BVY-91-118, Requests Further NRC Review of Flaws Found in Feedwater Check Valves at Facility in Light of Recent Failure Analysis Conducted by BNL Under Contract to Nrc.Nrc Concurrence Requested That Replacement of Valve V27B Unnecessary
| ML20086N967 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 12/19/1991 |
| From: | Pelletier J VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| BVY-91-118, NUDOCS 9112260099 | |
| Download: ML20086N967 (44) | |
Text
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VERMONT YANKEE
- NUCLEAR P WER CORPORATION fI t Ferry noad. Brattleboro, vT 05s01-7002
( dj/K ) vice r' resident, Enginetring (8023 2s7-5271 December 19, 1991 BVY 91118 United States Nucinar Regulatory Commission ATTN: Document Control Desh Washington, DC 20555
References:
(a) License No. DPR 28 (Docket No. 50 271) b) Letter, VYNPC to USNRC, BVY 89-031, dated March 28,1989 c) Letter, USNRC to VYNPC, NVY 89-067, dated April 5,1989 d) Letter, VYNPC to USNRC, BVY 90-033, dated March 16,1990 e) Letter, USNRC to VYNPC, NVY 90-073, dated April 19, 1990 (f) letter, VYNPC to USNRC, BVY 90-097, dated October 4,1990
-(g) Letter, USNRC to VYNPC, NVY 90-178, dated October 10, 1990 (h) Technical Report MT-L1529 3 " Evaluation of Cracks Found in Stellite Valve Guides at Vermont Yankes Nuclear Power Station," September 1991
Subject:
Feedwater Systern Check Valves
Dear Sir:
The purpose of this letter is to request further NRC review with regaru to flaws found in feedwater check valves at Vermont Yankee and, in lig1t of recent failure analysis conducted at Brookhaven National Laboratory under contract to NRC, to obtain NRC concurrence that repair or replacement of feedwater check valve V27B Is unnecessary.
During the Vermont Yankee Cycle 13, Spring 1989 outage, visual cracking was observed in the stellite wear pads of feedwater check valve V28B. Ultrasonic examination was performed to determine flaw depth and it was determined that the flaws exceeded the acceptance standard of the ASME Code,Section XI.
The flaw depth was conservatively sized at 0.65 inches deep and the valve body thickness at the location of the flaw was 2 inches. As required by Section XI, the three remaining valves of the same design (V28A, V278, and V96B) were inspected; similar visual cracking was observed, but the ultrasonically determined flaw depths were within ASME Code acceptance limits. Based on our inspection of the valves and experience with similar situations, it was concluded that the cracks were fabrication related and not serviced Induced.
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V'ECRMONT YANKEE NUCLEAR POWER CORPORATION US Nuclear Regulatory Commission ,
December 19, 1991 Page 2 As required by the ASME Code, a fracture mechanics evaluation was aerformed to demonstrate the suitability of valve V288 for continued service. T1e results of the fracture mechanics evaluation were submittod to the NRC via Reference (b). In Reference (c) and as a condition for startup, the NRC required that Vermont Yankee commit to repair or replace valve V288 during the Cycle 14, Fall 1990 outage. We also installed moisture sensitive tape to monitor for any '
possible leakage that might develop in the unlikely event that flaw Geath were to grow through wall during the operating cycle. Supplemental information relative to the flaw evaluation of valvo V28B was submitted via Reference (d).
The NRC acceptea the results of the Vermont Yankee evaluation of the V28B flaw, and issued a Safety Evaluation Report (SER) In Reference (e). In the SER, the NRC stated that, "The licensee has not provided sufficient information for us to conclude that the cracks remainlng in the three other valves will be stable '
and static flaws. We request that the licensee either replace / repair or in accordance with Section 50.55a(g)(6)(li) provide an augmented inspection program for the three other valves [V28A, V278, and V968)..."
During the Cycle 14, Fall 1990 outage, in accordance with our commitment and the NRC SER, we realaced the two inboard check valves, V28A and V288.
These valves were rep aced with swing check valves of the same design as the existing first outside containment valve in each foedwater line, valves V27A and V96A. An augmented inspection program, consisting of uitrasonic examinations, was also conducted on the two remaining lift check valves, V27B And V96B during the Fall 1990 outage. These examinations were performed using larger, lower frequency transducers with improved capabilities for examining cast carbon steel. These inspections sized flaws in valve V968 at 0.15 Inches, which is within the acceptance limits of ASME Section XI. However, in valve V278 the maximum flaw depth was estimated at 0.50 inches, which exceeds the Section XI acceptance limit. The wall thickness at the location of this flaw is 2.6 inches and the flaw is contained totally within the width of the stellite waar pad.
As stated above, V27B is identical to V28B that was considered in the previous flaw evaluation. The flaw in .V27B is srnaller than the V288 flaw evaluated; therefore, the fracture mechanics analysis performed for V288 and accepted by the NRC bounds the flaw depth of V278. The results of the UT inspections rand the determination that the V28B flaw evaluation from References (b) and (d) bounds the flaw discovered in V27B was reported to the NRC verbally and in Reference (f). As a condition for startup, the NRC required a commitment from Vermont Yankee to install moisture sensitive tape for leak monitoring purposes during the following operating cycle and to repair or replace V27B during the Cycle 15, Spring 1992 outage. NRC acceptanco of the Vermont Yankee analysis was received in Reference (g). The V28B flaw evaluation, that bounds the V27B flaw is provided agaln for ease of review.
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VERMOAT YANKEE NUCl.E AR POWER CORPORATION US Nuclear Regulatory Commission December 19, 1991 Page 3 Subsequent to the Cycle 14 outage, NRC requested that Vermont Yankee provide the now removed check valve V28B, containing the flaw, to an NRC contractor (Brookhaven Natiorial Laboratory) for a failure enal'.31s. The fa!!ure analysis
[Refere..ce (h)] concluded that, "The cracking was probably the result of the original welding process being inadequate to produce a ' defect free' weld deposit." This supports our 1989 determination that the cracking was fabrication related, and not service induced. In Attachment 1 we address the findings of the Brookhaven Report in support of our conclusion that repair or replacement of valve V27B is unnecessary Additionally, Attachment 2 responds to some concerns raised by Mr. Bob Hermann, Materials and Chemical Engineering Branch, Section Chief of thc NRC staff during a December 12, 1991 telecon.
While our current commitment to the NRC involves only repair or replacement of V278, Vermont Yankee has been developing plans to replace both feedwater check valves V27B and V968. Feedwater check valve replacement is a significant undertaking. The current cou estimate is greater than $700,000 and the ALARA estimate is 2 person-rem. The current draft cf the 1991 refueling outage schedule estimates the replacement effort will take 23 days and may impact the critical path of the outage. Tha work would be scheduled to begin as perscan weekasbasis, possible after'wo utilizing plant 12-hour shutdown, shifts.andAny willminor take place on aslippage schedule seven day!
wn impact the critical pati 1, affecting both the vessel hydrostatic test and the Type A Integrated Leak Rate Test of Primary Containment. The elimination of this job will allow increased management attention on o'her critical outage evolutions.
Vermont Yankee believes that the failure analysis provided by Brookhaven in Reference (h), in combination with out previously submitted fracture mechanics evaluations, provides assurance that the flaws are stable and static. Since replacement of feedwater check valves V27B and V96B would impose an unnecessary cost and dose burden on Vermont Yankee, it is requested that NRC provide concurrence that these valves need not be replaced, and accept the conch:sions of Vermont Yankee's failure analysis without condition.
We trust this information adequately supports our request and is sufficient to allow NRC timely review; however, should you have questions on this matter, please contact us. In order to minimize the impact in the current outage plans, we rcquest a response by January 31, 1991.
Very truly yours, Vermont Yankee Nuclear Power Corporation
.W ames P. Pelietier
> Vice President, Engineering
- Attachments cc
- USNRC Region i Administrator USNRC Resident inspector - VYNPS USNRC Project Manager VYNPS l
, VERMONT YANKEE NUCLEAR POWER CORPORATION ATTACHMENT 1 ,
BROOKHAVEN NATIONAL LABORATORY FINDINGS Flaw Size Valves V28A, V28B, V27A and V96B were all Inspected by the same inspector (a qualified Level 10 with EPRI sizing certification) using the same procedure and equipment. Flaw tiepths in V28A and V28B were alzed between 0.3 inches and 0.5 inches. The inspector believed he was "over-calling" the depth. He placed a b;gh vinfidence on detection and a lower confidence on accurate sizing. The dest /uctive analysis of V28A and V28B showed that the flaws were within the width and thickness of the stellite, with the flaw depth being controlled by the tnickness of the stellite. The largest flaw was approximately 0.25 inch deep. This determination supports our conclusion that the flaws are conservatively sized and results in the fracture mechanics evaluation being even more conservative than previously believed.
Flaw Growth In 1989 the USNRC expressed concern that the valves could experience rapid flaw growth. A comparison was made to the steam generator shell
. cracking that was observed in some pressurized water reactors. The Brooknaven Report did observe " apparent crack propagation into the base metal in some of the cracks [which) suggests the possibility of corrosion cracking after a leak path to the valve inside serface had been established." The pussible corro lon crack depths are less than 1 mm (less than 0.04 inches) in depth. Given the length of time the valves have been in service (since 1971) and the thickness of the valve body (two
. inches), we consider this posSible corrosion cracking to be insignificant.
In the fracture mechanics evaluation submitted by Vermont Yankee in 1989, fatigue crack growth was evaluated and detamined to.be negligible.
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1 Vh:RMOhT YANKEE NUCLEAR POWER CORPORATION ATTACHMENT 2 DISCUSSION OF ISSUES RAISED DURING NRC TELECON Stellite Weldino Process The Brookhaven Report is crit! cal of the process used to weld the stellite onto the valve body. "All of the cracks appear to be associated with some sort of welding defect (e.g. Inclusions, lack of fusion, vold). The cracking is predominantly interdendritic, typical of hot cracking" (which means it is fabrication related and not service induced). During the failure analysis process, six two inch diameter plugs were core drilled out of the valve bodies (two from V28A and four from V288). The plugs were cut into sections and the samplos bent to open the crack to observe the crack faces. During this process one of the samples had a piece of stellite pop free, due to lack of fusion. We do not consider this relevant. The piece of stellite didn't separate from the base metal until the sample was reduced to a small size by cutting it free from the valve body. In the valve body the ploco was restrained and integral with the surrounding stellite. The stellite piece did not " pop free" until the specimen was subjected to loading totally unrepresentative of the service loading in the valve.
However, even if a piece of stellite would break free in the plant, it would not pose a safety concern. The purpose of the two chec< valves is to direct HPCI and RCIC cooling water into the vessel in the event of a loss of normal feedwater flow. There are additional check valves further upstream in the feedwater system that can serve the same function (valves V63-1 A,18 and 1C on FSAR Figure 11.8-1). If these backup valves fall to function, manual or automatic reactor depressurization can be initiated to reduce reactor pressure and allow core spray to be actuated.
Post Weld Heat Treatment The Brookhaven Report concludes that, based on hardness measurements of the stellite, that the valve bodies were not post weld heat treated following the application of the stellite. This is correct as the ASME Code does not require post weld heat treatment for that situation.
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I EVALUATION OF TEEDWATER CHECK VALVE V28B FLAWS I
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EUMMARY OF FINDINGS ,
Q Feedwater Check Valve V28B was opened for verification of f ree piston mcvement as part of the Vermont Yankee In-Service Test Program. While the valve was open, a visual examination of the inside surfaces of the valve was performed as required by ASME Section XI. Visual cracking was observed in the Stellite I No. 6 wear pads in the piston guide portion of the valve (see Figure 1) The cracking was located within the approximately 1 inch wide Stellite No. 6 wear pads. The wear pads were created by machining a shallow groove in the bore of the piston housing and weld depositing Stellite No. 6 alloy. The bore was then machined to provide a cylindrical bore for the piston.
Ultrasonic inspections were performed on the valve. The thickness of the valve in the region of the flaws is approximately 2 inch to 2.1 inch. In valve V28B two flaws were identified, one wi a maximum flaw depth of I .65 inch and the other was a 0.40 inch deep .uw on the other sd.de of the valve. The deeper flaw was radiographed, confirming the fact that the flaw was contained entirely within the width of the Stellite No. 6 pad. The ultrasonic examiners reported that casting inclusions made it difficult to discern the crack tip from the inclusions (i.e., the flaw could be shallower than reported). The casting inclusions were also seen in the radiograph.
Since the reported flaw depths were in excess of Section XI acceptance criteria, detailed flaw evaluations were required.
I Discussion of Flawg All of the flaws were contained entirely within the width of the Stellite No.
6 wear pads (see Figure 1). Since the majority of the flaws are very shallow, this implies a very slow or nonexistent growth mechanism. Stellite No. 6 is a brittle material, and Stellite No. 6 cracking is not an uncommon occurrence.
Inspection of the Stellite No. 6 showed visible evidence of "between bead" cracking and slag inclusions. (The Stellite No. 6 serves no pressure retaining function.)
I Possible flaw initiation mechanisms have been evaluated. The thermal expansion coefficients of carbon steel and Stellite No. 6 are similar, so differential thermal expansion stresses and consequently thermal fatigue crack
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t e . .r propagation will be low. It 4.a highly likely that the majority or cracks remain in the Stellite No. 6 or extend olightly into the residual stress
( region resulting from the Stellite No. 6 weld deposit. (The Stellite No. 6 e thickness is approximately 90 mils). Pressure cycling has been evaluated and shown to produce negligible flaw growth.
f The two deeper flaws in Valve V2BB m.2y bu linked up to casting inclusions, or may even be simply in front of nonconnected casting irc.lusions.
Vermont Yankee concludes that all the flaws are stable, sratic flaws resulting from Stellite No. 6 cracking coincident wir.h casting defecta. The indications f are in a region of the valve body that will see high pressure induced stresses, and it is possible that the flaws developed during the original valve hydrotest at 3,250 psig. In any event th+ flaws are evaluated considering the full reported flaw depth.
Flaw Evaluatiann
{ The region of the valve body containing the flaw is more complex than a simple cylinder, and the valve body material (AS'IM A216 WCB cast carbon steel) is not a low alloy pressure vessel steel, so the " cookbook" flaw evaluation f
techniques of ASME Section XI cannot be utilized. Instead, a detailed fracture mechanics evaluation was performed utilizing a bench-marked, industry-accepted computer code (pc-CRACK, developed by Structural Integrity Associates).
In order to better represent the stress condition in the intersection region
- of the valve where the flaws are located, a two-dimensional finite element model was developed using ANSYS. The details and conservatisms associated with this model are discussed in Appendix A of this report.
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Since the region of the valve containing the flaw is not a standard geometry, several fracture mechanics cases bounding the actual case were performed. The table below sumarizes the cases evaluated and the yK for a flaw 0.65 inch deep. Flaw evaluation reports are contained in Appendix B.
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The flaws were evaluated for two conditionst using a design pressure for the piping, system of 1,900 psig; and, using an operating pressure for the system i of 1,100 psig. The design condition could only occur when the downstream t-manual isolation valve is shut, subjecting the piping to the combined shut of f heads of the condensate and feedwater pumps. This is classified as a test condition.
The following table lists the conditions that were evaluated and the Ky values at 1,100 psig and 1,900 psigt f,
K7 at a = 0.65 inches
! Elaw Evaluation Model 1.900 nsig L lOQ_ prig (Units of kai - /in)
Elliptical Flaw in Cylinder (a/1 = 0.2) 10.1 5.8 Elliptical Flaw in cylinder (a/1 = 0.5) 6.7 3.9 Fully Circumferential Flaw in Cylinder 13.9 8.0 Infinite Longitudinal Flaw in Cylinder 17.4 10.1 Elliptical Flaw in Flat Plate Subject to 11.2 6.5 Bending e.nd Tension Limit 24.5 12.2 The above listed K7 va?aes must be compared against the KIc ** "** ' #
valve material. Since no impact testing was performed on the salve bodies at the time of manuf acture, typical data from published reports have been used.
, ASME Code Case N-463 provides a lower bound KIe **1"* I # f* 'I" ****1 g piping, such as A106, Grade B. NRC Contractor Report NUREG/CR-300v and NRC Report NUREG-0577 show that A216 WCB has superior toughness properties compared to A106, Grade B; therefore, it is conservative to use the lower 7570R
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r L bound value f rom Code Case N-463; the lower bound K s 6.7 kai An.
Ic Jtilizing the appropriate factor of safety for the normal and test condition h provides an acceptance criteria of K Ic equals 24.5 kai /in for the test condition and 12.2 kai /in for the normal condition. As can be seen, both conditions satisfy their respective acceptance criteria by a significant
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margin.
As a worst case evaluation, we have considered the possibility that a flaw grows through wall. Figure 2 shows the hypothetical flew sizes for the different evaluation models, as compared to the assumed initial flaw. The limiting mode of operation for this condition is RCIC injection. RCIC f initially draws from the condensate storage tank. After the condensate
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storage tank is drawn down, suction is switched to the torus. For conservatism, we have assumed the RCIC injection water could be at 80'F. the
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minimum reported condensate storage tank temperature in winter. Figure 3
- shows the calculated K values for the various flaw models compared to g
typical K data at 80'F extracted from NUREG/CR-3009. Significant margin k
(Note that the center cracked plate flaw Kg is artificially high
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since the model does not allow a varying stress field. The peak stress had to be applied across the full thickness of the plate.)
CONCLUSION Flaws were detected during in-service inspection in the feedwater check valve. The valve had two flaws exceeding ASME Section XI acceptance criteria. Fracture mechanica evaluations were performed and compared against conservative materials properties.
In all cases up to and including a through wall flaw, stable crack behavior is demonstrated and, therefore, gross failure will not occur. As added assurance t
of safe plant operation, an enhanced leakage monitoring pcogram will be implemented. The specifica are discussed in Appendix D. In addition, repairs or replacement of V28B will be performed at the next scheduled refueling outage.
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' Finally, even if the feedvater check valve were to be conservatively assumed to experience a gross failure during operation, the resulting could be equivalent to a f eed - .er line break inside containment, which is witbin the L
plant accident analysis, so no unreviewed safety question exists.
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= 0.2 7 I a/ ^" "~c F'"'
a/ = 0.5 l o:e Cen~:er-Crac<ec
,__ c ~ r W - ~
L1J
_L 100.00 -
O :
Z :
8 Typical KIC Valve _for A_216_q at_Minimu_m Valve _Tegperature (T = 80 F for RCIC Injection Mode)
[ -
I O :
m :
l center cracxed P1 ate m 60.00 f I
Thru-wall Flaw 3 Inches Wide
")
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r1 Plate under bending
' and tension
" l 2
' Elliptical flaw in cylinder 40.00 7 - / a/1 = o.2 4
- /
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.+ J . /
/ Elliptical flaw in cylinder l , / a/1 = 0.5
( .
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+ 20.00 -
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O 0.00 J,......... .. ...... .........i.........i 1.50 2.00 E 0.00 ~~
0.50 1.00 Jey:, nc,es m ,ru-Wa! .
s , ,.,
L.
E L, .
- I APPDiDIX A STRESS ANALYSIS OF FEEDWATER CHECK VALVE V28B I
I I
E. J. Betti I J. C. Fitzpatrick I -
l 7570R
- i APPENDIX A Stress Analysis of Feedwater_ChgIk_ Valve V283 L
A. MILICABLE_ LOADINGS
- 1. Pressure
- a. Design pressure 1,900 psig (G-191167). This valve is based on feedwater pump dynamic head plus condensate pump head (G-191139).
- b. Operating pressure in thit gion is limited to reactor operating prescure plus the pressure crop across the feedwater spargers.
During full feedwater flow, 1,250 psig is assumed, 200 psig over RPV operating pressure. During low feedwater (<10% of rated) 1,100 psig feedwater operating pressure is assumed.
- 2. Mechanical Lonsla Mechanical loads are from the attached piping. The section of the check valve in question has a 20 inch OD with a 2 inch wall thickness. The attached piping is 16 inch SCH 120; t = 1.218 inch.
I The section modulus of the valve is as a minimum 2.4 times larger than the attached pipe in the region of the flave. From combined dead weight, thermal, and seismic piping moments at the valve I (VYC-634), valve stress was calculated to be less than 1,200 psi in the side region of the valve in the area of the detected flaws. In this region, the valve profile is flat. Therefore, localized through I wall bending is not a concern.
- 3. Rater Hamer and Valve Impact I Both vater hamer or piston impact-induced stress were considered small. This system has not been subject to water hammer events.
Also, with the exclusion of a double-ended pipe break upstream of the I check valves, the Feedwater System is not subject to rapid pressure decreases which could result in rapid valve closure. Finally, the piston structure is much lighter than the valve body. Therefore, in I the event of rapid valve closure, the piston, not the valve, would absorb the majority of impact energy,
- 4. Ihemal Transient-Induced Stresa Full power and partial power transients do not result in severe temperature transients in the region of the 28B check valve. The I largest potential thermal gradient that this valve could experience would be during a zero power hot standby condition when feedwater is in the low flow control mode (<10% of rated flow).
The B feedwater line is also used for RCIC, clean-up water return, and CRD return. The following is a stenary of the system capacities:
o RCIC - 416 gpm capacity at 80*F o CUW - 130 gpm at 430*F o CRD - 60 gpm ut 115'F o FDW - 7,700 gpm at 375'F I
7570R
\ o ' s APPENDI]LA (Continued)
The region of the valve in question is subject to membrane and through wall bending due to pressure. Hot-to-cold transients tend to decrease the through wall bending while cold-to-hot transients would add to pressure stress. Therefore, the following-cold-to-hot transient was selected for investigation:
Feedwater at 10% flow, 100'F with CW water at 100% flow, 430*T.
(Combined temperature of 152'F). Interruption of feedwater flow, continue 100% CW flow at 430'F. The pressure is assumed to be at 1,100 psig.
. B. STRESS MODEL FOR ANALYSIS From field walk down of the feedwater check valve and in situ dimensions, it was apparent both membrane and local bending were important in the flaw region. The two dimensional constant strain model shown in Figure Al was used for both stress and thermal analysis. The ANSYS finite element code was used to perform calculations and plots. This simplified, two dimensional model provides approximations of local membrane and bending stress in the flaw region.
C. STRESS PROFILE FOR FRACTURE MECHANICS ANALYSIS The first case evaluated with the model was the effect of 1,900 psig internal pressure, the design pressure of the valve. This condition resulted in compreesive forces on the insi.le face in the flaw region. A cection stress profile for a 1,000 psig case is shown in Figure A2. The 1,900 psig stress profile was interpreted from these results.
From the finite element model results, an enveloping stress profile for fracture mechanica study was developed. The cc:npressive stress profile on the inner face was changed to a constant 7,300 psi tensile stress to I approximately mid-thickness. Toward the outside wall, a linearly increasing stress profile from the model was used. Changing the compressive partion-of the stress profile to tensile provides a conservative " design" envelop for fracture mechanics evaluation.
To assure that the " design" stress profile is conservative, the maximum stress profile under thermal transient conditions was also studied (see I- Figure A3). The transient stress was combined with mechanical stress from piping and pressure stress and plotted on Figure A4 for comparison with the " design" stress profile. Figure A4 demonstrates that the I " design" profile was an appropriate choice for fracture mechanics evaluation.
I 7570R I
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-7500 DIST l I
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1 l 2 l 2.5 O I o.5 l 1 1.5 0.75 1.25 1.75 2.25 0.25 FEEDWATER CHECK VALVE. V_Y ____
. g m- - - , m 1 --- i. .r FICuae A3m: Tem renn rare e2coM ..rdr_cgLSeclido u,n go ,,g ]
Ab 50 E9 A' [L Al.) s< f~UfA97 l INTff UP7l0DE)DH?fV'Ur.'M*E POST 26 TEMP flo c zy -1 DIST=0.66GS g]O'f- 5 TEMP g YF -0.5
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FEEDWATER CHECK VALVE. VY _ _ _ _ _
M M M M M -
1 W' U 'LE-h~bb : jyAL 37QSg N_rc1vgh_.{P2@@^> G QAtd y 3n , '_
E vro) 7.' 0 *O 5 71r or1loto_Elo<.. r&*F1:u n r r< Rcw gy 12-5 17haurr: _ A r re.c 5 rr t' C.J) A r i n e STEP-1 U ITER-90 12500 ] SECTION PLOT
! NODI-35
~
10000
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[ STRESS GLGBAL 7500 ~
TOTAL DIST-0.666G
- G XF -0.5 g, 4 5000
- YF =0.5 ZF =0.5 p'
2500 f,p
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-10000 Il.6 ds t j -12500 l g l l g j j -]DIST o 0.5 1 1.5 2 2.5 l
I O.25 0.75 1.25 1.75 2.25 FEEDWATER CHECK VALVE. VY , _ _ _ __
l
-iANKEE ATOMIC ELECTRIC COMPANY cucuutiono exaE w L
JDJECT.
_I""" wonK ORDER HO PREPARED BY _h > DATE 20 87 REVIEWED BY DATE _
Ficurts A4- AxDs Tho gh LJ uu k r.eca e FLAW L
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B1Cc/A~rce { / Loc /7 0 n / Loo i s.c o / 2. c o TEN)' -!P.> i ~//5 0 3 ~3700 /400 5600 6700
~7T-M - S!d 4433 ')?~^ /3/76 Tv7AL' -li gg y -23lj Japo jgiy o p icy b
I L
F L
I APPENDIX B FLAW EVALUATION RESULTS I
I .
I I
I J. C. Fitzpatrick J. R. Iloffman i
I l
7570R
'a .
c.
A - Plot of Stress Profiles i
'- B - Comparison of Flaw Profiles p_ C - Least Square Curve Fit Profile L
D - Elliptical Flaw in cylinder - a/l = 0.2 E - Elliptical Flaw in Cylinder - a/l = 0.5 F - Full Circumferential Flaw in Cflinder G - Infinite Longitudinal Flaw in Cylinder i H - Elliptical Flaw in Flat Plate I - Center Cracked Plate J - Extrapolation of Elliptical Flaw to a Thru-Wall Flaw K - Fatigue Crack Growth for Semi-Elliptical Flaw 1.5 Inches Deep at Three Times Applied Strecs I
I I
I I
I I
I o
M M '. .M .
M M M 'W M S
L
- p .
I
/
Thru-wall Stress
. for Center Cracked j 20.00 : Plate .
~
Ccunparison
- / of
- Stress Profiles (f) j Used for y .
- / Evaluation
- of 15.00 i - / reed.sater Check Valve
/
V288 I -
/
fl)10.00i M
/
g rNCurve Fit Stre'ss Profile e-- e-- e w l Conservation Stress Profile km FE Analysis 5.00 i .
W W
m W
W 0.00 0.50 1.00 1.50 2.00 2.50 Jis :ance nc,es
~~
, ru-wa
" " = = ------.......
aw Sla3es
.Ccanorison of
~or T
rac~:ure Mecionics ~vo'uation 4
T eecwo:er Caec< Va ve V283
,m N
y il[/' a i 1
A33 vined Flas a/ = 0.5 Cen~:er-Crackec 3a:e 1
I i _--- _
- pc-CRACK e (C) COPYRIGHT 1984, 1987 STRUCTURAL INTEGRITY ASSOCIATES, INC.
SAN JOSE, CA (408)978-8200 VERSION 1.2
{
LEAST SQUARE CURVE FIT OF STRESS PROFILE
[:
lRMONT YANKEE FEEDWATER CHECK VALVE V28B TERM COEFFICIENT 8.0867E+00 I 'C0 C1 -5.335E+00 C2 5.8800E+00 C3 4.0177E-01 1
'OEFFICIEt;T OF DZTERMINATION R'2= 0.9845
)RRELATION COEFFICIENT = 0.9692 X VALUE Y VALUE Y CALC DIFF I0.0000E+007.5000E+00 8.0867E+00 -5.867E-01 1.2500E-01 7.5000E+00 7.5124E+00 -1.240E-02 2.5000E-01 7.5000E+00 7.1266E+00 3.7340E-01 7.5000E+00 6.9340E+00 5.6604E-01 13.7500E-01
- 5. 0 0 00 E-01 7.5000E+00 6.9392E+00 5.6000E-01 6.2500E-01 7.5000E+00 7.14 7 0 E+0 0 3.5298E-01 7.5000E+00 7.5621E+00 -6.213E-02 1.7.5000E-01 8.7500E-01 7.5000E+00 8.1892E+00 -6.892E-01 1.0000E+00 7.5000E+00 9.0331E+00 -1.533E+00 1 2500E+00 1.1515E+01 1.1390E+01 1.2535E-01 1.5530E+01 1.4670E+01 8.6043E-01 11.5000E+00 1.7500E+00 1.9545E+01 1.8910E+01 6.3452E-01
"!.0000E+00 2.3560E+01 2.4150E+01 -5.900E-01 END OF pc-CRACK I
I I
I I
I I
tm [)
~
(C) COPYRIGHT 1984, 1987
+ STRUCTURAL INTEGRITY ASSOCIATES, INC.
SAN JOSE, CA (408)978-8200
- VERSION 1.2 I; D LINEAR ELASTIC FRAUTURE MECHANICS EVALUATION I
VERMONT YANKEE FEEDWATER CHECK VALVE V28B C: RACK MODEL: ELLIPTICAL LONGITUDINAL CRACK IN CYLINDER (T/R=0.1,A/L=0.2) 4ALL iHICKNESS= 2.0000 STRESS COEFFICIENTS CASE ID CO C1 C2 C3 I THRUWALL FWNRC 23.5000 8.0867 -5.3350 0.0000 O.0000 5.8800 O.0000 0.4018 i CRACK ---------------STRESS INTENSITY FACTOR----------------
DEPTH CASE CASE THRUWALL FWNRC O.0320' 6.475 '2.198
'O.0640 9.216 .3.089 1 0.0960 0.1280 11.360 13.200 3.761 4.321 0.1600 14.851 4.811 0.1920 16.370 5.251 1 0.2240 17.789 5.655 0.2560 19.132 6.033 O.2880 20.413 6.390 i 0.3200 0.3520 21.645 22.836 6.732 7.062 m 0.3840 23.991 7.384 7.705
[ 0.4160 0.4480 25.133 26.268 8.027 O.4800 27.383 8.348 8.670 i 0.5120 0.5440 0.5760 28.479 29.561 30.629 8.993 9.320 0.6080 31.686 9.651 1 0.6400 0.6720 32.736 33.777 10.336 9.990 O.7040 34.811 10.690 i 0.7360 0.7680 35.836 36.856 11.053 11.426 0.8000 37.870 '1.810 0.8320 38.874 12.202 1 0.8640 0.8960 39.874 40.869 12.607 13.024 0.9280 41.861 13.456
pc-CPACK VERSION 1.0 PAGE 2
_ O.9600- 42.849 13.901 0.9920 43.835 14.362 1.0240 44.866 14.857 1.0560 45.913 15.375
.1.0880 46.960 15.912 1.1200 48.000 16.470 ,
1.1520 49.056 17.048 1.1840 50.104 17.649
( 1.2160 1.2480 51.151 52.19" 18.269 18.911 1.2800 53.240 19.576 1.3120 54.286 20.266 ,
[ 1.3440 55.333 20.982 1.3760 56.382 21.724
[ 1.4080 57,434 22.494 L- 1.4400 58.493 23.297 1.4720 59.554 24.130 1.5040 60.618 24.992 1.5360 61,683 25.885 h 62.750 26.810 1.5600 1.6000 63.820 27.768 I
I I
I 1
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- ' ' tm l3[
pc -C PAC K (C) COPYFIGHT 1984, 1987
- STPUCTUPAL INTEGRITY ASSOCIATES, INC.
SAN JOSE, CA (408)978-8200 VERSION 1.2 LINEAR ELASTIC FRACTURE MECHANICS EVALUATION 1
VERMONT YANKEE FEEDWATEP C ?ECK VALVE V20E4 CRACK MODEL ELLIPTICAL LONG1TUDINAL CRACK IN CYLINDER (T/R=0.1,A/L=0.5) l 4ALL THICKNESS = 2.0000 STPESS COEFFICIENTS l CASE ID CO C1 C2 C3 THPUWALL 23.5000 O.0000 O.0000 O.0000 ;
FWNRC G.0867 -5.3350 5.8800 0.4010 <
I CR AC K --------------- ST PE SS INTENSITY FACTOR------- --------
DEPTH CASE CASE THRUWALL FWNRC O.0300 4.726 1.602 0.0640 6.695 2.238 0.0960 8.214 2.710 0.1280 9.501 3.095 O.1600 10.641 3.427 0.1920 11.F7R 3.700 0.2240 '2.634 3.985 O.2560 13.528 4.229 0.2880 14.371 4.457 0.3200 15.173 4.672 0.3500 15.939 4.879 0.3840 16.674 5.070 0.4160 17.395 5.273 0.4480 18.075 5.466 I O.4800 0.5120 0.5440 18.744 19.395 20.030 5.657 5.848 6.041 0.5760 20.649 6.235 0.6080 21,253 6.433 0.6400 21.843 6.632 0.6720 22.420 6.837 I O.7040 0.7360 0.7680 22.9B7 23.544 24.091 7.047 7.263 7.486 O.0000 7.717 I 0.8320 0.8640 24.629 25.164 25.691 7.958 8.208 0.8960 26.211 8.468 0.9280 26.724 8.738
.pc-CPACM VEPS104 1.2 " AGE *
- 0.9600 27.2J2 9. ' : 59 l 0.9920 27.733 9.311
' 9.616 1.0240 28.232 1.0560 28.726 9.935 I 1.0880 29.215 10.268 L 1.1200 29.701 10.614 1.1520 'JO.182 10.976 1.1840 30.058 11.352 1.2160 31.133 11.744 1.2480 31.606 12.152 1 1.2800 32.076 12.577
=
1.3120 32.542 13.018 1.3440 33.005 13.478 1.3760 33.466 23.955 1.4080 33.922 14.452 1.4400 34.371 14.968 1.4720 34.818 15.504 1 1.5040 35.263 16.060 1.5360 35.704 1G.638 1.5600 36.144 17.236 l 1.6000 36.581 17.857 END Or pc-CRACK I
J
['
pc-CRACK (C) COPYRIGHT 1984. 1987 STRUCTURAL INTEGRITY ASSOCIATES. INC.
SAN JOSE, CA (408)978-8200 VERSION 1.2
{
LINEAR ELASTIC FRACTURE MECHANICS EVALUATION
[ IRMONT YANKEE FEEDWATER CHECK VALVE V28B
[RACKHODELCIRCUHFERENTIALCRACKINCYLINDER(T/R=0.2)
LLL THICKNESS = 2.0000 STRESS COEFFICIENTS CASE ID CO C1 C2 C3 FWNRC 8.0867 -5.3354 5.8800 0.4018 _
CRACK ---------------:1 TRESS INTENSITY FACTOR----------------
DEPTH CASE FWNRC 0.0320 2.799 0.0640 3.930 0.0960 4.781 1 0.1280 5.487 0.1600 6.100 0.1920 6.649 1 0.2240 7.181 0.2560 7.692 0.2880 8.178 0.3200 8.646 0.3520 9.100 0.3840 9.545 0.4160 10.017 =
1 0.4480 10.523 0.4800 11.032 0.5120 11.545 1 0.5440 12.064 0.5760 12.590 0.6080 13.138 0.6400 13.736 e 1 0.6720 14.349 0.?O40 14.978 0.7360 15.624 0.7680 16.288 0.8000 16.972 17.705 O.8320
. 0.8640 18.460 0.8960 19.239 0.9280 20.042 0.9600 20.871
PAGE 2 7-CKACK , VERSION 1.2 0.9920 21.727 1.0240 22.686 1 .' 88 4. 8
[ 1.1200 25.850 1.1520 26.981 1.1840 28.153 1.2160 29.406
[ 1.2480 30.747 1.2800 32.137 r 1.3120 33.578
( 1.3440 35.073 1.3760 36.623 1.4080 38.251 1.4400 40.002
( 1.4720 41.816 1.5040 43.695 45.642
~
1.5360 1.5680 47.659 1.6000 49.747 l
END OF pc-CRACK I
1 l
l l
l l
l l
1 l
g.... 1. h pc-CRACK (C) COPYRIGHT 1984, 1987 STkUCTURAL INTEGRITY ASSOCIATES. INC.
SAN JOSE. CA (408)978-8200 VERSION 1.2 LINEAR ELASTIC FRACTURE HECRANICS EVALUATION I: RHONT YANKEE FEEDWAT.ER CHECK VALVE V28D CK HODEL LONJITUDINAL CRACF. IN CYLIND.3.R(T/R=0.2) n...LL THICKNE3S= 2.0000 STRESS COEFFICIENTS CASL ID CO C1 C2 C3 FWNRC 8.0867 -5.3354 5.8800 0.4018 I CRACK ---------------STRESS INTENSITY FACTOR----------------
CASE I DEPTH FWNRC 0.0320 2.693 I 0.0640 0.0960 0.1280 3.848 4.763 5.561 I 0.1600 0.1920 0.2240 6.288 6.970 7.651 0.2560 8.324 I 0.2880 0.3200 8.986 9.642 0.3520 10.295 I 0.3840 0.4160 0.4480 10.948 11.630 12.346 0.4800 13.073 0.5120 13.810 0.5440 14.560 0.5760 15.323 I 0.6080 0.6400 0.6720 16.141 17.087 *C 18.060 I 0.7040 0.7360 0.7680 19.058 20.085 21.141 0.8000 22.227 I 0.8320 0.8640 23.523 24.860 0.8960 26.239 I 0.9280 0.9600 27.662 29.131 l
i
FAGE 2 i-C' RACK' VERSION 1.2 0.9920 30.647 I 1.0240 1.0560 1.0E80 32.401 34.278 36.217 1.1200 38.223 I 1.1520 1.1840 40.295 42.438 1.2160 44.778 I 1.2480 1.2800 1.3120 47.329 49.971 52.708 55.543 I 1.3440 1.3760 1.4080 58.480 61.637 1.4400 65.258 1.4720 69.006 1.5040 72.884 1.5360 76.896 I 1.5680 1.6000 81.048 85.343 END OF pc-CRACK I
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'. .. . tm b pc-CPACK I (C ) COPYRIGHT 1984, 1987 L STPUCTUPAL INTEGRITY ASSOCIATES, INC.
SAN JOSE, CA f408)978-82OO r VERSION 1.2 L
LINEAP ELASTIC FRACTUFE MECHANICS EVALUATION VEPMONT YANKEE FEEDWATER CHECK VALVE V288
[CRACKMODELELLIPTICALSUFFACECPACRPLATEUNDERMEMBRANEt4 BENDING STRESSED WALL THICKNESS = 2.0000
[ YIELDS 1 FESS = 30.e000
- PACK ASPECT RATIO (A/L)= 0.2500
( STRESS COEFFICIENTS
~
CASE ID CO C1 FWNPC 8.08G7 0.0000 C R ACK -- -------------S T P E S S INTENSITY FACTOP----------------
( DEPTH CASE FWNPC O 0200 1.854
( -
0.0400 2.623 0.0600 3.214 0.0000 3.713 O.1000 4.154 O.1200 4.3U2 0.1400 4. 9 ) '9 G.1600 5.261 0.1800 5.583 O.2000 5.887 0.2200 6.197 0.2400 6.496 0.2600 6.785 0.2800 7.067 O.3000 7.341 l C.3200 7,609 I O.3400 7.871 O.3600 8.!27 0.3800 8.379 1 0.4000 0.4200 8.627 8.844 O.4400 9.056 0.4600 9.264 I 0.4800 9.467 0.5000 9.667 0.5200 9.862 l
l
pc-CRACK VERSIOH 1.2 PAGE ,
O.5400 10.055 O.5600 10., tv8 0.5800 10.429 0.6000 10.612 r O.6200 10.049 L O.6400 11.004 0.6600 11.319 0.6000 11.553
[
0.7000 O.7200 11.707 12.019
, 0.7400 12.252
~
0.7600 12.484 0.7000 12.715 O.0000 12.947 I 0.8200 13.199 0.0400 13.452 0.8600 13.705 1 O.0000 13.959 0.9000 14.212 O.9200 14.467 i 0.9400 0.9600 14.721 14.976 0.9000 15.202 1.0000 15.487 wo 0, --
I I
I I
- ' tm pc -CP AC K i (C) COPYRIGHT 1904, 1987 5 T PUC TURAL INTEGRITY ASSOCIATEfi, INC.
SAN JOSE, CA 84081S73-8200 VEFSION 1.2 u
LINEAR ELASTIC FRACTURE MECHANICS EVALUATION r
VERMONT YANLEE FEEDWATEP CHECK VALVE V28B CRACK MODEL CENTER CRACK PLATE UNDER REMOTE TENSION STFESS AALF PLATE W1DTH= 10.0000 STFESS COEFFICIENTS CASE ID CO C1 FWNRC 8.0867 1 THRUWALL 23.5000 C R AC K ---------------S TFE SS INTENSITY FACTOR----------------
DEPTH CASE CASE FWNPC THRUWALL O.0400 2.867 8.331 0.0800 4.054 11.702 O.1200 4.966 14.430 0.1600 5.734 16.664 I 0.2000 6.412 18.632 0.2400 7.024 20.413 O.2800 7.588 22.051 I 0.3200 8.113 23.577 0.3600 8.607 25.011 O.4000 9.074 26.368 I 0.4400 9.519 9.944 27.661 28.897 0.4800 O.5200 10.352 30.084 0.5600 10.746 31.228 I 0.6000 11.126 32.333 0.6400 11.495 33.403 0.6000 11.852 34.442 lp O.7200 12.200 35.452 0.7600 12.538 36.437 g 0.0000 12.869 37.397 0.8400 13.192 38.336 l 0.8800 13.508 39.254 0.9200 13.017 40.153 0.9600 14.121 41.035 1.0000 14.419 41.901 1.0400 14.712 42.752 1.0800 14.999 43.588 I
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pc-CPAN VERSION 1.2 PAGE J 1 1200 15.283 44.412 '
1.1600 15.B62 45.223
- 1.2000 15.837 46.022 ,
I 1.2400 1.2800 16.100 16.376 46.810 47.500 40.356 1.3200 16.640 16.901 49.115 1.3600 1.4000 17.159 49.865 1.4400 17.415 50.607 I 1.4000 1.5000 1.5600 17.667 17.910 10.165 51.342 52.069 52.709 1.6000 10.411 53.502 1.6400 10.654 54.210 1.6900 18.896 54.911 1.7000 19.135 55.607 I 1.7600 1.0000 1.0400 19.3'73 19.609 19.843 56.298 56.904 57.665 20.076 58.342 I 1.8800 1.9200 1.9600 20.308 20.538 59.014 59.603 2.0000 20.767 60.348 END OF pc-CPACK I
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- Typical KIC Valve for A216 WCB at Minimum Valveyrature (T = 80 F for RCIC Injection Mode)
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(C) COPYRIGHT 1984, 1987 STRUCTURAL INTEGRITY ASSOCIATES, INC.
SAN JOSE, CA (408)978-8200 L VERSION 1.2 FATIGUE CRACK GROWTH ANALYSIS f~
L ERMONT YANXEE FEEDWATER CHECK VALVE V28B INITIAL CRACK SIZE = 1.5000
%LL THICKNESS = 2.0000 AX PRACK SIZE FOR FCG= 1.6000 FATIGUE CRACK GROWTH LAW (S)
I ASME SECTION XI BILINEAR LAW 3 FOR WATER ENVIRONMENT
= Kmin / Kmax FR< 0.25 THEN R' = 0.25 F R > 0.65 THEN R' = 0.65 ELSE R' = R QL = 26.9
- R' - 5.725 IATRAN QU = 3.75
- R' + 0.06
= (D
- QU / QL)*0.25 dx = Kmax - Kmin F dK < KTRAN THEN da/dN = CL
- QL
- dK"5.95 F dK > KTRAN THEN d&/dN = CU
- QU = dK*1.95 HERE:
CL = 1.020000E-12 CU = 1. 010 001E- 0 7 D = 9.902034E+04 IREFORTHECURRENTLYASSUMEDUNITSOF:
FORCE: kips LENGTH: inches I STRESS COEFFICIENTS l CASE ID CO C1 C2 C3 FWNRC 8.0867 -5.3354 5.8800 0.4018 JMBER OF CYCLE BLOCKS = 18 PRINT INCREMENT OF CYCLE BLOCK = 1.0 NUMBER OF CALCULATION PRINT FCG SUBBLOCK CYCLES INCREMENT INCREMENT LAW ID 1 1.0 1.0 1.0 SECT XI LAW I
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i ::-CRMK VERS 70!! 1 2 PAGE 2 m Km3x Dnin SUBBLOCK CASE ID SCALE FACTOR CASE ID fi.: Alt FACTOR
. 1 FWNRC 3 0000 FWNRC C.0000 CRACK HODEL: ELLIPTICAL LONGI'IUDINAL CRACK IN CYLINDER (T/R=0.1, A, L=0. 5)
CRAC K - ---- - -- ---- --- S TR ES S IllT ENS I TY FACTOR ~ - --- --- - --- ---
Dr.PTH CASE FWNRC 0.0320 1.602 0.0640 2.238 0.0960 2.710 I 0.1280 0.1600 3.095 3.427 0.1920 3.720 1 I 0.2240 0.2560 O.2880 3.985 4.229 4.457
)
l 0.3200 4.072 I 0.3520 0.3840 4.078 5.078 0.4160 5.273 I 0.4480 0.4800 0.5120 5.466 5.657 5.848 0.5440 6.041 0.5760 6.235 0.6080 6.432 0.6400 6.632 0.6720 6.837 0.7040 7.047 0.7360 7.263 I 0.7680 0.8000 0.8320 7.486 7.717 7.958 0.8640 8.208 0.8960 8.468 0.9280 8.738 0.9600 9.018 0.9920 9.310 1.0240 9.616 1.0560 9.935 I 1.0880 1.1200 1.1520 10.267 10.614 10.975 1.1840 11.352 1.2160 11.744 1.2480 12.151 1.2800 12.576 1.3120 13.018 1.3440 13.477 1.3760 13.955 1.4080 14.451
pc-CRACK VERSION 1.2 PAGE 3 1.4400 14.967 1.4720 15.503 1.5040 16.060 1.5360 16.637 r 1.5680 17.236 L 1.6000 17.856 f
L
~ TOTAL SUBBLOCK CYCLE CYCLE KMAX KHIN DELTAK R DADN DA A A/T LOCK 1 1.0 10 47.97 0.00 47.97 0.00 1.9E-04 0.0002 1.5002 0.75 I, LOCK 2 2.0 1.0 47.98 0.00 47.98 0.00 1.9E-04 0.0002 1.5004 0.75 BLOCK 3 3.0 1.0 47.99 0.00 47.99 0.00 1.9E-04 0.0002 1.5006 0.75 pLOCK 4 4.0 1.0 48.00 0.00 48.00 0.00 1.9E-04 0.0002 1.5006 0.75 l
J 5
[ LOCK 1.5010 C.75 I 5.0 1.0 48.01 0.00 48.01 0.00 1.9E-04 0.0 02 LOCK 6 6.0 1.0 48.02 0.00 48.02 0.00 1.9E-04 0.0002 1.5011 0.75 LOCK 7 7.0 1.0 48.03 0.00 48.03 0.00 1.9E-04 0.0002 1.5013 0.75 I
uLOCK 8 8.0 1.0 48.04 0.00 48.04 0.00 1.9E-04 0.0002 1.5015 0.75 BLOCK 9 9.0 1.0 48.05 0.00 48.05 0.00 1.9E-04 0.0002 1.5017 0.75 LOCK 10 .
10.0 1.0 48.06 0.00 48.06 0.00 1.9E-04 0.0002 1.5019 0.75
-CRACK VERSION 1.2 PAGE 4 ,
mILOCK 11 11.0 1.0 48.07 0.00 48.07 0.00 1.9E-04 0.0002 1.5021 0.75 I JLOCK 12 12.0 1.0 48.08 0.00 48.08 0.00 1.9E-04 0.0002 1.5023 0.75 BLOCK 13 13.0 1.0 48.09 0.00 48.09 0.00 1.9E-O' O.0002 1.5025 0.75 GLOCK 14 14.0 1.0 48.10 0.00 48.10 0.00 1.9E-04 0.0002 1.5027 0.75 15 17 LOCK 15.0 1.0 48.11 0.00 48.11 0.00 1.9E-04 0.0002 1.5029 0.75 ILOCK 16 16.0 1.0 48.12 0.00 48.12 0.00 1.9E-04 0.0002 1.5031 0.75 I 4 LOCK 17.0 17 1.0 48.13 0.00 48.13 0.00 1.9E-04 0.0002 1.5033 0.75 I clLOCK 18 18.0 .0 48.14 0.00 48.14 0.00 1.9E-04 0.0002 1.5035 0.75 END OF pc-CRACK I
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I g mmmC FRACTURE TOUGHNESS DATA FOR I A216 WCB CAST MATERIAL I
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NUREG/CR-3009 SAND 78-2347
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[ Fracture Toughness of L PWR Components Supports c
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Prepared by G. A. Knorovski, R. D. Krieg, G. C. Allen, Jr.
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Sandia National Laboratorios
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i Teble 3.2 Classification of Wrought Grades into Groups Plain carbon: A-7, A-53, A-106, A-201, A-212, A-28 3, A-28 4 ' I ~
A-285, A-306, A-307, A-501, A-515 P
Carbon-manganese A- 36 , A- 10 5 , A- 516 , A- 5 37 liigh-strength low-alloyt A- 4 41, A- 57 2, A- 5 8 8 , A-618 Low alloy (not quenched & tempered): A-302, A-322, A-353, A-387 5 Quenched & tempered: A-193, A-194, A-325, A-354, A-461, A-490, A-508, A-514, A-517, A-533, A-537, A-540, A-543, A-563, A-574.-
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Table 4.4 computation of ftDT Results 3
o t4DT + 1.30 11DT + 2a l Matacial 14DT Cast Steels 12*F 10'F 18'?
A-27, 3-216 1" - 6'F 35 17 57 69 +
(heat treated > 1"
__] condition) max. -20 7.-352 Wrought Steels 31 67 89 all " mild" steels
- 27 all " mild" steels 77 96
- except A-201 40 28 13 39 48 W C-Mn*(as-hot rolled) 22 8
-28 18 - 5 g (normalized) 25** 12** 41** 49**
HSLA* (as-bot rolled) 18** -27** -14**
(normalized) -50**
I low alloy non OLT A-302 A-353 8 28 45 64 max. -320 65**
A-387 Quenched & Tempered t max. 40'F I A-508 C12 A-514 A-517 max. -10'F max. -20'F max. 20*F A-533B C11 I, A-537 C12 max. -60'F max. -60*F A-543
- See table 3.2 for ASTM specs included in this category
- See discussion in Appendix B 4.4.3 Fracture Toughness Minimum values for fracture toughness of the material groups are indicated in Table 4.5. These are usually dynamic values or static values obtained at lower temperatures equivalenced via the Barsom temperature shift (see section 4.2). Data at the reference temera-ture, 75'F, was not always obtainable. If data was not obtainable, I,
m k --___ __
APPENDIX B - MATERIAL DATA B.1 Data Obtained The sources of material data for the various groups are listed in Tables B.1 through B.7. Included in these tables are data sour-ces which were not used in the body of the report. The actual data (NDT and K-type) have been plotted in Figs. B.1 through B.25. Tab-ulation of NDT data and standard deviations (where possible) are (I indicated in Table 4.4.
NDT data for several grades of steel were not located. Assign-ment into susceptibility groups for these materials were based on the minimum requirements of the appropriat<< standards under which the materials were procured (see Appendix C), as compared to materials for which data were obtained.
o B.2 Cast Steels Four grades of cast steels were listed in the utility cubmit-tais (not counting a stainless steel casting for Yankee, considered not to have a problem with respect to fracture toughness or lamellar taating). Two of the grades, A-27 Gr 70-40 and A-216 Gr WCB are carbon manganese-silicon types one, A-148 (Gr 80-40 and Gr 80-50) is not chemically specified (which indicates it may be either C-Mn or low-alloy depending upon the heat treatment and/or section size) and the last, A-352 Gr LC3, is a high ( 3-4%) nickel content heat-treated alloy requiring CVN testing. (Notes all % are by weight) l The A-352 Gr LC3 grade in either the double normalized and tempered, or quenched and tempered condition is expected to show excellent fracture toughness with NDT's in the range of -100*F for I 114 i
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1" section size (Fig. B.1). Some utility dat.1 (Ref. B-1) indicated thick section 14DT's in the -100 to -60'F range with a maximum value (one example) of -20'F.
A-27 Gr 70-40 and A-116 Gr WCB are both C-Mn-Si type alloys I varying only slightly in chemical composition allowables, and pri-marily in minimum yield strength (40 vs 36 kai, respectively). Of the two, the A-27 Gr 70-^0 allows less carbon (.25% vs .30%) but more nianganese (1.2% vs 1.0%) . A-216 Gr WCC is virtually identical to A-27 Gr 70-40 in this respect. A histogram of NDT values for A-27 Gr 70-40 heats mainly in the normalized and tempered condition 1
(five were normalized and four were quenched and tempered) plus five heats of A-216 Gr WCB is showr* in Fig. B.2. This is taken from a compilation made by the Steel Founder's Society of America (Ref.
B-2). The statistics of these data imply that 95% of all heats have NDT's below 20
- F. However, these data are taken from 1" thick test castings, and a section size effect may be expected. A second source of data (Ref. B-3) for these materials indicated that NDT was 35*P with a standard deviation (c) of 17'F for 12 specimens of varying i thickness (from 2-1/2" to 5") poured from two heats in the normalized This still indicates that 95% have their and tempered condition.
! NDT below 70*F, but not with an much margin as the 1 in. thickness I case. Finally, these two specifications allow the possibility of I
producing heats in the annealed condition, if the mechanical proper-ties can be me t. This would be expected to further degrade their fracture _ toughness properties since a coarser microstructure,would
, . result. This implies the only way to meet strength requirements would be by increasing carbon content.
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Finally, A-148 Gr 80-40 and Gr 80 50 (40 and 50 kai yield otrength, respectively) are more difficult to evaluate, since chemical specifications and data are lacking. The added strength rcquirements over A27 Gr 70-40 could be met in a number of ways via heat treatment, via additional carbon content, or via alloy c:ntent. Since additional carbon is usually the least expensive rcute, the implication is that these sub-grades of A-148 would have loss desirable NDT values than the previously discussed A-27 and A-216. However, A-148 was specified by only one plant and was part l of a wire rope system, which is probably not as critical a location on the other cast grades, which were typically in the sliding pedes-tal category of plants, In Fig. B.1 some NDT data (Ref. B-4) is available for normalized and tempered A-148 Gr 80-50 which indicate oxcclient HDT's around -10F; however, these heats contained approx-imately 2% Ni. Thus these data would be indicative of the best practices in meeting the mechanical property requirements.
K data were located for two heats of A-216 Gr WCC (Refs D-5, Ic B-6). These are shown in Figs. B.3. Applying a temperature shift of about 150'F, equivalent kid values at 75'F are roughly 40 kai /I'h.
These specimens were taken from immense (20"x20"x48")* castings, and C probably represent the worst possible section size effect.
I D.3 Weld Consumables I The weld metals are also in the cast steel category. It is i difficult to evaluate weld metal properties separately from the base g materials being $oined, since dilution ef fects can occur which signi-
'- ficantly change the chemical composition of the fused metal. Further-I 116 I
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r FIG. B.2 NOT FOR CAST GRADES (NDT)FOR-A-27 IS -7 F o 13 *F A-27 GR 70! 40 A-216 GR WCB F
NORMAllZED K
_ NORMAllZED & TEMPERED x 7
, C 15 -
P QUENCHED & TEMPERED E Q -
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-FRACTURE TOUGHNESS VERSUS TEST TEMPERATURE I
FIGURE B.3(a ) A-216 K lC DATA 152
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I A,, - 1x o g ENIIANCED LEAKAGE MONITORING ,ROGRAM FOR FEEDWATER CHECK VALVE V28B
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T .dp APPENDIX D ENHANCED LI1KAGE MCNI'IORING PROGRAM IVR FEEDWATER CHECK VALVE V28B To augment Vermont Yankee's existing Prinary Containment Leakage Monitoring System, a Local Leak Detection System has been installed on the FDW-28B valve. This additional system is a Techmark Leak Detection System to provide constant leakage monitoring during the next operating cycle. The system consists of three moisture sensitive tape (MST) transducers mounted on the mirror insulation below the valve (Y28B). To install the transducers, a " hole I was drilled through the insulation for the transducer sensor tube to be inserted. The sensor tube provides a path for moisture from under the insulation to contact the MST. The transducers have the ability to detect leakage as low as 0.1 gpn. 'Ihe transducers provide a multiplex signal to an indicator / control unit (TUM 700) nounted in the Reactor Building. The control unit interrogates
$11 sensors once per second and provides a digital display of sensor I location (s) for alarm or trouble conditions. The unit also provides remote alarm indication in the main control room.
The tocal Leak Detection System will be utilized by operations personne4 to initiate further achinistrative actions / controls which have been h eeleped as part of this enhanced plan. These achinistrative controls identify, in part, operator action upon receipt of an I a2crm on the MST unit, compensatory action if the unit experiences trouble as well as establishing additional leakage rate criteria h> %' th6t centained in Technical Specifications.
As ai t~1 above, the Local Leak Detection System is intended to augment the existing systems and provide acklitional assurance that any leakage from the FDW-28B valve will not go undetected. This system will prem.ie operators with an "early warning system" to initiate addccional measures of thie augmented program.
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