03 1.66E-03 3.16E-03 1.39E+02 Aux.

Building 4.53E+00 7.05E+00 4.27E+00 7.95E+00 1.73E+01 1.90E+00 6.70E+01 3.68E+00 2.40E+01 9.67E-01 3.42E+00 0.00E+00 5.02E-02 1.39E-01 6.56E-01 4.35E-01 1.06E+00 8.10E-01 0.00E+00 Turbine Building 1.23E+00 1.86E+00 1.09E+00 2.13E+00 4.53E+00 5.21 E-01 1.77E+01 9.80E-01 6.46E+00 2.58E-01 9.06E-01 0.00E+00 4.81 E-04 7.08E-03 1.70E-02 2.03E-02 1.47E-02 3.13E-02 0.OOE+00 Total 2.58E+01 6.99E+02 1.62E+01 3.85E+01 1.19E+03 4.88E+01 3.20E+03 8.52E+00 1.85E+02 1.54E+00 7.66E+00 3.40E+01 5.07E-02 1.53E-01 6.75E-01 4.58E-01 1.08E+00 8.45E-01 1.39E+02

[ H-3 (TPC)(3)

Unit 1 Only 3.70E+02 0.00E+00 0.OOE+00 3.70E+02 Cr-51 9.21 E-05 5.00E-04 0.00E+00 5.92E-04 Mn-54 5.30E-05 3.78E-04 0.OOE+00 4.31 E-04 Co-57 8.20E-06 0.00E+00 0.00E+00 8.20E-06 Co-58 2.50E-04 2.29E-02 0.OOE+00 2.32E-02 Co-60 2.61E-05 8.71 E-03 0.OOE+00 8.74E-03 Fe-59 2.70E-05 5.OOE-05 0.OOE+00 7.70E-05 Sr-89 1.30E-04 2.85E-03 0.0OE+00 2.98E-03 Sr-90 5.22E-05 1.09E-03 0.00E+00 1.14E-03 Zr-95 4.80E-08 1.OOE-03 0.OOE+00 1.00E-03 Nb-95 1.80E-05 2.43E-03 0.OOE+00 2.45E-03 Ru-103 1.60E-05 6.10E-05 0.OOE+00 7.70E-05 Ru-1 06 2.70E-08 7.50E-05 0.OOE+00 7.50E-05 Sb-125 0.OOE+00 6.09E-05 0.00E+00 6.09E-05 Cs-134 2.53E-05 2.24E-03 0.OOE+00 2.27E-03 Cs-136 3.21E-05 4.80E-05 0.OOE+00 8.01E-05 Cs-137 5.58E-05 3.42E-03 0.OOE+00 3.48E-03 Ba-140 2.30E-07 4.OOE-04 0.00E+00 4.OOE-04 Ce-141 1.30E-05 2.64E-05 0.OOE+00 3.95E-05 C-14 2.80E+00 4.50E+00 0.OOE+00 7.30E+00 (1) Includes release from GWPS (2) 4.28E+02 = 4.28 X 102 E (7-Tritium values for a Tritim Production Core N.9-Dlt 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 Purgel (Sheet 1 of 2) 1 1

Kr-85m Kr-85 Kr-87 Kr-88 Xe-1 31 m Xe-1 33m Xe-1 33 Xe-1 35m Xe-1 35 Xe-1 38 Br-84 1-131 1-132 1-133 1-134 1-135 Cs-134 Cs-1 36 Cs-137 Cr-51 Mn-54 Fe-59 Co-58 Co-60 Sr-89 Sr-90 Zr-95 Nb-95 Ba-140 Exp. Rel.

(Ci/yr) 2.58E+01 6.99E+02 1.62E+01 3.85E+01 1.19E+03 4.88E+01 3.20E+03 8.52E+00 1.85E+02 7.66E+00 5.07E-02 1.53E-01 6.75E-01 4.58E-01 1.08E+00 8.45E-01 2.27E-03 8.01 E-05 3.48E-03 5.92E-04 4.31 E-04 7.70E-05 2.32E-02 8.74E-03 2.98E-03 1.14E-03 1.00E-03 2.45E-03 4.OOE-04 1 39F+0l2 Design Des/Exp (Ci/yr)

Single Unit Design 10CFR20 Operation (pCi/cc)

(ECL)

C/ECL 12.28 33.08 7.45 12.33 2.91 43.24 111.07 5.04 6.97 5.43 2.50 52.41 4.00 26.85 1.65 7.91 40.60 165.20 153.22 0.29 0.47 3.48 5.37 1.38 22.45 13.49 1.71 2.34 0.311 3.17E+02 2.31 E+04 1.21 E+02 4.75E+02 3.45E+03 2.11E+03 3.55E+05 4.29E+01 1.29E+03 4.16E+01 1.27E-01 8.03E+00 2.70E+00 1.23E+01 1.78E+00 6.69E+00 9.20E-02 1.32E-02 5.33E-01 1.73E-04 2.03E-04 2.68E-04 1.24E-01 1.21 E-02 6.69E-02 1.54E-02 1.71 E-03 5.73E-03

.1.26E-04 1 _39QF02 1.10E-10 7.99E-09 4.18E-11 1.64E-10 1.19E-09 7.29E-10 1.23E-07 1.48E-11 4.46E-1 0 1.44E-11 4.38E-14 2.77E-1 2 9.33E-1 3 4.25E-1 2 6.14E-13 2.31 E-12 3.18E-14 4.57E-1i5 1.84E-13 5.96E-1 7 7.01 E-1 7 9.27E-1 7 4.30E-14 4.17E-15 2.31 E-14 5.33E-1i5 5.92E-1 6 1.98E-15 4.34E-17 4.80E-11I 1.OE-07 7.OE-07 2.OE-08 9.OE-09 2.OE-06 6.OE-07 5.0E-07 4.0E-08 7.OE-08 2.OE-08 8.0E-08 2.OE-10 2.OE-08 1.01E-09 6.0E-08 6.OE-09 2.OE-1 0 9.0E-10 2.OE-10 3.OE-08 1.OE-09 5.OE-10 1.OE-09 5.OE-11 1.01E-09 6.0E-12 4.0E-1 0 2.0E-09 2.0E-09 1 _lE-07 0.0010951 0.0114124 0.0020906 0.0182306 0.0005971 0.0012142 0.2456675 0.0003710 0.006375 0.0007188 5.478E-07 0.013875 4.67E-05 0.0042535 1.023E-05 0.0003851 0.0001589 5.079E-06 0.0009203 1.988E-09 7.005E-08 1.853E-07 4.298E-05 8.333E-05 2.313E-05 0.0008877 1.481 E-06 9.895E-07 2.171 E-08 0 oflff4mi 1

Dual Unit Operation C/ECL 0.0021902 0.0228248 0.0041812 0.0364612 0.0011942 0.0024284 0.4913350 0.0007420 0.012750 0.0014376 1.096E-06 0.027750 0.0000934 0.0085070 2.046E-05 0.0007702 0.0003178 1.016E-05 0.0018406 3.976E-09 1.401 E-07 3.706E-07 8.596E-05 1.667E-04 4.626E-05 0.0017754 2.962E-06 1.979E-06 4.342E-08 n.0009R22 I H-3 (TPC) 3.70E+02 1 3.70E+02 1.28E-10 1.OE-07 0.0012775 0.0012775 1 rod 1.53E+03 1 1.53E+03 5.29E-10 1.OE-07 0.0052869 0.0052869 2 rod 2.69E+03 1 2.69E+03 9.30E-10 1.OE-07 0.0092962 0.0092962 C-14 7.30E+00 1 7.30E+00 2.52E-12 3.OE-09 0.000841 0.001682 Ar-41 3.40E+01 1 3.40E+01 1.18E-11 1.OE-08 0.0011752 0.0023504 4 Total 0.3109694 0.6219388 Total (TPC) 0.3117657 0.6227352 1 rod 0.3157751 0.6267446 2 rod 0.3197845 0.6307539 K w

1.

INo. 10- Delete I 11.3-22 GASEOUS WASTE SYSTEMS

WATTS BAR WBNP-102 Table 11.3-7a Design (For 1% Failed Fuel) Expected Gas Release Concentrationl(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-1 02 Table 11.3-7b Design (For 1% Failed Fuel) Expected Gas Release Concentration/(Effluent Concentration Limit) With Continuous Filtered Containment Ventl (Sheet I ot 2)

Kr-85m Kr-85 Kr-87 Kr-88 Xe-1 31 m Xe-i 33m Xe-1 33 Xe-1 35m Xe-135 Xe-138 Br-84 1-131 1-132 1-133 1-134 1-135 Cs-134 Cs-136 Cs-137 Cr-51 Mn-54 Fe-59 Co-58 Co-60 Sr-89 Sr-90 Zr-95 Nb-95 Ba-140

"-,I-Exp. Rel.

(Ci/yr) 9.48E+00 6.78E+02 5.81 E+00 1.32E+01 1.09E+03 4.31 E+01 2.90E+03 4.68E+00 8.88E+01 4.34E+00 5.07E-02 1.53E-01 6.73E-01 4.57E-01 1.07E+00 8.42E-01 2.27E-03 8.01 E-05 3.48E-03 5.92E-04 4.31 E-04 7.70E-05 2.32E-02 8.74E-03 2.98E-03 1.14E-03 1.OOE-03 2.45E-03 4.00E-04 1 qQI=i.n9 Des/Exp 12.28

.33.08 7.45 12.33 2.91 43.24 111.07 5.04 6.97 5.43 2.50 52.41 4.00 26.85 1.65 7.91 40.60 165.20 153.22 0.29 0.47 3.48 5.37 1.38 22.45 13.49 1.71 2.34 0.31 1

Design (Ci/yr) 1.16E+02 2.24E+04 4.33E+01 1.63E+02 3.18E+03 1.86E+03 3.22E+05 2.36E+01 6.19E+02 2.36E+01 1.27E-01 8.OOE+00 2.69E+00 1.23E+01 1.77E+00 6.66E+00 9.20E-02 1.32E-02 5.33E-01 1.73E-04 2.03E-04 2.68E-04 1.24E-01 1.21 E-02 6.69E-02 1.54E-02 1.71 E-03 5.73E-03 1.26E-04 I *~IQF.(V),

Design (pCi/cc) 4.02E-11 7.75E-09 1.50E-11 5.63E-1i1 1.1OE-09 6.44E-1 0 1.11E-07 8.15E-12 2.14E-10 8.15E-12 4.38E-14 2.77E-12 9.30E-1 3 4.24E-12 6.1 OE-1 3 2.30E-12 3.18E-14 4.57E-1i5 1.84E-13 5.96E-1 7 7.01 E-1 7 9.27E-1 7 4.30E-14 4.17E-15 2.31 E-14 5.33E-15 5.92E-16 1.98E-15 4.34E-17 A %NflF.I11 10CFR20 (ECL) 1.OE-07 7.OE-07 2.OE-08 9.OE-09 2.OE-06 6.OE-07 5.OE-07 4.OE-08 7.OE-08 2.OE-08 8.OE-08 2.OE-10 2.OE-08 1.OE-09 6.OE-08 6.OE-09 2.OE-10 9.OE-10 2.OE-1 0 3.OE-08 1.OE-09 5.OE-1 0 1.OE-09 5.OE-11 1.OE-09 6.OE-12 4.OE-10 2.OE-09 2.OE-09 1 npF.-n7 Single Unit Operation C/ECL 0.0004024 0.0110743 0.0007480 0.0062505 0.0005489 0.0010735 0.2227110 0.0002038 0.0030561 0.0004073 0.0000005 0.0138277 0.0000465 0.0042433 0.0000102 0.0003837 0.0001589 0.0000051 0.0009203 0.0000000 0.0000001 0.0000002 0.0000430 0.0000833 0.0000231 0.0008877 0.0000015 0.0000010 0.0000000 n nnn*q 11 Dual Unit Operation C/ECL 0.0008048 0.0221486 0.0014960 0.0125010 0.0010978 0.0021470 0.4454220 0.0004076 0.0061122 0.0008146 0.0000010 0.0276554 0.0000930 0.0084866 0.0000204 0.0007674 0.0003178 0.0000102 0.0018406 0.0000000 0.0000002 0.0000004 0.0000860 0.0001666 0.0000462 0.0017754 0.0000030 0.0000020 0.0000000 n Nnnar9')

H-3 (TPC) 3.70E+02 1

3.70E+02 1.28E-10 1'OE-07 0.0012775 0.0012775 1 rod 1.53E+03 1

1.53E+03 5.29E-10 1.OE-07 0.0052869 0.0052869 2 rod 2.69E+03 1

2.69E+03 9.30E-10 1.OE-07 0.0092962 0.0092962 C-14 7.30E+00 1

7.30E+00 2.52E-12 3.OE-09 0.0008410 0.0016820 Ar-41 3.40E+01 1

3.40E+01 1.18E-11 1.OE-08 0.0011752 0.0023504 Total 0.2696131 0.5392262 Total (TPC) 0.2704095 0.5400226 1rod 0.2744189 0.5440320 2 rod 0.2784283 0.5480413 INo. 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.

\\\\,4No.

13 - Delete GASEOUS WASTE SYSTEMS 11.3-25

WATTS BAR WBNP-102 Table 11.3-7c Total Releases (2 118 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 Kr-85 Kr-87 Kr-88 Xe-131m xe-133m Xe-133 Xe-135m Xe-135 Xe-137 Xe-138 Ar-41 Br-84 1-131 1-132 1-133 1-134 1-135 H-3 3.72E+00 6.69E+02 4.48E-01 3.10E+00 1.07E+03 4.07E+01 2.82E+03 2.26E-02 5.83E+01 3.76E-04 1.69E-02 3.40E+01 8.16E-07 6.74E-03 1.36E-04 2.36E-03 4.26E-05 8.80E-04 13 IE+02 4.53E+00 7.05E+00 4.27E+00 7.95E+00 1.73E+01 1.90E+00 6.70E+01 3.68E+00 2.40E+01 9.67E-01 3.42E+00 0.OOE+00 5.02E-02 1.39E-01 6.56E-01 4.35E-01 1.06E+00 8.10E-01 n tOOE+O 1.23E+00 1.86E+00 1.09E+00 2.13E+00 4.53E+00 5.21 E-01 1.77E+01 9.80E-01 6.46E+01 2.58E-01 9.06E-01 0.00E+00 4.81 E-04 7.08E-03 1.70E-02 2.03E-02 1.47E-02 3.13E-02 n n00E+n0 9.48E+00 6.78E+02 5.81 E+00 1.32E+01 1.09E+03 4.31E+01 2.90E+03 4.68E+00 8.88E+01 1.23E+00 4.34E+00 3.40E+01 5.07E-02 1.53E-01 6.73E-01 4.57E-01 1.07E+00 8.42E-01 1 _39E+n2 I

H-3 (TPC) 3.70E+02 0.OOE+00 0.OOE+00 3.70E+02 Cr-51 Mn-54 Co-57 Co-58 Co-60 Fe-59 Sr-89 Sr-90 Zr-95 Nb-95 Ru-103 Ru-106 Sb-125 Cs-134 Cs-136 Cs-137 Ba-140 Ce-141 C-14 9.21 E-05 5.30E-05 8.20E-06 2.50E-04 2.61 E-05 2.70E-05 1.30E-04 5.22E-05 4.80E-08 1.80E-05 1.60E-05 2.70E-08 0.00E+00 2.53E-05 3.21 E-05 5.58E-05 2.30E-07 1.30E-05 2.80E+00 5.OOE-04 3.78E-04 0.00E+00 2.29E-02 8.71 E-03 5.OOE-05 2.85E-03 1.09E-03 1.OOE-03 2.43E-03 6.10E-05 7.50E-05 6.09E-05 2.24E-03 4.80E-05 3.42E-03 4.OOE-04 2.64E-05 4.50E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.001E+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00 0.OOE+00

/

5.92E-04 4.31 E-04 8.20E-06 2.32E-02 8.74E-03 7.70E-05 2.98E-03 1.14E-03 1.OOE-03 2.45E-03 7.70E-05 7.50E-05 6.09E-05 2.27E-03 8.01 E-05 3.48E-03 4.00E-04 3.95E-05 7.30E+00 (TPC) Tritium values for a Tritium Production Core (Unit 1 only) L-I

.X---

FN-o.-l-4 --D-e-let-e--]

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/mA3)

(1/mA2)

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 1.1Oe-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 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/mA3)

(1/mA2)

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-1 0 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

r-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 Fing Sector (Meters)

(s/mA3)

(1/mA3)

Feeding Factor Unrestricted Area Boundary N

1550 3.01 e06 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 2.81 e-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 2.81 e-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 2.01 e-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-1 0 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)

(slmA3)

(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-1 1 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

C)

C,,

IT' 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)

ITr C,,

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 E

0 2

55 53 78 252 2445 9497 38457 136395 ESE 0

2 7

53 38 482 9716 8837 10649 17404 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 C/i (D

0,U CD 0

<CDO

<CD CL

.,0 zz-o

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

WATTS BAR WBNP-102 No. 17-Replace with 0.479 Table 11.3-10 Watts Bar Nuclear Plant-Individual Doses From Gaseous Effluents (For 1 Unit without TPC)

Effluent Pathway Guideline*

Location Dose Noble Gases y Air dose 10 mrad Maximum Exposed 0.801 mrad/yr Individual1 13 Air dose 20 mrad Maximum Exposed 2.710 mrad/yr Individual 1

Total body 5 mrem Maximum Residence 2,3 0.571 mrem/yr Skin 15 mrem Maximum Residence 2' 3 1.540 mrem/yr lodines/

Thyroid 15 mrem Maximum Real 12.715 mrem/yr Particulates (critical organ)

Pathway4 1.62 0.38 1.02 1.70

- rNO. 17 - Replace with: Total Vegetable Ingestion 0.97 I Breakdown of lodine/Particulate Doses (mrem/yr)

Re witt

.17-place h

Cow Milk with Feedinq Factor of 0.33 2.44 Inhalation Ground Contamination Qi ihmnre n

0.174 0.0405 0.0603 L

LdiJi,...1.3I'.JI I INo. 17 -Replace with BeefIngesI rl 0.322 0.0499 0.0685 0.285 1.6954 0.0 2.7148 Total Appedix to 0 CF Par 50 IG*uidelines are defined ir Appendix I to 10 CFR Part 50.

No. 17 - Replace with "1280"

'Maximum exposure point is at 1250 meters in the Es ctor.

2Dose from air submersion.

3NAo.,irm 17-ns,,,,r, ce is at 1372 meters in the SE sector.

No. 17-Replac No. 17 - Replace with "child"

[

1 4Maximum exposed individual is aniaat 335 meters in the SSW sector.

a-e with "1979" E 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.OOE+00 3.52E-01 1.77E-01 8.93E-01 1.42E-00 Vegetable Ingestion O.OOE+00 1.18E-00 4.76E-01 1.26E-01 2.92E+00 Total man-rem 5.01 E-01 5.1OE+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 1.41 E-00 Beef Ingestion 0.OOE+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 2.3 1e-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 3.3 le+00 6.66e+00

'No.

18-New Data for Table 11.3.11 1

Completion and Operation of Watts Bar Nuclear Plant Unit 2 Table 3-19.

Receptors from Actual Land Use Survey Results Used for Potential Gaseous Rpl~aaep From WBN Unit 2 Receptor Receptor Sector Distance 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

Completion and Operation of Watts Bar Nuclear Plant Unit 2 Table 3-19 Receptors from 2007 Actual Land Use Survey Results Used for Potential Gaseous Releases From WBN Unit 2 Receptor Receptor Sector Distance 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 updated data in the following table WBN Total Annual Gaseous Discharge Per Operatin!

(curieslyear/reactor) g Unit Chapter 3 Table 3-20.

Containment Auxiliary ITurbine Total per, 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 8.51 E+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-1 38 3.32E+00 3.42E+00 9.06E-01 7.65E+00 Ar-41 3.40E+01 0.OOE+00 0.OOE+00 3.40E+01 Br-84 6.OOE-05 5.01 E-02 4.81 E-04 5.06E-02 1-131 7.29E-03 1.39E-01 7.08E-03 1.53E-01 1-132 1.60E-03 6.56E-01 1.70E-02 6.75E-01 1-133 3.55E-03 4.35E-01 2.03E-02 4.59E-01 1-134 1.66E-03 1.06E+00 1.47E-02 1.08E+00 1-135 3.16E-03 8.10E-01 3.13E-02 8.44E-01 H-3 1.37E+02 0.OOE+00 0.OOE+00 1.37E+02 H-3 (TPC) 3.70E+02 0.OOE+00 0.OOE+00 3.70E+02 Cr-51 9.21 E-05 5.OOE-04 0.OOE+00 5.92E-04 Mn-54 5.30E-05 3.78E-04 0.OOE+00 4.31 E-04 Co-57 8.20E-06 O.OOE+00 O.OOE+00 8.20E-06 Co-58 2.50E-04 2.29E-02 0.OOE+00 2.32E-02 Co-60 2.61 E-05 8.71 E-03 0.OOE+00 8.74E-03 Fe-59 2.70E-05 5.OOE-05 O.OOE+00 7.70E-05 Sr-89 1.30E-04 2.85E-03 O.OOE+00 2.98E-03 Sr-90 5.22E-05 1.09E-03 0.OOE+00 1.14E-03 Zr-95 4.80E-08 1.OOE-03 0.OOE+00 1.OOE-03 Nb-95 1.80E-05 2.43E-03 0.OOE+00 2.45E-03 Ru`103 1.60E-05 6.1OE-05 0.OOE+00 7.70E-05 Ru-106 2.70E-08 7.50E-05 O.OOE+00 7.50E-05 Sb-1 25 O.OOE+00 6.09E-05 0.OOE+00 6.09E-05 Cs-1 34 2.53E-05 2.24E-03 O.OOE+00 2.27E-03 Cs-136 3.21 E-05 4.80E-05 0.OOE+00 8.01E-05 Cs-1 37 5.58E-05 3.42E-03 0.OOE+00 3.48E-03 Ba-140 2.30E-07 4.OOE-04 0.OOE+00 4.OOE-04 Ce-141 1.30E-05 2.64E-05 O.OOE+00 3.94E-05 C-14 2.80E+00 4.50E+00 0.OOE+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 date in place of the data in the prior table Chapter 3 Table 3-20 WBN Total annual Gaseous discharge Per Operating Unit (curies/yearlreactor)

Containment Aux!lat Turbine Nuclide Building Buildin*

Building Total 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 5.81 E+00 Kr-88 3.1OE+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 4.31 E+01 Xe-133 2.82E+03 6.70E+01 1.77E+01 2.90E+03 Xe-1 35m 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 O.OOE+00 0.OOE+00 3.40E+01 Br-84 8.16E-07 5.02E-02 4.81 E-04 5.07E-02 1-131 6.74E-03 1.39E-01 7.08E-03 1.53E-01 1-132 1.36E-04 6.56E-01 1.70E-02 6.73E-01 1-133 2.36E-03 4.35E-01 2.03E-02 4.57E-01 1-134 4.26E-05 1.06E+00 1.47E-02 1.07E+00 1-135 8.80E-04 8.10E-01 3.13E-02 8.42E-01 H-3 1.39E+02 0.OOE+00 O.OOE+00 1.39E+02 H-3 (TPC) 3.70E+02 O.OOE+00 0.OOE+00 3.70E+02 Cr-51 9.21 E-05 5.OOE-04 O.OOE+00 5.92E-04 Mn-54 5.30E-05 3.78E-04 O.OOE+00 4.31 E-04 Co-57 8.20E-06 0.OOE+00 O.OOE+00 8.20E-06 Co-58 2.50E-04 2.29E-02 O.OOE+00 2.32E-02 Co-60 2.61 E-05 8.71 E-03 O.OOE+00 8.74E-03 Fe-59 2.70E-05 5.OOE-05 0.OOE+00 7.70E-05 Sr-89 1.30E-04 2.85E-03 0.OOE+00 2.98E-03 Sr-90 5.22E-05 1.09E-03 0.OOE+00 1.14E-03 Zr-95 4.80E-08 1.OOE-03 O.OOE+00 1.OOE-03 Nb-95 1.80E-05 2.43E-03 0.OOE+00 2.45E-03 Ru-103 1.60E-05 6.1OE-05 0.00E+00 7.70E-05 Ru-1 06 2.70E-08 7.50E-05 0.OOE+00 7.50E-05 Sb-125 0.OOE+00 6.09E-05 O.OOE+00 6.09E-05 Cs-1 34 2.53E-05 2.24E-03 O.OOE+00 2.27E-03 Cs-136 3.21E-05 4.80E-05 O.OOE+00 8.01E-05 Cs-137 5.58E-05 3.42E-03 0.OOE+00 3.48E-03 Ba-140 2.30E-07 4.OOE-04 0.OOE+00 4.OOE-04 Ce-141 1.30E-05 2.64E-05 0.OOE+00 3.95E-05 C-14 2.80E+00 4.50E+00 0.OOE+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 i-N I

Table 3-21.

WBN Doses From Gaseous Effluent For Unit 2 Without Tritium Production for Year 2040 Effluent Pathway Guideline' Location Dose Noble Gases 7 Air dose 10 mrad Maximum Exposed0.801 mrad/year individual2 0.0mrdya 13 Air dose 20 mrad Maximum 2.710 mrad/year Total body 5 mrem Maximum Residence 3,4 0.571 mrem/year lodines/

Particulate Skin 10 mrem 15 mrem Maximum Residence 3' 4 Maximum Real Pathway 5 1.540 mrem/year 2.715 mrem/year Thyroid (critical organ)

Breakdown of Iodine/Particulate Doses (mrem/yr)

Cow Milk with Feeding Factor of 0.65 Inhalation Ground Contamination Submersion Beef Ingestion2 Total 2.44 0.174 0.0405 0.0603 0.00 2.7148 1Guidelines are defined in Appendix I to 10 CFR Part 50.

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

4Dose from air submersion.

5Maximum 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 Tahil 1-01 Use this updated date in place of the data in the prior table Gaseous Effluent for Unit 2 Without Tritium r 2040 Table 3-21 WBN Doses From Production for Yea Effluent Pathway Guideline*

Location Dose Noble Gases y Air dose 10 mrad Maximum Exposed 0.479 mrad/year Individual1 Air dose 20 mrad Maximum Exposed 1.62 mrad/year Individual1 Total body 5 mrem Maximum Residence 2'3 0.38 mrem/year Iodines/

Particulate Skin 10 mrem Maximum Residence 2' 3 1.02 mrem/year 1.70 mrem/year Thyroid (critical organ) 15 mrem Maximum Real Pathway4 Breakdown of Iodine/Particulate Doses (mrem/yr)

Total Vegetable Ingestion Inhalation Ground Contamination Submersion Beef Ingestion 5

Total 0.97 0.322 0.0499 0.0685 0.285 1.6954 Guidelines are defined in Appendix I to 10 CFR Part 50.

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

2Dose from air submersion.

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

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

5Maximum 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 Watts Bar Nuclear Plant List of Commitments

1. In the footnote added to Table 11.2-5 by Amendment 102, the term "F/H1 D" in the formulation of Column 5 and "Mobi"le" 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 x/Q values at WBN receptors, and that GELC adequately estimates x/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 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 lodines/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 I Watts Bar Nuclear Plant Calculation WBN EEB EDQ1090-99005 Extending Channel Operational Test Frequency for Radiation Monitors E4-1

A.

NPG CALCULATION COVERS HE ET/CCRIS UPDATE Page 1 REV 0 EDMS/RIMS NO.

EDMS TYPE:

EDMS ACCESSION NO (N/A for REV. 0)

B26000215402 calculations (nuclear)

Calc

Title:

Extending Channel Operational Test Frequency for Radiation Monitors CALC ID TYPE ORG PLANT BRANCH NUMBER CUR REV NEW REV REVISION CURRENT CN NUC WBN EEB EDQ1090-99005 000 001 APPLICABILITY C

Entire calc

[

NEW CN NUC Selected pages Q No CCRIS Changes E]

ACTION NEW DELETE El SUPERSEDE ]

CCRIS UPDATE ONLY E (For caic revision, REVISION RENAME El DUPLICATE El (Verifier Approval Signatures Not Required)

CCRIS been reviewed and no CCRIS changes I_

req uired)

UNITS 001 SYSTEMS 090 UNIDS: N/A DCN.EDC.N/A

[ APPLICABLE DESIGN DOCUMENT(S) FSAR 12.3.4, 11.4 CLASSIFICATION PER 292999I D

QUALITY SAFETY RELATED?

UNVERIFIED SPECIAL REQUIREMENTS DESIGN OUTPUT SAR/TS and/orISFSI RELATED?

(If yes, QR = yes)

ASSUMPTION AND/OR LIMITING CONDITIONS?

ATTACHMENT?

SAR/CoC AFFECTED Yes 0 No E]

Yes(K No E]

YesE] No M]

YesE]

No [I Yes[]

No M]

Yes[]

Nof PREPARER ID IPREPARER PHONE NO PREPARING ORG (BIANCH~ I VERIFICATION METHOD NEW METHOD OF ANALYSIS acwhaley 423-452-4446 EEB Design Review E] Yes

[] No PREPARER SIGNATURE DATE CHECKER SIGNATU DATE Aaron C. Whaley

-7I A

Tho Jerem A-.

-'7"a'-'

VERIFIER SIGNATUZE DATE

._APPRO SIGNATURE DATE Jeremy A. Thompsd/~

W/

Thom aeRR STATEMENT OF PROBLEM/ABSTRACT The Channel Operational Tests (COT) have been performed at a quarterly frequency for many monitors. History has shown that frequency is overly conservative. This calculation analyzes the historical data and recommends changes in the COT frequency.

It was determined that the COT frequency can be extended. For monitors that are required by Tech Spec or ODCM, a COT frequency of once every 3 quarters was recommended and once per fuel cycle for all others.

MICROFICHE/EFICHE Yes E] No[

FICHE NUMBER(S)

E]

LOAD INTO EDMS AND DESTROY Eg LOAD INTO EDMS AND RETURN CALCULATION TO CALCULATION LIBRARY. ADDRESS: EQB-1M-WBN

[]

LOAD INTO EDMS AND RETURN CALCULATION TO:

TVA 40532 [10-20 08]

Page 1 of 2 NEDP-2-1 [10-20-2008]

LEGIBILITY EVALUATED AND ACCEPTED FOR ISSUE.

SIGNAMURE AT

Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 NPG CALCULATION COVERSHEETICCRIS UPDATE Page 1a CALC ID I TYPE I ORG I PLANT BRANCH I

NUMBER I

REV ALTERNATE

(*ALCULATION IDENTIFICATION CN NUC WBN I

EEB ED Q1 090-99005 001 ALTERNATE CALCULATION IDENTIFICATION BLDG ROOM ELEV COORD/AZIM FIRM Prnt Report Les N/A N/A N/A N/A Bechtel CATEGORIES: NA KEY NOUNS (A-add, D-delete)

ACTION KEY NOUN A/D KEY NOUN (AID)__

CROSS-REFERENCES (A-add, C-change, D-delete)

ACTION XREF XREF XREF XREF XREF XREF (A/C/D)

CODE TYPE PLANT BRANCH NUMBER REV A

P IN WBN EEB EEB-TI-28 A

P PE WBN EEB PER 292999 A

P PE WBN EEB PER 00-001766-000 A

P DN WBN EEB DCN 51426 A

P DN WBN EEB DCN 50483 A

P DN WBN EEB EDC 50574 A

P DW WBN EEB 1-47W610-90-1 A

P DW WBN EEB 1-47W610-90-2 A

P DW WBN EEB 1-47W610-90-3 A

P DW WBN EEB 1-47W610-90-4 A

P DW WBN EEB 1-47W610-90-5 A

P GN WBN EEB P.S.4.M.4.1 CCRIS ONLY UPDATES:

Following are required only when making keyword/cross reference CCRIS updates and page 1 of form NEDP-2-1 is not included:

PREPARER SIGNATURE DATE CHECKER SIGNATURE DATE PREPARER PHONE NO.

EDMS ACCESSION NO.

TVA 40532 [10-2008]

Page 2 of 2 N EDP-2-1 [10-20-20081 This sheet added by R1.

I Revision 001 1 Preparedl Aaron C. Whaley I Date l2/10/11 I Checked Jeremy A. Thompson I Datej 2/10/11 JSheetj la I

Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 Page 2 B.

NPG CALCULATION RECORD OF REVISION CALCULATION IDENTIFIER: WBN-EEB-EDQ1090-99005 Title Extending Channel Operational Test Frequency for Radiation Monitors Revision DESCRIPTION OF REVISION No.

000 Initial issue. FSAR Review: Referenced sections of the FSAR & Technical Requirements Manual, and Tech Specs were reviewed for impact. There is no present impact. This calculation gives a basis for change.

This calculation contains 24 pages with 2 attachments for a total of 47 pages.

00.1 This revision resolves PER 292999. Revision 0 contained a condition that did not meet the acceptance criteria specified in the calculation. Zero failures were allowed among the 41 samples selected in order to meet 95/95 criteria. One failure was recorded among the samples. Inadequate justification was given for this one failure.

This revision revises the acceptance criteria by expanding the number of samples taken to the 77 (all monitors presented in this calculation were sampled). By expanding the number of samples to 77, the 95/95 criteria can be met with a maximum of 3 failures within the sample set. The 95/95 criteria was met under these circumstances, therefore the conclusion of this calculation to change the COT frequency is supported. This revision does not change the conclusion of Revision 0. Only the justification supporting the conclusion has been modified in this revision. In addition to these changes, the word "of was inserted into a sentence on page 6 in order to make the sentence grammatically correct. Revision levels of the references were updated where appropriate.

There are~no computations in this calculation.

Pages Added: 1 a, Attachment 3 pages 1-3 Pages Changed: 1-13, 21 Pages Deleted: None Total Pages in this Calculation: 51 Including Appendixes (pages) 0 and Attachments (pages) 26 WBNSAR Section(s) 12.3.4 & 11.4 were Reviewed By Aaron C. Whaley and are not impacted by the results of this calculation.

WBN Technical Specifications/Bases and Technical Requirements Manual/Bases Sections(s) 3.3.3. 3.3.6. 3.3.7. 3.3.8. 3.4.15

, Drawings N/A and Table(s) 3.3.3-1. 3.3.6-1, 3.3,7-1.3.3.8-1 have been Reviewed By Aaron C. Whaley and are not impacte sults of this calculation.

TVA 40709 [10-20081 Page 1 of 1 NEDP-2-2 [10-20-20081 This sheet replaced by R1.

Revision 001 1 Prepared I Aaron C. Whaley I Date 12/10/11 1 Checked Jeremy A. Thompson I Date 2/10/11 1Sheet 2

Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 Page 3 NPG CALCULATION TABLE OF CONTENTS Calculat ion Identifier WBN-EEB-EDQI 090-99005 1

Revision:

1001 TABLE OF CONTENTS SECTION TITLE PAGE A. NPG Calculation Cover Sheet/CCRIS Update.....................................................

B. NPG Calculation Record of Revision 2

C. NPG Calculation Table Of Contents.............................................................................................................

3 D. NPG Calculation Verification Form..............................................................................................................

4 P u rp o se....................................................................................................................................................................

5 R efe re n c e s...............................................................................................................................................................

5 Design Input Data....................................................................................................................................................

5 Assumptions...............................................................................................................................................

6 Special Requirements/Limiting Conditions.................................................................................... 6 Computations and Analyses..........................................................................................

6 Supporting Graphics................................................................................................................................................

8 Summary of Results...............................................................................................................................

............... 21 C o n c lu s io n s........................................................................................................................

I....................................

2 1 System Engineer eReview.......................................................................................................................................

21 Attachments I

TVA40710 [10-2008]

Page 1 of 1 NEDP-2-3 [10-20-2008]

This sheet replaced by R1; Revision001 IPrepared Aaron C. Whaley I Date 2/10/11 1 Checked Jeremy A. ThompsonI Date 12/10/11 1Sheet1 3 1

Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 Page 4 C.

NPG CALCULATION VERIFICATION FORM Calculation Identifier WBN-EEB-EDQ1 090-99005 Revision 001 Method of verification used:

1. Design Review

2. Alternate Calculation E]

Verifier Datel""40

3. Qualification Test El Comments: See Reference 10 (Attachment 3) Design Verification Report, Design Verification Checklist, 25402-3DP-G04G-00027, R3 WBN, Instrumentation and Control, Calculation ID: WBN-EEB-EDQ1090-99005 TVA 40533 [10-2008]

Page 1 of 1 NEDP-2-4 [10-20-2008]

This sheet replaced by RI.

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Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 Purpose The purpose of this calculation is to determine whether the present Channel Operational Test (COT) frequency for radiation monitors can be reduced. Decreasing the frequency reduces the maintenance induced wear and reduces the manpower costs to the site. The present frequency is once per quarter. The present engineering calculations for radiation monitors with NE SSD's predict acceptable operation of the radiation monitors for a fuel cycle (18 months). All other radiation monitors' performance characteristics are based on NE SSD's.

References

1. G29 P.S.4.M.4.1 (R9) External Surface Cleanness of Austenitic Stainless Steel Piping and Components

2.

EEB-TI-28 Setpoint Calibrations (R7)

3.

FSAR 12.3.4 (Amendment 0), 11.4 (Amendment 0), Living FSAR (1557 SOO PKG)

4.

Tech Spec 1.1, SR 3.3.3.3, Table 3.3.3-1, SR 3.3.6.4, Table 3.3.6.1, SR 3.3.7.2, Table 3.3.7-1, SR 3.3.8-2, Table 3.3.8-1, SR 3.4.15.2 (Amendment 83)

5.

Technical Requirements Manual 1.1 (Rev. 45)

6.

ODCM Table 2.1-1, Table 2.1-2, Table 3.1 Rev. 23

7.

System Engineer Review of additional data (T69 000125 497)

8.

Control Diagrams 1-47W610-90-1 R37, -2 R51, -3 R35, -4 R59, -5 R40

9.

Procedures: NPG-SPP-6.7, NPG-SPP-3.6, NPG-SPP-6.6 and NEDP 12

10. Design Verification Report, Design Verification Checklist, 25402-3DP-GO4G-00027, Revision 3, WBN, Instrumentation and Control, Calculation ID: WBN-EEB-EDQ1090-99005

11. PER 292999

12. PER 00-001766-000

13. DCN 51426-A & DCN 50483-A

14. EDC 50574-A Design Input Data

1. COT Definitions:

- FSAR 12.3.4.1.3 & 12.3.4.2.6: The channel operational test (COT) is the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the operability of required alarm, interlock, display and trip functions.

The COT includes adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

- Tech Spec 1.1: A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

- Technical Requirements Manual 1.1: A COT shall be the injection of a simulated or actual signal into the channel as close to the sensor as practicable to verify the OPERABILITY of required alarm, interlock, display, and trip functions. The COT shall include adjustments, as necessary, of the required alarm, interlock, and trip setpoints so that the setpoints are within the required range and accuracy.

- ODCM Table 2:1-1 note (1) (typical example - each monitor may have unique functions that are required to be checked):

The CHANNEL OPERATIONAL TEST shall demonstrate that automatic isolation of this pathway occurs if the instrument indicates measured levels above the alarm/trip setpoint. The CHANNEL OPERATIONAL TEST also demonstrates control room annunciation occurs if any of the following conditions exist:

1. Instrument indicates measured levels above the alarm setpoint, or This sheet replaced by RI.

I Date 12/10/11 1 Checked IJeremy A. Thompson I Datel 2/10/11 ISheet 1 5 I Revision 1001 1 Prepared I Aaron C Whaley in10 IP~~~d ~n CWA~

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Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005

2. Circuit failure, or

3. Indication of downscale failure, or

4. Instrumentation controls not set in operate mode, or

5. Loss of sample flow.

2. Calculation Basis: Radiation monitors with NE-SSD's (Safety related, Tech Spec, and compliance that accuracy is a factor) have a calculational time basis of I fuel cycle (18 months + 25%). This was verified by a review of all the radiation monitor NE-SSDs. All other radiation monitors have a plant SSD and are based upon those with NE-SSDs. Therefore, all monitors have an engineering basis (directly via an NE-SSD or indirectly via similarity) of acceptable operation for 1 fuel cycle.

3. COT / Calibration / Maintenance History: EMPAC is a database that records the maintenance history on plant equipment.

This database was reviewed for the maintenance history on the COT performances. The list of radiation monitors in Table 3 was used as the basis. This table represents the radiation monitors that had data that could be retrieved from EMPAC at the time of the data study. EMPAC is a living database and changes are made with time. Until recentlythere was no set method of recording the work performed. Therefore, a set of the actual work performed sheets were retrieved in hardcopy to verify and interpret the information in EM PAC when the datarecorded was ambiguous. Only the latest available COT at the time of the study was used because this best represents the present instrument performance.

4. G29 - P.S.4.M.4.1 (R9) specifies a sampling methodology that produces a 95195 probability/confidence for an attribute in a population. This method was utilized to determine radiation monitor performance during COTs.

Assumptions None Special Requirements/Limiting Conditions None Computations and Analyses This calculation extends the COT frequency for the maintained monitors. The population size of monitors was determined to be 77 (See Table 3). Data was taken on the radiation monitors from EMPAC. The maintenance history recorded in EMPAC includes work orders when the radiation monitors were first brought into service. Therefore, there are work orders that show mis-wiring, missing components, etc. These work orders are not indicative of the operation of the present system. Therefore, the methodology chosen for this analysis used the latest available COT that data can be accurately understood. This means that if the work performed field in EMPAC is unambiguous that it was used directly. Hardcopy data was retrieved from RIMS when the EMPAC data was insufficient.

According to G29, 40 monitors must be sampled for a population of 77 monitors with zero allowed failures, 57 monitors must be sampled with 1 allowed failure, 70 monitors must be sampled with 2 allowed failures, and all 77 monitors must be sampled with 3 allowed failures in order to state with a 95/95 probability and confidence that the attribute of interest is indicative of the population. The attribute of interest for this statistical test is the "as-left" band. If the monitors are operating within this band for the COT test, then recalibration is not This sheet replaced by RI.

I Revision 001 I Preparedl Aaron C. Whalev I Date i 2/10/11 1 Checked Jeremv A. Thompson I Datel 2/10/11 ISheet i 6 I

Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 required. Since the engineering calculations prove that as long as a device is left within the "as-left" band, then they can operate for 18 months.

The criteria for using the "as-left" band as a successful performance is conservative for a COT since by definition the instrument would be correctly operating if it were found within the "as-found" band. EEB TI-28 defines the "as-found" band as the region where the instrument is performing as predicted. During a calibration and not a COT, if the instrument were found within the "as-found" band, then it would be recalibrated to within the "as-left" band. However, based upon a review of the EMPAC data it has been WBN's practice to recalibrate as part of a COT when the instrument is outside the "as-left" but within the "as-found".

The monitors were numbered 1 to 78. All monitors were sampled. The first COT for each of these sampled monitors that the performance could be determined either directly from the EMPAC "work performed" field orfrom the hardcopy RIMS data was chosen as the sample for each monitor. This data is the best example of how the monitors are presently maintained and operating.

The attribute evaluated is the accuracy or performance of each module checked. This includes as appropriate for each monitor

- Ratemeter Indications

- Recorder

- Flow Switches

- Alarm Setpoints

- Ratemeter Analog Outputs

- Flow Controllers Not all functions above are appropriate for all monitors. However, the COT tests each monitor for all calibratable features.

Table 2 tabulates the results of the above sample test. Out of the 77 samples, there were 3 failures recorded. Per Reference 1, the 95/95 criteria is supported even with 3 failures. With a 95/95 probability and confidence, the radiation monitors are operating within the setting tolerance assumed in the accuracy calculations. Therefore, the full 18 month uncertainty is still available.

This sheet replaced by R1.

I Revision 001 1Preparedl Aaron C. Whaley Date 2/10/11 1 Checked Jeremy A. Thompson I Date 2/10/11 ISheet 7

Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 Supporting Graphics Since this revision utilized all 77 monitors for observation, the monitors did not need to be randomly selected. Table I shows the monitors used in this calculation. Radiation monitor 2-RM-90-402 was listed originally in Revision 0, but has been replaced by 1-RM-90-112B (#48). This was done because no work order information could be found for 2-RM-90-402.

Table I Number Monitor Number Monitor Number Monitor 1

2 3

4 5

6 7

8 9

,10 11 12 13 14 15 16 17 18 19 20 21 22 23 24 25 26 0-RM-90-12 0-RM-90-13 1-RM-90-14 0-RM-90-15 0-RM-90-16 0-RM-90-17 I-RM-90-59 1-RM-90-60 1 -RM-90-61 1-RM-90-62 0-RM-90-101A 1 -RM-90-272 1 -RM-90-273-A I-RM-90-274-B 1 -RM-90-275 1 -RM-90-276 1-RM-90-277 1 -RM-90-278 1 -RM-90-280 1-RM-90-290 1 -RM-90-291 I-RM-90-292 1 -RM-90-293 I-RM-90-1 2-RM-90-1 I-RM-90-2 27 28 29 30 31 32 33 34 35 36 37 38 39 40 41 42 43 44 45 46 47 48 49 50 51 52 0-RM-90-3 0-RM-90-4 0-RM-90-5 I-RM-90-6 2-RM-90-6 1 -RM-90-7 2-RM-90-7 1 -RM-90-8 2-RM-90-8 0-RM-90-9 1 -RM-90-1 0 2-RM-90-1 0 0-RM-90-11 1-RM-90-119 1-RM-90-120 1 -RM-90-121 0-RM-90-122 0-RM-90-123 0-RM-90-103-B 1-RM-90-106 0-RM-90-102A 1-RM-90-112B 1-RM-90-112C 0-RM-90-118 1-RM-90-123 2-RM-90-123 53 54 55 56 57 58 59 60 61 62 63 64 65 66 67 68 69 70 71 72 73 74 75 76 77 78 0-RM-90-125-A Intentionally Blank 0-RM-90-126-B 0-RM-90-132 2-RM-90-400 0-RM-90-135 0-RM-90-138 0-RM-90-133-A 0-RM-90-134-B 0-RM-90-140-A 0-RM-90-141-B 0-RM-90-205-A 0-RM-90-206-B 0-RM-90-212 0-RM-90-225 0-RM-90-230 0-RM-90-231 1-RM-90-271-A 1-RM-90-404 1-RM-90-421 1 -RM-90-422 1-RM-90-423 1 -RM-90-424 1-RM-90-130-A 1 -RM-90-131-A 1-RM-90-400 This sheet replaced by R1.

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This sheet replaced by R1.

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I Date 12/10/11 I Checked Jeremy A. Thompson I Datel 2/10/11 ISheet 1 9 I

Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 This page left intentionally blank by R1 to maintain the same page numbers as RO. I This sheet replaced by R1.

I Revision 1001 I PreparedI Aaron C. Whaley I Date 1 2/10/11 I Checked Jeremy A. Thompson I Datel 2/10/11 1Sheet 1 10 1

Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 The following table lists the data obtained from EMPAC or RIMS hardcopy for the sampled monitors.

Table 2: Work Performed Number Monitor Date WO#

Work Performed Pass/Fail Reel/Frame 1

)-RM-90-12 03/29/1999 99-000322-000 All devices Left As Found (PM As Found condition 100)

Pass 2

0-RM-90-13 11/30/1998 98-012479-000 All devices Left As Found Pass Low flow switch and flow controller outside As-Left.

3 1-RM-90-14 12/07/1998 98-012935-000 All other devices LeftAs Found.

Fail E05791/1744 4

0-RM-90-15 11/05/1998 98-012004-000 All devices Left As Found Pass 8

1-RM-90-60 02/05/1999 98-014050-000 All devices Left As Found Pass 14 1-RM-90-274-B 01/22/1999 98-014058-000 All devices Left As Found Pass 16 1-RM-90-276 01/11/1999 98-014078-000 All devices Left As Found Pass 18 1-RM-90-278 01/19/1999 98-014067-000 All devices Left As Found Pass 20-1-RM-90-290 12/07/1998 98-012952-000 All devices Left As Found Pass 22 1-RM-90-292 02/16/1999 98-014045-000 All devices Left As Found Pass 24 1-RM-90-1 02/28/1999 98-015025-000 All devices Left As Found Pass 25 2-RM-90-1 02/28/1999 98-015026-000 All devices Left As Found Pass 26 1-RM-90-2 12/18/1998 98-012918-000 All devices Left As Found Pass 27 0-RM-90-3 12/23/1998 98-012949-000 All devices Left As Found Pass E05793/1024 29 0-RM-90-5 02/03/1999 98-016679-000 All devices Left As Found Pass 30 1-RM-90-6 01/11/1999 98-015028-000 All devices Left As Found Pass 31 2-RM-90-6 01/28/1999 98-016278-000 All devices Left As Found Pass 32 1-RM-90-7 02/10/1999 98-014065-000 All devices Left As Found Pass 33 2-RM-90-7 01/29/1999 98-014072-000 All devices Left As Found Pass 34 1-RM-90-8 02/11/1999 98-014018-000 All devices Left As Found Pass 35 2-RM-90-8 02/02/1999 98-014073-000 All devices Left As Found Pass 36 0-RM-90-9 02/05/1999 98-014064-000 All devices Left As Found Pass 37 1-RM-90-10 12/09/1998 98-013013-000 All devices Left As Found Pass 38 2-RM-90-10 04/06/1999 99-000208-000 All devices Left As Found (PM As Found condition 100)

Pass 39 0-RM-90-11 12/04/1998 98-012925-000 All devices Left As Found Pass 44 0-RM-90-123 02/01/1999 98-014079-000 All devices Left As Found Pass 46 1-RM-90-106A-A 01/11/1999 98-015535-000 All devices Left As Found Pass 49 1-RM-90-112C-B 01/29/1999 98-016539-000 All devices Left As Found Pass 51 1-RM-90-123 02/08/1999 98-015090-000 All devices Left As Found Pass 53 O-RM-90-125-A 03/06/1999 99-000338-000 All devices Left As Found Pass 59 0-RM-90-138 12/04/1998 98-012481-000 All devices Left As Found Pass 61 0-RM-90-134-B 02/16/1999 98-016593-000 All devices Left As Found Pass 62 0-RM-90-140-A 12/09/1998 98-013050-000 All devices Left As Found Pass 63 0-RM-90-141-B 02/18/1999 98-016898-000 All devices Left As Found Pass 64 0-RM-90-205-A 12/04/1998 98-013256-000 All devices Left As Found Pass 65 O-RM-90-206-B 02/11/1999 98-016654-000 All devices Left As Found Pass 68 0-RM-90-230 02/01/1999 98-014046-000 All devices Left As Found Pass 69 0-RM-90-231 02/08/1999 98-014049-000 All devices Left As Found Pass All devices Left As Found. Recorder 1-RR-090-0268 has 72 1-RM-90-421 12/09/1998 98-012924-000 WO 98-015690-000 against it for repair.

Pass 73 1-RM-90-422 02/08/1999 98-014066-000 All devices Left As Found Pass 74 11-RM-90-423 12/23/1998 98-012948-000 All devices Left As Found*

Pass E05793/0024 This sheet replaced by R1.

Revision 1001 1 Prepared I Aaron C. Whaley I Date 1 2/10/11 1 Checked j Jeremy A. Thompson I Date J 2/10/11 1 Sheet 11

Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 Table 2: Work Performed (continued)

Number Monitor Date WO#

Work Performed Pass/Fail Reel/Frame 5

0-RM-90-16 08/02/1999 99-008145-000 All devices Left As Found Pass H/L Flow Switches & L Flow Controller outside As-Left.

6 0-RM-90-17 02/04/1999 98-014054-000 All other devices Left As Found.

Fail E05797/2843 7

1-RM-90-59 12/17/1998 98-012959-000 All devices Left As Found Pass E05793/3851 9

1-RM-90-61 06/29/1999 99-006348-000 All devices Left As Found Pass E05863/1402 Lower Discriminator outside of As-Found limit.

10 1-RM-90-62 06/02/1999 99-004564-000 All other devices Left As Found.

Fail E05860/1843 11 0-RM-90-101A 06/28/1999 99-006364-000 All devices Left As Found Pass E05863/2513 12 1-RM-90-272 06/08/1999 99-004566-000 All devices Left As Found Pass E05855/0271 13 1-RM-90-273-A 01/07/1999 98-014183-000 All devices Left As Found Pass E05796/2898 15 1-RM-90-275 06/21/1999 99-004537-000 All devices Left As Found Pass E05860/2843 17 1-RM-90-277 06/08/1999 99-005053-000 All devices Left As Found Pass E05863/1713 19 1-RM-90-280 02/11/1999 98-014923-000 All devices Left AsFound Pass E05822/3198 21 1-RM-90-291 02/08/1999 98-014059-000 All devices Left As Found Pass E05797/2458 23 1-RM-90-293 02/09/1999 98-014055-000 All devices Left As Found Pass E05797/2443 28 O-RM-90-4 01/28/1999 98-014186-000 All devices Left As Found Pass E05796/2327 40 1-RM-90-119 05/03/1999 99-003256-000 All devices Left As Found Pass 41 1-RM-90-120 03/03/1999 99-000075-000 All devices Left As Found Pass 42 1-RM-90-121 05/06/1999 99-003298-000 All devices Left As Found Pass 43 0-RM-90-122 05/22/1999 99-004106-000 All devices Left As Found Pass 45 0-RM-90-103-B 06/10/1999 99-008316-000 All devices Left As Found Pass 47 0-RM-90-102A 06/18/1999 99-008317-000 All devices Left As Found Pass 48 1-RM-90-112B 06/08/2000 99-017717-000 All devices Left As Found Pass 50 0-RM-90-118 05/18/1999 99-004098-000 All devices Left As Found Pass 52 2-RM-90-123 05/22/1999 99-003571-000 All devices Left As Found Pass E05860/3620 55 0-RM-90-126-B 05/12/1999 99-004037-000 All devices Left As Found Pass 56 0-RM-90-132 01/25/1999 98-014068-000 All devices Left As Found Pass E05797/1054 57 2-RM-90-400 03/04/1999 99-000686-000 All devices Left As Found Pass 58 0-RM-90-135 05/05/1999 98-012902-000 All devices Left As Found Pass E05859/2252 60 0-RM-90-133-A 05/22/1999 99-004102-000 All devices Left As Found Pass 66 0-RM-90-212 08/05/1999 99-007998-000 All devices Left As Found Pass 67 0-RM-90-225 05/03/1999 99-003240-000 All devices Left As Found Pass 70 1-RM-90-271-A 06/10/1999 99-004555-000 All devices Left As Found Pass E05863/1259 71 1-RM-90-404 04/14/2005 03-012634-000 All devices Left As Found Pass 75 1-RM-90-424 06/14/1999 99-005315-000 All devices Left As Found Pass E05858/0001 76 1-RM-90-130-A 03/25/1999 99-001014-000 All devices Left As Found Pass 77 1-RM-90-131-A 03/10/1999 99-000718-000 All devices Left As Found Pass 78 1-RM-90-400 06/22/1999 99-006131-000 All devices Left As Found Pass This sheet replaced by R1.

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Extending Channel Operational Test Frequency for Radiation Monitors WBN-EEB-EDQ1090-99005 Rev. 0 Table 3 Disposition Listing from EMPAC and Compilation of Programs Unid FSAR FSAR not Other Tech Spec ODCM Sampled Maintained Required Required for Unit 1 WBN-0-LPR -090-0003 Y

Y Y

WBN-0-LPR -090-0004 Y

Y WBN-0-LPR -090-0005 Y

Y Y

WBN-0-LPR -090-0009 Y

Y Y

WBN-0-LPR -090-0011 Y

Y Y

WBN-0-LPR -090-0012 Y

-YY WBN-0-LPR -090-0013 Y

Y Y

WBN-0-LPR -090-0015 Y

Y Y

WBN-0-LPR -090-0016 Y

Y WBN-0-LPR -090-0017 Y

Y WBN-0-LPR -090-0101A Y

Y WBN-0-LPR -090-0101 B counted with 101A Y

Y WBN-0-LPR -090-0101 C counted with 101A Y(COT NA)

Y WBN-0-LPR -090-0102-A Y

Y Y

WBN-0-LPR -090-0103-B Y

Y I

Y WBN-0-LPR -090-0105 No SSD I

Prepared: Richard Brehm Checked: Edward Bradley Page 14,

Extending Channel Operational Test Frequency for Radiation Monitors WBN-EEB-EDQ]090-99005 Rev. 0 Listing from EMPAC and Compilation of Programs Unid FSAR FSAR not Other Tech Spec ODCM Sampled Maintained Required Required for Unit I WBN-0-LPR -090-0118 Y

Y Y

WBN-0-LPR -090-0122 Y

Y Y

WBN-0-LPR -090-0123 Y

Y Y

WBN-0-LPR -090-0125-A Y

Y Y

Y WBN-0-LPR -090-0126-B Y

Y Y

WBN-0-LPR -090-0132A Y

WBN-0-LPR -090-0132B Counted with 132 A Y

WBN-0-LPR -090-0132C Counted with 132 A WBN-0-LPR -090-0133-A Y

Y Y

WBN-0-LPR -090-0134-B Y

Y Y

Y WBN-0-LPR -090-0135 Y

Y WBN-0-LPR -090-0138 Y

Y Y

WBN-0-LPR -090-0140-A Y

Y Y

Y WBN-0-LPR -090-0141-B Y

Y Y

Y WBN-0-LPR -090-0205-A Y

Y Y

WBN-0-LPR -090-0206-B Y

Y Y

WBN-0-LPR -090-0211 No SSD WBN-0-LPR -090-0212 Y

Y Y

WBN-0-LPR -090-0217A Portable - No SSD WBN-0-LPR -090-0217B Portable - No SSD WBN-0-LPR -090-0217E Portable - No SSD WBN-0-LPR -090-0218A Portable - No SSD WBN-0-LPR -090-0218B Portable - No SSD WBN-0-LPR -090-0218D Portable - No SSD WBN-0-LPR -090-0218E Portable - No SSD Prepared: Richard Brehm Checked: Edward Bradley Page 15

Extending Channel Operational Test Frequency for Radiation Monitors WBN.EEB-EDQ 1090-99005 Rev. 0 Listing from EMPAC and Compilation of Programs Unid FSAR FSAR not Other Tech Spec ODCM Sampled Maintained Required Required for Unit I WBN-0-LPR -090-0219A Portable - No SSD WBN-0-LPR -090-0219B Portable - No SSD WBN-0-LPR -090-0219D Portable - No SSD WBN-0-LPR -090-0219E Portable - No SSD WBN-0-LPR -090-0225 Y

Y Y

WBN-0-LPR -090-0230 Y

Y Y

WBN-0-LPR -090-0231 Y

Y Y

WBN-0-LPR -090-0235 No SSD WBN-0-LPR -090-0236 No SSD WBN-0-LPR -999-0063 TLD - No SSD WBN-1-LPR -090-0001 Y

Y Y

WBN-1-LPR -090-0002 Y

Y Y

WBN-1-LPR -090-0006 Y

Y Y

WBN-1-LPR -090-0007 Y

Y Y

WBN-1-LPR -090-0008 Y

Y Y

WBN-1-LPR -090-0010 Y

Y Y

WBN-1-LPR -090-0014 Y

Y Y

WBN-1-LPR -090-0059 Y

Y WBN-1-LPR -090-0060 Y

Y Y

WBN-1-LPR -090-0061 Y

Y WBN-1 -LPR -090-0062 Y

Y WBN-1 -LPR -090-0.106A-A Y

Y Notes Only Y

Y WBN-1 -LPR -090-0106B-A Counted-with 106A Y

Notes Only Y

WBN-1 -LPR -090-0106C-A Counted with 106A Notes Only Y

WBN-1-LPR -090-0112A-B Y

Y Notes Only Y

Prepared: Richard Brehm Checked: Edward Bradley Page 16

Extending Channel Operational Test Frequency for Radiation Monitors WBN-EEB-EDQ 1090-99005 Rev. 0 Listing from EMPAC and Compilation of Programs Unid FSAR FSAR not Other Tech Spec ODCM Sampled Maintained Required Required for Unit I WBN-1-LPR -090-0112B-B Counted with 112A Y

Notes Only Y

WBN-1-LPR -090-0112C-B Counted with 112A Notes Only Y

Y WBN-1 -LPR -090-0119 Y

Y Y

WBN-1-LFR -090-0120 Y

Y Y

WBN-1-LPR -090-0121 Y

Y Y

WBN-1 -LPR -090-0123 Y

Y Y

WBN-1 -LPR -090-0124 No SSD WBN-1-LPR -090-0130-A Y

Y Y

Y WBN-1 -LPR -090-0131-B Y

Y Y

Y WBN-1 -LPR -090-0170 NOT No SSD WBN-1 -LPR -090-0210 No SSD WBN 1-LPR -090-0262 No SSD WBN-1-LPR -090-0271-A Y

Y Y

WBN-1 -LPR -090-0272-B Y

Y Y

WBN-1-LPR -090-0273-A Y

Y Y

WBN-1 -LPR -090-0274-B Y

Y Y

Y WBN-1-LPR -090-0275 Y

Y WBN-1 -LPR -090-0276 Y

Y Y

WBN-1 -LPR -090-0277 Y

Y WBN-1-LPR -090-0278 Y

Y Y

WBN-1-LPR -090-0280 Y

Y WBN 1-LPR -090-0290 Y

Y Y

WBN 1-LPR -090-0291 Y

Y WBN I-LPR -090-0292 Y

Y Y

WBN-1-LPR -090-0293 Y

Y Prepared: Richard Brehm Checked: Edward Bradley Page 17

Extending Channel Operational Test Frequency for Radiation Monitors WBN-EEB-EDQ 1090-99005 Rev. 0 Listing from EMPAC and Compilation of Programs Unid FSAR FSAR not Other Tech Spec ODCM Sampled Maintained Required Required for Unit I WBN-1-LPR -090-0400 Y

Y Y

WBN-1-LPR -090-0404 Y

Y WBN-1-LPR -090-0405 Same as 404 WBN-1-LPR -090-0421 Y

Y Y

WBN-1-LPR -090-0422 Y

Y Y

WBN-1-LPR -090-0423 Y

Y Y

WBN-1-LPR -090-0424 Y

Y WBN-1-LPR -090-0450 Same as 404 WBN-1-LPR -090-123 /A Same as 123 WBN-1-LPR -999-0099 TLD - No SSD WBN-1-LPR -999-0100A TLD - No SSD WBN-1-LPR -999-0100B TLD - No SSD WBN-1-LPR -999-0100C TLD - No SSD WBN-1-LPR -999-0104 TLD - No SSD WBN-1-LPR -999-0403 TLD - No SSD WBN-2-LPR -090-0001 Y

Y Y

WBN-2-LPR -090-0002 NOT No SSD WBN-2-LPR -090-0006 Y

Y Y

WBN-2-LPR -090-0007 Y

Y Y

WBN-2-LPR -090-0008 Y

Y Y

WBN-2-LPR -090-0010 Y

Y Y

WBN-2-LPR -090-0014 No SSD WBN-2-LPR -090-0059 NOT No SSD WBN-2-LPR -090-0060 NOT No SSD LWBN-2-LPR -090-0061 NOT No SSD Prepared: Richard Brehm Checked: Edward Bradley Page 18

Extending Channel Operational Test Frequency for Radiation Monitors WBN-EEB-EDQ 1090-99005 Rev. 0 Listing from EMPAC and Compilation of Programs Unid FSAR FSAR not Other Tech Spec ODCM Sampled Maintained

-Required Required for Unit 1 WBN-2-LPR -090-0062 NOT No SSD WBN-2-LPR -090-0099 NOT No SSD WBN-2-LPR -090-0100A No SSD WBN-2-LPR -090-0100B No SSD WBN-2-LPR -090-0100C No SSD WBN-2-LPR -090-0104 No SSD WBN-2-LPR -090-0106A-A NOT No SSD Notes Only WBN-2-LPR -090-0106B-A No SSD Notes Only WBN-2-LPR -090-0106C-A No SSD Notes Only WBN-2-LPR -090-0112A-B NOT No SSD Notes Only_

WBN-2-LPR -090-0112B-B No SSD Notes Only WBN-2-LPR -090-0112C-B No SSD Notes Only WBN-2-LPR -090-0119 NOT No SSD Y

WBN-2-LPR -090-0120 NOT No SSD Y

WBN-2-LPR -090-0121 NOT No SSD Y

WBN-2-LPR -090-0123 Y

Y WBN-2-LPR -090-0124 NOT No SSD WBN-2-LPR -090-0130-A NOT No SSD Y

WBN-2-LPR -090-0131-B NOT No SSD Y

WBN-2-LPR -090-0170 NOT No SSD WBN-2-LPR -090-0210 No SSD WBN-2-LPR -090-0260-A No SSD WBN-2-LPR -090-0261-B No SSD WBN-2-LPR -090-0262 No SSD WBN-2-LPR -090-0271-A NOT No SSD Prepared: Richard Brehm Checked: Edward Bradley Page 19

Extending Channel Operational Test Frequency for Radiation Monitors WBN-EEB-EDQ1090-99005 Rev. 0 Listing from EMPAC and Compilation of Programs Unid FSAR FSAR not Other Tech Spec ODCM Sampled Maintained Required Required for Unit I WBN-2-LPR -090-0272-B NOT No SSD WBN-2-LPR -090-0273-A NOT No SSD WBN-2-LPR -090-0274-B NOT No SSD WBN-2-LPR -090-0275 NOT No SSD WBN-2-LPR -090-0276 NOT No SSD WBN-2-LPR -090-0277 NOT No SSD WBN-2-LPR -090-0278 NOT No SSD WBN-2-LPR -090-0280 NOT No SSD WBN-2-LPR -090-0290 NOT No SSD WBN-2-LPR -090-0291 NOT No SSD.

WBN-2-LPR -090-0292 NOT No SSD WBN-2-LPR -090-0293 NOT No SSD WBN-2-LPR -090-0400 Y

No SSD Y

Y WBN-2-LPR -090-0402 Y

No SSD Y

Y WBN-2-LPR -090-0405 No SSD WBN-2-LPR -090-0421 NOT No SSD WBN-2-LPR -090-0422 NOT No SSD WBN-2-LPR -090-0423 NOT No SSD WBN-2-LPR -090-0424 NOT No SSD WBN-2-LPR -090-0450 No SSD WBN-2-LPR -090-123 /A Same as 123 WBN-2-LPR -999-0403 TLD-No SSD 77 Prepared: Richard Brehm Checked: Edward Bradley Page 20

Extending Channel Operational Test Frequency for Radiation Monitors BRANCH/PROJECT IDENTIFIER: WBN-EEB-EDQ1090-99005 Summary of Results The COT data was statistically sampled to determine the performance of the radiation monitors. The analysis has shown that the 95/95 level of probability and confidence has been met. Out of the 77 monitors observed, 3 failures were reported. The number of failures allowed to meet the 95/95 criteriafor a sample size of 77 is three, therefore the criteria is met. The three monitors that had failures were 1-RM-90-14, 0-RM-90-17, and 1-RM-90-62. These were Continuous Air Monitors (CAMs), and per DCN 51426 and DCN 50483, all three of these monitors were later removed from service for the following reasons:

-the monitors are not safety related

-they do not perform a primary safety function

-they are not required to mitigate a design basis accident

-they have not performed reliably (particularly the moving filter assembly)

These monitors were used for personnel protection and have been replaced by portable CAMs. Therefore, any calibration problems experienced by these monitors are no longer a concern.

Conclusions The analysis of the COT data has shown that the radiation monitors at WBN are operating within the "as-left" band at the current test frequency. The three observed failures out of the 77 samples is acceptable to meet 95/95 probability and confidence. In addition, the three monitors that experienced the failures have since been deleted from service. The calculational basis of 18 months is still completely valid starting at this time interval. Therefore, it is reasonable to extend the COT testing.

The monitors required by Tech Spec and ODCM can conservatively move the COT to 9 months. Keep in mind that this is 1/2 the calculational basis.

All other monitors should move the COT to the calculational basis of 18 months with the exception of flow switches on raw water systems and non VX-252 local indicators. PER 00-001766-000 initiated corrective actions for these devices. The calibration tolerances for the local indicators were relaxed per EDC 50574, and the calibration frequency of the ERCW flow switches was changed to 6 months. This PER has been closed.

Alternate Review (Ref. 7)

TVA's calibration program has feedback mechanisms built into the procedures to find instruments outside the norm and to fine tune engineering basis data (SPP 6.7 section 3.2.c.2, SPP 3.6, SPP 6.6 and NEDP 12). As part of this process the Radiation Monitor system engineer evaluated data on a device specific basis over the last 2+ years.

The review of reference 7 is summarized below.

The system engineer determined whether the instrument was found outside the "as-left" (OAL), outside the "as-found" (OAF),

inoperative (INOP), or within the "as-left" (no entry). Remember that the instrument is working as predicted if it is found within the "as-found" tolerance. Similar devices/features were combined, e.g. the analog ratemeters.

This calculation's review further categorized the data by instrument type and sub group. For instance, recorders were broken down by Manufacturer and Model and the liquid flow switches were grouped by calibration interval. These groupings are attached.

Ratemeter Modules:

There are two basic types of Ratemeter modules. One that measures pulses and one that measures current. There are similarities between these two types such as the indicator on the front and the bistable circuits.

This sheet replaced by R1.

I Revision 1001 I PreparedI Aaron C. Whalev I Date I 2/10/11 I Checked IJeremy A. Thompson I Datel 2/10/11 iSheet I 21 1

Extending Channel Operational TestFrequency for Radiation Monitors WBN-EEB-EDQ1090-99005 Rev. 0 Common Items

1. Indicator: Twenty-four (24) monitors were sampled. Two-hundred-forty (240) COTs were evaluated. There were no indicators found outside the "as-found" criteria. There were eight (8) instances where an indicator point was found outside the "as-left". This is a 100%

success rate.

2. Alert Bistable: Twenty-four (24) monitors were sampled. Two-hundred-forty (240) COTs were evaluated. There were no bistables found outside the "as-found" criteria. There was one (1) instance where a bistable was found outside the "as-left". This is a 100% success rate.

3. Hi Rad Bistable: Twenty-four (24) monitors were sampled. Two-hundred-forty (240) COTs were evaluated. There were no bistables found outside either the "as-found" or the "as-left"criteria. This is a 100% success rate.

Unique Items:

1. High Voltage: Eleven (11) monitor were sampled, One hundred ten, (110), COTs were evaluated. There were no instances where the High Voltage was outside either the "as-found" or the "as-left". This is a 100% success rate.

2. Low Discriminator: Eleven (11) monitors were sampled. One hundred ten (110) COTs were evaluated. There were four (4) instances where the discriminator was outside the "as-found" and two (2) instances outside the "as-left". The entire out of tolerance instances was on two (2) monitors. There is no engineering basis for this value. The system engineer has determined the setting. The discriminator sets the signal to noise ratio for the monitor. The pulse height of the radiation received at the ratemeter is a function of the gain of the detector tube, the gain of the preamp, and the attenuation of the cable run. The value has been taken as a rule of thumb and failures are not indicative of unacceptable operation of the monitor for this study Proper operation can only be determined by using a source on the detector. The system engineer determined that the 101 C monitor is not performing as expected and has initiated a work order to replace this module. Other than the iodine monitors there were no failures of the discriminator settings.

3. Upper Discriminator: All four (4) monitors that use this feature were evaluated. Forty-three (43) COTs were evaluated. There were seven (7) instances where the discriminator was outside "as-found" and one (1) instance outside "as-left". Six (6) of the OAF values were on one instrument. This function is not an engineered value. The value has been taken as a rule of thumb and failures are not indicative of unacceptable operation of the monitor for this study. Electronically, there is no difference in the upper and lower setpoints. As stated previously, proper operation can only be determined by using a source on the detector. The system engineer determined that the 101 C monitor is not performing as expected and has initiated a work request (WR C340313) to replace this module. Other than the iodine monitors there were no failures of the discriminator settings.

LOCAL INDICATOR:

Fifty-eight (58) COTs, eight (8) instruments, one (1) COT inop, three (3) COT OAF, five (5) instruments with no failures. There is nothing significantly different in these instruments from the MCR or the VX 252 indicators. PER 00-001766-00 was written to resolve this problem.

Prepared: Richard Brehm Checked: Edward Bradley Page 22

Extending Channel Operational Test Frequency for Radiation Monitors WBN-EEB-EDQ1090-99005 Rev. 0 DIGITAL RADIATION MONITOR:

1. Monitor 52 COTs, 6 instruments. A database value was found incorrect. This is not a calibration error.

2. Flow Switch 19 COTs, 2 Instruments, 1 COT found outside the "as-found", all others were within the "as-left". This is a small data set but still shows a 95% success rate.

COMPUTER:

Eighteen (18) computer points were sampled. One-hundred-eighty-one (181) COTs were evaluated. One (1) point was found outside the "as-found" during a COT and the next point in the same COT was outside the "as-left". All others COTs were within the "as-left". This is a 99.4% success rate. The one failure is insignificant.

RECORDERS:

1. Speedomax-M: Eleven (11) recorders were sampled. One-hundred-ten (110) COTs were evaluated. There were three instances where the recorder was outside "as-found", one (1) outside "as-left", and two (2) inop. All failures were on 3 recorders. All but one failure was corrected by "cleaning". The pen was seen to be stuck. This failure mode is best found by the source check and not a COT (Source checks are done at the discretion of the operations staff and procedurally at intervals less than COTs - Ref 4, 5, and 6). A review of maintenance history shows that indeed the operations staff has found these recorders inoperative by this method and cleaning was the fix. Additionally, these are obsolete and are slated to be replaced (WBN MIL 519 - Plant Recorder System Replacement).

2.

Speedomax-250: Two (2) recorders were sampled. Eighty-two (82) COTs were evaluated.

There was one (1) instance where the recorder was found outside the "as-found". All other COTs were within the "as-left". This is a 98.8% success rate. The failure is statistically insignificant.

3. Yokogawa-UR1800: One (1) recorders were sampled. Forty eight (48) COTs were evaluated. There were no instances where the recorder was found outside either the "as-found" or the "as-left". This is a 100% success rate.

Universal Flow Switch:

1. 92 day data : One (1) instrument, eleven (11) COTs, zero (0) OAF, four (4) OAL. This is a 100% success rate.

2.

18 month data: Seven (7) instruments, nineteen (19) COTs, four (4) OAF, three (3) OAL, four (4) instruments with no failures. The failures are on switches measuring ERCW system water. This is raw water and is a possible problem. Three (3) switches on Condenser Demineralizer or CCS system water showed no failures. The failures occurred on flow switches that are already on an 18 month calibration cycle. PER 00-001770-000 has been written to address this problem.

VX-252:

20 COTs, 2 Instruments, No Failures, All within "as-left".

CONCLUSION:

The system engineer data helps confirm that the radiation system as a whole is working well.

The feedback mechanism is in place (SPP 6.7 and NEDP 12) to evaluate and correct anomalous or mis-operating equipment.

Prepared: Richard Brehm Checked: Edward Bradley Page 23

Extending Channel Operational Test Frequency for Radiation Monitors WBN-EEB-EDQ1090-99005 Rev. 0 The analog ratemeters are working exceptionally well. There were no failures in the engineered features. One ratemeter was identified as being anomalous (but not inoperative) and is being replaced (WR C340313).

Yokogawa and the Speedomax 250 recorders are working well. Three of the eleven Speedomax M recorders sampled showed problems. However, the problems found are best determined by source checks and not COTs. The failure mode was a stuck pen and the resolution was to clean the slidewire or contacts. A source check that is already performed on a more frequent basis than COTs is the preferred method for finding these failures. Additionally, these recorders have been identified by the obsolete equipment project to be replaced.

Flow switches on raw water that are presently on 18 month intervals do show a problem. PER 00-001766-000 was written to address this problem.

Local indicators have shown failures when the MCR and VX-252 indicators have shown no problems. There is no fundamental difference in movements of these devices. PER 00-00 1766-000 was written to address this problem.

Prepared: Richard Brehm Checked: Edward Bradley Page 24

WBN-EEB-EDQ1090-99005 Rev. 0 Extending Channel Operational Test Frequency for Radiation Monitors Attachment I 71

'p P

CCC PERIODIC WO NO: 98-01.2949-000 COMP ID-WNUI-1.LPR -090-0423 DESCRXPTXOKt LOOP.

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RR132C P-30AM DIFF 13 DRIC P-30AM 01FF l

10 iR112C lRP-3OAM D1FF 11 LOCAL IND LOCAL IND CAL ONLY OAF NA NA NA NA NA NA NA NA NA NA NA NA 1

0 RECORDER SpeedoMax RECORDER RECORDER M OAL SpeedoMax SpeedoMax OMLY M OAF M INOP 0

0 0

0 1"

0 0

0" 0

0 0

0 0

1 2

o 1

0 0 0 0

N I

I I

I SpeedoMax M

RECORDER OAL 0

0:

, 0::

0":

1..

A 43!

LOCAL IND CAL LOCAL IND LOCAL IND LOCAL IND LOOP MODEL

1. OF COTS ONLY OAF INOP NO DATA 0R004 RP-1AM 10

1.

2 0

3 NA NA 0R011 RP-1AM 9

0 0

.0 0

NA NA 1R059 RP-1AM 9

w1 1

1 NA 2R210 r RP-1AMl 10 0

0 0

0 NA0 1R275 RP-IAM i

9 0

0 0

5 N

A NAI 1R7 P1MI0

0.

0 1

NA NA i_

i_

I8 I

_q LOOP MODEL

1. of COTs 1R271 RP 2C j

10 1R272 RP 2C 10 1R273 RP 2C 11 1R274 RP 2C 11 1R291 RP2C 11 1R293 RP 2C 9

LOOP MODEL

, # of COTS 1R002 RP 2AM 10

+

______ I s4~

LOCAL IND 0

I

________ I

________ I Page I

combined SpeedoMax M # Of RECORDER.

SpeedoMax Instrumentsw SpeedoMax. RECORDER RECORDER M#of ith NO SpeedoMax M OAL SpeedoMax SpeedoMax Instruments Failures M # of COTs

.ILY M OAF M JNOP 11 7

110

'1 3

2 Loa Iniatr wit NO Loa Iniao Inia TIdcao O niao of Instruments Failures

1. t of COTs OAL ONLY OAF COT INOP_____

t t

5 t

3 1

__________I____

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_ _ I it _

-I-i -A Page 2

0A Design Verification Report Extending Channel Operational Test Frequency for Radiation Monitors EDQ1090-99005 Revision 001 Design Verified (System, Structure, Component):

EDQ1090-99005 Revision 001 X I Interdisciplinary Review Off-Project Design Review Individual Critical Review Alternate Calculation Qualification Testing Document(s) Reviewed:

Calculation: EDQ1090-99005 Summary of Review (attach additional sheets if needed):

This calculation was reviewed for compliance with applicable procedures and technical instructions, appropriate references, completeness of documentation, and reasonableness of results.

This calculation is prepared in accordance with iVA Procedure NEDP-2, Bechtel Calculation Procedure 25402-3DP-GO4G-00037, and TVA Branch Technical Instruction EEB-TI-28, Setpoint Calculations, Revision 7. The calculation is verified in accordance with TVA Procedure NEDP-5, and Bechtel Procedure 25402-3DP-GO4G-0027.

Conclusions (attach additional sheets if needed)

Design Input Sources used in the uncertainty were-documented and appropriate. The methods used to determine uncertainty were consistent with procedural requirements defined above and the results obtained reasonable to demonstrate that the accuracy is adequate for the intended purpose.

/

Signature of Verifier Organization Position Report Prepared by Bechtel Control Systems Name Engineer Jeremy A. Thompson Sign Date February 10, 2011 Attachment No. 3: Page 1 of 3 Calculation ID: EDQ1090-99005 Refer to the electronic documents in TVA Business Support Library (BSL) for current revision.

25402-3DP-GO4G-00027 EFFECTIVE DATE: 12-23-09 Page I of 3

Design Verification Checklist (excerpted from ANSI N.45.2.11 [1974 Edition]

Extending Channel Operational Test Frequency for Radiation Monitors EDQ1090-99005 Revision 001 Design Verification Element Note: Any items checked "No" automatically imply the design is not Yes No NIA verified.

Remarks

• X Is the person performing the design verification qualified to originate the document?

X Is the design verification being performed by someone other than the supervisor of the originator?

If the supervisor of the originator is performing design verification, mark the answer "N/A" and provide justification in the document (see section 2.3.7 for requirements).

X Do the collective results of the design input/output substantiate the concept and approach chosen to ensure the design activity provides an adequate, accurate, and workable solution to the problem/question being resolved?

X Were the design inputs correctly selected and incorporated into The design inputs are design?

consistent with the requirements for this type calculation.

X Are assumptions necessary to perform the design activity There are no assumptions.

adequately described and reasonable? Where necessary, are assumptions identified for subsequent re-verifications when the detailed design activities are completed?

X Are the appropriate quality and quality assurance requirements specified?

X Are the applicable codes, standards and regulatory requirements including issue and addenda properly identified, and their requirements for design met?

X Have applicable construction and operating experiences been considered?

X Have the design interface requirements been satisfied?

X Were appropriate design methods and computer programs used?

X Is the design output reasonable compared to design inputs?

X Are the specified parts, equipment, and processes suitable for the required application? Are all applicable construction specifications referenced on the drawing(s)?

X Are the specified materials compatible with each other and the design environmental conditions to which the material will be exposed?

X Have adequate maintenance features and requirements been specified?

X Are accessibility and other design provisions adequate for initial installation and for performing needed maintenance and repair?

X Have adequate accessibility been provided to perform the in-service inspection expected to be required during the plant life?

X Has the design properly considered radiation exposure to the public and plant personnel (e.g., ALARA)?

I Attachment No. 3: Page 2 of 3 Calculation ID: EDQ1 090-99005 Refer to the electronic documents in TVA ibusins L,,.....,y.

25402-3DP-GO4G-00027 EFFECTIVE DATE: 12-23-09 Page 2 of.3

Design Verification Checklist (excerpted from ANSI N.45.2.11 [1974 Edition]

Extending Channel Operational Test Frequency for Radiation Monitors EDQ1090-99005 Revision 001 Design Verification Element Note: Any items checked "No" automatically imply the design is not Yes No N/A verified.

Remarks

• X Are the acceptance criteria incorporated in the design documents sufficient to allow verification that design requirements have been satisfactorily accomplished?

X Have adequate pre-operational and subsequent periodic test requirements been appropriately specified?

X Have adequate handling, storage, cleaning, and shipping requirements been specified?

X Are adequate identification requirements specified?

X Are requirements for record preparation review, acceptance, retention, etc., adequately specified?

X Has constructability been adequately considered?

• It is encouraged that the verifier provide a brief explanation of the considerations utilized in performing the design verification activity in the "Remarks" column Attachment No. 3: Page 3 of 3 Calculation ID: EDQ1090-99005 Refer to the electronic documents in TVA Business Support Library (BSL) for current revision.

25402-3DP-GO4G-00027 EFFECTIVE DATE: 12-23-09 Page 3 of 3