JAFP-14-0012, Inservice Testing Program 10 CFR 50.55a Request PRR-05

Inservice Testing Program 10 CFR 50.55a Request PRR-05

ML14057A553
Person / Time
Site:	FitzPatrick
Issue date:	02/21/2014
From:	Adner C Entergy Nuclear Northeast, Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
JAFP-14-0012
Download: ML14057A553 (28)

Text

1~nteJg)1  :~~~:: ~

James A. Fitzpatrick NPP P.O. Box 110 Lycoming, NY 13093 Chris M. Adner Regulatory Assurance Manager JAFP-1 4-0012 February 21, 2014 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555

Subject:

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 James A. FitzPatrick Nuclear Power Plant Docket No. 50-333 License No. DPR-59

Dear Sir or Madam:

Pursuant to 10 CFR 50.55a(a)(3)(ii), James A. FitzPatrick Nuclear Power Plant (JAF) requests NRC approval of the proposed alternative to the vibration criteria requirements of ASME CM Code ISTB Table ISTB-5100-1 and associated sections (ISTB-5121(e), ISTB-5122(b), and ISTB-6200(a)). The proposed alternative will be applicable for the duration of the fourth interval inservice testing program. Compliance with the pump testing requirements associated with exceeding the vibration level alert range absolute limit result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. This request is based on analysis of vibration and pump differential pressure data indicating no pump degradation is taking place. The details of the 10 CFR 50.55a(a)(3)(ii) request are enclosed.

The fourth interval inservice testing program began October 1, 2007. JAF requests approval of the enclosed 10 CFR 50.55a request by August 31, 2014.

There are no commitments made in this letter. Should you have any questions, please contact the Regulatory Assurance Manager, Mr. Chris M. Adner, at (315) 349-6766.

Verytr rs, Ch is M. Adner Regulatory Assurance Manager CMA:ds : James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 cc: USNRC, Regional Administrator, Region I USNRC, Project Directorate USNRC, Resident Inspector

JAFP-1 4-0012 Enclosure 1 James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 5055a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii),

Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii),

Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety

1. ASME Code Components Affected Pump Name Code Class Pump Category P&ID Drawing 10P-3A Residual Heat Removal Pump A 2 Group A FM-20A 10P-3B Residual Heat Removal Pump B 2 Group A FM-20A 10P-3C Residual Heat Removal Pump C 2 Group A FM-20A 10P-3D Residual Heat Removal Pump D 2 Group A FM-20A

2. Applicable Code Edition and Addenda

ASME OM Code 2001 Edition including 2003 Addenda

3. Applicable Code Requirement

ISTB Table ISTB-5100-1, Centrifugal Pump Test Acceptance Criteria

4. Reason for Request

Pursuant to 10 CFR 50.55a, Codes and Standards, paragraph (a)(3)(ii), relief is requested from the vibration criteria requirements of ASME OM Code ISTB Table ISTB-5100-1 during the Group A or biennial comprehensive pump test or any other time vibrations are taken to determine pump operational readiness.

Relief is requested from ISTB Table ISTB-5100-1, and associated sections (ISTB-5121(e), ISTB-5122(b), and ISTB-6200(a)) that refer to Table ISTB-5100-1, requirements to test the pump on an increased periodicity due to vibration levels exceeding the ISTB alert range absolute limit of 0.325 inches per second. Compliance with the specified requirement results in hardship or unusual difficulty without a compensating increase in the level of quality and safety. The increased periodicity of testing is an additional burden to the operations staff, plant scheduling, and adds unnecessary run time to all RHR pumps. This request is based on analysis of vibration and pump differential pressure data indicating that no pump degradation is taking place.

5. Proposed Alternative and Basis for Use James A. FitzPatrick is proposing to use an alternative vibration alert limit of greater than or equal to 0.408 in/s. This provides an alternative method that continues to meet the intended function of monitoring the pump for degradation over time while keeping the required action level unchanged.

Pump Testing Methodology The RHR pumps at James A. FitzPatrick are tested using a full flow recirculation test line back to the suppression pool for each pump surveillance test (including quarterly group A tests and bi-annual comprehensive pump tests). These pumps have a minimum flow line (per division) which is used only to protect the pump from overheating when pumping against a closed discharge valve. The min-flow line isolation valve for each division is initially open when the pump is started, and flow is initially recirculated through the min-flow line back to the suppression pool. Then, the full flow test line isolation valve is throttled open to establish flow through the full flow recirculation test line. The min-flow line is then isolated automatically, and all flow is directed through the full flow test line for the IST test.

Page 1 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii),

Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety The RHR system is operated in the same manner and under the same conditions for each IST test, regardless of whether JAF is operating or shutdown. Consequently, the pumps will experience the same potential for flow-induced, broad band vibration whenever they are tested, whether JAF is operating or shutdown. As a result, this relief is requested for the inservice testing of RHR pumps when vibration measurements are required or any other time vibrations are recorded to determine pump operational readiness (i.e., post-maintenance testing, other periodic testing, etc.).

NRC Staff Document NUREG/CP-0152 NRC Staff document NUREG/CP-0152, entitled Proceedings of the Fourth NRC/ASME Symposium on Valve and Pump Testing, dated July 15-18, 1996, included a paper entitled Nuclear Power Plant Safety Related Pump Issues, by the NRC staff. That paper presented four key components that should be addressed in a relief request of this type to streamline the review process. These four key components are as follows:

I. The licensee should have sufficient vibration history from the inservice testing which verifies that the pump has operated at the vibration level for a significant amount of time, with any spikes in the data justified.

II. The licensee should have consulted with the pump manufacturer or vibration expert about the level of vibration the pump is experiencing to determine if the pump operation is acceptable.

III. The licensee should describe attempts to lower the vibration below the defined code absolute levels through modifications to the pump.

IV. The licensee should perform a spectral analysis of the pump driver system to identify all contributors to the vibration levels.

The following is a discussion of how these four key components are addressed for this relief request.

I. Vibration History

a. Testing Methods and Code Requirements Elevated vibration levels on the RHR pumps, has been a condition that has existed since original installation. Prior to 1998, testing was measured in displacement (mils). These readings were taken in two directions, horizontal in-line with pump flow and horizontal perpendicular to pump flow. In 1998, JAF entered the third 10-year interval and implemented ASME/ANSI OM-1989, OM-6, for pump testing. During this interval, IST vibration data was taken in displacement (mils). Additional data was also taken in velocity (inches per second), with an additional data point in the axial direction. At the time, the velocity data points were used as info only. Upon the fourth 10 year interval at JAF, ASME OM Code 2001/2003 Addenda was adopted. With this adoption, it was determined that vibration measurements in velocity would be a much better indicator of pump condition and would be beneficial in terms of early identification of degradation. Therefore, data exists for two vibration points on each RHR pump from January 1986 to August 1998 in mils. Data from December of 1998 to Present is in the form of inches per second at three vibration data points. Various analyses of this data are included as figures with in this relief request.

Page 2 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii),

Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety

b. Review of Vibration History Data IST trends of RHR Pump vibration (Figures 1-4), which include data from 1998 through present, show readings to be consistently at or above current IST vibration alert criteria. From Figures 10-12, it can be seen that, 10P-3B and 10P-3C have exceeded the current OM Code alert criteria of 0.325 inches per second.

RHR pump Differential pressure trends (Figures 5-8) illustrate the differential pressure data during the same time period as the vibration figures 1-4. These graphs show a step change in flow around the 2009 time frame. This change is due to surveillance test changes in which test flow was lowered. The change was made after engineering analysis resulted in revised pump flow criterion. These trends do not show any signs of hydraulic degradation.

A review of the maintenance history for all four pumps and motors shows very minimal maintenance has been performed beyond the normally scheduled preventive maintenance. The only maintenance deemed to have the potential to effect vibration values is that of mechanical seal replacement.

10P-3A and 10P-3C both had mechanical seals replaced in 2009. Post work testing of these replacements did not show any change in vibration values.

Average run times for each RHR pump per cycle is approximately 200-300 hours. Run times of this nature, combined with the pumps and motors being built and maintained to the nuclear quality standards, are considered to have a low likelihood for significant wear related degradation.

c. Review of Spikes in Vibration Data Trends of recent vibration history (4th Interval) do not show any significant spikes above baseline levels. Instead, all values are seen as fairly consistent and there have been no significant degrading trends associated with vibration data for the past 15 years.

While the overall vibration trends have been steady, when compared to the OM Code recent vibration data points have exceeded the acceptance criteria.

RHR pump 10P-3B was the first pump to exceed the acceptance criteria in August of 2011. The V1 data point was reported as 0.34 in/s vice the 0.325 in/s requirement. Per ISTB-6200(a) of the OM Code, 10P-3B has been tested twice per quarter since this criteria failure. This increased frequency testing also affects 10P-3D, as surveillance testing procedures test both pumps in that particular division. In September of 2012, 10P-3C exceeded the acceptance criteria in both the V1 and V2 directions with values of 0.34 in/s each. 10P-3C, and subsequently 10P-3A have been on increased frequency testing since. Since the first instance in 2012, 10P-3C has exceeded the criteria in the V2 direction three additional times, once in August of 2013, once in November, and once in December with readings of 0.38, 0.34, and 0.35 in/s respectively.

Page 3 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii),

Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety II. Consultation- Pump Manufacturer/Vibration Expert During the initial investigation for the cause of the failed vibration acceptance criteria, Mancini Consulting Services (an industry pump expert) and Flowserve (the pump vendor) were consulted for input.

Each RHR pump motor is vertically mounted to the pump casing, with the piping entering and exiting the pump casing horizontally. The RHR pump motors weigh approximately 6500 pounds and operate at 1800 RPM. Other motor specs include: 1000 Hp, 3 Phase, 60 Hz, 4000 Volts. The pump casing, weighing 6100 pounds, is mounted on a reinforced floor pad.

The 20 inch suction piping enters the room level with the pump centerline. An additional 20 inch line tees into the suction piping approximately 5 feet from the pump. The 16 inch discharge piping leaves the pump on the same plane as suction piping but then elbows 90 degrees vertically 6 feet from the pump. This is then followed by a discharge check valve and an isolation valve on the vertical run. See figures 9 and 10 for the isometric layout of the suction and discharge piping near the RHR pumps.

Figure 11 shows the vibration monitoring points on the RHR motor/pump assembly. Points V1, V2, and Vaxial are the specified locations for IST data collection. V1 is taken in the vertical direction in line with pump flow. V2 is taken 90 degrees from V1 perpendicular to pump flow. Vaxial is taken 90 degrees from the V1 and V2 plane, on the underside of the motor.

Resonance testing was performed on all four RHR pump housings in the V1, V2, and Vaxial directions. Analysis has shown that the contributing cause of vibration in the V1 and V2 directions is from broadband peaks in the spectrum between 85.1 to 102.7 Hertz. These frequencies fall in the area of pump operation as these pumps are of a three vane design. In the case of the RHR pumps, vane pass frequency excitations are influencing vibration measurements.

III. Attempts to Lower Vibration Prior to completely understanding the cause of the vibration, it was thought that after the adjustment of testing flow in early 2009, vibrations would decrease.

Through the recent analysis of the vibration spectrum, the structural resonance and the running speed peaks can be seen. This analysis indicates that the running speed spectral peaks are consistent over years of testing. With the resonant frequency being considered as a significant contributor to exceeding the alert vibration range, options to reduce resonance reside in stiffening the pump or an internal design change (i.e. modifying to a 5 vane design).

JAF initially pursued a path to add additional stiffening to the pump and piping system. With the addition of supports, a new seismic analysis would be required for each RHR pump and the associated piping. Due to the complexity and resources needed for a new analysis, combined with the industry OE (see Fermi 2 approved submittal ML100491856) which shows that stiffening operations also lend to the potential for vibrations to be transferred to the surrounding piping, efforts were ceased.

Page 4 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii),

Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Major modifications, such as to add stiffening to the pump/motor system or changing the pump to a 5 vane design, when the pumps are not seen as degrading, are not deemed to result in an increase in the level of safety.

IV. Spectral Analysis ASME OM Code 2001 Edition/2003 Addenda section ISTB-6400 states if the reference value of a particular parameter being measured or determined can be significantly influenced by other related conditions, then these conditions shall be analyzed1 and documented in the record of tests. The footnote for analyzed states vibration measurements of pumps may be foundation, driver, and piping dependent. Therefore, if initial readings are high and have no obvious relationship to the pump, then vibration measurements should be taken at the driver, at the foundation, and on the piping and analyzed to ensure that the reference vibration levels will not prevent the pump from fulfilling its function.

Spectral analysis shows the total vibration broken down into individual frequencies over a span from 0-1500 Hz. Figures 15-18 show a recent analysis followed by a historical compilation for the data points that coincide with the IST surveillance testing. Broadband vibration is seen occurring at three times the pump running speed. These vibrations may exceed the Code alert criteria, which triggers the corrective action process and the need to increase the testing frequency.

Spectral data indicates that the overall vibration levels are primarily made up of a spectrum from 85 to 103 HZ due to the vane pass frequency induced by a 3 three vane pump at 1800 RPM. As this vibration stems from the design of the pump, all four RHR pumps are susceptible. This vibration is accentuated on 10P-3B and 10P-3C due to the similar piping configurations. Spectral data do not indicate any degradation to the bearings, pump, or motor that would lead to imbalance or misalignment.

Basis for Code Alternative Alert Values By this relief request, James A. FitzPatrick, is proposing to increase the absolute alert limit for vibration from 0.325 in/s to 0.408 in/s for all four RHR pumps in the V1 and V2 directions. The vane pass induced broadband vibration occasionally causes the overall vibration value to exceed 0.325 in/s, resulting in the pumps being placed on an increased testing frequency. A new alert acceptance criterion of 0.408 in/s coincides with the Warning level that is already developed per the Predictive Maintenance (PdM) program. The basis for the .408 in/s warning level came from the Technical Associates of Charlotte recommendations for vertical pumps. The set points recommended were 0.350 in/s and 0.525 in/s. The PdM program uses those set points as high and low criteria but the program also has two additional levels. These two additional levels split the difference between the suggested set points, resulting in 0.408 in/s and 0.466 in/s.

Expert analyses and maintenance reviews have shown that this vibration has not resulted in degradation to the pump or motor. Data trends show that overall vibrations have remained steady since 1998. These analyses and reviews demonstrate that the corrective action of doubling the testing frequency does not provide additional assurance as to the condition of the RHR pumps or the ability to perform their safety function.

Page 5 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii),

Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety The new alert criteria values allow an alternative measure that still meets the intended function of monitoring the pump for degradation, while leaving the action levels as mandated by the Code. The proposed criteria encompass the previous values that exceeded the alert level, which would eliminate the unnecessary actions associated with exceeding the Code Alert limits when the pump is not seen as degrading. Any corrective actions triggered by vibrations between 0.408 in/s to 0.7 in/s will result in the same actions as previously required when exceeding the alert limit of 0.325 in/s.

The vibration specialist at JAF routinely performs a spectral analysis on all data recorded during RHR pump inservice testing. This analysis is in addition to IST total vibration values. The analysis provides additional confidence on the ability to detect degradation at an early stage.

Each RHR Pump motor is also monitored through the Preventive Maintenance (PM) and Predictive Maintenance (PdM) programs. While these actions are intended to prevent degradation, any off-normal or unexpected conditions act as another indicator for early stages of degradation. PM and PdM activities include:

- Annual Non-Intrusive Thermography

- Annual Motor Bearing Sample

- 10 Year internal visual inspection Conclusions It has been determined through expert analysis that no internal pump or motor degradation is occurring due to the vane pass frequency induced vibrations. This phenomenon has become exposed since vibration readings have been taken in in/s. The available vibration and differential pressure data combined with the maintenance history support this conclusion.

Based on this information, James A. FitzPatrick concludes that the increased test frequency for the RHR pump motors that exceeded the 0.325 in/s Code alert limit does not provide additional information nor does it provide additional assurance as to the operational readiness of the pumps or their ability to perform their safety function.

Testing these pumps on increased frequency and performing associated corrective actions places an unnecessary burden on plant staff and resources. An alert limit of 0.408 in/s will provide margin above baseline and expected vibration values to prevent exceeding the Code alert limit due to vane pass frequency excitations.

6. Duration of Proposed Alternative This proposed alternative will be used for the remainder of the 4th 10 year interval

7. Precedents

1) Cooper Nuclear Station, Docket No. 05000298, Relief Request RP-06 dated 02-25-2004. TAC No. MB6821 (ML040560318)

2) Byron Station, Docket Nos. 05000454 & 455, Relief Request PR-2 dated 02-19-2002. TAC Nos. MB1852 & MB1853 (ML020070381)

3) Fermi Unit 2, Docket No. 05000341 Relief Request PRR-004, PRR-005, PRR-007, and PRR-010 dated 07-26-2010. TAC Nos. ME2552, ME2553, ME2554, &

ME2559 (ML101670372)

Page 6 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 1 - RHR 10P-3A 3rd and 4th interval vibration data in the V1 and V2 directions Page 7 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 2 - RHR 10P-3B 3rd and 4th interval vibration data in the V1 and V2 directions

• Note* Prior to 12/2009, the data points are inverted for V1 and V2. Database nomenclature (V1 and V2) did not match surveillance testing prior to 12/2009.

Data points were changed to match the surveillance test procedure in 12/2009, hence the inversion in the graph trend lines at that time..

Page 8 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 3 - RHR 10P-3C 3rd and 4th interval vibration data in the V1 and V2 directions

• Note* The two spikes in 10/2003 were taken when in/s readings were for information only. The data taken in mils was within the acceptance criteria at the time, so no corrective actions were taken. It is believed that the spike is an anomaly. The spike on V2 in 4/2001 was also an information only data point with no follow up actions.

This data point is not seen as a sign of degradation.

Page 9 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 4 - RHR 10P-3D 3rd and 4th interval vibration data in the V1 and V2 directions

• Note* The spikes on V1 and V2 in 3/2006, and V2 in 6/2006, were information only data point with no follow up actions. Data taken in mils was considered satisfactory. This data point is not seen as a sign of degradation.

Page 10 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 5 - RHR 10P-3A 3rd and 4th interval differential pressure readings

• Note* Downward spike in 12/2009 was due to reading an incorrect pressure indicator. Downward spike in 12/2012 was due to pressure indicator falling out of calibration.

Page 11 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 6 - RHR 10P-3B 3rd and 4th interval differential pressure readings Page 12 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 7 - RHR 10P-3C 3rd and 4th interval differential pressure readings

• Note* Downward spike in 12/2009 was due to reading an incorrect pressure indicator.

Page 13 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 8 - RHR 10P-3D 3rd and 4th interval differential pressure readings Page 14 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 9 - Isometric of Suction piping for 10P-3A and 10P-3C Page 15 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 10 - Isometric of Discharge piping for 10P-3A and 10P-3C

• Note* The pumps in the B train (10P-3B and 10P-3D) mirror this isometric, with 10P-3D in the same position as 10P-3A and 10P-3B corresponding to the 10P-3C location.

Page 16 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 11 - Vibration points for IST data Page 17 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 12 - Recent history of RHR 10P-3B in the V1 direction compared to current Alert Criteria Page 18 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 13 - Recent history of RHR 10P-3C in the V1 direction compared to current Alert Criteria Page 19 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 14 - Recent history of RHR 10P-3C in the V2 direction compared to current Alert Criteria Page 20 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 15 - B Pump Inboard Vertical - single test example Page 21 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 16 - B Pump Inboard Horizontal - single test example Page 22 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 17 - C Pump Inboard Vertical - single test example Page 23 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 18 - C Pump Inboard Horizontal - single test example Page 24 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 19 - 10P-3B Vibration data for V1 and V2 with trend lines Page 25 of 26

James A. FitzPatrick Nuclear Power Plant Inservice Testing Program 10 CFR 50.55a Request PRR-05 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(ii), Hardship or Unusual Difficulty without Compensating Increase in Level of Quality or Safety Figure 20 - 10P-3C Vibration data for V1 and V2 with trend lines Page 26 of 26