CNL-15-085, Response to NRC Request for Additional Information Regarding Proposed Technical Specification Change to Modify Technical Proposed Technical Specification Change to Modify Technical

Response to NRC Request for Additional Information Regarding Proposed Technical Specification Change to Modify Technical Proposed Technical Specification Change to Modify Technical

ML15156A563
Person / Time
Site:	Browns Ferry
Issue date:	06/03/2015
From:	James Shea Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML15156A562	List: CNL-15-085, Response to NRC Request for Additional Information Regarding Proposed Technical Specification Change to Modify Technical Proposed Technical Specification Change to Modify Technical
References
BFN TS-492, CNL-15-085 ANP-3408NP, Rev. 0
Download: ML15156A563 (25)

Text

Proprietary Information Withhold Under 10 CFR 2.390(d)(1) This letter is decontrolled when separated from Enclosure 1 Tennessee Valley Authority, 1101 Market Street, Chattanooga, Tennessee 37402 CNL-15-085

June 3, 2015 10 CFR 50.4 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:

Response to NRC Request for Additional Information Regarding Proposed Technical Specification Change to Modify Technical Specification 2.1.1, Reactor Core Safety Limits (BFN TS-492)

References:

1. Letter from TVA to NRC, "Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3 - Application to Modify Technical Specification 2.1.1, Reactor Core Safety Limits (BFN TS-492)," dated December 11, 2014 (ADAMS Accession No. ML14363A158)

2. Letter from NRC to TVA, "Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Request for Additional Information Related to License Amendment Request for Technical Specification Changes to Reactor Core Safety Limits," dated May 12, 2015 (TAC Nos. MF5412, MF5413, and MF5414)

(ADAMS Accession No. ML15126A530)

By letter dated December 11, 2014 (Reference 1), Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, to modify Technical Specification (TS) 2.1.1, Reactor Core Safety Limits, to revise the reactor dome pressure limit.

By letter dated May 12, 2015 (Reference 2), the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAI) from the Reactor Systems Branch. The due date for the response is June 5, 2015. Enclosure 1 contains AREVA report ANP-3408P, Revision 0, that provides the responses to the Reference 2 RAI. Enclosure 1 contains information that AREVA NP considers to be proprietary in nature and subsequently, pursuant to 10 Code of Federal Regulations 2.390, "Public inspections, exemptions, requests for withholding," paragraph (a)(4), it is requested that such information be withheld from public disclosure. Enclosure 2 contains the non-proprietary version of the Enclosure 1 report with the proprietary material removed, and is suitable for public disclosure. Enclosure 3 provides the affidavit supporting this request.

ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3

ANP-3408NP Revision 0, "AREVA RAI Responses for Browns Ferry Steam Dome Pressure for Reactor Core Safety Limits" (Non-proprietary)

ANP-3408NP Revision 0 AREVA RAI Responses for Browns Ferry Steam Dome Pressure for Reactor Core Safety Limits

May 2015 © 2015 AREVA Inc.

ANP-3408NP Revision 0 Copyright © 2015 AREVA Inc. All Rights Reserved

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page i Nature of Changes Item Section(s) or Page(s) Description and Justification 1 All Initial Issue

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page ii Contents Page

1.0 INTRODUCTION

............................................................................................... 1-1 2.0 NRC QUESTIONS AND AREVA RESPONSE

.................................................. 2-1

3.0 REFERENCES

.................................................................................................. 3-1

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page iii Nomenclature Acronym Definition BFN Browns Ferry Nuclear Plant BOC Beginning-of-cycle COLR Core Operating Limits Report EOFP End of Full Power FHOOS Feedwater Heaters Out

-of-service HGAP Pellet-to-Cladding Gap Coefficient LAR Licensing Amendment Request LPIS Low Pressure Isolation Setpoint MCPR Minimum Critical Power Ratio MOC Middle-of-cycle MSIV Main Steam Isolation Valve NRC Nuclear Regulatory Commission NSS Nominal Scram Speed OSS Optimum Scram Speed PRFO Pressure Regulator Failure Open RAI Request for Additional Information RTP Rated Thermal Power TS Technical Specification TSSS Technical Specification Scram Speed TVA Tennessee Valley Authority

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 1-1

1.0 INTRODUCTION

Tennessee Valley Authority (TVA) submitted a License Amendment Request (LAR) to change the Browns Ferry (BFN) Technical Specifications (TS) in support of steam dome pressure for reactor core safety limits. In response to the LAR, the US Nuclear Regulatory Commission (NRC) has issued an initial set of questions, in the form of Request for Additional Information (RAI), Reference 1.

Based on the information provided in this report, TVA will prepare a formal response to the NRC RAIs.

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 2-1 2.0 NRC QUESTIONS AND AREVA RESPONSE The NRC questions (i.e., RAIs) listed below are according to Reference 1: RAI-01: The LAR claims the GE14 fuel in the BFN Unit 1 is third cycle fuel, with large MCPR margins due to the depleted state of the fuel and the lower power locations of those bundles. Provide the normalized bundle power (ratio of bundle power to core

-averaged bundle power) at the beginning of the current cycle for: 1) the GE14 fuel with the highest bundle power and 2) the highest powered bundle in the core.

RAI-02: AREVA report ANP

-3245P Revision 1 (Attachment 5 to the LAR) presented an analysis of the Pressure Regulator Failure Open (PRFO) event in the BFN units. The analysis included sensitivity studies of the effect of key parameters that affect the minimum reactor steam dome pressure obtained during the PRFO event. The lowest steam dome pressure while the reactor power is still above 25% rated thermal power (RTP) is the relevant pressure to use in applying TS safety limits 2.1.1.1 and 2.1.1.2.

For the PRFO event represented in Tables 3.1 through 3.6 of ANP

-3245P Revision 1, clarify the occurrence of the minimum steam dome pressure with AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 2-2 respect to the full closure of the MSIV. Indicate the status of the MSIV, partially or fully closed, when the minimum steam dome pressure occurred.ANP-3245P Tabl e Sensitivity Parameter MSIV Position Percentage Open When Minimum Steam Dome Pressure is Reached (Unit 1)

Table 3.1 State Point 100 P / 105 F [ 100P / 81 F 65P / 110 F 65P / 40 F ] Table 3.2 Initial Conditions Nominal Temperature

, Increased Pressure

[ Nominal Temperature

, Reduced Pressure Reduced Temperature

, Increased Pressure Reduced Temperature

, Reduced Pressure FHOOS Temperature

] Table 3.3 MSIV Closure 3-second closure

[ 4-second closure 5-second closure

] Table 3.4 Cycle Exposure BOC [ MOC Licensing EOFP Coastdown ] Table 3.5 Scram Time TSSS [ NSS OSS OSS reduced by 10%

] Table 3.6 HGAP condition Nominal HGAP

[ HGAP +20% HGAP -20% ]

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 2-3 b) Table 3.7 of ANP

-3245P Revision 1 shows the minimum steam dome pressure for different initial state points (reactor power and core flow). Each row in the table is for a different combination of state points. Clarify the distinction between the pressures in the table with and without an asterisk. For the higher core flow cases (above 35% of rated core flow), indicate if the reactor thermal power is above or below 25% RTP when the minimum steam dome pressure occurred.

RAI-03: TS 2.1.1.2 specifies the safety limit (SL) on the minimum critical power ratio (MCPR). The proposed change in TS 2.1.1.2 expands the range of applicability of the SL on the MCPR to a lower pressure. The LAR requires extending the applicability of the SPCB/GE14 critical power correlation down to pressures as low as 585 psig. Explain the consistency of the approach discussed in ANP

-3245P Revision 1 (Attachment 5 to the LAR), for determining the critical power for GE14 fuel at pressures below 685 psig, with the NRC-approved AREVA methodology for applying AREVA critical power correlations to co

-resident fuel (as identified in the Core Operating Limits Report (COLR)).

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 2-4

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 2-5 RAI-04: In a PRFO event, the core inlet subcooling will decrease as the water saturation temperature decreases in response to the declining system pressure. Figure 4.1 of ANP-3245P Revision 1 shows that a lower inlet subcooling will reduce the critical heat flux. The SPCB critical power correlation also predicts a lower critical power for a lower inlet subcooling, as indicated in Figures 2.8 and 2.9 of the AREVA topical report EMF

-2209(p), Revision 3 (SPCB Critical Power Correlation, December 2009).The last paragraph on page 4

-2 of ANP-3245P Revision 1 (Attachment 5 to the LAR) states, "

." a) Explain how the varying inlet subcooling condition during a PRFO transient is accounted for in the application of the SPCB/GE14 correlation for pressures below 700 psia.

b) Explain in more detail the meaning of "preserving the same inlet subcooling."

Does it mean the actual inlet subcooling will be used (accounting for the effect of lower pressure) but the dome pressure will be assumed to stay at 700 psia?

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 2-6

RAI-05: ANP-3245P Revision 1 (Attachment 5 to the LAR) provides results for a series of sensitivity calculations in Tables 3.1 through 3.7. However the initial conditions are not stated for each series. For each series (i.e. Tables 3.1 through 3.7) provide the initial conditions for cycle exposure, core power, core flow, steam dome pressure, feedwater temperature, MSIV closure time, scram insertion speed, and core average gap conductance.

State point Cycle exposure HGAP (Btu/hr-°F-ft 2) Initial Conditions Dome Pressure /

Feedwater Temperature (psia / °F)

MSIV Closure Scram Speed Parameter evaluated BOC [ ] 100P - 1060 psia / 382.0°F 65P - 1031.40 psia / 342.6°F 3 sec TSSS AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 2-7 State point Cycle exposure HGAP (Btu/hr-°F-ft 2) Initial Conditions Dome Pressure /

Feedwater Temperature (psia / °F)

MSIV Closure Scram Speed 100 P / 81F BOC [ ] Parameter evaluated Nominal Temp, Increased Pressure

- 1060 psia / 382.0°F Nominal Temp, Reduced Pressure

- 1040 psia / 382.0°F Reduced Temp, Increased Pressure

- 1060 psia / 372.0°F Reduced Temp, Reduced Pressure

- 1040 psia / 372.0°F FHOOS Temperature

- 1030 psia / 317.0°F 3 sec TSSS State point Cycle exposure HGAP (Btu/hr-°F-ft 2) Initial Conditions Dome Pressure /

Feedwater Temperature (psia / °F)

MSIV Closure Scram Speed 100 P / 81F BOC [ ] FHOOS Temperature

- 1030 psia / 317.0°F Parameter evaluated OSS reduced by 10% State point Cycle exposure HGAP (Btu/hr-°F-ft 2) Initial Conditions Dome Pressure /

Feedwater Temperature (psia / °F)

MSIV Closure Scram Speed 100 P / 81F Parameter evaluated [ ] FHOOS Temperature

- 1030 psia / 317.0°F 5 sec OSS reduced by

10%

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 2-8 State point Cycle exposure HGAP (Btu/hr-°F-ft 2) Initial Conditions Dome Pressure /

Feedwater Temperature (psia / °F)

MSIV Closure Scram Speed 100 P / 81F BOC [ ] FHOOS Temperature

- 1030 psia / 317.0°F 3 sec Parameter evaluated State point Cycle exposure HGAP (Btu/hr-°F-ft 2) Initial Conditions Dome Pressure /

Feedwater Temperature (psia / °F)

MSIV Closure Scram Speed 100 P / 81F BOC Parameter evaluated

[ ] FHOOS Temperature

- 1030 psia / 317.0°F 5 sec OSS reduced by

10%

State point Cycle exposure HGAP (Btu/hr-°F-ft 2) Initial Conditions Dome Pressure /

Feedwater Temperature (psia / °F)

MSIV Closure Scram Speed 100P / 81F 90P / 70F 75P / 50F 65P / 40F 60P / 35F 50P / 35F 40P / 35F 30P / 35F BOC HGAP +20% [

] FHOOS Temperature

- 1030 psia / 317.0°F 5 sec OSS reduced by 10%

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure for Reactor Core Safety Limits Page 2-9 RAI-06: Explain why the pressures in Tables 3.4 and 3.6 of ANP-3245P Revision 1 (Attachment 5 to the LAR) are significantly lower than values in Tables 3.1, 3.2, 3.3 and 3.5.

RAI-07: Section 3.1.6 of ANP-3245P Revision 1 (A ttachment 5 to the LAR) discusses the sensitivity of the minimum steam dome pressure to the core average gap conductance (HGAP). Under steady-state conditions for a given power, the averaged fuel temperature will vary inversely with the HGAP while the fuel cladding surface temperatur es will not be a ffected. Thus the amount of heat transferred from the fuel to the coolant remains the same under steady-state conditions regardless of the value of the HGAP. A statement in the first paragraph of Sectio n 3.1.6 says, "." a) Explain if the statement is referring to steady-state or transient conditions and provide results from the analysis to substantiate the claim that a higher HGAP will result in more heat being transferred into the coolant.

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure for Reactor Core Safety Limits Page 2-10

b) Explain the impact of the HGAP on the timing of the turbine header pressure reaching the low-pressure isolation setpoint (LPIS).

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 2-11 [ ]

AREVA Inc.

ANP-3408NP AREVA RAI Responses for Browns Revision 0 Ferry Steam Dome Pressure f or Reactor Core Safety Limits Page 3-1

3.0 REFERENCES

1. Letter, F. E. Saba (NRC) to J. W. Shea (TVA), "Browns Ferry Nuclear Plant, Units 1, 2, and 3 - Request for Additional Information Related to License Amendment Request for Technical Specification Changes to Reactor Core Safety Limits (TAC Nos. MF5412, MF5413, and MF5414)," USNRC, May 12 , 201 5. (38-9240539-00 0)

ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 1, 2, AND 3

Affidavit for ANP-3408P Revision 0