ML18051A927
| ML18051A927 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 06/11/1984 |
| From: | Crutchfield D Office of Nuclear Reactor Regulation |
| To: | Vandewalle D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| Shared Package | |
| ML18051A928 | List: |
| References | |
| LSO5-84-06-015, LSO5-84-6-15, NUDOCS 8406130160 | |
| Download: ML18051A927 (37) | |
Text
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UNITED STATES NlJCLEAR REGULATORY COMMISSIO WASHINGTON, D. C. 20555 Docket No.
50-255 L.S05-84-06-015 Mr. David J. VandeWalle Nuclear Lic~n~ing Administrator Con~uners Power Company 1945 v:. Parnal1 Road Jacks6n, Michioan 49201
Dear Mr. VandeWalle:
June 11
- 1984
SUBJECT:
. 1983-84 STEAM GENERATOR INSPECTION Re:
- Palisades Piant ENCLOSURE
\\..Je have completed:_our review of your r~port, 11 1983/84 Steam Genera;tor Evaluation and Repair Program Summary, 11 dated April 19, 1984.
The NRC staff's safety evaluation report is enclos*ed.
This safety evaluation report concludes that you have (1) identified the e_.gent which caus*ed,*_:__~
the steam generator tube degradation, ( 2) pos tu-:[oted c. damage scena ri~
wh i ch
~ s consistent ~;th the known. facts per~.; ~t =to the observed_,_. - - ;....>
corros l on, ( 3) taken necessary 2ct10ns to prevent recurrence of the observed steam generator tube de.gradation, an-d -rA.) __ presented data which,:-
formed the basis for a plugging criteria of 51~ wall th~nning for the defects of limited c.xial and circumferential._:;.eX?:'ent at Palisades.
As discussed in Section 4.3 of the safety evaluation report, the NRC staff requested and you agreed to implement a more stringent primary to secondary leakage rate.
Your May 15, 1984 letter *stated that primary to secondary leakage will be limited to 0.lgpm as opposed to 0.3gpm spec1fied in.Technical Specification 3.1.5.d.
We have discussed this matter _with the Resident* Inspector at Palisades, and he concurs that your Standing Order No. 41: Rev. l, dated May 30, 1984 which limits ieakage to O.lgpm is sufficient and no modifications to your Technical Specifications are necessary.
Dayid J. VandeWalle 2 -
june 11, 19'84 Fi~2 1 1y, the results of the recent steam generator i~spections indicate that it would be appropriate to update Section 4.14 of the Palisades Technical Specifications.
Accordingly, we.request that you propose revised Technical Specifications within 60 days of 1he receipt of this letter.
Enclosure:
Safety Eva-luation Report
- cc-w/encl osure See next page
- Sincerely, n ( ~~-*--~*C ~h_\\o~d Ch. f f
uen-ni,--s--ri:-
ru~nT\\el,
ie
~~~~- ing Rea~~ Branch #5 D. ~:on of Licensing
~:--
J
Mr. David J. VandeWal1e cc M. I. Miller, Esquire Isham, Lincoln & Beale Suite 4200._
One First National Plaza C~icago, Illinois 60570 Mr. Paul A. Perry, Secretary
=~~2Jners Pow~r Company
- -"212 \\~est Michioan Ave.nue Jackson, Michi~an 49201 Judd L. Bacon, Esquire Consumers-Power Company 212 West Michigan Avenue Jackson, Michigan~ 49201 James G. Keppler, Regional Admin~strator Nuclear Regulatory Commission, Region III 799 Roosevelt Road Glen Ellyn, Illinois. 60137 Township Supervisor Covert Township Route 1, Box 10 Van Buren County, Michigan 49043 Office of the Governor Room 1 - Capitol Building Lansing, Michigan 48913 Palisades Plant ATTN:
Mr. Robert Montross Plant Manager Covert. Michioan 49043
~
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1.S. Environmentc:l:Protection ~gency Federal Activities Branch RegiG'h \\I Office ATTN:
Regional Radiation Representative 230 South Dearborn Street Chicago, Illinois 60604 Resident Inspector c/o U.S. NRC
~::li::e:des Plant Poute 2, P. 0". Box 155 Covert, Michigan 49043 Lee E. Jager, P.E., Chief Environmental and Occupational Health Services Administration Michigan Department of Public Health 3500 N. Looan Street Post Office Box 30035 Lansing, Michigan 48909
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ENCLOSURE SAFETY EVALUATIONS BY THE OFFICE OF NUCLEAR REACTOR REGULATION 1983/84 STEAM GENERATOR EVALUATION AND REPAIR REPORT PALISADES NUCLEAR PLANT DOCKEl NUMBER 50-255
__ J~_* Q_ INTRODUCTION By letter dated April 19, 1984, Consumers Power Company (CPC) submitte-d for our review, a report entitled, 111983/1984 Steam Gener*ator Eva l ua-tion and Repair Program Summary 11
, and on April 25, 1984, they provided responses to our request for additional information.
Infor.mati_9n meetings were_held in Bethesda on October 2~, 1983, February 21, 1984,
~
and April 10, 1984.
Members of the Headouarters and Region-1113iaff,
~
along with a consultant from -the Oak _Ri dg_e_ _Nati ona 1 Laboratory, ;ffi"et wi_th CPC on December 7, 1983 at the C_er~.ate Research Center in_
Ja~kson, Michigan in order to examine Eddy Current Test (ECT) data records and to observe a demonstration of the eddy current sy~tem equipment.
The 1983-1984 inspection was performed on 100% of the tubes with a
-- -~.
~ _,,_
4 x 4 panca~e probe.
The results of the final evaluation of the 1983-84 ECT data show a total of 181 tubes with cracks and 29 tubes with inter-granular attack (IGA) gr~ater than 50%.
With preventive plugging, a total of 277 tubes were plugged in both steam generators.
2.0 BACKGROUND
Pali s __ ades is a pressurized wc.ter react9r having two Combustion Engineering designed U-tube steam generators.
Each steam generator
- contains 8519 Inconel-600 tubes with a 0.750 inch outside diameter 2nd a 0.048 inch minimum wall thickness.
On -january 15, 1973 after approximately one year of iBtermittent operation at less than full pow.er, the Pali sades pl ant e~peri enced its first steam generator tube leak.
,1*
ECT examinations of ~he.~ubing
~
detected general wastage attack in the U~end area of tubes in_.t[ie first e 1 even rows from the divider pl at~
--. -: -e--
The attack was attri~ted to the use of a coordinated phosphate. secondary "water chemistry: *
~--
All tubes in these first eleven rows were plugged, and the plant r~turned to service in March, 1973.
The plant operated at essentially ful1 power until August 11, 1973, when it was shut down because of excess steam generator tube leakage.
ECT examinations in September, 1973 showed measurable wastage on nearly half of the steam-generator tubes.
Evaluation.of required tube strength showed that tubes with wastage less than 60% through-wall could re~ain in service.
All tubes with greater than 60% wastage defects were plugged.
ln May 1974, preoperational hydrotests identified two leaking tubes.
ECT showed that wastage had increased and_ that IGA accompanied by shallow ~itting was present between the support plates in the upper hot leg.
All tubes with indications of IGA and those with greater than 50% wastage were plugged.
The IGA was attributed to a mixture of reduced forms of sulfur and sodium phosphates.
Beca~se the IGA appeared to be growing rapidly while the plant was in cold shutdown, the plant was permitted to return to powet at levels up to 60% in o!der to flush or volatize any sulfur or soluble ~~os-phorous compounds from the steam genera~c>.r*s.
-~ ~
~-
- - ~
Upon return* to service in October 197-l. 1.b..e. sec9ndary water chem1 stry****
t re*atment was changed to an a 11-vo lat i 1 e treatment as recommended by the steam generator vendor in order to arrest wastage-type corr._osi on.
-~
Subsequent ECT examinations in 1975 ~hrough 1981 showed that wastage torrosion of the steam generator tubing had essentially ceased although minor tube denting was occurring as a result of the switch-over to all-volatile treatment.
In March, 1982, a primary-to-secondary leak-in excess of the Technical Specif_ication limit of 0._3 gpm occurred in steam generator 11A 11
- Initial ECT examination (1-.1ith the standard 11 bobbin 11 probe) of about 35 tubes which had been identified as possible leakers by observation of
4 moisture on the tubesheet, showed no indications of tube degradation other than those seen during previous inspections.
Pressurization of the s~condary side of the steam generator identified two leaking tub~s.
Subsequent ECT examinations of the faulted tubes with the "bobbin" probe and an advanced 11 pancake 11-style 4 x 4 ECT probe, which has better flaw detection capability for circumferential ~racks and IGA, shewed the defects to be through-wail with a ci rcumf e-i.ent i a 1 ori en ta-tion.
Thedefect in one tube was found to be high in the hraiaht,
,1*
vertical hot-leg section of.the tube at tub.:_ support plate.. No.:*:9, while the defect in the other tube was l~cated in the horizonta-=F*
.::==:---
~
section of a 11 Batwing.
11 Addi.ti ona 1 ECT -examinations were performed
~~
with both the 11 bobbi n 11 and 114 x 4 11 pr.ob~to provide assurance t_hat similar, pluggable defects did not exist in the remaining steam-generator tubes.
Also, in recognition of the fact that one of_ the defects was found near the 90-degree bend plus the fact that the 4 x 4 ECT probe was incapable of transversing bends, CPC committed to develop a probe similar to the 4 x 4 probe but capable cf trans-versing the _full length of the tube for the lSC.3 ste2m generator inspection.
In 1983-84, CPC performed a 100~ inspection of the steam generators with the new flexible 114 x 4 F 11 pancake probe.
InHial ECT data, report~d in October and November of 1983, indicated approximately 40,000 11 non-quantifiable 11 indications and a large number of probable cracklike indications.
- 3. 0 DISCUSSION An exte~sive effort was conducted by the Licensee to identify the extent and nature of the apparent indications and to identify remedial actions.
Portions of 55 steam generator tubes and approximately 70 tube-to-tube support plate intersections were removed from both steam oenerators for detailed metalluroical examination. chemical ana1vsis and-materials properties testing in order to correlate ECT indications with any observed degradation.
In addition, the 1 i censee conducted a_ P.rG-Qram based on 81 1 aborat:ory *
~
~
IGA samples in order to qualify the adequacy of the ECT techniqu~ to quantify the depth of IGA, to determine its sensitivity, and to~
- -.;..~*-*
establish its calibration. They confirmed that the IGA generated in the laboratory is metallurgically similar to that found in the-actual steam generator and that ECT measure~ents on laboratory samples corre-lated.with IGA samples removed from the steam generator.
As a result ?f this program, they have shown that the technique could correlate the volumetric loss of tube wall material due to JGA and express it as an equivalent average depth of intergranular attack within a statistical error of + 6% of wall penetration and to a threshold of detectability of 30% of wall.
Conclusions reached by the licensee from the ECT results and the metallurgical examination are summarized as follows:
(1) Three types of OD initiated degradation are present on the tubes, these were general !GA, pits - shallow and deep, and JGA spikes.
The deepest pit and the deepest !GA spike had 14% and 19% wall penetrations, respectively, A deep pit on one t~be had 16% wall penetration.
\\,
I,*.
(2)
The incipient !GA and pits were distributed over all ~he.tube sections, irrespective of their _pos-Ttion in relation to thG."'
antivibration bars (AVBs_).
The --~G~ spikes appear to be lim?.::ted
_7illl;..
to only the AVB intersections, whe~~CT i n'di cations were obtained.
(3)
The class of ECT indications which the ECT interpreters had originally classed as "non-quanti_fiable" were shown not to be related to actual observable defects.
Twelve tube-to-tube support plate i_ntersections with indications that were called either "non-quantifiable" or 11 no-detectable-defect 11 were metallurgically examined and no degradation greater than 25% through wall were found.
The eddy current signal associated io.1ith the 11 non-quantifiable 11 indications was apparently caused by surface deposits or dents.
One if'itersection io.1ith a "no-detectable indication "as found to contain a 25% degradation.
- (4)
Three intersections from the B steam generator cold ieg which had been identified by ECT as containing deep quantifiable defects were meta 11 urgi ca lly examined and.found to contain no degra-dat ion.
The ECT signal was apparently caused by dents o~ tube deformation.
(5)
Five intersections containing actual degradation (three cracklike.
(6) defects and two IGA) were metallurgically examined.. In all cases,
- ECT correctly called the degradation.
In the case of the two IGA defects 1 ECT provided an accurate determination of thetmaximum
- average depth.
=
-- -~
~=-
The actual crack and IGA degradatiori was found to be less t~an 75
--==-~
mils axial 2nd 130° circumferential extent.
- -* -,z
- -;.~'.'-..:
(7)
The actual degradation was found to contain relatively large amounts of sulfur and the tubing was found to be highly sensitized.
I.t is be 1 i eved that the corrosion was caused by a reduced form o"f*
sulfur acting on sensitized Inconel-600 tubing and that it occurred prior to 1974, since similar pits and IGA spikes were observed in the tubes from the same steam generator in the 1974 examination.
(8)
Most defects reside adjacent to dents at the upper or lo~er edge*of the tube support plate.
l (9)
Since sulfur in the form of sulfide is noncorrosive, the licensee chooses not to remove the remaining sulfur from the secondary side
.of the steam generators.
In summary, the results show a total of 181 tubes with cracks and 29 tubes with !GA defects greater than 50%.
The numbers for the 11 811 steam generator show 160 tubes with cracks and 22 tubes with IGA defects of greater than 50% depth.
Simi 1 ar numbers for the 11 /..
11 _steam generator are 21 tubes with cracks and 7 tubes with !GA defects greater than 50%.
/'
The licensee concluded t~at the st~am generator tube !GA aDd pittin~
- -=--
=--'-
degradation were caused by reduced form.pf sulfur introduced i n~.1974.
- 4. 0 EVALUATION 4.*l D~termination of Causative Agent(s)
~-
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-*;;r The licen~ee and ~ts consultants conducted extensive microstructural and fractographic examinations on Inconel-600 tubing specimens taken from the Palisades steam generators.
Degradation in the observed specimens exhibited a morphology characteristic of intergranular attack.
In some instances, in addition to !GA, cracking was also observed.
Austen~tic stainless steels and certain nickel-base alloys, such as Inconel-600, under certain conditions, are known to be susceptible to IGA and intergranular stress corrosion cracking (IGSCC).
The
9 -
occurrence of lGA requires the alloy to be in a sensitized condition or with a susceptible metal1urgical condition and the presence of an aggressive environment.
However, for lGSCC to occur, three conditions must be present simultaneously:
(1) a high tensile
- stress, (2) a susceptible alloy microstructure, and (3) an aggressive environment.
Microstructural characterization of specimens taken from pulled tubes from the Palisades steam generator confirmed that the tubing alloy i~
typical for Inconel-600. **In addition, the grain size (ASTM 5.a.nd 7}.
==
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is representative of mi 11 annealed A 11 oy.:,:600 mi crostructure wi tr..:-heavy_.. :
- ... - -~ :
carbide precipitation along grain bounda[ies.
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Modified Huey tesiing
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2nd electrochemical potentiokinetic reactivation (EPR) testing con-
--=-_~--~
ducted by the licensee on removed tube ~ect ions confi rrned that _the a 11 oy rnicrostructure has been heavily sensitized.
Based on this information, the staff agrees with the licensee's conclusion that the tubing in the Palisades S/Gs has been sensitized and is susceptible to IGA and JGSCC if containments are present in the S/G water.
Because the presence of I GP.- causing corrosive species is a necessary condition for the observed tube degradation and cracking, the i'icense~
and its consultants conducte~ numerous x-ray diffraction (XRD), scanning electron microscopy (SEM), and energy dispersive x-ray analysis (EDAX) on specimen taken from the defective tubes pulled from steam genera-tors.
The analytical results sho~ed that high concentrations of sulfur
10 -
existed.
Although the presence of chloride, phosphcrus, silicon, and calcium was also confirmed, sulfur, in the reduced form, is the only identified element that has been associated with IGA or IGSCC of 1nconel-600(l, 2, 3)
Sulfur-induced IGA and IGSCC of sensitized stainless steels is a familiar.
phenomenon(l)_
IGSCC of sensitized stainless steel by polythionic acid in defulfurizer hydrocrackers, and many other systems_._~of the petrochemical industry
. (2 ""')
~as been well documented
.~.
Furthermore, the 1 ~ulfur-induted IGA and IGSCC have been previously observed in the nuclear.. ind~stry and reproduced in laboratory tests by the 1 icensee.
Based on the above analysis, ihe staff aarees wi~h the license~
1 s conclusion that the type of corrosion observed at Palisades st~am generator tube was primarily sulfur-induced IGA and, in some c~ses, accompanied by IGSCC.
4.2 Postulated Damaoe Scenario Two types of corrosion related tube degradation were observed in the.
metallography.
Intergrannular attack (IGA) 1 and stress-assisted IGA or IGSCC.
Th~ licensee stated that the reduced sulfur and acidic pH, which occurred in the 1974 dry layup following sodium sulfite (Na2so3) chemistry additions for oxygen control in secondary water, which is no longer ~sed were responsible for the corrosion observed in 1974 and the.*
currently detected steam generator tube degradation.
Given the referenced chemistry of reduced sulfur aqueous solutions, th~ postulated corrosion mechanism is supported directly py the foll*owing facts and observations:
(1)
Removed tube degradation type, microstructure and sensitization-,
and reduced sulfur environment.
~
.-c--
-~ **:.
(2)
Other similar commercial US PWR corrosion experience.
~:*
(3)
CPC laboratory experience with generating IGA in sensitized Alloy-600.
(4)
A definable plant event when sulfur corrosion is known to have occurred (1974 dry layup following Na2so3 chemistry (oxygen control).
This scenerio is considered sufficient to account for local areas of IGA resulting from the dry layup event in 1973-1974.
However, some tubes were found, in certain instances, to contain cracks within the
1, f
12 IGA.
We attribute these cracks to the stresses imposec (operational-heatup/coo 1 dO\\*m axial stresses, bending/dent re 1 ated local stresses) in se.rviceC 3)_
Based upon the chemical analysis, and the correlation between the defects and sensitization, we conclude that the observed corrosion can-be attributed to reduced sulfur, acidic pH, and the resulting stress-assisted IGA.
The proposed damage scenario is consis...,.tnt with known facts perti-nent to the observed* corrosion.
\\1*
4.3 Determination of Interorannular Attack GP'Owth Rate To define an operating all owarice, the __ J ~.osee _ _developed repair:
criteria for the IGA detected during the 1983 steam generator :-
inspection.
This allowance consists of a minimum of two factors eddy current accuracy and flaw growth rate.
By comparing eddy current signals of pulled tube samples_ and laboratory IGA samples, and to a lessor degree on numerous historical data comparisons of in-generato~ suspected IGA indications, the licensee established a period in which IGA is believed to have occurred.
Detail analysis and determination on IGA growth rate are described in Section 3.3 of the licensee 1s April 16, 1984 submittal.
- -. - ~-
13 -
The licensee attempted to establish a zero growth rate through the comparison of historical ECT data on IGA detected in 1983 with the sen~itive 4 x 4 F probe and bobbin _probe data collected prior to 1982.
The licensee concede~that ~he ability of the bobbin probe to q~antify IGA is poor.
Therefore; their evaluation was qualitative and their argument to establish a zero growth rate was not convincing.
In additio~, the staff felt that the occurrence of tube le~ks in 1982 and the metallographic examination finding of cracks and IGA depths of up to 100~~ throughwa 11 in sarnp"l9:s =removed in** 1983-84. -=~
=
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was not cons i s*tent with the 1 i censee' 5-Contention of no defect -;=
'1iir gro~th since 1974.
The staff was unconvinced by the licensee's-*
argument that since plant chemistry records show a small level of leakage that fluctuated between 0.01/gal/min and zero from June 1975, that the tube leak event that occurred in 1982 was due to the opening up of a pre-existing (since 1974) throughwall-crack caused by tube lockup loads imposed by denting which held the tube rigid against the support plates.
For the above.stated reason we could not accept the licensee's*
argument and rule out the possibility of the development of a*-
throughwa11 defect during the next operational period.
We have therefore requested, and the licensee has agreed to implement a more stringent primary to secondary leakage rate limit than the 0.3gp_m limit imposed by the Technical Specifiction.
By letter dated.*
May 15, 1984, the licensee agreed to reduce the allowable primary to
- secondary leakage limit to 0.1 gallon per minute.
These restrictions will be imposed either by a Standing Order or a change to plant operating procedures.
The leakage limit for periods uf startup and-major 1 oad changes wi 11 remain as current iy specijii ed i*n the Technical Specification.
We find this acceptable.
/
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/*
=
4.4 Actions to Prevent Recurrence
- :~. -~ _*_ -
Considering the nature of the* observed steam gen*erator tube degrada- ---
~-.z tion, the probable cause, and the rate of progression, the lic~nsee h~s controlled the exposure of.the tube to air or oxygen as the main measure to prevent recurrence of the observed corrosion.
As such no specific changes in current operating practice are necessary.
Con-tinued emphasis on reducing periods of air exposure of tubing and controlling dissolved oxygen contents of the operating or layup
.. fluids should be sufficient, since th~se measures hav~ been effec-tively employed prior to 1978, when the degradation was stiffled.*
- Techniques for removal of the stable non-corrosive sulfur (presently as sulfide) has not been commercially demonstrated.
Oxidation of the sulfur to sulf~te during desulfurization process would risk additional tube corrosion due £6 the formation of the polythionate
- anions which are most likely responsible for the observed corrosion.
Ba~ed on the above analysis, the staff agrees with the lifensee's decision not to desulfurize the secondary side, and concludes that reducing tb~ exposure of the tubes to air or oxygen provides rea-sonable assurance that the observed tube degradation will not recur.
4.5 Conclusions
- ~---.
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We conclude that the licensee has (1) identified the agent which caused the SIG tube degradation; (2) postulated a damage scenario which is consistent with the known facts pertinent to the observed corrosion; and (3) taken necessary actions to prevent recurrence of the obse~ved SIG tube degradation.
The criteria of Regulatory Guide 1.121 provides for an operational allowance over the minimum steam generator tube wall required to preclude tube burst, consisting of two factors, an eddy currerit
'I
- uncertainty and an allowance for possible flaw growth during the operational period.
The licensee has adopted-a plugging criteria of 51.% t:-hrough wall with the stipulation that all cracklike defects are.*
plugged regardless of depth.
As discussed in Section 5.2 of this safety evaluation, the staff finds that a defect which is 1 imited to 0.380 inches axial, 360° circumferential and 82 percent throug~-wall is acceptable.
This provides an operational allowance of 31%.
\\,'e conclude that this combined operational allowance for eddy current uncertainty and degradation growth rate.along with the more stringent primary t'2_ secondary leakage rate allowance of O.lgpm~
provides reasonable assurance that the P_ali~des Nuclear P1.ant.,;..£~n be operated without undue risk to the publichealth and safety. -=-:--
- =
ooc-
17 -
5.0 STEAM GENERATOR TUBE INTEGRITY EVALUATION The requirements for steam generator tube plugging have been calculated in accordance with the guidelines set forth in Regulatory Guide 1.121 entitled "Basis for Plugging Degraded PWR Steam Generator Tubes",
Reference 5. The basic requirements are summarized below.
- 1.
for normal plant operation, primary tube stresses are limited such that a margin, of 3 is provided against exceeding the ultimate
- tensile strength of the tube material, and the yield strength of the tube material is not exceeded.
2.-
'For accident conditions, the requirements of paragraph NB-3225 of
~:...
Secti-0n III of the ASME Code are to be met.
In addition, it must be demonstrated that applied loads are less than the. burst strength of the tubes at operating temperature as detennined by testing.
- 3.
For a11 design transientsJ the cumulative fatigue usage factor must be less than unity.
In addition, leak before break must be demonstrated i.e., through wall cr2cks with a specified leakage.limit during normal operation do not propagate and resu1 t in tube rupture during, postulated accident con-di ti6ns.
~~~-
For the evaluation of the plugging limit for the Palisades tubes the following information was used:
Tubing Geometry:
Material Properties:
Operating Conditions:
0.750 11 Diameter 0.048 11 Thick.
Sy = 27.9 Ksi at 600°F SU = 80.8 Ksi at 600°F Normal: Pi = 2150 psi P ~7TDpsi 0 _:'"
A- ~1380 psi p "
-~--
--~--
Steam Line-Break (SLB).
jp = 2150 psi Loss of Coolant Accident (LOCA)
Maximum external pressure, zero internal pressure The demonstration of.tube integrity for unifonn thinning of up to 64 percent is contained in Reference 6.
For the purpose of establishing th~ tube integi::ity in bending for other than uniform thinning, a
19 comparison of the structural capabilities relative to unifol'iTl thinning,*
was used as the basis.
Since Regulatory Guide 1.121 const1tutes an operating rather than a
-aeS-ign requirement, the allowable stress limits are based on expected 1 owe r bound actua 1 ma teri a 1 properti es--Ca 0
s~Opposed to the Code-spec:-; fi ed mini~um values).
Expected strength properties were obtained from analyses of tensile test ~ata of actual *production tubing.
These calculations were performed to support the unifonn thinning plu~gi,ng margin evaluation reported by Reference 6.
The degradation exper-i enced in the Pa 1 i sades* --s'EEQm= genera tors is l"'imi t~=
E --
in both axial and azimuthal extent.
The siie o"ra-ctual defects removeo=
.:.~-
from the genera tor ex hi bi ts axi a 1 1 engths of*.-~'oxima.te ly O. 050 11 and
- ai~muthal lengths of about 90 to 100 degrees.
This range of defects can be conservatively bounded by considering the axial extent to be 0.075~.
and the azimuthal extent to be 135° (which ~onservatively envelopes the Eddy current testing (ET) limits of detectability of 130° for azimuthal tube degradation) for evaluation purposes.
Defects limited to the extent specified c6~1d be analyzed as cracks since Inconel 600 is a ductile material and the crack tip plastic zone could be on the order**of the axial extent dimension.
For a tight circumferential crack, crac~
tip blunting to the axial extent of Palisades defects would occur before crack extension.
For defects 1.onger than the 0.075 11 considered to be bounding, the burst behavior of the tubes may or may not be adequately
. 1'
- 20 described by treating the defects ~s crack-like.
In addition, since ET techniques are currently inadequate ~o accurately describe defect axial lengths below about 0.200 11
, it is considered necessary to evaluate the Palisades degradation without relying-on specifically characterizing the defects as having crack-like behavior:
The usual evaluation procedure ereployed to assess the effects of tube degradation assumes that-the degradation extent is ~round the full circumference of the tube and is unlim)ted in the axial direction,,(>2.0 in.).: Thus, the analysis is based.on evaluating a uniform tube ¥.1ith a thickness equal to the minimum remainina thickness of the dearaded tuD"e:""
..,I
'=-
~
This has, in the. past"~ been the usual approach oased on the expediency~
- t"-
of performi~g the analysis and the lack of a _significant data base which
~---
could be used to quantify the beneficial effeetof the degradation being
,.of limited extent.
Limiti.ng the extent of de~radation has the effect of also limiting the stresses in the material in the degraded area.
For degradation of limited axial extent, the magnitude of the hoop stress in the degraded area is restricted by the adjacent thicker material.
For deg*radation of limited azimuthal extent, the adjacent material restricts the magnitude of the axial stress. *The reinforcing effects of adjacent undegraded tube material can be *quantified using available testing inform2tion on tu~es with limited extent degradation, coupled with lower ~ound type collapse evaluations.
- i
- 21 Effects of depth of thinning along with axial and circumferential extents-of thinning on the burst strength of Incone1_600 tubing have been represented on a plot of axial extent of thinning versus depth of thinning by a family of curves whtch present ~he loci of all geometries
-h"a*ving _a given burst pressure.
For tubes thinned completely around the circu;;;ference (360°) over a certain axi--a-l~l"ength, an empirica*l equation from Reference 4 was used to calculate bursi pressure.
Results from limited axial extent, unifonn circumferential thinning
-~
burst tests are also reported in_ Reference 7.
In addition, an empirical equation was developed for relating the ratio e£ ttie burst pressure &.
=
the degraded tube to -the burst pres sure for: tJiE""Undegraded tube, the ;-
remaining strength fraction (RSF)-, to the rernanfing wall fraction (RWF~
--~
of the tube.
A comparison of this empirical ~ation with test data in~i~ates that for a RWF less than a specified value the burst pressure is overpredicted.
For normal tubes, and those in which the thinning is relatively long, the-mode of burst failure is characterized by the opening of an axially oriented split due to the hoop: stress being the maximum primary stress in the tube.
For degradation of limited extent, the hoop stress in the degraded area is restricted by the amount of hoop defonnation taking_
place in the undegraded region tif the tube.
The axial stress is howev-er, mainly dependent on ~he remaining thickness.
For small axial extent some notch strengthening will take place, but this will be of minor
22 -
significance.
For large depths and small axial extent the axial -stress will become larger than the hoop stress and the mode_of failure will be a circumferential separation rather than axial.
This, in effect, limits the amount of strengthening which can* be realized to a factor of 2, the ratio of the magnitude of hoop to axial stress in a tube of unifonn thickness.
The licensee has developed relations to predict the strengthening due this reinforcement that can be achieved by the unde-graded mat~rial.
In ord~r to validate the use of t~ese relations for application to Palisades, the data base in Reference 7 was e*xpanded to inc1Lide Westinghouse data for a.variety of tube sizes and data specific to the heats of material used for the Palisades=-tuoes.
lt was determined that *the Reference 7 results are- 'i!i =general conservative wim
-~-'
respect to the added Westinghouse* and Palisa.des;:data.. On this basis it':"
---*~--
was judged that thes-e relations are applicable--to Palisades and could be used to quantify the effect of limited axial extent on the burst pressure.
In order to evaluate actual tube burst performance during SLB, a toler-an~~ line with a 95 percent probability of forming a lower bound for 95 percent of the popu1ation underlying the Reference 7 data was* developed.
The final form of a plugging ljmit relation for tubes with limited axial extent was developed to satisfy the structural requirements of Regulate~
ry Guide 1.121~ The requirement for demonstrating adequacy using a factor of safety of 3 relative to the ultimate tensile strength is more
I*
- 1*
23 restrictive than the primary stress requirements relative to yield strength and accident condition loads.
For the yield s*trength requirement-of Regulatory Guide 1.121 it is noted that:*--.
SU = 2.87 Sy so sy = su;2.s7.>-su/3.o Thus, limiting primary membrane stress, Pm, based on a factor o~ 3, r.elat1ve to ultimate is *conservative as compared to limiting Pm to les*s than the yield stress.
~---
~
Si mil a r 1 y, the pressure differential during stearn--lein break (SLB) is
~
.....;:.-~-=
relatd to the normal operating pressure as:* - :--
The stress limit during SLB is the less~r of 2.4 Sm or 0.7 Su where Sm is*found as the lesser_ of 2 Sy/3 or Su/3 at temperature.
Using the Palisades tube properties, Pm
~s limited as given by the following equation:
P ~ 44.6 ksi =* 1.67 (Su/3) m
~ ---
24 -
Since the allowable stress is 67 percent larger than that during normal operation and the loading is only 56 percent larger, ~he requirement against the ultimate tensile strength is more restrictive.
For unlimited th.inning, it is shown in Reference E that the amount of uniform degradation which can be accommodated is 64 percent.
For limited ax~a1 extent thinning the wall thickness required to. provide the same margin against circu~ferentia1 burst is 0.0085, for an allowable degradation maximum of 82 percent.
In order to account for the rein-forci::ig effect for flaws of limited azimuthal extent, information was proviced considering the effect on burst presstfce of circumferential ~
cracks (representing zero axial extent), the -eff=ect of burst pressure~
axial pa rt through wa 11 and through wa 11 cra~ks ::Trepresenti ng ZE:ro azimuthal "extent), and 90° and 180° upper boum:Flimit solutions for
.rect~ngul~r patch type degradation (finite but limited axial and aximutha1 dimensions).
The evaluation for limited circumferential cracking relative to burst pressure is presented in the leak before break evaluation section of Reference 4.
Considering the pressure differential of 4140 psi, burst would be expected for a circumferential crack with an included angle of 145°.
Therefore, for cracks with less than an included angle of 145°, a margin of 3 against burst at normal operaing pressures ex~sts relative to actual test data.
The plugging limit developed for 360° thinning can be considered equally applicable for ~hinning limited in -azimuthal extent to 145°.
~-:_
-~.: -
\\
.I Analyti~al models have been developed to compute the burst strength of Inconel 600 tubes with thinned areas of limited axiat and circumferen-tial extent.
Effects of geometry on burst strength have been presented in terms of plots of axial extent of ihinnin~ and depth of thinning resulting in a given burst pressure.
Effects of the extent of cir-curnferentia1 thinning on burst strength have been represented by a number of *strength 1 ocus curves corresponding to different arc 1
.. engths of circumferential thinning. On the ba5is of these evaluations it has been demonstrated.Jhat a tube with through w~ll degradation up to D.075 inch~s in axial extent and up* to-145° in azimuthal extent is sufficient to withstand normal operating pressure with a fcrctor of safety of 3.
An additional requirement from Regulatory Gui_de 1.121 relative to burst*
--~-
at steaml ine break, i.e., beyond the evaluatiOii'"" of primary membrane stress, is that margin be provided against the ultimate burst pressure as determined by burst tests perfonned at operating temperature.
For this evaluation the lower tolerance limit line from the Reference 4 data was used to determine axial burst values expected for steamline break co~ditions. Since the lower tolerance limit is adequately below all of the additional data plotted and the original data, it provides a suffi-cient basis for evaluating axial burst pressures.
In addition, a lower tolerance value for the undegr?ded ~urst pressure was used based on
~
results reported in References 8 and 9.
This evaluation demonstrates that axial burst pressur~s for the limited axial extent plugging
criteria for steamline break are above the criteria established using a safety factor of 3 based on ultimate tensile strength.
Beyond the evaluation requirements on* primary membrane stresses and the steamline break ultimate test pressures, Regulatory Guide 1.121 also requires that degraded tubes be evaluated in accordance with the re-quirements of the ASME Code Section III, Paragraph NB~3225.. This paragraph of the ASME Code invokes the rules of Appendix F of the Code for the evaluation of faulted condition limits.
In genera 1, th_~r~ a re no primary bending stresses at degraded tube locations. This is either due to:
(l) the ab iii ty to with stand bending ~men ts is not necessa ~
~
~
~--:--
to satisfy requirements for equilibrium, or'°<-2ff°defonnations under
~
accident conditions are limited by either the anti-vibration devices or.
.the general deformation of the overall tube *tnrrrdle, thus d~fonnation of*
the degraded tube is displacement controlled.
However~ a comparison was made for the limited axial extent plugging margin criteria with the uniform thinning plugging margin criteria.
In this case the ultimate limit moments were calculated for each degradation condition.
As a li~iting condition for the limited axial extent degradation, a 135° through wall crack was considered. This crack represents the maximum expected azimuthal extent in the Palisades steam generators.
For this condition the ultimate bending moment for the limited axial extent plugging criteria is 12 percent higher than the limiting ultimate moment
~er a uniform..:hinning criteria of 64 percent.
In addition, a compari~
son of the section properties was made for degradation extending to 135°
- 27 and 82 percent throughwall.
In this case both the moment of inertia an~
the section modulus for the limited axial and azimut~al extent plugging margin criteria is greater than that for a tube degraded uniformly to 64 percent through wall.
--~=~-.:;.
Consideration of external collapse per the Code requirements can be made bas~d on collapse data presented in R~ference' 7 The ~eference document demonstrates that for 0.875 inch diameter by 0.050 i~ch thick tubes, which have a co 11 ~-p_se strength 1 ess than 84 percent of that for,. t~e Palisades 0.750 inch diameter by- 0.048 inch thick tubes, that the collcpse pressure is relatively unaffected for=:t:_niTonn thinning 3/8 ~
an inch 1 ong by 360° *azir.iutha l ly for degradatri ;;n-u~ to 60 percent
~
F-through wal 1.
For degradation at 80 percent--th;rough wal 1, the col lapse-:-
--=-- --=~
pressure was demonstrated to be t to 1 times* ""t:1i"e-undefected tube col- -
la.pse pressure.
For all collapse tests performed, which included degradation up to 80 percent of the wall and 1.5 inches in axia1 length, the collapse pressures found were well in excess of the external secon-dary pressure during LOCA.
In addition.the limited degraded tube is ju~ged to have more resistance to collapse than a tube which has been uniformly degraded *with unlimited aximuthal and axial extent up to 64 percent through wall.
In order to examine the fatigue resistance for the limited extent plugging margin criteria, the evaluation was made of the tolerable stress concentration factor for a tube with no degradation, but
28 -
including the effects of denting (to consider the bounding effect of a tube locked-in at the tube support plates), vibration and thenna1 stresses. It was found that a maximum allowable stress concentration factor (SCF) of 5.9 could be tolerated-to re~ult in a 40 year design
- ~..
- v~
life usage factor of 1.0.
For degradation of limited axial extent,
__.-excluding cracks, the maximum stress concentration factor required by the ASME Code is 5.0.
. for_ the case of th__rough wall cracking,, it is noted that there are,very few cycles of significant stress~levels. This results in low alternat-ing fracture stress intensity values.
The alte~ating fracture stres5-_
.i;;;.
~
.,;!;.r-:;-
a:::
=
~
- intensity is directly.. related to the alternati'fe stress perpendicular te.
~
--=---
the flanks of the crack multiplied by the square root of pi times i the -
---~--
crack* length, times a function of the crack length to thickness ratio.
I
.. For the case of the Palisades tubes, t_he alternating stress is judge.d __ to be of the order of 7 ksi.
For crack lengths which were on the order of twice th~ wall thickness,- which would be the case for degradation with stress corrosion cracks, the alternating stress intensity factor would be less than the threshold required for crack growth.
In addition, very deep cracks would be needed to'plastically yield the remaining ligament of the tube.
Crack growth would be on the order of the cyclic crack tip displacement,. in which case several mils of growth would represent a generous fatigue cratk growth allowance.
29 -
In sumrr~ry the Regulatory Guide 1.121 structural requirement for normal operating conditions (maintenance or a factor of safety of 3 against burst) and accident conditions (compliance with pararaph NB-3225 of Section III of the ASME Code) have been met by the tube with 360° thinning of 82 percent with axial extent limited to 0.380 inches.
Such a defect also satisfies the fatigue usage requirements for all applicable* design transients.
5.1 Leak Before Break Evaluations
~
Two separate approaches were considered applicatte to the Palisades
. *defec~s to demonstrat~ leak before break.
on. the observed morphology that indicates that the defects could be evalu*ated _as cracks.
The second is based on Considerations of finite axial extent in accordance with the range of the structural criteria discussed earlier.
These evaluations demonstrate that through wall cracking would develop prior to the crack reaching the critical circumferential length for a postulated SLB event.
The growth of part through wall cracks in tubes at Palisades have been observed to exhibit a limit~d aspect ratio which results in extension through the wall prior to reaching the critical SLB bursting 1 ength.
Furth~ rmore, based on geometrical considerations, circumferential crack extension*beyond 60° will lead to the axial bending stresses being a maximum at the crack front thus encouraging preferential growth through the wall.
Evaluation of leak before break
~.
for degradation with finite axial extent was perfonned considering the limited extent upper bound of 82 percent to be unifonn thinning extend-ing 360° circumferentially and of unlimited axial extent.
Final infer-mation is provided in the form of a pair of curves, one for cracks
%r:*
leaking* at the technical specification limit and the other for burst a SLB.
THe leak before break margin can be read as the distance between the curves as a function of crack extent and degradation depth.
Based on these analytical evaluations the licensee has demonstrated that through wall crack_jng would develop prior to the crack reaching,th,e length for a postulated SLB event.
=
5.2 Conclusions*.
~ =
~
~ ~,~
Based* en a review of the analyses and test dafa provided by the 1 i censee
.. in Section 4 of Reference 4 the staff concludes:
(1) The structural limit of 82 percent fer defects which are limited in axial extent to 0.075 and up to 135° in azimuthal extent is sufficient to meet the nonnal operating conditions, accident conditions and fatigue usage requ1rements outlined in Regulatory Guide 1.121. This includes the
~aintenance of a safety factor 'of 3 relative to ultimate strength which is the most restrictive primary membrane stress limit.
(2) Leak before break has been demonstrated for degradation of extent limited such that it can be considered to behave as a crack.
(3) A tube with 360°
- hinning of B2~percent with axial extent limited to 0.380 i~ches, has a safety factor against burst comparable to the margin afforded pressure vessels designed in accordance with Section III, of the ASME Band PV Code.
(4) Leak before break of circumferential cracks has been
I I
I
.\\... -'
31
~
demonstrated for uniform thinning degradation, extending 360° circumferentially and of unlimited axial extent, to a depth of 69 percent of the.tube thickness.
(5) The structural limit for uniformly thinned tubes with unlimited axiai and circumferential extents is 64 percent thinning.
In accordan~e with these ~onclusions, a defect which is limited to 0.380 inches axial, 360° circumferential and 82 pers;ervt through-wall is.
./
acceptable.
Therefore;* as discussed in Section 4.5 of this report, th~ threshold for corrective action was determined by reducirt£L tri.e 82 percent throug,b,,;J1al 1 by (1) an allowance for eddy current uncertainty., *-and (2) an allowance fot.R.ossible
~:.
~--
... flow growth during th-e oper2tional period.
=....:- -
The staff finds the -repair criterion of 51% tilr_~n-wa11 acceptable for-Palisades steam generators.
6.0 SUPPORT PLATE DEGRADATION CONSIDERATIONS Evidence of.denting related tube-to-tube support plate (TSP) interaction was observed during tube pulling and support plate removal operations at Palisades during the present outage.
In certain instances large forces were required to remove tube specimens.
Steam generator tube denting results in tubes being locked into tube support plates. The configura-tion of the TSP/tube bundle and corrosion mechanism leading to denting results in the development of compressive forces on the tubes and expansion of the tube support plate. This interaction can eventually lead to significant distortion of the TSP and cracking at areas 6f high stress concentration (e.g., TSP flow hples and slots).
.u I
( Secondary side vi SU?-1 inspection of TSPs 14, 13, 12 and portions of-11 at the accessible periphery and in. locations where TSP sections have been removed in the Palisades steam generators revea1ed no significant tube support plate degradation as a result of tube denting.
In addi-tion, ET dat~ indicates denting growt~ rate h~s decreased in recent years.
However, for conservatism, the structural analyses were
- er7ormed to provide the basis for the determination that Palisades steam generator tube bundle int~grity is maintained even if a single su~port plate is postulated to be missing, i.e. to lose its load carrying capability or fail to provide support for a tube.
A primary load capability evaluation of the Pabisa-Oes steam generator~~
=-:
tube bundle under an assumed genera 1 degradP:!j ~was conducted using postulated SLB, LOCA and SSE loading.
~
Tubin~ c0-l}apse potential and tne resulting 1 ass of tube fl ow area during a postiES.ted lOCA and SSE event were analyzed.
A detailed evaluati-0n of the Palisades steam generator tubes in terms of responses to such forcing functions as turbulence and vortex shedding mechanisms and fluid elastic stability criteria was completed (i.e., a normal operation fluid-structure interaction ev~Juation).
Fatigue effects resulting from the flow induced motions were aiso evaluated; Worst ca~e vertical and horizontal tube_ spans were analyzed with a single tube support plate missing.
Finally, tube wear-.
estimates were calculated assuming a postulated TSP fragment and partial TS? support.
- 3::.
,;1 33 6.1 cor~CLUS IONS Based on review of the analyses presented by the license relative to the postulated loss-of a single support p]ate, the potential tube flow area los_s~~due. to tube collapse during a LOCA and concurrent SSE, tube v-i-bration and postulated loose plate piece~-th~~~_taff concludes:
- The postulate~ loss of a single support plate does not adversely affect the tube bundle. integrity during LOCA, SSE or SLB accident
. ~ ***.*
condition loadings.
The potential tube flow area loss due to tubi= cQ_llapse during LOCt?.
and concurrent SSE_ is equivalent to less th*aIL._260 tubes (of which 63 are currently plugged) with.the postul.ateq_J_os*s of load carrying -
capability of a sjngle support plate.
~:-.
Tube vibration during normal operation should not be significant f~o~ flow-induced vibration mechanisms considering the postulated loss of a single support plate.
.Neither tube vibration nor postulated loose plate pieces lead to premature wear-through of a tube.
7.0 ACKNOWLEDGEMENT H. Conrad, P. Wu, and J. Rajan. prepared this report.
Dated: June 11, 1984
References f. ""
': /'.
- l.
Dravnieks A. and Samans C. H.,
11 Corrosion Control in Ultra Forming, 11 American Petroleum Institute, 37 (III), page 100, 1975.
- 2.
Samans, C._ H.
11 Stress Corrosion Cracking Susceptibility of Stainless Steels and Nickel Base Alloys in Polythionic Acids and Acid Copper Sulfate Solution, 11 Corrosion, 21, page 256, 1964.
- 3.
.!:.hmad, S. et al.,
11 Str~ss Corrosion Cracking of Sensitized 304 Austenitic Stainless Steel in Petroleum Refinery Environment, 11 Corrosion, Vol. 38, No.
6~ page 347~ 1982.
- 4.
Consumers Power Company Report "-1983/1984 Stearn Generator Evaluation and Repair Program, 11 April 1984. ~
USNRC Regulatory Guide 1.121, 11 Bas*es for ~.lugging De-graded PWR
~
Steam Generator tubes, 11 Issued for Corrrnent:-August 1976.
- 6.
Combustion Engineering Report, 11 Analysis to Determine Allowable Tube.Wall Degradation *for Palisades Steam Generators, 11 P. Anderson et al., Rev. 2, March 30, 1976.
- 7.
USNRC NUP.EG/CR-0718, 11 Steam* Generator Tube Integrity Phase 1
- Report, 11 J. Alzheimer et al., September 1979.
- 8.
CE Report CENC-1256, 11 Tube Burst and Leakage Test (Palisades),"
.~. K. Hayes-; February 1976.
'd.
Bc:ttelle (Colurr.bus) Topical Report, 11 Examination of Iconel 600 Tubing from the A Steam Generator of the Palisades Nuclear Plant, 11 1.. 1.,,,
107/1
--