CNL-16-114, Responses to Requests for Additional Information and Containment Accident Pressure Credit Elimination Updates on Proposed Technical Specifications Change TS-505 - Request for License Amendments - Extended Power Uprate .

Responses to Requests for Additional Information and Containment Accident Pressure Credit Elimination Updates on Proposed Technical Specifications Change TS-505 - Request for License Amendments - Extended Power Uprate .

ML16216A699
Person / Time
Site:	Browns Ferry
Issue date:	08/03/2016
From:	James Shea Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML16216A698	List: CNL-16-114, Responses to Requests for Additional Information and Containment Accident Pressure Credit Elimination Updates on Proposed Technical Specifications Change TS-505 - Request for License Amendments - Extended Power Uprate . NL-16-114, Browns Ferry, Units 1, 2 & 3, Responses to Requests for Additional Information and Containment Accident Pressure Credit Elimination Updates on Proposed Technical Specifications Change TS-505 - Request for License Amendments - Extended Powe
References
CNL-16-114
Download: ML16216A699 (252)

Text

Proprietary Information Withhold from Public Disclosure Under 10 CFR 2.390 This letter is decontrolled when separated from Enclosures 2 and 7 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-16-114 August 3, 2016 10 CFR 50.90 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296

Subject:

Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 24, Responses to Requests for Additional Information and Containment Accident Pressure Credit Elimination Updates

References:

1. Letter from TVA to NRC, CNL-15-169, "Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU)," dated September 21, 2015 (ML15282A152)

2. Letter from NRC to TVA, Browns Ferry Nuclear Plant, Units 1, 2, and 3 -

Request for Additional Information Related to License Amendment Request Regarding Extended Power Uprate (CAC Nos. MF6741, MF6742, and MF6743), dated June 13, 2016 (ML16146A635)

3. Letter from NRC to TVA, Browns Ferry Nuclear Plant, Units 1, 2, and 3 -

Regulatory Audit Plan for May 3 - 5, 2016, Audit at Excel Facility in Rockville, Maryland, in Support of Extended Power Uprate License Amendment Request (CAC Nos. MF6741, MF6742, and MF6743), dated April 21, 2016 (ML16105A297)

4. Letter from TVA to NRC, CNL-16-118, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 27, Responses to Requests for Additional Information, dated July 29, 2016 (ML16211A393)

U.S. Nuclear Regulatory Commission CNL-16-114 Page 2 August 3, 2016

5. Letter from TVA to NRC, CNL-16-048, Proposed Technical Specifications (TS) Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) - Supplement 7, Responses to Requests for Additional Information, dated March 24, 2016 By the Reference 1 letter, Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for the Extended Power Uprate (EPU) of Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3. The proposed LAR modifies the renewed operating licenses to increase the maximum authorized core thermal power level from the current licensed thermal power of 3458 megawatts to 3952 megawatts. The Reference 2 letter provided Nuclear Regulatory Commission (NRC) Requests for Additional Information (RAIs) related to containment systems. The due date, provided by the Reference 2 letter, for submittal of the responses to NRC RAIs SCVB-RAIs 1 (Revision 1), 8, 14, 24, 25, 26, 30, 32, 34, and Containment Accident Pressure (CAP) Credit Elimination LAR Supplement (Power Uprate Safety Analysis Report and LAR Attachment 39) was July 25, 2016. Due to the time required to complete the necessary analysis revisions and reviews, the due date for this submittal was extended to August 10, 2016, per communication with the NRC Project Manager. of this letter provides the responses to NRC RAIs SCVB-RAIs 1 (Revision 1),

8, 14, 24, 25, 26, 30, 32, and 34 from the Reference 2 letter. The response to SCVB-RAI 1 is revised to reflect changes made in Revision 1 of the BFN EPU LAR Attachment 39 (Enclosure 4 of this letter) to reflect the implementation of the BFN National Fire Protection Association 805 License Amendments, and for consistency with the responses to other SCVB-RAIs. of this letter provides a supplement to the Power Uprate Safety Analysis Report (PUSAR) (NEDC-33860P, Revision 0). The supplement contains changes that were presented at the NRC CAP Credit Elimination audit (Reference 3).

GE-Hitachi Nuclear Energy Americas LLC (GEH) consider portions of the information provided in Enclosure 2 of this letter to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390, Public inspections, exemptions, requests for withholding. An affidavit for withholding information, executed by GEH, is provided in . Enclosure 3 is a non-proprietary version of the supplement to the PUSAR provided in Enclosure 2. Therefore, on behalf of GEH, TVA requests that Enclosure 2 be withheld from public disclosure in accordance with the GEH affidavit and the provisions of 10 CFR 2.390. Enclosures 2 and 3 supersede and replace Sections 2.5.1.4.2, 2.6.5.2 (Fire Event ECCS NPSH and ATWS ECCS NPSH), 2.11.1.1, 2.11.1.2.2, 2.11.1.2.3, 2.11.1.3, and Tables 2.5-1, -2, Tables 2.6-3, -4, -4b, Tables 2.8-1, -2, -4, -5, -6, and Figures 2.5-1, 2.6-16, -18 of Attachments 6 and 7, respectively, of the BFN EPU LAR (Reference 1).

U.S. Nuclear Regulatory Commission CNL-16-114 Page 3 August 3, 2016 of this letter provides a supplement to Attachment 39 of the BFN EPU LAR (Reference 1). This supplement revises Attachment 39 to incorporate changes discussed at the NRC CAP Credit Elimination audit (Reference 3). Enclosure 4 supersedes and replaces 9 of the BFN EPU LAR (Reference 1). of this letter provides Revision 3 to the BFN EPU List and Status of Plant Modifications. The BFN EPU List and Status of Modifications is revised to reflect elimination of the need for the Emergency High Pressure Makeup Pump because this pump is no longer necessary to provide additional Emergency Core Cooling System net positive suction head margin under EPU conditions. Enclosure 5 supersedes and replaces Revision 2 of 7 of the BFN EPU LAR (Reference 4). of this letter provides the updated response (Revision 1) to the NRC RAI AFPB-RAI 4. The update has been made to reflect elimination of the need for the Emergency High Pressure Makeup Pump. Enclosure 6 of this letter supersedes and replaces the NRC RAI AFPB-RAI 4 response provided in Enclosure 1 of the Reference 5 letter. of this letter provides Revision 1 to the Zachry Nuclear Engineering, Inc.,

Engineering Evaluation 16-E04, Description of BFN RHR Heat Exchanger Test Data Evaluation, which is referenced in the response to NRC RAI SCVB-RAI 32 of Enclosure 1 of this letter. Zachry Nuclear Engineering, Inc., consider portions of the information provided in of this letter to be proprietary and, therefore, exempt from public disclosure pursuant to 10 CFR 2.390, Public inspections, exemptions, requests for withholding. An affidavit for withholding information, executed by Zachry Nuclear Engineering, Inc., is provided in Enclosure 10. Enclosure 8 is a non-proprietary version of the information provided in Enclosure 7. Therefore, on behalf of Zachry Nuclear Engineering, Inc., TVA requests that Enclosure 7 be withheld from public disclosure in accordance with the Zachry Nuclear Engineering, Inc., affidavit and the provisions of 10 CFR 2.390.

TVA has reviewed the information supporting a finding of no significant hazards consideration and the environmental consideration provided to the NRC in the Reference 1 letter. The supplemental information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. In addition, the supplemental information in this submittal does not affect the bases for concluding that neither an environmental impact statement nor an environmental assessment needs to be prepared in connection with the proposed license amendment.

Additionally, in accordance with 10 CFR 50.91(b)(1), TVA is sending a copy of this letter to the Alabama State Department of Public Health.

There are no new regulatory commitments associated with this submittal. If there are any questions or if additional information is needed, please contact Mr. Edward D. Schrull at (423) 751-3850.

U.S. Nuclear Regulatory Commission CNL-16-114 Page4 August 3, 2016 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 3rd day of August 2016.

i:;v.~w J. W. Shea Vice President, Nuclear Licensing

Enclosures:

1. Responses to NRC Requests for Additional Information SCVB-RAls 1 (Revision 1), 8, 14, 24, 25, 26, 30, 32, and 34

2. Supplement to BFN EPU LAR, Attachment 6, NEDC-33860P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate, (Proprietary version)

3. Supplement to BFN EPU LAR, Attachment 7, NED0-338 60, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate, (Non-proprietary version)

4. BFN EPU LAR, Attachment 39, Revision 1, RHR Heat Exchanger K-values Utilized in EPU Containment

5. BFN EPU LAR, Attachment 47, List and Status of Plant Modifications, Revision 3

6. Response to NRC Request for Additional Information AFPB-RAI 4, Revision 1

7. Engineering Evaluation 16-E04, Description of BFN RHR Heat Exchanger Test Data Evaluation , Revision 1 (Proprietary version)

8. Engineering Evaluation 16-E04, Description of BFN RHR Heat Exchanger Test Data Evaluation, Revision 1 (Non-proprietary version)

9. GE Hitachi Nuclear Energy Affidavit for NEDC-33860P, Revision 0

10. Zachry Nuclear Engineering , Inc. Affidavit for Engineering Evaluation 16-E04, Revision 1 cc:

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant State Health Officer, Alabama Department of Public Health (w/o Enclosures 2 and 7)

ENCLOSURE 1 Responses to NRC Requests for Additional Information SCVB-RAIs 1 (Revision 1), 8, 14, 24, 25, 26, 30, 32, and 34

ENCLOSURE 1 SCVB-RAI 1 (Revision 1)

During the review of Attachment X of the LAR regarding National Fire Protection Association (NFPA) 805 Transition Report (Reference 1), the NRC staff requested in an RAI (SCVB-RAl-5) that TVA describe the revised BFN, Units 1, 2, and 3, residual heat removal (RHR) heat exchanger performance monitoring program, which will assure that its fouling factor and tube plugging would not exceed their worst values assumed in calculating a K-value of 284.5 British thermal unit (BTU)/sec-degrees Fahrenheit (°F). In response to this RAI (Reference 2), TVA stated, "The revised performance monitoring program has not been developed at this time ... ",

and made a commitment to revise the program that monitors the RHR heat exchanger performance.

In the approval of the NFPA 805 LAR, the NRC imposed the following license condition, which was accepted by TVA as implementation Item No. 49 in Reference 3:

Revise the program that monitors BFN Residual Heat Removal (RHR) heat exchanger performance for consistency with the assumptions of the NFPA 805 Net Positive Suction Head (NPSH) analysis. The monitoring program shall include verification that the tested worst fouling resistance, with measurement uncertainty added, of all BFN Units 1, 2, and 3 RHR heat exchangers is less than the design value of 0.001517 hr-ft2-°F/BTU and the worst tube plugging is less than 4.57 percent.

In Attachment 6 (Reference 4) and Attachment 39 (Reference 5) to the extended power uprate (EPU) LAR, at the EPU design-basis accident (DBA) loss-of-coolant accident (LOCA) statepoint, the RHR heat exchanger K-value for one heat exchanger is reported to be 265 BTU/sec-°F for a design fouling resistance of 0.001521 hr-ft2-°F/BTU, which supersedes the fouling factor of 0.001517 hr-ft2-°F/BTU reported in the NFPA 805 LAR.

Section 2.1 of Reference 5 provides the following EPU RHR heat exchanger K-values used in the analyses:

265 BTU/sec-°F (DBA-LOCA, Small Break LOCA, Loss of Shutdown Cooling, Stuck Open Relief Valve and SBO [Station Blackout]), 302 BTU/sec-°F (Shutdown of Non-Accident Unit), and 287 BTU/sec-°F (fire event defense-in-depth demonstration case) are based on the EPU design fouling resistance, 0.001521 hr-ft2-°F/BTU.

307 BTU/sec-°F (fire event licensing basis) is based on the EPU nominal fouling resistance, 0.001097 hr-ft2- °F/BTU.

277 BTU/sec-°F for the ATWS-MSIVC-EOC event corresponds to a nominal fouling resistance of 0.001220 hr-ft2-°F/BTU.

Describe the performance monitoring program to monitor the as-found worst RHR heat exchanger fouling factor and plugged tubes. As mentioned above, the description of this program was previously requested for the NFPA 805 LAR approval and is being again requested for the EPU LAR submittal. The monitoring program must verify the EPU design fouling resistance, 0.001521 hr-ft2-°F/BTU and EPU nominal fouling resistance and 0.001097 hr-ft[2]-°F/BTU, as given above.

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ENCLOSURE 1 The description of the program should include the following:

(a) Scope of monitoring (b) Frequency of monitoring (c) Acceptance criteria for the fouling factor should be less than nominal fouling resistance of 0.001097 hr-ft2-°F/BTU (for fire event) with uncertainty included (d) Acceptance criteria for plugged tubes - must be less than or equal to 4.57 percent tubes (e) Accepted industry standards and guidelines used for heat exchanger performance testing (f) Test setup (g) Instrumentation with its accuracy (h) Method of suppression pool heatup (i) Data acquisition system (j) Uncertainty analysis (k) Data reduction method for calculation of the fouling factor (l) Method of as-found heat-exchanger inspection for determining the number of plugged tubes and the effective heat transfer area REFERENCES 1 Tennessee Valley Authority (TVA), Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3, Transition to 10 CFR 50.48(c) - NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition, Transition Report, dated March 2013 (ADAMS Accession Number ML13092A392).

2 Letter from TVA to NRC, dated June 13, 2014, Response to NRC Request for Additional Information Regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (TAC Nos.

MF1185, MF1186, and MF1187) - Attachment X and Fire Modeling (ADAMS Accession Number ML14167A175).

3 Letter from TVA to NRC, dated October 20, 2015, Update to License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (TAC Nos. MF1185, MF1186, and MF1187) - Revised Implementation Item 49 (ADAMS Accession No. ML15293A527).

4 Attachment 6 to EPU LAR, NEDC-33860P, Revision 0, Safety Analysis Report for Browns Ferry Nuclear Plant, Units 1, 2, and 3, Extended Power Uprate (proprietary) (ADAMS Accession No. ML15282A264 (non-public)).

5 Attachment 39 to EPU LAR, RHR Heat Exchanger K-values Utilized in EPU Containment Analyses (ADAMS Accession No. ML15282A235).

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ENCLOSURE 1 TVA Response:

General

Description:

Beginning in January 2012, Residual Heat Removal (RHR) heat exchanger performance tests were performed in conjunction with quarterly Reactor Core Injection Cooling (RCIC) surveillance testing (see item h, below, for further details). The RHR heat exchanger testing involves installation of temporary temperature and temporary differential pressure instruments (connected to instrument taps from permanently installed Residual Heat Removal Service Water (RHRSW) flow orifices or RHR flow nozzles) to collect the data necessary to compute heat exchanger tube side and shell side heat transfer rates. The RHR heat exchanger testing uses a Tennessee Valley Authority (TVA)-approved procedure to perform the test. Under contract, an outside vendor operating under a 10 CFR 50 Appendix B Quality Assurance (QA) program provides the test instrumentation, including data acquisition system, collects and processes the data, and provides test reports documenting the results. The data analysis and preparation of vendor test reports are performed by a vendor also operating under a 10 CFR 50 Appendix B Quality Assurance Program in accordance with approved procedures. The procedures include steps to compare process and tube side heat transfer rates and to statistically evaluate test data such that results conservatively account for the uncertainties associated with each test. Thus, the accuracy of each test, which varies from one test to another, is reflected in the test results which are then compared to the acceptance criteria.

This RAI response details describes how the Browns Ferry Nuclear Plant (BFN) Generic Letter (GL) 89-13 program is expected to be changed once incorporates the National Fire Protection Association (NFPA) 805 license condition requirements becomes effective. The GL 89-13 program is currently scheduled to be revised by June 16, 2016, to support NFPA 805 transition implementation. TVA will inform the NRC when tThe GL 89-13 program has been revised through incorporation of NFPA 805 license condition requirements into applicable implementing procedures. Copies of these implementing procedures, 0-TI-322, RHR Heat Exchanger Performance Testing, and 0-TI-522, Program for Implementing NRC Generic Letter 89-13, are provided in Attachments 1 and 2, respectively, of this RAI response.

A more detailed response addressing each specific item is provided below:

(a) The Current Licensed Thermal Power (CLTP) scope of monitoring complieswill be established once the with the NFPA 805 license condition is implemented (on or before June 16, 2016). Extended Power Uprate (EPU) implementation will not require any changes to the CLTP scope of monitoring. In both cases, the scope of monitoring will includes verification that: (1) tested worst fouling resistance of all BFN Units 1, 2, and 3 RHR heat exchangers is less than the design value acceptance criteria; (2) worst tube plugging is less than the tube plugging acceptance criteria.

(b) An RHR heat Heat exchanger Exchanger performance Performance testing Monitoring program Program will be maintained through the BFN Preventive Maintenance (PM) program. Previously (prior to June 16, 2016)Prior to implementation of the RHR Heat Exchanger Performance Monitoring Program required by the NFPA 805 license condition, the GL 89-13 commitment for the RHR heat exchangers was met through BFN PM Program routine cleaning and inspection. Current plans are to have tested eEach RHR heat exchanger was tested at least once by June 16, 2016. Thereafter,Subsequent testing of each RHR heat exchanger will be tested performed periodically at an interval that is nominally four years, but initially will not exceed five years. The performance testing PMs provide criteria for reassessing the performance testing frequency based E1-3

ENCLOSURE 1 upon test results. Additionally, the heat exchangers will be cleaned on an 8-year frequency at a maximum. This 8-year cleaning frequency is based on supporting PM-required RHR heat exchanger tube eddy current testing, which is a PM requirement outside of the BFN GL 89-13 program. More frequent RHR heat exchanger cleanings will occur if the fouling rate, (as trended,) indicates the need to take corrective actions in order to maintain the heat exchanger condition within the fouling resistance acceptance criteria (see item (c) response, below).

CLTP or EPU Frequency of Monitoring -

Performance Testing Each RHR heat exchanger will havehas been performance tested at least once CLTP (expectation as of June 16, 2016) and will be tested periodically at an interval that initially will not exceed five years.

Each RHR heat exchanger will be tested periodically at an interval that is nominally four years, but will not exceed five years.Each RHR heat exchanger will EPU Implementation have performance testing at a periodicity determined by RHR heat exchanger test results and the fouling rate trended by TVA engineering.

CLTP or EPU Frequency of Monitoring -

Visual Inspection and Cleaning Each RHR heat exchanger will be CLTP (expectation as of June 16, 2016) cleaned on an 8-year frequency.

Each RHR heat exchanger will be cleaned once every 8-years at a maximum. or mMore frequently RHR heat exchanger cleaning will occur if the EPU Implementation fouling rate, as trended, by TVA engineering indicates the need to take corrective actions in order to maintain the heat exchanger condition within the fouling resistance acceptance criteria.

(c) CLTP fouling resistance acceptance criteria will bewas established upon implementation of the NFPA 805 license condition (on or before June 16, 2016). Upon EPU implementation, the fouling resistance acceptance criteria will be changed, as described in EPU License Amendment Request (LAR) Attachment 39. Specifically, the NFPA 805 design fouling resistance of 0.001517 hr-ft2-°F/Btu (corresponding to Design Basis Accident-Loss of Coolant Accident (DBA-LOCA) event with RHR heat exchanger K of 265 Btu/sec-°F) will be replaced with the EPU design fouling resistance of 0.0015210.001562 hr-ft2-°F/Btu (corresponding to the limiting EPU DBA-LOCANFPA 805 fire event with RHR heat exchanger K of 265 290 Btu/sec-°F).

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ENCLOSURE 1 DBA-LOCA event CLTP or EPU Fouling Resistance Acceptance Criteria EPU Implementation EPU Design (DBA-LOCA event):

0.0015210.001562 hr-ft2-°F/Btu*

• The EPU DBA-LOCA design and licensing basis minimum required heat removal rate (System Operability Limit per American Society of Mechanical Engineers (ASME)

Operation and Maintenance (OM)-2015, Part 21, Section 9.1) is 80,136,000 Btu/hr per heat exchanger, with two heat exchangers in service. BFN design calculations establish tThe RHR heat exchanger testing program fouling resistance acceptance criteria is based on the limiting EPU DBA-LOCA NFPA 805 fire eventdesign and licensing basis minimum required heat removal rate.

The RHR heat exchanger performance test result fouling resistance (including test uncertainty) will be compared to the DBA-LOCA eventfouling resistance acceptance criterion (fouling resistance of 0.0015210.001562 hr-ft2-°F/Btu). This acceptance criterion was determined from a deterministic containment analysis based on conservative inputs.

Upon EPU implementation, BFN will follow the guidance from the ASME OM-2015, Part 21, Section 9.1, System Operability Limits, for the DBA-LOCA event. Specifically, the EPU DBA-LOCA design and licensing basis minimum required heat removal rate, as shown described in the notes for EPU LAR Attachment 39, Table 4, will be the System Operability Limit. Additionally, RHR heat exchanger performance test fouling resistance, including test uncertainty, will be trended for comparison to the DBA-LOCA event fouling resistance acceptance criterion, in a manner consistent with ASME OM-2015, Part 21, Section 6.10, except for the second paragraph of Section 6.10.2. Because the BFN testing program will compare the test results to the acceptance criteria at the time the program is established, BFN will take specific exception to the second paragraph of Section 6.10.2 which requires trending these parameters for a minimum of three test or monitoring points prior to comparison to the applicable acceptance criteria (i.e., trending will be performed prior to having a minimum of three test or monitoring points). This is the only exception to Section 6.10.

Fire Event CLTP or EPU Fouling Resistance Acceptance Criteria CLTP (expectation as of June 16, 2016) CLTP Fire event: 0.001517 hr-ft2-°F/Btu EPU Fire event: 0.0010970.001562 hr-EPU Implementation ft2-°F/BTU*

• The EPU fire event design and licensing basis minimum required heat removal rate is 128,203,200124,966,800 Btu/hr with one heat exchanger in service. BFN design calculations establish tThe RHR heat exchanger testing program fouling resistance acceptance criteria for the fire event at EPU conditions is based on the EPU fire event design and licensing basis minimum required heat removal rate.

The RHR heat exchanger performance test result nominal fouling resistance (including test uncertainty) will be compared to the fire event acceptance criterion (fouling resistance of 0.0010970.001562 hr-ft2-°F/Btu*).

(d) The CLTP allowable tube plugging acceptance criteria (4.57%) applicable to the NFPA 805 licensing condition will beas established in applicable implementing program E1-5

ENCLOSURE 1 procedures once the NFPA 805 license condition is implemented (on or before June 16, 2016). Upon EPU implementation, BFN design calculations establish the RHR heat Heat exchanger Exchanger testing Performance Monitoring program Program tube plugging limit acceptance criteria will remain at 77 tubes (4.57% of 1700 tubes) mechanically plugged.

(e) Accepted industry standards and guidelines used for heat exchanger performance testing will beare identified in applicable CLTP program documents once the NFPA 805 license condition is implemented (on or before June 16, 2016). EPU will not require any change from the standards and guidelines used for CLTP heat exchanger performance testing.

CLTP or EPU Accepted Industry Standards/Guidelines Performance testing will beis conducted consistent with the guidance described in Electric Power Research Institute (EPRI) 3002005340, Service Water Heat Exchanger Testing Guidelines, May 2015.

Testing performed between January, 2015 and May 2015 was conducted consistent with the guidance provided in CLTP (expectation as of June 16, 2016) the previous version of this document, EPRI TR-107397, Service Water Heat Exchanger Testing Guidelines, Final Report, March 1998. Testing conducted in January 2012 also followed the EPRI TR-107397 guidelines as well as EPRI NP-7552, Heat Exchanger Performance Monitoring Guidelines, dated December 1991.

Performance testing will be conducted consistent with the guidance described in EPRI 3002005340, Service Water Heat Exchanger Testing Guidelines, May 2015.

Parameter (fouling resistance) trending EPU Implementation and comparison to the DBA-LOCA event acceptance criteria will be performed consistent with the guidance provided in ASME OM-2015, Part 21, Section 6.10, Parameter Trending, except as noted in response (c).

(f) Test set-up included installation of temporary surface mounted temperature sensors on the heat exchanger process (RHR) and cooling water (RHRSW) inlet and outlet pipes.

The piping insulation was removed and eight temporary surface-mounted temperature sensors were uniformly spaced at 45° increments around the circumference of each outlet pipe. Piping insulation was also removed from each inlet pipe and four temporary surface-mounted temperature sensors were uniformly spaced at 90° increments around the circumference of each inlet pipe. The pipe insulation was then reinstalled over the E1-6

ENCLOSURE 1 temporary surface-mounted temperature sensors to reduce the influence from external environmental conditions. The temporary surface-mounted temperature sensor leads were bundled and routed to the data acquisition unit.

Dimensions of the system piping where RHRSW and RHR temperature sensors were mounted are contained in the detailed vendor test report. RHRSW inlet temperature sensors were mounted on 16 OD, 0.375 wall thickness, carbon steel pipe. RHRSW outlet temperature sensors were mounted on 12.75 OD, 0.375 wall thickness, carbon steel pipe. RHR inlet and outlet temperature sensors were mounted on 20 OD, 0.50 wall thickness, carbon steel pipe.

Temporary differential pressure instruments were connected to instrument taps from permanently installed RHRSW flow orifices and RHR flow nozzles.

The GL 89-13 implementing procedures require the following.

- Temporary surface mounted temperature instrumentation for RHR and RHRSW inlet and outlet piping shall meet the guidance identified in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

- Temporary differential pressure (DP) instrumentations connected to the instrument taps from the permanently installed RHRSW flow orifices and RHR flow nozzles shall meet the guidance provided in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

Future RHR heat exchanger performance testing will be performed in a manner similar to or the same as that described above and performed to date, or in a manner that is consistent with the EPRI guidance identified in response (e), above.

(g) Temporary test instrumentation (e.g., surface-mounted temperature sensors, delta-P meters, current converter and data acquisition system) and thus, the associated instrumentation accuracy, is provided to BFN under contract from an outside vendor where 10 CFR Part 50 Appendix B and 10 CFR Part 21 apply. The vendors Quality Assurance System complies with applicable requirements of International Standardization Organization (ISO)/International Electrotechnical Commission (IEC)-17025-2005, American National Standards Institute (ANSI)/National Conference of State Legislatures (NCSL) Z540-I-1994 and ISO 9001: 2008. The instruments are calibrated against standards traceable to the National Institute of Standards & Technology (NIST) or compared to nationally or internationally recognized consensus standards. The reported calibration uncertainty has a confidence level of 95% (k=2). The temporary surface-mounted temperature sensors used during the tests that have been performed to date had a calibration accuracy of 0.1 °F. Calibration certificates for the pre-test and post-test calibrations are included in the vendor test report.

Temporary delta-P meters were used to record the flow rates for RHR and RHRSW. The meters had a pre-test and post-test accuracy of 0.05% of full scale. The composite systematic uncertainty for the instruments used in measurement of RHRSW and RHR flow rates for the RHR heat exchanger performance tests performed since January 2012 is documented in each vendor test report. This value was calculated using 95% confidence analysis techniques. The resulting composite systematic uncertainty from each test for both RHRSW and RHR flow rates has been less than +/-5% of measured flow.

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ENCLOSURE 1 Supplied temporary test instrumentation and associated instrumentation accuracy meets the Industry Standards/Guidelines identified in response (e), above. Specifically, EPRI 3002005340, Section 1.6.9 states, ASME PTC 12.5, Single Phase Heat Exchangers was revised in 2005 and provides comprehensive guidance to plan, conduct and analyze results for accurate performance tests of single phase heat exchangers. The Code details information for calculation techniques and methods to determine steady state performance at both test conditions and reference conditions. Guidelines are also provided for instrumentation and accuracy. ASME PTC 12.5, Section 3.1.1 states, As a benchmark, the calibration uncertainty for temperature measurements shall be less than +/-0.2°F

(+/-0.1°C), the total flow measurement uncertainty shall be less than +/-5% of measured flow The instrumentation used in the tests performed to date meets these guidelines.

Installation of temporary test instrumentation is performed under an approved TVA procedure using the BFN work order process. The GL 89-13 implementing procedures require the following.

- Temporary instruments shall be calibrated against standards traceable to the National Institute of Standards and Technology or compared to nationally or internationally recognized consensus standards.

Future RHR heat exchanger performance testing is expected to be performed in a manner similar to or the same as that described above and performed to date.

(h) One of the attributes of an effective RHR heat exchanger performance test is to maximize the heat exchanger heat removal rate. This requires maximizing the difference between the RHR (heat exchanger shell side) inlet temperature and the RHRSW (heat exchanger tube side) inlet temperature. The RHR (heat exchanger shell side) inlet temperature is dependent upon the suppression pool (torus) temperature while the RHRSW (heat exchanger tube side) inlet temperature is dependent upon the ultimate heat sink (river) temperature. These water bodies experience a maximum temperature differential during winter months, with a reduced temperature difference in spring and fall months, and a minimum temperature differential during summer months. Consequently, the best time of the year for maximizing the heat load on the RHR heat exchanger is to perform the test during winter months.

Beginning January 2012, RHR heat exchanger testing has been performed in accordance with an approved TVA procedure in conjunction with quarterly RCIC system surveillance testing. The intent of performing these tests in a back-to-back fashion was to allow the heat input rate from the RCIC turbine exhaust to the suppression pool (torus) to be matched to the removal rate from the suppression pool through the RHR heat exchanger while in the suppression pool cooling mode of operation. A perfect match of the heat input and removal rates would result in no change to the suppression pool temperature over the duration of the test data collection period. This condition would result in optimal steady state RHR (heat exchanger shell side) inlet temperatures during the test and would also serve to reduce the uncertainty associated with the test data. However, in practice, it is not feasible to match the heat input and heat removal rates exactly. Consequently, during the test data collection period there is some suppression pool heating or cooling, even though establishment of test conditions matches heat input and heat removal rates to the extent practical. In conclusion, there is actually no intent to change the suppression pool temperature during the RHR heat exchanger test data collection period.

E1-8

ENCLOSURE 1 Future RHR heat exchanger performance testing will be performed in a manner similar to or the same as that described above and performed to date.

(i) The data acquisition system for all tests performed since January 2012 was provided to TVA by an outside vendor operating under their own 10 CFR 50 Appendix B Quality Assurance program. This vendor provided the test instrumentation, including data acquisition system. The test instrumentation and data acquisition system is capable of instrumenting two heat exchangers at the same time and still have spares. All instruments and software are labeled and configured for specific use in each test location. This system complies with the requirements listed in the EPRI document TR-107397. A data acquisition software package works with the heat exchanger instrumentation system and produces data files that may be loaded directly into Proto-HX. Personal computers with data collection software are provided. The data collection software is written and validated under the vendor software Quality Assurance program. Time stamped data is collected from each sensor.

The GL 89-13 implementing procedures require the following.

- The data acquisition system, including the associated software, shall comply with the guidance in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

- Computer programs used in the thermal performance analysis shall meet the requirements of 10 CFR 50, Appendix B, and 10 CFR 21.

Future RHR heat exchanger performance testing will be performed in a manner similar to or the same as that described above and performed to date, or in a manner that is consistent with the EPRI guidance identified in response (e), above.

(j) The RHR heat exchanger performance testing that has been performed at BFN since January 2012 has used an outside vendor operating under a 10 CFR 50 Appendix B Quality Assurance program to provide the test instrumentation, collect and process the data, and provide test reports documenting the results. The data analysis and preparation of vendor test reports are performed by a vendor also operating under a 10 CFR 50 Appendix B Quality Assurance Program in accordance with approved procedures that include steps to compare process and tube side heat transfer rates and to statistically evaluate test data such that results account for the uncertainties associated with each test.

The method of calculation follows the EPRI guidelines (see response (e), above) in terms of determining the uncertainty contributors of precision and bias errors for thermal performance test evaluations. The uncertainty analysis methodology in Proto-HX determines the sensitivity coefficients through a numerical approach using the central differencing method (i.e., symmetric uncertainties). The EPRI guideline (see response (e),

above) provides an overview of this approach. The variables considered are the test data (flow rates and temperatures) and the film coefficients.

The GL 89-13 implementing procedures require the following.

- The uncertainty analysis methodology shall comply with the approach described in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

- Computer programs used in the thermal performance analysis shall meet the requirements of 10 CFR 50, Appendix B, and 10 CFR 21.

Future RHR heat exchanger performance testing will be performed in a manner similar to or the same as that described above and performed to date.

E1-9

ENCLOSURE 1 (k) The data reduction method for calculating the fouling resistance involves averaging the data from the RHR heat exchanger test to identify nominal values for each parameter.

These nominal values are then analyzed to determine the condition of the heat exchanger with respect to the overall fouling factor. Further analysis of the test data identifies the uncertainty in each parameter measurement. These uncertainties are then used to establish the overall uncertainty in the test result. The test result is then compared to the acceptance criterion to determine if the test is satisfactory.

The data reduction is performed by the vendor under the same processes as described in item (j), above. In the testing that has been performed to date, the vendor calculation follows the heat transfer test method using the heat transfer at design limiting conditions as the performance parameter. This method is outlined in EPRI Test Report 107397, Service Water Heat Exchanger Testing Guidelines, dated March 1998.

These initial testing results (vendor test reports) will behave been revised to recompute the performance parameter, including the performance parameter (fouling resistance) uncertainty, so as to report results consistent with the units specified in the NFPA 805 license condition for the fouling resistance, hr-ft2-°F/Btu. The GL 89-13 implementing procedures require the following.

- Data reduction shall comply with the approach described in the EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

Future data reduction methods for calculating the fouling resistance both prior to and following EPU implementation will follow the same or similar data reduction method for calculating the fouling resistance or in a manner that is consistent with the EPRI guidance identified in response (e), above.

(l) In the context of the method of as-found heat-exchanger inspection for determining the number of plugged tubes and the resulting effective heat transfer area, it is noteworthy how this information actually impacts the heat exchanger post-test analysis. The vendor test reports contain the following conservative assumption: It is assumed that all tubes were unobstructed during the test (i.e., none of the tubes were plugged by macro fouling during the test). It is an inherent part of the PROTO-HX method of analysis to distribute the tube-side flow equally to all tubes and to use the specified heat transfer area in the fouling calculation. This assumption is acceptable since lost area due to unknown macrofouling will show up as extra fouling resistance.

The method of as-found heat-exchanger inspection for determining the number of unavailable tubes and the effective heat transfer area is performed in accordance with an approved TVA procedure. An attachment to the procedure provides a GL 89-13 Heat Exchanger Visual Inspection and Evaluation Form. The form requires recording the number of tubes plugged, the number of tubes fully blocked (>90%), and the number of tubes partially obstructed (75% - 90%). Guidance for determining the number of tubes equivalent blocked is that tubes found with >90% of the area obstructed are considered fully blocked and 50% of the tubes with 75% to 90% of the area obstructed are considered fully blocked. The acceptance criterion is the number of tubes plugged, fully blocked or equivalent blocked must be less than the maximum plugging limit. This information on as-found potential tube blockage is only applicable for consideration in determining heat exchanger past-operability for an event where the potential tube blockage was introduced into the system since the performance of the last heat exchanger test. In some cases, this E1-10

ENCLOSURE 1 same as-found blockage could have been present during previous heat exchanger testing.

The consideration of the effects of potential blockages inside the heat exchanger prior to the last heat exchanger performance test is addressed in the first paragraph of this response (item (l)).

EPU implementation will not change the method of as-found heat-exchanger inspection for determining the number of unavailable tubes and the effective heat transfer area.

However, upon EPU implementation, design basis calculations will be revisedacceptance criteria will be included in the Updated Final Safety Analysis Report (UFSAR), to identify, for any given fouling resistance, the maximum allowable number of tubes that could be are unavailableplugged or otherwise obstructed and still meet the DBA-LOCA system operability limit minimum heat removal requirement. This determination will facilitate a more immediate assessment of the as-found condition of any RHR heat exchanger where the fouling resistance is known or can be projected based on the fouling rate and the as-found number of tubes determined to be fully blocked from the as-found heat exchanger inspection. This acceptance criteria, included in the UFSAR, would only be used for past-operability/functionality determinations.

E1-11

ATTACHMENT 1 Procedure 0-TI-322, RHR Heat Exchanger Performance Testing, Revision 3

Browns Ferry Nuclear Plant Unit 0 Technical Instruction 0-TI-322 RHR Heat Exchanger Performance Testing Revision 0003 Quality Related Level of Use: Continuous Use Effective Date: 07-15-2016 Responsible Organization: SBE, System Engineering - BOP Prepared By: David Drummonds Approved By: John Colvin

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 2 of 33 Current Revision Description Pages Affected: 9 Type of Change: Revision Tracking Number: 004 Section 4.4, Page 9, Fourth NOTE, Added details in Section 4.4 on Data Acquisition System (DAS).

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 3 of 33 Table of Contents

1.0 INTRODUCTION

.......................................................................................................... 5 1.1 Purpose ........................................................................................................................ 5 1.2 Scope............................................................................................................................ 5 1.3 Frequency ..................................................................................................................... 5 1.4 Conditions ..................................................................................................................... 5 1.5 Applicability ................................................................................................................... 5

2.0 REFERENCES

............................................................................................................. 6 2.1 Performance References .............................................................................................. 6 2.2 Developmental References........................................................................................... 6 2.3 Commitments ................................................................................................................ 6 3.0 PRECAUTIONS AND LIMITATIONS ........................................................................... 6 3.1 Precautions ................................................................................................................... 6 3.2 Limitations..................................................................................................................... 7 4.0 PREREQUISITE ACTIONS .......................................................................................... 7 4.1 Preliminary Actions ....................................................................................................... 7 4.2 Special Tools ................................................................................................................ 7 4.3 Approvals and Notifications .......................................................................................... 8 4.4 Field preparations ......................................................................................................... 8 5.0 ACCEPTANCE CRITERIA ......................................................................................... 13 5.1 Mechanical Plugging ................................................................................................... 13 5.2 Fouling Resistance ..................................................................................................... 13 6.0 PERFORMANCE ........................................................................................................ 14 6.1 Responsibilities ........................................................................................................... 14 6.1.1 Corporate GL 89-13 Program Owner ............................................................ 14 6.1.2 Site GL 89-13 Program Owner ..................................................................... 14 6.1.3 Lead Performer ............................................................................................. 14 6.1.4 Heat Exchanger Engineer ............................................................................. 14 6.1.5 Maintenance ................................................................................................. 14 6.2 M&TE .......................................................................................................................... 15 6.3 FIRST HEAT EXCHANGER TEST ............................................................................. 15

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 4 of 33 Table of Contents (continued) 6.4 SECOND HEAT EXCHANGER TEST ........................................................................ 19 7.0 POST PERFORMANCE ACTIVITY ............................................................................ 23 7.1 Restoration ................................................................................................................. 23 7.2 Summarizing Results .................................................................................................. 23 8.0 RECORDS .................................................................................................................. 24 : Technical Instruction Review Form ........................................................ 25 : Typical RTD Test Sensor Locations ........................................................ 26 : RHR Surface Mounted Temperature Installation ................................... 27 : RHRSW Surface Mounted Temperature Sensor Installation ................................................................................................. 30 : Ambient Temperature Sensor Configuration ......................................... 33

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 5 of 33

1.0 INTRODUCTION

1.1 Purpose The purpose of this test is to periodically monitor the performance of the Browns Ferry Nuclear Plant Unit Residual Heat Removal (RHR) heat exchangers in accordance with the GL 89-13 program (0-TI-522). The test demonstrates that the overall fouling resistance bounds the values assumed in containment cooling analyses.

Inlet and outlet temperatures and flows are recorded on both the RHR and RHRSW sides of one RHR Heat Exchanger at a time with temporary instrumentation. The data is recorded on a data acquisition system (DAS). Data recorded in this test is analyzed to determine the overall fouling resistance.

1.2 Scope This test applies to all of the RHR heat exchangers. The test is written to test 2 heat exchangers, one at a time, on the same RHR loop during the performance of the procedure.

1.3 Frequency Frequency of this test is determined by 0-TI-522 - Program for Implementing NRC Generic Letter 89-13.

1.4 Conditions A. This test is performed while the unit is in power operation.

B. This test is performed in conjunction with RCIC surveillance testing. RCIC operation provides a heat load to the Suppression Pool for the test.

C. It is ideal to perform this test in cooler seasons when river intake temperature is less than 65°F in order to allow a 20°F, or greater, delta-T between the suppression pool and the RHRSW inlet.

1.5 Applicability A. Performance testing of RHR heat exchangers.

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 6 of 33

2.0 REFERENCES

2.1 Performance References A. 1/2/3-OI-74 B. 1/2/3-OI-23 C. ASME PTC 12.5 Single Phase Heat Exchangers, Section 3.1.1 2.2 Developmental References A. EPRI TR-107397, Service Water Heat Exchanger Testing Guidelines, dated March 1998 B. EPRI 3002005340, Service Water Heat Exchanger Testing Guidelines, May 2015 C. Calculation MDQ0009992012000094, NFPA-805, Containment Parameters, and AREVA Fuel PCT Analysis D. Calculation MDQ0023980143, RHR Heat Exchanger Tube Plugging Analysis for Power Uprates E. ASME OM-2015, Part 21, Operation and Maintenance of Nuclear Power Plants, Inservice Performance Testing of Heat Exchangers in Light-Water Reactor Power Plants F. 0-TI-522, Program for Implementing NRC Generic Letter 89-13.

2.3 Commitments A. BFN license condition 2.C.13, Units 1, 2 and 3 B. TVA letter to NRC dates 10/20/15 (L44151020001), Table S-3, Item 49 3.0 PRECAUTIONS AND LIMITATIONS 3.1 Precautions A. Data must be recorded only after flow rates and temperature fluctuations have stabilized.

B. Operate systems 23 and 74 in accordance with their respective Operating Instructions.

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 7 of 33 3.1 Precautions (continued)

C. Maintain constant RHR and RHRSW flow rates during the data collection period.

D. Minimize changes in Suppression pool bulk temperature during the data collection period.

3.2 Limitations None.

4.0 PREREQUISITE ACTIONS 4.1 Preliminary Actions A. Procure testing services including RFP and PO, if applicable B. Verify work orders are in place for scaffolding and insulation work as required.

C. Verify chemicals used are on the ACL [Approved Chemical List]

D. Complete Transient Combustible Permit, if required.

E. Prepare Cyber Security evaluation of pertinent test equipment.

F. Verify placement of Cyber Security tags on pertinent test equipment.

4.2 Special Tools A. Vendor supplied data acquisition system (DAS)

B. Surface mounted temperature sensors C. dP transmitters D. Omega Therm 201 paste E. Bands F. Specialty tape G. Rubber gasket material [used to protect the temperature sensors when securing them to the pipe with band clamps]

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 8 of 33 Date ________

4.3 Approvals and Notifications

[1] OBTAIN Shift Manager (SM) or Unit Supervisor (US) approval to start this test. ________

US/SM

[2] [NRC] NOTIFY the three Unit Operators (UO) that this test is starting [RPT 82-16, LER 259/8232].

(U1) ________

(U2) ________

(U3) ________

4.4 Field preparations NOTE dP transmitter connections are not all the same on all three units. Temporary differential pressure (DP) instrumentations connected to the instrument taps from the permanently installed RHRSW flow orifices and RHR flow nozzles shall meet the guidance provided in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

NOTE Install 4 RTDs on inlet piping locations and 8 RTDs on outlet piping locations. Temporary surface mounted temperature instrumentation for RHR and RHRSW inlet and outlet piping shall meet the requirements identified in the EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015, guidance.

NOTE Verify calibrations of dP transmitter [flow instrument] and RTDs [temperature instruments]

have not expired prior to installation. Temporary instruments shall be calibrated against standards traceable to the National Institute of Standards and Technology or compared to nationally or internationally recognized consensus standards.

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 9 of 33 Date ________

4.4 Field preparations (continued)

NOTE The Data Acquisition System (DAS), including the associated software, shall comply with the requirements in the EPRI 3002005340 , Service water Heat Exchanger Test Guideline, May 2015, guidance.

[1] INSTALL temporary FLOW and TEMPERATURE instrumentation required to support the planned testing.

NOTE If testing heat exchangers A and C perform steps [1.1] through [1.5], otherwise NA steps

[1.1] through [1.4]. Attachment 2 shows typical RTD test sensor locations.

[1.1] INSTALL temporary RTD test sensors on RHR and RHRSW inlet and outlet piping for RHR Heat Exchanger A __________

[1.2] INSTALL temporary RTD test sensors on RHR and RHRSW inlet and outlet piping for RHR Heat Exchanger C __________

[1.3] INSTALL temporary ambient temperature instrumentation in the area of RHR Heat Exchangers A and C __________

[1.4] VERIFY installation and locations for temporary FLOW instrumentation, temporary RTD test sensors, and temporary ambient temperature instrumentation for RHR Heat Exchangers A and C __________

[1.5] RECORD instrumentation details:

Heat Pipe Pipe Pipe Wall Pipe Paint Paint RTD Sensor Insulation Thermal Exchanger OD Material Thickness Painted Type Thickness Sensor Insulation Type Properties A [inches] [Yes/No] Mounting [Material, [Thermal Plates Used Specifications, Conductivity/

[Yes/No] Thickness, Resistance, Manufacturer] Specific Heat]

RHR Inlet

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 10 of 33 Date ________

4.4 Field preparations (continued)

RHR Outlet RHRSW Inlet RHRSW Outlet Heat Pipe Pipe Pipe Wall Pipe Paint Paint RTD Sensor Insulation Thermal Exchanger OD Material Thickness Painted Type Thickness Sensor Insulation Type Properties C [inches] [Yes/No] Mounting [Material, [Thermal Plates Used Specifications, Conductivity/

[Yes/No] Thickness, Resistance, Manufacturer] Specific Heat]

RHR Inlet RHR Outlet RHRSW Inlet RHRSW Outlet NOTE If testing heat exchangers B and D perform steps [1.6] through [1.10], otherwise NA steps

[1.5] through [1.8]. Attachment 2 shows typical RTD test sensor locations.

[1.6] INSTALL temporary RTD test sensors on RHR and RHRSW inlet and outlet piping for RHR Heat Exchanger B __________

[1.7] INSTALL temporary RTD test sensors on RHR and RHRSW inlet and outlet piping for RHR Heat Exchanger D __________

[1.8] INSTALL temporary ambient temperature instrumentation in the area of RHR Heat Exchangers B and D __________

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 11 of 33 Date ________

4.4 Field preparations (continued)

[1.9] VERIFY installation and locations for temporary FLOW instrumentation, temporary RTD test sensors, and temporary ambient temperature instrumentation for RHR Heat Exchangers B and D __________

[1.10] RECORD instrumentation details:

Heat Pipe Pipe Pipe Wall Pipe Paint Paint RTD Sensor Insulation Thermal Exchanger OD Material Thickness Painted Type Thickness Sensor Insulation Type Properties B [inches] [Yes/No] Mounting [Material, [Thermal Plates Used Specifications, Conductivity/

[Yes/No] Thickness, Resistance, Manufacturer] Specific Heat]

RHR Inlet RHR Outlet RHRSW Inlet RHRSW Outlet Heat Pipe Pipe Pipe Wall Pipe Paint Paint RTD Sensor Insulation Thermal Exchanger OD Material Thickness Painted Type Thickness Sensor Insulation Type Properties D [inches] [Yes/No] Mounting [Material, [Thermal Plates Used Specifications, Conductivity/

[Yes/No] Thickness, Resistance, Manufacturer] Specific Heat]

RHR Inlet RHR Outlet RHRSW Inlet RHRSW Outlet

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 12 of 33 Date ________

4.4 Field preparations (continued)

[2] RECORD the temporary instrumentation on Attachment 3, and Attachment 4. ________

NOTE Current tube plugging information can be obtained from the heat Exchanger Engineer.

[3] RECORD tube plugging

[3.1] Record the number of tubes mechanically plugged for each RHR heat exchanger to be tested.

[3.2] Calculate percentage of mechanically plugged tubes, P, for each RHR heat exchanger to be tested, where N=

number of tubes mechanically plugged.

HEAT EXCHANGER TUBES MECHANICALLY FORMULA PERCENTAGE OF PLUGGED AS SHOWN IN TUBES PLUGGED, P PROCEDURE 0-TI-522 ATTACHMENT 1 P = N

• 100 / 1700 P = _____ %

P = N

• 100 / 1700 P = _____ %

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 13 of 33 Date ________

4.4 Field preparations (continued)

[4] VERIFY the temporary instrumentation listed in Attachment 3 and Attachment 4 have been calibrated and the calibration has NOT expired. ________

[5] VERIFY RWPs have been issued as required for performance of this instruction. ________

[6] COORDINATE with Operations to have this test run in parallel with placing RCIC in service. ________

[7] COORDINATE with Operations to maximize suppression pool (SP) bulk temperature. SP temperature should be at least 20°F greater than RHRSW intake temperature prior to starting the test. If the temperature differential cannot be met, the test may continue at the discretion of the test coordinator. _________

[8] VERIFY that all test equipment is connected and test ready. _________

5.0 ACCEPTANCE CRITERIA NOTE Test results will be provided to engineering personnel for prompt evaluation following testing in order to provide preliminary assessment of the data. The parameter determined by this test is overall fouling resistance. This value is calculated from the test data using PROTOHX software.

5.1 Mechanical Plugging Percent tubes mechanically plugged less than 4.57%.

5.2 Fouling Resistance Overall Fouling Resistance with measurement uncertainty included less than

.001517 hr-ft2-°F/BTU.

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 14 of 33 Date ________

6.0 PERFORMANCE Deviations to be documented in the test log and considered for future revision of the procedure.

6.1 Responsibilities 6.1.1 Corporate GL 89-13 Program Owner A. Provides oversight for the BFN GL 89-13 Program.

6.1.2 Site GL 89-13 Program Owner A. Responsible for ensuring GL 89-13 Program requirements are met B. Assists as necessary in performance of GL 89-13 performance testing C. Runs, or works with vendor, to run PROTOHX program for test results D. Dispositions any acceptance criteria failures 6.1.3 Lead Performer A. Runs DAS in the field.

6.1.4 Heat Exchanger Engineer A. Provides current tube plugging information for RHR heat exchangers.

6.1.5 Maintenance A. Installs and removes temporary test instrumentation.

B. Removes and installs pipe insulation.

C. Cleans piping.

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 15 of 33 Date ________

6.2 M&TE (continued) 6.2 M&TE

[1] RECORD M&TE in the table below (N/A this table if no TVA M&TE is going to be used).

TEST TVA ID NO. CAL RANGE NOTES EQUIPMENT DUE DATE 6.3 FIRST HEAT EXCHANGER TEST

[1] CREATE a test group on the ICS computer to acquisition the data specified by Lead Performer for the heat exchanger being tested with the following parameters, if available. __________

Name of Test Group _________

[1.1] RHR FLOW

[1.2] RHRSW FLOW

[1.3] Bulk Suppression Pool Temperature

[1.4] RHR Inlet Temperature

[1.5] RHR outlet Temperature

[1.6] RHRSW Inlet Temperature

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 16 of 33 Date ________

6.3 FIRST HEAT EXCHANGER TEST (continued)

[1.7] RHRSW Outlet Temperature

[2] RECORD which RHR heat exchanger is to be tested ________

Component(s) Tested

[3] REQUEST/VERIFY Operations place the RHR pump and heat exchanger to be tested in Suppression Pool Cooling per appropriate sections of 1/2/3-OI-74. ________

[4] REQUEST Operations to adjust the RHR and RHRSW System to obtain NOTE The values in [4.1] and [4.2] may be adjusted, if necessary, to obtain zero cooldown.

[4.1] RHR flow set at 7000-8000 gpm per 1/2/3-OI-74 ________

[4.2] RHRSW flow set at 3900 to 4000 gpm per 1/2/3-OI-23 ________

[5] VERIFY with Operations that the tested heat exchanger train is in service in the Suppression Pool Cooling Mode. ________

[6] VERIFY all temperature sensors are reading consistently with one another and consistently with system operating configuration. ________

[7] VERIFY the DAS is recording data. ________

[8] MONITOR the onset of stable test conditions. ________

[8.1] VERIFY RHRSW flowrate through the tested heat exchanger is in the range of 3900 - 4000 gpm and is relatively stable. ________

[8.2] VERIFY RHR flow is in the range of 7000 - 8000 gpm and is relatively stable. ________

[8.3] VERIFY the RHR inlet temperature, as verified from the DAS using one minute average of all inlet sensors, has met the stability criterion of +/-0.20°F per 5 minute interval. _________

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 17 of 33 Date ________

6.3 FIRST HEAT EXCHANGER TEST (continued)

[8.4] VERIFY the RHRSW inlet temperature, as verified from the DAS using one minute average of all inlet sensors, has met the stability criterion of +/-0.20°F per 5 minute interval. ________

NOTE The following step checks the heat balance error (HBE) of the test data using temperature data and flow data from the DAS. HBE is calculated by the DAS and indicated on the screen.

[9] MONITOR the HBE from the DAS until the desired HBE is obtained.

NOTE The heat balance error is the difference between the heat transfer rates of the tube side to the shell side. A comparison of the heat balance error to the uncertainty of the heat balance error of the data provides an independent quality check of the test data.

[10] OBTAIN the starting heat balance data from the DAS and RECORD below.

RHRSW (Tube Side) QRHRSW RHR (Shell Side) QRHR Heat Load, Q

[ m& (Cp ) Toutlet Tinlet ]

Heat Balance Error (Btu/hr)

[QRHRSW-QRHR]

Heat Balance Error (%)

Q RHRSW Q RHR x 100 Q RHRSW

[10.1] VERIFY HBE meets the test data quality criterion of

+/- 10% prior to beginning the formal 30 minute test data collection run. ________

[11] BEGIN the formal 30 minute test data collection run at the discretion of the Test Coordinator, ________

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 18 of 33 Date ________

6.3 FIRST HEAT EXCHANGER TEST (continued)

Start Date Time

[11.1] RECORD the 1 minute average value of each sensor grouping via the DAS and label the data set as the START of the formal 30 minute data run. ________

[11.2] NOTIFY the Control Room that formal 30 minute data collection run has begun and that system manipulations should be kept to a minimum. ________

[11.3] MONITOR RHR inlet temperature to ensure the stability criterion of less than or equal to +/-0.20°F/5min continues to be met (as measured from the DAS and using the 1 minute average of all RHR inlet sensors). ________

[11.4] MONITOR RHRSW inlet temperature to ensure the stability criterion of less than or equal to +/-0.20°F/5min continues to be met (as measured from the DAS and using the 1 minute average of all RHR inlet sensors). ________

[11.5] MONITOR RHRSW flow through the test heat exchanger to ensure it is remains in the range of 3900 -

4000 gpm. ________

[11.6] MONITOR RHR flow to ensure it remains in the range of 7000 - 8000 gpm .

[11.7] VERIFY Steady state test data has been recorded for a period of 30 minutes or longer and RECORD the stop time. ________

Stop Date Time

[11.8] RECORD the 30 minute average value of each sensor grouping via the DAS and label the data set as the END of the formal 30 minute data run. ________

[12] OBTAIN the ending heat balance data from the DAS and RECORD below.

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 19 of 33 Date ________

6.3 FIRST HEAT EXCHANGER TEST (continued)

RHRSW (Tube Side) QRHRSW RHR (Shell Side) QRHR Heat Load, Q

[ m& (Cp ) Toutlet Tinlet ]

Heat Balance Error (Btu/hr)

[QRHRSW-QRHR]

Heat Balance Error (%)

Q RHRSW Q RHR x 100 Q RHRSW

[13] COPY the test data file on the Laptop computer to a flash drive. The data file name should specify the component and date of the test. ________

File Name:

[14] NOTIFY Control Room that data collection for the identified heat exchanger has been completed. ________

6.4 SECOND HEAT EXCHANGER TEST

[1] CREATE a test group on the ICS computer to acquisition the data specified by Lead Performer for the heat exchanger being tested with the following parameters, if available. __________

Name of Test Group _________

[1.1] RHR FLOW

[1.2] RHRSW FLOW

[1.3] Bulk Suppression Pool Temperature

[1.4] RHR Inlet Temperature

[1.5] RHR outlet Temperature

[1.6] RHRSW Inlet Temperature

[1.7] RHRSW Outlet Temperature

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 20 of 33 Date ________

6.4 SECOND HEAT EXCHANGER TEST (continued)

[2] RECORD which RHR heat exchanger is to be tested ________

Component(s) Tested

[3] REQUEST/VERIFY Operations place the RHR pump and heat exchanger to be tested in Suppression Pool Cooling per appropriate sections of 1/2/3-OI-74. ________

NOTE The values in [4.1] and [4.2] may be adjusted, if necessary, to obtain zero cooldown.

[4] REQUEST Operations to adjust the RHR and RHRSW System to obtain

[4.1] RHR flow set at 7000-8000 gpm per 1/2/3-OI-74 ________

[4.2] RHRSW flow set at 3900 to 4000 gpm per 1/2/3-OI-23 ________

[5] VERIFY with Operations that the tested heat exchanger train is in service in the Suppression Pool Cooling Mode. ________

[6] VERIFY all temperature sensors are reading consistently with one another and consistently with system operating configuration. ________

[7] VERIFY the DAS is recording data. ________

[8] MONITOR the onset of stable test conditions. ________

[8.1] VERIFY RHRSW flowrate through the tested heat exchanger is in the range of 3900 - 4000 gpm and is relatively stable. ________

[8.2] VERIFY RHR flow is in the range of 7000 - 8000 gpm and is relatively stable. ________

[8.3] VERIFY the RHR inlet temperature, as verified from the DAS using one minute average of all inlet sensors, has met the stability criterion of +/-0.20°F per 5 minute interval. _________

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 21 of 33 Date ________

6.4 SECOND HEAT EXCHANGER TEST (continued)

[8.4] VERIFY the RHRSW inlet temperature, as verified from the DAS using one minute average of all inlet sensors, has met the stability criterion of +/-0.20°F per 5 minute interval. ________

NOTE The following step checks the heat balance error (HBE) of the test data using temperature data and flow data from the DAS. HBE is calculated by the DAS and indicated on the screen.

[9] MONITOR the HBE from the DAS until the desired HBE is obtained.

NOTE The heat balance error is the difference between the heat transfer rates of the tube side to the shell side. A comparison of the heat balance error to the uncertainty of the heat balance error of the data provides an independent quality check of the test data.

[10] OBTAIN the starting heat balance data from the DAS and RECORD below.

RHRSW (Tube Side) QRHRSW RHR (Shell Side) QRHR Heat Load, Q

[ m& (Cp ) Toutlet Tinlet ]

Heat Balance Error (Btu/hr)

[QRHRSW-QRHR]

Heat Balance Error (%)

Q RHRSW Q RHR x 100 Q RHRSW

[10.1] VERIFY HBE meets the test data quality criterion of

+/- 10% prior to beginning the formal 30 minute test data collection run. ________

[11] BEGIN the formal 30 minute test data collection run at the discretion of the Test Coordinator, ________

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 22 of 33 Date ________

6.4 SECOND HEAT EXCHANGER TEST (continued)

Start Date Time

[11.1] RECORD the 1 minute average value of each sensor grouping via the DAS and label the data set as the START of the formal 30 minute data run. ________

[11.2] NOTIFY the Control Room that formal 30 minute data collection run has begun and that system manipulations should be kept to a minimum. ________

[11.3] MONITOR RHR inlet temperature to ensure the stability criterion of less than or equal to +/-0.20°F/5min continues to be met (as measured from the DAS and using the 1 minute average of all RHR inlet sensors). ________

[11.4] MONITOR RHRSW inlet temperature to ensure the stability criterion of less than or equal to +/-0.20°F/5min continues to be met (as measured from the DAS and using the 1 minute average of all RHR inlet sensors). ________

[11.5] MONITOR RHRSW flow through the test heat exchanger to ensure it is remains in the range of 3900 -

4000 gpm. ________

[11.6] MONITOR RHR flow to ensure it remains in the range of 7000 - 8000 gpm .

[11.7] VERIFY Steady state test data has been recorded for a period of 30 minutes or longer and RECORD the stop time. ________

Stop Date Time

[11.8] RECORD the 30 minute average value of each sensor grouping via the DAS and label the data set as the END of the formal 30 minute data run. ________

[12] OBTAIN the ending heat balance data from the DAS and RECORD below.

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 23 of 33 Date ________

6.4 SECOND HEAT EXCHANGER TEST (continued)

RHRSW (Tube Side) QRHRSW RHR (Shell Side) QRHR Heat Load, Q

[ m& (Cp ) Toutlet Tinlet ]

Heat Balance Error (Btu/hr)

[QRHRSW-QRHR]

Heat Balance Error (%)

Q RHRSW Q RHR x 100 Q RHRSW

[13] COPY the test data file on the Laptop computer to a flash drive. The data file name should specify the component and date of the test. ________

File Name:

[14] NOTIFY Control Room that data collection for the identified heat exchanger has been completed. ________

7.0 POST PERFORMANCE ACTIVITY 7.1 Restoration

[1] VERIFY disconnected all temporary instrumentation. ________

[2] REMOVE all test instrumentation and support equipment from the RHR heat exchanger areas. ________

7.2 Summarizing Results A. Acceptance Criteria Test results will be provided to engineering personnel for prompt evaluation following testing in order to provide preliminary assessment of the data. The parameter determined by this test is overall fouling resistance less uncertainty. This value is calculated from the test data using software which meets the following such as PROTOHX.

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 24 of 33 7.2 Summarizing Results (continued)

NOTES

1) Test data shall be analyzed in accordance with the EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015, guidance. The analysis determines the overall fouling resistance for the heat exchanger and also determines the associated uncertainty in the test result (fouling resistance).

2) The uncertainty analysis methodology shall comply with the approach described in the EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

3) Data reduction shall comply with the approach described in the EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015, guidance.

4) Computer programs used in the thermal performance analysis shall meet the requirements of 10 CFR 50, Appendix B, and 10 CFR 21.

% TUBES OVERALL FOULING PLUGGED RESISTANCE from PROTOHx from step 4.1.2 ACCEPTANCE CRITERIA 4.57% .001517 hr-ft2-°F/BTU HEAT EXCHANGER GL 89-13 Engineer Verification of Acceptance Criteria _________

Acceptance Criteria Met Yes No B. NOTIFY operations the systems can be returned to normal operation after collecting test data and receipt of preliminary Proto-Hx calculation.

8.0 RECORDS A. Attach preliminary test evaluation

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 25 of 33 Attachment 1 (Page 1 of 1)

Technical Instruction Review Form Date ________

Instruction criteria satisfied? Yes No Results reviewed:

System Engineer Date Reason for test:

Required by schedule Plant condition (explain in Remarks)

After maintenance (explain in Remarks) Other (explain in Remarks)

Another system inoperable Signature attests that I understand the scope and purpose of this instruction and that, to the best of my knowledge, it was properly performed in accordance with instruction in that: the recording, reduction, and evaluation of data were complete and correct; acceptance criteria were met or justification for exceptions provided; deficiencies were evaluated and dispositioned; and instruction was fully complete except as noted.

Sys Eng Cognizant Org. Cognizant Reviewer Signature Date Remarks:

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 26 of 33 Attachment 2 (Page 1 of 1)

Typical RTD Test Sensor Locations NOTE Install 4 RTDs on inlet piping locations and 8 RTDs on outlet piping locations.

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 27 of 33 Attachment 3 (Page 1 of 3)

RHR Surface Mounted Temperature Installation RHR HEAT EXCHANGER ____________________

NOTES

1) The indicated module corresponds to the calibrated pairing of sensor to specific module and should serve as a guide to pre-test connections to the laptop computer.

2) Angle locations are referenced clockwise in the direction of flow with 0° for the top of a horizontal pipe or plant north for a vertical pipe.

3) If interferences prevent the sensor orientation shown below, the pattern should be rotated in either direction to facilitate as equal a distribution about the pipe circumference as possible.

Make note of any adjustments.

RHR Inlet RHR Outlet Azimuth Sensor Serial No. Module No. Azimuth Sensor Serial No. Module No.

(1) (1) 0 0 90 45 180 90 270 135 180 225 270 315

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 28 of 33 Attachment 3 (Page 2 of 3)

RHR Surface Mounted Temperature Installation NOTE dP transmitter connections are not all the same on all three units.

RHR HEAT EXCHANGER ____________________

RHRSW Inlet RHRSW Outlet Azimuth Sensor Serial No. Module No. (1) Azimuth Sensor Serial No. Module No. (1) 0 0 90 45 180 90 270 135 180 225 270 315 RHR FLOW RHRSW FLOW Flow Sensor Serial No. Module No. (1) Flow Sensor Serial No. Module No. (1)

Element Element

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 29 of 33 Attachment 3 (Page 3 of 3)

RHR Surface Mounted Temperature Installation

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 30 of 33 Attachment 4 (Page 1 of 3)

RHRSW Surface Mounted Temperature Sensor Installation Date ________

RHR HEAT EXCHNAGER ____________________

NOTES

1) Sensor connection to specific modules will not occur until just prior to heat exchanger testing. The indicated module corresponds to the calibrated pairing of sensor to specific module and should serve as a guide to pre-test connections to the laptop computer.

2) Angle locations are referenced clockwise in the direction of flow with 0° for the top of a horizontal pipe or plant north for a vertical pipe.

3) If interferences prevent the sensor orientation shown below, the pattern should be rotated in either direction to facilitate as equal a distribution about the pipe circumference as possible.

Make note of any adjustments.

RHRSW Inlet RHRSW Outlet Azimuth Sensor Serial No. Module No. (1) Azimuth Sensor Serial No. Module No. (1) 0 0 90 45 180 90 270 135 180 225 270 315 RHR HEAT EXCHNAGER ____________________

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 31 of 33 Attachment 4 (Page 2 of 3)

RHRSW Surface Mounted Temperature Sensor Installation RHRSW Inlet RHRSW Outlet Azimuth Sensor Serial No. Module No. (1) Azimuth Sensor Serial No. Module No. (1) 0 0 90 45 180 90 270 135 180 225 270 315 RHR FLOW RHRSW FLOW Flow Sensor Serial No. Module No. (1) Flow Sensor Serial No. Module No. (1)

Element Element

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 32 of 33 Attachment 4 (Page 3 of 3)

RHRSW Surface Mounted Temperature Sensor Installation

BFN RHR Heat Exchanger 0-TI-322 Unit 0 Performance Testing Rev. 0003 Page 33 of 33 Attachment 5 (Page 1 of 1)

Ambient Temperature Sensor Configuration Date ________

Applicable Heat Exchanger:

A. Ambient Sensor 1 Serial Number:

Module Number:

Location

Description:

Applicable Heat Exchanger:

B. Ambient Sensor 2 Serial Number:

Module Number:

Location

Description:

ATTACHMENT 2 Procedure 0-TI-522, Program for Implementing NRC Generic Letter 89-13, Revision 7

Browns Ferry Nuclear Plant Unit 0 Technical Instruction 0-TI-522 Program for Implementing NRC Generic Letter 89-13 Revision 0007 Quality Related Level of Use: Information Use Effective Date: 06-08-2016 Responsible Organization: PGM, Engineering Program Group Prepared By: David L. Drummonds Approved By: Patrick M. Derriso

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 2 of 26 Current Revision Description Pages Affected: 7, 8, 9, 10, 11, 12, 14, 15,16, Attachment 2 Type of Change: Revision Tracking Number: 008 The purpose of this revision is to update the procedure for NFPA-805 requirements identified subsequent to Revision 6.

Page 7, Section 2.1 - Added procedures 0-TI-345(RHRSW), 0-TI-345(EECW), 1-TI-134, 2-TI-134, and 3-TI-134.

Page 8, Section 2.1, change NETP-110 to NPG-SPP-09.26.1. Section 2.3, Add RIMS number for GL 89-13.

Page 9, Section 2.3, Add ADAMS number for 9/15/2014 TVA letter and 10/20/2015 TVA letter. Revise NOTE. Section 3.1 indicate None.

Page 10, Sections 3.2 and 4.0 indicate None.

Page 11, Subsection 6.1.2.E, add license condition. Section 6.1.3, Deleted subsections F, G, and H related to heat exchanger performance testing.

Page 12, Section 6.1.6, Add intake pump pit de-silting to the scope for diving services.

Page 14, Section 6.3, Revise NOTE to 1] state that the program has been revised to include monitoring of the RHR Heat Exchangers and 2] change the wording NFPA-805 licensing condition to NFPA-805 license condition.

Page 14&15, Sub-Sections 6.3.A, 6.3.B, and 6.3.C, general revision. Subsection 6.3.K, Deleted, U1/U2 Emergency Chillers removed from the GL 89-13 program.

Page 16, Section 6.4.C, general revision. , Clarification - The Control Bay Chillers for Units 1 and 2 are air cooled, plus add performance test PM numbers.

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 3 of 26 Table of Contents

1.0 INTRODUCTION

.......................................................................................................... 5 1.1 Purpose ........................................................................................................................ 5 1.2 Scope............................................................................................................................ 5

2.0 REFERENCES

............................................................................................................. 5 2.1 Performance References .............................................................................................. 5 2.2 Developmental References........................................................................................... 8 2.3 Commitments ................................................................................................................ 8 3.0 PRECAUTIONS AND LIMITATIONS ........................................................................... 9 3.1 Precautions ................................................................................................................... 9 3.2 Limitations................................................................................................................... 10 4.0 PREREQUISITE ACTIONS ........................................................................................ 10 5.0 ACCEPTANCE CRITERIA ......................................................................................... 10 6.0 PERFORMANCE ........................................................................................................ 10 6.1 Responsibilities ........................................................................................................... 10 6.1.1 Corporate GL 89-13 Program Owner ............................................................ 10 6.1.2 Site GL 89-13 Program Owner ..................................................................... 10 6.1.3 System Engineers ......................................................................................... 11 6.1.4 Chemistry ...................................................................................................... 11 6.1.5 Mechanical Maintenance .............................................................................. 11 6.1.6 Diving Services ............................................................................................. 12 6.1.7 Chemical Treatment Vendor ......................................................................... 12 6.2 GL 89-13 Implementing Action I - Biofouling............................................................... 12 6.3 GL 89-13 Implementing Action II- Heat Transfer Testing ............................................ 13 6.4 GL 89-13 Implementing Action III - Routine Inspection and Maintenance .................. 15 6.5 GL 89-13 Implementing Action IV - Single Failure Walkdown..................................... 16 6.6 GL 89-13 Implementing Action V - Procedure Review ................................................ 16 7.0 POST PERFORMANCE ACTIVITY ............................................................................ 16 8.0 RECORDS .................................................................................................................. 16 : Heat Exchanger Tube Plugging Calculations / Tubes Plugged ..................................................................................................... 17

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 4 of 26 Table of Contents (continued) : Heat Exchanger Performance Testing and Cleaning/Inspection PM Reference ......................................................... 19 : RHR Heat Exchanger Flow Monitoring PMs ........................................... 24 : SI - Pump Flow Test Cross-Reference Table .......................................... 25

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 5 of 26

1.0 INTRODUCTION

1.1 Purpose This procedure defines the program for Browns Ferry Nuclear Plant (BFN) to meet the requirements of NRC Generic Letter (GL) 89-13.

1.2 Scope The GL 89-13 program identifies procedures and documents that implement the actions required to assure the integrity and capability of the service water systems at BFN. GL 89-13 defines a service water system as the system or systems that transfer heat from safety-related structures, systems, or components to the Ultimate Heat Sink (UHS). These systems at BFN include Emergency Equipment Cooling Water (EECW) and Residual Heat Removal Service Water (RHRSW).

2.0 REFERENCES

2.1 Performance References 0-AOI-27-1, Component Biofouling 0-OI-23, Residual Heat Removal Service Water System 0-SI-3.2.4(DG A), EECW Check Valve Test on Diesel Generator A 0-SI-3.2.4(DG B), EECW Check Valve Test on Diesel Generator B 0-SI-3.2.4(DG C), EECW Check Valve Test on Diesel Generator C 0-SI-3.2.4(DG D), EECW Check Valve Test on Diesel Generator D 0-SI-4.5.C.1(4), EECW System Annual Flow Rate Test 1-SI-3.2.4(CS I), EECW Check Valve Test on Core Spray Division I 1-SI-3.2.4(CS II), EECW Check Valve Test on Core Spray Division II 1-SI-3.2.4(RHR I), EECW Check Valve Test on Residual Heat Removal Division I 1-SI-3.2.4(RHR II), EECW Check Valve Test on Residual Heat Removal Division II 2-SI-3.2.4(CS I), EECW Check Valve Test on Core Spray Division I 2-SI-3.2.4(CS II), EECW Check Valve Test on Core Spray Division II 2-SI-3.2.4(RHR I), EECW Check Valve Test on Residual Heat Removal Division I

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 6 of 26 2.1 Performance References (continued) 2-SI-3.2.4(RHR II), EECW Check Valve Test on Residual Heat Removal Division II 3-SI-3.2.4(CS I), EECW Check Valve Test on Core Spray Division I 3-SI-3.2.4(CS II), EECW Check Valve Test on Core Spray Division II 3-SI-3.2.4(RHR I), EECW Check Valve Test on Residual Heat Removal Division I 3-SI-3.2.4(RHR II), EECW Check Valve Test on Residual Heat Removal Division II 3-SI-3.2.4(ACU), EECW Check Valve Test on Unit 3 Electric Board Room ACUS 3-SI-3.2.4(CBC 3A), EECW Check Valve Test on Control Bay Chiller 3A 3-SI-3.2.4(CBC 3B), EECW Check Valve Test on Control Bay Chiller 3B 3-SI-3.2.4(SDBR), EECW Check Valve Test on SDBR Chillers 3-SI-3.2.4(DG A), EECW Check Valve Test on Diesel Generator A 3-SI-3.2.4(DG B), EECW Check Valve Test on Diesel Generator B 3-SI-3.2.4(DG C), EECW Check Valve Test on Diesel Generator C 3-SI-3.2.4(DG D), EECW Check Valve Test on Diesel Generator D 0-SI-4.5.C.1(A1), RHRSW Pump A1 IST Group A Quarterly Pump Test 0-SI-4.5.C.1(A2), RHRSW Pump A2 IST Group A Quarterly Pump Test 0-SI-4.5.C.1(B1), RHRSW Pump B1 IST Group A Quarterly Pump Test 0-SI-4.5.C.1(B2), RHRSW Pump B2 IST Group A Quarterly Pump Test 0-SI-4.5.C.1(C1), RHRSW Pump C1 IST Group A Quarterly Pump Test 0-SI-4.5.C.1(C2), RHRSW Pump C2 IST Group A Quarterly Pump Test 0-SI-4.5.C.1(D1), RHRSW Pump D1 IST Group A Quarterly Pump Test 0-SI-4.5.C.1(D2), RHRSW Pump D2 IST Group A Quarterly Pump Test 1-SI-4.5.C.1(A), RHR HX A Valves Quarterly IST Test 1-SI-4.5.C.1(B), RHR HX B Valves Quarterly IST Test 1-SI-4.5.C.1(C), RHR HX C Valves Quarterly IST Test 1-SI-4.5.C.1(D), RHR HX D Valves Quarterly IST Test

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 7 of 26 2.1 Performance References (continued) 2-SI-4.5.C.1(A), RHR HX A Valves Quarterly IST Test 2-SI-4.5.C.1(B), RHR HX B Valves Quarterly IST Test 2-SI-4.5.C.1(C), RHR HX C Valves Quarterly IST Test 2-SI-4.5.C.1(D), RHR HX D Valves Quarterly IST Test 3-SI-4.5.C.1(A), RHR HX A Valves Quarterly IST Test 3-SI-4.5.C.1(B), RHR HX B Valves Quarterly IST Test 3-SI-4.5.C.1(C), RHR HX C Valves Quarterly IST Test 3-SI-4.5.C.1(D), RHR HX D Valves Quarterly IST Test 0-TI-63, RHRSW Flow Blockage Monitoring 1-TI-134, Core Spray And Residual Heat Removal Room Coolers Air Flow Verification 2-TI-134, Core Spray And Residual Heat Removal Room Coolers Air Flow Verification 3-TI-134, Core Spray And Residual Heat Removal Room Coolers Air Flow Verification 0-TI-154, Coupons and Monitoring for Corrosion and Deposit Control 0-TI-322, RHR Heat Exchanger Performance Testing 0-TI-345 (EECW), EECW Pump Curve Data Acquisition 0-TI-345(RHRSW), RHRSW Pump Curve Data Acquisition 0-TPP-ENG-389, Raw Water Fouling and Corrosion Control 0-TI-594, Aging Management Program Basis Document, Closed-Cycle Cooling Water System Program 0-TI-616, Aging Management Program Basis Document, Open-Cycle Cooling Water System Program 0-TI-618, Aging Management Program Basis Document, Unit 1 Periodic Inspection Program

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 8 of 26 2.1 Performance References (continued)

CHTP-108, Technical Chemistry Standards for SPP-9.7 CI-137, Raw Water Chemical Treatment CI-137.5. Raw Water Chemical Treatment Molluscicide Control NETP-108, Heat Exchanger Testing and Maintenance Program NPG-SPP-09.26.1, Chiller and A/C Testing and Maintenance Program NPG-SPP-03.3, NRC Commitment Management NPG-SPP-06.2, Preventive Maintenance NPG-SPP-09.0.6, Conduct of Engineering Programs NPG-SPP-9.7, Corrosion Control Program NPG-SPP-09.7.1, Corrosion Control - General, Localized and Galvanic, and Stress Corrosion Control Program NPG-SPP-09.7.2, Flow Accelerated Corrosion Control Program NPG-SPP-09.7.3, Raw Water Corrosion Program NPG-SPP-09.7.4, Boric Acid Corrosion Control Program NPG-SPP-09.14, Generic Letter (GL) 89-13 Implementation NPF-SPP-09.16.1, System, Component, and Program Health 2.2 Developmental References EPRI 3002005340, Service Water Heat Exchanger Testing Guidelines, May 2015.

2.3 Commitments NRC Letter to TVA dated 9/29/1988 RIMS L44 880929 803 NRC Generic Letter 89-13 dated 7/18/1989 RIMS A02 890731 011 TVA letter to NRC dated 03/16/1990 RIMS L44 900316 801 TVA letter to NRC dated 12/31/1990 RIMS R08 901231 988 TVA letter to NRC dated 09/01/1993 RIMS R08 930901 932 TVA letter to NRC dated 08/17/1995 RIMS R08 950817 800

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 9 of 26 2.3 Commitments (continued)

[4/18/1995] Justification for Use of Alternative Action to GL 89-13 RIMS R92 950418 800 TVA letter to NRC dated 3/3/1997 RIMS R08 970303 970 TVA letter to NRC dated 3/2/1998 RIMS R08 980302 978 TVA letter to NRC dated 11/8/1999 RIMS R08 991108 666 TVA letter to NRC dated 10/19/2001 RIMS R08 011019 876 Letter from TVA to NRC dated 9/15/2014 Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3 - Response to Generic Letter (GL) 89 Correction to BFN Response to GL 89-13 Actions II and III. (ADAMS No. ML14260A364).

TVA Letter to NRC dated 10/20/2015 RIMS L44 151020 001 ADAMS ML15293A527 NCO040006041, License Renewal commitments Table 2, Item 16, Open Cycle Cooling Water System NCO040006083, License Renewal commitments Table 1, Item 15, Open Cycle Cooling Water System NCO040006092, License Renewal commitments Table 1, Item 16, Open Cycle Cooling Water System NCO040006050, License Renewal commitments Table 1, Item 21, Fire Water System Program FSAR Appendix O, Section O.1.16, Open-Cycle Cooling Water System Program NOTE This program has been revised to include monitoring of the RHR Heat Exchangers in order to satisfy a license condition described in section 3.9.3.6 of the Safety Evaluation for the transition to a Risk-Informed, Performance-Based Fire Protection Program (NFPA 805) in accordance with 10CFR50.48(c). Reference ML15212A796 3.0 PRECAUTIONS AND LIMITATIONS 3.1 Precautions None.

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 10 of 26 3.2 Limitations None.

4.0 PREREQUISITE ACTIONS None.

5.0 ACCEPTANCE CRITERIA Acceptance Criteria is addressed in individual testing or inspection procedures for each component.

6.0 PERFORMANCE 6.1 Responsibilities 6.1.1 Corporate GL 89-13 Program Owner A. Develop any necessary NPG Administrative procedure(s) for the implementation of this Program.

B. Provide governance and oversight that site procedures conform to GL 89-13 Requirements in accordance with NPG-SPP-01.4.

C. Provide technical guidance and implementation assistance to the site GL 89-13 Program Engineers.

D. Maintain current status with industry GL 89-13 issues and rule-making changes.

E. Support sites with any personnel training in GL 89-13 requirements as necessary.

F. Participate in relevant industry events, self assessments, and benchmarking.

6.1.2 Site GL 89-13 Program Owner A. Develop any necessary site Administrative procedure(s) for implementation of this Program.

B. Coordinate all GL 89-13 activities with the appropriate site organizations.

C. Coordinate with the owners of each site action to ensure that appropriate records of completed testing, inspection, and maintenance activities are retained as appropriate for the work implementing document.

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 11 of 26 6.1.2 Site GL 89-13 Program Owner (continued)

D. Provide technical guidance and implementation assistance to the site groups involved in GL 89-13 activities.

E. Ensure that affected site procedures are annotated for reference to GL 89-13 and, affected Preventative Maintenances (PMs) are annotated as regulatory[commitment or license condition] to prevent a procedure or PM from being changed without realizing it is a regulatory commitment or license condition.

F. Support site personnel training in GL 89-13 requirements as necessary.

G. Ensure site PMs that affect the GL 89-13 program are reviewed.

H. Maintain the Program Health Report as required.

I. Perform periodic self-assessment of the GL 89-13 program to ensure compliance to the Generic Letter as well as identification of industry best practices.

J. Refer to NPG-SPP-03.3, NRC Commitment Management, to revise any site commitments to GL-89-13.

6.1.3 System Engineers A. Test for clam and zebra veligers B. Monitor flow of RHR heat exchangers C. Monitor flow of EECW safety-related components D. Verify design flow of RHRSW and EECW pumps E. Inspect safety-related heat exchangers F. DELETED.

G. DELETED.

H. DELETED.

6.1.4 Chemistry A. Monitor Closed Cooling Water (CCW) systems B. Support corrosion coupon monitoring 6.1.5 Mechanical Maintenance A. Clean safety-related heat exchangers

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 12 of 26 6.1.6 Diving Services A. Inspect and de-silt, if necessary, RHRSW intake pump pit 6.1.7 Chemical Treatment Vendor A. Inject treatment chemicals into RHRSW sluice gates.

6.2 GL 89-13 Implementing Action I - Biofouling Action I of the GL 89-13 is paraphrased to read: For open cycle service water systems, implement and maintain an ongoing program of surveillance and control techniques to significantly reduce the incidence of flow blockage problems as a result of bio-fouling."

The specific BFN actions to implement this are listed below.

A. The RHRSW intake pump pit is to be inspected by divers, via the BFN PM Program. The PM provides for reassessing this inspection frequency based on the results of inspections and pump flow testing. Reference PM 500103065.

B. Bio-fouling, corrosion, and MIC within the EECW and RHRSW systems are controlled by chemical injection (as required) into the RHRSW sluice gates.

Biocides, dispersant, and inhibitors are added to these systems. Reference CI-137, Raw Water Chemical Treatment, and CI-137.5, Raw Water Chemical Treatment Molluscicide Control.

C. Testing for clam and zebra mussel veligers is performed when conditions are conducive to spawning. Reference CHTP-108 D. Flow monitoring of the RHR heat exchangers is performed in accordance with 0-TI-63, RHRSW Flow Blockage Monitoring. This activity is scheduled by the BFN PM Program. The PM provides for reassessing this monitoring frequency based on the results of monitoring. Reference PM 500108601, 500133228, 500116540, 500116541, 500126929, and 500126932.

E. Flow monitoring of EECW safety-related components is verified as part of the American Society of Mechanical Engineers Section OM Code requirements.

Reference 0-SI-3.2.4(DG A), 0-SI-3.2.4(DG B), 0-SI-3.2.4(DG C), 0-SI-3.2.4(DG D), 1-SI-3.2.4(CS I), 1-SI-3.2.4(CS II), 1-SI-3.2.4(RHR I), 1-SI-3.2.4(RHR II), 2-SI-3.2.4(CS I), 2-SI-3.2.4(CS II), 2-SI-3.2.4(RHR I), 2-SI-3.2.4(RHR II), 3-SI-3.2.4(CS I), 3-SI-3.2.4(CS II), 3-SI-3.2.4(RHR I), 3-SI-3.2.4(RHR II), 3-SI-3.2.4(ACU), 3-SI-3.2.4(CBC 3A), 3-SI-3.2.4(CBC 3B), 3-SI-3.2.4(SDBR), 3-SI-3.2.4(DG A), 3-SI-3.2.4(DG B), 3-SI-3.2.4(DG C), and 3-SI-3.2.4(DG D).

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 13 of 26 6.2 GL 89-13 Implementing Action I - Biofouling (continued)

F. The RHRSW and EECW pumps are tested to verify the design flow through normally assigned flow paths. Reference 0-SI-4.5.C.1(4), EECW System Annual Flow Rate Test and 0-SI-4.5.C.1(A1), 0-SI-4.5.C.1(A2), 0-SI-4.5.C.1(B1), 0-SI-4.5.C.1(B2), 0-SI-4.5.C.1(C1), 0-SI-4.5.C.1(C2), 0-SI-4.5.C.1(D1), and 0-SI-4.5.C.1(D2), RHRSW Pump and Header Operability and Flow Test and 1-SI-4.5.C.1(A), 1-SI-4.5.C.1(B), 1-SI-4.5.C.1(C), 1-SI-4.5.C.1(D), 2-SI-4.5.C.1(A), 2-SI-4.5.C.1(B), 2-SI-4.5.C.1(C), 2-SI-4.5.C.1(D), 3-SI-4.5.C.1(A), 3-SI-4.5.C.1(B), 3-SI-4.5.C.1(C), and 3-SI-4.5.C.1(D), RHR Heat Exchanger Flow Test.

G. Monitoring of available Browns Ferry chemistry reports to verify Closed Cooling Water (CCW) systems are maintained to appropriate Action Levels as specified in EPRI Report 1007820.

6.3 GL 89-13 Implementing Action II- Heat Transfer Testing Action II of the GL 89-13 is paraphrased to read: Conduct a test program to verify the heat transfer capability of all safety-related heat exchangers cooled by service water. Frequent regular maintenance of heat exchangers is acceptable in lieu of testing. "

BFN will maintain an inspection and cleaning program in accordance with the BFN PM Program to verify the heat transfer capability of the safety-related heat exchangers cooled by EECW and RHRSW. The basis for reasonable assurance of adequate heat transfer is based on the condition of the tubes as compared to the instructions and guidance in the inspection criteria found on the inspection form NPG-SPP-09.14-1 [TVA Form 41343] and by comparing the number of plugged, blocked, and partially blocked tubes with the calculated tube plugging limit found in Attachment 1of this procedure. The PMs provide for reassessing this inspection frequency based on the results of inspections, not to exceed five years.

Thermal performance testing was performed in 1994 and 2001 to validate the inspection and cleaning philosophy. The specific BFN actions to implement this for each heat exchanger are listed below. Criteria for inspections is detailed in 0-TPP-ENG-389 and NPG-SPP-09.7.3. See Attachment 2 of this procedure for PM references.

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 14 of 26 6.3 GL 89-13 Implementing Action II- Heat Transfer Testing (continued)

NOTE NFPA-805 License Condition Performance testing of the RHR heat exchangers satisfies a license condition described in section 3.9.3.6 of the Safety Evaluation for the transition to a Risk-Informed, Performance-Based Fire Protection Program (NFPA 805) in accordance with 10CFR50.48(c). The implementation of this requirement provides for periodic heat exchanger inspections and performance testing to ensure that the worst fouling resistance, with measurement uncertainty added, is less than 0.001517 hr-ft2-°F/BTU, and the worst tube plugging is less than 4.57% in all RHR heat exchangers.

Other program heat exchangers do not fall under this condition, i.e. inspection and cleaning of program heat exchangers other than the RHR heat exchangers is acceptable.

The NFPA 805 license condition is not a GL 89-13 program commitment, but is being met by revising the GL 89-13 program to perform routine performance testing on BFN RHR Heat Exchangers.

A. The RHR heat exchanger performance testing program will be maintained through the Browns Ferry Nuclear (BFN) Preventive Maintenance (PM) program. Each RHR heat exchanger will be tested periodically at an interval that initially will not exceed five years. The performance testing PMs provide criteria for reassessing the performance testing frequency based upon test results. Additionally, the heat exchangers will be cleaned on an 8-year frequency. This 8-year cleaning frequency is based on supporting eddy current testing, a PM requirement outside of the BFN GL 89-13 program. More frequent cleanings will occur if the fouling rate (as trended by TVA engineering) indicates the need to take corrective actions in order to maintain the heat exchanger condition within the fouling resistance acceptance criteria.

B. Testing will be conducted when seasonal conditions provide the most accurate test results related to river and Torus water differential temperatures. Heat exchangers will be tested in divisional pairs. Tests will be conducted coincident with RCIC quarterly surveillances in order to have a controlled heat addition source for the Torus. Testing intervals will be sequenced such that fouling rates can be assessed and fouling predictions can be made to ensure that heat exchanger fouling does not encroach on the limits specified in the analysis.

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 15 of 26 6.3 GL 89-13 Implementing Action II- Heat Transfer Testing (continued)

C. Additionally, the heat exchangers will be cleaned no less than once every 8 years. This 8-year cleaning frequency is based on supporting eddy current testing, a PM requirement outside of the BFN GL 89-13 program. More frequent cleanings will occur if heat exchanger conditions (as trended by TVA engineering) warrant such actions.

D. Performance testing is conducted in a manner consistent with the guidance described in EPRI 3002005340, Service Water Heat Exchanger Testing Guidelines.

E. Diesel Generator Coolers - Inspect and clean the cooling water side of each Diesel Generator cooler.

F. RHR Pump Seal Coolers - Inspect and clean the cooling water side of each RHR pump seal cooler.

G. Control Bay Chillers - Inspect and clean the cooling water side of each Control Bay Chiller.

H. Shutdown Board Room Chillers/Electric Board Room ACU - Inspect and clean the cooling water side of each Shutdown Board Room Chiller/Electric Board Room ACUs.

I. RHR Pump Room Coolers - chemically flush the cooling water side of each RHR Pump Room Cooler.

J. Core Spray Room Coolers - Chemically flush the cooling water side of each Core Spray Room cooler.

K. DELETED.

6.4 GL 89-13 Implementing Action III - Routine Inspection and Maintenance Action III of the GL 89-13 is paraphrased to read: Ensure by establishing a routine inspection and maintenance program for open-cycle service water system piping and components that corrosion, erosion, protective coating failure, silting, and bio-fouling cannot degrade the performance of the safety-related systems supplied by service water. "

A. Flow monitoring is performed on a periodic basis as described in Action I.

B. RHRSW and EECW systems receive inspection and cleaning as described in Action II.

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 16 of 26 6.4 GL 89-13 Implementing Action III - Routine Inspection and Maintenance (continued)

C. The RHR Heat exchangers have coated inner surfaces [waterboxes and floating heads] and are required to be inspected in order to verify coating conditions are acceptable. The RHR heat exchanger coatings inspections are performed under the PM program in accordance with 0-TPP-ENG-389 and NPG-SPP-09.14 using TVA form 41313.

D. Corrosion coupons are installed and monitored (0-TI-154).

6.5 GL 89-13 Implementing Action IV - Single Failure Walkdown Action IV of the GL 89-13 is paraphrased to read: Confirm that the service water system will perform its intended function in accordance with the licensing basis for the plant."

BFN response to this action is recorded in TVAs response to the NRC on RIMS document RIMS L44 900316 801.

6.6 GL 89-13 Implementing Action V - Procedure Review Action V of the GL 89-13 is paraphrased to read: Confirm that maintenance practices, operating and emergency procedure, and training that involve the service water system are adequate to ensure that the safety-related equipment cooled by the service water system will function as intended, and the operators of this equipment will perform effectively."

BFN response to this action is recorded in TVAs response to the NRC on RIMS document RIMS L44 900316 801.

7.0 POST PERFORMANCE ACTIVITY 8.0 RECORDS No records are generated by this procedure.

Appropriate records of completed testing, inspection, and maintenance activities will be retained as appropriate for the work implementing document.

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 17 of 26 Attachment 1 (Page 1 of 2)

Heat Exchanger Tube Plugging Calculations / Tubes Plugged Current Max tubes  % of tube Number of Calculation # or UNID HEAT EXCHANGER that can be plug limit plugged guidance document plugged reached tubes 0-HEX-082-000A1 DG-A COOLING WATER HEX A1 3 (0.2%) Young HEXs 17 tubes per MDQ0067880201 pass 0-HEX-082-000A2 DG-A COOLING WATER HEX A2 9 (1.2%) RIMS R1409019106 0-HEX-082-000B1 DG-B COOLING WATER HEX B1 0 (0.0%) 0.00% Wiegmann & Rose 38 tubes per HEXs MD-Q0082-0-HEX-082-000B2 DG-B COOLING WATER HEX B2 0 (0.0%) pass 000016 0.00%

RIMS R14001016138 0-HEX-082-000C1 DG-C COOLING WATER HEX C1 0 (0.0%) 0.00% Wiegmann & Rose 38 tubes per HEXs MD-Q0082-pass 000016 0-HEX-082-000C2 DG-C COOLING WATER HEX C2 0 (0.0%)

0.00%

RIMS R14001016138 0-HEX-082-000D1 DG-D COOLING WATER HEX D1 0 (0.0%) 0.00% Wiegmann & Rose 38 tubes per HEXs MD-Q0082-0-HEX-082-000D2 DG-D COOLING WATER HEX D2 0 (0.0%) pass 000016 0.00%

RIMS R14001016138 1-HEX-074-0900A RHR HEAT EXCHANGER 1A 27 (1.5%) 35.07%

1-HEX-074-0900B RHR HEAT EXCHANGER 1B 18 (1.1%) 23.38%

MD-Q0023-980143 77 (4.57%)

RIMS W78 030630 006 1-HEX-074-0900C RHR HEAT EXCHANGER 1C 23 (1.4%) 29.87%

1-HEX-074-0900D RHR HEAT EXCHANGER 1D 20 (1.2%) 25.97%

2-HEX-074-0900A RHR HEAT EXCHANGER 2A 15 (0.9%) 19.48%

2-HEX-074-0900B RHR HEAT EXCHANGER 2B 16 (0.9%) 20.78%

MD-Q0023-980143 77 (4.57%)

RIMS W78 030630 006 2-HEX-074-0900C RHR HEAT EXCHANGER 2C 17 (1.0%) 22.08%

2-HEX-074-0900D RHR HEAT EXCHANGER 2D 26 (1.5%) 33.77%

3-HEX-074-0900A RHR HEAT EXCHANGER 3A 11 (0.6%) 14.29%

3-HEX-074-0900B RHR HEAT EXCHANGER 3B 0 (0.0%) 0.00%

MD-Q0023-980143 77 (4.57%)

RIMS W78 030630 006 3-HEX-074-0900C RHR HEAT EXCHANGER 3C 12 (0.7%) 15.58%

3-HEX-074-0900D RHR HEAT EXCHANGER 3D 23 (1.2%) 29.87%

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 18 of 26 Attachment 1 (Page 2 of 2)

Heat Exchanger Tube Plugging Calculations / Tubes Plugged Current Max tubes  % of tube Number of Calculation # or UNID HEAT EXCHANGER that can be plug limit plugged guidance document plugged reached tubes 3-HEX-082-000A1 DG-A COOLING WATER HEX 3A1 0 (0.0%)

0.00%

3-HEX-082-000A2 DG-A COOLING WATER HEX 3A2 0 (0.0%)

3-HEX-082-000B1 DG-B COOLING WATER HEX 3B1 0 (0.0%)

0.00%

(new HEX inst. 6/2008) 3-HEX-082-000B2 DG-B COOLING WATER HEX 3B2 0 (0.0%)

0.00%

(new HEX inst. 6/2008) 3-HEX-082-000C1 DG-C COOLING WATER HEX 3C1 0 (0.0%) Wiegmann & Rose 0.00%

(new HEX inst. 6/2008) HEXs MD-Q0082-38 tubes per pass 000016 3-HEX-082-000C2 DG-C COOLING WATER HEX 3C2 0 (0.0%)

0.00% RIMS R14001016138 (new HEX inst. 6/2008) 3-HEX-082-000D1 DG-D COOLING WATER HEX 3D1 0 (0.0%)

0.00%

(new HEX inst. 7/2008) 3-HEX-082-000D2 DG-D COOLING WATER HEX 0 (0.0%)

0.00%

3D2 (new HEX inst. 7/2008)

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 19 of 26 Attachment 2 (Page 1 of 5)

Heat Exchanger Performance Testing and Cleaning/Inspection PM Reference RHR Heat Exchangers [I/C] RHR Pump Seal Coolers [I/C]

UNID PM number UNID PM number 1-HEX-074-0900A 500108600 1-HEX-074-0005 500108596 1-HEX-074-0900B 500133230 1-HEX-074-0016 500108597 1-HEX-074-0900C 500108605 1-HEX-074-0028 500108598 1-HEX-074-0900D 500108607 1-HEX-074-0039 500108599 2-HEX-074-0900A 500116539 2-HEX-074-0005 500116535 2-HEX-074-0900B 500116542 2-HEX-074-0016 500116536 2-HEX-074-0900C 500116544 2-HEX-074-0028 500116537 2-HEX-074-0900D 500116546 2-HEX-074-0039 500116538 3-HEX-074-0900A 500126928 3-HEX-074-0005 500126924 3-HEX-074-0900B 500126931 3-HEX-074-0016 500126925 3-HEX-074-0900C 500126933 3-HEX-074-0028 500126926 3-HEX-074-0900D 500126935 3-HEX-074-0039 500126927

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 20 of 26 Attachment 2 (Page 2 of 5)

Heat Exchanger Performance Testing and Cleaning/Inspection PM Reference RHR Room Coolers [I/C] Core Spray Room Coolers [I/C]

UNID PM number UNID PM number 1-CLR-064-0068 500136529 1-CLR-064-0072 500136530 1-CLR-064-0069 500136470 1-CLR-064-0073 500136531 1-CLR-064-0070 500136532 2-CLR-064-0072 500114619 1-CLR-064-0071 500136472 2-CLR-064-0073 500114621 2-CLR-064-0068 500114611 3-CLR-064-0072 500124957 2-CLR-064-0069 500114613 3-CLR-064-0073 500124959 2-CLR-064-0070 500114615 2-CLR-064-0071 500114617 3-CLR-064-0068 500124949 3-CLR-064-0069 500124951 3-CLR-064-0070 500124953 3-CLR-064-0071 500124955

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 21 of 26 Attachment 2 (Page 3 of 5)

Heat Exchanger Performance Testing and Cleaning/Inspection PM Reference Electric Board Room ACU [I/C] Control Bay Chillers [I/C]*

UNID PM number UNID PM number 3-ACU-031-7205 500133646 3-CHR-031-1943 500124854 3-ACU-031-7206 500133647 3-CHR-031-1951 500124859 Shutdown Board Room Chillers [I/C] U1/2 Emergency Chiller [I/C]

UNID PM number UNID PM number 3-CHR-031-0623 500124849 **

3-CHR-031-0635 500124850 3-CHR-031-0651 500124851 RHRSW Pump Pit [I/C]

3-CHR-031-0667 500124852 UNID PM number 0-MISC-023 500103065

• The Control Bay Chillers for Units 1 and 2 are air cooled.

• • U1/2 Emergency Chiller removed from GL 89-13 Program, Reference CR 831431-004.

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 22 of 26 Attachment 2 (Page 4 of 5)

Heat Exchanger Performance Testing and Cleaning/Inspection PM Reference H2O2 Analyzer Coolers [I/C] Diesel Generator Coolers [I/C]

UNID PM number UNID PM number

• 0-HEX-082-000A1 500102570 0-HEX-082-000A2

• 0-HEX-082-000B1 500102572 0-HEX-082-000B2

• 0-HEX-082-000C1 500102574 0-HEX-082-000C2

• 0-HEX-082-000D1 500102576 0-HEX-082-000D2

• 3-HEX-082-000A1 500126940 3-HEX-082-000A2 3-HEX-082-000B1 500126942 3-HEX-082-000B2 3-HEX-082-000C1 500126944 3-HEX-082-000C2 3-HEX-082-000D1 500126946 3-HEX-082-000D2

• Hydrogen Oxygen Analyzers are no longer cooled by raw water following U2R16 (DCN 69446, March 2011).

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 23 of 26 Attachment 2 (Page 5 of 5)

Heat Exchanger Performance Testing and Cleaning/Inspection PM Reference RHR Heat Exchanger Performance Testing - EPU/NFPA-805 UNID PM number 1-HEX-074-0900A / 1-HEX-074-0900C 94733 1-HEX-074-0900B / 1-HEX-074-0900D 94808 2-HEX-074-0900A / 2-HEX-074-0900C 94809 2-HEX-074-0900B / 2-HEX-074-0900D 94810 3-HEX-074-0900A / 3-HEX-074-0900C 94812 3-HEX-074-0900B / 3-HEX-074-0900D 94815

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 24 of 26 Attachment 3 (Page 1 of 1)

RHR Heat Exchanger Flow Monitoring PMs Date ________

RHR Heat Exchanger UNID PM 1-HEX-74-0900 A & C 500108601 1-HEX-74-0900 B & D 500133228 2-HEX-74-0900 A & C 500116540 2-HEX-74-0900 B & D 500116541 3-HEX-74-0900 A & C 500126929 3-HEX-74-0900 B & D. 500126932

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 25 of 26 Attachment 4 (Page 1 of 2)

SI - Pump Flow Test Cross-Reference Table Date ________

Unit SI Number Description 0 0-SI-3.2.4(DG A) EECW Check Valve Test on Diesel Generator A 0 0-SI-3.2.4(DG B) EECW Check Valve Test on Diesel Generator B 0 0-SI-3.2.4(DG C) EECW Check Valve Test on Diesel Generator C 0 0-SI-3.2.4(DG D) EECW Check Valve Test on Diesel Generator D 0 0-SI-4.5.C.1(4) EECW System Annual Flow Rate Test 1 1-SI-3.2.4(CS I) EECW Check Valve Test on Core Spray Division I 1 1-SI-3.2.4(CS II) EECW Check Valve Test on Core Spray Division II 1 1-SI-3.2.4(RHR I) EECW Check Valve Test on Residual Heat Removal Division I 1 1-SI-3.2.4(RHR II) EECW Check Valve Test on Residual Heat Removal Division II 2 2-SI-3.2.4(CS I) EECW Check Valve Test on Core Spray Division I 2 2-SI-3.2.4(CS II) EECW Check Valve Test on Core Spray Division II 2 2-SI-3.2.4(RHR I) EECW Check Valve Test on Residual Heat Removal Division I 2 2-SI-3.2.4(RHR II) EECW Check Valve Test on Residual Heat Removal Division II 3 3-SI-3.2.4(CS I) EECW Check Valve Test on Core Spray Division I 3 3-SI-3.2.4(CS II) EECW Check Valve Test on Core Spray Division II 3 3-SI-3.2.4(RHR I) EECW Check Valve Test on Residual Heat Removal Division I 3 3-SI-3.2.4(RHR II) EECW Check Valve Test on Residual Heat Removal Division II 3 3-SI-3.2.4(ACU) EECW Check Valve Test on Unit 3 Electric Board Room ACUS 3 3-SI-3.2.4(CBC 3A) EECW Check Valve Test on Control Bay Chiller 3A 3 3-SI-3.2.4(CBC 3B) EECW Check Valve Test on Control Bay Chiller 3B 3 3-SI-3.2.4(SDBR) EECW Check Valve Test on SDBR Chillers 3 3-SI-3.2.4(DG A) EECW Check Valve Test on Diesel Generator A

BFN Program for Implementing NRC 0-TI-522 Unit 0 Generic Letter 89-13 Rev. 0007 Page 26 of 26 Attachment 4 (Page 2 of 2)

SI - Pump Flow Test Cross-Reference Table 3 3-SI-3.2.4(DG B) EECW Check Valve Test on Diesel Generator B 3 3-SI-3.2.4(DG C) EECW Check Valve Test on Diesel Generator C 3 3-SI-3.2.4(DG D) EECW Check Valve Test on Diesel Generator D 0 0-SI-4.5.C.1(A1) RHRSW Pump A1 IST Group A Quarterly Pump Test 0 0-SI-4.5.C.1(A2) RHRSW Pump A2 IST Group A Quarterly Pump Test 0 0-SI-4.5.C.1(B1) RHRSW Pump B1 IST Group A Quarterly Pump Test 0 0-SI-4.5.C.1(B2) RHRSW Pump B2 IST Group A Quarterly Pump Test 0 0-SI-4.5.C.1(C1) RHRSW Pump C1 IST Group A Quarterly Pump Test 0 0-SI-4.5.C.1(C2) RHRSW Pump C2 IST Group A Quarterly Pump Test 0 0-SI-4.5.C.1(D1) RHRSW Pump D1 IST Group A Quarterly Pump Test 0 0-SI-4.5.C.1(D2) RHRSW Pump D2 IST Group A Quarterly Pump Test 1 1-SI-4.5.C.1(A) RHR HX A Valves Quarterly IST Test 1 1-SI-4.5.C.1(B) RHR HX B Valves Quarterly IST Test 1 1-SI-4.5.C.1(C) RHR HX C Valves Quarterly IST Test 1 1-SI-4.5.C.1(D) RHR HX D Valves Quarterly IST Test 2 2-SI-4.5.C.1(A) RHR HX A Valves Quarterly IST Test 2 2-SI-4.5.C.1(B) RHR HX B Valves Quarterly IST Test 2 2-SI-4.5.C.1(C) RHR HX C Valves Quarterly IST Test 2 2-SI-4.5.C.1(D) RHR HX D Valves Quarterly IST Test 3 3-SI-4.5.C.1(A) RHR HX A Valves Quarterly IST Test 3 3-SI-4.5.C.1(B) RHR HX B Valves Quarterly IST Test 3 3-SI-4.5.C.1(C) RHR HX C Valves Quarterly IST Test 3 3-SI-4.5.C.1(D) RHR HX D Valves Quarterly IST Test

ENCLOSURE 1 SCVB-RAI 8 Section 2.6.5.2 of PUSAR fourth and fifth paragraphs under heading "Large Break LOCA Short-Term Phase ECCS NPSH [Net Positive Suction Head]," states:

As stated above, the actual delivered pump flow rate will be between the safety analysis flow rate (where there is positive NPSH margin) and the pump run-out flow rate (where there may be negative NPSH margin). Because it is assumed that the operators take actions to control the RHR and CS pumps at ten minutes (see following subsection for evaluation of Large Break LOCA Long-Term Phase ECCS NPSH), this condition, should it occur, would exist for no more than ten minutes.

During this ten minute period it is prudent to address two aspects of pump operation at these conditions: (1) whether the pump(s) could actually be operating with less than NPSHr [NPSH required]3%; and (2) whether the pump(s) could sustain any damage during this ten minute period. The SECY-11-0014 guidance (Reference 97) addresses these two concerns and the Boiling Water Reactors Owners Group (BWROG) provided in-depth assessments in two BWROG reports for the Browns Ferry RHR pumps: "Pump Operation at Reduced NPSHa [NPSH available] Conditions" (Reference 100) and "BWROG CVIC Report Task 4, Operation in Maximum Erosion Rate Zone" (Reference 101).

The above paragraphs gives the impression that the NPSHa is less than NPSHr3% (i.e., pumps may be operating with a negative NPSH margin during the short-term). This is in conflict with data in Tables 2.6-4a and the graphs presented in Figures 2.6-11a and 2.6-12a, which show that both RHR and CS pumps have positive margin with respect to effective required NPSH (NPSHreff). In case the above tables and the figures are correct, it is not clear why the assessment in the BWR Owners' Group report "Containment Accident Pressure Committee (344), Task 3 - Pump Operation at Reduced NPSHa [available NPSH] Conditions (Sulzer Model CVIC Pump)," BWROG-TP-13-009, Revision 0, June 2013 (ADAMS Accession No. ML14077A097), is used to justify pump operation with a negative NPSH margin during the short-term. Confirm that in the short-term, with runout flow rates, the RHR and CS pumps operate with positive NPSH margin with respect to NPSHreff.

TVA Response:

During the short-term operating period (initial ten minutes), credited pumps, defined as those performing the safety function of delivering flow to the reactor core, would be operating with positive net positive suction head (NPSH) margin (NPSHa > NPSHeff).

E1-12

ENCLOSURE 1 SCVB-RAI 14 Refer to Section 2.6.2 of PUSAR regarding the postulated break in a 4-inch jet pump instrument line nozzle at EPU conditions, and UFSAR Section 12.2.2.6.

UFSAR Section 12.2.2.6, eighth paragraph describes the effect of jet forces resulting from the 4-inch break line as follows:

An analysis has also been performed of the effects of jet forces resulting from a double-ended break of the 4-inch line, assuming the jet forces from the break were to impinge directly on the removable plug. The resulting load would be 11 kips [kilo pounds], which is less than the capability of the locking bars and hinges, the capability of the shield wall, and the capability of the reactor vessel and its support skirt.

Provide a description of the analysis and results of the effect of the jet forces on the removable plug from a double-ended break of the 4-inch line under the most limiting thermal-hydraulic conditions in the reactor at EPU conditions. Justify that the reactor conditions (such as fluid density, jet velocity, etc.) assumed in the analysis are most limiting. What is the load capability of the locking bars and hinges, the capability of the shield wall, and the capability of the reactor vessel and its support skirt? Confirm these load capabilities bound the loads resulting from the jet forces from the double-ended break of the 4-inch line under EPU conditions.

TVA Response:

The calculation that formed the basis for the current UFSAR Section 12.2.2.6 evaluation concerning the effect of the jet impingement loading from the 4-inch jet pump (JP) instrument line break was not located. The current UFSAR Section 12.2.2.6 discussion concerning the jet impingement loading for the 4-inch JP instrument line break was the incorporation of the TVA response to Atomic Energy Commission question 12.18 during the original licensing of BFN.

TVA performed an evaluation for the assumed instantaneous double-ended break of the 4-inch JP instrument nozzle. This included recreation of the jet force loads at various reactor conditions consistent with the conditions shown in Section 2.6.2 of EPU LAR Attachment 6, Power Uprate Safety Analysis Report (PUSAR). The evaluation included an additional condition for 102% of the BFN original licensed thermal power of 3293 Megawatts-thermal (MWt). These conditions are shown in Table SCVB-RAI 14-1.

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ENCLOSURE 1 Table SCVB-RAI 14-1 Parameter Unit 102% 102% 102% 102% 102% 55.4%

OLTP CLTP CLTP EPU EPU EPU Rated Rated MELLLA Rated Rated MELLLA Flow Flow NFWT Flow Flow Line NFWT NFWT [2] NFWT RFWT RFWT

[2] [2] [2] [2] [1]

Core Power MWt 3359 3527 3527 4031 4031 2189 Core Flow Mlbm/hr 102.5 102.5 83.0 102.5 102.5 38.2 Reactor psia 1023 1053 1053 1070 1070 991 Dome Pressure Downcomer BTU/lbm 520.4 523.6 516.4 525.4 518.7 484.8 Enthalpy Notes:

(1) 55.4% LPU condition is 102% of the power level at the intersection of the Maximum Extended Load Line Limit Analysis (MELLLA) line and minimum pump speed line (1.02

• 54.3%).

(2) OLTP = Original Licensed Thermal Power = 3293 MWt CLTP = Current Licensed Thermal Power = 3458 MWT EPU = Extended Power Uprate Power Level = 3952 MWt NFWT = Analysis at normal feedwater temperature RFWT = Analysis with reduced feedwater temperature The jet impingement force was calculated using Equation D-2 of Appendix D of ANSI/American Nuclear Society (ANS) 58.2-1988 (Reference 1).

Fj = Ae (CT*Po - Pamb) where:

Fj = Jet impingement force (pound force (lbf)),

Ae = Break area (square inches),

CT = Thrust coefficient - based on method described in Appendix B, Section (c) of Reference 1 (Henry-Fauske model for subcooled water),

Po = Stagnation pressure at break (reactor dome pressure plus water head above break) (psia)

Pamb = Ambient pressure (psia) - for the break analysis a constant ambient pressure of 14.7 psia inside the sacrificial shield wall is assumed.

The effect of RFWT is to lower the enthalpy in the reactor pressure vessel downcomer region.

This results in the determination of a higher thrust coefficient and higher calculated jet impingement force. The summary of the jet impingement force calculations for the various conditions are shown in Table SCVB-RAI 14-2.

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ENCLOSURE 1 Table SCVB-RAI 14-2 102% 102% 102% 102% 55.4%

102%

OLTP CLTP EPU EPU EPU CLTP Parameter Unit Rated Rated Rated Rated MELLLA MELLLA Flow Flow Flow Flow Line NFWT NFWT NFWT NFWT RFWT RFWT Core Power MWt 3359 3527 3527 4031 4031 2189 Core Flow Mlbm/hr 102.5 102.5 83.0 102.5 102.5 38.2 Reactor Dome psia 1023 1053 1053 1070 1070 991 Pressure Downcomer BTU 520.4 523.6 516.4 525.4 518.7 484.8 Enthalpy lbm Stagnation psia 1034 1064 1064 1081 1081 1002 Pressure -Po Thrust NA 1.276 1.279 1.302 1.28 1.302 1.386 Coefficient -CT Jet Impingement lbf 11801 12177 12406 12390 12605 12433 Force - Fj Figure SCVB-RAI 14-1 shows the dimensions of the sacrificial shield wall door for the JP instrument nozzles. Figure SCVB-RAI 14-2 shows the dimensions of the JP instrument nozzle, safe end and seal penetration. Observation of the two figures shows that the opening in the shield wall door has a diameter of approximately two feet and the diameter of the JP instrument nozzle safe end is approximately four inches in diameter. The distance between the JP instrument nozzle safe end weld and the shield door shielding inside surface is approximately 8.5 inches. Therefore, it is possible that the jet from the JP instrument nozzle safe end break would not impinge on the shielding or the door assembly if the jet does not expand sufficiently to have the outer edge of jet impinge of the shielding or door assembly. The jet area was calculated in order to determine the amount of force applied to the shielding and shield door assemblies. The calculation was performed using the methodology contained in Appendix C of Reference 1 for a circumferential break with full separation. The results of the calculation show that the jet radius for all break statepoints (See Table SCVB-RAI 14-2 for statepoints) is less than distance from the JP nozzle centerline to the edge of the hole in the shield door.

Consequently, the jet would not impinge upon the shielding or the door assembly and the jet impingement force can be assumed to be zero.

It was conservatively assumed that the jet expands such that it touches the inside surface of the hole in the shielding and door and a skin friction drag load acts on the door assembly. The drag force was calculated using a conservative skin friction drag coefficient, Cd =0.1, which is taken as a factor of 10 larger than the largest drag coefficient for a flat plate. The drag load on the shield doors for each condition is shown in Table SCVB-RAI 14-3.

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ENCLOSURE 1 Table SCVB-RAI 14-3 102% 102% 102% 102% 55.4%

102%

OLTP CLTP EPU EPU EPU CLTP Parameter Unit Rated Rated Rated Rated MELLLA MELLLA Flow Flow Flow Flow Line NFWT NFWT NFWT NFWT RFWT RFWT Core MWt 3359 3527 3527 4031 4031 2189 Power Drag Load on shield lbf 607 625 625 635 635 588 door The structural evaluation of the shield door was performed using the maximum jet drag load of 635 lbf shown in Table SCVB-RAI 14-3 concurrent with the deadweight load of the doors and shielding. The shield door including the latch, hinges and pins are constructed from SA-283 Grade B steel. The material properties for SA-283 Grade B at the maximum average drywell temperature of 150°F are as follows.

E (Elastic modulus): 29x106 psi Sy (Yield Strength): 25.4 ksi Su (Tensile Strength): 50 ksi Maximum Elongation: 0.28 Density: 0.280 lbf/in3 Poissons Ratio: 0.3 Flow Stress: (Sy + Su)/2 = 37700 psi The shield door area from Figure SCVB-RAI 14-1 is 1668.9 square inches. The maximum jet force is applied in the structural analysis as a static pressure load of 635/1668.9 = 0.38 psi.

The acceptance criteria for the shield wall doors is that the doors do not fail and become missiles in the primary containment. The structural evaluation acceptance criteria is that the von Mises total strain (elastic + plastic) is less than the maximum elongation of 28% specified in the ASME Boiler and Pressure Code for SA-283, Grade B.

The results of the structural evaluation show that the total von Mises strain for the shield door latch is 0.001634 and the total von Mises strain for the hinge and pin assembly is 0.000841, which is below the maximum elongation value of 0.28. Therefore, the door can withstand the jet impingement load for the double ended break of the 4-inch JP instrument line at EPU conditions.

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ENCLOSURE 1 Evaluation of Shield Wall, Capability of the Reactor Vessel and the Reactor Vessel Support Skirt The yield point of the ASTM A36 structural members in the shield wall is 36 ksi. The load capability of the shield wall is a minimum of 19 psi. TVA confirms these load capabilities bound the loads resulting from the jet forces from the double-ended break of the 4-inch JP instrument line under EPU conditions.

Figure SCVB-RAI 14-3 shows the orientation of the JP instrument nozzle with respect to the recirculation system inlet and outlet nozzles. Observation of Figure SCVB-RAI 14-3 shows that the JP instrument nozzle is located at an elevation below that of the recirculation inlet and outlet nozzles. The 4-inch JP nozzle is significantly smaller than the recirculation inlet and outlet nozzles. Therefore, any forces and moments on the Reactor Pressure Vessel (RPV) and the RPV support skirt from a break of the JP instrument nozzle will be bounded by the forces and moments developed from a break in the recirculation suction or discharge nozzle. The bounding jet forces used in the BFN EPU structural evaluation of the RPV are from the recirculation outlet nozzle. Table 2.2-6 of the BFN EPU LAR Attachment 6 shows that the ASME code limit stress for the RPV support skirt is 80.1 ksi and the ASME code limit stress for the RPV is 80 ksi. Since the EPU stresses reported in Table 2.2-6 of EPU LAR Attachment 6 are below the ASME code limit, the loading from a double-ended break of the 4-inch JP instrument nozzle under EPU conditions will not result in an overload condition for the RPV support skirt or the reactor vessel. TVA confirms that the load capabilities of the RPV and RPV support skirt bound the loads resulting from the jet forces from the double-ended break of the 4-inch JP instrument line under EPU conditions.

Reference

1. ANSI/ANS 58.2 - 1988, American National Standard Design Basis for Protection of Light Water Nuclear Power Plants Against the Effects of Postulated Pipe Rupture.

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ENCLOSURE 1 Figure SCVB-RAI 14-1: Sacrificial Shield Wall Door Dimensions for JP Instrument Nozzles E1-18

ENCLOSURE 1 Figure SCVB-RAI 14-2: JP Instrument Nozzle and Safe End E1-19

ENCLOSURE 1 Figure SCVB-RAI 14-3: Orientation of BFN RPV Nozzles E1-20

ENCLOSURE 1 SCVB-RAI 24 Section 2.5.3.2 of PUSAR under heading "Browns Ferry Current Licensing Basis" states:

Browns Ferry's current licensing basis regarding GL 89-13 is discussed in TVA's response to the NRC by letter dated March 16, 1990, "Response to Generic Letter 89-13 Service Water Problems Affecting Safety-Related Equipment.

In the TVA response letter to GL 89-13, dated March 16, 1990, the response to NRC recommended action II states:

Inspect and clean the cooling water side of the RHR heat exchangers at least once per cycle.

In Reference 7, response to SCVB-RAI 1 (b), under title "Frequency of Monitoring -

Performance Testing," states:

Each RHR heat exchanger will have been performance tested at least once and will be tested periodically at an interval that initially will not exceed five years.

In Reference 7, response to SCVB-RAI 1 (b), under title "Frequency of Monitoring - Visual Inspection and Cleaning" for EPU implementation states:

Each RHR heat exchanger will be cleaned once every 8-years or more frequently if the trended fouling rate indicates the need to take corrective actions in order to maintain the heat exchanger condition within the fouling resistance acceptance criteria.

As stated above, the inspection and cleaning interval of the RHR heat exchangers in TVA's response to GL 89-13 was at least once every cycle (i.e., at least 2 years) as committed in TVA letter dated March 16, 1990. Currently the inspection and cleaning is interval is 4 years. The inspection and cleaning interval after EPU implementations is being proposed to increase to 8 years.

a. Provide the basis for increasing the inspection and cleaning interval from 2 years to the current interval of 4-years, and

b. Provide the basis for the proposed increase in interval from 4 years to 8 years after EPU implementation. Justify the increase in intervals based on the trending history of the previous as-found inspection and cleanliness condition results of the RHR heat exchangers, such as the as-found number of partial and wholly plugged tubes, and/or any other parameter that reflected the cleanliness of the as-found heat exchanger.

TVA Response:

a. The basis for increasing the inspection and cleaning interval for the RHR heat exchangers as provided in the commitment change described in Enclosure 5 of Reference 1 was as follows.

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ENCLOSURE 1 A Chemical Treatment Program was put into service in September of 1996. The chemicals are dispersed throughout the RHRSW System to control corrosion, clam infestation, and water fouling. Inspections conducted, after implementation of the Chemical Treatment Program and prior to submittal of the commitment change described in Reference 1, showed a relatively small amount of debris and indicated that the heat transfer surfaces were in good condition. Based on the Chemical Treatment Program and the previous inspection results, TVA revised the inspection and cleaning commitment, as described in the Reference 1 letter, to the following, "Inspect and clean the cooling water side of these heat exchangers periodically as determined by the preventive maintenance (PM) program." Four years was chosen to be the new inspection and cleaning frequency for the RHR heat exchangers at that time based on the PM program results.

Upon implementation of EPU, the frequency of inspection and cleaning will be as described in the response to SCVB-RAI 24b. below.

b. The inspection and cleaning frequency will be determined by a Residual Heat Removal (RHR) Heat Exchanger Performance Monitoring Program being developed for implementation following EPU and will be maintained through the BFN PM Program.

Each RHR heat exchanger has been tested at least once prior to June 16, 2016.

Subsequent performance testing of each RHR heat exchanger will be performed periodically at an interval of nominally four years, not to exceed five years. The performance testing acceptance criteria will be used to demonstrate margin between the actual heat removal capability assumed in the containment analyses because it assumes more heat exchanger tubes are plugged in a given heat exchanger.

Performance test results, in conjunction with trended fouling rate, will be used to determine if the margin to the acceptance criteria will be exceeded before the next scheduled inspection and cleaning. The heat exchangers will be cleaned on a maximum frequency of eight years. This 8-year cleaning frequency is based on supporting PM-required eddy current testing of the RHR heat exchanger tubes. More frequent RHR heat exchanger inspection and cleaning will occur if the fouling rate, as trended, indicates the need to take corrective actions in order to maintain the heat exchanger condition within the fouling resistance acceptance criteria. These requirements will be included in the Updated Final Safety Analysis Report as described in the response to SCVB-RAI 26.

Reference

1. Letter from TVA to NRC, "Browns Ferry Nuclear Plant (BFN) - Units 1, 2, and 3 -

Summary Report for January 1, 1998, through May 31, 1999," dated November 8, 1999 (ML993270093)

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ENCLOSURE 1 SCVB-RAI 25 In Reference 4, the implementation Item 49, which is a license condition states:

Revise the program that monitors BFN Residual Heat Removal (RHR) heat exchanger performance for consistency with the assumptions of the NFPA

[National Fire Protection Association] 805 Net Positive Suction Head (NPSH) analysis. The monitoring program shall include verification that the tested worst fouling resistance, with measurement uncertainty added, of all BFN Units 1, 2, and 3 RHR heat exchangers is less than the design value of 0.001517 hr-ft2 -

°F/BTU and the worst tube plugging is less than 4.57 percent.

Provide the proposed revision to the above licensee condition for EPU implementation including the frequency of cleaning and testing of the RHR heat exchangers.

TVA Response:

The NFPA 805 license condition associated with implementation item 49 is proposed to be deleted. The RHR Heat Exchanger Performance Monitoring Program has already been established as required by the NFPA 805 license condition. These NFPA 805 licensing condition requirements have been incorporated into the applicable implementing procedures.

Copies of these implementing procedures, 0-TI-322, RHR Heat Exchanger Performance Testing, and 0-TI-522, Program for Implementing NRC Generic Letter 89-13, are provided as part of the revised response to SCVB-RAI 1, included in this EPU LAR supplement.

In order to support deletion of the NFPA 805 license condition associated with implementation item 49, it is proposed to add RHR Heat Exchanger Performance Monitoring Program requirements to the Administrative Controls section of the BFN Technical Specifications (TS).

Specifically, a new TS 5.5.14, Residual Heat Removal (RHR) Heat Exchanger Performance Monitoring Program, is proposed to be added. This TS would require the establishment of a program to ensure the RHR heat exchangers are maintained in a condition that meets or exceeds the minimum performance capability assumed in the EPU containment analyses, which support not taking credit for containment accident pressure in the NPSH analyses. The TS would require RHR heat exchanger performance testing and overall uncertainty in the fouling resistance to be performed in accordance with the guidelines in the EPRI report, EPRI 3002005340, Service Water Heat Exchanger Guidelines, dated May 2015. The TS requires the program to include the following.

a. Provisions for periodically monitoring RHR heat exchanger performance, including frequency of monitoring and methodology for considering uncertainty of the result.

b. Acceptance criteria for RHR heat exchanger worst fouling resistance and number of plugged tubes.

c. Limitations and compensatory actions if degraded performance is observed.

d. Controls for changes to program requirements.

The TS would require the details of the program to be described in the UFSAR.

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ENCLOSURE 1 Placing the RHR Heat Exchanger Performance Monitoring Program in the BFN TS, with detail of the program included in the UFSAR, provides assurance BFN RHR heat exchanger performance will be maintained consistent with BFN analysis and licensing bases. Changes to the BFN analysis and licensing bases associated with the RHR Heat Exchanger Performance Monitoring Program details included in the UFSAR will be controlled in accordance with 10 CFR 50.59, Changes, tests, and experiments. The proposed program details to be included in the UFSAR are provided in the response to SCVB-RAI 26.

Including the fouling resistance and tube plugging acceptance criteria in the UFSAR enables BFN to address the impact of potential heat exchanger degraded conditions, associated fouling resistance or tube plugging, on past operability/functionality within the TVA CAP. For example, basing thermal performance evaluations supporting past operability/functionality could be based on actual plant conditions that existed in the past (e.g., number of tubes plugged). The current NFPA 805 license condition, with explicit limits, does not facilitate this type of past thermal performance evaluation.

The proposed Technical Specification 5.5.14 wording is as follows.

5.5.14 Residual Heat Removal (RHR) Heat Exchanger Performance Monitoring Program This program is established to ensure that the RHR heat exchangers are maintained in a condition that meets or exceeds the minimum performance capability assumed in containment analyses, which support not taking credit for containment accident pressure in the NPSH analyses. The RHR heat exchanger testing and determination of overall uncertainty in the fouling resistance shall be in accordance with the guidelines in EPRI report, EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015. This program establishes the following attributes.

a. The program establishes provisions to periodically monitor RHR heat exchanger thermal performance. The program includes frequency of monitoring and the methodology considers uncertainty of the result.

b. The program establishes and controls acceptance criteria for RHR heat exchanger worst fouling resistance and number of plugged tubes.

c. The program establishes limitations and allows for compensatory actions if degraded performance is observed.

d. Changes to the program shall be made under appropriate administrative review.

e. Details of the program including program limitations, compensatory actions for degraded performance, testing method, data acquisition method, data reduction method, overall uncertainty determination method, thermal performance analysis, acceptance criteria, and computer programs used that meet the 10 CFR 50 Appendix B, and 10 CFR 21 requirements are described in the UFSAR.

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ENCLOSURE 1 SCVB-RAI 26 Provide the proposed revision to the GL 89-13 response applicable to the RHR heat exchanger for EPU implementation.

TVA Response:

The Browns Ferry Nuclear Plant (BFN) maintains a residual heat removal (RHR) Heat Exchanger Performance Monitoring Program designed to ensure the heat transfer capability of the RHR heat exchangers meet or exceed the minimum performance capability assumed in containment analyses, which support not taking credit for containment accident pressure in the NPSH analyses. The scope of the RHR Heat Exchanger Performance Monitoring Program is limited to the RHR heat exchangers. The program is implemented by the associated Generic Letter (GL) 89-13 program procedure. Requirements of the GL 89-13 program procedure are implemented through the BFN preventive maintenance (PM) program.

Upon EPU implementation, the RHR Heat Exchanger Performance Monitoring Program is proposed to be included in the BFN Technical Specifications. RHR Heat Exchanger Performance Monitoring Program requirements will be described in the UFSAR. The changes to the RHR Heat Exchanger Performance Monitoring Program for EPU are summarized below and are reflective of the revised EPU LAR Attachment 39. The following details will be reflected in the UFSAR.

1. The EPU DBA-LOCA minimum required heat removal rate is 80,136,000 Btu/hour (hr) per heat exchanger with two heat exchangers in service. The EPU fire event minimum required heat removal rate is 124,966,800 Btu/hr with one heat exchanger in service.

2. The EPU thermal performance test acceptance criteria for an RHR heat exchanger is 0.001562 hr-ft2-F/Btu with no more than 77 tubes (4.57% of 1700 tubes) mechanically plugged. The test acceptance criteria is used to demonstrate margin between actual heat removal capability and minimum required heat removal capability assumed in the containment analyses because it assumes more tubes plugged than the actual number of tubes plugged in the given heat exchanger. Performance test results in conjunction with trended fouling rate, is used to determine if the margin to the acceptance criteria will be exceeded before the next scheduled inspection and cleaning.

3. The nominal (measured) test result (fouling resistance) including the test and measurement uncertainty will be used for comparison to the acceptance criteria.

4. The program includes the following requirements:

a. Each RHR heat exchanger was performance tested at least once prior to the NFPA 805 implementation date, June 16, 2016.

b. Upon EPU implementation, each RHR heat exchanger will be performance tested at a nominal interval of four years, not to exceed five years.

c. The RHR heat exchangers will be periodically cleaned at an interval that will not exceed 8 years. This 8-year cleaning frequency is based on supporting PM-required eddy current testing of the RHR heat exchanger tubes. More frequent E1-25

ENCLOSURE 1 RHR heat exchanger inspection and cleaning will occur if the fouling rate (as trended) indicates the need to take corrective actions in order to maintain the heat exchanger condition within the fouling resistance acceptance criteria.

5. Thermal performance testing results, including uncertainty, is trended consistent with ASME OM 2015, Part 21, Section 6.10, to facilitate determining fouling rate. BFN takes exception to the second paragraph of Section 6.10.2 in that the test program currently compares the test fouling resistance results to the acceptance criteria. BFN did not trend the fouling resistance for a minimum of three tests or monitoring points prior to comparison to the acceptance criteria.

6. Temporary surface mounted temperature instrumentation for RHR and RHRSW inlet and outlet piping meets the guidance provided in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

7. Temporary differential pressure (DP) instrumentations connected to the instrument taps from the permanently installed RHRSW flow orifices and RHR flow nozzles meet the guidance provided in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

8. Temporary instrumentation includes a temporary data acquisition system (DAS). The DAS software translates instrument output into data files that may be loaded into analytical software. Time stamped data is collected from each temporary instrument sensor. The DAS, including the associated software, complies with the guidance provided in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

9. Temporary instruments are calibrated against standards traceable to the National Institute of Standards and Technology or compared to nationally or internationally recognized consensus standards.

10. Test data is analyzed in accordance with EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015. The analysis determines the overall fouling resistance for the heat exchanger and also determines the associated uncertainty in the test result (fouling resistance).

11. The uncertainty analysis methodology complies with the approach described in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

12. Data reduction complies with the approach described in EPRI 3002005340, Service Water Heat Exchanger Test Guidelines, May 2015.

13. Computer programs used in the thermal performance analysis are required to meet the 10 CFR 50 Appendix B, and 10 CFR 21 requirements. PROTO-HX is the computer program used for thermal performance analyses in the RHR Heat Exchanger Performance Monitoring Program.

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ENCLOSURE 1

14. Test results in conjunction with trended fouling rate are projected and compared to the test fouling resistance acceptance criterion to schedule the next performance of the cleaning or testing PM in a manner consistent with ASME OM 2015, Section 9, Acceptance Criteria.

15. Compensatory measures will be established based upon actual plant conditions.

Compensatory measures include entering the condition into the TVA CAP, cleaning of the heat exchangers after inspections, determining if more frequent inspections and cleaning are required, and evaluating past operability/functionality when tube plugging and macrofouling acceptance criteria from inspection procedures are not met. The methodology for performing as-found and as-left inspections are provided in TVA Standard Programs and Processes procedures. As-left inspections are procedurally required and ensure that tubes found blocked/obstructed by macrofouling will not be left in the as-found condition. The number of tubes found obstructed by macrofouling during heat exchanger (as-found) inspections is compared to acceptance criteria for the total number of tubes that can be mechanically plugged, which is identified in implementing procedures. Upon EPU implementation acceptance criteria will also be provided for the total number of tubes that can be blocked/obstructed (includes tubes mechanically plugged and tubes obstructed by macrofouling) for a given fouling resistance (see Table SCVB-RAI 26). These acceptance criteria are used to determine whether a subsequent engineering evaluation is required and are limited to use in evaluating only as-found conditions. Degraded conditions are entered into the TVA CAP.

16. Changes to the program requirements above will be controlled in accordance with 10 CFR 50.59, Changes, tests, and experiments. Change to the program requirements may be made without prior NRC approval provided the changes do not require a change to the Technical Specification requirements and the changes do not require NRC approval pursuant to 10 CFR 50.59.

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ENCLOSURE 1 Table SCVB-RAI 26: RHR Heat Exchanger - Fouling Resistance and Total Tube Plugging Allowed Total Number of Tubes Allowed Blocked Fouling Resistance* (Number of Tubes Mechanically Plugged +

Number of Tubes Blocked by Macrofouling)

(hr-ft2-°F/Btu) (each) 0.001694 0 0.001677 10 0.001660 20 0.001648 27 0.001643 30 0.001626 40 0.001609 50 0.001592 60 0.001575 70 0.001558 80 0.001541 90 0.001524 100 0.001507 110 0.001490 120 0.001473 130 0.001456 140 0.001439 150 0.001422 160 0.001405 170 0.001389 180 0.001372 190 0.001355 200 0.001338 210 0.001321 220 0.001304 230 0.001287 240 0.001270 250 0.001253 260 0.001236 270 0.001219 280 0.001203 290 0.001186 300 0.001169 310 0.001152 320 0.001135 330 0.001118 340 0.001101 350 0.001085 360 0.001068 370 0.001051 380 0.001034 390 0.001017 400 0.001001 410 0.000984 420 0.000975 425

• For fouling resistances less than 0.000975 hr-ft2-°F/Btu the total number of tubes allowed blocked without an engineering evaluation is limited to no more than 425 tubes, which is one-half of the number of tubes in the heat exchanger inlet pass (850 tubes).

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ENCLOSURE 1 SCVB-RAI 30 As per Electric Power Research Institute (EPRI) Report 107397, in the nuclear industry, test uncertainty is often determined with 95 percent coverage. Per American Society of Mechanical Engineers Operation and Maintenance (ASME OM)-21, "A 95% confidence level shall be applied to the calculated result for the purpose of comparing the testing or monitoring results to the acceptance criteria."

a. Provide a detailed discussion of how the uncertainty analysis approach described in Section 3.9.2 of EPRI Report 107397 was applied in determining the fouling resistance results for the 2C, 3A, 3C, 2C, and 2A RHR heat exchangers given Table 4 of Reference 6.

b. The fouling resistance result that should be less than the acceptance criteria is defined as [fouling resistance result = measured fouling factor + test and measurement uncertainty]. Provide the values of test and measurement uncertainties that were added to the measured fouling resistances for determining the fouling resistances of the RHR heat exchangers given in Table 4 of Reference 6.

TVA Response:

a. The uncertainty analysis approach is addressed in the response to SCVB RAI-32.

Table 4 of the EPU LAR Attachment 39 previously included test result fouling resistances and associated heat transfer capability for comparison to the fouling resistance acceptance criteria for the RHR heat exchanger performance tests that had been completed at the time of the submittal of the BFN EPU LAR. Since that time, all 12 BFN RHR heat exchangers have been tested. The results of these tests are provided in the response to SCVB RAI-34. EPU LAR Attachment 39 is also revised in this EPU LAR Supplement to include the test information provided in the response to SCVB RAI-34. The updated fouling resistance results reflect the application of the uncertainty approach addressed in the response to SCVB RAI-32.

b. The testing fouling resistance and the uncertainty in fouling resistance are provided in Tables SCVB RAI-34-1, SCVB RAI-34-2 and SCVB RAI-34-3. The terms testing fouling resistance and the uncertainty in fouling resistance in Tables SCVB RAI-34-1, SCVB RAI-34-2 and SCVB RAI-34-3 are equivalent to the terms measured fouling factor and test and measurement uncertainty, respectively, in SCVB RAI-30b. As reflected in Tables SCVB RAI-34-1, SCVB RAI-34-2, and SCVB RAI-34-3, the fouling resistance results (test fouling resistance + uncertainty in fouling resistance) are less than the EPU acceptance criteria for fouling resistance.

As discussed above, Table 4 of EPU LAR Attachment 39 previously included test result fouling resistances and associated heat transfer capability for comparison to the fouling resistance acceptance criteria for the RHR heat exchanger performance tests that had been completed at the time of the submittal of the BFN EPU LAR. Since that time, all twelve BFN RHR heat exchangers have been tested. The results of these tests are provided in the response to SCVB RAI-34. EPU LAR Attachment 39 is also revised in this EPU LAR Supplement to include the test information provided in the response to SCVB RAI-34.

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ENCLOSURE 1 SCVB-RAI 32 TVA, in response to SCVB RAl-1(e), states that the performance testing will be conducted consistent with the guidance described in EPRI 3002005340, "Service Water Heat Exchanger Testing Guidelines," May 2015.

The response to SCVB RAl-1 (j) states:

The data analysis and preparation of vendor test reports are performed by a vendor also operating under a 10 CFR 50 Appendix B Quality Assurance Program in accordance with approved procedures that include steps to compare process and tube side heat transfer rates and to statistically evaluate test data such that results conservatively account for the uncertainties associated with each test. The method of calculation follows the EPRI guidelines (see response (e), above) in terms of determining the uncertainty contributors of precision and bias errors for thermal performance test evaluations. The uncertainty analysis methodology in PROTO-HX determines the sensitivity coefficients through a numerical approach using the central differencing method (i.e., symmetric uncertainties). The EPRI guideline (see response (e), above) provides an overview of this approach. The variables considered are the test data (flow rates and temperatures) and the film coefficients.

Provide the following information:

a. Describe the statistical evaluation method of the test data with discussion of how the uncertainty analysis methodology described in Chapter 4 of EPRI 3002005340 is implemented for evaluating the Random Standard Uncertainty and Systematic Standard Uncertainty combined together.

b. Justify that the Combined Standard Uncertainty (combination of Random Standard Uncertainty and Systematic Standard Uncertainty) in the test result is conservative as mentioned in the above response.

TVA Response:

a. BFN thermal performance testing of RHR Heat Exchangers was conducted and evaluated following the guidelines of EPRI TR-107397. EPRI TR-107397 was updated in 2015 and issued as EPRI 3002005340. In the future, BFN will conduct thermal performance tests of the RHR heat exchangers following the guidelines of EPRI 3002005340. Both of these EPRI guidelines describe methods to reduce the test data to a set of nominal values and estimate the test uncertainty at a 95% confidence interval using statistical analysis. The methodology used in conducting the BFN tests and performing the test data statistical analysis relies on a combination of hand calculations and the software program PROTO-HX.

A detailed discussion of the test data statistical evaluation method, including discussion of how the Random Standard Uncertainty and Systematic Standard Uncertainty are combined using the uncertainty analysis methodology from Chapter 4 of EPRI 3002005340, is provided in Zachry Nuclear, Inc. Engineering Evaluation 16-E04, E1-30

ENCLOSURE 1 Description of BFN RHR Heat Exchanger Test Data Evaluation, Revision 1 (see Enclosures 7 and 8 of this EPU LAR Supplement).

b. The EPRI guidance is based on the ASME Performance Test Codes, specifically, ASME PTC 12.5-2000, Single Phase Heat Exchangers and ASME PTC 19.1-1985 Part 1, Measurement Uncertainty: Instruments and Apparatus (replaced by ASME PTC 19.1-2005, Test Uncertainty). Section 2.6.3 of ASME PTC 19.1-1985, Part 1 contains the following statements:

Monte Carlo simulation of the uncertainty intervals has been used to generate approximate coverage values based on assuming various bias error distributions, bias limits, and precision indices. The results of such studies follow:

UADD provides approximately 99% coverage while URSS provides approximately 95% coverage when neither bias errors nor precision errors are negligible compared to the other.

If either the bias error or the precision error is negligible compared to the other, both UADD and URSS provide 95%

coverage.

An uncertainty analysis at another confidence level might be done in a manner similar to the 95% analysis. For example, a coverage of approximately 90% might be achieved by judging bias errors on a 9-out-of-10 basis, using Student ts based on the 90% confidence level, and using the RSS model. However, this method has not been demonstrated by Monte Carlo simulation at this time.

The statistical analysis techniques described in EPRI TR-107397 and EPRI 3002005340 are based on the methods described in ASME PTC 19.1. The uncertainty analyses that were used to evaluate test data using PROTO-HX is consistent with these methods and are based on a root-sum-of-squares (RSS) method at a 95% confidence interval. The 95% confidence interval is achieved by application of the two-tailed Students t score to the random uncertainty terms. The test report uncertainties determined using the EPRI methodology for all 14 tests are reported in EPU LAR Attachment 39, Tables 5, 6 and 7.

Therefore, the Combined Standard Uncertainty (combination of Random Standard Uncertainty and Systematic Standard Uncertainty) in the test result reflect a methodology to produce a 95% confidence interval. It is this 95% confidence interval that was characterized as conservative in the response to SCVB-RAI 1(j).

E1-31

ENCLOSURE 1 SCVB-RAI 34 Provide the following latest as-found test results of all BFN Units 1, 2, and 3 RHR heat exchangers:

UNIT 1 Unit 2 UNIT 3 Parameter 1A 1B 1C 1D 2A 2B 2C 2D 3A 3B 3C 3D Test date Number of years in operation after last cleaning Hot side inlet temperature (°F)

Hot side outlet temperature (°F)

Hot side flow rate (pound mass/hour)

Cold side inlet temperature (°F)

Cold side outlet temperature (°F)

Test heat transfer rate (BTU/hr)

Test U (BTU/ hr-ft2-°F)

Test UA (BTU/hr-°F)

Test fouling resistance (hr-ft2-

°F/BTU)

Uncertainty in fouling (hr-ft2-

°F/BTU) resistance (hr-ft2-

°F/BTU)

Acceptance criteria for fouling 2

resistance (hr-ft -°F/BTU)

Margin in fouling resistance (hr-ft2-°F/BTU)

If inspected, number of fully blocked tubes If inspected, number of partially blocked tubes Acceptance criteria for blocked tubes TVA Response:

The thermal performance test results for all Browns Ferry Nuclear Plant (BFN) Units 1, 2 and 3 residual heat removal (RHR) heat exchangers tested since January, 2012 are provided in Tables SCVB-RAI 34-1, SCVB-RAI 34-2 and SCVB-RAI 34-3 and the associated Notes.

E1-32

ENCLOSURE 1 Table SCVBRAI 341: Test Results for BFN Unit 1 RHR Heat Exchangers UNIT 1 Parameter 1A 1B 1C 1D Test date 12/2/15 3/23/16 12/2/15 3/23/16 Number of years in operation after last 0.3 1.6 1.6 1.9 cleaning Hot side inlet temperature (°F) 77.64 73.43 79.14 74.55 Hot side outlet temperature (°F) 73.51 68.55 73.99 69.64 Hot side flow rate (pound mass/hour) 3,631,734 3,668,936 3,622,515 3,671,967 Cold side inlet temperature (°F) 57.76 57.20 57.75 58.08 Cold side outlet temperature (°F) 72.04 65.98 71.72 66.64 Cold side flow rate (pound mass/hour) 1,034,623 1,969,221 1,313,440 2,007,073 Test heat transfer rate 14,837,671 17,530,731 18,444,721 17,506,428 (BTU/hr)

Test U 273.2765 329.4458 291.2366 314.0870 (BTU/hrft2°F)

Test UA 1,716,835 2,080,849 1,834,042 1,981,481 (BTU/hr°F)

Test fouling resistance 0.000463 0.000543 0.000558 0.000720 (hrft2°F/BTU)

Uncertainty in fouling resistance 0.000292 0.000275 0.000275 0.000275 (hrft2°F/BTU)

Acceptance criteria for fouling resistance 0.001517 (1) 0.001517 (1) 0.001517 (1) 0.001517 (1)

(hrft2°F/BTU)

Margin in fouling resistance 0.000762 0.000699 0.000684 0.000522 (hrft2°F/BTU)

EPU Acceptance criteria for fouling resistance 0.001562 0.001562 0.001562 0.001562 (hrft2°F/BTU)

EPU Margin in fouling resistance 0.000807 0.000744 0.000729 0.000567 (hrft2°F/BTU)

Not Not If inspected, number of fully blocked tubes 5 Not Inspected Inspected Inspected If inspected, number of partially blocked Not Not 7 Not Inspected tubes Inspected Inspected 77 77 77 77 Acceptance criteria for blocked tubes (2) (2) (2) (2)

E1-33

ENCLOSURE 1 Table SCVBRAI 342: Test Results for BFN Unit 2 RHR Heat Exchangers UNIT 2 Parameter 2A 2A 2B 2C 2C 2D Test date 4/11/15 1/8/15 1/5/16 4/11/15 1/8/15 1/5/16 Number of years in operation 0.1 1.6 0.6 0.1 3.9 2.1 after last cleaning Hot side inlet temperature 80.74 77.16 74.58 83.19 76.84 75.10

(°F)

Hot side outlet temperature 76.69 68.68 66.14 78.22 68.27 66.75

(°F)

Hot side flow rate (pounds 4,030,078 3,537,939 3,770,401 4,022,707 3,612,076 3,804,645 mass/hour)

Cold side inlet temperature 65.19 41.13 46.96 64.88 40.68 47.06

(°F)

Cold side outlet temperature 73.89 62.20 61.46 75.98 62.77 62.06

(°F)

Cold side flow rate (pounds 1,706,146 1,346,317 2,035,618 1,686,190 1,311,739 1,993,996 mass/hour)

Test heat transfer rate 15,224,364 28,920,419 30,302,215 19,062,260 29,695,875 30,614,461 (BTU/hr)

Test U 291.0500 239.9253 328.7694 337.5143 256.0849 330.5212 (BTU/hrft2°F)

Test UA 1,841,610 1,518,120 2,079,046 2,133,077 1,618,446 2,077,712 (BTU/hr°F)

Test fouling resistance 0.000946 0.001152 0.000535 0.000477 0.000868 0.000515 (hrft2°F/BTU)

Uncertainty in fouling 0.000274 0.000284 0.000262 0.000258 0.000289 0.000262 resistance (hrft2°F/BTU)

Acceptance criteria for fouling 0.001517 0.001517 (1) 0.001517 (1) 0.001517 (1) 0.001517 (1) 0.001517 (1) resistance (hrft2°F/BTU) (1)

Margin in fouling resistance 0.000297 0.000081 0.000720 0.000782 0.000360 0.000740 (hrft2°F/BTU)

EPU Acceptance criteria for fouling resistance 0.001562 0.001562 0.001562 0.001562 0.001562 0.001562 (hrft2°F/BTU)

EPU Margin in fouling 0.000342 0.000126 0.000765 0.000827 0.000405 0.000785 resistance (hrft2°F/BTU)

If inspected, number of fully Not 305 3 Not Inspected Not Inspected 11 blocked tubes Inspected (3)

If inspected, number of Not Not Not 6 19 12 partially blocked tubes Inspected Inspected Inspected (3)

Acceptance criteria for 77 77 77 77 77 77 blocked tubes (2) (2) (2) (2) (2) (2)

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ENCLOSURE 1 Table SCVBRAI 343: Test Results for BFN Unit 3 RHR Heat Exchangers UNIT 3 Parameter 3A 3B 3C 3D Test date 1/25/12 10/30/15 1/25/12 10/30/15 Number of years in operation after last 1.9 1.4 4.0 4.0 cleaning Hot side inlet temperature (°F) 75.29 85.63 74.73 87.27 Hot side outlet temperature (°F) 68.83 81.60 68.67 83.10 Hot side flow rate (pounds mass/hour) 3,827,744 3,487,834 4,014,733 3,469,054 Cold side inlet temperature (°F) 53.02 65.39 53.02 65.52 Cold side outlet temperature (°F) 65.11 80.53 64.38 81.84 Cold side flow rate (pounds mass/hour) 1,979,900 896,677 2,036,549 877,040 Test heat transfer rate 24,291,564 13,679,428 23,630,760 14,349,628 (BTU/hr)

Test U 327.7767 257.1204 314.4073 252.4582 (BTU/hrft2°F)

Test UA 2,078,920 1,641,405 1,992,946 1,589,838 (BTU/hr°F)

Test fouling resistance 0.000569 0.000553 0.000746 0.000625 (hrft2°F/BTU)

Uncertainty in fouling resistance 0.000266 0.000299 0.000260 0.000296 (hr-ft2-°F/BTU)

Acceptance criteria for fouling 0.001517 0.001517 0.001517 0.001517 resistance (hr-ft2-°F/BTU) (1) (1) (1) (1)

Margin in fouling resistance 0.000682 0.000665 0.000511 0.000596 (hr-ft2-°F/BTU)

EPU Acceptance criteria for fouling resistance 0.001562 0.001562 0.001562 0.001562 (hr-ft2-°F/BTU)

EPU Margin in fouling resistance 0.000727 0.000710 0.000556 0.000641 (hr-ft2-°F/BTU)

If inspected, number of fully blocked Not 43 53 12 tubes Inspected (4)

If inspected, number of partially Not 22 77 12 blocked tubes Inspected Acceptance criteria for blocked tubes 77 77 77 77 (2) (2) (2) (2)

E1-35

ENCLOSURE 1 Notes for Tables SCVB-RAI 34-1, SCVB-RAI 34-2 and SCVB-RAI 34-3 (1) NFPA 805 license condition acceptance criterion.

(2) NFPA 805 license condition allowable tube plugging limit (4.57%) limits the number of tubes can be plugged (mechanically) to 77.

(3) During a July, 2013 inspection of a Residual Heat Removal (RHR) heat exchanger, macrofouling of the inlet side of the heat exchanger from relic Asiatic clamshells was identified. The corrective actions included adding frequent cleaning of the RHRSW Pump Pit to the PM program. A March 18, 2015 inspection of RHR heat exchanger 2C, which found 305 tubes potentially fully obstructed, was the last of the 12 heat exchangers inspected subsequent to the October, 2013 RHRSW Pump Pit cleaning. That is, it was the last inspection which could have found shells that had been introduced into the heat exchanger prior to the October, 2013 RHRSW Pump Pit cleaning. The previous cleaning of RHR heat exchanger 2C was performed in January, 2011. RHR heat exchangers which have been inspected more than once, subsequent to performing the RHRSW Pump Pit cleaning PM, have shown a significant reduction in the number of shells found. This provides objective evidence that the RHRSW Pump Pit cleaning PM is effective in reducing the accumulation of shells found during subsequent RHR Heat Exchanger inspections. The issue which caused 305 tubes to be plugged is historical and the corrective actions associated with shells in the RHRSW Pump Pit have resolved the issue.

E1-36

ENCLOSURE 3 Supplement to BFN EPU LAR, Attachment 7, NEDO-33860P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate, (Non-proprietary version)

NEDO-33860 Revision 0 September 2015 Non-Proprietary Information - Class I (Public)

SAFETY ANALYSIS REPORT FOR BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3 EXTENDED POWER UPRATE Copyright 2015 GE - Hitachi Nuclear Energy Americas LLC All Rights Reserved

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Term Definition ECP Electrochemical Potential EECW Emergency Equipment Cooling Water EFDS Equipment and Floor Drainage System EFPY Effective Full Power Years EHC Electro-Hydraulic-Control EHPMP Emergency High Pressure Makeup Pump ELTR1 Generic Guidelines for General Electric Boiling Water Reactor Extended Power Uprate Licensing Topical Report ELTR2 Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate Licensing Topical Report EMA Equivalent Margin Analysis EOC End of Cycle EOI Emergency Operating Instruction EOL End of Life EOC-RPT End of Cycle-Recirculation Pump Trip EPRI Electric Power Research Institute EPU Extended Power Uprate EQ Environmental Qualification ESF Engineered Safety Feature ESFVS Engineered Safety Feature Ventilation System ESW Emergency Service Water FAC Flow Accelerated Corrosion FCF Flow Correction Factor FCV Flow Control Valve FFWTR Final Feedwater Temperature Reduction FHA Fuel Handling Accident FIV Flow Induced Vibration xv

NEDO-33860 Revision 0 Non-Proprietary Information - Class I (Public) effects of EPUs. Section 6.7 of the CLTR addresses the effect of EPU on the fire protection program. The results of this evaluation are described below.

As explicitly stated in Section 6.7 of the CLTR, ((

))

Therefore, the reactor and containment responses and operator actions will be evaluated on a plant-specific basis for EPU.

This section addresses the effect of EPU on the fire protection program, fire suppression and detection systems, and reactor and containment system responses to postulated fire events.

Once the NFPA 805 (Reference 66) fire protection transition is implemented, Browns Ferry will meet all CLTR dispositions. The topics addressed in this evaluation are:

Browns Ferry Topic CLTR Disposition Result Meets CLTR Fire Suppression and Detection Systems Plant Specific Disposition Meets CLTR Operator Response Time Plant Specific Disposition Meets CLTR Peak Cladding Temperature Plant Specific Disposition Meets CLTR Vessel Water Level Plant Specific Disposition Meets CLTR Suppression Pool Temperature Plant Specific Disposition The higher decay heat associated with EPU results in higher heat input into the suppression pool which, without mitigation, will result in higher suppression pool temperatures. The higher decay heat may also result in lower vessel water levels or higher Peak Cladding Temperatures (PCTs),

depending on the plant-specific analysis basis. As a result of these effects, fire suppression and detection systems, operator response time, peak clad temperature (PCT), and suppression pool temperature need to be addressed. implemented Tennessee Valley Authority (TVA) is implementing the Nuclear Energy Institute methodology NEI 04-02, Revision 2, Guidance for Implementing a Risk-informed, Performance-based Fire Protection Program Under 10 CFR 50.48(c) (NEI 04-02) (Reference 67), to transition Browns 2-205

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Ferry Nuclear Plant (BFN) Units 1, 2 and 3 from its current fire protection licensing basis to the new requirements as outlined in NFPA 805. The BFN NFPA 805 Fire Safe Shutdown Analysis consists of a deterministic analysis and a performance based analysis. The deterministic analysis (NFPA 805 Section 4.2.3) identifies and evaluates one success path for each fire area to meet the nuclear safety performance criteria of Section 1.5. Section 2.5.1.4.2, Fire Event, addresses the deterministic analysis. For instances where the nuclear safety performance criteria are not met, a performance based analysis (NFPA 805 Section 4.2.4) is performed to demonstrate that risk is acceptable and that defense in depth and safety margin are maintained. The performance based analysis is addressed in LAR Attachment 44.

Safe Shutdown Systems, equipment, and compensatory measures will be sufficient to support EPU. EPU is found to not affect the elements of the fire protection program related to: (1) fire suppression and detection systems, (2) fire zones/areas, (3) fire barriers, and (4) fire protection responsibilities of plant personnel. Administrative controls, associated with fire protection in the Technical Specifications, the Fire Protection Report, and the Nuclear QA Plan, will be adequate for EPU conditions.

As a risk reduction action in the NFPA 805 Transition LAR (Reference 65), a non-safety Emergency High Pressure Makeup Pump (EHPMP) will be installed. The EHPMP will supply water to the RPV from the CST. Injecting with this pump will also add additional volume to the suppression pool. The EHPMP will not be credited for Containment Accident Pressure elimination; however, the analysis shows that a net positive suction head improvement for the safe shutdown pumps would be realized. (See Section 2.6.5.2) Other EPU modifications will be assessed and assured not to adversely affect the ability to achieve and maintain the fuel in a safe and stable condition in the event of a fire. A Original NFP 805 analyses were performed at EPU conditions and therefore operator action times cannot be compared to CLTP conditions. To ensure that PCT remains less than the acceptance criterion in the most limiting scenario, one LPCI pump must be manually aligned for injection within 20 minutes. The EPU requires no new operator actions for fire safe shutdown of the plant and there are no actions required inside the primary containment.

The reactor and containment responses to the postulated fire events at EPU conditions are described in Section 2.5.1.4.2. The results show that for the limiting thermal-hydraulic cases, peak fuel cladding temperature, vessel water level, and suppression pool temperature meet the acceptance criteria and there is sufficient time for the operators to perform the necessary actions to meet the NFPA 805 requirement to achieve and maintain the fuel in a safe and stable condition in the event of a fire.

Therefore, once the NFPA 805 fire protection transition is implemented, Browns Ferry will meet all CLTR dispositions.

2.5.1.4.2 Fire Event The limiting NFPA 805 fire events were analyzed under EPU conditions. The fuel heat-up analysis was performed using the NRC accepted AREVA LOCA methodology 2-206

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(RELAX/HUXY). The containment analysis was performed using the GEH SHEX model.

These analyses determined the effect of EPU on fuel cladding integrity, reactor vessel integrity, and containment integrity as a result of the fire event. The two bounding cases described below are identified as Case 1 and Case 4. See Tables 2.5-1, 2.5-2, and 2.5-3 for the inputs and results of the fire event analyses.

Case 1: The bounding safe shutdown case for PCT has Multiple Spurious Operation (MSO) of 11 of the 13 MSRVs which depressurize the reactor, and one RHR pump aligned in the LPCI/ASDC mode at 20 minutes. The analysis shows that the calculated PCT of 1,330°F is acceptable from a deterministic perspective (< 1,500°F) (See FUSAR Section 2.5.1.4). 207.7 Case 4 (See Figure 2.5-1): The bounding safe shutdown case for peak suppression pool temperature has reactor depressurization beginning at 25 minutes using three MSRVs. As the reactor is depressurized, condensate pumps replenish reactor inventory until hotwell inventory is depleted. After condensate is secured, one RHR pump is aligned into LPCI/ASDC mode. One RHRSW pump is initiated at two hours. Peak SP temperature reaches 208.0°F and this meets the containment integrity acceptance criteria of < 281°F and the torus attached piping limit of

<223°F (See Section 2.2.2.2.2.2). Analyses show that containment accident pressure credit is not required to ensure adequate pump net positive suction head (NPSH) to mitigate a fire event (see Section 2.6.5.2 and LAR Attachment 39).

The results of Case 4, and the evaluations in Section 2.6.5.2, FUSAR Section 2.5.1.4, and LAR 9, demonstrate that the peak fuel cladding temperature, vessel water level, and suppression pool temperature meet the acceptance criteria and the time available for the operators to perform the necessary actions is sufficient. Therefore, EPU has no adverse effect on the ability of the systems and personnel to mitigate the effects of a fire event and satisfies the requirement of achieving and maintaining the fuel in a safe and stable condition in the event of a fire.

Conclusion TVA has evaluated fire-related safe shutdown requirements and has accounted for the effects of the increased decay heat on the ability of the required systems to achieve and maintain safe shutdown conditions. The evaluation indicates that the FPP will continue to meet the requirements of 10 CFR 50.48, final GDC-3, and draft GDC-4 following implementation of the proposed EPU. Therefore, the proposed EPU is acceptable with respect to fire protection.

2.5.2 Fission Product Control 2.5.2.1 Fission Product Control Systems and Structures Regulatory Evaluation The NRCs acceptance criteria are based on GDC-41, insofar as it requires that the containment atmosphere cleanup system be provided to reduce the concentration of fission products released to the environment following postulated accidents.

Specific NRC review criteria are contained in SRP Section 6.5.3.

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Table 2.5-1 NFPA 805 Fire Event Key Inputs Input Parameters Values Reactor Thermal Power 3,952 MWt RPV Dome Pressure 1,055 psia Decay Heat ANS 5.1-1979 without 2 uncertainty adder and with GEH SIL 636 recommendations Initial Suppression Pool Liquid Volume 124,200 ft3 (Note 1) 122,940 Initial Suppression Pool and Wetwell Airspace 95°F (Note 2)

Temperature 92 Initial Wetwell Pressure 14.4 psia Initial Drywell Pressure 15.5 psia Initial Drywell Temperature 150°F Initial Wetwell Relative Humidity 100%

Initial Drywell Relative Humidity 20%

Drywell and Wetwell and Pool Heat Sinks Modeled Yes Drywell Heat Load Modeled Yes 88 RHR Service Water Temperature 92°F (Note 2)

RHR Heat Exchanger K Factor per Loop 307 Btu/sec-°F (Note 3) 290 Number of RHR Loops Available 1 Number of RHR Pumps in One RHR Loop 1 ASDC RHR Flow Rate 7,500 gpm Condensate Available for Injection 90,000 gallons INSERT 1A 2-253

Insert 1A Notes (1) Suppression Pool Volume corresponding to Browns Ferry Technical Specification low suppression pool water level with differential pressure control in service.

(2) Nominal values based on Browns Ferry plant data over seven year period from January 2008 through December 2014. Data analysis for this parameter shows that Browns Ferry operates at least 95% of time below this value.

(3) RHR heat exchanger K factor based on RHR flow of 7,500 gpm, RHRSW flow of 4,500 gpm, RHRSW temperature of 88°F and conservative RHR heat exchanger fouling resistance. See LAR Attachment 39 for calculation of K factor.

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Table 2.5-2 NFPA 805 Case 4 (EPU) Fire Event Evaluation Results Item Parameters Values 24.1 24.2 1 Peak DW Pressure (psia)

~21,570 seconds 24,640 276.5 2 Peak DW Temperature (ºF) 276.3

~1,500 seconds 24.6 24,640 3 Peak WW Airspace Pressure (psia)

~21,570 seconds 209.0 210.2 4 Peak WW Airspace Temperature (ºF) 54,110

~50,390 seconds 208.0 207.7 5 Peak Pool Temperature (ºF)

~18,280 seconds 19,850 2-254

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Insert A NEDO-33860 Revision 0 Non-Proprietary Information - Class I (Public)

Fire Event ECCS NPSH In the containment response analysis for a Browns Ferry fire event as described in Section 2.5.1.4.2, a single RHR pump is the only ECCS pump assumed to operate in order to achieve fire event safe shutdown. The containment response to a fire event was performed at EPU RTP. In the ECCS NPSH evaluation, a RHR flow rate of 7,615 gpm is used. For Browns Ferry, the limiting fire event scenario terminates following initiation of Alternate Shutdown Cooling (ASDC) and safe and stable conditions are achieved. Zero percent uncertainty is applied in the determination of NPSHReff for this special event. ECCS suction strainer debris loading and holdup volume are not included in the NPSH evaluation because there is no assumption of a pipe break or operation of containment sprays during the event. A nominal initial suppression pool level was assumed, which is consistent with the NRC guidance contained in Reference 97. Technical Specification minimum The maximum suppression pool temperature, NPSHa, NPSH margin, and the operating time with a margin ratio less than 1.6 are listed in Table 2.6-4. The RHR pump flow rate used in the ECCS NPSH evaluation are conservatively higher than those used in the safety analysis that provides the suppression pool temperature response. A t ime-history plot of the NPSHa for the limiting fire event is provided in Figure 2.6-16. This case demonstrates positive NPSH margin and thus, CAP credit is not required. However, the small margin prompted a further sensitivity case to show increased margin.

The sensitivity case involved an analysis where a postulated 1,000 hp electric-driven Emergency High Pressure Makeup Pump (EHPMP) could be used as defense-in-depth to inject water from the CST through the FW piping and into the RPV while the RHR pump was operating in ASDC mode. This effectively provides a means of pumping CST inventory through the RPV and into the torus to increase the suppression pool mass, providing more mass to accept the heat input from the RPV while at the same time increasing the suppression pool level which would increase the RHR pump NPSHa. This case used the Browns Ferry TS value of 95°F for RHRSW temperature. This case also used the EPU design RHR heat exchanger K-value of 287 BTU/sec-°F. Further details concerning the determination of the RHR heat exchanger K-value are contained in LAR Attachment 39.

The results from the sensitivity case described above are provided in Table 2.6-4b. The sensitivity case is provided for comparison purposes only. The improvement in the NPSH margin by using the EHPMP is 2.9 feet compared to the results contained in Table 2.6-4 for the fire event.

Station Blackout ECCS NPSH The Browns Ferry SBO event described in Section 2.3.5 postulates that on-site and off-site power are lost for the entire four hour coping duration. The containment response to SBO was performed at EPU RTP. Core cooling is maintained with high pressure injection systems (HPCI and/or RCIC) taking suction from the CST and excess reactor steam is vented to the suppression pool using MSRVs. At the end of the four hour coping period, RHR pumps are operated in 2-315

NEDO-33860 Revision 0 Non-Proprietary Information - Class I (Public) suppression pool cooling mode. NPSH concerns for the SBO event are related to the suppression pool level, pump suction strainer and suction piping friction losses, and peak suppression pool temperature at the end of the four hour coping period when suppression pool cooling is initiated.

For the SBO event, the only ECCS pumps operating with suction from the suppression pool are the RHR pumps. The assumed RHR pump flow for the SBO NPSH evaluation is 6,600 gpm. Zero percent uncertainty is applied in the determination of NPSHReff for this special event. ECCS suction strainer debris loading and holdup volume are not included in the NPSH evaluation because there is no assumption of a pipe break or operation of containment sprays during the event. The maximum suppression pool temperature, NPSHa, NPSH margin, and the operating time with a margin ratio less than 1.6 for the SBO scenario is listed in Table 2.6-4.

The pump flow rate used in the ECCS NPSH evaluation is conservatively higher than that used in the safety analysis that provides the suppression pool temperature response. The HPCI pumps are also credited for the SBO event, operating for a maximum of 30 minutes with suction from the CST only. A time-history plot of the NPSHa is provided in Figure 2.6-17.

and pressure regulator failure ATWS ECCS NPSH open (PRFO) EOC events are 171.8 As discussed in Section 2.8.5.7, the limiting event with respect to peak suppression pool temperature is the ATWS-LOOP event (two RHR pumps / heat exchangers) which results in a are peak suppression pool temperature of 173.3°F at EPU RTP. The most limiting non-LOOP (four RHR pumps / heat exchangers) ATWS event is main stem isolation valve closure (MSIVC)

EOC, which experiences a peak suppression pool temperature of 171.7°F. The ATWS events events were analyzed at EPU RTP. Similar to the previous discussion concerning the effect of total both of pump flow on ECCS pump suction piping and strainer friction loss (RSLB DBA-LOCA versus RDLB LOCA short-term discussion), when the combined transient effects of suppression pooland PRFO temperature, suppression pool level and ECCS pump suction losses are considered, the ATWSEOC events resulting in the least NPSH margin is the non-LOOP event (MSIVC EOC) where the pump suction piping losses exceed the gain from a lower peak pool temperature. The NPSH margin is 14.2 feet for this non-LOOP event whereas the NPSH margin for the LOOP event is 15.5 feet. Consequently, with no CAP credit, there is substantial NPSH margin for all the ATWS events. these events For the ATWS event, the only ECCS pumps operating from the suppression pool are the RHR pumps. HPCI supplies makeup to the RPV with suction from the CST. The CS pumps are not credited for the ATWS event. The assumed RHR pump flow for the ATWS event NPSH analysis is 6,600 gpm. Zero percent uncertainty is applied in the determination of NPSHReff for this special event. ECCS suction strainer debris loading and holdup volume are not included in the NPSH evaluation because there is no assumption of a pipe break or operation of containment sprays during the event. The Browns Ferry RPV pressure relief system uses only MSRVs that are piped to discharge headers (T-quenchers) below the torus water level. There are no un-piped spring safety valves that discharge directly into the drywell and could contribute to ECCS suction strainer debris loading during an ATWS event. The maximum suppression pool temperature, NPSHa, NPSH margin, and the operating time < 1.6 margin ratio for the ATWS 2-316

2.6-18a for MSIVC Time-history plots NEDO-33860 Revision 0 EOC and Figure Non-Proprietary Information - Class I (Public) 2.6-18b for PRFO are EOC event are listed in Table 2.6-4. The pump flow rates used in the ECCS NPSH evaluation are conservatively higher than those used in the safety analysis that provides the suppression pool temperature response. A time-history plot of the NPSHa is provided in Figure 2.6-18.

Shutdown of the Non-Accident Unit Following LOOP and Accident in the Accident Unit ECCS NPSH The suppression pool temperature response during shutdown and cooldown of the non-accident Browns Ferry units during an accident, including DBA-LOCA (on the accident unit) concurrent with loss of offsite power and loss of a 4 kV shutdown board is discussed in Section 2.6.5.1.

This event results in the non-accident Browns Ferry unit entering into shutdown cooling mode in order to achieve cold shutdown conditions. Evaluation of the shutdown of the non-accident unit was performed at 102% EPU RTP. For the non-accident unit safe shutdown NPSH analysis, a single RHR pump in the non-accident unit is assumed to operate at 10,000 gpm, which is conservatively higher than the RHR pump flow rate (9,700 gpm) assumed in the safety analysis that provides the suppression pool temperature response. For the non-accident unit NPSH analysis CS is assumed to operate at 3,173 gpm to provide RPV inventory makeup when HPCI is not available. HPCI is assumed available for part of this event. HPCI can be operated either with suction from the CST or HPCI operation can be secured prior to the suppression pool temperature reaching the 140°F qualification limit for HPCI. Zero percent uncertainty is applied in the determination of NPSHReff because this is a non-design basis event.

ECCS suction strainer debris loading and holdup volume are not assumed in the NPSH evaluation because there is no pipe break or operation of containment sprays during the event.

The maximum suppression pool temperature, NPSH margin, and the operating time with a margin ratio less than 1.6 is listed in Table 2.6-4. Time history plots of NPSHa for the RHR and CS pumps are provided in Figure 2.6-19a and Figure 2.6-19b, respectively.

ECCS NPSH Summary EPU RTP operation increases the reactor decay heat, which increases the heat addition to the suppression pool following a DBA-LOCA or other events. The peak suppression pool temperature for the analyzed accidents and transients is within the design capability of the ECCS pumps. Adequate NPSHa is demonstrated, and no credit for CAP is needed. The ECCS pump operating time with a margin ratio less than 1.6 is much less than 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br /> for any event.

The debris generated and transported following a LOCA that can cause ECCS strainer head loss includes fiber, reflective metal insulation, qualified coatings, dirt/dust, rust flakes, sludge, and unqualified coatings. The ECCS suction strainers are passive, stacked-disc strainers, which were designed, manufactured and tested by GE Nuclear Energy. The ECCS strainer design debris load, which was used as an input to the strainer design, is documented Reference 38. The quantity and characterization of the strainer debris loading is based on the methodology in Reference 102. The Browns Ferry design basis ECCS suction strainer debris loading was evaluated and is not affected by EPU.

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Table 2.6-3 Browns Ferry Peak Suppression Pool Temperatures for Postulated ATWS, Station Blackout, and NFPA 805 Events Event Peak Suppression Pool Temperature Limiting ATWS (Loss of Off-Site Power) 173.3ºF Station Blackout 203.7ºF NFPA 805 Fire 208.0ºF 207.7 2-330

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Table 2.6-4 ECCS Pump EPU NPSH Summary Non-RSLB Small Loss of RDLB Accident Fire Event Long Steam Shutdown SORV SBO ATWS Short Term Unit Event Term Break Cooling Shutdown Design Abnormal Abnormal Abnormal Event Special Special Special Basis Accident Accident Operational Operational Operational Type Event Event Event Accident Transient Transient Transient Parameter Units Number of operating RHR NA 2 2 2 2 2 2 1 4(1) 4 1 pumps RHR pump flow rate in safety (gpm) 6,500 11,000 9,000 6,500 6,500 6,500 9,700 6,500 6,500 7,500 analysis RHR pump flow in NPSH (gpm) 6,600 11,169 9,138 6,600 6,600 6,600 10,000 6,600 6,600 7,615 analysis Number of operating CS pumps NA 2 4 2 1 1 1 N/A N/A N/A CS pump flow rate in safety (gpm) 3,125 3,550 3,125 3,125 3,125 3,125 N/A N/A N/A analysis CS pump flow in NPSH analysis (gpm) 3,173 3,604 3,173 3,173 3,173 3,173 N/A N/A N/A Total flow rate in ring header for (gpm) 19,546 55,030 19,546 16,373 16,373 13,173 26,400 26,400 7,615 NPSH discussion Suction strainer debris loading NA Yes Yes Yes No No No No No No assumed?

Note:

1. The SBO analysis sequence of events shown in Table 2.3-4b states that two RHR pumps and two RHR heat exchangers are placed in service at the end of the four hour coping period. The SBO NPSH evaluation assumed a more limiting case where four RHR pumps are placed in service at the end of the four hour coping period. The configuration of four running RHR pumps results in a higher suction piping and strainer friction loss term and a more limiting NPSH margin determination than for a two running RHR pump configuration.

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NEDO-33860 Revision 0 Non-Proprietary Information - Class I (Public) 171.8 207.7 Table 2.6-4 ECCS Pump EPU NPSH Summary (continued)

RHR Pump NPSH Summary Non-RSLB RDLB Small Loss of Accident Fire Event Long Short Steam Shutdown SORV SBO ATWS Unit Event Term Term Break Cooling Shutdown Design Abnormal Abnormal Abnormal Event Special Special Special Basis Accident Accident Operational Operational Operational Type Event Event Event Accident Transient Transient Transient Parameter Units Peak SP Temperature (PPT) (°F) 179.0 152.0 181.5 178.3 161.8 185.1 203.7 171.7 208.0 15.2 ha, atmospheric pressure above SP (feet) 34.2 33.9 34.3 34.2 34.0 34.3 34.5 34.1 34.6 (1)

RHR hs, water static height (feet) 14.7 14.1 14.4 14.9 14.5 15.9 16.1 15.4 15.3 RHR hf, suction pipe and strainer 32.4 (feet) 2.8 12.4 2.8 2.5 2.5 2.3 3.6 3.6 1.31 friction loss hvap, vapor pressure @ PPT (feet) 17.5 9.2 18.5 17.2 11.7 20.0 29.8 14.7 32.6 RHR pump available NPSH (feet) 28.7 26.4 27.4 29.4 34.3 27.9 17.2 31.2 16.04 16.11 (NPSHa = ha + hs - hf - hvap)

RHR pump required NPSH (feet) 17.0 18.0 17.0 17.0 17.0 21.0 17.0 17.0 16.0 (NPSHR3%)

RHR pump NPSH uncertainty (%) 21 21 21 0 0 0 0 0 0 RHR pump NPSHReff (feet) 20.6 21.8 20.6 17.0 17.0 21.0 17.0 17.0 16.0

({1+NPSHuncertainty} x NPSHR3%)

RHR pump NPSH margin (feet) 8.1 4.7 6.8 12.4 17.3 6.9 0.2 14.2 0.04 0.11 (NPSHa - NPSHReff)

RHR pump minimum NPSH ratio NA 1.7 1.5 1.6 1.7 2.0 1.3 1.0 1.8 1.0 (NPSHa/NPSHR3%)

Time RHR pump NPSH ratio < 1.6 (hours) 0 <1 0 0 0 <1 <3 0 < 16 Note:

1. The water static height is the difference between the SP level and the RHR pump suction centerline elevation. The Browns Ferry SP (torus) zero elevation is at a plant elevation of 521.5 feet. The RHR pump suction centerline elevation is at a plant elevation of 521.6 feet. These values are applicable to all three Browns Ferry units.

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Table 2.6-4 ECCS Pump EPU NPSH Summary (continued)

CS Pump NPSH Summary Non-RSLB Small Loss of Fire RDLB Accident SBO ATWS Event Long Steam Shutdown SORV Event Short Term Unit Note 1 Note 1 Term Break Cooling Note 1 Shutdown Design Abnormal Abnormal Abnormal Event Special Special Special Basis Accident Accident Operational Operational Operational Type Event Event Event Accident Transient Transient Transient Parameter Units Peak SP Temperature (PPT) (°F) 179.0 152.0 181.5 178.3 161.8 185.1 N/A N/A N/A ha, Atmospheric pressure above (feet) 34.2 33.9 34.3 34.2 34.0 34.3 N/A N/A N/A SP CS hs, Water static height (2) (feet) 15.0 14.4 14.7 15.2 14.8 16.2 N/A N/A N/A CS hf, suction pipe and strainer (feet) 5.7 12.3 5.7 1.9 1.9 1.9 N/A N/A N/A friction loss hvap, vapor pressure @ PPT (feet) 17.5 9.2 18.5 17.2 11.7 20.0 N/A N/A N/A CS pump available NPSH (feet) 26.1 26.8 24.8 30.4 35.2 28.6 N/A N/A N/A (NPSHa = ha + hs - hf - hvap)

CS pump required NPSH (feet) 20.0 20.0 20.0 20.0 20.0 20.0 N/A N/A N/A (NPSHR3%)

CS pump NPSH uncertainty (%) 21 21 21 0 0 0 N/A N/A N/A CS pump NPSHReff (feet) 24.2 24.2 24.2 20.0 20.0 20.0 N/A N/A N/A

({1+NPSHuncertainty} x NPSHR3%)

CS pump NPSH margin (feet) 1.9 2.6 0.6 10.4 15.2 8.6 N/A N/A N/A (NPSHa - NPSHReff)

CS pump minimum NPSH ratio NA 1.3 1.3 1.2 1.5 1.76 1.4 N/A N/A N/A (NPSHa/NPSHR3%)

Time CS pump NPSH ratio < 1.6 (hours) < 18 <1 < 16 <1 0 <1 N/A N/A N/A Notes:

1. Core spray pumps do not operate during these events.

2. The water static height is the difference between the SP level and the CS pump suction centerline elevation. The Browns Ferry SP (torus) zero elevation is at a plant elevation of 521.5 feet. The CS pump suction centerline elevation is at a plant elevation of 521.3 feet. These values are applicable to all three Browns Ferry units.

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Table 2.6-4a ECCS Pump EPU NPSH Summary - Supplemental Evaluation Event RDLB Short Term Parameter Units Number of operating RHR pumps NA 2 2 RHR pump flow in NPSH analysis (gpm) 10,945 9,842 Number of operating CS pumps NA 4 CS pump flow in NPSH analysis (gpm) 3,830 Total flow rate in ring header for NPSH discussion (gpm) 56,894 Suction strainer debris loading assumed? NA Yes RHR Pump Evaluation Peak SP Temperature (PPT) (°F) 152.0 ha, Atmospheric pressure above SP (feet) 33.9 RHR hs, Water static height (feet) 14.1 RHR hf, suction pipe & strainer friction loss (feet) 13.5 hvap, vapor pressure @ PPT (feet) 9.2 RHR pump Available NPSH (feet) 25.3 (NPSHa = ha + hs - hf - hvap)

RHR pump Required NPSH (NPSHR3%) (feet) 20.0 RHR pump NPSH uncertainty (%) 21 RHR pump NPSHReff ({1+NPSHuncertainty} x (feet) 24.2 NPSHR3%)

RHR pump NPSH margin (NPSHa - NPSHReff) (feet) 1.1 RHR pump minimum NPSH ratio (NPSHa/NPSHR3%) NA 1.3 Time RHR pump NPSH ratio < 1.6 (hours) <1 2-334

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Table 2.6-4a ECCS Pump EPU NPSH Summary - Supplemental Evaluation (Continued)

Event RDLB Short Term Parameter Units CS Pump Evaluation Peak SP Temperature (PPT) (°F) 152.0 ha, Atmospheric pressure above SP (feet) 33.9 CS hs, Water static height (feet) 14.4 CS hf, suction pipe & strainer friction loss (feet) 13.6 hvap, vapor pressure @ PPT (feet) 9.2 CS pump Available NPSH (NPSHa = ha + hs - hf - hvap) (feet) 25.5 CS pump Required NPSH (NPSHR3%) (feet) 21.0 CS pump NPSH uncertainty (%) 21 CS pump NPSHReff ({1+NPSHuncertainty} x NPSHR3%) (feet) 25.4 CS pump NPSH margin (NPSHa - NPSHReff) (feet) 0.1 CS pump minimum NPSH ratio (NPSHa/NPSHR3%) NA 1.2 Time CS pump NPSH ratio < 1.6 (hours) <1 2-335

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Table 2.6-4b Fire Event ECCS Pump Sensitivity Cases Event Fire Event with HPMU Pump Case Type Sensitivity Event Type Special Event Parameter Units INPUTS for Containment Analysis RHR heat exchanger K-value (BTU/sec-°F) 287 Initial torus volume (nominal initial SP level) (ft3) 125,400 RHRSW temperature (°F) 95 OUTPUT from Containment Analysis Peak SP Temperature (PPT) (°F) 206.2 NPSH ANALYSIS Number of operating RHR pumps NA 1 RHR pump flow rate in safety analysis (gpm) 7,500 RHR pump flow in NPSH analysis (gpm) 7,615 Number of operating CS pumps NA 0 CS pump flow rate in safety analysis (gpm) N/A CS pump flow in NPSH analysis (gpm) N/A Total flow rate in ring header for NPSH discussion (gpm) 7,615 Suction strainer debris loading assumed NA No Peak SP Temperature (PPT) (°F) 206.2 ha, Atmospheric pressure above SP (feet) 34.6 RHR hs, Water static height (feet) 17.0 RHR hf, suction pipe and strainer friction loss (feet) 1.3 hvap, vapor pressure @ PPT (feet) 31.4 RHR pump Available NPSH (NPSHa = ha + hs - hf - hvap) (feet) 18.9 RHR pump Required NPSH (NPSHR3%) (feet) 16.0 RHR pump NPSH uncertainty (%) 0 RHR pump NPSHReff ({1+NPSHuncertainty} x NPSHR3%) (feet) 16.0 RHR pump NPSH margin (NPSHa - NPSHReff) (feet) 2.9 2-336

NEDO-33860 Revision 0 Non-Proprietary Information - Class I (Public) Insert B Figure 2.6-16 Fire Event - RHR NPSH versus Time 2-373

Insert B NEDO-33860 Revision 0 Insert C Non-Proprietary Information - Class I (Public)

Figure 2.6-18 ATWS - RHR NPSH versus Time 2-375

Insert C Figure 2.6-18a ATWS - RHR NPSH versus Time

Insert C (continued)

Figure 2.6-18b ATWS - RHR NPSH versus Time

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Table 2.8-1 Browns Ferry Key Inputs for EPU ATWS Analysis Input Variable CLTP EPU Basis Reactor Power (MWt) 3,458 3,952 Rated Value Analyzed Power (MWt) 3,293 3,952 Bounding initial condition at EPU.

OLTP value used maximizes the effect of EPU.(1)

Analyzed Core Flow (Mlbm/hr / % Rated) 76.9/ 75 101.475/ 99 Bounding initial condition on Maximum Extended Load Line Limit (MELLL) upper boundary.

Results in highest power level after recirculation pump trip.

Reactor Dome Pressure (psig) 1,035 1,035 Rated Value MSIV Closure Time (seconds) 4.0 4.0 Nominal Value High Pressure ATWS-RPT Setpoint (psig) 1,177.0 1,177.0 Analytical Limit more conservative than Technical Specification AV of 1,175 psig.

MSL Low Pressure Isolation Setpoint (psig) 825 825 Technical Specification AV RCIC Flow Rate (gpm) 540 600 Technical Specification value for EPU. Nominal Value for CLTP.

HPCI Flow Rate (gpm) 4,500 5,000 Technical Specification value for EPU. Nominal Value for CLTP.

Number of MSRVs / MSRVs OOS 13 / 1 13 / 1 Plant Configuration. Design Unchanged Number of MSRVs OOS 1 1 Plant Configuration. Design Unchanged Each MSRV Capacity at 1,090 psig 0.87 0.87 Plant Configuration. Design (Mlbm/hr) Unchanged SRV Analytical Opening Setpoints (psig) 1,179 to 1,199 1,174 to 1,194 Note 2 SLCS Injection Location Lower Plenum Lower Plenum Plant Configuration. Design Unchanged SLCS Injection Rate (gpm) 39.0 50.0 Note 3 Number of SLCS Pumps Credited for 1 1 Plant Configuration. Design Injection Unchanged Boron-10 Enrichment (Atom %) 68.1 94.0 Note 4 2-440

NEDO-33860 Revision 0 Non-Proprietary Information - Class I (Public) 259 /277 Table 2.8-1 Browns Ferry Key Inputs for EPU ATWS Analysis (continued)

Input Variable CLTP EPU Basis Sodium Pentaborate Concentration (% by 8.0 8.7 Note 5 Weight)

SLCS Liquid Transport Time (seconds) 60.0 28.5 Note 6 3

Initial Suppression Pool Liquid Volume (ft ) 123,000 122,940 Nominal value. EPU value conservative.

Initial Suppression Pool Temperature (°F) 95 95 Technical Specification RHR Heat Exchanger Effectiveness Per 223 277 Note 7 Loop (BTU/sec-°F)

Number of RHR Suppression Pool Cooling 2 4 Plant Configuration. Note 8 Loops (All Events Except Loss of Offsite Power Event)

Number of RHR Suppression Pool Cooling 2 2 Plant Configuration. Note 9 Loops During a Loss of Offsite Power Event RHR Startup Delay (seconds after T = 0) 660 660 Note 10 RHR Service Water Temperature (°F) 95 95 Technical Specification Decay Heat Correlation May-Witt May-Witt Note 11 Steam Extraction Points for Feedwater Note 12 Note 12 Plant Configuration Heaters Main Turbine Bypass Valve Capacity 3.5 3.5 Plant Configuration (Mlbm/hr)

Notes :

(1) To maximize the effect of EPU, a baseline is established at the OLTP level, assuming the current licensed equipment performance assumptions and plant parameters.

(2) In the ODYN analysis methodology, the MSRV setpoints for the ATWS analysis are statistically spread around the upper analytical limit. The EPU values are consistent with the values used for the RPV ATWS analysis contained in the FUSAR.

(3) The CLTP analysis used the current Browns Ferry Technical Specification value for SLCS flow rate. The EPU analysis uses a nominal SLCS pump flow rate. The EPU value is more conservative (lower) than the plant action level for SLCS testing specified in Browns Ferry Procedure SR 3.1.7.7 - Standby Liquid Control System Functional Test.

(4) As part of the Browns Ferry EPU implementation, TVA will perform a plant modification to increase the SLCS storage tank B-10 enrichment to 96 atom-%. 94 atom-% enrichment is credited in the EPU analysis.

(5) The CLTP analysis used the Technical Specification minimum concentration value. The EPU analysis uses a nominal concentration value that remains below the maximum allowed concentration stated in Browns Ferry Technical Specification SR 3.1.7.4.

(6) The CLTP analysis used a conservative generic transport time. The EPU analysis used a Browns Ferry plant specific value that considers the actual pipe lengths from the SLCS storage tank to the RPV lower plenum, the SLCS pipe internal diameters, and the EPU SLCS flow rate of 50 gpm. The calculated transport time was multiplied by a factor of 1.2 for additional conservatism.

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r NEDO-33860 Revision 0 of 259 BTU/sec-oF 3,800 Non-Proprietary Information - Class I (Public) conservative (7) The EPU heat exchanger effectiveness assumes a RHR flow rate of 6,500 gpm and RHR SW flow rate of 4,000 gpm through each in-service RHR heat exchanger for events that assume operation of 4 RHR loops (see Note 8 below). The heat exchange fouling resistance, used to calculate the RHR heat exchanger effectiveness value used in the ATWS analysis, is a nominal value that is supported by Browns Ferry plant specific RHR heat exchanger testing. Details concerning the fouling resistance used and the determination of the RHR heat exchanger K-value for the ATWS analysis are presented in Browns Ferry EPU LAR Attachment 39. of 277 BTU/sec-oF The EPU heat exchanger effectiveness assumes a RHR flow rate of 6500 gpm and RHR SW flow rate of 4500 gpm through each in-service RHR heat exchanger for the event that assumes operation of 2 RHR loops (see Note 9 below). The heat exchange fouling resistance, used to calculate the RHR heat exchanger effectiveness value used in the ATWS analysis, is a conservative value that is supported by Browns Ferry plant specific RHR heat exchanger testing. Details concerning the fouling resistance used and the determination of the RHR heat exchanger K-value for the ATWS analysis are presented in the Browns Ferry EPU LAR Attachment 39. 3,800 (8) The RHR suppression pool cooling configuration does not change for EPU. An RHR loop is defined as one RHR pump, one RHR heat exchanger and RHR SW flow of 4,000 gpm through the RHR heat exchanger. For ATWS events other than LOOP, the plant operators would be directed by plant EOIs to maximize suppression pool cooling. Because there is no concurrent event on the non-ATWS unit, four RHR loops are assumed available for suppression pool cooling in the ATWS unit.

(9) The RHR suppression pool cooling configuration does not change for EPU. An RHR loop is defined as one RHR pump, one RHR heat exchanger and RHR SW flow of 4,500 gpm through the RHR heat exchanger. For the LOOP ATWS event, operators will be directed by EOIs to maximize suppression cooling. Because there is also a LOOP (without ATWS) on the remaining two Browns Ferry units, only two RHR loops are assumed available for suppression pool cooling on the ATWS unit.

(10) The RHR startup delay time assumes no operator action for containment cooling for the first 10 minutes of the event with an additional 60 seconds for RHR to reach full effectiveness.

(11) The May-Witt decay heat correlation is used in the suppression pool temperature calculation following reactor shutdown. The May-Witt decay heat correlation yields a conservative pool heat-up compared to the 1979 ANS 5.1 + 2 curve.

(12) The steam extraction points for feedwater heaters are downstream of the MSIVs, such that FW heating is lost following MSIV isolation. The specific extraction points are as follows:

HP turbine exhaust to FW heater number 1 (highest pressure FW heater)

LP turbine stage 7 to FW heater number 2 LP turbine stage 8 to FW heater number 3 LP turbine stage 10 to FW heater number 4 LP turbine stage 12 to FW heater number 5 (lowest pressure FW heater) 2-442

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Table 2.8-2 Browns Ferry Containment Results for ATWS Analysis MSIVC Event Acceptance Criteria Acceptance EPU Result 171.8 Criteria Peak Suppression Pool Temperature (F) 281.0 171.7 Peak Containment Pressure (psig) 56.0 8.0 PRFO Event Acceptance Criteria Acceptance EPU Result Criteria 171.8 Peak Suppression Pool Temperature (F) 281.0 171.6 Peak Containment Pressure (psig) 56.0 8.0 8.1 LOOP Event Acceptance Criteria Acceptance EPU Result Criteria Peak Suppression Pool Temperature (F) 281.0 173.3 Peak Containment Pressure (psig) 56.0 8.7 IORV Event Acceptance Criteria Acceptance EPU Result Criteria Peak Suppression Pool Temperature (F) 281.0 142 143.3 Peak Containment Pressure (psig) 56.0 4.0 4.2 2-443

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Table 2.8-4 PRFO Sequence of Events EPU BOC Event EPU EOC Event Event Response Time (sec) Time (sec)

Turbine Control Valves (TCV) and Bypass 0.1 0.1 Valves Start Open MSIV Closure Initiated by Low Steam Line 15.7 14.9 Pressure MSIVs Fully Closed 19.7 18.9 Peak Neutron Flux 21.8 19.4 Opening of the First Relief Valve 21.6 20.8 High Pressure ATWS Setpoint 21.8 21.0 Recirculation Pumps Trip 22.2 21.6 Peak Heat Flux 22.5 21.7 Peak Vessel Pressure 28.6 27.7 Feedwater Reduction Initiated 46.0 46.0 BIIT Reached 56.0 55.0 SLCS Pumps Start (1) 141.8 141.0 RHR Cooling Initiated 660 660 RPV Water Level Increased after Hot 827-1,027 827-1,027 Shutdown Boron Weight Injected Peak Suppression Pool Temperature 889 889 988 Hot Shutdown Achieved (Neutron Flux 1,077 1,067 Below 0.1% for More Than 100 seconds)

Note:

1. SLCS injection is the later time of either 1) two minutes after the high-pressure recirculation pump trip or 2) when the suppression pool temperature reaches the BIIT. For Browns Ferry, there is no automatic actuation of the SLCS. Actuation of the SLCS occurs by operator manipulation of key-lock switches on the main control room front panel.

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Table 2.8-5 LOOP Sequence of Events EPU EOC Event Time Event Response (sec)

Main Turbine Generator Tripped 0.0 Recirculation Pumps Trip 0.0 Feedwater Pump Coastdown Initiated due to Tripping on 0.0 LOOP of Motor Driven Condensate and Condensate Booster Pumps Peak Neutron Flux 0.5 Peak Heat Flux 0.6 Opening of the First Relief Valve 0.9 High Pressure ATWS Setpoint 1.1 MSIV Isolation Initiates 2.0 MSIVs Fully Closed 6.0 Peak Vessel Pressure 7.1 BIIT Reached 40.0 40 SLCS Pumps Start (1) 121.0 RHR Cooling Initiated 660 121 RPV Water Level Increased after Hot Shutdown Boron 806-1,006 Weight Injected Hot Shutdown Achieved (Neutron Flux Below 0.1% for 1,236 More Than 100 seconds)

Peak Suppression Pool Temperature 9,192 Note:

1. SLCS injection is the later time of either: 1) two minutes after the high-pressure recirculation pump trip or 2) when the suppression pool temperature reaches the BIIT. For Browns Ferry, there is no automatic actuation of the SLCS. Actuation of the SLCS occurs by operator manipulation of key-lock switches on the main control room front panel.

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Table 2.8-6 IORV Sequence of Events EPU EOC Event Time Event Response (sec)

Inadvertent Opening of One Relief Valve 0.0 Peak Neutron Flux 0.0 Peak Vessel Pressure 0.0 Peak Heat Flux 0.2 BIIT Reached 434 Recirculation Pumps Tripped 434 Feedwater Reduction Initiated 434 SLCS Pumps Start (1) 434 RHR Cooling Initiated 660 RPV Water Level Increased after Hot Shutdown Boron 1,119-1,319 Weight Injected Hot Shutdown Achieved (Neutron Flux Below 0.1% for 1,273 More Than 100 seconds)

Peak Suppression Pool Temperature 5,379 6,127 Note:

1. SLCS injection is the later time of either: 1) two minutes after the high-pressure recirculation pump trip or 2) when the suppression pool temperature reaches the BIIT. For Browns Ferry, there is no automatic actuation of the SLCS. Actuation of the SLCS occurs by operator manipulation of key-lock switches on the main control room front panel.

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Figure 2.8-1 EPU MELLLA BOC MSIVC 2-448

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Figure 2.8-2 EPU MELLLA BOC MSIVC 2-449

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Figure 2.8-3 EPU MELLLA BOC MSIVC 2-450

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Figure 2.8-4 EPU MELLLA BOC PRFO 2-451

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Figure 2.8-5 EPU MELLLA BOC PRFO 2-452

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Figure 2.8-6 EPU MELLLA BOC PRFO 2-453

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Figure 2.8-7 EPU MELLLA EOC MSIVC 2-454

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Figure 2.8-8 EPU MELLLA EOC MSIVC 2-455

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Figure 2.8-9 EPU MELLLA EOC MSIVC 2-456

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Figure 2.8-10 EPU MELLLA EOC PRFO 2-457

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Figure 2.8-11 EPU MELLLA EOC PRFO 2-458

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Figure 2.8-12 EPU MELLLA EOC PRFO 2-459

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Figure 2.8-13 EPU MELLLA EOC LOOP 2-460

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Figure 2.8-14 EPU MELLLA EOC LOOP 2-461

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Figure 2.8-15 EPU MELLLA EOC LOOP 2-462

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Figure 2.8-16 EPU MELLLA EOC IORV 2-463

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Figure 2.8-17 EPU MELLLA EOC IORV 2-464

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Figure 2.8-18 EPU MELLLA EOC IORV 2-465

NEDO-33860 Revision 0 Non-Proprietary Information - Class I (Public) by approximately 15% of the CLTP injection flow but will have minimal effect on the plant operators.

Hot Shutdown Boron Weight (HSBW) and Cold Shutdown Boron Weight (CSBW) - The percentage of tank level that must be injected will change due to the increase in boron enrichment.

Decay Heat Removal Pressure (DHRP) - The DHRP will be affected by the increase in decay heat loading. The pressure will increase by approximately 12 psig but will have minimal effect on the plant operators.

The following EOIs and Severe Accident Management Guidelines (SAMG)s are planned to be revised as a result of EPU (The modifications mentioned can be found in EPU LAR 7.):

EOI-1 FLOWCHART and Bases, RPV CONTROL, are affected by the installation of the Emergency High Pressure Makeup Pump (EHPMP) modification.

EOI-1A FLOWCHART and Bases, ATWS RPV CONTROL, are affected by changes to the Cold and Hot Shutdown Boron Weights and the installation of the EHPMP modification.

EOI-2 FLOWCHART and Bases, PRIMARY CONTAINMENT CONTROL, are affected by the changes to the Heat Capacity Temperature Limit and the Pressure Suppression Pressure.

EOI-5, CURVES AND CAUTIONS, is affected by changes to the Heat Capacity Temperature Limit and the Pressure Suppression Pressure.

C-2 FLOWCHART and Bases, EMERGENCY RPV DEPRESSURIZATION, are affected by the changes to the Decay Heat Removal Pressure.

C-2A FLOWCHART and Bases, ATWS EMERGENCY RPV DEPRESSURIZATION, are affected by the changes in the Cold Shutdown Boron Weight and the Decay Heat Removal Pressure.

C-4 FLOWCHART and Bases, RPV FLOODING, are affected by changes to the Decay Heat Removal Pressure and by installation of the EHPMP modification.

C-4A FLOWCHART and Bases, ATWS RPV FLOODING, are affected by changes to the Decay Heat Removal Pressure and by installation of the EHPMP modification.

Emergency High Pressure Makeup Pump, will be developed to allow operators to make-up to the RPV from the CST with a newly installed non-safety-related pump.

(Future new procedure to be developed as part of implementation of NFPA 805.)

Hardened Wetwell Vent, will be developed to allow operators to vent the suppression pool air space through a hardened vent pipe that exhausts above the Reactor Building roof. (Future procedures will be developed to implement the requirements for the mitigation of beyond-design basis events.)

SAMG-1, PRIMARY CONTAINMENT FLOODING, is affected by changes to the Pressure Suppression Pressure, the Minimum Debris Retention Injection Rate, and by installation of the EHPMP modification. and 2-509

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Abnormal Operating Procedures (AOPs) at Browns Ferry are defined as Abnormal Operator Instructions (AOIs), Annunciator Response Procedures (ARPs), Fire Safe Shutdown (FSSs), and select sections of Operator Instructions (OIs).

AOI-3-1, Loss of Reactor Feedwater or Reactor Water Level High/Low, will be revised to no longer require operators to immediately lower reactor power level to 80%, by reducing Recirculation Pump flow, in order to avoid a low reactor water level scram.

The Condensate/Condensate Booster/Feedwater pumps were all changed from 1/3 capacity to 1/2 capacity pumps. The loss of certain single pumps or pump combinations, at EPU conditions, can require a power reduction of as much as 7% to re-establish desired NPSH ratios or to reduce the Condensate Booster Pump horsepower (still within its service factor) back to within its nameplate rating.

AOI 47-3, Loss of Condenser Vacuum, was revised to reflect the modification that replaced condenser low vacuum pressure switches with pressure transmitters. These transmitters provide Turbine Trip and Bypass Valve Trip inputs to the Electro-Hydraulic Control System and provide condenser vacuum indication in the CR via an existing recorder. Now the trips and operator indication will originate from the same transmitter and are therefore aligned.

AOI-57-1A, Loss of Offsite Power (161 and 500 kV)/Station Blackout (SBO), will be revised for the following:

1) To incorporate a time sensitive operator action:

a. Crosstie of the Containment Atmospheric Dilution system to the Drywell Control Air System approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the SBO scenario.

2) To incorporate changes to the SBO HCTL and SBO Pressure Suppression Pressure curves for the SBO scenario.

AOI 57-1E, Grid Instability, will be revised to take into account the uprated Turbine/Generator output capability for each of the units:

Unit 1: 1,330 MWe at 0.95 Power Factor Units 2 and 3: 1,332 MWe at 0.93 Power Factor OI-47, Turbine Generator System, required revision to reflect that the rewound Main Generator will be required to be tripped within 60 seconds of a loss of Stator Cooling Water. The previous time requirement was 70 seconds.

EOIs and AOPs will also be rescaled as required to reflect the power uprate.

2.11.1.2 Changes to Operator Actions Sensitive to Power Uprate Most abnormal events result in automatic plant shutdown (scram). Some abnormal events result in SRV actuation, ADS actuation and/or automatic ECCS actuation. All analyzed events result in safety-related SSCs remaining within their design limits. EPU does not change any automatic safety function. Changes to subsequent operator actions are as follows.

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NEDO-33860 Revision 0 Non-Proprietary Information - Class I (Public) 2.11.1.2.1 Changes for Design Basis Accidents (DBA) and Events An EHPMP will be installed as an additional option for operators to provide water to the RPV in the EOI and SAMG procedures.

AOI-57-1A, Loss of Offsite Power (161 and 500 kV)/Station Blackout (SBO), will be revised to incorporate a time sensitive action as a result of EPU:

1) Crosstie of the Containment Atmospheric Dilution system to the Drywell Control Air System approximately 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> into the SBO scenario.

This time sensitive operator action is a simple task, requires a small time duration to perform

(< 10 minutes), is performed in the control room (CR), and will easily be able to be successfully performed within the 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> required timeframe. As such, time validation of this action is not necessary.

2.11.1.2.2 Fire Safe Shutdown (FSS) Events The purpose of the FSS procedures is to provide supplemental instruction, including the manual recovery actions needed, in conjunction with the EOIs should an EOI initiating condition exist, to ensure safe shutdown of Unit 1, Unit 2 or Unit 3, or all three units if conditions require, in the event of a disabling fire event. Existing plant procedures are written to support an Appendix R fire at EPU conditions. The NFPA 805 FSS procedures will allow the plant to use all available equipment until proven unreliable with the exception of annunciators. Credited equipment is equipment which has the highest probability to remain functional in the fire event. Use of the credited, or preferred, equipment is desired in that it is least likely to be affected due to a fire related event. 7 of the EPU LAR provides a listing and discussion of the modifications planned for EPU. The effect of these modifications on the Browns Ferry Fire Protection Program will be evaluated, in accordance with TVAs configuration change process, prior to EPU implementation. Per the process, these modifications will be evaluated to assure the changes do not affect the approved Fire Protection Program and will not adversely affect the ability to achieve and maintain safe shutdown in accordance with the current Browns Ferry license conditions and procedures.

As the FSSs are symptom based, the implementation of EPU does not change how the FSSs will be implemented or executed. The EHPMP will be available as an additional option for operators to provide water to the RPV. No operator actions need to be performed more quickly as a result of EPU implementation.

2.11.1.2.3 Anticipated Transient Without Scram Event An EHPMP is not credited but will be available as an additional option for operators to provide water to the RPV. There will also be changes to the cold and hot shutdown boron weights and the decay heat removal pressure. No operator actions need to be performed more quickly as a result of EPU implementation.

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NEDO-33860 Revision 0 Non-Proprietary Information - Class I (Public) 2.11.1.2.4 Conclusion The changes to Browns Ferry operator actions, as a result of the EPU, are small in number. There is only one time sensitive operator action. This action is a simple task, requires a small time duration to perform (< 10 minutes), is performed in the control room and will easily be able to be successfully performed within the two hour required timeframe. The changes to operator actions will be reflected in the procedures and the operators will receive appropriate classroom and/or simulator training prior to EPU implementation. There are no new or revised operator workarounds as a result of EPU.

2.11.1.3 Changes to Control Room Controls, Displays and Alarms Changes to the CR are prepared in accordance with the plant design change process. Under this process, a Human Factors engineering review is performed for changes associated with the Browns Ferry CR. The change process also requires a review by Operations and Training personnel. Results of these reviews, including simulator effect and training requirements, are incorporated into the engineering change package and tracked to completion by the design change process.

The following changes have been/will be made to the CR Controls, Displays and / or Alarms resulting from EPU:

Controls will be installed in the Control Room that will allow the operators to start an EHPMP that will take suction from the CST and discharge to the RPV.

The Turbine First Stage Pressure Scram Bypass setpoint and associated alarm will be changed.

Condenser pressure transmitters have been installed and the Turbine Trip and Turbine-Bypass Trip signal and alarms originate from these instruments as does the control room indication for the operators.

Changes will be made to the Rod Worth Minimizer to reflect EPU conditions. A new runback for the Recirculation Pumps to 75% speed will occur on a scram signal to prevent water level from reaching the Level 2 setpoint.

Removed the SJAE auto-start capability and replaced the HS-150/152 three-position switches with two-position switches.

The control switches for the former Moisture Separator Drain Pumps that were removed now operate Moisture Separator Isolation Valves. A seventh control switch allows condensate to be injected into the Moisture Separator Drain Line.

Controls for an additional Bus Duct Cooler Fan have been installed.

TS instruments for instrument and control systems are affected by EPU as described in the Enclosure to the EPU LAR and Attachments 2 and 3.

2.11.1.3.1 Conclusion The changes to Browns Ferry CR interfaces as a result of the EPU do not significantly affect operator human performance. Operator training for changes to CR interfaces, alarms, and 2-512

ENCLOSURE 4 BFN EPU LAR, Attachment 39, Revision 1, RHR Heat Exchanger K-values Utilized in EPU Containment

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses The following summarizes the changes to EPU LAR Attachment 39 associated with this LAR Supplement:

• Use of nominal inputs for the fire event, special event containment analyses. For the three special events discussed in SECY-11-0014, ATWS, Station Blackout, and Fire event, nominal inputs are used only for the fire event in the BFN EPU containment analyses.

• Another change associated with the fire event is to eliminate previous discussions of EPU design and EPU nominal fouling resistances and instead use a single value for the EPU design fouling resistance (acceptance criterion) for all events, 0.001562 hr-ft2-

°F/Btu, that is based on the most limiting event, the NFPA 805 fire event.

• Another change associated with the fire event is the use of test results, with test uncertainty included, for comparison to the EPU design fouling resistance (acceptance criteria)

• A reduction in the ATWS non-LOOP event RHR heat exchanger K-value from 277 Btu/sec to 259 Btu/sec due to reduced RHRSW flow through each heat exchanger when all four heat exchangers in one unit are placed in service.

• Reporting the results of all fourteen RHR heat exchanger tests for the twelve RHR heat exchangers.

• Removal of the Emergency High Pressure Makeup Pump, previously discussed in association with the fire event, from this LAR Supplement.

The following provides a more detailed summary and justification for the changes to EPU LAR 9 associated with this LAR Supplement:

• For the NFPA 805 fire event containment analysis, use the following nominal inputs:

o 88°F RHRSW temperature instead of 92°F o 92°F initial suppression pool temperature instead of 95°F

• For the NFPA 805 fire event containment analysis, use an RHR heat exchanger K-value of 290 Btu/sec-°F instead of 307 Btu/sec-°F

• Use of the above inputs in the NFPA 805 containment analysis and a more rigorous benchmarking to design documents (heat exchanger specification sheets and drawings) resulted in:

o a change in the peak suppression pool temperature to 207.7°F instead of the previous 208.0°F o an allowable fouling resistance of 0.001562 hr-ft2-°F/Btu with the NFPA 805 licensing condition tube plugging limit maintained at 4.57% (77 tubes plugged)

• A reduction in the ATWS non-LOOP event RHR heat exchanger K-value from 277 Btu/sec to 259 Btu/sec due to reduced RHRSW flow through each heat exchanger when all four heat exchangers in one unit are placed in service.

• The ATWS (non-LOOP) limiting event was previously reported as MSIVC-EOC with a peak suppression pool temperature of 171.7°F. With the change in the ATWS (non-LOOP) event RHR heat exchanger K-value from 277 Btu/sec-°F to 259 Btu/sec-°F, the peak suppression pool temperature changes to 171.8°F for both the MSIVC-EOC and the PRFO-EOC events.

1.0 Introduction This attachment provides EPU event-specific RHR heat exchanger K-values and heat removal rates for postulated events. The heat exchanger K-value determinations in this attachment Att 39-1

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses confirm that the specific K-value used in each containment analysis is appropriate and suitable for the expected EPU conditions.

The Browns Ferry Nuclear Plant original design value for the RHR heat exchanger effectiveness K-value is 223 BTU/sec-°F.

In Reference 4.32, TVA previously submitted a License Amendment Request (LAR) to transition the licensing basis for the Browns Ferry Nuclear Plant to National Fire Protection Association standard 805 (NFPA 805) which, in (Ref. 4.32) Attachment X and subsequent responses to NRC requests for additional information (RAIs) (Ref. 4.43), included a change to the RHR heat exchanger K-value from 223 BTU/sec-°F to 265 BTU/sec-°F and provided the basis for the change to the K-value. The extended power uprate (EPU) design value used for the containment analyses under the design basis accident (DBA) loss of coolant accident (LOCA),

for containment heat removal is 265 BTU/sec-°F, which is consistent with the NFPA 805 LAR (CLTP) K-value of 265 BTU/sec-°F. However, this EPU LAR attachment also identifies changes to specific values reported in the NFPA 805 LAR. These changes are addressed in the following paragraph.

In the NFPA 805 transition LAR (Ref. 4.32) and associated RAI responses (Ref. 4.43) the RHR heat exchanger overall fouling resistance corresponding to a K-value of 265 BTU/sec-°F and a peak suppression pool temperature of 187.4°F (DBA-LOCA temperature) was reported to be 0.001517 hr-ft2-°F/BTU. This fouling resistance corresponds to a K-value of 284.5 BTU/sec-°F at the CLTP NFPA 805 fire event conditions. However, for conservatism, a K-value of 270 BTU/sec-°F is used in the CLTP NFPA 805 fire event analyses. This attachment provides a re-calculated DBA-LOCA EPU design fouling resistance associated with a K-value of 265 BTU/sec-°F for a peak suppression pool temperature of 179.0°F. This change in the peak suppression pool temperature from 187.4°F to 179.0°F results in a slightly different EPU design fouling resistance, 0.001521 hr-ft2-°F/BTU.

Additionally, the fire event analysis was performed at EPU conditions using nominal inputs of 88°F RHRSW temperature and 92°F initial suppression pool temperature. The previous (CLTP) fire event analysis had been performed using 92°F RHRSW temperature and 95°F initial suppression pool temperature. The NFPA 805 fire event tube plugging limit, 77 tubes mechanically plugged, was maintained for EPU. With the above changes in the nominal inputs and with 77 tubes mechanically plugged the EPU fire event allowable fouling resistance was determined to be 0.001562 hr-ft2-°F/BTU. Consequently, tThe values of DBA-LOCA peak suppression pool temperature and design fouling resistance that were provided in the NFPA 805 LAR (Ref. 4.32) are superseded by the EPU values provided herein, 179.0°F and 0.0015210.001562 hr-ft2-°F/BTU, respectively, and reflecting updated EPU analyses.

Att 39-2

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses EPU licensing basis RHR heat exchanger K-values are as follows:

EPU event K-value (BTU/sec-°F)

DBA-LOCA 265 Small Break LOCA 265 Loss of Shutdown Cooling 265 Stuck Open Relief Valve 265 Station Blackout (SBO) 265 ATWS (LOOP event)* 277 ATWSATWS (all ATWS events other 259 than LOOP)*

Shutdown of Non-Accident Unit 302 Fire 307*290

• The limiting ATWS event, with respect to ECCS pump NPSH, is where 4 RHR heat exchangers are in service with a RHRSW flow rate of 3800 gpm. However, a K-value of 277 is still used for the limiting peak suppression pool temperature ATWS event where 2 RHR heat exchangers are in service with an RHRSW flow rate of 4500 gpm. In addition, the fire event was analyzed using the Emergency High Pressure Make-Up (EHPMU) pump and an RHR heat exchanger K-value of 287 BTU/sec-°F. This represents a defense-in-depth demonstration and is not a licensing basis analysis.

The Browns Ferry Nuclear Plant Updated Final Safety Analysis Report (UFSAR) (Ref. 4.54) provides additional descriptive information concerning the Residual Heat Removal (RHR) heat exchangers in UFSAR sections 1.6.2.12 (Residual Heat Removal System (Containment Cooling)), 4.8.6.2 (Containment Cooling), 4.8.6.3 (Low Pressure Coolant Injection), Table 4.8-1 (Residual Heat Removal System Equipment Design Data) and Table 5.2-1 (Principal Design Parameters and Characteristics of Primary Containment).

2.0 Computations and Analysis 2.1 Methodology The methodology utilizes an Excel spreadsheet platform, and standard engineering formulas, to solve for the RHR heat exchanger K-value based on the input parameter values.

The formula for heat exchanger effectiveness is a function of the overall heat transfer coefficient, the effective heat transfer area, and other parameters identified below. The Browns Ferry Nuclear Plant RHR heat exchangers are identical single shell pass, two tube pass Tubular Exchanger Manufacturers Association, Inc. (TEMA) type CES heat exchangers. The effectiveness for this type heat exchanger is:

Att 39-3

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses 1

NTU (1+CR 2 )

0.5 e = 2

• 1 + CR + (1 + CR )

• 2 0.5 1+e (Equation 1 - Ref. 4.1)

(1+CR 2 )

0 . 5 NTU 1e Where:

e = heat exchanger effectiveness CR = heat capacity ratio

= Cmin/Cmax Cmin = minimum mass flow rate times fluid heat capacity product, BTU/hr-oF Cmax = maximum mass flow rate times fluid heat capacity product, BTU/hr-oF NTU = number of transfer units

= UA/Cmin U = overall heat transfer coefficient, BTU/hr-ft2-°F A = effective heat transfer area, ft2 Equation 1 was used to determine fouling resistance and/or K-values for various state points for the analyzed event sequences.

Tables 1, 2, and 3, below, provide the results from specific evaluations using the overall fouling resistance and the corresponding heat exchanger K-value for each of the fouling resistances in Table 4. The specified fouling resistances cover a wide range from the original supplied RHR heat exchanger design fouling resistance (0.0028000.002801 hr-ft2-

°F/BTU) to a clean RHR heat exchanger fouling resistance (0.000000 hr-ft2-°F/BTU).

Specifically, tThe EPU design fouling resistance, 0.0015210.001562 hr-ft2-°F/BTU; the EPU nominal fouling resistance, 0.001097 hr-ft2-°F/BTU and the fouling resistances determined from six different Browns Ferry Nuclear Plant RHR heat exchanger tests performed to date are some for the NFPA 805 fire event with 77 tubes assumed plugged is of the most more significant of the fouling resistancess.

Table 4 provides a comparison summary of heat exchanger heat transfer rates capability at different fouling resistances which can be compared to the followingand corresponding K-values for the different EPU K-values that were used in the event containment analyses:

• 259 BTU/sec-°F (ATWS with 4 RHR heat exchangers in service), 277 BTU/sec-°F (ATWS with 2 RHR heat exchangers in service), 265 BTU/sec-°F (DBA-LOCA, Small Break LOCA, Loss of Shutdown Cooling, Stuck Open Relief Valve and SBO), 302 BTU/sec-°F (Shutdown of Non-Accident Unit), and 287 290 BTU/sec-°F (fire event defense-in-depth demonstration case) are based onless than or equal to the heat exchanger heat transfer capability at the EPU design fouling resistance, 0.0015210.001562 hr-ft2-°F/BTU.

307 BTU/sec-°F (fire event licensing basis) is based on the EPU nominal fouling resistance, 0.001097 hr-ft2-°F/BTU.

277 BTU/sec-°F for the ATWS-MSIVC-EOC event corresponds to a nominal fouling resistance of 0.001220 hr-ft2-°F/BTU.

2.2 Analysis RHR heat exchanger K-values used in the containment analyses are identified for each event within the applicable Browns Ferry Nuclear Plant Power Uprate Safety Analysis Report Att 39-4

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses (PUSAR) section. Unless otherwise noted and discussed below for ATWS and fire events, aAll events were evaluated at the EPU design fouling resistance (same heat exchanger material condition). For the same EPU design fouling resistance, differences in RHR and Residual Heat Removal Service Water (RHRSW) flow rates from one event to another may result in different RHR heat exchanger K-values (see Table 4).

The DBA LOCA discussion provides the basis for the change in the pre-EPU and the EPU design fouling resistance and the corresponding EPU K-value of 265 BTU/sec-°F. The Shutdown of the Non-Accident Unit analysis K-value, 302 BTU/sec-°F, is bounded by the heat transfer capability at also based on the EPU design fouling resistance, 0.0015210.001562 hr-ft2-°F/BTU (see Table 4). The EPU fire event analysis K-value was based on the EPU nominal design fouling resistance with 77 tubes plugged. , 0.001097 hr-2 ft -°F/BTU and tThe ATWS event analyses K-values are also bounded by the heat exchanger heat transfer capability at were based on a the nominal design fouling resistance or K-value that lies between the EPU design and EPU nominal fouling resistances. Details are provided in the following discussion of the events.

DBA-LOCA - PUSAR Section 2.6.5.1 The EPU design basis fouling resistance, 0.001562 hr-ft2-°F/Btu, is based on the conditions from the limiting EPU DBA-LOCANFPA 805 fire event. The EPU design fouling resistance at DBA-LOCA conditions where Specifically, the RHR flow rate is of 6500 gpm at a peak suppression pool temperature (heat exchanger inlet temperature) of 179.0°F, the RHRSW flow rate isof 4000 gpm at 95°F RHRSW temperature, with up to 4.575 percent tube plugging, results in an and an RHR heat exchanger K-value of of 265266.4 BTU/sec-°F, as shown in Table 1. This K-value, 266.4, is used to calculate the DBA-LOCA EPU heat transfer capability, 80,559,360 Btu/hr shown in Table 4. are used to determine the EPU design basis fouling resistance. This K-value and heat transfer capability exceeds the DBA-LOCA K-value (265 Btu/sec-°F) and heat transfer capability (80,136,000 Btu/hr) assumed in the DBA-LOCA containment analysis. Equation 1 is used to solve for fouling resistance, based on a given K-value or, to calculate a K-value based on a specified fouling resistance, provided the other state points are known.

The EPU containment response analyses were performed for the long-term DBA-LOCA using a RHR heat exchanger K-value of 265 BTU/sec-°F compared to 223 BTU/sec-°F used in previous analyses. The previous value of 223 BTU/sec-°F was based on the RHR heat exchanger specification sheets, which specified the design fouling resistances. Browns Ferry Nuclear Plant plant-specific RHR heat exchanger testing (Ref. 4.43, including SCVB RAI-5) has shown there is substantial margin between the design and actual (tested) heat exchanger fouling resistances.

Some of this large margin was used to increase RHR heat exchanger heat transfer capability from that specified in the original design to a K-value of 265 BTU/sec-°F. The corresponding EPU design fouling resistance at the DBA-LOCA conditions identified above is 0.0015210.001562 hr-ft2-°F/BTU.

Under certain scenarios, higher RHR and/or RHRSW flows may be achieved than those used to establish the EPU design basis fouling resistance. Using minimum flow rates in determining the RHR heat exchanger heat removal capability embeds conservatism into the containment analysis results. A lower containment heat removal rate results in higher Att 39-5

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses suppression pool temperatures; thus, the analysis is conservative with respect to suppression pool temperature.

Other events, including the small break LOCA, loss of RHR shutdown cooling, stuck open relief valve (SORV), and station blackout, were evaluated using an RHR heat exchanger K-value of 265 BTU/sec-°F and are described in PUSAR Section 2.6.5.2.

Shutdown of the Non-Accident Unit - PUSAR Section 2.6.5.1 An RHR heat exchanger K-value of 302 BTU/sec-°F was used in the analysis of the shutdown of the non-accident unit. This value represents is bounded by the K-value resulting from a design basis fouling resistance, RHR flow of 9700 gpm at the peak suppression pool temperature of 185.1°F, RHRSW flow of 4500 gpm at 95°F, and 4.575 percent tube plugging. These parameters are consistent with the analysis inputs/results described in PUSAR Section 2.6.5.1.

Fire Event - PUSAR Section 2.5.1.4.2 The fire event analysis assumes one RHR heat exchanger available for event mitigation. All other RHR heat exchangers are assumed to be not available which restricts containment heat removal capability. Of all the events analyzed, because of restrictive/conservative assumptions, such as the availability of only one RHR pump and heat exchanger, the fire event results in the highest peak suppression pool temperature. In-situ RHR heat exchanger performance testing (results previously submitted under Browns Ferry Nuclear Plant NFPA 805 LAR (Ref. 4.32) Attachment X and associated RAI responses (Ref. 4.43))

demonstrate that the actual heat removal capability of RHR heat exchangers exceeds that assumed in the fire event containment analysis.

An RHRSW flow of 4500 gpm is achievable for the fire event. The combination of increased RHRSW flow and the use of a realistic or nominalthe design heat exchanger fouling resistance, 0.0010970.001562 hr-ft2-°F/BTU, corresponds to an increase in the EPUa fire event heat exchanger K-value to of 307 290 BTU/sec-°F with 77 tubes mechanically plugged. As discussed in PUSAR Section 2.5.1.4.2, a fire event containment analysis was performed using the K-value of 307 290 BTU/sec-°F which resulted in a peak suppression pool temperature of 208.0207.7°F.

SECY-11-0014 (Ref. 4.6) provides guidance on the use of realistic or nominal inputs for special event such as a fire event. For the Browns Ferry Nuclear Plant EPU, the fire event is the only event where nominal inputs were used. For the NFPA 805 fire event the nominal inputs used were an RHRSW (river) temperature of 88°F, instead of the Tech Spec value of 95°F, and an initial suppression pool temperature of 92°F, instead of the Tech Spec value of 95°F. Other inputs in the fire event containment analysis were conservatively assumed at Tech Spec values (e.g. initial suppression pool level). The basis for using 88°F as the RHRSW temperature instead of the UHS temperature of 95°F specified in TS surveillance requirement 3.7.2.1 is that historical plant operating data shows that actual temperatures are less than or equal to 88°F more than 95% of the time. The RHRSW temperature of 95°F is used in the design basis LOCA containment analyses.

The justification of UHS or RHRSW temperature of 88°F and an initial suppression pool temperature of 92°F as realistic values is that evaluation of 7 years of Tennessee River and suppression pool water temperature data between January 1, 2008 and January 1, 2015, Att 39-6

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses determined that these temperatures bound plant operating temperatures more than 95% of the time. This serves as an appropriate basis for establishing a realistic upper bound for these temperatures.

SECY-11-0014 (Ref. 4.8) provides guidance on the use of realistic or nominal inputs for special events such as a fire event. For the Browns Ferry Nuclear Plant EPU, when a nominal K-value was selected or established, the heat exchanger test data available for comparison to the selected realistic or nominal value was limited to that reported in the NFPA 805 LAR (Ref. 4.3) Attachment X and associated RAI responses (Ref. 4.4). These reported as tested fouling resistances of 0.0005164 and 0.000674 hr-ft2-°F/BTU, are shown on Figure 1, and resulted in a large difference between the test results and the EPU design (0.001521 hr-ft2-°F/BTU) fouling resistance. To provide confidence in establishing a reasonable realistic or nominal heat exchanger fouling resistance, industry literature was consulted for establishing the realistic or nominal fouling resistance.

EPRI Test Report 107397, Service Water Heat Exchanger Testing Guidelines, (Ref. 4.2) section 3.4.2 provides guidance in establishing an acceptance criterion for test uncertainty to determine whether a given test result is acceptable. The guidance suggests that an initial acceptance criterion can be established in terms of the heat transfer rate, at +/- 25 percent of the difference between the design and clean (zero fouling resistance) heat transfer rates.

This provided a reasonable bound for establishing the expected deviation from a design heat transfer rate to an actual or tested heat transfer rate. The realistic or nominal fouling resistance was determined by taking 25 percent of the difference between the EPU fire event design heat transfer rate and the clean heat transfer rate and adding that to the EPU design heat transfer rate to obtain the heat transfer rate corresponding to the realistic or nominal fouling resistance. The results of this computation are provided in Table 4 where this process was followed to determine the fire event K-value, 308 BTU/sec-°F. The value chosen for the RHR heat exchanger fire event nominal K-value was 307 BTU/sec-°F, which is within the range from 287 BTU/sec-°F (EPU design) to 308 BTU/sec-°F.

The nominal fire event K-value of 307 BTU/sec-°F represents an approximate 7 percent increase from the EPU fire event design K-value of 287 BTU/sec-°F (computed using the EPU design fouling resistance, 0.001521 hr-ft2-°F/BTU, 7500 gpm RHR flow at 208.0°F, 4500 gpm RHRSW flow at 92.0°F and 4.57 percent of the tubes plugged). This approximate 7 percent increase in the EPU fire event design K-value to the EPU fire event nominal K-value compared well with an NRC staff assessment (Ref. 4.6) where conservative and realistic RHR heat exchanger K-values were determined and compared at 147 and 158 BTU/sec-°F, respectively. The increase from 147 BTU/sec-°F to 158 BTU/sec-°F constitutes an approximate 7.5 percent increase in the conservative or design basis K-value, which compares well to the Browns Ferry Nuclear Plant 7 percent increase described above.

Using the nominal fire event K-value of 307 BTU/sec-°F, Equation 1 was used to solve for the resulting EPU nominal fire event fouling resistance, 0.001097 hr-ft2-°F/BTU (see Table 3). This fouling resistance is used for comparison to Browns Ferry Nuclear Plant RHR heat exchanger test results graphically presented in Figure 1. Figure 1 provides the results from the tests performed to date.

Comparison of the Browns Ferry Nuclear Plant RHR heat exchanger test results (see Tables 5, 6 and 7) to the EPU design and nominal fouling resistances shown provided in Figure 1Table 4 leads to two clearthe observations: that (1) the results from the 2A RHR heat exchanger as-found test (two years since last cleaning) closely match the EPU fire event Att 39-7

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses nominal fouling resistance and, (2) the 2A RHR heat exchanger as-left test results (0 years since last cleaning) when compared with the 2A RHR heat exchanger as-found test (two years since last cleaning) results and the other three heat exchanger test results show that the 2A RHR heat exchanger performance, when expressed in terms of fouling resistance, is not as good as the other heat exchangers. These (identical in design) heat exchangers when subjected to identical operating conditions would be expected to demonstrate essentially identical performance; however, the as-tested performance of the 2A RHR heat exchanger differs from that of the other heat exchangers. This apparent anomalous performance of the 2A RHR heat exchanger has been entered into the Browns Ferry Nuclear Plant corrective action program to identify the condition and evaluate the issue to resolution.

The difference between the RHR heat exchanger test results (including the 2A RHR heat exchanger) and the EPU design fouling resistance serves to demonstrate the conservatism embedded in the event analyses where the RHR heat exchanger K-value is determined from the EPU design fouling resistance.

The difference between the RHR heat exchanger test results (excluding the 2A RHR heat exchanger) and the EPU fire event nominal fouling resistance serves to demonstrate that the EPU fire event nominal fouling resistance is appropriate for use in fire event analyses.

The fire event sensitivity case with the EHPMU pump reported in PUSAR Section 2.6.5.2, PUSAR Tables 2.6-4 and 2.6-4a did not use a realistic or nominal heat exchanger fouling resistance but instead used a K-value, 287 BTU/sec-°F, corresponding to the EPU design 2

fouling resistance, 0.001521 hr-ft -°F/BTU. The sensitivity case demonstrates defense-in-depth for a fire event.

The current licensed thermal power (CLTP) analysis in Table 4 summarizes the fire event analysis parameters and the K-value used in the analysis submitted in the NFPA 805 LAR.

As indicated in Table 4, the CLTP NFPA 805 analysis was performed assuming 4400 gpm RHRSW flow and an RHR heat exchanger K-value of 270 BTU/sec-°F. For EPU the RHRSW flow rate is 4500 gpm and the RHR heat exchanger K-value is 307 290 BTU/sec-°F, as discussed above. No modifications are required to effect the change in the RHRSW flow rate from 4400 gpm to 4500 gpm as the CLTP analysis was performed using a conservatively low flow rate compared to the actual RHRSW flow capability during a fire.

Anticipated Transient Without Scram (ATWS) - PUSAR Sections 2.6.5.2 and 2.8.5.7 The limiting ATWS event with respect to peak suppression pool temperature is the ATWS loss-of-offsite power (LOOP) event (two RHR pumps and two RHR heat exchangers credited) which resulted in a peak suppression pool temperature of 173.3°F. The most limiting non-LOOP (four RHR pumps and four RHR heat exchangers credited) ATWS event is the Main Steam Isolation Valve Closure-End of Cycle (MSIVC-EOC) or Pressure Regulator Failure Open (PRFO) event both of which experiences a peak suppression pool temperature of approximately 171.7171.8°F. When the combined transient effects of pool temperature, pool level and pump suction losses are considered, the limiting ATWS event from a net positive suction head (NPSH) perspective is a non-LOOP event (MSIVC-EOC or PRFO-EOC). For Browns Ferry Nuclear Plant EPU ECCS pump NPSH, MSIVC-EOC or PRFO-EOC are remains the most limiting ATWS events.

An RHR heat exchanger K-value of 277 BTU/sec-°F was used in each the ATWS-LOOP analysis. The K-value corresponds to an RHRSW temperature of 95°F, 4500 gpm RHRSW Att 39-8

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses flow, RHR (heat exchanger shell inlet) temperature of 173.3°F (peak suppression pool temperature for ATWS-LOOP event as discussed in PUSAR Sections 2.6.5.2 and 2.8.5.7.2),

6500 gpm RHR flow and 4.575 percent tube plugging, at the EPU design fouling resistance, 0.0015210.001562 hr-ft2-°F/BTU, is 277.8278.6 BTU/sec-°F. These conditions are applicable for the ATWS-LOOP event where only two heat exchangers in the ATWS unit are credited with mitigating the event.

Where four RHR heat exchangers in the same unit are credited with mitigating the ATWS event, system hydraulic resistances in common RHRSW piping on the discharge side of the RHR heat exchangers and in the discharge piping where the two loops combine into common piping, results in restrict the achievable RHRSW flow being restricted such that onlyto slightly greater than 4000 3800 gpm is passed through each of the four heat exchangers when four heat exchangers are simultaneously placed in service for a single Browns Ferry Unit. Therefore, the RHR HX K-value is computed at a conservatively low RHRSW flow rate of 3800 gpm. This explains the reduction in the RHRSW flow rate per heat exchanger from 4500 gpm per heat exchanger in the two heat exchanger alignment described above, to 4000 3800 gpm per heat exchanger in the four heat exchanger alignment. The effect of this condition on the EPU ATWS analyses is described below.

The analyses K-value of 277 259 BTU/sec-°F used in the non-LOOP ATWS events, corresponds to ais bounded by the heat transfer capability, 260.5 BTU/sec-°F, at the design 2

fouling resistance of, 0.0012200.001562 hr-ft -°F/BTU. This was determined using Equation 1 with RHRSW temperature at 95°F, 4000 3800 gpm RHRSW flow, RHR (heat exchanger shell inlet) temperature of 171.7171.8°F, 6500 gpm RHR flow, and 4.575 percent tube plugging. This reduced fouling resistance of 0.001220 hr-ft2-°F/BTU is justified because it is 2

conservative relative to the EPU nominal fouling resistance of 0.001097 hr-ft -°F/BTU (see Figure 1). In addition, the difference between the RHR heat exchanger test results (including the 2A RHR heat exchanger) and this reduced fouling resistance value serves to demonstrate thatThus, the this fouling resistance value of 259 is appropriate for use in ATWS event analyses.

2.3 RHR Heat Exchanger Performance Monitoring An NFPA 805 lIn Attachment X,Section X.4 of Reference 4.3, TVA stated icensing condition requires that the Browns Ferry Nuclear Plant RHR heat exchangers would be subject toare included in a performance monitoring program to provide assurance that heat exchanger fouling that could affect the required heat transfer rate is detected and corrected in a timely manner. Reference 4.7 also included a commitment (i.e., Commitment 2) to implement a The RHR Heat Exchanger Performance Monitoring Program heat exchanger monitoring program to provides assurance that the RHR heat exchanger performance is maintained for consistencyconsistent with analytical assumptions associated with the adoption of the NFPA 805 standard. Because the revised performance monitoring program has not been developed at this time, the commitment specifies implementation within six months following NRC issuance of the license amendment approving adoption of the NFPA 805 standard for the Browns Ferry Nuclear Plant. In addition, TVA will revise the program that monitors Browns Ferry Nuclear Plant RHR heat exchanger performance for consistency with the assumptions used in analyses supporting EPU prior to EPU license amendment implementation. TVA intends to include the RHR Heat Exchanger Performance Monitoring Program in Technical Specifications where specific program attributes are identified. These program attributes include, as requirements in the program, periodic heat exchanger Att 39-9

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses inspections and heat exchanger performance testing requirements to ensure fouling factor assumptions and tube plugging assumptions in the EPU containment analyses remain valid.

3.0 Summary and Conclusion Based on the analyses described above, the RHR heat exchanger fouling resistances and K-values used in the EPU event analyses are appropriate:

• The EPU design fouling resistance is 0.0015210.001562 hr-ft2-°F/BTU for the limiting NFPA 805 (fire) event. This was determined using a heat exchanger K-value of 265 290 BTU/sec-°F with an RHR flow rate of 6500 7500 gpm at RHR (heat exchanger shell side inlet) temperature of 179.0207.7°F, RHRSW flow rate of 4000 4500 gpm at RHRSW inlet temperature of 9588°F with 4.57 percent tubes plugged; 2

• The EPU nominal fouling resistance is 0.001097 hr-ft -°F/BTU; and

• RHR heat exchanger as-found test results are provided in Tables 5, 6 and 7. These results in combination with Table 4, serve to demonstrate that the actual heat transfer capability at the EPU design and nominal fouling resistance meets or values areexceeds the heat transfer capability (K-value) credited in the containment analysis. Therefore, the K-values used in the event analyses are appropriate for use in the associated event analyses.

In addition, periodic RHR heat exchanger performance testing will be used to demonstrate that actual heat exchanger performance exceeds the performance credited in the analyses.

4.0 References 4.1 Fundamentals of Heat and Mass Transfer by Frank P. Incropera and David P. Dewitt, John Wiley & Sons, 3rd Edition, 1990 4.2 EPRI Test Report 107397, Service Water Heat Exchanger Testing Guidelines, Electric Power Research Institute, March 1998 4.3 TVA Letter to NRC, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) (Technical Specification Change TS-480), March 27, 2013 (ADAMS Accession Number ML13092A392) 4.43 TVA Letter to NRC, Response to NRC Request for Additional Information Regarding the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (TAC Nos. MF1185, MF1186, and MF1187) -

Attachment X and Fire Modeling, June 13, 2014 (ADAMS Accession Number ML14167A175) 4.54 Browns Ferry Nuclear Plant Updated Safety Analysis Report, Sections 1.6.2.12, 4.8.6.2, 4.8.6.3, Table 4.8-1, and Table 5.2-1 4.65 NRC Memorandum to ACRS Members, Certification of the Meeting Minutes from the nd Advisory Committee on Reactor Safeguards 572 Full Committee Meeting Held on May 6-8 2010 in Rockville, Maryland, August 16, 2010 (ADAMS Accession Number ML101830190)

Att 39-10

BFN EPU LAR Attachment 39, Revision 1 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses 4.7 TVA Letter to NRC, Response to NRC Request to Supplement License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (TAC Nos. MF1185, MFl186, and MF1187), May 16, 2013 (ADAMS Accession Number ML13141A291) 4.86 SECY-11-0014 - Use of Containment Accident Pressure in Analyzing Emergency Core Cooling System and Containment Heat Removal System Pump Performance in Postulated Accidents, U.S. Nuclear Regulatory Commission, January 31, 2011, (ADAMS Accession No. ML102590196)

Att 39-11

BFN EPU LAR Attachment 39 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses Table 1 Overall Fouling Resistances and K-values for Events with RHR Flow Rate 6500 gpm RHRSW Overall Fouling Tube Trhr2 Tsw3 K Event Flow Resistance Plugging5 F F gpm hr-ft2-°F/Btu Btu/sec-°F percent 1

DBA-LOCA (RSLB ) 179.0 95 4000 0.000000 333.7337. 4.575 9

DBA-LOCA (RSLB) 179.0 95 4000 0.000461 310.7 4.57 DBA-LOCA (RSLB) 179.0 95 4000 0.0005164 308.1 4.57 DBA-LOCA (RSLB) 179.0 95 4000 0.000674 300.7 4.57 DBA-LOCA (RSLB) 179.0 95 4000 0.000820 294.0 4.57 DBA-LOCA (RSLB) 179.0 95 4000 0.000930 289.1 4.57 DBA-LOCA (RSLB) 179.0 95 4000 0.001107 281.6 4.57 DBA-LOCA (RSLB) 179.0 95 4000 0.001097 282.0 4.57 DBA-LOCA (RSLB) 179.0 95 4000 0.0015210.00156 265.0266. 4.575 2 4 DBA-LOCA (RSLB) 179.0 95 4000 0.0028000.00280 223.1225. 4.575 14 0 ATWS- LOOP 173.3 95 4500 0.000000 355.6359. 4.575 5

ATWS- LOOP 173.3 95 4500 0.000461 329.2 4.57 ATWS- LOOP 173.3 95 4500 0.0005164 326.2 4.57 ATWS- LOOP 173.3 95 4500 0.000674 317.8 4.57 ATWS- LOOP 173.3 95 4500 0.000820 310.3 4.57 ATWS- LOOP 173.3 95 4500 0.000930 304.8 4.57 ATWS- LOOP 173.3 95 4500 0.001107 296.3 4.57 ATWS- LOOP 173.3 95 4500 0.001097 296.8 4.57 ATWS- LOOP 173.3 95 4500 0.0015210.00156 277.8278. 4.575 2 6 ATWS- LOOP 173.3 95 4500 0.0028000.00280 231.9233. 4.575 14 4 ATWS-MSIVC EOC 171.717 333.9328.

95 40003800 0.000000 4.557 and PRFO EOC 1.8 0 ATWS-MSIVC EOC 171.7 95 4000 0.000461 310.9 4.57 ATWS-MSIVC EOC 171.7 95 4000 0.0005164 308.2 4.57 ATWS-MSIVC EOC 171.7 95 4000 0.000674 300.8 4.57 ATWS-MSIVC EOC 171.7 95 4000 0.000820 294.2 4.57 ATWS-MSIVC EOC 171.7 95 4000 0.000930 289.3 4.57 ATWS-MSIVC EOC 171.7 95 4000 0.001107 281.7 4.57 ATWS-MSIVC EOC 171.7 95 4000 0.001097 282.1 4.57 ATWS-MSIVC EOC 171.7 95 4000 0.001220 277.0 4.57 ATWS-MSIVC EOC 171.717 0.0015210.00156 265.1260.

95 40003800 4.575 and PRFO EOC 1.8 2 5 ATWS-MSIVC EOC 171.717 0.0028000.00280 223.2220.

95 40003800 4 4.575 and PRFO EOC 1.8 1 9 Att 39-12

BFN EPU LAR Attachment 39 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses Notes

1. RSLB = recirculation suction line break

2. Trhr = temperature - residual heat removal

3. Tsw = temperature - service water

4. Heat exchanger original design fouling resistance

5. The current condition (number of tubes plugged) of each RHR heat exchanger is bounded by the design value of 4.575 percent tubes plugged. Work processes prohibit returning an RHR heat exchanger (HX) to service following maintenance with more than 4.57 percent tubes plugged without proper design and licensing basis review/evaluation being performed.

Att 39-13

BFN EPU LAR Attachment 39 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses Table 2 Overall Fouling Resistances and K-values for Events with RHR Flow Rate 9700 gpm RHRSW Overall Fouling Tube Trhr1 Tsw2 K Event Flow Resistance Plugging4 F F gpm hr-ft2-°F/Btu Btu/sec-°F percent Shutdown of Non-Accident Unit 185.1 95 4500 0.000000 398.3405. 4.575 1

Shutdown of Non-Accident Unit 185.1 95 4500 0.000461 365.0 4.57 Shutdown of Non-Accident Unit 185.1 95 4500 0.0005164 361.3 4.57 Shutdown of Non-Accident Unit 185.1 95 4500 0.000674 350.9 4.57 Shutdown of Non-Accident Unit 185.1 95 4500 0.000820 341.7 4.57 Shutdown of Non-Accident Unit 185.1 95 4500 0.000930 335.0 4.57 Shutdown of Non-Accident Unit 185.1 95 4500 0.001107 324.6 4.57 Shutdown of Non-Accident Unit 185.1 95 4500 0.001097 325.2 4.57 Shutdown of Non-Accident Unit 185.1 95 4500 0.0015210.001562 302.4304. 4.575 7

Shutdown of Non-Accident Unit 185.1 95 4500 0.0028000.002801 248.4251. 4.575 3

1 Notes

1. Trhr = temperature - residual heat removal

2. Tsw = temperature - service water

3. Heat exchanger original design fouling resistance

4. The current condition (number of tubes plugged) of each RHR heat exchanger is bounded by the design value of 4.575 percent tubes plugged. Work processes prohibit returning an RHR HX to service following maintenance with more than 4.57 percent tubes plugged without proper design and licensing basis review/evaluation being performed.

Att 39-14

BFN EPU LAR Attachment 39 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses Table 3 Overall Fouling Resistance and K-values for Events with RHR Flow Rate 7500 gpm RHRSW Overall Fouling Tube Trhr1 Tsw2 K Event Flow Resistance Plugging4 F F gpm hr-ft2-°F/Btu Btu/sec-°F percent FIRE - CLTP 205.7 92 4400 0.000000 366.4373.6 4.57 FIRE - CLTP 205.7 92 4400 0.000461 338.3 4.57 FIRE - CLTP 205.7 92 4400 0.0005164 335.1 4.57 FIRE - CLTP 205.7 92 4400 0.000674 326.3 4.57 FIRE - CLTP 205.7 92 4400 0.000820 318.3 4.57 FIRE - CLTP 205.7 92 4400 0.000930 312.5 4.57 FIRE - CLTP 205.7 92 4400 0.001107 303.6 4.57 FIRE - CLTP 205.7 92 4400 0.001097 304.1 4.57 FIRE - CLTP 205.7 92 4400 0.0015210.001562 284.2287.7 4.57 FIRE - CLTP 205.7 92 4400 0.0018560.001975 270.0 4.57 FIRE - CLTP 205.7 92 4400 0.0028013 236.2239.9 4.57 FIRE - EPU with EHPMU pump 206.2 95 4500 0.000000 370.8 4.57 FIRE - EPU with EHPMU pump 206.2 95 4500 0.000461 342.0 4.57 FIRE - EPU with EHPMU pump 206.2 95 4500 0.0005164 338.8 4.57 FIRE - EPU with EHPMU pump 206.2 95 4500 0.000674 329.7 4.57 FIRE - EPU with EHPMU pump 206.2 95 4500 0.000820 321.5 4.57 FIRE - EPU with EHPMU pump 206.2 95 4500 0.000930 315.6 4.57 FIRE - EPU with EHPMU pump 206.2 95 4500 0.001107 306.5 4.57 FIRE - EPU with EHPMU pump 206.2 95 4500 0.001097 307.0 4.57 FIRE - EPU with EHPMU pump 206.2 95 4500 0.001521 286.7 4.57 FIRE - EPU with EHPMU pump 206.2 95 4500 0.0028003 237.9 4.57 208.0207.

FIRE - EPU - No EHPMU pump 9288 4500 0.000000 370.8377.9 4.57 7

FIRE - EPU - No EHPMU pump 208.0 92 4500 0.000461 342.1 4.57 FIRE - EPU - No EHPMU pump 208.0 92 4500 0.0005164 338.8 4.57 FIRE - EPU - No EHPMU pump 208.0 92 4500 0.000674 329.7 4.57 FIRE - EPU - No EHPMU pump 208.0 92 4500 0.000820 321.6 4.57 FIRE - EPU - No EHPMU pump 208.0 92 4500 0.000930 315.7 4.57 FIRE - EPU - No EHPMU pump 208.0 92 4500 0.001107 306.5 4.57 FIRE - EPU - No EHPMU pump 208.0 92 4500 0.001097 307.0 4.57 208.0207.

FIRE - EPU - No EHPMU pump 9288 4500 0.0015210.001562 286.7290.0 4.57 7

208.0207. 0.0028000.002801 FIRE - EPU - No EHPMU pump 9288 4500 3 237.9241.4 4.57 7

Notes

1. Trhr = temperature - residual heat removal

2. Tsw = temperature - service water

3. Heat exchanger original design fouling resistance

4. The current condition (number of tubes plugged) of each RHR heat exchanger is bounded by the design value of 4.57 percent tubes plugged (77 out of 1700 tubes). Work processes prohibit Att 39-15

BFN EPU LAR Attachment 39 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses returning an RHR HX to service following maintenance with more than 4.57 percent tubes plugged without proper design and licensing basis review/evaluation being performed.

Att 39-16

BFN EPU LAR Attachment 39 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses Table 4 Overall Fouling Resistance, K-value and Heat Transfer Rate Summary for Comparison to Heat Exchanger Test Results ATWS-MSIVC-EOC or

SUMMARY

DBA-LOCA (RSLB) Shutdown of Non-Acc Unit Fire Event ATWS-PRFO-EOC EPU EPU EPU CLTP EPU RHR (HX shell (gpm) 6500 9700 6500 7500 7500 side) flow rate RHRSW (HX tube (gpm) 4000 4500 3800 4400 4500 side) flow rate RHR (HX shell side) inlet (°F) 179.0 185.1 171.8 205.7 207.7 temperature RHRSW (HX tube side) inlet (°F) 95.0 95.0 95.0 92.0 88.0 temperature Assumed Percentage of (%) 5 5 5 4.57 4.57 Tubes Plugged Number of RHR (each) 2 1 4 1 1 HXs in Service Fouling Q K Q K Q K Q K Q K Resistance per HX per HX per HX per HX per HX per HX per HX per HX per HX per HX (hr-ft2-°F/Btu) (Btu/hr) (Btu/sec-°F) (Btu/hr) (Btu/sec-°F) (Btu/hr) (Btu/sec-°F) (Btu/hr) (Btu/sec-°F) (Btu/hr) (Btu/sec-°F)

Clean (zero fouling resistance) Heat Transfer Capability 0.000000 102,180,960 338 131,398,236 405 89,712,230 328 152,921,952 374 162,844,668 378 EPU Heat Transfer Capability 0.001562 80,559,360 266 98,832,492 305 72,023,040 261 117,761,364 288 124,966,800 290 0.001975 110,516,400 270 Original Design Heat Transfer Capability 0.002801 68,040,000 225 81,446,796 251 61,074,432 221 98,195,868 240 104,024,088 241 Notes

1. Heat Transfer Capability K-values shown in this table are rounded from the values shown in Tables 1, 2, and 3.

Att 39-17

BFN EPU LAR Attachment 39 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses Table 5: Test Results for BFN Unit 1 RHR Heat Exchangers Parameter UNIT 1 1A 1B 1C 1D Test date 12/2/15 3/23/16 12/2/15 3/23/16 Number of years in 0.3 1.6 1.6 1.9 operation after last cleaning Hot side inlet temperature 77.64 73.43 79.14 74.55

(°F)

Hot side outlet temperature 73.51 68.55 73.99 69.64

(°F)

Hot side flow rate (pound 3,631,734 3,668,936 3,622,515 3,671,967 mass/hour)

Cold side inlet temperature 57.76 57.20 57.75 58.08

(°F)

Cold side outlet temperature 72.04 65.98 71.72 66.64

(°F)

Cold side flow rate (pound 1,034,623 1,969,221 1,313,440 2,007,073 mass/hour)

Test heat transfer rate 14,837,671 17,530,731 18,444,721 17,506,428 (BTU/hr)

Test U 2 273.2765 329.4458 291.2366 314.0870 (BTU/hr-ft -°F)

Test UA 1,716,835 2,080,849 1,834,042 1,981,481 (BTU/hr-°F)

Test fouling resistance 0.000463 0.000543 0.000558 0.000720 (hr-ft2-°F/BTU)

Uncertainty in fouling resistance 0.000292 0.000275 0.000275 0.000275 (hr-ft2-°F/BTU)

Acceptance criteria for 0.001517 0.001517 0.001517 0.001517 fouling resistance 2 (1) (1) (1) (1)

(hr-ft -°F/BTU)

Margin in fouling resistance 0.000762 0.000699 0.000684 0.000522 (hr-ft2-°F/BTU)

EPU Acceptance criteria for fouling resistance 0.001562 0.001562 0.001562 0.001562 (hr-ft2-°F/BTU)

EPU Margin in fouling resistance 0.000807 0.000744 0.000729 0.000567 (hr-ft2-°F/BTU)

Acceptance criteria for 77 77 77 77 blocked tubes (2) (2) (2) (2)

Att 39-18

BFN EPU LAR Attachment 39 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses Table 6: Test Results for BFN Unit 2 RHR Heat Exchangers Parameter UNIT 2 2A 2A 2B 2C 2C 2D Test date 4/11/15 1/8/15 1/5/16 4/11/15 1/8/15 1/5/16 Number of years in operation after last 0.1 1.6 0.6 0.1 3.9 2.1 cleaning Hot side inlet 80.74 77.16 74.58 83.19 76.84 75.10 temperature (°F)

Hot side outlet 76.69 68.68 66.14 78.22 68.27 66.75 temperature (°F)

Hot side flow rate 4,030,078 3,537,939 3,770,401 4,022,707 3,612,076 3,804,645 (pounds mass/hour)

Cold side inlet 65.19 41.13 46.96 64.88 40.68 47.06 temperature (°F)

Cold side outlet 73.89 62.20 61.46 75.98 62.77 62.06 temperature (°F)

Cold side flow rate 1,706,146 1,346,317 2,035,618 1,686,190 1,311,739 1,993,996 (pounds mass/hour)

Test heat transfer rate 15,224,364 28,920,419 30,302,215 19,062,260 29,695,875 30,614,461 (BTU/hr)

Test U 2 291.0500 239.9253 328.7694 337.5143 256.0849 330.5212 (BTU/hr-ft -°F)

Test UA 1,841,610 1,518,120 2,079,046 2,133,077 1,618,446 2,077,712 (BTU/hr-°F)

Test fouling resistance 0.000946 0.001152 0.000535 0.000477 0.000868 0.000515 (hr-ft2-°F/BTU)

Uncertainty in fouling resistance 0.000274 0.000284 0.000262 0.000258 0.000289 0.000262 (hr-ft2-°F/BTU)

Acceptance criteria for fouling resistance 0.001517 0.001517 0.001517 0.001517 0.001517 0.001517 (1) (1) (1) (1) (1) (1)

(hr-ft2-°F/BTU)

Margin in fouling resistance 0.000297 0.000081 0.000720 0.000782 0.000360 0.000740 (hr-ft2-°F/BTU)

EPU Acceptance criteria for fouling resistance 0.001562 0.001562 0.001562 0.001562 0.001562 0.001562 (hr-ft2-°F/BTU)

EPU Margin in fouling resistance 0.000342 0.000126 0.000765 0.000827 0.000405 0.000785 2

(hr-ft -°F/BTU)

Acceptance criteria for 77 77 77 77 77 77 blocked tubes (2) (2) (2) (2) (2) (2)

Att 39-19

BFN EPU LAR Attachment 39 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses Table 7: Test Results for BFN Unit 3 RHR Heat Exchangers Parameter UNIT 3 3A 3B 3C 3D Test date 1/25/12 10/30/15 1/25/12 10/30/15 Number of years in operation after last 1.9 1.4 4.0 4.0 cleaning Hot side inlet 75.29 85.63 74.73 87.27 temperature (°F)

Hot side outlet 68.83 81.60 68.67 83.10 temperature (°F)

Hot side flow rate 3,827,744 3,487,834 4,014,733 3,469,054 (pounds mass/hour)

Cold side inlet 53.02 65.39 53.02 65.52 temperature (°F)

Cold side outlet 65.11 80.53 64.38 81.84 temperature (°F)

Cold side flow rate 1,979,900 896,677 2,036,549 877,040 (pounds mass/hour)

Test heat transfer rate 24,291,564 13,679,428 23,630,760 14,349,628 (BTU/hr)

Test U 2 327.7767 257.1204 314.4073 252.4582 (BTU/hr-ft -°F)

Test UA 2,078,920 1,641,405 1,992,946 1,589,838 (BTU/hr-°F)

Test fouling resistance 0.000569 0.000553 0.000746 0.000625 (hr-ft2-°F/BTU)

Uncertainty in fouling resistance 0.000266 0.000299 0.000260 0.000296 (hr-ft2-°F/BTU)

Acceptance criteria for fouling resistance 0.001517 0.001517 0.001517 0.001517 2 (1) (1) (1) (1)

(hr-ft -°F/BTU)

Margin in fouling resistance 0.000682 0.000665 0.000511 0.000596 (hr-ft2-°F/BTU)

EPU Acceptance criteria for fouling resistance 0.001562 0.001562 0.001562 0.001562 (hr-ft2-°F/BTU)

EPU Margin in fouling resistance 0.000727 0.000710 0.000556 0.000641 (hr-ft2-°F/BTU)

Acceptance criteria for 77 77 77 77 blocked tubes (2) (2) (2) (2)

Att 39-20

BFN EPU LAR Attachment 39 RHR Heat Exchanger K-values Utilized in EPU Containment Analyses Notes for Tables 5, 6 and 7:

(1) NFPA 805 license condition acceptance criterion.

(2) NFPA 805 license condition allowable tube plugging limit (4.57%) limits the number of tubes that can be plugged (mechanically) to 77.

Att 39-21

ENCLOSURE 5 BFN EPU LAR, Attachment 47, List and Status of Plant Modifications, Revision 3

Browns Ferry Units 1, 2 and 3 EPU Modifications, Revision 3 The modifications required to support Extended Power Uprate (EPU) for Browns Ferry Nuclear Power Station (BFN) Units 1, 2 and 3 have been compiled and are shown in Table 1. The modifications reported as Complete in Table 1 are fully implemented, for all other modifications a schedule for full implementation is provided. All EPU modifications, either completed or being prepared, are in accordance with the TVA Plant Modifications and Engineering Change Control process.

Further evaluations may identify the need for additional modifications or obviate the need for some modifications. As such, Table 1 listings are not a formal commitment to implement the modifications exactly as described or per the proposed schedule.

Additionally, various minor modifications and adjustments to plant equipment, which may be necessary, are not listed.

Att 47-1

Table 1: BFN EPU Planned Modifications and Current Schedule Scheduled Modification Description Completion (Note 1)

Replacement New steam dryers will be installed with increased structural design margin Unit 1 - Fall 2018 Steam Dryer to accommodate EPU operation. Unit 2 - Spring 2019

• Replacement steam dryers are curved hood six-bank dryers analyzed Unit 3 - Spring 2018 for fatigue resulting from flow induced vibration and hydrodynamic loads.

• Main steam line strain gages were previously installed to obtain measurements at CLTP conditions which were used to design the replacement steam dryers.

• New main steam line strain gages will be installed to replace the existing strain gages which have reached end of life to obtain measurements during power ascension testing of the replacement steam dryers.

Main Turbine Replace the High Pressure Turbine rotor. Incorporate GE's Advanced Design Steam Path which is designed for the increased flow associated with EPU.

Unit 1 - Fall 2018

• Replace High Pressure Turbine diaphragms and rotor buckets. Unit 2 - Spring 2019 Unit 3 - Spring 2018

• Modify the cross around relief valves (CARVs) to permit increased set Unit 1 - Complete pressure. Unit 2 - Complete Unit 3 - Complete

• Replace and/or recalibrate Main Steam system flow and pressure Unit 1 - Complete instruments. Unit 2 - Complete Unit 3 - Complete Turbine Sealing Increase the size of the Steam Packing Unloader Valves (SPUVs) and Unit 1 - Complete Steam associated piping to enable the turbine sealing system to accommodate Unit 2 - Complete EPU flow requirements. Unit 3 - Complete

• Increase SPUVs and piping from 8-inch to 10-inch components.

• Replace and rescale steam flow and steam pressure transmitters.

Att 47-2

Scheduled Modification Description Completion (Note 1)

Condensate Upgrade Condensate pumps with new impellers and motors to Unit 1 - Complete Pumps accommodate the increased flows that will be required for EPU operation. Unit 2 - Complete

• Replace impellers in each pump (3 pumps per Unit). Unit 3 - Complete

• Replace 900 HP motors with 1250 HP motors.

• Add orifice plate to the Condensate Recirculation line to reduce pressure drop across the flow control valve to minimize cavitation and vibration.

• Replace existing pump discharge check valves with different style check valves having lower pressure drop and better transient response.

• Replace pump suction strainers with stronger mesh screen to prevent screen deformation with the increased EPU flow conditions.

• Change motor protection relay settings.

• Recalibrate/replace pump and motor instrumentation.

Condensate Replace the Condensate Booster (CB) pumps and motors to increase Unit 1 - Complete Booster Pumps pump capacity to accommodate the increased flows that will be required Unit 2 - Complete for EPU operation. Unit 3 - Complete

• Replace CB pumps with higher capacity pumps.

• Replace air-cooled 1750 HP motors with water-cooled 3000 HP motors.

• Replace existing pump discharge check valves with different style check valves having lower pressure drop and better transient response.

• Change motor protection relay settings.

• Recalibrate/replace pump and motor instrumentation.

Condensate Pump Provide additional cooling/ventilation in vicinity of the Condensate and Unit 1 - Complete and Condensate Condensate Booster pumps to accommodate the increased heat load Unit 2 - Complete Booster Pump resulting from larger air-cooled Condensate Pump motors and supplement Unit 3 - Complete Area Ventilation cooling requirements for the hydrogen water chemistry (HWC) main control panel.

• Replace 3-position switches for operation of the Air Handling Units (AHUs) with 4-position switches that will allow parallel operation of the AHUs.

• Addition of a balancing damper to the Condensate Pump motors to provide better balancing of air flow.

• Addition of a branch duct and balancing damper to the HWC main control panel.

Att 47-3

Scheduled Modification Description Completion (Note 1)

Feedwater Pumps Upgrade the Feedwater system to provide increased Feedwater flow for and Turbines EPU operation.

• Replace pumps with higher capacity pumps. Unit 1 - Complete

• Replace turbine rotor, diaphragms and buckets. Unit 2 - Complete

• Replace turbine/pump coupling. Unit 3 - Complete

• Upgrade seal water injection subsystem.

• Update Feedwater control system software for EPU conditions. Unit 1 - Fall 2018 Unit 2 - Spring 2019 Unit 3 - Spring 2018 Moisture Modify the internals of the moisture separators to increase moisture Unit 1 - Complete Separators removal and accommodate increased flows at EPU conditions. Unit 2 - Complete

• Change vanes and added perforated plate on moisture separators. Unit 3 - Complete

• Modify internal drains as needed.

Feedwater Upgrade Feedwater Heaters to support EPU operating conditions.

Heaters Unit 1 - Complete

• Re-rate the number 1, 2 and 3 Feedwater Heater shells to meet higher Unit 2 - Complete pressures, temperatures and flows under EPU conditions by Unit 3 - Spring 2018 modification of selected nozzles and replacement of shell relief valves to meet ASME code requirements.

• Replace level control instrumentation on the number 1, 2 and 3 Feedwater Heaters to reduce susceptibility to flow induced turbulence (pressure transients).

• Provide additional welds and bracing to the pass partition plates for Unit 1 - Complete Nos. 1, 2, 3, and 5 Feedwater Heaters. (Number 4 Feedwater Unit 2 - Complete Heaters pass partition plates will be addressed with replacement of Unit 3 - Complete the tube bundle and channel head.)

• Due to the increase in tube-side design pressure with the increase Unit 1 - Complete head capacity of the Condensate Booster pumps, replace channel Unit 2 - Complete head relief valves for No. 3 Feedwater Heaters with valves having Unit 3 - Spring 2018 higher setpoints, and install a reinforcement ring on the manways for the number 3 and number 5 Feedwater Heaters.

• On each of the number 3 Feedwater Heaters, replace the upper shell Unit 1 - Complete and install an extraction steam inlet duct to minimize heater shell Unit 2 - Spring 2017 erosion and preclude tube damage from steam jet impingement. Unit 3 - Complete

• Replace tube bundle and channel head in the number 4 Feedwater Unit 1 - Fall 2018 Heaters with a design less susceptible to damage from flow induced Unit 2 - Spring 2019 vibration. Unit 3 - Spring 2018 Main Condenser Replace Main Condenser Extraction Steam bellows #2, #3, #4 and #5 with Unit 1 - Complete Extraction Steam bellows accommodating higher design temperatures and pressures for Unit 2 - Complete Bellows EPU. Unit 3 - Complete Att 47-4

Scheduled Modification Description Completion (Note 1)

Condensate Install a 10th condensate demineralizer (and associated valves and Unit 1 - Complete Demineralizers controls) on each unit to accommodate the increased condensate flow Unit 2 - Complete associated with EPU operation.

Unit 3 - Complete Steam Packing Increase the capacity of the steam packing exhauster bypass line to Unit 1 - Complete Exhauster Bypass accommodate increased flow under EPU conditions. Unit 2 - Complete

• Install larger piping and flow control valve. Unit 3 - Complete Torus Attached Modification to reinforce an existing pad at an ECCS ring header branch Unit 1 - N/A Piping connection to address higher pipe stresses associated with EPU Unit 2 - Complete conditions. Required only on Units 2 and 3 as sufficient stress margin Unit 3 - Complete exists on Unit 1.

Main Steam Modify one Unit 2 Main Steam pipe support due to increased loads Unit 1 - NA Supports resulting from turbine stop valve closure at EPU steam flow rates. All other Unit 2 - Complete existing Unit 2 Main Steam pipe supports, and all Main Steam pipe Unit 3 - NA supports on Units 1 and 3, were determined to have sufficient design margin to accommodate the increased turbine stop valve closure loads.

Reactor Upgrade the reactor recirculation system for EPU core flow operating Recirculation conditions.

Pumps & Motors Unit 1 - Complete

• Perform analyses/evaluations to increase the design ratings for the recirculation pumps and motors. Unit 2 - Complete

• Upgrade the Variable Frequency Drive (VFD) control system. Unit 3 - Complete

• Perform pump and motor instrumentation upgrades - jet pump head, RCW flow, motor winding temperatures, VFD protective relay settings.

• Revise Upper Power Runback setting for EPU conditions. Unit 1 - Fall 2018 Unit 2 - Spring 2019 Unit 3 - Spring 2018 Jet Pump Sensing Install jet pump sensing line clamps to reduce pipe vibration under EPU Unit 1 - Complete Line Clamps conditions. Unit 2 - Complete Unit 3 - Complete Main Generator Uprate main generator to 1330 MVA (Unit 1) / 1332 MVA (Units 2 & 3). Unit 1 - Complete System

• Install rewound stator to support higher generator output capacity. Unit 2 - Spring 2019 Unit 3 - Complete

• Replace/modify stator water cooling (SWC) instruments and change SWC flow, pressure, DP and temperature settings to support increased stator water cooling requirements.

Main Generator Increase generator hydrogen pressure from 65 psig to 75 psig to support Unit 1 - Fall 2018 Hydrogen EPU operation. Unit 2 - Spring 2019 Pressure

• Change pressure regulating valve settings and pressure alarm setting. Unit 3 - Spring 2018

• Replace pressure switches as needed for new operating range.

• Change generator field over-excitation relay settings.

• Eliminate hydrogen flow integrator to mitigate hydrogen leakage.

Att 47-5

Scheduled Modification Description Completion (Note 1)

Isophase Bus Duct Modify isophase bus duct cooling system to remove increased bus duct Unit 1 - Complete Cooling heat under EPU conditions. Unit 2 - Complete

• Replace cooling fans and motors. Unit 3 - Complete

• Replace cooling coils.

Main Bank Upgrade main bank transformers to account for the higher power output Unit 1 - Complete Transformers from the main generators at EPU conditions. Unit 2 - Complete

• Replace three 500 MVA transformers per unit. Unit 3 - Installation

• Replace one Units 1 and 2 500 MVA spare transformer. complete, post-modification testing

• Install new dedicated Unit 3 500 MVA spare transformer.

of the Unit 3 Spare Transformer pending Vibration Install mounting brackets/supports and temporary instrumentation for Unit 1 - Fall 2018 Monitoring vibration monitoring during EPU power ascension in accordance with Unit 2 - Spring 2019 Attachment 45 (Flow Induced Vibration Analysis and Monitoring Program).

Unit 3 - Spring 2018 Main Steam Modify MSIVs to support steam flow increase at EPU conditions. Unit 1 - Complete Isolation Valves

• Install longer stroke actuators to move the poppet further out of the Unit 2 - Complete (MSIV) flow stream. This modification reduced valve pressure drop to Unit 3 - Complete accommodate EPU conditions.

• Perform additional modifications to improve performance of the MSIVs including new bonnets, nose guided poppets (trimmed profile), and larger diameter valve stems.

Electro-Hydraulic Revise EHC software to address changes in plant parameters required to Unit 1 - Complete Control (EHC) support EPU.

Unit 2 - Complete Software

• Electrical Overspeed set point, Intermediate Pressure, Power Load Unbalance, Turbine First Stage Pressure, and Megawatt (MW) Control Unit 3 - Complete Technical Technical Specification Instrument respan and setpoint changes for EPU Unit 1 - Fall 2018 Specification

• Turbine 1st stage pressure scram bypass permissive setpoint change Unit 2 - Spring 2019 Instrument

• Main steam line high flow isolation channel respan Unit 3 - Spring 2018 Respan

• APRM flow biased and setdown instrument respan and setpoint change Balance of Plant Respan balance of plant (BOP) instruments for EPU. Unit 1 - Complete Instrument

• Update hydrogen water chemistry programmable logic controller (PLC) Unit 2 - Complete Respan software for control of hydrogen and oxygen injection at EPU. Unit 3 - Complete

• Replace and respan hydrogen water chemistry flow instruments.

• Replace and respan extraction steam pressure instruments.

• Replace and respan feedwater heater pressure and level instruments.

• Recalibrate setpoints for reactor feedwater low suction and steam jet air ejector stage I/II/III low pressure switches.

• Respan high pressure turbine exhaust intermediate pressure.

• Replace and respan offgas condenser cooling water temperature instruments.

Att 47-6

Scheduled Modification Description Completion (Note 1)

Condenser Upgrade condenser instrumentation for improved reliability and Instrumentation performance monitoring under EPU conditions.

• Replace/relocate condenser A/B/C hotwell pressure transmitters to Unit 1 - Complete improve inputs to the integrated computer system (ICS).

• Add condenser circulating water (CCW) inlet/outlet temperature inputs Unit 2 - Complete to the integrated computer system (ICS). Unit 3 - Complete

• Respan condenser A/B/C CCW outlet flow channels and add to ICS.

• Revise reactor feed pump turbine (RFPT) trip to two out of three logic.

• Modify Steam jet air ejector (SJAE) to remove the trip on low condenser vacuum and eliminate auto-start of standby SJAE.

• Install nine new condenser vacuum pressure transmitters per unit (3 on each condenser) and provide signals to electro-hydraulic control Unit 1 - Fall 2018 (EHC) system. Unit 2 - Spring 2019

• Move condenser A/B/C low vacuum alarm, low vacuum turbine trip Unit 3 - Spring 2018 and low vacuum bypass trip functions to EHC logic (previously performed by pressure switches).

• Perform hardware and software changes to EHC system to support new alarm and trip functions.

Steam Jet Air Revise setpoints for SJAE condensate pressure switches to prevent Unit 1 - Complete Ejector (SJAE) inadvertent SJAE isolation. Unit 2 - Complete Pressure switches Unit 3 - Complete Main Steam Install Acoustic Vibration Suppressors (AVS) inside the Main Steam 6" Unit 1 - Complete Acoustic Vibration diameter blind flanged branch lines to reduce acoustic loading on the Unit 2 - Complete Suppressors steam dryer.

Unit 3 - Complete Standby Liquid The shutdown capability of the SLC system is being increased to support Unit 1 - Fall 2018 Control (SLC) the Containment Accident Pressure Credit Elimination during an ATWS Unit 2 - Spring 2019 System event as discussed in PUSAR Section 2.8.4.5.3 (Attachment 6) by Unit 3 - Spring 2018 increasing the Boron-10 enrichment.

Emergency High As part of the transition to National Fire Protection Association Standard Unit 1 - Fall 2016 Pressure Makeup (NFPA) 805, BFN is installing a non-safety related emergency high Unit 2 - Spring 2017 Pump pressure pump in each unit to provide makeup from the Condensate Unit 3 - Spring 2018 Storage Tank to the Reactor Pressure Vessel. This modification is not required for EPU operation but is addressed in PUSAR Section 2.6.5.2 (Attachment 6), Containment Accident Pressure (CAP) Elimination.

Although not needed for CAP Credit Elimination, use of the makeup pump will provide additional NPSH margin during the Fire Event.

Att 47-7

Scheduled Modification Description Completion (Note 1)

Hardened Wetwell In response to EA-13-109, Order Modifying Licenses with Regard to Unit 1 - Fall 2016 Vent Reliable Hardened Containment Vents Capable of Operation Under Unit 2 - Spring 2017 Severe Accident Conditions, the Hardened Wetwell Vent (HWWV) will be Unit 3 - Spring 2018 modified to provide individual vent lines for each BFN unit.

As discussed in PUSAR Section 2.6.1.4 (Attachment 6), the existing HWWV capacity would be reduced to 0.88% of rated thermal power under EPU conditions. However, with the implementation of this modification in response to EA-13-109, the capacity of the HWWV will be restored to 1%

of EPU thermal power.

Static Excitation As a result of the increased electrical generation at EPU, the Excitation Unit 1 - Fall 2020 System System on Units 1, 2, and 3 will be upgraded by installing a Static Unit 2 - Spring 2023 Excitation System. The system will include a dual channel digital Unit 3 - Spring 2024 automatic voltage regulator (AVR) for complete redundancy, with each channel consisting of an auto and manual back-up mode. (See also Note 2 below.)

Alternate Leakage Provide a highly reliable Alternate Leakage Treatment Pathway that routes Unit 1 - Fall 2018 Treatment 99.5% of the leakage from the primary containment MSIVs directly to the Unit 2 - Spring 2019 Pathway condenser. The modification on each unit will include replacement of five Unit 3 - Spring 2018 motor operated valves on the Main Steam drain lines with air operated valves. Also, an additional Main Steam drain line valve will be installed, with an air operator, to address single failure criteria. All the new air operated valves will fail open on loss of electrical power or control air.

Notes:

1) The expected completion timeframes reported in Table 1 correspond to the following refuel outages: For BFN Unit 1, Fall of 2016 is RFO-U1R11, Fall of 2018 is RFO-U1R12 and Fall 2020 is RFO-U1R13. For BFN Unit 2, Spring of 2017 is RFO-U2R19, Spring of 2019 is RFO-U2R20 and Spring 2023 is RFO-U2R22. For BFN Unit 3, Spring of 2018 is RFO-U3R18 and Spring 2024 is RFO-U3R21.

2) The Static Excitation System is not required to be installed prior to EPU operation. During the interim period of EPU operation preceding installation of the Static Excitation System, transmission system grid stability will be maintained through use of a detailed temporary operating guide.

Att 47-8

ENCLOSURE 6 Response to NRC Request for Additional Information AFPB-RAI 4, Revision 1

ENCLOSURE 6 AFPB-RAI 4 The staff notes that LAR Attachment 6 to the SAR for BFN, Units 1, 2, and 3, EPU, NEDC-33860P, Revision 0, September 2015, Section 2.5.1.4.1, Fire Protection Program, states that, Other EPU modifications will be assessed and assured not to adversely affect the ability to achieve and maintain the fuel in a safe and stable condition in the event of a fire Further, Section 2.11.1.2.2, Fire Safe Shutdown (FSS) Events, states:

Attachment 47 of the EPU LAR provides a listing and discussion of the modifications planned for EPU. The effect of these modifications on the Browns Ferry Fire Protection Program will be evaluated, in accordance with TVAs configuration change process, prior to EPU implementation. Per the process, these modifications will be evaluated to assure the changes do not affect the approved Fire Protection Program and will not adversely affect the ability to achieve and maintain safe shutdown in accordance with the current Browns Ferry license conditions and procedures The NRC staff notes that modifications associated with the EPU have not yet been completed to address the impact on the fire protection program. The staff requests that the licensee discuss how the results of plant modifications would impact the fire protection program and the plants compliance with the fire protection program licensing basis (10 CFR 50.48(c)).

TVA Response:

All of the modifications listed in Attachment 47 of the EPU LAR have been reviewed for impact on the Fire Protection Program and the plants compliance with the fire protection program licensing basis, 10 CFR 50.48(c).

The Hardened Wetwell Vent modifications, which have not been completed, will result in a positive impact, i.e., plant risk improvement, on the 10 CFR 50.48(c) compliant fire protection program when they are implemented.

All other modifications listed in Attachment 47 of the EPU LAR do not have an adverse impact on the Fire Protection Program, including the plants compliance with the fire protection program licensing basis, 10 CFR 50.48(c).

E6-1

ENCLOSURE 8 Engineering Evaluation 16-E04, Description of BFN RHR Heat Exchanger Test Data Evaluation, Revision 1 (Non-proprietary version)

ZN/ Document Type: QAPD Zachry Nuclear, Inc.

ENGINEERING EVALUATIO N 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION REVISION 1 QA CLASSIFICATION: SAFETY RELATED

(,, (/-z. /1l Dae 6/~7J6 Date Date Computer Code & Version (if applicable): None Zachry Nuclear, Inc. Property Code (if applicable): N/A Client: Graftel, LLC Zachry Nuclear, Inc. Job No. : 4069 Page 1 of 15 Total number of pages including Attachments - 53 Form: N0302F01 Revision: 00-00 Date: 11-2-2011 Page 1 of 1

ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD REVISION HISTORY Revision Revision Description 0 Original Issue 1 Revised Section 4.6 to provide additional details regarding the determination of sensitivity coefficients.

Corrected spelling of Sieder-Tate in Section 4.4.

Other minor changes made at the request of Browns Ferry personnel.

Added the PROTO-HX User Documentation to the reference list as Reference 5.6.

All changes in the body of the evaluation are indicated with change bars in the right margin.

Added Attachment A to provide supporting information from the PROTO-HX User Documentation (Reference 5.6).

Page 2 of 15 Revision 1 Form: N0302F02 Revision: 00-00 Date: 11-2-2011 Page 1 of 1

ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD TABLE OF CONTENTS EVALUATION TITLE SHEET ....................................................................................................................... 1 REVISION HISTORY .................................................................................................................................. 2 TABLE OF CONTENTS .............................................................................................................................. 3 1.0 PURPOSE .................................................................................................................................... 4 2.0 EXECUTIVE

SUMMARY

.................................................................................................................. 4

3.0 BACKGROUND

............................................................................................................................. 5 4.0 EVALUATION ............................................................................................................................... 6

5.0 REFERENCES

............................................................................................................................ 15 TOTAL NUMBER OF PAGES IN EVALUATION BODY ............................................................ 15 Total ATTACHMENTS Pages A. Excerpts from PROTO-HX User Documentation (Reference 5.6) 36 B. Evaluation Review and Verification Information 2 TOTAL NUMBER OF PAGES IN ATTACHMENTS .................................................................. 38 TOTAL NUMBER OF PAGES IN EVALUATION ..................................................................... 53 Page 3 of 15 Revision 1 Form: N0302F04 Revision: 00-00 Date: 11-2-2011 Page 1 of 1

ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD 1.0 PURPOSE This evaluation provides a description of the methods Zachry uses to evaluate data collected during thermal performance tests of the Residual Heat Removal System (RHR) heat exchangers at Browns Ferry Nuclear Plant (BFN). A brief overview of the data collection method is provided along with a detailed discussion of the statistical reduction methods used to determine the nominal heat exchanger performance and application of uncertainty principles to provide an estimate of the test error at a 95% confidence interval. This evaluation is being provided in support of BFNs response to NRC Request for Additional Information (RAI) SCVB RAI-32.

2.0 EXECUTIVE

SUMMARY

As stated above, this evaluation provides a description of the methods Zachry uses to evaluate thermal performance test data for the BFN RHR heat exchangers in support of BFNs response to NRC SCVB RAI-32. Specifically, Zachry was requested to provide the following information:

(a) Describe the statistical evaluation method of the test data with discussion of how the uncertainty analysis methodology described in Chapter 4 of EPRI TR-3002005340 (Reference 5.1) is implemented for evaluating the Random Standard Uncertainty and Systematic Standard Uncertainty combined together.

(b) Justify that the Combined Standard Uncertainty (combination of Random Standard Uncertainty and Systematic Standard Uncertainty) in the test result is conservative.

The following discussion is provided in response to item (a) above:

Zachry performs a statistical reduction of the test data and assesses the uncertainty of the test result through the major contributing error sources typically observed in service water heat exchanger testing as described in Sections 4.2 through 4.7. The methods Zachry uses are consistent with the guidance provided in EPRI TR-3002005340 (Reference 5.1) and EPRI TR-107397 (Reference 5.2).

The methodology for evaluating the variability in the test data is currently expressed as the precision error in accordance with Reference 5.2. First, the test data is reduced to time-series and spatial-series parameter averages. Next, the sampling error is determined via the standard deviation of the sample size. The sampling error is representative of the random error as described in Reference 5.2. This approach remains valid in Reference 5.1, but the precision error is referred to as the random standard uncertainty.

The methodology for evaluating the constant off-set of errors observed in the data is currently expressed as the bias error (B) and is the limit of the fixed elemental error. The instrument, spatial, and methodology biases are combined through the square-root-sum-squares (RSS) to determine the combined bias error for the parameters in accordance with Reference 5.2. The analytical uncertainty is also attributed as a bias error source, but is only accounted for with the analytical method being used. In accordance with Reference 5.1, the constant off-set errors are expressed as the systematic standard uncertainty and can be evaluated on a standard deviation level similar to the random standard uncertainty. The instrument, spatial and methodology Page 4 of 15 Revision 1 Form: N0302F05 Revision: 00-00 Date: 11-2-2011 Page 1 of 1

ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD errors have been previously estimated on the basis of a 95% confidence interval; therefore, half of the bias error is functionally equivalent to the systematic standard error, b, as described in Reference 5.1.

To perform the uncertainty analysis, a numerical determination of the sensitivity coefficients is implemented. The uncertainty in the result can be based on two approaches: (i) combining the systematic and random uncertainties to determine the error; or (ii) evaluating the uncertainty in the result with respect to the systematic and random uncertainties individually. The approach to be selected will be consistent with the systematic uncertainties being evaluated as independent error sources or dependent error sources. Both approaches result in a sensitivity coefficient and an uncertainty in the result for each parameter of interest (i.e. shell flow rate, shell inlet temperature, etc.). The overall uncertainty in the test result is the combined uncertainties for each parameter through the square-root-sum-of squares.

The following discussion is provided in response to item (b) above:

The combined standard uncertainty of the result represents the uncertainty in the test result at the standard deviation level as discussed in Section 4.7. This value is expanded by the Students t score based on the desired confidence level, 95% in this case, and the effective degrees of freedom for the test. The resulting expanded uncertainty of the result represents the uncertainty in the test result at a 95% confidence level. Assuming symmetric uncertainties, the test result could be reported as the mean value (µ) plus or minus the uncertainty (Utest). The test result applies the uncertainty in a conservative fashion. For example, if the performance parameter of interest is fouling resistance, the test result is reported as µ + Utest. If the performance parameter of interest is the heat transfer rate at Design Limiting Conditions, the test result is reported as - Utest. Reporting the result in this manner ensures a conservative accounting of the test uncertainty.

3.0 BACKGROUND

BFN began conducting thermal performance tests of its RHR heat exchangers in 2012 to verify that the heat exchangers are capable of meeting their Design Basis function requirements in support of EPU. Thermal performance tests of the BFN RHR heat exchangers conducted to date have been performed and evaluated following the guidelines of EPRI TR-107397 (Reference 5.2). These guidelines were updated in 2015 and issued as EPRI TR-3002005340 (Reference 5.1). BFN will conduct future tests of the RHR heat exchangers following the Reference 5.1 guidelines (Reference 5.3).

Page 5 of 15 Revision 1 Form: N0302F05 Revision: 00-00 Date: 11-2-2011 Page 1 of 1

ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD 4.0 EVALUATION 4.1 General Test Method In a typical six-point test of a water-to-water heat exchanger, the following parameters are measured using either permanently installed plant equipment or recorded by a data acquisition system (DAS) via temporary test instrumentation:

1. Tube-side inlet temperature (°F);

2. Tube-side outlet temperature (°F);

3. Tube-side flow rate (gpm);

4. Shell-side inlet temperature (°F);

5. Shell-side outlet temperature (°F); and

6. Shell-side flow rate (gpm).

Measurements are typically recorded at a predetermined frequency over the duration of the test.

Longer test durations provide a more accurate estimate of the true heat exchanger performance by virtue of the resulting larger data set. However, the benefit gained in collecting more than 31 data points is not significant. This practical limitation of a larger data set has led to typical tests being conducted over a 30 minute period with data recorded every minute. Once the data has been recorded, statistical reduction methods are used to determine a nominal value for each recorded parameter along with an estimate of the uncertainty in each measurement at a 95%

confidence interval. Evaluation of measurement uncertainty takes into account random and systematic sources of error as described in Sections 4.3 and 4.4, respectively.

The following measurement techniques have been used to record data for the BFN RHR heat exchanger performance tests: tube-side inlet temperature has been measured using four temporary surface mounted temperature sensors secured to the outside of the Service Water (SW) piping; tube-side outlet temperature has been measured using eight temporary surface mounted temperature sensors secured to the outside of the SW piping; tube-side flow rate has been measured using a temporary pressure transmitter connected to the plants permanently installed differential pressure meter; shell-side inlet temperature has been measured using four temporary surface mounted temperature sensors secured to the outside of the RHR piping; shell-side outlet temperature has been measured using eight temporary surface mounted temperature sensors secured to the outside of the RHR piping; and shell-side flow rate has been measured using a temporary pressure transmitter connected to the plants permanently installed differential pressure meter.

A resistance to heat transfer is determined from the average measurements based on a mass and energy balance of the tube and shell sides along with a mean temperature difference (MTD) evaluation. This resistance is a function of the measured parameters and is, therefore, subject to the uncertainty of the measurements. Sections 4.2 through 4.7 provide additional detail regarding the statistical reduction techniques and uncertainty analysis used to determine the demonstrated performance of the RHR heat exchangers.

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ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD 4.2 Determination of Nominal Test Values Zachry uses the following statistical reduction techniques in a hand calculation to determine the nominal values of the measured parameters.

For parameters measured with a single instrument (e.g., tube-side and shell-side flow rates), the time average value of the data points is calculated as follows:

Equation 1 where:

time-average value of the measurements number of data points in the data segment (i.e., number of time steps) single measurement at time step "i" For parameters measured with multiple instruments (e.g., inlet and outlet temperatures), the spatial average value of the data points is calculated as follows:

Equation 2 where:

spatial-average reading of the measurements number of data points in the data segment (i.e., number of time steps) number of locations at which the parameter was measured (i.e.,

number of sensors) single measurement at time step i and location j 4.3 Determination of Random Uncertainty Repeated measurements of a test parameter over the duration of a test will produce random errors that are observed as scatter in the data. This random error is approximated by a Gaussian distribution about the mean, , as demonstrated in Figure 1.

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ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD

µ Figure 1: Random Uncertainty For parameters measured with a single instrument, the standard deviation of the data sample (Sx) is calculated as follows:

Equation 3 where:

single measurement at time step "i" time-average value of the measurements (calculated using Equation 1) number of data points in the data segment (i.e., number of time steps)

For parameters measured with multiple sensors, the standard deviation of the data sample (Sx) is calculated as follows:

Equation 4 Page 8 of 15 Revision 1 Form: N0302F05 Revision: 00-00 Date: 11-2-2011 Page 1 of 1

ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD where:

average value for all measurement locations at time step "i" spatial-average reading of the measurements (calculated using Equation 2) number of data points in the data segment (i.e., number of time steps)

Finally, the standard deviation of the sample mean is calculated as follows for all measurements regardless of how many sensors are used:

Equation 5 where:

standard deviation of the sample mean standard deviation of the data sample (calculated using Equation 3 for single sensor measurements or Equation 4 for multiple sensor measurements) number of data points in the data segment (i.e., number of time steps)

The standard deviation of the sample mean provides an estimate of the random uncertainty in each measured parameter at a 95% confidence interval. This approach is consistent with the methodology of EPRI TR-107397 (Reference 5.2) and TR-3002005340 (Reference 5.1) in that the sampling error of the data is translated as the source of the random error. It is noted that Reference 5.1 recommends using a pooled sample standard deviation for parameters measured using multiple sensors (i.e., temperatures). However, Reference 5.1 also states that either method is acceptable as the random component is relatively insignificant for most heat exchanger testing performed under steady conditions making the difference between the two methods insignificant.

Zachry uses a hand calculation to determine the random uncertainty of each of the measured parameters.

4.4 Determination of Systematic Uncertainty Systematic uncertainty is attributed to those factors that are expected to remain constant for each measurement over the duration of the test (e.g., instrument drift). Because these sources of uncertainty remain constant over the duration of the test, they can be harder to detect.

Systematic uncertainty is demonstrated graphically in Figure 2.

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ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD

µ Figure 2: Systematic Error The four major contributors to systematic uncertainty are:

1. Instrument bias;

2. Method bias;

3. Spatial bias; and

4. Analytical uncertainty.

Each of these contributors to systematic uncertainty is discussed in greater detail in Sections 4.4.1 through 4.4.4.

4.4.1 Instrument Bias The instrument bias is influenced by the instrument calibration uncertainty as well as the uncertainty introduced by transmitters and the DAS itself. Instrument calibration uncertainties for the temperature sensors, current sensors, and differential pressure transmitters used for the BFN RHR heat exchanger tests are provided by instrument calibration records. These individual components are combined as the square root of the sum of the squares (RSS) using Equation 6 to provide an estimate of the instrument bias at a 95% confidence interval.

Equation 6 Page 10 of 15 Revision 1 Form: N0302F05 Revision: 00-00 Date: 11-2-2011 Page 1 of 1

ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD where:

parameter instrument bias uncertainty individual bias contribution of the ith element out of J total contributors 4.4.2 Method Bias The method bias of a measurement is assessed from an evaluation of the engineering or physical principles causing the error. For example, fluid temperatures are measured by surface mounted sensors attached to the outside of the pipe conveying the fluid. The resulting temperature measurement is an approximation of the fluid temperature, the uncertainty of which is affected by conductive heat transfer through the liquid film, fouling layer on the inside of the pipe, and the pipe wall itself as well as convective heat transfer with the surrounding environment. Zachry uses a hand calculation to determine the method bias of the temperature measurements.

The method bias of the flow measurements is caused predominately by the uncertainty associated with calculation of the discharge coefficient of the flow elements themselves.

The uncertainty of the discharge coefficient is affected by such factors as frequency of calibration and proximity to flow disturbances (i.e., elbows or valves) installed in the upstream and downstream piping. It is noted that the current sensor and DP transmitter calibration uncertainties are also accounted for in the method bias analysis for the flow measurements. Zachry uses a hand calculation to combine the individual sources of method bias for the flow measurements as the RSS in the same manner that Equation 6 is used for instrument bias.

4.4.3 Spatial Bias Parameter spatial bias uncertainties are a result of using point measurements to determine a bulk parameter. Spatial bias applies to temperature measurements for the BFN RHR heat exchanger performance tests. The spatial bias of the temperature measurements is calculated as follows:

%&' ( "# #$

"# #$ Equation 7 where:

two-tailed Student's t value (95% confidence, ) degrees of parameter spatial bias uncertainty

%&' (

freedom where ) = L -1) standard deviation of the time-averaged values of the parameter at each location (calculated using Equation 8) number of locations at which the parameter was measured (i.e.,

number of sensors)

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ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD Equation 8 where:

time-averaged value of the parameter at location j (calculated using Equation 1) average value of the parameter at all locations (calculated using Equation 2) number of locations at which the parameter was measured (i.e.,

number of sensors) 4.4.4 Analytical Uncertainty Uncertainties in the empirical correlations used to calculate the film heat transfer coefficients are the dominant contributors to the analytical uncertainty associated with the heat transfer method as described in References 5.1 and 5.2. The uncertainties in the heat transfer correlations used by PROTO-HX are specified as follows:

! "# $%!&'() * ++&' , $*-$+#. *#/ 0# &.+12 Reference 5.4 34  ! 5 #6#*(78 # Reference 5.5 9

34 Reference 5.4 The methods described herein assume steady-state conditions exist during the test.

Application of these methods to data gathered under transient conditions would introduce an additional source of analytical uncertainty in the determination of the test result. The BFN RHR heat exchanger tests are conducted in such a manner as to provide steady-state conditions. Fluid inlet temperatures are evaluated to confirm the existence of steady-state conditions. Therefore, evaluation of the uncertainty due to transient conditions is not typically required.

4.4.5 Combining Individual Components of Systematic Uncertainty Zachry uses a hand calculation to combine the individual components of systematic uncertainty as a RSS using Equation 9 to provide an estimate of the total systematic uncertainty at a 95% confidence interval consistent with the guidance provided in References 5.1 and 5.2.

<

= > :?@ AB >: Equation 9 Page 12 of 15 Revision 1 Form: N0302F05 Revision: 00-00 Date: 11-2-2011 Page 1 of 1

ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD 4.5 Expanded Uncertainties of Measured Parameters Zachry uses a hand calculation to combine the random and systematic uncertainties of each parameter using Equation 10 consistent with the guidance provided in References 5.1 and 5.2.

R L M *%&' ( + Equation 10 where:

systematic uncertainty for parameter x at a 95% confidence interval (calculated using Equation 9) standard deviation of the sample mean for parameter x (calculated using Equation 5)

%&' (

freedom where ) = N -1) two-tailed Student's t value (95% confidence, v degrees of number of data points in the data segment (i.e., number of time steps) 4.6 Determination of Sensitivity Coefficients PROTO-HX performs a parametric analysis of the test data in which the result is re-analyzed multiple times with each analysis varying one parameter at a time, first in the plus direction and then in the minus direction, while holding all other parameters at their nominal value. This creates 12 iterations to address the measurement error in both the positive and negative directions. Four additional iterations are performed to account for the analytical uncertainty associated with the film coefficient as discussed in Section 4.4.4. The methodology of the parametric analysis is in agreement with the numerical approach for determining sensitivity coefficients as described in References 5.1 and 5.2.

PROTO-HX performs the following steps for each iteration of the parametric analysis involving a change in one of the six measured parameters (excerpts from Reference 5.6 are provided in Attachment A for information):

1. A test heat load (QTest) is calculated using Equation 154 from Reference 5.6.

2. An inside film coefficient at the test conditions is calculated using the Petukhov-Kirilloc correlation as described in Section 9.7 of Reference 5.6.

3. An outside film coefficient at the test conditions is calculated as described in Section 9.8 of Reference 5.6.

4. An overall heat transfer coefficient (UTest) is calculated using Equation 148 from Reference 5.6.

5. A test fouling resistance is calculated using Equation 162 from Reference 5.6.

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ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD PROTO-HX calculates a fouling resistance uncertainty or variance using Equation 170 from Reference 5.6 for each iteration of the parametric analysis involving analytical uncertainty (i.e.,

uncertainty in the film coefficients). It is noted that the overall heat transfer coefficient (UTest) is solely a function of the measured parameters and, therefore, is held constant at the nominal test conditions for each iteration involving analytical uncertainty.

The output of the parametric analysis (i.e., fouling resistances from each iteration) is used as input to a hand calculation to determine a sensitivity coefficient for each parameter as follows:

TUV T S 2 ER Equation 11 where:

S sensitivity coefficient for parameter x TUV T absolute value of the change in the result (in this case, fouling R

resistance) overall uncertainty for parameter x (calculated using Equation 10)

For example, consider the following sample test data and parametric results for variations in shell flow rate:

Parameter Value Nominal Shell Flow 7,341 gpm Shell Flow Uncertainty +/- 384 gpm Nominal Test Fouling Resistance 0.000543 hr-ft2-°F/Btu Fouling Resistance at Shell Flow Plus Uncertainty 0.000515 hr-ft2-°F/Btu Fouling Resistance at Shell Flow Minus Uncertainty 0.000573 hr-ft2-°F/Btu From the data above, a sensitivity coefficient for shell flow is calculated using Equation 11 as follows:

TZ[ZZZF F Z[ZZZF\]T ^_ `% ab d %c ^_ `% ab d %c S \[FFE 2j2 ZWk 2

-O$$WX$PY E ]ef2ghi ghi Page 14 of 15 Revision 1 Form: N0302F05 Revision: 00-00 Date: 11-2-2011 Page 1 of 1

ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD 4.7 Combining Sources of Uncertainty in the Overall Test Result Finally, an overall test uncertainty is calculated as a RSS combination of the product of the sensitivity coefficients and parameter uncertainties as follows:

k RO S R Equation 12 This uncertainty is an estimate of the overall test uncertainty at a 95% confidence interval.

Assuming symmetric uncertainties, the estimated heat exchanger performance at the 95%

confidence interval is the nominal value (µ) plus or minus the uncertainty (Utest). For the case of the BFN RHR heat exchangers where fouling is the performance parameter of interest, the nominal fouling resistance (Rf) is calculated using Equation 162 of Reference 5.6 at the nominal test conditions and the uncertainty (Utest) is calculated using Equation 12 above. The resulting estimate of fouling resistance at the 95% confidence interval is Rf +/- Utest. However, the test result is reported as Rf + Utest to ensure a conservative accounting of the test uncertainty.

5.0 REFERENCES

5.1 EPRI TR-3002005340, Service Water Heat Exchanger Testing Guidelines, May 2015 5.2 EPRI TR-107397, Service Water Heat Exchanger Testing Guidelines, March 1998 5.3 Tennessee Valley Authority Letter CNL-16-055, Proposed Technical Specifications (TS)

Change TS-505 - Request for License Amendments - Extended Power Uprate (EPU) -

Supplement 8, Response to Request for Additional Information, dated March 28, 2016 5.4 Thomas, L.C., Heat Transfer - Professional Version, 2nd ed. Tulsa, OK: Capstone Publishing Corporation, 1999 5.5 Incropera, F.P, and DeWitt, D.P., Fundamentals of Heat and Mass Transfer, 4th ed. New York:

John Wiley & Sons, 1996 5.6 Zachry Nuclear Engineering, Inc., User Documentation for Heat Exchanger Modeling Software PROTO-HX Shell and Tube Module, UD-93948-02ST, Version 5.00, Revision H (excerpts included in Attachment A)

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ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD Attachment A Excerpts from PROTO-HX User Documentation (Reference 5.6)

Version 5.00 User Documentation

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 Copyright This manual and the software described in it are copyrighted with all rights reserved. Under copyright laws, neither this manual nor the software may be copied, in whole or part, without the written consent of Zachry Nuclear Engineering, Inc., except in the normal use of the software or to make a backup copy. This exception does not allow copies to be made for others.

1994, 1995, 1996, 1999, 2004, 2011 Zachry Nuclear Engineering, Inc.

14 Lords Hill Rd Stonington, CT 06378 USA Telephone: 860.446.9725 Facsimile: 860.535.9200 Website: http://www.zhi.com Trademark PROTO-HX is a trademark of Zachry Nuclear Engineering, Inc.

Zachry Nuclear Engineering, Inc. Page i of v

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 9.0 Engineering Theory and Basis The following provides a listing of the key equations used in the solution process for a heat exchanger model developed using PROTO-HX Shell and Tube module. The equations have been taken predominately from Reference (2) supplemented by other references of Section 10 as appropriate. Consult Attachment B for a listing and definition of all terms presented.

9.1. Governing Equations QTube = M t c p,t (Tt , o - Tt ,i ) Equation 1 QShell = M s c p,s (Ts ,i - Ts ,o ) Equation 2 QTest = U x A x (DTM ) = U x A x ( F x DTLMTD ) Equation 3 Q = Pf x U x A x (DTM ) = Pf x U x A x ( F x DTLMTD ) Equation 4 9.2. Mass Flow Rates lbm lbm gal 60 min 1 ft 3 Mt ç ÷ = r t ç 3 ÷V& t ç ÷ç ÷ç ÷ hr ø ç ÷ min hr ç 7.48 05 gal ÷ Equation 5 ft ø ø ø ø lbm gal 60 min 1 ft 3 lbm gal 60 min 1 ft 3 M t i = r t i çç 3 ÷÷V& t i ç ÷ç ÷çç ÷÷ = M t o = r t çç 3 ÷÷V& t o ç ÷ç ÷çç ÷÷ ft ø min ø hr ø 7.4805 gal ø o ft ø min ø hr ø 7.4805 gal ø Equation 6 rt o ÷ V t i çç

& =

÷V t o iø r t Equation 7 lbm lbm gal 60 min 1 ft 3 Ms ç =

÷ r s çç 3 ÷÷V& s ç ÷ç ÷çç ÷÷ hr ø ft ø min ø hr ø 7.4805 gal ø Equation 8 9.3. Heat Transfer Area Ao, gross = p d o Lt N t N s Equation 9 Lu = 2 x ( Lsl + 0.3 x Dotl ) Equation 10 Zachry Nuclear Engineering, Inc. Page 73 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 Ao,eff = p d o Leff N t N s Equation 11 Ao,eff Avendor Equation 12 AF = =

Ao, gross Ao, gross Ao = ( Ao, gross )( AF ) = p do Leff Nt , n N s ( AF ) Equation 13 Ai = ( Ai , gross )( AF ) = p di Lt Nt , n N s Equation 14 Ao Ao AR = = Equation 15 Ai Ai do AR = (for bare tubes) Equation 16 di Avendor Equation 17 AR =

Ai 9.4. Log Mean Temperature Difference

( T s ,i - T t , o ) - ( T s , o - T t , i )

DTLMTD =

é ( T s ,i - T t , o )

ln ú

( T s , o - T t ,i ) û Equation 18 9.5. Log Mean Temperature Difference Correction Factor T t , o - T t ,i (T ) - (T2 )i P = = 2 o T s ,i - T t ,i (T1 )i - (T2 )i Equation 19 T s ,i - T s , o (T ) - (T1 )o R = = 1 i T t , o - T t ,i (T2 )o - (T2 )i Equation 20 R -1 1- P d= =

é 1- P P ln ú R 1 1 - P x R û R ¹1 Equation 21 Zachry Nuclear Engineering, Inc. Page 74 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 9.5.1 Counter Flow F = 1.0 9.5.2 Cocurrent Flow Note: For simplification purposes, PROTO-HX no longer supports the analysis of a Cocurrent flow arrangement.

9.5.3 TEMA-E Shell F = 1.0 for single tube pass Otherwise:

h F= Equation 22

é 2 - P(1 + R - h )

d ln ú 2 - P(1 + R + h )û h = R 2 +1 Equation 23 2

Pmax = Equation 24 1 + R +h Using Ns identical units of the TEMA-E shell type in series:

h F= Equation 25

é 2 - P(1 + R - h )

d ln ú 2 - P(1 + R + h )û 1

1 - (X ) Ns Po Equation 26 P= =

R - (X )

1 Ns N s - (Po (N s - 1)) R =1 R ¹1 Po R - 1 X= Equation 27 Po - 1

ç Z N s ÷ - 1 Equation 28 2Ns Pmax= N ø =

ç Z ÷ - R s 2Ns + 2 R =1

ø R ¹1 Zachry Nuclear Engineering, Inc. Page 75 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 h - R +1 Z= Equation 29 h + R -1 h = R 2 +1 Equation 30 9.5.4 TEMA-F shell F = 1.0 for 2 tube passes; for more than 2 tube passes F < 1.0 Additional baffle correction factor Fb,:

Z Fb = d

é1 + ZX ln 1 - ZX úû Equation 31 P

X= Equation 32 2 - PR - P Ab U b Y= Equation 33 Ao U Z = 4 R 2Y + (R - 1) 2 Equation 34

-1 2 L / 12 U b = çç + b ÷ Equation 35 as lb ÷ø Correction = ( F ) x ( Fb ) = (1) x ( Fb ) = Fb Equation 36 9.5.5 TEMA-G Shell J -1 P= a Equation 37 J + 2 Rç e ÷

ø 1

J=

1- D

[(1 + G + 2RG )e a + 2RD ] Equation 38 1 - e -b 1 Equation 39 G= =

2 R - 1 R ¹ 0.5 2 Fd R 0.5 Zachry Nuclear Engineering, Inc. Page 76 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 1 - e-a Equation 40 D=

2R + 1 2R + 1 a= Equation 41 4 Fd 2R - 1 b= Equation 42 2 Fd 2R + 1 Pmax = = 1.0 R £ 0.5 Equation 43 2 R 2 + R + 1 R > 0.5 9.5.6 TEMA-J Shell Single tube pass:

-( R +0.5) 1+ 1 2R - 1 2R + f f P = 1- ç ÷ -( R -0.5)

= 1-2R + 1 ø 2R - f R ¹0.5 2 + ln f Equation 44 R0.5 1

f = exp ç ÷ Equation 45 Fd ø 2R - 1 Pmax =1 - ç ÷ = 1.0 R£0.5 Equation 46 2 R + 1 ø R>0.5 Any even number of tube passes:

-1 Rf R f 1 P = çç R + - ÷÷ f - 1 f - 1 ln f ø Equation 47 9.6. Overall Heat Transfer Coefficient DTM DT = Q R Q = = U x Aref x F x (DTLMTD )

R Equation 48 1

U x Aref = Equation 49 R

1 U=

( R )Aref Equation 50 Zachry Nuclear Engineering, Inc. Page 77 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 1

Rhi = Equation 51 hi Ai 1

Rho = Equation 52 ho Ao tw d - di 1 Rw = = çç o ÷÷ Equation 53 k w Aw 2k w ø Aw A - Ai Equation 54 Aw = çç o ÷÷ ln( Ao / Ai ) ø bare tube:

d - di d - di Equation 55 Aw = p Nt Lt çç o ÷÷ or Aw = p Nt Lt çç o ÷÷ ln(do / di ) ø ln(do / di ) ø 1 R fi d - di 1 R fo 1 R= + + çç o ÷÷ + +

hi A i A i 2k w ø Aw A o ho A o Equation 56 1

U =

1 ç Aref A

÷ + R fi ç ref d o - d i Aref

÷+ç A

÷ + R fo ç ref 1 Aref

÷ç ÷+ ç ÷ Equation 57 hi ç A i ÷ ç A ÷ ç 2k w ÷ç Aw ÷ ç Ao ÷ h o ç Ao ÷

ø i ø ø ø ø ø 1

U =

1 Ao d - di A o 1

ç ÷ + R fi ç A o ÷ + ç o ÷çç ÷÷ + R fo + Equation 58

ç ÷ ç ÷ ç ÷ hi A i ø A i ø 2k w ø Aw ø ho Ao R f = R fi çç ÷÷ + R fo Ai Equation 59

ø for bare tubes do R f = R fi çç ÷÷ + R fo di Equation 60

ø for tubes with low profile fins R f = R fi ( AR )+ R fo Equation 61 Zachry Nuclear Engineering, Inc. Page 78 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 1 1 Rf = -

U service U clean Equation 62 Table 3 Thermal Conductivity of Materials Conductivity Conductivity Alloy Material (Btu/hr-ft-°F) Alloy Material (Btu/hr-ft-°F) 706 Copper Nickel, 10% 26.0 445 Phosphorized Admiralty 64.0 710 Copper Nickel, 20% 21.0 443 Arsenical Admiralty 64.0 715 Copper Nickel, 30% 17.0 122 Deoxidized Copper 196.0 304 Austenitic Stainless Steel 9.4 142 Arsenical Copper 112.0 304L Austenitic Stainless Steel 9.4 230 Red Brass, 85% 92.0 310 Austenitic Stainless Steel 8.2 280 Muntz Metal 71.0 316 Austenitic Stainless Steel 9.4 Ti-0.2 Pd Titanium 9.5 316L Austenitic Stainless Steel 9.4 200 Nickel Alloy 44.5 321 Austenitic Stainless Steel 9.3 Monel 400 12.6 347 Austenitic Stainless Steel 9.3 1060 Aluminum 128.0 348 Austenitic Stainless Steel 9.3 3003 Aluminum 111.0 687 Aluminum Brass 58.0 5052 Aluminum 80.0 608 Aluminum Bronze 46.0 Copper 225.0 9.7. The Inside Film Heat Transfer Coefficient hi d i Nu t =

12 k t Equation 63 12 k t hi = çç ÷÷ Nu t di ø Equation 64 9.7.1 Turbulent Flow Regime Option 1: The Petukhov-Kirillov Correlation (10,000 < Ret<5,000,000 and 0.5 < Pr < 200)

( f/ 2) Re t Prt Nu t =

1.07+12.7 f/ 2 Prt -1 ( 2/ 3

) Equation 65 1

f=

(1.58 ln Re t - 3.28)2 Equation 66 Nut,corrected = (Nut ) x (e ) Equation 67 n

Pr e = çç t ÷÷ Prt , w ø Equation 68 where n = 0.11 for shell - side hot n = 0.25 for tube - side hot Zachry Nuclear Engineering, Inc. Page 79 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 Option 2: The Sieder-Tate Correlation (10,000 < Ret and 0.7 < Prt < 16,700 and Lt /di > 10) 0.14 0.8 mt 1/3 ç Nut = 0.027 Ret Pr t ÷

çm ÷ Equation 69 t,w ø

 Transition Flow Regime 5HWDQG3UWDQG/WGL! 

0.14 1/3

é mt Nu t = Pr t ú '

m t,w ûú

é é -1/3 4-2

é 4-0.027 Ret 0.8 - A - B çç Lt ÷÷ ú çç 10 Ret ÷÷ EXP - C çç 10 Ret ÷÷ú ú d i ø úû 7900 ø 7900 øû úû Equation 70 where: A = 38.25734951 B = 74.21557464 C = 1.1365 9.7.3 Laminar Flow Regime 5HWDQG3UDQG/WGL!

0.14 -1/3 1/3 mt 1/3 ç Lt Nut = 1.86 ( Ret ) ( Pr t ) ç ÷ çç ÷÷ m ÷ di ø Equation 71 t,w ø 9.7.4 Wall Temperature Calculation QTube Tt ,w = Tt ,av + (for tube side as cold side) Equation 72 hi Ai QTube Tt , w = Tt , av - (for tube side as hot side) Equation 73 hi Ai 9.7.5 Reynolds Number and Prandtl Number Calculations Straight tube:

N N wc = Integer ç t ÷

çN ÷ Equation 74 p ø U-tube:

N N wc = Integer ç t ÷x2

çN ÷ Equation 75 p ø Zachry Nuclear Engineering, Inc. Page 80 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 60 r t V& t M wc = Equation 76 7.4805 N wc 2

p di p Awc = ç ÷ = di 2

4 12 ø 576 Equation 77 2

p di p 2 At = ç ÷ (N wc )= ç d i ÷(N wc ) = (Awc )(N wc )

4 12 ø 576 ø Equation 78 M wc

ç ÷ = r t AwcVt 3600 ø Equation 79 M wc Vt =

3600 r t Awc Equation 80 Mt Vt =

3600 r t At Equation 81 Vt ( d i /12)

Ret = Equation 82 nt mt nt=

3600 r t Equation 83 3600 r t V t ( d i /12) 300 r t V t d i M wc d i Re t = = =

mt mt 12 m t Awc Equation 84 c p ,t m t Pr t =

kt Equation 85 9.8. Outside Film Heat Transfer Coefficient ho = hoff a Equation 86 9.8.1 Shell-Side Velocity M

M bb = ç s ÷ for TEMA G and TEMA J shells Equation 87 2 ø Zachry Nuclear Engineering, Inc. Page 81 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 M bb = M s for all other shells Equation 88 Option 1: The Geometry Method

é Amin = Lbc Lbb + D ctl ( Ltp - d o )ú Equation 89 144 Ltp,eff û Lbb = Ds - Dotl Equation 90 Dctl = Dotl - d o Equation 91 Table 4 Effective Tube Pitch as a function of Layout Angle LAYOUT ANGLE (q tp) Ltp,eff 30° Ltp 45° 0.707 Ltp 90° Ltp M bb V s= Equation 92 3600 r s ,av Amin Option 2: The Back-Calculation Method M bb Amin = Equation 93 3600 r s (V s ,vendor)

M bb m& s = Equation 94 A min V s ( d o /12)

Res = Equation 95 ns ms ns= Equation 96 3600 r s 3600 r s V s ( d o /12) m& s d o M bb d o Re s = = = Equation 97 ms 12 m s 12 m s A min c p, s m s Pr s =

Equation 98 ks Zachry Nuclear Engineering, Inc. Page 82 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 9.8.2 The Ideal Outside Film Heat Transfer Coefficient

-0.14 a a 2/3 m 1.33 j = (Prs ) çç s ÷÷ = ( a 1 ) çç ÷ ( Re s )a 2

÷ Equation 99 c p , s m& s m s,w ø Ltp / d o ø a3 a= Equation 100 1+ 0.14( Re s )a 4 a 0.14

é 1.33 é ms

-2/3 a = ( a1 ) çç ÷ ( Res ) ú x c p ,s m& s (Prs ) ç a2 ÷ ú Equation 101 L / d ÷ ú çm ÷ ú tp o ø û s,w ø û Table 5 Ideal Outside Film Coefficient Input Parameters LAYOUT Res a1 a2 a3 a4 ANGLE (q tp) 30° 105 - 104 0.321 -0.388 1.450 0.519 104 - 103 0.321 -0.388 103 - 102 0.593 -0.477 102 - 10 1.360 -0.657

<10 1.400 -0.667 45° 10 - 104 5

0.370 -0.396 1.930 0.500 104 - 103 0.370 -0.396 103 - 102 0.730 -0.500 102 - 10 0.498 -0.656

<10 1.550 -0.667 90° 10 - 104 5

0.370 -0.395 1.187 0.370 104 - 103 0.107 -0.266 103 - 102 0.408 -0.460 102 - 10 0.900 -0.631

<10 0.970 -0.667 0.14

- 2/3 ms Equation 102 a = j c p , s m& s (Pr ) ç ÷

çm ÷ s,w ø 9.8.3 Wall Temperature Correction QShell Ts , w = Ts , av - (for shell side as hot side) Equation 103 ho Ao Zachry Nuclear Engineering, Inc. Page 83 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 QShell Ts , w = Ts , av + (for shell side as cold side) Equation 104 ho Ao 9.8.4 Correction Factor Calculation Using the Back-Calculation Method 1

ho,vendor =

1 1 d o

çç ÷÷ - çç ÷÷çç ÷÷ - R fi çç d o ÷÷ - çç d o ÷÷ ln çç d o ÷÷ - R fo Equation 105 U vendor ø hi ø d i ø d i ø 24 k w ø d i ø ho,vendor hoff = Equation 106 a

9.8.5 Correction Factor Calculation Using the Geometry Method hoff = JL JC JB JS Equation 107 The Baffle Leakage Correction Factor, JL S sb = 0.00436 Ds L sb (360 - q ds ) Equation 108 360 é Bc Equation 109 q ds = ç ÷2 arccos 1 - 2ç ÷ú 2p ø 100 øû 360 é arccos 1 - 2ç Bc ÷ú Equation 110 q ds =

p 100 øû Lbch Equation 111 Bc = 100çç ÷÷ Ds ø For straight tubes:

Equation 112 p

S tb = [( d o + Ltb )2 - d o2 ]( N T )(1 - F w )

4 For U tubes:

p

[( d o + Ltb )2 - d o2 ](2 N T )(1 - F w )

Equation 113 S tb =

4 Zachry Nuclear Engineering, Inc. Page 84 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 1 2 é q 2p q 2p A = ç ÷ r çç ctl ÷÷ - sin çç ctl ÷÷ú Equation 114 2 ø 360 ø 360 øû q p sin ç ctl ÷ q

Fw = ctl - 180 ø Equation 115 360 2p Ds é Bc ü q ctl = 2 arccos í 1 - 2ç 100 ÷ú ý Equation 116 Dctl øû S sb +

rs = rlm = S sb S tb Equation 117 S sb + S tb 144 A min J L = 0.44(1 - rs ) + [1 - 0.44(1 - rs )] exp (-2.2rlm ) Equation 118 The Segmental Baffle Window Correction Factor, JC FC = 1 - 2FW Equation 119 J C = 0.55 + 0.72FC Equation 120 The Bypass Correction Factor, JB D s é1 - 2 B c Equation 121 N tcc = ç ÷ú L pp 100 øû Lbch Equation 122 Bc = 100ç ÷ Ds ø Table 6 Lpp as a function of Layout Angle LAYOUT ANGLE (q tp) Lpp 30° 0.866Ltp 45° 0.707Ltp 90° Ltp N ss r ss =

N tcc Equation 123 Lbc ( -

Sb= D s D otl ) Equation 124 144 Zachry Nuclear Engineering, Inc. Page 85 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 Sb F sbp = Equation 125 Amin

[

J B = exp - C bh F sbp(1 - 3 2 r ss ) ] Equation 126 The Non-Equal Inlet/Outlet Baffle Spacing Correction Factor, Js Straight tubes:

é 12 L t N b = Integer - 1ú Equation 127 Lbc û U-tubes:

é 12 L t N b = Integer - 1ú Equation 128 2 Lbc û Lbi Lbo L i = L o = Equation 129 Lbc Lbc

( N b - 1) + ( L i )(1-n) + ( L o )(1-n)

Js= Equation 130

( N b - 1) + ( L i ) + ( L o )

9.9. Heat Exchanger Model Solution Methodology

( - ) - ( T ho - T ci ) Q LMTD = T hi T co = Equation 131

é( - ) UA o F ln T hi T co ú

( T ho - T ci ) û Q é( - )

( T hi - T ho ) - ( T co - T ci ) = ln T hi T co ú Equation 132 UA o F ( T ho - T ci ) û Q

Q = M h c p h ( T hi - T ho ) ( T hi - T ho ) = Equation 133 M h c ph Q

Q = M c c p c ( T co - T ci ) ( T co - T ci ) = Equation 134 M c c pc Zachry Nuclear Engineering, Inc. Page 86 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 1 1 1 é -

- = ln T hi T co ú Equation 135 M h c p h M c c pc UA o F T ho - T ci û

é T hi - T co éUA o F( M c c p - M h c p )

c h

ú = exp ú Equation 136 T ho - T ci û M h M c c p h c pc úû

éUA o F( M c c p - M h c p )

c h b = exp ú Equation 137 M h M c c p h c pc úû b T ho = T hi - T co + b T ci Equation 138 M c c p c ( T co - T ci ) = M h c p h ( T hi - T ho )

Equation 139 M h c ph T co = T ci + ( T hi - T ho ) Equation 140 M c c pc M hcp T hi çç 1 -

h÷

+ ( b - 1)

÷ T ci Equation 141 M c c p cø T ho =

ç b - M h c ph ÷

ç M c c pc ÷ø M h c ph g= Equation 142 M c c pc T hi (1 - g ) + T ci ( b - 1) , g ¹ 1 T ho = Equation 143 (b -g )

Q M hc p h( T hi - T ho )

LMTD = ( T hi - T co ) = ( T ho - T ci ) = = Equation 144 UA o F UA o F C = M h c p h = M c c pc Equation 145 Zachry Nuclear Engineering, Inc. Page 87 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 T hi (C) + T ci ( UA o F)

T ho = ,g = 1 Equation 146 (C + UA o F) 9.10. Fouling Calculations QTest = UTest x Ao x F x DTLMTD Equation 147 QTest U Test =

Ao x F x DTLMTD Equation 148 QTest = QTube = M t c p,t (Tt ,o - Tt ,i ) Equation 149 QTest = QShell = M s c p ,s (Ts ,i - Ts ,o ) Equation 150 2 2 2 UM s UTs,i - UTs,o UQShell = QShell çç ÷÷ + ç ÷ ç ÷

ç (T - T ) ÷ + ç (T - T ) ÷ Ms ø s,i s,o ø s,i s,o ø Equation 151 2 2 2 UM t UTt,o - UTt,i UQTube = QTube çç ÷÷ + ç ÷ +ç ÷

ø (Tt,o - Tt,i ) ø (Tt,o - Tt,i ) ø

ç ÷ ç ÷ Mt Equation 152 2 2 UQShell UQTube QTest = QTube 2ç ÷ + Q ç ÷

çU 2 ÷ Shell ç 2 2 ÷ Equation 153 QShell + UQTube ø UQShell + UQTube ø QTubeUQ2Shell + QShellUQ2Tube QTest =

UQ2Shell + UQ2Tube Equation 154 QTest = QShell = QTube = Qh = Qc Equation 155 HBE = Qh - Qc Equation 156 Q - Qc HBE (%) = çç h ÷÷ x100 Qc ø Equation 157 2 2 2 2 2 2 Qh UM s UTs,i - UTs,o U M t UTt,o - UTt,i U HBE = 100 ç ÷ +ç ÷ ç ÷ ç ÷ ç ÷ ç ÷ Qc çM ÷ ç (T - T ) ÷ + ç (T - T ) ÷ + ç M ÷ + ç (T - T ) ÷ + ç (T - T ) ÷ s ø s,i s,o ø s,i s,o ø t ø t,o t,i ø t,o t,i ø Equation 158 Zachry Nuclear Engineering, Inc. Page 88 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 2 2 Q UQShell UQTube UHBE = 100 h ç

ç Q

÷ +ç

÷ ç Q

÷

÷ Qc Shell ø Tube ø Equation 159 HBE (%) £ UHBE Equation 160 2 2 2 2 2 2 QShell UV&s UTs,i - UTs,o UV&t UTt,o - UTt,i UHBE = 100 ç ÷ +ç ÷ ç ÷ ç ÷ ç ÷ ç ÷ QTube ç V& ÷ ç (T - T ) ÷ + ç (T - T ) ÷ + ç V& ÷ + ç (T - T ) ÷ + ç (T - T ) ÷ s ø s,i s,o ø s,i s,o ø t ø t,o t,i ø t,o t,i ø Equation 161 1 1 d d 1 Rf = - çç Ao ÷÷çç ÷÷ - çç o ÷÷ ln çç o ÷÷ - çç ÷÷ U Test Ai ø hi ø Test 24k w ø d i ø ho ø Test Equation 162 9.11. Uncertainty Calculation UTest = ( U )

x x 2

Equation 163 RPx + Px - RPx R x x = =

(Px + Px ) - Px Px Equation 164 x =

(R Px + Px ) (

- RPx - RPx - Px - RPx )= R Px + Px - RPx - Px

=

DRx (Px + Px ) - (Px - Px ) 2 Px 2 Px Equation 165 RPx + Px - RPx - Px DRx x = =

2Ux 2Ux Equation 166 DQ x x =

2Ux Equation 167 2 2

é DQx* DQ*

UTest = (xUx ) = çç 2

÷÷(Ux )ú = çç 2 x ÷÷ 2Ux ø Equation 168

û ø UTest UTest (%) = x 100 Equation 169 Q*

1 1 Ao d d 1 Rf = - çç ÷÷ - çç o ÷÷ ln çç o ÷÷ -

U Test [(1 +/- d hi )hi ] Ai ø 24k w ø d i ø [(1 +/- d ho )ho ] Equation 170 Zachry Nuclear Engineering, Inc. Page 89 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 d h = 6%

i with Petukhov-Kirillov Equation 171 d h = 25%

i with Seider-Tate Equation 172 d h = 25%

o Equation 173 Zachry Nuclear Engineering, Inc. Page 90 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 10.0 References (1) Zachry Nuclear Engineering - Software Requirements Specification (SRS) for Shell and Tube Heat Exchanger Thermal Performance Modeling Software - PROTO-HX, SRS-93948-02ST, Version 5.00, Revision K (2) Heat Exchanger Design Handbook , Hemisphere Publishing Co. 1983 and 1999 (3) Fundamentals of Heat and Mass Transfer, F.P. Incropera and D.P. DeWitt, J. Wiley and Sons, 1981 (4) Not used (5) Heat Transfer - Professional Version, Lindon C. Thomas, Capstone Publishing Corporation, Second Edition, 1999 (6) ASME Standards and Guides for Operation and Maintenance of Nuclear Power Plants, Part 21, Appendix C, 1994 (7) Process Heat Transfer , G.F. Hewitt, G.L. Shires, T.R. Bott, CRC Press, London England, 1994 (8) Principles of Enhanced Heat Transfer, R.L. Webb, J. Wiley and Sons, 1994, pp. 395-431 and pp. 520-525 (9) Proto-Power Corporation Calculation 96-016 Revision -, Calculating the Tube Inside Heat Transfer Coefficient, hi, for Reynolds Numbers Corresponding to the Transition from Laminar to Turbulent Flow for Shell-and-Tube Heat Exchanger, dated 3/13/96 (10) Proto-Power Corporation Calculation 93-048 Revision A, Fluid Properties - Fresh Water - Range 32°F to 500°F (11) Proto-Power Corporation Calculation 93-049 Revision A, Fluid Properties - Salt Water - Range 32°F to 320°F - Salinity 35 g/kg (12) Proto-Power Corporation Calculation 96-008 Revision -, Safety and Turbine Auxiliaries Cooling System (STACS) PROTO-HX Heat Exchanger Models, Attachment P, Brackish Water (12 parts per thousand salinity) Thermodynamic Property Curve Fits (13) Proto-Power Corporation Calculation 96-008 Revision -, Safety and Turbine Auxiliaries Cooling System (STACS) PROTO-HX Heat Exchanger Models, Attachment O, Water + 30% Ethylene Glycol Thermodynamic Property Curve Fits (14) Proto-Power Corporation Calculation 93-057, Revision A, Fluid Properties - Mobilguard 450 Lube Oil - Range 32°F to 300°F (15) EPRI Report TR-107397, Service Water Heat Exchanger Testing Guidelines, March 1998 (16) The Century Heat Exchange Tube Manual; Century Brass, Waterbury, CT, 1977 (17) Zachry Nuclear Engineering - Software Quality Assurance Plan (SQAP) Heat Exchanger Thermal Performance Modeling Software - PROTO-HX, SQAP-93948-02 (18) Zachry Nuclear Engineering - Software Design Description (SDD) for Shell and Tube Heat Exchanger Thermal Performance Modeling Software - PROTO-HX, SDD-93948-02ST-Basis, Version 5.00, Revision I.

(19) Engineering Formulas, K. Gieck and R. Gieck, seventh edition, McGraw-Hill, 1997 (20) FAX, Jerry Taborek to Chris DAngelo dated March 26, 1993 (21) Karl A. Gardner, Mean Temperature difference in Multipass Exchangers: Correction Factors with Shell Fluid Unmixed, Industrial Engineering Chemistry, December, 1941, page 1495 Zachry Nuclear Engineering, Inc. Page 91 of 91

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 Attachment B Nomenclature Zachry Nuclear Engineering, Inc. Attachment B, Page 1 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 General Terms a coefficient for calculation of the Colburn heat transfer factor a1 coefficient for calculation of the Colburn heat transfer factor a2 coefficient for calculation of the Colburn heat transfer factor a3 coefficient for calculation of the Colburn heat transfer factor a4 coefficient for calculation of the Colburn heat transfer factor A coefficient for defining the tube-side convection film coefficient in the transitional flow regime Ab surface area of longitudinal baffle in a TEMA-F heat exchanger (ft2)

Ai , gross gross heat transfer area on inside surface (ft2)

Ai,gross gross heat transfer area on inside surface (in2)

AF area factor, the ratio of the effective heat transfer area to the gross outside surface area Ai effective heat transfer area on inside surface (ft2)

Ai effective heat transfer area on inside surface (in2)

Amin shell-side minimum cross-sectional flow area between central baffles (ft2)

Ao,eff effective heat transfer area on outside surface (ft2)

Ao,eff effective heat transfer area on outside surface (in2)

Ao, gross gross heat transfer area on outside surface (ft2)

Ao,gross gross heat transfer area on outside surface (in2)

Zachry Nuclear Engineering, Inc. Attachment B, Page 2 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 Ao effective heat transfer area on outside surface (ft2)

Ao effective heat transfer area on outside surface (in2)

AR area ratio, the ratio of the outside surface area to the inside surface area Aref reference heat transfer area (ft2)

At tube-side flow area (ft2)

Avendor heat transfer surface area specified by the heat exchanger manufacturer (ft2)

Aw logarithmic mean area of the tube wall between the inside and outside diameters (ft2)

Aw logarithmic mean area of the tube wall between the inside and outside diameters (in2)

Awc cross sectional flow area per water circuit (i.e., per tube) (ft2)

B coefficient for defining the tube-side convection film coefficient in the transitional flow regime Bc baffle cut, the baffle cut height expressed as a percentage of the shell inside diameter

(%)

C coefficient for defining the tube-side convection film coefficient in the transitional flow regime Cbh coefficient used in the bundle bypass correction factor (fixed at 1.25 limiting its application to non-laminar flows with Res > 100) cp,s specific heat of shell-side fluid at the bulk average temperature (Btu/lbm-°F) cp,t specific heat of tube-side fluid at the bulk average temperature (Btu/lbm-°F) di tube inside diameter (ft) di tube inside diameter (in)

Zachry Nuclear Engineering, Inc. Attachment B, Page 3 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 do tube outside diameter (ft) do tube outside diameter (in)

D simplifying term used for calculation of the LMTD correction factor Dctl diameter of the circle circumscribed through the centers of the outermost tubes of the tube bundle (in)

Dotl tube circle diameter (in)

Ds shell inside diameter (in)

F log mean temperature difference correction factor f simplification term used in the calculation of the inside film heat transfer coefficient using the Petukhov-Kirillov correlation Fb TEMA-F baffle correction factor Fc fraction of tubes in crossflow between baffle tips Fsbp ratio of bypass to crossflow area Fw fraction of tubes in a baffle window G simplifying term used for calculation of the LMTD correction factor HBE heat balance error, the difference between the tube-side and shell-side heat load calculated from test data HBE% heat balance error expressed as a percent referenced to the heat load calculated from the cold-side test data hi inside (tube-side) film heat transfer coefficient (Btu/hr-ft2-°F) ho outside (shell-side) film heat transfer coefficient (Btu/hr-ft2-oF) hoff outside (shell-side) film heat transfer coefficient correction factor Zachry Nuclear Engineering, Inc. Attachment B, Page 4 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 j Colburn heat transfer factor (dimensionless)

J simplifying term used for calculation of the LMTD correction factor JB tube bundle bypass correction factor JC segmental baffle window correction factor JL baffle leakage correction factor JS non-equal inlet/outlet baffle spacing correction factor ks thermal conductivity of shell-side fluid at bulk average temp (Btu/hr-ft-°F) kt thermal conductivity of tube-side fluid at bulk average temp (Btu/hr-ft-°F) kw thermal conductivity of the tube wall (Btu/hr-ft-°F)

Lt tube length (ft)

Lt tube length (in)

Lb thickness of TEMA-F longitudinal baffle plate (in)

Lbb bypass channel diametral gap (in)

Lbc central baffle spacing (in)

Lbch baffle cut height (in)

Lbi inlet baffle spacing (in)

Lbo outlet baffle spacing (in)

Leff effective tube length (ft)

Leff effective tube length (in)

Zachry Nuclear Engineering, Inc. Attachment B, Page 5 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 Li dimensionless length ratio, ratio of inlet baffle spacing to central baffle spacing LMTD log mean temperature difference (oF)

Lo dimensionless length ratio, ratio of outlet baffle spacing to central baffle spacing Lpn tube layout characteristic - tube pitch in direction normal to flow (in)

Lpp tube layout characteristic - tube pitch in direction parallel to flow (in)

Lsb clearance between shell inside diameter and baffle outside diameter (in)

Lsl straight length of tube (the straight section of a U-tube tube bundle) (ft)

Lsl straight length of tube (the straight section of a U-tube tube bundle) (in)

Ltb diametral clearance between baffle hole and tube outside diameter (in)

Ltp tube pitch (in)

Ltp,eff effective tube pitch (in)

Lu representative total length of the tubes in a U-tube bundle (in) m& s maximum shell-side crossflow mass velocity (lbm/ft2 hr)

Mbb shell-side mass flow rate between central baffles (lbm/hr)

Ms shell-side mass flow rate (lbm/hr)

Mt tube-side mass flow rate (lbm/hr)

Mwc tube-side mass flow rate per water circuit (lb/hr) n coefficient for non-equal inlet/outlet baffle spacing correction factor (fixed at 0.6 limiting its application to non-laminar flows with Res > 100)

Nb number of baffles Zachry Nuclear Engineering, Inc. Attachment B, Page 6 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 NP number of tube passes Ns number of shells in series Nss number of sealing strips in one baffle Nt number of tubes Ntcc number of tube rows crossed between baffle tips in one baffle section Nut Nusselt number for the tube-side fluid Nut,corrected Nusselt number for the tube-side fluid corrected for the temperature change across the tube-side film Nwc number of water circuits P thermal effectiveness parameter Pf extrapolation performance factor Pmax the highest (asymptotic) value of the thermal effectiveness parameter Po thermal effectiveness parameter based on the overall terminal temperatures for Ns shells in series Prs Prandtl number for the shell-side based on the bulk average temperature (dimensionless)

Prt Prandtl number for the tube-side fluid based on the bulk average temperature (dimensionless)

Prt,w Prandtl number for the tube-side fluid based on the inside tube wall temperature (dimensionless)

Px value of parameter x in numerical perturbation Q heat transfer rate (Btu/hr)

Zachry Nuclear Engineering, Inc. Attachment B, Page 7 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 Qh hot-side heat transfer rate (Btu/hr)

Qc cold-side heat transfer rate (Btu/hr)

Q* extrapolated heat transfer rate (Btu/hr)

QShell heat transferred to/from the shell-side fluid (Btu/hr)

QTest heat transfer rate used to represent test conditions (Btu/hr)

QTube heat transferred to/from the tube-side fluid (Btu/hr) r radius of tube bundle (inches)

R heat capacitance ratio Res Reynolds number of the shell-side fluid based on the bulk average temperature (dimensionless)

Ret Reynolds number of the tube-side fluid based on the bulk average temperature (dimensionless)

Rf overall fouling thermal resistance (hr-ft2-°F/Btu)

Rfi inside surface fouling thermal resistance (hr-ft2-°F/Btu)

Rfo outside surface fouling thermal resistance (hr-ft2-°F/Btu)

Rhi inside film convective thermal resistance (hr-ft2-°F/Btu)

Rho outside film convective thermal resistance (hr-ft2-°F/Btu) rlm leakage ratio related to baffle leakage correction factor RPx analysis result as a function of parameter x rs leakage ratio related to baffle leakage correction factor Zachry Nuclear Engineering, Inc. Attachment B, Page 8 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 rss sealing strip correlation parameter Rw tube wall thermal resistance (hr-°F/Btu)

Sb bypass area within one baffle (ft2)

Ssb shell-to-baffle leakage area (in2)

Stb tube-to-baffle hole leakage area (in2)

T1 shell-side temperature (oF) for correlation to heat exchanger configuration figures accompanying the DTLMTD correction factor equations T2 tube-side temperature (oF) for correlation to heat exchanger configuration figures accompanying the DTLMTD correction factor equations Tc,i cold stream inlet temperature (oF)

Tc,o cold stream outlet temperature (oF)

Th,i hot stream inlet temperature (oF)

Th,o hot stream outlet temperature (oF)

Ts shell-side temperature (oF)

Ts,av shell-side average temperature (average of inlet and outlet temperatures) (oF)

Ts,i shell-side inlet temperature (oF)

Ts,o shell-side outlet temperature (oF)

Ts,w shell-side temperature at the tube outside wall surface (oF)

Tt tube-side temperature (oF)

Tt,av tube-side average temperature (average of inlet and outlet temperatures) (oF)

Tt,i tube-side inlet temperature (oF)

Zachry Nuclear Engineering, Inc. Attachment B, Page 9 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 Tt,o tube-side outlet temperature (oF)

Tt,w tube-side temperature at the tube inside wall surface (oF) tw tube wall thickness (ft) tw tube wall thickness (in)

U overall heat transfer coefficient (Btu/hr-ft2-°F)

U

• extrapolated overall heat transfer coefficient (Btu/hr-ft2-°F)

Ub overall heat transfer coefficient for TEMA-F longitudinal baffle plate (Btu/hr-ft2-°F)

Uclean overall heat transfer coefficient for a heat exchanger in the clean condition (i.e., overall fouling = 0) (Btu/hr-ft2-°F)

UM s uncertainty in the shell-side mass flow rate (lbm/hr)

UM t uncertainty in the tube-side mass flow rate (lbm/hr)

UQ uncertainty in the shell-side heat transfer rate (Btu/hr)

Shell UQ uncertainty in the tube-side heat transfer rate (Btu/hr)

Tube Uservice overall heat transfer coefficient for a heat exchanger in the service or fouled condition (i.e., overall fouling = design value) (Btu/hr-ft2-°F)

UTest overall heat transfer coefficient calculated from test data (Btu/hr-ft2-°F)

UTest uncertainty in the test result UTest (%) uncertainty in the test result expressed as a percentage of the nominal result UTs, i uncertainty in the shell-side inlet temperature (oF)

Zachry Nuclear Engineering, Inc. Attachment B, Page 10 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 UTs, o uncertainty in the shell-side outlet temperature (oF)

UT t, i uncertainty in the tube-side inlet temperature (oF)

UT t, o uncertainty in the tube-side outlet temperature (oF)

Uvendor manufacturers overall heat transfer coefficient at design conditions (Btu/hr-ft2- oF)

UV& s uncertainty in the shell-side volumetric flow rate (gpm)

UV& t uncertainty in the tube-side volumetric flow rate (gpm)

Ux uncertainty in parameter x uncertainty in the heat balance error (%)

U HBE V& s shell-side volumetric flow rate (gpm)

Vs shell-side maximum mean velocity (ft/sec)

Vs,vendor shell-side maximum mean velocity specified by vendor for design conditions (ft/sec)

V&t tube-side volumetric flow rate (gpm)

Vt tube-side mean velocity (ft/sec)

X simplifying term used for calculation of the LMTD correction factor Y simplifying term used for calculation of the LMTD correction factor Z simplifying term used for calculation of the LMTD correction factor Greek Terms a ideal outside film heat transfer coefficient for pure cross flow over an ideal tube bank (Btu/hr-ft2-°F)

Zachry Nuclear Engineering, Inc. Attachment B, Page 11 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 as outside film heat transfer coefficient for the TEMA-F longitudinal baffle plate (Btu/hr-ft2-

°F) simplifying term used in solving for Tho simplifying term used for calculation of the LMTD correction factor d hi uncertainty in the inside (tube-side) film heat transfer coefficient d ho uncertainty in the outside (shell-side) film heat transfer coefficient Px unit change in input parameter x DQ*x change in extrapolated test result as a function of a change in input parameter x (Btu/hr per Dx)

R x change in the output result due to a change in the input parameter x DT temperature difference (°F)

DTM mean temperature difference (°F)

DTLMTD log mean temperature difference (°F) e temperature correction factor for the Nusselt number used to correct the inside film heat transfer coefficient found using the Petukhov-Kirillov correlation f simplifying term used for calculation of the LMTD correction factor lb thermal conductivity of the TEMA-F longitudinal baffle plate material simplifying term used for calculation of the LMTD correction factor ms absolute viscosity of shell-side fluid at bulk average temperature (lbm/ft-hr) ms,w absolute viscosity of shell-side fluid at outside tube wall temperature (lbm/ft-hr) mt absolute viscosity of tube-side fluid at bulk average temperature (lbm/ft-hr)

Zachry Nuclear Engineering, Inc. Attachment B, Page 12 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 mt,w absolute viscosity of tube-side fluid at inside tube wall temperature (lbm/ft-hr) vs kinematic viscosity of shell-side fluid at bulk average temperature (ft2/sec) vt kinematic viscosity of tube-side fluid at bulk average temperature (ft2/sec)

SR summation of thermal resistances that separate tube-side and shell-side bulk fluid streams (hr-ft2-°F/Btu) ctl centri-angle of the baffle cut intersection with the central tube limit (degrees)

)

q ctl centri-angle of the baffle cut intersection with the central tube limit (radians) ds centri-angle of the baffle cut intersection with the inside shell wall (degrees)

)

q ds centri-angle of the baffle cut intersection with the inside shell wall (radians) tp Tube pitch angle (degrees) qx sensitivity coefficient for parameter x relating a change in test result to a change in test parameter x in uncertainty perturbation analysis rs density of shell-side fluid at a specified temperature (lbm/ft3) rt density of tube-side fluid at a specified temperature (lbm/ft3)

Subscript Terms av bulk average condition s b baffle-related bb between baffles c cold stream fi fouling (inside surface)

Zachry Nuclear Engineering, Inc. Attachment B, Page 13 of 14

PROTO-HX Shell and Tube Module User Documentation, Version 5.00 fo fouling (outside surface) h hot stream i inlet conditions min minimum o outlet conditions off outside film factor s shell-side parameter t tube-side parameter w tube wall conditions wc water circuits x parameter designation in uncertainty analysis numerical perturbation Superscript Terms

• designates the calculation result at extrapolation conditions a simplifying term used for calculation of the LMTD correction factor for TEMA-G shell b simplifying term used for calculation of the LMTD correction factor for TEMA-G shell Zachry Nuclear Engineering, Inc. Attachment B, Page 14 of 14

ZACHRY NUCLEAR, INC.

ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZNI Document Type: QAPD Attachment B Evaluation Review and Verification Information Attachment B, Page 1 of 2 Revision 1

ZACHRY NUCLEAR, INC.

ZACHRV ENGINEERING EVALUATION 16-E04 DESCRIPTION OF BFN RHR HEAT EXCHANGER TEST DATA EVALUATION ZN/DocumentType:QAPD ENGINEERING EVALUATION VERIFICATION FORM

1. VERIFICATION METHOD: 2. EXTENT OF VERIFICATION:

Yes No Complete Evaluation {including attachments) has been reviewed to A. Approach Checked [8'.J determine impact of revision on un-revised areas.

A. IDV of Complete Evaluation {including attachments) D B. Logic Checked [8'.J

c. Arithmetic Checked [8'.J B. IDV of revised areas of Evaluation only. [8'.J D*

D. Alternate Method D* [8'.J c. Other {describe below): D

{Attach documentation or forward to QA)

E. Other D* [8'.J

• Describe below.

3. DOCUMENTATION OF VERIFICATION:

The approach, logic, and methodology of the evaluation are acceptable. The requirements of Paragraph 7.7 of N0302 (as applicable) have been met. The overall evaluation is found to be valid and conclusions to be correct and reasonable.

IDV Signature:~~'--___,.,......,.._ _ --+- Printed Name: _J_e_ff_L_u_n_d._y_ _ _ _ _ Date: ~fa 1pf Item Comment Resolution

1. Section 4.6: Equation 170 of the UD describes Clarified the discussion concerning the the variance in fouling resistance as the film analytical uncertainty in Section 4.6.

coefficients are adjusted according to their uncertainties.

2. Section 4.6: The overall heat transfer Clarified the discussion concerning the coefficient is held constant for the analytical analytical uncertainty in Section 4.6.

uncertainty runs because it is a function of the measured parameters only which are held constant for these runs. Add this to the discussion of analytical uncertainty.

3. Editorial comments provided in a hard copy Comments incorporated.

mark-up.

Attachment B, Page 2 of 2 Revision 1 Form: N0302F03 Revision: 00-00 Date: 11-2-2011 Page 1 of 1

ENCLOSURE 9 GE Hitachi Nuclear Energy Affidavit for NEDC-33860P, Revision 0

NEDC-33860P Revision 0 GEH Proprietary Information - Class II (Internal)

GE-Hitachi Nuclear Energy Americas LLC AFFIDAVIT I, James F. Harrison, state as follows:

(1) I am Vice President, Fuel Licensing, Regulatory Affairs, GE-Hitachi Nuclear Energy Americas LLC (GEH), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in GEH proprietary report, NEDC-33860P, Safety Analysis Report for Browns Ferry Nuclear Plant Units 1, 2, and 3 Extended Power Uprate, Revision 0, dated September 2015. GEH proprietary information within text is identified by a dotted underline within double square brackets. ((This sentence is an example.{3})) Figures and large objects containing GEH proprietary information are identified with double square brackets before and after the object. In all cases, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GEH relies upon the exemption from disclosure set forth in the Freedom of Information Act (FOIA), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for trade secrets (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of trade secret, within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2.d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2.d 1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GEH's competitors without license from GEH constitutes a competitive economic advantage over other companies;

b. Information which, if used 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 of a similar product;

c. Information which reveals aspects of past, present, or future GEH customer-funded development plans and programs, resulting in potential products to GEH;

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

NEDC-33860P Revision 0 Affidavit Page 1 of 3 1

NEDC-33860P Revision 0 GEH Proprietary Information - Class II (Internal)

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.

(5) To address 10 CFR 2.390(b)(4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GEH, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GEH, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties, including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GEH.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist, or other equivalent authority for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GEH are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2), above, is classified as proprietary because it contains detailed results and conclusions regarding supporting evaluations of the safety-significant changes necessary to demonstrate the regulatory acceptability of the analysis for a GEH Boiling Water Reactor (BWR). The analysis utilized analytical models and methods, including computer codes, which GEH has developed, obtained NRC approval of, and applied to perform evaluations of Power Uprates for a GEH BWR. The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset.

The development of the evaluation process along with the interpretation and application of the analytical results is derived from the extensive experience database that constitutes a major GEH asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GEH's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GEH's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost. The value of the technology base goes beyond the extensive physical database and analytical NEDC-33860P Revision 0 Affidavit Page 2 of 3

NEDC-33860P Revision 0 GEH Proprietary Information - Class II (Internal) methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

The research, development, engineering, analytical and NRC review costs comprise a substantial investment of time and money by GEH.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GEH's competitive advantage will be lost if its competitors are able to use the results of the GEH experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GEH would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GEH of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on this 17th day of September 2015.

James F. Harrison Vice President, Fuel Licensing Regulatory Affairs GE-Hitachi Nuclear Energy Americas LLC 3901 Castle Hayne Road Wilmington, NC 28401 James.Harrison@ge.com NEDC-33860P Revision 0 Affidavit Page 3 of 3

ENCLOSURE 10 Zachry Nuclear, Inc. Affidavit for Engineering Evaluation 16-E04, Revision 1

AFFIDAVIT STATE OF CONNECTICUT SS.

COUNTY OF NEW LONDON

1. My name is Christopher D'Angelo. I am Senior Director, Analysis Division, for Zachry Nuclear Engineering, Inc. and as such I am authorized to execute this Affidavit.

2. I am familiar with the criteria applied by Zachry Nuclear Engineering to determine whether certain Zachry Nuclear Engineering information is proprietary. I am familiar with the policies established by Zachry Nuclear Engineering to ensure the proper application of these criteria.

3. I am familiar with the Zachry Nuclear Engineering information contained in Engineering Evaluation 16-E04, Revision 1, "Description of BFN RHR Heat Exchanger Test Data Evaluation," dated June 2016 and referred to herein as "Document." Information contained in this Document has been classified by Zachry Nuclear Engineering as proprietary in accordance with the policies established by Zachry Nuclear Engineering for the control and protection of proprietary and confidential information.

4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by Zachry Nuclear Engineering and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. The following criteria are customarily applied by Zachry Nuclear Engineering to determine whether information should be classified as proprietary:

(a) The information reveals details of Zachry Nuclear Engineering's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for Zachry Nuclear Engineering.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for Zachry Nuclear Engineering in product optimization or marketability.

(e) The information is vital to a competitive advantage held by Zachry Nuclear Engineering, would be helpful to competitors to Zachry Nuclear Engineering, and would likely cause substantial harm to the competitive position of Zachry Nuclear Engineering.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b), 6(c), and 6(e) above.

7. In accordance with Zachry Nuclear Engineering's policies governing the protection and control of information, proprietary information contained in this Document have been made available, on a limited basis, to others outside Zachry Nuclear Engineering only as required and under suitable agreement providing for nondisclosure and limited use of the information.

8. Zachry Nuclear Engineering policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.

9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

c~iQ ' ~

Christopher D'Angelo Senior Director, Engineering Analysis SUBSCRIBED before me this die~ day of-:'.StA--\~, 2016.

Notary Public, State of Connecticut My Commission Expires: February 28, 2017 PAMELA KAY KLARMANN Notary Public Connecticut My Commission Expires Feb 28. 2017