ML20198P616

Forwards Response to RAI Re TS Change TS-386,which Increases Main Steam Safety/Relief Valve Setpoint Tolerance

ML20198P616
Person / Time
Site:	Browns Ferry
Issue date:	11/03/1997
From:	Abney T TENNESSEE VALLEY AUTHORITY
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
TAC-M97412, TAC-M97413, TAC-M97414, TVA-BFN-TS-386, NUDOCS 9711070247
Download: ML20198P616 (8)

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I Tennessee Valley Authority, Post Offee Su 2000, Decatur, Alabamo 35009-2000

. November.3, 1997 4

TVA BFN TS-386

'U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-Gentlemen:

-In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260 50-296 BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 - ,

REQUEST FOR ADDITIONAL INFORMATION REGARDING INCREASE IN MAIN STEAM SAFETY / RELIEF VALVE (S/RV) SETPOINT TOLERANCE In (TAC. NOS. M97412, M97413, AND M97414) (TS-386) }

This. letter provides supplemental information for the review of' Technical Specificaticns (TS) change TS-386 as requested by NRC in a letter dated June 26, 1997. TS-586 was submitted on December 11, 1996, andproposeschangesto)'([3.

the existina custom TS to increase the S/RV setpoint tolerance from approximately 1% to 3%. As stated in +

.the original submittal, TS-386 also provides additional

-technical' justification for the corresponding 3% S/RV i setpoint tolerance proposed in TS-362. TS-362 was submitted-on_ September 6, 1996, and is the proposed conversion package.to Improved Standard Technical Specifications format.

'Re'sponses.to_the five NRC requests are provided in the

-Enclosure. There are no commitments in this letter.

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-U.S. Nuclear Regulatory Commission Page 2 November 3, 1997 If you have further questions on this TS change, please contact me at (205) 729-2636.

Sincerely, l Qj' Uw T. E.

ney-Manager o Licensi g and Industry. ffairs Enclosure cc (Enclosure):

Mr. Mark S. Lesser, Branch Chief U.S. Nuclear Regulatory Commission Begion II Atlanta Federal Center 601 Forsyth St., Suite 23T85 Atlanta, Georgia 30303 NRC Resident Inspector Browns Ferry Nuclear Plant 10833 Shaw Road Athens, Alabama 35611 Mr. Joseph F. Williams, Project Manager U.S. Nuclear Regulacory Commission One White Flint, North 11555 Rockzille Pike Rockville, Maryland 20852

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• _ ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY: NUCLEAR. PLANT-(BFN)-

-UNITS-1,.2,-AND 3 l

TECHNICAL SPECIFICATIONS (TS) CHANGE TS-386 INCREASED -SETPOINT. TOLERANCE MAIN STEAM SAFETY / RELIEF VALVES y (S/RVs)

RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION DATED JUNE 26, 1997 Below are-responses'to the five NRC questions provided in the subject request-for additianal information on TS-386. TS-386 was.uuhmitted on December 11,_1996, and proposes changes to the-existing.TS te support increasing the main steam S/RV setpoint tolerance from approximately 1% to 3%. A detailed engineering evaluation was provided in TS-386 and the effects of an increased S/RV setpoint tolerance have been evaluated in 4 the cycle-specific reload-analyses for Units 2 and 3 as referenced in the TVA response to question 5.a.

LTS-386 also provided added technical justification for the

, corresponding. 3% S/RV.setpoint-tolerance proposed in TS-362.

TS-362 wan submitted on September 6, 1996, and is the proposed conversion package to Improved Standard Technical Specifications format. The TVA responses provided below are likewise applicable to TS-362.

NRC Question 1-Enclosure 5, page 13 of the December 11, 1996 submittal states that the hydrodynamic main steam safety / relief valves (S/RVs) '

loads resulting from the-3% set;sint tolerance are less than 1%

higher than the " loads used in design". However, it is the NRC staff's experience via review cf other analyses, such as those Esupporting power;uprate at several boiling water reactors that-theLloads vary almost linearly with respect to the opening.

pressure of the S/RVs. Therefore, it is expected that an

' increase in the S/RV setpoint to 3% would result in a 3%

increasefin1the S/RV hydrodynamic loads, ai 'What S/RV setpoints and setpcint tolerances were cssumed forEdeveloping the above " loads used in design"?

b. Describe the methodology for determining the S/RV

. hydrodynamic. loads and the analysis results which indicate

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only a 1% load increase for the 3% setpoint tolerance. It is also stated that the resulting 1% increase in loads is negligible when combined with other loads; therefore, the effect of load increases on the plant structures were not evaluated.

c. What is the available margin such that these lead increases are negligible?

,TVA Response Below are the responses to questions 1.a, 1.b, and 1.c. It should be noted that the revised calculations were performed at

+ 3% of the highest nominal S/RV setpoint. Previous analyses were performed at + 1% of the highest nominal S/RV setpoint.

Therefore, the new setpoint evaluation represents a net increase of 2% in analyzed S/RV pressure setpoints rather than the 3% cited by the staff in question 1.

a. The S/RV " loads used in design" were those described in the BFN Long Term Torus Integrity Program (LTTIP) Plant Unique Analysis Report (PUAR) und considered in the associated NRC Safety' Evaluation Report (SER) [BFN FSAR Appendix C References 12 and 22]. The S/RV setpoints and secpoint tolerances used in the LTTIP analysis and design of modifications were 1125 psig at a + 3% tolerance for all valves (i.e., S/RV opening pressure of 1136 psig) which translates to a S/RV mass f]ow rate of 308 pounds mass per ,

second through each S/RV (Rererence PUAR Appendix H, page H-BNL-16-4R2]. There are 13 S/RVs on each BFN unit. Of the 13 S/RVs, 4 are set at 1105 psig: 4 are set at 1115 psjg; and 5 are set at 1125 psig in accordance with TS 2.2.A.

Therefore, the LTTIP analysis was conservative since the highest S/RV setpoint and tolerance was used to establish the flowcate for all 13 valves.

b. The LTTIP analyses for S/RV nydrodynamic loads were based on loads defined in accordance with the general design criteria in PUAR Section 4.2.2. Of the controlling load combinations, including Loss of Coolant (LOCA), thermal, earthouake, deadweight, and pressure, the S/RV blowdown j loads represent only about 35% of the total load combinatdan. Thus, a 2% increase in S/RV blowdown loads can be expected to increase-the total loading by less than 1% on a first estimate basis.

Additionally, an in-plant S/RV confirmatory test was conducted for BEN Unit 2 after installation of LTTIP E-2

modifications, as described in Section 4.2.2.1, Section 4.6,

- and Appendix C of the PUAR. This test showed that significant margin existed between the analytically predicted (calculated) stresses and actual stresses. S/RV load reduction factors were conservatively justified and applied for the torus support system and the vent system downcomers. These load reduction factors were approved in the LTTIP SER.

To analytically account for the change in S/RV setpoint tolerance from 1% to 3%, the opening pressures and S/RV mass flow rates were linearly extrapolated by the ratio 1.03/1.01. This corresponds to an opening pressure of 1159 psig and a mass flow rate of 314 pounds mass per second.

The S/RV confirmatory test results were also linearly extrapolated based on the ratio of the test conditions (opening pressure ot 1000 psig and mass flow rate of 268 y pounds mass per second) to the + 3% setpoint tolerance conditions (1159 psig and 314 pounds mass per second).

Other differences in the tested conditions and the + 3%

setpoint tolerance conditions were also conservatively considered in the extrapolation methudology. These included differences in torus water level and drywell-to-wetwell differential pressure.

Margins betwcen analytically predicted S/RV loads and the extrapolated S/RV test results were reduced by the TVA evaluation, as permitted by the LTTIP PUAR and SER. As noted above, the S/RV loading is only one factor in the controlling equations for stress, so the increase has a proportionally small overall effect. As expected, TVA's evaluation concluded that the overall combined loads on torus structures, systems, and components increased by less than 1% due to the S/RV setpoint tolerance change.

c. The statement in the December 1996 submittal was: "When S/RV loads are combined with other design basis loads including dead weight, thermal, LOCA, and earthquake, the total increase is negligible and the affected components meet design basis requirements."

The key load . , combinations are associated with the original LOCA and S/RV hydrodynamic loads analysis, and the in-plant S/RV confirmatory test which demonstrated significantly lower stresses than analytically predicted.

The TVA evaluation determined that the increase in overall loads resulting from the small increase in S/RV flow rate (314 versus 308 pounds mass per second) associated with the E-3

i setpoint tolerance increase js very small when combined with the other torus loads. The resulting increase in loads was evaluated against existing design basis allowables defined in the PUAR, which includes all LTTIP load combinations. It was concluded that the small increase in S/RV loads yields very small combined stress increases. The combined stresses remain within existing design basis stress allowables established by the LTTIP PUAR and SER.

NRC Question,2 Enclosure 5, page 13 of the December 11, 1996 submittal states that the maximum differential due to the ir. creased S/RV setpoint was evaluated for valves in the Generic Letter 89-10 program. Please verify that that the calculations which demonstrate the capability of these valves to function adequately have been revised for these differential pressures.

TVA Response As discussed in the TS-386 submittal, the Generic Letter (GL) 89-10 motor operated valves (MOVs ) affected by the proposed new S/RV tolerance were evaluated for increased differential pressure loads and found to be acceptable. Subsequently, on October 1, 1997, TVA submitted a TS amendment request (Power Uprate Project) to increase the authorized reactor thermal power by 5%. Since this Power Uprate Project will require a general revision of many of the calculations for GL 89-10 valves, including those listed in TS-386, the master MOV calculations are not being updated to account for the 3% S/RV setpoint tolerance at this time. The Power Uprate Project is in progress and is scheduled to be first implemented in Fall 1998 on Unit 3.

In the interim, the engineering evaluation performed for the proposed 31 S/RV TS setpoint tolerance change has been linked to the subject MOV calculations uti]izing TVA's configuration control system. This ensures that Engineering personnel are cognizant of the 3% S/RV tolerance evaluation pending final MOV calculation revisions under the Power Uprate Project.

NRC Question 3 What is the basis for the high-pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) design requirement to provide design flow to the reactor ressel at pressures up to 1120 psig? In other words, how was the 1120 psig value originally determined?

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TVA Response The 1120 psig value is documented in the vendor (GE) design

-specifications for the HPCI and RCIC pumps, and is shown on the original system process diagrams as the maximum reactor pressure for HPCI/RCIC operating modes. A review of the original design and safety analysis documentation did not identify a specific system operating limitation associated with the 1120 psig. Rather, the 1120 psig value appears to have been selected as a conservative higher pressure bound for system design considerations.

NRC Question 4 The engineering analysis describes the calculations performed to determine the required additional flow, power, and rpm needed for HPCI and RCIC.

a. Do the HPCI and RCIC turbines have a maximum allowable pressure?

b. If so, what is the value and is the value exceeded due to increased S/RV tolerance?

TVA Response

a. Yes, the HPCI and RCIC system turbines have a maximum allowable pressure.

b. The design pressure for both the HPCI and RCIC inlet nozzle and steam chest is 1250 psig which is not exceeded by the increased S.'RV tolerance.

NRC Ouestion 5 TVA's submittal includes plant-specific analyses supporting the increased S/RV setpoint tolerance for Unit 2. The cover letter states that analyses supporting the increased S/RV setpoint tolerance is being performed for the next Unit 3 reload cycle.

a. Verify that the necessary plant-specific analyses have been completed for the current Unit 3 operating cycle.

b. Confirm that any differences between Units 2 and Unit 3 that exist in approved alternative operating modes (e.g.,

increased core flow, extended load line limit, or final feedwater reduction) have been addressed for Unit 3.

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ac . Describe how TVA will ensure that the increased setpoint tolerance will be reflected in future operating cycles-for each unit?

TVA Response

a. The unit-specific analysis-for the current Unit 3 operating cycle has been completed and a copy of the report is included in Appendix N of the Updated Final Safety Analysis Report. The Unit 2 reload analysis is also in Appendix N for the cycle just completed. General Electric performs the cycle-specific analyses for BFN.

b. A listing of the operating flexibility options in effect for Units 2 and 3 are listed in Section 8 of the reload analyses referenced in item 5.a above. The selected operating flexibility options may and often do vary from cycle to cycle for either unit. As a result, it can be expected that the reload analyses for the two units may have differences in the operating flexibility options.

Since the reload analysas are unit-specific and cycle-specific, differences between the two units are immaterial.

Both the Unit 2 and the Unit 3 analyses were performed using a 3% S/RV tolerance with the selected flexibility options. The results of these analyses are shown in the above referenced reload reports,

c. An input to the cycle-specific reload analyses are system setpoints as defined in the TS and TS be.ses. If the requested TS is approved by NRC, the 3% S/RV setpoint will become the minimum input value used for the cycle-specific analyses for transients affected by S/RV setpoints. In addition, prior to performing the cycle-specific analyses, TVA and General Electric review the analysis assumptions to verify the suitability of the ar:_ ysis inputs.

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