ML083530619

Technical Specifications Change TS-431 - Extended Power Uprate - Response to Round 22 Request for Additional Information - Unit 1 Large Transient Testing (Ltt) -

ML083530619
Person / Time
Site:	Browns Ferry
Issue date:	12/15/2008
From:	Godwin F Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MD5262, TVA-BFN-TS-431
Download: ML083530619 (7)

Text

Tennessee Valley Authority, Post Office Box 2000, Decatur, Alabama 35609-2000 December 15, 2008 TVA-BFN-TS-431 10 CFR 50.90 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Mail Stop OWFN, P1-35 Washington, D. C. 20555-0001 In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority )

BROWNS FERRY NUCLEAR PLANT (BFN) - UNIT 1 - TECHNICAL SPECIFICATIONS (TS) CHANGE TS-431 - EXTENDED POWER UPRATE (EPU) - RESPONSE TO ROUND 22 REQUEST FOR ADDITIONAL INFORMATION (RAI) - UNIT 1 LARGE TRANSIENT TESTING (LTT-) - (TAC NO. MD5262)

By letter dated June 28, 2004 (ADAMS Accession Nos. ML041840109), TVA submitted a license amendment application to NRC for EPU operation of BFN Unit 1. The proposed amendment would change the operating license to increase the maximum authorized core thermal power level of Unit 1 by approximately 14 percent to 3952 megawatts.

In a submittal dated October 22, 2007 (ML072960311), TVA provided supplemental justification for not performing main turbine trip/generator load reject LTT on Units 2 and 3 at EPU conditions. However, at the time, due to the limited operating experience on Unit 1 following its restart in May, 2007 from the extended outage, a regulatory commitment was provided to perform a turbine trip/generator load reject large transient test from EPU rated power conditions within 60 days of operation at EPU power on Unit 1. Subsequently, on October 3, 2008, TVA made an additional submittal (ML072960311) that summarized the Unit 1 restart testing and operating experience on Unit 1 Cycle 7, and requested NRC concurrence on withdrawing the Unit 1 LTT commitment. On November 12, 2008, NRC sent a Round 22 RAI which included RAI questions (SRXB-75 and SRXB-76) on TVA's justification to not perform a main turbine trip/generator load reject transient test on Unit 1.

A response to these two RAI questions is provided in the enclosure to this letter.

I U.S. Nuclear Regulatory Commission Page 2 December 15, 2008 TVA has determined that the additional information provided by this letter does not affect the no significant hazards considerations associated with the proposed TS change. The proposed TS change still qualifies for a categorical exclusion from environmental review pursuant to the provisions of 10 CFR 51.22(c)(9).

No new regulatory commitments are made in this submittal. If you have any questions regarding this letter, please contact me at (256)729-2636.

I declare under penalty of perjury that the foregoing is true and correct. Executed on this 15 th day of December, 2008.

Sincerely, F. R. Godwin Manager of Licensing and Industry Affairs

Enclosure:

Response to Round 22 Request For Additional Information (RAI) - Unit 1 Large Transient Testing (LTT)

U.S. Nuclear Regulatory Commission Page 3 December 15, 2008 Enclosure cc (Enclosure):

State Health Officer Alabama State Department of Public Health RSA Tower - Administration Suite 1552 P.O. Box 303017 Montgomery, Alabama 36130-3017 Ms. Eva Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)

One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739 Eugene F. Guthrie, Branch Chief U.S. Nuclear Regulatory Commission Region II Sam Nunn Atlanta Federal Center 61 Forsyth Street, SW, Suite 23T85 Atlanta, Georgia 30303-8931 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611-6970

ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 1 TECHNICAL SPECIFICATIONS (TS) CHANGE TS-431 EXTENDED POWER UPRATE (EPU)

RESPONSE TO ROUND 22 REQUEST FOR ADDITIONAL INFORMATION (RAI) - UNIT 1 LARGE TRANSIENT TESTING (LTT)

NRC RAI SRXB-75 (Unit 1 only)

Provide a (using tables and/or figures) comparing the values of key plant parameters such as reactor pressure and reactor water level observed during the October 12, 2007, main turbine trip event from approximately 100 percent (CLTP) at 3458 megawatts thermal (MWt) with the results of the following:

a. An ODYN calculation performed at CLTP for the October 12, 2007, main turbine trip event;

b. An ODYN calculation at extended power uprate (EPU) power (3952 MWt) for a main turbine trip at a comparable core exposure as the October 12, 2007, event; and,

c. The main turbine trip event of June 9, 2007, at 80-percent CLTP (2761 MWt), and the corresponding ODYN calculation performed for that event.

In addition, provide the calculated maximum delta critical power ratio for a), b) and c).

This study should show the degree of conservatism between the maximum reactor peak pressure observed during the actual events (June 9 and October 12, 2007) and the calculated maximum vessel pressure for rated CLTP, 80-percent CLTP and rated EPU power using the transient analysis computer code used for pressurization events. In addition, the results will allow for a comparison between the calculated change in delta critical power ratio between a simulation of the actual events at CLTP (calculation a), at 80-percent CLTP (calculation c) and at EPU (calculation b).

TVA Response to SRXB-75 The requested information for the October 12, 2007, main turbine trip from approximately 100%

Current Licensed Thermal Power (CLTP) is provided in the following table. A transient plot of the October 17, 2007, trip generated from the plant process computer showing several key plant parameters is provided in the response to SRXB-76. Also included in the table are the results of a turbine trip simulation for the October 12, 2007, CLTP transient (calculation a) using the ODYN code and with the power stepped up to the EPU power (calculation b). The requested information for the June 9, 2007, main turbine trip from approximately 80% CLTP (calculation c) is also provided in the table. As expected, the ODYN cases are conservative with respect to the actual plant peak pressures and the calculated peak pressures increase with higher initial power. Delta Critical Power Ratio likewise increases with higher initial power. Reactor pressure is the chief output of interest for ODYN simulations and less agreement is seen in the table between the calculated and observed short-term minimum water levels. In part, this is due to differences in the computer modeling and the physical measurement of the level. Also, in the ODYN simulations for 100% CLTP and EPU, Recirculation Pump Trip (RPT) occurred whereas for the plant event, the pressure did not reach the high pressure RPT setpoint. RPT initiation will result in a higher water level than a comparable no RPT case.

Comparison of Plant EventSand ODYN Simulation Cases Plant Parameter 6/9/07 6/9/07 10/12/07 10/12/07 10/12/07 Plant Event ODYN Case Plant Event ODYN Case ODYN Case 80% CLTP 80% CLTP 100% CLTP 100% CLTP EPU Peak Reactor Pressure (psig) 1099 1106 1130 1156 1181 Reactor Pressure Increase (psi) 79 86 94 122 148 Minimum Reactor Water Level -5 -15 -41 -27 -20 (inches from vessel zero reference)

Delta Critical Power Ratio na .01 na .02 .03 NRC RAI SRXB-76 (Unit 1 only)

Explain how past Unit 1 operating experience, including the turbine trip event from 100 percent of CLTP on October 12, 2007, provides information representative of how the unit would respond to a load rejection from 100 percent of EPU conditions. In particular, address how well the modified feedwater (FW) and condensate system would maintain reactor level consistent with the design response (i.e., avoiding unnecessary emergency core cooling system actuation on low level and high pressure coolant injection isolation on high level) considering the higher power level (and associated higher FW flow) and the different timing of signals for a main-generator-initiated load rejection as opposed to a main turbine trip. Identify any component-level testing and computer modeling of plant transient response that supports proper operation of the FW system in responding to the load rejection from the EPU power level.

TVA Response to SRXB-76 During Unit 1 Cycle 7 operation, Unit 1 experienced two turbine trips from high power levels. The first turbine trip was on June 9, 2007, from approximately 80% CLTP. TVA provided a summary of the plant performance during the event to NRC in a submittal dated July 3, 2007 (ML071870024). The second turbine trip occurred on October 12, 2007, and was from rated CLTP. The reactor scram was uncomplicated and reactor water level was recovered by the feedwater and condensate system. The Unit 1 plant responses during the two turbine trips were evaluated against previous turbine trip/load reject experience from high power on Units 2 and 3, including a recent load reject from CLTP on Unit

2. Unit 2 previously installed the same EPU feedwater and condensate system modifications as Unit

1. The transient characteristics and Unit 1 integrated system response were determined to be typical when compared with previous turbine trip/load reject events at BFN. BFN operating experience at CLTP includes both turbine trip and load reject events and there is no significant difference in integrated plant response. Turbine trips and generator load rejects are both rapid reactor pressurization transients and the timing is similar since the turbine stop valve (TSV) and turbine control valve (TCV) closure times are comparable. At BFN, turbine trip or load reject signals result in a trip of both the TSVs and TCVs.

2

The attached figure shows the reactor water level, reactor pressure, feedwater flow, and steam flow following the October 12, 2007, turbine trip. Fast closure of the TSVs and TCVs results in a short-term increase in reactor pressure and an accompanying decrease in reactor water level due to void collapse from the pressure increase. Closure of the TSVs and TCVs also results in a direct reactor scram, which shuts down the reactor and causes a rapid reduction in steam production. The turbine Electro-Hydraulic Control (EHC) system opens turbine bypass valves to take control of reactor pressure soon after the turbine trip and reduces reactor vessel pressure to a set pressure. The peak reactor pressure was 1130 psig, which is below the nominal setpoint (1135 psig) of the lowest set group of safety relief valves (SRVs) and no SRVs operated. The minimum water level remained several inches above the High Pressure Coolant Injection (HPCI) system initiation point at Reactor Vessel Level - Low Low, Level 2. The feedwater control system responded to the reduced low water level and recovered water level to normal operating ranges within about a minute after the scram and the feedpumps subsequently tripped on high water level at about 110 seconds.

With the modified feedwater/condensate system modeled on the BFN simulator, load rejects, were executed from EPU conditions, which provided insights on plant response to transients.

Rated EPU power level is approximately 14% higher than the CLTP. The primary differences between turbine trip/load reject events at EPU versus CLTP will be an increase in the short-term peak vessel pressure peak, which causes the short-term minimum reactor water level to be lower. The ODYN case results shown in SRXB-75 provide a conservative estimate of the pressure increase. A higher peak pressure increases the likelihood of SRV actuation during the initial pressurization.

However, the initial pressure transient will still be of short duration since the scram would reduce steam production to within bypass valve capacity within a few seconds and the EHC system will take control of reactor pressure very early in the transient. The short-term minimum reactor water level is expected to be lower and could encroach on the Low Low, Level 2 setpoint, which, if reached, would result in a HPCI initiation. The HPCI system would not be needed in this event since it is well within the capability of the feedwater system to restore level. Although initiation of HPCI during a turbine trip/load reject event is not operationally desirable, it does not represent a risk to plant operation. If HPCI was not secured by operators, it would automatically shut down on high water level to prevent water intrusion into the steam lines. Avoidance of HPCI starts was considered a Level 2 criterion for turbine trip/load reject transient test in the original startup test program and the possibility of HPCI start on turbine trip/load reject events is documented in the BFN corrective action program.

The favorable operating experience observed during the two turbine trips on Unit 1 during Cycle 7 operation and experience on Unit 2 during Cycle 15 with the modified feedwater/condensate system has shown that the EPU modifications are capable of satisfactorily responding to turbine trip/load reject events at CLTP. Aside from differences described above, the integrated response of the plant feedwater and reactor pressure control systems to turbine trip/load reject events is similar for both CLTP or EPU conditions and well within the capability and capacity of the modified feedwater and condensate system, and the turbine EHC system at EPU conditions. Therefore, the operating experience of the EPU modified feedwater/condensate systems is representative of the expected performance at EPU conditions and supports TVA's request to not perform a turbine trip/load reject LTT at EPU conditions in Unit 1.

3

reactor water level, reactor pressure, feedwater flow, and steam flow following the October 12, 2007, turbine trip

. DatAWare istory 12-Oct-2007 08:02:00 to 12-Oct-2007 08:14:00(...)

16+A 1200.0 55.0 -------------- -- -- - -- - -

- P T tic 1

-13. 1140.0 35.0 ...... - - --------------

i

---

----- -------I- iLI 904 1080.0 15.0 ---- ---- - -

6.4 1020.0 -5.0 ---------------

3-2 960.0 -25.0 2 - - - - - - - - - - - - - 2 - - - ---

-i - - - - - - - -

44 4 4 , 4 4 4 4 4 4 0.0 900.0 45.0 ----------------

10:-02.007 CT-: t-2t107 12-Ott-2007 I 2-Oct-2007 12-O"t-2007 I 2-lt-002007 I _-or't-20fl7 08: 02:00 C3DT 08:04 :00 CDT 08:06:00 CDT 08:08:00 CDT 08:10:00 CDT 08:12:00 CDT 08:14:00 CDT Description Low-Y Hi-V Units (1) 3-58A (B1:U1) RX LEVEL-EMERGENCY SYSTEMS RANNGE -45 55 IN (2) 3-2A-, (BI:UI) REACTOR PRESSURE A 900 1200 PSIG (3ý) (Al(: NO (BIUI) RIW

• • OW TO ESACT'OR 0 16

04) C.A.LCO41 (BI:UI) N SMI' LEAVING REACTOR 0 16 MLBHR 4