ML103610140

From kanterella
Revision as of 13:16, 11 March 2020 by StriderTol (talk | contribs) (StriderTol Bot insert)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
Response to U.S. EPR Design Certification Application RAI No. 422, Supplement 5
ML103610140
Person / Time
Site: Framatome
Issue date: 12/22/2010
From: Sloan S
AREVA NP
To:
Document Control Desk, Office of New Reactors
References
NRC:10:116
Download: ML103610140 (50)
v • d • e


Text

AR:E VA December 22, 2010 NRC:10:116 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Response to U.S. EPR Design Certification Application RAI No. 422, Supplement 5 Ref. 1: E-mail, Getachew Tesfaye (NRC) to Martin C. Bryan (AREVA NP Inc.), "U.S. EPR Design Certification Application RAI No. 422 (4792), FSAR Ch. 3," August 3, 2010.

Ref. 2: E-mail, Martin C. Bryan (AREVA NP Inc.) to Getachew Tesfaye (NRC), "Response to U.S.

EPR Design Certification Application RAI No. 422, FSAR Ch. 3," September 2, 2010.

Ref. 3: E-mail, Martin C. Bryan (AREVA NP Inc.) to Getachew Tesfaye (NRC), "Response to U.S.

EPR Design Certification Application RAI No. 422, Supplement 1, FSAR Ch. 3," September 9, 2010.

Ref. 4: E-mail, Martin C. Bryan (AREVA NP Inc.) to Getachew Tesfaye (NRC), "Response to U.S.

EPR Design Certification Application RAI No. 422, Supplement 2, FSAR Ch. 3," September 27, 2010.

Ref. 5: E-mail, Martin C. Bryan (AREVA NP Inc.) to Getachew Tesfaye (NRC), "Response to U.S.

EPR Design Certification Application RAI No. 422, Supplement 3, FSAR Ch. 3," November 2, 2010.

Ref. 6: E-mail, Martin C. Bryan (AREVA NP Inc.) to Getachew Tesfaye (NRC), "Response to U.S.

EPR Design Certification Application RAI No. 422, Supplement 4, FSAR Ch. 3," November 22, 2010.

In Reference 1, the NRC provided a request for additional information (RAI) regarding the U.S. EPR design certification application. Reference 2 provided a schedule for technically correct and complete responses to RAI No. 422. Reference 3 provided technically correct and complete responses to two of the 61 questions. Reference 4 provided a revised schedule for technically correct and complete responses. Reference 5 provided a revised schedule for technically correct and complete responses to 6 of the remaining questions. Reference 6 provided a revised schedule for technically correct and complete responses to 16 of the remaining 61 questions to allow additional time to interact with NRC.

The enclosed response provides technically correct and complete responses to 11 of the remaining 61 questions. AREVA NP considers some of the material contained in the attached response to be proprietary. As required by 10 CFR 2.390(b), an affidavit is attached to support the withholding of the information from public disclosure.

An AREVA:and Siemens company -r.

3j315 'Old For~s't Road. P.O. B~ox 10935, Lynchburg, VA 24506-0935 Tel.: '(4'34) 832-3000 F"x. (434) 832-3840 FOR.M41979VA:10111200O6)

Document Control Desk Page 2 NRC:10:1 16 The following table indicates the respective pages in the enclosed response that contain AREVA NP's response to the subject questions.

Question # Start Page End Page RAI 422 - 03.09.02-125 2 3 RAI 422 - 03.09.02-128 4 5 RAI 422 - 03.09.02-129 6 7 RAI 422 - 03.09.02-130 8 9 RAI 422 - 03.09.02-132 10 12 RAI 422 - 03.09.02-133 13 13 RAI 422 - 03.09.02-135 14 15 RAI 422 - 03.09.02-136 16 18 RAI 422 - 03.09.02-137 19 19 RAI 422 - 03.09.02-139 20 22 RAI 422 - 03.09.02-141 23 25 The schedule for the technically correct and complete responses to the remaining 50 questions is unchanged and is provided below.

Question # Response Date RAI 422 - 03.09.02-82 February 24, 2011 RAI 422 - 03.09.02-84 February 24, 2011 RAI 422-- 03.09.02-85 February 24, 2011 RAI 422- 03.09.02-87 January 12, 2011 RAI 422 - 03.09.02-88 January 12, 2011 RAI 422- 03.09.02-89 January 12, 2011 RAI 422 - 03.09.02-90 January 12, 2011 RAI 422- 03.09.02-91 January 12, 2011 RAI 422 - 03.09.02-92 January 12, 2011 RAI 422 - 03.09.02-93 January 12, 2011 RAI 422 - 03.09.02-94 January 12, 2011 RAI 422 - 03.09.02-95 January 12, 2011 RAI 422 - 03.09.02-96 January 12, 2011 RAI 422 - 03.09.02-97 January 12, 2011 RAI 422 - 03.09.02-98 February 24, 2011 RAI 422 - 03.09.02-99 February 15, 2011 RAI 422 - 03.09.02-100 February 15, 2011 RAI 422 - 03.09.02-101 February 15, 2011 RAI 422 - 03.09.02-102 February 15, 2011 RAI 422 - 03.09.02-103 February 15, 2011 RAI 422 - 03.09.02-104 February 15, 2011 RAI 422 - 03.09.02-105 February 15, 2011 RAI 422 - 03.09.02-106 February 15, 2011 RAI 422 - 03.09.02-107 February 15, 2011 RAI 422 - 03.09.02-108 February 15, 2011 RAI 422 - 03.09.02-109 February 24, 2011 RAI 422 - 03.09.02-110 February 15, 2011

Document Control Desk Page 3 NRC:10:116 Question # Response Date RAI 422 - 03.09.02-111 February 15, 2011 RAI 422 - 03.09.02-112 February 15, 2011 RAI 422-03.09.02-113 February 24, 2011 RAI 422 - 03.09.02-114 February 24, 2011 RAI 422 - 03.09.02-115 February 24, 2011 RAI 422 - 03.09.02-116 February 24, 2011 RAI 422 - 03.09.02-117 February 24, 2011 RAI 422- 03.09.02-118 February 24, 2011 RAI 422-03.09.02-119 February 24, 2011 RAI 422--03.09.02-120 February 24, 2011 RAI 422--03.09.02-121 February 24, 2011 RAI 422 - 03.09.02-122 February 24, 2011 RAI 422 - 03.09.02-123 February 24, 2011 RAI 422 - 03.09.02-124 January 20, 2011 RAI 422 - 03.09.02-126 January 20, 2011 RAI 422 - 03.09.02-127 January 20, 2011 RAI 422 - 03.09.02-131 February 24, 2011 RAI 422 - 03.09.02-134 January 20, 2011 RAI 422- 03.09.02-138 January 12, 2011 RAI 422 - 03.09.02-140 January 20, 2011 RAI 422 - 03.09.02-142 January 12, 2011 RAI 422 - 03.09.02-144 February 24, 2011 RAI 422 - 03.09.02-145 February 24, 2011 If you have any questions related to this submittal, please contact me by telephone at 434-832-2369 or by e-mail to sandra.sloan*,areva.com.

Sincerely, Sandra M. Sloan, Manager New Plants Regulatory Affairs AREVA NP Inc.

Enclosures cc:" G. Tesfaye Docket No.52-020

AFFIDAVIT COMMONWEALTH OF VIRGINIA )

) ss.

COUNTY OF CAMPBELL )

1. My name is Sandra M. Sloan. I am Manager, Regulatory Affairs for New Plants, for AREVA NP Inc. and as such I am authorized to execute this Affidavit.
2. I am familiar with the criteria applied by AREVA NP to determine whether certain AREVA NP information is proprietary. I am familiar with the policies established by AREVA NP to ensure the proper application of these criteria.
3. l am familiar with the AREVA NP information contained in letter NRC:10:116, "Response to U.S. EPR Design Certification Application RAI No. 422, Supplement 5," and referred to herein as "Document." Information contained in this Document has been classified by AREVA NP as proprietary in accordance with the policies established by AREVA NP for the control and protection of proprietary and confidential information.
4. This Document contains information of a proprietary and confidential nature and is of the type customarily held in confidence by AREVA NP and not made available to the public. Based on my experience, I am aware that other companies regard information of the kind contained in this Document as proprietary and confidential.
5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding of proprietary information is made in accordance with 10 CFR 2.390. The information for which withholding from disclosure is

requested qualifies under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information".

6. The following criteria are customarily applied by AREVA NP to determine whether information should be classified as proprietary:

(a) The information reveals details of AREVA NP's research and development plans and programs or their results.

(b) Use of the information by a competitor would permit the competitor to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive advantage for AREVA NP.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive advantage for AREVA NP in product optimization or marketability.

(e) The information is vital to a competitive advantage held by AREVA NP, would be helpful to competitors to AREVA NP, and would likely cause substantial harm to the competitive position of AREVA NP.

The information in the Document is considered proprietary for the reasons set forth in paragraphs 6(b) and 6(c) above.

7. In accordance with AREVA NP's policies governing the protection and control of information, proprietary information contained in this Document has been made available, on a limited basis, to others outside AREVA NP only as required and under suitable agreement providing for nondisclosure and limited use of the information.
8. AREVA NP policy requires that proprietary information be kept in a secured file or area and distributed on a need-to-know basis.
9. The foregoing statements are true and correct to the best of my knowledge, information, and belief.

SUBSCRIBED before me this day 2 010.

Kathleen A. Bennett NOTARY PUBLIC, COMMONWEALTH OF VIRGINIA MY COMMISSION EXPIRES: 8/31/2011 Reg. #110864 KATHLEEN ANN BENNETT Notary Public Commonwealth of Virginia 1 "110864

.My Commission Explres Aug 31, 2011

Response to Request for Additional Information No. 422(4792), Revision 0, Supplement 5 8/3/2010 U. S. EPR Standard Design Certification AREVA NP Inc.

Docket No.52-020 SRP Section: 03.09.02 - Dynamic Testing and Analysis of Systems Structures and Components Application Section: 3.9.2 QUESTIONS for Engineering Mechanics Branch 2 (ESBWR/ABWR Projects)

(EMB2)

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 2 of 25 Question 03.09.02-125:

This is related to RAI 03.09.02-48d.

The staff noted that CVAP section 4.3.2.1 describes-the development of the turbulent pressure forcing function that was the turbulent pressure spectrum prior to performing the HYDRAVIB model scale test. CVAP Figure 4.28 demonstrates that the forcing function is conservative over the range of nondimensional frequencies from 0.5 to 10. Comparing CVAP Figure 4.28 to 4.29 based upon the inflection point at F=2 (25 Hz), indicates that the conservative range of the forcing function covers a frequencies from about 6 Hz to 125 Hz. The staff noted that this appears to indicate that the lower frequency limit is acceptable based upon the first modal frequency of the FDD of 68 Hz, the applicant is requested to confirm this comparison and discuss the comparison of the HYDRAVIB and the estimated nondimensional pressure spectrum below F=0.5.

Response to Question 03.09.02-125:

The response of the first modal frequency of the flow distribution device (FDD) ( [ ] Hz) is based on the level of energy in the power spectral density (PSD) curve at this frequency, which confirms the comparison. Because the PSD used in the analysis bounds the PSDs developed during the HYDRAVIB testing at a frequency of [ ] Hz, the response of the first mode is conservatively predicted.

As stated in the Response to Question 03.09.02-126, the dimensional PSD used with the analytical evaluation of the FDD is based on a velocity of [ ] ft/sec, which is approximately [ ] percent larger than the actual velocity through the FDD. This difference in velocity corresponds to a dimensional PSD value that is [ ] times larger at all frequencies. The degree of conservatism in the PSD at dimensionless frequencies less than F=0.5 is estimated as follows:

Referring to Table 03.09.02-129-1 for the lower bound nondimensional frequency of 0.005, the dimensional pressure PSD term computed with a [ ] ft/sec is equal to:

Gp(f) = [ ] (psi2 /Hz)

The dimensional pressure PSD in the lower plenum of the reactor vessel (RV), which is determined from the definition in Technical Report ANP-10306P, Section 4.2.2.4.3, is used to compare the difference in the PSD at the lower bound frequencies. The equivalent dimensionless frequency for the definition of the lower plenum PSD is 2(0.005) or 0.010. For a nondimensional frequency 0.01 and a velocity of [14.3] ft/sec through the FDD, the dimensional pressure PSD is equal to:

AREVA NP Inc.

I Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 3 of 2 5 The dimensional pressure PSD for the lower plenum (PSD(f) = 3.76 x 101 psi 2 / Hz) does not completely bound the PSD that is used in the analytical evaluation (G(f) = 2.13 x 101 psi 2 / Hz) at the dimensionless frequencies near the 0.01 value. The first modal frequency for the FDD is

[ ] Hz (F=2.6), which is larger than 0.01. The unconservatism in the PSD at this low frequency does not impact the results of the current analysis. Considering the conservatisms described in this response for the PSD at higher frequencies, the analytical evaluation of the FDD is conservative.

FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report," Revision 0 will not be changed as a result of this question.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 4 of 25 Question 03.09.02-128:

This is related to RAI 03.09.02-48d.

The applicant estimated a bandwidth of the acoustic signal based upon the viscous damping of the fluid. The frequency of the shaft rate tone and its harmonics will be governed by the stability of the rpm of the RCP. Variations in the RCP rpm will cause corresponding variations in the frequency of the shaft rate tone. Operating the RCP at 23 Hz will result in the pump blade passing frequency increasing to 138 Hz which corresponds to the 138 Hz plate mode of the FDD. Additionally, a 10 percent RCP over speed will place the rpm of the RCP at 22 Hz which results in the center frequency of the blade passing frequency increasing to 132 Hz which is very close to the 138 Hz plate mode of the FDD. The staff noted that the bandwidth of the RCP shaft rate harmonics are not wide enough to cover this contingency. Therefore, the applicant is requested to discuss how the analysis accounts for frequency bias and uncertainty in the modeling and in the forcing function which includes these potential effects.

Response to Question 03.09.02-128:

Since the modal frequencies and mode shapes of the flow distribution device (FDD) are determined using the material properties established in ASME code and its nominal (design) dimensions, no frequency bias or uncertainty is included in the modal solution of the FDD. To address these, additional analyses are performed to assess the sensitivity of the FDD response to off-resonant and resonant forced vibrations created by the reactor coolant pump (RCP) acoustic pressure fluctuations.

The off-resonant response of the FDD is determined with a forcing function representative of a

[ ] psi (0-peak) RCP acoustic pressure amplitude at a [ ] Hz (the first blade passing frequency of the U.S. EPR RCP). The resonant response of the FDD is determined with the same magnitude of pressure but applied at the resonant frequency of the FDD plate mode ( [

] Hz). The numerical results from these two models will be reported and compared in Technical Report ANP-10306P, Section 4.3.4.2 (or Table 4-16-A).

The results of the analytical evaluation demonstrate that the resonant response of the FDD to the [ ] psi (0-peak) amplitude of the RCP-generated pressure fluctuation does not result in high cycle fatigue failure. If this resonant condition is revealed during the hot functional test (HFT) of the reactor pressure vessel (RPV) internals and could jeopardize the performance and structural integrity of the FDD, appropriate actions will be taken to confirm that the causes are mitigated.

Technical Report ANP-10306P, Section 4.3.2 and its sub-sections will be revised to separate the analytical methodologies used for the evaluation of flow excitation resulting from random turbulence from those used for the RCP acoustic pressure fluctuations. Technical Report ANP-10306P, Section 4.3.3 and its sub-sections will be revised to include an acceptance criterion for flow excitation resulting for RCP acoustic pressure fluctuations. Technical Report ANP-10306P, Section 4.3.4 and its sub-sections will be revised to separate the results obtained for the response of the FDD due to random turbulence from those obtained for the RCP acoustic pressure fluctuations.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 5 of 25 FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report," Revision 0 will be revised as described in the response and indicated on the enclosed markup.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 6 of 25 Question 03.09.02-129:

This is related to RAI 03.09.02-48d.

The staff noted that the amplitude of the pressure does not appear to change when a comparison of CVAP Figures 4.28 and 4.29 is made. The amplitude in both figures at F=0 (0 Hz), F=2, (25 Hz) and F=10 (125 Hz) appear to have the same numerical value. The applicant is requested to verify the axis label on figure 4-29. If correct, the applicant is requested to provide the numerical values of the normalizing variables.

Response to Question 03.09.02-129:

Technical Report ANP-10306P, Figures 4-28 and 4-29 are correct. The numeric values used to create the AREVA NP power spectral density (PSD) presented in these figures, as well as the figure plotting data, are provided in Table 03.09.02-129-1.

The PSDs in Technical Report ANP-10306P, Figure 4-28 are plotted against the Strouhal number (fDhN) based on the hydraulic diameter of the downcomer annulus. The PSD in Figure 4-29 is plotted against a Strouhal number (fRh/V) based on the hydraulic radius or the gap of the downcomer annulus, so the x-axis scales for these figures differ by a factor of two.

FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report," Revision 0 will not be changed as a result of this question.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 7 of 25 Table 03.09.02-129-1-Raw Data for the PSD used with the Analysis of the FDD in Figure 4-28 of the Technical Report ANP-10306P Notes:

1. The two dimensionless frequencies (F) reported in this table are based on the two different definitions provided in Technical Report ANP-10306P, Sections 4.2.2.4 and 4.3.2.1.
2. The dimensionless PSD defined in this table is based on the definition in Technical Report ANP-10306P, Section 4.3.2.1.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 8 of 25 Question 03.09.02-130:

This is related to RAI 03.09.02-48d.

The staff noted that in CVAP Section 4.5.1.1.2, the applicant detailed the analysis of the columns for vortex shedding induced vibration. The staff agrees that the procedure to estimate the response of the columns to off-resonant conditions by applying the fluctuating lift and drag forces with an appropriate amplification factor is acceptable. However, the applicant is requested to provide the fluctuating lift and drag forces for use with this procedure along with the methods used to derive them.

Response to Question 03.09.02-130:

The fluctuating lift (FL) and drag (FD) forces per unit length applied to the upper internal support columns are determined using the equations provided in Technical Report ANP-10306P, Section 4.5.1.1.2 and shown below.

FL = (p V 2 / 2)(D)(CL)Sin(2zfLt)

FD = (p V 2 /2)(D)(CD)Sin(27fDt)

Where:

D is the column support diameter.

pV 2/2 is the dynamic pressure acting on the outside diameter (OD) surface of the column support along the axial length of the support column.

CL and CD are the lift and drag coefficients. The lift coefficient for an array of cylinders is 0.07; the drag coefficient is equal to one order of magnitude less than the lift coefficient per Reference 1, Chapter 6.

fL and fD are the vortex-shedding frequencies that are determined with the relationship described in Technical Report ANP-10306P, Section 4.5.1.1.2 and an appropriate Strouhal number and limiting cross-flow velocity.

The static lift and drag forces are determined for the nodes along the length of the column supports considering the thermal hydraulic conditions unique to a particular elevation in the upper internals. This distribution of static forces along the length of the support columns is converted to a single equivalent static force that is multiplied by the amplification factor "Ao(f),"

which is determined from the relationship provided in Technical Report ANP-10306P, Section 4.5.1.1.2.

A Strouhal number equal to 2.26 and a limiting cross-flow velocity of [ ] ft/sec are used to determine the vortex-shedding frequencies of the normal column support. The equivalent static lift and drag forces are:

Feq-lift =ý [ ] lbs and Feq-drag = [ ] lbs

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 9 of 25 The dynamic lift and drag forces are then determined from the product of the equivalent static force and the amplification factors. The amplification factors for each of the first five modal frequencies for the normal column support are:

As stated in Technical Report ANP-1 0306P, Section 4.5., AREVA NP will revise the column support analysis to evaluate a detailed velocity and density profile along the length of the column supports. The revised results for the upper internals components will be provided with the Response to Question 03.09.02-131. See the Response to Question 03.09.02-132 for additional information about this revised velocity and density profile along the length of the column supports.

FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report," Revision 0 will not be changed as a result of this question.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 10 of 25 Question 03.09.02-132:

This is related to RAI 03.09.02-48d.

The staff reviewed the applicant's use of the FSM to insure that the pitch velocity is less than the critical velocity given by Conner's equation and determined that it is acceptable. The applicant uses a factor of safety for this ratio of 30%. The applicant is requested to discuss this factor of safety in relation to the bias and errors in the calculations that underlie the estimation of the pitch velocity in order to justify the value.

Response to Question 03.09.02-132:

The minimum fluid-elastic stability margin (FSM) for the control rod guide assembly (CRGA) column support in the upper internals is [ ] , which represents a safety margin of [ ]

percent for the column supports rather than a safety margin of 30 percent. The additional conservatisms inherent in the [ ] percent margin of safety are the estimation of the pitch velocity and the selection of the Connor's constant and the damping ratio.

As stated in Technical Report ANP-10306P, Section 4.5.1.1.3, a minimum value for the Connor's constant of [ ] and a damping ratio of [ ] percent viscous are applied in the analysis, both of which decrease the critical velocity and result in a lower computed FSM.

These values of the Connor's constant and damping ratio are also in agreement with the minimum recommendations of the 2004 ASME Code Section III, Appendix N-1331.3.

As stated in Technical Report ANP-10306P, Section 4.5.3, and in the Response to Question 03.09.02-130, AREVA NP will revise the column support analysis to evaluate a detailed velocity and density profile along the length of the column supports. As stated in Technical Report ANP-10306P, Section 4.5.3, the thermal hydraulic conditions in the upper plenum are determined with a one-dimensional model. The three-dimensional computational fluid dynamic (CFD) model and thermal hydraulic analysis of the upper plenum has been completed. The accuracy of the CFD model has been validated against flow tests performed during the ROMEO mockup flow testing. The velocity and density distribution along the length of the support columns that is obtained with the three-dimensional model of the upper plenum and associated structures will improve the pitch velocity prediction. A comparison of the velocity and density profile from the one-dimensional and three-dimensional models is provided in Figure 03.09.02-132-1 for location S6 of the CRGA (Refer to Technical Report ANP-10306P, Figure 5-10) to show the additional conservatism in the flow-induced vibration (FIV) analysis. The thermal hydraulic condition for locations S4, J9, T7, and R3 have been computed and also bound the thermal hydraulic conditions that were originally evaluated.

The factor of safety inherent in the computations for the critical velocity of the CRGA column support using the minimum values of Connor's constant and the damping ratio provides at least a [ ] percent margin to fluid-elastic instability. Additional conservatism exists in the minimum computed FSM value of [ ] for the CRGA column support by the conservative prediction of the pitch velocity from the one-dimensional thermal hydraulic model.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 11 of 25 FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor InternalsTechnical Report," Revision 0 will not be changed as a result of this question.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 12 of 25 Figure 03.09.02-132-1-Comparison of Cross Flow Gap Velocity for CRGA Column Support (S6) with Four RCP Operation - Full Power Normal Operating Conditions

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 13 of 25 Question 03.09.02-133:

This is related to RAI 03.09.02-48d.

The applicant detailed the analysis procedure for the upper plenum internals random turbulence induced vibrations in CVAP Section 4.5.1.1.4,. This is also known as turbulent buffeting and results from cross flow conditions above the upper core plate as flow progresses to the outlet nozzles. The methodology is standard and follows the work of Pettigrew and Gorman, referenced in the ASME Boiler and Pressure Vessel Code, 2004. Further, the applicant discuss the mean square vibratory amplitude in terms of a single sided random force PSD. The applicant is requested to provide a discussion of the derivation of this force PSD form the pressure PSD obtained from the Random Lift Coefficient of CVAP Figure 4-34.

Response to Question 03.09.02-133:

The force power spectral density (PSD) is determined from the product of the pressure PSD, as defined in Technical Report ANP-1 0306P, Section 4.5.1.1.4, and the area of the finite element on which the pressure PSD is acting. A force PSD is computed for each finite element node exposed to cross-flow conditions along the length of the structure. The area for each node is determined based upon one-half of the area of each of the two elements adjacent to the node in question.

FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report," Revision 0 will not be changed as a result of this question.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 14 of 25 Question 03.09.02-135:

This is related to RAI 03.09.02-48d.

The staff noted that the applicant uses acceptable values for the correlation length (2*OD) and for the structural damping coefficient (1 percent) for the instrumentation guide tube. The applicant is requested to provide a discussion of the use of the correlation length in the analysis procedure and the calculation of the Joint Acceptance integral.

Response to Question 03.09.02-135:

The acceptance integral, which is dependant upon the correlation length (k), is a measure of the matching of the spatial distribution of the forcing function and the mode shapes of the guide tube. Because the width of the guide tube is small, the forcing function is assumed to be 100 percent coherent and in phase across its width. Along the length of the guide tube, the correlation length equal to ( [ ]) defines the length for which the forcing function is coherent between two different points along that axis. Outside of this limit, the excitation created from the turbulence is considered insignificant and is not computed.

The acceptance integral between the modes "m" and "n" at a frequency "f' is defined by the integral:

Jm.(f) = f.m(x')Gp(X',X", f ) 0,(X")dx'dx" D

The acceptance integral computed by the AREVA NP computer program PCRandom follows the "coherence integral" method outlined in Reference 2. The coherence integral method is a general finite element formulation that replaces the acceptance integral method and the double integral over the entire structure for each mode pair combination with a mode independent "coherence integral" over each finite element and then the appropriate summations over the entire model. The coherence integral formulation is defined by the integral; J.n(f) ZjZlin.(i) 0"(Xj) f fJGP (x' X,x"f )dchidx" DI Di The cross spectral density function "Gp(x',x",f)" is expressed as a function of the power spectral density "Gp(x',f)" and equal to; Gp(x', xf) = G (x', f)F(x',xVIf) Gp(x" with the coherence function equal to functional form;

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 15 of 25 r1(x', x" f) = e[rla[+ib(U*r)]

a- A(xi, xjf) b 2;Tf U(xi,x 1,f)2 r = 11-XI Further simplification identified in Reference 2 leads to the following expression for the coherence integral summation; sn(fs = ZZ .(Xz)0n(Xj,) IdG(x,,f) VG(x,,f)Cs(f)

For x, < xj C. = 4AyjAy e[(a+ibUx)(xjx,)]

  • Lsinh[(a + ibUx)(Axi /2)]
  • sinh[(a + ibUx)(Axj /2)]

For x, = xj 2aAx Ay 2 ..A 2

[(a++/-bUxJ)(Ax/2)]* sinh[(a + ibUx)(Ax, /2)]

"- = a2 + b 2U 2 (a+ibU) 2

+ 2Ay [(a-ibUx)(&x /2)],

  • sinh[ (a-_...(tbxibUx)(Ax (a - ib(Ub)
  • /2)]2 1 Where; Axi = is the length of element "i" Ayi = is the width length of element "i" FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report," Revision 0 will not be changed as a result of this question.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 16 of 25 Question 03.09.02-136:

This is related to RAI 03.09.02-48d.

The applicant described the CRGA (tie rods, c-tubes, rod cluster control assembly (RCCA) in CVAP Section 4.6.3.1. The internal components of the CRGA and which portions of the structure will be susceptible to FIV are described in CVAP Section 4.6.3. The applicant is requested to provide a diagram of the CRGA showing the relationship of the various components, the position relative to the UCP, and the USP as well as which components are exposed to the flow. Such a diagram is required to adequately assess the statements on CVAP page 4-149 regarding which components are not evaluated for FIV.

Response to Question 03.09.02-136:

Figure 03.09.02-136-1 depicts the relative orientation of the control rod guide assembly (CRGA) with the control rod guide tube (CRGT). Figure 03.09.02-136-2 depicts the relative orientation of the reactor cluster control assembly (RCCA) with the upper core plate (UCP) and the top nozzle of the fuel assembly (FA) when the RCCA is fully withdrawn during full power normal operating conditions. Refer to Technical Report ANP-1 0306P, to Figure 4-39, for a plan view of the CRGA guide plates.

Figure 03.09.02-136-1 also shows the locations along the length of the CRGA for which the thermal hydraulic conditions identified in Technical Report AN P-1 0306P, Table 4-21, are evaluated. The cross-flow gap velocity acting on the RCCA below the UCP is insignificant since the flow in this region is axial through the UCP and is not explicitly evaluated.

The CRGT protects all but the lower portion of the CRGA from the extreme cross-flow conditions in the upper plenum. The lower region of the CRGT is open to allow the majority of the flow from the fuel bundle passing through the UCP to exit into the upper plenum of the reactor pressure vessel (RPV). Some of the flow entering the lower region of the CRGT does not exit into the upper plenum but continues an upward path through the CRGT into the upper head. Primary flow into the upper head region is also aided by the spray nozzles. The axial flow is downward toward the UCP for some of the CRGAs.

The Technical Report ANP-10306P, Section 4.6.3.1 will be revised to include a figure similar to Figure 03.09.02-136-1 of this response.

FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report," Revision 0 will be revised as described in the response and shown in the attached markup.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 17 of 25 Figure 03.09.02-136-1-CRGA and CRGA Column Support

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 18 of 25 Figure 03.09.02-136-2-RCCA Interface with UCP and Top Nozzle of FA Notes:

1. Dimensions are in inches

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 19 of 25 Question 03.09.02-137:

Follow-up to RAI 245, Question 03.09.02-49 The applicant responded to RAI 03.09.02-49 in Response to RAI 245, by citing the CVAP. The applicant described justification for not performing an explicit transient analysis of the reactor vessel (RV) lower internals in CVAP Sections 4.2.8 and 4.3.5. The staff noted that justification given for not explicitly analyzing transients is that the primary loading for the lower internals is high cycle fatigue and that transient analysis is not warranted. To bolster the argument, the applicant estimates the response of the lower internals to a 10 percent reactor coolant pump over speed by considering the effect of increasing the coolant flow through the reactor by 10 percent and scaling the turbulent forcing functions on the increased dynamic pressure and results in an increase in the turbulent forcing functions of (1.1)2. The staff noted that this type of increase would occur if all four pumps were simultaneously operating at a 10 percent overspeed. However, in CVAP Section 4.5.1, the applicant indicates that the four RCP line-up 10% overspeed condition may not-be the highest transient to consider. The applicant suggests that the two pump operation will result in higher response by the upper internals. The applicant is requested to discuss this statement and explain why the 10 percent overspeed for all four RCPs is the bounding case for the RCP operation.

Response to Question 03.09.02-137:

As stated in Technical Report ANP-10306P, Sections 4.2.8, 4.3.5, and 4.5.4, the contribution to high cycle fatigue created by turbulence in the primary flow with four reactor coolant pump (RCP) operations at 110 percent capacity (or rated flow) for 60 effective full power years (EFPYs) is more limiting than the high cycle fatigue incurred from short-duration transients (e.g.,

different combinations of RCP operation that typically occur during plant heat up).

Acceptable stress levels in the reactor pressure vessel (RPV) internal components during these transient conditions when the amplitudes of vibration may be larger than those during full power normal operating conditions, with four RCPs operating, will be verified with the hot functional test (HFT). The response of the column supports to the turbulent conditions in the upper internals resulting from two RCP operations, as well as other combinations of RCP operation, will be measured during HFT to confirm acceptable vibratory behavior according to the acceptance criteria established in Technical Report ANP-10306P, Section 4.5.2.2. The fatigue damage incurred during these short-duration transients will be calculated from the HFT test results to confirm that damage is insignificant. In addition, the fatigue results of the HFT for these transients will be used to confirm that the development of an analytical response of the column supports to the flow excitations resulting from each of these RCP transient conditions is not required to demonstrate the integrity of the upper internals of the reactor vessel (RV), as described in Technical Report ANP-10306P, Section 4.5.4.

FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report," Revision 0 will not be changed as a result of this question.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 20 of 25 Question 03.09.02-139:

Follow-up to RAI 245, Question 03.09.02-51 The staff issued RAI 03.09.02-51 as a follow-up to RAI 03.09.02-32 requesting the analysis of the full-scale CRGA components that demonstrates acceptable vibrational behavior.

The applicant responded to RAI 03.09.02-51 in Response to RAI 245, by stating that the details of the analyses and testing that will indicate acceptable behavior, including the acceptance criteria, details on the validations of the test plan, and the instrumentation and test conditions that will be used in the U.S. EPR preoperational testing, to confirm the acceptable CRGA design, are provided in Sections 4.6 and 5.0 of the Technical Report ANP-1 0306P.

The staff reviewed the applicants response to RAI 03.09.02-51 and noted that the applicant references CVAP section 4.6 for the analysis of the CRGA design. The acceptance criteria for the CRGA was reviewed by the staff in response to RAI 03.09.02-48E.

The analysis techniques follow those for the rest of the upper internals found in CVAP Sections 4.6.3 that were reviewed by the staff in response to the applicant's answer to RAI 03.09.02-48D.

The applicant referenced CVAP Section 5.0 for the test details, the instrumentation details and the test conditions. The instrumentation for the CRGA are found in CVAP section 5.2.2.3 and all are intended to be temporary. Only the portions of the HFT that pertain directly to the CRGA are discussed in the response to RAI 03.09.02-51. The remainder of the test requirements are discussed in the review of the applicant's response to RAI 03.09.02-48G and RAI 03.09.02-54.

The staff reviewed the applicants description of the intended instrumentation suite as described in CVAP Section 5. The set of instrumentation specified for determining the response of the CRGA column and CRGA internals is sufficient to obtain the upper bound response provided most of the instrumentation survives throughout testing.

The staff noted that one of the CRGAs which is located in a region anticipated to offer the highest cross flows has been chosen to measure the reaction of the CRGA column. This column is located at S6, which is on the outer periphery, near the center outlet jet port for loop 4. This CRGA column has been instrumented with four strain gages near the connection to the Upper Support Plate (USP). Two of these are intended to be redundant. However, if only two of these fail on opposite sides of the column, one of the principle directions will be lost. One of the guide plates, located at midspan of the CRGA at S6, has been instrumented with two accelerometers, oriented to measure in the horizontal plane at 90 degrees to each other.

Based on the CVAP description, the applicant is requested to:

A. Confirm that the instrumentation on the CRGA at S6 is intended to be oriented in the mean flow direction and transverse to the mean flow direction. Under this assumption, if either accelerometer fails, one of the principal directions will be lost.

B. Discuss the impact of failing transducers and the potential negative impact to achieving the stated HFT goal of characterizing the behavior of the CRGAs and columns (see CVAP Section 5.2.2.3).

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 21 of 25 C. Discuss options or plans for increasing the number of transducers installed or replacing failed transducers during the test phase. Note that the guide plate of a second CRGA, located at J9, in the center of the core support plate, is instrumented with two radial accelerometers, oriented at 90 degrees to each other.

D. Discuss the methodology, analysis, or rationale used for selecting the guide plate in these two CRGAs, one at the periphery and the second in the center.

Response to Question 03.09.02-139:

a) The orientation of the strain gages on the control rod guide assembly (CRGA) tube and the accelerometers on the guide plate for location S6 will be orientated at approximate 45 and 135 degree angles relative to the mean flow to achieve greater instrumentation redundancy.

b) As stated in the response to (a), this redundancy provides some backup for failed transducers. The degree of negative impact on the behavior of the CRGA depends on when the instrumentation fails. Ifthe first six of eighteen tests identified in Technical Report ANP-10306P, Table 5-4 are completed with the transducers functioning properly, the impact would not be negative.

This degree of success is because the largest response of the CRGA occurs during this series of tests. The response of the CRGA is largest during these tests because of the larger forcing function that is created from the larger dynamic pressure term at the lower temperatures and pressures of the primary fluid. The slight variation of the natural frequencies that occurs between 70OF and 578'F does not have a significant effect on the response of the CRGA since the fluid-structure coupling mechanism for random turbulence is weak. This is because the flow field induced by the structural motion is linearly superimposed on the incident flow field. A strongly coupled fluid-structural system, characterized by large structural motion, induces fluid velocities and distorts the incident flow field. Vortex-shedding induced vibration and fluid-elastic instability of heat exchanger tubes are examples of strongly coupled fluid-structure systems. In strongly coupled fluid-structural systems, the natural frequencies of the components are fundamental to the sensitivity of the flow-induced vibration (FIV) mechanism and the strength of the response.

For the weakly coupled fluid-structural system associated with random turbulence, the relationship between the natural frequencies and the response of the structure is not as strongly associated.

The variation in the natural frequencies that occurs between 70'F and 578'F does not significantly alter the response of the structure. The confirmation of the forcing function at the 140OF and 390 psia test conditions provides sufficient test data to confirm the forcing function and the response of the CRGA at the higher temperatures and pressures for tests, seven through eighteen of Technical Report AN P-1 0306P, Table 5-1.

c) The number of transducers cannot be increased because of the limitations created by the size of the reactor vessel (RV) head penetrations through which the instrumentation cables are routed to the acquisition system. The redundancy priority given to the instrumentation to characterize the RPV components through the hot functional test (HFT) is manifested through the instrumentation scheme shown in Technical Report ANP-1 0306P, Table 5-1, and is the most appropriate. As stated in Technical Report AN P-1 0306P, Section 5.5, the

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 22 of 25 greatest priority is given to the characterization of the core barrel pendulum mode of the lower internals.

If the testing identified in Response (b) to Question 03.09.02-139 is not accomplished, it may be necessary to suspend the HFT either to replace failed transducers or to relocate the transducers at the CRGA location J9 to location S6. These priorities and contingents cannot be determined until the failed transducers are identified.

d) The guide plate instrumentation confirm the modal characterization and fatigue tests, identified in Technical Report ANP-10306P, Sections 4.6.1 and 4.6.2, that were performed for the CRGA. These tests provide the justification of the CRGA structural integrity to the high cycle fatigue. The high cycle fatigue of the CRGA is created by its interaction with the CRGA tube, which is exposed to the flow excitation created from random turbulence acting on its outer surfaces.

The results obtained with the HFT should be bound by the test identified in Technical Report ANP-1 0306P, Sections 4.6.1 and 4.6.2 to validate the CRGA design for the thermal hydraulic conditions prevalent in the full-scale design during normal operating and transient conditions.

FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of this question.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report," Revision 0 will not be changed as a result of this question.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 23 of 25 Question 03.09.02-141:

Follow-up to RAI 245, Question 03.09.02-54 The applicant responded to RAI 03.09.02-54 in Response to RAI 245, by stating that a response to this question will be provided in the CVAP Technical Report ANP-1 0306P.

The staff noted that applicant was requested to provide a discussion of the factors that influence the comparison of the test results to the analysis and how they will be incorporated into the testing program. The applicant states that the experimentally obtained vibration amplitudes, frequencies and stresses will be compared to the analytical values and the appropriate criteria developed in CVAP Section 4. The applicant is requested to explain:

a. what is meant by sufficient safety margin
b. what will constitute a match between the experimental and analytical frequencies or amplitudes
c. quantitatively how the errors will be combined with the measurements to indicate sufficiently good comparisons
d. how fluid-elastic instability, acoustic resonance and vortex-shedding lock-in will be quantitatively determined.

The applicant provided a discussion of the test acceptance criteria and many of the parameters that can influence. the collection of the data in CVAP sections 4, 5.4 and 5.5.

The applicant has stated that deviation between theoretical predictions and measured values may result in changes to the theoretical evaluation if the differences have an impact on the integrity of the RV internals. The applicant is requested to explain how it will determine that a particular deviation between experimental results and analytical results impact the integrity of the RV internals.

Response to Question 03.09.02-141:

a) The term "sufficient margin of safety" means that based upon the margin of safety obtained with the analytical results for the full-scale design, the U.S. EPR reactor pressure vessel (RPV) internal components will not experience excessive vibrations or experience high cycle fatigue failure during the hot functional test (HFT), even if the most limiting analytical solution is inaccurate within 200 percent, consistent with the acceptance criteria established in Technical Report AN P-1 0306P, Section 5.5.

b) In general, a difference of 10 percent in the comparison of the frequencies obtained with the analytical solution and the HFT measurements will be considered acceptable for the RPV internal components. This degree of difference will not be acceptable for some of the RPV components whose behavior is critical to vibratory characterization of the RPV internals and fuel bundle; e.g., the fundamental frequency of the core barrel pendulum mode. This mode is principal in the base motion that the fuel assemblies experience, and good agreement of this frequency is fundamental to the characterization of the vibratory behavior. It is not as crucial that the core barrel shell frequencies compare as closely since they cannot impart excitation to the fuel bundle.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 24 of 25 The amplitudes of vibration for each modal frequency are primarily related to the strength of the forcing function. As stated in the Response to Question 03.09.02-35, the forcing function measured during HFT will be revised to obtain the necessary agreement of results, if necessary.

c) The HFT acquisition system design minimizes errors in the measurements as stated in Technical Report ANP-10306P, Section 5.4.4. The calculation of the errors associated with the sensors, signal conditioning, and data acquisition system will be combined using the square root sum of the squares methodology.

d) The existence or detection of either fluid-elastic instability, acoustic resonance, or vortex-shedding lock-in during HFT are apparent because of the ability of these flow-induced vibration mechanisms to induce a significant response. As stated in the Response to Question 03.09.02-139 (b), these mechanisms exhibit strong coupling between the fluid and structure. Their '"resonant" response is sensitive to the natural frequencies of the structure and exhibits a maximum response at these frequencies.

The design of the RPV internals is such that these flow excitation mechanisms are not active at normal operating and transient conditions. Therefore, the comparison of the analytical results and the field data obtained from HFT, or at least the confirmation of the margin obtained with the analytical results, is not possible.

The accuracy of the analytical solution and the margin is inherent in the design criteria provided with ASME Section III, Appendix N-1300, 2004, to which the U.S. EPR RPV internal components conform. The design criteria provided therein for fluid-elastic instability and vortex shedding lock-in are based upon testing that has been correlated with analytical solutions. The RPV need not be tested at a flow rate that would cause these flow-induced vibration mechanisms to become active to confirm the degree of margin in the analytical solution.

Regarding acoustic resonances, the design criteria provided in Technical Report ANP-10306P, Appendix A.2.1, will verify that this mechanism is not active in the RCS. The screening criterion is based upon field testing of 40 valves. Once the piping attached to the RCS has been designed, the degree of margin to the onset of acoustic resonances can be assessed.

FSAR Impact:

The U.S. EPR FSAR will not be changed as a result of thisquestion.

Technical Report Impact:

ANP-10306P, "Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report," Revision 0 will not be changed as a result of this question.

AREVA NP Inc.

Response to Request for Additional Information No. 422, Supplement 5 U.S. EPR Design Certification Application Page 25 of 25

References:

1. Au-Yang, M.K. Flow-Induced Vibration of Power and Process Plant Components, ASME Press, 2001
2. B. Brenneman, "Random Vibrations due to Small Scale Turbulence with the Coherence Integral Method," Journal of Vibrations, Stress and Reliability in Design, Volume 109, April 1987.

Comprehensive Vibration Assessment Program for the U.S. EPR TM Reactor Internals Technical Report Markups

AREVA NP Inc. ANP-1 0306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-98 4.3 Flow DistributionDevice (FDD)

The assessment for the vibratory behavior of the FDD is based on 1/8 scale flow testing and full scale theoretical analysis. The primary objectives of the testing on the FDD are as follows:

0 To verify that an adequate distribution of the flow is created through the fuel bundle for various RCP operating combinations.

0 To verify that vortices are not created by the FDD to avoid flow excitation from vortex shedding.

a To assess the hydrodynamic mass effects of the FDD by determining the in-air and the in-water experimental natural frequencies.

4.3.1 Testing Performed for the FDD Flow testing for the FDD was performed as part of the HYDRAVIB test program. The 1/8 scale FDD mockup is used to identify the horizontal, vertical and torsional modes for both the in-air and in-water environments at ambient temperature. The FDD was instrumented with accelerometers and excited in different directions to determine the modal frequencies and mode shapes of the FDD. The modal frequencies for the 1/8 scale mockup of the FDD are reported in Table 4-11.

4.3.2 Theoretical Analysis of the Flow Distribution Device RAI 422 A full scale theoretical evaluation of the FDD is performed. The flow-induced vibratio103.09.02-128 phenomenon of concern for the FDD is random turbulence excitation resulting from the downcomer flow going through and past the FDD. The narrow band acoustic pressure fluctuations associated with the RCP rotational speed([ Hz)and the pump bladepassing frequency ([ ) Hz) are also a potential source of excitation. This a.rroW band pressuar flucuaton s uporimpsed upon the Ipressure PRIM for turbuldence inthe lower plenum; to acco-unit for this sourcoe of exciotation to the EDDO.

The measures identified in Section 4.2.5.2.2 for the lower internals will also be implemented for the FDD to assess the excitation of the FDD to other sources of acoustic pressure fluctuations (e.g., acoustic resonance and loop acoustics).

AREVA NP Inc. ANP-10306P Revision G1 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-99 The base excitation of the FDD resulting from the CB beam mode (~4 "lHz, See Table 4-5 and Table 4-6) is well separated from the beam mode of the FDD (I ] Hz, See Table 4-12). [

and is not represented in the theoretical analysis.

4.3.2.1 Analysis Methodology and FIV Inputs 4.3.2.1.1 Random Turbulence Induced Vibrations IRAI 422 The turbulence response of the FDD is calculated assuming exc ....

103.09.02-128 shell as well as the square grid plates. The response of the FDD to random turbulence due to parallel flow excitation of the thermal hydraulic conditions in the RV is calculated using the AREVA NP PC-based computer program "PCRANDOM." This program finds the response of a FEM to pressure PSDs based on finite element implementation of the acceptance integral method of Reference 7 and Reference 8.

The analytical method implemented to estimate the RMS response of the FDD follows the guidelines established in the ASME Appendix N-1300 (Reference 9a) and the equation shown below for parallel flow excitation of two dimension structures which, is taken from Reference 4, Equation 8.50.

-~

AGp(fr)V*(---

A~~a)x --

)Jarcrfa)

<Y 2 (X )>=Ia 647r3 m*2f, (fa Where:

< y 2 (x) > = the mean square vibratory amplitude.

A = the surface area of the element or structure.

Gp = the single-sided fluctuating pressure power spectral density (PSD) in (force/area)2/Hz, due to turbulence.

2 (x) = the mode shape function of the structure.

a = modal index, two directions.

J,, = the joint acceptance integral of the structure.

ma = the generalized mass.

f. = the modal frequency in Hz.

ANP-10306P AREVA NP Inc.

Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-102 Therefore, a single phase PSD is assumed for the evaluation of the FDD and later confirmed to be bounding as shown in Figure 4-28.

The PSD applied to the FDD FIV evaluation is representative of an upper bound curve for the turbulence in the downcomer of another RV design, which is taken from Reference 4, Figure 8.17 and reproduced in Figure 4-29.

SG(f) 3 0.155e-3 .°" for 0 < F < 1.0 qof-p2 V 3RH

= 0.027e-1. 26 F for 1.0*< F *5.0 Where:

F= fRH V

Gp (f)= single sided dimensional pressure PSD p = density of primary fluid RH = hydraulic radius of RV annulus or the gap V = velocity of the primaryflow RAI 422 03.09.02-128 The non-dimensional PSD is conservatively converted to a dimensional pressure PSD "Gp(f)"

considering the primary flow velocity and hydraulic radius of the RV downcomer as opposed to the RV lower plenum. The pressure PSD that is applied in the FIV analysis of the FDD is shown in Figure 4-29.

I B dl .1-,

Te account for the acoust1i p1escure l Tluctuatlone associated With the GI' rotatona 'I-_-_ i I

i Hz) and the pump blade passing fr*equec (r I H;z) wn assumed apressure

'~ ~ ~ ]psi.-FMs

~ ~ -e-[ ~ ... ~ ""*

fh4t44ti~-ef ...

I psi 4*(0 peak) :1" is,,superimposed i, ,,;S: ..

upon the pressure PS2D for

. L'_ I*,M'""

~~-A VA- .. ew .. % rr F9 I U; ...... ~a~

The banAdwýid-th for the pressure flcutosAro basod upon the 1,42 power width of the spectral peak as-fGl'QW4A9 Af 0 2fo i=J SolVing forF and

Af coc *l~~r;.*/ I viscous6

,*ra damping L/ *A,*r** yields

,*1 the

÷, frequency

, r~r.*, bandwidths 1-, I*-rrL,*;_ H at -- I

A.... ~ uwýtt~it ui nt ^n 1ýu
: e n +i nulLil nIuj Al PAP ARUUIU~t AL Q~ + A 3 RPAP S.. -~.. -F

+F6q61eRG1A_9_ G. +..A- . -

1ý ff, r1%

.......... ... ... ....Irr....... ) .....-

.... ab le. 4 1....-T . 2)

AREVA NP Inc. ANP-10306P Revision G1 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-103 proximn0ity to the RC-P shaft and- -bladoparssing frequencies. Theroeforo, the bro-adbhand 1PSO foi turbulenceRA wasr altered to considor theseP narrow band acous6tic pressu re fluc-rtuations, Tho final definition for the pressuro- PS-D that irs applied In tho FIV analysis of the FObD is6 s-hownVIC in Figure 4 29.

The convective velocity and the correlation length of the forcing function The correlation lengths and convective velocities are used to calculate the matrix of coheren

/~RAI1422 integrals based upon the definition for each pair of nodes in the FE model. The convective A4 velocity is conservatively assumed to be [ ] of the nominal flow velocity ([ ft/sec30902128 through the FDD to create the greatest coincidence of phase relationship with the forcing functions and the modal frequencies of the FDD. The correlation length is [

I inches) and does not have any frequency or spatial dependency.

4.3.2.1.2 Vibrations Induced by RCP Acoustic PressureFluctuations The response of the FDD to the RCP acoustic pressure fluctuations is determined by traditional techniques of forced vibrations using the PC-based computer Progqram EBDynamics, which is listed in the U.S. EPR FSAR, Tier 2, Chapter 3, Appendix C. The PC-based program EBDynamics program finds the time domain solution of coupled structure and fluid finite element models. The program also has the capability to find the solution for a structure-only model or a fluid-only model.

The EBDynamics model of the FDD reads the modal frequencies, stiffness and mass matrices of the finite element model of the FDD that are created with the execution of CASS in Section 4.3.2.1.1. An assumed acoustic pressure fluctuation of E j psi (0-peak) is evaluated at the RCP rotational speed ] Hz) and the pump blade passing frequency [ 1Hz). The modal frequencies of the FDD (See Table 4-12 and Figure 4-27) that are most susceptible to the excitation of the RCP pressure fluctuations are:

JHz (modes I and 2) and ft. M fbeam[

fpigte_=[ ]Hz.(mode) 2 IRAI 03.09.02-1 28

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-104 These modal frequencies are close to the RCP shaft and blade passingq frequencies, and their mode shapes are capable of being excited by acoustic pressure loading. The harmonic force that is representative of these RCP acoustic pressure fluctuations is applied to the EBDynamic model of the FDD.

To address the bias and uncertainties associated with the modal frequencies of the FDD, additional analyses are performed to determine the resonant response of FDD to the acoustic pressure fluctuations generated by the RCP blade passing frequencies. This resonant condition is most probable during the 10% RCP overspeed transient condition when the RCP blade passing frequency could increase tof .Hz (1.1 , Hz), which is in close proximity to the E J Hz plate mode of the FDD. If the actual frequency of the FDD plate mode varies due to differences in the material properties and the as-built dimensions, the two frequencies may be closer than the degree of separation that is identified, and potentially a resonant condition could exist.

RAI 422 03.09.02-128

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-111 Figure 4-29-Pressure PSD Loading for FDD (Log vs. Linear Scale)

RAI 422 03.09.02-128

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-112 4.3.3 FIV Acceptance Criteria for FDD 4.3.3.1 Acceptance Criteriafor Displacements Displacement limits are typically imposed to prevent impacts with the adjacent structures.

Because such interfacing structures do not exist for the FDD, displacement limits for the FDD are not applicable.

4.3.3.2 Acceptance Criteria for High Cycle Fatigue The FDD is primarily fabricated from stainless steel type F304LN. The acceptance criterion for the stress and high cycle fatigue limits of the FDD is identical to the criteria established for the RV lower internals (Section 4.2.6.2) for loading generated from random turbulence. The allowable stress for fatigue curve "A" at 1013 cycleis [ ]psi, rms is applicable to the FDD structure.

Since, the response of the FDD to the RCP acoustic pressure fluctuations are in units of (0-peak), the ASME fatigue curve "A" shown in Figure 4-21 is applicable. The allowable high cycle fatigue stress is[ ](0-peak) at 1011 cycles.

R.AI.422 03.09.02-128

AREVA NP Inc. ANP-10306P Revision G1 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-113 4.3.4 Response of the FDD The response of the FDD to the turbulence in the RV lower plenum resulting from the full power, steady normal operating condition and the acoustic pressure fluctuations associated with the RCP blade passing frequencies is reported in this section. The response of the FDD during HFT (Test #17, See Table 5-4) is assessed based upon the results for the full power normal operating condition and the ratio of the dynamic pressure term between the two operating conditions or (

The results show that the FIV acceptance criteria established in Section 4.3.3 for the FDD are satisfied.

422 03.0RA.

4.3.4.1 Over-all.Response of FDD to Random Turbulence 103.09.02-128 The response PSD (displacement) of the FDD at seven nodal locations around the circumference of the cylinder is shown in Figure 4-30. The figure shows the response at each of the natural frequencies of the FDD with a major response occurring at the fundamental frequency ([ ]). A significant response of the shell mode I

]) is also shown. The response of the FDD at the blade passing frequency ([ )

Hz) is evident even though the response amplitude is not very large (because the natural frequencies of the FDD are not very close to the blade passing frequency).

The RMS responses at the seven node locations are listed in Table 4-13. The table shows that the responses atr therep a ]. At the other two locations

]

The statistical frequencies for the FDD are zero crossing frequencies. The zero crossing frequency is defined as the number of times per second the response curve crosses the zero response line (assuming the mean has been subtracted out). This effective frequency is determined using the relation shown below considering the weighted integrals of the single-sided displacement response PSD, Gd(f):

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-114 f02 0fof2 Gd(fldf G ( definition of the zero crossing frequency.

fJGd(f)df 0

The AREVA NP computer program "PCRANDOM" uses this definition to determine'the zero crossing frequencies. The crossing frequencies reported in Table 4-13 for the discrete locations along the circumference of the FDD cylindrical shell are below the natural frequencies of the FDD. The reason for this discrepancy is based on the method by which the effective zero crossing frequencies are computed by PCRANDOM through the weighting of the integral of the single-sided displacement response PSD, Gd(f). Because the effective zero crossing frequency is determined by the integration of the response PDS between 0 and 200 Hz, the weighting of the PSD at the lower frequencies (below the fundamental frequency of the FDD) skews this computation. RAI 422 computation. __ 1 03.09.02-128 44444.3.4.1.1 DD Support Column Stress The support column moments at the fixed locations are reported for six of the 24 support columns in Table 4-14. Because the FDD is a symmetrical structure, the moments from the six columns (one quadrant) define the behavior of the FDD. The highest bending stress in the support column is approximately[ I psi, rms. Assuming a conservative value of [ Jfor the FSRF to account for discontinuities, the maximum rms stress is [ Jpsi, rms.

Considering the beam modal frequency of the FDD ( ] Hz, See Table 4-12), a 60 year life and a capacity factor of 100 percent, an endurance limit of [ ] rms is obtained from Figure 4-21. A factor of safety of [ I for the FDD column supports, and high cycle fatigue failure of this component is not expected. RAI 422

< 1~03.09.02-128 1 4.3*34. 3.4.1.2 4upport Column Reaction Forces The reaction forces and moments are reported at the fixed nodal locations for six of the 24 support columns as shown in Table 4-15. These are the same node locations at which the support column stresses are reported in Table 4-14. The results show that the reaction forces at the support column attachment locations[

]

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessme; ntP;,nrm fnrl _*111 eactor Internals Technical Report . ..--- RAI 42Page 4-115 1~03.09.02_128 4.4444.3.4.1.3 ,fDD CylindricalShell Stresses The principal stresses at the centroid of the cylindrical shell plate elements are obtained to evaluate the potential for fatigue damage to the FDD. For each of the plate elements shown in Figure 4-32, the two principal stresses (Si and S2) are calculated at the centroid of each element. These alternating stresses are calculated at both the inside'and outside surfaces.

Table 4-16 shows the principal stresses and demonstrates[

I is not expected.

4.3.4.2 Response of the FDD to the RCP Acoustic PressureFluctuations The response of the FDD to the RCP acoustic pressure fluctuations for the off-resonant[ ]

Hz) and resonant[ ] Hz) conditions is provided in this section. The resonant response of the FDD plate mo[ ]Hz) to the acoustic pressure fluctuations qenerated at the same blade passinq frequency is provided to evaluate the bias and uncertainties associated with the modal frequencies and the response of the FDD to this source of excitation. The summary of results provided in Table 4-16-A compares the off-resonant and resonant response of the FDD that is created from the two forcinq functions;

  • AP: ]psi(0-peak) at 1z AP : -peak) at Hz The stress results provided in this Table 4-16-A are for the most limitinq locations and do not include the stress amplification effects of the structural discontinuities of the FDD. Assuminq a conservative value of 4.0 for the FSRF, a maximum alternatinq stress of[ ]psi (0-peak)is obtained for the resonant loadinq condition (AP[ I psi at[ ]Hz). Since this stress amplitude is much less than the endurance limit of 23,300 psi at 1011 cycles for SS 304-LN, hiqh cycle fatique failure is not expected for the FDD structure RAI 422 03.09.02-128

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-116 Table 4-13-RMS Response at Different Shell RAI 422 03.09.02-128 Notes for Table 4-13:

1. See Figure 4-31 for node locations.

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-117 Table 4-14-Maximum Stress for the FDD Support Columns F 1 RAI 422 03.09.02-128

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-118

AREVA NP Inc. ANP-1 0306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-119 RAI 422 Table 4-15-Support Column Reaction Forces k103.09.02-128 Table 4-16-FDD Cylindrical Shell Stresses (psi, rms)

Notes for Table 4-16:

1. See Figure 4-32 for element locations.

IRAI 03.09.02-128

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-120 RAI 422 03.09.02-128

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-123 RAI 422 4.3.5 Conclusions 03.09.02-128 The resultslfrom the analysis for turbulence how that the FIV performance of the FDD is acceptable for the full power, steady state normal operating conditions. The stresses in the cylindrical shell and square flow channels are significantly below the allowable stress. The high cycle fatigue effects of FIV are not applicable for this component. Additionally, the reaction loads at the supports of the FDD that result from the FIV response of the FDD are minimal. It is not possible for the bolted connections between the FDD and the LCP or at the tie rod connections between the LCP and the HR to separate and create additional cyclic loading on these members.

The result from the analysis of the FDD for the forced vibrations created by a RCP acoustic pressure fluctuation show that both the off-resonant and the resonant response of the FDD are insignificant during full power, steady state normal operating conditions. The response of the FDD to the resonant conditions between the RCP blade passing frequency and the FDD frequency (mode 6, Hz) demonstrate that any bias or uncertainty that may exist in the modal frequencies or the response predicted for the FDD to this source of excitation will not create unacceptable vibrations.

Based on the large margin of safety for the FDD at the full power normal operating conditions, there is ample margin for this component during the following RV transient conditions:

  • 10 percent RCP overspeed transient conditions that may occur during full power normal operating conditions. RAI 422 103.09.02-128
  • Different combinations of RCPs operation (i.e., one, two, or three RCP combinations).

Explicit analytical evaluations of the short term transients are not performed. These transient conditions are evaluated as follows:

Because a 10 percent RCP overspeed transient condition will produce a 10 percent increase in the primary flow through the FDD, this corresponds to a 21 percent increase in the response of the FDD or a scaling factor of 1.21 (1.10)2 based on the relationship for the dynamic pressure

AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-156 Table 4-21-Thermal-Hydraulic Inputs for the CRGA and the RCCA Note(s) for Table 4-21:Table-4-2-1

1. Refer to Fiqure 4-39-A for the locations along the length of the CRGA at which these thermal hydraulic conditions are applied.

RAI 422 03.09.02-136

I AREVA NP Inc. ANP-10306P Revision 01 Interim Comprehensive Vibration Assessment Program for U.S. EPR Reactor Internals Technical Report Page 4-158 Fiqure 4-29-A RAI 422 CRGA and CRGA Column Support 03.09.02-136

Retrieved from "https://kanterella.com/w/index.php?title=ML103610140&oldid=2554457"