ML20040C086
| ML20040C086 | |
| Person / Time | |
|---|---|
| Site: | San Onofre  |
| Issue date: | 09/25/1981 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML13319A635 | List:
|
| References | |
| FOIA-81-313 NUDOCS 8201270251 | |
| Download: ML20040C086 (1) | |
Text
{{#Wiki_filter:.. - -_., _.. a.. _.-i..:- . 1: E - 3EaE',. w._ I"T:.:7_^ TIC' San Oncfre Unit 1 has used a coordinated socium phosphate treatment for the secondary coolan:, since initial start up in 1958. Wastage type tube cegradation aas f' irs detected in 1972, however the rate of tube cegradation was low and no extensive wastage occurred frem September 1973 to April 1975; i.e., there were only a few tubes that exceeded the 50% plugging limit. During the April 1975 inservice inspection of steam generator tubes at San Onofre Unit No. I minor tube denting was observed. Consequently, as a result of a small radius steam generator tuce failure at Surry Unit No. 2, the Sout,hern California Edison Company E (the licensee) in conjunction with Virginia Electric and Power Company, Florida Power and Light Company, Consolidated Edison Company, and with the participation of the Westinghouse Electric Corporation formulated a specific steam generator inspection program to assure the integrity of steam generator tubes at these PWR facilities that have experienced tube denting and the U-bend cracking of the small radius steam generator tubes. The inspection at San Onofre Unit No.1 included: (1) hand hole entry inspections (2) upper bundle e..try inspections, (3) eddy current - I probing inspections, and (4) a three inch access opening. drilled in l Steam Generator "C" for observation and measurement of the flow clots in the upper tube support plate. BACKGROUND Water Chemistry i l For many years a sodium phosphate treatment for PWR secondary coolant was widely used for U-tube design steam generators that removed precipitated or suspended solids by blowdown. It was successful as a scale inhibitor, i 8201270251 810925 PDR FOIA UDELL81-313 PDR
in::n ;-C. u::r.; ex:en er.:ec stres: ::rr:sion trackir.;. Tr.e : 10 1 :- was attributed to free caustic which can be formed when the Na/PC. ra-io exceeds the reco= ended limit of 2.6. In addition, s:::e of the incoluble metallic' phosphates, formed by the reaction of sodium phosphate: with the dissolved solids in the feedwater, were not adequately removed by bicw-down. These precipitated phosphates tended to accumulate as sludge on the tube sheet and tube supports at the central portion of the tube bundle where restricted water flow and high heat flux occurs. Phosphate concentration (hideout) at crevices in areas of the steam generator, noted above, caused localized wastage resulting in thinning of the tube wall. Largely to correct the wastage and caustic stress corrosion cracking encountered with the phosphate treatment, most PWRs (except for San Onofre ~ Unit No. l}with a U-tube design steam generator using a phosphate treatment for the secondary coolant have now converted to an all volatile chemistry (AVT). In 1975, radial deformation, or the so-called " denting," of steam generator tubes occurred in several PWR facilities after 4 to 14 months operation, following the conversion from a sodium phosphate treatment to an AVT chemistry for the steam generator secondary coolant. For this reason San Onofre Unit No.1 did not convert to an AVT chemistry, although tube denting did occur even with the phosphate treatment. Tube denting occurs predominately in rigid regions or so-called "hard spots" in the tube support plates. These hard spots ar.e located in the tube lanes between the six rectangular flow slots in the support plates j near the center of the tube bundle and around the peripherial locations of the support plate where the plate is wedged to the wrapper and shell.
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- _::;r Tr.e ;.-.; E:.;r :i :er.-ir.; n ; :een L::ri:u:ed :: the accelera:ed ::r-rosion of the carbon steel suppor plates at the tube / tube support plate intersections (annuli). The corrosion product (magnetitie) from the car::n steel plate has expanded volumetrically to exer: sufficient compressive forces to dent the tube and crack the tube support plate ligaments between the tube holes and water circulation holes, caused by an in-plane expansion of the support plate due to the magnetite growth.
As a result of the tube support plate expansion, the rectangular flow slots began to " hourglass"; i.e., the central portion of the parallel flow slot walls have moved closer so that some of the flow slots are now narrower in the center than at.the ends. ~. l' U-Bend Cracks On September 15, 1976, during normal operation, one U-tube in the innermost row parallel to the rectangular flow slots in steam gener-I ator A at Surry Unit No. 2 rapidly developed a substantial primary to secondary leak (about 80 gpm). After removal of the damaged tube ' and subsequent labcratory analysis, it was established that the leak resulted from an axial crack, approximatley 4-1/4 inches in length, f in the U-bend apex due to intergranular stress corrosion cracking l that initiated from the primary side. Since the initial paralled flow slot wall in the top support plate has moved closer, the support plate material around the tubes nearest the central portion of these O
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It is this additional increase in strain at tne apex of the U-bend which is believed to be required to initiate stress corrosion cracking cf the Inc;nel 600 alloy tubing ex:csed to PWP. primary c001 ant. However, the San Onofre Unit No. 1 steam generators have not experienced hourglassing in the top tube support plate that would lead to U-bend cracking of small radius tubes. Laboratory exaEination of 71 U-bends removed from flow slot locations in rows 1, 2, and 3 of the Surry Units Nos.1. and 2 and Turkey Point Unit No. 4 steam generators has shown that-intergranular cracking at the U-bend apex was found only in the row 1 tubes. t Of the 71 tubes removed from these operating reactors, which are the most severely affected, no cracks have been found in tubes with computed equivalent strains less than 13.5% after approximately 11,000 hours of effective full power operation since detection of the first tube dent. However, this same equivalent operating time lead to the tube failure at Surry Unit No. 2, where the equivalent strain was estimated to be >14.3%. This indicates a strain level at which rapid development of stress corrosion cracking may occur in U-bends of steam generators of this design. i ~ l Recent test work also indicates that long incubation periods are needed for the development of stress corrosion cracking at some strain rates.-5/ Tests indicated that at 12.5% outer fiber strain, 1/ 2-l
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~ Altncugh tnese tests results are no: directly applicable to the PWR steam generator tubing at Surry, they do confirn the observed operating experience that (1) a long. incubation time is required to initiate intergranular cracking in Inconel 600 material, and (2) a high strain is required for crack propagation. In this regard, the staff requested that the. licensees of affected plants address the following concern: "Hourglassing" may continue and close.the flow slots in the top support plate increasing the strain at the U-bend apex of the tubes in rows 2 and beyond. In response to this concern, and to supplement plugging of row 1, VEPCO' was.the.only: utility: ta;jnstall staijless. steel'3Q4. alloy blocks in each af the six flow slots in the top support plate of all three Surry Unit No.1 l steam generators. These blocks will prevent further closure of the flow slots and. inward displacement of the legs of the U-bends, thereby preventing further anticlastic straining at the U-bend apex of these tubes in rows 4 and beyond. As a result, intergranular stress corrosion cracking of those tubes at the U-bends in rows 2 and beyond is not anticipated during near term (next year) normal operation.
- However, the flow slot blocking devices would cause:
(1) an increase in strain I
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nt 2:;i:,. ::::- channa', and ::ea ganera::r shell due :: the peripnerial expansion of tne support plate. The net overall effect of flow slot blocking device: w:uid be similar :c c:= pie:e :iosure of the finw slots. H: wever, VEpCO had also increased s. elective tube plugging in the hard spo: regions for the prevention of tube leaks at dented locations. Succort plate Exoansion Continued growth of the magnetite in the tube-tube support plate annuli results in a non-uniform increase in strain in the support plates and corresponding in-plane expansion. In this regard, the staff recuested , that the licensee's of affedted plants address the following concerns: "1. Severe cracking of the support plate may result due to.the contin-uing in-plane expansion of the support plate, 2. The rate of in-plane expansion in any support plate could increase the severity of tube denting in "hard spot" regions. Severe denting would restrain the tubes in the support plate and the plate may have a tendency to buckle or otherwise deform and thus exert additional bending loads on tubes. 3. With the closure of all. th.e flow slots in any one support plate ~ i additional loads could be transmitted (due to the in-plane expansion of the plate) to the wedges, wrapper., channel spacer, 1 tubes, and the steam generator vessel and
- 4. Thermal-hydraulic performance could be affected >dth the closure of all the flow slots in any support plate."
_! :d.vi:rz: - Ea ~re:-ine Or Novem:er 17, 1976 Southern California Edison Ccmpany (SCEC) reported IEE, F.egi:r. 'l, tr.::, durin; the ir.:;e:ticr. of the San Oncfre Ur.it No.1 ::eam ;enera tors, excessive wear or mechanical fretting of anti-vicration bars was found in one of the steam generators. A failure of these bars could result in excessive flow induced vibrati.on that might affect tube integrity, especially for those plants where the tube denting, phenomenon was observed at the top support plate. Subsequent investigation revealed that the anti-vibration bar design of San Onofre Unit No.1 and Connecticut Yankee is uniqu_e in comparison with other Westinghouse plants. Differences in the design are summarized as follows: a. Materials - carbon steel for San Onofre Unit 1 and Connecticu?. Yankee; Inconel 600 for new models (44 and 51). b. Bar Cross-section - 3/8 inch round bars; changed to^ square bars in the new models. c. Clearances - (L-35 mils); was changed to (L-20 mils) for new models where L is the tube spacing. i d. Changes in V-bar configuration and spacing. DISC'JSSION i On Fttruary 1,1977, the licensee submitted results of the steam generator inspection program at San Onofre Unit No.1 and the corrective actions to assure continued integrity of the steam generators. e 4 ~"*A-* %+=
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-E_ Tr.c ::rr :: ice 2::i:r.: c:r.:i::cd Of: 1. instaliation of. new antivibration bars of differen: si:e, materiti, and location, 2. Prever,tive tube plugging for.all effected tubes, not supcor:ed by the new antivibration bars, witn ECT indications less than 20% which provides a safety margin of 60% for tube wall cegradation between inservice inspection periods, i 3. plugging of all steam generator tubes with ECT indications of 50% or more at other defect locations, and 4. a total of 99 tubes plugged, 30 in "A" steam generator,18 in "B" steam generator, 51 in "C" steam generator. Implicit in this (or any) plugging limit are three factors. The first relates to the maximum allowable tube wall degradation or partially through wall crack-depth that should not be exceeded at all times. The second concerns the operating al.lowance, which is subtracted from the maximum wall degradation stated above, to ensure that the tube defect will remai.n below the maximum allowable during the operating - period between inspections. In consideration 01 the fact that tube degradation may. progress, during.this period, the operating allowance provides margin for this eventuality. The third considers tube integrity f during combined loadings produced by a toss-of-Coolant Accident (LOCA), a main Steam Line Break (MSLB), and a Safe Shutdown Earthquake (SSE). These requirements are consistent with Regulatory Guide 1.121, " Bases for Plugging Degraded PWR Steam Generator Tubes" August,1976. b
-p_ Easec :n :ne esul Of 'es:ingncese's analysis, :se licenses nas pree: sed 50% tube wall thinning (40% of original wall remaining) as the maximum cecrada-ion that can be tolerated anc still preserve the streng:n recuired to meet all requirements exce:t at the t:p su: port plate intersections, where maximum bending stresses are expected during a LOCA. At these intersections, the maximum degradation that can be tolerated and still be able to withstand LOCA induced bending loads is calculated to be 53% of the wall thickness; i.e., the minimum wall thickness recuired is 47%. On the basis of inspection results, the licensee has proposed a series of operating allowances, which vary depending on the type and locations of the particular anticipated defect location, to establish a tube plugging criteria for San Onofre linit No.1 (summarized in Table I). The licensee propo'se's that the next steam get.eretor inspection be conducted during the next refueling outage. EVALUATION The results of the October-December 1976 inspection indicate that tube degradation by wastage occurred only in the "A" steam generator, which had 12 tubes with ECT-indications (50-59%) two inches above the tube sheet. ~ There was no indication of tube degradation from caustic stress corrosion . cracking in any of the three steam generators. The rate of wastage observed in the San Onofre steam generators is not unexpected for plants that have operated with a phosphate treatment for the secondary coolant, t am,..-- .-een_. w-
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. as ciscussec n.tne a:ove cackground. The c:ntinued eas a;e may be caused by periodic localized acidic conditions on the tube surfact in conjunction with a low Na/PO. ratio (<2.3) in the sludge layers that have accu =ulated on the tube sheet during the period of pnesphate treatment. San Onofre has had periods of condenser inleakage and tne chlorides frcm the sea water (for condensate cooling) that concentrate in the sludge layer tend to lower the solution pH at the tube / sludge layer interface. Since the San Onofre condensers have been retubed with copper-nickel (90-10) T and titanium tubes, we believe that this will significantly reduce the potent.ial for the inflow of impurities into the secondary coolant, and thus decrease the probability of acidir environment in the steam generators thereby decreasing the rate of wastage, pitting, and the potential for accelerated tube dhnting or tube support plate deformation. Furthermore, San Onofre has not had a history of caustic stress corrosion cracking and the Na/PO ratio for the secondary coolant has been maintained at 4 2.6-2.3 since early 1973, tnerefore, we believe the probability of this type of tube degradation before the next refueling outage is minimal. Although the October-December 1976 eddy current inspection results indicate no evidence of cracking at the apex of small radius tubes in rows 1 through 5, tube wall thinning in all three steam generators was detected in the U-bend reginn at the.anHvibration bar locations. Approximately 1900 tubes in each steam generator were inspected. Tube defects in steam generator "A" occur above row 35 of which three' tubes had wall thinning >50%; in steam generator "B" there were 13 tubes with .=
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.c : s : r.2:r :nc :c-i-.try O' t.. u:: bundie. The licensee has submitted analytical data in support of the proposed plugging Timits described in the above discussion. 'n'e have reviewed this data and have performed incependent evaluations to determine the adequacy of tne proposed plugging criteria and themaximwn allowable tube wall degradation. Regarding the tube plugging criteria applied to the San Onofre steam generators for continued service, we have taken the position that (1) the tubes with detected wastage acceptable for service must not be stressed beyond the yield strength of tube material during the full range of normal reactor operations, (2) the fa.ctor of; safety against failure by bursting i under normal operating conditions must remain at no less than 3 at any tube location, and (3) _st,resses induced by the combinations of loads due to postulated accident. conditions and the Safety Shutdown Earthquake (SSE) should not exceed the limits for plant faulted conditions of Section III of the ~ XSME ' Cod's.~ The factor'of safety of 3 against ductile failure is consistent with the safety margin inco.porated in the Class 1 design rules of Section III of the ASME Coce appi1 cable-to the San Onorre steam generators This positien has been delineated in the new Regulatory Guide 1.121, { which was published in August 1976 for comment. The licensee demonstrated that the maximum allowable tube wall thinning would be 53% at the upper tube support plate and 60% at the antivibration bar locations and tube sheet region. The analysis were made under e .:. = - -
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The SSE bending stresses were derived frcm a dy:nmic seismic reanalysis of the San Onofre Unit M.1 steam genera:crs to a 0.57g grcund motion earthcuake. The resultant stresses in an assumed 55% uniformly thinned tube subjected to the ccmbination of the hydrodynamic loads caused by rarefraction waves, steam generator LOCA shaking stresses, LOCA pressure stress, and the SSE lead were shown to be below the Code allowables for plant faulted conditions. The tubes are assumed to be uniformly thinned; while in reality, degradation occurs only locally. The staff agrees that the ~ assumption of uniform thinning in the analyses is conservative and acceptable. With regard to tube degradation between inservice inspection, field data covering five years of operation it San Onofre Unit No.1 indicates that tube degradation at antivibration bars is 5% per year near the center of the tube bundle. However, the licensee has chosen a conservative 10% until the next planned reiueling outige. A similar tube degradation allowance for wastage at the tube sheet will apply. Since the tubes with large bend radii, near the bundle periphery, are not supported by the-antivibration bars, special consideration was given to these tubes,.and a margin of 60% will.be applied to these tubes. In addition, the licensee has applied a.3% margin for tube degradation at the upper tube support plate, although there has never been any evidence of tube thinning at these locations. In sumary, uniformed thinned tubes with up to 53% to 60% degradation meet all strength requirements, with a safety margin against failure ... ~.
. by curs:in; :na; es.:ee: :nree (3) f r ne whcie range Of n:r :.1 :;erating conditions. Adecua:e margins are provided for postulate: accident c:nditicn:. Based on the San Oncfre Unit No. 1 operating his ry for tne pas; nine year;, ,,e ::n:ur that the varicus Operating allewances proposed by the licensee provide a conservative allowance for tube degradatun until tne next inservice inspection. Therefore these proposed minimum wall thicknesses are acceptable, as is the operation allowance applied to the plugging criteria to ensure tha_t this ninimum wall is maintained until the next refueling outage. In regard to the tube wall thinning at the antivibration bar locations, there w'as also thinning of the 0.25 inch diameter carbon steel antivibration bars in all three steam generators. [aboratoryexamination of three tubes and two antivibration bars from the "C" steam generator revealed that thinning of the tube and antivibration bar was the result of mechanical fretting at points of contact due to excessive tube vibration and impact, leading to increased clearances between the tubes and the antivibration bars. There was no evidence of cracking or corrosion attack on either the tube or the antivibration bar. In the U-bend region, where the flow is two phase; a vortex shedding mechanism is not expected to occur that will damage' tubing. Consequently, the licensee analyzed the tubes for fluidelastic excitation. The results of the tube dynamic analyses showed that the effects of cross-flow over tubes with a large bend radius in rows 32 through 48 make the tubes unstable when there is excessive clearance between the tube and the antivibration bar. Furthermore, limiting the diametric clearance between the tube and the ativibration n =- . :- ~ - =.
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antivibration bar wear to no more than 3t per year. In order to arres; the accelerating tuce w2ar in the U-bend area, new antivibration bars fabricated frcm chrome-pla:ed Inconel 600 with a square cross section instead of the original carbon steel bars with a circular cross section were installed between the existing antivibration bars. The new antivibration bars penetrate the bundle to a sufficient depth to provide additional support for all tubes between columns 18 through 83 in rows 23 to 48 that have potential for excessive vibration damage. Based.on the evaluations and corrective actions discussed .above, we concur that further tube and.antivibration bar degradation due to excessive vibration damage will-be arrested. We also concur that further thinning of the existing carbon steel antivibration bars will be arrested and the potential for excessive vibration of.the bars themselves is not anticipated. In the case where an old antivibration bar might fail, resulti.ng in a loose piece, analysis has shown that the. normal operational flow forces will not lift the fragment. Th,er,efo r,e, j. impact of the fragment against adjacent tubes during normal operation i i is not expected. For accident conditions,. projectile impact tests at velocities simulating a worst case steam line break (MSLB) accident indicate that the tubes will not fail upon fragment impact. ~ ~ The eddy current probing (ECT) d'emonstrated that tube denting in the i San Onfore steam generators is minimal. Steam generator "B" had no {' denting at any of the tube / tube support plate intersections, and any 9 1T... p.emee
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f.ates in these steam generat::rs. The severity of the OD/ID indentations is n:minally 0.001 inch on the diameter,'and the rate of denting is.no more than 0.5% per year. Consecuently, the San Onofre steam generators have not experienced any leaks associated with dented tubes, since the first observation of denting in April 1975. Results of the hand hole entry, upoer bundle entry, and the 3-inch j hole (at the top support plate in the "C" steam generator) inspections at flow slots in the top support plates confirm the tube denting observations; - i.e., minor tube support plate deformation and flowslot "hourglassing" 4 _is confinedto the bottom twc supoort plates in steam generators "A" and "C". There was no evidence of either phenomena in steam generator "B". Physical measurements of the nearest three flow slot widths in the upper support plate of generator "C" were made, Based on these measurements, the flow slots in the top support plate have suffered no deviations from the manufactured condition, i.e., a reasured widthof 2-3/810.0625' inches versus a manufactured width of 2-3/8 1 0.0120 inches. Measurements of the hourglassed flow slots in the lower two support plates in generators A & C were not made. Based on the information submitted by the licensee, the NRC staff concurs in the following: 1. Since tube support plate defonnation or flow slot "hourglassing" does not exist in the top two support plates, of the San Onofre steam generators, there is a low probability for U-bend cracking ....,.-..;;,y..
.~: of reali radius U-bend tutes in rows 1 tnrough 5 or beyond. 2. The retubing of the San Onofre condensers, in addition to the phosphate treatment, minimi:es the environment conducive to magnetite grow:n at the tube /tute support plate intersections and significantly reduces the potential for rapid support plate deformation and severe tube denting. 3. The present degree of support plate expansion in the bottom two support plates in the proposed period of operation will have insignificant ffects on the wrapper and the stemn. generator vessel Therefore, the wrapper and the vessel integrity during normal c;erating und accident conditions will not be affected by continued support plat'e expansion. Due to hourglassing of the flow slots.: minor additionar loads are transmitted to the steam generator shell through the load path of the support olate, wedge, wrapper'and channel spacer. Based on preliminary " crush" tests performed by Westinghouse the maximum load that can be developed along this load. path is 60 000 pounds. 2 Analysis of the bearing stress along this load path indicates that all stresses are.less than the yield strength. Such stresses on the steam generator,shell are highly localized and self limiting and will not adversely affect the integrity of the shell under accident conditions. 4. Since the total area of all six flow slots is only a small fraction of the total area for flow circulation.,the effect of slight flow slot hourglassing on the thermal hydraulic performance of the steam W
-1= ger.erator will be negligible. Tnere will ce a slign: cecrease in the circulation ratio and the liquid flow velocities, with an increase in raw steam quality. But these are so small that they may be disregarded. We therefore conclude tnat: a. Tubes with ECT indications less than those indicated in the plugging criteria of Table 1 retain adequate strength to withstanc the most severe accident conditions of a main steam line break and of a LOCA in combination with the SSE for the proposed operating period. ~ b. During normal operation the factor of safety against burst is greater than three and the maximum stress levels for the limiting case do not exceed the elastic range of the tube materials under normal operating conditions. I c. Additional tube plugging is not necessary to provide protection against the occurrence of substantial number of rapid leaks or of a darge number of undetected ' incipient failures, which could result in adverse safety consequences in the event of a reactor accident i or an un-anticipated operation transient. I d. There is reasonable and conservative assurance of tube integrity ? for the proposed period of operation to provide adequate protection to the public health and safety. l .m==. =~
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al:ncu:n ne cegree of den:in: d:e; n;; appear :: be severe, tne staff does not agree that the preposed power operation until the next refueling outage can be fully justified and believe that a interim insce: tion should be conducted at the end of twelve (12) months after start up. The licensee has been unable to ~' quantify the effects on tubes at intersections or assure that continuing growth of magnetite will not occur. Therefore, concern over a possible increase in tube denting at the tube / tube support plate intersection for the lower two support plates cannot be completely alleviated.
- Also, because of the existaoce of dented tubes subjected to support plate strains, there is some concern that the severity of denting may increase and the integrity of some un-plugged dented tubes cannot be maintained
~. during postulated accidents. There aret.however...several factors which suooort a period of twelve months operation before the reinspection of the San Onofre steam generators; i.e., qualitative and preliminary quantatitive integrity data, the low consequences of the relatively limited tube leakage that would be expected under postulated accident conditions and the very low probability of an initiating accident coincident with a large number of tube failures. The qualitatiee and preliminary quantitative integrity data is sumarized as follows: a. Preliminary analyses of the support plate expansion (with a flow slot hourglassino) indicated small hard spot strain increases - - - - ~ - --
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b. All leaks associated with dented tuces experienced a the c;her reactors to date have been small, well belcw ccmm:nly acceptable leakage limit:. i Possible through-wall cracks in the dented regions, i.e., tube /tuoe c. support plate intersections, are constrained by the support plates; therefore, cracks should not burst during postulated accidents, 4 until the crack grows substantially beyond the tube support plate I region. d. Through wall cracks at dented locatiqns, with the amount of leak-ages experienced to date at other reactors, have been stable during s normal operation (no rapid failures), and are not anticipated to become unstable during postualted accidents. i Even though some non-through-wall cracks may exist and may crack e. '~ through during post 21sted accidents, the. associated leakage rate with such an event would be similar to tnat resulting from through wall cracks found during normal operation and the crack would not i !j; be unstable. .= j' CONCLUSION We have concluded, based on the considerations discussed above, that t I f (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, and (7) f such activities will, be conducted in compliance with the Commission's regulation and the issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the.public. a ..s.. ..sw
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Place (1) (2) Other Min. Tube Wall Thickness Required (3) 477. 407. 407. 407. Allowance for Degrada-{' ion between Inspections +> 37. 107. 607. 107. I= perfection Depth a: or Beycnd Which Any measured Tube Plugging will be Perfor=ed 507. 507. indication 507. Notes: ~ (1) Colt =ns 18-83 (2) Rows 30-36, Colu=ts 12-17,.84-89 ( Percent of Tube Wall Remaining (4) Percent of Tube Wall Thinning Allowed Between Inspections m O 9 i G =
F E:E:E';:FI 1. H.A. ::mian, Et a'.. Effe:: of Micr:s:ru::ure en Stress C:rr:sion Cra: king of Alicy 600 in High Purity Water. Corrosion, Vol. 33, . 26, (January 1977). 2. R.L. C wan an: G.P.. Gcrd:n. Intergranular Stress Corr:sion Cracking and Grain Ecundary Com:osition of Fe-Ni-Cr Alloys, Prcprint G-14 of paper presented at Stress Corrosion Cracking and Hydrogen Em:rittlement of Iron Base Alloys Conference, Firminy, France (June 1973). 3. J. Blanchet H. Coriou and et al. Influence of Various Parameters on Intergranular Cracking of Inconel 600 and X-750 in Pure Water et Elevated Temperature, Preprint G-13 of paper presented at Stress Corrosion Cracking and Hydrogen Embrittlement of Iron Base Alloys Conference,Firminy, France,(June 1973). i ' l 4. F.W. Pement and N.A. Graham. Stress Corrosion Cracking in High Purity Water, Scientific Paper 74-186-TUCOR-P1, Westinghouse Research Laboratories,- (June 23,1974). 5. Ph. Berge, H.D. Bui J.R. Donati and_D. Villard, Corrosion, Vol. 32, ,p.357,(September 1976). i l 1 4 4 6 i i -~-... ---}}