ML20056F294
| ML20056F294 | |
| Person / Time | |
|---|---|
| Site: | Mcguire |
| Issue date: | 08/16/1993 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20056F253 | List: |
| References | |
| PIP-2-M93-0717, PIP-2-M93-0717-R00, PIP-2-M93-717, PIP-2-M93-717-R, NUDOCS 9308260321 | |
| Download: ML20056F294 (44) | |
Text
{{#Wiki_filter:-- _ -. i l ) l I i l McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. 0 '
. Subiect: Evaluation of Reactor Vessel OD Flaw (PIP 2nM93-0717)
Pace: 1 By: 4XA4/ Date: ts (tu/o Ck:Q_ D1 - Date: (IEli% ! g , 1 1.0 STATEMENT OF PROBLEM ! i The purpose of this calculation is to evaluate the McGuire Unit 2 Reactor Vessel flaw at the lower head to ring segments girth weld, as identified during the 10 year In Service Inspection during refueling l outage 2-EOC-8. The indication being evaluated has been I identified and documented in the Problem Investigation Progrhm (PIP-2-M93-0717). This calculation is performed to satisfy the requirements of ASME B&PV Code, Section XI, subsection IWB-3600
" Analytical Evaluation of Flaws".
1 2.0 OA CONDITION QA Condition 1 - Nuclear Safety Related The reactor vessel is an ASME Section III, Class 1 pressure boundary. 3.0 DESIGN METHOD USED The flaw is characterized and evaluated using Linear Elastic Fracture Mechanics methods. The nil-ductility (RTndt) temperature of the limiting material is determined based on actual material tests and USNRC Standard Review Plan, Branch Technical Position 5-2. The loads are based on conservative boundings of all Normal and Upset loads. Faulted loads and Hydrostatic Test load cases are also addressed. The formulas or 39ME Section XI, Appendix A are used to calculate the stress intensities.. 4.0 APPLICABLE CODES AND STANDARDS ASME Boiler and Pressure Vessel Code, Section III, Subsection NB (Class 1 Vessels) 1971 Edition, including addenda thru the Winter 1971. ASME Boiler and Pressure Vessel Code, Section XI, Subsections IWA, IWB (Class 1 Vessels) and Appendix A , 1980 Edition, including addenda thru the Winter 1980. 5.0 OTHER DESIGN CRITERIA Reg. Guide 1.99, " Radiation Embrittlement of. Reactor Pressure Vesselr', ! Rev. 2 dated May 1988 4 US NRC Standard Review Plan, Section 5.3.2 (Pressure- Temperature l Limits), and Branch Technical position MTEB 5-2 (Fracture Toughness ! Requirements) l l 9308260321 930818 E I PDR ADOCK 05000370 % P PDR k
t I i McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O Subiect: Evaluation of Reactor Vessel OD Flaw (PIP 2eM93-0717) Pace- 2 Bv: .-S A// Date: th./* 3 Ck: O VM Date: Aff 6 A% O } 6.0 APPLICABLE DESIGN INPUTS i The following are the applicable design criteria used in this 1 analysis. The criteria of ANSI N45.2.11 have been reviewed and included as applicable to this work. Function: The reactor vessel provides a class 1 reactor coolant pressure boundary. Design Conditions: Design life is defined as 32 effective full power 1 I years (40 years
- 0.8 availability).
Environmental Conditions: The vessel lower head does not see a significant neutron flux, and embrittlement of the ferritic steels is considered unlikely. To ensure conservatism, the radiation shift of the limiting material was calculated and shown to be negligible. Structural Requirements: The reactor vessel structural integrity is required for plant operating modes 1 through 6. Material Properties: The mechanical and semistry values used as the basis of reports. this calculation are from the original Rotterdam Material test 7.0 INITIAL / FINAL CONDITIONS Not Applicable 8.0 CRITERIA CITED IN THE FSAR Sections: 5.2.4 " Fracture Toughness" l 5.4.3.7 " Reactor Vessel Material Surveillance Program Requirements" 9.0 CRITERIA CITED IN THE TECHNICAL SPECIFICATIONS 3.4.9.1 " Reactor Coolant System, Pressure / Temperature Limits" 3.4.10 " Reactor Coolant System, Structural Integrity" 10.0 ASSUMPTIONS As stated. l
McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O Subiect: Pace: 3 Evaluation By: of Reactor Vessel OD Flaw.(PIP 2mM93-0717) s// d L__.
~
Date:slalo Ck :A__.1 -- _EM Date:s/r6 M3 c *1 > '
11.0 REFERENCES
11-1 WCAP-11029,. Analysis of Capsule V from the Duke Power Company McGuire Unit 2 Reactor Vessel Radiation Surveillance Program, January 1986 11-2 WCAP-13516, Analysis of Capsule U from the Duke Power ; Company McGuire Unit 2 Reactor Vessel Radiation Surveillance i Program, October 1992 11-3 Westinghouse Electric Corporation Equipment Specification 676413, rev 2, " Reactor Vessel" No. , 11-4 Westinghouse Electric Corporation Equipment Specification ' 952564, rev 1, " Addendum to Equipment Specification 676413, rev. No. 2, Standard Four Loop Plant, McGuire II & Catawba I Reactor ' Vessels", MCM 2201.21-0001 12.0 ATTACHED REFERENCES 1 12-1 Rotterdam Dockyard Drawing " Location, Identification and Thermal History of Vessel Parts" i j 12-2 Certified Material Test Reports, Rotterdam Dockyard Vessel 30664 12-3 Rotterdam Dockyard ! Drawing "173 Inch PWR Vessel ' Westinghouse' l General Arrangement", Drawing No. 30738-1510 sheet 1 of 2, rev. D l 12-4 Ultrasonic Examination Results of the Lower Head-to-Bottom Head Weld (2RRV-WO1) for the McGuire Unit II July 1993 Reactor Pressure Vessel Examination, BWHT 13.0 EVALUATION 13.1 Determination of Reference Transition / Nil Ductility Temperature (RT Welbdt) for the McGuire Unit 2 Reactor Vessel Lower Head Girth Regions: ; The lower head consists segment (Item 02). The girth ofweld a bottom done (Item 01) welded to a ring is identified as sean location and identification (Wol). The of the materials used in these materials l are shown in the attached Rotterdam drawing " Location, Identification and Thermal History of Vessel Parts, Weld and Surveillance Material" and Weld Material Qualification reports. ( ! l
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l
I McGuire Nuclear Station Unit 2 File No: MCC 120) 01-00-0027 Rev. O h Subiect: Evaluation of Reactor Vessel OD Flaw ~1 PIP 2-M93-0717) i Pace: 4 Bv: :dlNi- Date:S!! win Ck:U _ Vl %.__Date: f(//6 /9 3 _. ; W ' Weld Seam Wol: The following defines the method of determining the weld seam RT ndt-The nil ductility temperature is based on Charpy V-notch tests performed at a single temperature, and USNRC Branch Technical Position 5-2. Radiation shift is verified to be negligible for the limiting material (Weld seam WO1). l Material: Heat Number: 899680 Flux: Grau L.O. (LW 320) Lot P.23 L Test Reports: ; Rotterdam Lab No. P710 dated July 1972 Charpy V-Notch results (conducted at -12*C, or 10*F): I (acceptance criteria was 5.2 Kgm/cm2 or 30 ft-lbs)
- Impact Touchness Lat. Expansion % Shear '
10.0 Kgm/cm2 (57.8 ft-lbs) 0.043 in LE 47% shear ' 7.5 Kgm/cm2 (43.4 ft-lbs) 0.039 in LE 47% shear 6.8 Kgm/cm2 (39.3 ft-lbs) 0.055 in LE 55% shear Weld History: The weld Seam WO1 was weld repaired using three different combinations of heats of weld wire. The combinations, and i test report references are as follows: I Heat Test Report No. 7011/ 6143 L747 ) 7565/ 7158 0726 I 7359/ 6708 N750 The test reports show that in all 899680, test report P710) is cases the base weld (Heat limiting in terms of material { i toughness. The vessel was post weld heat treated following each i weld repair. RTndt Determination: USNRC Branch Technical Position 5-2 (Fracture Toughness Requirements) ! is used as guidance in determining RTndt. Based on section Bl.1(4) ,
"If limited Charpy V-Notch tests were performed at a single temperature to confirm that at least 30 ft-lbs was obtained, that temperature may be used as an estimate of the RTndt provided at least 45 ft-lbs was obtained if the specimens were longitudinally oriented..."
l l
e McGuire Subiect: Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. 0 Pace: 5 Evaluation Bv: ' of.
/ A // Reactor Vessel ,OD Flaw dPIP 2-M93-0717) ~
Date: J/N/._, Ck:( A ~._ E J L ( / CD9!;/S[13 ' N Date: c( ( f 6 /U ~
' ~
The test temperature of -12*C of RTndt, (10*F) can be used as an estimate this case since a minimum the specimen is a weld, of 39.3 ftrlbs was obtained. Note that in ; (weak) direction. and there is no longitudinal !t Therefore: RTndt(initial) = -12*C (10*F) i The radiation i shown to be insignificant, shiftasfor this material in this region of the vessel is follows: ! r Weld the Seam (WO1) is located approximately 200 inches (520 cm) below of 2). core midplane (refer to Rotterdam Drawin 30738-1510, sheet 1 l; The axial variation in fast neutron (>g 1.0 MeV) fluence is provided by Westinghouse WCAP 11029, figure 6-5. At approximately ! 280 cm be below(0.1%) 0.001 the coreofmidplane the (relative) fluence is noted to the peak. l The latest surveillance capsule fluence results (Capsule U, WCAP-13516) show the peak { (>1. 0MeV) to be 2.04E19 n/cm2 neutron cm below core midplane Thus, would be approximately the fluence at 280 i i (0.001
- 2E16 2E19 = 2E16). This fluence is considerably belown/cm2 threshold _(1E17 n/cm2) for the embrittlement. The fluence consideration at of neutron -
the weld seam below the core midplane) would be considerably lower. Wo1 (500 cm , The RT i predicENbradiationshift using the formulas at this fluence level can be ' follows: of Reg. Guide 1.99, Rev. 2, as cRTndt = CF
- f (0.28 - 0.10 log f)
(Where j CF = 41 all(Ni values
= 0.75, are defined and obtained using RG 1.99, Rev. 2) '
f= Cu = 0.03) 0.002E19 n/cm2 (EOL = 32 EFPY assuming no thru wall, attenuation of WO1 from core midplane) assuming axial location ap, prox. half actual distance ORTndt = 1.3 'F (considering conservatisms - insignificant) Since this RTndt initial) the lowest impac(t toughnessis based on actual material test results,and ft-lbs) by more than the accuracy (39 ft-lbs) was greater than the limit (30 of the test method calabration to 5% = E23; margin to the initial 1.5 - 2.0Additionally, values. ft-lbs) there is no nee (ASTM d to apply significant radiation shift, since there is no shift margin. there is no need to apply a radiation RTndt = 10*F Bottom Head Dome, Item 01: The following defines the method of determining the RT ndt- bottom head dome i i i I l l l l
McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. 0 - Subiect: Evaluation. of Reactor Vessel OD Flaw-4 PIP 2-M93-0717) ! Pane: 6 Bv: N/NC Date:4 Ju l'73 Ck:AJ . DK Date:N/n,M3 c m - The nil ductility temperature is based on drop weight test results, l a series of Charpy V-notch tests at various temperatures and USNRC Branch ! Technical Position 5-2. Material: SA-533 Gr. B, Cl. 1 ] Heat Number: 55292-3 ' Test Reports: Rotterdam Lab No. L573 H.D.T. Temperature (Drop Weight Tests: ASTM E208) = -40C (-40*F) ! Charpy V-Notch results (conducted at +4.4C, or 4 0
- F) :
(acceptance criteria was 5.2 Kgm/cm2 or 30 ft-lbs) {* Impact Touchness Lat. Expansion % Shear 23.8 Kgm/cm2 ! (137 ft-lbs) 0.087 in LE 95% shear i 16.6 Kgm/cm2 (96 ft-lbs) 0.071 in LE 82% shear 15.5 Kgm/cm2 (89 ft-lbs) 0.071 in LE 72% shear Additional CVN Tests: ; Temperature Impact Lateral Exp. (*C) (*F)
-90 (Kgm/cm2) (Ft-Lbs) (in)
-130 1.4 8.1 0.008
-77 -107 2.1 12.1 0.008 ;
-70 -94 6.5 37.6 0.032 i
-60 -76 6.4 37.0 0.028 I
-50 -58 5.9 34.1 0.028 3
-40 -40 6.3 36.4 0.028
-30 -22 8.6 49.7 0.039 )
-20 -4 10.3 59.6 0.047
-10 14 12.3 {
71.1 0.055 ! O 32 12.8 74.0 0.059 10 50 15.6 90.2 0.067 l 20 68 16.1 93.1 0.071 30 86 18.5 107 40 0.083 1 104 20.5 118 0.087 50 122 21.6 125 0.087 60 140 24.0 138 0.091 s At T ndt + 60 F= 20*F (-40*F + 60*F = 20*F), the Charpy V-Notch tests are aII greater than 71 ft-lbs impact energy (required to be greater than 50 j i ft-lbs) and cJreater than 55 mils lateral expansion (required to be greater than 35 mils), and therefore the Tndt may be used as the t RTndt-RTndt = -40*F i \
]
I h
- - l 1
- l l
l i
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McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O ! Subiect: Pace: 7 Evaluation. Bv: 1/ N of Reactor Vessel OD Flaw -LPIP 2.-N93-0717) i Date: tifIO Ck:!_ L- f R Date: Fs/f t /Ct3 Ring Segment Item 02-03: The following defines the method of determining the rl.g segment 02-03 i RTndt-The nil ductility temperature is based on drop weight test results, a series of Charpy V-notch tests at various temperatures and USNRC Branch i Technical Position 5-2. i Material: SA-533 Gr. B, Cl. 1 Heat Number: 55126-2-1 ; i Test Reports: Rotterdam Lab No. L589 ! N.D.T. Temperature (Drop Weight Tests: ASTM E208) = -40C (-40*F) Charpy V-Notch results (conducted at +4.4C, or 40*F): ] (acceptance criteria was 5.2 Kgm/cm2 or 30 ft-lbs) I Impact Touchness Lat. Expansion % Shear 18.6 Kgm/cm2 (107 ft-lbs) 0.075 in LE 67% shear 19.3 Kgm/cm2 (111 ft-lbs) 0.083 in LE 72% shear 15.8 Kgm/cm2 (91.4 ft-lbs) 0.067 in LE 67% shear Additional CVN Tests: Temperature Impact Lateral Exp. (*C) (*F) (Kgm/cm2) (Ft-Lbs) (in)
-100 -148 1.1 6.. 0.004
-90 -130 1.8 10.4
-80 0.008
-112 1.5 8.6 0.008
-70 -94 4.6 26.6
-60 -76 0.024 5.3 30.6 0.028
-50 -58 6.8 39.3
-40 -40 0.032 8.5 49.1 0.039
-30 -22 9.1 52.6
-20 -4 0.039 13.8 79 0.063
-10 14 14.9 86 0 32 0.063 18.3 105 0.075 0 32 16.9 97 0.071 10 50 25.1 145 20 68 0.091 24.6 142 0.094 30 86 20.5 118 40 104 0.087 21.4 123 0.087 50 122 21.9 126 60 140 0.087 23.3 134 0.099 At T ndt + 60*F = 20*F are (-4 0
- F + 60 *F = 20 *F) , the Charpy V-Notch tests all greater than 86 ft-lbs (required to be greater than 50 ft-lbs) and cJreater than 63 mils lateral expansion (required tc be greater than 35 mils), and therefore the Tndt may be used as the RTndt- .
RT ndt : -40*F 1 l
McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O Subiect: Evaluation.of Reactor Vessel OD Flaw JPIP 2-M93-0717) Pace: 6 By: ##st -%-- Date: 6/W/93 Ck: V h.h Date: S-1643 13.2 Characterization of Flaw The indication is located at 129 to 130 degrees on the lower hemisphere circumferential weld at the surface of the vessel outer diameter. The indication has been determined by B&W Nuclear Services to be planar, oriented transverse to the weld, .5 inch deep, and 2.4 inches in length. A complete report from B&W Nuclear Services describing flaw determination and sizing methods is included as Attachment 4. 13.3 General Discussion of Loads Applicable load cases for consideration are Gravity, Seismic, Pressure, and Thermal Transients. By inspection, stresses due to Gravity and Seismic are negligible. Pressure and Thermal Transients conditions are defined in references {11-3], [11-4] and the Technical Specification. Thermal Transient descriptions and numbers of cycles are given in Table 13.3-1. Appropriate combinations of Pressure and Temperature stress fields are evaluated for crack stability and growth in sections 13.4 through 13.7. . Details of methods and references are provided there. Additional residual stresses are considered in combination with stresses produced by loads. Residual stresses are conservatively estimated in accordance with ASME Section XI, Appendix E, Table E-2, 1986 edition; that is +/- 10 ksi. Tension is assumed on the exterior surface since the last welding passes were done there. s
McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O Subiect: Evaluation of Reactor Vessel OD Flaw (PIP 2-M93-0717) Pace: 9 Bv: #M Date: t9/e//f3 Ck: dm D a-v Date:8 4-95 i Table 13.3-1 Thermal Transients Reference Westinghouse Equipment Specification 952564 Rev. 1 Figure Description Cycles Condition 1 Plant Heatup 100'F/hr 200 Normal 1 Plant Cooldown 100*F 200 Normal 2 Plant Loading 5%/minu/hrte 18300 Normal 2 Plant Unloading 5%/ minute 18300 Normal 3 10% Step Load Increase 2000 Normal 3 10% Step Load Decrease 2000 Normal 4 Large Step Load Decrease 200 Normal 13 Turbine Roll Test 10 Normal 14 Hydro Before Startup 5 Normal 15 Hydro Test 0 2485 psig 50 Normal 5 Loss of Load 80 Upset 6 Loss of Power 40 Upset 7,8,9 Loss of Flow in 1 Loop 80 Upset 10 Reactor Trip From Full Power 400 Upset 11 Reactor Coolant Pipe Break 1 Faulted 12 Steam Line Break From No Load 1 Faulted Total 41867 Total w/o Loading & Unloading 9 5%/ minute 5267
McGuire Nuclear Station Unit 2 File No: MCC 1203.01-00-0027 Rev. 0 ) Subiect: Paae: to Evaluation of Reactor Vessel OD Flaw .fPIP 2-M93-0717) ' Bv: /M/=X Date: f//Y/f] Ch: v) A .& Date: A 44-9] 13.4 Evaluation 1: Heatup, Leak Test, Hydro, and Cooldown
- This evaluation considers conditions of Normal Heatup, Leak Test, Hydro, and Cooldown.
Tables 13.4-1 and 13.4-2 and their notes describe, in summary form, the results of this evaluation. Further discussion is given here. In this evaluation Transient stresses of combined cases of Pressure and Thermal temperature fields values. (and residual) are used to calculate Ky PRESSURE and TEMPERATURE STRESS FIELD and METAL TEMPERATURE DETERMINATION Since K and K material values are a function of material temperakaure, ankcK 1 values are a function of pressure, it is necessary that the evaluated. appropriate pairings of pressure and temperature are Pressure and temperature combinations for Plant Heatup, Leak Test and t Cooldown are defined by Figures 3.4-3 and 3.4-5 and discussed on page 3/4 4-30 of the Technical Specification. (Figures reproduced here as Enclosures 13.4-1, 13.4-2, and 13.4-3.)and text are ! Figure 13.4-1 shows these curves labeled as " Envelope of Tech Spec Limits for Normal for Normal Cooldown". Heatup & Leak Test" and " Envelope of Tech Spec Limits l Hydro Tests (described performed for McGuire 2. as figures 14 and 15 of Table 13.3-1) are not l Instead, pressures and temperatures as defined by ASME Section XI, Table IWB-5222-1, 1980 edition are used. Figure 13.4-1 also shows these combinations. Since only those beneath the Tech Spec limitations are valid for use, and these are enveloped by the Tech Spec conditions, no specific pressure temperature combinations for Hydro are evaluated. This evaluation is performed for temperatures beginning at 85 *F and increasing to 557 *F. Pressure values for each temperature are the envelope of Heatup, Leak Test, and Cooldown. Pressure Stress is computed using thin wall theory. This location is sufficiently remote from significant structural discontinuities to preclude the necessity for stress multipliers. i
a i McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. 0 . Subiect: Evaluation of Reactor Vessel OD Flaw (PIP 2-M93-0717) Pace: si Bv: @/ /=f Date: 8//y,/r1 Ck: VMOWs Date: 9-/643 The effects of uneven heating through the vessel wall must be included in the evaluation. These effects increase in severity with the rate of heating or cooling, and since it is heating that will cause tensile ! stress on the exterior surface (additive with pressure stress), heating ' is evaluated. Table 13.4-3 shows the temperature changes and rates for Hydro and Heatup, and shows that Heatup will be used. (Reference [11-4] specifies that the Heatup rate is 100 *F/ hr, whereas the Tech Spec limits this rate to 60 *F/hr as used here. e Reference [11-4) also begins a heatup at 70 *F, but this evaluation will begin pressure ! application and heatup from isothermal conditions at 85 'F., in line l with the curves presented in the Tech Spec.) t The stress effects of this heatup were determined using the computer t program TRANS2A, described in section 14.0. The following summarizes ; the input. Section Properties: SECTION MATERIAL INSIDE WALL AMBIENT WALL NUMBER NAME TYPE (DIAMETER THICKNESS TEMPERATURE DIVISIONS (IN) (IN) (F) l 1 LOWRHEAD CMS 176.000 5.690 70.000 20 The effects of not considering the cladding are judged negligible. ! The input boundary conditions specified a ramped temperature function I at the inside surface beginning at 85 'F at time zero and ending at 557 *F at 28320 seconds (60 "F/ hour) . f The solution was numerically determined using one time step for each l degree of temperature rise, or about every 37 seconds. ' I a The computer program internally computes a heat transfer coefficient ' based on temperatures, geometries, and flows as for the inside of a pipe. Having a high heat transfer coefficient would be conservative for determining a through wall stress gradient, but a low value would be proper for bounding the temperature lag for the metal and hence the , K la and K Ic- i The flow in this case was set equivalent to full design flow of 144.4x10 6 lbm/hr [ref. FSAR Chapter 4, Appendix 4, Table 4-1], or ! i 378000 gpm at 557 'F and 2500 psia. Note that the program assumes this - is flowing inside a pipe 176 inches inside diameter, thus the heat l transfer coefficient is lower bound. The resulting heat transfer i coefficient ranged from approximately 300 to 750 BTU /sf/hr/*F. ' f 1 i 7
4 McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O Subiect: Evaluation of Reactor Vessel OD Flaw (PIP 2-M93-0717) Pace: st- By: Wff' Date: 6//'// M Ck: 4 t>a.w) Date: B -/6 47 Figure 13.4-2 and its lower range, Figure 13.4-3, show the water f temperature and the inside, average, and outside temperature responses
~
vs. time. From these it is seen that the heat transfer coefficient is also sufficiently high to determine the stresses due to uneven heating since the maximum temperature difference between the fluid and the inner wall is less than five degrees. l Figure 13.4-4 shows the outer wall temperature as a function of the water temperature for water temperatures less than 105*F. In the determination of K Ia and K Ic, the metal temperature shown here is used ' at these water temperatures. For water temperatures at or above 105 'F, a metal temperature 18 degrees less than the water temperature is used. This reflects the maximum difference from Figure 13.4-2. ! Since no rotation takes place in the vessel wall, the resulting stress : due to uneven heating at any point at any time is E
- alpha
- OT /
(1-nu) where E is Young's modulus, alpha is the linear coefficient of ; thermal expansion, cT is the difference between the temperature at the point and the average through wall temperature at that time, and nu is Poisson's ratio. t To find this stress, Figure 13.4-5 shows a plot of the subject oT at the exterior surface as a function of the water temperature for water temperatures less than 105 'F. In the determination of this stress, i
- the cT values shown here are used at these water temperatures. For ,
water temperatures at or above 105 'F, the maximu.a OT for the . transient, 5 *F, is used. ! The resulting stress is conservatively assumed to be all membrane in calculating Ky. , Figure 13.4-6 shows the profile of stress through the section at the end of the transient. (Note the near flat slope of the profile near the exterior surface, reinforcing the decision to characterize this stress as all membrane in accordance with ASME Section XI Figure A-3200-1.) 2 RESIDUAL STRESS l The 10 ksi residual stress (as discussed in section 13.3) is applied as bending stress in accordance with ASME Section XI Figure A-3200-1. K y DETERMINATION Calculation of K 7 and comparison to allowables for the appropriate combinations of stresses are shown in summary form in Tables 13.4-1 and 13.4-2.
SUMMARY
Tables 13.4-1 & 13.4-2 show that the worst comparison of calculated Ky to allowable is 86%, even with all the conservatisms employed. I
McGuire Sub7ect: Nuclear Station Unit 2 File _No: MCC 1201. 01-00-0027 Rev. O r Pace: is Evaluation By: 4W-X of Reactor Vessel OD Flaw .(PIP 2-M93-0717) . Date: Bugna Ck: 4t D ov Date: E u _ 1 Table 13.4-1 1980 ASME Section XI Appendix A Analysis of Flaw Indications Normal Heatup & Cooldown conditions for Temperatures < 100*F Units inch, kip, 'F - Notes Water Temp fli 85.00 90.00 95.00 100.00 Metal Temp f25 85.00 90.00 95.00 100.00 Press J l 568.00 Sb 00 594.00 609.00 a J) 0.50 0.50 0.50 0.50 1 J l 2.40 2.40 2.40 2.40 t (3) 5.69 5.69 5.69 5.69 a/t 0.09 0.09 0.09 0.09 a/l 0.21 0.21 0.21 0.21 RTNDT (4) 10.00 10.00 10.00 10.00-T-RT NDT 75.00 80.00 85.00 90.00 KI (5) 64.02 66.82 69.83 Ka Ic (5) 126.12 135.89 146.70 73.07 158.63 i, om pres (6) 4.41 4.51 4.61 4.73 oT memb 0 0.84 1.78 2.45 c OT, O 0.26 0.54 0.75 o*mresid a 0 0 0 0 L m 4.41 4.77 5.15 5.47 ab pres (6) 0 0 0 0 oT bend 71 0 0 0 0 ob OT , 7? O O O O ob resid 8 > 10.00 10.00 10.00 10.00 ob 9) 10.00 10.00 10.00 10.00 i M 1.10 1.10 M 103 1.10 1.10 c$ Qys 10p 11 ) 0.92 50.00 0.92 50.00 0.92 50.00 0.92 50.00 l 121 1.33 1.33 1.33 1.32 EI 13h 15.29 15.72 16.19 16.57 V2K7 /KIc (14 0.17 0.16 0.16 0.15 V10Ky /hIa (14)) 0.75 0.74 0.73 0.72 - Notes: (1) Reactor coolant(IWB-5222-1). requirements temperature & pressureEnclosures Reference for Normal 13.4-2 Heatup&s-3100*F, Leak Test, Hydro i (2) Reactor wall outsidesee fluid temperature, metal temperature. Figure 13.4-4 'outside metal temperature lags behind the l (3) Crack depth, length & vessel wall thickness, Reference 12-4 & " Characterization of Flaw" section 13.2 (4) Determination 13.1 of Reference Transition / Nil Ductility Temperature, reference section (5) Available fracture toughness based on crack arrest & fracture initiation, respectively, for the corresponding crack tip temperature (ksiVin) as defined in ASME Section XI, Appendix A, Figure A-4200-1 (reference section 4.0 . (6) Membrane pressure stress = PD/4t l where D = 176.75*. Bending compon(thin wall theory ent of pressure hoop stress stress in a) spherical shell)
= 0.
(7) Average minus outside temperature from TRANS2A analysis (microfiche attachment M1 & Fig. 13.4-5), used to calculate o n(EnoT/(1-u) where E = 29.9E3, a = 7.12E-6 in/in/*F,than larger p = M0.3). Conservatively define all transient stress b I" KI ccmputation, reference ASME Section XI Appendix to be membrane (M,is A, article A-3300). (8) Residual stress is conservatively assumed to be 10 ksi bending per the 1986 edition ' of the ASME Section XI Appendix E, Table E-2. (9) Total membrane transient / bending
& residual stresses.stress to be used in Ky determination. Sum of pressure, (10) Correction factors for membrane & bending stress as defined in ASME Section XI, Appendix A, article A-3300.
(11) Yield stress of material @ temperature. See Figures A-3300-3 & A-3300-5 (M = 1.1 & Mb = 0.9). Appendix I, 1977. Reference Figure 13.4-7 & ASME Section III,
.(12) Shape factor for flaw as defined in Figure A-3300-1 of ASME Section XI Appendix A. l (13) Stress intensity factor as defined in ASME Section XI Appendix A, article A-3300. 3
+ oKwMb M (a/Q)
.(14) K7 = o ,to MmVnV(a/Q) r Rotio allowables; y<K
& faulted conditions, arerhu/v10fornormalconditionsandK <K ired by ASME Section XI articley IWB-N12. /V2 for emergency
_ ~ l McGuire Nuclear Station Unit 2 File No: MCC 1201.03-00-0027 Rev. O i
Subject:
Evaluation of Reactor Vessel OD Flaw,(PlP 2-M93-0717) i Paae: Ig Bv: m/;-f
, Date: A,//v, ,/.f t Ck : dhild cr<is Date: 0-/4-93 l Table 13.4-2 i
1980 ASME Section XI Appendix A Analysis of Flaw Indications i Normal Heatup & Cooldown & Leak Test Conditions for Temperatures > 100*F , Units inch, kip, 'F , Note I Water Temp (1) 105.000 130.000 150.000 175.000 200.000 215.000 230.000 235.000 Metal Temp (2) 86.890 111.890 131.890 156.890 181.890 196.890 211.890 216.890 l Press (1) 621.000 724.000 832.000 1021.000 1291.000 2000.000 2363.750 2485.000 ! a (3) 0.500 0.500 0.500 0.500 0.500 0.500 0.500 0.500 i 1 (3) 2.400 2.400 2.400 2.400 2.400 2.400 2.400 2.400 i t (3) 5.690 5.690 5.690 5.690 5.690 5.690 5.690 5.690 I a/t 0.088 0.088 0.088 0.088 0.088 0.088 0.088 0.088 a/l O.208 0.208 0.208 0.208 0.208 0.208 0.208 0.208
- RTHDT (4) 10.000 10.000 10.000 10.000 10.000 10.000 10.000 10.000 i T-RT NDT 76.890 101.890 121.890 146.890 171.890 186.890 201.890 206.890 K Ia (5) 65.058 81.774 100.268 132.368 178.492 200.000 200.000 200.000 K ye (5) 129.702 192.304 200.000 200.000 200.000 200.000 200.000 200.000 t o, pres (6) 4.823 5.622 6.461 7.929 10.026 15.532 18.356 19.298 i oT memb (7) 5.170 5.170 5.170 5.170 5.170 5.170 5.170 5.170 omoT (7) 1.572 1.572 1.572 1.572 1.572 1.572 1.572 1.572 og resid (8) 0 0 0 0 0 0 0 0 og (9) 6.395 7.195 8.033 9.591 11.598 17.104 19.929 20.870 j.
ob pres (6) 0 0 0 0 0 0 0 0 (7) 0 0 oboT 0 0 0 0 0 0 . oT memb (7) 0 0 0 0 0 0 0 0 l ob resid (8) 10.000 10.000 10.000 10.000 10.000 10.000 10.000 10.000 I ob (9) 10.000 10.000 10.000 10.000 10.000 10.000 10.000 10.000 j
- =
M (10) 1.100 1.100 1.100 1.100 1.100 1.100 1.100 1.100 ; M ,b (10) 0.920 0.920 0.920 0.920 0.920 0.920 0.920 0.920 o (11) 50.000 49.855 49.130 48.550 47.825 47.825 47.100 46.815 Q,Y [ (12) 1.322 1.320 1.317 1.311 1.302 1.277 1.260 1.253 l K7 (13) 17.694 18.670 19.701 21.511 24.119 31.069 34.754 36.005 V2Ky /K yc (14) 0.193 0.137 0.139 0.152 0.171 0.220 0.246 0.255 V10Ky /K y , (14) 0.860 0.722 0.621 0.514 0.427 0.491 0.550 0.569 Notes: [ (1) Reactor coolant temperature & pressure for Normal Heatup > 100*F, Leak Test, Hydro ; requirements (IWB-5222-1). Reference Figure 13.4-1 - (2) Reactor wall outside metal temperature. Outside metal temperature lags behind the . fluid temperature by a maximum of 18.1 *F. See Figure 13.4-2 (3) Crack depth, length & vessel wall thickness, Reference 12-4 & " Characterization of ' Flaw" section 13.2 (4) Determination of Reference Transition / Nil Ductility Temperature, reference section 13.1 (5) Available fracture toughness based on crack arrest & fracture initiation, respectively, for the corresponding crack tip temperature (ksiVin) as defined in ASME Section XI, Appendix A, Figure A-4200-1 (reference section 4.0). (6) Membrane pressure stress = PD/4t (thin wall theory hoop stress in a spherical shell) where D = 176.75". Bending component of pressure stress = 0. (7) Maximum of average minus outside temperature from TRANS2A analysis (microfiche ! attachment M1 & Figure 13.4-6), used to calculate om .(EaoT/(1-p) where E = 29.9E3, a
= 7.12E-6 infin/*F, p = 0.3). Conservatively define all transient stress to be membrane (M is larger than Mb I" K computation, reference ASME Section XI Appendix A, article A ,-3300). I i (8) Residual stress is conservatively assumed to be 10 ksi bending per the 1986 edition '
of the ASME Section XI Appendix E, Table E-2. (9) Total membrane / bending stress to be used in K y determination. Sum of pressure, ' transient & residual stresses. a (10) Correction factors for membrane & bending stress as defined in ASME Section XI, Appendix A, article A-3300. See Figures A-3300-; & A-3300-5 (M = 1.1 & Mb = 0.9). . (11) Yield stress of material O temperature. Reference Figure 33.4-9 & ASME Section III, 1 Appendix I, 1977. ! (12) Shape factor for flaw as defined in Figure A-3300-1 of ASME Section XI Appendix A. , d (13) Stress intensity factor as defined in ASME Section XI Appendix A, article A-3300. < Ky = o M mVnV(a/Q) + o M #"YI"/9) (14) Ratio E,o allowables; bh yb< KIa/V30 for normal conditions and K y <K //2 for emergency
& faulted conditions, are required by ASME Section XI article IWB-b812. i l
I McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O Sub7ect: Evaluat3on ot Reactor Vessel OD Flaw.fPIP 2-M93-0717) P_aoe: (5 By: 4/*r y Date: 6/M/? Ck : smO 4cr Date: d -/6 -5.5 Table 13.'4-3 , Thermal Transients with Teold Increasing Reference Westinghouse Equipment Specification 952564 Rev. 1 Transient Envelope 1: Startup & Hydro Figure Description Tc Rise otine (*F) (sec) 1 Plant Heat Up 100*F 487 17532 15 Hydro Test 0 2485 ps/hr ig 300 10800 (use Plant Heatup 9 60 *F/hr from 472 *F 28320 sec 85 *F to 557 *F, 472*F in 28320 sec in accordance with Tech spec) ; i l 1 I b t i
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L McGuire-Nuclear 56blect: Station Unit 2 File No: MCC 1201.01-00-0027 Rev. 0 Pace: 30 Evaluation Bv: or Reactor Vessel OD Flaw fRIP 2-H93-0717) met Date: A/Ag/o_> CR: che h Date:8-4->7 Table 13.5-2 1 1 Thermal Transients with Teold Increasing Reference Westinghouse Equipment Specification 952564 Rev. 1 Transient Envelope 2: All except Heatup, Cooldown & Hydro l Figure Description Tc Rise otime (*F) (sec) ; 3 10% Step Load Decrease 14 55 4 Large Stcy) Load Decrease 13 75
- 5 Loss of Load 34 30 7 Loss of Flgw in 1 Loop 44 15 10 Reactor Trip From Full Power 3 8 Enveloped Heating Transients 44 'F 8 sec P
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L t l l i l ! McGuire Nuclear Station Unit 2 File No: Mcc 1201.01-00-0027 Rev. O Subiect: EvaluationCof Reactor Vessel OD Flaw (PIP 2-M93-0717) Pace: 3'E By: #Af/ X Date: R//g9) Ck: c4 Wet */r Date: 4-/6-13 i 1306 Evaluation 3: Other Dounding Pressure-Temperature-Transient Cases To be further assured of having bounded all combinations of stress and temperature, this evaluation considered the worst combination of cold l metal and high stress defined at any point in all Normal, Upset, and l Faulted Transients. This evaluation is performed to consider cases in ! I which the metal is cooled in contrast to the previous evaluations in j which the metal is heated. Table 13.6-1 and its notes describe in summary form the results of this evaluation. Further discussion is given here. * ' Table or an 13.6-2 shows all transients with either a decrease in temperature ! increase in pressure and the resulting combination. Figure 13.6-1 graphically compares these combinations with the combinations for Normal Heatup and Leak Test. i These combinations are shown evaluated in Table 13.6-1. To add conservatism, a local temperature stress equivalent to that used in ' evaluation 2 is included. (Examination of Tables 13.5-2, 13.6-2 and reference [11-4) shows this to be appropriately bounding. ) l
SUMMARY
l Table 13.6-1 shows that the worst comparison of calculated KI to i allowable is 63.1%, even with all the conservatisms employed. ' l
l
- l jMcGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O <
Sub7ect: Evaluation of Heactor Vessel OD Flaw JPIP 2-M93-0717)
- Pace
- W Bv: i
/29A,L Da te : ,O//u /v , Ck: ,34me9aa r
, a g a _a - - Date:S.-h - D-Table 13.6-1 i 1980 ASME Section XI Appendix A Analysis of Flaw Indications I Dounding Pressure / Temperature values I
i Units inch, kip, 'F ! l Note Figure per Table 13.3-1 3 3 4 5& 6 7 8 10 12 ( Temp (1) l Press (1) 571.000 543.000 545.000 555.000 513.000 547.000 547.000 212.000 i a (2) 2325.000 0.500 2310.000 0.500 2350.000 0.500 2500.000 0.500 1875.000 1875.000 1870.000 2500.000 i 0.500 0.500 0.500 0.500 1 (2) 2.400 2.400 2.400 2.400 2.400 2.400 2.400 2.400 l t (2) 5.690 5.690 5.690 5.690 5.690 5.690 5.690 5.690 a/t 0.088 0.088 0.088 0.088 0.088 0.088 0.088 a/1 0.208 0.208 0.088 0.208 0.208 0.208 0.208 0.208 0.208 RTNDT (3) 10.000 10.000 10.000 10.000
)
10.000 10.000 10.000 10.000 ' T-RTNDT K y (4) 561.000 533.000 535.000 545.000 503.000 537.000 537.000 202.000 K, yc (4) 200.000 200.000 200.000 200.000 200.000 200.000 200.000 200.000 - l 200.000 200.000 200.000 200.000 200.000 200.000 200.000 200.000 o (5) 18.056 17.939 18.250 19.415 14.561 o ET (6) 4.090 4.090 4.090 4.090 4.090 14.561 4.090 14.522 4.090 19.415 4.090 I o ,resid (7) 0 0 0 l m 0 0 0 0 0 o, (8) 22.146 22.029 22.340 23.505 18.651 18.651 18.612 23.505 I ob Pres (5) 0 0 0 0 0 0 0 0 ob oT (6) 0 0 0 0 0 0 0 0 ob resid (7) 10.000 10.000 10.000 10.000 10.000 10.000 10.000 10.000 ob (8) 10.000 10.000 10.000 10.000 10.000 10.000 10.000 10.000 1 (9) 1.100 1.100 1.100 1.100 1.100 1.100 1.100 1.100 (9) 0.920 0.920 0.920 0.920 0.920 0.920 0.920 0.920 o (10) 42.100 42.800 42.800 42.300 9, Y 43.000 42.800 42.800 47.000 (11) 1.222 1.226 1.224 1.212 1.251 1.250 1.250 1.237 l Ky (12) 38.056 37.835 38.258 39.905 l 33.297 33.309 33.258 39.495 ' V2Ky/Kyc (13) 0.269 0.268 0.271 0.282 0.235 0.236 0.235 0.279 i lV10Ky /K y , (13) 0.602 0.598 0.605 0.631 0.526 0.527 0.526 0.624 i l l Notes: (1) Reactor 13.5-2. coolant temperature & pressure for miscellaneous transients. Reference Table (2) Crack depth, length & vessel wall thickness, Reference 12-4 & " Characterization of Flaw" section 13.2 (3) Determination 13.1 of Reference Transition / Nil Ductility Temperature, reference section (4) Available fracture toughness based on crack arrest & fracture initiation, respectively, for the corresponding crack tip temperature (ksiVin) ASME Section XI, Appendix A, figure A-4200-1 (reference section 4.0).as defined in
-( 5 ) Membrane pressure stress = FD/4t where D = 176.75". Bending compon(ent of pressure stress = 0. thin wall theory hoop stress in a spherical she (6) Use a value of stress for thermal transients equivalent to Table 13.5-1
(?) Residual stress is conservatively assumed to be 10 ksi bending per the 1986 edition of the ASME Section XI Appendix E, Table E-2. (8)- Total membrane / bending stress to be used in K ydetermination. Sum of pressure, transient & residual stresses. (9) Correction factors for membrane & bending stress as defined in ASME Section XI, Appendix A, article A-3300. f (10) Yield stress of material O temperature. See Figures A-3300-3 & A-3300-5 (M = 1.1 & Mb = 0.9). Appendix I, 1977. Reference Figure 13.4-9 & ASME Section III, (11) Shape factor for flaw as defined in Figure A-3300-1 of ASME Section XI Appendix A. (12) Stress intensity factor as defined in ASME Section XI Appendix A, article A-3300. K y= o Mm Ratio EoVrV(a/Q) allowables;+ b b<nV(a/0) MV (13) A K emergency & faulted cobditions,la/VIO for normal are required by ASME conditions Section and K <IWbE3612. XI artlcle K /V2 for L i
L l l l l McGuire liuclear Station Unit 2 File fio: MCC 3201.01-00-0027 Rev. 0 ! Evaluation of Reactor Vessel OD Flaw (PIP 2-M93-0717) Subiect: Pace: % By: /NApC Date: RWA./n Ck:4h Date:f46-5J t i i Table 13.6-2 l l Thermal Transients l Bounding Pressure / Temperature Values for Selected Design Transients Reference Westinghouse Equipment Specification 952564 Rev. 1 ! Figure Description oT' T cP P (*F) (*F) (psig) (psig) 3 10% Step Load Increase 14 571 75.00 2325 i 3 10% Step Load Decrease -14 543 60.00 2310 ! 4 Large Step Load Decrease -12 545 100.00 2350 5 [ Loss of Load 0 557 250.00 2500 ; 6 Loss of Power -2 555 250.00 2500 l 7 Loss of Flow in 1 Loop -44 513 ' -375.00 1875 8 Loss of Flow in 1 Loop -10 547 -375.00 1875 , 10 Reactor Trip From Full Power -10 547 -380.00 1870 ! 12 Steam Line Break From No Load 212 2500 11 Reactor Coolant Pipe Break (Enveloped by Normal Heatup by inspect.) l l l' l l i l i 1 i j l 1
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McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. 0 l' Subiect: Evaluation of_ Reactor Vessel OD Flaw ,(PIP 2-M93-0717)
-Pace: 30 Bv: NFX- Date: #//g/92 Ck: d N Date : #-/6-TJ 13.7 Crack Growth Evaluation Table 13.7-1 shows the evaluation of crack growth, evaluated in two !
parts. ; i Only those Normal and Upset stresses which cycle will contribute to I crack growth. These are limited to those caused by pressure and uneven l temperature fields. l 1 The pressure cycling is grouped into two parts. Refer to Tables 13.3-1, 13.4-2, and 13.6-2. A full pressurization occurs only 200 times. The maximum thermal stress accompanying this is 1.5 ksi. The next highest pressure excursion is 380 psig. An exterior temperature difference of 15 *F, which bounds all cases, is assumed to accompany this pressure excursion, resulting in a thermal stress of 4.56 ksi. The number of cycles for this condition is bounded by 5267, found by summing all cycles in Table 13.3-1 and removing those for loading and unloading at St/ min, which have no stress contribution. Table 13.7-1 shows that the crack growth is bounded by 0.001 inches, a negligible amount. 1 l l l l l
McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O ! Subiect: Evaluation of Reactor Vessel OD Flaw (PIP 2-M93-07171 i Page: 3'l Bv: Mry Date: 67/%/f3 Ck: ,-/# M M Date: 6'-/6 -7.7 - Page 1 of 2 ! Table 13.7-1 1980 ASME Section XI Appendix A Analysis of Flaw Indications h Bounding Cyclic Stress for crack Growth Units inch, kip, 'F Reactor Note N&U Heatup ' Temp (1) 557.000 557.00 oPress (1) 380.000 2485.00 a (2) 0.500 0.500 1 (2) 2.400 2.400 , t (2) 5.690 5.690 ' a/t 0.088 0.088 a/l 0.208 0.209 1 RTNDT (3) 10.000 10.000 T-RTNDT 547.000 547.00 i K Ia (4) 200.000 200.00 L K Ic (4) 200.000 200.00 com pres (5) 2.951 19.298 ! oT memb (6) 15.000 ; a n oT (6) 4.562 1.572 o m resid ! (7) 0 0 i oo m (8) 7.513 20=,878 ab Pres (5) 0 0 oT bend (6) 0 0 ! ab oT (6) 0 0 ! ab resid (7) 0 0 l oob (8) 0 0 - 3 Mm (9) 1.100 1.100 [ Mb (9) 0.920 0.920 z a (10) 42.516 42.516 i bs (11) 1.339 1.254 ' cK y (12) 8.952 25.293 j N (13) 5267.000 200.00 ca (14) 0.000 0.001 s Total Crack Growth 0.001 inchas i i t See notes on the following page. !
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i McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O Subiect: Pace: 40 Evaluation BV: 4%CA of Reactor vessel OD Flaw.(PIP 2-M93-0717) 4 Date: 8/Agf/3 Ck: tJAf&M Date : 6'-O F3 i Page 2 of 2 Table 13.7-1 1980 ASME Section XI Appendix A Analvsis of Flaw Indications Bounding Cyclic Stress for C tek Growth ! Notee: I (1) Bounding cyclic reactor coolant pressure temperature for Eormal & Upset and associated ; 13.3-1, 13.5-2 & 13.6-2. transients. Reference Tables (2) Crack depth, length & vessel wall thickness, Reference 12-4 & (3)
" Characterization of Flaw" section 13.2 !
Determination reference sectionof Reference 13.1 Transition / Nil Ductility Temperature, * (4) Available fracture toughness based on crack arrest & fracture ! initiation, respectively, for the i corresponding crack tip temperature (ksiVin) as defined in ASME Section XI, Appendix A, l Figure A-4200-1 (reference section 4.0). (5) oMembrane pressure stress = PD/4t (thin wall theory hoop stress in a spherical shell) where D = 176.75". Bending component ot ; pressure stress = 0. (6) Use a bounding value of 15 Tables 13.4-1, -2 & 13.5-1. *F for Normal & Upset transients. See , Use Heatup transient. See Table 13.4-2. a bounding value of 1.572 *F for ! (7) Residual stress i; non cyclic. j (8) . Total membrane i Sum of pressure / bending stress to be used in oKy determination. (9) .orrection factors transient & residual stresses.
.SME Section XI, for membrane & bending stress as defined in Appendix A, article A-3300. See Figures (10) A-330^-3 & A-3300-5 Yield stress (Mm of material @=temperature.
1.1 & Mb = 0.9). Reference Figure 13.4-7
& ASME Section III, Appendix I, 1977. 1 (11) Shape factor for flaw as defined in Figure Section XI Appendix A. A-3300-1 of ASME (12) QStress intensity factor as defined in ASME Section XI Appendix A, article A-3300. K7=o m (13) Plant Loading 5%/ minute E,M VnV(a/Q) + ob bM YUYI"/Q) wall temperature. Use an as a negligible effect on the outside envelope of all transient cycles excluding Plant Loading / Unloading 05%/ minute. -See (14) Crack Growth is determined using ASME Section XI, Table 13.3-1.
Apendix A, article A-4000, Figure A-4300-1. The subsurface crack growth law was used on this surface crack since the crack is on the outside in an air environment. I i i
i I i i McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O
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l Subiect: Evaluation of Reactor Vessel OD Flaws (PIP 2-M93-0717) Pace: 4J Bv: 4A Date: d -/MJ Ck:d I'JR Date: N/6/Q T VL S -l b O
15.0 CONCLUSION
f i This calculation evaluated the McGuire Unit 2 Reactor Vessel for an i external flaw indication as described in PIR 2-M93-D717. The vessel has been evaluated using the linear elastic fracture mechanics analysis methods of the ASME Boiler .and Pressure Vessel Code Section XI, Appendix A and the acceptance criteria of IWB-3612.
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i The evaluation included all Transient and Heatup-Cooldown conditions l described in the Equipment Specification as modified or otherwise limited by the Technical Specification. Specifically, it has been _ evaluated for a bounding Heatup event commencing from isothermal ! conditions at 85 *F and linearly ramping to 557 'F at 60 *F/hr. ! This analysis was not intended to determine any actual margins, but to [ show compliance with Code criteria. The worst case comparison to Code ; allowables was found to be 86%. !
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Further note that ana' lysis of the reactor vessel heatup using the allowable pressure from the. Technical Specification Pressure- l Temperature (P-T) curves (Figures 3.4-3 and 3.4-5) results in a , conservative bounding of actual Unit operation. The low temperature ! i overpressure protection (LTOP) system prevents the reactor coolant ! system from being pressurized above the P-T curve up to the LTOP enable ; temperature of 300F. The Unit startup procedure- (OP/2/A/6100/01) ! administratively maintains LTOP up to 320F. Up to the enable i temperature, the pressurizer PORV's limit pressure by.the 365 - 395 low ! pressure mode setpoint. The peak pressure, including instrument j accuracies, PORV opening times and valve accumulation is calculated to ' be 483 psig (ref. Duke Power pressure is adequate to maintain the calculation MCC-1223.03-00-0033). This , reactor vessel below the P-T limits for a 60 *F/Hr heatup, down to temperatures as low as 85 *F. Maintaining the same PORV setpoint up to the enable temperat*. ' of 300 *F keeps actual operation well below the P-T curves _. the Technical the Specification and makes the pressure stresses very low during heatup portion when vessel temperature is low and brittle fracture is a concern. For the above reasons, the analysis using actual P-T limits rather than the LTOP imposed peak pressure is very conservative. This calculation concludes that the subject flaw does not limit operation beyond any previously set criteria for the 40 year life of the unit. 1
1 i l i McGuire Nuclear Station Unit 2 File No: M C C__,1 2.0 1 0 1 - 0 0 - 0 0 2 7 Rev. O : Subiect: Evaluation of Reactor vessel OD Flav iPIP 2-M93-0717) ! Pace: 44 Bv: 4/Af Date: A,//Q2 ck: 44v3 % Da t e : B -/< - r> j 13.8 Net Section Evaluation Acceptable by inspection. I i
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t I McGuire Nuclear Station Unit 2 File No: MCC 1201.01-00-0027 Rev. O T i Subiect: Evaluation of Reactor Vessel OD Flaw JPIP 2-M93-0717) Pace: M BV: @ f/c X Date: A//v/rt Ck: cdwh ( Date: 8-/6 93 1
14.0 DESCRIPTION
OF COMPUTER PROGRAMS USED 6 i TRANS2A i i
' TRANS2A is a computer program which determines radial temperature distribu-tions and gradients in a pipe wall experiencing fluid temperature excursions TRANS2A determines these temperature distributions by solution of the unsteady one-dimensional (
axisymmetric heat transfer equation. The fluid boundary properties (temperature, flow) ! at the end of each calculation step are used to determine the heat transfer for that step. For aid in Class I piping analysis values of the thermal gradients, AT y and AT ad j the average temperatures (T and/or T ) bare calculated in accordance with ASME2BrePVC Section III Article NB-3650. ,To be of more aid to the analyst in choosing values of the { l average and temperature gradient data to be input to the combined stress analysis, ! { TRANS2A evaluates the actual histories of the thermal stress terms according to the equations of Section III, Article NB-3650 with as many as ten sets of stress indices and ~ ) j summarizes them in a table by extreme and time of occurrence.
- TRANS2A has been extensively tested and compared with independent results for j I
sample problems. TRANS2A temperature distributions agree favorably with calcula- ! tions using TRANSIA and the EDS proprietary finite element program TAPAS. Calcula- ! tions for thermal gradient terms aT1 and AT compare 2 favorably with results from ' TRANS1A and with values derived from charts published by McNeill and Brock in
" Engineering Data File - Charts for Transient Temperatures in Pipes",1971.
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