ML032731422

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Reactor Pressure Vessel Flaw Evaluation
ML032731422
Person / Time
Site: Nine Mile Point Constellation icon.png
Issue date: 09/19/2003
From: William Holston
Constellation Energy Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NMP1L 1776
Download: ML032731422 (22)
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Text

P.O. Box 63 Lycoming, New York 13093 Constellation Energy Group September 19, 2003 Nine Mile Point NMP1L 1776 Nuclear Station U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, D.C. 20555

SUBJECT:

Nine Mile Point Unit 1 Docket No. 50-220 Facility Operating License No. DPR-63 Reactor Pressure Vessel Flaw Evaluation Gentlemen:

In accordance with the requirements of the American Society of Mechanical Engineers (ASME)

Boiler and Pressure Vessel Code, Section XM, Paragraph IWB-3610 (1989 Edition), Nine Mile Point Nuclear Station, LLC (NMPNS) is submitting for NRC staff review and approval the attached structural flaw evaluation of a subsurface flaw indication found in a Nine Mile Point Unit 1 reactor pressure vessel closure head meridional weld.

Ultrasonic examinations performed during Refueling Outage 17 identified a subsurface flaw indication in closure head meridional weld RV-WD-005. The indication was characterized per Section Xl, Paragraph IWA-3320 and Figure IWA-3320-1 of the ASME Code as a subsurface planar flaw. Supplemental inspections and reviews conducted by NMPNS provide reasonable assurance that the flaw indication is related to the original fabrication, not service induced.

Flaw characteristic calculations were performed to determine acceptability on the basis of IWB-3500, Paragraph IWB-3510.1. The flaw was initially dispositioned as unacceptable. Further analytic evaluation, performed in accordance with the methods prescribed in 1WB-3600, demonstrate that the weld containing the flaw is acceptable for continued service. The attached calculation SOVESSELM035, Revision 0, documents the closure head weld flaw evaluation.

Very truly yours, William C. Holston Manager Engineering Services WCH/JJD/bjh Attachment

Page 2 NMP1L 1776 cc: Mr. H. J. Miller, NRC Regional Administrator, Region 1 Mr. G. K. Hunegs, NRC Senior Resident Inspector Mr. P. S. Tam, Senior Project Manager, NRR (2 copies)

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NUCLEAR ENGINEERING CALCULATION

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Project: NINE MILE POINT NUCLEAR STATION COVER SHEET

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Title Calculation No.

RPV CLOSURE HEAD WELD FLAW EVALUATION SOVESSELM035 (Sub)system(s) Building Floor Elev. Index No.

RXVE RX 340 l SO Originator(s)

G.L. STEVENS (STRUCTURAL INTEGRITY ASSOCIATES, INC.)

Reviewer(s) / Approver(s)

K. K. FUJIKAWA (SIA)I G.L. STEVENS (SIA) I ACCEPTED BY R.CORIERI Eval., DER, or Prep'd Reviewed Rev Description Change No. By Date By Date App Date 0 INITIAL ISSUE 2003-1344 GLS 7/21/03 KKF 7/21/03 GLS 7/21/03

° CEG ACCEPTANCE - - _ _ I cZT7 l Computer Output/Microfilm Filed Separately (Yes / No I NA): NA Safety Class (SRI NSR I Qxx): SR Superseded Document(s): N/A Document Cross Reference(s) - For additional references see page(s) :19 Ref Doc No Document No. Type Index Sheet Rev 1 SO.ORXVESTUD02 CALC 0 01 2 SOVESSELM026 CALC 0 01 General Reference(s):

Structural Intergrity Associates, Inc. Report SIR-03-036 Remarks:

None Confirmation Required (Yes / No): No Final Issue Status: Turnover Required See Page(s): IAPP ( Yes /WA): NIA 10CFR50.59 Evaluation Number(s): Component ID(s) (As shown In MEL):

Copy of Applicability Determination or 50.59 Screen RPV-NRO2 Attached? Yes 3 No[]

Key Words: REACTOR, ASME, FLAW, INDICATION, FRACTURE MECHANICS, HEAD, FLAW HANDBOOK, OYSTER CREEK C:\Documents and Settings\CORIERIR\My Documents\RPV\RPV HEAD FLAW EVAL\SOVESSELM035.doc NEP-DES-08

Structural Integrity Associates, Inc. REVISION.

PAGE t1 O..

6595 S.Dayton Street Suite 3000 Greenwood Wlage, CO 80111.6145 Phone: 303-792-0077 Fax: 303-792-2158 www.strucitntcom gstevens~structintcom July 22, 2003 SIR-03-036, Rev. 0 GLS-03-019 Roy Corieri Constellation Energy Group Nine Mile Pt. Nuclear Station Nine Mile Point - U2 Warehouse 348 Lake Road Oswego, NY 13126

Subject:

Nine Mile Point Unit 1 RPV Closure Head Flaw Evaluation

Dear Roy:

This letter report documents Structural Integrity Associates' (SI's) evaluation of the flaw in the Nine Mile Point Unit 1 RPV closure head.

INTRODUCTION An indication was identified in the Nine Mile Point 1 (NMP-1) closure head. Per the NDE Report NMP1-03-001 [1], the indication is located in the vicinity of axial weld RV-WD-005 of the closure head. The closure head weld seam is seen in View lA-lA of the Nine Mile Point drawing [3]. The indication was characterized per ASME Section XI, Paragraph IWA-3320 and Figure IWA-3320-1 as a subsurface planar flaw. Flaw characteristic calculations were performed to determine acceptability on the basis of IWB-3500, Paragraph IWB-3510.1. The flaw was dispositioned as unacceptable. See Figure 1 detailing the flaw characterization per IWB-3500.

SI performed the same flaw characterization and evaluated the flaw in accordance with ASME Section Xl, IWB-3600, Analytical Evaluation of Flaws. Per the Nine Mile Point Unit 1 purchase order [2] for performing flaw evaluations in case flaws were deemed unacceptable per IWB-3500, the Flaw Evaluation Handbook for Oyster Creek [4] was to be used.

METHODOLOGY This subsection summarizes the methodology used to develop the Oyster Creek Flaw Evaluation Handbook

[4].

Charltte. NC San Jose, CA N.Stonington, CT Sunrise, FL Rockvile, MD Uniontown, OH 704-573-1369 408-978-8200 860-599-5o

954572-2902 301-231-7746 330899-9753

kALC NO Mr. Roy Corieri REVISION __n July 22, 2003 Page 2 . SIR-03-036, Rev. O/GLS-03-019 Overview ofSection XI Evaluation The 1986 edition of Section XI of the ASME Boiler & Pressure Vessel Code was generally used to develop the allowable flaw sizes, modified with additional more conservative criteria, as discussed below. The rules for evaluation of flaws in reactor vessels are contained in IWA-3000, IWB-3500 and IWB-3600 of Section XI [11]. Appendix A of Section XI provides specific methodology that may be used for detailed fracture mechanics evaluations. The following provides an overview of the Section XI evaluation approach.

In the first step of vessel flaw evaluation, the indications from vessel inspections must be characterized per the requirements of Section XI Article IWA-3000. This requires that the indications be bounded by a rectangular shape with depth (a for surface flaws and 2a for subsurface flaws) and length (1)that will completely contain the suspected material flaws. Closely adjacent flaws must be linked together based on proximity criteria contained in IWA-3000. Similarly, flaws closely adjacent to the base metal surface must be considered surface flaws, based on criteria presented in IWA-3000.

The next step in the vessel flaw evaluation is to compare the flaw with the evaluation standards included in Table IWB-3510-1. This table provides the size of allowable planar flaws that may be accepted without further evaluation. Table IWB-3510-1 defines allowable sizes for surface and subsurface flaws as a function of wall thickness (t), flaw aspect ratio (a/l) and flaw depth ratio (a/t), where t is the measured base metal thickness.

If the indication is larger than may be accepted by IWB-3510-1, then additional analytical evaluation is allowed per IWB-3600. These evaluations are based on the total wall thickness including cladding with characterization and surface proximity rules shown in Figure IWB-3610-1. As shown in the figure, flaws located closely adjacent to the surface must be evaluated as surface flaws. However, flaws located completely within the vessel cladding are acceptable with no further evaluation per IWB-3610 (bXl). Key points of the evaluation include:

  • The ASME Code Section Xl criteria allow acceptance by specifying a factor of safety on either the size of the critical flaw, or a factor of safety on the stress intensity factor.
  • Separate evaluations are required for Normal/Upset and Emergency/Faulted conditions, with different factors of safety for each condition.

Appendix A of Section XI provides a detailed procedure for vessel flaw evaluation. For evaluation of flaws in shell-like structures such as the reactor vessel wall, the methodology of Appendix A of Section XI is directly applicable. To perform the analysis, the following factors must be considered:

  • The flaw must be characterized and resolved into a shape that can be evaluated. This includes determination of the depth ratio (a/t) and the aspect ratio (al I or 2 a/ ) of the flaw (Figure IWB-3610(bXl)). For subsurface flaws, the eccentricity ratio (e/t or 2e/t) must be determined, where e is the distance from the center of the vessel wall (including cladding) to the center of the flaw (Figure A-3300-4, of Appendix A of Section XI).

t SiicturalIntegrity Associates, Inc.

CALC NO xn-Mr. Roy Corieri REVISION July 22, 2003 Page 3 PAGE NO SIR-03-036, Rev. O/GLS-03-019

  • Stresses and the temperature at the location of the flaw must be determined for all loading conditions.
  • The flaw stress intensity factor must be calculated, either by using the equations, charts, and tables of Appendix A of Section XI or through use of other, more sophisticated, documented analytical techniques.
  • The material properties must be defined at the location of the flaw, including the effects of irradiation. Any through-wall variation of material properties should be considered.
  • The crack growth that can occur during the evaluation interval must be determined (e.g., to the next inspection or to end-of-life).
  • The predicted flaw size at the end-of-evaluation period must be less than that allowed by Section XI.
  • The primary stress limits of the original code of design (NB-3000) must also be met assuming a local area reduction of the pressure-retaining membrane that is equal to the area of the characterized flaws.

Specific Details of Vessel Shell Evaluation Stress Intensity Factors Appendix A of Section XI provides a basic methodology for evaluating vessel flaws. However, there is limited guidance for the determination of stress intensity factors for cracks extending through cladding. In addition, the guidelines are very limited for determining the stress intensity at the surface for surface flaws.

The following describes how the stress intensity factors were determined for the Oyster Creek RPV evaluation.

Appendix A Methods For all stresses, except for those due to cladding for an internal surface flaw, the methods of Appendix A are used for the deepest point of surface flaws and for subsurface flaws.

Surface Stress Intensity Factors (except for cladding)

For surface stress intensity factors of surface flaws, Mm and Mb, as defined in Appendix A of Section XI, have been determined based on the Raju/Newman membrane and bending solutions [12] for the worst case of internal and external cracks for a vessel with thickness-to-radius (tlr) ratio of 0.1. Referencel2 derives surface stress intensity factors for t/r values of 0.1 and 0.25. Based on a comparison of the influence coefficients for t/r values of 0.1 and 0.25, it is noted that tlr has a very minor effect on the stress intensity factor solutions. Therefore, the calculated Mm and Mb values are applicable for Oyster Creek, which has a t/r ratio of approximately 0.06. The resulting surface stress intensity factor is applied at the cladding-to-base metal interface for the vessel inside surface. For flaws with an aspect ratio (a/l) of zero, the surface stress intensity factor is assumed to be zero since an infinitely long crack does not have a surface point.

r Structural IntegrityAssociates, Inc.

Mr. Roy Corieri CALC NO T July 22, 2003 Page 4 PAGE NO SIR-03-036, Rev. O/GLS-03-019 Cladding Stress Intensity Factor For cladding stresses for a long inside-surface flaw, the stress intensity factor at the deepest point of the flaw is determined by integration of the stress over the crack face for an edge-cracked plate using the methods from Tada and Paris [13].

Ki=-la m(x)

  • t(x)d (1) where: oa (x) = cladding stress distribution in cladding and base metal as a function of distance (x) from clad surface a = crack depth m(x)= 3.52( -a*) 4.3 5 -5. 2 8 a (2)

M! (X) (2)~*)O

+ +0.83- 1.76ae[1l-(J-a*)th]

where: a x/a t = alt t wall thickness As shown in a paper by Kuo,

Deardorff,

and Riccardella [14], this type of solution yields a stress intensity factor that shows a reasonable comparison to "exact" solutions for flaw depth ratios (a/t) up to about 0.33.

Above this ratio, the solution is unrealistic and increases significantly whereas solutions based on more sophisticated techniques vary little for deeper cracks. Thus, it is assumed that the stress intensity factor may be described by the following:

For a *am.n.

K; = (3)

For a > amin K, =lesserof K.

or

- a Kein l-(4) t3 StructuralIntegrityAssociates, Inc.

Mr. Roy Corieri REVISIN ° _ July 22, 2003 Page 5 PAGE NO SIR-03-036, Rev. 0/GLS-03-019 where: K, = minimum K, in base material a = crack depth size amin - a at K,_

To account for the flaw aspect ratio, the stress intensity factor is corrected using the crack shape factor, Q, of Appendix A of Section XI as follows:

K - Kr ( J (5) where, Q = shape factor for flaw with aspect ratio of (a/l) = 0 Q = shape factor for flaw with aspect ratio (all) being evaluated In determining the shape factor ratio, the stress ratio (the other factor affecting Q and Q0) is determined based on membrane plus bending stress ( a,. + rb ) for the flaw. However, the cladding induced stresses at the crack are excluded since, in the base material, these stresses are either small or compressive.

ASME Section XI does not require that the stress intensity factor in the cladding be evaluated. However, the stress intensity factor for the cladding-to-base metal interface location is calculated based on determination of the surface stress intensity factor. To calculate the surface stress intensity factor, the cladding stress intensity factor (obtained by Equations 1 and 5 above) for a flaw depth equal to the thickness of the cladding (with the same aspect ratio as the deeper flaw being evaluated) is determined. It is then modified based on the ratio between the membrane stress intensity correction factors for thez surface (using the Raju/Newman, Mm) and the crack tip (using the Appendix A, M.). This formulation is considered conservative for this analysis. Note that ASME Section XI does not require the stress intensity factor due to the cladding to be considered. Thus, this methodology is considered more conservative compared to Appendix A,Section XI of the ASME Code.

For very long flaws with an aspect ratio of zero (a/l = 0), there is no surface point. Therefore, the stress intensity factor due to cladding at the "surface" is set to zero.

Fracture Toughness The fracture toughness of the reactor pressure vessel material is obtained from Section XI, Appendix A in terms of KIa and Kl, which represent the critical values of the stress intensity factor, K1, based on crack arrest and static crack initiation, respectively.

The conservative lower bound Kr. and Kh are defined as functions of the local temperature of the vessel wall as follows:

Kle = 33.2 + 20.734 e0.02(r-RTND2 and, Kha = 26.8 + 12.445 eo.*14 5(T-RTNDl

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CALC NO -"VeSeL4*Q'X Mr. Roy Corieri GREVISION July 22, 2003 MrPaRoygCoreeri Page 6 PAGE NO SIR-03-036, Rev. 0/GLS-03-019 The analyzed vessel wall local fracture toughness, at the location of the associated crack stress intensity factor, is determined with consideration of local temperature (as a function of wall depth), initial RTNDT, local fluence, margins and chemistry factors in accordance with the methods of Regulatory Guide 1.99 Revision 2 [15]. The approach is as follows:

ART = RTNDTi + RTNDTShift + Margin (6) where:

ART = adjusted reference temperature, 'F RTNDT.i = initial RTNr, -F Margin = required margin = 2 I , F (RTNDTShift, a,, and a, are defined below)

The margin is determined based on the standard deviation of the initial RTNDT (ai) and that of the RTNDT shift (cA,). The standard a. is 280F for welds and 170 F for base metal [15], except that a, need not exceed 0.5 times the computed shift in RTNDT.

RTNDTShift = (CF) * (FF) (7) where: CF = chemistry factor, 'F FF = fluence factor, dimensionless EF = fF.28 4/ og1)- (8) where: f = local fluence, neutrons/cm 2 x 1019 (E>lMeV)

The local fluence, f at any position in the wall may be calculated from:

f =c 0.24x

.- (9) where: ff = fluence at inside surface, neutrons/cm 2 x 1019 (E>lMeV) x distance from inside surface, inches The fluence at the surface is a function of the amount of irradiation exposure time:

fS,,f f x EFPY (10)

EFPY,,

where: fd = reference surface fluence, neutrons/cm 2 x 1019 (E>lMeV)

EFPYf = effective full power years associated with fe EFPY = effective full power years for evaluation This allows the adjusted reference temperature, ART, to be calculated for all the vessel regions at any depth, at any time, and for each specific weld or plate being evaluated.

- - - Structural IntegrityAssociates, Inc.

Mr. Roy Corieri REVSIN 0 July 22, 2003 Page 7 PAGRE NO 0 SIR-03-036, Rev. 0/GLS-03-019 Fatigue Crack Growth Considerations The fatigue crack growth law for ferritic steels in water environment is provided by Section XI of the ASME Code [11]. This law is applicable to the inside surface flaws of the reactor vessel. For allowable subsurface and outside surface flaws, a crack growth curve for ferritic steel in air environment is assumed.

-A conservative estimate of the crack growth for air is defined in Section XI Appendix A of the 1992 Edition of ASME Code. For inside surface flaws, the water environment crack-growth curve used is conservatively based on R > 0.65, where R is the ratio of the minimum crack tip stress intensity factor to the maximum stress intensity factor (Kmin/Knax). For subsurface and outside surface flaws, the crack growth curve for air environment is conservatively based on Rel.

The stress intensity factor at the allowable flaw size for each flaw is conservatively used in the crack growth evaluation. For flaws accepted by the evaluation standards of Table IWB-3500-1, there is no requirement to consider crack growth.

The primary source for crack growth in the vessel low alloy steel is the cyclic loading due to startup/shutdown of the reactor. For purpose of crack growth, the number of cycles from the time of the vessel inspection to the end of the evaluation period is estimated. A conservative estimate of 240 startup/shutdown cycles were considered in the determination of crack growth.

Subsurface Flaw Characterization Subsurface flaws of various eccentricity ratios (e/t) are evaluated. Per the requirements of Table IWB-35 10-1 or Figure IWB-3610-1, a near surface subsurface flaw must be evaluated as a surface flaw when the distance between the surface and the nearest point of the flaw (s) is less than 0.4d, where d is half the depth of the subsurface flaw. Based on these requirements, the maximum subsurface flaw size that does not have to be evaluated as a surface flaw is found with the following equation:

(a) O.S;e /1t (11) where: t = thickness of vessel base material (for IWB-3500 evaluation), or total thickness of vessel wall including cladding (for IWB-3600/Appendix A evaluation).

e = flaw eccentricity, measured from center of vessel wall, (determined with or without cladding as appropriate), negative if toward inner vessel wall.

Definition of Allowable Flaw Size and Shape The development of the flaw acceptance diagrams requires evaluating hypothetical flaws to determine the allowable flaw depth for the complete range of flaw aspect ratios and eccentricities (for subsurface flaws).

The evaluation is started with the maximum flaw depth that can be analyzed. For surface flaws, the maximum depth of 0.8 times the total vessel wall thickness is used since it is the limit of applicability of most fracture mechanics solutions. The flaw is checked for acceptability based on the Section XI, IWB-5g StructuralIntegrityAssociates, Inc.

Mr. Roy Corieri PALC NO s July 22, 2003 Page 8 -? - SIR-03-036, Rev. 0/GLS-03-019 3600 criteria for crack size and stress intensity factor. Then, the flaw depth is incrementally decreased based on a specified crack size increment and the acceptability check is repeated for the resulting crack size.

In some cases, such as when there is a large bending component to the through-wall stress distribution and, the fracture toughness through the wall is not constant due to irradiation embrittlement and/or local temperature and/or, cladding stresses are a significant contribution to the stress intensity factor, there may be more than one acceptable flaw geometry through the vessel wall. Thus, there may be some flaws near the surface that are not acceptable even though flaws with larger depths may be acceptable. Also, if, due to high surface stresses, the stress intensity factor is larger at the surface point of the flaw rather than at its deepest point, a longer flaw (smaller aspect ratio, e.g., infinitely long flaw) may be acceptable when a similar depth flaw which is shorter would not be acceptable. These cases do not reflect the underlying idea of the acceptance criteria of IWB-3500, which implies that a smaller flaw is more acceptable than a larger one. Therefore, the allowable flaw size at a location can be reported three different ways:

  • Option 1: Report the largest flaw depth that is acceptable for each of the flaw aspect ratios evaluated. This is analogous to evaluating an actual flaw by assuming a larger bounding flaw size.

This approach is the most logical although it appears to be less conservative for the special stress and material conditions discussed above.

  • Option 2: Report the largest near-surface flaw depth that is acceptable for each of the flaw aspect ratios evaluated, regardless of the fact that flaws with larger depths are also acceptable. This approach is more conservative as compared to Option 1.
  • Option 3: Report the largest near-surface flaw depth that is acceptable as in Option 2 but use a smaller flaw aspect ratio (assume a longer flaw). This approach is specifically applicable to cases where, due to the fact that the surface stress intensity factor is controlling, a longer flaw is more acceptable than a shorter flaw of the same depth.

In the evaluations performed in this report, the first option has been chosen since the stress intensity factor solutions for surface stresses and cladding are believed to be very conservative. In addition, when the stress/material conditions at the surface are controlling, the limiting acceptable flaw is shallow and less likely to grow to any significant depth. Moreover, if such a flaw were to grow, it would eventually become large enough to meet the acceptance criteria for the larger flaw. In most cases, more sophisticated analysis, as allowed by Section XI, Appendix A, A-3300 (c), could result in lower values of stress intensity factors.

The stress intensity factor within the cladding does not have to be evaluated for acceptability per Section XI requirements. Thus, the allowable inside surface flaw size will always be equal to the cladding thickness plus that allowed by the acceptance standards of IWB-3500.

Vessel Shell Analysis Implementation The flaw acceptance analysis has been prepared using a computer program developed and verified by SI for this specific purpose. APPENDA (Appendix A Analysis) [16] is a computer program written to perform reactor pressure vessel flaw evaluation in accordance with Appendix A of Section XI and Subarticle IWB-3600 of Section XI of the ASME Boiler and Pressure Vessel Code [11]. It uses the methodology described

$3StructuralIntegrityAssociates, Inc.

Mr. Roy Corieri CALC NO July 22, 2003 Page 9 FAGE NO . lO SIR-03-036, Rev. 0/GLS-03-019 above, and determines allowable inside surface, outside surface and subsurface flaws. It is intended to provide a rapid assessment of all possible flaws to allow construction of flaw acceptance diagrams that may be used to provide guidance in reactor vessel inspections.

APPENDA performs an evaluation to determine the acceptable size of surface and subsurface flaws in accordance with the requirements of ASME Code,Section XI, Appendix A and Subarticle IWB-3600 [l1].

In addition, the acceptability of relatively smaller flaws is evaluated in accordance with Section XI, Table IWB-35 10-1 [11] for planar flaws. The program output includes the acceptable flaw size for the complete range of flaw aspect ratios and flaw eccentricities (for subsurface flaws). The key features of the program include:

  • ability to include an arbitrary stress distribution for pressure, bending, thermal, and residual stresses, including load multiplier factors for each.
  • evaluation of cladding stresses, with several methods to handle the effects of the cladding stresses at the surface for inside surface flaws.
  • ability to evaluate flaws based on either the maximum acceptable size, minimum acceptable size, or the minimum acceptable size assuming a smaller aspect ratio, Ca/l .
  • consideration of the varying safety factors associated with Normaflpset condition, Emergency/Faulted condition or regions near local discontinuities (per IWB-3613 (a)). A
  • automatic determination of the wall fracture toughness distribution given initial material properties and accumulated surface fluence at the end of the evaluation period.
  • conservative assessment of flaw growth to the end of the evaluation period.

A separate utility program MAPPA (MMultiple Appendix Analysis) provides an evaluation of multiple input cases (load conditions) and determines the controlling loading condition (or combination of conditions) for a location based on a number of individual evaluations using APPENDA.

Code Reconciliation

  • current evaluation is generally based on methodology from the 1986 edition of Section XI of the The ASME Code. The Oyster Creek reactor pressure vessel was constructed in accordance with the 1959 ASME Code,Section I, with Addenda through Summer 1963 and the 1963 ASME Code, Section HI. The following steps of the evaluation contain criteria taken from ASME Code Edition issued after the Oyster Creek's Code of Construction:
1. For this evaluation, the 1995 Code with 1996 Addenda is used as the basis for fatigue crack growth curves. This provides a more conservative curve for an air environment (subsurface or outside surface flaws) that is dependent on the R ratio (KW/Kd).
2. In this evaluation, the methodology of Reg. Guide 1.99, Rev. 2 [15] is used for determining the total shift in RTNw as affected by uncertainties, fluence and materials composition. Use of this approach is consistent with current regulatory requirements for evaluation of reactor vessel materials.

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Mr. Roy Corieri ALC NO S July 22, 2003 Page 10 '~%REVISIN _ SIR-03-036, Rev. 0/GLS-03-019

3. For purposes of stress analysis, material properties were obtained from the 1995 edition of the ASME Code (Section II, Part D), which provides more detailed temperature dependent material properties.

The 1995 edition of the ASME Code with 1996 Addenda was the most recent version of the Code approved by 10CFR50 when the Oyster Creek flaw handbook was developed.

A default maximum flaw size was determined such that the nominal stress would increase to approximately 1.5 times the nominal stress if a long flaw existed at the location. The additional limitation was based on primary stress limits in Subsection NB-3000 of Section III of the ASME Code. The stress limits were added since IWB-3610(d)(2) requires that the primary stress limits of NB-3000 (of ASME Section III) be satisfied for the size of the evaluated flaw. For actual flaws found in a reactor pressure vessel, this should never become limiting because NB-3000 allows local primary membrane stresses to approach 1.5 Sm provided that the extent of the region with stress exceeding 1.1 Sm does not exceed V (where R is the mean vessel radius and t is the thickness). This compares to the requirement for the design equations for pressure sizing where the stress must be maintained below Sm. Based on this ratio, if the pressure stress is near the allowable stress, the additional primary stress criterion might become governing for axial flaws with depths approaching one-third of the wall thickness that have any significant extent. Since the stresses acting on circumferential flaws are about one half of that for axial flaws, greater flaw depths would be allowed for flaws with a circumferential orientation, assumed in this evaluation to be limited to 50% of the wall thickness, except at regions near discontinuities where the one-third of wall thickness default maximum size is used.

Therefore, the maximum allowable flaw sizes defined by Section III stress limits were evaluated for each region and incorporated in into the flaw acceptance diagrams.

ANALYSIS AND RESULTS The flaw handbook for Oyster Creek RPV [4] was used to evaluate the flaw in the closure head of NMP- 1, due to the availability of a detailed finite element model for Oyster Creek and since the RPVs between the two plants are nearly identical. The validity of using the Oyster Creek RPV flaw handbook was based on a comparison of the associated RPV geometries and stresses. Each of these comparisons is discussed in the sections that follow.

Geometry Comparison A comparison of the two RPV geometries and materials was performed and is summarized below:

  • Vessel closure head materials, SA-302 Gr. B, are the same for both plants [4, 6].
  • The limiting RTNr for the top head material (for Region B) for Oyster Creek is 450 F [4]. Per Constellation input, the limiting RTNm for the top head material for NMP-l is 401F.
  • The dimensions of the Oyster Creek finite element model are shown in Figure 2, as extracted from Reference [8]. These dimensions are identical to those found on the final machining drawing applicable to NMP-1 [7].

t3 StructuralIntegrityAssociates, Inc.

Mr. Roy Corieri t,., July 22, 2003 Page 1II ISIOI SIR-03-036, Rev. 0/GLS-03-019 Based on the above, the Oyster Creek flaw handbook results are applicable and bounding for application to NMP-I from a materials and geometry point-of-view.

Stress Comparison A comparison of the stresses for the two RPVs was performed and is summarized in Table l. The stresses in Table 1 come from four sources: (1) the original Combustion Engineering (CE) RPV stress report for Oyster Creek, (2) the original CE RPV stress report for NMP-1 [6], (3) the RPV flaw handbook for Oyster Creek [4], and (4) the RPV head tensioning evaluation report for NMP-1 [5]. The stresses listed in Table 1 were extracted from similar locations in all evaluations for a consistent comparison, and bound the stresses at the flaw location (but are not coincident with the flaw location).

The following conclusions can be made with respect to the results shown in Table 1:

  • Considering the different models, techniques and boundary conditions used between all of the analyses, there is very reasonable agreement between all of the stress reports.
  • The revised bolting procedure evaluation caused increased boltup stresses in the top head region.
  • The stresses from the revised bolting procedure evaluation are slightly higher than those used in the RPV flaw handbook. These higher stresses should be considered before application of the Oyster Creek flaw handbook to NMP- 1.

Since the stresses from the revised bolting procedure evaluation were not bounded by those used in the flaw handbook [4], additional refined stress values were obtained at the flaw location of interest. Reference [10]

documents the stresses from the revised bolting procedure evaluation to be 5.6 ksi on the outside surface, and 1.5 ksi on the inside surface at the flaw location. These stresses are well below the stresses tabulated in Table I that were used in the Reference [4] flaw evaluation handbook.

Based on the above, the Oyster Creek flaw handbook results are applicable for this flaw and bounding for application to NMP-1 from a stress point-of-view.

Results Figure 3 details the flaw characterization calculation as well as flaw evaluation based on Reference [4].

According to the revised Oyster Creek flaw acceptance diagram for the closure head axial welds (Appendix B, Figure B-2 Subsurface Flaw [e/t = -0.4]), the flaw RV-WD-005 is acceptable per IWB-3600. Figure 4 shows a plot of Allowable Flaw Depth (in.) versus Flaw Aspect Ratio (2a/l). Flaw RV-WD-005 is below the acceptable flaw limit determined using the methods of IWB-3600. The allowable flaws shown in the revised Oyster Creek flaw acceptance diagrams include allowance for fatigue crack growth with an assumed 240 cycles of startup/shutdown.

CONCLUSION Based on the analysis documented herein, RPV top head flaw RV-WD-005 is considered acceptable for continued operation at NMP-1.

!t Strucftural Integrity Associates, Inc.

Mr. Roy Corieri W'~F NOf --- la July 22, 2003 Page 12 SIR-03-036, Rev. 0/GLS-03-019 If you have any questions, please feel free to give me a call at (303) 792-0077.

Prepared By: Reviewed By:

G. L. Stevens, P.E. K K. Fujikawae E.

Associate Associate Approved By: / 9i

i. .

G. L. Stevens, P.E.

Senior Associate cc: NMP-05Q-106 (ODL), -404 V StructuralIntegrity Associates, Inc.

Mr. Roy Corieri AlOC NO T July 22, 2003 REVISISON Page 13 ~c

- - 71i - SIR-03-036, Rev. O/GLS-03-019 Table 1. Comparison of Boltup Hoop Stresses NMP-1 CE Oyster Creek Oyster Creek Dominion Head Tensioning Reference ' Stress Report CE Stress Flaw Handbook Report [5]

[6] Report [9] [4] (at Cut Line 6)

(at Cut I) (at Cut I) (Region B, 900 F) Design Case Revised Process Inside Surface 3.7 3.7 2.1 6.4 7.6 Hoop Stress (ksi)

Outside Surface 14.6 14.6 17.6 15.2 18.1 Hoop Stress (ksi) I I_

3 StSructural IntegrilyAssociates, Inc.

ot 'tMrn "trT Mr. Roy Corieri .~ALC NO X f July 22, 2003 R~EVISION a Page 14 PAG NO SIR-03-036, Rev. 0/GLS-03-019 Nv NIAGARA - Evaluation of Rapodtatl 111dcationa NU MOHAWK NoE Repoit N oP31-0300I Nine Mue Poi Un I

~~3.L 2.5 SM ~ Workbocmnet fiSdOJ System: 00 Q PF lrl= V w bdlctonafI. I Appilcabie ASME Code Standard(a) used for evalimtln

  • Sectio .v.... Editon .JB0 Addenda A/
  • (Spcify how width AXhodetmonnined. p DIffatinraphktcaly.etc.)

Ti4~fenhs Comnponent Thickneass: P, o

Ch~ancrtdwof Slaw Inperatla g p: IwA -3320 -

Flaw chapactuzatlcds calculatios: (Attach sketch Vnecessaty)

(Attach sapaat calibrtion repmUtfor evauation examinatkins)

~~/C~ $'.,Z 0. 03/r a177 pjjuAwL~Do bifor* sxru~w /IWO 7J5' X.). (ASPI 1-7 uD-D3- OZ44)

O1404uiM. C.C. C-j~r~nqcint7. F"L WI41 b14'r'U431 1A*tlt,4117 07 7Wa' g~e'c aaeu PoMVM,~rbigrWAdd

  1. A - w' ~Afrat Agt.,we. 1Nme ..,ir %to AA.#.sw tedw tjSs r-t 4joje".-,4J 4e AC ..

.&446~t p i5gk Of ftdSit..

o~p, Comparbson to Pertineont ealuaption standard to wo~a! hcidwcto jize:.

maS ew'crpow IS

'voT a /74 '1. r#9 $(6$iLr O Examinner 1: 4,4J, (Z.';.A Lawhp = Dafta I-z/-d 3 Examtiner-2 Q 0 CI .tALAIII LevWe 3I 11I.1 7-24-.!

Pieoous Duag

. ap:ZDats: DenRay ed RAdnve 7 Ye z/mo_

Data: tRanw al Figure 1. NMP-I Flaw Characterization V StructuralIntegrityAssociates, Inc.

ALC NO 9QEFU-4 July 22, 2003 Mr. Roy Corieri 4EVI1ON 4n Page 15 SIR-03-036, Rev. 0/GLS-03-019 Figure 2. Oyster Creek Finite Element Model Dimensions V Structlural Interpity Associates, Inc.

Mr. Roy Corieri 44uMMET July 22, 2003 Page 16 DREVIAN- SIR-03-036, Rev. 0/GLS-03-019 Flaw No. I NDE Report NMPI-03-001 System: 00 RPV Exam Item RV-WD-005 Evaluate as a Subsurface Flaw Per Figure IWA-3310-1 I= 7 inches Criteria for flaw to be characterized as Subsurface 2d = 0.3 inches S = 0.4a ? 0.4a= 0.06 S= 0.2 inches TRUE Surface flaw is characterized as Subsurface flaw.

Per Table IWB-3510-1 2a = 0.3 inches a= 0.15 inches

,t=

  • 4.8 inches all= 0.0214 Y =Sla= 1.3333 Per note 4 of IWB-3510-1 if Y > 1, then use Y =1.

Allowable alt, % = 2.086 % interpolated per Table IWB-351 0-1 Actual aft = 0.031 3.1250 %

Acceptability of flaw Is Actual (alt) < Allowable (aft) FALSE Per Oyster Creek Flaw Handbook ted = 0.21875 inches t= 4.8 inches Clad ID to Flaw CL = 0.569 inches negative sign since flaw nearer to ID.

e= 1.83 no dad elt = 0.38 -0.38 alt = 0.085 8.46 %

et = 1.94 with dad eat = 0.39 -0.39 art = 0.081 8.09 %

use -OA alt= 0.071 7.14 %

2a1 = 0.043 inches Max Allowable Depth per Fig. B-2 IW1B4600 0.328 inches From 2a/1 = 0.1. for Region B Flaw Acceptance Diagram B-2 Acceptability of flaw Is 2a11 < 2af (QWB-3600) TRUE Figure 3. Flaw Evaluation t StructuralIntegrityAssociates, Inc.

  • frrtcfMNT CALC NO ' m e Mr. Roy Corieri REVISION July 22, 2003 Page 17 .PAGE NO SIR-03-036, Rev. O/GLS-03-019 Subsurface Axial Flaw [eft = - 0.4]

-MJ-3500 IW I X _RV-WD-005 S

r ISI a'

0 0.1 0.2 0.3 - 0.4 0.5 - 0.6 0.7 0.8 0.9 Flaw Aspect Ratio oM Figure 4. Flaw Acceptance Diagram C Stfrutural Integr ty Associates, Inc.

Mr. Roy Corieri CALC NO S July 22, 2003 Page 18 RE ISIO SIR-03-036, Rev. O/GLS-03-019

References:

1. NMP Report No. NMPI-03-001, SI File No. NMP-05Q-212.
2. Constellation Energy Group Purchase Order No. 02-42537-001, dated 03/03/04.
3. Niagara Mohawk Drawing No. F-45183-C, Revision 1, "Weld Map - Reactor Vessel," Sheet 28, SI File No. NMP-05Q-210.
4. Structural Integrity Associates Report Number SIR-00-109, Revision 1, "Flaw Acceptance Handbook for Oyster Creek Reactor Pressure Vessel Shell-Weld Inspections," SI File No. NMP-05Q-1 14.
5. Dominion Engineering, Inc. Document Number R-3661-00-1, Revision 0, February 2003, "Reactor Vessel Tensioning Optimization Stress Report Nine Mile Point Nuclear Generating Station Unit 1," SI File No.NMP-05Q-213. (HmPI cGoAut-A low SoapRJvEsrvbOL,jReg./)
6. Combustion Engineering Report No. CENC-1142, "Analytical Report for Niagara Mohawk Reactor Vessel," SI File No. NMP-05Q-210. (NuPi P11CUMTVA'
  • SOVELA40Z6, REV. /) h
7. Combustion Engineering Drawing No. E-231-575-3, "Closure Head Final Machining," SI File No.

NMP-05Q-210.

8. Structural Integrity Associates Calculation No. GPUN-27Q-301, Revision 0, "Development of RPV Finite Element Model," SI File No. NMP-05Q-1 10.
9. Combustion Engineering Report No. CENC-1 143, "Analytical Report for Jersey Central Reactor Vessel," SI File No. NMP-05Q-1 16.
10. Dominion Engineering, Inc. Letter No. L-3661-00-2, Revision 0, "Flaw Location Specific Membrane and Bending Stresses for Nine Mile Point Unit 1 Reactor Vessel Head," March 31, 2003, SI File No.

NMP-05Q-214. (1A N LPIJ<1tuowe S OOXVE STUDO2. -l) rh Sk3

11. "Rules for In-service Inspection of Nuclear Power Plant Components,"Section XI of the ASME Boiler and Pressure Vessel Code, 1989 Edition, American Society of Mechanical Engineers, New York, July 1,1989.
12. Raju, I. S., and Newman, J. C., "Stress-Intensity Factors for Internal and External Surface Cracks in Cylindrical Vessels," Journal of Pressure Vessel Technology, 104/298, November 1982.
13. Tada, Paris and Irwin, "Stress Analysis of Cracks," Del Research Corporation, 1973.
14. Kuo, A. Y.,

Deardorff,

A. F., and Riccardella, P. C., "Thermal Stress Intensity Factor of an Axial Crack in a Cladded Cylinder," presented at 1993 ASME Pressure Vessel & Piping Conference.

15. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, Revision 2, May 1988.

-l SiruStural IntegrityAssociates, Inc.

Mr. Roy Corieri CACNO July 22, 2003 Page 19 REVISION SIR-03-036, Rev. 0/GLS-03-019 PAGE NO _ 7

16. APPENDA and MAPPA, "Computer Programs for Performing Flaw Tolerance Analysis of Reactor Vessel Shells," Structural Integrity Associates (QA- 1800), Version 1.1.

Strufcturall IntegrityAssociates, Inc.

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