ML20216G440

Forwards Request for Addl Info Re 971001 Proposed Rev to Plants,Units 2 & 3 TSs to Permit Operation of Units at Uprated Power Level of 2458 Mwt

ML20216G440
Person / Time
Site:	Browns Ferry
Issue date:	03/13/1998
From:	De Agazio A NRC (Affiliation Not Assigned)
To:	Zeringue O TENNESSEE VALLEY AUTHORITY
References
TAC-M99711, TAC-M99712, NUDOCS 9803190412
Download: ML20216G440 (8)

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• March 13,1998 Mr. O. J. Zeringue Chief Nuclear Officer And Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3: REQUEST FOR ADDITIONAL INFORMATION RELATING TO TECHNICAL SPECIFICATION CHANGE NO. TS-384 - POWER UPRATE OPERATION (TAC NOS. M99711 AND M99712)

Dear Mr. Zeringue:

By letter dated October 1,1997, the Tennessee Valley Authority (TVA) proposed revisions to the Browns Ferry Nuclear Plant Units 2 and 3 Technical Specifications to permit operation of the units at the uprated power level of 3458 Mwt. The staff has determined thLt additional information is required to complete the review of the proposed action. A description of the information required is provided in the Enclosure. This request for additional information is in addition to that sent on February 18,1998.

Please provide the required information no later than April 15,1998.

Sincerely,

/s/

Albert W. De Agazio, Senior Pro.iect Manager Project Directorate 11-3 Division of Reactor Projects -l/Il Office of Nuclear Reactor Regulation Docket Nos. 50-259,50-260, and 50-296 Serial No. BFN-98-006

Enclosure:

Request for Additional Information i

cc w/ encl: .iee next page .

DISTRIBUTION: _ i Docket File B. Clayton G. Thomas PUBLIC A. DeAgazio OGC J. Zwolinski(A) P. Kang ACRS F. Hebdon D. Shum L. Plisco, Region 11

. DOCUMENT NAME: G:\BFN\99711#2.RAI To receive a copy of this document, indicate in the box: "C" = Copy without attachment / enclosure "E" = Copy with attachment / enclosure "N" = No copy 0FFICE PM:PDII 3 fyho/t L.V LA:PDil-3 l D:PDII-3 )) l(j l l NAME ADeAgazio M / " BClayton .dtn .(9 FHebdon W DATE 03/ # /98 03/f5 /98 03/('?/98 03/ /98 03/ /98 _

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%, * * * * * # March 13,1998 Mr. O. J. Zeringue Chief Nuclear Officer And Executive Vice President Tennessee Valley Authority 6A Lookout Place 1101 Market Street Chattanooga, Tennessee 37402-2801

SUBJECT:

BROWNS FERRY NUCLEAR Pl. ANT, UNITS 2 AND 3: REQUEST FOR ADDITIONAL INFORMATION RELATING TO TECHNICAL SPECIFICATION CHANGE NO. TS-384 - POWER UPRATE OPERATION (TAC NOS. M99711 AND M99712 )

Dear Mr. Zeringue:

By letter dated October 1,1997, the Tennessee Valley Authority (TVA) proposed revisions to the Browns Ferry Nuclear Plant Units 2 and 3 Technical Specifications to pe!mit operation of the units at the uprated power level of 3458 Mwt. The staff has determined that additional information is required to complete the review of the proposed action. A description of the information required is provided in the Enclosure. This request for additional information is in addition to that sent on February 18,1998.

Please provide the required information no later than April 15,1998.

Sincerely, asad ,

Albert W. De Agazio, Senior Project Manager Project Directorate ll-3 Division of Reactor Projects -!/II Office of Nuclear Reactor Regulation Docket Nos. 50-259, 50-260, and 50-296 Serial No. BFN-98-006

Enclosure:

Request for Additional Information cc w/end: See next page 1

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o Mr. O. J. Zeringue BROWNS FERRY NUCLEAR PLANT Tennessee Valley Authority cc:  ;

Mr. J. A. Scalice, Senior Vice President Mr. Mark J. Burzynski, Managar i Nuclear Operations Nuclear Licensing Tennessee Valley Authority Tennessee Valley Authority 6A Lookout Place 4J Blue Ridge 1101 Market Street 1101 Market Street Chattanooga, TN 37402-2801 Chattanooga, TN 37402-2801 Mr. Jack A. Bailey, Vice President Mr. Timothy E. Abney, Manager Engineering & Technical Services Licensing and Industry Affairs Tennessee Valley Authority Browns Ferry Nuclear Plant 6A Lookout Place Tennessee Valley Authority 1101 Market Street P.O. Box 2000 Chattanooga, TN 37402-2801 Decatur, AL 37402-2801 Mr. C. M. Crane, Site Vice President Regional Administrator, F ;gion il Browns Ferry Nuclear Plant U.S. Nuclear Regulatory Commission Tennessee Valley Authority 61 Forsyth Street, SW., Suite 23T85 P.O. Box 2000 Atlanta, GA 30303-3415 Decatur, AL 35609 Mr. Leonard D. Wert General Counsel Senior Resident inspector Tennessee Valley Authority U.S. Nuclear Regulatory Commission ET 10H Browns Ferry Nuclear Plant 400 West Summit Hill Drive 10833 Shaw Road Knoxville, TN 37902 Athens, AL 35611 Mr. Raul R. Baron, General Manager State Health Officer Nuclear Assurance Alabama Dept. of Public Health Tennessee Valley Authority 434 Monroe Street 4J Blue Ridge Montgomery, AL 35130-1701 1101 Market Street Chattanooga, TN 37402-2801 Chairman Limestone County Commission Mr. Karl W. Singer, Plant Manager 310 West Washington Street Browns Ferry Nuclear Plant Athens, AL 35611 Tennessee Valley Authority P.O. Box 2000 Decatur, AL 35609

REQUEST FOR ADDITIONAL INFORMATION BROWNS FERRY NUCLEAR PLANT, UNITS 2 AND 3 DOCKET NUMBERS 50-260 AND 50-296 A. SPENT FUEL STORAGE

1. Since the spent fuel pool (SFP) heat loads will increase because of plant operations at the proposed increased power level, provide the following information:

a. Provide / compare the heat loads and corresponding peak calculatad SFP ,

temperatures (for plant operations at current power level and at proposed j uprate power level) during planned refueling and unplanned full core off-load. A single failure of SFP cooling system need not be assumed for the unplanned full core off-load.

b. Is full core off-load the general practice for planned refuelings?

c. How many SFP cooling system trains will be available/ operable prior to a planned refueling outage or an unplanned full core off-load?

2. Discuss the provisions (actions) established in plant operation procedures to provide the controls necessary to ensure that the limiting condition for operation, LCO 3.10.C.2 temperature limit of 150*F will not be exceeded.

3. In the unlikely event that there is a complete loss of SFP cooling capabikty, the )

SFP water temperature will rise and eventually will reach boiling temperature.

Provide the time to boil (from the pool high temperature alarm caused by loss-of-pool cooling) and the boil-off rate (based on the heat load for the unplanned full core off-load scenario). Also, discuss sources and capacity of make-up water and the methods / systems (indicating system seismic design Category) used to )

provide the make-up water.  !

4. The environmental qualification (EQ) of mechanical equipment inside and outside containment has not been addressed. Please demor' strate that plant operations at the proposed uprated power level will have no impact on the EQ of mechanical equipment inside and outside containment.

B. ELECTRICAL POWER. AND AUXILIARY SYSTEMS

1. With the thermal power uprated from 3293 Mwt to 3458 Mwt at Browns Ferry Nuclear Plant (BFNP) Units 2 and 3, provide the net electrical power output i increase for each unit resulting from the proposed power uprate. Discuss the Enclosure

potential impact the additional heat has on the main generator and its auxiliary equipment due to power uprate. Specifically, address in this discussion, the main generator stator and rotor, exciter and voltage regulator, hydrogen cooling system, and the generator protective relays .

2. Discuss the impact that power uprate has on all levels of the station auxiliary electrical onsite distribution and offsite power system by providing bus voltages and power flow changes from before and after power uprate load flow studies.

This discussion should include a specific assessment for the main step-up transformer, startup transformer, unit auxiliary transformer, emergency diesel generators, and the iso-nhase buses.

3. State whether the added generation will affect the offsite system grid voltage profiles and system grid stability. With the added generation, does BFNP require re-analysis of Branch Technical Position PSB-1 and changes in the degraded grid setpoints? Provide summaries of the grid stability cases reviewed and attendant findings.

4. Explain why station loads under normal and emergency operational conditins are computed using equipment nameplate data except for the core spray and residual heat removal pump motors where the actual brake horsepower for the flow conditions is used.

5. As a result of lessons leamed from the Main Yankee independent Safety Assessment inspection, all licensees are required to review and evaluate whether the power uprate would alter the original licensing basis of General Design Criterion (GDC)-17 and station backout (SBO) requirements. Please I provide BFNP's assessment regarding GDC-17 and SBO requirements.

6. In the TVA submittal for Generic Letter 89-10, it is noted that BFNr2 may require replacement of some valve motor operators which may add smal' additional loads to the emergency diesel generators. BFNP has committed to verify the emergency diesel generator capacity. Provide the findings for this commitment.

C. ENVIRONMENTAL QUALIFICATION FOR SAFETY-RELATED ELECTRICAL EQUIPMENT For each component / equipment type (or one representative / bounding example of a component / equipment type) where expected environmental conditions at the uprate power level exceeds the environmental conditions tested to, provide the following:

1. A description showing the relationship between environmental conditions (i.e., temperature vs. time) tested to, the expected environmental conditions at current power levels, and the expected environmental conditions at the power uprate level from time 0 (initiation of accident) to the time the l

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component / equipment type is required to remain opeiable for post LOCA operation.

2. An evaluation demonstrating qualification for each segment of the uprate power level temperature response that is not enveloped by the environmental conditions (i.e., temperature) tested to.

3. Where (or if) margins derived through the use of the Arrhenius methodology are utilized ses part of the basis for concluding continued qualification, provide the Arrhenius calculation at the current (if applicable /available) and uprate power levels. Define the margins available for the current and uprate power levels and describe and justify the reduced margin for the uprate power level.

D. REACTOR AND REACTOR SYSTEMS

1. The standby liquid control system pump discharge pressure is increased by 50 psig (Page E1-6, Enclosure 1 to TVA letter dated October 1.1997), but the increase in reactor operating pressure due to power uprate is only 30 psig.

Allowances for system test inaccuracies were supposed to be in the original values. What is the basis for the 50 psig increase?

2. It is stated (Page E1-8, item # 5, Enclosure 1 to TVA letter dated October 1, 1997) that "Due to human factors consideration, the value of 1175 psig was chosen." The value of 1175 psig was chosen instead of 1176.5. Why wasn't the value of 1176 or 1177 chosen instead of 11757

3. Why are the maximum operating pressures for high-pressure coolant injection (HPCI) and reactor core isolation coolant increased by 54 psi and not 30 psig (Page E1-11, item # 10, Enclosure 1 to TVA letter dated October 1,1997)?

4. It is stated (NEDC-32751P, Section 1.2.1, Enclosure 5 to TVA letter dated October 1,1997) that some analyses are performed at 100% rated power.

Identify the portions of the analysis where the power levels were assumed at i 100 % power and explain why analyzing at 100 % power is acceptable for the uprated conditions.

5. It is stated (see NEDC-32751P, Section 2.4, Stability, Enclosure 5 to TVA letter dated October 1,1997) that the BFNP will rely on the revised interim corrective actions for both units until the LTS Option 111 is implemented. What is the schedule for implementing Option Ill?

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6. Confirm that credit is not taken for ;he relief flow (see NEDC-32751P, Section 3.4, Reactor Overpressure Protection, Enclosure 5, to TVA letter dated I

October 1,1997). Also, specify the safety / relief valve set points used in the analysis. Identify the NRC-approved model used in the analysis. ,

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7. Higher pump speed is expected (Section 3.4, Reactor Recirculation System, Enclosure 5, to TVA letter dated October 1,1997), but it is not clear how much 1

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4 I increase is expected. Specify the expected increase in speed. Describe the plant experience with higher pump speed operation such as increased core flow and pump vibration problems, if any, i

8. In NEDC-31897P-A. " Generic Guidelines for General Electric Boiling Water Reactor Power Reactor," May 1992, the staff required that plant-specific submittals must address the modifications described in General Electric (GE)

Service Information Letter (SIL) No. 377. Section 3.8, Reactor Core Isolation Cooling, Enclosure 5, to WA letter dated October 1,1997, does not address the SIL 377 modifications.

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9. Confirm that the reliability of the HPCI System will be monitored in accordance with the criteria which might have been developed to comply with the maintenance rule 10 CFR 50.65 (Section 4.2.1, High Pressure Coolant Injection (HPCI)), Enclosure 5, to TVA letter dated October 1,1997).

10. Analysis power assumed for the GEMINI analyses is only 100 % (Table 9-1, Parameters used for Transient Analysis, Enclosure 5, to TVA letter dated October 1,1997). Justify the analyses at 100 % instead of 102%. Also, the steam flow for the power uprate analysis is assumed at 100 % instead of 106 %.

Justify the analysis with only 100 % steam flow.

11. Please refer to Section 10.5, Required Testing, Enclosure 5, to TVA letter dated October 1,1997.

a. Tests will be required on the recirculation system to demonstrate flow control over the entire pump speed range to enable a complete calibration of the flow controlinstrumentation. These tests should also assure that no undue vibration occurs at uprate conditions.

b. Startup tests on HPCI during the initial startup after being licensed at j uprated power will be required. 1

12. As a result of power uprate, a number of variables and limits utilized in the .

Emergency operating Procedures (EOP) may be affected (Section 11.1.2.3, I Emergency Operating instructions, Enclosure 5, to WA letter dated October 1, 1997). GE report NEDC-32751P states that "The plant EOls will be reviewed for j any effects of power uprate, and the EOls will be updated as necessary." i Confirm that TVA performed a review of the EOP variables and limit curves for the uprate conditions.

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13. The following items in the acceptance criteria are not addressed: fuel integrity, radiological consequences, containment pressure, reactor oscillations and long  ;

term shutdown and cooling (Section 9.3.1, ATWS, Enclosure 5, to TVA letter j '

dated October 1,1997). What is the calculated peak containment pressure?

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14. Class 1E battery capacity and compressed air system are not addressed for the scoping analysis (Section 9.3.2, Station Blackout, Enclosure 5, to TVA letter dated October 1,1997).

15. The staff stated in its Safety Evaluation (SE) for NEDC-31984P (letter W T.

Russell to P. W. Marriott, July 31,1992) that " Individual licensees should share to existing radial power shape limitations when designing core reloads for uprated conditions." Confirm that this requirement is followed.

E. MAINE YANKEE LESSONS LEARNED

1. The submittal included proposed changes to the technical specifications.

However, the submittal did not provide any matrix or plan indicating which sections of the Final Safety Analysis Report (FSAR) will be superseded by current extended power aprate re-analysis. Provide a list or matrix that identifies which subsections of the FSAR will be superseded and identify the corresponding sections of the current submittal. The actual updating of the FSAR will be governed by the current regulations, and the affected FSAR subsections should be documented.

2. Provide a list of all the computer codes used to perform the re-analysis and indicate if the particular code was approved for the specific application. Respond to the following requests which pertain to the codes used in the power uprate.

a. Review the approving SE for the each code and state whether your application of the code complies with any limitations, restrictions or conditions specified in the approving SE. Demonstrate that your applications of the computer codes in the reanalysis conforms with all assumptions and restrictions given by the corresponding approving SE.

b. In addition, review the SEs for the extended power uprate generic reports and indicate if you complied with all restrictions stated in the approving SE.