ML102290258

Final Safety Analysis Report - Response to Preliminary Requests for Additional Information and Requests for Additional Information

ML102290258
Person / Time
Site:	Watts Bar
Issue date:	07/31/2010
From:	Bajestani M Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML102290258 (141)

Text

Withhold from Public Disclosure Under 10 CFR 2.390 Tennessee Valley Authority, Post OfficeOffice Box 2000, Spring 37381-2000 Spring City, Tennessee 37381-2000 31, 2010 July 31, 2010 U.S. Nuclear Nuclear Regulatory Regulatory Commission ATTN: Document Control Desk Mail Stop: OWFN OWFN P1-35 Washington, D.C. 20555-0001 Watts Bar Nuclear Plant,' Unit 2 NRC Docket Docket No. 50-391

Subject:

Watts Bar Nuclear Nuclear Plant (WBN) Unit 2 - Final Safety Analysis Report (FSAR) - Response Preliminary Requests for Additional Response to Preliminary Additional Information Information and Requests For Additional Information Information This letter letter responds to a number of both preliminary requests for additional information information (RAIs)

(RAls)

RAIs regarding the Unit 2 Final Safety and RAls Safety Analysis Report Report (FSAR). provides the response to preliminary RAIs and RAIs involving multiple FSAR provides the response to preliminary RAls and RAls involving multiple FSAR chapters.

Enclosure provides the new commitments commitments contained in this letter.

The electronic electronic files of documents provided by the response documents noted as being provided response to specific RAIs RAls are contained contained on the enclosed Optical Storage Media (OSM).

(OSM). Enclosure 3 lists the electronic files files and the file sizes.

Attachments to Enclosure Enclosure 3 contain information information proprietary proprietary to various companies companies as denoted denoted in specific Attachments. TVA requests requests that this vendor vendor proprietary proprietary information be withheld from public disclosure disclosure in accordance accordance with 10 CFR § 2.390.

Nuclear Regulatory Commission U.S. Nuclear Page 2 July 31, 2010 31, 2010 I declare under penalty of perjury that the foregoing is true and correct. Executed on I declare under penalty of perjury that the foregoing is true and correct. Executed on 31tmday of July, 2010.

the 31 If you have any questions, please contact Bill Crouch at (423) 365-2004.

If Sincerely,

'Masoud estani

'estani Watts Pa nit 2 Vice President

Enclosures:

1. Response to Preliminary Additional Infor~ation Preliminary Requests for Additional Information

~.

2. Commitments List of Regulatory Commitments

3. List of Files Provided Provided on Enclosed Optical Storage Media (OSM) .

cc (Enclosures):

U. S. Nuclear Regulatory U. Regulatory Commission Commission Region II II Marquis One Tower Tower Peachtree Center Ave.,

245 Peachtree Ave., NE Suite 1200 Suite 1200 Atlanta, Georgia 30303-1257 30303-1257 Resident Inspector NRC Resident Inspector Unit 2 Watts Bar Nuclear Nuclear Plant Plant Nuclear Plant Road 1260 Nuclear Spring City, Tennessee Tennessee 37381

u.s.

U.S. Nuclear Regulatory Commission Page 3 July July31,2010 31, 2010 bcc (Enclosures):

Lakshminarasimh Raghavan Lakshminarasimh U.S. Nuclear Regulatory Commission MS 08H4A One White Flint North North 11555 Rockville 11555 Rockville Pike Pike Rockville, Maryland Maryland 20852-2738 20852-2738 Stephen Campbell Nuclear Regulatory Commission U.S. Nuclear Commission MS 08H4A One White Flint North 11555 Rockville 11555 Rockville Pike Pike Rockville, Maryland 20852-2738 Maryland 20852-2738 Patrick D. Milano, Milano, Senior Project Manager Project Manager Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission 08H4 MS 08H4 One White White Flint North 11555 Rockville Rockville Pike Pike Rockville, Maryland 20852-2738 20852-2738 Loren R. Plisco, Deputy Regional Administrator for Construction U. S. Nuclear Nuclear Regulatory Regulatory Commission Region Region IIII Marquis Marquis One Tower Tower Peachtree Center 245 Peachtree Center Ave., 1200 Ave., NE Suite 1200 Georgia 30303-1257 Atlanta, Georgia

U.S. Nuclear Regulatory Commission Nuclear Regulatory Page 4 31,2010 July 31, 2010 WDC:TLE:CLH:DLB WDC:TLE:CLH:DLB bcc (Enclosures):

Lakshminarasimh Raghavan Lakshminarasimh Commission U.S. Nuclear Regulatory Commission MS 08H4A One White Flint North North 11555 11555 Rockville Pike Pike Rockville, Maryland Maryland 20852-2738 20852-2738 Stephen Campbell Nuclear Regulatory U.S. Nuclear Regulatory Commission Commission' MS 08H4A One White White Flint North Rockville Pike 11555 Rockville Pike Maryland 20852-2738 Rockville, Maryland 20852-2738 Patrick D. Milano, Milano, Senior Project Manager Project Manager U.S. Nuclear Nuclear Regulatory Regulatory Commission Commission MS 08H4 08H4 North One White Flint North 11555 Rockville Pike Rockville Pike Maryland 20852-2738 Rockville, Maryland 20852-2738 Loren R. Plisco, Deputy Regional Regional Administrator Administrator for Construction Construction U. S. Nuclear U. Commission Nuclear Regulatory Commission Region IIII Marquis One Tower 245 Peachtree Peachtree Center Ave., 1200 Ave., NE Suite 1200 Georgia 30303-1257 Atlanta, Georgia G. P. Arent, LP 5A-C*

M. Bajestani, EQB 11B-WBN*

M. B-WBN*

R. R. Baron, EQB 1B-WBN*

A. S. Bhatnagar, LP 6A-C*

M. K.

M. K. Brandon, ADM 1 1L-WBN*

L-WBN*

W. D. Crouch, EQB 11B-WBN*

B-WBN*

D. E. Grissette, ADM 1V-WBN*

1V-WBN*

R. M.

M. Krich, LP 3R-C*

A. L.

L. Sterdis, LP 5A-C*

E. J. Vigluicci, WT 6A-K*

K.

K. W. Whittenburg, SP 2B-C*

EDMS, WT 3B-K (wi (w/ Enclosures)

• These CCs did not receive

• These CCs did not receive the the attached attached documents.

documents. The The attached attached documents documents can be can be obtained by contacting the WBN Unit 2 Licensing Licensing office.

ENCLOSUREI 1 ENCLOSURE

Response

Response to Preliminary RAIS and RAIs Regarding Unit 2 FSAR RAls Regarding Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - UnitUnit 2, Docket No. 50-391 Preliminary RAIs RAls (taken from NRC phone call of 05/18):

3.5.1.3.6 3.5.1.3.6 - 1. In section 3.5.1.3.6, a sentence Insection sentence was deleted that started like "This "This analysis analysis...."

The deletion of a reference reference to an analysis analysis seems seems to the NRC NRC to be a technical change. The change change was not reflected Enclosure 1 as a technical change, reflected in Enclosure and the NRC questioning why it would not have NRC is questioning have been included.

included.

Response

Response: This item was previously addressed addressed as the response to RAI 3.5.1-2 3.5.1-2 on page E1-7 El-7 of TVA letter to NRC dated 3, 2010 (ADAMS Accession No. ML June 3,2010 MLI1016004770) 016004770) as follows:

"2. On page 3.5.22, second sentence of the first paragraph paragraph stated:

stated: "In"In this analysis, only Unit 2 containment and the containment the control building appear to be prominent prominent enough to be be threatened.

threatened.

Please discuss how these threats will be addressed addressed and indicate what section section of the FSAR addresses addresses this issue?

Response: Amendment 98 deleted the sentence sentence since itit was was intermediate issue and not the conclusion an intermediate conclusion of the analysis."

Prior to submittal of Amendment 98 to the Unit 2 FSAR, it was was decided by a telecom with the NRC reviewer decided reviewer that this sentence sentence would be removed since it created created confusion. TVA chose to delete delete itit in Amendment Amendment 98 as an editorial change (The conclusions conclusions of the analysis remain as shown in the FSAR.), and chose chose not to flag it it as a change in the cover letter since the the explanation explanation was going to be in the RAI response response noted above.

3.2-2 3.2 2. The following questions refer to Table 3.2-2:

a. On page 3.2-27, an asterisk was eliminated eliminated from the reference reference to TVA C. Was this a technical Class C. change or the deletion of an unused asterisk?

technical change asterisk?

Response: Amendment Amendment 98 to the Unit 2 FSAR replaced "TVA Class C*" C*"

with "TVA Class C" C" in Note Note (18)

(18) for Table 3.2-2.

This change implemented to make change was implemented make this designation designation consistent consistent with the associated associated TVA Design Criteria Criteria and the the definition definition of vendor-supplied vendor-supplied equipment as defined defined in Note Note 1) of Table Table 3.2-4. Thus, this was considered considered to be an editorial change.

change.

E1-1 E1-1

ENCLOSURE 1 ENCLOSURE Response to Preliminary Response Preliminary RAIS and RAIs Regarding Unit 2 FSAR RAls Regarding FSAR Tennessee Authority - Watts Bar Nuclear Tennessee Valley Authority Nuclear Plant - Unit 2, Docket No. 50-391

b. The description reference in the safety description for "Note 22" says itit is not used, but the reference class column for the Unit I1 PWST says changed to "Note 25."

Response: Amendment 100 Amendment 100 to the Unit 2 FSAR re-instate Note 22 FSAR will re-instate which applies applies only to the Unit 1 I Primary Primary Water Water Storage Tank.

Note 26 applies to the Unit 2 Primary Primary Water Storage Storage Tank. Two Two separate separate notes are required to identify identify the differences differences between the Unit 1 and Unit 2 tanks.

E1-2 E1-2

ENCLOSURE ENCLOSURE 1 Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Regarding Unit 2 FSAR Authority - Watts Bar Nuclear Tennessee Valley Authority Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 RAIs for FSAR 3.10 (taken from e-mail from NRC dated 05/04/2010)

Preliminary RAls Preliminary 3.10 Section 3.10 3.10 - 1. Reference Reference is made to IEEE Tables 3.10-1 and 3.10-2. However, IEEE 344-1987 in Tables 344-1987 is not mentioned in Section 3.10.1, IEEE 344-1987 3.10.1, "Seismic Qualification Qualification Criteria."

Nor is this discussed discussed in the referenced referenced Section 3.7.3.16.

Clarify IEEE-1 987 is used in aa similar manner to how you discuss the use of Clarify how IEEE-1987 IEEE 344-1971 and IEEE IEEE 344-1975.

Response: Amendment Amendment 98 to the Unit 2 FSAR deleted equipment heading deleted equipment heading "PAMS Cabinet Components and Main Control Room Cabinet and Components Components" and its reference to IEEE 344-1987 from Table 3.10-1. 3.10-1.

(Qualification Of Instrumentation Table 3.10-2 (Qualification Instrumentation And Control Equipment) lists Qualification Qualification Method Method 9 as IEEE 344-1987.

Qualification Method Qualification incorrectly added at Amendment 95.

Method 9 was incorrectly reference to it in the Table 3.10-2 There is no reference 3.10-2 equipment equipment listing inin Amendment 95 or any subsequent subsequent amendment.

amendment.

Amendment 100 to the Unit 2 FSAR will delete Qualification Amendment Qualification Method 9, IEEE 344-1987 344-1987 from Table 3.10-2.

3.10 -- 2. Table 3.10.1, Equipment," in WBN-2 FSAR Instrumentation and Electrical Equipment,"

3.10.1, "WBNP Instrumentation Section 3.10 contains three new rows related to certain equipment equipment and their their qualification methods and test methods. The first new row in Table qualification Table 3.10.1 states 3.10.1 states that the "Control Instrument Loops" (Unit (Unit 2) located at "multiple locations" were were qualified by "Test" qualified "Test" using "multiaxis" Qualification "multiaxis" test method performed by "Nuclear Qualification Services."

Clarify ififthe "Test" "Test" method and the "Test""Test" results were reviewed reviewed by the NRC staffstaff and provide aa reference conclusion. IfIf they were not reference that documents the review conclusion.

reviewed by the NRC staff, submit the results of the test for the staffs review.

reviewed The second new row in Table 3.10.1 states that "Panels 2-L-1 2-L-11A1A and 2-L-11B" 2-L-1 1B" were qualified qualified by "Analysis."

"Analysis."

1) mentioned in the second Clarify ififthe Analysis mentioned second new row in table 3.10.1 was was performed in-house by the TVA staff; if not, complete the Table 3.10.1 giving giving the name of the company, which performed performed the Analysis.

2) "Analysis" method and the Analysis results were reviewed by Also, clarify ifif the "Analysis" the NRC staff and provide a reference documents the review conclusion.

reference that documents If If they were not reviewed reviewed by the NRC staff, submit the results of Analysis for the staffs review.

3) The third new row in Table 3.10.1 3.10.1 states that the qualification the qualification method for the equipment (PAMS Cabinet and Components and Main Control Room Components) is "Analysis (to be performed)."

E1-3 E1-3

ENCLOSURE 1 ENCLOSURE Response to Preliminary RAIS and RAIs Regarding Unit 2 FSAR RAls Regarding Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 performed, submit the results Provide a target date when this analysis will be performed, of the analysis for the staff's review, and amend the FSAR as needed.

Response

Response: Instruments Foxboro Spec 200 Instruments TVA does not does not know if if the test results have been reviewed reviewed by the NRC staff. This hardware is widely used in multiple nuclear multiple nuclear facilities and may have been reviewed previously. To ensure compliance, the requested documentation is provided as requested documentation as Attachment 1.

1) and 2) second new row in Table 3.10.1 states that "Panels The second 2-L-11A 2-L-111B" were qualified by "Analysis." The 2-L-1 1A and 2-L-1 The performed in house, and the analysis analysis is being performed analysis reviewed by the NRC staff. The results have not been reviewe.d The analysis will be submitted to the NRC by by November November 30, 2010.

3) Amendment 99 to the Unit 2 FSAR removed this item Amendment item 3.10-1. Due to hardware from Table 3.10-1. hardware changes, the the qualification will be by analysis and testing. The vendor qualification vendor is scheduled scheduled to provide documentation to TVA provide this documentation December December 27, 2010, and itit will be submitted to the NRC 27,2010, NRC 2011.

January 14, 2011.

by January 3.10 - 3. locations in FSAR In several locations In 3.10-11, 3.10-12, and FSAR Section 3.10 (e.g., pages 3.10-11,3.10-12, 3.10-18), the word "LATER" is inserted before before a Reference Reference or a report.

If this word LATER refers to future action, provide If provide a target date to provide these qualification Tests / Analysis included reports and the results of the qualification reports included in these reports for the staff's staffs review.

Response: The word LATER is used for the following references:

(26) Westinghouse seismic Westinghouse qualification report for installing seismic qualification installing hardware in Unit 2 NIS cabinets.

Metrics hardware Gamma Metrics EQ-EV-39-WBT, Revision This item is EQ-EV-39-WBT, Revision 1 (Seismic Evaluation Evaluation Of Nuclear Instrumentation System Console Nuclear Instrumentation Console 2-M-13 With Gammametrics Equipment Gammametrics Equipment For Watts Bar Unit 2, proprietary version of this document is March 2009). The proprietary provided as Attachment 2. A non-proprietary non-proprietary version and affidavit for withholding will be provided by November affidavit November 30, 2010. Amendment 100 100 to the Unit 2 FSAR will reflect this this information.

information.

E1-4 E1-4

ENCLOSURE11 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 (29) Ametek qualification report for containment Ametek seismic qualification containment pressure pressure transmitters.

This item is vendor document document number Report No. TR-1136 TR-1 136 (Qualification Documentation (Qualification Documentation Review PackagePackage For Ametek Aerospace Aerospace Gulton-Statham Gulton-Statham Products Nuclear Qualified Products Nuclear Qualified Pressure Transmitter Series Enveloping --- Gage Pressure Transmitter Gage Pressure Pressure Differential Pressure Transmitter Series PG 3200, Differential Transmitter Series Series POPD 3200, Differential Differential High Pressure Transmitter Series PDH 3200, Draft Range Pressure Pressure Transmitter Series Series DR 3200, Remote Remote Diaphragm Diaphragm Seal Differential Differential Pressure Transmitter Series PO 3218, Remote Remote Diaphragm Differential High Pressure Diaphragm Seal Differential Pressure Transmitter Transmitter Series PDH 3218). The proprietary proprietary version of the document document is provided as Attachment 3. A non-proprietary non-proprietary version and affidavit affidavit for withholding withholding will be provided by December December 17, 17, 2010. Amendment Amendment 100 100 to the Unit 22 FSAR will reflect this information.

information.

(30) Qualification of Weed Pressure Transmitter.

Seismic Qualification This item is vendor document number number 16690-QTR, 16690-QTR, Revision 0 (Qualification Test Report For Environmental And And Seismic Qualification Seismic Qualification Of Weed Model Model DTN2010 DTN201 0 Pressure proprietary version of this document Transmitters). The proprietary document is provided as Attachment non-proprietary version Attachment 4. A non-proprietary version and affidavit for withholding will be provided provided by November 30, 2010. Amendment November Amendment 100 to the Unit 2 FSAR FSAR will reflect this will reflect this information.

information.

3.10 - 4. The numbering numbering of the Unit 2 list on page page 3.10-4 is not consistent with the the referenced by the text below the list.

numbering referenced

1) Correct the numbering numbering to clearly identify identify the references references associated the associated with the items in the list.

Response: Amendment Amendment 98 to the Unit 22 FSAR corrected the numbering of the list. Since these were editorial changes, the amendment amendment level remained the same.

E1-5 E1-5

ENCLOSURE 1 ENCLOSURE Response to Preliminary RAIS and RAIs Response RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant Plant - Unit 2, Docket No. No. 50-391

2) The Nuclear Nuclear Instrumentation Instrumentation System Power Range Electronics appears to be a Electronics appears a new item added to the list for Unit 2. ClarifyClarify which reference documents its its qualification qualification testing. Provide the results of the test or analysis.

Response

Response: Amendment 98 to the Unit 2 FSAR corrected the numbering of Amendment the list. Since Since these were editorial changes, the amendment amendment level remained the same.

There is no change to the Power Range Electronics.Electronics. The The electronics are electronics are the same and use the same same qualification qualification documents as Unit 1. As aa result, no qualification documents qualification documents documents are submitted submitted for the electronics.

Amendment 95 to the Unit 2 FSAR added the Nuclear Nuclear Instrumentation Instrumentation System Power Power Range Range Electronics. This was was done to differentiate differentiate the qualification qualification of the cabinets cabinets from the the electronics. In Unit 2, Westinghouse Unit Westinghouse updated updated the cabinet cabinet installation of the Gamma qualification to support the installation qualification Gamma MetricsMetrics hardware. In Unit 1, the cabinet qualification qualification analysis was done done by TVA. Having Westinghouse Westinghouse perform perform the Unit 2 analysis analysis resulted resulted in Reference Reference 26 being added to the referencereference section.

Reference 26 was inadvertently However, Reference inadvertently omitted from the the 3.10.1 text discussion of the equipment 3.10.1 equipment qualified by by Westinghouse.

Westinghouse.

Amendment Amendment 100 to the Unit 2 FSAR FSAR will update update the FSARFSAR wording wording as shown below:

qualification testing/analysis "Seismic qualification testing/analysis of Items 1 I through through 99 is documented in References documented References [1] through [10] and [26]. [261. Reference Reference

[10] presents the theory and practice, as well as justification, justification, for the use of single axis sine beat test inputs used in the seismic seismic qualification of electrical equipment. In addition, it is noted that qualification Westinghouse Westinghouse has conducted conducted aa seismic qualification qualification "Demonstration Test Program" (reference(reference Letter Letter NS-CE-692, NS-CE-692, C.

Eicheldinger (W), to D. B. Vassallo (NRC), 7/10/75) to confirm Eicheldinger equipment operability operability during during a seismic seismic event. This program is documented documented in References References [12] through [14] (Proprietary)

(Proprietary) and References [16] through [19] (Non-Proprietary).

References (Non-Proprietary). Seismic Seismic qualification testing of Item 10 qualification 10 to IEEE 344-1975 is documented documented in References References [21],

[21], [22], [23],

[23], [31] and [32]. Reference Reference [26]

documents the Westinghouse Westinghouse qualification qualification by analysis of the the Nuclear Instrumentation Nuclear Instrumentation System cabinetcabinet 2-M-13 2-M-13 with Gamma Gamma Metrics Metrics Source Source and Intermediate Intermediate Range hardware hardware installed."

E1-6 E1-6

ENCLOSURE 1 ENCLOSURE I

Response

Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Authority - Watts Bar Nuclear Plant - Unit 2, Docket Tennessee Valley Authority Tennessee Docket No. 50-391 The proprietary Reference 26, Westinghouse proprietary version of Reference Westinghouse report EQ-EV-39-WBT is provided as Attachment 2. The EQ-EV-39-WBT The non-proprietary version and the affidavit for withholding will be non-proprietary be submitted November 30, 2010.

submitted by November 3.10 - 5. Page No. 3.10-4 in several new in FSAR Section 3.10.1 lists several instrumentation new items of instrumentation and electrical equipment requiring seismic electrical equipment seismic qualification. 3.10-6 qualification. ItIt is stated on page 3.10-6 of Section Section 3.10.1 that seismic qualification testing of items 11 and 12 is seismic qualification documented in a new Reference documented Reference 25 for the staff's Reference 25. Provide a copy of Reference staff's review.

Response: The proprietary proprietary version Qualification version of Thermo Fisher Scientific Qualification Report No. 864, Rev. 0 (Class 1E Qualification Qualification of Source Source Range, Range and Wide Intermediate Range Intermediate Wide Range provided as Range Channels) is provided proprietary version contains some Attachment 5. Note that this proprietary some marked-up marked-up editorial corrections for cross reference information.

information. A A corrected non-proprietary version, and an corrected proprietary version, the non-proprietary affidavit for withholding will be provided by November affidavit 15, 2010.

November 15,2010.

E1-7 E1-7

ENCLOSUREI 1 ENCLOSURE Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar NuclearNuclear Plant - Unit Unit 2, Docket No. 50-391 Preliminary Preliminary RAls RAIs for FSARFSAR Sections 5.2.3 5.2.3 through 6.1.1 6.1.1 (taken from e-mail from NRCNRC dated 05/13/2010) 05/13/2010) 5.2.3-1 5.2.3-1 Background

Background

In FSAR FSAR Section Section 5.2.3.4, Chemistry Chemistry of Reactor Coolant, the applicant applicant added aa description of the process description process for zinc addition. The applicant applicant indicated that zinc would would be added for the purpose of reducing radionuclide radionuclide content in the primary system corrosion films, and that the residual zinc corrosion zinc content content would be maintained at a a concentration of 2-8 parts per billion (ppb). FSAR Table 5.2-10 also indicates zinc concentration zinc will be limited to less than 40 ppb during normal normal power operation.

The staff reviewed several reports documenting documenting industry experience experience with zinc zinc addition addition in Pressurized Pressurized Water Water Reactors Reactors (PWR's), which indicate that there is no no concern with crud deposition for plantsplants with low-duty low-duty or medium-duty cores cores (Reference 1, 2), and, in fact, zinc addition typically (Reference typically leads to thinner, more evenly distributed crud on fuel. However, there is currently insufficient distributed operating insufficient operating experience experience with zinc addition in plants plants with high-duty high-duty cores to be able to conclude conclude injection would that zinc injection would not cause aa problem problem with crud deposition in such plants.

Core duty is a measure measure of the amount of subcooled nucleate nucleate boiling (SNB) occurring in the core. Plants with high-duty cores cores are those with high fluid fluid temperatures temperatures and high surface heat flux at the fuel clad causing causing a portion of the the total heat transfer to the coolant to occuroccur by SNB. Although favorable for thermal efficiency, the combination temperature and SNB leads to more surface combination of high temperature boiling, which is known to enhance enhance the formation formation of corrosion product product deposits deposits (crud) at the cladding surface. The tendency for SNB can be quantified by means means of the High Duty Core Index Index (HOCI),

(HDCI), calculated calculated in accordance accordance with Appendix F of Reference Reference 3. Cores with an HOCI HDCI of ~> 150 150 are considered considered to be high duty plants, medium medium duty duty plants have HDCI 120-149, and a plant with HOCI HOCI of 120-149, HDCI <~ 119 is considered considered a low-duty plant. Staff calculations based based on thermal-hydraulic thermal-hydraulic data from FSAR Chapter Chapter 4 indicate the WBNP-2 core may be considered high-duty.

There There may be alternate determine the amount alternate methods to determine amount of SNB other the HOCI, HDCI, such as detailed thermal thermal hydraulic hydraulic computer computer models.

models.

Potential problems with crud deposition deposition could include excessively excessively thick fuel crud, crud,oror uneven crud thickness thickness that could lead to crud induced induced power shift (CIPS), also known known as axial offset anomaly. Reference Reference 2 also indicates that fuel clad corrosioncorrosion cannot be completely ruled out for high-duty high-duty cores exposed exposed to zinc addition even though no problems have been been observed Reference 2 recommends observed to date. Reference recommends a fuel surveillance surveillance program program for high-duty high-duty plants implementing implementing zinc addition.

addition.

Information Requested Information

1. Is the WBNP core design considered considered a high-duty high-duty core when when the HOCI HDCI is calculated in accordance accordance with Appendix F of Reference F Reference 3, or an alternate alternate method of evaluation?

evaluation?

E1-8 E1-8

ENCLOSUREI ENCLOSURE 1

Response

Response to Preliminary RAIS and RAIs RAls Regarding Unit 2 FSAR FSAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket No. No. 50-391 Response: Westinghouse does not recommend the use of HDCI measure HOCI as a measure of plant or core "duty." formulation is overly sensitive to outlet "duty." The formulation tem peratu re.

temperature.

Using the bounding values listed in the FSAR, Watts Bar Unit 2 Using HOCI falls into the "High Outy HDCI Duty Core" category. Appendix Appendix F of Reference 3 suggests using best estimate Reference estimate values HDCI.

values to calculate HOCI.

When best estimate power, flow, and outlet temperatures are used HOCI calculation, the resulting HOCI in the HDCI HDCI for Watts Bar Unit 2, Cycle 1 is in the "Medium Duty Outy Core" category.

Requested Requested Information Information

2. If If the WBNP core is considered high-duty, describe describe the measures measures to be taken to to increase the risk of CIPS and or clad ensure that zinc addition does not increase clad corrosion, and that the overall corrosion, minimized.

overall risk of adverse fuel effects is minimized.

Possible measures could include, but are not limited to:

surveillance program monitoring Implementation of a fuel surveillance

a. Implementation monitoring crud buildup buildup and clad corrosion;

b. Additional Additional chemistry monitoring monitoring

c. Application Application of operating operating experience experience with similar core designs.

Response

Response: Westinghouse substantial fuel operating Westinghouse has substantial operating experience experience with zinczinc addition in low, medium, and high duty cores. The industry addition experience experience with zinc addition in high high duty cores has increased increased substantially since Reference substantially Reference 22 was issued.

issued. At Watts Bar, Unit 1 has already operated for two cycles with zinc injection already operated the injection using the same fuel type as will be used in Unit 2 and at aa slightly higher power level.

As zinc addition has been applied outside applied to plants with boiling duty outside the prior fuel operating experience base for zinc addition, operating experience Westinghouse undertaken fuel examinations Westinghouse has undertaken examinations including clad corrosion measurements.

measurements. These measurements measurements have shown no no increase corrosion resulting from zinc addition and increase in cladding corrosion and include high duty plants using zinc addition. The current operating include operating experience and fuel surveillance experience experience bounds surveillance experience bounds the duty expected for Watts Bar Unit 2. Therefore, no Watts Bar Unit 2 addition in surveillance is required to implement zinc addition specific fuel surveillance Watts Bar Unit 2 since the plant is bounded bounded by other applicable applicable fuel surveillance campaigns.

surveillance performs a fuel crud risk analysis for each Westinghouse performs Westinghouse each operating operating plant cycle and includes includes the consideration consideration of zinc addition in that risk analysis ifif zinc is used.

used. Westinghouse Westinghouse also has zinc addition guidelines guidelines that are provided to each zinc addition PWR using guidelines describe Westinghouse fuel. These guidelines Westinghouse describe the chemistry monitoring requirements monitoring requirements needed for zinc addition. In addition to E1-9 E1-9

ENCLOSUREI 1 ENCLOSURE Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 monitoring monitoring zinc concentration concentration in the coolant, corrosion product concentrations are also required to be measured.

concentrations measures measured. These measures have proven proven successful successful in avoiding avoiding any fuel performance performance issues issues associated with zinc addition associated addition in PWR cores using Westinghouse Westinghouse fuel.

5.3.1-1 5.3.1-1 Background

Background

The regulatory acceptance criteria for a reactor regulatory acceptance reactor vessel material surveillance material surveillance program are the requirements program requirements of Section III of Appendix H of 10 CFR Part 50.

Complying with the acceptance criteria satisfies the requirements of 10 10 CFR CFR Part 50, Appendix A, General General Design Criteria (GDC) 32 regarding regarding an appropriate appropriate material material surveillance surveillance program for the reactor reactor vessel. 1010 CFR Part 50 Appendix H, H, paragraph paragraph III.B.3 requires:

"A proposed withdrawal withdrawal schedule must be submitted with a technical technical justification justification as as specified in § 50.4. The proposed schedule must be approved prior to specified implementation."

Additionally, Generic Pressure Temperature Limit Generic Letter 96-03, "Relocation of the Pressure Curves Temperature Overpressure Curves and Low Temperature Overpressure Protection Protection System Limits," contains contains seven criteria that must be met for the relocation of the pressure-temperature pressure-temperature (P-T)(P-T) limits from the technical technical specifications specifications to aa pressure-temperature pressure-temperature limits report (PTLR). One of these criteria is:

"The Reactor Reactor Vessel Material Surveillance Program Material Surveillance Program shall comply with Appendix H to 10 CFR Part 50. The reactor vessel material material irradiation surveillance surveillance specimen removal schedule schedule shall be provided, provided, along along with how the specimen examinations specimen examinations shall be used to update the PTLRPTLR curves."

In Amendment In Amendment 97, FSAR Section 5.4.3.6 was modified modified to state that the tentative schedule for removal of the capsules for post-irradiation post-irradiation testing is as shown in in Table Pressure and Temperature Table 4.0-1 of the Pressure Temperature Limits Reports (PTLER). The actual proposed schedule was deleted from the FSAR. However, the PTLR proposed (WCAP-17035-NP, Reference (WCAP-17035-NP, Reference 4) does not contain this information.

information.

Requested Information Requested Information

1. Provide the proposed Table 4.0-1 which incorporates

1. incorporates the schedule for withdrawal of the surveillance surveillance capsules in the PTLR.

El-10 E1-10

ENCLOSURE 1 ENCLOSURE Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 Authority - Watts Bar Nuclear Response: following table will be added to the Unit 2 PTLR.

The following TABLE 4.0-1 Watts Bar Unit 2 Surveillance Surveillance Capsule Removal Schedule(a)

Capsule Removal Schedule(a)

Expected Capsule Expected Capsule Fluence Fluence Orientation Orientation Lead Removal (n/cm 22,E > 1.0 Capsule Capsule of Capsule Capsule Factor Time MeV)

U Dual 340 5.13 1st Refuel 0.50 x 1019 1st Refuel U Dual 34° 5.13 0.50 x 1019 Outage Outage W 34' Single 34° 5.18 5.18 6.1 EFPY 6.1 EFPY 3.17 xx 1100 19

,3.17 19(b)

(b) 6.2EFPY 3.17 x 1019 19 to 6.2 EFPY 3.17 x 10 to X Dual 340 34° 5.13 to 12.5 19(c) 6.34 x 1019 (e)

EFPy(e)

EFPY(c) 6.34X10 Z 340 Single 34° 5.18 Standby Standby -----

Dual Dual V 31.5° 4.40 Standby Standby -----

31.5° Dual Y 31.5' 4.40 Standby Standby -----

31.5° Notes:

information is taken from the withdrawal (a) This information withdrawal schedule schedule contained in WCAP-9455, Revision contained Revision 3 (Ref. 3).

(b) Approximate (b) Fluence at Approximate Fluence vessel inner at vessel inner wall at End-of-Life wall at End-of-Life (32 EFPY).

withdrawn between 6.2 EFPY and Capsule X should be withdrawn (c) Capsule corresponds to a capsule 12.5 EFPY, which corresponds f1uence of not less capsule fluence less than once (3.17 x 10 19 1019 n/cm 22 (E > 1.0 1.0 MeV)) or greater greater than twice (6.34 x 10101919 n/cm 22 (E > 1.0 MeV))

n/cm End-of-Life MeV)) the peak End-of-Life recommendations of vessel fluence. This is consistent with the recommendations vessel ASTM E185-82.

E 185-82.

included in the Unit 2 System Description The Unit 2 PTLR is included Description for the the Reactor Reactor Coolant System (WBN2-68-4001). This system description will be revised revised to reflect required revisions to the PTLR by by September 17, September 17, 2010.

El-11 E1-11

ENCLOSURE11 ENCLOSURE Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, DocketDocket No. 50-391 5.3.1-1 Background

Background

The regulatory regulatory acceptance surveillance acceptance criteria for a reactor vessel material surveillance program are the requirements Section III of Appendix H of 10 requirements of Section 10 CFR Part 50.

Complying with the acceptance acceptance criteria requirements of 10 CFR criteria satisfies the requirements General Design Criteria Part 50, Appendix A, General Criteria (GDC) 32 regarding an appropriateappropriate material surveillance program for the reactor vessel. 10 CFR Part 50 Appendix H, material surveillance H, paragraph III.B.3 requires:

paragraph "A proposed proposed withdrawal submitted with a technical justification withdrawal schedule must be submitted justification as as specified in § 50.4. The proposed schedule specified approved prior to schedule must be approved to implementation."

Additionally, Generic Letter 96-03, "Relocation of the Pressure Generic Letter Temperature Limit Pressure Temperature Curves and Low Temperature Overpressure Protection Temperature Overpressure Protection System Limits," containscontains seven criteria that must be met for the relocation of the pressure-temperature pressure-temperature (P-T) limits from the technical technical specifications pressure-temperature limits report specifications to a pressure-temperature (PTLR). One of these criteria is:

"The Reactor Material Surveillance Reactor Vessel Material Program shall comply with Appendix H Surveillance Program to 10 CFR Part 50. The reactor vessel material irradiationirradiation surveillance surveillance specimen removal schedule shall be provided, along with how the specimen examinations examinations shall be used to update the PTLR curves."

In Amendment 97, FSAR Section 5.4.3.6 was modified to state that the tentative tentative schedule schedule for removal of the capsules for post-irradiation post-irradiation testing is as shown in Table Table 4.0-1 of the Pressure and Temperature Temperature Limits Reports (PTLER).

(PTLER). The actual proposed schedule was deleted from the FSAR. However, the PTLR (WCAP-17035-NP, 17035-NP, Reference 4) does does not contain this information.

Requested Information Requested Information surveillance program in the description of the reactor vessel material surveillance

2. Include a description the PTLR, including a discussion of how the specimen examinations shall be used specimen examinations to update the P-T curves.

Response: material irradiation surveillance The reactor vessel material surveillance specimens shall examined to determine changes in be removed and examined in material properties. The results of these specimen examinations shall be specimen examinations be used to update the P-T Curves.

The pressure surveillance program pressure vessel steel surveillance program is in compliance compliance with 10 CFR 50, Appendix Appendix H, H, "Reactor Vessel Material Surveillance Surveillance Program Requirements." The material test requirements Program requirements and the the acceptance standard utilize the reference nil-ductility temperature, acceptance temperature, determined in accordance RT NOT, which is determined RTNDT, accordance with ASTM E208. The The empirical relationship between relationship between RTNDT fracture toughness of RT NOT and the fracture the reactor vessel steel is developed developed in accordance accordance with Appendix Appendix Protection Against Failure," to G, "Fracture Toughness Criteria for Protection Section XI of the ASME Boiler and Pressure Vessel Code. Code. The The E1-12 E1-12

ENCLOSURE11 ENCLOSURE Response to Preliminary RAIS and RAts RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 surveillance capsule removal removal schedule meets the requirements of ASTM E185-82.

Westinghouse provided Westinghouse provided a proposed revision of the PTLR. Per this this markup, the Reactor Material Surveillance Reactor Vessel Material Surveillance Program Program has has been updated updated as indicated indicated above. The Unit 2 PTLR is included in the Unit 2 System Description for the Reactor Coolant System (WBN2-68-4001). This system description description will be revised to reflect required revisions to the PTLR by September 17, 17, 2010.

5.3.1-2 5.3.1-2 Background

Background

10 10 CFR Part 50, Appendix Appendix H,H, Paragraph Paragraph III.B.1 requires that the design of the the surveillance program and the withdrawal surveillance withdrawal schedule schedule to meet the requirements requirements of the the edition of ASTM E E 185 that is current on the issue date of the American American Society of Mechanical Engineers (ASME) Code to which the reactor vessel was purchased, Mechanical purchased, or or later editions editions through ASTM E 185-1982.

185-1982.

FSAR FSAR Section 5.2.4 indicates indicates that changes in fracture toughness of the core region forgings, weldments weldments and associated heat affected zones (HAZ) due to radiation radiation damage will be monitored monitored by a surveillance surveillance program which is based based on ASTM E

E 185-82, (Ref. 5) and 10 10 CFR Part 50, Appendix H. FSAR Section 5.2.4 further indicates indicates that the surveillance surveillance program will be in compliance with these documents documents with the exception that all of the RV irradiation irradiation surveillance surveillance capsules will receive a neutron neutron flux which is at least 4 times the maximum RV neutron neutron flux. (i.e.,

(i.e., the lead factor for all the capsules will be at least 4).

ASTM E 185-82 recommends that the surveillance 185-82 recommends surveillance capsule lead factors (the ratio of the instantaneous instantaneous neutron neutron flux density at the specimen specimen location location to the maximum maximum calculated calculated neutron neutron flux density at the inside surface of the RV wall) be in the range range of one to three.

More recent versions of ASTM acknowledge that it ASTM E 185 acknowledge it may not be possible to position capsules in low lead factor locations locations due to the design of the RV internals.

ASTM E E 185-02 recommends recommends that plants with lead factors greater greater than five should should provide a method method of verifying the validity of the accelerated accelerated irradiation data. This This verification verification may be accomplished accomplished by the inclusion inclusion of a reference reference material.

Requested Information Requested Information

1. If surveillance capsules will have a lead factor greater than five, describe If any surveillance describe how the validity of the accelerated accelerated irradiation data will be verified.

Response

Response: Four of the six Unit 2 surveillance capsules have have lead factors greater than five. The validity of the accelerated accelerated irradiation data will be be assessed assessed by comparing the Unit 2 surveillance surveillance data to the the Regulatory Guide Regulatory Guide 1.99, Revision Revision 2 (RG (RG 1.99, Rev. 2) predictions.

The validity of the data will also be assessed assessed by comparing the the surveillance data to results for similar forging and weld Unit 2 surveillance material to ascertain that the observed trends are consistent. The The El-13 E1-13

ENCLOSURE 1I

Response

Response to Preliminary RAls RAIS and RAls RAIs Regarding Unit 2 FsAR FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 RG 1.99, Rev. 2 predictions predictions and the available available data on similar similar materials from other plants represent data irradiated under materials under a wide wide range of lead factors, so the Unit 2 surveillance surveillance data will be verified verified by trending itit against against the RG 1.99, Rev. 2 predictions predictions and the data surveillance materials.

for similar surveillance Due to the lead factors being greater than 5, all capsulescapsules will bebe removed early in life. Unit 2 will be able to store capsules capsules in the the spent fuel pool for future reinsertion to assure that the vessel continues to be monitored throughout the licensed life (and potential license extensions).

Requested Information Requested Information

2. For those surveillance surveillance capsules with lead factors greater than 3, justify that the the surveillance specimens in surveillance in the capsules will provide metallurgically meaningful provide metallurgically data, in terms of the expected expected design life and/or licensed life of the RV, including including possible license license renewal renewal terms, based on the fluences fluences these capsules capsules are projected projected to receive.

Response: Two of the six Unit 2 surveillance capsules capsules have lead factors greatergreater than three, but less than five. The validity validity of the accelerated irradiation irradiation data will be assessed comparing the Unit 2 assessed by comparing surveillance surveillance data to the Regulatory Guide 1.99, Revision 2 (RG (RG 1.99, Rev. 2) 2) predictions. The validity of the data will also be be

. assessed comparing the Unit 2 surveillance assessed by comparing surveillance data to results for similar forging and weld material material to ascertain that the observed trends are consistent. The RG 1.99, Rev. 22 predictions and the the available available data on similar materials from other represent data other plants represent data

..irradiated irradiated under a wide range of lead factors, so the Unit 2 surveillance surveillance data will be verified by trending itit against the RG 1.99, Rev. 2 predictions predictions and the data for similar surveillance surveillance materials.

Due to the all six of the Unit 22 lead factors being greater than 3, all being greater capsules capsules will be removed early early in life. Unit 2 will be able to store capsules in the spent fuel pool for future reinsertion reinsertion to assure that the vessel continues continues to be monitored throughout throughout the licensed life licensed life (and potential license extensions).

5.3.1-3 5.3.1-3 Background In Amendment 97, in FSAR Section Section 5.2.4.2, the description description of the orientation of the the Charpy V-Notch V-Notch specimens used to determine the initial initial USE of the RV beltline beltline region was changed "transverse" to "tangential and axial." Section changed from "transverse" Section 5.4.3.6 5.4.3.6 of the FSAR FSAR was also modified modified in Amendment Amendment 97 to change the description description of the the specimen specimen orientation from "longitudinal and transverse" to "tangential and axial."

E1-14 E1-14

ENCLOSURE 1 ENCLOSURE1 Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket Docket No. 50-391 ASTM E 185-82 Section Section 6.2 states:

"The "The tension and Charpy specimens from base metal shall be oriented so that the Charpy specimens the major axis of the specimen in parallel surface and normal to the principal parallel to the surface rolling direction direction for plates, or normal to the major working working direction direction for forgings asas described Section III of the ASME Code. The axis of the notch of the Charpy described in Section specimen for base metal and weld metal shall be oriented perpendicular perpendicular to the the surface of the material; for the HAZ specimens, the axis of the notch shall be as surface close to perpendicular perpendicular to the surface as possible so long as the entire length of the the notch is located within the HAZ."

notch HAZ."

The requirements of ASTM E E 185-82 with respect to specimen specimen orientation orientation are consistent with the ASME Code, Section consistent Section III, III, NB-2322.2. The Nuclear Nuclear Regulatory Commission's (NRC)

Commission's (NRC) Branch Branch Technical Technical Position 5-3, "Fracture Toughness Toughness direction", defined as "transverse to the Requirements," uses the term "weak direction", the direction of maximum direction maximum working." NRC BTP 5-3 generally generally uses the terms terms describe the weak and strong specimen "transverse" and "longitudinal" to describe "transverse" orientations.

PTLR (Ref. 4) indicates that the tangential Table B-1 in the PTLR Note (b) to Table tangential direction is the strong direction direction is the weak direction. The FSAR direction and the axial direction changes appear changes consistency with the PTLR.

appear to be for consistency Requested Information Requested Information beltline materials, clarify For the WBNP-2 RV beltline orientation of the Charpy clarify the orientation specimens for the base metal in terms of the language language used used in ASTM E 185-82.

Response

Response: direction can be described Weak direction orientation for forgings described as axial orientation forgings specimen is parallel to (transverse for plates). The major axis of the specimen to the surface surface and normal to the major major working direction. The axis of notch is oriented the notch oriented perpendicular perpendicular to the surface of the material.

material.

Strong Strong direction described as tangential orientation for direction can be described (longitudinal for plates). The major axis of the specimen forgings (longitudinal specimen is surface and parallel to the major working direction.

normal to the surface direction.

perpendicular to the surface of the The axis of the notch is oriented perpendicular the material.

material.

upper shelf energy values Data for the beltline initial upper values in Table B-1B-1 was only available for specimens specimens tested in the strong direction (hence the 65% reduction per NUREG-0800, NUREG-0800, Rev. 1). However, the the surveillance program contains Charpy specimensspecimens oriented oriented both in axial (weak) and tangential (strong) orientations so that both the axial orientations may be tested in the future.

E1-15 El-15

ENCLOSURE 1 ENCLOSURE Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. No. 50-391 5.3.1-4 Background Backqround Per 10 CFR 50.61, 50.61, the pressurized pressurized thermal shock reference reference temperature temperature (RTPTS)

(RT PTS) for the RV beltline materials materials must be calculated calculated for the end-of-license end-of-license (EOl)

(EOL) fluence. 10 10 CFR 50.61 definesdefines the EOL f1uence fluence as the best-estimate best-estimate neutron neutron projected fluence projected for a specific specific vessel beltline material material at the clad-base-metal interface on the inside surface interface surface of the vessel at the location location where where the material receives the highest fluence on the expiration date of the operating receives the highest fluence on the expiration date of the operating license. license.

In order to comply with GDC GOC 1414 and 31 related to ensuring ensuring that the RCPB will behave behave in a nonbrittle manner nonbrittle manner and that the probability of rapidly propagating propagating fracture is minimized, the RV materials materials must meet 10 CFR Part 50 Appendix G, Appendix which requires that the Upper Shelf Energy Energy (USE) remain above 50 ft-Ibs above ft-lbs through the expiration of the plant license unless it can be demonstrateddemonstrated that that lower lower values of Charpy USE will provide margins of safety against against fracture fracture equivalent to those required required by Appendix Appendix G of Section XI of the ASME Code. Regulatory Guide Guide (RG) 1.99 provides the guidance (RG) guidance on evaluating evaluating the drop in USE due to neutron irradiation. The projected projected fluence at EOl EOL must be usedused to evaluate evaluate the EOl EOL USE.

Appendixes Band B and C to the PTLR PTlR (Ref. 4) provide the projected projected USE and RTPTS RT PTS based on EOl values based EOL fluence values for 32 Effective Full Power Power Years (EFPY).

However, many many nuclear plants are now operating operating to capacity factors of 90% or or more, which could make the projected projected EOL nonconservative if EOl fluence values nonconservative WBNP-2 achieves achieves similar efficiency.

Requested Requested Information Information nuclear power plants Given that nuclear plants are now typically typically operating operating at capacity capacity factors of 90% or greater, justify that a best estimate EOl EOL neutron fluence fluence based on 32 EFPY is appropriate conservative, given that 32 EFPY is based on a capacity factor appropriate and conservative, factor 80% over of 80% over a 40-year license. If If a greater EFPY value should be postulated, postulated, provide updated Appendixes Appendixes Band B and C to the PTLR PTlR which recalculate recalculate the USE and and RT PTS values RTPTS values for the WBNP-2 RV based on new EOl EOL neutron fluence values.

neutron fluence Response: For reactors with actual operating experience, aa plant-specific plant-specific calculation is performed for fuel cycles that have been completed to provide provide aa best-estimate best-estimate fluence, and fluence projections for future operation are generated generated on an assumed mode of operation. operation. For reactors with no actual operating operating experience, such as Watts Bar Bar Unit Unit 2, design basis fluence calculations calculations are completed based on the assumption of operation with a conservative conservative out/in out/in (non-low-leakage)

(non-low-leakage) fuel loading pattern for the entire licensed licensed lifetime lifetime of the reactor. This design basis vessel fluence EOL, whether fluence at EOl, based based on the assumed assumed 80% capacitycapacity factor factor or a higher capacity factor, will be conservative conservative compared best-estimate vessel compared with best-estimate fluence analyses. Therefore, the design basis EOl EOL fluence values values are appropriately appropriately conservative and do not need to be adjusted to account account for a higher capacity factor.

El-16 E1-16

ENCLOSURE 11 ENCLOSURE Response to Response to Preliminary Preliminary RAIS RAIS and and RAIs RAls Regarding Regarding Unit Unit 22 FSAR FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket Docket No. No. 50-391 50-391 5.3.1-5 5.3.1-5 Background

Background

10CFR 10CFR 50 50 Appendix Appendix G, G, requires requires the the values values of RT NDT and of RTNDT and Charpy Charpy USE USE for for the the RVRV beltline beltline materials, including welds, plates and forgings to account for the effects of materials, including welds, plates and forgings to account for the effects of neutron radiation, including neutron radiation, including the the results results of the surveillance of the surveillance programprogram of Appendix H of Appendix H of of this this part.

part.

In In order order to to determine determine the the changes changes in in reactor reactor vessel fracture toughness vessel fracture toughness from from neutron neutron irradiation, the RV wall fluence must be determined. NRC irradiation, the RV wall fluence must be determined. NRC Regulatory Regulatory Guide 1.190, "Calculational Guide 1.190, "Calculational and Dosimetry Methods and Dosimetry Methods for for Determining Determining Pressure Pressure Vessel Neutron Fluence," describes acceptable approaches for Vessel Neutron Fluence," describes acceptable approaches for fluence fluence determinations.

determinations.

FSAR FSAR Section 5.4.3.6.2 describes Section 5.4.3.6.2 describes the the methodology methodology of the calculation of the calculation of of the the fast fast neutron flux received by the surveillance samples.

neutron flux received by the surveillance samples. In FSAR Section 5.4.3.6.2.1, In FSAR Section 5.4.3.6.2.1, "Reference "Reference Forward Forward Calculation, Calculation, "" in in Amendment Amendment 97, 97, the the applicant applicant included included new new material on material on the transport calculation.

neutron transport the neutron calculation.

Requested Requested Information Information The The first first paragraph paragraph of of FSAR FSAR Section Section 5.4.3.6.2.1 5.4.3.6.2.1 refers refers to to legacy legacy references references for for Unit 2 Unit 2 that predate the development of ENDF-BIVI, and hence are unacceptable that predate the development of ENDF-B/VI, and hence are unacceptable for for current current calculations, according to calculations, according to RGRG 1.190 recommendations. The 1.190 recommendations. The second second paragraph, however, paragraph, however, refers refers to forward transport to forward transport calculations calculations performed performed using using more more up-to-date methods for analysis of Capsule Wand subsequent capsules. Please up-to-date methods for analysis of Capsule W and subsequent capsules. Please clarify clarify the the difference difference betweenbetween thesethese two two paragraphs.

paragraphs.

Response

Response: The The first paragraph contains first paragraph discussion of contains aa discussion of the the previous previous methodology methodology used in the calculation of the reference forward used in the calculation of the reference forward design design basis basis analysis. This methodology used the BUGLE-93 crosssection analysis. This methodology used the BUGLE-93 crosssection which was library, which library, based on was based ENDF-BNI data.

on ENDF-BNI data.

It It should should be be noted noted thatthat the the reference reference forward forward design design basis basis analysis analysis was updated using the BUGLE-96 cross-section library, which was updated using the BUGLE-96 cross-section library, which is is also also based based on on ENDF-BNI ENDF-BNI data. data. The The calculated calculated design design basisbasis fluxes fluxes for these two for these two cross-section cross-section libraries libraries areare virtually virtually identical.

identical.

The The second second paragraph paragraph is is aa discussion discussion of of the the updated updated Westinghouse Westinghouse fluence methodology fluence methodology which which was used in was used in the the analysis analysis of Unit 11 of Unit Capsule Capsule W. This is the the current current Westinghouse Westinghouse methodology methodology which which will will bebe used used toto calculate calculate updated updated fluence fluence values values whenwhen surveillance surveillance capsules capsules are pulled and are pulled analyzed in and analyzed in the the future.

future.

5.3.1-6 5.3.1-6 Background

Background

With respect With respect to to special processes used special processes used to to fabricate fabricate the the RV (e.g. welding)

RV (e.g. welding)

SRP Section 5.3.1 essentially states that the requirements of GDC SRP Section 5.3.1 essentially states that the requirements of GDC 11 and and 30 30 and and 10 CFR 50.55a regarding quality standards are 10 CFR 50.55a regarding quality standards are met by compliance with the met by compliance with the provisions provisions of of the ASME Code, the ASME Code, Section Section III, III, for fabrication of for fabrication of components, components, when when the the appropriate code symbols appropriate code symbols are affixed are affixed and appropriate certifications made by appropriate certifications made by the the manufacturer manufacturer or installer.

E1-17 E1-17

ENCLOSURE ENCLOSURE 1 Response to Preliminary Response Preliminary RAlsRAIS and RAls RAIs Regarding Unit 2 FSAR FsAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket Docket No. 50-391 With respect to special methods for nondestructive nondestructive examination, examination, SRP Section Section 5.3.1 further further states that the requirements requirements of GDC GOC 1 and 30 and 10 CFR 50.55a regarding regarding quality standards standards are met by compliance with the ASME Code,Section III, III, for nondestructive testing, and that the acceptance fabrication nondestructive acceptance criteria for examination examination of the RV and its appurtenances appurtenances by nondestructive nondestructive examination examination are those specified in ASME Code Section III, NB-5000.

Section III, NB-5000.

In FSAR Section 5.4.4.2, "Penetrant Examinations," in Amendment In Amendment 97 the the applicant applicant changed changed the description description of the liquid penetrant examinations examinations of the core support block attachment welds. The description description previously previously stated the core core support block attachment welds were inspected inspected by dye penetrant after first layer of weld metal and after each 2 inches of weld metal. The description now states the the core support block attachment attachment welds were inspected by dye penetrant penetrant after after each % 1/2 inch of weld metal.

The code of record for the WBNP-2 WBNP-2 RV is the ASME Code, 1971 edition edition through through 1973 addenda. ASME Code,Section III, III, Sub article article NB-4433 of the code of record requires that structural attachment attachment welds be full penetration penetration welds except for temporary temporary attachments and minor supports. ASME Section III, NB-5260 III, Subarticle NB-5260 states that structural attachment attachment welds to pressure-retaining pressure-retaining material shall be be examined examined by either the magnetic magnetic particle particle or liquid penetrant method. The ASME penetrant method. ASME Code does requirements for the increment does not provide any requirements penetrant increment of liquid penetrant examination, examination, ifif performed, for structural attachment attachment welds. However, inspecting the core support block attachment attachment welds after each % A inch of weld is consistent consistent with the requirements requirements of ASME Code,Section III, III, NB-5245 for fillet welded and partial penetration welded joints. ASME Code, Section Il1,NB-5245 requiresSection III, requires such welds to be examined progressively progressively using either either the magnetic particle or liquid liquid penetrant methods, with the increments penetrant increments of examination being the lesser of one half half of the maximum maximum welded welded joint dimension measured measured parallel parallel to the center center line of thethe connection or % A in. (13 (13 mm).

Provided that the core support block attachment Provided attachment welds are full penetration penetration welds, all ASME code requirements requirements are met and the specified nondestructive nondestructive examination examination requirements do not conflict with the ASME code requirements.

requirements requirements.

Requested Information Requested Information

1. Are the core support block attachment welds full penetration welds?

1. welds?

Response: The core support block welds are full penetration penetration welds, but they are not welded through the full section section of the blocks. There is a slot between the block and the vessel shell running the full width of the the each block.

E1-18 E1-18

ENCLOSUREI 1 ENCLOSURE

Response

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

2. Provide the basis for the nondestructive nondestructive examination requirements for the core examination requirements core support block attachment attachment welds.

Response

Response: The progressive liquid penetrant is performed performed as an alternative to ultrasonic examination (UT) because the configuration configuration of the core support block attachment welds is not suitable for performing performing meaningful UT. In In that respect, the situation situation is similar to a partial penetration weld, and progressiveprogressive liquid penetrant penetrant examination examination is considered acceptable alternative examination.

considered the acceptable examination.

NOTE: One comment regarding NOTE: regarding Paragraph 4 of the RAI states that the WBNP-RV Record is the ASME Code 1971 Code of Record 1971 Edition through 1973 Addenda.

Reactor Vessel Design Specification, the Code of incorrect; per the Reactor This is incorrect; Record is 1971 Edition through Winter 1971 1971 Addenda.

5.3.2-1 5.3.2-1 Background

Background

10 CFR Part 50 Appendix pressure-temperature (P-T) limits must Appendix G requires that pressure-temperature be at least as conservative conservative as limits obtained obtained by following the methods analysis methods of analysis and the margins of safety of AppendixAppendix G of Section XI of the ASME Code. The The applicant applicant provided proposed proposed P-T limits in in Reference Reference 4. In In the process of performing calculations to check the WBNP-2 P-T curves, for the confirmatory calculations performing confirmatory the 1 00°F/hour 100°F/hour heatup, the allowable pressure pressure calculated by the staff at lower temperatures is significantly temperatures significantly less than the temperature calculated by the applicant temperature calculated applicant using the methodology methodology of Reference Reference 6. The staff used used the methodology of the the ASME Code,Section XI, Appendix G to calculate the allowable allowable pressures.

pressures. The The staff also observed that the thermal stress intensity factors Kit calculated calculated by the the applicant applicant were generally significantly less than those calculated using the the equations equations of the ASME Code, Section Section XI, Appendix G. G. The staff also observed that the metal temperatures temperatures given in in the PTLR PTLR are higher than the corresponding corresponding metal temperatures temperatures calculated calculated using ASME ASME Code Section Xl XI Appendix G G 0

Figure G-2214.2, up to a coolant temperature temperature of 150°F, 150 F, above which the ASME ASME temperatures are higher. The combination metal temperatures combination of the higher Kit and lower metal higher Kit temperatures temperatures (resulting (resulting in in a lower K Kic) allowable pressure being lc ) results in a lower allowable calculated using the ASME Code Appendix G methods, although the ASME values values applicant's values as temperatures converge with the applicant's temperatures increase, and are within 1%  %

of the applicant's applicant's values at 170°F.170 0 F.

The staff used the simple equation from ASME Code Section XI, Appendix G, The staff used the simple equation from ASME Code Section XI, Appendix G, paragraph G-2214.3 paragraph G-2214.3 to calculate calculate the maximummaximum Kit Kit as a function of heatup heatup rate:

Kit =

Kit = 0.753 xX 103 HU xx tt2.5 10.3 xx HU 25 Where:

HU is the heatup rate in °F/hr °F/hr t==wall thickness in inches inches The staff used ASME Code Section XI, XI, Appendix G, Figure G-2214-2 to determine determine E1-19 E1-19

ENCLOSURE 1 ENCLOSURE Response to Preliminary Response Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant Plant - Unit 2, Docket No. 50-391 the temperature temperature difference from the coolant at a specific wall depth.

The PTLR provides the above ASME Code Section Xl, XI, Appendix G G equation equation for K1t from paragraph Kit paragraph G-2214.3 G-2214.3 as equation (6).

ASME Code Section XI Xl Appendix G, paragraph G-2214.3 also provides the paragraph G-2214.3 the following alternative alternative equation for the thermal thermal stress intensity of an outsideoutside surface surface defect during heatup defect heatup (reproduced Equation (8) in the PTLR:

(reproduced as Equation Kit Klt = (1.043C 0.630C11+ 0.481 C2 ++ 0.401 C3)* 'lrra (1.043C0o + 0.630C ...JTTa The coefficients Co, Co, C 1,, C2 and C3 are determined determined from the thermal stress stress distribution distribution at any specified specified time during the heatup heatup or cooldown cooldown using:

u(x) = Co + Ci(x la) + C2(x la) 2

+ C3 (x la) 3 PTLR Section 3.2 notes that equations 3, 7, and 8 were implemented implemented in the the OPERLIM computer code, which is the program used to generate the the pressure-temperature (P-T) limit curves. Section pressure-temperature Section 3.2 of the PTLR further states that that the P-T curve methodology methodology is the same as that describeddescribed in Section 2.6 of Reference Reference 6 (equations 2.6.2-4 2.6.2-4 and 2.6.3-1).

2.6.3-1).

However, Equation 2.6.2-4 (for the steady state analysis) analysis) and 2.6.3-1 (for the finite finite heatup and cooldown cooldown rate analyses) of Reference 6 provide Reference provide a different method of calculating Kit Kit than described described by the above as equations, as described described by the the following following equations:

K1 p =I .1 MKOp l-rra/Q (2.6.2-4)

Q = (D2 -0.212(o/oay)

Kit = [GM1.1 Mk+ UbMb]/-rra/Q (2.6.3-1)

(2.6.3-1)

Q = (D2 _0.212(Om+OJYy)

Where:

o,== constant Om constant membrane membrane stress component component from the linearized linearized thermal thermal hoop hoop stress stress distribution, d istri bution, Cb =

Ob = linear linear bending stress component bending stress component from from the the linearized linearized thermal thermal hoop hoop stress stress distribution, distribution, Mk ==correction correction factor for membrane stress stress Mb =

Mb correction factor for bending stress, as a function function of relative flaw depthdepth (a/t)

Q ==flaw shape factor modified for plastic zone size,

=

$ = is the elliptical integral of the 2 nd 2 nd kind ($ =

(0 = 1.11376 1.11376 for the fixed aspectaspect ratio of 3 of the code reference reference flaw),

0.212 = plastic zone size correction correction factor, E1-20

ENCLOSUREI 1 ENCLOSURE Preliminary RAIS and RAIs Response to Preliminary Response Regarding Unit 2 FSAR RAls Regarding Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket No. No. 50-391 op Up == pressure pressure stress, ucyy = yield stress, 1.1 = = correction factor for surface surface breaking breaking flaws, a= = crack depth of % 1/

t, and KiP =

Kip = pressure pressure stress intensity factor.

The staff notes that when the applicant's temperature values were applicant's Kit and metal temperature were input to the ASME Code Section XI, XI, Appendix G equations equations for allowable pressure, the values obtained are identicalidentical to the applicant's applicant's allowable pressures.

pressures. However, the staff requires additional information on how the applicant's Kit and metal additional information temperatures temperatures were determined.

Requested Information Requested Information In order to complete our review of the P-T limits, the staff requests the following In following information:

equations was used to determine

1. Clarify which set of equations determine the Kit values used as as input to the P-T curve calculation.

Response

Response: The set of equations used to determine KIt values are defined determine the Kit in defined in Appendix G to Section XI X1 of the ASME Code, paragraph G-2214.3, (b).

part (b).

For an inside surface defect defect during cooldown:

Klt == (1.00359C Kit (1.00359C0o ++ 0.6322C 0.6322C11++ 0.4753C 0.4753C 2 ++ 0.3855C 0.3855C 3 )* )* ý/-a

--JTT8 For an outside surface surface defect defect during heatup:

(1.043C0o + 0.630C Klt = (1.043C Kit 0.630C11+ 0.481 C 2 ++ 0.401 C 3)* 41a

)* --JTT8 The coefficients CO, C1, determined from the thermal C1, C2 and C3 are determined specified time during the heatup stress distribution at any specified heatup or cooldown using:

2 3 a(x) = Co + C1(x la) + C2(x la) + C3 (x /a)

Where x is a variable that represents the radial distance distance (in.) (in.) from appropriate (i.e., inside or outside) surface the appropriate surface and a is the the maximum maximum crack depth (in.). (in.).

incorporated in the OPERLIM computer These equations are incorporated computer code.

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ENCLOSURE 1 ENCLOSURE1 Preliminary RAls Response to Preliminary RAIS and RAls Regarding Unit 2 FSAR RAIs Regarding FsAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 Requested Information Requested Information

2. If equation (8) of the PTLR was used, elaborate elaborate on how how the constants Co, C C1,1 ,

C2 and C3 C2 determined, and how the thermal stress distribution C3 were determined, distribution a(x) was o(x) was determined.

determined.

Response

Response: Equation (8) of WCAP-17035-NP WCAP-17035-NP was used. The temperature temperature through the wall was calculated distribution through distribution function of time and calculated as a function and position for bothboth heatup heatup and cooldown. The one-dimensional transient heat transient conduction equation was used heat conduction used to determine the the through-wall temperature through-wall distribution (see Section 2.6.1 of temperature distribution WCAP-14040-A, Revision Revision 4) as a function of time during during the heatup or cooldown.

Using the through-wall through-wall temperature temperature distribution, the thermal thermal stress stress determined using the equations distribution was then determined contained in equations contained Section 2.6.1 of WCAP-14040-A, WCAP-14040-A, Revision 4 for computing computing thermal stresses. Then, a polynomial fitting method was used to determine determine the values for Co, C C1,1 , C2 ,, and C C3, 3 , as defined defined in Equation (8) of WCAP-17035-NP.

WCAP-17035-NP.

methods described above The methods above are incorporated incorporated into the OPERLIM computer code.

Requested Information Information Reference 2 equations 2.6.2-4

3. If Reference 2.6.2-4 and 2.6.3-1 were were used, describe describe how thethe membrane (om) membrane (arn) and bending (Ob) stresses were determined.

(Ob) stresses determined. Specifically, Specifically, how was the initial stress profile determined prior to the linearization procedure?

linearization procedure?

Also, what value was used for the yield stress? stress?

Response: Equations 2.6.2-4 2.6.2-4 and 2.6.3-1 were used in the calculations. calculations. TheThe Appendix A to Section XI of the ASME Code were methods of Appendix were used membrane and bending stresses. The initial in determination of membrane determined using the equation stress profile was determined equation given by by Timoshenko (Reference Timoshenko (Reference 14 14 of WCAP-14040-A, Revision 4), which is 2.6.1-4 in WCAP-14040-A, Revision 4. The value used for Equation 2.6.1-4 stress is a constant yield stress constant 50 ksi.

incorporated into the OPERLIM computer code.

These methods are incorporated These Information Requested Information determination of the crack

4. With respect to the determination crack tip metal temperatures:

temperatures:

Describe the boundary conditions that were

a. Describe were assumed, particularly at the the vessel outer diameter, and, Describe the methodology for calculating the temperatures.

b. Describe E1-22 E1-22

ENCLOSURE11 ENCLOSURE Preliminary RAIS and RAIs Response to Preliminary Response Regarding Unit 2 FSAR RAls Regarding Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 Response: The vessel inner surface surface is assumed assumed to have a very high convection convection coefficient coefficient (7000 BTU I/ (hr ** ft2 ** OF). The very high convection coefficient coefficient essentially allows unimpeded unimpeded transfer of heat from the the coolant to the inside surface of the vessel. The outside surface is assumed to be adiabatic. Therefore, it is perfectly insulated so that perfectly insulated heat does not transfer transfer from the outside surface surface to the air.

The temperatures temperatures are calculated calculated using the one-dimensional transient conduction equation that is contained in Section 2.6.1 transient heat conduction of WCAP-1 4040-A, Revision 4. A through-wall WCAP-14040-A, through-wall temperature temperature distribution was calculated for each each time step during each cooldown or heatup heatup ramp of interest.

These methods methods and convection incorporated into the convection coefficients are incorporated the OPERLIM computer computer code.

5.3.2-2 5.3.2-2 Background

Background

In order to comply with GDC 15 as itit relates to the reactor coolant system (RCS)

In being designed designed with sufficient sufficient margin to assure that the design conditions of the the RCPB are not exceeded during any condition of normal operations, including including anticipated anticipated operational occurrences, and 10 CFR Part 50 Appendix G with respect operational occurrences, to fracture toughness requirements requirements for the RV, SRP Section 5.2.2 provides provides guidance guidance for the design design of the low-temperature low-temperature overpressure overpressure protection protection (L (LTOP)

TOP) system or the cold overpressure overpressure mitigation mitigation system (COMS). The L LTOP TOP system or or COMS COMS should be in accordance accordance with the requirements requirements of NRC Branch Technical Pressurized-Water Reactors Position (BTP) 5-2, "Overpressurization Protection of Pressurized-Water Reactors While Operating Operating at Low Temperatures," and that the LLTOP TOP system or COMS COMS should should be operable operable during startup and shutdown shutdown conditions below the enable enable temperature paragraph 11.2 of BTP 5-2.

temperature defined in paragraph Technical Specification Specification B 3.4.12, "Cold Overpressure Overpressure Mitigation Mitigation System (COMS),"

refers to the COMS arming temperature temperature specified specified in the PTLR. However, the the COMS arming temperature is not specified specified in the PTLR. Technical Specification Specification B 3.4.12 also states that "The PTLR provides the maximum maximum allowable actuation logic setpoints for the power power operated relief valves (PORVs)."

(PORVs)." However, the PORV setpoints are setpoints not provided are not provided in in the the PTLR.

PTLR.

Requested Information Requested Information

1. temperature for WBNP-2. The PTLR must be Provide the COMS arming temperature be information.

revised to incorporate this information.

Response

Response: COMS COMS is armed when any RCS cold leg temperature temperature is

~-<225°F. The Unit 2 PTLR is included in the System included in Reactor Coolant System (WBN2-68-4001)

Description for the Reactor (WBN2-68-4001) which will be revised to incorporate this information by September September 17, 2010.

E1-23 E1-23

ENCLOSUREI 1 ENCLOSURE Response Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit Unit 2, Docket No. 50-391 Requested Information Requested Information

2. Provide the PORV setpoints for WBNP-2. The PTLR must be revised to to incorporate incorporate this information.

Response: The PORV for the A train Pressurizer is 340A, and the PORV Response: The PORV for the A train Pressurizer is 340A, and the PORV for the B train Pressurizer Pressurizer is 334B.

The PORV setpoints for Unit 2 are given in in the following table:

Watts Bar Unit 2 PORV Setpoints Setpoints vs. Temperature PCV 334 PCV 340A Temperature Temperature Setpoint Setpoint (OF)

('F) (psig)

(psig) (psig)

(psig) 70 455 425 100 455 425 125 455 425 145 510 460 150 510 460 175 510 460 200 775 720 225 775 720 350 2335 2335 The Unit 2 PTLR PTLR is included in the System Description Description for the the Reactor Coolant System (WBN2-68-4001)

(WBN2-68-4001) which will be be revised to incorporate incorporate this information information by September 17, 2010.

6.1.1-1 6.1.1-1 Background

Background

In order to comply with GDC 41 as it relates relates to control of the concentration of hydrogen in the containment hydrogen containment atmosphere following postulated postulated accidents accidents to assure that containment integrity is maintained, maintained, SRP Section Section 6.1.1 recommends recommends that hydrogen hydrogen generation generation resulting from the corrosion of metals by containment sprays containment sprays during a design-basis accident accident should be controlled as described described in RG 1.7, "Control of Combustible Gas Concentrations Concentrations in Containment," Regulatory Position Position C.6 (note, the SRP recommendation recommendation was based based on Revision 2 to RG 1.7, Regulatory Position C.6 is now Regulatory Position CA now Regulatory C.4 in Revision 3 to RG 1.7).

Regulatory Position RG 1.7, Regulatory Position CAC.4 states:

Materials within the containment containment that would yield hydrogen gas by corrosion corrosion from the emergency emergency cooling or containment containment spray solutions solutions should be identified, identified, and their use use should be limited as much as practicable.

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ENCLOSURE I1 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and and RAIs RAls Regarding Regarding Unit Unit 22 FSAR FSAR Tennessee Valley Authority Tennessee Watts Bar Nuclear Authority - Watts Nuclear Plant - UnitUnit 2, 2, Docket Docket No. No. 50-391 50-391 requires that all 10 CFR 50.44 requires water-cooled reactor construction all water-cooled permits or construction permits operating licenses under operating under Part 50 issued October 16, 2003 issued after October comply with the 2003 comply the following:

containments must have an All containments an inerted atmosphere, or must inerted atmosphere, limit hydrogen must limit hydrogen concentrations in containment concentrations following an during and following containment during an accident releases an accident that releases equivalent amount of equivalent amount of hydrogen hydrogen as generated from aa 100 percent as would be generated percent fuel clad-coolant reaction, clad-coolant distributed, to less than uniformly distributed, reaction, uniformly 10 percent than 10 percent (by volume) and (by volume) and maintain containment maintain structural integrity containment structural appropriate accident integrity and appropriate accident mitigating mitigating features.

combustible gas control system of the FSAR Section 6.2.5 states that the combustible the containment air return system, the hydrogen containment hydrogen analyzer analyzer system (HAS) (HAS) and the the hydrogen mitigation system (HMS) hydrogen (HMS) conform to 10 CFR 50.44 requirements.

requirements.

FSAR Section 6.2.5.1 further states that:

In an accident more severe In severe than the design-basis loss-of-coolant accident design-basis loss-of-coolant accident (LOCA),

(LOCA),

combustible combustible gas is predominantly generated predominantly generated within containment containment as a result of the the following:

following:

clad-coolant reaction between the fuel cladding and the reactor (1) Fuel clad-coolant coolant.

core-concrete interaction (2) Molten core-concrete severe core melt sequence interaction in a severe sequence with a failed reactor vessel.

It appears that corrosion It materials has been removed corrosion of materials removed from the design bases of the HMS in FSAR SectionSection 6.2.5.

Additionally, FSAR FSAR Section 6.2.1.3.3 includes the following Section 6.2.1.3.3 following as an input assumption containment pressure analysis:

for the containment Hydrogen gas was added to the containment in the amount of 25,230.2 Standard Cubic Feet (SCF) over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. Sources accounted for were radiolysis in the core and sump post-LOCA, corrosion of plant materials (aluminum, zinc, and painted surfaces found in containment), reaction of 11%

surfaces  % of the Zirconium Zirconium fuel rod cladding in in hydrogen gas assumed to be dissolved in the reactor coolant system the core, and hydrogen water. (This bounds tritium producing core designs.)

generation of hydrogen If the potential for generation If hydrogen gas due to corrosion of reactive metals is is generation of hydrogen due to the fuel clad-coolant compared to the generation insignificant compared core-concrete interaction reaction and the molten core-concrete reaction sequence interaction in a severe core melt sequence unnecessary to limit with a failed RV, itit may be unnecessary metals in limit or quantify reactive metals in order to comply with 10 CFR 50.44.

However, SRP Section 6.1.1 still recommends that hydrogen generation generation due to corrosion of reactive metals be addressed. The information in in FSAR Section also implies a design basis assumption on the amount of aluminum, zinc 6.2.1.3.3 also zinc and coatings containing reactive metals.

E1-25 E1-25

ENCLOSURE11 ENCLOSURE Response to Preliminary RAIS and RAls Response RAIs Regarding Regarding Unit 2 FSAR FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 Requested Information Requested Information

1. Was the potential for generation of hydrogen due to corrosion of reactive reactive metals such as zinc or aluminum considered considered in the design design of the combustible combustible gas control system, or other analyzes analyzes such as containment pressure? IfIf so, describe describe how this contribution contribution was evaluated.

2. If If the contribution contribution of hydrogen from reactive metals was not evaluated, evaluated, justify why this contribution was not evaluated.

evaluated.

3. If If the contribution contribution of hydrogen hydrogen from reactive metals was evaluated, evaluated, discuss the the measures taken to ensure that the use of materials materials that could yield hydrogen gas by corrosion from the emergency emergency cooling or containment spray spray solutions is limited as much as practicable, and is maintained within design basis limits.

References References Overview Report

1. Overview Report on Zinc Addition in Pressurized Pressurized Water Reactors-2004, Reactors-2004, 1009568 Final Report, December December 2004, Electric PowerPower Research Institute Research Institute

2. Pressurized Pressurized Water Reactor Primary Water Reactor Primary Water Zinc Application Guidelines Guidelines 1013420 1013420 Final Report, December 2006, ElectricElectric Power Research Institute Research Institute

3. PWR Axial Offset Offset Anomaly (AOA) Guidelines, Revision, Revision ,1008102, 1008102, Final Report, June 2004, Electric Power Research Institute Power Research Institute WCAP-17035-NP, "Watts Bar Unit 2 Heatup and Cooldown

4. WCAP-17035-NP, Cooldown Limit Curves for for Normal Normal Operation Operation and PTLR Support Support Documentation," Revision 2, December 2009 (ADAMS Accession No. ML ML1100550651) 00550651)

5. ASTM E-185-82, "Conducting Surveillance Surveillance Tests for Light-Water Light-Water Cooled Nuclear Power Nuclear Reactor Vessels" Power Reactor

6. WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Cooldown Limit Curves," Revision 4, May 2004 (ADAMS Accession No. ML0501202094)ML0501202094)

Response

Response to 1: Yes. Reactive metals were considered considered in the concentration concentration of hydrogen hydrogen generated generated during post-LOCA post-LOCA conditions conditions calculations.

ItIt should be noted that Unit 2 is being licensed licensed consistent consistent with Regulatory Guide 1.7, "Control of Combustible Regulatory Combustible Gas Gas Concentrations in Containment," Revision 3. As such, unlike Concentrations unlike Unit 1, 1, Unit 2 will not have Hydrogen Hydrogen Recombiners Recombiners (there will be be Hydrogen Igniters). Reactive metals metals were also appropriately appropriately considered considered in the Westinghouse containment analyses. The The quantity quantity of reactive reactive metals metals considered considered was conservatively conservatively assumed to be approximately 130% of the Unit 1 baseline baseline inventory.

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ENCLOSUREI 1 ENCLOSURE

Response

Response to Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 Response to 2: As discussed in the response response to question question 1, the contribution of reactive metals was considered.

considered.

Response

Response to 3: Watts Bar procedures procedures require evaluation accounting of evaluation and accounting reactive metals (aluminum and zinc) in containment containment to minimize minimize the production production of post accident hydrogen. procedures hydrogen. These procedures require that, "Materials within the containment that would would yield yield hydrogen gas due to corrosion from the emergency hydrogen emergency cooling or containment spray solutions should be identified, identified, and their use should be limited as much as practical."

As indicated indicated in Response 1, Unit 2 calculations are based on on 130% of the Unit 1 baseline approximately 130% baseline inventory. Near thethe end of the Unit 2 construction construction completion completion project, a Unit 2 baseline inventory inventory will be established using techniques techniques developed for Unit 1. This will include a combination combination of containment walkdowns, and design containment design basis document document and work package package reviews. Once the baseline baseline inventory is established, itit will be compared compared to Unit 2 calculations calculations to assure that the the assumptions used are conservative. After the baseline assumptions baseline inventories are established, inventories established, future additions and removals removals will be controlled by station station procedures.

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ENCLOSURE 1I ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAls RAts Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Plant - Unit 2, Docket Nuclear Docket No. 50-391 Preliminary Preliminary RAlsRAIs for FSAR 6.2.4 and 6.2.6 6.2.6 (taken from e-mail from NRC dated 05/10/2010) 05/10/2010)

FSAR Sections Sections 6.2.4 andand 6.2.6 6.2.6

1. Figure 6.2.4-13, "Type XV, Personnel appears to show the interior door within the Personnel Access", appears the airlock doors opening outward, away shield building wall with the airlock away from primary primary containment.

containment.

FSAR FSAR Section 6.2.4.2.3, 6.2.4.2.3, "Penetration Design", states that "A special hold-downhold-down device is provided provided to secure secure the inner door in a sealed position during during leak rate testing of the space between between the doors." This would suggest that the doors in each primary containment containment airlock open inward.

inward. Clarify the orientation orientation of the personnel personnel access airlock doorsdoors

Response

Response: Both personnel personnel access access airlock doors (interior (interior and exterior) open inward towards towards containment containment (refer to Section A-A A-A of Unit Unit 2 FSAR Figure Figure 6.2.4-13).

2. FSAR Section 6.2.4.2.3, "Penetration Design", states that "The ice blowing line penetration has a blind flange with double O-rings installed on the outside of the containment as shown a-rings installed in Figure 6.2.4-16. Sealing between between the outside outside and the annulus penetration penetration through thethe shield is provided by a blind flange fitted with a gasket installed on the inside and outside of the Shield Building penetration.

penetration. FSAR Section 6.2.4.3.1, 6.2.4.3.1, "Possible Leakage Paths",

Paths",

includes in the Type B leakage leakage paths from containment to the annulus the ice blowing blowing line line O-ring a-ring and blind flange through line leak and refers to Figures 6.2.4-16 and 6.2.4-23.

Figures 6.2.4-16 Figure 6.2.4-16, "Type XVIII, XVIII, Ice Blowing Line", shows a line having having a flange with single single O-ring gasket inside the containment a-ring containment building as well as in the annulus between the between the containment building building and and shield building. Figure 6.2.4-23, "Ice Blowing and Negative Negative Return Lines - Blind Flange Flange Details", shows a single flange with double square cross section gaskets located located in the annulus. Clarify the configuration configuration of the ice blowing and negative negative return penetration barriers.

Response

Response: Amendment 98 to the Unit Amendment corrected Figure 6.2.4-16 to show that the Unit 2 FSAR corrected the ice penetration (X-79A) has ice blowing line penetration has a blind flange with double O-rings a-rings installed installed on the outside of containment. Figure 6.2.4-23 which provides provides thethe details for the ice blowing and negative return penetration penetration barriers (X-79A and X-79B) correctly showsshows double O-rings that fit into a square groove double a-rings groove in thethe flange. This figure also shows that these a-ring O-ring gaskets are compressed by are compressed the blind flange when when it is installed, installed, thereby providing a double thereby providing double barrier against containment leakage.

containment E1-28 E1-28

ENCLOSURE ENCLOSURE 1

Response

Response to Preliminary RAIS to Preliminary RAIS and and RAIs Regarding Unit 2 FSAR RAls Regarding FSAR Tennessee Valley Authority Tennessee Valley Authority - Watts Bar Bar Nuclear Plant - Unit Nuclear Plant Unit 2, Docket Docket No.

No. 50-391

3. Table 6.2.4-1, "Watts Bar Nuclear Table 6.2.4-1, Nuclear Plant Containment Penetrations and Barriers", has Containment Penetrations has several inconsistencies several needs clarification, inconsistencies for which the staff needs clarification, including:

including:

a) Table 6.2.4-1 contains numerous abbreviated 6.2.4-1 contains information as well as a notations for information abbreviated notations notes notes column column but doesdoes not appear to have aa legend or notes not appear notes table to support understanding of the data presented.

understanding presented.

Response: Amendment 99 Amendment 99 to the the Unit 22 FSAR reattached reattached the four four pages pages of of abbreviations and abbreviations and notes notes to the end of Unit 22 FSAR FSAR Table 6.2.4-1 6.2.4-1 as as sheets 65 through sheets 65 through 69 of 69.

69.

b) numbers as 50 and 51 while the details penetration X-6 lists the valve numbers Entry for penetration details sketch shows valves 51 and 58.

Response: Amendment 98 to the Unit 22 FSAR corrected Amendment corrected the sketch forfor containment penetration containment 6.2.4-1 to show the valve penetration X-6 of Table 6.2.4-1 valve numbers as 50 and 51.

numbers 51.

c) penetrations X-23, X-28, X-85A, X-86A, X-86B, Entry for penetrations Entry X-86C, X-92C, X-105, and X-86B, X-86C, X-106 X-1 show them to be spare 06 show spare but also shows both App J Type C as well as Type Type A applicable. Table 6.2.6-2, "Containment Isolation Valves Subjected to tests being applicable. to penetrations X-23, X-28, X-85A, X-86A, Type C Testing", lists isolation valves for penetrations X-86B, X-86C, X-92A, X-92B, X-92C, X-105, and X-106.

Response: Amendment 99 to the Unit 2 FSAR corrected remove corrected Table 6.2.4-1 to remove the Appendix Appendix J Type C identified for spare containment C test identified containment penetrations.

penetrations.

Amendment 98 to the Unit 2 FSAR corrected corrected Table 6.2.6-2 6.2.6-2 to remove remove all valves listed for spare penetrations penetrations X-1 X-105 05 and X-106 and the the penetrations.

Amendment 100 to the Unit 2 FSAR will revise Table 6.2.6-2 to remove remove penetrations X-23, X-28, X-85A, X-86A, X-86B, X-86C, X-92A, spare penetrations X-92B and X-92C and associated valves.

d) connection valves (in series) 42B/1 and Entry for penetration X-25C shows both test connection 42B/2 as being closed for ILRT testing but only 42B/1 42B1/1 as being tested by the ILRT.

Response: As noted in the response response to question 3.e below, these test connection connection valves should not have been listed in Unit 2 FSAR Table 6.2.4-1 and Amendment 100.

are being removed from the table via Amendment 100.

Containment penetration X-25C has been spared and the Note: Containment the instrument line for PdT 30-42 now enters containment via via penetration X-60B.

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ENCLOSURE11 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket Docket No. 50-391 e) Entries for instrumentation penetrations (X-25C, X-26C, X-85C, X-86D, instrumentation penetrations X-860, X-97) list the the connection valves while the entries for other penetrations test connection penetrations do not specifically specifically list valve data for test connection valves.

Response: The subject test connection connection valves are all less than 1-inch 1-inch nominal diameter, administratively administratively locked closed whenever containment containment integrity is required and form a double double barrier in the associated associated test connection line. Section 3.3.1 of ANSI/ANS-56.8-2002 ANSI/ANS-56.8-2002 indicates indicates valves valves that meet these criteria criteria do not require Appendix J, Type B or C testing.

connection valves, when These test connection when installed in the mid 1990s, were added to the entries for the associated associated containment instrumentation instrumentation penetrations on FSAR Table penetrations Table 6.2.4-1 via Amendment 66. However, as test connection connection valves that are not subject to Appendix J, Type B or C testing, they should not be listed in FSAR FSAR Table 6.2.4-1.

6.2.4-1.

Amendment 100 100 to the Unit 2 FSAR will revise Table 6.2.4-1 6.2.4-1 to remove remove the listing of these test connection valves from containment containment instrumentation instrumentation penetrations penetrations X-26C, X-57B, X-60B, X-97, X-98 and X-102. This will ensure the entries for these penetrations penetrations are consistent with the entries for other penetrations penetrations that do not specifically specifically list valve data for test connection connection valves.

Additionally, since these test connection connection valves are not subject to Type B or C testing, Amendment Amendment 100 to the Unit 2 FSAR FSAR will revise revise Table 6.2.6-3 connection valves for containment 6.2.6-3 to remove the test connection containment instrumentation instrumentation penetrations penetrations X-26C, X-57B, X-60B, X-60B, X-97, X-98 and X-102.

Note: Amendment 98 to Unit 2 FSAR Table 6.2.4-1 made the following following changes changes in penetrations for the indicated instrument instrument PdTs:

PdT 30-42 was moved penetration moved from penetration X-25C to penetration X-60B.

X-60B.

PdTs 30-44 and 30-311 were moved from penetration X-85C to penetration X-57B.

penetration .

PdT 30-45 was moved from penetration penetration X-86D X-860 to X-102.

X-1 02.

f) Entries Entries for fire protection protection system penetrations penetrations X-31 and X-78 show a process fluid "A"and normal position codes of "0" code of "A" "0"for the inboard and outboard outboard isolation isolation valves.

Response: The reactor building, including including the annulus, is provided provided with dry standpipe systems. Thus, process fluid code of "A" "A"(air) is correct forfor fire protection line penetrations penetrations X-31 and X-78.

Amendment Amendment 100 100 to the Unit 2 FSAR FSAR will revise Table 6.2.4-1 to show shutdown positions of check valves the normal and shutdown valves 26-1296 and 26-1260 as closed closed for penetrations penetrations X-31 and X-78, respectively.

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ENCLOSURE1 ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Nuclear Plant - Unit 2, Docket No.

Authority - Watts Bar Nuclear No. 50-391 g) Entry for penetration X-107 penetration X-1 applicable App J test for relief valve 07 shows no applicable valve 74-505 74-505 when the other inboard boundary valves are shown as tested by App J Type A test.

inboard boundary Response: The RHR system is in service service during the ILRT and therefore, the valves valves penetration X-1 07 are not Type A tested.

associated with penetration Amendment 100 to the Unit 2 FSAR will revise Amendment revise FSAR Table 6.2.4-1 to:

" show valve 74-2 is open during the ILRT; and

" remove remove the Type A test identified for valves 74-2, 74-2,74-8 74-8 and 63-185.

h) Entry for penetration X-1 X-118 18 indicates there are two blind flanges, one in the the containment building and one in the shield building, containment building building, although the detail sketch only only shows the flange in the shield building. The detail listinglisting also does not show a normal, shutdown, post-accident, or ILRT position position for the flange in the shield building.

Response: Amendment 100 to the Unit 2 FSAR will revise Table 6.2.4-1 for penetration X-1 18 to remove the blind flange inside containment from penetration X-118 the table since the flange is not a Code item, and thus cannot be be containment boundary. The revision will show that the credited for containment the flange inside the shield building is closed under normal conditions, can be open for shutdown conditions, is closed under post-accidentpost-accident conditions conditions and can be open for ILRT.

ILRT. The blind flange installed on penetration X-1 penetration X-11818 inside the shield shield building is designed double designed with double O-ring a-ring gaskets and a leak rate test connector. These a-ring gaskets O-ring gaskets compressed are compressed by the blind flange when it it is installed thereby, providing a double barrier against containment containment leakage.

penetration X-54 lists aa blank flange in both the containment building and in Entry for penetration i) Entry in building but the detail sketch the shield building sketch shows only the flange in the shield building building and Figure 6.2.4-15, "Type XVII, Incore Instrumentation Instrumentation Thimble Assembly Renewal Line", shows only one flange in the shield building. The valve (barrier) data listing for post-accident but that the App J ILRT position is one flange indicates that it is closed post-accident "0O".

"0".

Response: Unit 22 FSAR Table incorrectly lists two blind flanges for Table 6.2.4-1 incorrectly for penetration penetration X-54, one in the shield shield building building and one in the auxiliary building.

building. Note 16 to Table 6.2.4-1 incorrectly indicates 6.2.4-1 incorrectly indicates two blind flanges provided in the annulus for penetration are provided penetration X-54, one as a primary containment isolation barrier and the other as a secondary containment containment secondary containment isolation isolation barrier.

Amendment 100 to the Unit Unit 2 FSAR will revise Table 6.2.4-1 to remove remove the blind flange inside the auxiliary building from the listing for penetration will be penetration X-54. Note 16 for this penetration containment penetration containment be revised also.

The blind flange installed on penetration penetration X-54 inside the shield building designed with double is designed. double O-ring a-ring gaskets and a leak rate test connector.

E1-31 El-31

ENCLOSURE ENCLOSURE 1

Response

Response to Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 These O-ring gaskets are compressed by the blind flange when itit is installed, thereby, providing providing a double barrier against containment containment leakage. The blind flange in the shield building on containment containment penetration X-54 is closed post-accident. During the ILRT, the blind blind flange is opened, and a pressurization assembly is attached pressurization assembly attached to the the pressurization of the containment penetration to support the pressurization containment for the the ILRT.

j) Entry for penetration penetration X-39A shows isolation isolation valves 63-64 and 63-868 63-868 while while Table 6.2.6-2 6.2.6-2 shows isolation valves 63-64 and 77-868.

Response

Response: Amendment Amendment 100 to the Unit 2 FSAR will revise Table 6.2.6-2 6.2.6-2 to identify valve 77-868 77-868 as63-868.

k) Entry for penetration penetration X-39B shows isolation isolation valves68-305 and 68-849 68-849 while while Table 6.2.6-2 6.2.6-2 shows isolation valves68-305 and 77-849.

Response: Amendment 100 to the Unit 2 FSAR will revise Table 6.2.6-2 to identify Response: Amendment 100 to the Unit 2 FSAR will revise Table 6.2.6-2 to identify valve 77-849 as68-849.

I)

I) Entry for penetration penetration X-41 details sketch shows relief valve 1-77-2875.

1-77-2875.

Response: Amendment Amendment 100 100 to the Unit 2 FSAR will revise Table 6.2.4-1 to identify the valve shown in the sketch for penetration penetration X-41 as 77-2875.

m) Entry for penetration penetration X-47A shows isolation valves61-191,61-192,61-191, 61-192, and 61-533 61-533 while while 61-191,61-192, and 61-788.

Table 6.2.6-2 lists valves61-191,61-192, Response: Amendment 100 to the Unit 2 FSAR will revise Table Table 6.2.6-2 to replace replace valve 61-788 61-788 with valve 61-533.

n) n) Entry for penetration X-47B shows isolation valves61-193, 61-194, and 61-680 while 61-193,61-194, while Table 6.2.6-2 lists valves61-193,61-194,61-193, 61-194, and 61-935.

Response: Amendment Amendment 100 to the Unit 2 FSAR will revise Table Table 6.2.6-2 to replace replace valve 61-935 61-935 with valve 61-680.

o)

0) Entry for penetration X-76 shows isolation valves33-713 and 33-714 while while Table 6.2.6-2 lists isolation isolation valves valves33-732 and 33-733 33-733 Response: Unit 2 FSAR Amendment Amendment 98 corrected corrected Table Table 6.2.4-1 such that both thethe valve sketch sketch and the valve data identify the valves as33-732 33-732 and 33-733.

El-32 E1-32

ENCLOSURE 1 ENCLOSURE I Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Unit 22 FSAR Tennessee Valley Authority .- Watts Bar Nuclear Nuclear Plant*

Plant - Unit 2, Docket No. 50*391 50-391 p) p} Entry for penetration penetration X-97 shows no isolation valves and only test connection valves valves 30-133B/1 and 30-133B/2 while Table 6.2.6-2 lists isolation valves30-134 and 30-135

Response

Response: Table 6.2.4-1 is correct correct in that there are no containment containment isolation isolation valves valves for penetration X-97. Amendment Amendment 98 to the Unit 2 FSAR removed penetration X-97 from Table 6.2.6-2 (Containment penetration Isolation Valves (Containment Isolation Valves Subjected Subjected to Type C Testing).

q) Entry for penetration penetration 26C lists in-line valves 30-43A30-43A and 30-310A along with test connection 30-43C1, 30-43C2, 30-31 OC1, connection valves 30-43C1, OC1, and 30-31 30-310C2 OC2 while Table 6.2.6, "Valves Exempted From Type C Leak Testing", Testing", lists only the test connection connection valves.

Table 6.2.4-1 also shows valves30-43A and 30-310A being open post-accident post-accident and for ILRT but also being tested by the ILRT ILRT Response: 30-43A and 30-31 OA are normally open manual isolation valves Valves30-43A valves that are not required to close post-accident to provide containment post-accident provide containment isolation. As such, these valves valves should not have been listed in Unit 2 FSAR Table Table 6.2.4-1 for containment penetration penetration X-26C. These valves valves are also not subject to any Appendix J Type B B or C testing, and therefore, should not be included in Unit 2 FSAR Table 6.2.6-3.

Amendment Amendment 100 100 to the Unit 2 FSAR will revise Tables 6.2.4-16.2.4-1 and 6.2.6-3 to remove the test connection valves and valve numbering numbering from the sketches for penetration 26C."

4. Section 6.2.4.1, "Design Bases", item (5) states that "Relief valves may be used as isolation Section 6.2.4.1, valves, provided the relief valve setpoint valves,provided setpoint is greater greater than 1.5 times the containment containment design design internal pressure." The information in Table 6.2.4-1 6.2.4-1 shows all relief valves used as containment isolation valves, other than those on the four main steam lines, to be be discharging to the primary containment, whether discharging whether or not they are installed inside or outside of containment. Confirm Confirm ififthis is correct.

Response

Response: Unit 2 FSAR Table 6.2.4-1 identifies five relief valves as containment containment isolation valves: 2-RV-62-662 2-RV-62-662 (X-15), 2-RV-63-28 2-RV-63-28 (X-30), 2-RV-70-703 2-RV-70-703 (X-35),

(X-35),

2-RV-77-2874 (X-41) and 2-RV-74-505 2-RV-77-2874 (X-41) 2-RV-74-505 (X-107).

The requested requested information for each each of these valves is as follows:

1. 2-RV-62-662 (X-15):

2-RV-62-662 (X-1 5): configuration configuration control drawings drawings (CCDs) 2-47W809-1 and 2-47813-1 2-47813-1 show 2-RV-62-662 2-RV-62-662 discharging discharging inside containment containment to the the pressurizer relief tank.

2. 2-RV-63-28 2-RV-63-28 (X-30): EDCR 53580 installs 2-RV-63-028.

2-RV-63-028. Drawing Drawing Revision Authorization, 53580-001 is the markup for Authorization, DRA 53580-001 CCD 2-47W811-1.

2-47W811-1. On this drawing, drawing, the valve is located located outside of containment. The flow diagram continues on CCD 2-47W813-1 2-47W813-1 which shows the relief valve discharge entering entering containment through X-24; itit then goes to the pressurizer relief tank. The relief valve discharge discharge line line for thermal relief valve 2-RV-63-28 2-RV-63-28 (X-30) is routed to a 4-inch relief valve discharge discharge header header located located outside containment. The relief valve valve E1-33 E1-33

ENCLOSURE I1 ENCLOSURE Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. No. 50-391 discharge header discharge header is routed to the PRT and enters containment containment through through penetration penetration X-24. The containment containment isolation isolation provisions provisions for penetration X-24 consist of a closed system system outside containment containment and check valve valve 68-559 inside containment.

3. 2-RV-70-703 2-RV-70-703 (X-35): EDCR installsinstalls 2-RV-70-703.

2-RV-70-703. As shown on CCD 2-27W859-3, 2-RV-70-703 2-RV-70-703 is located inside containment and discharges discharges inside containment containment to the waste disposal disposal system.

4. 2-RV-77-2874 (X-41): EDCR 2-RV-77-2874 2-RV-77-2874. DRA EDCR 53948 installs 2-RV-77-2874.

53948-005 53948-005 is the markup for CCD 2-47W851-1.

2-47W851-1. On this drawing,drawing, thethe valve is located located inside containment and discharges discharges to the equipment equipment drain sump.

5. 2-RV-74-505 2-RV-74-505 (X-107): As shown on CCD 2-47W810-1, 2-47W810-1, 2-RV-74-505 2-RV-74-505 is located located inside containment and discharges containment to the discharges inside containment the pressurizer relief tank.

5. Section Section 6.2.6.2, "Containment Penetration Penetration Leakage Rate Test", description description of exemptions exemptions (1)(2), (1)(3), and (1)(5)

(1)(2), (1)(3), (1)(5) state, as does Note 3 to Table 6.2.6-3, that water water testing testing of water water sealed sealed valves is "as "as specified in 1010 CFR 50, Appendix J." This section section also indicates indicates that Appendix JJ Option B is to be implemented specified in the plant implemented as specified Technical Specifications.

plant Technical Specifications.

Appendix J Option Option B does not specifically specifically describe water testing as does Option Option A.

Appendix JJ Option B requires the plant technical specifications specifications will include, by general reference, reference, the regulatory guide or other implementation implementation document used to develop a performance-based leakage-testing performance-based leakage-testing program.

program. Clarify Clarify the source of the requirement requirement regarding water testing of valves.

Response: The references references to water testing in 1(2), 1(3) 1(3) and Note 3 pertain to piping piping integrity and Section XI of the ASME Boiler and Pressure Pressure Vessel Code.

System pressure tests are described described in Article IWA-5000 IWA-5000 of the code.

Exclusion Exclusion from Type C testing is based on the provisions of a seal pressure greater than 1.1 Pa Pa and a 30-day seal water inventory. This is as described described in in 10 CFR 50, Appendix J. Type C testing is performed performed with air or nitrogen.

nitrogen.

6. Section 6.2.6.2, "Containment Penetration Penetration Leakage Rate Test", item (1), "Method 1, Pressure Decay" indicates that either air or nitrogen can be used as the test medium and that the leakage leakage rate be calculated using the specified formula. The formula does not appear to provide a conversion conversion for the results when using nitrogen.

nitrogen. Clarify ifif test results are converted to equivalent equivalent air leakage leakage when using nitrogen. This section also uses the the abbreviation abbreviation "Pac" while "Pa" is used used elsewhere. "Pa""P," is defined defined inin Appendix J while "Pac" is abbreviations are being used.

not. Describe how both abbreviations used.

Response

Response: Nitrogen Conversion Nitrogen The Pressure Decay Decay Method Method does not use partial vapor pressures nor does it use a conversion conversion for pure nitrogen.

nitrogen. It It uses the Ideal Gas Law found in in ANSI/ANS-56.8-1994:

ANSI/ANS-56.8-1994:

E1-34

ENCLOSURE1 ENCLOSURE 1

Response

Response to Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 TV P P21 Tstp t [ýl T2Pstp Where:

Li = Local leak rate, cfm 3

TV = volume, ft

Initial pressure, psia psia P z = Final pressure, psia P2= psia T

Tl = Initial temperature, temperature, OR oR T z = Final temperature, T2= temperature, OR oR Tst T stp = Standard atmospheric atmospheric temperature, oR OR stp = Standard Pstp P Standard atmospheric atmospheric pressure, psia psia t= = Test duration, duration, min.

For Type Band B and C tests, the test volume is initially pressurized pressurized with air or nitrogen nitrogen above the design accident pressure pressure of 1515 psig. The final test pressure also remains above the design accident pressure. Pressures Pressures and temperatures are recorded temperatures recorded at the beginning beginning and at the end of each test. This This is a 'state point to state point' test that does not need to be adjusted for partial vapor pressures.

Watts Bar Unit 2 Technical Specifications Specifications Bases 3.6.1 (Containment)

Regulatory Guide 1.163 for Appendix J testing. The Regulatory references Regulatory Position in that document is that NEI 94-01 provides methods acceptable acceptable toto the NRC staff for complying with the provisions of Option B in 10 CFR 50 50 Appendix J. NEI 94-01, 94-01, in turn, references references ANSI/ANS-56.8-1994 ANSI/ANS-56.8-1994 where this this Pressure Decay Method is described (Section Pressure Decay Method is described (Section 6.4). 6.4).

Pac VS Pa abbreviation P, The abbreviation Pa is found in 10 CFR 50 Appendix JJ Option B and is defined as the calculated calculated peak containment internal pressure related to the containment internal the loss-of-coolant accident design basis loss-of-coolant accident as specified in the Technical abbreviation Pac Specifications. The abbreviation Pac is found in ANSI/ANS-56.8-1994 ANSI/ANS-56.8-1994 and is defined as the calculated calculated peak containment internal pressure related to the internal pressure the DBA. In In later later revisions of 56.8, Pac Pac was changed to Pa- Pa . Also, as seen in the the NRC Safety Safety Evaluation Report Report to revision revision 2-A of NEI 94-01, 94-01, Pac was changed to PPa.

a . Pac was a carry-over carry-over from the 19941994 revision revision of 56.8. Amendment Amendment 100100 to the Unit 2 FSAR will change Pac to P P,.

a .

E1-35 E1-35

ENCLOSURE 1 ENCLOSURE Response to Preliminary RAIS and RAls Response RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 7.

7. Section 6.2.6.2, "Containment Penetration Penetration Leakage Leakage Rate Test", item (1), "Method 3, "Waterflow", has been deleted deleted but the paragraph paragraph preceding the Method 1 I description description still refers to water as being aa pressurizing pressurizing medium.

medium. Clarify Clarify whether or not water could be used pressurizing medium.

as the pressurizing Response: Watts Bar 2 does does not perform perform Type B or C testing testing using water as the the medium. These tests areare performed performed using the pneumatic fluids nitrogen or air. Water seals are provided provided on some of the penetration lines and a seal water inventory inventory leakage leakage rate test may be performed performed in lieu of aa Type C airair leakage rate test. Amendment Amendment 100 to the Unit 2 FSAR FSAR will remove water water from the paragraph that discusses it being a pressurizing pressurizing medium.

medium.

8. Table 6.2.6-2, "Containment Isolation Valves Subjected Subjected to Type C Testing",

Testing", has several inconsistencies for which the staff needs inconsistencies needs clarification, including:

a) Penetration Penetration X-52 entry lists valves 1-70-100 1-70-790.

1-70-100 and 1-70-790.

Response: Amendment Amendment 100 to the Unit 2 FSAR will remove the Unit 1 identifiers identifiers from isolation valves "1 100" and "1 "1-70-100" 790" in Table "1-70-790" Table 6.2.6-2. These These valve valve numbers numbers were verified with CCOCCD 2-47W859-3.

b) Penetration X-56A entry lists valves 1-67-113 1-67-113 and 1-67-10540.

1-67-1054D.

Response: Amendment Amendment 100 100 to the Unit 2 FSAR FSAR will remove remove the Unit 1 identifiers identifiers from isolation valves "1-67-113" and "1-67-1054Y" "1-67-10540" in Table 6.2.6-2.

These valve numbers were verified with CCO CCD 1-47W859-3 1-47W859-3 and EDCR EOCR 52796.

c) Penetration X-65 entry lists valves31-309, 31-308, Penetration 31-308, and 31-3407 while Table 6.2.4-1 6.2.4-1 details details sketch shows valves31-309, 31-309, 31C-308, 31 C-308, and 31-3407.

Response: Amendment 100 to the Unit 2 FSAR will remove remove the "C" designator from "31 C-308" in Table 6.2.4-1 for penetration X-65. ItIt was also noted penetration X-64 had an unnecessary that penetration unnecessary "C" in "31 C-306." It will also be removed. These valve numbers were verified verified with CCD CCO 2-47W865-5.

d) Penetration Penetration X-66 entry lists valves valves31-326,31-327,31-326, 31-327, and 31-3392 while Table 6.2.4-16.2.4-1 detail sketch sketch shows shows valves31-326, 31C-327, and 31-3392.

31-3392.

Response: Amendment 100 to the Unit Amendment Unit 2 FSAR will remove the "C" designator from "31C-327" "31C-327" in Table 6.2.4-1 for penetration X-66. This valve valve number was verified with CCOCCD 2-47W865-5.

e) Penetration X67 entry lists valves31-330, Penetration 31-330,31-329,31-329, and 31-3378 31-3378 while Table 6.2.4-1 detail sketch sketch shows valves31-330, 31C-329, and 31-3378.

31-3378.

Response

Response: Amendment Amendment 100 100 to the Unit 2 FSAR designator FSAR will remove the "C" designator from "31C-329" "31C-329" in Table 6.2.4-1 for penetration penetration X-67. This valve valve number was verified with CCOCCD 2-47W865-5.

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ENCLOSURE ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Authority - Watts Bar Nuclear Tennessee Valley Authority Tennessee Docket No. 50-391 Nuclear Plant - Unit 2, Docket

9. Table Exempted From Type C Testing", has two inconsistencies, for which Table 6.2.6-3, "Valves Exempted the staff needs needs clarification:

a) Penetration X-19A 72-044 while Table 6.2.4-1 X-19A lists valves63-072 and 72-044 valves 6.2.4-1 shows valves 63-72 63-72 and 72-44.

Response: Amendment 100 Amendment FSAR will change the valve identifiers for 100 to the Unit 22 FSAR 6.2.6-2 from "63-072" and "72-044" to penetration X-19A in Table 6.2.6-2 penetration "63-72" "63-72" and '72-44,"

"72-44," respectively. These valve numbers were verified with CCO 2-47W812-1.

CCD 2-47W812-1.

b) Penetration X-19B lists valves63-073 and 72-045 while Table 6.2.4-1 shows valves Penetration valves 63-73 and 72-45.

Response: Amendment 100 to the Unit 2 FSAR will change change the valve identifiers identifiers for penetration penetration X-19B in Table Table 6.2.6-2 from "63-073" and "72-045" to "63-73" "63-73" and "72-45," respectively. These valve numbers numbers were verified verified with CCO CCD 2-47W812-1.

2-47W812-1.

E1-37 E1-37

ENCLOSURE11 ENCLOSURE Response to Preliminary Preliminary RAIS and RAts Regarding Unit 2 FSAR RAls Regarding Tennessee Valley Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 Valley Authority - Watts Bar Nuclear Preliminary RAls for FSAR Preliminary RAIs FSAR 6.2.5 6.2.5 (taken from e-mail from NRC dated 05/10/2010) 05/10/2010)

FSAR Section 6.2.5 6.2.5 -- 1. Provide description of how each of the criteria 1 through 8 of Regulatory Provide a description Regulatory Guide Guide (RG) 1.7 Revision 3, Section C.2.1 for commercial (RG) commercial grade hydrogen analyzer analyzer are met.

Response

Response: Unit 2's Hydrogen Hydrogen Analyzer Analyzer is manufactured manufactured by Meggitt Safety Systems, Inc. (MSSI).

(MSSI).

1. Survivability Survivability system is comprised The sampling system comprised of a combination of components qualified by test for 1E safety related system and components commercial components. Wetted commercial Wetted components the components within the sample loop (components sample (components exposed to containment sample) are containment sample) qualified by test. Exceptions include the use primarily qualified use of commercially available commercially available solenoid valves and aa condensate condensate trap.

Materials of construction Materials these two items consist of stainless construction for these> stainless steel (SST) and ethylene propylene propylene diene monomer monomer (EPDM) for extensively the seals. Both of these materials have been tested extensively within our 1E for use within E systems. Although the sample pumppump housing is smaller, all of the components that are exposed to those used in the 1E sample pump.

sample are identical to those the sample The pump motor is also smaller (1/2 hp) but continues to provide margin. Testing under the worse case combination of conditions has proved maximum load the pump imposes is proved the maximum

<1/3 hp.

Mechanical integrity of the system was demonstrated by test Mechanical integrity of the system was demonstrated by test (Seismic Test Report ER 11441) 11441) in 2009.

2. Power Power operating code is stored on an Electrically Erasable The operating Erasable Programmable Read-Only Memory Programmable Memory (EEPROM)

(EEPROM) included with provided with the CPU. Power the CPU. Battery backup is also provided Power 480V Reactor Vent Board 2A which is a provided from the 480V is provided diesel backed backed source.

3. Quality Assurance Quality Assurance Meggitt Safety Systems operates Meggitt documented Quality operates under aa documented Quality Assurance program that meets Assurance meets the requirements of 10 CFR 50, Appendix Appendix B and is structured around the guidelines structured around guidelines provided byby ANSI/ASME NQA-1, ANSIIASME Amended Rule Continuous NQA-1, 2000. Amended Continuous Air Air manufactured within Monitor System (CAMS) was designed and manufactured this QA program.

program.

E1-38

ENCLOSUREI 1 ENCLOSURE Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Plant - Unit 2, Docket Nuclear Docket No. 50-391 equipment in the Meggitt is aa major supplier of Safety related equipment the Mexico, Europe and Korea. This equipment US, Mexico, encompasses equipment encompasses hydrogen monitoring hydrogen monitoring systems, and silicon dioxide insulated instrument and control signal cables (including 10 CFR 50, instrument Appendix R Replacement Replacement cables).

In the course of being an equipment In services supplier of equipment and services Meggitt has been safety related equipment, Meggitt been audited by many US Nuclear Procurement utilities with the most recent being a Nuclear Procurement Issues Committee (NUPIC)(NUPIC) audit of the program.

program. Meggitt is committed to maintaining committed maintaining this prominent position as a quality supplier critical, post accident monitoring supplier of critical, monitoring equipment.

Meggitt Safety Systems is an ISO 9001, 9001, 2001 certified supplier.

Display/Recording

4. Display/Recording The system includes an internal trend function for the most recent 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period and results of the most recent 5 automaticautomatic calibrations. Storage beyond described herein is outside beyond that described outside the scope of the system. Analog output signals are provided suitable for user's chart recorders.

Range

5. Range The MSSI analyzer's internal operating range is 0 to 100% 100%

Hydrogen. specifies the desired operating Hydrogen. The user specifies operating range of the analog output signal to be as low as 5% full scale up to 100% full scale increments of 1 scale in minimum increments 1%.

6. Servicing/Calibration Servicing/Calibration Meggit periodic testing Meggit specifies periodic including an testing of the system including instrument channel function that is based on field experience.

instrument channel experience.

administrative control. All of the Access codes provide for administrative the performed during plant power maintenance activities may be performed maintenance operation.

Factors

7. Human Factors interface for the system is a touch screen display The human interface display for screens and their organization which the screens organization have been well received by the industry. The sample station and system control electronics have been designed to provide good access control electronics access maintenance activity as well as convenient test points.

for maintenance Rapid resolution of issues is enhanced enhanced by diagnostic features diagnostic features included within the software. A key feature includes listings of current and historical alarms with recommended historical alarms recommended steps for their Converter (DAC) Drive resolution. Logic and Digital to Analog Converter Drive convenient method for verifying screens provide a convenient logic verifying control logic El-39 E1-39

ENCLOSURE11 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit Unit 2, Docket No. 50-391 and analog signal functions.

8. Direct Measurement Measurement MSSI CAMS use electrochemical electrochemical sensing for detecting detecting hydrogen concentration. The cells are specifically hydrogen (or oxygen) concentration. specifically designed to respond only to hydrogen (or oxygen). This is designed achieved through selection of the electrode materials and the achieved the electrolyte. For example, hydrogen entering the hydrogen hydrogen sensor will be ionized, ionized, releasing electrons. The number number of electrons electrons being exchanged exchanged is directly proportional to the number number of hydrogen molecules entering hydrogen molecules entering the sensor. Since the number number of molecules molecules entering the sensor is directly directly related to hydrogen partial pressure, the exchange exchange of electrons electrons between between electrodes electrodes provides provides a signal directly proportional to hydrogen partial directly proportional pressure.

6.2.5 6.2.5 - 2. Describe the approach for demonstrating demonstrating equipment equipment survivability in the beyond accident environment design basis accident conditions inside the containment. RG 1.7 environment conditions Revision Revision 3, Section C.2.1, item (1) "Equipment Survivability identifies Section C.2.1, identifies that the the acceptable approaches acceptable approaches for demonstrating demonstrating equipment survivability are described equipment survivability described in in Chapter 19 19 of references 9 and 11 given in the RG. RG.

Response: In Supplements 4 and 5 to the Sequoyah (SQN)

In Supplements (SON) SER (NUREG-001Q1), the NRC staff found the interim hydrogen (NUREG-0011), hydrogen ignition ignition system to be an acceptable acceptable means for hydrogen hydrogen control for degraded core accidents. The operating licenses of SQN SON 1 & &2 were conditioned, however, on the basis that TVA continue research programs on hydrogen programs hydrogen control measures for containment containment integrity and equipment equipment survivability.

TVA, in cooperation with Duke Power and American Electric Electric Power, continued continued studies and eventually decided decided that deliberate ignition deliberate systems were the best option for controlling hydrogen. In In the early 1980s, TVA developed, developed, and submitted submitted to the NRC, an extensive extensive evaluation of equipment subjected subjected to the hydrogen hydrogen burn. This was was documented in letters dated June documented June 2, 1981; December December 1, 1981; and December 9, December 1982. TVA compared analytically 9,1982. experimentally analytically and experimentally determined determined thermal responses of essential essential equipment equipment with theirtheir qualification qualification temperatures. The NRC NRC concluded concluded in SSER6 that the the issues of hydrogen hydrogen control and equipment survivability survivability during postulated degraded-core degraded-core accidents satisfactorily resolved, accidents were satisfactorily subject to the addition of 4 more igniters in upperupper containment. SON SQN then added 4 igniters bringing the license license condition to a close.

Watts Bar Unit 1 also added 4 igniters and performed performed analyses on on the production accumulation of hydrogen production and accumulation hydrogen within containment.

Watts Bar Unit 1 however however based based much of its analysis on its sister. sister.

plant, SQN.

SON. All three units are 4-loop Westinghouse Westinghouse pressurized water reactors with ice condenser containment containment systems that are E1-40 E1-40

ENCLOSUREI 1 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAIs Regarding Unit 2 FSAR RAls Regarding Tennessee Valley Authority - Watts Bar Nuclear Plant Plant - Unit 2, Docket No. 50-391 nearly identical in design. Two plant differences differences noted by the NRC were the higher spray system design flow rate and the the non-existence of vacuum breakers at Watts Bar. Based on the non-existence the similarity of the igniters, the air return fans and the hydrogen monitors, the NRC staff concluded that a plant specific specific analysis of degraded-core accidents was not necessary degraded-core necessary for Watts Bar Unit 1.

The staff found that Watts Bar Unit 1 met the requirements of 10 CFR 10 CFR 50.44 50.44 inin SSER8.

SSER8.

Since the Watts Bar Unit 2 design is nearly identical to the SON SQN units and Watts Bar Unit 1, 1, the approach approach that SQN SON used to demonstrate equipment survivability demonstrate survivability is applicable applicable to Watts Bar Unit 2 Sequoyah's in lieu of a plant-specific analysis. An overview of Sequoyah's extensive survivability analysis extensive survivability analysis follows.

Equipment Essential Equipment The selection selection of equipment that must survive a hydrogen hydrogen burn was was based on that component's component's function during and after an accident.

The four equipment categories categories were:

(1)

(1) Systems mitigating mitigating the consequences consequences of the accident (2) Systems needed needed for maintaining maintaining integrity of the containment containment pressure boundary (3) needed for maintaining Systems needed maintaining the core inin aa safe condition condition (4) monitoring the course of the accident Systems needed for monitoring accident The list of equipment was then limited to equipment most sensitive sensitive to temperature temperature change. Items Items with low heat capacity, items that contained heat sensitive components contained located in components or items located in containment were determined containment determined to bound all the items originally on the the list. These items were selected selected for an evaluation of their thermal response in a hydrogen burn environment:

(1) Mitigating Systems Systems 1.1 Hydrogen Igniters 1.2 Air Return Fans (ARFS) 1.3 Associated Power and Control Cables Cables 1.4 Hydrogen Recombiners Recombiners (2) Maintaining Containment Systems Maintaining Containment Pressure Boundary 2.1 Air Locks and Equipment HatchesHatches 2.2 Containment Isolation Isolation Valves including Hydrogen Hydrogen Sample Valves Valves 2.3 Electrical Penetrations Electrical Penetrations 2.4 Gaskets and Seals for FlangesFlanges E1-41 El-41

ENCLOSURE11 ENCLOSURE Response to Preliminary RAIS and RAls Response RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar NuclearNuclear Plant - Unit 2, Docket No. 50-391 2.5 Electrical Boxes Electrical Boxes (3) Systems Maintaining Maintaining Core Safety 3.1 Reactor Vessel Vent Valves (PORVs)

Reactor (4) Monitoring Systems Systems 4.1 Steam Generator, Pressurizer Pressurizer and Sump Water Water Level Transmitters Transmitters 4.2 Core Exit Thermocouples Thermocouples 4.3 Reactor Reactor Coolant System Pressure Transmitters Transmitters 4.4 Hot Leg RTDs RTDs 4.5 Cold Leg RTDsRTDs 4.6 Reactor Vessel Level System 4.7 Associated Cables (in conduits and exposed)

Associated exposed) 4.8 Junction Boxes Junction Boxes 4.9 Operators Operators on Solenoid Valves Solenoid Valves 4.10 Analyzers Hydrogen Analyzers (1)

(1) Assemblies Igniter Assemblies (2) Transmitters Barton Transmitters (3) Cables Igniter Power Cables (4) Thermocouple Thermocouple Cables Cables (5) Resistance Temperature Resistance Temperature Detector (RTD) CablesCables The NRC staff reviewed the criteria criteria for selecting selecting the equipment and the rationale for bounding bounding the remaining remaining equipment and found them to be acceptable.

acceptable.

Environment Response Thermal Environment Response Analysis Analysis The CLASIX computer code was used used for modeling the thermal environment. Hydrogen Hydrogen accumulation was assumed as a result of aa small break LOCA in conjunction with the loss of emergency emergency core coolant injection injection but with both trains of sprays and air return fans fans operating. The hydrogen operating. hydrogen was assumed to reach 8 volume percent percent when ignition ignition was initiated with eacheach burn assumed to reach 85%

completion. Flame propagation propagation was assumed to have a velocity of 1 fps throughout throughout containment containment with a constant constant adiabatic adiabatic flame flame temperature of 1400 0 F. lower temperature 1400°F. 6 burns were assumed in the lower compartment and 26 burns were assumed in the upper compartment upper plenum.

plenum. No burns were assumed assumed in the upperupper compartment. The average time time between between burns burns in the lower compartment compartment was 200 seconds and gas gas temperature reached temperature reached 8841F.

884°F. The average time betweenbetween burns in the the upper upper plenum plenum was 90 seconds and gas temperaturetemperature reached E1-42 El-42

ENCLOSURE1 1 ENCLOSURE Response to Preliminary RAIS and RAIs Response RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket Docket No. 50-391 1146 0 F. Acceptance 1146°F. Acceptance criterion was based on the qualification qualification temperature temperature of the equipment equipment and the duration for which the the temperature was maintained.

temperature maintained.

The models used in the analysis were verified by comparing comparing calculated calculated results derived from other accepted accepted computer computer codes.

Heat transfer transfer inside equipment was determined determined using the the HEATING HEATING 5 computer code (ORNL). The COCO computer program used for pressure was used pressure transmitter response. Thermocouple Thermocouple response was compared compared to the studies performed by Fenwal Fenwal in 1980.

1980. Sandia performed performed independent independent verification of TVA's analyses analyses and concluded they yielded conservative conservative results. The thermal response of igniter cable was determined determined experimentally experimentally at Singleton Laboratory.

Pressure Effects Pressure Effects containment during The pressure profile inside containment hydrogen burn was during a hydrogen was obtained from a CLASIX analysis analysis with aa 12 fps flame speed. With deliberate ignition, the highest deliberate ignition, highest pressure pressure did not exceed pressures exceed pressures used during during qualification qualification testing. It was also shown that the strain on the blades blades of the air return fans, because because of the pressure differential between the upperupper and lower compartments, was was precluded by the backdraft precluded backdraft dampers.

Conclusions Staff Conclusions After reviewing reviewing TVA's analyses and experimental experimental investigations, the investigations, the NRC staff concluded that all the equipment equipment required to ensure safe shutdown and containment integrity was able to survive the shutdown the environment created environment created by the burn of hydrogen hydrogen in aa degraded degraded corecore accident. The staff acknowledged acknowledged that the peak containment containment pressure, assuming assuming a broad broad range range of accident accident scenarios scenarios with conservative conservative assumptions, would remain remain well below the containment containment pressure capacity. The staff'sstaffs concern that the igniters might not hydrogen in a spray environment was initiate a lean mixture of hydrogen was resolved by the addition of 4 igniters.

6.2.5 -- 3. RG 1.7, Section Section C.3 states that all containment containment types should have an analysis of the effectiveness effectiveness of the method used for providing providing a mixed atmosphere.

atmosphere. This This analysis should demonstrate demonstrate that combustible gases will not accumulate within a within a compartment compartment or cubicle to form a combustible or detonable detonable mixture mixture that could cause loss of containment containment integrity. In addition, the footnote footnote 2 in the RG states "The

The NRC staff believes that current lumped parameterparameter analytical codes may overestimate mixing processes (in overestimate (in particular, natural convection). Applicants Applicants should substantiate substantiate the applicability of these codes to their analyses through through sensitivity sensitivity studies, validation validation with data, or other other means."

performed and the results obtained Describe the analysis performed obtained which demonstrates demonstrates the the effectiveness effectiveness of the proposed proposed method for providing providing a mixed atmosphere atmosphere during aa beyond design basis accident. In addition, describe the approach beyond approach taken to demonstrate demonstrate that mixing is not overestimated overestimated by the code used for the analysis.

E1-43

ENCLOSURE 1 ENCLOSURE1 Response to Preliminary Preliminary RAls RAIs Regarding Unit 2 FSAR RAIS and RAls FsAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket No. 50-391

Response

Response: In Supplement 3 to the Sequoyah In Sequoyah (SON) (NUREG-001 1),

(SQN) SER (NUREG-0011),

indicated that hydrogen released during analyses indicated degraded core during a degraded core accident reasonably well mixed by the time itit left the lower accident would be reasonably lower compartment. These analyses analyses were cursory in nature nature and did not quantitatively characterize quantitatively hydrogen mixing and distribution within characterize hydrogen within the ice condenser containment. During this time, the Electric Power (EPRI) was engaged in aa hydrogen research Institute (EPRI)

Research Institute program attempting to more effectively show that large hydrogen gradients would not occur during a degraded gradients degraded core accident. As part of this research, the Hanford Engineering Development Laboratory Engineering Development (HEDL) performed performed a series of large scale tests.

conducted at HEDL's Containment The mixing tests were conducted Containment Systems Systems Facility (CSTF). This facility had a vessel Test Facility that was 67 ft'tall fttall and and 25 ft in diameter. The upper compartment was prone to better mixing because containment sprays therefore because of containment modeling therefore the modeling emphasis was on the lower compartment. Atmospheric Atmospheric temperatures, concentrations were measured at temperatures, velocities and gas concentrations at several distribution points. The tests were designed several designed to characterize characterize hydrogen distribution for two release scenarios:

scenarios: (1) aa 22 inch pipepipe horizontal orientation, and (2) aa 10 inch pressurizer break with a horizontal relief tank rupture disc opening with a vertical upward orientation, emergency core cooling injection. Two different release both with no emergency release investigated and the tests were performed rates were investigated performed with and without air return fans. Helium was used as the test medium medium in most cases for site safety.

The test results showed good mixing mixing in the lower compartment when operated throughout the air return fans operated throughout the accident. In In all cases, with forced recirculation, the maximum hydrogen hydrogen or helium helium between all points was less than 3 volume concentration difference between concentration volume percent and was generally on the order of 2%. The concentration percent concentration differences were noted to stop increasing differences increasing even before the releaserelease period was over and were less than 1 volume percent within percent within minutes after stopping the source gas.

5 minutes recirculation (air return fans The HEDL test results with no forced recirculation fans inconclusive. During the gas-steam release, the inoperable) were inconclusive. the maximum concentration difference between concentration difference between all measurement measurement points was 2 volume percent. Following the release, however, the test compartment developed compartment developed a vacuum as the steam condensed. This This coupled with the lack of forced recirculation reverse migration coupleel created difference of as much as 7 volume percent.

concentration difference created a concentration noted, however, that later in the test (-20 minutes after ItIt was noted, stopping the release), with or without air return fans, the test volume stopping volume was well mixed with less than 1 volume percent percent concentration concentration difference.

difference.

El -44 E1-44

ENCLOSURE 1 ENCLOSURE Response to Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. No. 50-391 Based on the results of the HEDL tests, the NRC NRC staff concluded that the formation of significant significant hydrogen concentration gradients hydrogen concentration gradients in containment containment was unlikely if the air return fans operated during during the the accident. They concurred that operation operation of the igniters would maintain the hydrogen concentrations maintain concentrations at or below the flammability flammability limit for the duration of the accident. The staff agreed agreed with the TVA position that detonation detonation was not a credible phenomenon phenomenon because because (1) no rich hydrogen hydrogen concentrations concentrations would be reached with the the combination of mixing and deliberate combination deliberate ignition; (2) there there were no no sources to initiate high-energy sources initiate aa detonation; detonation; and (3) there were no no geometrical geometrical channel obstructions that could cause flame flame acceleration acceleration yielding a deflagration deflagration to detonation detonation transition (DDT).

The staff found that the concurrent concurrent loss of air return fans and igniters igniters was highly unlikely because highly unlikely because of the redundancy redundancy of both systems.

They also concluded concluded that with the combination combination of air return fans and deliberate deliberate ignition, the DDT phenomenon was also highly unlikely. unlikely.

documented in SSER6.

This was documented Following these tests, TVA was asked asked to model broader accident model broader scenarios to account scenarios account for variable variable steam and hydrogen hydrogen releases releases (i.e.,

(Le., steam inerting, inerting, bursts of steam or hydrogen, hydrogen, etc.). These These scenarios included included an intermediate intermediate line break with the loss of ECCS; a small line break with the loss of containment heat removal; aa loss loss of main feedwater feedwater concurrent concurrent with a total loss of AC power; power; and a concurrent loss of main and auxiliaryauxiliary feedwater with the loss of ECCS. TVA effectively effectively modeled hydrogen release rates of 6 Ib hydrogen release lb per second with and without ice. These CLASIX CLASIX computer code studies studies were then submitted to the NRC.

The NRCNRC staff compared compared the results of the CLASIX studies studies with an independent independent study performed by Brookhaven National National Laboratory (BNL). They also compared compared the release rates to those proposed proposed in 46 FR 62281, 62281, "Interim Requirements Requirements Related to HydrogenHydrogen Control."

The staff found that the release rates and the accident accident sequences representation of degraded were an adequate representation degraded core situations. They concluded that the scenarios used to develop concluded steam/hydrogen develop steam/hydrogen source terms were acceptable.

acceptable. They also concluded that the the CLASIX code was an adequate adequate tool for modeling the ice condensercondenser containment in a degraded containment degraded corecore condition. This was also documented in SSER6.

documented The tests and studies documented documented in Supplement Supplement 66 of the Sequoyah SER demonstrate demonstrate that combustible accumulate within combustible gases will not accumulate within compartment or cubicle to form a combustible or detonable a compartment detonable mixture mixture that could cause a loss of containment containment integrity.

Additionally, the CLASIX code was found to be an acceptable acceptable tooltool with which to model hydrogen in containment. The Watts Bar Unit Unit 2 containment containment structure is nearly identical to the Sequoyah containment containment structure as are the air return fans and deliberate deliberate ignition ignition system. ItIt is proposed, that since Watts Bar and Sequoyah El-45 E1-45

ENCLOSURE11 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 are sister plants, the HEDL HEDL tests and CLASIX CLASIX studies be directly applicable to Watts Bar Unit 2.

applicable 6.2.5 - 4. Refer to the first paragraph Describe the "Phase B isolation paragraph of Section 6.2.5.2.b. Describe signal" referred referred to in this section.

Response

Response: This is the containment "Phase B B isolation isolation signal" that is described in Unit 2 FSAR FSAR Sections Sections 7.3.1.1.1 and 7.3.1.1.4. Amendment 100 100 to the Unit 2 FSAR will add a reference reference to 7.3.1.1.1.

7.3.1.1.1.

6.2.5 -- 5. RG 1.7 Revision 3, Section C.1 provides guidance guidance for survivability of systems, structures and components components (SSCs) installed to mitigatemitigate the hazards hazards from the the generation of combustible gas during a beyond-design-basis beyond-design-basis accident environment.

FSAR Section 6.2.5.2, third paragraph paragraph states "Ductwork not protected protected byby embedment is designed to withstand the LOCA environment" - i.e., does not state embedment state beyond-design-basis beyond-design-basis environment. Verify that the SSCs installed installed for the mitigation mitigation of the hazards of combustible gas are designeddesigned to operate in the beyond-designed beyond-designed basis environment or provide provide justification justification for not meeting meeting the guidance.

guidance.

Response

Response: In the early 1980s, TVA In TVA developed, and submitted submitted to the NRC, an extensive evaluation evaluation of equipment subjected subjected to the hydrogen burn in in a degraded degraded core condition.

condition. This was documented documented in letters dated June 2, 1981; December 2,1981; December 1,1981; 1, 1981; and December 9, 1982. 1982. TVA analytically and experimentally compared analytically experimentally determined determined thermal responses of essential responses essential equipment with their qualification qualification temperatures.

The selection selection of equipment that had to survive aa hydrogen burn was was based based on that component's component's function during and after an accident.

The four equipment equipment categories were:

(1) systems mitigating consequences of the accident; mitigating the consequences (2) systems needed for maintaining integrity of the containment containment pressure boundary; (3) systems needed for maintaining maintaining the core in a safe condition; and (4) systems needed for monitoring the course of the accident.

The list of equipment was then limited to equipment most sensitive sensitive temperature change. Items with low heat capacity, items to temperature items that contained heat contained heat sensitive components or items locatedlocated in containment containment were determined determined to bound all the items originally originally on the the list. These items were selected selected for an evaluation of their thermal response in a hydrogen hydrogen burn environment:

The NRC staff reviewed the criteria for selecting the equipment and the rationale for bounding the remaining equipment equipment and found them to be acceptable.

acceptable. The NRC concluded in in Supplement Supplement 6 to the the Sequoyah SER that the issues of hydrogenhydrogen control and equipment equipment El-46 E1-46

ENCLOSURE ENCLOSURE 1 Response to Response to Preliminary Preliminary RAIS and RAIs RAIS and RAls Regarding Regarding UnitUnit 22 FSAR Tennessee Authority - Watts Tennessee Valley Authority Nuclear Plant Watts Bar Nuclear Plant - Unit Docket No. 50-391 Unit 2, Docket survivability during survivability during postulated accidents was degraded-core accidents postulated degraded-core was satisfactorily resolved. Since the Watts Bar Unit satisfactorily design is nearly Unit 2 design nearly approach that SQN identical to the SQN units, the approach identical SQN used used to to demonstrate equipment demonstrate equipment survivability applicable to Watts Bar survivability is applicable Bar Unit 2 in lieu lieu of of aa plant-specific plant-specific analysis.

Unit 2 FSAR Sections 6.2.5 and 6.8 have Unit revised to remove have been revised remove references references to thethe LOCA replace them with the LOCA and to replace the beyond-design-basis beyond-design-basis accident.

E1-47

ENCLOSURE1 1 ENCLOSURE Response to Preliminary Preliminary RAIS and RAts RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 RAIs for FSAR Preliminary RAls Preliminary FSAR 11 (taken from e-mail from NRC dated 03/23/2010) 03/23/2010)

Section 11 11 11 - 3. Table 11.2-7 11.2-7

a. Provide a basis for concluding Provide concluding that the doses to members of the public public presented presented in the table for the year 2040, are bounding and conservativeconservative for for current plant operation.

operation:

Response: Based on the 20002000 Census, the population population within 50 miles in in estimated to have been 1,064,513, 2000 is estimated 1,064,513, indicating indicating that the the area area around projected.

around the site has been growing faster than projected.

Based on this trend, the population population in the year 2040 is projected to be 1,519,000 within 50 miles. Taking the ratio of the year 2040 population to the year 2000 population, population, results in aa growth rate of 1.42.

1.42. The Year 2040 doses doses for the water supplies supplies are bounding bounding conservative since the population doses were multiplied by and conservative population growth factor of 1.42.

the population

b. individual doses listed in the table are to the maximum Verify that the individual maximum exposed individual in each group.

Response: The individual individual doses in the table were verified using aa Watts Bar computer software software program, "Quarterly Water Water Dose Assessment" (QWATA) that calculates calculates release doses. QWATA uses the the methodology specified methodology specified in Reg. Guide 1.109 to calculate radiation Guide 1.109 doses from potable potable water, aquatic shoreline deposits to aquatic food, and shoreline exposed individuals the maximum exposed individuals in each each of four age groups.

11 - 4. land-use census that is reflected in Table 11.3-9 is still valid or Verify that the land-use or provide aa basis for concluding concluding that the analysis based on this information information is bounding and conservative.

bounding conservative.

Response: When the NRC reviewed the FSAR, the table of interest was was Table amendment to the Unit 2 FSAR Table 11.3-9; a later amendment FSAR resulted in thisthis table being re-numbered as Table Table 11.3-8.

The land-use land-use data contained in Table 11.3-8 data currently contained 11.3-8 is not valid valid...

Amendment Amendment 100 to the Unit 2 FSAR will correct the land-use census land-use census data in Table 11.3-8.

11.3-8.

E1-48

ENCLOSURE 1 ENCLOSURE Response to Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Valley Authority Authority - Watts Bar Nuclear Nuclear Plant Plant - Unit Unit 2, Docket No. 50-391 11 - 7. Describe Describe the source term used to calculate calculate the doses listed in Table Table 11.3-11.

11.3-11.

Amendment 95 resulted in lower lower values for the Total Body and Skin doses. Describe Describe what factors changed with Unit 2 operation that resulted resulted in these lower revised values.

Response

Response: The source source term used used to calculate the doses listed in Table Table 11.3-11 11.3-11 is the same as that discussed in the response response for RAI 11 - 3.

Factors that resulted in the differences differences in dose values in Table 11.3-10 Table 11.3-10 (Amendment 99 version) version) versus the dose values in Table 11.3-11 11.3-11 (Amendment 95 version) version) are as follow:

For Amendment 95, the information information contained in Table 11.3-11. was for Table 11.3-11 two (2) units operating with one one (1) unit containing containing TPBARs.

For Amendment 99, the information information in Table 11.3-10 11.3-10 was for only 1,unit 1 unit operating operating without TPBARs.

Another factor that resulted in different doses was the land use survey used in Amendment Amendment 95 contained higherhigher feeding factors for the time time cows were grazing grazing on pasture.

updated XIO, Further, Amendment 99 used updated X/Q, D/O, D/Q, and joint frequency distribution distribution tables for the period period from January 1986 1986 to December 2005, whereas whereas Amendment Amendment 95 used XlO, X/Q, D/Q, D/O, and joint jOint frequency distribution frequency distribution tables for the period from January January 1974 1974 to December December 1993.

Finally, the 50-mile population was updated updated since Amendment Amendment 95 and was used in determining determining values values for Table 11.3-10 11.3-10 in Amendment Amendment 99.

E1-49

ENCLOSURE1 ENCLOSURE 1 Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant Plant - Unit Unit 2, Docket No. 50-391 Preliminary RAIs Preliminary RAls for FSAR 11.3 11.3 and 11.4 (taken from e-mail from NRC dated 04/28/2010)

Section 11.3 11.3 11.3 - 02 11.3 The references references in Section 11.3.2 to Tables 11.3-4/5 appear Section 11.3.2 appear incorrect. A A review of the tables indicate indicate the reference reference to 11.3-4 11.3-4 should be 11.3-3 11.3-3 and 11.3-5 should be be 11.3-4 and 11.3-5 11.3-5 (these two tables appear appear to be sheet 1 and sheet 2 of the samesame table or some of the sheets are missing) missing)

Response: Amendment Amendment 98 to the Unit 2 FSAR FSAR corrected corrected "Table 11.3-5" to "Table 11.3-4" and "Table 11.3-4" to "Table 11.3-3." Since these were were considered considered to be editorial editorial changes, no change bars were provided, and the amendment level remained the same.

11.3-03: Section 11.3.3.1 under "Waste Gas Compressors" did not include a revision base for Section 11.3.3.1 "Each unit is sized for 40 gpm."

the deletion of: ,"Each Response: Amendment 95 to the Unit 2 FSAR Amendment FSAR deleted deleted this information because because the equivalent equivalent Unit 1 UFSAR UFSAR portion did not include include this information.

information.

This change change was considered to be editorial. Additionally, this this information is provided in Unit 2 FSAR Table information Table 11.3-1 (Gaseous WasteWaste Processing System Component Component Data).

11.3-05: Section 11.3.3.2, "Instrumentation Section 11.3.3.2, "Instrumentation Design." last paragraph.

paragraph. Confirm that this this accurately describes paragraph accurately describes the operation operation of this instrument. Specifically Specifically address address the difference between how the Unit 2 instrument difference between operates compared instrument operates compared to how the the Unit 1 instrument instrument operates as described in the Unit 1 UFSAR.

Response: The last paragraph of Unit 2 FSAR FSAR 11.3.3.2, "Instrumentation Design,"

accurately describes accurately describes the operation of the automatic sequential gas gas analyzer. As stated in the "Auxiliary Services" portion of Unit 2 FSAR 11:3.2, 11 ;3.2, the auxiliary auxiliary services portion portion of the Gaseous Gaseous Waste Waste Processing ProceSSing System includes two automatic automatic gas analyzers. The The automatic sequential gas analyzer monitors several sample automatic sample points and the second analyzer analyzer monitors the operating gas compressor.

analyzers are common These two analyzers common plant equipment equipment for both Unit 1 I and and Unit 2 operation. Since these are are common monitors, there no there is no difference between Unit 1 and Unit 2.

difference Section 11.4 Section 11.4 11.4-01::

11.4-01 11.4-1, "Process and Effluent Radiation Monitors Table 11.4-1, Monitors - Liquid Media,"

Media," includes the the "Steam Generator Blowdown Liquid Sample Monitor" and the "Boric Acid Evaporator Generator Blowdown Evaporator Condensate Monitor." Both of these monitors have Condensate have a footnote that states "Deleted by Amendment 95." Explain why these monitors are being deleted and why are they still included included in the table, ifif not to be installed in in the plant.

Response: The Steam Generator Blowdown Liquid Sample Monitor was isolated Steam Generator in Unit 1 by DCN 29903. Monitor RE-90-124 was to be used solely to determine which SG has a leak during determine during an SGTR SGTR event. In In place place of the monitor, grab samples provide a quicker determination.

quicker determination.

EI-50 E1-50

ENCLOSURE 1 ENCLOSURE Response to Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Nuclear Plant - Unit 2, Docket No. 50-391 Tennessee Valley Authority - Watts Bar Nuclear The Boric Acid Evaporator Evaporator Condensate Condensate Monitor was already deleted in Unit 2 by Unit 1 DCNs 21582 and 51426. The monitors were abandoned in place by DCN 21582 and physically abandoned physically removed by by DCN 51426. Abandoning of the monitors is part of the original original Unit 1 licensing bases.

Both Unit 1 DCNs were used as design input to EDCREDCR 52339. While While no physical included a FSAR physical work was required, EDCR 52339 included change package change package that removed monitor entries from the Unit 2 removed the monitor FSAR.

Amendment Amendment 98 to the UnitUnit 2 FSAR deleted the entries.

E1-51 El-51

ENCLOSURE11

,-ENCLOSURE Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 RAIs for FSAR 12 (taken from e-mail from NRC dated Preliminary RAls 03/25/2010) dated 03/25/2010)

Chapter Chapter 12 12 - 4. As required by 10 CFR 20.1406, describe the Watts Bar Unit 2 design features and operating operating procedures that will minimize, to the extent practicable, contamination of the the facility and the environment environment to facilitate decommissioning.

Response: Design features and operating operating procedures procedures that: 1) minimize minimize contamination contamination to the facility, 2) facilitate the eventual decommissioning decommissioning of the plant, and 3) minimize radioactive radioactive waste are contained in Unit 2 FSAR Section Section 12.3, "Radiation Protection Design procedures that are Features". Plant physical attributes and procedures described in this FSAR Section are consistent with 100 CFR 20.1406.

consistent with*1 12 - 9. Provide a description of the radiation monitoring in areas where where reactor fuel is handled or stored sufficient demonstrate compliance with the requirements of sufficient to demonstrate 10 CFR 70.24 or 10 CFR 50.68 50.68 Response: The referenced referenced CFRs are requirements related to criticality criticality monitors monitors for areas where reactor reactor fuel is handled or stored. NRC issued an exemption requirements of 10 CFR 70.24 as part of the exemption from the requirements the operating licensing. See the following excerpt from section Unit 1 operating section 2.D.(2) of the Unit 1 operating license, which has been incorporated 2.0.(2) into the Unit 1 Technical Specifications:

Specifications:

"2.D.(2)

"2.0.(2) The facility was previously granted an exemption from the criticality monitoring requirements criticality monitoring requirements of 10 CFR 70.24 Nuclear Material License No.

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

Since the new fuel and spent fuel storage areas are common to both units, it is concluded concluded that based on the above, criticality monitors are not required for Watts Bar in areas where the fuel is handled or stored. This is also consistent with our application application for Special Nuclear Nuclear Material Material License dated November November 12, 2009 (ADAMS Accession Accession No:

ML100120487).

ML100120487).

El -52 E1-52

ENCLOSUREI 1 ENCLOSURE Response to Preliminary RAIS and RAls Response RAIs Regarding Regarding Unit 2 FSARFSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 Preliminary RAI for FSAR 14 (taken from e-mail from NRC dated 03/25/2010)

Chapter 14 14 - 1. In NUREG-0847, In NUREG-0847, SSER 16, dated SeptemberSeptember 1995,1995, the staff stated in Section 14.2, "Preoperational "Preoperational Tests," Item 11, 11, that ..."Before

... "Before issuance issuance of an operating operating license license for Unit 2, however, the applicant would have to demonstrate demonstrate the capability capability of each common common station service transformer transformer to carry the load required to supply ESF loads of one unit under LOCA conditions, in addition to power required required for shutting down the the non-accident non-accident unit." However, Table 14.2-1 14.2-1 (Sheet 48 of 89) of Amendment 97 to the the Watts Bar FSAR FSAR for the AC Power Distribution Distribution System Test Summary, does not incorporate this additional language language in the Test Method section of the test description.

description. Provide a discussion discussion specifically specifically addressing this SSER condition condition for Unit 2, given the scenario scenario of having both units operational.

Response: Amendment 100 to the Unit 2 FSAR will correct correct the test description description and Acceptance Acceptance Criteria.

E1-53 E1-53

ENCLOSUREI ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAlsRAIs Regarding Unit 2 FSAR FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 RAIs for Various Portions of the FSAR [from NRC letter dated 06/23/2010 (ADAMS RAls for Various Portions of the FSAR [from NRC letter dated 06/23/2010 (ADAMS ML101450084)]

Accession No. ML 101450084)]

Reactor Reactor Systems Systems (SRXB)

SRXB 1. (5.2.2.4.2) Section 5.2.2.4.2 of FSAR Amendment Pressure Transient Amendment 97, Pressure Analyses, includes an evaluation evaluation of low temperature temperature overpressure transients transients (Section 5.2.2.4.2.1), at-power 5.2.2.4.2.1), but no evaluation of at-power overpressure transients. Please provide provide an evaluation evaluation of at-power consistent with the guidelines of Section 5.2.2 overpressure transients, consistent NUREG-0800.

of NUREG-0800.

Response: Amendment 100 to the Unit 2 FSAR will add the Response: Amendment 100 to the Unit 2 FSAR will add the following:

5.2.2.4.2.1 At-Power Overpressure Transients Overpressure Transients For overpressure overpressure protection during power operation, operation, thethe relief valves are provided provided with sufficient sufficient capacity capacity to preclude preclude actuation of the safety valves during during normal operational transients, when assuming operational assuming the following following conditions:

a. Reactor Reactor is operating at the licensed licensed core thermal power level.

level.

b. RCS ReS and core parameters parameters are at values within within normal normal operating range that produce produce the highest anticipated pressure.

anticipated instrumentation and controls

c. All components, instrumentation controls function normally.

The two PORVs are designed pressurizer designed to limit the pressurizer pressure to a value below the high-pressure high-pressure reactor trip trip setpoint setpoint for all design transients up to and including a 50%

50% step load decrease decrease with steam steam dump actuation.

actuation.

Isolated output signals from the pressurizer pressurizer pressure protection channels are used to control pressurizer pressurizer spray and the PORVs PORVs in the event of an increase increase in ReS RCS pressure. The PORVs are pilot actuated valves which respond to pressure signals or to manual manual control. They provide a means for venting noncondensible noncondensible gases or steam from the pressurizer pressurizer which may impair impair stabilization of the ReS RCS following a design basis event.

They also provide a means to depressurize depressurize the RCS ReS following a steam generator generator tube rupture event by reducing primary to secondary secondary break flow as well as as increasing safety injection flow to refill the pressurizer.

increasing E1-54 E1-54

ENCLOSURE I1 ENCLOSURE Preliminary RAIS and RAIs Response to Preliminary RAls Regarding Regarding Unit 2 FSAR Nuclear Plant - Unit 2, Docket Authority - Watts Bar Nuclear Tennessee Valley Authority Docket No. 50-391 The pressurizer safety valves prevent prevent RCS pressure from exceeding 110% 110% of system design pressure, in in compliance with ASME Nuclear compliance Nuclear Power Plant Components Code. These are totally enclosed pop-type, spring loaded valves and are self actuated actuated by by pressure action and back-pressure direct fluid pressure back-pressure compensation designed to ASME Boiler and Pressure Code,Section III. II1. The combined capacity of two of the combined capacity the three safety valves is greater than or equal to the the maximum surge rate resulting from the complete maximum complete loss of concurrent with the complete load due to a turbine trip concurrent complete loss of main feedwater, all without a reactor trip or any other control.

control.

A rise in coolant temperature can cause an insurge to coolant temperature Pressurizer spray provides a method the pressurizer. Pressurizer method to decrease the rate of steam production in the pressurizer decrease pressurizer condenses the steam at aa faster rate as spray injection condenses than itit is generated. The spray line enters the the pressurizer at the top and terminates terminates in the spray nozzlenozzle inside the unit. The spray rate is regulated by a PIO PID controller which has remote controller remote overrides. In In parallel parallel with the spray valves are manual throttle valves.

Temperature sensors in each Temperature each spray line alert thethe insufficient operator of insufficient bypass flow. The spray rate isis pressurizer pressure from reaching selected to prevent pressurizer reaching the PORV setpoint during during a step load reduction of ten percent from full load.

percent provided with heaters and their pressurizer is provided The pressurizer primary function is to heat and maintain water in the the pressurizer temperature corresponding pressurizer at the saturation temperature corresponding operating pressure. The heaters to the operating heaters limit the pressure decrease resulting from a drop in decrease average coolant in average coolant temperature which, during unloading, causes an outsurge from the pressurizer. The heaters are actuated outsurge automatically during insurges and outsurges and they automatically also have manual overrides.

pressurizer relief tank (PRT) is to The function of the pressurizer discharge from the pressurizer condense and cool the discharge condense safety and relief valves. Steam is discharged safety discharged into the the PRT through a sparger pipe under the level of the water.

The tank is designed condense and cool a discharge designed to condense discharge of steam equal to 110% 110% of the volume above above thethe full-power pressurizer water level set.

E1-55 E1-55

ENCLOSUREI ENCLOSURE 1 Response to Preliminary Preliminary RAls RAIS and RAlsRAIs Regarding Unit 2 FSAR FsAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket Docket No. 50-391 Nuclear Nuclear Performance and Code Review Review (sNPB)

(SNPB)

All references references to Watts Bar Unit 1 (WB1)

(WB1) are from the approved UFSAR UFSAR Amendment Amendment 7. All references references to Watts Bar Unit 2 2 (WB2) are are from Amendment Amendment 95 which is currently under review.

Chapter 4.1 SNPB 4.1 - 2.

sNPB Identify Identify which control rods are being used, WBN Unit 2 FSAR makes multiple multiple reference to both Ag-In-Cd control rods and B4 C with Ag-In-Cd tips. If reference If WBN Unit 2 is transitioning from B4C with Ag-In-Cd tips (as approved in WBN Unit 1) to solely Ag-In-Cd Ag-In-Cd control rods, provide justification justification for such a transition.

Response: Westinghouse Westinghouse Field Change Change Notice Notice FCN-WBTM-10794 FCN-WBTM-10794 will install Ag-In-Cd rod control cluster assemblies (RCCAs) at Watts Bar Unit 2. The previous design was the L-106A-HDR, heavy drive rod with B4C RCCAs. The new design is the the standard standard L-106A drive rod with a modified coupling to mate with Ag-In-Cd RCCAs. The new drive drive rods will be installed after RCCA installation but prior to the reactor vessel head installation. The associated safety analyses analyses reported reported in the the Unit 22 FSAR assume the use of the Ag-In-Cd control rod design.

Amendment Amendment 100 100 to the Unit 2 FSAR will revise applicable applicable portions of the FSAR to reflect the change change in control rods.

Chapter 4.2.2 4.2.2 sNPB 4.2.2 SNPB 4.2.2 - 1. In WBN Unit 2 Amendment In Amendment 95 Section Section 4.2.2.2 4.2.2.2 under under the heading 'Upper

'Upper Core Support Assembly' (p 4.2-25), do the support columns also contain the the thermocouple thermocouple supports?

Response: No.

No. The core exit thermocouples thermocouples are located located at the top of the the Instrument Thimble Assemblies (IITA)

Incore Instrument (IITA) which are inserted from the bottom of the core. The liT IITAs As are supported supported by the the fuel assembly assembly into which which they are inserted.

inserted.

Chapter 4.2.3 4.2.3 SNPB 4.2.3 - 2.

sNPB Does WBN Unit 2 have any part-length CRDMs? CRDMs?

Response: Watts Bar 2 will not use part-length part-length CRDMs. Westinghouse Westinghouse has has completed completed the modifications modifications to Unit 2 to remove 8 part-length CRDMs drive rods and to install 8 guide tube covers at the top CRDMs of the reactor internals under the part length CRDM housings.

El-56 E1-56

ENCLOSUREI ENCLOSURE 1 Response to Preliminary Preliminary RAIS and RAls RAts Regarding Unit 2 FSAR FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 SNPB SNPB 4.2.3 - 3. Amendment 95 for WBN Unit 2 Table 4.1-1 (page 4.1-6) indicates FSAR Amendment indicates the the control rods for WBN Unit 2 are Ag-In-Cd. FSAR Amendment Amendment 7 for WBN Unit 1 Table 4.1-1 indicates the control rods for WBN Unit 1 are B44C with Ag-In-Cd tips. Amendment 95 for WBN Unit 2 Section Section 4.2.3.2.1 under under the the heading 'Rod

'Rod Cluster Control Control Assembly' indicates the control rods are identical to WBN Unit 1, B4C pellets, which are stacked on top of the extruded Ag-In-Cd slugs. Are the control control rods in WBN Unit 1 and WBN Unit 2 the the same, B B4C with Ag-In-Cd tips, or does WBN Unit 2 have different control rods rods indicated in Table 4.1-1 of its FSAR? Additionally, make as indicated make the appropriate updates to WBN Unit 2's FSAR.

Response: See the response response to RAI SNPB SNPB 4.1 - 2.

Additionally, Amendment Amendment 95 to the Unit 2 FSAR corrected the the design of the control rods to reflect the Ag-In-Cd Ag-In-Cd design.

Unit 1 is in the process of transitioning transitioning to this design such that both units will be using the Ag-In-Cd Ag-In-Cd design.

Plant Systems Systems (SBPB)

SBPB 10.4.1-1 10.4.1-1 In the 1982 Safety Safety Evaluation Evaluation (SE) for Watts Bar (NUREG-0847),

(NUREG-0847), the NRC NRC staff observed observed that the three pressure zones zones for the main condenser are designed to produce a turbine back pressure pressure of 1.5 (low pressure pressure -lP),

-LP),

2.15 (intermediate (intermediate pressure -IP),

-IP), and 3.065 (high (high pressure -HP) inches of mercury for Units 1 andand 2. In Amendment Amendment 7 of the WBN Unit Unit 1 FSAR, those values are are changed to 1.63 lP,LP, 2.38 IP, and 3.40 HP. For the proposedproposed FSAR FSAR for WBN Unit 2, the values for the three pressure zones are 1.92, 2.70, and and 3.75 respectively. The NRC staff requests requests that TVA explain why the values for WBN Unit 1 and the proposed proposed WBN Unit 2 pressure pressure zones for the mainmain condenser are different from the values condenser values indicated indicated in the original original SE.

Response: The three zone pressures (1.5, 2.15, and 3.065 inches of mercury) in the 1982 SE were the original original design design pressures for the Unit 1 and Unit 2 main main condensers. The pressures pressures were provided by the condenser vendor based on 100% 100% condenser condenser cleanliness and the condenser condenser heat duties associated associated with thethe licensed thermal power (OlP).

original licensed (OLP). The design pressures pressures of 1.63, 2.38, and 3.40 inches of mercury mercury (identified in the the Condensate System Design Description Condensate N3-2-4002) are based Description N3-2-4002) on 70 *Fof CCW inlet inlet temperature, temperature, 95% clean tubes tubes and HEI Design Mode, 1995 Revision.

In the period since 1982, 1982, several modifications modifications have been made made to Unit 1, which have have affected affected main performance:

condenser performance:

1) re-tubing the condenser with Sea-Cure Sea-Cure stainless steel tubes tubes to remove copper from the secondary secondary side for steam generator generator preservation; preservation; 2) rerouting of the condenser zone cascadingcascading scheme (Zone "A" "A" bypass to Hotwell)

Hotwell) to mitigate mitigate the impact of lower than expected cooling tower performance to maintain lower maintain E1-57 E1-57

ENCLOSURE1 ENCLOSURE 1 Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket Docket No. 50-391 hotwell temperature temperature to the condensate condensate demineralizers demineralizers below process limits); and 3) aa 1.4%

process 1.4% power uprate was approved power uprate which increased increased the heat duty on each each condenser condenser zone. In In addition, with a 115-foot 115-foot length, stainless steel tubing the the condenser condenser cleanliness cleanliness cannot be maintained maintained at 100% 100%

cleanliness (typically 80 - 85%). Note that the Amendment 6 cleanliness heat balance (Figure 10.1-2) heat balance 10.1-2) which is included included in Unit 1 UFSAR Amendment 7 utilizes zone pressures pressures of 1.92, 2.38, and 3.75 inches inches of mercury. As shown on the figure, these zone zone pressures pressures are based on a condenser cleanliness cleanliness of 80%.

The Unit 2 condenser has also been re-tubed with the same same material (Sea-Cure) as Unit 1, 1, but itit will not receive the Zone "A" "A" modification. Following the bypass modification bypass modification. modification on 1, supplemental Unit 1, supplemental condenser condenser cooling water (SCCW) from the the Watts Bar lake Lake was mixed with the Unit 1 cooling tower discharge. This reduced the condenser condenser cooling water inlet temperature and thereby helped temperature helped reduce the hotwell temperature. Unit 2 will receive significant significant benefit from SCCW; therefore, the Zone "A" "A" bypass modification modification will not be made. In In the proposed Unit 2 FSAR, the condenser zone backpressures backpressures were developed consistent consistent with the secondary secondary side thermal model generated generated for a future (post-license)

(post-license) 1.4% 1.4% power uprate, uprate, rather than OlP.

OLP. In addition, the Westinghouse Westinghouse supplied HP LP turbines and the Moisture-Separator-Reheaters and lP Moisture-Separator-Reheaters (MSRs) have been replaced replaced with more efficient units supplied by Siemens. With no operational operational data for Unit 2, the zonezone pressures utilized utilized for Unit 1 were provided to Siemens for the the, Unit 22 replacement turbine design. These pressures These same zone pressures were carried over to the initial heat balances for Unit 2. There There are other minor design differences between the turbine systems systems that also contribute to the unit differences.

differences.

SBPB 10.4.1-2 10.4.1-2 In Amendment Amendment 7 to the WBN Unit 1 FSAR, TVA shows shows the Birmingham Birmingham Wire Wire Gauge (BWG) for the balance of tubes to be 22 BWG, whereas whereas in the the proposed WBN Unit 2 FSAR, this value proposed value is shown as 12 BWG. The NRC staff NRC staff requests that TVA explain explain why WBN Unit 2 has a lower BWG value than WBN Unit 1.

Response: Amendment 98 to the Unit 2 FSAR corrected corrected "12 BWG" to "22 BWG." Since Since this was a correction correction of a typographical typographical error error (22 BWG is correct correct per EDCR 52320), the amendmentamendment level on the page did not change.

SBPB 10.4.1-3 10.4.1-3 In the original original SE for WBN Unit 1 I and 2, copper-nickel copper-nickel tubes were used to minimize minimize corrosion and erosion of condenser condenser tubes. In Amendment Amendment 7 to the the WBN Unit 1 FSAR and the proposed WBN Unit 2 FSAR, the material is listed listed as SEACURE SEACURE for tubes. The NRC staff requests requests that TVA confirm that the the SEACURE material SEACURE material is the replacement replacement for the copper-nickel copper-nickel tubing that was was originally used for the condenser condenser tubes.

E1-58 E1-58

ENCLOSURE 1 ENCLOSURE Response to Preliminary RAIS and RAls Response RAIs Regarding Regarding Unit 2 FSAR FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 Response: TVA confirms TVA confirms that the Unit 2 Main Condenser Condenser has, in fact, been retubed with SEACURE SEACURE stainless steel material identical identical to that that used on the Unit 1 Main Main Condenser. Unit 2 Engineering Engineering Document Document Construction Construction Release (EDCR) 52320 effectively effectively replicates actions taken for Unit 1 under TVA Design Change Change Notice (DCN)

(DCN) M-38974-A M-38974-A which retubed the Unit 1 Main Main Condenser.

SBPB 10.4.1-4 10.4.1-4 In the original SE, the NRC staff noted noted that the main condenser design allows allows for water water storage approximately 33 ~

storage capacity for approximately /2 minutes minutes of full-load operation.

proposed WBN Unit 22 FSAR, this value is the same. However, in In the proposed Amendment Amendment 7 of the WBN Unit 1 FSAR, the condenser condenser is shown to handle handle water storage storage capacity of 3 minutes at full-load operation. The NRC staff staff requests that TVA explains deviation of the WBN Unit 1 storage explains the deviation storage capacity from the proposed WBN Unit 2 storage storage capacity capacity for the main main condenser.

Response

Response: The "3 minutes of full-load operation" information information contained contained in Amendment Amendment 7 to the Unit 1 UFSAR was determined determined to be be inconsistent with design documentation.

inconsistent documentation. Amendment Amendment 88 to the the Unit 1 UFSAR corrected Unit corrected "3 minutes of full-load operation" to read "approximately 3 ~2 minutes of full-load operation."

SBPB 10.4.2-1 In the original SE (NUREG-0847), there are listed three mechanical vacuum electrical heating pumps, an electrical heating coil, coil, a HEPA HEPA filter, and a carbon absorber absorber to comprise of the main condenser evacuation main condenser evacuation system (MCES). In Amendment 7 of the WBN Unit Unit 1 I FSAR and proposed proposed FSAR for WBN Unit 2, TVA does does not not provide provide the full description of the components components of the MCES, nor clarifies if the the MCES MCES still comprises of the same components components as listed in the original SE, with the exception of the vacuum pumps. The NRC staff requests that TVA provides a description description of the components components for the MCES for WBN Units 1 and 2.

Response: The Main Condenser Condenser Evacuation Evacuation System is shown in Unit 2 10.4-7, 10.4-9, 10.4-10, and 10.4-12.

FSAR Figures 10.4-7,10.4-9,10.4-10, 10.4-12. These These figures show the System Components, the flow diagrams, the the electrical electrical control diagrams, and the electrical electrical logic diagram for for the Condensate Condensate System including including the Main Condenser Condenser Evacuation System. The description description of these components components for the Condensate Condensate System is as follows:

1. The Condensate system boundaries boundaries are defined defined to include include equipment:

the following eqUipment:

a. The main condenser to the interface interface with the turbine turbine interface with the Condenser exhaust, and to the interface Condenser Circulating Circulating Water Water (CCW) system connections at the the waterboxes.

waterboxes.

b. The condenser condenser vacuum pump including including the suction suction piping from the main condenser and the MFPT MFPT E1-59

ENCLOSURE11 ENCLOSURE Preliminary RAIS and RAls Response to Preliminary Regarding Unit 2 FSAR RAIs Regarding Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Tennessee Valley Authority Tennessee condensers, and the vacuum discharge discharge to the turbine turbine building roof.

c. The MFPT condenser condenser channel and tube side, and up to the MFPT exhaust and up to the shell draindrain connections.

d. The condensate storage tanks including condensate storage including all condensate transfer interconnecting piping and the condensate interconnecting pump.

e. The hotwell pumps.

f.f. The condensate demineralizer pumps.

condensate demineralizer

g. The condensate condensate booster pumps.

h. The channel and tube side of the GSC, SGB Heat Heat Exchangers and low pressure Exchangers intermediate pressure and intermediate pressure feedwater heaters.

pressure feedwater

2. Condenser At Watts Bar Units 1 and 2, the Main Condenser Evacuation System is comprised of: Condenser Vacuum Vacuum Radiation Monitors Pumps, Radiation exhaust of Condenser Monitors for the exhaust Condenser Vacuum Pumps and Vents to roof.

Vacuum 10/1992) removed M-03153-A (Dated 10/1992)

DCN # M-03153-A internals removed the internals Absorber Train in the MCES and HEPA Filter, Carbon Absorber for HEPA unplugged the Duct Heater. This DCN allowed unplugged allowed the removal of the HEPA and Charcoal Charcoal Filter Absorber Cartridge from its housing in the MCES. ECDR # 53306 allowed allowed the same same modifications modifications for WBNP-2 MCES.MCES. Flow Diagrams and Physical Drawings will be updated to reflect this change Physical change for Unit 22 as part of the EDCR implementation implementation process.

10.4.2-2 SBPB 10.4.2-2 In Amendment 7 to the WBN Unit 1 FSAR, TVA indicates that there are two In Amendment radiation monitors for the vacuum pump exhaust. However, TVA types of radiation vacuum pump exhaust in the proposed describes the vacuum describes proposed WBN Unit 2 FSAR FSAR asas having three types of radiation having radiation monitors. The NRC staff requests that TVA explains the reasoning for WBN Unit 2 having three radiation monitors explains monitors versus versus WBN Unit 1 only having two radiation radiation monitors.

El-60 E1-60

ENCLOSURE 1 ENCLOSURE Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 Response: Unit Unit 1 originally had three physical monitors containing four containing four detectors monitoring the Condenser detectors Condenser Vacuum Pump Exhaust:

RE-90-99, RE-90-99, RE-90-119, and RE-90-404. Unit 1 deleted 2-RE-90-99 with DCN 35431.

2-RE-90-99 35431. The basis for removal was:

"Calculation WBNAPS3-048 WBNAPS3-048 Rev.8 Rev.8 concluded that the the multiple detectors used to monitor multiple noble gas detectors monitor the the condenser vacuum pump exhaust condenser exhaust stream provide provide sufficient range overlap overlap without without the RE-90-99 RE-90-99 mid-mid-range channel.

channel. This channel channel is therefore declared redundant redundant and may be removed from the Radiation Monitoring Monitoring System without loss of required monitoring monitoring capability."

The Unit 1 RE-90-404 is an accident accident monitor that is in in actuality made up of two separate overlapping separate radiation detectors with overlapping ranges. This dual detector detector arrangement arrangement is what allowed allowed Unit 1 midrange midrange monitor RE-90-99 to be deleted. Because monitor Because the monitor currently currently used for RE-90-404 RE-90-404 in Unit 1 is obsolete, two new single-detector single-detector monitors are being installed installed in Unit 2. The two monitors monitors are required required to maintain necessary range overlap maintain the necessary overlap...

Installing two separate separate monitors equipment monitors requires that new equipment identifiers be assigned.

assigned. This results in the appearance appearance that Unit 2 has three detectors and Unit 1 has only two.

E1-61

ENCLOSUREI 1 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 RAIs for FSAR Chapters RAls Chapters 8 and 9 [from [from NRC (ADAMS Accession 07/12/2010 (ADAMS NRC letter dated 07/12/2010 Accession No. ML101530354)]

ML101530354)]

8.1- Electric Power Section 8.1- Power-- Introduction Introduction 8.1 - 1. Final Safety Analysis Report (FSAR) Section 8.1.5.3, "Compliance to Regulatory Regulatory Electrical and Electronics Guides and Institute of Electrical Engineers {IEEE Electronics Engineers (IEEE Standards),"

Standards},"

Regulatory Guide (RG) states that WBN Unit 2 complies with Regulatory (RG) 1.32, 1.32, Revision 0, "Criteria for Power Systems for Nuclear Power Plants;" RG 1.81, 1.81, "Shared and Shutdown Electric Electric Systems for Multi-Unit Nuclear Power Plants," Revision 1, IEEE Multi-Unit Nuclear IEEE 308-1971 ," "Criteria for Class 1E Electric Systems for Nuclear Power Generating Std 308-1971," Generating Stations," in in meeting meeting NRC regulations regulations in Title 10, CodeCode of Federal Regulations (10 Federal Regulations Appendix A General CFR) Section 50, Appendix General Design Criteria 17. FSAR Section Criteria 5 and 17. 3.1.2 Section 3.1.2 states that the preferred and emergency electric power power systems are shared. Since Since NRC staff has not previously reviewed the capability of the preferred preferred and emergency electric power systems operation, provide an executive summary of the systems for dual-unit operation, the analysis to support the following design requirements.

requirements.

Response

Response: reflected on the key diagrams and the single line diagrams, the As reflected the Auxiliary Power System (APS) was originally designed for Watts Bar Auxiliary two unit operation. However, due to indefinite deferral deferral of Unit 2, detailed adequacy of APS was performed detailed analysis of the adequacy performed for Unit 1 only during the Unit 1I licensing licenSing effort. Therefore, the current APS APS analysis analysis evaluates evaluates the system to support Unit 1 I only although although it takes takes account Unit into account Unit 2 busses, boards and loads required required for Unit 1I operation operation and safe safe shutdown.

shutdown.

The purpose of the AC APS analysis is to determine determine its adequacy to support support two unit operation. The analysis was performed performed usingusing Transient Analyzer Program (ETAP) Version 5.5.6N.

Electrical Transient Electrical 5.5.6N. TheThe evaluation includes steady state and transient voltages, equipment evaluation short circuit currents, equipment capability and bus loading. The The configurations and includes analysis was performed for various configurations includes dynamic motor starting, and short circuit steady state conditions, dynamic circuit current and degraded degraded voltage analysis. The detailed detailed analysis was was equipment and components limited to safety related equipment components poweredpowered from safety related boards. The scope of this analysis included the the safety the Common Station Service Transformers Service Transformers (CSST A, B, C and D),

D),

6.9kV Shutdown Boards, 6.9kV Start Buses, 6.9kV Common Common Boards, 6.9kV Unit Boards, the downstream 6900V-480V transformers, downstream 6900V-480V 480V Distribution Systems loads, and all interconnections.

interconnections. The The answers to specific NRC questions below is based based on this analysis:

a. dual-unit trip as aa result of abnormal operational occurrence A dual-unit occurrence Response: analysis is enveloped by the analysis performed This analysis performed while while accident in one unit and concurrent postulating an accident postulating concurrent orderly orderly shutdown of the second unit (See details in the response to shutdown RAI 8.1 - 1.c.).

RA18.1 1.c.).

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ENCLOSURE1 ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Nuclear Plant - Unit 2, Docket No. 50-391 Authority - Watts Bar Nuclear

b. Accident in one unit and concurrent shutdown of the second unit (with and

b. Accident in one unit and concurrent shutdown of the second unit (with and without offsite power)

Response

Response: Analysis with offsite power available:

Unit 1 and Unit 2 share the two independent Unit independent offsite power normal operation sources for normal sources shutdown of the plant.

operation and safe shutdown capability of the auxiliary power system for two To review the capability unit operation, the following analysis was performed:

• Normal plant configuration with CSSTs C and 0 Normal supplying D supplying shutdown boards and CSSTs A and B power to the 6.9kV shutdown supplying power to non-Class non-Class 1E loads: In In this scenario, secondary winding of CSST C and 0D supplies power each secondary to one shutdown board only (one Train of one unit).

One offsite circuit out of service resulting in outage outage of either either CSSTs C&B or D&A: With an outage of one set of transformers, all loads powered transformers are powered by those transformers automatically transferred automatically transferred to the remaining remaining set of transformers. Thys, boards are powered from Thus, all shutdown boards from one CSST C or 0 D and each winding will supply power to to two shutdown boards (one train of each each unit)

Substituting CSST B or A in case of an outage of CSST C Substituting or 0D respectively: InIn this case, CSST B the B or A will feed the manually transferred boards in addition to the transferred shutdown boards the other normally fed non-Class non-Class 1E loads (substitution of source is limited to the use either CSST B or A as offsite source either B or A at any given given time, provided both CSSTs A and B are Bare available).

Normally when the units are operating, the unit boards are Normally powered from the USSTs and transfer to CSSTs A &

powered & B on unit trip. Therefore, ifif a shutdown shutdown board is transferred transferred to the corresponding board and an accident corresponding unit board accident occurs, the the ESF loads on the transferred transferred shutdown board would start on the USST. Since the USSTs are are unitized (there are two USSTs per unit), only one shutdown board board can bebe transferred to one USST. Analysis has been performed performed to verify that the USSTs are capable to start all ESF loads on the transferred shutdown board in addition to the non-class transferred shutdown non-class 1I E loads on the associated associated RCP and unit boards.

• The analysis analysis was performed considering "block start" of all performed considering required to start on receipt of the 6.9kV and 480V motors required the accident unit. It also included accident signal for the accident included those those loads on both units that are associated associated with the normal automatically tripped.

operation and are not automatically E1-63 E1-63

ENCLOSURE11 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSARFSAR Tennessee Nuclear Plant - Unit 2, Tennessee Valley Authority - Watts Bar Nuclear 2, Docket No. 50-391 performed with a grid voltage drop of The above analysis is performed 1 kV (164-153kV) 11 shutdown boards are powered (164-153kV) when the shutdown powered from CSSTs C&D and 9kV with a minimum voltage of 153kV when powered from CSST A or B. In In case of ESF loads starting on USSTs, the analysis is performed considering generator as considering the generator as swing bus at 23kV since the generator remains connected connected to thethe working as a synchronous condenser.

grid and working The auxiliary power system was determined adequate to determined to be adequate support the above scenarios operation. The voltage scenarios for two unit operation. voltage recovery times were within the time limits so that 6.9kV recovery shutdown board degraded voltage relays reset and do not separate 6.9kV shutdown boards from the offsite power source.

Analysis with onsite power (Diesel Generators)

Generators)

There are four diesel generators generators (DGs), one each dedicated one'each dedicated DGDG for 6.9kV shutdown board feeding one train of Unit 1 or Unit 2 ESF loads. The ESF loads are not shared on the diesels. The The analysis with onsite power system was limited to the verification of each DG loading under accident considering accident conditions considering Pre-operational testing loading of 6.9kV motors. Pre-operational sequential loading performed on Unit 1 DG with a safety injection performed injection signal and and starting of random loads demonstrated the capability of the DGs DGs to provide adequate adequate voltage to all required loads. Due to reasonable to similarity of Units 1 and 2 diesel generators, it is reasonable dynamic voltage response of Unit 2 assume that the dynamic pre-operational testing will be the same as previously pre-operational previously determined preoperational testing. Unit 2 determined during Unit 1 preoperational pre-operational testing will validate the diesel response to pre-operational to sequencing of loads on the Unit 22 diesels.

sequencing

c. Accident in one unit and spurious Engineered Features actuation in the Engineered Safety Features the other unit (with and without offsite power)

Response: The ESF loads for each Unit's Train A and B are powered from from a dedicated exception of the dedicated DG and battery (load group) with exception the Turbine Driven Auxiliary Turbine Feedwater (TDAFW) Pumps and vital Auxiliary Feedwater power (See Attachment Attachment 6.). In In the case of the TDAFW pump, create controls cross to the opposite unit's battery in order to create the controls a "Special Train." InIn the case of the 120VAC vital power, a division of power is supplied from each of the four batteries in batteries in order to create required four divisions. Each DG and create the required Battery can support the ESF loads on aa safety injection signal.

The auxiliary power system and supporting analysis complies complies requirements of position C.2.b of Regulatory with the requirements 1.81. The design of the onsite power system is unitized Guide 1.81.

applied due to a spurious such that ESF loads that would be applied spurious E1-64

ENCLOSURE 1I ENCLOSURE

Response

Response to Preliminary RAIS and RAls RAts Regarding Unit 2 FSAR Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit Unit 2, Docket No. 50-391 accident accident signal on one unit would not adversely affect the the capability capability of the other unit to supply supply accident accident loads. The The electrical electrical loads for shared systems such as emergency emergency raw component cooling are fully applied in the cooling water and component the loading analysis for both units. Board feeder alignments alignments for power distribution boards are required to be in the normal position for operability operability such that cross-unit power connections connections would would not be an operable operable condition.

condition. There is no design requirement per the FSAR to be able to support an accident accident inin one unit and a spurious ESF actuation in the other unit when supplied from off-site power. Therefore, Therefore, analysis with one unit unit accident and the spurious in accident spurious actuation of ESF loads in the the performed.

second unit has not been performed.

8.1 - 2. Explain Explain how the industry and WBN Unit 1 operatingoperating experience experience and the NRC generic NRC generic communications communications have been reviewed and incorporatedincorporated in the electrical design, maintenance, maintenance, surveillance surveillance testing, and operations for WBN Unit 2.

Response

Response: The electrical electrical power power system is under the operating unit's control. The The operating experience (OE) and NRC generic operating unit reviews operating experience generic communications communications under the Operating Operating Experience Experience Program, SPP-3.9.

Under this program, external and internal OE as well as generic generiC communication is reviewed communication developed.

reviewed and actions developed.

Section 8.2-8.2- Offsite Offsite Power System 8.2 - 1. The common station service service transformers transformers (CSSTs) are described described in Section 8.2.1.2 8.2.1.2 of the FSAR. It is stated that their calculated calculated loading is well below their winding winding ratings for all conditions. FSAR Page 8.1-13 8.1-13 Position Position C2 states, "The shared safety systems are designed designed so that one load group (Train 1A & 2A or Train 1 1BB & 2B) can mitigate a design design basis accident accident in one unit and accomplish an orderly orderly shutdown of the other unit." These CSSTs are are shared between between the WBN UnitUnit 1 and WBN Unit Unit 2.

In view of the WBN Unit 2 loads being applied to the CSSTs along with WBN Unit Unit 1 loads, the NRC staff requests requests the following information:

Note: For ease ease of providing providing a response, the NRC question question is split into three parts as as described described under each bullet below:

• Provide an executive executive summary of the calculations calculations and analyses analyses which detail the the loading for both units units (or added added loads of WBN Unit 2 to the existing loads of Unit 1).

WBN Unit

Response

Response: TVA performed the APS loading loading analysis analysis with the offsite power system. CSSTs C and D 0 provide two shared shared sources of offsite powerpower to the 6.9kV boards boards to provide provide power to both trains of Unit 1 and Unit 2 ESF loads. Train A of both units is powered from CSST C and train B is powered powered from CSST D. D. When both CSSTs are available, available, secondary winding feeds only one 6.9kV shutdown each transformer secondary board board (one train of a unit).

El-65 E1-65

ENCLOSURE 1 ENCLOSURE Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket Nuclear Docket No. 50-391 In addition to CSST C and D, In D, CSST A and B, B, which currently provide provide maintenance feed for the 6.9kV shutdown boards a maintenance boards via 6.9kV unit boards, are retrofitted with automatic automatic on load tap changers changers by DCN 52336 and may be used during transformer maintenance maintenance to provide a qualified offsite source source to one train of safety related ESF loads (EDC 55945 55945 - under issue). CSST A will be used as a substitute for CSST D D when when CSST D is out of service, and similarly similarly CSST B will be used to substitute for CSST C. Thus, two CSSTs will always be available to provide provide two independent independent sources sources of offsite offsite power to the 6.9kV shutdown boards. However, since CSSTs A A and B B also feed the plant non-safety non-safety related loads, the use of CSST A A qualified offsite source will be limited to only one and B to provide a qualified one train of ESF loads with both CSSTs A and B available. CSST A and B will be used used either as an immediate immediate or as aa delayed delayed second offsite offsite source. A delayed delayed second source is permissible and meets the the requirements requirements of RG 1.32.1.32. The calculation also analyzes analyzes when one one train of ESF loads is supplied from CSSTs A and B.

The worst case loading on anyone any one secondary secondary winding of CSSTs C and D with Unit 1 in accident condition (Unit (Unit 22 defueled) defueled) is calculated as 6.02MVA.

6.02MVA. With only Unit 1 operating, Unit 2's 6.9kV shutdown boards are lightly loaded (2.84MVA).

(2.84MVA). With Unit 2 operating, operating, the worst case loading on the secondary winding feeding the accident unit and and non-accident non-accident unit loads is calculated calculated as 6.12MVA 6.12MVA and 5.07MVA respectively.

Analysis has also been performed performed to verify that sufficient power is available available to safety loads to mitigate an accident in one unit and to to safely shutdown shutdown the other unit for loss of one offsite power source, or a CSST or a Common Switchgear. In In this, case one secondary winding of the remaining CSST will feed two 6.9kV shutdown boards boards (i.e., Train A of both units) and the second secondary secondary winding will feed two 6.9kV shutdown boardsboards for Train B of both units. The worst case loading on each loading each secondary secondary winding with one unit in accidentaccident and the the concurrent controlled other unit in concurrent controlled shutdown is calculated as as 11.30MVA with primary 11.30MVA primary winding load of 22.44MVA. Both primary and secondary winding loads are well within the CSST self-cooled secondary self-cooled rating rating of 22MVA and 33MVA respectively.

Note: ERCW ERCW pumps are being replaced. Upon final testing, slight slight adjustment adjustment may be required for ERCW pump loading which is less than O.013MVA.

0.01 3MVA.

With one train of ESF loads loads powered from CSST A or B, the worst worst case steady state loading loading on CSST A or B is calculated calculated as 67MVA which is within the FA rating of 76MVA.

The worst case case loading on the (Y) (Y) secondary secondary and primary windingwinding (Y-winding feeding shutdown boards boards via unit boards) when one Train Train of ESF loads for each unit is simultaneously simultaneously powered from CSST A E1-66 E1-66

ENCLOSURE 1 ENCLOSURE

Response

Response to Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 or B (substituting for CSST D or C, respectively) is 39.79MVA 39.79MVA and 67.61 MVA which is within the FA rating of 48MVA and 76MVA respectively. '

Refer to Attachment Attachment 6 for CSSTs loading details for various various configurations.

configurations.

The 6.9kV unit boards, which provide alternate alternate supply to the 6.9kV shutdown boards, are normally normally powered powered from the USSTs. The The connected to the 500kV system for 30 seconds after a USSTs remain connected unit trip. Therefore, itit is possible that the ESF loads for the accident forthe accident unit will be connected connected to the USSTs ifif CSST A or B is being used as as the offsite source. InIn such a case, the ESF loads could "block start" on the USSTs. Therefore, an analysis has also been performed performed to verify that the USSTs are adequate adequate to support the safe shutdown while powering the shutdown shutdown boards. ItIt is determined determined that in this case case also, the degraded degraded voltage relays get reset in less than 5 seconds seconds without tripping the offsite offsite power. The loading of the USSTs USSTs is also calculated to be well within its FA rating of 20MVA.

calculated The rating of all 6900V/480V transformers, transformers, buses and boards boards feeding feeding safety related loads are adequate adequate and no overloading overloading was identified.

Define the bounding conditions for maximum loading that demonstrates demonstrates that the the winding ratings are not exceeded.

Response: Bounding Bounding conditions evaluated evaluated in the AC APS analysis is one unit in in accident accident and the other other unit in controlled shutdown with loss of one one offsite power power source or a CSST (C or D) D) or a common station switchgear (C or D).

switchgear D). ItIt is concluded concluded from the analysis analysis that the winding ratings are not exceeded.

Provide a summary of the calculation demonstrates the design margin calculation that demonstrates margin in thethe CSSTs with a design-basis accident accident (DBA) in one unit and a concurrentconcurrent shut down of the other unit. The summary of the design design calculations must include include inputs, assumptions, and a summary summary of output results (with acceptanceacceptance criteria) including any load creep for both units.

including

Response

Response: As stated in response to the question question under the first bullet, adequate adequate design margins are available available in the CSSTs under the worst case configuration. Both CSSTs C and D operate within their OA ratings ratings even when all shutdown boards boards are powered from one transformer are powered transformer due to outage outage of one offsite power source. As stated earlier, both CSST A and B operate within their FA ratings when used as alternate B operate alternate to CSST D and C respectively. See Attachment Attachment 7 for inputs, assumptions and summary of the output results with acceptance acceptance criteria.

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ENCLOSUREI 1 ENCLOSURE Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 8.2 - 2. FSAR Section 8.2.1 states that to provide a stable voltage, CSSTs C and 0D have have high-speed load tap changers (LTCs) automatic high-speed (LTCs) on each secondary, which adjust voltage based on the normally connected shutdown boards.

connected shutdown Provide executive summary of the calculations/analyses Provide an executive calculations/analyses that details the plant loadflow/voltage studies and operations of the load tap-changing tap-changing units including including a detailed discussion of the control voltage setting, the voltage control band, the time-detailed delays for LTC operation, etc. The summary of the calculations calculations must include include inputs, summary of the output results (with acceptance assumptions, and summary acceptance criteria).

Response

Response: The load flow/voltage analysis analysis of the APS system to support two unit unit operation is performed performed under various configurations configurations for steady state and transient conditions (See the matrix in Attachment Attachment 8 for the the configurations analyzed.). This transient analysis (motor starting) is configurations performed considering one unit in accident performed accident and the other unit in in controlled shutdown mode. The short circuit analysis is performed performed upper band of the LTC and considering the highest grid voltage using upper voltage of 169kV.

For the CSST C and D 0 automatic automatic LTCsLTCs are provided on each low low voltage winding of the CSSTs. The optimum setting for the L LTCs TCs was was previously established established prior to Unit 1 start. The LTC settings are set to regulate the 6.9kV shutdown shutdown board bus voltage to 7071V (102.5%). (102.5%).

The lower and upper upper setpoints of the dead band are 7132V (103.4%)

(103.4%)

and 701 7010VOV (101.6%)

(101.6%) respectively. The analysis is performed to evaluate the minimum voltage requirements requirements with LTC setting of 701 OV and a dead band of +/-

7010V +/- 82.2V. The initial time delay for the LTC is 2 seconds and a step time delay delay of 1 second second with each tap step of 1.25%.

CSSTs A and B have been retrofitted retrofitted with an LTC provided on the the primary primary (high voltage) winding of the CSSTs. The LTC settingsare settings are set to regulate regulate the 6.9kV shutdown board bus voltage to 7071 7071VV (102.5%).

(102.5%). The lower and upper upper setpoints setpoints of the dead band are 7132V (103.4%) and 7010V (101.6%)

(101.6%) respectively. The analysisanalysis is performed to evaluate evaluate the minimum minimum voltage requirements with LTC LTC setting of 701 OV and aa dead band of +/- 82.2V. The initial time delay delay for the LLTCs TCs is 1 second and step time delay of 2 seconds with each tap step of 1.05%.

1.05%.

Based on the analysis, adequate adequate voltages are available at the 6.9kV and 480V buses/boards. The transient voltage on the 6.9kV shutdown shutdown boards falls below the degradeddegraded voltage due to "block start" of ESF loads on the accident unit 6.9kV shutdown shutdown boards, but the the degraded voltage relays reset in 5 seconds seconds or less without tripping the the offsite power. See Attachment Attachment 8 for inputs, assumptions assumptions and summary of the output results with acceptance acceptance criteria and Attachment Attachment 8 for summary summary of the board voltages and degraded voltage relay reset time. Loading Loading of all 6900V-480V 6900V-480V transformers, transformers, safety related boards boards and MCCs was determineddetermined to be within the the E1-68

ENCLOSURE ENCLOSURE 1 Response to Response to Preliminary Preliminary RAISRAIS and and RAls RAIs Regarding Regarding UnitUnit 22 FSAR FSAR Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 equipment rating and no overloading was identified.

equipment identified.

Section 8.3.1 8.3.1 - Alternating Alternating Current (AC) Power Current CAC) Power Systems (On (Onsite) site) 8.3.1 - 1.

8.3.1 FSAR Sections 8.2.2 and 8.3.1 describe the degraded degraded voltage (27DAT, DBT, DCT) and loss of voltage voltage relays (27LVA, LVB, LVC). LVC). The degraded-voltage degraded-voltage relays as as described described in the FSAR have aa voltage setpointsetpoint of 96 percent percent of 6.9 kV and a timetime delay of 10 seconds. FSAR Section delay states that to provide aa stable Section 8.2.1 states stable voltage, CSSTs C and D have automatic high-speed CSSTs high-speed L TCs on each secondary, which adjust LTCs voltage based on the normally connected shutdown shutdown boards. The recent NRC Component Design Basis Inspections (CDBI) (CDBI) indicated indicated issues associated associated with calculations to support the degraded calculations degraded voltage setpoints. In view of the CDBI findings, provide the below listed information information with regard to WBN Unit 2.

a. Provide an executive executive summary of the results of calculations/analyses calculations/analyses which detail detail plant load flow and voltage drop studies, and operations of the load tap-changing units including aa detailed discussion detailed discussion of the voltage setting, the control voltage the voltage voltage control band, time-delays associated with LTC operation, etc. The The summary of the calculations summary calculations must include inputs, assumptions, assumptions, and summary of the output results (with acceptance acceptance criteria).

Response: See the response to RAJ RAI 8.2 - 2.

b. Provide a summary summary of the analyses analyses (steady state and transient) that demonstrates demonstrates that the above degraded voltage trip set points are adequate to safety-related equipment required for design basis events and also protect all safety-related to provide the required minimum voltage voltage at the equipment terminal to start and run all loads consistent with the accident accident analysis assumptions assumptions without without crediting the L TCs of the CSSTs.

LTCs Response: As a result of NRC NRC Electrical Electrical Distribution System Functional Inspections (EDSFI)

Inspections (EDSFI) at various various plants, concerns were raised about the adequacy degraded undervoltage adequacy of degraded undervoltage relays and time time delay delay setpoints. instrumental in forming an industry setpoints. TVA was instrumental working group with otherother utility members and developing developing aa set of guidelines for the required analysis to determine determine proper degraded degraded voltage setpoints and time delays. These These recommendations recommendations were used used to develop the TVA degraded voltage analysis methodology. TVA analyzed the degraded voltage voltage protection scheme for Unit Unit 1 based on the the aforementioned aforementioned methodology. The same approach was was adopted adopted to perform the analysis analysis for Unit 2. This approach approach is consistent with IEEE 741-1997, "IEEE Standard consistent Criteria for the Standard Criteria the Protection of Class 1E E Power Systems and and Equipment Equipment in Nuclear Power Generating Stations" Annex Power Generating Annex A, "Illustration of concepts concepts associated with degraded voltage voltage protection".

Capability Capability to start Class 1E motors motors for Unit 2 has been evaluated evaluated in two ways (similar to Unit 1) as follows:

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ENCLOSUREI 1 ENCLOSURE Response to Preliminary RAIS and RAls RAts Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391

• To evaluate maximum evaluate the plant response to a DBE, maximum (block-start) loading is applied applied in conjunction with the the maximum grid drop down down to the minimum expected grids expected grids voltage of 153kV. This dynamic motor starting analysis analysis demonstrates demonstrates that even though the shutdown board voltage voltage drops drops into the degraded degraded voltage relay operating rangerange during during the momentary momentary voltage dip, the voltage recovers voltage recovers above above the reset value within the degraded degraded voltage relay time delay. This analysis also demonstrates demonstrates that all equipment equipment required to mitigate receive mitigate an accident receive sufficient sufficient voltage to start and accelerate accelerate within the required required time.

• The analysis is performed performed with a maximum maximum grid drop of 11 kV in case of normal alignment from CSST C &D and and with grid drop of 9kV with alternate alignment from CSSTs CSSTs A&B.

A&B.

• Degraded voltage analysis evaluates Degraded evaluates the capability to individually start and run class 1 E motors motors at steady state conditions. This analysis ensures that all motors have have adequate starting voltage at the upper boundary boundary of the the degraded voltage relay setpoint setting (6672V) and and adequate running voltage voltage at the lower boundary of the the degraded voltage relay setpoint setting (6555V). The upper upper boundary was chosen boundary chosen because this is the lowest voltage lowest voltage that guarantees guarantees offsite power supply recovery from a DBE DBE transient. The upper upper boundary boundary setpoint setpoint has been revised to 6681V; therefore, the analysis analysis performed performed with 6672V is conservative.

c. Provide executive summary Provide executive calculations/analyses for settings of the loss-of-summary of calculations/analyses voltage Relays (27LVA, LVB, LVC) provided provided at 6.9 kV shutdown shutdown boards. The The summary of the calculations calculations should include include criteria, assumptions, and output results results

Response

Response: The Loss of Voltage (LOV) relay voltage setpoint upper upper limit is established established to be less than, with margin, voltage margin, the safety bus voltage equivalent to the design equivalent design calculated worst case case transient voltage voltage dip during during accident accident loading sequence. The setpoint should be be less than the TPS calculated calculated security boundary voltage voltage which is based based on the grid voltage voltage that is one failure away away from collapse.

The LOV relay voltage setpoint lower limit is selected by by evaluating evaluating operation operation of the APS under under steady state (running)

(running) conditions, with the 6.9kV shutdown board voltage as low as as possible possible while keeping all connected connected safety related motor loadsloads above their stall voltage (greater than 70.7% of rated motor voltage for NEMA Design B motors). The lower limit must also be greater than the 6.9kV shutdown shutdown board voltage equivalent to E1-70 E1-70

ENCLOSURE11 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar NuclearNuclear Plant - Unit 2, Docket No. 50-391 having the lowest switchyard switchyard voltage that could be sustained without instability instability or collapse.

The lowest lowest boundary boundary of the LOV voltage relay time delay should be long enough to ride through short circuits and other short short time system transients (lightning strikes, switching switching transients, etc.) taking into account the total sensing and clearing times for the above type of events. The time delay upper boundary should be less than the safety analysis time allowed for loss of of voltage detection.

Based Based on the above criteria, the LOV relay lower lower and upper limits were calculated calculated to be 5968V and 6060V respectively. The The associated associated time delay relays lower lower and upper limits are at 0.4 seconds and 1.14 1.14 seconds, respectively.

The lowest 6.9kV 6.9kV shutdown shutdown board voltage which will prevent prevent safety related motors from stalling was determined to be 5600V.

The settings of LOV relays are not impacted by Unit 2 since the the design calculated calculated worst case transient voltage dip during accident sequence used to determine accident loading sequence determine the upper boundary setpoint is bounding bounding based on the Unit 1 and Unit 2 analysis. Also the stall voltage of 5600V was determined determined to bebe acceptable acceptable for Unit 22 motors.

8.3.1 - 2. According to the FSAR, Figure 8.1-2A, the CSSTs, C and D normally supply Train 'A' 0 normally (1A and 2A) and Train 'B' (1 (1 (1Band B and 2B) shutdown shutdown loads, respectively the respectively relating to the two units. In In case of loss of either CSST, the loads fed from the corresponding corresponding CSST shutdown buses are automatically automatically transferred to other CSST via the automatic automatic transfer transfer scheme. .

Provide description description of the automatic transfer scheme from normal to alternate source (whether fast - how many cycles etc.)etc.) as itit relates to WBN Unit Unit 2. Also, explain the the transient behavior of loads that were already runningrunning on the shutdown boards.

Response

Response: The 6.9kV shutdown shutdown boards have been providedprovided with an automatic automatic fast bus transfer scheme. Since all Unit 1 and Unit 2 shutdown boards boards are required to support Unit 1 operation, the existingexisting bus bus transfer scheme scheme is operational and is not affected by Unit 2.

The existing existing automatic automatic bus transfer scheme has been evaluated.

evaluated. The The fast transfer scheme for the 6.9kV shutdown boards boards from normal toto alternate source is a trip signal given to the normally alternate normally closed breaker with a "b" contact on the closed breaker initiating the close command for the alternate alternate breaker. The transfer scheme uses an early "b" contact. An early "b" contact operates at the time the arcing contacts separate. Since the closed breaker breaker gives a close command to the the second second breaker upon opening of the arcing contacts, the "dead bus bus E1-71

ENCLOSUREI 1 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 time" is the time required for second breaker to close. The fast bus time" bus transfer can be either operator initiated initiated (manual)

(manual) or automatic (fault initiated).

initiated).

The 6.9kV shutdown AM-7.2-500 breakers. Per the shutdown boards have AM-7.2-500 the data sheet, the "dead bus time" for operator initiated fast vendor data (manual) using early "b" contact is 3.1 cycles. The times transfer (manual) times given in the vendor data sheet are nominal times with a tolerance tolerance of

+/- 1 cycle.

sequence of events is the For a fault initiated transfer (automatic), the sequence the protective relay actuation, fault followed by the protective actuation, lockout relay (LOR) operation and breaker operation overcurrent relay operating breaker operations. The overcurrent approximately 4 ms and the LOR relay operates in 13 - 14 time is approximately 14 ms.

The minimum time for the breaker breaker contacts to open is approximately approximately resulting in a fault isolation time of 40 ms resulting of 57 ms to 58 ms. The longer the time period that a board is connected connected to aa fault the less energy the the transferred. The decrease operating motors will have when transferred.

operating kinetic decrease of kinetic decrease in magnitude of the motor's generated energy results in a decrease ac voltage and the resultant volts/hertz that the motors would would experience on transfer. Industry standards experience standards issued recommend recommend the the (calculated in per unit on the motor base) should resultant volts/hertz (calculated exceed 1.33 and this criterion was used to evaluate not exceed evaluate adequacy adequacy of the fast bus transfer scheme at Watts Bar.

The Fast Bus Transfer dynamic fast bus transfer Transfer Analysis is a dynamic analysis using ETAP computer software 6.5N1. The analysis software version 6.5N1. analysis considers two fast bus transfers, operator initiated (manual) and and initiated) with three different loading conditions:

automatic (fault initiated) operation, (2) accident SI-phase (1) normal operation, SI-phase B, and (3) light loaded (one motor per bus). Based on the analysis, itit is concluded that the the automatic and manual 6.9kV fast bus transfers between that the automatic between the the normal and alternate acceptable for all board loading breakers are acceptable alternate breakers loading conditions.

conditions.

8.3.1 - 3. sequence of loads applied following a loss of preferred FSAR, Table 8.3-3 shows sequence preferred (offsite) power (from the time of closing of the generator breaker breaker connecting thethe shutdown board). However, in Section generator to the shutdown diesel generator Section 8.3.1.1 8.3.1.1 (under subheading 'System Operation', standby (onsite) power Operation', itit is stated that "The standby power automatic sequencing system's automatic sequencing logic is designed automatically connect designed to automatically the connect the required sequence should required loads in proper sequence should the logic logic receive an accident accident signal prior to, concurrent with, or following a loss of all preferred (offsite) nuclear units and preferred ali nuclear statement with respect to WBN Unit 2, explain the power." Regarding the above statement the design of the automatic sequencing sequencing logic.

Response

Response: standby (onsite) power system's automatic The design of standby automatic sequencing logic for Unit 2 is identical to the logic for Unit 1. The load sequencer consists of discrete individual load's discrete relays which are part of the individual load's breaker control control circuit. InIn the event of a loss of offsite power, the the block starting breaker closure circuit path is opened opened and the sequence E1-72 E1-72

ENCLOSURE11 ENCLOSURE Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Plant - Unit 2, Docket Nuclear Docket No. 50-391 starting path is closed using the BOX or BOY relay. Once voltage is restored to the board from the diesel diesel generator, the sequencing timer for that load is initated by the UVX or UVY relay. Once the timer is timed out, the breaker receives a close signal.

8.3.1 - 4.'

4., In In Section Section 8.3.1.1 of the FSAR (under the subheading FSAR (under subheading "Equipment Capacities"), itit is stated stated that "Tables 8.3-4 through 8.3-7 presentpresent the bus rating, connected connected load, and maximum demand demand load for each electrical distribution board board in the standby (onsite) power system."

Because of anticipated Because anticipated two unit operation, the NRC staff requires the informationinformation on connected and maximum connected maximum demand loads to assess the capacity capacity and capability capability of the the onsite distribution system for WBN Unit 2.

Response

Response: Unit 2 FSAR Tables 8.3-4 through 8.3.7 describe describe the board/bus rating rating in kVA and do not represent the connected load or the maximum maximum demand. This kVA rating is calculated demand. calculated by kVA = =4/3 VI, where V and I

-v3 VI, are the rated voltage voltage and the rated current, respectively.

respectively. Unit 2 6.9kV/480 boards have the same rating as the Unit 1 boards. It 6.9kV/480 It is verified in the APS analysis that loading on all safety related boards is overloading has been identified.

within their rating, and no overloading identified. TheThe loading on all boards, when powered powered from the standby onsite power power system (diesel generators) is enveloped enveloped by the loading in the the calculation referred above.

calculation 8.3.1 - 5. In Section 8.3.1.1 of the FSAR (under the subheading "Standby Diesel Generator Generator Operation"), itit is stated that "For test and exercise exercise purposes, a diesel diesel generator generator may be manually paralleled paralleled with a normal or alternate alternate (offsite) power power source. A loss of offsite power will automatically automatically override the manual manual controls and establish the establish the appropriate appropriate alignment."

Regarding the above statement statement with respect to WBN Unit 2, please please explain what is meant by "appropriate alignment."

Response: The term "appropriate alignment" is explained as follows:

During testing of the diesel diesel generator, should a loss of offsite power and accident accident occur, an accident accident signal will trip the DG feeder breaker.

Tripping the DG breaker will automatically place the DG in in asynchronous mode of operation. As soon as the offsite power asynchronous power supply breaker to the 6.9kV shutdown shutdown board is tripped and the the undervoltage undervoltage load stripping relays operate, the DG feeder breaker breaker to the board will close and load sequencing sequencing logic will be initiated to load load the accident accident loads.

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ENCLOSUREI 1 ENCLOSURE Response to Preliminary RAIS and RAls Response RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 8.3.1 - 6. In order In generator capacity order to verify the adequacy of the diesel generator Section capacity stated in Section 8.3.1.1 of the FSAR for WBN Unit 2 loading, provide the following information: information:

(a) expected diesel generator load profile (considering Worst case expected (considering both auto and manual loads) occurring in one unit and loads) during first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of accident occurring shutdown of other unit and assuming single failure of one diesel generator generator or assuming single failure of "A" generators of same "A" train or "B" train (two diesel generators same train out).

Response: There are four diesel generators which supply onsite onsite power toto four 6.9kV 6.9kV shutdown 1A-A, 1B-B, 2A-A and 2B-B with shutdown boards 1A-A,.1 generator dedicated one diesel generator dedicated to one shutdown board (Unit 1 and Unit 2; Train A and and B). The ESF loads for Unit 1I and Unit 2 are separated different shutdown separated on different shutdown boards. Therefore, the Therefore, the worst load on any diesel generator generator will be that of Train A or or Train B of one unit only. Since there is a dedicated dedicated diesel generator for each train and one train of ESF loads is considered to be adequate adequate to safety shutdown shutdown an accident accident unit, an outage outage of one train (two diesel generators) generators) will not affect thethe ability to safely safely shutdown the unit.

performed to verify A calculation was performed adequacy of the DGs to verify adequacy power all the required ESF loads for an event. The calculation calculation was performed using the same ETAP database database which was was used to perform perform APS analysis. Each diesel generator generator was was evaluated for the worst case loading under case accident loading under LOOP andand LOOP+SI Phase A or B. The calculated LOOP+SI the loading for the calculated loading automatically sequenced loads and required manual loads automatically loads is determined to be within the DG ratings for the analyzed determined analyzed DBEs.

Table below depicts the worst case loading of all scenarios and the margin available.

Maximum Steady-State Loading.

Maximum hrs to 2 hrs Loading, 0 hrs hrs Minimum Minimum Margin Margin 1A-A 1 A-A 11B-B B-B 2A-A 2B-B Rating (%)

(%)

kVA 4847.06 4721.16 4688.42 4688.42 4831.87 6050 19.9 19.9 Event/Time EventiTinie SIB SIB SIB SIB SIB 1810s 1810s 1810s 1810s 1810s 1810s 1810s El-74 E1-74

ENCLOSURE 1 ENCLOSURE Response to Preliminary Preliminary RAIS and RAIs Regarding Unit 2 FSAR RAls Regarding Tennessee Valley Authority .- Watts Bar Nuclear Plant*- Unit Nuclear Plant Unit 2, Docket No. 50*391 50-391 Maximum Steady-State Maximum Steady-State Loading, Loading, 2 hrs to End)

End)

Minimum Minimum Margin Margin 1A-A 1A-A 1B-B 1B-B 2A-A 2B-B Ratinq RatinQ (%)

(%)

kW 4207.45 4077.12 4079.96 4198.79 4198.79 -4400 4400 4.97*

4.97*

Event/Time EventlTime SIB SIB SIB SIB SIB 7200s 7200s 7200s 7200s 7200s kVA 4847.06 4847.06 4721.16 4721.16 4688.42 4831.87 5500 11.8 11.8 Event/Time EventlTime SIB SIB SIB SIB SIB 72005 7200s 72005 7200s 72005 7200s 7200s 7200s

• Excludes load of 125V DC spare charger. Also all pumps are considered Excludes operating for the entire period.

operating period.

Maximum Transient Loading, 0 to 180 sec Maximum Minimum Minimum Margin Margin 11A-A A-A 1B-B 1B-B 2A-A 2B-B 2B-B - Rating Rating (%)

(%)

kW 3937.65 3520.30 3459.14 3878.36 4785 17.7 17.7 Event/Time EvenUTime SIA 35s355 35s SIA 355 35s SIA 355 35s SIA 355 Transient Loading, 180 sec to End Maximum Transient Maximum End 11A-A A-A 1B-B 1B-B 2A-A 2A-A- 2B-B Rating Minimum Minimum Margin Margin

(%)

(%)

kW 4736.24 4498.05 4481.74 4755.44 5073 6.2 Event/Time EventlTime SIB 184s SIB 1845 SIB 184s SIB SIB 184s 184s 184s Increase (Excitation), 0 sec to End Maximum Step Load Increase Maximum Minimum Minimum Margin Margin 11A-A A-A 1B-B 1B-B 2A-A 2B-B Rating (%)

(%)

kVA 4111.29 4397.35 3885.44 3944.21 8000 45.0 Event/Time EventlTime SIB Os Os SIB Os Os SIB Os Os SIB Os Os E1-75

ENCLOSURE 1 ENCLOSURE

Response

Response to Preliminary RAIS and RAIs Regarding Unit 2 FSAR RAls Regarding Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit Nuclear Plant Unit 2, Docket No. 50-391 (b) Confirm factors such as cable losses, pump run-out conditions, power factor, off-nominal frequency and off-nominal voltage, motor efficiency efficiency have been accounted accounted in the diesel generator generator load profile profile calculations.

Response

Response: The ET AP DG loading analysis considers ETAP considers the cable losses, power factor, and efficiency efficiency as documented documented in the load data data editors of ETAP.

editors ETAP. The off-normal off-normal frequency and voltage voltage (for diesel loading) is not considered in the diesel generator loading loading analysis. TVA's TVA's position on this is that TVA has added administrative limits to the plant plant operating procedures for both operating procedures DG voltage voltage and speed range. These These administrative limits are so tight that there would be negligible negligible impact impact on the DG loading due to off-normal frequency frequency and voltage.

8.3.1 - 7.

8.3.1 In Section Section 8.3.1.2.3 8.3.1.2.3 of the FSAR (under the subheading "Underground"Underground Cable Cable Installation"), it is stated that "Cables are designed Installation"), operate in wet conditions. The designed to operate The Class 1 E cables required required to operate the plant in the flooded condition condition are continuous continuous or provided with a waterproof waterproof splice in aa manhole. Cables have been tested at the the factory by the manufacturer manufacturer according according to TVA specifications, which invoke Insulated invoke Insulated Cables Cables Engineers Engineers Association Association (ICEA, formerly formerly IPCEA) standards standards for cables installed in wet environments." Clarify whether whether the WBN Unit 2 underground underground cables are designed designed for submerged submerged or flooded conditions.

conditions.

Response

Response: Unit 2 underground underground cables are designed for submergedsubmerged or flooded condition. All Unit 2 safety related underground cables that were routed in duct banks to the remote remote facilities like intake pumping station and diesel generator generator buildings were turned over to Unit 1 when Unit 1 was licensed in 1996 and have since been in service supporting service supporting Unit 1.

Section 8.3.2 - DC Power SystemsSystems (Onsite) 8.3.2 - 1. On page 8.3-60 of the FSAR, a description description is given on load assignments assignments with respect to divisional divisional requirements.

requirements. The staff requests additional information on assignment assignment of loads for maintaining maintaining separation separation between loads of different different divisions channels divisions and channels as follows:

a. Provide a detailed detailed discussion on the divisional requirements requirements (i.e., thethe requirements for two and four divisions of separation).

Response: The divisional requirements requirements for Class 1 E E electrical systems are delineated in a Watts Bar DesignDesign Criteria. The Reactor Reactor Protection Engineered Safeguard Protection System (RPS), Engineered Safeguard Features Features Actuation Actuation System (ESF), and Essential Supporting Supporting Auxiliary Systems (ESAS) and and Electrical Electrical Power Power Systems must function to initiate shutdown shutdown of the reactor and initiate engineered engineered safety features, ifif required, under under the conditions produced produced by design design occurring basis event, occurring before, during, during, or after the abnormality requiring requiring protective actions. The number of divisions (channels (channels or trains) is determined determined by the number of independent number independent sources sources E1-76 E1-76

ENCLOSUREI 1 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS RAls and RAls Regarding Unit 2 FsAR RAIs Regarding FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 of power required pictorially required for a given function. Attachment 6 pictorially feature for Unit 1 and Unit 2. Loads are depicts this design feature according to the load's divisional assigned to the systems according requirements. Auxiliary Power System requirements. The Auxiliary System consists of two two trains, A & & B. 125VDC vital dc and 120VAC B. 125VDC 120VAC instrument instrument power assigned to the four redundant loads are assigned redundant channels (I, Ill, and II, III, (I, II, IV). The 125 VDC loads primarily associated associated with Unit 1 I are are assigned to Channels I and II primarily II while the loads primarily associated assigned to Channels associated with Unit 2 are assigned Channels III and IV. IV.

b. Describe methodology that you used for distributing the nondivisional Describe the methodology nondivisional loads loads among the four channels.

Response

Response: As stated in the FSAR (page 8.3-56), non-divisional non-divisional loads loads associated with Unit 1 are assigned to Channels associated Channels I or IIII and non-divisional loads are assigned to Channels III or IV.

Unit 2 non-divisional IV.

distribution of loads is done to minimize the impact of a This distribution loss of a battery to more than one unit.

FSAR page 8.3-64 states the following:

FSAR "A battery service test, conducted in accordance battery service accordance with the procedures procedures of Section 6.6 6.6 of IEEE Standard Standard 450-1980 or modified performance modified performance test based on Section 5.4 of batteries under conditions as close to design IEEE 450-1995, is also used to test the batteries as practical."

a. In order to credit the modified In modified performance replacement for the performance test as a replacement the service service test itit must completely completely envelope envelope the service test. Provide the duty performance tests (in cycle load profile for both the service and modified performance (in graphic form) to show that the modified performance test completely envelopes envelopes the service generator (DG) service test for each of the vital and diesel generator (DG) system batteries' design duty cycles (i.e., DBAs, station blackout batteries' blackout (SBO), and Appendix R).

Response

Response: limitation imposed by IEEE Standard Due to duty cycle limitation 450-1995, Section 5.4, Watts Bar has not developeddeveloped a modified modified performance test duty cycle or an implementing performance implementing procedure.

b. The latest version of IEEE endorsed by the NRC is IEEE Standard 450 that is endorsed IEEE 450-2002. The NRC has not endorsed IEEE Standard IEEE Standard 450-2002. Standard 450-technical basis for selecting the IEEE Standard 1995. Provide the technical 1995. Standard 450-1995 450-1995 instead of the IEEE Standard Standard 450-2002.

Response: IEEE Standard 450-1995 is cited in the FSAR and Technical IEEE Specifications Specifications Bases for Unit 1 relative to requirements for modified modified performance applicable to Unit 2.

performance tests tests and is applicable E1-77 E1-77

ENCLOSURE ENCLOSUREI 1 Response to Preliminary RAIS and RAls Regarding Unit 2 FSAR RAIs Regarding Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391

c. Clarify whether the battery service test(s) verify the design duty cycles cycles for DBAs, SBO, and Appendix Appendix R scenarios.

Response: service test duty cycle is determined in The service in a Unit 11 1/ Unit 2 vital battery sizing calculation from a review of battery load duty battery Battery I,1,11, associated with Battery cycles associated II, 111, & V for the SBO, IV &

III, IV LOCA/LOOP and Appendix R cases. A composite load duty LOCAILOOP batteries was established cycle for the batteries established manually manually by identifying identifying maximum positive plates required the maximum required plus the maximum current maximum current required during the first minute. Design margin was included in in calculations for the battery the load values obtained from the calculations duty cycle used for testing.

The duty cycle chosen represents represents the worst-case worst-case duty cycle for an SBO event wherein wherein worst-case worst-case means that battery load case case controls the battery size (more positive controls positive plates) and voltage voltage (minimum (minimum voltage). The LOCA/LOOP LOCAILOOP cases cases were considered for first minute only because the total duty cycle is much smaller than the SBO duty cycle. The "Appendix R" cases do not have because the duty an impact because duty cycles bounded by SBO.

cycles are bounded Based on review of the load study case results, Load Study Cases for Battery I LOCA and Battery Cases Battery III SBO representrepresent thethe case loading for the first minute and complete duty cycle worst case cycle 1 - 240 minutes minutes respectively batteries 1, respectively for batteries I, 11, II, 111, III, and IV && V.

8.3.2 - 2.

8.3.2 FSAR page 8.3-66 8.3-66 states the following:

125V dc Class 1 E electrical systems were designed, The 12SV components fabricated, designed, components fabricated, and are or will be installed installed meeting requirements of the NRC 10 CFR Part 50 meeting the requirements SO Appendix A General Appendix General Design Criteria, IEEE Standard 308-1971, IEEE Standard 308-1971, NRC Regulatory Guides (Revision 0)

Guides 1.6 (Revision 0) and 1.32 (Revision 0), and otherother applicable criteria as as enumerated herein.

herein.

a. how the system design meets the guidance Explain how guidance provided provided in RG 1.32 and IEEE Standard 308-1971 with regard to sharing DC power sources at a multi-unit nuclear power plant site.

multi-unit Response: chargers are demonstrated The vital batteries and battery chargers through analysis to have the capability capability to supply shared loads loads would result from an accident including those that would including accident in one unit unit and safe shutdown of the second unit. Sharing Sharing of dc power at a addressed in Regulatory multi-unit power plant site is addressed Guide 1.81.

Guide addressed in Unit 2 FSAR section 1.81. This is addressed section 8.3.2.2 8.3.2.2 under 125-volt dc system for compliance under Analysis of Vital 12S-volt compliance to to Regulatory Regulatory Guide 1.81,1.81, Position C2 and Section 8.3.2.4, Redundant DC Power Systems. The safety Independence of Redundant Independence assigned to the vital 125-V loads are assigned 12S-V batteries so that sharing significantly impair their ability to perform will not significantly perform their safety functions including in the event of an accident accident in one unit and E1-78 E1-78

ENCLOSUREI 1 ENCLOSURE Response to Preliminary Response Preliminary RAIS and RAIs RAls Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 orderly shutdown shutdown and and cool-down of the remaining unit while while considering the effects of a single failure. ESF loads loads are are assigned to batteries batteries I and and IIII for Unit 1 and to III and IV for Unit 2 such that an accident Unit accident in one unit will not adversely adversely impact battery sources sources for ESF loads of the other unit. The TDAFW controls are supplied by battery III (normal) or battery IV battery IV (alternate) for Unit 1 and battery I (normal)

(normal) or battery II II (alternate) for Unit 2. The loading loading impact impact of these for both units units is considered considered in the analysis.

The Watts Bar 125V 125V DC power system system meets meets the requirements requirements of Section C.b of Regulator Regulator Guide 1.32 (Battery Charger Charger Supply). The battery chargers capacity to support the chargers have the capacity the steady-state operation of the connected steady-state connected loads required during during normal normal operation operation while maintaining maintaining its battery in a fully charged state and have sufficient capacity capacity to restore the battery from the the design design minimum discharged discharged state to its fully charged statestate within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> hours in Loss of Coolant Accident (LOCA) and and 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> in Station Blackout (SBO) events while supplying supplying applicable applicable steady-state steady-state loads.

b. Provide an executive summary summary of the results of calculations calculations used for determining the size of the inverters, determining inverters, chargers, batteries, and fuses (the scope scope of this request includes the normal normal and 125 Volt (V) DG battery systems).

Include key inputs and assumptions and a discussion discussion of margins in youryour response.

Response: Vital Battery Batterv and Batterv Battery Chargers Analysis of the Watts Bar 125V125V DC vital battery system has has determining the size of battery and battery been performed for determining chargers for the design basis conditions. The analysisanalysis developed load profiles for each battery for two unit operations operations under the following design basis conditions:

a) Station Blackout (SBO): loss of both offsiteoffsite and onsite onsite acac sources without without accident accident (four hours) b) Accident: accident accident with loss of offsite power chargers power and chargers plus a single failure (30 minutes, LOCAILOOP LOCA/LOOP in one unit & &

orderly shutdown of the second unit) c) Normal Operation: charger Normal Operation: charger loading loading during normal operation d) d) Appendix-R: Loss of ac power to chargers and inverters; maintain maintain reactor at hot shutdown shutdown for a period of two hours.

Appendix R event loading is bounded by SBO.

This calculation calculation establishes establishes minimum available voltage and minimum available short circuit current at the battery boards and downstream downstream 125V E1-79 E1-79

ENCLOSUREI 1 ENCLOSURE Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Authority - Watts Bar Nuclear Tennessee Valley Authority Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 switchgear and control panels DC control buses at switchgear panels to verify their their adequacy adequacy for continuous operation operation within their voltage and current ratings. Design inputs and assumptions current assumptions for this this calculation calculation are included in Attachment 9.

The analysis and methodology accordance with IEEE 485, methodology is in accordance Recommended Practice "IEEE Recommended Storage Practice for Sizing Lead Storage Batteries for Generating Stations and Substations" and IEEE IEEE 946, "IEEE Recommended Recommended Practice for Design of Safety DC DC Generating Stations." The analysis Auxiliary Power Systems for Generating analysis is performed performed using ETAP-DC Version Version 5.5.6N.

Loading Loading The capacity Batteries I,1,11, 125V Vital Batteries capacity of the existing 125V II, 111, III, IVIV and V with 16 positive plates is adequate adequate to support support two unit unit operation under all design conditions and the size of the existing existing chargers is adequate battery chargers battery adequate for recharging the batteries while while feeding the applicable continuous simultaneously feeding continuous loads. Both batteries chargers meet the sizing requirements as batteries and battery chargers as stated in the Unit 2 FSAR Section 8.3.2. The results from the the most recent revision (R8) of the sizing calculation calculation are summarized as follows:

summarized Battery Size Positive Plates Plates (PP)

Battery Available CaIc.

Calc. PP CaIc. PP Calc. CaIc.

Calc. PP PP No. PP SBO LOCA+LOOP LOCA+LOOP APP.

APP.R R I1 16 16 15.90 15.90 5.40 11.21 II II 16 15.23 5.30 10.78 10.78 III 16 15.90 4.75 10.12 10.12 IV IV 16 14.97 4.87 9.59 9.59 V 16 15.90 5.40 11.21 Load shedding on vital inverters 1-1, 1-1, 2-1, 2-1, 1-11 1-11 and 2-11 is required required to be completed within 30 minutes minutes in accordance accordance with Unit 2 8.3.2.1.1, Load Time of Application, FSAR Section 8.3.2.1.1, conclude Application, to conclude that the selected battery is adequate for the four (4) hour SBO SBO specific loads which may be turned coping duration. The list of specific off to shed load is documented documented in the inverter loading calculation and design output drawings.

E1-80

ENCLOSURE 1 ENCLOSURE Response to Preliminary RAIS and RAls Regarding Unit 2 FSAR RAIs Regarding Tennessee Tennessee Valley Authority Authority - Watts Bar NuclearNuclear Plant Plant - Unit Unit 2, Docket No. 50-391 Margins Margins Margins Margins for aging and minimum temperature have been applied minimum temperature accordance with IEEE in accordance IEEE 485 methodology. Design margin margin has has been identified identified on each battery battery board for accommodation accommodation of future loads. An additional additional margin included calculation is included in the calculation load shedding that load shedding effects are credited credited at 45 minutes rather than 30 minutes. The worst case loading for all batteries, considering conservative considering conservative application application of connected connected loads and design margin allowances, allowances, requires less than the rated 16 positive positive plates per cell.

cell. Provision is made made in the calculation calculation for evaluating and incorporating effects of future design incorporating the effects design changes on design margins.

Voltages Voltages Available Available voltage voltage at the battery battery terminals and downstream control buses exceeds exceeds the required minimum voltage when they are fed from normal or alternate batteries with all battery are battery cellscells are in service. The available maximummaximum voltages at busses busses and devices was evaluated evaluated and determined to be acceptable. The The batteries are capable batteries capable of supporting aa minimum minimum terminal voltagevoltage of 114V (117.8V, Battery 114V (117.8V, Battery V) for the first minute minute and 105V (108.5V, (1 08.5V, Battery V) at the end of the discharge discharge period.

Short Circuit Circuit Available short circuit currents at battery board I,1,11, II, 111, III, IV, Vand V and associated downstream downstream safety related control buses buses were determined determined and evaluated evaluated with respect to interrupting rating capacity capacity of the protective protective devices. The evaluation determined determined that the interrupting ratings of the breakers and and fuses for the the existing design on the battery boards and downstreamdownstream control buses are adequate.

adequate.

Service Service Test Duty Cycle Cycle A service service test duty cycle cycle which bounds worst-case worst-case load load conditions for all batteries for operation operation of Units Units 1I and 2 is established by this calculation. Refer to Unit 2 FSAR established Table 8.3-12.

Vital DC fuses Fuses Fuses for the 125V 125V dc system include battery board fuses for system include battery main feed, vital battery charger and vital inverter battery inverter feeds, non-safety related load bus groups, and control fuse assemblies non-safety assemblies consisting of 250 sets of 5-amp signal fuses used used to supply miscellaneous miscellaneous control circuits for solenoid valves and control relays. Additional fuses analyzed analyzed includes includes fuses for medium E1-81

ENCLOSURE1 ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAls RAts Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 and low voltage switchgear switchgear control circuits, 5-amp 5-amp signal fuses used used for power distribution in relay racks for relay control circuits circuits and fuses used for supplemental supplemental circuit protection or isolation in in various control verifies fuses are applied control circuits. This analysis verifies applied within their ratings for continuous operation, have adequate adequate dc dc voltage ratings and interrupting interrupting capacity, provide cable provide cable protection where required and that they are appropriately appropriately coordinated coordinated with upstream protective protective devices.

A supplemental analysis specifically for Unit 2 circuits hot A supplemental analysis specifically for Unit 2 circuits hot already service for Unit I1 has been performed.

already in service performed.

Vital Inverters Inverters 120V AC Vital Inverter 120V Inverter loading loading analysis considering effects considering the effects of Unit 2 loads has been performed.

performed. The purpose of this this adequacy of the 120VAC analysis is to verify the adequacy 120VAC Vital Instrument Power System to supply loads powered Instrument powered from 120VAC 120VAC Vital Instrument Instrument Power Boards Boards 2-1, 2-11, 2-111 and 2-IV.

2-1,2-11,2-111 Each power board is powered powered by a static inverter inverter rated at 20kVA 20kVA @ 0.8 to 1.0 1.0 power factor. The boards each contained contained 48 branch circuit breakers supplied from the main bus in groups groups of 12 through fused sub-distribution buses. The analysis analysis performs the following listed objectives:

• • Determine and document branch Determine branch circuit loads and verify the the adequacy of the breaker trip ratings.

• 0 Calculate sub-distribution bus loads and verify adequacy adequacy of fuse ratings.

• 9 Calculate overall power board steady-state load and power power factor and verify capability of inverter inverter and power board main bus to service the load.

Limits Load Limits Limitations are placed placed on the Unit 1 and Unit 2 120VAC 120VAC Vital Instrument Instrument Power Boards 1-1, 1-11, 1-111, I-IV, 2-1, 1-1,1-11,1-111, 1-IV, 2-1, 2-11, 2-11, 2-111 and 2-IV due to restraints imposed by battery capability for two unit operation, and these limits are tabulated below. Unit 2 FSAR Table 8.3-11 has been revised to show the following load limits limits for the Unit 1 and Unit 2 120VAC Vital Instrument Power Instrument Power Boards.

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ENCLOSURE11 ENCLOSURE

Response

Response to Preliminary Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 Load Limits (kVA)

Channel Channel 1-1 1-11 1-11 1-111 1-IV 1-IV Rating 20 20 20 20 20 Load Limit 14 14 10.5 10.5 10.5 Limit

• SBO Load Limit* 10 10 10 10.5 10.5 10.5 Channel Channel 2-1 2-11 2-111 2-IV 2-IV Rating 20 20 20 20 Load Limit 14 14 14 10.5 10.5 10.5 SBO Load Limit ** 10 10 10.5 10.5 10.5

• Shedding Shedding of non-required non-required loads within 30 minutes of the onset of a SBO event may be necessary necessary to reduce reduce actual actual loading on distribution 1-1, 1-11, 2-1 distribution boards 1-1,1-11,2-1 and 2-1.

2-1. Loads which may be shed to achieve the SBO SBO load limit are identified on load summary summary table with the the designation "LS" adjacent to the breaker designation "LS" breaker number and are listed as design design output on drawings.

Branch circuit breakers are properly properly sized and the adequacy adequacy of the breaker breaker trip ratings is verified for continuous continuous full load' load and inrush currents.

70A sub-distribution sub-distribution bus fuses are properly sized to supply all bus loads.

Loading for all of the 120VAC 120VAC Vital Instrument Power Boards is Power Boards within the inverter inverter ratings and allowable load limits. The Vital Inverters Inverters are properly properly sized to supply supply power for power board loads. The results of the most recent revision (R131) (R1 31) of the the inverter loading calculation calculation are as follows:

E1-83

ENCLOSURE11 ENCLOSURE Response to Preliminary RAIS and RAls Response RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 Calculated Loading Inverter Ratings and Calculated Loading (KVA)

Channel 2-1 2-11 2-111 2-IV 2-IV Rating 20 20 20 20 Load Limit 14 14 10.5 10.5 10.5 Calculated Calculated Load 8.9 9.5 8.7 8.9 8.9 SBO Load Limit 10 10 10.5 10.5 10.5 Calculated Calculated SBO Load 6.1 7.1 8.7 8.9 8.9 Design margins are identified identified and the documentation documentation provides aa means for evaluating and incorporating the effects of future future design changes.

Vital AC fuses Fuses for the 120V ac system includeinclude vital inverter output fuses, distribution board load group group fuses, fuses in series with circuit breakers that provide back-up breakers penetration protection and fuses back-up penetration used for supplemental supplemental circuit protection protection or isolation isolation in various various control circuits. This analysis verifies fuses are applied within within their ratings for continuous continuous operation, have adequate adequate voltage voltage ratings and interrupting capacity, provide cable protection where where required and that they are appropriately required appropriately coordinated coordinated with upstream upstream protective protective devices.

The analysis was revised to incorporate evaluation for Unit 2 The analysis was revised to incorporate evaluation for Unit 2 circuits not previously previously analyzed.

analyzed.

Generator (DG)

Diesel Generator (DG) Battery and Battery CharqerCharger System The 125V dc DG control power system is required to provide provide control power for the control and field flashing of the DG sets under under normal conditions and aa loss of all ac power. Analysis for the DG control power system has evaluated adequacy of the evaluated the adequacy the size of battery and battery chargers. The analysis establishes establishes a duty cycle for a worst case SBO scenario scenario conSidering considering diesel diesel circuit loads based based on sequence application for an assumed sequence of application three start attempts: (1) a failed start attempt at initial diesel emergency emergency start signal, (2) a second failed attempt attempt at 2929 minutes, and (3) a third attempt at the end of the four-hour SBO coping period. The analysis conservatively conservatively assumes that fuel oil pumps and lube oil pumps for both engines each each start simultaneously simultaneously such that starting currents for two pump motors motors E1-84

ENCLOSURE 1I ENCLOSURE

Response

Response to Preliminary RAIS and RAls RAts Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 are applied at the same time. No special operator actions are credited credited to reduce circuit load between between start attempts. The The calculation calculation demonstrates demonstrates that the batteries batteries and battery battery chargers chargers adequately sized.

are adequately The diesel generator generator battery system configuration is shown on FSAR Figure 8.3-24.

Margins for Diesel Generator Margins Generator Battery Battery Systems Systems The cell sizing calculation calculation utilizes utilizes IEEE-485 methodology and margins for aging and minimum temperature.

considers margins temperature. Design margin remains based on a required 2.75 positive plates versus margin versus the provided provided 3 for the C&D KCR-7 batterybattery cells.

Battery Size Positive Plates (PP)

Battery No. Available PP Calc. PP SBO 11A-A A-A 3 2.75 2.75 1B-B 3 2.75 2A-A 3 2.75 2B-B 3 2.75 2.75 The battery battery chargers chargers are rated 20 amps and can carry carry normal load load currents while recharging recharging the battery within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> for the the analyzed analyzed battery duty cycle.

The calculation calculation provides a means for evaluating evaluating and incorporatingthe incorporating the effects of future design design changes.

Fuses for Diesel Diesel Generator Battery Systems Systems 125V Diesel Generator battery Fuses for the 125V battery system include include distribution distribution panel panel fuses for the battery main main feed, and panel instruments. Additional fuses analyzed include engine include fuses for engine control and lube oil pump control circuits and alarm circuitcircuit calculation verifies that the fuses are isolation fuses. The calculation applied applied within their ratings for continuous operation, have have adequate adequate dc voltage ratings and interrupting capacitycapacity and protective devices.

provide selective coordination with other protective All of the analyzed analyzed circuits circuits are already already in service service for Unit Unit 1. The The analysis analysis was reviewed reviewed for applicability applicability to Unit 22 and it was was determined determined that no additional additional analysis analysis was required. Design required. Design inputs and assumptions assumptions are included in Attachment Attachment 9.

E1-85 E1-85

ENCLOSURE ENCLOSURE 1 I Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 8.3.2 - 3.

8.3.2 FSAR page 8.3-67 states the following:

The normal normal or preferred preferred power source to each distribution board board is from the battery charger, which is supplied from either one of two 480V ac shutdown distribution distribution boards. The battery battery serves as an emergency emergency source in the event the battery charger charger inadequate for the load required. Table 8.3-12 provides maximum source is lost or is inadequate maximum loading for each board for normal, normal, loss of all ac power, and accident accident conditions.

After reviewing reviewing Table Table 8.3-12, the staff noticed noticed that itit does not include the maximum loading values for normal conditions.

conditions.. Provide the maximum maximum loading loading values under normal normal conditions.

Response

Response: The bounding bounding value for maximum charger load other than battery maximum charger charging, which would include normal normal operation, is 120 amps as as specified in Note 2 of Table specified Table 8.3-12. The vital battery system analysis analysis determines determines the loads for each battery including load including the normal load current to be supplied by the chargers. Based Based on the latest revision of the analysis, the maximum normal continuous continuous load for any channel channel is less than 95 amps.

8.3.2 8.3.2 - 4. FSAR FSAR page 8.3-69 states the following:

"Seismic Category I(L) battery battery charger charger V is intended solely to maintain vital battery V in its fully charged state and to recharge recharge itit following its use or testing. At no time will battery charger V be used to supply vital battery system loads. The fifth battery charger charger does not supply dc system loads; therefore, the overvoltage overvoltage and failure alarm alarm relays do not serve any safety or protective function and consequently consequently are notnot required for alarms."

a. Describe how Vital Battery V and its associated components components will be protected against against potential overvoltage conditions when being used used as a temporary replacement for Vital BatteryBattery I,1,11, II, 111, III, or IV.

IV.

Response: When Vital BatteryBattery V (62 cells) is used used to replace Vital Batteries Batteries I,1,11, II, 111, III, or IV IV (60 cells), Battery Charger Charger V is disconnected disconnected from Vital Battery Board V and Vital Battery Board V is aligned through transfer switches and disconnects disconnects to connect Vital Battery V to the main distribution bus of the replaced battery.

The n'ormal normal or spare charger for the replaced channel can be be used to maintain maintain charge on Vital Battery V and supply normal loads. The typical typical charger alignment for this configuration configuration is that the normal charger is isolated from the main distribution bus bus and used to support the disconnected, replaced battery. The disconnected, replaced The spare spare charger charger is aligned aligned to the battery board main distribution bus and output voltage verified per procedure, Standard Operating Instruction SOI-236.5 SOI-236.5 (125V DC VITAL BATTERY BOARD V) to be within the range of 137 to 140 volts, the float voltage voltage for Battery V. This setting is within the allowed maximum voltage of 140V for the affected channel.

maximum channel. Equalizing Equalizing E1-86

ENCLOSUREI ENCLOSURE 1 Response to Response to Preliminary Preliminary RAIS RAIS and and RAIs RAls Regarding Regarding Unit Unit 22 FSARFSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit Docket No.

2, Docket No. 50-391 50-391 Vital Battery VV when Vital Battery when connectedconnected to Vital Battery to Vital Battery Boards Boards I,I, 11III, II, III, or IV isis not or permitted since not permitted since the.the equalize voltage would equalize voltage would exceed exceed the allowed the allowed maximum maximum voltage voltage for for the the channel.

channel.

b.

b. Explain Explain howhow thethe fifth battery is fifth battery is maintained maintained in in aa fully charged state fully charged state and and itsits associated associated equipment equipment is is supplied supplied power power when when used used as as aa temporary temporary replacement replacement for for Vital Battery 1, Vital Battery I, 11,II, 111, III, or IV. In or IV. In your response, include your response, include aa discussion on discussion on thethe capability capability of of the battery charger the battery charger to to recharge recharge the battery and the battery and supply expected supply expected loads. loads.

Response

Response: See See the the response response to to item 4a.above. The item 4a.above. The battery battery charger charger sizing sizing calculation calculation is based on is based on restoring restoring the discharged ampere-hours the discharged ampere-hours plus plus carrying carrying the the normalnormal steady-state steady-state loads. loads. The The Vital Vital battery sizing analysiS considers sizing analysis considers both both the the normally normally aligned aligned battery battery andand Vital Vital Battery Battery V for each V for each channel channel and design basis and design basis condition.

condition.

The The results results of of the the analysis analysis demonstrates demonstrates that that the the ampere ampere hours hours discharged from Vital Battery V is essentially the same as discharged from Vital Battery V is essentially the same as Vital Vital Batteries 1, Batteries I, 11,II,111 III and and IV; IV; therefore, therefore, the the battery battery charger charger sizing sizing calculation calculation is applicable and is applicable and valid valid when when Battery Battery V V is is aligned aligned to to the board.

the board.

8.3.2 8.3.2 -- 5.

5. Provide Provide the title for the title for Section 8.3.2.5 of Section 8.3.2.5 of the the FSAR (located on FSAR (located on page page 8.3-71).

8.3-71).

Response

Response: The The RAI RAI was was thethe result result of of aa review review of of the the red-line red-line version version of of Amendment Amendment 95 to the Unit 2 FSAR. The red-line shows the 95 to the Unit 2 FSAR. The red-line shows the deletion deletion of header 8.3.2.5 of header 8.3.2.5 and and the the first paragraph under first paragraph under this this header.

header. The The software software used used to generate the to generate the markup markup leaves leaves the the header header numbernumber (i.e.,

(i.e., 8.3.2.5) 8.3.2.5) until until the the changes changes are are incorporated.

incorporated.

The issued version The issued version of of page page 8.3-66 contained in 8.3-66 contained in Amendment Amendment 95 95 toto the the Unit Unit 2 FSAR properly reflects the deletion of both the header title and 2 FSAR properly reflects the deletion of both the header title and the the associated associated header header number number (i.e., (i.e., "8.3.2.5").

"8.3.2.5").

8.3.2 8.3.2 - 6. FSAR FSAR page page 8.3-72 8.3-72 states states the the following:

following:

"The "The limiting limiting conditions conditions studies studies was was the the loss loss of of offsite offsite power concurrent with power concurrent with the the failure of failure of one battery. Table one battery. 8.3-13 shows Table 8.3-13 shows the the results results of this study."

of this After reviewing After reviewing the the FSAR, FSAR, the staff could the staff could not not locate locate thisthis Table. Provide the Table. Provide the Table 8.3-13 (or the Table 8.3-13 the results results of of this study) for staff staff review.

Response

Response: Amendment Amendment 100 100 to to the the Unit Unit 22 FSAR FSAR will will add add Table 8.3-13 to Table 8.3-13 to the the FSAR.

FSAR.

E1-87 E1-87

ENCLOSURE ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, DocketDocket No.

No. 50-391 8.3.2 8.3.2 - 7. performance characteristic Provide the performance characteristic curves that illustrate the capability of the the Class 1E Batteries to respond to and supply the most severe severe loading loading conditions conditions at the plant. In In your response, include the performance performance characteristic characteristic curves such as as voltage profile curves, discharge discharge rate curves, and temperature temperature effect effect curves.

Response: Load study cases for each vital battery have been performed performed for all design basis conditions. Battery performance performance curves for voltage voltage profile and discharge discharge rate for the most severesevere loading condition condition are are included as Attachment Temperature effects Attachment 10. Temperature effects are applied inin accordance accordance with Table 1, 1, Cell Size Correction Factors for for Temperature, Temperature, of IEEEIEEE Standard Standard 485.

8.3.2 8.3.2 - 8. FSAR page 8.3-19 states the following:

"The

The diesel generator generator 125V 125V dc battery system's chargers chargers have the capacity capacity to continuously continuously supply all steady-state steady-state loads and maintain the batteries batteries in the design design maximum charged maximum charged state or to fully recharge recharge the batteries from the design minimum minimum discharge discharge state within an acceptable acceptable time interval, irrespective irrespective of the status of the the plant during during which these demands demands occur."

occur."

a. Define the term 'acceptable

'acceptable time interval'.

interval'.

Response: Adequacy of the size of the Diesel Generator Adequacy Generator (DG)

(DG) Battery Chargers Chargers has been analyzed.

analyzed. DG battery chargers are sized to supply all steady state loads and to fully recharge recharge the batteries batteries from design minimum minimum discharge state to fully charged state within less than eight (8) hours. This DG battery recharge recharge time time compares favorably to the recharge recharge time of 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> for 125 125 VDC Vital battery battery following an SBO and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> followingfollowing a LOCA. The period of eight (8) hours is therefore judged to be be an "acceptable time interval."

interval." .

E1-88 E1-88

ENCLOSURE 1 ENCLOSURE1 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSARFSAR Tennessee Tennessee Valley Authority Authority - Watts Bar NuclearNuclear Plant - Unit 2, Docket No. 50-391 8.3.2 8.3.2 - 9. FSAR page 8.3-19 states the following:

"Each of the diesel generator battery system has sufficient capacity to supply required loads for the four-hour station blackout blackout (SBO) period."

Provide the technical basis for the 4-hour period and discuss in detail the required loads the DG battery system will be supplying during the four-hour SBO period Response: Watts Bar SBO event analysis requires that the DG battery supply DG loads without the benefit of aa charger charger for a 4-hour period. This period. This includes includes three diesel attempts: first attempt at t == 0 (initial diesel start attempts:

emergency start), second attempt at t = 29 minutes, and the third emergency attempt at the end of four-hour Because Watts Bar is a four-hour period. Because 4-hour coping plant, this is the technical basis for DG battery sizing.

The following lists the DG battery duty cycle loads:

1.

1. generator control circuit Diesel generator

2. pumps Diesel fuel oil pumps

3. Diesel lube oil pumps pumps

4. Diesel generator generator field flash circuit circuit Blackout Section 8.4 - Station Blackout guidance on Station Blackout (SBO) (10 The staff review guidance (10 CFR 50.63) is given in NUREG NUREG 800, Chapter 8, Section Section 8.4. The staffs staff's review of FSAR Amendments Amendments No. No. 95 and 97 finds that they do not contain information information on Section Section 8.4 for addressing an SBO event in WBN Unit 2. The The NRC staff issued a safety evaluation (SE) report dated March 18, 18, 1993, (TAC Nos. M68624 and and M68625) supplemental SE dated September M68625) and a supplemental September 9, 1993, on WBN compliance with 10 CFR 50.63. The NRC staff believes that the original review of WBN Unit 2 compliance compliance with SBO was performed an SSO performed under TAC No. M68625. Since WSN WBN Unit 2 is now seeking an OL 17 years after the initial review for conformance conformance to the SBO rule, the NRC NRC staff requests that TVA update and/or validate validate the original information, or provide a new submittal on how WBN Unit 2 meets the SBO rule. TVA should also update FSAR Section 8.4 to include include the relevant relevant information information on SBO. The information to be submitted to the staff on SBO SSO for WBN WSN Unit 2 should include the following:

8.4 - 1. The specified coping duration duration to withstand and recover from a SBO based based on thethe factors listed in 10 10 CFR 50.63 and the expected frequency of grid-related grid-related loss of offsite power in the last 20 years.

Response: Watts Bar's SBO design basis is defined defined in Watts Bar (WBN)

(WBN)

Design Criteria WB-DC-40-64, WB-DC-40-64, "Design Basis Events Design Criteria,"

Section 4.41, 4.41, "Loss of All AC Power (Station Blackout Blackout (SBO))."

Compliance with the SBO design basis is documented Compliance documented in WBN calculation EPMMA041592, EPMMA0411592, "Station Blackout Blackout Coping Evaluation," and and referenced information.

referenced information. The SSO SBO coping time, determined determined in E1-89

ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Docket No. 50-391 Nuclear Plant - Unit 2, Docket accordance with NUMARC 87-00 guidelines, is four (4) hours. Areas accordance with NUMARC 87-00 guidelines, is four (4) hours. Areas evaluated include the following:

Condensate Inventory for Decay Heat Heat Removal Using the methodology described in NUMARCNUMARC 87-00, Revision 1, Section 7.2.1, calculate the Condensate Storage Tank (CST) 7.2.1, calculate inventory required to maintain the RCS at Hot Standby without inventory without cooldown for four (4) hours and assess the adequacyadequacy of CST inventory (200,000 gallons).

inventory Auxiliary Control Air (ACA)

Assess adequacy adequacy of ACA system supply supply to the TDAFW pump level control valves and SG-PORVs SG-PORVs since these these are the only SBO shutdown shutdown components components which uses ACA air.

Reactor Reactor System (RCS) Inventory Evaluate RCS inventory loss (seals, letdown) letdown) during during four (4) hour hour SBO coping period.

Class 1 E Battery Capacity Evaluate the Class 1E 125 volt batteries batteries capacity to provide DC power to SBO shutdown shutdown components during the coping period.

Other Battery Systems Other Systems Evaluate the EDG 125 volt DC batteries for their ability to successfully start EDGs at end of SBO SBO to ensure ensure restoration of AC power, assuming two (2) attempted EDG (2) attempted EDG starts at the beginning of the the SBO event.

Appropriate Appropriate Containment Containment Integrity Evaluate mechanical penetration/fluid Evaluate the mechanical penetration/fluid system containment containment isolation isolation valves against the exclusion exclusion criteria criteria of NUMARC NUMARC 87-00, Revision 1, Section 7.2.5. Those valves which did not meet those Revision those exclusion criteria are analyzed further. These These valves are associated associated with penetrations penetrations X-19A X-1 9A and B, X-44 X-44,, and X-1 07. .

E1-90

ENCLOSURE ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 8.4 - 2. Provide a summary summary of the strategies and analysis for coping with SBO for the the specified duration. This discussion should provide sufficient information, information, including including baseline baseline assumptions, on the systems and equipment equipment required for coping with an SBO for the specified duration duration without ac power for the following:

a. The core and reactor system conditions and the ability ability to maintain adequate adequate reactor coolant system (RCS) inventory inventory to ensure that the core is covered and cooled. Discuss and provide information information on RCS inventory taking into consideration consideration shrinkage, leakage from pump seals, and inventory loss from letdown or other normally open lines.

Response

Response: The success criterion for RCS inventory inventory is that the core remains remains covered covered throughout throughout the SBO coping period (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />),

accounting for: 1) shrinkage; 2) letdown; 3) normal system accounting leakage; and, 4) reactor coolant pump (RCP) seal leakage.

1) Watts Bar is a "Hot Standby" plant, so shrinkageshrinkage is not significant. Even Even if if a cool-down cool-down to 350'F 350°F occurred occurred within within the SBO coping period,period, the core would remain covered.

2) The letdown containment loss-of-containment isolation valve closes on loss-of-AC power.

3) The normal system leakage is limited to 10 gpm (identified (identified leakage) by Technical Specifications Specifications 3.4.13.

4) RCP seal leakage leakage is assumed assumed to be 25 gpm per RCP. This This conservative with respect is conservative respect to the 21 gpm provided in WCAP 10541,10541, "Westinghouse Owner's Group Report, 'RCP 'RCP Performance Following Seal Performance Following a Loss of All AC Power.'"

Power."' This This value has also been assessed assessed with respect to seal leak-off leak-off line failure as reported by Revision 2 of WCAP 10541. 10541. TheThe seal leak-off leak-off line will not fail at WBN.

WBN.

The total RCS inventory at the end of the SBO coping period is approximately approximately 8,600 fe ft 3 (assuming no shrinkage) shrinkage) versus a reactor vessel volume of approximately approximately 5,000 fe.ft3. Therefore, the Therefore, the core remains covered.covered.

b. Discuss and provide information information on the capacity condensate storage capacity of the condensate storage tank to ensure ensure that there will be sufficient water inventoryinventory to remove decay decay heat during the specified SBO duration.

Response

Response: Per Technical Specification 3.7.6, the CST shall have at least Technical Specification 200,000 gallons reserved reserved for the Auxiliary Feedwater Feedwater System.

ItIt will take approximately 75,500 75,500 gallons of CST inventory inventory to remove decay heat (without cooldown) cooldown) during the SBO coping period. In In the unlikely event that plant cooldown is required, then the required inventory inventory from the CST is 197,200 gallons.

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ENCLOSURE 1 ENCLOSURE Response to Preliminary RAIS and RAIs Response RAls Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit Unit 2, Docket No. 50-391 c.

c. Discuss and provide information information on the compressed compressed air capacity to ensure that air operated valves required for decay decay heat removal have sufficient sufficient reserve reserve air air appropriate containment and appropriate containment integrity will be maintained maintained for the specified specified duration.

duration.

Response: The only Air Operated Operated Valves (AOVs) requiredrequired during the SBO coping period are the Turbine Driven Auxiliary FeedwaterFeedwater Pump Pump (TDAFWP) Level Control Valves (LCVs). (LCVs). These valves are normally supplied supplied by the Auxiliary Control Air System (ACAS).

This system will not be available available during an SBO event.

Modifications to install Modifications install bottled valves bottled nitrogen to supply these valves were made to the plant. The bottled nitrogennitrogen has sufficient sufficient capacity to supply the TDAFWP TDAFWP LCVs during the SBO coping coping period. The nitrogen period. nitrogen bottles are normally installed.

installed. They areare sized for five (5) LCV cycles. The Appendix Appendix R event is moremore limiting than the SBO, so the nitrogen bottle sizing is based on Appendix Appendix R.

d. Discuss and provide information on the adequacy adequacy of the battery capacity to support loads required for decay heat removal for the specifiedspecified SBO duration and emergency diesel generator generator field flashing for recovering recovering onsite power sources.

Response: The SBO design design basis for Watts Bar is one unit in an SBO condition and the other unit with one operable operable diesel generator.

All battery sizing calculations calculations consider consider dual unit operation.

125 VDC Vital Power System

1) 125 As stated in Unit 2 FSAR 8.3.2.1.1, 8.3.2.1.1, the vital 125V DCDC control power system batteries are designed designed to support an SBO event for four (4) hours. The coping duration for Watts Watts Bar cannot be met with all the normalnormal loads. Loads that are not required to mitigate mitigate an SBO will be removed within within 30 minutes into the event to increase the discharge discharge time of the battery. The battery capacity capacity analysis demonstrates demonstrates the the capacity to provide the remaining loads for up to four (4) evaluation includes allowances hours. The evaluation allowances for aging, design temperature derating.

margin and temperature

2) 250 VDC Battery System As stated in Unit 2 FSAR 8.2.1.4, the 250 VDC batteries are designed to support an SBO event for four (4) hours. The The duration for Watts Bar cannot be met with all the coping duration the normal loads. Loads that are not required to mitigate normal mitigate an SBO will be removed between one (1) and three (3) hours hours into the event to increase the discharge discharge time of the battery.

The battery battery capacity analysis analysis and demonstrates demonstrates the capacity to provide the remaining remaining loads for up to four (4) hours. The The El-92 E1-92

ENCLOSUREI 1 ENCLOSURE Response to Preliminary RAIS and RAIs Response Regarding Unit 22 FSAR RAls Regarding Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit Unit 2, Docket No. 50-391 evaluation includes evaluation allowances for aging, design margin includes allowances margin and temperature temperature derating.

derating.

3) 125 VDC Emergency Emergency Diesel (EDG) Power Diesel Generator (EDG) Power System As stated in Unit 2 FSAR FSAR 8.3.1.1, 8.3.1.1, the EDG batteries are designed to support an SBO event for four (4) hours.

designed hours. The The batteries are required required to provide power power and indication indication to to allow starting of the EDG to recover allow recover from the event. The The coping duration can be achieved coping achieved with the battery and all loads connected, provided maximum of three provided that a maximum start three (3) start sequences sequences are are attempted. The batteries will have sufficient capacity remaining to "flash the generator generator field" with the third and final start occurring at the end of the coping period. The The battery capacity analysis analysis demonstrates the capacity capacity to remaining loads for up to four (4) hours. The provide the remaining The evaluation includes allowances allowances for aging, design margin and temperature temperature derating.

e. electrical cabinets and provide information Discuss the integrity of electrical information on the the effects of the loss of ventilation to other equipment, such as the turbine driven turbine driven emergency feed water pump, valves, the battery room and other equipment emergency equipment mitigating an SBO event. Discuss and credited for mitigating and provide information on provide the information the effects of loss of ventilation ventilation in all dominant concern and on the dominant areas of concern the equipment credited equipment during an SBO event.

credited during Response: Detailed Detailed room temperature temperature evaluations evaluations consistent with RG 1.55 and NUMARC NUMARC 87-00, Revision 1, guidelines have been performed performed in areas containing equipment required to cope with an SBO event. The following areas were evaluated:

1) 250 V Battery and Board Rooms Board Rooms

2) Control Room Complex

3) Cable Spreading Spreading Room

4) 125 V Battery and Board RoomsRooms

5) Rooms 480 V Board Rooms

6) Pipe Chase Pipe Chase Area Area

7) North and South South Main Rooms Main Steam Valve Rooms

8) Feedwater Pump Turbine Driven Auxiliary Feedwater Pump room

9) 6.9 kV & 480 V Shutdown Board Room A A Sequoyah and Watts Bar Unit 1, general, and consistent with Sequoyah In general, electrical heat loads in these areas are assumed electrical heat assumed to be reduced by 50% of their normal values. An assumption of 50% reduction is considered conservative as all AC power has been considered to be very conservative E1-93 E1-93

ENCLOSURE ENCLOSURE 1 Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 lost with only battery power remaining. In In the rooms where where actual heat calculated, the results have been heat loads have been calculated, less than 50% of normal.

normal. If If the corresponding corresponding evaluation predicted predicted excessive excessive temperatures, detailed evaluations temperatures, more detailed evaluations of actual electrical electrical heat loads were performed. Most HVAC HVAC systems are AC powered and are therefore lost during an SBO SSO event. The Turbine Driven Auxiliary FeedwaterFeedwater Pump room is serviced by a DC powered powered ventilation system. Resulting Resulting room temperatures are within equipment temperatures qualification limits and equipment qualification support human habitability habitability requirements requirements(if (if any).

8.4 - 3. Provide information on site specific Provide information procedures and training on the following:

specific procedures

a. Coping with an SBO for the specified duration; Response: The Unit 1 procedure procedure for loss of shutdown shutdown board power is ECA-0.0 (Loss of Shutdown Power, currently ECA-O.O This currently Rev. 20). This procedure procedure provides actions for responding to a loss of shutdown power. It also directs the restoration of shutdown shutdown board power based on the cause cause for the loss of shutdown board board power.

Step 6.a, Response Response Not Obtained (RNO), directs the operator to utilize AOI-35 (Loss of Offsite Offsite Power) or AOI-40 AOI-40 (Station Blackout), or AOI-43 AOI-43 (Loss of Shutdown Boards). If the loss of shutdown shutdown power is due to a Station Blackout, the operator will refer to AOI-040.

ECA-0.0 (Loss of Shutdown ECA-O.O Shutdown Power) is being being drafted for Unit 2 and will be issued to support Unit Unit 2 Startup.

The PURPOSE section of the Unit 1 and common Procedure Procedure AOI-40 AOI-40 (Station Blackout, currently Rev. 013, states the the following:

"This Instruction Instruction provides guidance for restoration provides guidance restoration of shutdown AC power via the DIGsD/Gs or backfeed backfeed from the 500Kv 500Kv system.

Provides Provides operator actions actions to reduce load on the 125V vital, and and 250V station batteries for complete complete loss of all AC power to to extend the useful life of the DC backup power system(s)."

AOI-40 AOI-40 also provides coping strategies for station 125 Vdc and 250 Vdc batteries.

Section 3.2 of this instruction instruction includes includes restoration steps to return components and station service plant components service to normal configuration.

AOI-40 has been drafted for the Unit 2 procedure; procedure; it will be be issued issued will be issued to support support Unit 2 Startup.

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ENCLOSURE ENCLOSURE11 Preliminary RAIS and RAIs Response to Preliminary RAls Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket No. 50-391

b. Restoration of ac power following an SBO event of specified duration; Restoration duration; and

Response

Response: The PURPOSE section of the Unit I1 and common common Procedure Procedure AOI-40 (Station Blackout, currently Rev. 013, states the the following:

"This Instruction Instruction provides guidance for restoration provides guidance restoration of shutdown AC power via the DIGsD/Gs or backfeed backfeed from the 500Kv 500Kv system.

Provides operator actions to reduce load on the 125V vital, and Provides operator 250V station batteries batteries for complete complete loss of all AC power to extend the useful life of the DC backup power system(s)."

Section 3.2 of this instruction instruction includes restoration steps includes restoration steps to return components and station service plant components service to normal configuration.

AOI-40 has been drafted for the Unit 22 procedure; itit will be be issued to support Unit 2 Startup.

c. Preparation Preparation for severe weather weather conditions conditions to reduce the likelihood and consequences consequences of loss of offsite offsite power and to reduce the overall risk of an SBO SBO event.

Response

Response: Procedure AOI-8 (Tornado Watch or Warning, currently 051) provides Rev. 051) Operations' response to this event or provides Operations' or conditions at the site.

The PURPOSE PURPOSE of AOI-8 states the following:

"This Instruction provides actions to be taken in the event aa

'This Tornado Tornado Watch or a Tornado Warning is issued."

AOI-8 also provides guidance damage to the facility guidance to prevent damage from the potential effects effects of tornadoes at or near near the site. The The procedure includes actions such as anchoring anchoring cranes, verifying /I establishing required damper /I door /I hatch positions, stabilization of fuel movements, and securing loose items items outside outside of buildings.

AOI-8 AOI-8 has been drafted for the Unit 2 procedure; procedure; itit will be issued to support Unit 2 Startup.

El-95 E1-95

ENCLOSUREI ENCLOSURE 1 Preliminary RAIS and RAls Regarding Response to Preliminary Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Nuclear Plant - Unit Unit 2, Docket No. 50-391 Section 9.5.3 - Lighting System l.a.

9.5.3 - 1.a. Provide Provide a summary discussion of the typical luminance ranges for normal typical luminance normal and and emergency lighting in all areas/rooms of the plant to ensure that the functional emergency capability of the lighting system design provides illumination illumination level in accordance accordance with the IESNA [Illuminating Engineering Society of North America] Lighting America] Lighting Handbook Handbook for Central Stations or NUREG 700. Discuss the technical technical basis ififthe the design illumination levels are not in conformance conformance with the guidelines guidelines of IESNA Lighting Lighting Handbook for Central Stations Stations and NUREG NUREG 700.

Response: Luminance for normal Luminance normal and emergency emergency lighting in all areas/rooms areas/rooms of the plant is delineated delineated in TVA design design Standard for Lighting Lighting Standards and Practices.

Practices. This document document is based on the following following references:

Engineering Society of North America Illuminating Engineering America (IESNA)

Handbook-1993 Lighting Handbook-1993 NUREG 0700, "Guidelines for Control Design USNRC NUREG Reviews," Appendix Appendix E-2 E-2 The design illumination levels are in conformance conformance with the above references. illumination levels in footcandles for various references. Average illumination various plant areas/rooms areas/rooms are based on the above documents documents and are given in this standard standard as follows:

IIluminance Illuminance Area/Room Area/Room (fc)

Normal Lighting Normal Lighting Aisles, Corridors and Stairways 10 - 2020 Auxiliaries, pumps, tanks, compressors compressors 20 Battery and battery battery board room 20 Communication s room Communication 40 40 Conference room Conference 30-50 Equipment mechanical and miscellaneous Equipment rooms, mechanical miscellaneous75-100 E1-96

ENCLOSUREI 1 ENCLOSURE Response to Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit Unit 2, Docket No. 50-391 9.5.3 - 1.b.

9.5.3 Section 9.5.3 of the WBN Unit 2 FSAR does not describe describe the illumination illumination levels for the work areas areas or type of tasks in the Main Control Room (MCR), (MCR), safety-related safety-related panels in the MCR and and remote shutdown consoles. Provide aa description description of the the normal lighting in these areas. Discuss the technical illumination levels for normal illumination technical basis ifif the design design illumination levels levels do not conform conform to the guidelines of IESNA LightingLighting Handbook for Central Handbook Central Stations Stations or NUREG NUREG 700700..

. Response:

Response: The Watts Bar Units 1 and 2 MCR and Auxiliary Control Room (ACR) lighting system was modified 1989 to 1991 modified in the 1989 frame 1991 time frame requirements of the following documents:

to comply with the requirements Illuminating Society of North America (IESNA)

Illuminating Engineering Society (IESNA)

Lighting Handbook-1981 application volume and 1981 Handbook-1981 application reference 1981 reference volume.

USNRC USNRC NUREG 0700, "Human Factors Engineering," Sections Sections 6.1.5.3 and 6.1.5.4 and Appendix Appendix E-2.

TVA Design Standard DS-E17.1.1, Rev. 2, "Lighting Design Standard DS-E17.1.1, Standards Standards and Practices."

Acceptance Acceptance Criteria (minimum average illumination illumination levels in in footcandles) for the MCR and the ACR were based based on the above above documents.

documents.

MCR AND ACR MCRANDACR Illuminance (fc)

Illuminance Normal Lighting Minimum Maximum Maximum Vertical Vertical Face of Switchboard Switchboard (66 inches 20 50 above the floor)

Benchboard (Horizontal Benchboard (Horizontal Level) 20 50 Rear of the switchboard switchboard (vertical 10 --

60 inches above floor)

Unit Operators' Desk 50 100 100 After the modifications modifications were implemented, implemented, a survey of both the MCR MCR and the ACR was conducted conducted to determine the actual illumination illumination levels and to ascertain ascertain that the acceptance acceptance criteria was met. The The results of this survey are documented documented in Watts Bar analysis for Main Main and Auxiliary Auxiliary Control Rooms; the analysis concludes that the analysis concludes the illumination illumination levels meet the acceptance acceptance criteria.

El-97 E1-97

ENCLOSUREI 1 ENCLOSURE

Response

Response to Preliminary RAIS and RAIs Regarding Unit 2 FSAR RAls Regarding FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 9.5.3 - 1.c.

1.c. Discuss ififthe emergency safety-related panels in the MCR and emergency lighting in the MCR, safety-related remote shutdown shutdown consoles provides illumination illumination levels in in these areas equal equal to recommended by the IESNA greater than those recommended IESNA Lighting Handbook for Central NUREG 700 for at least 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Stations or NUREG Response: Emergency and Standby lighting in the MCR and remote shutdown Emergency consoles meets the requirements requirements of IESNA Lighting Handbook Handbook for Central Central Stations or NUREG NUREG 700 and is designed designed to provide the the following illumination levels:

Standby Lighting: All Tasks, Each Train - 10 10 fc Emergency Lighting:

Emergency 3 fc Acceptance Acceptance Criteria for the task area luminance ratio:

Luminance Luminance Area Ratio Ratio Task area versus adjacent darker 3:1 surroundings surroundings Task area versus adjacent adjacent lighter lighter 1:3 1 :3 surroundings surroundings E1-98

ENCLOSURE11 ENCLOSURE Response to Preliminary Response Preliminary RAIS and RAls Regarding Unit 2 FSAR RAIs Regarding Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Docket No. 50-391 Nuclear Plant - Unit 2, Docket RAIs RAls for FSAR SECTIONS 3.9.1, 3.9.2, 3.9.3, and 5.5.1 [taken from NRC letter dated 3.9.1, 3.9.2, 07/02/2010 (ADAMS 07/02/2010 Accession No. MLI (ADAMS Accession 01530474)]

ML101530474)]

EMCB 3.9-1 The NRC staff noted a number number of instances in the review of Sections 3.9.1, 3.9.1, 3.9.2, 3.9.3 and their corresponding tables and figures of Amendment Amendment No. 97 to the WBN Unit 2, Final Safety Analysis Report (FSAR) (FSAR)

(Reference 1) where editorial modifications (Reference necessitated in modifications may be necessitated in subsequent revisions to the WBN Unit 2 FSAR. Please review the following subsequent following NRC staff notations and rectify, as necessary.

1) On page 3.9-18 of Reference continuing to page 3.9-19, the first Reference 1, continuing two paragraphs Section 3.9.2.5.6, "Results and Acceptance paragraphs of Section Acceptance paragraphs of the following Criteria," are duplicates of the first two paragraphs following section (3.9.2.5.7), also titled "Results and Acceptance Acceptance Criteria."

Response

Response: Amendment 98 to the Unit 2 FSAR corrected corrected the noted discrepancy discrepancy by deleting information.

deleting the repeated information.

Since this was an editorial change, the amendment Since amendment level remains the same.

2) On page 3.9-36 3.9-36 of Reference Reference 1, superfluous superfluous spaces exist between between the the word "Table" "3.9-17."

"Table" and "3.9-17."

Response: The page of concern was page page 3.9-30 in the issued version of Amendment 97 to the Unit 2 FSAR.

Amendment corrected the noted discrepancy Amendment 98 corrected discrepancy by deleting superfluous spaces. Since this was an deleting editorial change, the amendment level remains the the same.

. 3) On page 3.9-44 of Reference Reference 1, the primary membrane membrane plus primary bending stress limit should be "1.1 S" versus the current "1.1.S."

"1.1 S" Response: The page of concern was page 3.9-37 in the issued Amendment 97 to the Unit 2 FSAR. The version of Amendment The second line on the version of page 3.9-37 in the issued Amendment 98 still contains "1.1.1.S."

version of Amendment Amendment 1100 00 to the Unit 2 FSAR will replace "1.1.1 .S" with "1.1.1.S" "1.1 S."

with "1.1 S."

4) On page 3.9-63 of Reference Reference 1, the title of Table 3.9-5 3.9-5 should should be be revised to state that the limits are "Maximum "Maximum Deflections" versus the the current wording of "Maximum Defections."

Response

Response: Amendment 98 to the Unit 2 FSAR corrected the title of FSAR corrected Table 3.9-5 to read "Maximum Deflections" instead of "Maximum Defections." Since this was an editorial "Maximum change, the amendment amendment level remains the same.

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ENCLOSURE ENCLOSURE 1 Preliminary RAIS and RAts Response to Preliminary RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar NuclearNuclear Plant - Unit 2, Docket No. 50-391

5) On page 3.9-63 3.9-63 of Reference Reference 1, Note 1 references references Westinghouse Westinghouse Commercial Commercial Atomic Power (WCAP)-5890 with a corresponding corresponding superscript of number 21, superscript indicating that this refers to Reference 21, indicating Reference 21. 21.

Page 3.9-58 of Reference Reference 1 indicates indicates that this WCAP WCAP report is Reference 22, not Reference Reference 21. 21. IfIf this is not erroneous, please provide additional additional justification in conjunction conjunction with RAI 3.9.2-3 below.

Response

Response: comment was based on a review of the The NRC's comment the red-lined red-lined version for Amendment Amendment 97 to the Unit 2 FSAR. The noted discrepancy in the noted apparent discrepancy the reference number reference number is because Reference (13) is marked Reference (13) for deletion deletion on page 3.9-57.

The issued issued version of Amendment 97 to the Unit 2 FSAR FSAR correctly correctly shows Reference Reference (21)(21) on page 3.9-50 as being Westinghouse Westinghouse Commercial Atomic Power Power (WCAP)-5890 which agrees with Note 1 [refers to Reference (21)] which is on page 3.9-55.

Reference

6) Reference 1, the third note On page 3.9-77 of Reference corresponding to note corresponding Table 3.9-16 should be revised to correct correct the misspelling misspelling of "Non-pressure" and "other justifiable" versus the current wording of "Non-pressur" and "othe justifiable."

"Non-pressur" justifiable."

Response: The page of concern concern was page 3.9-69 in the issued version of Amendment Amendment 97 to the Unit 2 FSAR.

Amendment 98 (page 3.9-69) replaced "Non-pressur" "Non-pressur" with "Non-pressure."

The third line of NoteNote 3 still contains "othe" "othe" instead of "other."

"other." Amendment Amendment 1 100 00 to to the Unit 2 FSAR will the Unit replace "othe" "othe" with "other."

"other."

EMCB 3.9.1-1 3.9.1-1 In Supplemental Supplemental Safety Evaluation Report Report (SSER) 6 (Reference (Reference 3), thethe NRC staff noted noted that the licensee's licensee's piping piping evaluation evaluation for a postulated postulated main feedwater header rupture transient, which results in a water feedwater header water hammer eventevent due due to a rapid check valve closure, included an assumption assumption that certain certain feedwater piping system supports failed when the loads exceeded feedwater exceeded their calculated calculated capacities; this was listed as an open item in SSER 6 (tracked as as Outstanding 20(a)). In SSER Outstanding Issue 20(a>>. SSER 13 (Reference (Reference 6), the staff noted that the analyses performed, which postulated analyses performed, postulated pipe support failures, was was acceptable acceptable based on the difficulty difficulty involved with making subsequent subsequent pipepipe support modifications and the low probabilistic support modifications probabilistic nature involved the involved with the water hammer transient. Additionally, as part of the closure of this open item, SSER 13 also included included a copy of a report performed performed by Brookhaven Brookhaven National Laboratories National Laboratories (BNL) regarding this issue. BNL was contracted contracted by by the NRC to evaluate the licensee's piping analyses performed performed to demonstrate demonstrate compliance compliance with the criteria of Appendix Appendix F of the American American Society of Mechanical Mechanical Engineers Engineers Boiler & & Pressure Pressure Vessel Code. BNL concluded that the licensee's licensee's piping analyses performed for the feedwater feedwater El-100 E1-100

ENCLOSURE1 ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 containment were sufficient loops inside containment sufficient and demonstrated demonstrated that the piping piping system would maintain its structural integrity integrity when subjected to the dynamic dynamic loading associated associated with the water hammer hammer event.

describe the applicability Please describe applicability of the conclusions conclusions made by the NRC staff staff (BNL) regarding the piping analyses described and the contractor (BNL) described above above as as they relate to the current WBN Unit 2 refurbishment refurbishment efforts. Please indicate indicate whether whether the same issues exist with the inability to modify certain piping piping supports containment and whether the piping analyses for the WBN supports within containment Unit 2 feedwater feedwater loops are the same as those analyses performed in in .

support of WBN Unit 1. If If these analyses are dissimilar, please summarize summarize and provide justification for any portions of the analyses analyses that are not exactly the same and whether whether the results of these dissimilar analyses demonstrate demonstrate that the feedwater piping loops meet the acceptance acceptance criteria of the code of record for this piping system.

Response

Response: Analysis methodology, piping geometry and pipe support locations including pipe whip restraint locations locations are similar to to Unit 1. Also, Unit 1 pipe support designs stiffnesses designs and stiffnesses were used as input into the Unit Unit 2 pipe support designs.

Each Each Unit 2 Main Feedwater Loop has a separate pipe stress stress calculation. As-designed As-designed Whip Restraint cold gaps and Generator nozzle Steam Generator nozzle displacements displacements were used the used in the analysis as design input. Six snubbers (from three of the four four loops) loops) that exist in in the Unit 1 analysis are are being deleted in in Unit 2; however, these snubbers snubbers were assumed to fail in the the Unit 1 Check Valve Slam analysis. The Unit 2 analysis has has accounted accounted for the removal of these snubbers.

EDCR 52430 (Modification of pipe supports on Main 52430 (Modification Main Feedwater System -003) has been issued to perform modifications on pipe supports as required to meet the the acceptance acceptance criteria of the code of record record and attain similarity with Unit 1 pipe support support designs.

El-101 E1-101

ENCLOSURE1 ENCLOSURE 1 Preliminary RAIS and RAIs Response to Preliminary RAls Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Tennessee Nuclear Plant - Unit 2, Docket Docket No. 50-391 EMCB 3.9.2-1 In Section 3.9.2.3 of Reference 1, it is indicated In Section indicated that Sequoyah Nuclear Nuclear Plant Unit 1 and Trojan Nuclear Power Plant (Trojan) ""...have ... have been prototype data applicable to Watts Bar" for the instrumented to provide prototype instrumented the evaluating the flow induced purposes of evaluating induced oscillatory effects on the oscillatory pressure effects the reactor vessel internals. Additionally, it is concluded, based based on scale model test results and ""...preliminary Trojan...,"

... preliminary results from Trojan ... ," that plants with neutron shielding pads exhibit less core barrel vibration than plants with thermal shields. Based on the fact that Trojan ceased ceased operations in the the year 1992, please discuss the applicability applicability of the statements above, which are currently included in Reference Reference 1. If If these data was captured captured during during Trojan's operational operational state, please describe how this operating operating experience experience has been applied to the design or operational characteristics of any of the operational characteristics the reactor vessel vessel internals. Additionally, please indicate whether whether additional results, other than the "preliminary results" mentioned Reference 1, were mentioned in Reference utilized to provide additional information comparison between information regarding the comparison neutron shielding pads and plants with thermal shields as they plants with neutron relate to core barrel excitation.

relate. excitation.

Response

Response: The testing on Trojan was done initially and was completed without the need to follow plant operation operation through the design life of the plant. Since no operational operational data was collected, none was applied to the Watts Bar internalsinternals design.

operations in the year 1992 Therefore, Trojan ceasing operations 1992 has no no affect on the results obtained from Trojan, and there are no no additional test results required from Trojan's operational state additional through 1992. Also, there are no additional results, other than the "preliminary results" mentioned Reference 1, that mentioned in Reference provide additional information regarding the were utilized to provide the comparison between plants with neutron shielding pads and comparison plants with thermal shields as they relate to core core barrel excitation.

excitation.

3.9.2-2 EMCB 3.9.2-2 The analyses methods described in Section 3.9.2.5 of Reference 1, methods described "Dynamic System Analysis of the Reactor Internals Under Faulted Conditions," were approved for use by a previous Conditions," previous license amendment license amendment request submitted for WBN Unit 1. These methods incorporate incorporate the use of the MULTIFLEX, LATFORCE, LATFORCE, FORCE-2 and WECAN WECAN computer codes to model the complex, non-linear thermal-hydraulic thermal-hydraulic loadings induced induced on the the reactor vessel vessel internals under under upset upset loading conditions.

a. Please confirm that the inputs used to analyze these conditions for WBN Unit 2 are the same inputs as those used to analyze analyze the the loadings induced induced on the WBN Unit 1 reactor vessel internals.

Response

Response: analyze these conditions for The inputs used to analyze for Watts Bar Unit 2 are the same inputs as those used to to analyze the loadings induced on the Watts Bar Unit 1 reactor vessel internals.

E1-102 El-102

ENCLOSUREI 1 ENCLOSURE

Response

Response to Preliminary RAIS and RAls RAts Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Authority *- Watts Bar Nuclear Nuclear Plant*

Plant - Unit 2, Docket Docket No. 50-391 50*391

b. If If any variances variances exist between between the WBN Unit 1 and WBN Unit 2 inputs inputs for these codes, including primary primary and secondary secondary loadings, flow flow parameters, parameters, mass models, finite element formulations, or other input input parameters, provide justification justification for the variation and its effects effects on the the ability of the WBN Unit 2 reactor vessel internals to meet the the acceptance acceptance criteria provided provided in Table 3.9-5.

Response: No variations exist between between the Watts Bar Unit 1 and Watts Bar Unit 2 inputs for these including these codes, including secondary loadings, flow parameters, primary and secondary mass models, finite elementelement formulations, or other other input parameters.

c. Additionally, please clarify clarify whether the references references to "Watts Bar Bar 11" Unit 1" on pages 3.9-15, 3.9-19, and 3.9-20 (2) are correctly correctly referring referring to WBN Unit 1 for purposes purposes of comparing comparing analyses or whether whether these these instances are incorrect incorrect (i.e.,

(i.e., these references should state WBN Unit 2 and not WBN Unit 1).

Response

Response: The references references to Watts Bar Unit 1 are for purposes purposes of Watts Bar Unit 1 analyses which are also applicable applicable to to Watts Bar Unit 2.

EMCB 3.9.2-3 3.9.2-3 Table 3.9-5 of Reference Reference 1, "Maximum Def[l]lections Under Design Basis "Maximum Def[I]lections Basis Event (in)," provides provides the maximum maximum allowable and no loss-of-function limits loss-of-function limits for the reactor vessel vessel internals under designdesign basis loading conditions.

conditions.

Note 1 to Table 3.9-5 indicates that WCAP-5890 WCAP-5890 provides limiting limiting criteria for internals deflection based on stress levels induced in the internals internals internals structures.

a. Please Please discuss discuss whether the acceptance acceptance criteria provided in Table 3.9-5 are based on WCAP-5890. If If these criteria criteria are based based on on this WCAP report, please please provide the bases for the regulatory acceptance of this report.

acceptance

Response

Response: Amendment 100 Amendment 100 to the Unit 2 FSAR FSAR will replace the the values in Unit 2 FSAR Table 3.9-5 with the values from Unit 1 FSAR Table 3.9-5, as required.

In replacing this table, itit was observed In observed that the the no-loss-of-function limit for the upper no-Ioss-of-function upper package package axial deflection deflection is 1.5 inches. This value should be changed to 0.15.

As noted in Note 1. 1. of Table 3.9-5, "The "The allowable allowable limit deflection deflection ... correspond correspond to stress levels for internals internals structures structures well belowbelow the limiting limiting criteria given by the the curves in WCAP-5890 ...

collapse curves .. ".

E1-103 El-103

ENCLOSURE ENCLOSURE 1 I Response to Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Plant - Unit Unit 2, Docket No. 50-391 This WCAP was developed developed to document the basis for ultimate strength criteria to ensure ensure no loss of function.

Westinghouse has no record that WCAP-5890 has has been formally submitted submitted to the NRC for review.

However, WCAP-5890 WCAP-5890 is referenced referenced in the Unit 1 FSAR FSAR which was reviewed and approved by the NRC.

The regulatory acceptance acceptance for Table 3.9-5 from Unit 1 is applicable to Unit 2.

b. If If these criteria are based on a methodology other than the WCAP WCAP report, please provide additional information regarding the additional information the development of these deflection development deflection limits and the bases for the regulatory acceptance of this alternate acceptance alternate methodology.

methodology.

Response

Response: As noted in the response response to RAI EMCB 3.9.2-3 3.9.2 a.,

the values in Unit 2 FSARFSAR Table 3.9-5 will be replaced replaced with the values from Unit 11 FSAR Table 3.9-5, as as required. As described required. response to RAI described in the response EMCB EMCB 3.9.2-3 3.9.2 a., the methodology methodology is based on the the WCAP as a strength based based limit and incorporates incorporates functional considerations considerations which may be more more limiting.

This approach approach is the same as used and accepted for Unit 1. Since Amendment 100 to Unit 2 FSAR FSAR Table 3.9-5 will make its values the same as Unit 1I FSAR Table additional development Table 3.9-5, there is no additional development information to be provided regarding Unit 2.

EMCB 3.9.2-4 3.9.2-4 a. Please Please provide justification for the variance variance between between the WBN Unit 1 and WBN Unit 2 allowable and no loss-of-function deflection limits as as this variance variance relates relates to the upper barrel expansion expansion and compression compression limits and the no loss-of-function loss-of-function limit for the upper package axial deflection.

deflection.

This justification should include information regarding include information regarding whether whether there are variations variations in the analyses methodologies methodologies for determining the WBN Units 1 and 2 reactor vessel internals internals faulted loads (as requested in EMCB 3.9.2-2).

Response

Response: As noted noted in the response to RAI EMCB 3.9.2-3 3.9.2 a., the the values in Unit 2 FSAR Table 3.9-5 will be replaced with the the values from Unit 1 FSAR FSAR Table 3.9-5, as required.

b. Additionally, this justification justification should indicate whether whether there are are variations in the acceptance acceptance criteria for the WBN Units 1 and 2 deflection deflection limits.

Response: As noted in the response to RAI EMCB 3.9.2-3 3.9.2 a., the the values in Unit 2 FSAR Table 3.9-5 will be replaced with the the values from Unit 1 FSAR Table Table 3.9-5, as required, and and .

therefore, there is no change to the acceptance acceptance criteria.

E1-104 El-104

ENCLOSURE11 ENCLOSURE

Response

Response to Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 EMCB 3.9.3-1 3.9.3-1 In SSER 4 (Reference In (Reference 2), the NRC staff noted that aa sampling sampling program was was initiated by TVA to determine determine whether the compressive compressive stresses imposed imposed on short column pipe supports exceeded exceeded the buckling criteria criteria margin established established by by the NRC. The NRC staff accepted accepted the sampling program and determined that TVA had adequately addressed addressed the NRC design criteria for Class 2 and 3 pipe pipe supports; supports; this resolved Outstanding Outstanding Issue 2. 2.

a. Please confirm the applicability of the sampling programprogram discussed discussed in in Reference Reference 3 as itit relates to Class 2 and 3 pipe supports at WBN Unit 2.

(Note: Reference 3 in the transmittal of EMCB RAI (Note: Reference RAI 3.9.3-1 is SSER SSER 6 which does not discuss a sampling program.

program. ItIt was discussed with Licensing Licensing and concluded concluded that the sampling program program referred to is thethe sampling program discussed in SSER 4).

Response

Response: SSER 4 references references TVA's letter to the NRC dated May 14, 1984, and a review of the sampling program program described in that letter letter as the basis for concluding concluding that Class 2 and 3 pipe supports at Watts Bar comply with the the applicable NRC design criteria. The May May 14, 1984, 14,1984, response was not defined defined as being unit specific specific and was was written in response response to a request from the NRC (letter dated April 17, 1984) 1984) to provide information information for both Watts Bar Bar Unit 1 and Unit Unit 2. A list of the pipe supports included in the the sample program described described in TVA's May 1984, letter May 14, 1984, letter was attached attached to a TVA letter to the NRC NRC dated November November 10, 1982. This listing of supports 10, 1982. supports included several Unit 2 supports.

Watts Bar Design Criteria WB-DC-40-31.9, Criteria Criteria for Design Design of Piping Supports Piping Supports and and Supplemental Steel in Category/I Structures, Category Structures, Section Section 3.8 (Appendix "B" Table B-2) describes the allowable stress limits used describes used in the design of pipe supports. This section section of the design criteria criteria incorporates the requirements of the TVA letter to the NRC incorporates NRC dated May 14, 1984, as indicated by source note 14 on 14, 1984, page 50 of WB-DC-40-31.9, WB-DC-40-31.9, R21. R21. Design Criteria Criteria WB-DC-40-31.9 WB-DC-40-31.9 is used used for the design of pipe supports for both Unit 1 and Unit 2 at Watts Bar.

Therefore, the sampling program Therefore, program and the information information provided in TVA's letter letter to the NRC 14, 1984, NRC dated May 14,1984, which supports supports the conclusion provided in SSER 4 is used to support the design and qualification qualification of Class 2 and 3 pipe supports in in both Watts Bar Unit 1 and Unit 2.

El-105 E1-105

ENCLOSURE 1I ENCLOSURE Response to Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391

b. sampling program was not used If this sampling used in support of the WBN Unit 2 refurbishment effort, please refurbishment please discuss the current current criteria criteria used for for demonstrating demonstrating that these these pipe supports maintain sufficient sufficient margin margin against critical buckling buckling of short short column pipe supports.

Response

Response: The response to RAI EMCB 3.9.3-1 3.9.3-1a.a. notes that the the program was used. Thus, this question is not applicable.

3.9.3-2 EMCB 3.9.3-2 In SSER In SSER 6 (Reference (Reference 3), the NRC staff noted its concerns regarding the the earthquake experience licensee's use of earthquake experience data to seismically seismically qualify Category I(L) I(L) piping and identified this concern as Outstanding Issue 19(h). In Issue 19(h). In SSER 8 (Reference 5), the NRC (Reference NRC staff noted that the licensee had developed a screening screening criteria to identify identify items in Category I(L) I(L) piping systems that may require further further based on this earthquake evaluation based experience data. Additionally, earthquake experience Additionally, the licensee licensee indicated that bounding stress cases would be performed indicated demonstrate the performed to demonstrate the screening criteria. The NRC staff found this screening conservatism of these screening screening criteria adequate demonstrating the seismic ruggedness of Category adequate for demonstrating Category I(L) piping.

piping.

a. Please confirm that this screening has been performed performed for the WBN Unit 2 refurbishment efforts.

refurbishment efforts.

Response: The I(L)

I(L) piping seismic evaluation evaluation program program is currently currently being performed. screening criterion used performed. The screening used for the the (Integrated Interaction Unit 2 completion is WB-DC-20-32 (Integrated Interaction Program Screening and Acceptance Acceptance Criteria) which is the the screening criteria that was used for the Unit 1 and same screening common.

b. If this screening method was not utilized in the seismic qualification qualification of the the WBN Unit 2 Category I(L) please discuss the criteria that has been I(L) piping, please seismically qualify these piping systems and discuss used to seismically used discuss the the regulatory acceptance regulatory acceptance bases for this alternate alternate criteria.

Response

Response: As stated in the response response to portion a. of this request, change in the screening there has been no change screening method.

EMCB 3.9.3-3 3.9.3-3 In addition screening methods used addition to the screening Category I(L) used for Category systems I(L) piping systems described in RAI 3.9.3-2, SSER describes TVA's criteria SSER 8 also describes criteria used for thethe I(L) piping supports. The NRC staff noted in SSER Category I(L) evaluation of Category SSER 8 that TVA had indicated indicated it would utilize a factor evaluation of factor of safety of three in their evaluation concrete expansion expansion anchor bolts for these pipe supports. The NRC staff staff accepted the use of this safety factor value for validating the existing design accepted design of concrete expansion expansion anchors used used in this piping system based on TVA's TVA's implementation of recommendations implementation additional concrete recommendations including additional inspection, concrete inspection, anchor spacing, and concrete distance in conjunction with the existing concrete edge distance existing anchor bolts. The NRC staff also noted in SSER 8 that for future Category I(L) I(L) safety factors for these piping systems found in the former piping, the required safety former Office of Inspection and Enforcement Enforcement (IE) utilized.

(IE) Bulletin 79-02, should be utilized.

El-106 E1-106

ENCLOSUREI 1 ENCLOSURE Response to Preliminary Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar NuclearNuclear Plant - Unit 2, Docket No. 50-391

a. Please discuss whether whether the existing, applicable applicable Category I(L) piping I(L) piping supports supports at WBN Unit 2 have been evaluated evaluated in the manner described described in SSER SSER 8.

Response: The category I(L) piping supports for Unit 22 have been evaluated in the same manner manner as the category category I(L)I(L) piping piping supports for Unit 1. That is, the existing supports were were evaluated with the use of aa factor of safety safety of three, which included additional concrete inspection, anchor spacing, anchor spacing, and concrete concrete edge distance distance attributes. The design of new category I(L) pipe supports is performed performed in accordance accordance with Watts Bar design criteria WB-DC-20-32 WB-DC-20-32 (Integrated (Integrated Interaction Interaction Program Program Screening Screening and Acceptance Acceptance Criteria) which is consistent consistent with the required required safety factors found in Bulletin 79-02.

b. If these supports have been evaluated evaluated in a dissimilar manner, please please provide justification justification for the departure departure from the methods methods described described in Reference 4.

Reference Response: As stated in the response to portion a. of this request, there has been been no change in the manner of evaluation.

evaluation.

5.5.1 a.

EMCB 5.5.1-1 Please discuss whether TVA has committed to perform an augmented inservice inspection of the reactor coolant pump (RCP) flywheel.

inservice flywheel.

Response

Response: TVA to NRC letter dated March 4,2009 dated March 4, 2009 (ADAMS Accession Accession ML090700378) submitted No. ML090700378) submitted Development Development Revision A of Unit 2's Technical Technical Specifications Specifications (TS) and Technical Requirements Manual.

Requirements Manual.

TS 5.7.2.10 (Reactor (Reactor Coolant Pump Flywheel Inspection Inspection Program) states, "This program shall provide for the the inspection inspection of each each reactor reactor coolant pump flywheel per the the recommendations of Regulation recommendations Regulation Position co4.b c.4.b of Regulatory Guide Guide 1.14, Revision 1, August 1975."

TRM Surveillance Surveillance requirement TSR 304.5.1 3.4.5.1 states, "Inspect each reactor reactor coolant pump flywheel according to the the recommendations Regulatory Position Co4.b recommendations of Regulatory C.4.b of Regulatory Guide Guide 1.14, Revision 1, August 1975." Its frequency is "According to the recommendations recommendations of C.4.b of Regulatory Regulatory Position Co4.b Regulatory Guide 1.14, Revision 1."

El-107 E1-107

ENCLOSURE1 ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and and RAIs RAls Regarding Unit 22 FSAR Regarding Unit FSAR Tennessee Tennessee Valley Authority - Watts Valley Authority Watts Bar Nuclear Plant Bar Nuclear Plant - Unit 2, Docket No.

2, Docket No. 50-391 EMCB EMCB 5.5.1-1 5.5.1 b. If commitment has been made, please If no commitment justification that please provide justification that the the potential potential for excessive vibration on the excessive vibration reactor coolant the reactor flywheels will coolant pump flywheels addressed to minimize adequately addressed be adequately the possibility of RCP shaft or minimize the flywheel failure.

flywheel Response: The response to RAI EMCB 5.5.1 a. documents the Response: The response to RAI EMCB 5.5.1 a. documents the response is needed commitment; no response commitment; needed to this this RAI.

References References

1) Letter from M. Generation Company, LLC, to NRC D. Jesse, Exelon Generation M. D. NRC Document Document Control Desk, "Watts Bar Plant (WBN)

Bar Nuclear Plant Safety Analysis Report (FSAR),

(WBN) - Unit 2 - Final Safety (FSAR),

Amendment 97," dated January Amendment 11, 2010. (ADAMS Accession January 11, Accession Nos.: ML 100191421 (letter),

ML100191421 ML1 00191684 (Section 3.8.5-3.11))

ML100191684 3.8.5-3.11))

2) NUREG-0847, Supplement 4, "Safety Evaluation Report NUREG-0847, Supplement Operation of Watts Report Related to the Operation Watts Bar Nuclear Plant, Units 1 and 2," dated Bar Nuclear 31, 1985.

March 31, dated March Accession No.:

1985. (ADAMS Accession No.:

ML072060524)

ML072060524)

Supplement 6, "Safety Evaluation NUREG-0847, Supplement

3) NUREG-0847, Report Related Evaluation Report Operation of Watts Related to the Operation Watts Bar Nuclear Plant, Units 1 and dated April 30, 1991.

and 2," dated Accession No.:

1991. (ADAMS Accession No.:

ML072060464)

ML072060464)

Report Related to the Operation Evaluation Report Supplement 7, "Safety Evaluation

4) NUREG-0847, Supplement Operation of Watts Watts Nuclear Plant, Units 1 and 2," dated Bar Nuclear 1991. (ADAMS Accession No.:

September 30, 1991.

dated September ML072060471))

ML072060471

5) NUREG-0847, Evaluation Report Related Supplement 8, "Safety Evaluation NUREG-0847, Supplement Operation of Watts Related to the Operation Watts January 31, Nuclear Plant, Units 1 and 2," dated January Bar Nuclear 31, 1992. (ADAMS Accession Accession No.:

ML072060478)

ML072060478)

Evaluation Report Related to the Operation of Watts NUREG-0847, Supplement 13, "Safety Evaluation

6) NUREG-0847, Nuclear Plant, Units 1 and 2," dated April 30, Bar Nuclear 1994. (ADAMS Accession No.:

30 ,1994.

ML072060484)

M L072060484)

El-108 E1-108

ENCLOSURE 1I Response to Preliminary RAIS and RAls Response RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 RAIs Portions of the FSAR [from RAls for Various Portions [from NRC NRC letter dated 06/24/2010 (ADAMS (ADAMS Accession No. MMLI101540250)]

L 101540250)]

Systems (SBPB)

Plant Systems SBPB 3.6-01 3.6-01 Section Section 3.6A.2.2.2 "Blowdown Thrust Loads" contains the following equation:

VE = [2gc(Po - PA )/PEI' The equivalent equivalent equation in the WBN Unit 1 I FSAR, Amendment Amendment No.7 No. 7 is:

VE= [2gc(PO -PA)]'/pE (see page 3.6A-1 7)

TVA is requested requested to clarify the differences differences in the two equations.

equations.

Response: The Unit 22 FSAR is correct. Amendment 8 corrected the Unit 1 UFSAR UFSAR such that itit now agrees with the Unit 2 FSAR.

SBPB 5.2.5-01 Previously there existed an intersystem leakage path "upper head injection Previously injection (UHI)." This system is no longer system (UHI)." longer described described in in the FSAR. TVA is requested to confirm this system is no longer included included in the WBN plant and provide the basis for deleting the system.

provide the basis for deleting the system.

Response: Deletion of UHI Deletion UH/

The UHI system was never utilized at Watts Bar. Both the Unit 1 and Unit 2 UHI Systems have been removed. removed. "Watts Bar Nuclear Nuclear Plant (WBN)

(WBN) - Assessment Report Report NA-WB-94-0020" was aa determination of whether determination Injection Deletion whether the Upper Head Injection Deletion (UHID)

(UHID) - Capital Program (CP) was adequatelyadequately implemented.

implemented.

Attachments 1, 2 and 3 of the report list the DCNs, WPs and and WOs that were executed during the removal of the upper head injection system. In In SSER 7, the staff reviewed reviewed the request request for change and found itit acceptable the design change acceptable to delete delete the UHI system from both units.

Licensinq Basis for the Deletion Licensing Basis UHI Deletion of UHI The UHI was eliminated eliminated to increase increase operational flexibility. The The system benefits were overshadowed overshadowed by frequent frequent operational problems such as:

- Rupture or leakage Rupture leakage of membrane membrane in the gas crossover crossover line line separating the water and nitrogen separating nitrogen accumulators.

accumulators.

- Level switch/transmitter switch/transmitter problems problems for volume delivery within within tolerances (specifically accuracy tolerances accuracy of Barton levellevel switches, installation installation error of sensing lines and calibration calibration procedures).

- Violation of chemistry Violation requirements (for water accumulator chemistry requirements accumulator nitrogen nitrogen entrainment entrainment and boron concentrations).

- Violation of system gas pressure requirements.

El-109 E1-109

ENCLOSURE ENCLOSURE 1

Response

Response to Preliminary Preliminary RAIS and RAls Regarding Unit 2 FSAR RAIs Regarding Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 The approach approach for justifying removal of UHI,UHI, however, relied on on computer code technology computer technology to show that the acceptance acceptance criteria criteria specified in 10 specified 10 CFR 50.46 for the LOCA analysis was met.

Westinghouse small-break LOCA analysis using Westinghouse performed the small-break NRC-approved methods. The NOTRUMP NRC-approved NOTRUMP code code (Westinghouse Topical Reports WCAP-1 WCAP-1 0080 and WCAP-1 WCAP-1 0081) 0081 ) was used used for calculation of transient the calculation transient depressurization depressurization of the reactor system, core power, water-steam water-steam mixture height height and steam flow past the the uncovered uncovered portion of the core. The LOCTA LOCTA code (Westinghouse (Westinghouse Topical Report Topical Report WCAP-8305) was used used for the peak cladding temperature temperature analysis. The staff staff concluded concluded that the Watts Bar Watts Bar small-break LOCALOCA analysis results were within the acceptance acceptance criteria specified in 10 CFR 50.46.

performed a large-break Westinghouse also performed large-break LOCA analysis analysis supporting its request for removal for the UHI system. In the the analysis, only the double-ended, double-ended, cold-leg, guillotine (DECLG) guillotine (DECLG) analyzed because they resulted in the highest breaks were analyzed breaks highest peak cladding temperatures. The analysisanalysis was performed performed using aa modified modified revision of the 1981 1981 Westinghouse Westinghouse ECCS evaluation evaluation model model (WCAP-9220-P-A, Rev. 1). evaluation model 1). This evaluation model used the revised PAD fuel thermal safety safety model (WCAP-8720) for calculating calculating the initial fuel conditions; conditions; the SATAN-VI SATAN-VI code code (WCAP-8302) for the thermal (WCAP-8302) thermal hydraulic calculation during the hydraulic calculation the blowdown blowdown period; period; the transient transient WREFLOOD WREFLOOD (WCAP-8170)

(WCAP-8170) and BASH (WCAP-1 0266 and addendum)

(WCAP-10266 addendum) codes for calculating the the refill and reflood transient periods; the LOCBART LOCBART code code (WCAP-8305)

(WCAP-8305) for calculating calculating the peak cladding temperature; temperature; and

-. the LOTIC "the LOTIC code (WCAP-8355) for calculating the ice condenser containment containment pressure transient. The staff found that the results results showed showed that peak peak cladding metal-water reaction and cladding temperature, metal-water cladding cladding oxidation were within the acceptance acceptance criteria specified in criteria specified 10 CFR 50.46 for LOCA analysis.

WBT-D-1460, WBT-D-1460, "Final Small Small Break LOCA Summary Report,"

January 22, January 2010 and WCAP-17093-P, 22,2010 WCAP-17093-P, Revision 0, "Best Estimate Analysis Analysis of the Large-Break Large-Break Loss-of-Coolant Loss-of-Coolant Accident Accident for Watts Bar UnitUnit 2 Nuclear Power Plant Nuclear Power Plant using the ASTRUM Methodology," December Methodology," December 2009, are the current analyses analyses generated for Unit 2. These analyses analyses both show considerable considerable margin to 10 margin 10 CFR 50.46 peak clad temperature temperature limits without the the UHI system. .

E1-110 El-110

ENCLOSURE1 ENCLOSURE 1 Response Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 SBPB 9.2.6-1 9.2.6-1 In the Safety Evaluation Report (SER) related related to the operation of WBN Units 1 and 2, NUREG-0847, Supplement 12, dated NUREG-0847, Supplement dated October 1993, the NRC staff wrote wrote Section 9.2.6, Condensate in Section Condensate Storage Storage Facilities:

In Section 9.2.6 of the SER, the NRC NRC staff indicated that the two condensate condensate storage tanks reserved reserved 200,000 gallons gallons of condensate condensate for each each unit's auxiliary feedwater (AFW) system. In FSAR unit's auxiliary FSAR Amendment No. No. 72, TVA revised this reserved amount to 210,000 210,000 gallons. The basis for the storagestorage capacity capacity is not affected this affected and this correction is made correction made for clarification clarification purposes purposes only. This does not not change any of the NRC NRC staffs conclusions reached in the SER or conclusions reached supplements related supplements related to the condensate condensate storage facilities or the AFW system. The NRC staffs effort was tracked by TAC M85037 and M85037 and M85038.

M85038.

In the proposed FSAR for WBN Unit 2, Section 9.2.6.2 System Description, TVA proposed FSAR proposal states:

proposal The condensate facility, shown shown in Figure 10.4-7, consists of one one condensate transfer pump and two condensate condensate condensate storage tanks connected in parallel (one tank for each unit) and associated piping, connected controls, and instrumentation.

instrumentation. The tanks are located in the plant yard adjacent adjacent to the east wall of the TurbineTurbine Building. The auxiliary feedwater pumps take suction feedwater condensate storage suction directly from the condensate storage supply treated water for cooldown of the reactor tanks to supply reactor coolant coolant minimum of 200,000 gallons system. A minimum system. -qallonsin each tank is reserved for for auxiliary feedwater system. This quantity is assured by means of the auxiliary standpipes through standpipes through which other systems are supplied.supplied.

The NRC requests TVA to justify why the change NRC staff requests change to 210,000 210,000 gallons gallons waswas not incorporated.

incorporated.

Response

Response: Preliminary RAI 9.2.6 provided by the NRC in an e-mail of 04/23/2010 asked a similar question. TVA provided 04/23/2010 provided the following following answer on page E1-17 answer El-17 of a TVA letter dated 06/03/2010 (ADAMS (ADAMS Accession No. ML101600477):

Accession ML101600477):

"Amendment 89 revised the value value from "210,000 gallons" to to "200,000 gallons." At that time, the FSAR FSAR was for both Unit Unit 1 and Unit 2. Thus, the revision applied to both units.

and Unit units.

amendment resulted from FSAR The amendment FSAR change change number number 0889.

The reason for the change was to correct the condensate condensate storage tank minimum reserve volume for auxiliary feedwater feedwater use, basedbased on Calculation Calculation HCG-LCS-043085, Rev. 4."

El-111 E1-111

ENCLOSURE 1 ENCLOSURE Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 SBPB 9.3.1-1 9.3.1-1 TVA provided a document titled, "FSAR Cross Referenced Referenced to SER sorted by by SER, then by FSAR." In this documentdocument under the line item SER Section 9.3.1, 9.3.1, "Compressed Air System," the scope identified identified new essential air compressors compressors were installed.

installed. The compressed compressed air system is a shared shared system between WBN Units 1 and 2. During During a review of the proposed proposed FSAR for WBN Unit 2, the NRC staff did not detect any changes.

NRC staff requests The NRC requests TVA to explain explain whether whether there were any changes needed to be made to the proposed FSAR for WBN Units 1 and 2, based upon the the installation installation of new essential essential air compressors.

compressors.

Response: Evaluations of dual unit essential essential air system (Auxiliary Compressed Air System (ACAS)), demands demands have determined determined that the currently installed ACAS compressors have sufficientsufficient capacity to support dual unit operation.

operation. There are no plans to either replace the existing existing ACAS air compressors or to add additional compressors.

10.3.0 SBPB 10.3.0 TVA provided aa document document titled, "FSAR "FSAR Cross Referenced Referenced to SER sorted by Main Steam Steam SER, then by FSAR." In this documentdocument under SER Section 10.3.0, 10.3.0, "Main Steam System Supply System," TVA identifies that this section includes includes a review of the the following FSAR sections:

• 10.3 10.3 MAIN MAIN STEAM STEAM SUPPLY SUPPLY SYSTEM

• • 10.3.0 10.3.0 Main Steam Supply Supply System 10.3.1 Design System 10.3.1 Bases Design Bases

• • 10.3.0 10.3.0 Main Steam Supply System 10.3.4 10.3.4 Inspection and Testing Requirements Requirements 10.3.0 10.3.0 Main Steam Supply Supply System 10.4.11 Steam Generator System 10.4.11 Generator Wet Layup Layup System During a review of the FSAR, the NRC staff noted that Section 10.4.11, "Steam Section 10.4.11, "Steam Generator Generator Wet Layup System," was not included. included.

The NRC staff requests TVA to justify the omission of the FSAR Section 10.4.11, to include disposition 10.4.11, disposition of safety-related safety-related components that were a part of this system (e.g., containment isolation isolation valves, piping and components).

Response: The Steam Generator Wet Layup System (SGWLS)

Steam Generator (SGWLS) is no longer used used on Watts Bar Unit 1, and it will not be used on Unit 2. The The designed to help protect the steam generator (SG) system was designed internals internals from corrosion during periods of cold shutdown. This shutdown. This protection protection was previously previously provided by the thorough mixing of the the ammonium ammonium hydroxide and hydrazine layup solutions in each SG by utilizing the system. The previous "alternate" method of previous "alternate" corrosion corrosion control was accomplished by injecting the chemicalschemicals into the SGs during during cold shutdown; sampling using existing SG sample sample lines; and performing the mixing of the solutions by bubbling bubbling nitrogen nitrogen (N22)) through the bottom of the SGs. An E1-112 E1-112

ENCLOSURE 1 ENCLOSURE

Response

Response to Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 engineering analysis was performed engineering performed which determined determined that N2 N2 sparging for chemical chemical mixing mixing in the SGs was sufficient for the wet layup process. The alternate method method was adopted as the primary and only means for SG wet layup and a Unit 1 design change design change (DCN 51724) was performed performed to delete the SGWLS SGWLS sincesince it was was no longer required.

That wet layup layup method will also be utilized on Unit Unit 2 and a design change is being performed performed on Unit 2 (DCN 53864 and EDCR 54263). All Unit 2 associated associated piping, components components and valves valves will be removed, modified or abandoned abandoned in place. The The associated UnitUnit 2 containment containment isolation valves will be removed and the piping capped. The discussion of the SGWLS was was previously removed previously removed from the Unit 1 UFSAR UFSAR and has been removed from the Unit 22 FSAR since the system will not be be utilized.

utilized. Neither RG 1.70 nor NUREG-0847 NUREG-0847 requires a discussion discussion of the system. The abandonment abandonment of the SGWLS on Unit 1 was was not associated associated with the replacement of the SGs.

SBPB 10.4.7 10.4.7 document titled, "FSAR Cross Referenced TVA provided a document Referenced to SER sorted by Condensate Condensate SER, then by FSAR." In In this document document under SER Section 10.4.7, 10.4.7, "Condensate and and Feedwater System," TVA identifiesidentifies that this section section includes includes a review of thethe Feedwater Feedwater following FSAR FSAR sections:

Systems FSAR 5.5.9 Main Steam Line and Feedwater Piping FSAR 5.5.9 Main Steam Line and Feedwater Piping

. 10.4.7 FSAR 10.4.7 Condensate Feedwater Systems Condensate and Feedwater Systems

• 10.4.10 FSAR 10.4.10 Heater Drains and Vents Vents During a review of the FSAR, the NRC staff noted that Section Section 10.4.10, "Heater Drains and Vents," shows up in the table of contents, but the text section is not included.

The NRC staff requests requests TVA to justify the omission of the FSAR Section 10.4.10, to include include disposition of any safety-related safety-related components that were a part of this system.

Response: Heater Drains Drains and Vents (HDV)(HDV) are designed to remove all condensate condensate from the feedwater feedwater heaters, heaters, moisture separators, nd 1ýt stage (low pressure) reheaters, 2 121 21d stage (high pressure) condensers and gland steam reheaters, main feed pump turbine condensers condensers.

condensers. These drains drains and vents are not required required for the safe shutdown shutdown of the plant or to mitigate consequences of an mitigate the consequences accident.

During the re-constitution re-constitution of the Unit 2 FSAR, one of the the underlying tenets was to have the Unit Unit 2 FSAR correlate as as closely closely as possible to the Unit 1 UFSAR. Section 10.4.10 was 10.4.10 was removed from the Unit 1 UFSAR by Amendment 1; documented in Change Package 1569 1569 and the Safety Evaluation for E1-113 E1-113

ENCLOSURE 1 ENCLOSURE

Response

Response to Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 EDC E-50038-A.

RG 1.70, Rev. 3, "Standard Format and Content of Safety Analysis for Nuclear Power Plants (LWR Edition)" and and NUREG-0847 NUREG-0847 (series) "Safety Evaluation Reports Reports Related Related to Watts Bar Nuclear Plant" do not require aa discussion discussion about about HDVs.

Amendment 94 to the Unit 2 FSAR removed 10.4.10 10.4.10 from the the Table of Contents.

Component Performance Performance And Testing Testing (CPTB)

CPTB Title 10 of the Code of Federal FederalRegulations, Regulations,Section Section 50.55a (10 CFR 50.55a),

Inservice requires that inservice testing (IST)

(1ST) of certain certain American Society of Mechanical Testing -1.

- 1. Engineers (ASME) Code Class 1, 2, and Engineers and 3 pumps and valves be performed performed in accordance with the ASME accordance ASME Code for Operation Operation and Maintenance Maintenance of Nuclear Nuclear Power Plants (OM Code) and applicable addenda.

Paragraph 10 CFR 50.55a(f)(4 Paragraph )(i) requires:

50.55a(f)(4)(i)

"Inservice tests to verify operational operational readiness of pumps and valves, whose whose function is required for safety, conducted during the initial initial 120-month 120-month interval requirements in the latest must comply with the requirements latest edition and addenda addenda of the Code Code incorporated by reference in paragraph incorporated paragraph (b) of this section section on the date 12 months months before the date of issuance issuance of the operating license license under this part, or 12 12 months before the date scheduled for initial initial loading loading fuel under under aa combined combined license under part 52 of this chapter license chapter (or the optional optional ASME Code cases listed in Regulatory Guide 1.192, that is incorporated NRC Regulatory incorporated by reference reference in paragraph paragraph (b) of this section), subject to the limitations modifications listed in paragraph limitations and modifications (b) of this section."

In Amendment Amendment No. 97 to the Watts Watts Bar Unit 2 FSAR, TVA TVA states that IST1ST of ASME ASME Code Class 1, 2, and 3 pumps and valves will be conducted conducted to the extent extent practical in accordance practical accordance with 2001 Edition of ASME OM Code with Addenda Addenda through 2003. Justify Justify how 1010 CFR 50.55a(f)(4)(i) 50.55a(f)(4)(i) is met.

Response

Response: This portion of Amendment Amendment 97 to the Unit 2 FSAR FSAR was / is in error. The Code of Record Record (COR) listed listed in the affected section is the COR for UnitUnit 1. The COR forthe for the initial Unit Inservice Test Unit 2 Inservice Ten-Year Interval will be in accordance Ten-Year Interval accordance with 10 CFR 50.55a(f)(4)(i). Based on the current Unit 2 schedule and .the 50.55a(f)(4 )(i). Based the state state of current current rulemaking, rulemaking, TVA anticipates anticipates that the Unit 2 COR COR will be the 2004 Edition Edition through 2006 addenda of the ASME ASME Code for Operation Maintenance of Nuclear Operation and Maintenance Nuclear Power Plants Plants (OM (OM Code). However, since since the required COR is based on an unknown date for the Unit 2 operating license, TVA cannot cannot bebe certain certain of the actual COR at this time.

Accordingly, Amendment 100 to the Unit 2 FSAR will revise the the E1-114 E1-114

ENCLOSURE ENCLOSURE 1 Response Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 sentence regarding the COR to read as follows until the date for sentence regarding an operating license becomes becomes more certain: "Inservice testing of ASME Code Class 1, 2, and 3 pumps and valves will be be conducted conducted to the extent practical practical in accordance accordance with the latest edition and addenda of the ASME OM Code incorporatedincorporated byby reference in 10 CFR 50.55a(b) on the date 12 reference 12 months before the before the date of issuance of the operating operating license for Unit 2, as required by 10 CFR 50.55a(f)."

Once the date for issuance issuance of an operating license becomes becomes more certain, certain, TVA will update this sentence in accordance accordance with 1 50.55a(f)(4)(i).

100 CFR 50.55a(f)(4 )(i).

CPTB TVA also indicates indicates in Amendment Amendment No. No. 97 that exceptions to the OM Code Code Inservice Inservice requirements are noted IST program noted in the 1ST program submittal Exceptions submittal made to NRC. Exceptions Testing - 2. to the Code requirements requirements are allowed by NRC regulations, but they must be be identified identified in the 1ST IST program specifically specifically for WBN Unit Unit 2 along with proposed alternatives alternatives and relief requests. In proposing proposing alternatives alternatives or requesting requesting relief, TVA must demonstrate demonstrate that: (1) the alternatives alternatives will provide an acceptable acceptable level of quality and safety, (2)

(2) compliance hardship or unusual compliance would result in hardship unusual difficulty difficulty compensating increase in the level of quality without a compensating quality and safety, or (3)

(3) conformance conformance would be impractical impractical for its facility. The regulations regulations in 10 CFR 50.55a authorize the Commission to approve 50.55a approve alternatives and to grant relief from OM Code requirements requirements upon making the necessary necessary findings.

findings. NRC guidance guidance contained in Generic LeUer contained (GL) 89-04, "A Guidance on Developing Acceptable Letter (GL) Acceptable Inservice Testing Programs," provides Inservice provides alternatives alternatives to Code requirements requirements that acceptable to the NRC staff. Further are acceptable Further guidance guidance for developing an 1ST IST program is given in NUREG-1482, NUREG-1482, Revision Revision 1, "A Guidance Guidance for Inservice Inservice Testing Testing Nuclear Power at Nuclear Motor-Operated Valve Power Plants," GL 89-10, "Safety-Related Motor-Operated Valve Testing and Surveillance,"

Surveillance," and GL 95-07, "Periodic Verification of Design-Basis Design-Basis Safety-Related Motor-Operated Capability of Safety-Related Motor-Operated Valves."

Response: understands that any exceptions TVA understands exceptions or alternatives alternatives to the ASME ASME OM Code requirements require NRC approval and will submit any such exceptions alternatives for approval exceptions or alternatives approval in accordance accordance with NRC guidance and regulations.

NRC guidance El-115 E1-115

ENCLOSURE 1 ENCLOSURE1

Response

Response to Preliminary Preliminary RAIS and RAIs RAls Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 RAls (taken from e-mail from NRC dated 04/28/2010):

Preliminary RAIs 5.5.11-01 Section 5.5.11 does not provide provide any information describing the design design codes/standards/etc. for the pressurizer relief tank. A review of Table 3.2-2, "Summary of Criteria Criteria - Mechanical Mechanical System Components," finds the pressurizer pressurizer relief tank listed on Page 1 of 18, which would be an acceptable acceptable place to provide provide the information. information provided information. However, there is no information provided on the pressurizer pressurizer relief tank line in the columns. ItIt appears that the table columns are out of alignment with the listing of components in the first column.

The staff also notes that there is no information information provided on the Pressurizer Safety Valves Valves line. Again, this appears appears to be because because of the table column column misalignment.

Response

Response: Amendment 98 to the Unit 2 FSAR corrected the alignment FSAR corrected issues in Table 3.2-2.

FSAR 3.2.1 (Seismic Classifications)

Unit 2 FSAR Classifications) presents presents the the applicable Standards for the Structures, applicable Codes and Standards Structures, Systems and Components. Unit 2 FSAR Tables 3.2-2 (Summary of Criteria -

Components.

Mechanical System Components)

Mechanical Components) and 3.2-2a (Classification (Classification of Systems Systems Having Having Major Concerns Related to a Primary Design Concerns Major DeSign Safety Function) provide the Codes and Standards Safety the Standards for the Pressurizer Relief Pressurizer Relief Tank /Piping Pressurizer Safety Valves

/Piping and Pressurizer Valves and Piping. SheetSheet 1 of Unit 22 FSAR Figure 5.1-1-1 (Powerhouse 5.1-1-1 (Powerhouse Unit 1I - Flow Diagram Diagram - Reactor Coolant System) shows shows the codecode class breaks for the Pressurizer Pressurizer Safety Valves and for the the Pressurizer Relief Tank.

Pressurizer 5.5.11-02 Section 5.5.11.3, Design Evaluation, states "The "The rupture disc on the relief tank a ... " This is not clear in that ""...rupture have a..." ... rupture disc.." one indicates there is one disc .. " indicates

... relief tank have rupture disc while ""...relief have..."

... " implies there is more than one rupture rupture disc. ItIt should be noted that the Unit 1I UFSAR UFSAR contains the same words.

NUREG-0847, Watts Bar Unit 1 and Unit 2 SER, Section 5.4.4 states "Tank NUREG-0847, overpressurization protection overpressurization protection is provided provided by two rupture discs."

Clarify the number of rupture discs that are provided with the pressurizer relief tank.

Response

Response: There are two parallel rupture discs on the relief tank in in agreement with NUREG-0847, agreement NUREG-0847, Watts Bar Unit 1 and Unit 2 SER, Section 5.4.4 (Ref. FSAR Figure 5.1-1, 5.1-1, sheet 1 and Drawing Drawing 2-47W813-1).

2-47W813-1 ).

Amendment 100 to the Unit 2 FSAR will revise the first sentence sentence paragraph of 5.5.11.3 (Design Evaluation) to state:

of the second paragraph "The two rupture discs on the relief tank have a total relief capacity equal to or greater than the combined capacity capacity capacity of the the three pressurizer pressurizer safety valves."

El-116 E1-116

ENCLOSURE I1 ENCLOSURE

Response

Response to Preliminary RAIS to Preliminary RAIS and and RAIs Regarding Unit RAls Regarding Unit 2 FSAR Tennessee Authority - Watts Bar Valley Authority Tennessee Valley Bar Nuclear Plant - Unit Nuclear Plant Docket No. 50-391 Unit 2, Docket RAIs for FSAR Section 9.1 - SBPB [taken from NRC letter dated 07/07/2010 RAls for FSAR Section 9.1 - SBPB [taken from NRC letter dated 07/07/2010 (ADAMS Accession (ADAMS Accession No. M MLLI101620047)]:

01620047)]:

Background:

Many handling related structures, systems, and components storage and handling Many of the fuel storage components withinwithin the the WBN Auxiliary Building WBN Building are shared between the two shared between units, including the spent fuel pool, the two units, the spent fuel cooling cleanup system, and the spent cooling and cleanup handling equipment. The WBN Unit spent fuel handling Unit 2 FSAR describes FSAR conformance with the NRC degree of conformance describes the degree General Design NRC General Criteria (GDC)

Design Criteria (GDC) ofof Title 10, Code Code of of Federal Regulations (10 Federal Regulations Appendix A. The ability (10 CFR) Part 50, Appendix ability of shared perform their safety systems to perform safety functions credible combinations functions for credible combinations of normal normal and accident accident requirements of 10 CFR 50.34(b), applicants for states is addressed in GDC 5. Pursuant to the requirements operating licenses operating licenses must must include description and analysis of the structures, include in the FSAR aa description systems, and components of the facility, and the evaluations required and components required to showshow that safety functions will be accomplished.

accomplished.

SBPB 9.1 - 1. Amendment Nos. 40, 48, The NRC issued Amendment 67, and 77 to WBN Unit 1 operating 48,67, license license on September (ML022540925), October September 23, 2002 (ML022540925), October 8, 2003 2003 (ML032880062),

(ML032880062), January (ML073520546), and May 4, 2009 January 18, 2008 (ML073520546),

amendments authorized (ML090920506). These amendments (ML090920506). irradiation of tritium production authorized irradiation production absorber rods (TPBARs) within WBN Unit 1 core burnable absorber burnable core and transfer transfer of these these TPBARs through the shared irradiated TPBARs shared WBN spent fuel pool.

WBN spent pool. In granting these these evaluations of the effect considered evaluations amendments, the NRC staff considered effect of storage of these TPBARs within the shared WBN spent fuel pool on heat generation generation and prevention. However, the WBN Unit 2 FSAR through criticality prevention.

criticality through Amendment Amendment address the presence of the TPBARs within the shared spent No. 97 does not address fuel pit and the effect of these TPBARs on safety safety functions related to fuel fuel storage. Update appropriate information demonstrating Update the FSAR to provide appropriate demonstrating accomplished considering the effects that safety functions would be accomplished TPBAR effects of TPBAR storage in the shared spent fuel pool Response: The effects of TPBARS have been been properly considered in the the calculation of Spent Fuel Pool (SFP) decay decay heat loads. Although irradiate TPBARs in Unit 2, SFP there are no current plans to irradiate decay conservatively consider their use. See decay heat calculations conservatively response to RAI SBPB 9.1 - 2. for additional SFP decay the response aspects of TPBAR storage in the information. Other aspects heat information. the common SFP, e.g., criticality, have been previously addressed addressed in in Unit 1 analyses and licensing activities and are not impacted by licensing activities by Unit 2 operation. As indicated in in the response to RAI RAI SBPB 9.1 - 2, the Unit 2 FSAR will be appropriately appropriately updated.

updated.

E1-117 E1-117

ENCLOSURE1 ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAts RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 SBPB 9.1 - 2. Section 9.1.3.3.3, "Pool and Fuel Temperatures," of the WBN Unit 2 FSAR describes that, with a 12 day decay time, the maximum heat load associated describes with a full core discharge discharge is 28.1 E+06 Btu/hr while the the. maximum heat load for aa full core discharge discharge following a normal refueling outage case is 32.6E+06 Btu/hr.

statement is essentially identical This statement identical to the corresponding corresponding Section Section of the the WBN I1 Updated Updated Safety Analysis Report, which was potentially potentially based solely on operation of WBN Unit 1. Since the operation of a second operation second unit would increase increase the frequency of fuel discharges discharges to the spent fuel pool, the heat load values may not be representative representative of dual-unit dual-unit operating conditions. Confirm the expected heat loads for representative dual-unit dual-unit scenarios and describe describe the methodology, including decay heat models, used to determine determine the heat load.

Response: The expected heat loads for dual-unit dual-unit operating operating conditions with the current installed installed SFP capacity capacity of 1386 1386 locations locations completely filled are as follows:

• • 12-day 12-day decay time, full core discharge: 39.06E+0639.06E+06 BTU/hr BTU/hr

• • Full core discharge discharge following normal outage case:

25.62E+06 25.62E+06 BTU/hrBTU/hr Amendment Amendment 100 to the Unit 2 FSAR will update update the FSAR with these values.

Methodology Discussion Methodology These These SFP decay decay heat loads are calculated in accordance accordance with ANS Standard Standard 5.1,5.1, "Decay Heat Power in Light Water Water Reactors,"

and USNRC Regulatory Guide 3.54, "Spent Fuel Heat Generation Regulatory Guide Generation in an Independent Independent Spent Fuel Pool Storage Installation."

Normal Offload Conditions Normal Conditions The SFP will reach an equilibrium equilibrium situation, alternating refueling situation, alternating refueling offloads between between the two units every every 180 and 355 days. In this this analysis, the pool is filled and the decaydecay heat is calculated based on this offload offload schedule, beginning with Unit 1. To fill the pool to capacity, an additional additional 154 154 assemblies assemblies were given the same age age as the initial Unit Unit 1 batch and added to the pool. The total decay heat heat in the SFP at full pool conditions conditions is obtained obtained by adding the the heat heat load from each each cycle (referred to as the background background heat load) to the total decay heat produced by the most recently heat produced offloaded offloaded core (referred (referred to as the final offload).

The decay heat produced produced by the final offload and the background background decay decay heat is computed accordance with the methodology computed in accordance methodology described described above. All data used accounts for additional additional decay decay heat due to TPBARs. For each offload, the number of fuel fuel assemblies was also taken into account. Unit I1 offloads are assemblies assumed assumed to be 96 assemblies assemblies and Unit 2 offloads offloads are are assumed assumed to be 80 assemblies.

E1-118 E1-118

ENCLOSURE1 ENCLOSURE 1 Response to Response Preliminary RAIS to Preliminary RAIS and and RAIs Regarding Unit 2 FSAR RAls Regarding Watts Bar Authority - Watts Valley Authority Tennessee Valley Tennessee Bar Nuclear Docket No.

Nuclear Plant - Unit 2, Docket No. 50-391 Emergency Offload Conditions Emergency Offload Conditions For the For the emergency offload scenario, emergency offload background heat load scenario, the background load changes slightly. Current Unit calculation changes calculation licensing basis Unit 1 licensing basis requires requires 36 before the shutdown before 36 days to elapse after the first unit shutdown the second unit second shutdown; a period of no greater is shutdown; unit is greater than days is than 60 days elapse before then allowed to elapse completion of the second before completion second (emergency) core offload.

(emergency) decay time for the offload. Thus, the decay the most recent "normal" offloaded recent along with the days. This, along offloaded fuel is 96 days. the slight change in fuel age age for the balance balance of the pool, is factored the pool, factored into determination of the background into the determination heat load (computed as background heat as above above in the Normal Normal Offload Scenario). The decay heat heat for the the emergency offload core emergency taken at 60 days decay.

core is then taken Emergency Offload decay heat Therefore, the total Emergency Therefore, heat is the sum of the "background" decaydecay heat plus the most recently normal discharge batch discharge batch (96 assemblies decayed assemblies decayed for 96 days) plus the the emergency offload emergency (193 assemblies decayed assemblies decayed for 60 days).

E1-119 E1-119

ENCLOSURE11 ENCLOSURE Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 RAIs [taken RAls [taken from NRC letter dated dated 06/29/2010 06/29/2010 (ADAMS Accession No. ML101620006)]:

(ADAMS Accession ML101620006)]:

Performance and Code Review (SNPB)

Nuclear Performance references to WBN Unit 1 are from the approved All references approved FSAR Amendment Amendment No.7. No. 7. All references to WBN Unit 2 are from Amendment Amendment No. 95.

Chapter 4.3.2 SNPB SNPB 4.3.2 - 1. Discuss the initial core loading strategy for WBN Unit 2.

Response: The initial core design for WBN Unit 2 was developed to Response: The initial core design for WBN Unit 2 was developed to meet a number of criteria. These included:included:

• Achieve a minimum of 10% 10% margin margin to peaking peaking limits.

• • Limit the most positive moderator temperature temperature coefficient coefficient during the cycle to less than -1.0 pcm/°F.pcmrF.

Allow hot, full power axial flux difference difference limits of +7%

+7%

and and-12%.

-12%.

• Meet all safety analysis analysis limits with margin.

• Provide 405 days of full power capability to support the the planned planned operation of cycle 1.

Support Support a cycle 2 design with 500 days of full power power capability capability ifif cycle 1 is shut down after 240 or 435 effective effective full power days of operation.

• Keep the power of the fuel at the edge of the core high high enough to reduce the potential that hot leg temperature temperature streaming will adversely impact the measured measured core flow flow in cycle 1.

1.

Minimize fuel cost over cycles 1 through 3.

Minimize Reload designs designs are typically comprised approximately comprised of approximately 84 fresh fuel assemblies assemblies with a large number of fresh Integral large number Fuel Burnable Absorber Absorber (IFBA) rods and relatively relatively few few discrete burnable discrete burnable absorbers. The remainderremainder of the reload design is comprised of fuel that has been operated operated in one or more previous previous cycles.

With all fresh fuel, the initial cycle design characteristics characteristics differ differ from those of a reload design. A large number number of Wet Burnable Absorber (WABA) rods and fewer IFBA Annular Burnable IFBA rods will be used used in in the initial cycle design because because thethe WABA absorber absorber is more effective effective than IFBA for maintaining maintaining a nonpositive moderator temperature nonpositive moderator temperature coefficient.

The moderator temperature temperature coefficient coefficient of an initial cycle cycle design is more positive at aa given boron concentration concentration than the moderator temperature coefficient moderator temperature coefficient of a reload design design duedue El-120 E1-120

ENCLOSURE11 ENCLOSURE

Response

Response to Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Authority - Watts Bar Nuclear Tennessee Valley Authority Nuclear Plant - Unit Unit 2, Docket No. 50-391 to lack lack of fission products and the decreased decreased competition competition for neutrons in the design with all fresh fuel. The increased worth of the soluble boron in in the all fresh fuel design increases the reactivity reactivity effects that arise from a change change inin moderator density.

The Unit 2 initial initial cycle design is very similar to the Unit 1 initial cycle design. Both designs use feed fuel enrichments enrichments of 2.1, wlo in similar numbers. While the RFA-2 2.1, 2.6, and 3.1 w/o fuel assembly design that will be loadedloaded for the initial Unit 2 cycle differs from the Vantage 5H fuel assembly design-that design that was loaded in the initial Unit 1 cycle, the two fuel assembly designs have have very similar neutronic characteristics. This This increases confidence similarity increases confidence that Unit 2 design will operate as designed.

The Unit 2 cycle cycle design is preliminary preliminary pending completion completion of the safety analysis scheduled to be analysis review. The review is scheduled be completed by the end of 2010.

SNPB SNPB 4.3.2 --22 In Table Table 4.3-1 (p 4.3-40) define the two numbers given given for the following in the Fuel Assemblies section:

a. Diameter

~iameter of Guide Thimbles (upper part)

Response: These numbers These numbers are the inside and outside diameters diameters of the guide thimbles. The "10" "ID"and "00" "OD" labels were inadvertently removed.

inadvertently Amendment 100 100 to the Unit 2 FSAR will reinstate reinstate the the "10" and "00" "ID" "OD" labels.

b. ~iameter of Guide Diameter Guide Thimbles Thimbles (lower (lower part)

Response: These numbers are the inside and outside outside diameters diameters of the guide guide thimbles. The "ID" and "OD" "OD" labels were were inadvertently removed.

inadvertently removed.

Amendment 1100 00 to the Unit 2 FSAR will reinstate thethe "ID" "10" and "OD" "00" labels.

c. Diameter Instrument Guide

~iameter of Instrument Guide Thimbles Thimbles Response: These numbers are the inside and outside diameters of the guide thimbles. The "ID" "10" and "OD" "00" labels were were inadvertently inadvertently removed.

removed.

Amendment 100 to the Unit 2 FSAR will reinstate the the "ID" "10" and "OD" "00" labels.

E1-121 El-121

ENCLOSURE 1 ENCLOSURE1 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority Nuclear Plant - Unit 2, Docket Authority - Watts Bar Nuclear Docket No. 50-391 SNPB SNPB 4.3.2 - 10. In WBN Unit 2 Amendment Amendment No. 95 section 4.3.2.4.2, 4.3.2.4.2, what is 4EF 4EF (p 4.3-20)?

4.3-20)?

Response

Response: Amendment 98 to the UnitUnit 2 FSAR corrected corrected "4EF" "4EF" to read 0

"4°F."

"4 F." Since this was an editorial editorial change, aa change change bar waswas not provided, provided, and the amendment amendment level was not revised.

SNPB 4.3.2 - 11.

11. In WBN Unit 2 Amendment No. No. 95 section 4.3.2.8.5 and section 4.3.3.2, LEOPARD is referenced as reference LEOPARD reference 17, however, according the according to the references section, Reference references Reference 17 was deleted deleted by Amendment Amendment 92. Why was was deleted?

amendment deleted?

the amendment

Response

Response: Amendment 98 to the Unit 2 FSAR added the required Amendment required reference for LEOPARD.

LEOPARD.

Chapter 4.3.3 4.3.3 SNPB 4.3.3 - 2.

SNPB In WBN Unit 22 Amendment Amendment No. 95 section 4.3.3.3, should the referencereference to Section 4.3.2.2.7 be to Section Section 4.3.2.2.7 Section 4.3.2.2.6 instead?

instead?

Response: Amendment 99 to the Unit 2 FSAR Amendment "4.3.2.2.7" with FSAR replaced "4.3.2.2.7" "4.3.2.2.6."

"4.3.2.2.6."

Chapter 4.4.1 Chapter SNPB 4.4.1 -1.

SNPB - 1. In WBN Unit 22 Amendment Amendment No. 95 section 4.4.1.1 4.4.1.1 under under the heading

'Discussion' (p

'Discussion' (p 4.4-1), change change 'DBN'

'DBN' to 'DNB'.

'DNB'.

Response

Response: Amendment 100 Amendment FSAR will replace 100 to the Unit 2 FSAR replace "DBN" with "DNB."

Chapter Chapter 4.4.2 SNPB SNPB 4.4.2 - 2. In WBN Unit 2 Amendment 95 page 4.4-11, 4.4-11, there are multiple multiple locations on hand side of the page where the right hand where equation numbers are pasted in the equation numbers the middle of paragraphs middle paragraphs blocking the view of the words. Correct these errors.

Response

Response: Amendment 98 to the Unit Amendment corrected the noted Unit 2 FSAR corrected discrepancies. reviewed the pages discrepancies. TVA reviewed pages currently numbered 4.4~9 through 4.4-23 and ensured as 4.4-9 ensured they were corrected.

SNPB 4.4.2 -3.

SNPB - 3. Confirm that WBN Unit 2 is limited to cores with only RFA-2 fuel and will not Confirm use any other type of fuel until an approved transition transition core methodology methodology is submitted.

E1-122 E1-122

ENCLOSURE ENCLOSURE11 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Regarding Unit 22 FSAR Tennessee Valley Authority Authority - Watts Bar NuclearNuclear Plant - Unit Unit 2, Docket No. 50-391

Response

Response: fabrication contract for WBN specifies that the The fuel fabrication the RFA-2 fuel design will be supplied.

supplied. There There are no provisions provisions for the supply of aa different different fuel design. There There is no plan to change change the fuel design.

design. If TVA decides to change change the fuel design, TVA will comply design, comply with applicable fuel transition transition requirements.

SNPB SNPB 4.4.2 - 5. In WBN Unit 22 Amendment Amendment No. 95 page page 4.4-18, there are are multiple locations locations on the right hand side of the pagepage where where equation numbers numbers are pasted in the middle of paragraphs paragraphs blocking the view of the words. Correct Correct these these errors.

Response: Amendment 98 to the Unit 22 FSAR corrected Amendment corrected the noted discrepancies. TVA reviewed pages discrepancies. pages currently numbered as as 4.4-9 through through 4.4-23 and ensured they were corrected.

corrected.

Chapter 4.4.3 SNPB 4.4.3 - 3.

SNPB In WBN Unit 2 Amendment Amendment No. 95 page 4.4-25, there is one location location on thethe right hand side of the page page where equation numbers equation numbers are pasted pasted in the the middle of paragraphs paragraphs blocking the view of the words. Correct this error.

Response: Amendment 98 to the Unit 2 FSAR corrected Amendment corrected the noted discrepancies. TVA reviewed the pages currently currently numbered as 4.4-9 through 4.4-23 4.4-23 and ensured ensured they were corrected.

corrected.

Chapter Chapter 4.4.5 SNPB SNPB 4.4.5 4.4.5 - 1. In WBN Unit 2 Amendment No. 95 section 4.4.5.1, 4.4.5.1, the figure referred to is Figure 4.4-5. In WBN Unit Unit 22 Amendment Amendment No. 98 section 4.4.5.1, 4.4.5.1, the figure figure referred to is Figure 4.4.6. While there is aa change change in the figure number between between the two amendments, AmendmentAmendment No. 98 has has the page marked as as

'WBNP-95'

'WBNP-95' which means there have been been no changes since Amendment Amendment No. 95. What areare the criteria criteria that would signify a change change and cause cause the the page page to be marked marked 'WBNP-98'?

'WBNP-98'?

Response

Response: A portion portion of the changes incorporated incorporated per Amendment Amendment 98 to the Unit 22 FSAR was the addition of a new Figure Figure 4.4-4.

This resulted in the renumbering of old Figure 4.4-5 to 4.4-6.

As aa result, the reference to "Figure 4.4-5" was changed to "Figure 4.4-6." The change change to the correct correct figure number number in in 4.4.5.1 was an editorial change only; thus, a change in amendment number amendment number was not required for the applicable applicable page page (i.e., 4.4-32 4.4-32 in the Amendment Amendment 98 version).

E1-123 E1-123

ENCLOSURE 1 ENCLOSURE

Response

Response to Preliminary RAIS and RAls RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 RAIs for FSAR 15.5 [from NRC letter dated 07/12/2010 (ADAMS Accession No.

RAls for FSAR 15.5 [from NRC letter dated 07/12/2010 (ADAMS Accession No. MLI101600278)]

ML101600278)]

Accident Dose Branch FSAR 15.5 Branch - FSAR 15.5 15.5 - ADB-1.a.

15.5 For calculations calculations of atmospheric atmospheric dispersion factors (X/Q (x/Q values) using the the ARCON96 methodology, please ARCON96 please provide provide the input files (electronic (electronic files forfor data input into computer computer codes) and aa discussion discussion of the assumptions used to to generate the X/Q generate x/Q values.

Response

Response: Attachment Attachment 11 provides provides the ARCON96 meteorological input ARCON96 meteorological input data in electronic format. Other inputs available in inputs are not available electronic electronic format, but areare provided in hard copy format in Attachment 11. 11. Also included Attachment 11 included in Attachment 11 is a list of assumptions and a discussion discussion of the assumptions assumptions and methodology methodology employed. The atmospheric dispersion factors atmospheric dispersion factors (X/Q (x/Q values) calculated calculated using the above inputs are listed listed in Unit 2 FSAR Table 15.5-14 Unit 15.5-14 (Atmospheric Dilution FactorsFactors At The Control Building).

15.5 - ADB-1.b.

ADB-1.b. Include one or more scaled figures with true north clearly shown, when appropriate, from which distance, height, and direction inputs can be appropriate, be reasonably reasonably approximated. ProvideProvide the scale of each figure. Highlight Highlight all postulated postulated sources and receptors, including the location of the control room postulated release locations.

envelop with respect to the postulated locations.

Response: The requested figures are provided in Attachment Attachment 12.

15.5 - ADB-1.c.

15.5 explain how distance Please explain distance inputs into the ARCON96 ARCON96 calculations were estimated (e.g., horizontal estimated horizontal straight line distances). explain how the distances). Please explain the procedure used to estimate the distances properly factored in differences distances properly differences in heights between each sourcesource and receptor pairpair Response: Distances between the sources and receptors used in in the the dose analysis were horizontal horizontal straight straight line distances calculated by triangulation dimensions shown on the drawings triangulation using the dimensions drawings in Attachment Attachment 12. distances is

12. Use of the horizontal distances conservative since the hypotenuse distance conservative between the distance between the source and receptor would be a greater greater distance, resulting resulting inin smaller X/Q x/Q values.

15.5 - ADB-1.d.

15.5 ADB-1.d. Were any sources modeled as diffuse diffuse or high energy releases? If so, what is determination of the inputs specific to those cases?

the basis for determination Response: sources were modeled as diffuse or high energy releases.

No sources 15.5 - ADB-2. x/Q values were used in the dose assessments to model unfiltered Which X/Q unfiltered inleakage inleakage into the control room envelope and why is use of these X/Q values x/Q values appropriate?

appropriate?

El-124 E1-124

ENCLOSURE1 ENCLOSURE 1 Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket Docket No. 50-391 Response: The same X/Q The same values used x/Q values used inin the the dose dose analysis analysis are are applied applied to to unfiltered inleakage. Use of these the unfiltered these x/Q X/Q values is appropriate appropriate since the intake air vent is a direct direct path path into the MCR and other specific unfiltered leakage specific leakage paths into the MCR MCR have not been identified.

identified. In addition, the maximum x/Q X/Q values used in thethe analysis represent dose analysis represent leakage leakage path locations relatively close to the release point, where where unfiltered inleakage could originate unfiltered inleakage originate over the total habitability envelope envelope surface area, the majority of which is further from the release point than the air intake vents vents used in the dose analysis. The more remote remote inleakage inleakage locations would produce produce smaller X/Q values as compared x/Q values compared to to near the release point. Thus, use of the maximum locations near maximum calculated X/Q calculated x/Q values are conservative conservative relative to unfiltered unfiltered inleakage. A total of 51 cfm unfilteredunfiltered inleakage inleakage into the MCR is assumed in the dose analysis. Tracer gas testing of the MCR assumed MCR indicated the actual inleakage is less than 6 cfm unfiltered inleakage actual unfiltered

[TVA letter to NRC NRC dated 08/04/2004 08/04/2004 (ADAMS Accession No. ML042230173)]. Consequently, the dose analysis ML042230173)]. Consequently, analysis is very conservative relative to unfiltered inleakage.

conservative relative inleakage.

15.5 - ADB-3.a. Please explain if anyany source/receptor sourceireceptor pairs other other than those resulting resulting in the the X/Q values listed in Table 15.5-14 x/Q 15.5-14 were considered.

considered. If so, which source/receptor pairs and X/Q values were compared source/receptor compared to determine determine the the limiting control room x/Q values for each design basis limiting control room X/Q values for each design basis accident? accident?

Response: Unit 22 FSAR Table 15.5-14 lists X/Q Table 15.5-14 x/Q values used used in the FSAR FSAR chapter 15 Main Control Room (MCR) dose analysis for: 1)

LOCA/FHA, SGTR/MSLB/LOSS of AlC LOCAlFHA, 2) SGTRlMSLB/LOSS A/C POWER, and 3)

WGDT Rupture.

Rupture. The Loss of Coolant (LOCA) (LOCA) and Fuel Handling Accident (FHA) result in releases from the Shield Handling Accident Shield Building Stack. The Steam Generator Building Generator Tube Rupture (SGTR),

Main Steam Line Break (MSLB), and Loss of A/C Break (MSLB), AlC Power Power accidents accidents results in releases from the steam generator/main generator/main steam system relief valves located near the roof of the valve steam valve vaults. The Waste Gas Decay Decay Tank (WGDT) rupture accident accident releases from the Auxiliary Building results in releases Building Vent Stack.

isolated during The MCR is isolated during the above above accidents, butbut makeup/pressurization air is supplied to the MCR through makeup/pressurization through one one of two intakes. The above sources and receptors are located drawing 47W200-1 provided in Attachment on drawing Attachment 12. The X/Q x/Q values were determined from each release point to each MCR values were MCR air intake for both units, considering considering various building building configurations.

configurations. The worst case x/Q values were selected for case X/Q for use in the MCR dose analysis and were included in FSAR Table 15.5-14. However, the selection of X/Q Table 15.5-14. X/Q values diddid consider that the Operating Operations to procedures require Operations Operating procedures to switch the MCR intake to the less contaminated contaminated (smaller X/Q) x/Q) air supply supply location 88 hours0.00102 days <br />0.0244 hours <br />1.455026e-4 weeks <br />3.3484e-5 months <br /> post-accident.

El-125 E1-125

ENCLOSURE1 ENCLOSURE 1 Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Regarding Unit 2 FSAR Authority - Watts Bar Nuclear Tennessee Valley Authority Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 15.5 - ADB-3.b.

15.5 ADB-3.b. Please explain how limiting releases were determined (quantitatively or subjectively).

Response

Response: The limiting accident determined based on the accident releases were determined the applicable NRC guidance applicable documentation. For example, the guidance documentation. the LOCA analysis followed the guidelines in NRC Regulatory Guide 1.4. The FHA followed the guidance guidance in in Regulatory Guide 1.25. The SGTR and MSLB Guide guidance MSLB analysis followed guidance NUREG-0800, SRP 15.1.5. The Waste in NRC NUREG-0800, Waste Gas Decay guidelines in Regulatory Tank rupture analysis followed guidelines Guide 1.24 for the design case and in NUREG-0800, Guide 1.24 NUREG-0800, Section 11.3 for the realistic Section 11.3 realistic analysis. No specific NRC specific NRC guidance guidance has beenbeen issued for Loss of A/C AlC Power dose dose analysis. However, this analysis employed appropriateappropriate conservative assumptions.

conservative 15.5 - ADB-3.c. If If only only three source/receptor pairs three source/receptor were considered, pairs were considered, asas implied by the implied by the X/Q x/Q values listed in Table 5.5-14, explain why they were the limiting cases. For For example, was this determined examination of plant determined by examination plant drawings or plant walk-downs?

walk-downs?

Response: indicated in the response As indicated response above, x/Q computations above, the X/Q computations considered numerous sources and receptors,receptors, and the worst case X/Q x/Q values were selected selected for use in the dose analysis.

15.5 15.5 - ADB-3.d.

ADB-3.d. accident scenarios and generated X/Q Do the postulated accident x/Q values model thethe limiting doses considering considering multiple release scenarios, including including those due due to loss of offsite power or other single failures?

single failures?

Response: The dose analysis and associated release scenarios scenarios were based on NRC guidance guidance documentation documentation and considered considered worst case single failures. Loss of off-site power is assumed for all accidents, in addition to the worst case case single failure.

15.5 - ADB-4.

15.5 Please Please provide electronic copy of the PAVAN computer provide an electronic computer code input, if assumptions used provide a list of all inputs and assumptions available. Otherwise, provide used in the PAVAN PAVAN calculations. A copy of the summary pages of the PAVAN outputs is acceptable to show inputs.

acceptable El-126 E1-126

ENCLOSURE1I ENCLOSURE 1 Response to Preliminary Response Preliminary RAIS and RAIs RAls Regarding Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Plant - Unit 2, Docket No. SO-391 50-391 Response: TVA did not use the PAVAN code to calculate xiQ values for the calculate x/Q the exclusion boundary and LPZ. Per the first paragraph of Unit 2 exclusion FSAR 2.3.4.1 (see Amendment Amendment 99 version), "Revised estimates estimates of atmospheric atmospheric diffusion expressed expressed as dispersion dispersion factors (X/Q) calculated for accident have been calculated accident releases releases considered considered as ground-level releases from the Wafts ground-level Watts Bar Nuclear Nuclear Plant for specified time intervals and distances. values distances. The revised X/Q values

. are based on an updated onsite meteorological data base for onsite meteorological 1974 through through 1993 and RG 1.145 calculation methodology..... "

calculation methodology The methodology methodology is discussed discussed in further detail in Unit 2 FSAR Section 2.3, and the results are reported in Unit 2 FSAR reported in FSAR Tables 2.3-66A (Atmospheric Dispersion Factors (X/q), Sec/m 33, Dispersion Factors For Design Basis Accident Accident Analyses Based On OnsiteOnsite Meteorological Meteorological Data For Watts Bar Nuclear Plant) and 15A-2 15A-2

[Accident

[Accident Atmospheric (sec/m 3)].

Atmospheric Dilution Factors (sec/m3)).

15.5 - ADB-5.

1S.S ADB-S. The choice of wind speedspeed categories used in the PAVAN computer computer code code calculations appears calculations appears to result in some clustering of the datadata in the lower lower categories. NRC Regulatory Regulatory Issues Summary Summary (RIS)

(RIS) 2006-4, "Experience with Implementation Implementation of Alternative Alternative Source Source Terms," states that input to PAVAN should have a large number of wind speed speed categories categories at the lower wind wind speeds in order to produce the best results. Therefore, please please provide provide justification that the wind speed categories categories used in the PAVAN calculations calculations have produced adequate adequate estimates estimates of the exclusion area and low population population x/Q values zone X/Q values for the Watts Bar site.

Response: As stated inin the response to RAI1S.S RAI 15.5 - ADB-4., TVA did not use the PAVAN code to calculate x/Q X/Q values for the exclusion exclusion boundary and LPZ.

E1-127

ENCLOSURE 1 ENCLOSURE Response to Preliminary Preliminary RAIS and RAts RAls Regarding Regarding Unit 2 FSAR Tennessee Valley Authority Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391 RAls (taken from NRC e-mail of 06/08):

Preliminary RAIs Quality and Vendor Branch RAIs RAls for FSAR FSAR Section 14.2 14.2 EQVB EQVB 14.2-1.

14.2-1. FSAR Chapter Chapter 14, Table 14.2-1, 14.2-1, Sheet 48 of 90, "[Alternating

"[Alternating Current] AC AC Power Distribution System Test Summary," TVA deleted the requirement Power Distribution requirement to verify item 5 "Demonstrate manualmanual and automatic transfer schemes operate operate in in accordance with design drawings" under "Test Method." The NRC staff did not accordance find a basis or justification justification for deletion deletion of this test.

a. The NRC staff requests TVA to provide a basis or justification justification why this test is not required required for start of WBN Unit 2 to demonstrate demonstrate the capability of the the manual and automatic manual automatic transfer schemes for the AC power distributiondistribution system for dual unit operation.

operation.

Response

Response: 1. The automatic automatic and manual transfer features of the the Unit 2 6.9kV Shutdown Boards have Unit have been tested since since Unit Unit 1 began operation operation and are are tested tested every months every 18 months on all four 6.9kV Shutdown boards in accordanceaccordance withwith 0-SI-211-1 demonstrates fulfillment of 0-SI-211-1 which demonstrates TS SR 3.8.1.8.

3.8.1.8.

2. Unit 2 startup and operation will be the same as as Unit 1l's.

'So Thus, there are no planned design changes design changes to the transfer transfer schemes of the Unit 2 6.9kV Shutdown Boards.

3. For Unit 2, the manual transfer from normal to alternate alternate feeder is performed performed remotely bycontrol remotely by control room handswitch and locally by handswitches handswitch handswitches provided on on the compartment compartment doors. This is the same as the Unit 1 methodology.

4. Automatic Automatic transfers transfers of both the Unit 1 and Unit 2 6.9kV Shutdown Boards from normal feeder to to Standby power supply feeder are initiated alternate or Standby initiated as follows:

a. A fault condition on the normal power supply supply feeder common station service service transformer transformer (CSST) or a line fault on the Preferred Preferred Offsite Offsite power supply, 161 kV line, originating in the Watts Watts Bar Hydro Hydro switchyard switchyard results in transfer transfer to the the Alternate power supply feeder. Note 4 on TVA 1-45W760-21 1-1 and 2-45W760-211-1 Drawings 1-45W760-211-1 2-45W760-21 1-1 provides additional detailed detailed information information on the the automatic transfer transfer schemes for the 6.9kV Shutdown Boards.

E1-128 E1-128

ENCLOSURE ENCLOSURE 1 Response to Preliminary Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR FSAR Tennessee Valley Authority Tennessee Nuclear Plant - Unit 2, Docket No. 50-391 Authority - Watts Bar Nuclear b.

b. A Loss of Voltage Voltage or Degraded Degraded Voltage condition Power supply of the Preferred Offsite Power supply on either the the Normal or Alternate feeder, as sensed at the Normal the 6.9kV Shutdown Board(s), results in transfer transfer to the the Onsite Standby Standby power power supply. This transfer scheme will be tested on the Unit 2 6.9kV scheme Shutdown Boards Shutdown operations in Boards prior to Unit 2 operations in accordance with 2-PTI-262-01 accordance 2-PTI-262-02.

2-PTI-262-01 and 2-PTI-262-02.

5. Both the automatic and manual transfer schemes, from manual transfer Normal feeder to the Alternate feeder, are not the Normal affected affected by loads or loading on the 6.9kV Shutdown Board electrical buss. The transfer transfer scheme scheme is solely dependent upon the initiating condition and source of dependent the condition as noted in 4.a. above.

6. paragraph 4.a. above, To test the condition noted in paragraph the automatic actuated by automatic transfer scheme test is actuated by simulating an electrical fault on the Normal feeder feeder Offsite Power supply by rotating the appropriate appropriate protective electrical relay disc closing a contact protective contact in thethe transfer circuit initiating initiating an automatic transfer from the the normal feeder to the alternate alternate feeder.

7. protective relay that initiates the transfer The protective transfer from Normal Alternate feeder initiates a transfer Normal feeder to Alternate transfer on both the Unit 1I and the Unit 2 6.9kV Shutdown Boards Boards simultaneously due to the common power supply.

simultaneously

b. description of the transfer scheme Also, provide a description include whether scheme to include re-sequenced on, or ififthe loads are block running loads are shed and then re-sequenced loaded.

Response: 1. automatic transfer schemes are described The automatic described in in paragraph 4. of the response to RAI EQVB paragraph 14.2-1.a.

EQVB 14.2-1.a.

2. For all automatic transfers due to transformer or line line Alternate faults, the transfer from Normal feeder to Alternate feeder is classified as a Fast transfer and does not interrupt any power or loads on the 6.9kV Shutdown interrupt Shutdown Boards. No loads are shed, and loads on the buss are not required to be shed and sequenced sequenced back onto the the buss.

3. For all Loss of Voltage or Degraded Voltage conditions conditions which initiate an automatic transfer to the Onsite Onsite Standby power supply, the following actions take place Standby place E1-129 E1-129

ENCLOSURE 1 ENCLOSURE Response Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit Unit 2, Docket No. 50-391 once the transfer is initiated:

initiated:

a. Any normal normal or alternate power sourcesource breaker breaker is opened and all Normal or Alternate feeder breakers are locked out, loads are shed, the the affected diesel affected generator (DG) diesel generator (DG) is started, voltage voltage and frequency are checkedchecked on the DG and, if if within limits, the emergency DG breaker is closed.

Once Once proper voltage is returned to the buss, the the

.loads loads will then sequence sequence on. See drawing drawing 45W760-211-16 for sequence 45W760-211-16 sequence times.

b. There There is no block loading of loads onto the buss buss once load shedding has occurred when supplied by the Onsite Standby power power source. By design, there are some some loads not shed. For example, the the 6.9kV load breakers breakers providing providing power to the 480V AC Shutdown Shutdown Board Transformers Transformers are not shed; they remain closed, and will reenergize upon the the DG breaker closing in to the 6.9kV Shutdown Board.

c. If If TVA is taking credit for the Technical Specification Specification (TS) Surveillance Surveillance Requirement (SR) 3.8.1.8 currently performed for WBN Unit 1 every Requirement 18 18 months for deleting deleting this test requirement then describe describe how the loads loads of WBN Unit 2 are included included in this surveillance.

surveillance.

Response

Response: 1. As far as performing performing the manual and automatic automatic transfers for the Unit 2 6.9kV Shutdown Shutdown Boards, there are no design changes being made to the transfer transfer schemes. The schemes for Unit 2 remain the same as as Unit Unit 1 and are currently currently being tested at the required surveillance frequency.

surveillance

2. Loads that will be added to the Unit 2 6.9kV Shutdown Boards as a result of completing Unit 2 will be tested Boards under under 2-PTI-262-01 2-PTI-262-01 and 2-PTI-262-02 2-PTI-262-02 for the Loss of Degraded Voltage Voltage and Degraded Voltage transfer transfer schemes that result in the Onsite Standby Standby power supply supply providing providing power power to the shutdown shutdown boards.

These preop tests will verify that, when required, the the shutdown shutdown boards load shed, start the Onsite standby power supply, sequence sequence on the required loads, with or without an accident signal present, and verify that the the Onsite Standby power supply meets all required required loading design design calculations calculations for Unit 2 operation.

operation.

3. The surveillance (0-SI-211-1)

(0-SI-211-1) does not take into consideration consideration loading on the 6.9kV Shutdown boards loading boards E1-130 E1-130

ENCLOSURE 1 ENCLOSURE1 Response to Preliminary Response Preliminary RAIS and RAls RAIs Regarding Unit 2 FSAR FSAR Tennessee Valley Authority Tennessee Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. No. 50-391 when testing is conducted.

conducted.

4. The manual transfers of the boards are performed manual transfers under normal loading loading as well as the automatic transfer transfer initiated by simulation simulation of the Preferred Preferred Offsite Offsite power supply CSST or line fault as noted rioted in paragraph 4.a. of the response to RAI EQVB EQVB 14.2-1.a.

14.2-1.a.

Summary It is appropriate appropriate to not require the preoperational preoperational testing of the Unit 2 6.9kV Shutdown Shutdown Boards Boards manual and automatic transfer transfer due to a transformer transformer or line line fault. For the Unit 2 6.9kV Shutdown Shutdown Boards, the Loss of Voltage and Degraded Voltage automatic transfer to the Onsite Standby Standby power supply will be tested with the new loading accordance with 2-PTI-262-01 loading for Unit 2 operation in accordance 2-PTI-262-01 and and 2-PTI-262-02.

2-PTI-262-02.

References References

• • 45W760-211 45W760-211 series drawings drawings

• • 45W760-82 45W760-82 series drawings drawings

• Month 6.9 KV Shutdown Boards Transfer 0-SI-211-1 (18 Month Transfer From Normal Normal To To Alternate Alternate Supply)

• • 2-PTI-262-01 (Unit 2-PTI-262-01 (Unit 2 Integrated Integrated Safeguards Safeguards Test, Train 2A)

• • 2-PTI-262-02 (Unit 2-PTI-262-02 (Unit 22 Integrated Integrated Safeguards Safeguards Test, Train 2B) 2B) 14.2-2.

EQVB 14.2-2. FSAR Chapter 14, Table 14.2-1, Sheet 48 of 90, "AC Power Distribution System 14.2-1, Sheet Test Summary," (renumbered item 6) to state that it will Summary," TVA revised Item 7 (renumbered verify the capability of each common station service transformertransformer (CSST) to carry the load required engineered safety feature '(ESF) required to supply engineered (ESF) loads for its respective load group under respective under WBN Unit 2 loss-of-coolant loss-of-coolant accident (LOCA) conditions. The NRC staff finds this commitment to be different than the staffs staff's acceptance documented acceptance documented in the Supplements Supplements 14 and 16 to the NUREG NUREG -0847, "Safety Evaluation Report Related to the Operation Operation of Watts Bar Nuclear Plant Units 1 and 2." In Supplement NUREG-0847, the NRC staff accepted Supplement 16 to NUREG-0847, accepted TVA's position position to demonstrate demonstrate the capability capability of each CSST to carry the load load required to supply ESF loads on one unit (WBN Unit 1) under under LOCA condition since TVA, at that time, was not seeking an operating operating license for WBN Unit 2.

The NRC documented its acceptance NRC staff documented acceptance of TVA's position based on the the commitment that before before issuance issuance of an Operating Operating License for WBN Unit 2, TVA would demonstrate the capability would have to demonstrate capability of each CSST to carry the load load required to supply ESF loads of one unit (WBN Unit Unit 2) under under LOCA conditions in in addition to power required for shutting down the non-accident non-accident unit (WBN Unit 1).

Therefore, the NRC staff requests requests that TVA revise its test commitment commitment to verify the capability capability of each CSST with LOCA conditions in WBN Unit 2 in addition to power required for normal shutdown shutdown (non-accident (non-accident loads) of WBN Unit Unit 1, or provide provide an explanation explanation and justification commitment was justification why the original commitment was revised.

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ENCLOSURE11 ENCLOSURE Response to Preliminary RAIS and RAIs Response Regarding Unit 2 FSAR RAls Regarding Tennessee Valley Authority Authority - Watts Bar NuclearNuclear Plant - Unit 2, Docket No. 50-391 Response: See the response response to RAI 14 - 1.

EQVB 14.2-3. FSAR Chapter 14, Table 14.2-1,14.2-1, Sheet 49 of 90, "AC Power Distribution Distribution System Test Summary," TVA deleted deleted the requirement requirement to verify item 2 under Acceptance Acceptance Criteria. The NRC staff did not find a basis or justification for deletion of this this test.

a. The NRC staff requests TVA to provide a justification justification why this test is not not required required for start of WBN Unit 2.

Response: This test will be performed performed by 2-PTI-262-01 2-PTI-262-01 (Unit 2 Integrated Safeguards Integrated Safeguards Test, Train 2A) and 2-PTI-262-02 2-PTI-262-02 Integrated Safeguards (Unit 2 Integrated Safeguards Test, Train 2B). 2B).

Amendment Amendment 100 to the Unit 2 FSAR will add item 2, "Power supply to safety related loads will automatically automatically and manually transfer transfer to the onsite (standby) diesel units from the normal alternate supply or manually from the diesel normal or alternate diesel generator units back to the normal or alternate supply supply as described by FSAR Section 8.3.1" back to Table 14.2-1, 14.2-1, Sheet 49.

b. TVA should describe describe how the loads of WBN Unit 2 are addressed addressed with respect to the capability capability of the manual manual and automatic transfer transfer schemes for for the AC power distribution distribution system between between onsite (standby) diesels units units from normal normal or alternate alternate supply for dual-unit operation.

Response

Response: Unit 2 startup and operation operation will be the same as Unit l's. 1'so Thus, there are no planned planned design design changes to the transfer schemes. The only changechange inin loading loading will be the addition of previously in service and tested with Unit 2 loads not 'previously with Unit 1. The additional additional Unit 2 Engineered Engineered Safety Features Features loads are accounted accounted for by tests 2-PTI-262-01 (Unit 2 2-PTI-262-01 (Unit Integrated Safeguards Test, Train 2A) and 2-PTI-262-02 2-PTI-262-02 (Unit (Unit 2 Integrated Safeguards Test, Train 2B).

El-132 E1-132

ENCLOSURE11 ENCLOSURE Preliminary RAIS and RAls Response to Preliminary RAIs Regarding Regarding Unit 2 FSAR Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket No. 50-391 EQVB 14.2-4. FSAR Chapter 14, Table 14.2-1, Sheet 44 of 89, "Diesel Generator 14.2-1, Sheet Generator Test Summary," TVA deleted the requirements to verify items 1 through 5 under under Test Method.

Method. The NRCNRC staff did not find a basis or justification for deletion deletion of these these tests. The NRC NRC staff requests TVA to provide provide a justification why these tests are not required required for start of WBN Unit 2. If these tests are currentlycurrently being performed under the WBN Unit 1 TS SRs, then provide aa summary summary of the impact on these tests for the WBN Unit 22 diesel generators generators due to dual- unit operation.

operation. Also, identify the WBN Unit 1 TS surveillances surveillances that are currently being performed to being performed to accomplish these tests.

Response

Response: 1.

1. These tests were previously performed as part of declaring declaring the DGs functional and operable for Unit 1 operation. All four four DGs were designated designated as required required for Unit 1 operations.

2. Preop testing for these Test Methods was completed for all DG systems and support support systems systems on Unit 1. The DGs are being maintained maintained operable operable per the Unit 1 TS, including including surveillance surveillance requirements, for single unit operation.

3. In changing to a dual unit operation, operation, the only impact on the the current testing testing methodology methodology is to add the Unit 22 loads loads not currently being tested and to revise the appropriate appropriate surveillance instructions.

instructions. Unit 2 preop preop testing will test the the additional loads on the Unit 2 DGs to confirm designdesign calculations.

4. The Unit Unit 2 DGs will be tested again to account account for the the additional loading loading required required for Unit 2 operations prior to beingbeing declared declared operable per the Unit Unit 2 TS for Unit 2 operations.

The testing will satisfy Unit 1 TS SRsSRs during during the Unit 2 preop testing in order to be able to call the Unit Unit 2 DGs and Unit 2 Shutdown Boards TS operable for Unit 1 coming out of Unit 1 Shutdown

. RF10. The Unit 2 PTIs PTls will contain contain all SRs required required for Unit 1 and later Unit 2 SRs, as currently currently known.

Summary ItIt is appropriate to not require the five items listed under under Test Method due due to the previous testing and acceptance acceptance for UnitUnit 1 operations. The Unit 2 DGs have been maintained operable per the Unit 1 I TS since that acceptance acceptance testing. The only impact impact on the UnitUnit 2 DGs due due to dual unit operations operations is the new new Unit 2 loads loads that will be added to the DG. DG. Integrated Safeguards Safeguards testing for Unit 2 operations operations will retest the Unit 22 DGs and 6.9kV Shutdown Boards Boards to ensure the components components will fulfill their required safety function as well as all design features.

El-133 E1-133

ENCLOSURE ENCLOSURE 1 Response to Preliminary Preliminary RAIS and RAls RAts Regarding Unit 2 FSAR Tennessee Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391

References:

• • Safeguards Test, Unit 1-PTI-262-01, Integrated Safeguards 1-PTI-262-01, Unit 1

• • 2-PTI-262-01, Integrated Safeguards Test, Unit 2-PTI-262-01, Unit 2, Train 2A

• • Safeguards Test, Unit 2, Train 2B Integrated Safeguards 2-PTI-262-02, Integrated

• • 18 Month Loss of Offsite Power - DG 2A-A 0-SI-82-5, 18

• 0-SI-82-6, 18 Month Offsite Power - DG 2B-B Month Loss of Offsite 2B-B

• • 0-SI-82-12-A, Monthly DG Start and Load Test DG 2A-A

• • 0-SI-82-12-B, Monthly DG Start and Load Test DG 2B-B 0-SI-82-12-B,

• 0-SI-82-15, 24 Hour Load Run - DG 2A-A 0-Sl-82-15,

• • 0-SI-82-16, 24 Hour Load Run - DG 2B-B

• • 184 Day Fast Start and Load Test DG 2A-A 0-SI-82-19-A, 184 0-SI-82-19-A,

• • 184 Day Fast Start and Load Test DG 2B-B 0-SI-82-20-B, 184 0-SI-82-20-B, 2B-B

• • 0-SI-215 series - Series of surveillances 0-SI-215 batteries surveillances on the DG batteries and chargers.

EQVB 14.2-5.

14.2-5. 14.2-1, Sheet 45 of 89, "Diesel Generator FSAR Chapter 14, Table 14.2-1, Generator Test requirement to verify items 10 under Test Method.

Summary," TVA deleted the requirement Method.

a. If this test is currently performed on WBN Unit 2 diesel generators If generators under under the WBN Unit 1 TS SR 3.8.1.14 requirements, then confirm that the WBN Unit 1 TS SR 3.8.1.14 accomplishes this test.

3.8.1.14 accomplishes

Response

Response: 0-SI-82-15.(24 0-SI-82-15 0-SI-82-16 (24 Hour Load Run - DG 2A-A) and 0-SI-82-16 (24 Hour Load Run - DG 2B-B) are performed performed to satisfy Unit 1 TS SR 3.8.1.14 requirements. This TS SR has a frequency of 18 months and may be performed performed anytime in in Modes 1 through 4, if Modes if required.

required.

This 24-hour test fulfills the requirements of the items items mentioned in Test Method mentioned Method 10.

b. Also, confirm performed for verifying WBN Unit confirm that this surveillance performed Unit 2 generator capacity diesel generator design-basis accident capacity envelops the design-basis loads accident (DBA) loads of WBN Unit 2 plus the power required for the WBN Unit 1 loads.

Response

Response: The capacity capacity of the Unit 2 diesel generators generators to envelope the the Unit 2 design-basis (DBA) loads plus the common design-basis accident (DBA) demonstrated by tests (shared) Unit 1 loads will be demonstrated 2-PTI-262-01 Integrated Safeguards 2-PTI-262-01 (Unit 2 Integrated Safeguards Test, Train 2A) and 2-PTI-262-02 (Unit 2 Integrated Safeguards Safeguards Test, Train 2B).

El-134 E1-134

ENCLOSURE 1 ENCLOSURE1 Response to Response to Preliminary Preliminary RAIS RAIS and and RAIs RAls Regarding Regarding Unit 2 FSAR FSAR Tennessee Valley Tennessee Valley Authority Authority - WattsWatts Bar Nuclear Plant Bar Nuclear Plant - Unit Unit 2,2, Docket Docket No.No. 50-391 50-391 EQVB EQVB 14.2-6.

14.2-6. The FSAR The FSAR Table 14.2-1, 14.2-1, Chapter Chapter 14 (Sheets 44 44 and and 45 of 89), "Diesel Generator Test Summary," does not list Generator list aa test that that the WBN Unit Unit 22 diesel generators automatic generators automatic trip are bypassed bypassed on automatic automatic or or emergency emergency startstart signal except for for engine engine over speed speed and and generator generator differential differential current.

a. The NRC The NRC staff requests requests TVATVA to provide a justification justification why this this test is is not not required for start of WBN Unit Unit 2.

Response

Response: Currently, surveillances Currently, performed by Unit surveillances performed Unit 1 per 0-SI-82-3 0-SI-82-3 (18 (18 Month Month Loss of Offsite Power Power With Safety Injection -

Safety Injection-DG 1-AA) and 0-SI-82-4 O-S 1-82-4 (18 Month Loss of Offsite (18 Month Power Offsite Power With Safety Injection - DG 1-BB)

Safety Injection 1-BB) check the engine engine and and generator trips are disconnected generator disconnected as required every 18 months. When the new surveillances 18 surveillances for Unit 22 (i.e.,

(i.e.,

2-SI-82-5, and 2-SI-82-6) are written, 2-SI-82-5, written, the the Unit 2 DG checks checks will be removed from the Unit 1 surveillances.

surveillances.

feature will also be This feature be verified verified as part of 2-PTI-262-01 (Unit 2 Integrated Integrated Safeguards Safeguards Test, Train Train 2A) and and 2-PTI-262-02 (Unit 2-PTI-262-02 (Unit 22 Integrated Safeguards Test, Train 2B).

Integrated Safeguards 2B).

compliance with Reg. Guide To be in compliance Guide 1.9 and System Description N3-82-4002 (Standby Diesel Generator Generator System) Amendment Amendment 100 to the Unit 2 FSAR FSAR will add aa requirement requirement to test the above above features to Table 14.2-1. 14.2-1.

b. If If this test is currently currently performed performed on WBN Unit 2 diesel generators generators under under the WBN Unit 1 TS SRs, then identify identify the WBN Unit 1 TS SR that that accomplishes this test.

accomplishes Response: The current Unit 1 TS SRs are applied to both the Unit 1 and the Unit 2 diesel generators.

generators. Unit 2 is tested to the the requirements as Unit 1 (i.e., TS SR 3.8.1.13).

same requirements El-135 E1-135

ENCLOSURE22 ENCLOSURE List of Regulatory Regulatory Commitments Commitments Tennessee Valley Authority - Watts Bar Nuclear Nuclear Plant - Unit 2, Docket Docket No. 50-391

1. Amendment Amendment 100 to the Unit 2 FSAR will be revised as noted in the applicable applicable preliminary RAI / RAI responses.

2. The analysis for Panels 2-L-11A 2-L-1 1A and 2-L-11 NRC by 2-L-1 1 B will be submitted to the NRC November November 30, 2010.

3. qualification for PAMS Cabinet and Components The qualification Components and Main Control Room Components will be submitted Components submitted to the NRC by January January 14, 14, 2011.

2011.

4. A non-proprietary non-proprietary version of and an affidavit for withholding withholding for EQ-EV-39-WBT, EQ-EV-39-WBT, Revision 1 will be submitted to the NRC by November November 30, 2010.

5. A non-proprietary non-proprietary version of and an affidavit for withholding withholding for TR-1136 TR-1 136 will be submitted submitted to the NRC by December December 17, 2010.

6. A non-proprietary non-proprietary version of and an affidavit for withholding withholding for 16690-QTR, 16690-QTR, Revision Revision 0 will be submitted submitted to the NRC by November November 30, 2010.

7. A corrected corrected proprietary proprietary version of,of, a non-proprietary non-proprietary version of, and an affidavit for withholding for Thermo Fisher Thermo Fisher Scientific Qualification Scientific Qualification Report No. 864, Rev. 0 will bebe submitted to the NRC by NovemberNovember 15, 2010.

8. Unit 2 System Description Description for the Reactor Reactor Coolant (WBN2-68-4001) will be Coolant System (WBN2-68-4001) be revised to reflect reflect required revisions to the PTLR by September September 17, 2010.

9. Once the date for issuance issuance of an operating operating license becomes more certain, certain, TVA will update the sentence sentence regarding the Code of Record Record in accordance accordance with 10 CFR 50.55a(f)(4 50.55a(f)(4)(i).

)(i).

ENCLOSURE ENCLOSURE 3 List of Files Provided Provided on Enclosed Optical Optical Storage Media (OSM)

(OSM)

Tennessee Valley Authority Authority - Watts Watts Bar Nuclear Plant - Unit 2, Docket No. 50-391 File Name Fill Size - Bytes Bytes 001 - ATTACHMENT ATTACHMENT 1 - Foxboro Test ResultsResults 338,284,559 338,284,559 002 - ATTACHMENT ATTACHMENT 2 - EQ-EV-39-WBT, EQ-EV-39-WBT, Revision Revision 1 (Proprietary)

(Proprietary) 938,156 938,156 003 - ATTACHMENT ATTACHMENT 3 - TR-1136 (Proprietary)

TR-1136 (Proprietary) 14,939,714 14,939,714 004 - ATTACHMENT ATTACHMENT 4 - 16690-QTR 16690-QTR (Proprietary) 21,740,229 005 - ATTACHMENT ATTACHMENT 5 - Qualification Qualification Report 864 (Proprietary) 9,983,274 006 - ATTACHMENT ATTACHMENT 6 - Pictorial Pictorial Depiction of APS 400,199 400,199 007 - ATTACHMENT ATTACHMENT 7 - AC APS Analysis 213,153 213,153 008 - ATTACHMENT ATTACHMENT 8 - AC APS Analysis 142,157 142,157 ATTACHMENT 9 - 125V DC Vital Battery System Analysis 009 - ATTACHMENT 1,408,247 1,408,247 010 - ATTACHMENT 10 - 125V DC Vital Battery System System Analysis 361,406 361,406 011 - ATTACHMENT ATTACHMENT 11 - Atmospheric Atmospheric Dispersion Factors Supporting 1,651,483 1,651,483 Information Information 012 - ATTACHMENT ATTACHMENT 12 - Drawings to Support Accident Dose Review Review 1,321,171