L-2006-115, Appeal of NRC Final Significance Determination for a White Finding and Notice of Violation

Appeal of NRC Final Significance Determination for a White Finding and Notice of Violation

ML061250321
Person / Time
Site:	Turkey Point
Issue date:	04/27/2006
From:	Jones T Florida Power & Light Co
To:	Document Control Desk, NRC/RGN-II
References
IR-06-010, L-2006-115
Download: ML061250321 (9)

Text

APR 2 7 20E FPI '

L-2006-115 10CFR50.4 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, DC 20555-0001

Subject:

Florida Power and Light Company (FPL)

Turkey Point Units 3 and 4 Docket Nos. 50-250 and 50-251 Appeal of NRC Final Significance Determination for a White Finding and Notice of Violation (Turkey Point Nuclear Plant - NRC Inspection Repor No. 05000250.251/200610)

Reference:

1) NRC Letter, C. A. Casto to J. A. Stall, Turkey Point Nuclear Plant -

Integrated Inspection Report 05000250/2005005 and 05000251/2005005; Preliminary White Finding, dated January 27, 2005[6]

2) FPL Letter T. 0. Jones to USNRC, Document Control Desk, Responsc to NRC Integrated Inspection Report Preliminary White Finding Apparent Violation, L-2006-066, dated March 13, 2006

3) NRC Letter, William D. Travers to J. A. Stall, Final Significance Determination for a White Finding and Notice of Violation (Turkey Point Nuclear Plant - NRC Inspection Report No. 05000250,251/200610), dated April 17, 2006 Gentlemen:

Reference 1) provided the details of an NRC inspection finding characterized as a preliminary white finding, regarding an inoperable Auxiliary Feedwater (AFW) Pump, due to an incorrectly installed bearing. Reference 2) provided FPL's response to the preliminary white finding, in which FPL agreed with the NRC determination of a Technical Specification 3.7.1.2 violation for the Auxiliary Feedwater (AFW) System and a contributing violation of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, for the AFW pump failure. Reference 3) provided the NRC's final significance determination for the white finding and a corresponding Notice of Violation.

FPL has completed its review of NRC Inspection Report No. 05000250,251/200610 and the Notice of Violation VIO 05000250,251/2006010-01. The NRC has correctly reflected FPL's position, including the reason for the violation, the corrective actions taken to correct the violation and prevent recurrence, and the date when full compliance was achieved as stated in the FPL March 13, 2006 letter and the NRC Inspection Report 05000250,251/2005005.

Accordingly, no written statement or explanation pursuant to 10 CFR 2.201, Notice of Violation, is required by FPL.

an FPL Group company

L-200-1 15 FPL ha, reviewed the additional information titled "NRC Evaluation' of Risk Significant Factors"', which was provided as Enclosure 2 to Reference 3). As set forth in further detail in Attachrment 1 to this letter, FPL respectfully wishes to appeal the NRC's final significance determination and requests a reduction of the significance of the inspection finding (from white to green). FPL's contentions fall into the category of "Actual (verifiable) plant hardware, procedures and equipment configurations that were not considered by the staff' as set forth in NRC Inspection Manual Chapter 0609.

If there are any questions regarding this letter, please contact Walter Parker at 305-246-6632.

Sincerely Yours, Terry C. Jones Vice President Turkey Point Nuclear Plant Attachments: 1) Final Significance Determination Appeal of Notice of Violation VIO 05000250,251/2006010-01 cc: Senior Resident Inspector, USNRC, Turkey Point Director, Division of Reactor Projects NRC Regional Administrator

L-2006-1 15 ATTACHMENT 1 Final Significance Determination Appeal of Notice of Violation V10 05000250,251/2006010-01 Based upon IMC 0609, Attachment 2, ProcessforAppealing NRC Characterizationof Inspection Findings (SDP Appeal Process),FPL requests the NRC to reduce the final significance determination (White) of the NRC inspection finding involving the inoperable auxiliary feedwater (AFW) pump. As provided in IMC 0609, a request to reduce the significance of an inspection finding will be considered as having sufficient merit for review by this appeal process only if the licensee's contention falls into one of the following categories:

a. Actual (verifiable) plant hardware, procedures, or equipment configurations were not considered by the staff.

b. The staff's significance determination process was inconsistent with the applicable SDP guidance or lacked justification.

FPL's appeal is based upon category "a" in that it involves actual (verifiable) plant hardware, procedures and equipment configurations that were not considered by the NRC staff.

Enclosure 2 of the NRC Final Significance Determination letter to FPL, dated April 17, 2006.,

itemized in thirteen enumerated paragraphs (nos. I - 13) the risk significant factors and issues that form the bases for the NRC's final risk significance determination. As described below, FPL's request for reduction of the final significance determination is limited to the issues described in the NRC's April 17, 2006 letter, paragraphs numbered 1 and 10 (Risk Factor #1 and

1. 10). FPL considers these items as the most significant contributors to the NRC's risk significance determination.

FPL Position: Risk Factor # I This consideration presents actual (verifiable) plant hardware, equipment capability and system configurations that were not considered by the staff in its SDP.

The NRC evaluation indicates that for AFW "Recognizing that surveillance performance does not represent actual demand conditions (longer duration, higher pressure, and flow increasing shaft loading), no correlation for bearing performance between test operation and that which would be applicable for an actual demand was provided." In addition, the NRC evaluation stated that ".. ..the post reactor trip performance of AFW pump B on March 22, 2005, was evaluated as the break point between failure in less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. This is because the information provided indicated that the pump operated in a post trip condition for greater than I hour."

As determined by FPL's root cause, the failure of the bearing was due to lack of lubrication. The lack of lubrication causes wear of the bearing and is directly related to the rotational speed of the pump: the higher the speed, the more potential bearing wear. As the bearing wears, the rotating element support is diminished which causes the element to sag. Since this AFW pump is a high 1

L-2006-1 15 speed machine and with an off-center rotating element, this causes a rotation in an oval pattern.

This off-center rotation causes an increase pump vibration.

This phenomenon is time dependent, since pump operating time will be required to wear the bearing. As operating time increases, the vibration increases. This behavior is consistent with actual observations.

Surveillance testing of AFW is performed via 3/4-OSP-075.1, -075.2 (monthly surveillances) or 3/4-OSP-075.6, -075.7 (quarterly surveillance utilizing back-up nitrogen). Common to all procedures is the configuration and flow characteristics are a direct representation of actual demand conditions.

For a specific AFW pump, surveillance testing is performed with the AFW pump in its normal Technical Specification required standby alignment and the test is initiated by opening the associated steam supply motor operated valve (MOV). Opening the associated steam supply MOV simulates the automatic AFW pump start and the AFW Flow Control Valves' opening actuation signals. When the steam supply MOV travels off the closed seat, limit switches on the steam supply MOV complete the control circuit to open the AFW Flow Control Valves to direct flow to all three steam generators. Due to the operational characteristics of the turbine governor ramping to rated speed, the three AFW Flow Control Valves travel full stroke open until pump discharge pressure exceeds steam generator pressure. Upon initiation of AFW flow to the three steam generators, the characteristic flow response is an initial overshoot, due to the AFW Flow Control Valves being full open, and subsequent slow decrease to design flow of 130 gpm to each steam generator as flow control action occurs (total flow = 390gpm - the same as in design basis conditions). This response is observed during all surveillance testing and is bounded for all design basis events for which AFW flow is credited. In this manner, the testing is the same as actual injection during event response.

Since the pump and turbine operating speed is a setting of the terry turbine governor and is maintained at the 5900 rpm design operating speed, variations in the AFW flow or required AFW injection pressure (which could be higher during an AFW demand event) does not impact the turbine / pump rotational speed. Since the bearing wear rate is a function of the pump speed, no difference in bearing wear would be expected if the pump is operated in a surveillance steam generator injection mode versus an actual event injection mode.

Surveillance testing is also performed at various steam generator pressures with similar system response. Normal surveillance testing occurs during power operation at approximately 100%

reactor power. Steam generator pressure is approximately 760 psig at the location of the AFWV turbine/pump during AFW system operation. However, surveillance testing is also performed in Mode 2 following refueling outages. Steam generator pressure recorded at the location of the AFW turbine/pump on April 8, 2006, following the recent Unit 3 refueling outage, was 970 psig during AFW system operation. The critical aspect observed during various test conditions is that the discharge pressure at the AFW pump is unaffected by steam generator conditions, due to AFW Flow Control Valve response to the flow demand signals.

Based upon the critical aspects of PTN's surveillance test method discussed above, and 1) direct injection of design flow rate Condensate Storage Tank water into the Steam Generators, 2) testing the AFW System in the same manner as would occur during an actual AFW Actuation 2

L-2006-115 Signal, and 3) plant conditions during the testing that provide adequate representation of design basis event conditions, it is concluded that there is a direct correlation for bearing performance between test operating conditions and conditions that would be required for an actual AFW flow demand. Since the surveillance testing and event response both subject the pump to the same hydraulic conditions relative to pressure, flow/shaft loading, the time dependent degradation assumption for the B AFW Pump is maintained.

Considering the above system configuration and performance information, it is concluded that the AFV surveillance performed on October 10, 2005, with a total operating time of 2.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, is representative of operating in an injection mode during an actual event. Because there are NO differences in the AFW operating parameters during surveillance testing or actual event response, this surveillance test justifies that the pump would have operated adequately for greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> as late as October 10, 2005.

Based on the above, the exposure period is calculated to be approximately 1 / 12 th of the year or 8.3%, rather than the 63% used by the NRC.

FPL Position: Risk Factor # 10 This consideration presents information regarding the actual (verifiable) plant system configuration and plant procedures that were not considered by the staff in its SDP. NRC disposition for item 10 of the Final SDP evaluation of the Turkey Point B AFW Pump event wvas based on an incorrect analysis of operator response to a loss of 125 volt vital DC bus event.

Specifically, the analysis performed by NRC stated that "The procedure for responding to a loss of the A. DC bus does not direct or provide a transition to the procedure that contained the instructions for operating AFW pump A locally". It appears that the only procedure to be used following a loss of 125 volt vital DC bus from 100% power was assumed to be the off-nornal operating procedure (ONOP) 3/4-ONOP-003.4, ",Loss of DC bus 3D01 and 3DO1A (3A)".

Whereas, a loss of either A or B 125 volt vital DC bus will cause the associated train reactor trip breaker to open on a loss of power to the breaker's undervoltage coil. Opening of either reactor trip breaker causes a reactor trip.

The plant response to a reactor trip is to enter the Turkey Point Emergency Operating Procedure network, beginning with 3/4-EOP-E-0, Reactor Trip or Safety Injection. Direction and training to restore AFW is available using the EOP network. The licensed operator training lesson plan 690225:3, "Loss of a Vital DC bus" reflects the fact that the ONOP for the loss of a 125 volt vital DC bus is not referenced until after the plant is stabilized per 3/4-EOP-ES-0.1, Reactor Trip Respons;e. This aspect of the classroom training lesson plan is reinforced during simulator training using simulator practice scenario 760209201, "Loss of SFP Level & Cooling / MSR Intercept Valve Closure / RCP Oil Cooler Leak / Loss of 3D01". The following sequence of procedure use applies.

3

L-2006-1 15 314-EOIP-E-0, Reactor Trip or Safety Injection Operators perform steps 1 through 4 from memory which verify:

1) Reactor trip

2) Turbine trip

3) Power to Emergency 4kV buses

4) Safety injection (SI) actuation if required Assuming SI is not required, the operators then perform the actions from the 3/4-EOP-E-0 "Response Not Obtained" (RNO) column for step 4. Specifically the operators are directed to:

1) Stabilize the plant as follows or as directed by the Shift Manager:

Ensure Open SGFP A AND B Recirculation Valves.

Maintain Tavg between 5430 F to 550'F.

. Maintain pressurizer pressure between 2220 to 2250 psig.

. Maintain AFW flow greater than 345 gpm until S/G levels between 6% to 50%.

Monitor steam dumps for proper operation.

• Close feedwater isolation MOVs.

These six (6) steps are the first actions performed, following the four (4) immediate steps for a reactor trip response.

Since the crew is directed to maintain AFW flow in the above step, the operator monitors indications of AFW flow and SIG level. It is normal for a trip from 100% power for SIG level to drop below 10% narrow range which would automatically initiate AFW. If automatic AFW initiation did not occur, due to the loss of a vital 125 volt DC bus, then the operator would report that cordition to the unit SRO, who is following the procedure, and manual initiation of AFW is then directed at the appropriate step. This action takes place whenever a loss of all AFW would occur, initially or after some period of AFW pump run time.

Following the plant stabilization steps above, the next two (2) RNO steps in 3/4-EOP-E-0 direct the crew to:

2) Monitor Critical Safety Functions using 3/4-EOP-F-0, CRITICAL SAFETY FUNCTION STATUS TREES.

3) Go to 3/4-EOP-ES-0.1, REACTOR TRIP RESPONSE, Step 1.

If all steam generator levels were below 6% and AFW flow was less than 345 gpm, due to AFZW system malfunction, then 3/4-EOP-F-0 directs immediate entry into 3/4-EOP-FR-H.1, Response to Loss of Secondary Heat Sink. For a loss of all AFW, these conditions would most likely exist at the pDint of transition from 3/4-EOP-E-0 to 3/4-EOP-ES-0.1 or shortly thereafter. Regardless, with loss of all AFW, entry into 3/4-EOP-FR-H.1 is directed by 3/4-EOP-F-0.

4

L-200(-115 3/4-EOP-FR-H.1, Response to Loss of Secondary Heat Sink Step 1 - For a reactor trip and loss of all feedwater from mode 1, a secondary heat sink will be determined as required.

Step 2 - The crew is directed to try to establish AFW to at least one steam generator.

a) First, S/G blowdown is verified isolated.

b) Next, the cause for loss of AFW is evaluated. In this case it is most likely related to a loss of DC power induced failure of an AFW steam supply valve to open or AFW' flow control valve failure.

c) The operator is directed to 'Try to restore AFW flow".

d) "Check total feed flow to S/Gs - greater than 345 gpm". RNO step d.2) states:

IF feed flow to at least one S/G NOT verified, THEN dispatch operator to locally restore AFW flow.

3/4 EOP-FR-H.1, steps 2.c and 2.d do not specifically direct the use of 3/4-ONOP-075, Auxiliary Feedwater System Malfunction, for restoration of AFW, but are the cues to enter the ONOP for guidance regarding manual local actions available to restore the AFW system flow. The Control Room operators continue in 3/4-EOP-FR-H.1, while field operators perform manual local actions to restore AFW outlined below.

The licensed operators are trained to refer to 3/4-ONOP-075 for specific guidance on restoration of AFW while the EOP network is in use. While 3/4-ONOP-075 is not used as a step-by-step procedure concurrently with EOP network procedures (other than for 3/4-EC'P-ECA-0.0, "Loss of All AC Power") operators are trained to use step 3 and Attachments 3.,4 and 5 for guidance in restoring AFW. Use of 3/4-ONOP-075 in this manner during non-loss of all AC power scenarios is covered during licensed operator simulator training using the following scenarios:

750207601 - Ruptured Steam Generator with Loss of AFW 750208301 - Ruptured Steam Generator with Loss of AFW & SSGFPs 760207301 - Condenser Tube Leak/Loss of All Feedwater/ATWS/Loss of Heat Sink 760208301 - Ruptured Steam Generator with Loss of AFW and SSGFPs 3/4-ONOP-075, Auxiliary Feedwater System Malfunction One of the entry conditions for this ONOP is the loss of Auxiliary Feedwater due to a failure of the pumps, valves, and/or associated piping system. While EOPs are in use, this ONOP is used as a reference document for methods to restore AFW, when the minimum required 345 gpm flow has not been established. The licensed operators in the control room will normally direct field operatcr(s) to locally perform the portions of 3/4-ONOP-075 deemed necessary to restore AFW to service. Note that the Nuclear Station Operator (field operator) training program provides classroom training and in-field practice for the local actions required by 3/4-ONOP-075 to restore AFW. A copy of 3/4-ONOP-075 is maintained in the area of the AFW pumps for ready access by field operators directed to perform these local actions. The following exceprts from 3-ONOP-075 are used for guidance by licensed operators in determining what local actions should be taken to restore AFW:

5

1I L-200-1 15 Step 3a - Direction to locally open AFW steam supply MOVs if still closed Step 3b - Direction to use Attachment 3 for local operation of AFW flow control valves if they are found closed Step 3cl - Direction to use Attachment 4 to locally reset the AFW pump overspeed trip if necessary to restore an AFW pump to operation.

Step 3c2 - Direction to use Attachment 5 to locally operate the AFW pump trip & throttle valve if necessary to restore an AFW pump to operation 3/4-EOP-FR-H.1, Response to Loss of Secondary Heat Sink (continued)

While field operator actions to restore AFW to operation are in progress using the guidance of 3/4-ONOP-075, the crew continues to use other methods to restore SIG feedwater per 3/4-EOP-FR-H.1:

Step 3 - All RCPs are stopped to reduce heat input to the reactor coolant system Step 4 - An attempt is made to restore main feedwater flow from a S/G feed pump via the feedwater flow control bypass valves (locally operated if A 125 volt DC bus lost).

Step 5 - Check is made to see if feedwater flow is restored Step 6 - Standby feedwater flow is aligned from either A standby S/G feedwater pump (motor-driven) or B standby S/G feedwater pump (diesel-driven). These pumps can be started remotely or locally. After pump start, one manual isolation valve is locally opened for the affected unit and flow is established to the S/Gs via the feedwater flow control bypass valves (locally operated if A 125 volt DC bus lost).

Anothe:: method available involves depressurizing any S/G to < 385 psig and establishing feed with a condensate pump.

If any SIG wide range level drops below 22% during the process of heat sink restoration, thern the crew immediately establishes feed and bleed using steps 13 through 21.

Techniques and equipment used to restore feedwater flow in the aforementioned procedures are used frequently by the operators during equipment surveillance tests and training. In particular, the Operators are very familiar with use of the AFW and standby SIG feedwater systems due to the weekly surveillance testing outlined below. Additionally, the standby S/G feedwater system is normally used as the source of SIG feedwater when a unit is in modes 3 or 4.

Equipment Testing 3/4-OS'P-075.1, Train 1 AFW Operability Verification & 3/4-OSP-075.2, Train 2 AFW Operability Verification. These surveillances are performed once a week on a rotating basis so that each train is tested from each unit monthly, on a staggered basis. Each surveillance involves starting & operation of the AFW system for the train and unit covered by that surveillance procedure. Each AFW pump is tested individually and must develop 390 gpm flow to S/Gs within 95 seconds and maintain that flow for at least five minutes. Each AFW pump must run for at least 15 minutes.

6

L-2006-115 O-OSP-074.3, Standby Steam Generator Feedwater Pump Availability Test. This surveillance is performed weekly for the B (diesel-driven) standby feedwater pump and monthly for the A (motor-driven) standby feedwater pump. It involves starting the respective pump and establishing recirculation flow and pump Ap within established limits provided in Enclosure I of the procedure, when the unit is at power. The ability to deliver flow to S/Gs is tested when the affected. unit is shutdown (mode 3 and below).

Operator Testing The following Job Performance Measures are a part of the Licensed Operator Continuing Training program and are included in the bank used for annual operator testing required for maintaining an NRC RO or SRO license:

407500 2300 - Control S/G Level Locally with AFWV Control Valve (3/4-ONOP-075, Attachment 3) 4075011100 - Reset AFW Pump Overspeed Trip Latch (3/4-ONOP-075, Attachment 4) 4075012300 - Manually Control Steam to AFW Pump with Trip & Throttle Valve (3/4-ONOP-075, Attachment 5) 14074017101 -Place B Standby Steam Generator Feed Pump In Service (0-OP-074.1)

The information provided herein represents procedures and system configurations that were not considered by the staff.

7