ML20155C282
| ML20155C282 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 10/30/1998 |
| From: | Caine T, Mehta H, Van Diemen S GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20155C271 | List: |
| References | |
| GENE-187-30-059, GENE-187-30-0598-NP, GENE-187-30-59, GENE-187-30-598-NP, NUDOCS 9811020174 | |
| Download: ML20155C282 (29) | |
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Attachment 1 ta JPN-98 xxx General Electric Nuclear Energy Report GENE-187-30-0598 NP, B13-01920-30, Revision 0, "CRD Bolting Flaw Evaluation for James A. FitzPatrick Nuclear Power Plant",
t October 1998 i (non-proprietary) l l
l New York Power Authority JAMES A. FITZPATRICK NUCLEAR POWER PLANT Docket No. 50-333 DPR-59 c
9811020174 981023 PDR ADOCK 05000333 P PDR
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. h GE Nuclear Energy TECHNICAL SERVICES BUSINESS GENE-187-30-0598 NP GE Nuclear Energy Revision 0 175 Curtner Avenue, San Jose, CA 95125 DRF # B13-01920-30 Class I l October 1998 ;
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1 CRD Bolting Flaw Evaluation for l
James A. FitzPatrick Nuclear Power Plant i
l October 1998 1
Prepared for New York Power Authority Prepared by GE Nuclear Energy San Jose, California NEW YORK POWER AUTHORITY DOCUMENT RE\TN STATUS STATUSW/
s y uxamD 2 O ACCEPTED AS NoTED RESUSWrTAL NOT REQUMED 3 O ACCEPTED AS NoTED RESUBMTTAL REQUMED 4"O NoT AoCEPTED Post.alon to poceed does not conseus accepance w appromul of tMgn doissa, cahnssons, analyss, inn memods w rnanwess onwaped w seiscw by me sesor and does not reasse supper tem u compience wm "W n _
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CE Nucleer Energy GENE 187 304598 NP a
E CRD Bolting Flaw Evaluation for James A. FitzPatrick Nuclear Power Plant October 1998 1
Prepared by: i vtpv ,
Sylvilkan Diemen, Project Integrator !
Structural Mechanics & Materials ,
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Verified by: . b
" H.S. Mehta, Principal Engineer Stmetural Mechanics & Materials Approved by: [
T.A. Caine, Manager Structural Mechanics & Materials
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GE Nuclect EnerU GENE-187-30-0598 NP I
l IMPORTANTNOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully ~
The only undertakings of the General Electric company (GE) respecting information in this document are contained in the contract between New York Power Authority and GE, C95-20013, CO#47, Task #49, effective 2N/98, as amended to the date of transmittal of this l document, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than New York Power Authority, or for any purpose other than that for which it is intended is not authorized: and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy or usefulness of the information l contained in this document, or that its use may not infringe privately owned rights.
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".', - l' GE Nuclear Energy GENE.187 30-0598 NP TABLE OF CONTENTS Pace
- 1. INTRODUCTION & B AC KG ROU N D ........................................................... 1 -1
- 2. REVIEW OF METALLURGICAL EVALUATIONS ....................................... 2-1 2.1 Plant A . _ - . . . 2-1 2.2 Plant B .. _
. . . 2-1 2.3 Plant C - ._ .. .
- _ 2-2 2.4 Plant D -
.. . . _ _ . .. 2-2 2.5 Plant E ......._..-.__.........22 2.6 Plani F- _ .
... . . 2-3 2.7 Plant G _ _ - _ . . _ _ _ . .. 2-3 2-4
2.8 JAFNP
2.9 Conclusions Fmm Review of Metallurgical Enminations _. - -
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- 3. ALLOWABLE FLAW DEPTH EVALUATION .............................................. 3 1 3.1 Fracture Mechaales Assessment : - _ .. 3-1 l
3.2 Required Area to Meet Section III ASME Code Criteria .. . . 3-2 3.3 Allowable Maw Depth Based on Section XI Acceptance Standards _ - .- 3-3 3.4 Allowable Maw Depth Summary. _ . 3-3 3 9 Comparison with Observed Crack Depths. -_ . 3-3 1
- 4. ASME SECTION XI EXAMINATION CONSIDERATIONS........................... 4-1
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! 4.1 IWB-2500-1 Examluations __ .
.. 1 4.2 Code Case N-547-- .-
4.3 Inspection, Evaluation / Replacement Practice During CRD Housing Refurbishment ._4-2 i
- 5. S U M M ARY & CON CLU StON S ................................. ................................. 5-1
- 6. R E F E R E N C E S . . . . ... . . .. . . . . ... . . . . . . . . ... ... . . ... . .. .. . . . . . . . . . . . . . . .. .. ... . ....
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. l List of Floures I
1-2 l Figure 1 CRD Cap serew Geometry -
Figure 2-1: Indications os CRD cap screw from Plant B; crack is approximately 0.045long. .?5 l i
Figure 2-2: Cracking on CRD cap screw from Plant B; appearance is similar to corrosion pits. 2-5 Fipre 2-3: 250X view of crack from Figure 2-1; note blunted tip. _ -24 Figure 2-4: Crack from Figure 2-2; crack has appearance of corrosion pit. . 2-7 Figure 2-5: Head-to shank transition region #1 (Plant C) -
.. 2-8 Figure 24: Head-to shank transition region #2 (Plant C) . ._ 2-8 l 2-9 Figure 2-7: Cracking from Figure 2 .
2-9 Figure 2-8: Cracking fmm Fipre 24 .
Figure 2-9: Visual indications, Plant D _ _ _ 2-10 2-10 Figure 2-10: Visual indications on second cap screw from PIant D .
2-11 Figure 2-11: Cracking from Fipet 2 --
2-11 Figure 2-12: Cracking from Figure 2-10 :
List of Tables Table 3-1 Measured t.narpy Energies of CRD Cap screwing _ -
5 Table 3-2 Measured Crack Depths in CRD Cap screws: - - . _ . : - 34 t
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CE Nuclecr Energy GENE 187-30-0598 NP
- 1. INTRODUCTION & BACKGROUND During routine control rod drive (CRD) maintenance at several BWRs, visual examination of cap screws that connect the CRD to the housing flange revealed circumferential cracking and corrosion pitting in the shank directly below the cap screw head. The cracks were first discovered at a BWR plant in May 1988 and reported in GE RICSIL No. 019
[1]. SIL No. 483, Rev. 2 [2] updated the information on the evaluation of CRD cap screw
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indications discovered at'other GE BWRs and recommended corrective actions.
The original CRD cap screws at J.A. FitzPatrick Nuclear Plant (JAFNP) were 1-inch '
nominal diameter and were made of AISI 4140 which meets the requirements of SA193 Grade B7 of the ASME Code. Figure 1-1 shows the geometry of a CRD cap screw of original design. New York Power Authority (NYPA), the operator of the JAFNP, has been replacing the CRD cap screws in accordance with SIL 483, as the cap screws are removed during normal maintenance, since 1991. The replacement cap screws are of the new design as recommended in Reference 2.
The metallurgical evaluation of cracked cap screws from several BWR plants conducted by GE has shown that the observed cracking, mainly near the head to shank fillet, is very shallow and the crack tips have typically blunted appearance indicating insignificant crack growth expected during future operation. Also, the structural and fracture mechanics evaluations show significant margin.
The procedures in the ASME Code Section XI of record at JAFNP [3c] require a sample expansion if the indications in the removed cap screws exceed the acceptance standards of Table IWB-3517. The metallurgical and structural margin evaluations have shown that these requirements represent unnecessary hardship for the plant without significant added safety benefit.
l l The objective of this report is to develop the technical justification for not expanding the sample size when cap screws with cracks exceed the acceptance standards of Section XI.
The justification is based upon review of metallurgical evaluations from several plants and a structural margin evaluation.
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Figure 1-1 CRD Cap screw Geometry (from GE Dwg. I17C4515P2, Rey,6 [5])
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GE Nuclear Energy GENE 187-30-0598 NP
- 2. REVIEW OF METALLURGICAL EVALUATIONS To characterize the observed indications on CRD cap screws, metallurgical evaluations <
have been performed by several plants. The following paragraphs describe the results from eight (8) plants. These evaluations were performed by both GE and other laboratories. _ _ . .
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2.1 Plant A l
2.2 Plant B 2-1
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', - GE Nuclecr Energy GENE 187 30-0598 NP 2.8 JAFNP During visual examination of CRD cap screws, only six cap screws (out of 306 examined) were identified as having indications in excess of the ASME Section XI criteria. Two of the six cap screws were then sent out for metallurgical examination to characterize the !
flaws. These cap screws were selected based upon the longest linear indications and/or l I
severest general corrosion.
The observed cracks were wide with blunted tips, similar to the cracking observed in all the other plants. The cracks were also filled with oxide, which indicated a corrosion mechanism was operating. The tempered martensite microstructure was consistent with the material for the bolts. The maximum observed crack depth was 0.036".
The conclusion of the evaluation was that stresses may have had some effect on the l cracking, but that corrosion was the dominant mechanism. Again, this is a different l interpretation of the mechanism; however, the depth and crack morphology are similar to j the otherinvestigations.
2.9 Conclusions From Review of Metallurgical Examinations l
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GE Nuclect Energy GENE 187 304598 NP Figure 2-1: Indications on CRD cap screw from Plant B; crack is approximately 0.045"long.
Figure 2-2: Cracking on CRD cap screw from Plant B; appearance is similar to corrosion pits.
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GENE 187-304598 NP GE Nuclear Energy l
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t Figure 2-3: 250X view of crack from Figure 2-1; note blunted tip.
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Figure 2-4: Crack from Figure 2-2; crack has appearance of corrosion pit.
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Figure 2-5: Head-to shank transition region #1 (Plant C) )
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Figure 2-6: Head-to shank transition region #2 (Plant C) 2-8 t
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Figure 2-8: Cracking from Figure 2-6 1
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Figure 2-9: Visual indications, Plant D l
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i Figure 2-10: Visualindications on second cap screw from Plant D 2-10
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i Figure 2-11: Cracking from Figure 2-9 1
I Figure 2-12: Cracking from Figure.2-10 l.
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.I GE NucleCr Energy GENE-187-304598 NP
- 3. ALLOWABLE FLAW DEPTH EVALUATION I
The allowable crack depth is governed by two different criteria:
(i) Fracture mechanics assessment (ii) Needed area to meet Section IIIlimits 3.1 Fracture Mechanics Assessment A key material property in the fracture mechanics evaluation is the Ke value of the cap screw material. While the Km values for the cap screws were not directly measured, a l good estimate of it can be made from the measured Charpy energies reported in the l certified material test reports (CMTRs) of a large batch of the cap screws.
The following relationship called the Rolfe-Novak-Barsom correlation [4], was used:
(Km/Sy )2 = 5 [(CVN/Sy )-0.05] (1) where Ko= Critical plane-strain stress intensity factor at slow loading rates, i ksifm.
S=
y 0.2% offset yield strength at the upper shelf temperature, ksi.
CVN = Standard Charpy V-notch impact test value at the upper shelf, ft-Ibs.
The CRD cap screws in use at the FitzPatrick station were ordered per the Reference 5 specifications. Three Charpy tests were conducted on each batch consisting of up to 2000 cap screws. A total of 13 batches were involved. GE CMTR records were searched to obtain these values and are given in Table 3-1. In addition, GE searched its records of Charpy values from the tests conducted on cap screws received for metallurgical evaluation from one of the BWR plant. The last entry in Table 3-1 which pertains to CVN value from FitzPatrick station, was obtained from Reference 8. All of the Charpy values are based on tests conducted at 40'F, except the FitzPatrick CVN test which was conducted at 10 F, and were determined to be 100% shear fractures indicating upper shelf conditions.
Results of the statistical evaluation of the Charpy data in Table 3-1 are shown at the bottom of that table. The mean Charpy value was calculated as 88.6 fi-lbs and the standard deviation was calculated as 9.2 ft-lbs. This would indicate a mean minus two sigma value of(88.6-2x9.2) or 70.2 ft-lbs. A mean minus tw'o sigma value approximately represents a confidence level in excess of 97% that the Charpy value of any cap screw will.
not be below 70.1 ft-lbs. The lowest value noted in Table 3-1 is 68 ft-lbs.
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. The cap screw material meets the requirements of ASME Code bolting material SA193, l B7 for which the specified minimum yield strength is 105 ksi. Based on this value of yield I strength and the assumed minimum value of 68 fi-lbs for the Charpy energy, Equation (1) predicts a Ke value of181.5 ksiVin.
The fracture mechanics based IWB-3600 evaluations in ASME Section XI [3] typically assume safety factor values of 410 or 3,16 for normal (Level A), upset (Level B) and test conditions and 42 or 1.44 for emergency (Level C) and faulted (Level D) conditions. -
Thus, the allowable value of Ko for normal / upset / test conditions is (181.5/3.16) or 57.4 j ksiVin. i The applied values of stress intensity factor, K, were calculated using the 360 circumferential flaw solution from Tada and Paris flaw handbook [6]. The K, is given by:
K, = o {V(xc)} { l.122 - 1.302(c/b) + 0.988(c/b)2 - 0.308(c/b)'}/{(1-c/b)"}
(2) where, c= Applied stress based on nominal un-cracked cross-section c= Circumferential crack depth b= radius ofcap screw The applied stress a was calculated as 54.9 ksi [7]. This value includes the stress due to preload, thermal differential expansion and the shear stress resulting from friction. A tightening torque of 375 ft-lbs. was assumed in calculating the applied stress. A trial-and-error solution gave an allowable circumferential crack depth of 0.15 inch (i.e., the K, is equal to the allowable value of 57.4 ksiVin at c = 0.15 inch).
Based on the preceding the allowable value of crack depth for the cap screws is 0.15 inch from fracture mechanics considerations.
3.2 Required Area to Meet Section ill ASME Code Criteria Reference 7 describes the analysis of the CRD bolted joint using typical ASME Code,Section III, Subsection NB, procedures. The minimum cap screw cross-section area to meet the Code requirements was calculated as 1.61 in'. This is equivalent to the sum of cross-sectional areas of three cap screws. In other words, only three uncracked cap screws out of eight cap screws present at a CRD flange, are sufficient to meet the Code structural margins.
l Assuming that all cap screws at a CRD flange experience c' racking, the minimum cross-section area needed for each cap screw is (1.61/8) or 0.2025 in' or approximately a core area of 0.25 inch radius. The radius of the cap screws in the shank rer, ion where cracking 3-2
. . .- . _. . . . ~ - _ - - - . _ . _ _ . - _ - . . _ _ - - _ . _
GE Nuclect Energy GENE 187-30-0598 NP has been observed is 0.41 inch. Therefore, uniform fully circumferential cracks of(0.41- '
. 0.25) or 0.16 inch depth in all eight cap screws at a CRD flange are acceptable while still maintaining the ASME Code required structural margins.
3.3 Allowable Flaw Depth Based on Section XI Acceptance Standards l l
Table IWB-3515-1 of ASME Section XI provides the acceptance standards for indications that may be detected during the volumetric examinations of cap screws greater than 2-inch nominal size. When a fully circumferentialindication (aspect ratio approx. Equal to zero) is postulated, the allowable depth in Table IWB-3515-1 is 0.075 inch. The Code does not provide any guidance for cap screws less than 2-inch diameter. For the CRD cap screw which is 1-inch diameter (nominal size), a linear-clastic fracture mechanics (LEFM) analysis was conducted in Reference 8 and an equivalent allowable depth of 0.071 inch was extrapolated based on the values given in Table IWB-3515-1 for botting greater than 2-inch in size.
3.4 Allowable Flaw Depth Summary The preceding fracture mechanics and Code required area calculations indicate that CRD flange integrity is assured with required ASME Code margins even with 360*
i circumferential cracks of up to 0.15 inch depth in all cap screws. The acceptance standards which require no analysis, allow a crack depth of 0.071 inch. Thus, the maximum allowable flaw depth is 0.15 inch.
3.5 Comparison with Observed Crack Depths A comparison of the allowable flaw depth of 0.15 inch with the observed crack depths in CRD cap screwsis presented next.
4 GE Nuclear Energy has conducted several metallurgical analyses of the cracked cap screws as described in Section 2. The measured crack depths obtained in such analyses were reviewed and are summarized in Table 3-2. The crack depths for the FitzPatrick plant CRD cap screws were obtained from Reference 8. As noted at the bottom of Table 3-2, the mean value of the measured crack depths was 0.025 inch with a standard deviation of 0.015 inch. The deepest crack depth reported was 0.065 inch. The largest crack depth for FitzPatrick plant cap screws reported in Reference 8 is 0.036 inch. If one were to calculate the mean plus three standard deviation (99% probability) value, it would
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be (0.025 + 0.015x3) or 0.070 inch.
The deepest observed crack depth measured in cracked CRD cap screws removed from service at JAFNPP is 0.036 inch.
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Based on the preceding comparison, it can be concluded that the observed crack depths in l
- CRD cap screws removed from service are much less than the allowable crack depth that l still maintains the structural margins consistent with the ASME Section XI requirements.
The signi6cance of this conclusion in terms of the ASME Section XI inspection program for the CRD cap screws is discussed in the next section. ,
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..:. GE Nuclear Energy GENE-187-30-0598 NP Table 3-1 Measured Charpy Energies of CRD Cap screws 35
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l Table 3-2 Measured Crack Depths in CRD Cap screws i
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- 4. ASME SECTION XI EXAMINATION CONSIDERATIONS There are two ways in which the examination of CRD cap screws can be triggered. One is when the CRD housing is disassembled for some reason such as refurbishment.
Examination category B-G-2, Item No. B7.80 in Table IWB-2500-1 specifies a visual VT-1 examination of cap screws, studs, and nuts when the CRD housing is disassembled. The acceptance standard is specified in IWB-3517. The other way is if a CRD exhibits leakage '
during the system pressure test (IWA-5250). The need for sample expansion in the first case is the subject of discussion in this section.
4.1 IWB-25001 Examinations If the indications exceed the acceptance standards, additional examinations are required.
Paragraph IWB-2430(a) states, " Examinations performed in accordance with Table IWB-2500-1 that reveal indications exceeding the acceptance standards of Table IWB-3410-1 shall be extended to include additional examinations at this outage. . ..."
The interpretation from the preceding paragraph is that additional examinations in Category B-G-2 may be required following the evaluation results of examinations performed in accordance with Table IWB-2500-1.
It is surmised that the intent of the Code in requiring the sample expansion was to assure that the extent and the nature of cracking exceeding the acceptance standards observed in a category during the current examination scope is not found elsewhere or more severe in the remaining components of the same category. Metallurgical examination results of cracked CRD cap screws from nine plants reviewed in Section 2 showed that this type of cracking is shallow and well understood. Furthermore, the fracture mechanics and Code vea requirements calculations presented in Section 3 showed that there is a significant structural margin with respect to any likely depth of cracking that may be present. These j
results provide a strong technical justification for eliminating the need for sample expansion.
In this connection, a recent Code Case N-547 [9] is relevant and is discussed next.
4.2 Code Case N-547 This Code Case pertains to alternative examination requirements for pressure retaining bolting of CRD housings: .
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i j ' l- GE Nucle:r Energy GENE 187-30-0598 NP huluiry: What alternative to the requirements of Table IWB-2500-1, Category B. l
, , G-2 may be usedfor VT-1 visual examination of CRD housing bolts, studs, arr!
nuts? .
Reply: It is the opinion of the Committee that the VT-1 visual examination of j CRD housing bolts, studs, arulnuts is not required This Code Case, which is applicable from 1980 Edition with the Winter 1980 Addenda to and including 1995 Edition, erminates the need for the VT-1 examinations of the CRD cap screws. By logical ext: cion, it also eliminates the need for sample expansion j triggered by the requirements of Paragraph IWB-2430(a). This Code Case was incorporated into the Code in the 1995 Addenda.
4.3 Inspection, Evaluation / Replacement Practice During CRD Housing Refurbishment The current inspection, evaluation / replacement practice during the CRD housing refurbistunent at JAFNP consists of replacing the cap screws with new design cap screws as suggested in Reference 2. If a previously installed cap screw is reused, VT-1 and augmented surface examinations prior to installation are performed per the GE SIL [2].
Given the metallurgical evaluations reported in Section 2, and stmetural margins shown and typical indication depths observed as discussed in Section 3, this practice is technically sound and meets the intent of the Section XI requirements for verification of acceptability for replacements (IWA-7220 for 1980 Edition or IWA-4150 for 1989 Edition of Section XI).
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- 5. '
SUMMARY
& CONCLUSIONS 1 This report summarizes the results of previous metallurgical analyses of cracked CRD cap l screws conducted by GE and presents the results of structural margin evaluation. The ,
j i results can be summarized as follows:
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(a) The review of previous metallurgical evaluations of cracked CRD cap screws shows l that the observed cracking is typically shallow (less than 0.065 inch) and the expected crack growth in future is insignificant. __ ._
(b) The allowable flaw depth is 0.15 inch considering both the Code cross-section area requirements and the LEFM considerations. This value considerably exceeds the deepest observed crack depth of 0.065 inch.
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- (c) The current inspection, evaluation / replacement practice during the CRD housing refurbishment at JAFNP consists of replacing the cap screws with new design cap screws as suggested in Reference 2. If a previously installed cap screw is reused, VT-1 snd augmented surface examinations prior to installation are performed per the GE SIL [2]. This evaluation / replacement practice, coupled with results in (a) and (b), and the recently enacted Code Case N-547, which eliminates the requirement for VT-1 visual examination of the CRD cap screws, provide suflicient technicaljustification for not expanding the sample size when cracked cap screws are found during the l refurbishment of the CRD housings. l 1
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- l 6.. REFERENCES.
[1] GE RICSIL No.19,"CRD Cap Screw Crack Indications," May 19,1988.
[2] GE SIL No. 483, Revision 2,"CRD Cap Screw Crack Indications," August 5,1992.
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[3] ASME Pressure & Vessel Code. l i
(a)Section III, Rules for Construction of Nuclear Power Plant Components, Div. l~.
I (b)Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, d
1980 Edition with Winter 1981 Addenda (2 interval).
(c)Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components,
! d 1989 Edition (3 interval). l
[4] J.M. Barsom and S.T. Rolfe, " Fracture and Fatigue Control in Structures - l Applications of Fracture Mechanics," Second Edition,1987, Prentice Hall, ;
Englewood, NJ.
[5] GE Drawing No. I17C4515P2, Rev. 6; Applied Practice 209A4270, Rev.1.
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[6] Tada, H., Paris, P. and Irwin, G., "The Stress Analysis of Cracks Handbook,"
Second Edition, Paris Productions and Del Research Corporation, St. Louis, i Missouri,1985.
[7] GE Document 22A2016, Rev. 2, dated January 15,1970.
[8] " Evaluation of CRD Cap Screw Defects James A. FitzPatrick Nuclear Plant,"
Report in NYPA File 62-B-0160, June 1997. ,
[9] ASME Section XI, Division Code Case N-547, " Alternative Examination Requirenmnts for Pressure Retaining Bolting of Control Rod Drive (CRD)
Housings," Approved August 24,1995.
[10] GE Internal Communication from E.Y. Gibo to A.L. Jenkins;
Subject:
Browns Ferry 2 CRD Flange Leakage, April 13,1991.
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