ML20236Q659

Summary of 980709 Meeting W/Tva Re Browns Ferry Nuclear Plant License Amend Request for Power Uprate.List of Attendees & Request for Addl Info,Encl

ML20236Q659
Person / Time
Site:	Browns Ferry
Issue date:	07/10/1998
From:	Raghavan L NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
TAC-M99711, TAC-M99712, NUDOCS 9807200392
Download: ML20236Q659 (8)

Text

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July 10,1998 LICENSEE: Tennessee Valley Authority FACILITY: Browns Ferry, Units 1,2 and 3

SUMMARY

OF MEETING WITH TENNESSEE VALLEY AUTHORITY -

MEETING OF JULY 9,1998 (TAC NOS. M99711 AND M99712)

On July 9,1998, the Nuclear Regulatory Commission (NRC) staff met with representatives of the Tennessee Valley Authority (TVA or the licensee), the licensee for Browns Ferry Nuclear Plant, Units 1,2 and 3. Attachment i lists the meeting attendees.

The discussions were technical in nature regarding BFN license amendment request for power

, uprate. Attachment 2 describes issues for which the NRC staff requested the licensee to provide additional information. The licensee agreed to provide the requested information. The meeting did not result in any regulatory action.

Original signed by L. Raghavan, Senior Project Manager Project Directorate ll-3 Division of Reactor Projects - 1/Il Office of Nuclear Reactor Regulation Docket Nos. 50-259.

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LIST OF ATTENDEES MEETING WITH REPRESENTATIVES OF TENNESSEE VALLEY AUTHORITY ROCKVILLE. MARYLAND JULY 9.1998 Name Organization L. Raghavan NRR Fred Hebdon NRR Raj Goel NRR Vonna Ordaz NRR Steve LaVie NRR Peter Kang NRR George Thomas NRR Frank Collins NRR David Shum NRR John Wu NRR Tom Scarbrough NRR Paul Shemanski NRR Tony Ulses NRR Tim Abney TVA Jim Shaw TVA

. Ed Hartwig TVA Eric Frevold. TVA David Langley TVA J. McLamy TVA Henry Jones TVA Raymond Wright .TVA

. H. Mehta GE Nuclear H.Hoang GE Nuclear Dan Pappone GE Nuclear Joe Quirk GE Nuclear John Chase GE Nuclear Michael Dick GE Nuclear I

Attachment 1

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MEETING WITH REPRESENTATIVES OF TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT UNITS 2 AND 3 ROCKVILLE. MARYLAND JULY 9.1998 REQUEST FOR ADDITIONAL INFORMATION

1. Pioina and Comoonent analvsis The following questions are related to your letter dated May 22,1998,

a. Please provide the Editions of ASME Code for evaluation of the control rod drive mechanism, reactor recirculation and residual heat removal piping.

b Table 1(c)-2 shows that the stress in the supporting skirt exceeds the Code allowable limits. The skirt support was acceptable by elastic-plastic analysis. Please provide a desenption of the elastic-plastic analysis, Code-allowable limits and calculation results that derenstrate the reactor pressure vessel support skirt to be acceptable.

c. Table 1(c)-2 also indicates that the fatigue usage factors (CUFs) for the feedwater nozzles are 1.0 and 0.984 for the current rated power and the proposed uprated power conditions, respectively. Please provide description of how these two CUFs were calculated including the location and all transients which were considered in the CUF calculation.

d. On page E-21, you indic. ate that the power uprate temperature and pressure are bounded by the pre-upraw conditions used in the existing piping analysis. Therefore, the existing stress reports are bounding for the power uprate. This is inconsistent with Section 3.12.1, General E4ctric (GE), Licensing Topical Report NEDC-32751P, " Power Uprate Safety Analysis for fhe Browns Ferry Nuclear Plant (BFNP), Units 2 And 3,"

dated September 1997 (Proprietary), which states that operation at the uprated conditions, would increase the piping and piping component stresses due to slightly higher operating temperature, pressure, and flow rates internal to the pipes. Please provide the margin between the existing calculated stresses and the Code allowable limits for each line in Figure 3-4 and compare the margin to the stress increases in figures 3-4 and 3-5 based on specific increases in temperature, pressure and flow rate.

Also please provido an evaluation of the piping systems attached to the torus shell with regard to the incrrease in the pool temperature at the power uprate condition.

e. In your response S, you did not address issues relating to pressure locking and thermal binding of valves (Generic Letter (GL) 95-07 issues) Please provide an evaluation of the power uprate effects on the potential pressure locking and thermal binding of safety-related power-opemted valves. Also discuss the potential for over-pressurization of isolated water-filled piping sections per GL 96-06.

Attachment 2

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f. In its GL 89-10 inspection at Browns Ferry, April 27 to May 1,1998, the U.S. Nuclear Regulatory Commission staff determined that you had not updated the motor operated valve (MOV) calculations for Unit 2 to reflect the power uprate conditions. Please provide a schedule for revising these calculations (e.g., MOV, and other valve and pump calculations) to reflect the power uprate conditions for Browns Ferry Unit 2 and identify any expected adjustment or modifications. Please confirm that all required modifications will be accomplished prior to the implementation of the proposed power uprate for Unit 2 or 3.

g. Please clarify whether any MOV modifications (in addition to the torque switch adjustments for the four GL 89-10 MOVs) are planned for the Unit 3 power uprate.

Also, indicate if any other power-operated valves (such as air-operated valves or hydraulic-operated valves) were adjusted or modified based on the power uprate conditions.

h. Please discuss the post-accident containment temperature increase (from 322'F to 336*F) as a result of the " GOTHIC" analysis (your letter of March 16,1998) and its effects on MOV output (GL 89-10 issue), pressure locking and thermal binding (GL 95-07 issue) and potential over-pressurization of isolated water-filled piping sections (GL 96-06 issue).

2. Soent Fuel Pools

a. Please provide the heat load and corresponding peak calculated spent fuel pool (SFP) temperature for both planned and unplanned full core offloads at the current power level and the proposed power uprate level and confirm whether these heat loads and corresponding SFP temperatures include a single failure of SFP cooling (e.g., one of two trains of SFP cooling).

b. Your May 20,1998 letter states that no specific calculations were made for the peak SFP normal operation and unplanned full core offloads and that the design basis for the SFP cooling system remains the same for the pre-and post-power uprate conditions. If no calculations were performed for the proposed power uprate level, please discuss your basis for assuring that both the heat load and the peak SFP temperature would not increase for the proposed power uprate conditions.

3. Reactor Systems

a. NEDC-32751P, Section 4.3 ECCS performance evaluation: please clarify the following statement: "The SAFER /GESTR code is the pre-uprate analysis for BFNP, and therefore an update from a previous analysis ECCS (emergency core cooling system)

Code is not a part of the uprate license amendment." Our understanding is that Tennessee Valley Authority (TVA) completed the loss-of-coolant-accident (LOCA) analysis for uprate conditions in 1996, as given in NEDC-32484P. Please confirm that power uprate will not change the limiting break, single failure, or the break spectrum as compared to the existing analysis.

b. Please provide a baseline coroparison run using SAFER /GESTR of pre- and post-power uprate conditions.

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c. NEDC-32751P, Section 6.5 SLCS: The pump disch::rge pressure is increased from 1275 psig to 1325 psig. Please discuss why the pump discharge rel;ef valve setpoint is not changed and what the setpoint of the relief valve is now. If the new pump discharge pressure is close to the present setpoint, there may be inadvertent lifting of the relief valve.

d. NEDC-32571P, Section 9.3.1 ATWS: Please confirm that the model ODYN was used for the plant-specific analyses.

e. Your May 20,1998 letter, item D.4, references licensed power. Please confirm that the power conditions are the uprated conditions.

f. In your May 20,1998 letter, you addressed the issues relating to Maine Yankee lessons learned and described different codes used in your power uprate evaluation. Your response does not indicate whether any third party or independent review of GE calculations was performed. Please discuss the process that was used to verify GE calculations are based on approved methodologies and consistent with all constraints.

g. Please identify new codes that were used in the power uprate analysa and confirm that they were used in accordance with any conditions associated with tM use of these codes. These codes should also identified in the technical specifice ons bases.

4. Electrical Power and Auxiliary Systems The folicwing questions are related to your May 20,1998 letter,

a. Although no hardware changes or modifications are needed for power uprate, the electrical power requirements for the condensate, condensate booster, and recirculation pumps are expected to increase. You response (Item B.2) did not quantify what increases in the electrical loads are required for these pumps, and concludes that the pre-uprate electrical calculations would be valid for power rprate. Please provide the basis for your conclusion including any supporting analysis to show that the onsite electrical distribution system voltage is adequate to handle the increases in the electrical loads required by the power uprate or demonstrate that the electrical load increases for the above pumps are minimal. Also, submit the one line-diagrams (from load flow cases) which illustrate the load and voltage changes before and after power uprate cases under the worst expected grid voltage. As part of its review, the staff will examine the bus loadings and voltage changes for the onsite and the offsite electrical power system.

b. In a previous request for additional information (RAl), the staff requested the list of grid stability cases performed to support the power uprate and summary of thf., findings for each case. Please provide the list and discuss new stability limits that would result in an Operations Standing Order (response item B.3).

c. In response to staff's question that an increase of 57.5 MW generation to each unit could have an impact on grid voltage profile, you indicated that (response item B.3)

PSB-1 does not require reanalysis since the methodology and software have not changed, and the degraded grid setpoint is not affected by the added generation because there are no load changes. The staff believes that the offsite grid voltage may be impacted as result of power uprate and by increases in the onsite loads which, in

4 turn, could affec' previous PSB-1 analysis and the degraded grid setpoints. Please reassess the degraded grid setpoint calculation and PSB-1 based on the new grid voltage to ensure its adequacy, or show that there is no impact on the previous PSB-1 analysis.

d. Besides containment spray and residual heat removal pumps, please identify and discuss acceptability of other eiectrical equipment for which name plate horsepower values are not used.

e. Based on the review of the radiological doses for the safety-related electrical equipment before and after power uprate, please confirm that all safety-related electrical equipment is bounded by the original design basis.

5. Radiolootal Issues in the February 18,1998 RAI, the staff regarding parameters for the design basis accidents (DBA) analyses. TVA responded in a letter dated April 1,1998. TVA has determined the radiological doses due to DBAs for power uprate conditions by scaling the pre-uprate doses upward by 5%. Although TVA provided great detail on how the scaling factor was developed and applied, the input data to each analysis was not provided. In its letter of May 7,1998, the staff identified a large number of apparent discrepancies between the Updated Final Safety ,

Analysis Report (UFSAR) and your October 1,1997 analyses and requested that TVA provide I a tabulation of analysis parameters. TVA responded in a ictter dated June 12,1998. After reviewing the response, the staff has the following additional questions:

a. Please resolve the following discrepancies between the data provided in your June 12, 1998 letter and the UFSAR and other regulatory documents.

1. Item 1.b.(l) of the table identifies that the iodine concentration in the containment sump water is 25% of the core inventor / Tbh is inc9nsistent with regulatory guidance that the activity should be based on 50% core inventory.

2. Item 3.c specifies that main steam isolation valves (MSIVs) will close in 5.5 seconds.

The analysis described in the UFSAR assumes an isolation time of 10.5 seconds.

3. Item 3.d lists main steam line break steam and water release quantities of 19,874 and 43,740 lbm, respectively. The UFSAR analysis lists 25,000 and 160,000 lbm, respectively.

4. Item 4.a states that there are 48,132 fuel rods in the BFN core (764 bundles). This l appears to assume 63 rods per bundle. In the generic GE 8 x 8 fuel there are two water rods and 62 rods that contain fuel, for a total of 47,368 rods in the core. The UFSAR analysis assumes 62 rods per bundle. In this application, a larger number of rods is less conservative.

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5. Your response, item 4.h. states that 5500 cubic feet of release occurs from the reactor building prior to isolation and transfer to standby gas treatment system  !

(SGTS). This is apparently based on a 15-second ventilation flow at the rate of 22,000 cfm. This flow rate is inconsistent with the second bulleted item under item 5 which states that the air flow from the reactor building prior to isolation is 95,000 cfm.

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6. The new power after uprate would be 3458 Mwt. Your submittal indicates that safety analyses have been performed at 102% of the uprated power level. This is inconsistent with Section 8.3.2 of your submittal which states that "... fission product inventories are prepared based on irradiation of BFNP fuel for 1400 days at the uprated power of 3458 Mwt...." However, RG 1.49, Power Levels of Nuclear Power Plants, paragraph C.3, provides that analyses of the offsite radiological consequences of postulated design basis accidents should be performed for an assumed core power level equal to 1.02 times the proposed licensed power level.

l b. Please review the above discrepancies and any other UFSAR discrepancies in the data I

relating to power uprate issues and determine whether they involve an unreviewed safety questions. Please inform us of your schedule for resolving these discrepancies.

This is necessary for the staff to make a finding on the acceptability of the power uprate values obtained by the scaling approach.

c. The first bulleted item under item 5 states that the reactor zone volume is 1,335,000 cubic feet prior to secondary containment isolation. TVA has previously stated that the secondary containment is being treated as a single zone. Please explain (1) the l- applicability of the phrase " prior to isolation" and (2) why the secondary containment l volume of 1,931,500 cubic feet used in other analyses is not applicable.

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6. Human Factors lasues

a. In response to the r.taff's RAl, by letter dated April 28,1998, you discussed issues relating to operator actions that ere particularly sensitive to the power uprate, including operator response times, or performance. You stated that you have reviewed all l

operator responses used in your probalistic safety snalysis (PSA) and confirmed that the effect on operator response times [due to power uprate) at Browns Ferry are consistent with the GE generic findings. For certain operator actions which would be sensitive to power uprate, the required operator response time has decreased. You

concluded that because these operator actions are controlled by emergency operating procedures, the ' slight reduction in response times noted for the power uprate condition l "will not significantly affect the operator's ability to safely complete the required actions."

For certain scenarios, operator responses are required to be achieved in less than 1 5 minutes. In general, minimal reduction in response times should not significantly l affect the operator's ability to complete their actions, but the staff is concemed that I those actions assumed to be performed in 5 minutes or less by the PSA may not be achievable under realistic conditions. The staff refers to guidance contained in ANSI /ANS Standard 58.8, " Time Response Desiga Criteria for Safety-Related Operator Actions" (1994), which indicates that safety-related operator actions that must be  !

initiated within 5 minutes or less (for events that occur with an estimated frequency of -

104) "shall be initiated by automatic protection systems.". Therefore, please provide evidence that operators can perform the required tasks under accident conditions in the times assumed by the PSA.

b. Your April 28,1998 letter also indicated that one manual action, i.e., termination of the High Pressure Coolant injection [HPCl] system injection following an Appendix R fire event has a reduction in response time from 10 minutes to 7 minutes. Your letter explained that the reduction in response time is a result of using different models to l

6 predict pre-and post-uprate operator action times (i.e., the GE SAFE model for pre-uprate predictions and the SAFER model for post-uprate). Your letter further ;ndicated that this action which involves closing a valve from the 250V DC reactor MOV board located just outside the main control room, has been performed in a shorter time than allowed by the SAFER model, i.e.,7 minutes.

1. Your letter, page El-4 states: 'TVA has previously demonstrated that this action, close one valve from the 250V DC reactor MOV board, located on the same elevation just outside the main control room, can be performed within the shorter time predicted by the SAFER model." Please provide a reference for this demonstration including the following factors considered: Environmental conditions expected; procedural guidance for the required actions; support personnel and/or equipment required to carry out the required actions; information requirements including qualified instrumentation.

2. The SAFER modelis used to analyze LOCA conditions and fuel heatup activities.

Your October 1,1997 letter, (enclosure 5) indicates that GE applied the model to analyze an Appendix R fire event and stated that, " Sufficient time is available for the operator to perform the necessary actions" (p.6-9). How does the SAFER model evaluate human actions? What is the basis for GE's conclusion that operators have sufficient time to perform necessary actions?

3. It is our understanding that the SAFER model was used for the power uprate analysis and the SAFE model for the pre-uprate analysis. According to your April 28,1998 letter, the SAFE model predicted a 10-minute required operator action time to shut down HPCI. The SAFER model predicted 7 minutes to shut down HPCI.

Which action time is TVA taking credit for?

4. In addition, what are credible errors that operators could make in taking this action?

What are the consequences of the operator failing to accomplish the action and how will recovery from the failure (s) be accomplished? How does WA know that operators can successfully recover from credible errors, i.e., provide evidence that operators can recover from credible errors.

c. Please describe all changes the power uprate will have on the operator training program and the plant simulator. Provide a copy of the post-modification test report (or test abstracts) to document and support the effectiveness of simulator changes as required by ANSI /ANS 3.51985, Section 5.4.1 within 60 days of implementing power uprate on each unit.