NL-19-0009, Supplemental Response to Alternative Request HNP-ISI-ALT-05-04
| ML19008A483 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 01/08/2019 |
| From: | Gayheart C Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19009A112 | List: |
| References | |
| HNP-ISI-ALT-05-04, NL-19-0009 | |
| Download: ML19008A483 (41) | |
Text
~ Southern Nuclear Cheryl A. Gayheart Regulatory Allairs Director 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5316 tel 205 992 7601 fax cagayhea@ southernco.com Enclosure 2 to this letter contains Proprietary Information to be withheld from public disclosure per 10 CFR 2.390. When separated from Enclosure 2, this transmittal document is decontrolled.
January 8, 2019 Docket Nos.: 50-321 NL-19-0009 50-366 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Edwin I. Hatch Nuclear Plant- Units 1 and 2 Supplemental Response to Alternative Request HNP-ISI-ALT-05-04 Ladies and Gentlemen:
By letter dated June 21, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18172A281), and as supplemented by letter dated November 29, 2018 (ADAMS Accession No. ML18333A382), Southern Nuclear Operating Company (SNC) submitted proposed Alternative Request No. HNP-ISI-ALT-05-04 (Proposed Alternative) to certain requirements of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code (ASME Code), for the fifth 10-year lnservice Inspection (lSI) Program for the Edwin I. Hatch Nuclear Plant, Units 1 and 2 (Hatch) . Specifically, pursuant to Title 10 of the Code of Federal Regulations (1 0 CFR) paragraph 50.55a(z)(1 ), SNC requested approval to implement alternative BWRVIP Guidelines in lieu of ASME Code Section XI Table IWB-2500-1 Examination Category B-N-1 and B-N-2 requirements.
Based on a follow-up call held on December 20, 2018 between the Nuclear Regulatory Commission (NRC) staff and SNC, it was requested that SNC transmit the latest Unit 1 core shroud flaw evaluation calculation. Enclosures 2 and 3 contain the proprietary and non-proprietary versions, respectively, of the requested calculation. Enclosure 1 contains the supporting affidavit signed by Electric Power Research Institute (EPRI), the owner of the proprietary information in Enclosure 2. This affidavit sets forth the basis on which the information in Enclosure 2 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390 of the Commission's regulations. Accordingly, it is respectfully requested that the information, which is proprietary to EPRI, be withheld from public disclosure in accordance with 10 CFR 2.390 of the Commission's regulations.
U. S. Nuclear Regulatory Commission NL-19-0009 Page2 This letter contains no NRC commitments. If you have any questions, please contact Jamie Coleman at 205.992.6611.
Respectfully submitted, CAG/RMJ
Enclosures:
- 2. Hatch Unit 1 Core Shroud Leakage Calculation (Proprietary)
- 3. Hatch Unit 1 Core Shroud Leakage Calculation (Non-Proprietary)
Cc: Regional Administrator, Region II NRR Project Manager- Hatch Senior Resident Inspector- Hatch RTYPE: CHA02.004
Edwin I. Hatch Nuclear Plant - Units 1 and 2 Supplemental Response to Alternative Request HNP-ISI-ALT-05-04 Enclosure 1 EPRI Affidavit Requesting Withholding of Proprietary Information
~~~I I ELECTRIC POWER
-=*-*~ RESEARCH INSTITUTE NEIL WILMSHURST Vice President and Chief Nuclear Officer Ref. EPRI Docket No. 99902016 January 7, 2019 DocumentControiDesk Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555-0001
Subject:
Request for Withholding of the following Proprietary Information Included in:
Southern Nuclear Company, E. I. Hatch Nuclear Plant Core Shroud Leakage Evaluation of Hatch Nuclear Plant, Unit 1, Core Shroud Leakage Rate Calculation Revision 1, dated January 2019 To Whom It May Concern:
This is a request under 10 C.F.R. §2.390(a)(4) that the U.S. Nuclear Regulatory Commission ("NRC")
withhold from public disclosure the report identified in the enclosed Affidavit consisting of the proprietary information owned by Electric Power Research Institute, Inc. ("EPRI") identified above in the attached report. Proprietary and non-proprietary versions of the Report and the Affidavit in support of this request are enclosed.
EPRI desires to disclose the Proprietary Information in confidence to assist the NRC review of the enclosed submittal to the NRC by Southern Company. The Proprietary Information is not to be divulged to anyone outside of the NRC or to any of its contractors, nor shall any copies be made of the Proprietary Information provided herein. EPRI welcomes any discussions and/or questions relating to the information enclosed.
If you have any questions about the legal aspects of this request for withholding, please do not hesitate to contact me at (704) 595-2732. Questions on the content of the Report should be directed to Andy McGehee of EPRI at (704) 502-6440.
Sint\Y/
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Together . . . Shaping the Future of Electric ity 1300 West W.T. Harris Boulevard, CharloHe, NC 28262-8550 USA
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- nwilrnshurstOepri.com
~~~~ ~ ELECTRIC POWER
~~-~~ RESEARCH INSTITUTE AFFIDAVIT RE: Request for Withholding of the Following Proprietarylnformation Included In:
Southern Nuclear Company, E. I. Hatch Nuclear Plant Core Shroud Leakage Evaluation of Hatch Nuclear Plant, Unit 1 I Core Shroud Leakage Rate Calculation Revision 1 dated January 2019 I
I, Neil Wilmshurst, being duly sworn, depose and state as follows:
I am the Vice President and Chief Nuclear Officer at Electric Power Research Institute, Inc. whose principal office is located at 3420 Hillview Avenue, Palo Alto, California ("EPRI") and I have been specifically delegated responsibility for the above-listed Report contains EPRI Proprietary Information that is sought under this Affidavit to be withheld "Proprietary Information". I am authorized to apply to the U.S. Nuclear Regulatory Commission ("NRC") for the withholding of the Proprietary Information on behalf of EPRI.
EPRI Proprietary Information is identified in the above referenced report with text inside double brackets. Examples of such identification is as follows:
((This sentence is an example Tables containing EPRI Proprietary Information are identified with double brackets before and after the object. In each case the proprietary notation refers to this affidavit and all the bases included below, which provide the reasons for the proprietary determination. EPRI requests that the Proprietary Information be withheld from the public on the following bases: Withholding Based Upon Privileged And Confidential Trade Secrets Or Commercial Or Financial Information (see e.g. 10 C.F.R. §2.390(a)(4))::
- a. The Proprietary Information is owned by EPRI and has been held in confidence by EPRI. All entities accepting copies of the Proprietary Information do so subject to written agreements imposing an obligation upon the recipient to maintain the confidentiality of the Proprietary Information. The Proprietary Information is disclosed only to parties who agree, in writing, to preserve the confidentiality thereof.
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- d. EPRI's classification of the Proprietary Information as trade secrets is justified by the Uniform Trade Secrets Act which California adopted in 1984 and a version of which has been adopted by over forty states. The California Uniform Trade Secrets Act, California Civil Code §§3426 - 3426.11, defines a "trade secret" as follows:
"'Trade secret' means information, including a formula, pattern, compilation, program device, method, technique, or process, that:
(1) Derives independent economic value, actual or potential, from not being generally known to the public or to other persons who can obtain economic value from its disclosure or use; and (2) Is the subject of efforts that are reasonable under the circumstances to maintain its secrecy."
- e. The Proprietary Information contained therein are not generally known or available to the public. EPRI developed the Information only after making a determination that the Proprietary Information was not available from public sources. EPRI made a substantial investment of both money and employee hours in the development of the Propretary Information. EPRI was required to devote these resources and effort to derive the Proprietary Information. As a result of such effort and cost, both in terms of dollars spent and dedicated employee time, the Proprietary Information is highly valuable to EPRI.
- f. A public disclosure of the Proprietary Information would be highly likely to cause substantial harm to EPRI's competitive position and the ability of EPRI to license the Proprietary Information both domestically and internationally. The Proprietary Information and Report can only be acquired and/or duplicated by others using an equivalent investment of time and effort.
I have read the foregoing and the matters stated herein are true and correct to the best of my knowledge, information and belief. I make this affidavit under penalty of pe~ury under the laws of the United States of America and under the laws of the State of California. Executed at 1300 W WT Harris Blvd, Charlotte, NC being the premises and place of business of Electric Power Research Institute, Inc. Neil Wilmshurst
(State of North Carolina) (County of Mecklenburg) Subscribed and swo111 to :or ~ed) before me on this .2_day of 20.!9 by _.£M.>ot4o.._l~/~4.a...<):..,..o,l.""'b-L/2ZJ..A5-'-'l&"""""~Z:;ti:.>..LJ- _ _ __,, proved to me on the basis of satisfactory evid ce to be the Sla/;) person(s) who agpeared before me. s;gnaru~a 0 My Commission Expires .it2day of /J~Jd
Edwin I. Hatch Nuclear Plant - Units 1 and 2 Supplemental Response to Alternative Request HNP-ISI-ALT-05-04 Enclosure 3 Hatch Unit 1 Core Shroud Leakage Calculation (Non-Proprietary)
l)Structural Integrity Associates, Inc.'" File No.: 1500880.303 Project No.: 1801566 CALCULATION PACKAGE Quality Program: [8::1 Nuclear 0 Commercial PROJECT NAME: Update of Hatch Shroud Flaw Evaluation Calculation CONTRACT NO.: Purchase Order SNC54443-0009, Rev. 0 CLIENT: PLANT: E. I. Hatch Nuclear Plant - Southern Company E. I. Hatch Nuclear Plant CALCULATION TITLE: Core Shroud Leakage Evaluation of Hatch N uclear Plant, Unit 1 - Core Shroud Leakage Rate Calculation Project Manager Preparer(s) & Document Affected Revision Description Approval Checker(s) Revision Pages Signature & Date Signatures & Date 0 1 - 20 Initial Issue Res[!onsible Engineer A A-12 Daniel V. Sommerville Stephen M. Parker 1111116 1111116 Res[!onsible Verifier GoleMukhim 1/ 11 / 16 1 2, 8, 10, 11 ' Added formatting for Res[!onsible En&ineer 13- 16, A-3 , A-7, A-8, proprietary information. ~ 4v~ ~ v~ Matthew Walter A-9 Matthew Walter 1/7/2019 1/7/2019 Res[!onsible Verifier
~
Chris Lohse 1/7/2019 Page 1 of 21 F0306-0 IRI
e Structural Integrity Associates, Inc.* PROPRIETARY INFORMATION NOTICE THIS DOCUMENT CONTAINS EPRI PROPRIETARY INFORMATION. SUCH INFORMATION IS IDENTIFIED IN THE LIST OF REFERENCES AND IS IDENTIFIED IN THE TEXT BY DOUBLE BRACKETS, A REVISION BAR IN THE RIGHT HAND MARGIN, AND RED, ITALIC FONT AS SHOWN IN THE FOLLOWING EXAMPLE (( THIS INFORMA TION IS EPRI PROPRIETARY}} . NOTE: FOR THIS NON-PROPRIETARY VERSION, INFORMATION IN RED BRACKET HAS BEEN REDACTED AS SHOWN IN THE FOLLOWING EXAMPLE, (( } }. File No.: 1500880.303 Page 2 of21 Revision: I F0306-0IRI
e Structural Integrity Associates, /nc.e Table of Contents I.O INTRODUCTION ......................................................................................................... 4 2.0 OBJECTIVE ................................. ................................................................................. 4 3.0 METHODOLOGY ........................................................................................................ 4 3 .I End oflnterval Crack Lengths ........................................................................... 5 3.1 .1 Characterization of Flaws .......... ... .. ...... .. ............ ............... .. ............... .... .. ..... .. . 5 3.1.2 Material Properties ........... ..... .. ....... .......... ..... .............. .. ... .... .. .... ............ .. .. ... ... 5 3.1.3 Inspection Uncertainty ... ...... .. .. ........ ........... .. .. ... ..... .. .... ... ..... ... ....... .... .. .. .. ......... 5 3.1.4 Crack Growth .... ... ...... ....... ..... .. ....... ............... ................................................... 5 3.1 .5 PlasticZoneSize ............................ ........ ............... .. ........................................... 6 3.2 Crack Opening Area ...................................... ...... .............................................. 7 3.3 Leakage Rate Multiplier .................................................................................... & 3.4 Leakage Rate and Loss Coefficient Calculation ................................................ & 4.0 DESIGN INPUTS .......................................................................................................... 9 5.0 ASSUMPTIONS .......................................................................................................... II 6.0 CALCULATIONS AND RESULTS ........................................................................... I2
7.0 CONCLUSION
S ................................................ ......................................................... I3
8.0 REFERENCES
.............................................................................................. .............. I3 Appendix A CALCULATIONS AND SUPPORTING FILES ............................................ A-1 List of Tables Table 1. Uninspected Region Leak Rate Multiplier ................................................................ I6 List of Figures Figure I. Single Through-wall Axial Crack in an Internally Pressurized Cylinder ............... 17 Figure 2. Example of an Elliptical-Shaped Crack .................................................................. 17 Figure 3. HNP 1 Core Shroud Configuration [22] .................................................................. 18 Figure 4. Previous HNP1 Core Shroud Inspection Results [22] ............................................ I9 Figure 5. HNP1 V6 Weld Co-linear Indication Results ......................................................... 20 Figure 6. FEA Model Boundary Conditions ........................................................................... 21 Figure 7. FEA Model Pressure Loads ..................................................................................... 21 File No. : 1500880.303 Page 3 of21 Revision: 1 F0306-0 IR I
e Structural Integrity Associates, /nc.a
1.0 INTRODUCTION
Multiple reportable indications were detected in the E. I. Hatch Nuclear Plant, Unit 1 (HNP1) core shroud during the Spring 2014 refueling outage. Numerous axially oriented indications were reported on the outside and inside surfaces of the core shroud in the vicinity ofthe core shroud circumferential weld H4 as well as multiple co-linear flaws in the core shroud axial welds V5 and V6. These flaws were evaluated and leakage calculations were performed to determine their acceptability for continued operation [ l]. The purpose ofthis evaluation is to update the leakage calculations performed in Reference [I] to consider updated methodologies and the behavior of long co-linear cracks. In Reference [2], a plant-specific analysis was performed to investigate whether very long flaws would result in crack opening areas (COA) that are not adequately predicted using the existing handbook fracture mechanics solutions. The resulting finite element analysis (FEA) calculated larger COAs for large cracks than those calculated using the existing handbook solutions. The Reference [2] study was extended to investigate the behavior of co-linear flaws identified in the HNPI core shroud [3]. The results of these plant-specific studies are incorporated in this updated core leakage calculation. The leakage calculations also incorporate the latest methodology developed in References [4, 13] which accounts for the effects of crack morphology and treatment of through-wall cracking in uninspected regions of the core shroud, respectively. 2.0 OBJECTIVE The objective ofthe work documented in this calculation package is to revise the HNPI core shroud combined leakage evaluation for the H4-V4 weld intersection and V6 weld through-wall indications documented in Reference [1], considering the following:
- 1. The results of the analyses performed in References [2, 3] to account for the effects of long single and multiple co-linear cracks.
- 2. A three cycle operating interval.
- 3. The results of the updated core leakage methodology documented in References [4, 13].
3.0 METHODOLOGY The leakage calculation is performed using methods consistent with those given in BWRVIP-76, Rev. 1 [5]. The leakage calculation is performed using the following process:
- 1. Calculate end of interval crack length.
- 2. Calculate end of interval COA.
- 3. Determine leakage rate multiplier
- 4. Calculate through wall leakage rate.
Each item is addressed separately as follows. File No. : 1500880.303 Page 4 of21 Revision: l F0306-0IRI
e Structural Integrity Associates, Inc.* 3.1 End of Interval Crack Lengths To predict the crack length at the end of three cycles of operation (6 years), the following is considered: I. Location and dimensions ofthe flaws.
- 2. Applicable material properties.
- 3. Flaw sizing uncertainty associated with the core shroud weld inspections.
- 4. Crack growth rate due to degradation mechanisms.
- 5. Plastic zone size due to small-scale yielding at the crack-tip Each ofthese aspects is described separately below.
3.1.1 Characterization of Flaws The number, orientation, and dimensions of the reported indications are obtained from the inspection notification reports (INRs) provided by General Electric Hitachi Nuclear Energy Americas, LLC (GEH) [6]. The INRs are also used to infer the likely initiation and growth mechanism for the reportable indications which is necessary in order to identify which crack growth mechanisms to consider in the evaluation. The as-measured lengths ofthe through-wall indications identified in the HNPI core shroud H4-V4 weld intersection and V6 weld [6.j, 6.k] are the initial crack lengths used for this calculation. 3.1.2 Material Properties Tensile properties are selected at a temperature and fluence level such that the plastic zone sizes are bounding. BWRVIP-IOO-A [7] and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section II, Part D [8.a] are also used to select the appropriate tensile properties for the fluence level and temperature. The bounding fluence along the entire core shroud height at 49.3 effective full power years (EFPY) [9] is conservatively used for this evaluation. The current refueling outage is l R27 and the end of design life EFPY is reported to be 49.3 EFPY in Reference [9] which corresponds to refueling outage I R36. 3.1.3 Inspection Uncertainty Inspection uncertainties provided in the inspection method demonstration documentation, appropriate to the inspection method and delivery system, are conservatively applied to the length of all reportable indications [I 0]. 3.1.4 Crack Growth Consistent with the methods provided in BWRVIP-76, Rev. I [5] and the clarifying guidance given in References [4, II] intergranular stress corrosion crack growth (IGSCC) is calculated for the evaluation interval and added to each end of each reportable indication. Fatigue crack growth is not a relevant mechanism for the core shroud [5, Section K.3]; therefore, IGSCC is the only relevant crack growth mechanism. File No.: 1500880.303 Page 5 of2I Revision : I F0306-0IRI
e Structural Integrity Associates, Jnc.c The IGSCC length crack growth rate provided in BWRVIP-76, Rev. 1 [5] and BWRVIP-14-A [11] is used for all flaws: da =S.O*lO-s . in dt hr For a three cycle operating interval (six years), the resulting total crack growth is 2.63 inches. 3.1.5 Plastic Zone Size The radius of the plastic zone size is added as an additional crack length at each end of each flaw. The plastic zone size correction is estimated, for conditions of small scale yielding, using the following equation [ 12, pg. 16]: 2 ry-- 1 * ( -K1-) Equation (1) 2*n <J'Yield Where is the Mode I stress intensity factor, ksi-in° 5 a is used to adjust for plane strain or plane stress conditions at the crack tip, where: Plane Strain: a=1/6rr Plane Stress: a=l/2rr OYield is yield stress, ksi For the off-axis parallel flaws identified in the V4/H4 weld intersection, the stress intensity factor, K1, is determined using a linear-elastic fracture mechanics (LEFM) solution as presented in Reference [12, pg. 582] for a single longitudinal through-wall crack in a cylinder (see Figure 1). Use of the single flaw equations to calculate K1 for parallel flaws is conservative since the K1 solutions for parallel flaws result in smaller values [12, pg. 256]. The equations used are: K1 = SF
- a*Jrra
- F(A.) Equation (2)
Where: SF is the structural factor for operating conditions. 2.77 for Normal (Level A) and Upset (Level B) conditions. 1.39 for Emergency (Level C) and Faulted (Level D) conditions, unitless R (J' = p- is the hoop stress, ksi t' a half crack length (see Figure 2), in. F(A.) = (1 + 1.25A2 /fz 0 ::::; A::::; 1
= 0.6 + 0.9A 1 ::::; A ::::; 5 A.= a/.../Rt File No.: 1500880.303 Page 6 of21 Revision: 1 F0306-0IRI
e Structural Integrity Associates, Inc.* For the co-linear flaw identified in the V6 weld, the stress intensity factor, Kt, used is equal to the fracture toughness, K1c, for the irradiated core shroud material. As Kt cannot be greater than K1c for this calculation, this approach is conservative. 3.2 Crack Opening Area The COA is determined for the crack configuration presented in Figure 1 and Eq. (3) below [12, Part VII]. The handbook solution for COA has been shown to result in lower values than those from a finite element analysis [2] for large cracks. Therefore, the handbook solution results are adjusted by applying the factors FFEA and FcL to account for the effects of large single and multiple co-linear cracks, where applicable [2, 3]: A=;, (2rrRt)
- G(il.)
- FFEA
- FcL Equation (3)
Where: A is the crack opening area, in 2 CJ = p-Rt is the hoop stress, ksi p is the internal pressure difference across the core shroud, ksi R is the core shroud mean radius, in. t is the core shroud thickness, in. E (plane stress) E' = E ( (l-v 2 ) (plane strain) Note: For this application, plane stress is conservative. Therefore, E' is equal to the modulus of elasticity [12, Part VII]. G(A.) = A.2 + 0.625A.4
= 0.14 + 0.36A.2 + 0.72A.3 + 0.40SA.4 A.= a/...fRt a is the half crack length (see Figure 2), in.
FFEA is the adjustment factor to account for the difference in COA calculated using FEA [2] for long cracks. This factor is equal to one plus the percent difference between the FEA and LEFM COA calculations, unitless FcL is the adjustment factor to account for the effect of co-linear cracking on the COA, unitless File No.: 1500880.303 Page 7 of21 Revision: 1 F0306-0IRI
e Structural Integrity Associates, Inc.<< 3.3 Leakage Rate Multiplier Guidance is available in Reference [13] that can be used to determine the extent of off-axis, through-wall cracking assumed to exist in uninspected core shroud weld regions. Table I excerpted from Reference [13, Table 1], provides leak rate multipliers that can be used to determine the total calculated leakage rate at the end of interval for the entire weld. The examination coverage is defined as the inspected weld length divided by the total weld length. Leakage rate multipliers are available for EVT-1 and UT examinations at various percentiles of examination coverage. The multiplier used in this evaluation is provided in Section 6.0. 3.4 Leakage Rate and Loss Coefficient Calculation The volumetric leakage rate through a through-wall crack can be determined using the methodology given in Reference [4]: Equation (4) (( }} Equation (5) Where: 'Y is the coolant density, lbf/ft3 g is the gravitational acceleration, in/s2
~p is the pressure differential across the length of the crack, psi T is the thickness of the reactor internal component, in 8 is the average COD, equal to COA I L. This is an approximation based on a rectangular COA shape which results in an average COD larger than that for an elliptically shaped COA, in COA is the crack opening area (See Section 3.2), in 2 L is the total crack length, in is the number of turns per unit length, ((
}}, unitless This equation is consistent with the equation given in BWRVIP-76 [5] and includes the ability to consider head loss from additional factors not considered in BWRVIP-76 [5], such as IGSCC morphology. Based on discussions in Reference [4], NRC studies [14, 15, 16] have shown that crack morphology parameters along the crack flow path have a significant impact on predicted leak rates. The crack morphology parameters are the surface roughness, the number ofturns the crack makes, and the ratio of the actual flow path length to the thickness of the pipe. The two software programs most commonly used to compute leakage rates in the nuclear industry are PICEP [17] and SQUIRT [18].
File No.: 1500880.303 Page 8 of21 Revision: 1 ,..----- - -- - - -- - - - - - - - - - -- -, (( This information }} is EPRI Proprietary Information F0306*01Rl
S}Structural Integrity Associates, Inc.<< These software programs calculate fracture mechanics parameters such as the crack opening displacement (COD) and the leakage rates. These computer codes have been benchmarked against experimental data. Both PICEP and SQUIRT are based on the same thermal hydraulics model [19, 20]. Both of these codes have been mainly used for the analysis of Leak-Before-Break (LBB) where more leakage is desirable from a detection point of view. Considerable conservatism is added to the LBB leakage calculations by imposing a factor of 10 on the leakage rate calculation and a factor 2 on the crack length considered for leakage. The problem of leakage through the BWR reactor internals is much simpler in that there is no likelihood of water flashing in the path and hence the flow is single phase. This is true even if the reactor coolant is subcooled since the pressure drop between the inside and outside of the core shroud, for example, is on the order of 30 psi. Therefore, simple fluid mechanical relationships are employed to calculate the head loss due to friction and crack morphology. The approach used herein is consistent with that discussed in Reference [4]. A note of caution about units in Eq. (4). In FPS (Foot, Pound and Second) units, density is given in terms of a force unit i.e. y = p g. In MKS (Meter, Kilogram and Second) units, the density p is in mass units and hence the g terms cancel out and the denominator will just be p on the right hand side of Eq. (4). 4.0 DESIGN INPUTS The design inputs used for this calculation are identified below: Geometry: The shroud geometry is taken from Reference [21]:
- Shroud ID: 174.5 in. [2 I.a]
- Shroud Thickness, t: 1.5 in. [2 I.b]
Loads: The upper shroud RlPD values used for this analysis are taken from Reference [22], and are summarized as follows:
- Level C RIPD: 29.5 psi
- Level D RIPD: 29.5 psi For the purposes of this evaluation, the Level C and D RIPD values are used as they yield the largest crack opening displacement.
IGSCC Crack Growth Rate:
- The length COR: 5 X 1o-s in/hr (Section 3 .1.4)
File No.: 1500880.303 Page 9 of21 Revision: 1 F0306-0IRI
lJ Structural Integrity Associates, /nc.c Reactor Coolant Water Chemistry: HNPI implemented HWC in September I987, and started Noble Metal Chemical Addition (NMCA) in March of I999 [22, 23]. ((
}}.
Shroud Fluence: The peak core shroud fluence at 49.3 EFPY is: 3.77 x I0 21 n!cm 2 [9] Material Type: The shroud material is SA-240 TP304 stainless steel [2l.a]. Material Properties: For fluence values greater than ((
}}, and the following static initiation plane strain, mode I I fracture toughness is applicable:
(( }} The material flow stress and yield stress both increase with fluence [7]. It is conservative to use un-irradiated materials properties since this will result in a larger plastic zone size and a smaller allowable flaw size. Consequently, un-irradiated tensile properties are used [8.a]:
- cry (un-irradiated, 550 "F): I8.9 ksi
- Elastic Modulus of Type 304 Stainless Steel at 550°: 25.55 x I 06 psi
- Poisson's Ration of Type 304 Stainless Steel at 550°: 0.3 For the leakage rate calculations the fluid density is:
- Coolant density, r: 53.9 lbf/ft3 , water at 400 °F [25] (see Assumption 5.4)
Initial Flaw Distribution: The flaw lengths, depths, locations, and distributions are taken from the 20I4 INRs [6]. It is only necessary to consider through-wall indications for the leakage calculation, which were identified at the H4-V4 weld intersection [6.j] and the V6 weld [6.k] . Figure 3 and Figure 4 illustrate the HNPI core shroud configuration, weld locations, and identify historically reported indications. Three unique indications were identified in the V6 weld. Indications 2 and 3 are the same flaw identified on the inside and outside of the shroud wall, confirming that the indication is through-wall [6.k, pg 2 of2]. Co-linear part through-wall indications exists above and below the through-wall flaw. Indication I is 3.5 inches long and has an initial ligament length of I8.5 inches to indication 2. The start of indication 4 is at 90 inches and the end of indication 2 is at 86.3, resulting in an initial ligament length of 3.7 inches between the flaws. Due to the small ligament length and projected crack growth over three File No.: 1500880.303. -- - - - -- - - - - - - - - - -- - - . Page IO of2I Revision: I (( This information } } is EPRI Proprietary Information F0306-01Rl
e Structural Integrity Associates, /nc.a cycles, the cracking identified in indications 2, 3, and 4 are considered to be a single indication starting at 65.9 inches and ending at 96.I inches, with a total initial length of30.2 inches. See Figure 5 for details regarding the V6 weld flaw configuration. The combined length of indications 2, 3, and 4 is evaluated with indication I to consider the effects of co-linear flaws on the crack opening. Inspection Uncertainty: Evaluation factors to account for inspection uncertainties for the ultrasonic testing (UT) inspection data are taken from the applicable demonstrations for the inspection tooling identified in the INRs. The evaluation factors are taken from BWRVIP-03 [IO.a] and BWRVIP Letter 20I4-0I5 [IO.b]. Bounding length evaluation factors are taken from the applicable demonstrations. No depth evaluation factors are used since only through-wall flaws are considered for the leakage rate evaluation. The following length evaluation factors are used:
- TEIDE Tool UT Length Evaluation Factor: (( }}
- Shroud 00 UT tool Length Evaluation Factor: (( }}
Operating Cycle Duration: HNPI is on a 2 year operating cycle [22]. In this calculation, the evaluation is perfonned for 6 years. 5.0 ASSUMPTIONS The following assumptions are used in this evaluation. I. The 49.3 EFPY peak fluence is used to detennine the fracture toughness for all flaws. This assumption is conservative because it applies the maximumfluence projected at the end of the operating license to all locations. The current refueling outage is 1R2 7 and the end of design life EFPY is reported to be 49.3 EFPY in Reference [9], which corresponds to refueling outage 1R36.
- 2. A I 00% capacity factor is assumed for crack growth.
This assumption is conservative because it uses the largest number of hours possible, each year, for crack growth.
- 3. Flaws are allowed to grow through the horizontal welds.
Inspection data from the HNP 1 core shroud shows evidence offlaws growing through the weld [6]; therefore, this assumption is considered appropriate. File No.: 1500880.303 Page II of21
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Revision: I { { This information }} is EPRI Proprietary lnfonnation F0306-0IRI
l} Structural Integrity Associates, Inc.~
- 4. Leakage rates are calculated using a conservative fluid temperature of 400 °F.
In the event of emergency core cooling, initially the fluid from the feedwater system enters the core. The normal operatingfeedwater temperature for HNPI is 400 oF [28}. Supplemental cooling is provided by fluid from the suppression pool and condenser fluid, which operate at temperatures lower than the feedwater system. For this reason, use of the normal operating feedwater temperature is a conservative upper bound temperature for the purposes of determining the fluid density. 6.0 CALCULATIONS AND RESULTS Calculations are performed using the MathCAD software [26], ANSYS Mechanical APDL [27], and MS Excel. The calculations performed for this core leakage evaluation and associated supporting files are identified in Appendix A. The COA for the core shroud through-wall flaws are calculated using LEFM methods and adjusted using the results from Reference [2] by applying multiplication factors to account for the increased COA values calculated using FEA methods. The COAs for the flaws identified in the H4-V4 weld intersection are increased by 3.43% (FFEA= 1.0343) and the flaw in the V6 weld is increased by 12% (F FEA= 1.12) based on the values summarized in the supporting files in Reference [2]. These percent increases are based on the predicted crack length at the end of three cycles. After calculating the crack lengths at the end of three operating cycles, the flaw configuration for the co-linear flaws identified in the V6 weld was determined. The end of interval crack length of the through-wall flaw is approximately 38.0 inches with a ligament length of 10.7 inches to a 11.3 inch co-linear part through-wall flaw. The inspection results of these co-linear flaws, aligned end-to-end in series and parallel to the V6 axial weld, are shown in Figure 5. The COA of this co-linear flaw configuration is calculated in ANSYS by modifying the finite element model created in [3]. SI performed a plant specific FEA of the V6 co-linear flaws rather than using the scaling factor approach developed in Reference [3] because the end of interval flaws in the V6 weld are slightly longer than the range of flaws considered in Reference [3]. Rather than extrapolate the results in Reference [3] or revise the previous work documented in Reference [3], all of which remains valid, SI decided to use the model and methodology documented in Reference [3] to perform a plant specific, flaw specific analysis of the V6 co-linear flaw configuration. The boundary conditions and loads applied to the model are shown in Figure 6 and Figure 7, respectively. The resulting COA for the through-wall flaw is 0.460 im and is similar to that calculated for a single flaw configuration, including an adjustment factor FFEA= 1.12, which is 0.466 im. As the single flaw configuration results in a larger COA and leakage rate, it is conservative to disregard the effect that the co-linear part through-wall flaw has on the through-wall flaw. Therefore, no adjustment (i.e., FcL= 1.0) is needed for the V6 weld through-wall flaw to account for the effects of the co-linear flaw configuration. File No.: 1500880.303 Page 12 of21 Revision: 1 F0306-0IRI
e Structural Integrity Associates, /nc.e Leakage rates are calculated for the four through-wall flaws reported in the H4-V4 intersection examined in 20I4, which consists of a 28.8 inch window (-5.2%) [6.j, Sht. 5] out of -552 inch shroud circumference at this elevation using a leakage rate multiplier of26 as described in Section 3.3 and summarized in Table I. The through-wall indication in the V6 weld is also included in the total leakage rate calculation. Leakage rates are calculated for three cycles of operation. The cumulative leakage rate for the core shroud reported and assumed through-wall flaws after three cycles of operation is 49.I GPM.
7.0 CONCLUSION
S The cumulative leakage rate calculated for all reported and assumed through-wall cracks in the HNP I core shroud, after three operating cycles of 2 years each, is 49.1 GPM. The through-wall cracks contributing to the leakage rate are those located in the core shroud H4-V4 weld intersection and V6 weld. This leakage rate should be combined with other applicable leakage rates and compared to the current LOCA leakage margin.
8.0 REFERENCES
- 1. Sl Calculation Package 1200283.303, Rev. 1, "Hatch Nuclear Power Plant Unit 1 Core Shroud Axially Oriented Flaw Evaluation and Leakage Rate Calculation- H4 Weld, Base Material Flaws between H3 and H4 Weld, V7 Flaws, V8 Flaws."
- 2. Sl Calculation Package 1500880.301, Rev. 0, "Core Shroud Leakage Evaluation of Hatch Nuclear Plant, Unit I - Crack Opening Area Calculation."
- 3. Sl Calculation Package 1500880.302, Rev. 0, "Core Shroud Leakage Evaluation of Hatch Nuclear Plant, Unit 1 - Co-linear Flaw Crack Opening Calculation."
- 4. SI Calculation Package 1500205.305, Rev. 0, "Recommended Methodology for Calculating Leakage Flow Rates through IGSCC cracks in Austenitic Stainless Steel BWR Internals."
- 5. BWRVIP-76, Revision 1-A: BWR Vessel and Internals Project, BWR Core Shroud Inspection and Flaw Evaluation Guidelines. EPRI, Palo Alto, CA: 2015. 3002005566. EPRI Proprietary Information.
- 6. H1R26 Inspection Results:
- a. INR H1R26 IVVI-14 Shroud OD Indications observed during VT-3
- b. INR H 1R26 IVVI-14-06 -Core Shroud Plate Indications
- c. INR H1R26 IVVI-I4 Shroud !D-OD-Vertical Weld 4, Horizontal Weld 4 Intersection
- d. INR H1R26 IVVI-I4 Shroud ID Indications observed during VT-3
- e. INR H 1R26 IVVI-14 Shroud ID Vertical Weld V8 @ H5
- f. GE Hitachi Customer Notification Form CNF -SHRD-00 1 RO
- g. GE Hitachi Customer Notification Form CNF-SHRD-002 R1
- h. GE Hitachi Customer Notification Form CNF-SHRD-003 RO File No.: 1500880.303 Page 13 of21 Revision: 1 F0306-0IRI
S}Structural Integrity Associates, Inc:~
- 1. GE Hitachi Customer Notification Form CNF-SHRD-004 RO J. GE Hitachi Customer Notification Form CNF-SHRD-005 R1
- k. GE Hitachi Customer Notification Form CNF-SHRD-006 RO I. GE Hitachi Customer Notification Form CNF-SHRD-007 RO
- m. GE Hitachi Customer Notification Form CNF-SHRD-008 RO
- 7. BWRVIP-100-A: BWR Vessel and Internals Project, Updated Assessment of the Fracture Toughness of Irradiated Stainless Steel for BWR Core Shrouds. EPRI, Palo Alto, CA: 2006. 1013396. EPRI Proprietary Information.
- 8. ASME Boiler & Pressure Vessel Code
- a. Section II Part D, 200 I Edition.
- b. Section XI, 20 I 0 Ed.
- 9. Transware Report No. SNC-HA1-003-R-001, Revision 0, "Edwin I. Hatch Unit 1 Fluence Evaluation at End of Cycle 26 and 49.3 EFPY," SI File No. 1500880.202P. TransWare Proprietary Information.
- 10. Inspection Uncertainty:
- a. TR-105696-R17 (BWRVIP-03) Revision 17: BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines. EPRI, Palo Alto, CA: 2014.
3002003091. EPRI Proprietary Information.
- b. BWRVIP Letter 2014-015, New NDE Demonstrations, February 3, 2014, SI File No.
1200283.207. EPRI Proprietary Information.
- 11. BWRVIP-14-A: BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Stainless Steel RPV Internals. EPRI, Palo Alto, CA: 2008. 1016569. EPRI Proprietary Information.
- 12. Tada, H., Paris, P., Irwin, G., The Stress Analysis of Cracks Handbook, 3rd. Ed., ASME Press. 2000.
- 13. SI Calculation Package 1500205.303, Rev. 0, "Guidance for Assumption of 'Off-Axis' Flaws in Uninspected Regions of Core Shrouds."
- 14. P.M. Scott, R.J. Olson and G.M. Wilkowski, "Development ofTechnical Basis for Leak-Before-Break Evaluation Procedures," NUREG/ CR-6765, May 2002, SI File No. 1500205.201.
- 15. D.S. Kupperman et al., "Barrier Integrity Research Program," NUREG/CR-6861, ANL-04/26, December 2004, Sl File No. 1500205.201.
- 16. S. Rahman, N. Ghadiali, D. Paul and G. Wilkowski, "Probabilistic Pipe Fracture Evaluations for Leak-Rate-Detection Applications," NUREG/ CR-6004 BMI-2174, April 1995, SI File No.
1500205.201.
- 17. D. Norris, B. Chexal, "PICEP: Pipe Crack Evaluation Computer Program," EPRI Report NP 3596-SR Rev 1, July 1987, Sl File No. 1500205.201.
File No.: 1500880.303 Page 14 of21 Revision: 1 F0306-0IRI
e Structural Integrity Associates, Inc.~
- 18. Paul, D.D. et al., "Evaluation and Refinement of Leak-Rate Estimation Model, "NUREG/CR 5128, Rev.1, June 94, SI File No. 1500205.201.
- 19. Henry, R.E., and Fauske, H.K., "Two Phase Critical Flow at Low Qualities," Nuclear Science and Engineering, il_, 79-98, 1970, SI File No. 1500205.201.
- 20. Henry, R.E., "The Two-Phase Critical Discharge of Initially Saturated or Subcooled Liquid,"
Nuclear Science and Engineering, il_, 336-342, 1970, SI File No. 1500205.201.
- 21. Hatch Nuclear Plant Unit 1 Core Shroud Drawings:
- a. Southern Nuclear Company Drawing No. 1-BN-6-6, Revision 1, "Core Shroud Weld Section and Details," SI File No. HTCH-02Q-205.
- b. Southern Nuclear Company Drawing No. 1-BN-6-5, Revision 2, "(Inside View) Core Shroud Weld Identification Layout," Sl File No. HTCH-02Q-205.
- 22. Design Input Request, Revision 1, SI File No. 1200283.200.
- 23. Southern Nuclear Company Nuclear Management Guideline, NMP-CH-005-GL01 , Version 5.0, "E.
I. Hatch Water Chemistry Strategic Plan," Apri12012, SI File No. 1200283.206.
- 24. BWRVIP-62 Revision 1: BWR Vessel and Internals Project. Technical Basis for Inspection Relieffor BWR Internal Components with Hydrogen Injection. EPRI, Palo Alto, CA: 2011. 1022844. EPRI Proprietary Information.
- 25. NIST Chemistry WebBook, http://webbook.nist.gov/chemistrylfluid/.
- 26. Mathcad Version 14.0 M035 (14.0.3.374), Parametric Technology Corporation.
- 27. ANSYS Mechanical APDL and PrepPost, Release 14.5 (w/Service Pack 1), September 2012.
- 28. General Electric Report, NEDC-32800 GE-NE-A13-00402-9A, Revision 0, "Extended Power Uprate Evaluation Task Report for Edwin Hatch Plant Unit 1," SI File No. GPC0-42Q.241P. GE Proprietary Information.
File No.: 1500880.303 Page 15 of21 Revision: 1 F0306-0IRI
S)Structural Integrity Associates, lnc.c Table 1. Uninspected Region Leak Rate Multiplier File No.: 1500880.303 Page 16 of21 Revision: 1 ...-----------------------, (( This infomwfion }} is EPRI Proprietary Information F0306-01Rl
e Structural Integrity Associates, Inc.* Figure 1. Single Through-wall Axial Crack in an Internally Pressurized Cylinder Figure 2. Example of an Elliptical-Shaped Crack File No.: 1500880.303 Page 17 of21 Revision: 1 F0306-0IRI
S) Structural Integrity Associates, /nc.e
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File No.: 1500880.303 Page18of21 Revision: I F0306-0IRI
~Structural Integrity Associates, /nc.c
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$-1~ CIIENEIIM. EUCTJIJC) 1-BN SA Figure 4. Previous HNPl Core Shroud Inspection Results [22]
File No.: 1500880.303 Page 19 of21 Revision : I F0306-0IRI
S) Structural Integrity Associates, /nc.c 230° I H-4
#4 - Length = 3.5 in /
-1U lf7.lln T
30.2 in
#2&3 - Length = 20.3 in '
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#1 -Length = 3.5 in Vt V-6 Figure 5. HNPl V6 Weld Co-linear Indication Results File No.: 1500880.303 Page 20 of21 Revision: 1 F0306-0IRI
l}Structumllntegrity Associates, Inc.* Figure 6. FEA Model Boundary Conditions (The cyan and orange symbols along the perimeter of the model indicates symmetry constraints) ELtHE!n'S Ill PIUS - NOR."!
- 8~7. !58 -660. 74~ - 463.532 - 266.319
- 69.1065
-759 . 352 -562.139 -364.9215 -167.113 29.5 Hatch Shroud VS/V6 WQld ~Qakaq9 Figure 7. FEA Model Pressure Loads File No.: 1500880.303 Page 21 of21 Revision: 1 F0306-01Rl
e Structural Integrity Associates, lnc.c Appendix A CALCULATIONS AND SUPPORTING FILES File No.: 1500880.303 Page A-1 of A-12 Revision: 1 F0306-0IRI
e Structural Integrity Associates, Inc.* File Name Description ANSYS Mechanical APDL input file that calculates the Hatch-Colinear.inp input Loop.inp file for a ligament length of 12.3 inches. Input script is provided in Appendix A . ANSYS Mechanical APDL input file that defines builds the FEA model, defines loads and boundary conditions, and solves the resulting displacements of the crack openings. Input script is provided in Appendix A. Hatch-Colinear.inp This file was adjusted by changing the value of the crack length definition parameters "crack", " cocrackl ", and "cocrack2" to define the co-linear crack configuration in the HNPl V6 core shroud weld. ANSYS Mechanical APDL output file that reports the maximum displacement of Displacement.txt the crack opening the HNPl co-linear crack configuration. Output results are provided in Appendix A. Ligament Length MS Excel spreadsheet to calculate the ligament length between the co-linear Calc.xlsx cracks studied in the HNPl V6 core shroud weld. File No.: 1500880.303 Page A-2 of A-12 Revision: 1 F0306-0IRI
e Structural Integrity Associates, Inc.* Responsible Engineer. Stephen Parker Date: JO()CT2015 Plant: HNP1 Calculation
Title:
HNP1 core shroud leakage evaluation Objective: Perform* 1 Leakage calculation for the through-wall indications. Inputs: 174 Ri :- ..5 *in Inside radius of shroud at H4 location 2 t := 1.5-in Wall thickness of shroud at H4 location RIPDab ::= 11.6-psi Level AlB RIPD RIPDcd := 29..5-psi Level C/0 RIPD SFab :- 2.77 Level AlB SF SFcd ::= 139 Level C/0 SF <-ONLY THIS VALUE IS USED
-5 in CGR := 5-10 -- SCC length crack growth rate hr ay :- 18900-psi Yield strength of shroud material, unirradiated, 550 F.
6 E ::= 25.55 x 10 -psi Elastic modulus. 550 F LEF :-{ Length evaluation factor in 1!: ;z: 386.6--
..w. 2 Newton's proportionality constant s
lbf
-y :- 53.9- Density of water at 400 F ft3 f := .02 Friction factor for loss coefficient
{ } { } File No.: 1500880.303 Page A-3 of A-12 Revision: 1 { T11is infonnation } is EPRI Proprietary Information F0306-01Rl
SJ Structural Integrity Associates, /nc.c Calculations: 1_Calculate leakage at end of three operating cycles for the V4/H4 weld intersection through-wall flaws _ Flaws- 8 0" at V41H4 @ --15 inches 5.5" at V4/H4 @ --6 inches 6.2" at V4/H4 @ --1 inches 5.3" at V4/H4 @ -3 inches Rm= 88-in (Rm) a:= RIPDcd-- - t 8_0] 55 1 i := in
- 62
[ 53 13..554] 1 f := 1 i + 2-LE.F + 2-6-yr-CGR-365 day l- ~ 11.054
- - yr day 1f = -in
- 11.754
[ 10.854 Iterate to find plastic zone size correction factor on a flaw specific basis for the flaws in the V41H4 weld intersection: i := 0,1.. 3 Iteration 1: l_f 2 >.. = 0.59 ] 0.481 [
>.. := - - -
(Rm*t)Oj 0..512 0.472 o- l.l98]
.!(>-) :=
2 (1 + 125->.. ) .) if 0 :S >.. :S 1 F(>--) = l.l35 ((0.6 + 0.9->..)) if 1 :S >.. :S 5 1 [ 1.152 1.131 File No.: 1500880.303 Page A-4 of A-12 Revision: 1 F0306-0IRI
e Structural Integrity Associates, Inc.* 13296] Kl1(~.1_f) := SFcd*u*[ 11*( 1 2f)~0.5
~ *F(~) KI1(~-.1 f.)=
113&2
*psi -in 0.5 1 - 1
[ 11908 11233 2 0.079] ryl(~.1_f) := _1_ _(KI1(~.l f)) ry1 (~-.1 f.) = 0.058
*in 2 *11 uy 1 - 1* 0.063
[ 0.056 Iteration 2: 1f
-=- + ry1(~.1 f ) 0.597]
2 - 0.486 1f. =
~~.l_f) := - - - - -
(Rm*t)05
>.2{ ~-
1' - 1) 0.517 [ 0.477 05 F2(~.1_f) := {1 + 125*~~.1_f) 2) if 0 5; ~~.l_f) 5; t ((0.6 + 0.9- ~~ . l_f))) if 1 5; ~~ . l_f) 5; 5 1202] 1.138 n{ ~- 1' 1 f 1).
= 1.155
[ 1.133 Kll(~ . 1_f) := SFcd*u *[ ?r *( 2+ 1f )~0.5 ryl(~ . l_f) ~ *F2(~.1_f) 13421 ] 11468 0.5 f(t>/ ~ .* 1 f .) = *psi*in
"\ 1 - 1 12003
[ 11317 0.0&] ryl(~.l_f) := -*- *(Kll(~.l_f))2 ry2( ~-. 1 f -) = 0.059 0.064
*in 2*?1 uy 1 - 1
[ 0.057 1.879] 152 PC( ~ ' 1- f ) ;... ry2( ~ . 1 f ) - ry1 (~ . 1 f ) *100 PC{ ~- 1f. = 1.613 ry 1 (~ ,1_f ) 1' - 1) [ 1.495 File No.: 1500880.303 Page A-5 of A-12 Revision: 1 F0306-0IRI
e Structural Integrity Associates, Inc.* JgJ!>-.,1_f) := KI2()., l_f)
~M>-. . 1_£) := ry2(>-.,1_f)
Iteration 3: 1f
-=-
2
+ ry1(). ,1 f )
0597] 0.486 1 f)*=----- ).. 1 f . =
).?f ). /t':,W. * -
- o- >..2{ t' - 1) 051 7 (Rm*t) ..) [
0.477
- 05
,~>-..1_f) := ( 1 + 1.25*>.2(>-.,l_f)2) if 0 ~ >.2(>-.,l_f) ~ 1
((0.6 + 0.9->.2(>-.,1_f))) if 1 ~ >.2().,1_f) ~ 5 1.202] 1.138
=
F2{ ).. 1 f . t ' - 1) [ 1.155 1.133
~). , 1_f) := SFcd*CT*[ 'lf*( l 1f
+ ryl(>-.,1_f)
)~05
~ *F2(>-. ,1_f) 13423]
11470 05 JCnf >-..,1 f .) = *psi*in A\ 1 - 1 12005 [ 11318 0.08 ]
~>-..1_f) := _1 *(Kll(>-. ,1 f ))2 ry2( >-..,1 f .) =
0.059 0.064
-in 2*'lf cry - 1 - 1
[ 0.057 0.035] PC'./). 1 f ) := ry2(). ,1 f ) - ry1(>-. ,1 f) *100 0.023 MNJ> * - ry 1(). 1 f ) PC( ).. 1 f . 1' - 1)
=
[ 0.026 0.022 ry(>-. ,1_f) := ry2( ). ,1_f) 13.715] l_f_ry(>-. ,1_f) := 1_f + 2ry(). ,1_f) 11.171
- -ry(
1f ).. 1 f . = 1' - 1) [ 11.882 in 10.968 File No.: 1500880.303 Page A-6 of A-12 Revision: 1 F0306-0IRI
s; Structural Integrity Associates, /nc.c 2 + ry(>.,1_f) 1f 0j97] 0.486
>.f(>.,1_f) : - - - - - -
(Rm*t)0.5 ( I'1-f I). -
},.f }...
[ 0.517 0.477 2 4 G(>.,1_f) :- >.f(>.,1_f) + 0.625*>.f(>.,1_f) if 0 ~ >.f{}..,l_f) ~ 1 2 3 4 0.14 + 0.36->.f{>.,1_f) + 0.72*>.f(>.,1_f) + 0.405*>.f(>. ,1_f) if 1 < >.f(>.,1_f) ~5 0.436] 0.271 o( }... 1 f ... I ' - ,) 0.312 [ 0.26 To account for the difference in COA calculated using an FEA model versus LEFM handbook solutions, the COAs are increased by a percentage based on the crack size. A comparison of these calculation methods was performed in 1500880.301 [2] and the percent difference of COA calculated using the FEM for 14 inch cracks is 3.43% (see attached spreadsheet of 1500880.301 ). This correction factor is applied to the four off-axis flaws in weld H4. O.o25] 0.016 . 2 1.03430" 1 f ) :- -E-*2*'lf*Rm*t*G(>. ,1_f ) CO*.&~t }.. *- CO.A f>..,1 f .) '"' *Ill
' l_ I - l
[ 0.018 0.015 1.845 X 10- 3 1.411 X 10- 3 11}.. 1 f) :- COA(>.,1 f) Jri>..,1 f .) ... in 1_f_ry(>.,l_f) U26x 10- 3 iiJI , - \ I -I 1.379 X 10- 3 { } cr(>.-,1 I -I r.)... [ 0.092] 0.091 0.092 0.091 Q_V4H4{>.,l_f) := Cf(>.,1_f) COA(>.,1_f)-J l*g*~d 0j18] 0.319 n V4HM>..,1
'<- \. I -
f.)
- I
- gal 0.369 min
[ 0.306 File No.: 1500880.303 PageA-7 ofA-12 Revision: 1 { T11is infonnation } is EPRI Proprietary Information F0306-01Rl
{JStructural Integrity Associates, Inc.* To account for the uninspected region of the H4 core shroud weld, multiply the leak rate for the H4 flaws by the UT multiplier determined in Table 1 of 1500205.303 . This total plus the leak rate of the through-wall flaw in V6 gives. 28.8*in Pen:mt Coverage :- - - = 5.209*%
- 21f*Rm 3
Q_V4H4_Total := L Q_V4H~>.i,1_fJ26 = 39328gpm i= 0 2 Calculate leakage at end of three operating cycles for the V41H4 weld intersection through-wall flaws. Flaws- 20.3" at V6 <-Crack length increased to account for growth into co-linear crack. Updated initial crack length is 30.2 inches Rm:=Ri+.! Rm = 88-in
- 2 0' := RIPDcd. (Rm)
NW t i := o_1 302 1 i := ( 3.5 ) in N'tW 35.754) 1 f := 1 i + 2*LEF + 2*6*yr*CGR*365 clay *24*2!._ 1f = ( *in
- - yr clay - 9.054 The plastic zone size for the co-linear through-wall flaw in the V6 weld is conservatively calculated by using a stress intensity factor equal to the allowable fracture toughness (Kd.
For stainless steel with an accrued fluence of{ } the fracture toughness is equal to{ }
~>. . 1_f) 1 (KIC)2
- = - * -
2*1f ay sy(>.,1_f) = l.l14in
.!.!...!l:C>-.1_f) := l_f + 2sy(>.,1_f) 37.982). <- Through-wall crack length at 6 years l_f_sy(>., 1_f) .. ( l1.2&2 10 <- Part through-wall co-linear crack length at 6 years (does not contribute to leakage calculation.)
File No.: 1500880.303 Page A-8 of A-12 Revision: 1 { This infomwtion } is EPRI Proprietary Information F0306-01Rl
I)Structural Integrity Associates, Inc.<< 2 1f
+ ry()l.,l_f)
¥,(>. .1 f ) := - - - -
- (Rm*t)0.5 2 4
~>.,1_f) := >J'(>.,1_f) + 0.625->.f()l.,1_f) if 0 ~ >.f(>.,1_f) ~ 1 2 3 4 0.14 + 0.36- >J'(>.,1_f) + 0.72*>.f(>. ,1_f) + 0.405*>.f(>.,1_f) if 1 < >J'(>.,1_f) ~5 To account for the difference in COA calculated using an FEA model versus LEFM handbook solutions, the COAs are increased by a percentage based on the crack size. A comparison of these calculation methods was performed in 1500880.301[2] and the percent difference for 38 inch cracks is 12% and this correction is applied to the through-wall flaw in weld VG.
l.lla
~
t"'OAf).
*1- f) := -
E *2*7t*Rm*t*G(>. *1- f)
.V).
1 f ) :"" COA()I. ,1 f) ~"'o.1_r0) ... o.012in
~ *- 1_f_ty()l.,1_f)
{ } Q_V6(>. ,1_f) := Cf(>.,1_f) COA()I.,1_f)*J l *g*RIPDcd
""I
- 3. Calculate total leakage for the through-wall flaws of the V4/H4 and V6 welds.
Q_V4H4_Total = 39.328gpm Q_V4H4_Total + Q_V~"'o
- 1_f o) "' 49.155 gpm File No.: 1500880.303 Page A-9 of A-12 Revision: 1
{ This infonnation } is EPRI Proprietary Information F0306-01Rl
e Structural Integrity Associates, /nc.c ANSYS Mechanical APDL Input Filename: Loop.inp lig = 10.718
/INPUT,Hatch-Colinear,inp
*get,disp,NODE , 1,U,Y nsel , s,loc,z,cocrack1, cocrack2 nsel , r,loc,y,O
*get , ndcount,node,O,count
*get , num1ow,node,O,num,min numtemp=numlow maxd=O
*get,maxdtemp,node,numtemp,u,y maxd=maxdtemp
*do,tempcount,1,ndcount-1 numtemp=ndnext(numtemp )
*get,maxdtemp,node,numtemp,u,y
*if,maxd,1e,maxdtemp,then maxd=maxdtemp ndmax=numtemp
*endif
*enddo
/OUTPUT,Displacement,txt,,APPEND
/COM, %lig % %disp% %maxd% %numtemp %
/OUTPUT finish
/clear, start Filename: Hatch-Colinear.inp
!finish
!/clear,start
/FILNAME,Hatch-Leak,1
/prep7
/title, Hatch Shroud V5/V6 Weld Leakage
/com, Element Types et,1,shell63
!Material Properties from 1200283.303 MPTEMP,1,0 MPDATA,EX,1,,25550000 !Young's Modulus 25.55 Msi MPDATA,PRXY,1,,.3 !Poisson's Ratio 0.3 rad = 87.25 !Inside shroud radius thk = 1.5 !Thickness of shroud crack= 5.64 !HALF crack length
!lig = 12 . 3 !Ligament Length (can be uncommented for ligament-specific run )
p ressure = 29.5 cocrack1 = crack + lig File No.: 1500880.303 Page A-10 of A-12 Revision: 1 F0306-0IRI
e Structural Integrity Associates, Inc.~ cocrack2 = cocrackl + 38 capload = pressure*rad/(2*thk) R, 1 , thk, I I I I I
!Shell Quadrant Modeling k , l , O, O,O k , 2 , 0 , 0,150 FLST , 2,2,8 FITEM,2,0,0,0 FITEM,2,rad,O,O CIRCLE,PSlX, I I ,9 0 ,1, !90 Degree model LSTR,1,2 ADRAG I 1, I I I I I 2 !Crack Modeling for inner crack FLST,2,1,4,0RDE,l FITEM,2,4 FLST,3,1,3,0RDE,l FITEM,3,3 KGEN,2,P51X, , , , ,crack, , 0 KWPAVE, 7 ASBW, 1 !Crack modeling for outer crack FLST,2,1,4,0RDE,l FITEM,2,4 FLST,3,1,3,0RDE,l FITEM,3,3 KGEN , 2,P51X, , , , ,cocrackl, , 0 KWPAVE, 10 ASBW, 3 FLST,2,1,4,0RDE,l FITEM,2,4 FLST,3,1,3,0RDE,l FITEM,3,3 KGEN , 2,P51X, , , , ,cocrack2, , 0 KWPAVE, 13 ASBW, 1 !Mesh !Mesh Density 0 . 5 inch/element ESIZE , O. S , O, MSHAPE , 0 , 2D MSHKEY, l AMESH, 2 AMESH, 3 AMESH, 4 AMESH, S !All Symm File No.: 1500880.303 Page A-ll of A-12 Revision: I F0306-0IRI
S} Structural Integrity Associates, Inc.~ FLST,2,7,4,0RDE , 7 FITEM,2,1 FITEM,2,7 FITEM,2,9 FITEM,2,-10 FITEM,2,12 FITEM,2,-13 FITEM,2,16 DL,P51X, ,SYMM
!Cap Load FLST,2,1,4,0RDE,l FITEM,2,3 SFL,P51X,PRES,-capload !Internal Pressure FLST,2,4,5,0RDE,2 FITEM,2,2 FITEM,2,-5 SFA, P51X, I PRES, 29.5 !Pressure on Crack Faces FLST,2,2,4,0RDE,2 FITEM,2,6 FITEM,2,15 SFL,P51X,PRES,pressure FIN! /SOL SOLVE FIN! /POSTl ANSYS Mechanical APDL Output Filename: Displacement.txt Inner Outer Crack Crack Major Minor Minor Radius Radius Radius 10 . 718 5 . 391163809E-04 7 . 70870701E-03 63078 File No.: 1500880.303 Page A-12 of A-12 Revision: 1 F0306-0IRI}}