ML102800347

Final Safety Analysis Report (FSAR) - Response to Requests for Additional Information

ML102800347
Person / Time
Site:	Watts Bar
Issue date:	10/04/2010
From:	Bajestani M Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC ME2731, TAC ME4074, TAC ME3945
Download: ML102800347 (43)

Text

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 October 4, 2010 10 CFR 50.4(b)(6) 10 CFR 50.34(b) 10 CFR 2.390(d)(1)

ATTN: Document Control Desk U.S. Nuclear Regulatory Commission 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 Requests for Additional Information This letter responds to a number of requests for additional information (RAIs) regarding the Unit 2 FSAR. provides the responses to RAIs involving multiple FSAR chapters. provides the report referred to in the response to RAI 5.2.2 - 1.a. provides the curves referred to in the response to RAI 5.2.2 - 1.d.. provides the new commitments contained in this letter.

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 4 th day of October, 2010.

Sincerely, Masoud 'estani Watts, Unit 2 Vice President

U.S. Nuclear Regulatory Commission Page 2 October 4, 2010

Enclosures:

1. Response to RAIs Regarding Unit 2 FSAR

2. Report for Response to RAI 5.2.2 - 1.a

3. Curves for Response to RAI 5.2.2 - 1.d

4. List of New Regulatory Commitments

References:

1. NRC to TVA letter dated August 11, 2010, "Summary of August 3, 2010, Meeting With Tennessee Valley Authority Regarding Watts Bar Nuclear Plant, Unit 2, Final Safety Analysis Report" (ADAMS Accession No. MLI102180055)

2. TVA to NRC to TVA letter dated August 6, 2010, "Watts Bar Nuclear Plant (WBN) Unit 2 - Final Safety Analysis Report (FSAR) - Response to Requests for Additional Information" (ADAMS Accession No. ML102210440)

3. NRC to TVA letter dated September 20, 2010, 'Watts Bar Nuclear Plant, Unit 2 - Request for Additional Information Regarding Final Safety Analysis Report Related to Section 15 (TAC No. ME4074)" (ADAMS Accession No. ML102590244)

4. NRC to TVA letter dated August 19, 2010, 'Watts Bar Nuclear Plant, Unit 2 -

Request for Additional Information Regarding Final Safety Analysis Report Amendment Related to Section 11 (TAC No. ME3945)" (ADAMS Accession No. ML102240513)

5. NRC to TVA letter dated August 27, 2010, 'Watts Bar Nuclear Plant, Unit 2 -

Request for Additional Information Regarding Licensee's Final Safety Analysis Report Amendment Related to Chapters 11, "Radioactive Waste Management" and 12, "Radiation Protection" (TAC No. ME2731)"

(ADAMS Accession No. ML102360201) 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

ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 RAIs for FSAR 2.3.1.3, 2.3.2, 4.2, and 5.2.2 [taken from NRC letter dated 08/11/2010 (ADAMS Accession No. ML102180055):

Section 2.3.2. Local Meteorology 2.3.2 -9. Some of the precipitationmeasurements made at the Watts Bar Dam site appearto be more limiting than subsequent measurements made at the Watts Bar meteorologicaltower site. However, the limiting Watts Bar Dam site values were not carriedforward or otherwise discussedin Amendment No. 99. As a result, the summary historic recordpresentedin the FSAR may appearto be incomplete.

Therefore, in the next FSAR amendment, discuss the entire period of historic measurements. Confirm that the Watts Bar Dam site data were consideredand that the limiting values have been identified.

Response: Amendment 101 to the Unit 2 FSAR will revise Table 2.3-4 to include precipitation measurements for both Watts Bar Dam and Watts Bar Nuclear Plant (WBN) meteorological tower. Additionally, the corresponding FSAR text will be modified to reflect the table revision and to specifically state the limiting values.

2.3.2 - 10. The text of FSAR Section 2.3.2.2 of Amendment No. 99 states that the annual average precipitationfor 1974 through 2009 was approximately 43.4 inches while Table 2.3-4 lists a slightly different value of 45.43 inches. Confirm which number is correct. Further,either number is a decrease from the 1941 through 1970 value of 52.57 inches in prior FSAR amendments. Why should 52.57 inches not be identified as the limiting value?

Response: As stated in the response to RAI 2.3.2 - 9, Amendment 101 to the Unit 2 FSAR is revising Table 2.3-4. This amendment will also modify the FSAR text to be consistent with revised Table 2.3-4 and to include the 52.57 value. Also, maximum values for both Watts Bar Dam and WBN are listed because data are from different locations for different time periods.

2.3.2 - 11. Some of the humidity data presented in FSAR Tables 2.3-10 and 2.3-11 of prior FSAR amendments appearto be more limiting than subsequent measurements reportedin Amendment No. 99. However, the limiting values were not carried forward or otherwise discussed in Amendment No. 99. Forexample, in FSAR Table 2.3-10, the Januaryand July values of extreme minimum relative humidity appearto be lower in previous FSAR amendments than in Amendment No. 99. In FSAR Table 2.3-11, the extreme maximum of absolute humidity appears to be lower for Februaryand extreme minimum appears to be higher in July in Amendment No. 99 than in previous FSAR amendments. As a result, the summary historic recordpresented in the FSAR may appearto be incomplete. Therefore, in the next FSAR amendment, please discuss the entire periodof historic measurements. Confirm that data humidity measurements reportedin prior FSAR amendments were considered and that the limiting values have been identified.

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ENCLOSURE 1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Response: Amendment 101 to the Unit 2 FSAR will revise FSAR tables 2.3-10 and 2.3-11 to include the entire period of historical measurements.

TVA has confirmed that data for the full period has been considered in the supporting analyses and designs.

2.3.2 - 12. Reconfirm that the relevant meteorologicalparametersand analysis in FSAR Section,2.3, "Meteorology," were reviewed and updated, as appropriateand, that, although resolution of Items 1 through 11 above may result in changes to the FSAR, any changes associatedwith these items are either not sufficient to impact design assumptions or that design assumptions have been appropriatelyupdated.

Response: TVA has confirmed that data from Items 1 through 11 for the full period has been considered in the supporting analyses and designs.

2.3.2 - 13.- 1. Please note the following for possible editorialadjustment in the FSAR and make changes as appropriate:

Review and, as appropriateand feasible, reformat titles of tables associated with Section 2.3, "Meteorology," to clarify the listings in Section 2.3, List of Tables (pages2-vi through 2-xii). In some cases, the title listings:

a. appearto contain typographical errorsbecause of apparentword processormanipulation of footnote superscriptsin the actual titles,

b. are identical to one or more other table listings because apparentword processortruncation has omitted key information due to the arrangement of the information in the actual title, and

c. contain awkward line breaks in the title listings.

The following are examples:

- Tables 2.3-5, 2.3-6, 2.3-9 & 2.3 superscript

- Tables 2.3-13 & 2.3-14, and Tables 2.3-15 & 2.3 truncation

- Tables 2.3-17 through 2.3-40 -truncation, Table 2.3-32 - typographical error Tables 2.3-45 & 2.3-53, Tables 2.3-46 & 2.3-54, etc. - truncation

- Tables 2.3-61 & 2.3-61A, etc. - truncation, superscript,line breaks In addition, considerationshould be given to editorialmodification of several of the table titles to conform to common practice. For example:

Tables 2.3-61 through 2.3-63 and 2.3-64 through 2.3-66a - capitalize "q"in X/Q.

Tables 2.3-67 and 2.3-67a - capitalize "loca" Response: Amendment 101 to the Unit 2 FSAR will make the needed corrections.

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ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 2.3.2 - 13. - 2. Please note the following for possible editorialadjustment in the FSAR and make changes as appropriate:

The first paragraphof Section 2.3.3.3 states that Tables 2.3-45 through 2.3-52 and Tables 2.3-53 through 2.3-60 summarize data measured from 1974 through 1988 and 1977 through 1988, respectively. However, the titles of the tables state that the data were measured from 1974 through 1993 and 1977 through 1993, respectively. Table 2.3-13, which references data measuredfrom 1974 through 1988, is not mentioned in Section 2.3.3.3.

Please resolve this discrepancy between the text and tables.

Response: Amendment 101 to the Unit 2 FSAR will update Section 2.3.3.3 to use dates that match those contained in Tables 2.3-45 through 2.3-60. Table 2.3-13 is referenced in Section 2.3.2.2; however, this table currently does not display the correct information. Amendment 101 will correct this table also.

2.3.2 - 13. - 3. Please note the following for possible editorialadjustment in the FSAR and make changes as appropriate:

Table 2.3-2: Several of the column headings are out of alignment. The footnote superscriptfor the Daily Mean column is shown as "3" although it appears that the intended superscriptwas probably "a." in addition, should "HighestDaily Minimum" be "Lowest Daily Minimum"?

Response: Amendment 101 to the Unit 2 FSAR will make the needed corrections to Table 2.3-2.

2.3.2 - 13. - 4. Please note the following for possible editorialadjustment in the FSAR and make changes as appropriate:

Table 2.3-3: Several of the column headings are out of alignment.

Response: Amendment 101 to the Unit 2 FSAR will make the needed corrections to Table 2.3-3.

2.3.2 - 13. - 5. Please note the following for possible editorialadjustment in the FSAR and make changes as appropriate:

Table 2.3-4: Delete "Dam"from the title if the 1974 through 2009 data were collected at the Watts Bar meteorologicaltower location, rather than at the Watts Bar Dam site as was done prior to 1974.

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ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Response: Amendment 101 to the Unit 2 FSAR will make the needed correction to Table 2.3-4.

2.3.2 - 13. - 6. Please note the following for possible editorialadjustment in the FSAR and make changes as appropriate:

Table 2.3-5: Are the column superscripts,"*"intended to refer to item "b"in the footnotes? Also, does the period of record include 2001 although the reference title, Climatographyof the United States, No. 20, 1971-2000, appearsto cite data only through 2000?

Response: Amendment 101 to the Unit 2 FSAR will revise Table 2.3-5 to match superscripts and footnotes, and to match the period of record as stated in the source reference.

2.3.2 - 13. - 7. Please note the following for possible editorialadjustment in the FSAR and make changes as appropriate:

Table 2.3-10: Is the extreme minimum relative humidity for March 10.4 percent, ratherthan 1.04 percent?

Response: The correct value is 10.4 percent. Amendment 101 to the Unit 2 FSAR will correct this typographical error.

2.3.2 - 13. - 8. Please note the following for possible editorialadjustment in the FSAR and make changes as appropriate:

Table 2.3-11: Is the extreme annual minimum absolute humidity 0.4 percent?

Response: This value was a restatement of the extreme monthly values, and it was not intended to represent an annual value. To avoid confusing monthly and annual values, Amendment 101 to the Unit 2 FSAR will revise Table 2.3-11 and other tables as necessary to remove information from the "Annual" line that does not reflect actual annual values.

4.2 - 1. TVA described its planned approach toward answering the NRC staff's requests for additionalinformation regardingnuclearfuel in Section 4.2 of the FSAR. The NRC staff indicated that it had one issue with the information. In NRC Information Notice 2009-23, "NuclearFuel Thermal Conductivity Degradation,"the NRC staff stated that some fuel thermalmodels do not account for the effect of degradationin the thermal conductivity of uranium fuel pellets with increasingexposure. The NRC staff stated that TVA should review this issue because it could lead to nonconservative results for thermal conductivity.

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ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Response: In its August 6, 2010, letter to the NRC (ADAMS Accession No. ML102210440), TVA provided the following response to RAI SNPB 4.4.2:

"The licensed design models that utilize equation 4.4-1 for fuel thermal conductivity were approved by the NRC in the Safety Evaluation Report (SER) to Topical Report WCAP-15063-P-A, Revision 1, with Errata, "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)" (TAC No. MA2086).

PAD 4.0 was approved by the NRC, and it is documented in Section 4.3 of the SER that an audit calculation was done with the FRAPCON-3 code, which accounts for thermal conductivity degradation. This calculation showed that PAD 4.0 yielded conservative results with respect to the FRAPCON-3 code."

A review of the Information Notice (IN)2009-23 indicates it is concerned with fuel thermal models predating 1999. The INstates:

"Beginning in 1999, several reactor fuel vendors submitted improved fuel thermal models to the NRC for review and approval. These new models incorporate updates to the fuel thermal conductivity models that account for degradation caused by irradiation. The improved vendor models generally considered experimental qualification data that were substantially similar to the data considered in NUREG/CR-6534. However, the staff is aware that models that do not account for the effect of degradation are still used to perform safety analyses."

The SER for PAD 4.0 was received from the NRC on May 12, 2000, following the audit described above which indicated conservative results and therefore should have closed the IN 2009-23 issue for PAD. The conclusion of the SER stated:

"The staff has reviewed the Westinghouse improved fuel performance code PAD 4.0 as described in WCAP-15063-P, Revision 1, and concludes that PAD 4.0 is acceptable for fuel licensing applications up to rod average burnup 62,000 MWdMTU."

The NRC has recently approved a licensing action for an extended power uprate at another plant; PAD 4.0 was used in its safety analysis. Since the use of the PAD code is not unique to WBN, TVA facilitated a discussion with Westinghouse to allow the NRC to express reviewer concerns with the accuracy of the model at higher burnups. During that discussion, in the public meeting held September 15, 2010, at the NRC offices, Westinghouse noted that the code had been calibrated using data from higher burnup fuel, that E1-5

ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 the NRC contractor audit of PAD had examined this issue, and that at higher burnups, less power is produced in the fuel and that "burn-down" credit could be accounted for to additionally offset thermal conductivity degradation effects. The NRC reviewer was referred to Section 4.3 of the SER which discussed this issue using FRAPCON-3 as a comparison. The SER supports the conclusion that applications for WBN are well within the approved limits for the PAD 4.0 code.

The following RAI was divided into four portions for ease of response:

5.2.2 - 1.a. Regarding the information on TVA's analyses of various accidents and transients, the NRC staff noted that in support of FSAR Section 5.2.2, an analysis of a loss of load transientmust be completed that does not credit the first safety grade actuation signal.

Response: In a loss of load transient, the first expected safety-grade actuation signal would be the reactor trip. The turbine trip without reactor trip overpressure transient is discussed in FSAR Section 15.2.7.2, "Analysis of Effects and Consequences." The basis for this discussion is the Westinghouse Final Overpressure Protection Report for Watts Bar Nuclear Plant Unit 2, March 2010. This report is referenced in FSAR Sections 5.2 and 15.2 and is provided as Enclosure 2.

5.2.2 - 1.b. Further,the NRC staff discussed its comments on the analysis of an inadvertent actuation of the emergency core cooling system. The NRC staff informed TVA that it needs to consider the information in NRC Regulatory Information Summary 2005-29, 'Anticipated Transients That Could Develop into More Serious Events,"

dealing with the qualificationof the pressurizerpower-operatedrelief valves (PORVs) for water dischargeand the closure of the PORV block valves.

Response: Chapter 15 of the Unit 2 WBN FSAR provides an analysis consistent with the analysis of record for the licensing of Unit 1. WBN will consider the information in Regulatory Information Summary 2005-29 and notify the NRC of our plan of action for resolution of this concern by November 1, 2010.

5.2.2 - 1.c. The NRC staff also stated that an analysis was requiredin the FSAR for a malfunction of the chargingand volume control system.

Response: This same question was asked in RAI 15.0.0 - 1 in NRC letter dated 09/20/2010 (ADAMS Accession No. ML102590244)]. TVA will respond to this question under that RAI.

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ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 5.2.2 - 1.d. In addition, the NRC staff noted that the graph describing the response to a main steamline break with a loss of offsite power was terminated at 200 seconds. The NRC staff noted that the transientconditionshad not yet fully stabilized at that point in time and that it appearedthat there was an insufficient boron addition to control power. The NRC staff asked TVA to similarly review the results presented for other analyses to ensure that the results were not truncatedearly.

Response: The figures presented for the steamline break with a loss of offsite power (i.e., Unit 2 FSAR Figures 15.4-12A, 15.4-128, and 15.4-12C) include run times up to 200 seconds. Nuclear Power is the transient parameter which influences the departure from nucleate boiling ratio the most and had turned around before 200 seconds. The actual analysis run was executed to 600 seconds following the steamline break even though only the first 200 seconds are presented in the FSAR. Re-scaled figures from the original analysis (provided in Enclosure 3) demonstrate a noticeable decrease in the core reactivity and nuclear power from 200 to 600 seconds.

The figures provided for the remainder of the events have been examined, and it is determined that all other events do not have results that were truncated early.

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ENCLOSURE 1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 RAIs for FSAR Chapter 11 [from NRC letter dated 08/19/2010 (ADAMS Accession No. ML102240513)]

11 - 1. Table 11.2-5, Columns 6, 7, and 8.

a. Explain why these columns were added to the Final Safety Analysis Report (FSAR). Provide some detail as to theirpurpose, other than to match the Unit 1 FSAR.

Response: Columns 6, 7, and 8 were added to the FSAR to reflect the potential waste release concentrations for different modes of waste release.

Column 6 (no Condensate Demineralizer [CD] processing) indicates a yearly release of 30.03 Ci with no CD processing of waste and no limitations on steam generator blowdown concentrations. This operational mode is not normally used, since long-term use results in exceeding the 5 Ci/yr limit in 10 CFR 50, Appendix I.

Column 7 indicates that the total release, including untreated steam generator blowdown, is significantly below the 10 CFR 50, Appendix I limit of 5 Ci/yr if the steam generator blowdown concentration is restricted to the Lower Limit of Detection (LLD) of 5E-7 uCi/cc (Watts Bar Offsite Dose Calculation Manual [ODCM])

gross gamma during the release and no other CD waste is processed during the release. Column 7 does include other releases from waste holdup tanks which are treated using the Mobile Demineralizers (MDs).

Column 8 indicates steam generator blowdown can be released untreated and remain within the 10 CFR 50, Appendix I limit of 5 Ci/yr if the steam generator blowdown concentration is restricted to a maximum concentration of 3.65E-5 uCi/cc gross gamma during the release and no other CD waste is processed during the release. Column 8 does include other releases from waste holdup tanks which are treated using the MDs.

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ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391

b. Columns 7 and 8 are labeled "no CD [condensate polishing demineralizer]."

However, the magnitudes of the releases listed in both of these columns are closer to Column 5 (w/CD and mobile demineralizer)than they are to Column 6.

Verify these are untreatedreleases as indicated.

Response: As indicated in the response to RAI 11 - 1.a, Columns 7 and 8 reflect release of steam generator blowdown without treatment and no other releases from the CDs, but with restrictions on steam generator blowdown concentrations. Columns 7 and 8 also include other releases from waste holdup tanks which are treated using the MDs.

Column 5 reflects releases of the condensate after treatment by the CDs and the MDs and treatment of the steam generator blowdown and waste holdup tanks with the MDs.

Thus, Columns 7 and 8 reflect release of the steam generator blowdown untreated for limited concentrations with no other releases from the CDs and release of the waste holdup tanks after treatment. The only difference between Column 5 and Columns 7 and 8 is Column 5 includes the condensate with a MD factor of 1000. That contribution for each isotope is extremely small, and thus the columns should be on the same magnitude.

c. What is the basis for the 3.6 E-5 uCl/cc limit used to calculate the steam generatorblowdown (SGBD) contributionin Column 8?

Response: The steam generator blowdown concentration of 3.6 E-5 uC/cc is the limiting concentration that will permit untreated steam generator blowdown releases and remain within the 10 CFR 50, Appendix I limit of 5 Ci/yr. It is a back calculated value.

d. What is the isotopic distribution of the SGBD release contributions?

Response: The steam generator blowdown isotopic distribution is the same as the secondary system liquid isotopic distribution. See Table 1.

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ENCLOSURE 1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 TABLE I Secondary Water Concentration Nuclide (uCi/gm)

Class 1 Kr-85m 0.OOE+00 Kr-85 0.OOE+00 Kr-87 0.OOE+00 Kr-88 0.00E+00 Xe-131m 0.OOE+00 Xe-133m 0.OOE+00 Xe-133 0.OOE+00 Xe-135m O.OOE+00 Xe-135 O.OOE+00 Xe-137 O.OOE+00 Xe-138 0.00E+00 Class 2 Br-84 9.56E-08 1-1.31 1.41E-06 1-132 3.37E-06 1-133 4.03E-06 1-134 2.93E-06 1-135 6.19E-06 El-10

ENCLOSURE 1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 TABLE I Secondary Water Concentration Nuclide (uCi/gm)

Class 3 Rb-88 7.36E-07 Cs-134 4.58E-07 Cs-136 5.56E-08 Cs-137 6.11E-07 Class 4 N- 16 1.29E-06 Class 5 H-3 1.OOE-03 Class 6 Na-24 1.86E-06 Cr-51 1.56E-07 Mn-54 7.80E-08 Fe-55 5.88E-08 Fe-59 1.44E-08 Co-58 2.28E-07 Co-60 2.64E-08 Zn-65 2.52E-08 Sr-89 6.84E-09 El-11

ENCLOSURE 1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 TABLE I Secondary Water Concentration Nuclide (uCi/gm)

Sr-90 5.88E-10 Sr-91 3.52E-08 Y-90 5.88E-10 Y-91 m 4.34E-09 Y-91 2.52E-10 Y-93 1.50E-07 Zr-95 1.92E-08 Nb-95 1.32E-08 Mo-99 3.03E-07 Tc-99m 1.40E-07 Ru-103 3.72E-07 Ru-106 4.44E-06 Rh-103m 7.89E-03 Rh-106 4.44E-06 Ag-110m 6.36E-08 Te-129m 9.36E-09 Te-129 2.96E-07 Te-131m 6.60E-08 Te-131 3.97E-08 Te-132 7'98E-08 El-12

ENCLOSURE 1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 TABLE I Secondary Water Concentration Nuclide (uCi/gm)

Ba-137m 9.79E-03 Ba,140 6.25E-07 La-140 1.13E-06 Ce-141 7.32E-09 Ce-143 1.22E-07 Ce-144 1.92E-07 Pr-143 2.96E-03 Pr-144 1.92E-07 W-187 1.07E-07 Np-239 1.02E-07

e. Verify that releases consistent with Columns 7 and 8 (e.g., with SGBD releases) meet Title 10 of the Code of FederalRegulations (10 CFR) Part20.

Response: The releases in Columns 7 and 8 meet the requirements of 10 CFR 20. See the response to RAI 11 - 2.c.

11 -2. Revise Paragraph11.2.6.5 to discuss limitationsof untreatedreleases from the plant.

a. Table 11.2-5 does address compliance with 10 CFR Part20. The table presents total curie release during different radiologicalwaste configurationsand compares them with the limit of RM 50-2 (5 Ci/yr limit). Table 11.2-5a (untreated)and Table 1 1.2-5b (treated)seem to demonstrate noncompliance and compliance, respectively, with 10 CFR Part20.

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ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Response: Table 11.2-5 addresses compliance with 10 CFR 50, Appendix I.

The intent of Tables 11.2-5a and 11.2-5b is to address compliance with 10 CFR Part 20 for two different releases. The releases in Table 11.2-5a do not comply with 10 CFR Part 20, and thus is not an acceptable method of operation. The releases in Table 11.2-5b do comply with 10 CFR Part 20 and is an acceptable method of operation.

Amendment 101 to the Unit 2 FSAR will revise the text in Section 11.2.6.5 to clarify the use of these tables. Additionally, Tables 11.2-5c and 11.2-5d will be added to show that releases in Columns 7 and 8 of Table 11.2-5 comply with 10 CFR Part 20.

b. Tables 1 1.2-5a and b do not demonstrate that "releasesfrom the plant are in accordance with the design objectives" (found in FSAR Section 11.2. 1), since they do not include SGBD contributionsfrom tube leaks (objective (1). In addition, Table 1 1.2-5a clearly demonstratesthat liquid effluent processing is necessaryto meet 10 CFR Part20. Therefore, it is unclear how design objective (2) is demonstrated. Explain how FSAR Section 11.2.1 objectives (1) and (2) are met.

Response: Objective (1) of Unit 2 FSAR Section 11.2.1 is "Steam Generator tube leaks." Tables 11.2-5a and 11.2-5b are based on secondary coolant isotopic concentrations listed in Table 6 (adjusted for WBN specific parameters) of ANSI/ANS718.1, 1984. The secondary coolant concentrations in the table are based on 75 lbs/day primary-to-secondary system leakage. Thus, Objective (1) of FSAR Section 11.2.1 is met.

Objective (2) of Unit 2 FSAR Section 11.2.1 is "Malfunction in Liquid Waste Processing System." Tables 11.2-5a and 11.2-5b include releases from Auxiliary Building floor drains (200 gal/day),

Turbine Building floor drains (7,200 gal/day), and other leaks and drains (10 gal/day). Malfunctions in waste processing systems would result in a loss of fluid to the floor drains and would be collected in various waste holdup tanks. Note the waste tanks are collected for a 24-hour period prior to release and are normally surveyed prior to release in accordance with the Watts Bar ODCM.

Thus, radioactive waste is held up and not released if a potential exists to exceed the 10 CFR 50, Appendix I, or 10 CFR 20 limits.

Based on the above, Objective (2) is met.

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ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391

c. Statements added to paragraph11.2.6.5 by Amendment 98 need clarificationand justification.

i. The first sentence implies that operating with releases consistent with Table 11.2-5a is acceptable,which is incorrect. These releases exceed the limits of 10 CFR Part20.

ii. Also, nothing in this FSAR supports the statement that releasinguntreated waste below 5E-7 uCi/cc will meet 10 CFR Part20 Appendix B limits.

Response: Amendment 101 to the Unit 2 FSAR will revise/clarify statements added to 11.2.6.5 by Amendment 98 as follows:

• Revise 11.2.6.5 to clarify that releases in Table 11.2-5a exceed the 10 CFR 20 limitations and are not acceptable.

" Revise the FSAR text to indicate that the total release in Column 7 of Table 11.2-5, including untreated steam generator blowdown, is significantly below the 10 CFR 20 limit ifthe steam generator blowdown concentration is restricted to the LLD of 5E-7 uCi/cc (WBN ODCM) gross gamma during the release and no other CD waste is processed during the release. The release does include other wastes from waste holdup tanks which are treated using the MDs.

Revise the FSAR text to indicate the steam generator blowdown can be released untreated and remain within the 10 CFR 20 limits if the steam generator blowdown concentration is restricted to a maximum concentration of 3.65E-5 uCi/cc gross gamma during the release (Column 8 of Table 11.2-5) and no other CD waste is processed during the release. The release does include other wastes from waste holdup tanks which are treated using the MDs.

• Add Tables 11.2-5c and 11.2-5d to the Unit 2 FSAR to show the above releases remain less than the 10 CFR 20 limits.

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ENCLOSUREI1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 RAIs for FSAR Chapters 11 and 12 [from NRC letter dated 08/27/2010 (ADAMS Accession No. ML102360201)]

Chapter 11 11.4-1 The response to question 11.4-1 in the July 31, 2010, submittal, states that "the Steam Generator(SG) Blowdown Liquid Sample Monitor was isolated in Unit 1 by DCN 29903. Monitor RE-90-124 was to be used solely to determine which SG has a leak during an SG Tube Rupture event. In place of the monitor,grab samples provide a quicker determination." Describe how the routine Watts Bar Unit 2 steam generatorblowdown will be monitored for radiologicalreleases.

Response: Unit 2 steam generator blowdown will be monitored continuously via radiation monitors 2-RE-90-120 and 2-RE-90-121 providing real-time measurements of the gross gamma radioactivity in the blowdown liquid from the steam generators to the cooling tower blowdown.

These monitors are capable of detecting the radioactivity resulting from a primary to secondary side leakage in compliance with 10 CFR 50 Appendix A, General Design Criterion 30, 60, 64, and 10 CFR 50 Appendix 1, RG 1.21 and RG 1.45.

In addition to the steam generator blowdown monitoring via radiation monitors 2-RE-90-120 and 2-RE-90-121, grab samples will be obtained monthly from each steam generator and analyzed for gamma isotopic.

During periods when both 2-RE-90-120 and 2-RE-90-121 are inoperable, grab samples will be obtained and analyzed once per every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the secondary coolant Dose Equivalent 1-131 is greater than or equal to 0.01 uCi/g or once every 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> when the secondary coolant Dose Equivalent 1-131 is less than 0.01 uCi/g.

During periods when there is no indication of a primary to secondary leak, the monitors' setpoints are set in accordance with procedures established by the ODCM to less than twice background. Upon receipt of a high radiation alarm signal from either channel, the monitor shall generate a signal to terminate the steam generator flow directly to the cooling tower blowdown line and divert the flow through the CD System.

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ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Chapter 12 12 - 5.b. Amendment 88 indicated that the normal operatingsource term was based on AMERICAN NationalStandardsInstitute document 18.1 1984 (Rev. 1).

Amendment 97 revised this descriptionto make a distinction between the normal operating design source term (now referencingANSI 18.1 Rev 0, 1976) and the "Currentoperationalradiationlevels," based on ANSI 18. 1 Rev. 1. Clarify and explain the purpose of this distinction.

Response: Section 12.3 of Amendment 88 to the Units 1 and 2 FSAR stated, "Normal plant design radiation levels are based on ANSI/ANS-18.1, 1984, which..."

Amendment 97 to the Unit 2 FSAR revised the words to, "In the initial plant design, radiation levels were based on ANSI/ANS-18.1, 1984, Revision 0, which... Current operational radiation levels are based on ANSI/ANS-18.1, 1984, Revision 1, and..."

Amendment 98 to the Unit 2 corrected the year for Revision 0 of the ANSI standard and revised the words to, "In the initial plant design, radiation levels were based on ANSI / ANS-18.1, 1976, Revision 0, which... Current operational radiation levels are based on ANSI/ANS-18.1, 1984, Revision 1, and..."

The change of year (1976 versus 1984) in Amendment 98 to the Unit 2 FSAR was an administrative correction only with no technical intent change. This error was a carryover from Amendment 88 to the Units 1 and 2 FSAR.

The NRC question implies there was a change in terminology from "normal operating design source term" to "current operational radiation levels." As can be seen from the first quote versus either the second or third quote above, there was no change in terminology (i.e., no change from "source term" to "radiation levels"). The Amendment 97 and 98 versions include information regarding the initial plant design (circa 1980) which was based on Revison 0 of the ANSI Standard and the current radiation levels which are based the newer Revision 1 version of the Standard.

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ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 12 - 7. - The response to RAI # 7 is unresponsive to the request. In lieu of layout drawings as requested(consistent with Section 12.3.1 of Regulatory Guide 1.70), identify which of the following radiationprotection facilities are common to Watts Bar Unit 1 and indicate whether they are sized and situatedproperly to support two operating units.

The facilities are:

Foreach of these facilities not shared with Unit 1, demonstrate they are sized and situatedproperly to support Unit 2 operation.

1. Controlledaccess areas Response: The Unit 2 controlled access area will be common with the Unit 1 controlled access area. This facility is located in the Service Building and is sized and situated properly to support two operating units.

2. personneland equipment decontamination, Response: The Unit 2 personnel and equipment decontamination facilities will be common areas shared with the Unit 1 personnel and equipment decontamination facilities. The personnel decontamination facility is located in the Service Building and is sized and situated properly to support two operating units. The equipment decontamination facility is located in the Auxiliary Building and is sized and situated properly to support two operating units.

3. contamination control areas, Response:. The primary Unit 2 contamination control area (i.e., radiologically controlled area [RCA] exit personnel monitoring station) will be a common area shared with the Unit 1 contamination control areas.

This facility is located in the Service Building adjacent to the Auxiliary Building common entrance/exit point. This RCA exit point is equipped with personnel contamination monitors (PCMs),

personnel gamma monitors, and tool and equipment monitors.

PCMs are also provided at common locations inside the Auxiliary Building (i.e., 676', 713', and 757' elevations) for prompt personnel contamination monitoring when personnel exit contaminated areas.

Frisker stations are also set up at remote exit points from the RCA when the use of PCMs is not practicable. Remote exit points are used only after prior approval by Radiation Protection and in the event of an emergency.

Protective clothing laundry services are provided offsite by a contractor. Soiled protective clothing is temporarily stored onsite in designated containers until ready for shipment offsite. Laundered protective clothing returned from the laundry contractor is received at the plant Power Stores loading dock and temporarily stored in a E1-18

ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 room adjacent to the loading dock until issued for use. This facility is located in the Service Building and is sized and situated properly to support two operating units.

Hot machine shop type activities are not normally performed onsite. Instead, radioactive materials requiring machine shop support are normally sent to either the TVA Western Area Radiological Laboratory or to a contractor licensed to receive radioactive material. If machining is performed onsite, a temporary RCA is set up in the clean area machine shop and controlled by Radiation Protection. Contamination control barriers, containments, or tents, as appropriate, are used to prevent the spread of contamination and airborne radioactivity.

Storage facilities are established in the Auxiliary Building and the Service Building to maintain a central location for the inventory of standard portable decontamination equipment and supplies (i.e., high energy particulate air [HEPA] vacuums, HEPA ventilation units, pressure washers, spray wands, mops, etc.) that may be used during the normal decontamination and control of radioactive contamination in the plant. These facilities are sized and situated properly to support two operating units.

4. in plant traffic patterns to radiologicallycontrolledareas, Response: The Unit 2 in plant traffic patterns to RCAs will be virtually the mirror image of the Unit 1 in plant traffic patterns to RCAs. Access to Unit 1 and Unit 2 RCAs will share common corridors at each elevation of the Auxiliary Building. Common areas are sized and situated properly to support two operating units. Access to satellite Unit 2 RCAs (i.e., outside tanks) will be controlled in the same manner as Unit 1.

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ENCLOSUREI1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391

5. health physics facilities (including dosimetry issue and maintenance, respiratory protection issue and maintenance, in-vivo and in-vitro bioassay,protective clothing and radiationsurvey instrument issue, Radiation Protection Managerand technicalstaff office/work space),

Response: The Unit 2 Health Physics facilities (including dosimetry issue and maintenance, respiratory protection issue and maintenance, in-vivo and in-vitro bioassay, protective clothing and radiation survey instrument issue, Radiation Protection Manager [RPM] and technical staff office/work space) will be common with Unit 1 facilities. These facilities and offices are located in the Training Center (dosimetry issue and maintenance, and in-vivo and in-vitro bioassay), Service Building (respiratory protection issue and maintenance, protective clothing and radiation survey instrument issue), and Office Building (RPM and technical staff office/work space). These facilities are sized and situated properly to support two operating units.

6. onsite laboratoryfor analysis of chemical and radiologicalsamples, and Response: The Unit 2 onsite laboratory for analysis of chemical and radiological samples will be common with the Unit 1 onsite laboratory for analysis of chemical and radiological samples.

These facilities are located in the Auxiliary Building and the Service Building and are sized and situated properly to support two operating units.

7. radiologicalcounting room.

Response: The Unit 2 radiological counting room will be common with the Unit 1 radiological counting room. This facility is located in the Service Building and is sized and situated properly to support two operating Units.

12 - 10. - The response to Question # 10 subpartsa., c., and e., do not clearly indicate the capabilitiesof the airbornemonitoring system as revised by Amendment 97.

Consistent with the acceptance criteriafor section 12.3.4.b, of the StandardReview Plan (NUREG 0800), verify that the airborneradioactivitymonitoring system for Watts Bar Unit 2 is capable of the following;

1. The monitoring system should be capable of detecting ten maximum permissible concentrationhours increasein particulateand iodine radioactivityfrom any compartment that has a possibility of containingairborneradioactivityand that normally may be occupied by personnel.

Response: The airborne monitoring system described in Amendment 97 is common equipment shared with Unit 1. The airborne monitoring system is capable of detecting ten derived air concentration (DAC) hour increase in particulate airborne radioactivity within the E1-20

ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 compartments that the monitors reside. The subject airborne monitoring system is not designed to detect iodine airborne radioactivity and does not detect airborne radioactivity in all Unit 2 compartments that may be normally occupied by personnel.

These latter functions are performed by the installed Auxiliary Building ventilation exhaust monitor.

2. Each monitor location has a local audible alarm (and,or visual alarms for monitors locatedin high noise areas)and variable alarm set points.

Response: Each monitor has a local audible and visual alarm with variable alarm set points (Slow alarm: 3E-9 uC/cc; Fast alarm: 1 E-8 uC/cc).

3. Readout and annunciationof each monitor are provided in the control room.

Response: Remote readout and annunciation (i.e., high radiation and instrument malfunction) for each monitor described in Amendment 97 is provided in the Main Control Room on panel 0-M-12.

4. Emergency power is initiatedafter loss of offsite power.

Response: Emergency (diesel) power is provided to each monitor following loss of offsite power to ensure continuous monitor operation.

12 - 13. - The firstpart of this response is acceptable. However, the response goes on to modify the statedposition with a discussion of a commitment to ANSI/American Nuclear Society guidance document ANSI/ANS 3.1-1981. The following text in the response, listed as paragraphC. from SPP-5, is not completely consistent with ANSI/ANS 3.1-1981, and is not acceptableto the staff. Specifically, contraryto the ANSI criteria,paragraphC.

1. does not specify that the five years of applied radiationprotection experience be orofessional exoerience. and...

Response: According to TVA-NQA-PLN89-A, TVA will meet the requirements of Regulatory Guide 1.8, Revision 2 (04/1987) for all new personnel qualifying on positions identified in regulatory position C.1 after January 1, 1990.

Position C.1.k. of Regulatory Guide 1.8, Revision 2 (04/1987) states 'The radiation protection manager should have the qualifications described in Section 4.4.4 of ANSI/ANS3.1-1981 with the clarification that 3 of the 4 years of experience in applied radiation protection should be professional-level experience."

Section 4.4.4.b. of ANSI/ANSI 3.1 (1981) states, "...the responsible individual shall have four years of experience in applied radiation E1-21

ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 protection. At least three years of this experience shall be in applied radiation protection work in a nuclear facility dealing with radiological problems similar to those encountered in nuclear power plants, preferably in a nuclear power plant...."

TVA was unable to validate the NRC statement that ANSI/ANS-3.1-1981 includes standards or specifications that require the RPM to possess 5 years of radiation protection professional experience.

Section 3.3.3.2.C. of NPG-SPP-05.1* (Radiological Controls) states, "The Radiation Protection Manager shall have ... five years of experience in applied radiation protection. At least three of the five years shall be professional-level experience in applied radiation protection work in a nuclear facility dealing with radiological problems similar to those encountered in nuclear power plants, preferably in a nuclear power plant...."

The RPM experience requirements established by TVA are conservative and meet or exceed the standards or specifications of ANSI/ANS 3.1- 1981 and Regulatory Guide 1.8, Revision 2 (04/1987). However, provisions are made in TVA-NQA-PLN89-A for personnel (RPM) who qualified on the position prior to January 1, 1990, to meet the requirements of Regulatory Guide 1.8, Revision 1-R (05/1977), which does establish the standard for RPMs to possess 5 years of radiation protection professional experience.

• SPP-5.1 was converted to NPG-SPP-05.1 with an effective date of 08/20/2010.

2 ... allows individuals that do not meet the qualificationcriteriato be temporarily assigned to fill the RPM position.

Response: Section 3.1 of ANSI/ANSI 3.1 (1981) states, in part, "Personnel temporarily filling positions due to absences of the principal may not meet the literal requirements of this standard. Use of personnel to fill a position for which they do not meet the minimum requirements set forth in this standard, is permissible on a justifiable basis ordinarily not to exceed three continuous months, and shall not be used as a means of reducing the level of minimum qualifications which the following paragraphs establish as being acceptable. The personnel filling positions due to the absence of a principal shall, as a minimum, possess the qualifications of the next lower level in that field. This does not apply to positions requiring active NRC senior operator or operator licenses or where otherwise stated in this standard."

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ENCLOSURE 1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Section 4.4.4.d. of ANSI/ANSI 3.1 (1981) states, "The individual who temporarily replaces the radiation protection group leader shall have a Bachelor Degree in a science or engineering subject and two years experience, one of which shall be nuclear power plant experience. Six months experience shall be onsite."

Revision 0 of NPG-SPP-05.1 (Radiological Controls) contains the following:

Section 3.3.2.C. states: 'The Radiation Protection Manager shall have ... Individuals who do not fully meet the literal requirements for the position may be temporarily assigned to fill that position. Such assignments shall be justified and a time for the temporary assignment specified and documented.

Temporary assignments shall not reduce the collective experience requirements specified for the level."

• Section 3.3.2.D. of Revision 0 of NPG-SPP-05.1 states, "Ifthe Radiation Protection Manager is temporarily replaced, the following shall apply:

The individual who temporarily replaces the Radiation Protection Manager shall have a bachelor's degree in a science or engineering subject and two years experience, one of which shall be nuclear power plant experience. Six months experience shall be onsite (See Section 5.0 Definitions for clarification)."

• Section 5.0 defines On-Site Experience as: "Applicable work performed at the plant for which the individual seeks qualification. Work shall involve that plant's systems and procedures. Observation of others performing work is not experience. In those cases where the collective experience does not exceed the sum of the minimum for individual positions, support shall be provided by additional personnel so that the collective experience exceeds the sum of the minimum."

ANSI/ANS-3.1-1981, Sections 3.1 and 4.4.4 clearly provide provisions for the temporary replacement of the RPM with an individual who possesses less than the normal experience criteria.

TVA determined that the qualification requirements for personnel temporarily filing the RPM position as specified in ANSI/ANS-3.1-1981, Section 3.1, were not fully included in Revision 0 of NPG-SPP-05.1. Specifically, it failed to address the following:

limiting the duration of appointment for a period ordinarily not to exceed 3 continuous months; and E1-23

ENCLOSURE1 Response to RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391

• requiring the incumbent to possess, as a minimum, the qualification of the next lower level in that field.

PER 248640 was initiated to identify/track resolution of this failure.

Revision 1 of NPG-SPP-05.1 was effective 09/15/2010. It contains the following (bolding added to show area of revision):

° Section 3.3.2.C states, 'The Radiation Protection Manager shall have ... Individuals who do not fully meet the literal requirements of ANSI/ANS-3.1-1981 for the position may be temporarily assigned to fill that position. Such assignments shall be justified and a time, ordinarily not to exceed three continuous months, for the temporary assignment specified and documented. Temporary assignments shall not reduce the collective experience requirements specified for the level."

Section 3.3.2.D. states, 'The individual who temporarily replaces the Radiation Protection Manager shall have a bachelor's degree in a science or engineering subject or the equivalent and possess the qualifications of the next lower level in that field (RP Superintendent)..."

E1-24

ENCLOSURE 2 Report for Response to RAI 5.2.2 - 1.a Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391

OVERPRESSURE PROTECTION REPORT FOR WATTS BAR NUCLEAR POWER PLANT UNIT 2 AS REQUIRED BY ASME BOILER AND PRESSURE VESSEL CODE SECTION III, ARTICLE NB-7300 MARCH 2010 Prepared by: M. C. Smith Approved: K. A. Plute 0

Certified: _ _ _ _ _ _ _ _ _

C. J. McHugh 0 Professional Engineer - 039195 - E Commonwealth of Pennsylvania

1.0 Purpose of Report This report documents the overpressure protection provided for the Reactor Coolant System (RCS) in accordance with the ASME Boiler and Pressure Vessel Code,Section III, NB-7300. This report documents the overpressure protection provided in the Westinghouse NSSS scope. The methods described in Reference 2 are applicable to Watts Bar Unit 2. The details of the limiting Final Safety Analysis Report analysis described herein are based on the analysis and evaluations contained in References 4 and 5.

2.0 Description of Overpressure Protection 2.1 Overpressure protection is provided for the RCS and its components to prevent a rise in pressure of more than 10% above the system design pressure of 2485 psig, in accordance with NB-7400. This protection is afforded for the following events which envelope those credible events which could lead to overpressure of the RCS if adequate over pressure protection were not provided.

1. Loss of Electrical Load and/or Turbine Trip

2. Uncontrolled Rod Withdrawal at Power

3. Loss of Reactor Coolant Flow

4. Loss of Normal Feedwater

5. Loss of Offsite Power to the Station Auxiliaries 2.2 The extent of the RCS is as defined in 10CFR50 and includes:

1. The reactor vessel including control rod drive mechanism housings.

2. The reactor coolant side of the steam generators.

3. Reactor coolant pumps.

4. A pressurizer attached to one of the reactor coolant loops.

5. Safety and relief valves.

6. The interconnecting piping, valves and fittings between the principal components listed above.

7. The piping, fittings and valves leading to connecting auxiliary or support system boundaries as defined in the system design documents.

2.3 The pressurizer provides volume surge capacity and is designed to mitigate pressure increases (as well as decreases) caused by load transients. A pressurizer spray system condenses steam at a rate sufficient to prevent the pressurizer pressure from reaching the setpoint of the power-operated relief valves during a step reduction in power level equivalent to ten percent of full rated load.

The spray nozzle is located in the top head of the pressurizer. Spray is initiated when the pressure controlled spray demand signal is above a given setpoint. The spray rate increases proportionally with increasing compensated error signal until it reaches a maximum value. The compensated error signal is the output of a proportional plus integral controller, the input to which is an error signal based on the difference between actual pressure and a reference pressure.

The pressurizer is equipped with 2 power-operated relief valves which limit system pressure for a large power mismatch to avoid actuation of the fixed high pressure reactor trip. The relief valves are operated automatically or by remote manual control. The operation of these valves also limits the frequency of opening of the spring-loaded safety valves. Remotely operated stop valves are provided to isolate the power-operated relief valves if excessive leakage occurs. The relief valves are designed to limit the pressurizer pressure to a value below the high pressure trip setpoint for all design transients up to and including the design percentage step load decrease with steam dump but without reactor trip.

Isolated output signals from the pressurizer pressure protection channels are used for pressure control. These are used to control pressurizer spray and power-operated relief valves in the event of increase in RCS pressure.

In the event of unavailability of the pressurizer spray or power operated relief valves, and a complete loss of steam flow to the turbine, protection of the RCS against overpressure is afforded by the pressurizer safety valves in conjunction with the steam generator safety valves and a reactor trip initiated by the Reactor Protection System.

There are 3 safety valves with a minimum required capacity of 420,000 lb/hr for each valve at system design pressure plus 3% allowance for accumulation.

The pressurizer safety valves are totally enclosed pop-type, spring loaded, self-activated valves with back pressure compensation. The set pressure of at least one of the safety valves will be no greater than the system design pressure of 2485 psig in accordance with section NB-751 1. The 3 safety valves provide excess capacity (Figure 2) and are backed up independently by the power operated relief valves and pressurizer spray. The pressurizer safety valves and power operated relief valves discharge to the pressurizer relief tank

(PRT). Rupture disks are installed on the pressurizer relief tank to prevent PRT overpressurization.

There are five (5) main steam safety valves per steam generator each with a capacity of 220 lb/sec at their respective opening pressure plus 3%

accumulation. The nominal lift setpoints range from 1185 psig to 1224 psig. The main steam safety valves discharge to the atmosphere outside containment.

This report demonstrates that these capacities are adequate to maintain the peak primary and secondary pressures below 110% of their respective design pressures. Neither the primary nor secondary safety valves can be bypassed or isolated.

Figure 1 shows a schematic arrangement of the pressure relieving devices.

3.0 Sizing of Pressurizer Safety Valves 3.1 Pressurizer safety valve sizing calculations are discussed in Reference 2. The Reference 2 analyses are based on analysis of a complete loss of steam flow to the turbine with the reactor operating at 102% of Engineered Safeguards Design Power. In the analysis, feedwater flow is assumed to be maintained, and no credit is taken for operation of pressurizer power operated relief valves, pressurizer level control system, pressurizer spray system, rod control system, steam dump system or steam line power operated relief valves. The reactor is maintained at full power (no credit for reactor trip), and steam relief through the steam generator safety valves is considered. The total pressurizer safety valve capacity is required to be at least as large as the maximum surge rate into the pressurizer during this transient.

The sizing procedure results in a safety valve capacity well in excess of the capacity required to prevent exceeding 110% of system design pressure for the events listed in Section 2.1. The conservative nature of this sizing procedure is demonstrated in the following section.

3.2 Each of the overpressure transients listed in Section 2.1 has been analyzed and reported in the Final Safety Analysis Report. The analysis methods, computer codes, plant initial conditions and relevant assumptions are discussed in the FSAR for each transient.

Review of these transients shows that the Turbine Trip results in the maximum system pressure and the maximum safety valve relief requirements.

This transient is presented in detail below.

For a turbine trip event, the reactor would be tripped directly (unless below approximately 50 percent power) from a signal derived from the turbine stop

emergency trip fluid pressure and turbine stop valves. The turbine stop valves close rapidly (typically 0.1 seconds) on loss of trip fluid pressure actuated by one of a number of possible turbine trip signals. This will cause a sudden reduction in steam flow, resulting in an increase in pressure and temperature in the steam generator shell. As a result, heat transfer rate in the steam generator is reduced, causing the reactor coolant temperature to rise, which in turn causes coolant expansion, pressurizer insurge, and RCS pressure rise.

The automatic steam dump system would normally accommodate the excess steam generation. Reactor coolant temperature and pressure do not significantly increase if the steam dump system and pressurizer pressure control system are functioning properly. If the turbine condenser were not available, the excess steam generation would be dumped to the atmosphere and main feedwater flow would be lost. For this situation feedwater flow would be maintained by the Auxiliary Feedwater System to ensure adequate residual and decay heat removal capability. Should the steam dump system fail to operate, the steam generator safety valves will lift to provide pressure control.

In this analysis, the behavior of the unit is evaluated for a complete loss of steam load from 102 percent of full power without direct reactor trip; that is, the turbine is assumed to trip without actuating all the sensors for reactor trip on the turbine stop valves. The assumption delays reactor trip until conditions in the RCS result in a trip due to other signals. Thus, the analysis assumes a worst transient. In addition, no credit is taken for steam dump. Main feedwater flow is terminated at the time of turbine trip, with no credit taken for auxiliary feedwater to mitigate the consequences of the transient.

The turbine trip transients are analyzed by employing the detailed digital computer program LOFTRAN. The program has the capability to simulate the neutron kinetics, RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, steam generator, and steam generator safety valves. User input allows for the availability of the valves and control systems to operate as appropriate for a particular analysis. The program computes pertinent plant variables including temperatures, pressures, and power level.

Major assumptions are summarized below:

Initial operating conditions

a. The initial reactor power and RCS temperatures are assumed at their maximum values consistent with the steady state full power operation including allowances for calibration and instrument errors. The initial RCS pressure is assumed at a minimum value consistent with the steady state full power operation including allowances for calibration and instrument errors. This results in the maximum power difference for the load loss, and

the minimum margin to core protection limits at the initiation of the accident.

b. Moderator and Doppler coefficients of reactivity The analysis assumes both a least negative moderator coefficient and a least negative Doppler power coefficient, as this results in maximum pressure relieving requirements.

c. Reactor control From the standpoint of the maximum pressures attained it is conservative to assume that the reactor is in manual control. If the reactor were in automatic control, the control rod banks would move prior to trip and reduce the severity of the transient.

d. Steam release No credit is taken for the operation of the steam dump system or steam generator power operated relief valves. The steam generator pressure rises to the safety valve setpoint where steam release through safety valves limits secondary steam pressure at the setpoint value.

e. Pressurizer spray, power operated relief valves and safety valves No credit is taken for the effect of pressurizer spray and power operated relief valves in reducing or limiting the coolant pressure. Safety valves are operable. The pressurizer safety valves are assumed to lift at 2575 psia and be full open at 2580 psia.

f Feedwater flow Main feedwater flow to the steam generators is assumed to be lost at the time of turbine trip. No credit is taken for auxiliary feedwater flow since a stabilized plant condition will be reached before auxiliary feedwater initiation is normally assumed to occur; however, the auxiliary feedwater pumps would be expected to start on a trip of the main feedwater pumps.

The auxiliary feedwater flow would remove core decay heat following plant stabilization.

g. Reactor trip Reactor trip is actuated by the first Reactor Protection System trip setpoint reached with no credit taken for the direct reactor trip on the turbine trip.

Trip signals are expected due to high pressurizer pressure,

overtemperature AT, high pressurizer water level, and low-low steam generator water level.

The results of the Turbine Trip transient are shown in Figures 2 and 3.

Figure 2 shows the pressurizer pressure, the reactor coolant pump discharge pressure, which is the point of highest pressure in the RCS, and the pressurizer safety valve relief rate. Figure 3 shows steam generator shell side pressure, reactor coolant loop hot leg and cold leg temperature, and nuclear power. The reactor is tripped on a high pressurizer pressure signal for this transient.

The results of this analysis show that the overpressure protection provided is sufficient to maintain peak RCS pressure below the code limit of 110%

of system design pressure. The plot of pressurizer safety valve relief rate also shows that adequate overpressure protection for this limiting event is provided by the three installed safety valves.

4.0 References

1. ASME Boiler and Pressure Vessel Code,Section III, Article NB-7000, 1971 Edition Winter 1972 Addenda.

2. Topical Report - Overpressure Protection for Westinghouse Pressurized Water Reactors, WCAP-7769, Rev. 1, June 1972.

3. Pressurizer Safety Valves ASME Boiler and Pressure Vessel Code,Section III Class 1, Equipment Specification No. G-678838, Rev. 2, October 19, 1977.

4. Watts Bar Units 1 and 2 (WAT/WBT) Loss of Load / Turbine Trip Analysis for the 10% Steam Generator Tube Plugging Program, Calculation No.

CN-TA-96-125, Rev. 0, December 6, 1996.

5. Watts Bar Unit 2 (WBT) Completion Program Evaluation, Calculation No.

CN-TA-09-73, Rev. 1, November 17, 2009.

W fY VAINYvI M~WN StAUNU iAWyV VALMI VLtIV 14 tu~d Figure 1 Schematic Arrangement of Pressure Relieving Devices

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Figure 3

ENCLOSURE 3 Curves for Response to RAI 5.2.2 - 1.d Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391

,A E

0 C-0 CD ca- 0 100 200 300 400 500 600 700 Time (seconds)

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WATTS BAR NUCLEAR PLANT TRANSIENT RESPONSE TO STEAM LINE BREAK WITHOUT OFFSITE POWER NUCLEAR POWER AND REACTIVITY VERSUS TIME

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WATTS BAR NUCLEAR PLANT TRANSIENT RESPONSE TO STEAM LINE BREAK WITHOUT OFFSITE POWER CORE AVERAGE TEMPERATURE AND RCS PRESSURE VERSUS TIME

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. WATTS BAR NUCLEAR PLANT TRANSIENT RESPONSE TO STEAM LINE BREAK WITHOUT OFFSITE POWER FAULTED LOOP STEAM FLOW VERSUS TIME

ENCLOSURE 4 List of New Regulatory Commitments Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391

1. Amendment 101 to the Unit 2 FSAR will implement changes as noted in responses to the following RAIs: 2.3.2 - 9, 2.3.2 - 10, 2.3.2 - 11,2.3.2 1,2.3.2 2, 2.3.2 3, 2.3.2 4, 2.3.2 5, 2.3.2 6, 2.3.2 7, 2.3.2 8, 11 - 2.a, and 11 -

2.c.

2. As noted in the response to RAI 5.2.2 - 1 .b, TVA will consider the information in RIS 2005-29 and notify the NRC of our plan of action for resolution of this concern by November 1, 2010.