Text

Fuel Thermal Conductivity Degradation, Response to Supplemental Safety Evaluation Report Open Item 61
	ML13225A024
Person / Time
Site:	Watts Bar
Issue date:	08/06/2013
From:	Hruby R; Tennessee Valley Authority
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	CAW-13-3755
	Download: ML13225A024 (97)
	v • d • e

Enclosure 1 is to be withheld from public disclosure under 10 CFR § 2.390. When separated from this enclosure, this letter is decontrolled.

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381-2000 August 6, 2013 10 CFR 50.46 10 CFR 2.390(b)(4)

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

Subject:

WATTS BAR NUCLEAR PLANT UNIT 2 - FUEL THERMAL CONDUCTIVITY DEGRADATION, RESPONSE TO SUPPLEMENTAL SAFETY EVALUATION REPORT OPEN ITEM 61

References:

1. NRC Information Notice 2011-21, dated December 13, 2011, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation" (ADAMS Accession No. ML113430785)

2. TVA Letter to NRC dated October 16, 2012, "Watts Bar Nuclear Plant (WBN) Unit 2 - 10 CFR 50.46 - Estimated Increase in Peak Clad Temperature (PCT) Due to the Effect of Fuel Pellet Thermal Conductivity Degradation and Peaking Factor Burndown Not Being Considered - 30 Day Report" (ADAMS Accession No. ML12296A226)

3. Supplemental Safety Evaluation Report (SSER) 24, "Safety Evaluation Report Related to the Operation of Watts Bar Nuclear Plant, Unit 2, Docket No. 50-391, Tennessee Valley Authority,"

published September 2011 The purpose of this letter is to provide information that addresses the Thermal Conductivity Degradation (TCD) effect on the Large-Break Loss of Coolant Accident (LOCA) analysis for Watts Bar Nuclear Plant (WBN) Unit 2. The information provided in Enclosures 1 and 4 completes TVA's response to Supplemental Safety Evaluation Report (SSER) Appendix HH, Open Item 61.

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U.S. Nuclear Regulatory Commission Page 2 August 6, 2013 The Nuclear Regulatory Commission (NRC) approved Westinghouse Best-Estimate Loss-of-Coolant Accident (BELOCA) Automated Statistical Treatment of Uncertainty Method (ASTRUM) methodology is based on the Westinghouse Improved Performance Analysis and Design (PAD) 4.0 fuel performance code. PAD 4.0 was licensed without explicitly considering fuel TCD with burnup.

Fuel performance data that accounts for fuel TCD was used as an input to the updated WBN Unit 2 LOCA analysis discussed herein. The new PAD fuel performance data was generated with an updated PAD model that includes explicit modeling of TCD.

Therefore, the BELOCA analysis was updated to consider the fuel TCD effects cited in Reference 1. The discussion and results of this reanalysis are provided in Enclosure 1.

Mark-ups of Final Safety Analysis Report (FSAR) Section 15.4.1 are provided in. The FSAR changes will be incorporated in Amendment 110. contains information proprietary to Westinghouse. Accordingly, TVA respectfully requests that this proprietary information be withheld from public disclosure in accordance with 10 CFR 2.390. Enclosure 2 provides the supporting affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390.

NRC correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-1 1-3149 and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066. provides the non-proprietary version of Enclosure 1. Enclosure 4 provides marked-up sections of the WBN Unit 2 FSAR incorporating the new information. provides the new commitment made in this letter.

If you have any questions, please contact me at (423) 365-1260 or Gordon Arent at (423) 365-2004.

I declare under penalty of perjury that the foregoing is true and correct to the best to my knowledge. Executed on the 6 th day of August, 2013.

Respectfully, Raymond A. Hruby, Jr.

General Manager, Technical Services Watts Bar Unit 2

U.S. Nuclear Regulatory Commission Page 3 August 6, 2013

Enclosures:

1. Proprietary Attachment to Westinghouse Letter to TVA WBT-D-4396, dated July 23, 2013, "Thermal Conductivity Degradation (TCD) effect on the Large-Break LOCA Analysis for Watts Bar Nuclear Plant Unit 2"

2. Westinghouse Affidavit CAW-1 3-3755 for Withholding Proprietary Information from Public Disclosure

3. Non-Proprietary Attachment to Westinghouse Letter to TVA WBT-D-4396, dated July 23, 2013, "Thermal Conductivity Degradation (TCD) effect on the Large-Break LOCA Analysis for Watts Bar Nuclear Plant Unit 2"

4. WBN Unit 2 FSAR Chapter 15 Mark-up

5. New Commitment cc (Enclosures):

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

ENCLOSURE 2 Tennessee Valley Authority Watts Bar Nuclear Plant, Unit 2 Docket No. 50-391 WESTINGHOUSE AFFIDAVIT CAW-13-3755 FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE E2-1

S Westinghouse Westinghouse Electric Company Engineering, Equipment and Major Projects 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 720-0754 11 555 Rockville Pike e-mail: maurerbf@westinghouse.com Rockville, MD 20852 Proj letter: WBT-D-4396 CAW-1 3-3755 July 10, 2013 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

WBT-D-4396 P-Attachment, "Thermal Conductivity Degradation Effect on the Large-Break LOCA Analysis for Watts Bar Unit 2" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-13-3755 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Tennessee Valley Authority.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference CAW-1 3-3755, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 310, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Very truly yours, Bradley F. Maurer, Principal Engineer Plant Licensing Enclosures

CAW-1 3-3755 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

SS COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared Bradley F. Maurer, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

Bradley F. Maurer, Principal Engineer Plant Licensing Sworn to and subscribed before me this 10th day of July 2013 Notary Public COMMONWEALTH OF PENNSYLVANIA N OTARIAL SE AL

\\Renee Giampoie.

Notary Public Penn TownShiP, \\Nestmoreland county my commission Expires September 25,20

2 CAW-13-3755 (1) 1 am Principal Engineer, Plant Licensing, in Engineering, Equipment and Major Projects, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) 1 am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3)

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4)

Pursuant to the provisions of paragraph (bX4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i)

The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii)

The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a)

The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

3 CAW-1 3-3755 (b)

It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c)

Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

(d)

It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e)

It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f)

It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a)

The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b)

It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c)

Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

(d)

Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component

CAW-13-3755 may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e)

Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f)

The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii)

The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(iv)

The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v)

The proprietary information sought to be withheld in this submittal is that which is appropriately marked in WBT-D-4396 P-Attachment, "Thermal Conductivity Degradation Effect on the Large-Break LOCA Analysis for Watts Bar Unit 2" (Proprietary), for submittal to the Commission, being transmitted by Tennessee Valley Authority letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse for Tennessee Valley Authority use for Watts Bar Unit 2 to close out certain NRC open items related to the Large Break Loss-of-Coolant Accident analysis, and may be used only for that purpose.

This information is part of that which will enable Westinghouse to:

(a)

Provide input to the U.S. Nuclear Regulatory Commission for review of Watts Bar Unit 2 (WBT) Completion Project submittals.

5 CAW-13-3755 (b)

Provide information on fuel thermal conductivity degradation results for the LB BELOCA.

Further this information has substantial commercial value as follows:

(a)

Westinghouse plans to sell the use of the information to its customers for the purpose of meeting NRC requirements for licensing documentation.

(b)

Westinghouse can sell support and defense of the technology to its customer in its licensing process.

(c)

The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar calculations and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests to close out certain NRC open items related to the Large Break Loss-of-Coolant Accident analysis.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

Tennessee Valley Authority Letter for Transmittal to the NRC The following paragraphs should be included in your letter to the NRC Document Control Desk:

Enclosed are:

1. One (1) copy of WBT-D-4396 P-Attachment, Revision 0, "Thermal Conductivity Degradation Effect on the Large-Break LOCA Analysis for Watts Bar Unit 2" (Proprietary)

2. One (1) copy of WBT-D-4396 NP-Attachment, Revision 0, "Thermal Conductivity Degradation Effect on the Large-Break LOCA Analysis for Watts Bar Unit 2" (Non-Proprietary)

Also enclosed is the Westinghouse Application for Withholding Proprietary Information from Public Disclosure CAW-l 3-3755, accompanying Affidavit, Proprietary Information Notice, and Copyright Notice.

As Item 1 contains information proprietary to Westinghouse Electric Company LLC, it is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2.390 of the Commission's regulations.

Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2.390 of the Commission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-13-3755 and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 310, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

ENCLOSURE 3 Tennessee Valley Authority Watts Bar Nuclear Plant, Unit 2 Docket No. 50-391 NON-PROPRIETARY ATTACHMENT TO WESTINGHOUSE LETTER TO TVA WBT-D-4396, DATED JULY 23, 2013, "THERMAL CONDUCTIVITY DEGRADATION (TCD) EFFECT ON THE LARGE-BREAK LOCA ANALYSIS FOR WATTS BAR NUCLEAR PLANT UNIT 2" E3-1

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 1 of 26 NP-Attachment to WB.T-D-4396 Thermal Conductivity Degradation Effect on the Large-Break LOCA Analysis for Watts Bar Unit 2

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 2 of 26 Thermal Conductivity Degradation Effect on the Large-Break LOCA Analysis for Watts Bar Unit 2 1.0

Introduction:

This report is supplemental to FSAR Section 15.4.1.1 [1] and serves to communicate the results of an updated LBLOCA analysis performed to address the fuel Thermal Conductivity Degradation Industry Issue [4].

The Nuclear Regulatory Commission (NRC) approved Westinghouse Best-Estimate Loss-of-Coolant Accident (BELOCA) Automated Statistical Treatment of Uncertainty Method (ASTRUM) methodology [2] is based on the PAD 4.0 fuel performance code [3]. PAD 4.0 was licensed without explicitly considering fuel thermal conductivity degradation (TCD) with burnup. Explicit modeling of TCD in the fuel performance code leads directly to increased fuel temperatures (pellet radial average temperature) as well as other fuel performance related effects beyond beginning-of-life. Since PAD provides input to the large-break LOCA analysis, this will tend to increase the stored energy at the beginning of the simulated large-break LOCA event. This in turn leads to an increase in Peak Cladding Temperature (PCT) if there is no provision to credit off-setting effects.

The Watts Bar Unit 2 PCT was previously ([1], Table 15.4-18b) calculated to be 1552°F, the Maximum Local Oxidation (MLO) was calculated to be 1.04% and the Core-Wide Oxidation (CWO) was calculated to be 0.0%

with the currently licensed Westinghouse BELOCA methodology [2]. Thus, the analysis demonstrated margin to the analysis limits of:

PCT 2200°F MLO 17%

CWO 1%

Fuel performance data that accounts for fuel TCD was used as input to the updated Watts Bar Unit 2 BELOCA analysis discussed herein. The new PAD fuel performance data was generated with an updated PAD model that includes explicit modeling of TCD. Therefore, the BELOCA analysis was updated to consider the fuel TCD effects cited in NRC. Information Notice 2011-21 [4].

2.0 Input Parameters, Assumptions and Acceptance Criteria:

No updates to design inputs and plant operating ranges to gain large-break LOCA margin were utilized in this updated analysis to show compliance with the 10 CFR 50.46 acceptance criteria while maintaining a margin of safety to the prescribed limits. The acceptance criteria and results of the updated BELOCA analysis considering TCD effects are discussed in Section 4.0. The base input assumptions are provided in Tables 15.4-14,15,1.6,19,.23 and Figure 15.4-56 of the FSAR [1], with the limited exception of core peaking factors as noted below.

In order to mitigate the impact of the increasing effect of pellet TCD with bumup, the large-break LOCA evaluation of 2d Cycle fuel utilized reduced peaking factors from those shown in FSAR Table 15.4-19 [1 ]. The reduced peaking factors are limited to the following application:

Burmdown credit for the hot rod and hot assembly is taken for higher burnup fuel in the 2nd and 3d cycle of operation. The Watts Bar Unit 2 peaking factor values utilized in this updated analysis are shown in Table 1.

Note that the beginning to middle of life values are retained at their original [ 1] values.

©201', Westin,2house Electric. Company tIC All Rights Rescnved

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 3 of 26 3.0 Description of Analysis and Evaluations:

The purpose of this update to the analysis is to consider fuel performance inputs that explicitly model TCD to show compliance with the 10 CFR 50.46 acceptance criteria while maintaining a margin of safety to the prescribed limits. The updated BELOCA analysis considering the effects of TCD is supplemental to information provided in FSAR Section 15.4.1.1 [1].

]ac The updated analysis also credited peaking factor burndown to evaluate higher burnup fuel in its second/third cycle of irradiation. Evaluation of fuel in its second/third cycle of irradiation is beyond the first cycle considered in the approved ASTRUM Evaluation Model (EM), but was considered in the updated analysis when explicitly modeling TCD to demonstrate that conformance to the acceptance criteria is met for the second/third cycle fuel.

Physically, accounting for TCD leads to an increase in fuel temperature as the fuel is burned, while accounting for peaking factor bumdown leads to a reduction in fuel temperature as the fuel is burned. The compensating nature of these phenomena is considered in the updated analysis in order to appropriately capture the effect of TCD in the updated Watts Bar Unit 2 BELOCA analysis.

The analysis was updated by re-running all 124 cases from the original ASTRUM analysis (FSAR Section 15.4.1.1 [1]) in both the first cycle and the second cycle. Therefore the same non-parametric order statistics singular statement of a 9 5 th percentile at the 95-percent confidence joint probability for PCT, MLO and CWO of an ASTRUM re-analysis is ensured for the Watts Bar Unit 2 updated analysis. In addition, the uncertainty parameters seed treatment is

]a.c The confirmatory configuration and. the conservatively low containment backpressure from the original ASTRUM run set (FSAR Section 15.4.1.1 [ 1 ]) were also re-evaluated considering the effects of TCD. The 'imiting plant configuration was determined to remain the same. Details of the FSAR analysis confirmatory suite limiting plant configuration study were not previously presented in the FSAR Section 15.4. 1. 1, but are included herein as Table 5 to support this conclusion. First, it is reinforced that no miscellaneous plant configuration changes have been introduced, meaning the confirmation study evaluation only needs to focus to physical TCD effects.

[

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 4 of 26 TLC The conservatively low containment backpressure from the original analysis remains bounding since the core stored energy increases when explicitly modeling fuel TCD, which would tend to increase energy released through the break and hence increase the containment pressure.

IFBA fuel

]c

4.0 Results

The Watts Bar Unit 2 PCT-limiting transient is a double-ended cold leg guillotine break when considering fuel TCD and the peaking factor burndown provided in Table 1. Table 2 summarizes the results of the updated BELOCA analysis considering the effects of TCD. Table 3 provides the sequence of events for the PCT-limitmig transient from the updated analysis. Further IFBA and non-IFBA PCT results for both cycles are given in Table 4.

Figures 1-15 provide a. variety of transient responses for the limiting HOTSPOT PCT transient (2 d Cycle, run073, non-IFBA) and are generally self explanatory in nature. Figures 2-15 are from the associated WCOBRAITRAC response. Figure 1 multiplots the following:

HOTSPOT Clad Temperature at the limiting PCT elevation (observe the peak at 1766 0F)

WCOBRA/TRAC PCT. This PCT represents the highest clad temperature at any elevation.

These Figures are equivalent to FSAR Figures 15.4-41 through 15.4-55 [1].

FSAR Figures 15.4-40b,c,de,fg [I I remain equally appropriate for the TCD analysis because the containment backpressure study did not need repeated as discussed in Section 3.0.

The updated TCD analysis compares to the prior Analysis as follows:

Case TCD 1st Cycle Runset, Limiting run 100 TCD 2nd Cycle Runset, Limiting runG73 PCT (-F) 1580 (non-IFBA) 1766 (non-IFBA)

©'20 13 Westinghouse Electric Company LLC All Riabts Resenved

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 5 of 26 TCD Limiting IFBA Run, 2'd Cycle run073 1763 (IFBA)

FSAR [I] Limiting runO18 1552 (non-IFBA) 10 CFR 50.46 Requirements It must be demonstrated that there is a high level of probability that the limits set forth in 10 CFR 50.46 are met.

The demonstration that these limits are met is as follows:

(b)(1) The limiting PCT corresponds to a bounding estimate of the 95h percentile PCT at the 95-percent confidence level. Since the resulting PCT for the limiting case is 1766F,. the updated analysis confirms that 10 CFR 50.46 acceptance criterion (b)(l), i.e., "Peak Clad Temperature less than 2200'F," is demonstrated. The result is shown in Table 2.

(b)(2) The maximum local cladding oxidation corresponds to a bounding estimate of the 95th percentile MLO at the 95-percent confidence level. Since the resulting transient MLO for the limiting, case is 1.99%, the updated analysis confirms that 10 CFR 50.46 acceptance criterion (b)(2), i.e., "Maximum Local Oxidation of the cladding less than 17 percent," is demonstrated. The result is shown in Table 2.

(b)(3) The limiting core-wide oxidation corresponds to a bounding estimate of the 95' percentile CWO at the 95-percent confidence level. The limiting Hot Assembly Rod (HAR) total power census includes many lower power assemblies. The CWO value can be conservatively chosen as that calculated for the limiting HAR, 0.08%. A detailed CWO calculation is not needed due to the margin in the conservatively obtained result. Therefore, the updated analysis confirms that 10 CFR 50.46 acceptance criterion (b)(3), i.e., "Core-Wide Oxidation less than 1 percent," is demonstrated. The result is shown in Table 2.

(b)(4) 10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains amenable to cooling. This criterion has historically been satisfied by adherence to criteria (b)(1) and (b)(2), and by assuring that fuel deformation due to combined LOCA and seismic loads is specifically addressed. It has been demonstrated that the PCT and maximum cladding oxidation limits remain in effect for Best-Estimate LOCA applications. The approved methodology [5] specifies that effects of LOCA and seismic loads on core geometry do not need to be considered. unless grid crushing extends beyond the 44 assemblies on the core periphery. This situation is not calculated to occur for Watts Bar Unit 2 per FSAR 15.4.1.1.6 [1 ] prior to TCD considerations, and this conclusion is not affected by the modeling of fuel TCD. Therefore, acceptance criterion (b)(4) remains satisfied.

(b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that long-term core cooling be provided, following the successful initial operation of the ECCS. Long-term cooling is dependent on the demonstration of continued delivery of cooling water to the core. The actions, automatic or manual, that are currently in place at these plants to maintain long-term cooling remain unchanged due to the modeling of fuel TCD, as follows:

The primary impact of TCD in the fuel potentially important to Long Term Cooling (LTC) is an increase in initial fuel pellet temperature. This in turn leads to a higher amount of stored energy at the initiation of the LOCA event.

Initial stored energy is not important to LTC evaluations as these evaluations only consider decay heat removal during the sump recirculation phase of emergency core cooling system (ECCS) operation. The increased stored energy in the fuel due to higher fuel pellet temperature is a short term effect that does not persist into the LTC phase of ECCS performance evaluations; therefore, the heat source remains limited to decay heat for LTC evaluations. Consequential impacts of higher fuel pellet temperature such as higher fuel rod internal pressure also have no impact on LTC evaluations as fuel cladding temperatures are maintained well below the threshold for cladding rupture such that cladding burst and blockage does not occur during LTC. Based on the above, it is shown that no additional LTC analysis is required to assess TCD for Watts Bar Unit 2.

Westinghouse Non-Proprietary Class. 3 NP-Attachment to WBT-D-4396 Page 6 of 26 Results Summary An update to the analysis was performed considering fuel performance inputs that explicitly model TCD and the inherently associated peaking factors burndown credit for 2nd cycle analysis only (Table 1.), including [

]a.c IFBA fuel product to show compliance with the current 10 CFR 50.46 acceptance criteria while maintaining a margin of safety to the prescribed limits. For both the I S cycle study and. the 2 d cycle study,, the ASTRUM EM [2] required 124 runs were executed using the same random seed. In this way, the integrity of the TCD analysis is maintained and the difference between the new 95/95 estimate herein and the previous pre-TCD estimate FSAR [ I ] Section 15.4. 1.1 is the singular effect of TCD. Based on the results from the updated BELOCA analysis (see Table 2), it is concluded that Watts Bar Unit 2 continues to maintain a margin of safety to the limits prescribed by 10 CFR 50.46.

[ Additional Information:

Additional Information on treatment of burnup and fuel in its second and third cycle of irradiation The NRC-approved Westinghouse ASTRUM Methodology [2] assumes a LOCA to be [

]Jc Please refer to Section 11-2-2 of the ASTRUM Topical [21 for more information. Although a small amount of peaking factor burndown credit might have been available for some of the higher burnup cases of the I"S cycle runset (Table I), no credit was taken.

]'-. Since the analyzed burnup range covered up to the its first, second or third cycles of operation are covered.

I" design limit, fuel in Is' Cycle Study 2nd Cycle Study Hot Rod Bumup (MWD/MTU)

Minimum Maximum

((

a Hot Rod Burnup (MWD/MTU)

[

Overall Limiting PCT Case 073 (2nd Cycle)

Case 073 Is' Cycle 7- 013 Wesinhous Electric Conpany LLC Alt Riehts Reserved

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBTD-4396 Page 7 of 26

References:

1. WBT-D-1373. WEC to TVA ASTRUM FSAR Package.

(Recommend TVA convert this to the counterpart submittal to the NRC.)

2. Westinghouse Report WCAP-16009-P-A, "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment Of Uncertainty Method (ASTRUM)," January 2005. (Westinghouse Proprietary Class 2)

3. Westinghouse Report WCAP-15063-P-A, Revision I with Errata, "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," July 2000. (Westinghouse Proprietary Class 2)

4.

NRC Information Notice 2011-21, "Realistic Emergency Core Cooling System Evaluation Model Effects Resulting from Nuclear Fuel Thermal Conductivity Degradation," December 13, 2011. (NRC ADAMS Accession Number ML113430785)

5. Westinghouse Report WCAP-12945-P-A, "Code Qualification Document for Best Estimate LOCA Analysis," March 1998. (Westinghouse Proprietary Class 2)

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 8 of 26 Table 1 Watts Bar Unit 2 Best-Estimate Large-Break LOCA Updated Analysis Considering the Effects of Thermal Conductivity Degradation - Summary of Peaking Factor Burndow*Nn Utilized Hot Rod Burnup FdH (1)

FQ Transient FQ Steady-state (MWDiMTU)

(with uncertainties)

(writh uncertainties)

(without uncertainties) 0 1.65 (2) 2.50'(2) 2.00(2) 30000 1.65 (2) 2.50 (2).

2.00(2) 60000 1.525 2.25 1.800 62000 1.525 2.25 1.800 Note I The standard BELOCA assumption of [

jac Note 2: Same Value as FSAR Table 15.4-19 (Note FQ SS is titled 'SS depletion& therein)

Table 2 Watts Bar Unit 2 Best-Estimate Large-Break LOCA Updated Analysis Considering the Effects of Thermal Conductivity Degradation - Comparison of Results to Current 10 CFR 50.46(b) Acceptance Criteria Result Acceptance Criterion 95/95 PCT' 1766

< 2200°F 95/95 Transient MLO2 1.99

< 17%

95/95 CWO3 0.08

< 1%

Coolable Geometry Criterion Met Remains Coolable (See Section.4.0 (b)(4) herein)

Long-Term Cooling See Long Term Cooling TCD assessment in Section 4.0 (b)(5}

herein Notes:

1. Peak Cladding Temperature

2.

Maximum Local Oxidation, transient

3.

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 9 of 26 Table 3 Watts Bar Unit 2 Best-Estimate Large-Break LOCA Updated Analysis Considering the Effects of Thermal Conductivity Degradation - Sequence of Events for the Limiting PCT Transient Event Time After Break (sec)

Start of Transient 0

Safety Injection Signal 5

Accumulator Injection Begins 10 End of Blowdown 1I Bottom of Core Recovery 36 Accumulator Empty' 43 Safety InJection Begins 60 PCT Occurs

190, End of Analysis Time 400 Notes:

1.

Accumulator switches from liquid to injection

2.

Limiting Case is 073. Non-IFBA. of 2nd Cycle, study

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 10 of 26 Table 4 Watts Bar Unit 2 Best-Estimate Large-Break LOCA Updated Analysis Considering the.

Effects of Thermal Conductivity Degradation - Detailed Results MLO Non-IFBA IFBA Rank Ru#

Non-IFBA Hot Rod Run IFBA Hot Rod n MLO (%)

Burnup MLLO (%)

Burnup (MVWD/MITU)

(MWDJMTU) l' Cycle 1

073 1.83 14785 073 1.76 14785 2 nd Cycle 1

104 1.45 52673 104 1.99 52673 2

73 1.37 46909 93 1.48 61773 3

18 1.06 45156 73 1.34 46909 4

100 0.88 46759 18 1.08 45156 5

93 0.81 61773 118 0.97 37365 PCT on-IFBA Non-IFBA IFBA FBA Rank Run #

HOTSPOT Hot Rod Run #

HOTSPOT Hot Rod

("F)

Burnup PCT (-F)

Burnup PCT___ _

(MNWD/MTU)F)

(NIWD(ITU)

I" 1 Cycle 1

100 1580 14940 100 1579 14940 2d Cycle 1

73 1766 46909 73 1763 46909 2

100 1646 46759 104 1646 52673 3

104 1630 52673 100 1645 46759 4

42 1609 57001 118 1637 37365 5

118 1596 37365 42 1607 57001

' 2013 Westinghouse Electric Company LLC All Riphts Reserved

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 11 of 26 Table 5 Watts Bar Unit 2 Best-Estimate Large-Break LOCA Updated Analysis Considering the Effects of Thermal Conductisity Degradation - Original FSAR [11 WCOBRAJIRAC Confirmatory Studies Results Confirmatory Study Parameters WC/T Results I

F I

I Hot Rod PCT Hot Rod PCT Time Confirmatory Run #

SGTP (%)

PLOW LOOP?

(0 F)

(sec after break) 002 LOOP 1229 High (0.8) 004 No-LOOP 1200 High (10) 000H6h LOOP 1298 Low (0.2) 008 No-LOOP 1210 010 LOOP 1221 High (0.8) 012 No-LOOP 1196 Low (0) 014 LOOP 1209 Low (0.2) 016 No-LOOP 1183 3.1 2.9 265 248 3.0 2.9 2.9 2.7 Note:

Watts Bar does not have an RCS T,, window.

1.

Denotes WCOBRAITRAC limiting PCTtReference Transient

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 12 of 26 Wacts Bar Rtr0 7 3

..R~r 073 Unit 2' HOTSPOT PCT WOýT PCT WVBT) ASTRUM Ar lysls Ci-)-

E Tr Time After Break,,s (o '

4W, 6525~ 177 F ure 1 Watts Bar Unit 2 Limiting PCT Case (2 Cycle, RunO73, non-IFBA) HOTSPOT Clad Temperature at the Limiting Elevation and WC/T PCT

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 13 of 26 Wac:s Bcr Uni 2 AS#RKM VESSEL S IDE BREAK FLOW PUMP S I DE BREAK F L0 W TCD BELOCA Ar-avsI's CD f-n 0

~0C 400 Figure 2 Watts Bar Unit 2 Limiting PCT Case Break Flow

@ 2013 Westinghousc Electrw Company LLC All Rights Reservcd

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 14 of 26 Wcvtts Bcr Unit 2 ASTRUM PRESSURIZER PRESSURE TCD BELOCA !An c y siss cn c3) 0 100 1r0 Time After Breck (s) 200 250 29535,59S Figure 3 Watts Bar Unit 2 Limiting PCT Case Pressurizer Pressure

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 15 of 26 Wc'-s Bar Unit 2 AS -Rum TCD BELOCA Arclys-s LOOP 2

fINTACT LOOP)

PUMP VOID FRACTION LOOP 4

%8ROKEN LOOPS PUMP VOID FRACT ION

=0 1

CD 0

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 16 of 26 Waurs Bar JnI7 2 ASTRUV VAPOR FLOW RATE AT TOP 0

TC D CORE B E LOCA r.a ys',

AVERAGE CHANNEL 17 1000 800-

~r)

C)

LL 0

cC)

C) 600 -

400 -

200-0-

I I I I I I I I

I I

i I

I I

r I

I I

I

[

I I

LU*

c I n 28 30 Time After Break (sl 50G 295357599 Figure 5 Watts Bar Unit 2 Limiting PCT Case Core Vapor Flow at the Top of the Core for a Core Average Channel Q 2013 Westinghouse Electie (onvany LLC All Righits Reserved

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 17 of 26 Watts Bar,r'lt 2-ASDTRbM VAPOR FLOW RATE AT TOP OF T COE COR PE BE L 0CA HOT ASSEMB Anaaysis LY CHANNEL 15 q-CD)

CD 4

0 2

0 10 20 30 Time After Break 's) 40 29535759S Figure 6 Watts Bar Unit 2 Limiting PCT Case Core Vapor Flow at the Top of the Core for the Hot Assembly Channel

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 18 of 26 A/Ci s Bcr Unit 2 ASTPRUH TCD BELOA AnlvsiA LOWER PLENUM COLLAPSED LIUID LEVEL 6-C5l r-4-

0 100 2O0 300 Time After Break (s' 400 295357599 Figure 7 Watts Bar Unit 2 Limiting PCT Case Lower Plenum Collapsed Liquid Level C 2013 Weslinhouse Ekctric Company LLC All Rightsi Resen-ed

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 19 of 26 Wayts Bar UnLt 2 ASTRUM TCD BE -0C'A, Arcalysis INTACT LOOP 2

ACCJMULATOR MASS FLOW RATE P) rim 0:3 100

50 Time After Breok (s) 250 295357599 Figure 8 Watts Bar Unit 2 Limiting PCT Case Intact Loop Accumulator Flow

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 20 of 26 Watts Bar UnTt 2 ASTRUM TCD BELOCA AraclysIs INTACT LOOP 2 SI MASS FLOW RATE 1200 1000-8001 0~

Cn ZU) 600-400 200 I

I I

I i

i i

i G

10C 1

200 Time After Break I

4W0 295,35759A3 Figure 9 Watts Bar Unit 2 Limiting PCT Case Intact Loop Safety Injection Flow

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 21 of 26 Wortls Bcr Unit 2 ASTRUM TCD BELOCA Arclysis COLLAPSED L

QUID LEVEL IN CORE AVERAGE CHANNEL 7.5 I

C)

-C)

C)

-C)

C)

C)-

C)

C~)

100 200 Time After Break (sn; 40/0 Figure 10 Watts Bar Unit 2 Limiting PCT Case Core Average Channel Collapsed Liquid Level

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 22 of 26 Wcats Bar Un 2 ASTRUM TCD BELOCA A

'naKsis COLLAPSED LI OLI D LEVEL IN I NT 4CT LOOP 2

DOWNCOMER 35 30 25 G)

-J U) 0 15 0-100 200 3G&

Time After Break ýs) 44M0 29535759 Figure 11 Watts Bar Unit 2 Limiting PCT Case Loop 2 Downcomer Collapsed Liquid Level Q 2013 Weslmgbouse Eketnic Company LLC All Rights Rcscrvd

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 23 of 26 Warts Bar Unl t 2 ASTRUM VESSEL L I QU I D MASS TCD BELOCA ArGcys~s

~J-)

1 o00 200 300 Time After Break (s) 400 2935M759S Figure 12 Watts Bar Unit 2 Limiting PCT Case Vessel Fluid Mass C 2013 Wcstinghouc Eltami Compuny IIC All Rights Resen-ed

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 24 of 26 Watts Bar Unit PCT LOCAT 12 10 8-2 ASTRUM ON TOD BELOCA Anravsls LIZ 6

4 0

0 IGO 200 Time After Break (s) 40A Figure 13 Watts Bar Unit 2 Limiting PCT Case PCT Location

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 25 of 26 Bar Unit 2 ASTRULM TCD BELOCA Analysis LIOUID TEMPERATURE AT BOTTOM OF DOWNCOMER CHANNEL 2

SATURATION TEMPERATJRE AT BOTTOM OF DOWNCOMER CHANNEL 2

w 0)

E 0)

H-0 Time After Break (s) 400 29-5-57599 Figure 14 Watts Bar Unit 2 Limiting PCT Case Liquid and Saturation Temperature at Bottom of Downcomer Channel

Westinghouse Non-Proprietary Class 3 NP-Attachment to WBT-D-4396 Page 26 of 26 Watts Bar-Unit 2 ASTRUv TCD BELOCA Ac I a ys S HOT ROD PEAK CLADD ING TEMPERATURE HOT ASSEMBLY PEAK CLADDING TEMPERATURE GUIDE TUBES PEAK CLADDING TEMPERATURE SUPPORT COLJMNS PEAK CLADDING TEMPERATURE LCN-POWER REGION PEAK CLADDING TEMPERATURE W

C)

U 2OC lime After Break (s) 400 29535759~4 Figure 15 Watts Bar Unit 2 Limiting PCT Case Peak Cladding Temperature for all 5 Rods

ENCLOSURE 4 Tennessee Valley Authority Watts Bar Nuclear Plant, Unit 2 Docket No. 50-391 WBN UNIT 2 FSAR CHAPTER 15 MARK-UP E4-1

WATTS BAR TABLE OF CONTENTS Section 15.3.1.1 15.3.1.2 15.3.1.3 15.3.1.4 15.3.2 15.3.2.1 15.3.2.2 15.3.2.3 15.3.3 15.3.3.1 15.3.3.2 15.3.3.3 15.3.4 15.3.4.1 15.3.4.2 15.3.4.3 15.3.5 15.3.5.1 15.3.5.2 15.3.6 15.3.6.1 15.3.6.2 15.3.6.3 Title IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION ANALYSIS OF EFFECTS AND CONSEQUENCES REACTOR COOLANT SYSTEM PIPE BREAK RESULTS CONCLUSIONS - THERMAL ANALYSIS MINOR SECONDARY SYSTEM PIPE BREAKS IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION ANALYSIS OF EFFECTS AND CONSEQUENCES CONCLUSIONS INADVERTENT LOADING OF A FUEL ASSEMBLY INTO AN IMPROPER POSITION IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION ANALYSIS OF EFFECTS AND CONSEQUENCES CONCLUSIONS COMPLETE LOSS OF FORCED REACTOR COOLANT FLOW IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION ANALYSIS OF EFFECTS AND CONSEQUENCES CONCLUSIONS WASTE GAS DECAY TANK RUPTURE IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION ANALYSIS OF EFFECTS AND CONSEQUENCES SINGLE ROD CLUSTER CONTROL ASSEMBLY WITHDRAWAL AT FULL POWER IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION ANALYSIS OF EFFECTS AND CONSEQUENCES CONCLUSIONS 153e 15.3-1 15.3-2 15.3-3 15.3-5 15.3-5 15.3-5 15.3-5 15.3-5 15.3-6 15.3-6 15.3-6 15.3-7 15.3-7 15.3-7 15.3-9 15.3-10 15.3-10 15.3-10 15.3-10 15-3-10 15.3-10 15.3-11 15.3-12 Scope 15.4 for 15.4.1 markup 15.4.1.

15.4. 1.1 15.4.2 15.4.2.

15.4.23 15.4.3 CONDITION IV - LIMITING FAULTS MAJOR REACTOR COOLANT SYSTEM PIPE RUPTURES (LOSS OF COOLANT ACCIDENT)

I THERMAL ANALYSIS 2

HYDROGEN PRODUCTION AND ACCUMULATION MAJOR SECONDARY SYSTEM PIPE RUPTURE 1

MAJOR RUPTURE OF A MAIN STEAM LINE 2

MAJOR RUPTURE OF A MAIN FEEDWATER PIPE STEAM GENERATOR TUBE RUPTURE 15.4-1 15.4-1 15.4-2 15.4-12 15.4-12 15.4-12 15.4-19 15.4-23 15-iv Table of Contents

WATTS BAR LIST OF TABLES Section TABLE 15.4-12 TABLE 15.4-13 TABLE 15.4-14 TABLE 15.4-15 Scope for markup TABLE 15.4-16 TABLE 15.4-17 TABLE 15.4-18A Title PARAMETERS USED IN THE ANALYSIS OF THE ROD CLUSTER CONTROL ASSEMBLY EJECTION ACCIDENT PARAMETERS RECOMMENDED FOR DETERMINING RADIOACTIVITY RELEASES FOR ROD EJECTION ACCIDENT LARGE-BREAK LOCA CONTAINMENT DATA (ICE CONDENSER CONTAINMENT) USED FOR CALCULATION OF CONTAINMENT PRESSURE FOR WATTS BAR UNIT 2 LARGE-BREAK CONTAINMENT DATA - HEAT SINKS DATA (ICE CONDENSER CONTAINMENT)

MASS AND ENERGY RELEASE RATES USED FOR CALCULATION OF CONTAINMENT PRESSURE FOR WATTS BAR UNIT 2 WATTS BAR UNIT 2 BEST-ESTIMATE LARGE-BREAK LOCA SEQUENCE OF EVENTS FOR LIMITING PCT TRANSIENT PEAK CLAD TEMPERATURE INCLUDING ALL PENALTIES AND BENEFITS, BEST-ESTMATE LARGE-BREAK LOCA (BE LBLOCA) FOR WATTS BAR UNIT 2 WATTS BAR UNIT 2 BEST-ESTIMATE LARGE-BREAK LOCA RESULTS PLANT OPERATING RANGE ANALYZED BY THE BEST-ESTIMATE LARGE-BREAK LOCA ANALYSIS FOR WATTS BAR UNIT 2 y1\\

TABLE 15.4-18B TABLE 15.4-19 TABLE 15.4-20 OPERATOR ACTION TIMES FOR DESIGN-BASIS STEAM GENERATOR TUBE RUPTURE ANALYSIS TABLE 15.4-21 STEAM GENERATOR TUBE RUPTURE ANALYSIS SEQUENCE OF EVENTS TABLE 15.4-22 STEAM GENERATOR TUBE RUPTURE ANALYSIS MASS RELEASE RESULTS TOTAL MASS FLOW (POUNDS)

Scope for TABLE 15.4-23 markup E ABLE 15.5-1 TABLE 15.5-2 TABLE 15.5-3 MINIMUM INJECTED SAFETY INJECTION FLOW USED IN BEST-ESTIMATE LARGE-BREAK LOCA ANALYSIS FOR WATTS BAR UNIT 2 PARAMETERS USED IN LOSS OF A. C. POWER ANALYSES DOSES FROM LOSS OF A/C POWER PARAMETERS USED IN WASTE GAS DECAY TANK RUPTURE ANALYSES WASTE GAS DECAY TANK INVENTORY (ONE UNIT)

(REGULATORY GUIDE 1.24 ANALYSIS)

DOSES FROM GAS DECAY TANK RUPTURE TABLE 15.5-4 TABLE 15.5-5 15-viii List of Tables

WATTS BAR LIST OF FIGURES Section FIGURE 15.4-27 FIGURE 15.4-28 FIGURE 15.4-29 FIGURE 15.4-30 FIGURE 15.4-31 FIGURE 15.4-32 FIGURE 15.4-33 FIGURE 15.4-34 FIGURE 15.4-35 FIGURE 15.4-36 FIGURE 15.4-37 FIGURE 15.4-38 FIGURE 15.4-39 FIGURE 15.4-40A FIGURE 15.4-40B FIGURE 15.4-40C FIGURE 15.4-40D FIGURE 15.4-40E FIGURE 15.4-40F FIGURE 15.4-40G FIGURE 15.4-40H FIGURE 15.4-401 FIGURE 15.4-40J FIGURE 15.4-40K Title HOT SPOT FUEL AND CLAD TEMPERATURE VERSUS TIME; EOL HZP ROD EJECTION ACCIDENT REACTOR COOLANT SYSTEM INTEGRATED BREAK FLOW FOLLOWING A ROD EJECTION ACCIDENT DELETED BY AMENDMENT 97 DELETED BY AMENDMENT97 DELETED BY AMENDMENT9 DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 WATTS BAR UNIT 2 LOWER BOUND CONTAINMENT PRESSURE WATTS BAR UNIT 2 CONTAINMENT TEMPERATURES WATTS BAR UNIT 2 LOWER COMPARTMENT STRUCTURAL HEAT REMOVAL RATE WATTS BAR UNIT 2 ICE BED HEAT REMOVAL RATE WATTS BAR UNIT2 SUMP HEAT REMOVAL RATE WATTS BAR UNIT2 SUMP HEAT REMOVAL RATE CONTAINMENT LOWER COMPARTMENT PRESSURE, MAXIMUM SAFEGUARDS, UPFLOW BARREL/BAFFLE REGION COMPARTMENT TEMPERATURES, MAXIMUM SAFEGUARDS, UPFLOW BARREL/BAFFLE REGION LOWER COMPARTMENT STRUCTURAL HEAT REMOVAL RATE, MAXIMUM SAFEGUARDS, UPFLOW BARREL/BAFFLE REGION ICE BED HEAT REMOVAL RATE, MAXIMUM SAFEGUARDS, UPFLOW BARREL/BAFFLE REGION Scope for markup T1 15-xx List of Figures

WATTS BAR LIST OF FIGURES Section FIGURE 15.4-40L FIGURE 15.4-40M 4I Scope for markup FIGURE 15.4-41A FI*G6U.

15.4 1 11B FIGURE 15.4-42 FIGURE 15.4-43 FIGURE 15.4-44 FIGURE 15.4-45 FIGURE 15.4-46 FIGURE 15.4-47 FIGURE 15.4-48 FIGURE 15.4-49 FIGURE 15.4-50 FIGURE 15.4-51 FIGURE 15.4-52 FIGURE 15.4-53 FIGURE 15.4-54 FIGURE 15.4-55 FIGURE 15.4-56 Title HEAT REMOVAL BY SUMP, MAXIMUM SAFEGUARDS, UPFLOW BARREL/BAFFLE REGION HEAT REMOVAL BY SPRAY, MAXIMUM SAFEGUARDS, UPFLOW BARREL/BAFFLE REGION WATTS BAR UNIT 2 LIMITING PCT CASE HOTSPOT PCT AT THE LIMITING ELEVATION <

land WCfT PC WIATTS BAR UNI 2 LIMITING PCT CASE WCT PCT WATTS BAR UNIT 2 LIMITING PCT CASE BREAK FLOW WATTS BAR UNIT 2 LIMITIG PCT CASE PRESSf4ZER PRESS*

WATTS BAR UNIT 2 LIMITING PCT CASE BROKEN AND INTACT LOOP VOID FRACTION WATTS BAR UNIT 2 LIMITING PCT CASE VAPOR FLOW AT TOP OF CORE AVERAGE CHANNEL WATTS BAR UNIT 2 LIMITING PCT CASE VAPOR FLOW AT TOP OF HOT ASSEMBLY CHANNEL WATTS BAR UNIT 2 LIMITING PCT CASE LOWER PLENUM COLLAPSED LIQUID LEVEL WATTS BAR UNIT 2 LIMITING PCT CASE LOOP 2 ACCUMULATOR FLOW WATTS BAR UNIT 2 LIMITING PCT CASE LOOP 2 SAFETY INJECTION FLOW WATTS BAR UNIT 2 LIMITING PCT CASE CORE AVERAGE CHANNEL COLLAPSED LIQUID LEVEL WATTS BAR UNIT 2 LIMITING PCT CASE LOOP 2 DOWNCOMER COLLAPSED LIQUID LEVEL WATTS BAR UNIT 2 LIMITING PCT VESSEL FLUID MASS WATTS BAR UNIT 2 LIMITING PCT CASE PCT LOCATION WATTS BAR UNIT 2 LIMITING PCT CASE LIQUID AND SATURATION TEMPERATURE AT BOTTOM OF DOWNCOMER WATTS BAR UNIT 2 LIMITING PCT CASE PCT FOR ALL RODS WATTS BAR UNIT 2 BELOCA ANALYSIS AXIAL POWER SHAPE OPERATING SPACE ENVELOPE DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 DELETED BY AMENDMENT 97 TV FIGURE 15.4-57 FIGURE 15.4-58 FIGURE 15.4-59 FIGURE 15.4-60 List of Figures 15-xxi

WATTS BAR INo changes on this page WBNP-109 15.4 CONDITION IV - LIMITING FAULTS Condition IV occurrences are faults which are not expected to take place, but are postulated because their consequences would include the potential for the release of significant amounts of radioactive material. They are the most. drastic which must be designed against and represent limiting design cases. Condition IV faults are not to cause a fission product release to the environment resulting in an undue risk to public health and safety in excess of guideline values of 10 CFR Part 100. A single Condition IV fault is not to cause a consequential loss of required functions of systems needed to cope with the fault including those of the emergency core cooling system (ECCS) and the containment. For the purposes of this report the following faults have been classified in this category:

(1)

Major rupture of pipes containing reactor coolant up to and including double ended rupture of the largest pipe in the reactor coolant system (loss of coolant accident).

(2)

Major secondary system pipe ruptures.

(3) Steam generator tube rupture.

(4)

Single reactor coolant pump locked rotor-(5) Fuel handling accident.

(6)

Rupture of a control rod drive mechanism housing (rod cluster control assembly ejection).

The analysis of thyroid and whole body doses, resulting from events leading to fission product release, appears in Section 15.5. The fission product inventories which form a basis for these calculations are presented in Chapter 11 and Section 15.1. Section 15.5 also includes the discussion of systems interdependency contributing to limiting fission product leakages from the containment following a Condition IV occurrence.

15.4.1 Major Reactor Coolant System Pipe Ruptures (Loss of Coolant Accident)

Loss-of-coolant accidents (LOCAs) are accidents that would result from the loss of reactor coolant at a rate in excess of the capability of the reactor coolant makeup system. LOCAs could occur from breaks in pipes in the reactor coolant pressure boundary up to and including a break equivalent in size to the double-ended rupture of the largest pipe in the reactor coolant system (RCS). Large breaks are defined as breaks in the reactor coolant pressure boundary having a cross-sectional area greater than or equal to 1.0 ft2. Reference [34] documents this criterion. The large break LOCA analysis is performed to demonstrate compliance with the 10 CFR 50.46 acceptance criteria[351 for emergency core cooling systems for light water nuclear power reactors.

A large break LOCA is the postulated double-ended guillotine or split rupture of one of the RCS primary coolant pipes.

CONDITION IV - LIMITING FA UL TS 15.4-1

INo chanoes on this oaae I

IIIBJillll*lll*lPl@,

I'*

Jl MA I I.> IAMt WBNP-109 The boundary considered for loss of coolant accidents is the RCS or any line connected to the system up to the first closed valve.

The sequence of events following a limiting large break LOCA transient is presented in Table 15.4-17. Before the break occurs, the RCS is assumed to be operating normally at full power in an equilibrium condition, i.e., the heat generated in the core is being removed via the secondary system. A large beak is assumed to open almost instantaneously in one of the main RCS pipes. Calculations have demonstrated that the most severe transient results occur for a break in the cold leg between the pump and the reactor vessel.

Immediately following the cold leg break, a rapid system depressurization occurs along with a core flow reversal due to a high discharge of subcooled fluid into the broken cold leg and out the break. The fuel rods go through departure from nucleate boiling (DNB) and the cladding rapidly heats up, while the core power shuts down due to voiding in the core. The hot water in the core, upper plenum, and upper head flashes to steam, and subsequently the cooler water in; the lower plenum and downcomer begins to flash.

Once the system has depressurized to the accumulator pressure, the accumulators begin to inject cold borated water into the intact cold legs. During the blowdown period a portion of the injected ECOS water is calculated to be bypassed around the downcomer and out the break. The bypass period ends as the system pressure continues to decrease and approaches the containment pressure, resulting in reduced break flow and consequently reduced core flow.

As the refill period begins, the core begins a period of heatup and the vessel begins to fill with ECCS water. This phase continues until the lower plenum is filled and the bottom of the core begins to reflood and entrainment begins.

During the reflood period, the core flow is oscillatory as ECCS water periodically rewets and quenches the hot fuel cladding which generates steam and causes system repressurization. The steam and entrained water must pass through the vessel upper plenum, the hot legs, the steam generators, and the reactor coolant pumps before it is vented out the break. This flow path resistance is overcome by the downcomer water elevation head which provides the gravity driven reflood force. The pumped ECCS water aids in the filling of the downcomer and subsequently supplies water to maintain a full downcomer and complete the reflood period.

15.4.1.1 Thermal Analysis 15.4.1.1.1 Westinghouse Performance Criteria for Emergency Core Cooling System The reactor is designed to withstand thermal effects caused by a loss of coolant accident including the double ended severance of the largest reactor coolant system pipe. The reactor core and internals together are designed so that the reactor can be safely shutdown and the essential heat transfer geometry of the core preserved following the accident. The current internals is of the upflow barrel/baffle design. The 15.4-2 CONDITION IV-LIMITING FAULTS

WATTS BAR INo changes on this page I

WBNP-109 ECCS, even when operating during the injection mode with the most limiting single active failure, is designed to meet the acceptance criteria.

15.4.1.1.2 Method of Thermal Analysis When the Final Acceptance Criteria (FAC) goveming the loss-of-coolant accident (LOCA) for Light Water Reactors was issued in Appendix K of 10 CFR 50.46, both the Nuclear Regulatory Commission (NRC) and the industry recognized that the stipulations of Appendix K were highly conservative. That is, using the then accepted analysis methods, the performance of the Emergency Core Cooling System (ECCS) would be conservatively underestimated, resulting in predicted Peak Clad Temperatures (PCTs) much higher than expected. At that time, however, the degree of conservatism in the analysis could not be quantified. As a result, the NRC began a large-scale confirmatory research program with the following objectives:

1) Identify, through separate effects and integral effects experiments, the degree of conservatism in those models permitted in the Appendix K rule. In this fashion, those areas in which a purposely prescriptive approach was used in the Appendix K rule could be quantified with additional data so that a less prescriptive future approach might be allowed.

2)

Develop improved thermal-hydraulic computer codes and models so that more accurate and realistic accident analysis calculations could be performed. The purpose of this research was to develop an accurate predictive capability so that the uncertainties in the ECCS performance and the degree of conservatism with respect to the Appendix K limits could be quantified.

Since that time, the NRC and the nuclear industry have sponsored reactor safety research programs directed at meeting the. above two objectives. The overall results have quantified the conservatism in the Appendix K rule for LOCA analyses and confirmed that some relaxation of the rule can be made without loss in safety to the public. It was confirmed that some plants were being restricted in operating flexibility by the overly conservative Appendix K requirements. In recognition of the Appendix K conservatism that was being quantified by the research programs, the NRC adopted an interim approach for evaluation methods. This interim approach is described in SECY-83-472 [50]. The SECY-83-472 [501 represented an important step in basing licensing decisions on realistic calculations, as opposed to those calculations prescribed by Appendix K.

In 1998, the NRC Staff amended the requirements of 10 CFR 50.46 and Appendix K, "ECCS Evaluation Models", to permit the use of a realistic evaluation model to analyze the performance of the ECCS during a hypothetical LOCA. This decision was based on an improved understanding of LOCA thermal-hydraulic phenomena gained by extensive research programs. Under the amended rules, best-estimate thermal-hydraulic models may be used in place of models with Appendix K features. The rule change also requires, as part of the LOCA analysis, an assessment of the uncertainty of the best estimate calculations. It further requires that this analysis uncertainty be CONDITION IV - LIMITING FA UL TS 15.4-3

WATTS BAR WBNP-109 included when comparing the results of the calculations to the prescribed acceptance criteria of 10 CFR 50.46. Further guidance for the use of best-estimate codes is provided in Regulatory Guide 1.157[44].

To demonstrate use of the revised ECCS rule, the NRC and its consultants developed a method called the Code Scaling, Applicability, and Uncertainty (CSAU) evaluation methodology (NUREG/CR-5249[45]). This method outlined an approach for defining and qualifying a best-estimate thermal-hydraulic code and quantifying the uncertainties in a LOCA analysis. A LOCA evaluation methodology for three-and four-loop Pressurized Water Reactor (PWR) plants based on the revised10 CFR 5046 rules was developed by Westinghouse with support of EPRI and Consolidated Edison and has been approved by the NRC (WCAP-12945-P-A [46]).

More recently, Westinghouse developed an alternative methodology called ASTRUM, which stands for Automated Statistical TReament of Uncertainty Method (WCAP-16009-P-A [49]). This method is still based on the CQD methodology and follows the steps in the CSAU methodology (NUREG/CR-5249 [451). However, the uncertainty analysis (Element 3 in the CSAU) is replaced by a technique based on order statistics.

The ASTRUM methodology replaces the response surface technique with a statistical sampling method where the uncertainty parameters are simultaneously sampled for each case. The ASTRUM methodology has received NRC approval for referencing in licensing calculations in WCAP-16009-P-A [49].

The three 10 CFR 50.46 criteria (peak clad temperature, maximum local oxidation, and core-wide oxidation) are satisfied by running a sufficient number of WCOBRAJTRAC calculations (sample size). In particular, the statistical theory predicts that 124 calculations are required to simultaneously bound the 95th percentile values of three parameters with a 95-percent confidence level.

This analysis is in accordance with the applicability limits and usage conditions defined in Section 13-3 of WCAP-1 6009-P-A [491, as applicable to the ASTRUM methodology.

Section 13-3 of WCAP-16009-P-A [491 was found to acceptably disposition each of the identified conditions and limitations related to WCOBRAITRAC and CQD uncertainty approach per section 4.0 of the ASTRUM Final Safety Evaluation Report appended to IInsertA this topical report.

The methods used in the application of WCOBRA /TRAC to the large break LOCA with ASTRUM are described in WCAP-12945-P-A [461 and WCAP-16009-P-A [49]. A detailed assessment of the computer code WCOBRAfTRAC was made through comparisons to experimental data. These assessments were used to develop quantitative estimates of the code's ability to predict key physical phenomena in a Ilnsert B PWR large break LOCA. Modeling of a PWR introduces additional uncertainties which are identified and quantified in the plant-specific analysis. WCOD,,AJ'TRAC,MOD7A WCOBRAITRAC combines two-fluid, three-field, multi-dimensional fluid equations used in the vessel with one-dimensional drift-flux equations used in the loops to allow 15.4-4 CONDITION IV-LIMITING FAULTS

INo channes on this nane I

Illll*A A

l lit WA I I i tRAHL WBNP-109 a complete and detailed simulation of a PWR. This best-estimate computer code contains the following features:

1) Ability to model transient three-dimensional flows in different geometries inside the vessel

2) Ability to model thermal and mechanical non-equilibrium between phases

3) Ability to mechanistically represent interfacial heat, mass, and momentum transfer in different flow regimes

4) Ability to represent important reactor components such as fuel rods, steam generators, reactor coolant pumps, etc.

The two-fluid formulation uses a separate set of conservation equations and constitutive relations for each phase. The effects of one phase on another are accounted for by interfacial friction and heat and mass transfer interaction terms in the equations. The conservation equations have the same form for each phase; only the constitutive relations and physical properties differ. Dividing the liquid phase into two fields is a convenient and physically accurate way of handling flows where the liquid can appear in both film and droplet form. The droplet field permits more accurate modeling of thermal-hydraulic phenomena such as entrainment, de-entrainment, fallback, liquid pooling, and flooding.

WCOBRAITRAC also features a two-phase, one-dimensional hydrodynamic formulation. In this model, the effect of a phase slip is modeled indirectly via a constitutive relationship which provides the phase relative velocity as a function of fluid conditions. Separate mass and energy conservation equations exist for the two-phase mixture and for the vapor.

The reactor vessel is modeled with the three-dimensional, three-field model, while the loop, major loop components, and safety injection points are modeled with the one-dimensional model.

All geometries modeled using the three-dimensional model are represented as a matrix of cells. The number of mesh cells used depends on the degree of detail required to resolve the flow field, the phenomena being modeled, and practical restrictions such as computing costs and core storage limitations-The equations for the flow field in the three-dimensional model are solved using a staggered difference scheme on the Eulerian mesh. The velocities are obtained at mesh cell faces, and the state variables (e.g., pressure, density, enthalpy, and phasic volume fractions) are obtained at the cell center. This cell is the control volume for the scalar continuity and energy equations. The momentum equations are solved on a staggered mesh with the momentum cell centered on the scalar cell face.

The basic building block for the mesh is the channel, a vertical stack of single mesh cells. Several channels can be connected together by gaps to model a region of the reactor vessel. Regions that occupy the same level form a section of the vessel.

CONDITION IV - LIMITING FAULTS 15.4-5

WATTS BAR INo changes on this page J

WBNP-109 Vessel sections are connected axially to complete the vessel mesh by specifying channel connections between sections. Heat transfer surfaces and solid structures that interact significantly with the fluid can be modeled with rods and unheated conductors.

One-dimensional components are connected to the vessel. The basic scheme used also employs the staggered mesh cell. Special purpose components exist to model specific components such as the steam generator and pump.

A typical calculation using, WCOBRA/TRAC begins with the establishment of a steady-state initial condition with all loops intact. The input parameters and initial conditions for this steady-state calculation are discussed in the next section.

Following the establishment of an acceptable steady-state condition, the transient calculation is initiated by introducing a break into one of the loops. The evolution of the transient through blowdown, refill, and reflood proceeds continuously, using the same computer code (WCOBRAITRAC) and the same modeling assumptions. Containment pressure is modeled with the BREAK component using a time dependent pressure table. Containment pressure is calculated using the LOTIC-2 [5] code and mass and energy releases from the WCOBRAITRAC calculation.

The final step of the best-estimate methodology, in which all uncertainties of the LOCA parameters are accounted for to estimate a Peak Cladding Temperature (PCT),

Maximum Local Oxidation (MLO), and Core-Wide Oxidation (CWO) at 95-percent probability, is described in the following sections.

1) Plant Model Development:

In this step, a WCOBRAITRAC model of the plant is developed. A high level of noding detail is used in order to provide an accurate simulation of the transient.

However, specific guidelines are followed to ensure that the model is consistent with models used in the code validation. This results in a high level of consistency among plant models, except for specific areas dictated by hardware. differences, such as in the upper plenum of the reactor vessel or the ECCS injection configuration.

2) Determination of Plant Operating Conditions:

In this step, the expected or desired operating range of the plant to which the analysis applies is established. The parameters considered are based on a "key LOCA parameters" list that was developed as part of the methodology. A set of these parameters, at mostly nominal values, is chosen for input as initial conditions to the plant model.

A transient is run utilizing these parameters and and is known as the "initial transient". Next, several confirmatory runs are made, which vary a subset of the key LOCA parameters over their expected operating range in one-at-a-time sensitivities. Because certain parameters are not included in the uncertainty analysis, these parameters are set at their bounding condition-This analysis is 15.4-6 CONDMONIV-LIMMNG FAULTS

WATTS BAR WBNP-109 WATTS BAR WBNP-109 commonly referred to as the confirmatory analysis-The most limiting input conditions, based on these confirmatory runs, are then combined into the model that will represent the limiting state for the plant, which is the starting point for the assessment of uncertainties.

3) Assessment of Uncertainty:

The ASTRUM methodology is based on order statistics. The technical basis of the order statistics is described in Section 11 of WCAP-1 6009-P-A [49]. The determination of the PCT uncertainty, MLO uncertainty, and CWO uncertainty relies on a statistical sampling technique. According to the statistical theory, 124 WCOBRA /TRAC calculations are necessary to assess against the three 10 CFR 50.46 criteria (PCT,MLO,CWO).

The uncertainty contributors are sampled randomly from their respective distributions for each of the WCOBRA/TRAC calculations. The list of uncertainty parameters, which are randomly sampled for each time in the cycle, break type (split or double-ended guillotine), and break size for the split break are also sampled as uncertainty contributors within the ASTRUM methodology.

Results from the 124 calculations are tallied by ranking the PCT from highest to lowest. A similar procedure is repeated for MLO and CWO. The highest rank of PCT, MLO, and CWO will bound 95 percent of their respective populations with 95-percent confidence level.

4) Plant Operating Range:

The plant operating range over which the uncertainty evaluation applies is defined.

Depending on the results obtained in the above uncertainty evaluation, this range may be the desired range or may be narrower for some parameters to gain additional margin.

15.4.1.1.3 Containment Analysis The containment pressure analysis is performed with the LOTIC-2 [5] code. Transient mass and energy releases for input to the LOTIC-2 model are obtained from the WCOBRA/TRAC code. The transient pressure computed by the LOTIC-2 code is then used in WCOBRAITRAC for the purpose of supplying a backpressure at the break plane while computing the reflood transient. The containment pressure transients and associated parameters were computed by LOTIC-2 and are presented in Figures 15.4-40b through 15.4-40g. The data used to model the containment for the analysis is presented in TablesI5.4-14 and 15.4-15. Mass and energy release rates to containment can be found in, Table 15.4-1,6.

The impact of purging on, the calculated containment pressure was addressed by performing a calculation to obtain the amount of mass which exits through two available purge lines during the initial portion of a postulated LOCA transient. The CONDITION IV - LIMITING FAULTS 15.4-7

WATTS BAR WBNP-109 WATTS BAR WBNP-1 09 maximum air loss was calculated using the transient mass distribution (TMD) computer code model, which is described in Section 6.2.1.3.4, to be 1160 Ibm. The containment pressure calculations account for a loss of 1160 Ibm of air after initiation of the accident through modifying the compression ratio input to the LOTIC-2 code.

15.4.1.1.4 Results of Large Break Limiting Transient The Watts Bar Unit 2 PCT,-,-O,-,, W lmitintransient is a Geld leg split break

{(,fleetire br==,a. ;;;= - !.B 3,which analyzes conditions that fall within those listed in Table 15.4-19. Traditi y, cold leg breaks have been limiting for large break LOCA. Analysis ex ence indicates that this break location most likely causes conditions that rýult in flow stagnation to occur in the core. Scoping studies with WCOB C have confirmed that the cold leg remains the limiting break location (WC 2945-P-A[461).

Insert E Ihe large break LOCA transient can be divided into convenient time periods in which specific phenomena occur, such as various hot assembly heatup and cool down transients. For a typical large break, the blowdown period can be divided into the Critical Heat Flux (CHF) phase, the upward core flow phase, and the downward core flow phase. These are followed by the refill, reflood, and long-term cooling periods.

Specific important transient phenomena and heat transfer regimes are discussed below, with the transient results shown in Figure 15-4-41 through 15.4-55. (The limiting case was chosen to show a conservative representation of the response to a large reak LOCA.)

IFigure 15.4-41

1) Critical Heat Flux (CHF) Phase:

IPCT Immediately following the cold leg rupture, the break discharge rate is subcooled d

high (Figure 15.4-42). The region of the RCS with the highest initial tempera es (core, upper plenum, upper head, and hot legs) begin to flash to steam, t core flow reverses and the fuel rods begin to go through departure from nucle oiling (DNB).

The fuel cladding rapidly heats up (Figures 15.4 44A -nd 15.A 44b) while the core power shuts down due to voiding in the core. This phase is terminated when the water in the lower plenum and downcomer begins to flash (Figures 15.4-47 and 15.4-51).

The mixture swells and intact loop pumps, still rotating in single phase liquid, push this two-phase mixture into the core.

2) Upward Core Flow Phase:

Heat transfer is improved as the two-phase mixture is pushed into the core. This phase may be enhanced if the pumps are not degraded, or if the break discharge rate is low due to saturated fluid conditions at the break. If pump degradation is high or the break flow is large, the cooling effect due to upward flow may not be significant. Figure 15.4-44 shows the void fraction for one intact loop pump and the broken loop pump. This figure shows that the intact loop remains in single-phase liquid flow for several seconds, resulting in enhanced upward core flow cooling. This phase ends as the lower plenum mass is depleted, the loop flow becomes two-phase, and the pump head degrades.

15.4-8 CONDITION IV-LIMITING FAULTS

WATTS BAR WBNP-109

3) Downward Core Flow Phase:

The loop flow is pushed into the vessel by the intact loop pumps and decreases as the pump flow becomes two-phase. The break flow begins to dominate and pulls flow down through the core, up the downcomer to the broken loop cold leg, and out the break. While liquid and entrained liquid flow provide core cooling, the top of the core vapor flow (Figures 15.4-45 and 15.4-46) best illustrates this phase of core cooling.

Once the system has depressurized to the accumulator pressure (Figure 1"5.4-43), the accumulators begin to inject cold borated water into the intact cold legs (Figure 15.4-48). During this period, due to steam upflow in the downcomer, a portion of the injected ECCS water is calculated to be bypassed around the downcomer and out the break.

As the system pressure continues to fall, the break flow, and consequently the downward core flow, is reduced. The core begins to heat up as the system pressure approaches the containment pressure and the vessel begins to fill with ECCS water (Figure 15.4-52).

4) Refill Phase:

As the refill period begins, the core begins a period of heatup and the vessel begins to fill with ECCS water (Figure, 15.4-48 aRd-5449). This period is characterized by a rapid increase in cladding temperatures at all elevations due to the lack of liquid and steam flow in the core region. This period continues until the lower plenum is filled and the bottom of the core begins to reflood and entrainment begins.

5) Early Reflood Phase:

(Figure 15.4-49)

During the early reflood phase, the accumulators begin; to empt and nitrogen enters the system. This forces water into the core, which then boils, using system re-pressurization, and the lower core region begins to quenchM igure 15.4-50). During this time, core cooling may increase due to vapor genera, n and liquid entrainment.

During the reflood period, the core flow is oscillatory as; d water periodically rewets and quenches the hot fuel cladding, which generates team and causes system re-pressurization. The steam and entrained water mu pass through the vessel upper plenum, the hot legs, the steam generators, and t reactor coolant pumps before it is vented out the break. This flow path resistance is overcome by the downcomer water elevation head, which provides the gravity d 'en reflood force. From the later stage of blowdown to the beginning of reflood, th accumulators rapidly discharge borated cooling water into the RCS, filling the lo plenum and contributing to the filling of the downcomer. The pumped ECCS water aids in the filling of the downcomer and subsequently supplies water to maintain a full downcomer and complete the reflood period. As the quench front progresses up the core, the PCT location moves higher into the top core region. As the vessel continues to fill, the PCT location is cooled and the early reflood period is terminated.

6) Late Reflood Phase:

The late reflood phase is characterized by boiling in the downcomer. The mixing of ECCS water with hot water aA4 and steam from the core, in addition to the continued heat transfer from the hot vessel metal, reduces the subcooling of water in the lower CONDITION IV-LIMMNG FAULTS 15.4-9

WATTS BAR WBNP-109 I(Figure 15.4-41, HOTSPOT result).

plenum and downcomer. Figure 15.4-54 illustrates the reduction i lower plenum subcooling.

/

The saturation temperature is dictated by the containment back ressure. For WBN, which has a low containment pressure after the LOCA, boiling oes occur and has a significant effect on the gravity reflood. Vapor generated in th downcomer reduces the driving head which results in a reduced core reflood rateE.

he top core elevations experence a second reflood heatup, which exceeds the first.

15.4.1.1.5 POST ANALYSIS OF RECORD EVALUATIONS "In addition to the analyses presented in this section, evaluations and reanalyses may be performed as needed to address computer code errors and emergent issues, or to support plant changes. The issues or changes are evaluated, and the impact on the Peak Cladding Temperature (PCT) is determined. The resultant increase or decrease in PCT is applied to the analysis of record PCT. The PCT, including all penalties and

-'benefits is presented in Table\\15.4-18a for the large break LOCA. The current PCT is demonstrated to be less than the 10 CFR 50.46(b) requirement of 2200 OF.

In addition, 10 CFR 50.46 requires that licensees assess and report the effect of changes to or errors in the evaluation model used in the large break LOCA analysis.

These reports constitute addenda to the analysis of record provided in the FSAR until overall changes become significant as defined by '10 CFR 50.46. If the assessed changes or errors in the evaluation model results in significant changes in calculated PCT, a schedule for formal reanalysis or other action: as needed to show compliance will be addressed in the report to the NRC.

Finally, the criteria of 10 CFR 50.46 requires that holders and users of the evaluation models establish a number of definitions and processes for assessing changes in the models or their use. Westinghouse, in consultation with the PWR Owner's Group (PWROG), has developed an approach for compliance with the reporting requirements. This approach is documented in WCAP-13451 [361, Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting. TVA provides the NRC with annual and 30-day reports, as applicable, for Watts Bar Unit 2. TVA intends to provide future reports required by 10 CFR 50,46 consistent with the approach described in WCAP-13451.

15.4.1.

1.6 CONCLUSION

S - THERMAL ANALYSIS It must be demonstrated that there is a high level of probability that the limits set forth in 10 CFR 50.46 are met. The demonstration that these limits are met is as follows:

(b)(1) The limiting PCT corresponds to a bounding estimate of the 95th percentile at the 95-percent confidence level. Figure 15.4-41ýshows the predicted

-and the WCOBRATRAC HOTSPOT cladding temperature transient nt thp PrT Focatia,* for the limiting PCTT IPCT transient, both POT' case. The HOTSPOT POT plot includes local uncertainties applied to the Hot Rod. Figure 15.4 4.!b present." the WCOP.R.AJTRA.C PCT tr3nsient pr*edicted for te_ impff. PCT dass.

This.-

-'oes not account for any local

, whereas the 15.4-10 IWCOBRA/TRAC coNDiON IV-LIMITING FAULTS P-CT plot

WATTS BAR WBNP-109 uncertainties. Since the resulting HOTSPOT PCT for the limiting case is 11766L p 44.1F, the analysis confirms that 10 CFR 50.46 acceptance criterion (b)(1),

i.e., "Peak Clad Temperature less than 22000F, is demonstrated. The results are shown in Table 15.4-18b.

(b)(2) The maximum cladding oxidation corresponds to a bounding estimate of the 95th percentile MLO at the 95-percent confidence level. Since the resulting MLO for the limiting cas

.*4-percent, the analysis confirms that 10 CFR 9

ance criterion (b)(2), i.e., "Maximum Local Oxidation of the cladding less than 17 percent", is demonstrated. The results are shown in Table t5.4-18b.

(b)(3) The limiting core-wide oxidation corresponds to a bounding estimate of the 95th percentile CWO at the 95-percent confidence level. The limiting Hot Assembly Rod (HAR) total maximum

+,^n

" percent. A detailed 0.08

_ ion es advantage of the core power census that includes many lower power assemblies. Because there is significant margin to the regulatory limit, the CWO value can be conservatively chosen as that calculated for the limiting HAR. A detailed CWO calculation is therefore not needed because the outcome will always be less than the HAR value. Since the resultin C percent, the analysis confirms that 10 CFR 50.46 n0-u Lacceptance criterion (b)(3), i.e., "Core Wide Oxidation less than 1 percent", is demonstrated.

(b)(4) 10 CFR 50.46 acceptance criterion (b)(4) requires that the calculated changes in core geometry are such that the core remains amenable to cooling. This criterion has historically been satisfied by adherence to criteria (b)(1) and (b)(2), and by assuring that the fuel deformation due to combined LOCA and seismic loads is specifically addressed. It has been demonstrated that the PCT and maximum cladding oxidation limits remain in effect for Best-Estimate LOCA applications. The approved methodology (WCAP-12945-P-A

[46]) specifies that effects of LOCA and seismic loads on core geometry do not need to be considered unless grid crushing extends beyond the 44 assemblies in the low-power channel. This situation has not been calculated to occur for Watts Bar Unit 2. Therefore, acceptance criterion (b)(4) is satisfied.

(b)(5) 10 CFR 50.46 acceptance criterion (b)(5) requires that the long-term core cooling be provided following the successful initial operation of the ECCS.

Long-term cooling is dependent on the demonstration of continued delivery of cooling water to the core. While WCOBRAJTRAC is typically not run past full core quench, all base calculations are run well past PCT tumaround and past the point where increasing vessel inventories are calculated. The CONDITION IV-LIMITING FAULTS 15.4-11

WATTS BAR INo changes on this page WBNP-109 conditions at the end of the WCOBRA/TRAC calculations indicate that the transition to long term cooling is underway even before the entire core is quenched.

Based on the ASTRUM Analysis results (Table 15.4-18b), it is concluded that Watts Bar Unit 2 maintains a margin of safety to the limits prescribed by 10 CFR 50.46.

15.4.1.1.7 PLANT OPERATING RANGE Scope for markup I

The expected PCT and its uncertainty developed are valid for a range of plant operating conditions. The range of variation of the operating parameters has been accounted for in the uncertainty evaluation. Tables 15.4-19 summarizes the operating ranges as defined for the proposed operating conditions which are supported by the Best-Estimate LBLOCA analysis for Watts Bar Unit 2. Tables 15.4-14 and 15.4-15 summarize the LBLOCA containment data used for calculating containment pressure.

If operation is maintained within these ranges, the LBLOCA results developed in this report using WCOBRAITRAC are considered to be valid. Note that some of these parameters vary over their range during normal operation (accumulator temperature) and other ranges are fixed for a given operational condition (Tavg).

15.4.1.2 Hydrogen Production and Accumulation Pursuant to NRC final rule as defined in 10 CFR 50.44 and Regulatory Guide 1.7, the new definition of design-basis LOCA hydrogen release eliminates requirements for hydrogen control systems for mitigation of releases. "All PWRs with ice condenser type containments must have the capability to control combustible gas generated from metal-water reaction involving 75% of the fuel cladding surrounding the active fuel region (excluding the cladding surrounding the plenum volume) so that there is no loss of containment structural integrity. The deliberate ignition systems provided to meet this existing combustible gas source term are capable of safely accommodating even greater amounts of combustible gas associated with even more severe core melt sequences that fail the reactor vessel and involve molten core-concrete interaction.

Deliberate ignition systems, if available, generally consume the combustible gas before it reaches concentrations that can be detrimental to containment integrity." On the basis of this definition, no further analysis is required to support events considered to be outside the design basis. Deliberate ignition systems are described in FSAR Section 6.2.5 15.4.2 Major Secondary System Pipe Rupture 15.4.2.1 Major Rupture of a Main Steam Line 15.4.2.1.1 Identification of Causes and Accident Description The steam release arising from a rupture of a main steam line would result in an initial increase in steam flow which decreases during the accident as the steam pressure falls. The energy removal from the reactor coolant system causes a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in a reduction of core shutdown margin.

15.4-12 CONDITION IV-LIMITING FAULTS

INo c'hannes on this nlil I

IlfAI*A R

l IPl WA! I., lbA".

WBNP-109 released to the water relatively slowly, and it is considered inconceivable that cross flow will be sufficient to produce significant lattice forces. Even if massive and rapid boiling, sufficient to distort the lattice, is hypothetically postulated, the large void fraction in the hot spot region would produce a reduction in this ratio at the hot spot.

The net effect would therefore be a negative feedback. It can be concluded that no conceivable mechanism exists for a net positive feedback resulting from lattice deformation. In fact, a small negative feedback may result. The effect is conservatively ignored in the analysis.

15.4.6.3 Conclusions Even on a worst-case basis, the analyses indicate that the described fuel and clad limits are not exceeded. It is concluded thatthere is no danger of sudden fuel dispersal into the coolant. Since the peak pressure does not exceed that which would cause stresses to exceed the faulted condition stress limits, it is concluded that there is no danger of further, consequential damage to the reactor coolant system. The reference

[16] analyses have demonstrated that the number of fuel rods entering DNB amounts to less than 10%, thus satisfactorily limiting fission product release.

The environmental consequences of this accident is bounded by the loss of coolant accident. See Section 15.5.3, "Environmental Consequences of a Loss of Coolant Accident." The reactor coolant system integrated break flow to containment following a rod ejection accident is shown in Figure 15.4-28.

Following reactor trip, requirements for operator action and protection system operation are similar to those presented in the analysis of a small loss of coolant event, section 15.3.1.

REFERENCES (1)

Deleted by Amendment 97 (2)

Deleted by Amendment 97 Scope (3)

Deleted by Amendment 97 for(

markup (4)

Deleted by Amendment 97 (5)

Hsieh, T., and Raymund, M., "Long Term Ice Condenser Transient Analysis (LOTIC II)," WCAP-8355 Supplement 1, May 1975 and WCAP-8354 (Proprietary), July 1974.

(6)

Deleted by Amendment 97 (7)

Deleted by Amendment 63.

(8)

Deleted by Amendment 80.

(9)

Moody, F. S., "Transactions of the ASME, Journal of Heat Transfer," Figure 3, Page 134, February 1965.

15.4-44 CONDITION IV-LIMITING FAULTS

WATTS BAR INo changes on this page I

WBNP-109 (10) Deleted Amendment 80.

(11) Burnett, T. W. T., et. al., "LOFTRAN Code Description," WCAP-7907-P-A (proprietary) and WCAP-7907-A (non-proprietary), April 1984.

(12) Hunin, C., "FACTRAN, A FORTRAN IV Code for Thermal Transients in a UO2 Fuel Rod," WCAP-7908, July 1972.

(13) Liimataninen, R. C. and Testa, F. J., "Studies in TREAT of Zircaloy-2-Clad, UO2-Core Simulated Fuel Elements," ANL-7225, January - June 1966, p.

177, November 1966.

(14) Burnett, T. W. T, "Reactor Protection System Diversity in Westinghouse Pressurized Water Reactors," WCAP-7306, April 1969.

(15) Taxelius, T. G., "Annual Report - Spert Project, October 1968, September 1968," Idaho Nuclear Corporation IN-1370, June 1970.

(16) Risher, D. H., Jr., "An Evaluation of the Rod Ejection Accident in Westinghouse Pressurized Water Reactors Using Spatial Kinetics Methods,"

WCAP-7588, Revision 1-A, January 1975.

(17) Barry, R. F., and Risher, D. H., Jr., "TWINKLE - A MuLti-Dimensional Neutron Kinetics Computer Code," WCAP-7979-P-A, January 1975 (Proprietary) and WCAP-8208-A, January 1975 (Non-Proprietary).

(18) Deleted by Amendment 80.

(19) Bishop, A. A., et al., "Forced Convection Heat Transfer at High Pressure After the Critical Heat Flux," ASME 65-HT-31, August 1965.

(20) Deleted by Amendment 97.

(21) Deleted by Amendment 97.

(22) Deleted by Amendment 97.

(23) Deleted by Amendment 97.

(24) Deleted by Amendment 97.

(25) Deleted by Amendment 97.

(26) Deleted by Amendment 97.

(27) Deleted by Amendment 97.

(28) Deleted by Amendment 80.

(29) Deleted by Amendment 80.

CONDITION IV-LIMITING FAULTS 15.4-45

WATTS BAR WBNP-109 WATTS BAR WBN P-i 09 (30) C. W. Stewart, et al., "VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores, "Volumes 1-3 (Revision 3, August 1989), Volume 4 (April 1987), NP-2511-CCM-A, EPRI.

(31) Deleted by Amendment 80.

(32) Deleted by Amendment 97.

(33) "American National Standard for Decay Heat Power in Light Water Reactors,"

ANSI/ANS-5.1-1979, August 1979.

(34) Rupprecht, S. D, et. al., "Westinghouse Small Break LOCA ECCS Evaluation Model Generic Study with the NOTRUMP Code," WCAP-1 1145-P-A (Proprietary), WCAP-1 1372 (Non-Proprietary), October 1986.

(35) U.S. Nuclear Regulatory Commission, Code Federal Regulations - Energy 10, Chapter 1, Part 50, Section 50.46b< "Acceptance Criteria for Emergency Core Cooling Systems for Light Water Nuclear Power Reactors."

(36) "Westinghouse Methodology for Implementation of 10 CFR 50.46 Reporting",

WCAP-13451 October 1992.

(37) Devault, R. M., Smith, J. D., and Studer, P. G., "MONSTER - A Multi-Compartment Containment System Analysis Program User Manual," System I.D. 262303, March 1993.

(38) Deleted by Amendment 97 (39) Letter from Walsh, L. A., Westinghouse Owners Group, to Jones, R. C., U.S.

Nuclear Regulatory Commission, "Steam Generator Tube Uncovery Issue,"

OG-92-25., March 1992.

(40) "Report on the Methodology for the Resolution of the Steam Generator Tube Uncovery Issue," WCAP-13247 (Proprietary), March 1992.

(41) Letter from Jones, R. C., U.S. Nuclear Regulatory Commission, to Walsh, L.

A., Westinghouse Owners Group, "Steam Generator Tube Uncovery Issue,"

March 10, 1993.

(42) Watts Bar "Design Basis Events Design Criteria", Document WB-DC-40-64.

(43) Criticality Analysis Summary Report for Watts Bar Nuclear Plant," Document Number PFE-R07, Tennessee Valley Authority Nuclear Fuels Department (L38 961015 802).

(44) USNRC Regulatory Guide 1.157, "Best-Estimate Calculations of Emergency Core Cooling System Performances", May 1989.

15.4-46 CONDITION IV-LIMITING FAULTS

WATTS BAR WBNP-1O9 (45) Boyack, B., et al, 1989, "Qualifying Reactor Safety Margins: Application of Code Scaling Applicability and Uncertainty (CSAU) Evaluation Methodology to a Large Break Loss-of-Coolant-Accident", NUREG/CR-5249 (46) "Code Qualification Document for Best Estimate Loss of Coolant Accident Analysis," WCAP-12945-P-A, Volume 1 (Revision 2) and Volumes 2 through 5 (Revision 1), March 1998 (Westinghouse Proprietary).

(47) "Best Estimate Analysis of the Large Break Loss of Coolant Accident for the Watts Bar Nuclear Plant," WCAP-14839-P Revision 1, June 1998.

(48) Letter from W. J. Johnson of Westinghouse to R. C. Jones of the NRC, "Use of 2700 °F PCT Acceptance Limit in Non-LOCA Accidents, "NS-NRC 3466, October 1989.

(49) "Realistic Large-Break LOCA Evaluation Methodology Using the Automated Statistical Treatment of Uncertainty Method (ASTRU M)", WCAP-16009-P-A, January 2005 (Westinghouse Proprietary)

(50) "Emergency Core Cooling System Analysis Methods", SECY-83-472, Information Report from W. J. Dircks to the Commissioners, November 17.1983.

Insert G

CONDITION IV - LIMITING FAULTS 15.4-47

INo changes on this page WATTS BAR WBNP-109 Table 15.4-14 Large-Break LOCA Containment Data (Ice Condenser Containment) Used for Calculation of Containment Pressure for Watts Bar Unit 2 Parameter Value Net Free Volume Distribution Between Upper (UC), Lower (LC), Ice UC: 710,000 ft3 Condenser (IC) and Dead-Ended (DE) Compartments LC: 253,114 ft3

[C: 122,350 ft3 DE: 129,900 ft3 Initial Condition Containment Pressure 14.7 psia Maximum Temperature for the Upper (UC), Lower (LC and Dead-UC: 110°F Ended (DE) Compartments LC: 120'F DE: 120'F Minimum RWST Temperature (Containment Spray Temperature) 60°F Minimum Temperature Outside Containment 5YF Maximum Containment Spray Flow Rate 4000 gpmlpump Number of Spray Pumps Operating 2

Post-Accident Initiation of Spray System 25 sec Post-Accident Delay Time for Deck Fan Actuation 490 sec Deck Fan Flow Rate 41,690 cfm/fan Initial Ice Mass 2,450,000 Ibm 15.4-64 CONDITION IV-LIMITING FAULTS

INo changes on this page I

WATTS BAR WBNP-109 Table 15.4-15 Large-Break Containment Data - Heat Sinks Data (ice Condenser Containment)

Wall Compartment(1)

Area [ft2]

Thickness [ft]

Material 1

UC 5124.

1.6 concrete 2

UC 19992.

0.000525/1.6 coating/concrete 3

UC 4032.

0.02167/1.6 stainless steel/concrete 4

UC 11192.

0.00065/0.03908 coating/carbon steel 5

UC 47800.

0.00065/0.09252/1.0 coating/carbon steel/concrete 6

UC 273.

0.00065/0.1308 coating/carbon steel 7

LC 59000.

2.1 concrete 8

LC 17178.

0.000133/2.1 coating/concrete 9

LC 12988.

2.1 concrete 10 LC 2384.

0.02167/2.1 stainless steel/concrete 11 LC 25444.

0.00065/0.1089/1.0 coating/carbon steel/concrete 12 LC 12810.

0.00065/0.07593 coating/carbon steel 13 LC 2625.

0.00055/0.12083 coating/carbon steel 14 LC 1575.

0.00065/0.14167 coating/carbon steel 15 LC 12915.

0.00065/0.044167 coating/carbon steel 16 LC 12988.

2.1 concrete 17 LC 3439 0.1561 carbon steel Notes:

1. UC and LC are Upper and Lower Compartment, respectively.

CONDITION IV - LIMITING FAULTS 15.4-65

WATTS BAR No changes on this page WBNP-109 Table 15.4-16 Mass And Energy Release Rates Used for Calculation of Containment Pressure for Watts Bar Unit 2 (Page 1 of 2) lime After Break (sec)

Mass Flow Rate(Ibmlsec)

Energy Flow Rate (BTU/sec)

0.

9646.7 5369419.

1.

71201.5 39577048.

2.

50782-1 28747484.

3.

40475.1 23410743-

4.

34105A 20560588.

5.

30009.1 18713303.

6.

27906.0 17640053.

7.

26130.9 16632257.

8.

24651.1 15663961.

9.

22805.6 14511306.

10.

20004.8 13053678.

11.

17472.9 11605252.

12.

14601.0 10093803.

12.4 13464.4 9420184.

14.

12172.5 7614137.

15.

12554.4 6455205.

16.

11369.7 5308157.

17.

10902.4 4491501.

18.

10124.7 3756484-

19.

9258A 3127399.

20.

8178.7 2411114.

21.

7321.0 2120146.

22 7603.9 1977749.

23.

5474.9 1402837.

24.

4641.5 999621.

25.

6992.0 1356562.

26.

5955.4 1051498.

28.

4062.4 618361.

29.

3020.7 405978.

30.

1824.1 201868.

32.

1873.8 190499.

15.4-66 CONDITION IV-LIMITING FAULTS

INo changes on this page j

WATTS BAR WBNP-109 Table 15.4-16 Mass And Energy Release Rates Used for Calculation of Containment Pressure for Watts Bar Unit 2 (Page 2 of 2)

Time After Break (sec)

Mass Flow Rate(Ibmlsec)

Energy Flow Rate (BTUIsec)

33.

1882.1 180047.

34.5 1890.2 204412.

35.

1921.9 200973.

39.

2275.7 246561.

41.

1959.7 214441.

43.

2031.8 267974.

45.

2650.2 384113.

46.

7824.1 1100908.

47.5 2842.5 400880.

50.

1811.6 373546.

51.

1764.3 397686.

55.

2254.1 544982.

57.5 1383.9 503576.

60.

1621.8 592463.

65.

790.9 338984.

80.

686.2 251517.

110.

646.9 232801.

150.

643.9 307300.

190.

654.1 229705.

226.

374.2 116811.

300.

404.3 144644.

349.

503.8 176903.

CONDITION IV-LIMITING FAULTS 15.4-67

WATTS BAR WBNP-109 Table 15.4-17 Watts Bar Unit 2 Best-Estimate Large-Break LOCA Sequence Of Events for Limiting PCT Transient Event Time after break (sec)

Start of Transient 0.0 Safety Injection Signal

&,I 15 I Accumulator Injection Begins 42 1

End of Blowdown 24- &

Bottom of Core Recovery 404 36 Accumulator Empty(')

5G F43 Safety Injection Begins 6W 60-o PCT Occurs 2.09" 190I End of analysis time 400.0 Note:

1. Accumulator injection switches from liquid to nitrogen.

15.4-68 CONDITION IV-LIMITING FAULTS

WATTS BAR WBNP-109 Table 15.4-18a Peak Clad Temperature Including All Penalties and Benefits, Best-Estmate Large-Break LOCA (BE LBLOCA) for Watts Bar Unit 2 PCT for Analysis-of-Record (AOR)

-. 5 0F 1F766 PCT Assessments Allocated to AOR None N/A BE LBLOCA PCT for Comparison to 10 CFR 50.46 45°F 1766 Requirements

~AOR Value Table 15.4-18b Watts Bar Unit 2 Best-Estim 'e Large-Break LOCA Results ASTRUM Results va" h-,

Acceptance Criteria 95/95 PCT 455°F 117661

<2200°F 95/95 MLO 1444%

P

<17%

95/95 CWO 0

0.

<1%

CONDITION IV-LIMITING FAULTS 15.4-69

WATTS BAR WBNP-109 Table 15.4-19 Plant Operating Range Analyzed by the Best-Estimate Large-Break LOCA Analysis for Watts Bar Unit 2 (Page 1 of 2)

Parameter As-Analyzed Value or Range 1.0 Plant Physical Description a) Dimensions Nominal b) Pressurizer location!

Modeled on an intact loop c) Hot assembly location Anywhere in core intenor (1) d) Hot assembly type 17x17 RFA-2, ZIRLO Clad with IFMs e) Steam generator tube plugging level

-< 10% Any or All SGs 2.0 Plant Initial Operating Conditions 2.1 Reactor Power a) Core Power 3479.8 MWt +/-0% Uncertainty(2) b) Peak heat flux hot channel factor (FQ) 52.50 See Table 15.4-24 c) Peak hot rod enthalpy rise hot channel

-51.65 ISee Table 15.4-24 factor (FAH) d) Hot assembly radial peaking factor (P)

-<1.65/1.04 See Table 15.4-24 e) Hot assembly heat flux hot channel factor

-<2.50/1.04 See Table 15.4-24 (FQHA) f) Axial power distribution (PBOT,PM[D)

Figure 15.4-56 g) Low power region relative power (PLow) 0.2 -S PLOW < 0.8 h) Hot assembly burnup 7-5,000 MWD/MTU, lead rodK i) MTC 6<

0 at hot full power (HFP) j) Typical cycle length 20,000 MWD/MTU k) Minimum beginning of cycle core average

> 10,000 MWD/MTU burnup I) Maximum steady state depletion, FQ 2.0 See Table 15.4-24 2.2 Fluid Conditions a) TAVG 582.2 0F < TAVG < 594.20F b) Pressurizer pressure 2180 psia < PRCS - 2300 psia c) Loop flow TDF > 93,100 gpm/loop d) Upper head temperature

= TCOLD e) Pressurizer level (at full power) 1067 ft3 f) Accumulator temperature, 100°F :TACc - 120°F g) Accumulator pressure 585 psig S PAC S 690 psig h) Accumulator liquid volume 1005 ft3 < VACc -1095 ft3 15.4-70 CONDMONIV-LIMMNG FAULTS

WATTS BAR WBNP-109 Table 15.4-19 Plant Operating Range Analyzed by the Best-Estimate Large-Break LOCA Analysis for Watts Bar Unit 2 (Page 2 of 2)

Parameter As-Analyzed Value or Range i) Accumulator fL/D 5.6186 +/- 20%

j) Minimum accumulator boron 1900 ppm (4) 3.0 Accident Boundary Conditions a) Minimum safety injection flow Table 15.4-23 b) Safety injection temperature 60'F < St Temp s 105 0F c) Safety injection delay (5) 40 seconds (with offsite power) 55 seconds (with! LOOP) d) Containment modeling Tables 15.4-14, 15.4-15, and 15.4-16 and Figure 15.4-40b e) Single failure 1 RHR, 1 IHS[, and 1 CH/Si; Pump Operable; Containment pressure: all trains operational Notes:

1. 44 peripheral locations will not physically be lead power assembly.

2. The core average linear heat rate is set equal to a value corresponding to 3479.8 MWt (100.6 percent of 3459 MWt), and is not ranged in the uncertainty analysis. This power level approach bounds any future plant operation whose product of nominal full power and calorimetric uncertainty of -5 3479.8 MWt (for example, a nominal full power of 3479.8/1.005) MWt and 0.5% calorimetric uncertainty is bounded).

3a. P12at, note that the fuel tom IrAt-re rid rod internal 9r cI-re 1~tA i2 Gnl' ProVi'dod 6' to0 62,Q00 R mL A I*

Im A*L I

I

FNotUed, r.'.L.:M i

U. in aaailion, tne not astemo t ro940i11 not nav6 a Wurnu6P thns hghn in.4..l OR

4. The accumulator boron concentration used for the uncertainty analysis was 1900 ppm rather than 3000 ppm, which was the value transmitted to Westinghouse by TVA. This bounds the value transmitted by TVA and will have no impact on the results presented herein.

5. Conservatively high SI delay times were used to bound the values transmitted by TVA to Westinghouse.

CONDITION IV-LIMITING FAULTS 15.4-71

WATTS BAR WBNP-109 INo Changes to Table 15.4-23 1

Table 15.4-23 Minimum Injected Safety Injection Flow Used in Best-Estimate Large-Break LOCA Analysis for Watts Bar Unit 2 Pressure

[psia]

14.7 34.7 54.7 74.7 94.7 114.7 134.7 154.7 214.7 314.7 414.7 514.7 614.7 714.7 814.7 914.7 1014.7 1114.7 1214.7 1314.7 1414.7 1514.7 1614.7 1714.7 1814.7 1914.7 2014.7 2114.7 Charging Flow SI Flow RHR Flow

[gpm]

262.2 260.5 258.8 257.0 255.3 253.6 251.9 250.1 244.9 236.1 227.1 218.0 208.4 198.6 188.5 178.2 167.6 156.7 145.4 131.0 115.9 100.1 83.0 63.7 44.3 27.0 5.8 0.0

[gpm]

416.9 413.7 410.4 407.2 403.9 400.7 396.8 393.0 381.4 360.8 339.5 317.6 294.6 269.8 242.7 214.5 185.0 150.2 150.9 50.8 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0

[gpmj 2715.2 2284.6 1811.3 1367.7 1156.9 916.9 633.2 232.2 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 0.0 Total Flow

[gpm]

3394.3 2958.8 2480.5 2031.9 1816.1 1571.2 1281.9 875.3 626.3 596.9 566.6 535.6 503.0 468.4 431.2 392.7 352.6 306.9 251.3 181.8 115.9 100.1 83.0 63.7 44.3 27.0 5.8 0.0 ITable Insert I CONDITION IV-LIMmNG FAULTS 15.4-75

[No changes on this page I

WATTS BAR WBNP-iOS Figure 15.4-40a Deleted by Amendment 97 15.4-126 Conitionw IV-Limiting Faufts

WATTS BAR INo changes on this page

ý WBNP-109 LO 1 9 WA-tS RAR MIUCLEAf PLAlff

"%-SFEW MANAYSgM WEO

%WATM M BR LQT 2 LCFiur 15.440b Womet s

e Figure 15.4-40b Watts Bar Unit 2 Lower Bound Containment Pressure CondItion IV. L-mltirng Faults 15.4-127

INo changes on this page WATTS BAR WHNP-409

C~V.r Cou~p.~t~eut F.

PMO 0

D o

imu WA -18r eARq PAUCLEAR PLAKrT Fg.ure SA154V4 Watts BREs WATTSOK UW LNIf 2Ai Figure 154.440c Watts Bar Unit 2 Containment Temperatures 15.4-428 Cacond IV-UndJitng Faults

INo changes on this page WATTS BAR WBNP-109 I-~

ONNUM Figure 15.4-40d Watts Bar Unit 2 Lower Compartment Structural Heat Removal Rate Condition IV-Limttng Faults M544129

[No changes on this page WATTS BAR WBNP-109

CIs.

I 41 a

so too

-7 WA'7TS BAR D&JXZEAR WLAK

4NAL SAULir AMALYS48 J~O~4 OCE lEID FW-CT PUEOVAJ. PATE Figure 15.4-40e Watts Bar Unit 2 Ice Bed Heat Removal Rate M4.4130 1Corxftk IV-LUaJ~hM Fakudl

[No changes on this page I

WATTS BAR WBNP-109 faymy.

E

~2E!~

II f.

0 Twt Figure 15.4-40f Watts Bar Unit2 Sump Heat Removal Rate Conditon IV - UýMltg Faults M5.,-131

[No changes on this page WATTS BAR WBNP-109 2=4 Figure 15.4-40g Watts Bar Unit2 Sump Heat Removal Rate 15.44132 1Candltlo IV. LhoIlheg Fauts

WATTS BAR WBNP-109 Watts Bar Unit 2 ASTRUM BELOCA Analysis

- Ofspol PC?

£ am -

sm -

4M Update per WBT-D-4396 NP -

Attachment Figure 1 0

RD' 20

£0 Tim After Oreok (s) land WCrT PCT Figure 15.41/Wfatts Bar Unit 2 Limiting PCT Case Notspot PCT AtThe Limiting Elevation HOTSPOT Coditfio, IV-LUrnfing Falts 4-139

WATTS BAR WBNP-109 Watts Bar Unit 2 ASTRUM BELOCA Analysis i

tD PE&K CLAODINC EMI*EIIAUIE E

" -delete this figure WATTS BAR WIMaEAR P'..HW WAllS BAA LB42 UMLTW4 PCT CASE FRM.

f PCTYA5~L~ C1 P*.....

Fk3wrp;b-44m1AIE

-2ta IR~nm u'mIKR13~ M 1 4 15.4.140 CondlIUn IV - LUnitlng Faufts

WATTS BAR WBNP-409 E

V 03 7W"

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Attachment Figure 2 0

fo 150 Thme Af~ter BrWak (s)

Figure 15.4-42 Watts Bar Unit 2 Limiting PCT Case Break Flow Condition IV. Limiting Faults 15A-141

WATTS BAR WSNP4109 Watts Bar Unit 2 ASTRIUM BELOCA Analysis PRESS3I1EN PRESSURE Ii" 25m;,-

Update per WBT-D-4396 NP -

Attachment Figure 3 r

r Figure 15.4-43 Watts Bar Unit 2 Limitig PCT Case Press4-er Pressil 1&.4-142 Condition iV-Liwif/tg Fautts

WATTS BAR W~iT BARWrNID109 Watts Bar Unit 2 ASTRUM BELOCA Analysis 4.00 2

MC

~IT~ ~

O

~PVK FtAMr9" COO,..

UOMLS.

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Attachment Figure 4 T

ID I

-FM H

VeA3Tr BAR POLMLL W,

U11T rs ROFO c$MT 2

ý~

5 W.TA=T LOOP "Po PHA*CW1X Figure 15.4-44 Watts Bar Unit 2 Limiting PCT Case Broken And Intact Loop Void Fraction Cordition IV - Limlting Faults 15,4443

WATTS BAR WBNP-109 Watts Bar Unit 2 ASTR[JM BELOCA Analysis VAPOR ru.C3 RArC A7 TOP OF' CORE AWCRASE CHIANNEL tZ Update per WBT-D-4396 NP -

ttachment Figure 5

m

ATUrU 6A*

M2*J~LEAF3 t AMIF r.M.

GAFIrY AJ4DLVSS FIP4R WATTS BAR UgffX! 2 JIVXT"VdG PiC-T CASE* VAPOR Fk

^T F r 1Lo AOIo OAfore Aerg4 hne Figure 15.4-45 Watts Bar Unit 2 Limiting PCT Case Vapor Flow At Top Of Care Average Channel C-,ditlo Iv-Lb,'jtng F.ofls 15.4-144

WATTS BAR WBNP-109 WAiTS BAR WBNP-1OS Watts Bar Unit 2 ASTRUJM BELOCA Analysis VPRF'-3W R~ATE Ar rO? CF COME hr ASS~CKlLW CH~ARELE 95 Update per WBT-D-4396 NP -

Attachment Figure 6 Figure 15.4-46 Watts Bar Unit 2 Limiting PCT Case Vapor Flow At Top Of Hot Assembly Channel Condition IVP-Lhmitlng Faults 1&.4-145

WATTS BAR WBNP-109 Watts Bar Unit 2 ASTRUM BELOCA Analysis LOWER PLENUM COLLAPSED LIQUI!D LEVEL

-2 Update per WBT-D-4396 NP -

Attachment Figure 7 Tim 200 30W" Timne After Break (s)

Figure 15.4-47 Watts Bar Unit 2 Limiting PCT Case Lower Plenum Collapsed Liquid Level 15.4-146 Condion i' - Limiting Faulft

WATTS BAR WBNP-109 Watts Bar Unit 2 ASTRUM I3ELOCA Analysis DOIACT r.CCP 2 XCCUM.LAVAOM WASS VFLOW V&WE I

x J A 0LL~~

Update per WBT-D-4396 NP-Attachment Figure 8 11%m Ater Break (9)

WX773 GA" OMT 2

AUIGPC ASE SLOOPW Figure 15.4-48 Watts Bar Unit 2 Limiting PCT Case Loop 2 Accumulator Flow Codiilon I-Limtnfg Faufts I&I-IA7

WATTS BAR WBNP-109 Watts Bar Unit 2 ASTRUM BELOCA Analysis

-- *AOT LOOP 2

GS UASS FLOW *ATC i

1

--- -........ +/-

Update per WBT-D-4396 NP -

.Attachmen oo 2E*:

MO IR

.*+Figure 9

rM r

WVATTS EAft ?U.3LRS.AJ PLUir VE..AL 3A!~VM AMML~SIS ----

WATTS BK54AR 8UN 2

LM"1T~r~f lPG CAýS LcýM a SAS=ElY MECTIO IAI I~U Fv15Ms4a ii Figure 15.4-49 Watts Bar Unit 2 Limiting PCT Case Loop 2 Safety Injection Flow 1&.4-148 Condotjo, V.- LkmWI FW

-ts

WATTS BAR WBNP-109 Watts Bar Unit 2 ASTRUM BELOCA Analysis cG'-&AkPGC

&IC60 &CtVtEL ON =912C A*VCMACC C:XA1MEL ZZ 4S.

WATTS 00 ALt tUJ.EMR rLAMY FM.A1 - SAFETY ^*jArYS:S ~~

YEACTS BAR LOICT 2 LMA TIOM PCi CASE =FIEAVM1^

CWý GGILJUPSED LXEXSO LEVEL.

P-ýcrle 7,5,4-60 Figure 15.4-50 Watts Bar Unit 2 Limiting PCT Case Core Average Channet Collapsed Liquid Level Condltfon IV - Lkmifing Faults M4.149

WATTS BAR WBNP-109 Watts Bar Unit 2 ASTRUM BELOCA Analysis COLLAPSED LIQUID LEVEL 111 INTACT LOOP 2 DOWNCOMER C-2 Update per WBT-D-4396 NP -

Attachment Figure 11 19777 Figure 15.4-81 Watts Bar Unit 2 Limiting POT Case Loop 2 Downcomer Collapsed Liquid Level 15.4-150 Conditon IV - LLmJin FPauft.,

WATTS BAR WBNP-109 Watts Bar Unit 2 ASTR-6M BELOCA Analysis VESSCL LIGQUO0 WASS 7

t..

=*

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Update per WBT-D-4396 NP -

Attachment 2

{Figure 12 0

rwre Afterek (s I

WATWS BSAR MMU.EZA PLANTr FWJ.. S FEW ANAL'YSIS ftPCU Figure 15.4-52 Watts Bar Unit 2 Limiting PCT Vessel Fluid Mass Condition IV - LUmting Faults MS4-151

WATTS BAR WBNP-lO9 Watts Bar Unit 2 ASTRUM BFOCA Analysis PCV 90CATOOM

J,7 Update per WBT-D-4396 NP -

Attachment Figure 13 C

tdO 200 rMe, Mter anw;k (S) 4M W.AXTS BAR WSiX.AR PLMAJE IZ AT7S ARLO L

te L4~u -qT Figure 15.4-53 Watts Bar Unit 2 Limiting PCT Case PCT Location 15.4.152 Condition IV-Lmrriting Fautts

WATTS BAR WBNP-109 Watts Bar Unit 2 ASTRUM BELOCA Analysis LIQUID TEMPERATURE AT BOTTOM OF DOWILCOMER CHANNEL 2

SATURATION TEMPERATURE AT BOTTON OF DONC MER CHANNEL 2

V E

PT Update per WBT-D-4396 NP -

Attachment Figure 14 200 300 Time After Break (s)

IMMMtU Figure 15.4-54 Watts Bar Unit 2 Limiting PCT Case Liquid And Saturation Temperature At Bottom Of Downcomer CondIton IV-Untig Faufs 1&*4-153

WATTS BAR WBNP-109 Watts Bar Unit. 2 ASMhUM BELOCA Analysis

~OTt~O~

PIM9 CL~ADDRIP rENEPMUZEI Ro

~E~Y VA CAIM LKLATX INPOT4 IUS6CS Pe,:ý.tCLp*flG EWCWd4U Suppinrý CoLU(LKSH P;4

&OaUlGt TC~EN6r V(UNG LCM-POVZ"MC CSPAK CLADDONG tEMPERATUUE Ei

ýa tc3cr mcý 4w 7

Jn Ar-I'm J-ý 0

Time After Brecak (s')

0 4M m

r WA

-T GG~ A~SIM2'Lg~W I

WAtTGOU0,4M-L O

A M

UNI MT r

FALATOvANWL5I RE&4166 I11 Figure 15.4-55 Watts Bar Unit 2 Limiting PCT Case PCT For All Rods 15.4-154 15.4 Condit IV - Liml'ing Faults

WATTS BAR

[No changes to this page I WBNP-109*

0-55 FCAS 0.2 0.25 0.3 C-35 0.4 MAS 0zE 0-16 0.6 Pkww~r in the ffa~b tfdrd ut tha wos (PUIC)

FOOT 6t agramd power tu:Gn in the bmla Ui aortwie PMEM kittgramd pawuw fracMio n t. tmcmkU~d ofheco IWATTS CIAR M.CL&.Ml PLANT PIAI SA4T L6~~Aft4LMxm WI-OW SP-EMNWOM eE.OAPIT IA4.Y' I

Figure 15.4-56 Watts Bar Unit 2 BELOCA AnaLysis Axial Power Shape Operating Space Envelope Condition tV -Limiting Faults 1&4-155

FSAR INSERTS (2 pages follow) to LTR-LIS-13-375

Insert A (to page 15.4-4)

The Watts Bar 2 ASTRUM LBLOCA uses a plant-specific adaptation of the ASTRUM methodology that includes explicit modeling of fuel thermal conductivity degradation (TCD), as well as a larger sampling range for rod internal pressure (RIP) uncertainty.

Insert B (to page 15.4-4)

WCAP-16009-P-A [49] states that the ASTRUM methodology is based on the frozen code version WCOBRA/TRAC MOD7A, Revision 6. WCOBRA/TRAC MOD7A, Revision 8-T2 was used for the execution of ASTRUM Uncertainty Studies for Watts Bar Unit 2. The confirmatory analysis (paragraph "2)

Determination of Plant Operating Conditions") were executed with WCOBRA/TRAC MOD7A Revision 7.

The Nuclear Regulatory Commission (NRC) approved Westinghouse Best-Estimate Loss-of-Coolant Accident (BELOCA) ASTRUM methodology [49% is based on the PAD 4.0 fuel performance code [511. PAD 4.0 was licensed without explicitly considering fuel thermal conductivity degradation (TCD) with burnup.

Explicit modeling of TCD in the fuel performance code leads directly to increased fuel temperatures (pellet radial average temperature) as well as other fuel performance related effects beyond beginning-of-life. Since PAD provides input to the large-break LOCA analysis, this will tend to increase the stored energy at the beginning of the simulated large-break LOCA event. This in turn leads to an increase in Peak Cladding Temperature (PCT) if there is no provision to credit off-setting effects. In addition, a different fuel thermal conductivity model in WCOBRA/TRAC and HOTSPOT was used to more accurately model the fuel temperature profile when accounting for TCD.

In order to mitigate the impact of the increasing effect of pellet TCD with burnup, the large-break LOCA evaluation of second/third Cycle fuel utilized reduced peaking factors from those shown directly in FSAR Table 15.4-19. The reduced peaking factors are limited to the following application: Bumdown credit for the hot rod and hot assembly is taken for higher burnup fuel in the second/third cycle of operation. The Watts Bar Unit 2 peaking factor values utilized in this analysis are shown in Table 15.4-24. Note that the beginning to middle of life values are retained at their direct Table 15.4-19 values.

It should be noted that evaluation of fuel in its second/third cycle of irradiation is beyond the first cycle considered in the approved ASTRULM Evaluation Model (EMI, but was considered in the analysis when explicitly modeling TCD to demonstrate that conformance to the acceptance criteria is met for the second/third cycle fuel.

In addition to the standard uncertainty calculations, the Watts Bar 2 LBLOCA analysis sampled a larger rod internal pressure (RIP) uncertainty than originally included in the ASTRUM methodology [491. It was discovered that the as-approved sampling range did not bound the plant-specific rod internal pressure uncertainties for Watts Bar 2. Therefore, the approved sampling range was expanded to bound the Watts Bar 2 plant-specific data.

Insert C (to page 15.4-7)

The confirmatory configuration analysis was performed previous to the ASTRUM uncertainty calculations prior to the identification of the TCD issue and associated PAD data. However, as no to LTR-LIS-13-375

miscellaneous plant configuration changes were introduced, and the effects of TCD are minimal for the confirmatory analysis, the limiting plant configuration (Referred to as the Reference Transient) was judged to remain the same.

Insert D (to page 15.4-7)

The Table 15.4-16 mass and energy releases are taken from the 'Reference Transient' case of Section 15.4.1.1.2, which did not include the fuel TCD modeling. The conservatively low containment backpressure from this LOTIC-2 study is bounding since the core stored energy increases when explicitly modeling fuel TCD, which would tend to increase energy released through the break and hence increase the containment pressure.

Insert E (to page 15.4-8)

PCT and MLO/CWO transients are double ended cold leg guillotine breaks with an effective break area of 1.911, and 2.0968 respectively (note that the limiting MLO and CWO arise from the same case),

Insert F (to page 15.4-10)

An evaluation of IFBA fuel including the effects of pellet TCD was performed, and shows that IFBA fuel is limiting for MLO but not for PCT. The AOR PCT and MLO results in Tables 15.4-18a and 15.4-18b reflect the higher results of IFBA/non-IFBA.

Insert G (to page 15.4-47)

(51) "Westinghouse Improved Performance Analysis and Design Model (PAD 4.0)," WCAP-15063-P-A, Revision 1 with Errata (Proprietary), July 2000.

Table Insert l(follows Table 15.4-23)

Add Table 15.4-24 as follows:

Table 15.4-24 Summary of Peaking Factor Burndown Analyzed by the Best-Estimate Large-Break LOCA Analysis for Watts Bar Unit 2 Hot Rod Burnup FdH FQ Transient FQ Steady-state (MWD/MTU)

(with uncertainties)

(with uncertainties>

(without uncertainties) 0 1.65"t 2.50 200 c) 30000 1.65 "

2.50 2.00 ("

60000 1.525 2.25 1.800 62000 1.525 2.25 1.800 Note 1: Same Value as Table 15.4-19 (Note FQ SS is titled 'SS depletion' therein) to LTR-LIS-13-375

ENCLOSURE 5 Tennessee Valley Authority Watts Bar Nuclear Plant, Unit 2 Docket No. 50-391 NEW COMMITMENT The information provided in Enclosure 4 will be incorporated into the WBN Unit 2 FSAR by Amendment 110.

E5-1

ML13225A024

Text

Subject:

References:

Enclosures:

Subject:

Introduction:

4.0 Results

References:

1.6 CONCLUSION

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