ML19072A104

Draft SE for Topical Report PWROG-17011

ML19072A104
Person / Time
Site:	99902037
Issue date:	03/19/2019
From:	NRC/NRR/DLP/PLPB
To:	Westinghouse
Drake J, NRR/DLP/PLPB, 415-8378
References
EPID L-2018-TOP-0019
Download: ML19072A104 (12)

Text

1 DRAFT SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION 2

3 TOPICAL REPORT PWROG-17011-NP, REVISION 1 4

5 UPDATE FOR SUBSEQUENT LICENSE RENEWAL: WCAP-14535A, TOPICAL REPORT ON 6

7 REACTOR COOLANT PUMP FLYWHEEL INSPECTION ELIMINATION AND WCAP-15666-A, 8

9 EXTENSION OF REACTOR COOLANT PUMP MOTOR FLYWHEEL EXAMINATION 10 11 PRESSURIZED WATER REACTOR OWNERS GROUP 12 13 WESTINGHOUSE ELECTRIC COMPANY 14 15 16

1.0 INTRODUCTION AND BACKGROUND

17 18 On May 15, 2018, the Pressurized Water Reactor Owners Group (PWROG), formerly 19 Westinghouse Owners Group or WOG, submitted Topical Report (TR) PWROG-17011-NP, 20 Revision 1, Update for Subsequent License Renewal: WCAP-14535A, Topical Report on 21 Reactor Coolant Pump Flywheel Inspection Elimination and WCAP-15666-A, Extension of 22 Reactor Coolant Pump Motor Flywheel Examination (Ref. 1), dated May 2018 (referred to as 23 the TR in the remainder of this document) for the U.S. Nuclear Regulatory Commission (NRC) 24 staff review. Further clarifying information related to this TR was submitted on August 31 and 25 December 21, 2018 (Refs. 2 and 3), by Florida Power & Light Company (FPL) in response to a 26 request for additional information (RAI) for its subsequent license renewal (SLR) application for 27 Turkey Point, Unit 3 and 4. The purpose of this TR is to extend the applicability of 28 WCAP-14535A (Ref. 4) and WCAP-15666-A (Ref. 5) to the subsequent period of extended 29 operation (SPEO), i.e., from 60 years of operation to 80 years of operation. The original 30 inspection frequency for reactor coolant pump (RCP) flywheels was specified in Regulatory 31 Guide (RG) 1.14, Revision 1, Reactor Coolant Pump Flywheel Integrity, dated August 1975 32 (Ref. 6).

33 34 Prior to 1996, plants gathered more than 20 years of operating experience and inspection 35 results, and there were no service-induced flaws identified which would affect RCP flywheel 36 integrity. Therefore, considering the inspection history, the savings in inspection cost, and the 37 reduction in personnel radiation exposure, the WOG submitted a deterministic and probabilistic 38 fracture mechanics methodology in WCAP-14535A to the NRC in January 1996 for elimination 39 of the RCP flywheel inspection. As indicated in the NRC staffs safety evaluation (SE) of 40 WCAP-14535A, the NRC staff only evaluated the stress and deterministic fracture mechanics 41 part of the methodology and approved the extension of the RCP flywheel inspection interval 42 from 40 months as specified in RG 1.14, Revision 1 to 10 years.

43 44 Subsequently in 2001, the WOG submitted and the NRC staff approved an extension of the 45 RCP flywheel inspection interval from 10 years to 20 years for the PEO based on a probabilistic 46 fracture mechanics (PFM) and risk-informed methodology (Ref. 5).

47 ENCLOSURE

1

2.0 REGULATORY EVALUATION

2 3 The function of the RCP in the reactor coolant system (RCS) of a pressurized water reactor 4 (PWR) plant is to maintain an adequate cooling flow rate by circulating a large volume of 5 primary coolant water through the RCS. Following an assumed loss of power to the RCP motor, 6 the flywheel, in conjunction with the impeller and motor assembly, provides sufficient rotational 7 inertia to assure adequate primary coolant flow during a RCP coastdown, thus maintaining 8 adequate core cooling. A concern regarding the overspeed of the RCP and its potential for 9 failure led to the issuance of RG 1.14. This RG describes a method acceptable to the NRC staff 10 for implementing the requirements of General Design Criterion 4, Environmental and Missile 11 Design Basis, of Appendix A, General Design Criteria for Nuclear Power Plants, to Title 10 of 12 the Code of Federal Regulations Part 50, Licensing of Production and Utilization Facilities.

13 14 As mentioned in the introduction, the original inspection interval was specified in RG 1.14, 15 Revision 1. However, following the NRC staff approval for WCAP-14535A and WCAP-15666-A, 16 inspection of RCP flywheels were performed using the approved inspection interval stated in the 17 SEs for WCAP-14535A and WCAP-15666-A (Refs. 4 and 5).

18 19 Once approved, the TR will provide the basis to continue a 20-year inspection interval for the 20 RCP flywheels into the SPEO.

21 22

3.0 TECHNICAL EVALUATION

23 24 3.1 The Stress and Deterministic Fracture Mechanics Methodology Evaluation 25 26 The primary regulatory position of RG 1.14, Revision 1, regarding flywheel design concerns 27 three critical speeds: (a) the critical speed for ductile failure, (b) the critical speed for non-ductile 28 failure, and (c) the critical speed for excessive deformation of the flywheel. This regulatory 29 position specifies, as a design criterion, that the normal speed of the flywheel should be less 30 than one-half of the lowest of the critical speeds, and the loss-of-coolant accident (LOCA) 31 overspeed should be less than the lowest of these three critical speeds.

32 33 Section 2.3 of this TR documents the stress and fracture mechanics analyses that were 34 performed for two bounding RCP flywheel groups to address these RG issues. They are 35 identical to those in WCAP-14535A (WCAP-14535A contains analysis results for the complete 36 RCP flywheel groups) and WCAP-15666-A. Since fatigue crack growth (FCG) is the only time-37 limited aging mechanism and the total cycle assumption of 6000 in WCAP-14535A is very 38 conservative, this cycle assumption was used in WCAP-15666-A and again in this TR, 39 considering operating experience. Consequently, the NRC staffs acceptance of the WOGs 40 evaluation in Reference 5 regarding the following to meet the RG 1.14, Revision 1, 41 requirements remains applicable to this TR:

42 43

• Material Information 44

• Analysis for Critical Speed Based on Ductile Fracture 45

• Analysis for Critical Speed Based on Non-ductile Failure 46

• Compliance with the Excessive Deformation Failure Criterion 47

• Compliance with LOCA Overspeed Criterion 48

1 3.2 The PFM Methodology Evaluation 2

3 Important results from the deterministic fracture mechanics analysis show that the critical crack 4 length is 3.1 inches for Group 1 RCP flywheels under the lowest fracture toughness assumption, 5 and the crack growth after 80 years is 0.08 inch. The relatively small crack growth could be 6 used to qualitatively justify the continued adoption of the 20-year inspection interval in the 7 SPEO. However, Westinghouse supplements the deterministic evaluation with a qualitative and 8 quantitative risk assessment using a methodology consistent with RG 1.174, "An Approach for 9 Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to 10 the Current Licensing Basis" (Ref. 7).

11 12 Analytically, the failure of the flywheel is determined by the presence of a crack large enough 13 that, if subjected to the stresses caused by a given speed, will cause the flywheel to fail. The 14 length of a flaw that will cause failure is defined as the critical flaw size, which depends on the 15 rotating speed of the flywheel and the fracture toughness of the flywheel material. Given an 16 event with a specific flywheel speed, if a crack grows from an assumed initial flaw size to the 17 critical size during the evaluation time, then the flywheel is assumed to fail. Since all 18 parameters mentioned above have some uncertainty, this TR treats them as random variables 19 with distributions and uses a PFM methodology to estimate the conditional probability of failure 20 (PoF) of the flywheel for the event. These conditional PoFs for RCP flywheels are essential 21 inputs to the risk assessment.

22 23 The PFM methodology provided in this TR employed the core Monte-Carlo simulation modules 24 with importance sampling that have been adopted and validated in the Structural Reliability and 25 Risk Assessment (SRRA) model, supporting WCAP-14572, Revision 1-NP-A, Westinghouse 26 Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical 27 Report (Ref. 8). Due to the simplicity of the RCP flywheel geometry and loading, only six 28 flywheel-specific equations needed to be incorporated into the standard subroutines of the 29 computer code to support PFM analyses in this TR. The limited number of equations needing 30 modification also made identification and correction of any errors in coding easier. Therefore, 31 evaluation of the PFM methodology for RCP flywheels is reduced to evaluation of the input 32 constants and random variables (with mean and standard deviation () for a normal distribution 33 or a median and factor for a log-normal distribution).

34 35 3.2.1 The Driving Force for an Assumed Crack 36 37 Regarding the driving force for an assumed crack, the loading in the PFM methodology is 38 reflected in two normally distributed random variables: (1) speed change per transient and 39 (2) speed for design limiting events. The first random variable corresponds to the normal speed 40 with a mean of 1200 rpm and a of 120 rpm for FCG calculations. The NRC staff found this 41 acceptable because using the steady state condition parameter (in this case, the normal speed) 42 as a mean value in the FCG calculation is a standard approach by the industry and the NRC.

43 Further, the assumption that is 10 percent of the mean is commonly used when actual data 44 are not available for a statistical analysis. The second random variable corresponds to the 45 design limiting speed with a mean of 1500 rpm and a of 150 rpm for the first three events 46 identified in Table 3-7 of the TR and a mean of 3321 rpm for the fourth event identified in the 47 table. The NRC staff approved 1500 rpm as the design speed and 3321 rpm as the peak speed 48 for the double ended guillotine break (DEGB) LOCA event in WCAP-15666-A (Ref. 5). This 49 acceptance remains valid during the SPEO because these speeds are not time dependent.

50 Likewise, using 10 percent of the mean value as the value for the limiting event speed is 51 acceptable as stated above. The corresponding speeds for Calvert Cliffs are 1125 rpm and

1 1368 rpm. Since the TR is not responsible for the accuracy of its reported plant-specific data, 2 these speeds should be confirmed in the future Calvert Cliffs SLR application. This is further 3 discussed in Section 4.0.

4 5 The randomly selected speed, when combined with a randomly selected initial flaw length, 6 adjusted for crack growth based on a randomly selected crack growth rate (CGR), would 7 determine the change of stress intensity factor (K) per transient for the fatigue crack growth 8 evaluation and the applied K for the critical flaw evaluation. This TR used a log-normally 9 distributed initial flaw length with a median of 0.1 inch and an uncertainty factor of 2.153. These 10 are acceptable to the NRC staff because they are more conservative than the values used in 11 the SRRA program, supporting WCAP-14572, Revision 1-NP-A.

12 13 For FCG calculations, the following CGR equation is used, 14 15 da/dN = C0(K)n, 16 17 where C0 is treated as a log-normally distributed random variable with a median of 9.95x10-11 18 and an uncertainty factor of 1.414. The exponent n is treated as a constant with a value of 19 3.07, consistent with the ASME Code FCG equation for ferritic steels in an air environment. The 20 ASME Code FCG equation represents best-estimate values. However, the TR used one half of 21 the C0 in the ASME Code FCG equation as the median for this random variable and, thus, 22 underestimated the CGR by 50 percent (correct CGR = 2 x TR CGR). The NRC staff evaluated 23 and found that this non-conservatism is completely compensated for by considering the shrink-24 fit stresses, which were neglected in the TR CGR calculations. The response to an RAI 25 documented in WCAP-14535A (Ref. 4) indicated that, if the shrink-fit stresses were considered, 26 the K and the CGR would be significantly reduced due to decreased stresses at high speed 27 caused by reduced shrink-fit and increased stresses at zero speed caused by full shrink-fit.

28 Therefore, using the TR CGR would overestimate the CGR by 150 percent (correct CGR = 0.4 x 29 TR CGR). The NRC staff found that the TR CGR is acceptable because the net effect of 30 underestimating C0 and overestimating K will result in an overestimation of the CGR by 25 31 percent (correct CGR = 2 x 0.4 TR CGR). Consequently, the NRC staff determined that the 32 FCG methodology that was used to generate the conditional PoF values presented in Table 3-3 33 and Tables 3-6 to 3-9 of the TR is acceptable.

34 35 Another random variable related to FCG is the number of transients per operating cycle. The 36 TR used a mean of 100 transients and a of 10 transients for this random variable. This is very 37 conservative and acceptable because a plant is unlikely to experience 100 transients in an 38 operating cycle which would require frequent RCP starts and stops.

39 40 3.2.2 Fracture Resistance for an Assumed Crack 41 42 Regarding the fracture resistance of the flywheel to a crack after 80 years of growth, the PFM 43 methodology determines the KIC value using two normally distributed random variables:

44 (1) operating temperature of the flywheels and (2) nil-ductility transition reference temperature 45 (RTNDT) value of the flywheel material. The KIC value can then be determined by these two 46 variables using the ASME Code Appendix A KIC equation. The TR used a mean of 95 °F for 47 Westinghouse plants (70 °F for Calvert Cliffs) and a of 12.5 °F for the flywheel operating 48 temperature. The TR states that the containment building temperature is typically 100 °F to 49 120 °F. Therefore, using the above mean and for the flywheel ambient temperature would be 50 conservative and acceptable because (1) 95 percent of the time, the randomly selected ambient 51 temperature would be between 70°F (mean - 2) and 120°F (mean + 2) for Westinghouse

1 plants, giving lower KIC values for the biased lower temperatures; and (2) the corresponding 95 2 percent ambient temperature range is between 45 °F and 95°F for Calvert Cliffs, giving an even 3 lower KIC values. The TR used a mean of 30 °F and a of 17 °F for the RTNDT value. Using 4 these values are conservative, considering that the first flywheel specification dated 1969 5 requires the RTNDT values from both longitudinal and transverse Charpy specimens be less than 6 10 °F. To supplement this determination, the NRC staff examined the initial RTNDT values for 7 SA-533 Grade B material for reactor pressure vessels (RPVs) fabricated before 1975 8 (approximately the same vintage) and found the highest initial RTNDT value for this RPV material 9 is 60 °F, supporting the TRs selection of mean and for the RTNDT value.

10 11 The TR provides no description for the crack initiation toughness factor (F-KIC), which is 12 another random variable with a mean value of 1.0 and a of 0.1. In the PFM analysis, the 13 fracture toughness KIC can be determined by the randomly selected RTNDT value and the 14 operating temperature of the flywheel. Therefore, F-KIC must be related to the use of the 15 ASME Code KIC curve. Since the ASME Code KIC curve is a 95 percent lower bound curve, 16 applying F-KIC to it, or a mean curve based on the ASME Code KIC curve, is acceptable.

17 Reference 3 confirmed that the TR used the mean KIC curve.

18 19 3.2.3 Constant Variables Related to Inspections 20 21 Except for the flywheel inner radius, outer radius, and the FCG exponent, all other constant 22 variables are related to inspections. The TR used 0.1 for the probability of a flaw existing after 23 the preservice inspection. This is conservative because the preservice inspection was 24 performed to detect and to repair all relevant indications and it is unlikely to miss a flaw 25 emanating from the keyway which is 0.1 inch radially and 6.5 inches through-thickness.

26 27 The TR used 3 for the operating cycles for the first ISI and 4 for the operating cycles between 28 ISIs. Since Table 3-3 of the TR used ISI at 4-year intervals to characterize the PFM results, it 29 is clear that the operating cycle is conservatively set to one year in the PFM analyses. Using 30 these ISI parameters for the first 10 years of operation is consistent with RG 1.14, Revision 1.

31 The risk increase based on the PFM results between the base case having ISIs at the entire 32 80 years (20 ISIs) and the proposed case having ISIs for only the first 10 years (about 3 ISIs) is 33 evaluated in the risk assessment in Section 3.3 of the TR. This assumption is very conservative 34 because it maximized the difference in the number of ISIs between the two cases. Based on 35 this, the NRC staff determined that there is additional margin in the conditional PoF difference 36 between these two cases.

37 38 In summary, the NRC staff determined that the PFM methodology that was used to generate the 39 conditional PoF values presented in Table 3-3 and Tables 3-6 to 3-9 of the TR is acceptable.

40 41 3.3 Risk Assessment 42 43 The quantitative risk assessment discussed in the TR provides the justification for applying the 44 WCAP-15666-A 20-year flywheel inspection interval for 80 years of operation. Specifically, the 45 risk analyses confirms that applying the inspection extension to flywheels in operation up to 46 80 years has a negligible impact on risk (core damage frequency (CDF) and large early release 47 frequency (LERF)), i.e., it is within the risk acceptance criteria of RG 1.174. Section 3 of the TR 48 provides a discussion on the requirements of RG 1.174, and extends the previous flywheel 49 failure probability assessment in WCAP-15666-A to 80 years of operation.

50

1 The risk evaluation includes the likelihood that a crack will grow large enough to cause failure at 2 the following conditions or events: (1) normal plant operation resulting in a plant trip, (2) a 3 transient or LOCA event with no loss of electrical power to the RCP, (3) a transient or LOCA 4 event with loss of electric power to the RCP motor, and (4) after the DEGB LOCA with the 5 simultaneous loss of power to the RCP motor.

6 7 Section 3.4.3 of the TR provides descriptions and frequencies of the initiating events for the 8 different conditions listed in Table 3-5. The NRC staff noted apparent inconsistencies between 9 TR Tables 3-5, 3-7, and 3-8 and WCAP-15666-A, Tables 3-12 and 3-13 regarding conditions for 10 LOCA events versus non-LOCA events. In the RAI response to a SLR application for Turkey 11 Point Units 3 and 4 (Ref. 2), FPL clarified that the two documents represent the same conditions 12 and employ the same analysis assumptions. Furthermore, it clarified that all entries are 13 bounded by the design limiting transient. The NRC staff finds this clarification acceptable and 14 no changes are needed to further clarify information in the TR.

15 16 The risk assessment requires an estimate of the conditional PoF of the flywheel at a given event 17 (speed), an estimate of the event frequency, and the change in CDF and LERF given a flywheel 18 failure.

19 20 3.3.1 Flywheel Conditional Failure Probability 21 22 The method for calculating flywheel failure probabilities is based on the method described in 23 section 3.3.1 of the TR and evaluated in Section 3.2 of this SE. In the evaluation, the NRC staff 24 also relied on the more detailed information in WCAP-15666-A.

25 26 3.3.2 Core Damage Evaluation 27 28 The failure of the RCP motor flywheel during normal plant operation would directly result in a 29 reactor trip. However, the potential indirect or spatial effects associated with a postulated 30 flywheel failure present a greater challenge in terms of failure effects or consequences. The 31 flywheel has the potential to catastrophically fail, resulting in flywheel fragments, which are 32 essentially high energy missiles, which could impact other structures, systems, and components 33 (SSCs) important to plant safety. Failure of these other SSCs could potentially impact the 34 overall plant safety in terms of core damage (e.g., as a result of the loss of safety injection) or 35 large early release (as a result of potential impacts on containment structures or systems).

36 37 In order to address plant specific design differences on a generic basis, the risk assessment in 38 the TR conservatively assumed that failure of the RCP motor flywheel results in core damage 39 and a large early release, i.e., the flywheel failure frequency is equal to CDF and LERF.

40 Section 3.3 of the TR discusses the process for estimating the likelihood of the primary failure 41 mode of the RCP motor flywheel. Section 3.4 of the TR then combines this failure probability 42 estimation with the likelihood of various plant events and consequences to estimate the change 43 in risk for continuing the 20-year examination interval for RCP flywheels in the SPEO.

44 45 To investigate the consequences of RCP overspeed, the TR analyzed a spectrum of LOCA 46 events resulting in a range of flywheel transients. Results of this analysis indicated that the 47 limiting event was the DEGB with an instantaneous loss of power, this led to a peak flywheel 48 speed of 3321 rpm. It was also noted that the 3 ft2 break area case showed a decrease in 49 speed such that the normal operating speed is not exceeded. Based on the WCAP-15666-A 50 assessments, the following scenarios are associated with the primary mode of potential failure

1 in the Westinghouse RCP motor flywheel that are related to operating speed and potential 2 overspeed during various conditions:

3 4

• Failure during normal plant operation resulting in a plant trip (1200 rpm peak speed) 5 6

• Failure of the RCP motor flywheel associated with a plant transient or LOCA event with 7 no loss of electrical power to the RCP (1200 rpm peak speed) 8 9

• Failure of the RCP motor flywheel associated with a plant transient or LOCA event up to 10 3 ft2 with an instantaneous loss of electrical power to the RCP (1200 rpm peak speed) 11 12

• Failure of the RCP motor flywheel associated with a DEGB coincident with an 13 instantaneous loss of electrical power, such as loss offsite power (3321 rpm peak 14 speed). This case bounds and is conservatively applied to all flywheel transients for 15 LOCA break areas.

16 The TR states that the normal operating speed of the flywheel is 1189 rpm with a synchronous 17 speed of 1200 rpm. If there is a pipe rupture in the RCPs outlet piping, the high reactor coolant 18 pressure will force reactor coolant out through the RCP into the low pressure containment 19 structure and hydraulic torque would be applied to the shaft in the direction of increasing shaft 20 speed. If electrical power is maintained to the RCP motor, the motor will function as a dynamic 21 break and limit the increase in speed of the shaft to less than 1500 rpm. If, however, electric 22 power is lost to the RCP, the flywheel will accelerate. The maximum estimated flywheel speed 23 is 3321 rpm for a DEGB, the largest possible break in the RCP outlet piping with simultaneous 24 loss of electric power to the RCP motor.

25 26 The TR evaluated the likelihood of a flywheel developing a critical crack for the overspeed of 27 3321 rpm and the overspeed of 1500 rpm. The speed of 3321 rpm is used for the DEGB with 28 loss of electric power to the RCP motor (Event 4). The speed of 1500 rpm is used for all other 29 scenarios (Events 1 to 3). The probability of cracks reaching the critical sizes given different 30 inservice inspection (ISI) programs was estimated using the PFM methodology discussed in 31 Section 3.2. In order to develop estimates applicable to the current fleet of reactors, the TR 32 conservatively estimated the impact of the requested 20-year inspection interval by assuming 33 that, aside from the initial inspections during the first 10 years of plant life, there will be no more 34 inspections over the operating life of the units. The base case for comparison assumed ISIs for 35 the entire 80 years. As discussed in Section 3.2.3, this conservative evaluation is acceptable.

36 Given these four events, the conditional PoFs of the flywheel from the PFM results are 37 presented in Table 3-3 of the TR.

38 39 The staff noted that the PoF values in Table 3-3 for the Groups 1 and 2 flywheels are nearly 40 identical to those from WCAP-15666-A for 60 years of operation, but slightly lower. This is 41 reasonable because the PFM analyses in both TRs used 100 fatigue cycles per year. When ISI 42 was performed every 4 years for 60 and 80 years, the total number of ISIs is 15 and 20, 43 respectively. The slightly lower PoF for 80 years reflects the benefit from 5 additional ISIs.

44 However, the NRC staff also found that the TR needs to explain why the PoFs for the case with 45 ISI performed every 4 years for only the first 10 years in 80 years of operation are lower than 46 the corresponding PoFs in WCAP-15666-A for 60 years. Reference 3 indicated that through a 47 verification and validation process, Westinghouse found that changes were made to the core 48 PFM executable program that produced PoF results in the TR. These changes were not

1 managed through configuration control and resulted in calculations that should be discarded.

2 The Westinghouse effort also discovered a mistake in the mean KIC equation in the PFM 3 calculation. Therefore, Westinghouse provided revised PoF values to replace the values in 4 Table 3-3 of the TR. The NRC staff reviewed Reference 3 and considers the explanation of the 5 loss of configuration control and the discovery of a mistake in the core PFM program credible 6 and the corrective actions appropriate. Therefore, the revised PoF values can be used in the 7 subsequent risk evaluation. For clarity, the NRC staff reproduced the revised PoF values in the 8 following table.

9 10 Flywheel Group Maximum Speed Cumulative Probability of Flywheel Failure over 801 years reported in the TR Continuously With ISI With ISI at 4-Year at 4-Year Intervals Intervals Prior to 10 Years and Without ISI after 10 years Group 1 1500 RPM 1.99E-08 2.02E-08 Group 1 3321 RPM 5.88E-02 5.88E-02 Group 2 1500 RPM 1.26E-08 1.37E-08 Group 2 3321 RPM 1.66E-02 1.66E-02 11 12 13

• All PoF values are from Reference 3 which supersede the values in Table 3-3 of the TR.

14 15 The peak speed of the flywheel during normal operation (Event 1) is about 1200 rpm. Because 16 the PFM results shown in Figures 3-6 to 3-8 of Reference 5 indicated that the failure probability 17 of the RCP flywheels is not changing much with time, the PoF under normal operation during 18 any given year can be reasonably approximated by the cumulative probability at the end of 19 80 divided by 80 years.

20 21 The failure of a flywheel is an irreversible change of state, i.e., the flywheel cannot be repaired 22 and returned to service and there is no likelihood that more than one flywheel failure may occur 23 during the operating lifetime. Without the possibility of multiple failures during the operating 24 lifetime, the assumption - that the probability that a critical flaw exists prior to each transient is 25 equal to the probability that a critical flaw will develop by the end of the 80 year operating life -

26 is conservative and acceptable.

27 28 Rather than attempt to develop a full spectrum of break sizes, overspeeds, and associated 29 critical sizes, Table 3-5 in the TR stated that LOCAs up to a three square foot break in the 30 primary loop with a loss of electric power to the RCPs would not cause the flywheel to exceed 31 1500 RPM. Therefore, the frequency of LOCAs up to this size are grouped together with the 32 transient frequency.

33 1

The reported cumulative probability is the probability that the flywheel will fail at the indicated RPM. The flywheel will only reach 3321 RPM if there is a LBLOCA/LOOP event.

1 The TR estimated the frequency of large break LOCA (LBLOCA) events with break areas in 2 excess of three square foot break using NUREG-1829 and, therefore, the NRC staff finds that 3 the LOCA frequency is appropriate for use in support of this submittal.

4 5 A DEGB must be accompanied by a simultaneous loss of power to the RCP motor in order for 6 the flywheel speed to exceed 1200 rpm. Loss of power to the RCP motor is most likely caused 7 by a loss of station power caused by transfer from the offsite electrical grid to the onsite 8 emergency electrical grid, and failure of the emergency grid to properly load and operate. The 9 probability of the loss of station power is dependent on the LOCA because the changing 10 electrical configuration and loads induced by the LOCA may cause the loss of power.

11 Evaluation of the potential for loss of station power indicates that a reasonable estimate for the 12 probability of loss of station power following a LOCA is about 1.4E-2 (Ref. 9). The TR uses this 13 conditional probability.

14 15 The above discussion regarding Westinghouse RCP motor flywheels applies to Calvert Cliffs 16 flywheels also, but with the reduced overspeeds specific to Calvert Cliffs (i.e., 1125 RPM and 17 1368 RPM).

18 19 The NRC staff noted, however, that Tables 3-7 to 3-9 of the TR show that the event frequency 20 for the fourth condition is 1.4E-8/year. In WCAP-15666-A, Table 3-12, the corresponding event 21 frequency is 2.8E-8/year based on a maximum LOCA frequency (LOCAs with greater than 22 5000 gpm blowdown) of 2E-6/year and the probability of loss of station power following a LOCA 23 of 1.4E-2/year. In the RAI response to the SLR application for Turkey Point, Units 3 and 4 24 (Ref. 2), FPL indicated that the frequencies of LBLOCA events with break areas in excess of 25 3 ft2 (fourth condition) reported in WCAP-15666-A were based on Westinghouse fracture 26 mechanics calculations performed prior to NRC issuance in 2008 of NUREG-1829 (Ref. 10).

27 The frequencies of LBLOCA events with break areas in excess of 3 ft2 estimated in the TR were 28 updated based on NUREG-1829 and the mean failure rates associated with the larger LOCA 29 break sizes presented in Table 7.19 of NUREG-1829. NUREG-1829 states that, The results in 30 Table 7.19 are appropriate to use for PRA [probabilistic risk assessment] applications that 31 separately consider SGTRs [steam generator tube ruptures]. Specifically, in establishing the 32 large break frequency, the 14 inch and 31 inch diameter breaks were extrapolated to 80 years 33 and interpolated to determine a cumulative frequency for a 3 ft2 break. The NRC staff finds the 34 use of NUREG-1829 acceptable because it is consistent with current probabilistic risk 35 assessment practices and provides LOCA frequencies as a function of break size.

36 37 Consequence Estimate 38 39 The flywheel has the potential to catastrophically fail, resulting in flywheel fragments which are 40 essentially high energy missiles that could impact other SSCs important to plant safety. The TR 41 states that the initial investigations indicate that there is not much uniformity with respect to the 42 layout of critical targets that potential flywheel fragments could impact given its failure and, 43 therefore, a generic damage scenario is difficult to develop. The TR assumes that a flywheel 44 failure would lead directly to core damage and large early release. Therefore, the adequacy of 45 a generic scenario and the quality of the probabilistic risk assessment analysis used to support 46 the TR methodology are not issues and the consequence evaluation is acceptable.

47 48 Risk Estimates 49 50 The detailed results for all the scenarios for each of the three flywheel types are provided in 51 Tables 3-7 to 3-9 of the TR. Since revised Tables 3-7 to 3-9 are not provided in Reference 3,

1 the NRC staff used the revised PoF values in Reference 3 (reproduced in the above table) and 2 followed the same calculation procedures in Tables 3-7 to 3-9 of the TR to generate the final 3 CDFs for Groups 1 and 2 RCP flywheels in the following table to replace the outdated values in 4 Tables 3-7 to 3-9, where each CDF is the sum of the contributions from all four scenarios. It 5 should be noted, however, that the outdated CDFs for Calvert Cliffs RCP flywheels in 6 Tables 3-7 to 3-9 are provided in the table for reasons to be explained below.

7 8

Calvert Cliffs*

Group 1 Group 2 1.54E-08/year CDF and LERF with 2.125E-08 1.317E-08 ISI after 10 years 1.557E-8/year CDF and LERF 2.156E-08 1.430E-08 without ISI after 10 years 1.30E-10/year Increase in CDF and 3.10E-10 1.13E-09 LERF for one flywheel 5.20E-10/year Increase in CDF and 1.24E-09 4.52E-09 LERF for four flywheels 9

10

• CDF values for Calvert Cliffs are from Reference 1. CDF values for Groups 1 and 2 are based 11 on PoF values from Reference 3.

12 13 As can be seen in the table, the bounding estimated increase in risk is 4.52E-09/year for 14 Group 2 flywheels. This estimate is well below the very small change in LERF guideline of 15 1E-7/year in RG 1.174. The NRC staff noted that the revised CDF values for Groups 1 and 2 16 based on PoF values from Reference 3 are less than the corresponding values in Reference 1.

17 This indicated that if the revised PFM program in Reference 3 is used to generate the CDF 18 values for Calvert Cliffs, the CDF values will also be less than those in Reference 1, establishing 19 that the estimate for Calvert Cliffs is also well below the very small change in LERF guideline of 20 1E-7/year in RG 1.174.

21 22 4.0 LIMITATIONS AND CONDITIONS 23 24 There is no limitation or condition for the Westinghouse RCP flywheels, considering that the 25 flywheel operating and material data used in the generic analyses have already been 26 examined twice and accepted during plant-specific applications using WCAP-14535A and 27 WCAP-15666-A. However, the flywheel operating and material data used in the generic 28 analyses for Calvert Cliffs in this TR are new and need to be confirmed by the licensee:

29 30

• the normal operating speed for the RCP flywheels is 900 rpm 31

• the design limiting speed for the RCP flywheels is 1125 rpm 32

• the maximum overspeed following a design basis LOCA is 1368 rpm 33

• it is appropriate to use 70 °F as the medium temperature for design limiting event 34 (Table 3-2) in the PFM analysis

1 The TR requires applicants of this TR to confirm that 6000 cycles for 80 years of operation is 2 applicable on a plant-specific basis. This confirmation shall be made in all SLR applications to 3 fulfill the TR requirement. Please note that TR requirements are not considered as SE 4 limitations and conditions.

5 6

5.0 CONCLUSION

S 7

8 The change in risk estimate includes numerous conservative assumptions including:

9 10

• The use of the 1500 rpm for 1189 and 1200 rpm scenarios for Westinghouse RCP 11 flywheels and 1125 rpm for 900 rpm scenarios for Calvert Cliffs.

12 13

• The use of the probability that a critical crack exists at the end of the 80 year life as the 14 probability that the crack would exist during each operating year.

15 16

• The use of 100 start-ups and shutdowns per calendar year when simulating the FCG.

17 18

• Maximizing the difference in number of ISIs between the base case and the proposed 19 case.

20 21

• Characterizing the DEGB flow rate as 5000 gpm or higher.

22 23

• The failure of the flywheel will cause core damage and a large early release event with a 24 probability of 1.

25 26 Considering all the conservative assumptions listed above, the NRC staff determined that the 27 increase in CDF and LERF values for three groups of flywheels provide a bounding estimate of 28 the change in risk associated with the continued adoption of the 20-year inspection interval from 29 the PEO to the SPEO. The bounding estimate is below the very small change in LERF 30 guidelines in RG 1.174, and the NRC staff finds that the increase in risk is small and is 31 consistent with the Commissions Safety Goal Policy Statement.

32 33 The TR also addresses the other key principles of risk-informed licensing actions described in 34 RG 1.174. There are no changes to the evaluation of design basis accidents and the margin of 35 safety is being maintained. Nondestructive examinations and inspections will continue to be 36 conducted every 20 years. Therefore, the NRC staff finds the requested change to be well-37 defined, consistent with defense-in-depth philosophy, contains adequate margin of safety, and 38 incorporates a performance measurement strategy to monitor the change. The NRC staff also 39 finds that the risk evaluation is consistent with the risk-informed methodology and guidelines 40 described in RG 1.174 and that the potential change in risk caused by the continued adoption of 41 the 20-year inspection interval from the PEO to the SPEO is small and acceptable.

42 43 The request is a change from the current RG 1.14, Revision 1 guidance. The NRC staff finds 44 that the regulatory positions in RG 1.14, Revision 1 concerning the three critical speeds are 45 satisfied, and that the TR evaluation indicating that critical crack sizes are not expected to be 46 attained during the 20-year inspection interval in the SPEO is reasonable and acceptable. The 47 potential for failure of the RCP flywheel is, and will continue to be, negligible during normal and 48 accident conditions.

49

1

6.0 REFERENCES

2 3 1. Letter from Ken Schrader, Pressurized Water Reactor Owners Group, to USNRC, 4 May 15, 2018, enclosing Westinghouse Electric Company Report PWROG-17011-NP, 5 Rev. 1, Update for Subsequent License Renewal: WCAP-14535A, Topical Report on 6 Reactor Coolant Pump Flywheel Inspection Elimination, and WCAP-15666-A, 7 Extension of Reactor Coolant Pump Motor Flywheel Examination Non-Proprietary 8 Class 3, dated May 2018 (Agencywide Documents Access and Management System 9 (ADAMS) Package Accession No. ML18143B465).

10 11 2. Letter from William Maher, Florida Power & Light Company, to USNRC, August 31, 12 2018, enclosing Florida Power & Light Company response to NRC request for additional 13 information Set 1 regarding Turkey Point Units 3 and 4 subsequent license renewal 14 application (ADAMS Accession No. ML18248A257).

15 16 3. Letter from William Maher, Florida Power & Light Company, to USNRC, December 21, 17 2018, enclosing Florida Power & Light Company response to NRC request for additional 18 information Set 5 Response 4.3.5-2 Revision regarding Turkey Point Units 3 and 4 19 subsequent license renewal application (ADAMS Accession No. ML18362A146).

20 4. WCAP-14535A, Topical Report on Reactor Coolant Pump Flywheel Inspection 21 Elimination, November 1996 (ADAMS Accession No. ML18312A151).

22 23 5. WCAP-15666-A, Extension of Reactor Coolant Pump Motor Flywheel Examination, 24 October 2003 (ADAMS Accession No. ML18303A413).

25 26 6. U.S. Nuclear Regulatory Commission, Reactor Coolant Pump Flywheel Integrity, 27 Regulatory Guide 1.14, Revision 1, August 1975 (ADAMS Accession 28 No. ML003739936).

29 30 7. U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk 31 Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing 32 Basis, Regulatory Guide 1.174, Revision 3, January 2018 (ADAMS Accession 33 No. ML17317A256).

34 35 8. WCAP-14572, Revision 1-NP-A, Westinghouse Owners Group Application of Risk-36 Informed Methods to Piping Inservice Inspection Topical Report, February 1999 37 (ADAMS Accession No. ML042610469).

38 39 9. NUREG/CR-6538, Evaluation of LOCA with Delayed LOOP and LOOP with Delayed 40 LOCA Accident Scenarios, July 1997 (ADAMS Accession No. ML071630062).

41 42 10. NUREG-1829, Estimating LOCA Frequencies Through the Elicitation Process, April 43 2008 (ADAMS Accession Nos. ML081060300 and ML082250436).

44 45 Principal Contributors: S. Sheng 46 J. Dozier 47 48 Date: March 19, 2019