ML20141E613
| ML20141E613 | |
| Person / Time | |
|---|---|
| Site: | Oyster Creek |
| Issue date: | 01/02/1986 |
| From: | Donohew J Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8601080226 | |
| Download: ML20141E613 (286) | |
Text
{{#Wiki_filter:. ._. . . _. - ,_ . . - _ _ 1 ( UNITED STATES 8 o NUCLEAR RE'GULATORY COMMISSION r, j WASHINGTON, D. C. 20555 S f
\ # January 2,1986 l Docket No. 50-219 i
i l LICENSEES: GPU Nuclear Corporation
! Jersey Central Power and Light Company 3
FACILITY: Oyster Creek Nuclear Generating Station i .
SUBJECT:
DECEMBER 13, 1985 MEETING WITH GPU NUCLEAR CORPORATION (GPUN) TO l DISCUSS' DEFERMENT OF FEEDWATER N0ZZLE INSPECTIONS FP0M CYCLE 11R l OUTAGE 1 On Friday, December 13, 1985, a meeting was held at NRC, Bethesda, Maryland with GPU Nuclear (GPUN).-the licensee, on.the licensee's requested deferment ]! of feedwater nozzle inspections from Cycle 11 Refueling (Cycle 11R) outage to the Oyster Creek Cycle 12R Outage. The Cycle 11R outage is scheduled to. > I begin in April'1986. ' Attachment 1 is the list of individuals that attended the meeting. Attachment 2 is the material handed out by the licensee in the meeting for its presentation.- Attachment 3 is the material handed by the licensee
- ! to the staff to answer the staff's questions. This is-discussed further in these meeting minutes. The following is a sumary of the significant items -
l discussed and the actions, if-any, taken or proposed. i
! The feedwater nozzle inspections came out of the resolution of the NRC' staff Generic Technical. Activity (GTA)~ A-10, BWR Feedwater Nozzle and Control Rod !
{ Drive Return Line Nozzle Cracking. By 1980, inspections at 22 of the 23
; BWRs in the United States that have feedwater nozzle /sparger systems disclosed . some degree of cracking in the feedwater nozzles of-the reactor vessels at i
18 plants. NUREG-0619,' November 1980, sumarizes the work performed by NRC to resolve GTA A-10. Table 1 of the NUREG sumarizes the BWR feedwater - nozzle cracking problems and solutions by plant and Table 2 lists the routine inspection intervals for different configurations. The recommended inspection
- interval for Oyster Creek in Table is 2 refueling cycles for ultrasonic for the feedwater nozzles and 4 cycles for visual inspection of the sparger.
By letter dated November 20, 1985, the licensee provided technical justification for its requested deferment of the feedwater nozzle inspections. This defer-
- ment is one of several which the licensee has requested for the staff, by l letter dated July 26, 1985, to consider deferring from the Cycle 11R outage i becau*e of the large amount of work scheduled to be completed.in this rutage.
l This meeting'was requested by the licensee to discuss its requested deferment
- of feedwater nozzle inspections with the staff. The licensee began the meeting with a chronology and sumary of correspondence regarding NUREG-0619.
i (SeeAttachment2.) The-licensee stated that inspections of the feedwater nozzles were previously done in the Cycle 7R outage. The licensee removed the stainless steel clad
- and some base metal from the feedwater nozzles in the outage and installed an improved feedwater sparger design. It is comitted to inspections of the
. control rod drive (CRD) return line' nozzle'in the Cycle 13R outage. There have been no ultrasonic (UT) inspections of the feedwater nozzles. The i B601080226 860102
- PDR ADOCK OJOOO219 G PDR-
cleanup system piping rerouting and feedwater low flow control modifications were reviewed and approved by the staff and the work was done in the Cycle 10R outage. The UT inspections to meet Table 2, Routine Inspections Intervals, in NUREG-0619 were previously deferred from the Cycle 9R outage to the Cycle 11R outage. The licensee's contractor, MPR Associates, made a presentation. Its handout is also in Attachment 2. This material covered part of the licensee's presentation discussed above. The staff requested what finite element code was used. The contractor stated the code used was ANSYS. The contractor's work on the estimated maximum crack to be expected in the feedwater nozzle by Cycle 9R outage, Cycle 11R outage and Cycle 12R outage was based on the following: (1) the reference BWR duty cycle and (2) the best estimate crack growth. The difference in the feedwater nozzle crack propagation between the reference BWR duty cycle and the Oyster Creek. specific duty cycle is shown in figure 10 of the licensee's TDR No. 497, OCNGS RPV FW Nozzles: Structural Integrity After 1977 Repairs, November 17, 1983. This TDR is in . The reference BWR duty cycle is more conservative than the Oyster Creek specific duty cycle. The technical conclusion of the licensee is that the maximum crack in the feedwater nozzles should be no more than 0.107 inch. This is based on the material in Attachment 3. The licensee stated that it would be better to wait to the Cycle 12R outage to allow better UT inspection techniques to be developed to avoid an unnecessary removal of the feedwater nozzles and sparger from the reactor vessel. The licensee is currently developing UT techniques to provide the maximum sensitivity in the detection of flaws which are of concern to nozzle integrity. It is using a mockup to develop these techniques. The staff requested the following data from the licensee: (1) piston ring seal leakage test and full size two temperature test referred to on page 2-5 of the licensee's November 20, 1985, submittal; (2) high cycle fatigue evaluation referred to on page 3-3 of its submittal; (3) thermal analysis and finite analysis described on page 3-7 of its submittal and (4) calculations of heat transfer coefficients for the inside surface of the feedwater nozzles referred to on page 3-7 of its submittal. This data was provided by the licensee and is in Attachment 3. Original signed by John A. Zwolinski for' Jack N. Donohew, Jr. , Project Manager BWR Project Directorate #1 Division of BWR Licensing Attachments: DISTRIBUTION
- 1. List of Attendees Docket BGrimes, w/o attachments 2/3
- 2. Licensee's handout NRC PDR ACPS (10)
- 3. Licensee's answer's Local PDR NRC Participants to staff's questions PD#1 Rdg JZwolinski, w/o attachments 2/3 cc: R. Bernero JDonohew G. Lainas CJamerson, w/o attachments 2/3 B. D. Liaw OELD, w/o attachments 2/3 EJordan, w/o attachments 2/3 L:PD#1 DBL:PD#1 Donohew:jg JZwolinski i /P/86 \ / v/86
I cc: G. F. Trowbridge, Esquire Resident Inspector Shaw, Pittman, Potts and Trowbridge c/o U.S. NRC 1800 M Street, N.W. Post Office Box 445 i Washington, D.C. 20036 Forked River, New Jersey 08731 J.B. Liberman, Esquire Commissioner Bishop, Liberman, Cook, et al. New Jersey Department of Energy 1155 Avenue of the Americas 101 Commerce Street New York, New York 10036- Newark, New Jersey 07102 i Eugene Fisher, Assistant Director l Regional Administrator, Region I Division of Environmental Quality
! U.S. Nuclear Regulatory Commission Department of Environmental 631 Park Avenue . Protection King of Prussia, Pennsylvania 19406 380 Scotch Road i Trenton, New Jersey 08628 l
BWR Licensing Manager GPU Nuclear
!' '100 Interpace Parkway Parsippany, New Jersey 07054 Deputy Attorney General State of New. Jersey l Department of Law and Public Safety 36 West State Street - CN 112 Trenton, New Jersey 08625 ! Mayor I
Lacey Township { 818 West Lacey Road j Forked River, New Jersey 08731 D. G. Holland Licensing Manager Oyster Creek Nuclear Generating Station
; Post Office Box 388 Forked River, New Jersey 08731-P. B. Fiedler Vice F.esident & Director Oyster Creek Nuclear Generating Station Post Office Box 388 Forked River, New Jersey 08731 e
ATTACHMENT 1 MEETING WITH GPU NUCLEAR CORPORATION (GPUN) TO DISCUSS DEFERMENT OF FEEDWATER N0ZZLE INSPECTIONS FROM THE CYCLE 11 REFUELING OUTAGE (DECEMBER 13,1985) NAME ORGANIZATION J. Donohew NRC/NRR J. Tsao NRC/NRR W. Hazelton NRC/NRR D. Holland GPUN S. Giacobbe- GPUN S. Leshnoff GPU:1 J. Johnson MPR Associates K. Tosch State of New Jersey *
- Department of Environmental Protection .
ATTACHMENT 2 CHRONOLOGY AND
SUMMARY
OF REGULATORY CORRESPONDENCE REGARDING NUREG 0619 April 9,1981 to NRC Explained that feedwater spargers were replaced witn an approved type during 7R. Described station procedures as being designed to limit injection of unheated feedwater. Also described non-automatic nature of system satisfactory in preventing " hunting" and minimizing thermal cycling. Committed to 0619 Table 2 inspection frequency. May 22,1981 to NRC Presented arguments against need for modifications relative to feedwater control system and cleanup system rerouting. i fuLIaturt July 20,1981 to NRC to GPUN otr NRC responds to GPUN regarding unresolved safety issue (USI) A-10 "8WR Wlb ~ Feedwater Nozzle Crackiag". The staff concurred thatQ modifications Ghdmoy? to the feedwater flow control system or cleanup system are required. 6O Requested GPUN to commit to CRD Return Line Nozzle inspection requirements of NUREG 0619.
~
b.T/ith1_ August 25,1981 to NRC P. R. Clark comits to CRD return line nozzle inspection methods and frequency of 8.2(5) of 0619 and that the PT examination will be performed at or before 13R.
. March 19,1982 NRC to GPUN NRC acknowledges that CR0 Return Line Nozzle Inspection-will occur 13R or earlier.
November 29,1982 to NRC Explained changes in schedule of 0619 inspections: a) postpone VT from 10R to llR b) move up PT from 13R to llR c) move up VT from llR to 10R l l
In addition, it was explained that.VT results of sparger inspections during 8R and 9R were satisfactory and that new design is working well. November 20,1985 to NRC Explained desire to postpone VT. examination until 12R with internal PT and VT examinations to be performed if defects are found through UT methods. 13R is the earliest that VT and PT would be required by 0619
~
otherwise. GPUN presently plans to examine the CRD return line nozzle during 13R as was originally committed and as described in 0619. (0664A) 1 i
, .- -- - - , , - . , - . , - e .ey, ,
f GPU NUCLEAR CORPORATION OYSTER CREEK NUCLEAR GENERATING STATION JUSTIFICATION TO DEFER ULTRASONIC EXAMINATION OF FEEDWATER N0ZZLES'ONE ADDITIONAL FUEL CYCLE . DECEMBER 13, 1985 h
+
k i j e
l
SUMMARY
OF TECHNICAL JUSTIFICATION BASED ON RESULTS OF TESTS AND ANALYSES: . O UNDETECTED FLAWS IN FEEDWATER N0ZZLE WILL NOT GROW TO UNACCEPTABLE SIZE IN ONE ADDITIONAL FUEL CYCLE. 0 UT EXAMINATION CAN BE DEFERRED ONE ADDITIONAL FUEL CYCLE WITHOUT JEOPARDIZING INTEGRITY OF FEEDWATER N0ZZLE. , O O
. , , . , _ , . . - - , . . . . , - ,- . ,-- ,y- . - + - -,-- -- ---,,
DYSTER CREEK FEEDWATER N0ZZLE CRACKING NO. OF INDICATIONS BEFORE CLAD AFTER CLAD MAX DEPTH 1/ N0ZZLE REMOVAL REMOVAL (INCH) 450 36 5 0.54 1350 3 0 - 2250 4 4 0.46
. 3150 11 3 0.62
- 1. FROM ORIGINAL CLAD SURFACE.
3
- l i
i 4 I
CAUSE OF FEEDWATER N0ZZLE CRACKS PRIMARY CAUSE - THERMAL FATIG!IE O CRACK INITIATION - HIGH CYCLE FATIGUE TURBULENT MIXING OF ' COLD' FEEDWATER BYPASS LEAKAGE WITH ' HOT' REACTOR RECIRCULATION WATER a T (METAL) = 50% AT (FLUID) FREQUENCY 0.1 - 1 HZ CRACK INITIATION IN 3 MONTHS TO 3 YEARS CRACK ARREST AT 0.25 IN 0 CRACK GROWTH - LOW CYCLE FATIGUE i -- PRESSURE CYCLES - HYDRO, STARTUP, SHUTDOWN THERMAL CYCLES DUE TO ON/0FF CYCLING OF . FEEDWATER FLOW DURING LOW FLOW CONDITIONS, I.E., STARTUPS, SHUTDOWNS, HOT STANDBY
, l l
)
. _ ., I
PREVIOUS ACTIONS TAKEN AT OYSTER CREEK TO REDUCE FEEDWATER N0ZZLE CRACKING REMOVE STAINLESS STEEL CLADDING (AND THIN LAYER OF BASE METAL) 1977 REFUELING OUTAGE O REDUCE THERMAL STRESSES 0 ELIMINATE METAL AT HIGH STRESS REGIONS WHICH MAY HAVE SUFFERED FATIGUE DAMAGE O IMPROVE UT SENSITIVITY REMOVE DEFECTS 0 0 LOCALGRINDINGWITHPTINSPEkCTION g h BLENDCAVITYINTOFEEDWATERN0ZZLETO INSTALL REPLACEMENT FEEDWATER SPARGER OF IMPROVED DESIGN (FIG 1 AND FIG 2) 0 SINGLE PISTON RING AT THERMAL SLEEVE TO REDUCE BYPASS LEAKAGE. MAXIMUM LEAKAGE = 0.3 GPM. O DOUBLE FLOW BAFFLES (WITH LEAK 0FFS) TO PREVENT THERMAL MIXING. MAX aT (METAL) = 12-17% 6 T (FLUID). ( __ - _ _ _ _ .-. _ . . _ _ _ _ _ ___ _ __
TECHNICAL JUSTIFICATION N0 NEW CRACK INITIATION BY HIGH CYCLE FATIGUE MECHANISM 0 HIGH CYCLE FATIGUE EVALUATION METHOD IN ACCORDANCE WITH NEDE-21821 (ORDERED l OVERALL RANGE CYCLE COUNTING APPROACH) METAL TEMPERATURE CYCLING (FIG 3) TIME / TEMPERATURE / FLOW MAP (TABLE 1) RESULTS GAP AT aT (METAL) FATIGUE USAGE FLOW BAFFLE (%) PER YEAR IN 30 YEARS ZERO 12 5.18E-5 0.002 0.02 17 '4.55E-3 0.137 fV Y NMP-1 EXPERIENCE - PT INSPECTION OF ONE N0ZZLE INDICATED NO NEW CRACK INITIATION AFTER 4 YEARS OPERATION RESULTS VALID FOR BYPASS LEAKAGE UP TO 15 GPM. FACTOR 50 TIMES EXPECTED, Yd O INITIAL FLAW SIZE 0.062-IN - MAX FLAW THAT MAY HAVE BEEN LEFT IN FEEDWATER N0ZZLE DURING 1977 REPAIRS BASED ON PT INSPECTION CRITERIA
TECHNICAL JUSTlFICATION (CONTINUED) SLOW CRACK GROWTH RATE O FRACTURE MECHANICS ANALYSIS METHOD IN ACCORDANCE WITH NEDE-21821 INITIAL FLAW SIZE 0.062 IN REFERENCE BWR DUTY CYCLE 130 STARTUP/ SHUTDOWN (FIG 4) 349 SCRAMS TO LP HOT STANDBY-(FIG 5) 62 SCRAMS TO HP HOT STANDBY (FIG 6) OYSTER CREEK PLANT SPECIFIC DUTY CYCLE. DIFFERS MAINLY IN NO. THERMAL SHOCKS. STARTUP/ SHUTDOWN (FIG 7) 9 vs 12 SCRAM (FIG 8) 12 vs 72 (LP) AND 12 (HP) BEST ESTIMATE CRACK GROWTH LAW (FIG 9) ENVELOPES BWR DATA POINTS ADEQUATELY PREDICTS ACTUAL CRACK GROWTH RATES IN BWRs (FIG 10) PRESSURE AND THERMAL STRESSES BASED ON OYSTER CREEK PLANT SPECIFIC FINITE ELEMENT ANALYSIS. UNIT 1000 PSI PRESSURE (FIG 11) UNIT 4500F STEP CHANGE IN FEEDWATER TEMPERATURE (FIG 12 AND FIG 13) i
t TECHNICAL JUSTIFICATION (CONTINUED) RESULTS (FIG 14)
@Me.cD REFUELING FLAW SIZE OUTAGE YEAR (INCH) 7R 1977 0.062 11R 1986 0.095 12R 1988 0.107
- ESTIMATE RESULTS VAllD FOR BYPASS LEAKAGE UP TO 3 GPM.
FACTOR 10 TIMES EXPECTED. O INTEGRITY OF FEEDWATER N0ZZLE IS ".ATNTAINED
- - - .- . r, . - - , ._ - - . ,- . -. ,, _. .,_. ,, , _ , , , . , _ , , , _
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\ Feedwater Nozzle
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- s. N N..b Section A-A OYSTER CREEK REPLACEMENT FEEDWATER SPARGER I
FIG' 2 Safe End
/ Feedwater Nozzle / Piston Ring
(( ANNNNNNNNNN O Spring Thermal Sleeve . Platen Ring Detail 1 Feedwater Nozzle outer saffle See Detail Above , Inner Baffle s L5 r-w x s
/
s s s s s s s s
/ / / Y x R x g ,7 s -s -s x t Sparger
- ' Thermal Sleeve ,
Flow Baf fle Detall 9 OYSTER CREEK REPLACEMENT FEEDWATER SPARGER PISTON RING AND FLOW BAFFLE
FIG 4 l 900 FULL POWER m OPERATION z
~ ~
430 g 444 430 425 i , . it e 4aa - 300 13 wC 300 = 6' Id , , l
*: 2m -
9O ON/OFF FW FLOW t00 gg 3 R 100 100 CYCLING AT SIX
** cycies PER hour -I Q
- TIME thri .
21/2 he
, 2 hr HP HPHOT STAND gT STANO8Y ,
2 hr LP HOT STAND 8Y ! -->- 4-- 1 . . 1200 1046 , ,
} 1046 # s \
j1000 =
. g30 \ 930 1
w 5 == - . N e h soO - 8 4ao - U E 2o0 - o- ' TIME thr) Startup/ Shutdown Cycles (130 Per 40 Year Plant Life)
FIG 5 FULL OPE A. FULLPOWER TION WE R ATlON e00 ON/OFF FW FLOW CYCLING AT SIX -
! Boo cyew PER hour 430 425 2 4:5 ,, 'f EC # ~
Do ti -
=
. j i
. eo ago - - - - -
[ f
$b i g, _
q w, Q , 6 j sd ,, sumamI$Ell l lil,. i= - TIME thri 10 hr LP HOT STANDtY 2 hr HP I HOT STAND 5Y l l l 1200 - l 1200 - 1045 1046 10 o - 930 [ 3
- S g om -
- )
Ba E l 8m - G rm - , 120 0 TIME thr) i Scram to Low Pressure.Ilot Standby and Return to Full Power (349 Per 40 Year Plant Life)
FIG 6 FULL POWER OPERA. FULLPOWER TION OPE RATION 600 -- - --4 426 / 430 425
$ 54m -
$,8 '
ON/OFF FW FLOW wc 300 - l CYCLING AT SIX I$ ,
,1l l
cycles PER toWr ND l a o 2m - l 5 J *E 100 -
-, 100 TIME thr) 2 hr HP
> 4 HOTSTANDSY I
l 1200 - 1046 31000 930 s y s00 - o 3 - y e00 5 4m - t i 2m - 0 Tih8E thr) Scram to High Pressure Hot Standby and Return to Full Power (62 Per 40 Year Plant Life)
6 FIG 7 700 - 600 - w-485 9 CYCLES 9 CYCLES $$ 500 - 430 j 430 ,p 425 400 - - I g# 5 g2 300 - el 200 - I 100 10- . TIME (HR) 1400 -
~
P
- 1200 -
b 1045 1045 ff
\ 930 1000 -
930 [ E 800 - y 600 - e 400 - 200 - f i TIME (HR) OYSTER CREEK' PLANT SPECIFIC STARTUP/ SHUTDOWN CYCLE i e
.- -- _ - - - - - , ,. -, --.__,. - - +-.,
FIG 8 800 700 2C - CL 600 12 CYCLES hh g; 500 - 425 430 [ 425 2" w - I
- R.
O 400 Gz $ [ 300 - 200 - h 100 - 100 TIME lHR) 1400 -
}1200 1050 1045 3 j 2045 g 1000 -
N 930 / c 800 - 8 g 600 - 5 400 - 200 = MM TIME (HR) OYSTER CREEK PLANT SPECIFIC SCRAM CYCLE
l FIG 9 : ASME - SECT XI - b' 8EST PIT CURVE de/dN = 0.0707 m Id K,n 3.726 UPPER BOUND CURVE , de/dN = Q.3796 a 104 K,n 3.723
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. 55 g O E h LVDATA O LOW R 10.16 < R < 0.25)
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$ $ g PWR ENV) E HIGH R (0.63 < R < 0.70)
GE DATA O LOW R (R = 0.05) EE iaWR ENVi e stGs R iR . Q.7s a
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1 i 2 3 10 10 10 Ef FECTIVE STRESS INTENSITY F ACTORf (kal 8) 0 0 Ken = Km (1' Al .5 = AK/11 R1 5 , Carbon and Low-Alloy Steel Fatigue Crack Growth Data in Water Reactor Environments
FIG 10 OBSE RVED CRACK DSPTHE. O DORE REGDON O BLEND RAD 6US Q PLANTS WITH PARTIAL FlXES O g3 _ (WE LDE D SP ARGE R. INTE R-FERENCE *lT SPARGERS.OR NINE WILE HIGH FERRITE CLADO 68*GI PolNT SEST ESTIMATE 1.4 - AN ALYTICAL CURVE ltORE REGIONI i 1.3 - UPPER BOUND AN ALYTICAL CURVE (SOME REGIONI g , 1.1 - OVE RSE AS BWR 1.0 - [ E c.s - 2 O w 4 as - 3 g PILGRIM
. a b sEST ESTIM ATE 5 o,7 _
l AN ALYTICAL CURVE HUM 00LDT BAYO O '
- teLEND RADIUS)
I o.s - OYSTER CREEK O j f MILLSTONE - os -
) MONTICE LLO O O DRESDEN 2 0.4 -
/ MACH 90TTOM-2 O OUAD CITIES.2 O QU AD CITIES-1 O DPESOES3 O
VERMONT DJ YANKEE 0.2 - COOPER D OVERSEAS _ SROWNS FEhMY 1 OVERSEAS 0.1 - SWRS PEACH SOTTOM-3 h BWRS SMOWNS FERRY-2 I la )J ^ lg I I ; HhTCH1 I ! t o v v v vv o al ao ao Vao too. 12o NUMBER OF STARTUPfSHUTDOWN CYCLES l Comparison of Analytically Derived C ack Growth Curves and Maximum Crack Depths Observed in Operating Reactor Feedvater No :les - . . . ,_. . - _ , . - _ _ _ . - - - _ - . . _ - _ . _ _ - _ . - . - . ~ - _ _-
FIG 11 Section Thickness A-A 7.433"
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C D P4 O HOOP STRESS PROFILE FOR I000 PSI INTERNAL PRESSURE ,
FIG 12 h=0 h T bulk (No Axial Heat Flow) Region (Btu /Hrft *F (*F) 3 I] a I 150 100 II 575 325
.. III 1000 550
- i I
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._ (Insulated) h=0 (No Meridional
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9 FEEDWATER NOZZLE l HEAT TRANSFER COEFFICIENTS
e FIG 13 I i lh 120 , Section (Thickness
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Wdig , QQg TDR NO. 497 REVISION NO. n isUDGET TECHMCAL DATA REPORT ACTIVITY NO. 315302 PAGE 1 OF o PROJECT: gg DEPARTMENT /SECTION Tech. Func./ Enc'c. & Desicn RELEASE DATE 11-7 Al REVISION DATE DOCUMENT TITLE: OC;GS RPV P, :so:: . .c: Structural Integrity After 1977 Repairs ORIGINATOR SIGNATURE DATE APPROVAL (S) SIGNATURE A rt e DATE A n.4 M L- u!4/93 p -
/K,UIn, a.s-a3 APP / TOV,AL FpR EXTERNAL DISTRIBUTION DATE
/}'f V & '%/sc Does this TDR include recommendation (s)? bas ONo if yes.T R/T o DISTRIBUTION ABSTRACT:
puracse-
- e Capodar~no
* [G~$$Chardos An inspection of the Reactor Vessel Feedwater no::les was T. M. % sey riginally scheduled for the Cycle 10 outage. 3ecause of the p, 7, g.,g decision to not replace the core spny sparger during this
- D. Elam wtage, a decisien was made to defer the inspection of the
{ S. Fiedler feedwater no::les. This report evaluates the predicted I. R. Finfrock grwth of defects which went undetected during the 1977 F. S. Giacobbe repairs of no :le clad and defect renoval. Specifically, the
- J. L. Sullivan, Jr. purpose of t".is evaluation is to demonstrate that a 0.0625" R. F. Uilson (1/16") deep no::le inside corner region crack would not D. Ec.g propagate to a size requiring repair, viz. 0.5".
Results Using fracture mechanics methods verified accurate for 3',2 feedwater no::le applications and taking credit for design leakage perforance of the theral sleeves installed during the 1977 outage, it is shown that an initial 0.0625" deep flaw will not propagate to more than 0 070" deep in five years of use. This result was obtained using teh IS:E, ESVC, Section XI crack growth rule and employing a generic theral duty cycle. Using a themal duty cycle specifically representing CC operating practice results in prediction of even less growth, p Conclusion
/(?SO It is very unlikely that small defects left after the 1977 feedwater no :le repairs would propagate to a depth reggiring repair at this time.
Reccrmndations It is conservative to inspect the no::le region during the cycle 11 refueling outage.
~~
CCOVER PAGE ONLY Accoco30 4.aa
TABLE OF CONTENTS PAGE I. INTRODUCTION 1 I I. METHODS 2 III. METHODS 3 A. ANALYTICAL STRATEGY 3 B. MECHANICAL AND THERMAL STRESSES 4 C. CGR 4 D. THERMAL DUTY CYCLE 5 E. 0YSTER CREEK SPECIFIC DUTY CYCLE 5 III. RESULTS 6 I V. CONCLUSIONS 7 V. REODMMENDATIONS 7
~
VI. REFERENCES 8 (i)
E l 4 l TDR No. 497 1 Rev. O Page 2 of 8 I. Introduction . In 1977, inspection of the feedwater nozzle region of the reactor ! pressure vessel (RPV) identified blend radius and bore region cracks. Oye penetrant indications were ground out and blended smooth. A dozen other domestic reactors had similar or worse indications. l To prevent reoccurrence at Oyster Creek a piston ring seal thermal sleeve was installed in each of four feedwater nozzle penetrations. The piston ring seal is intended to reduce leakage between the thermal sleeve and the feedwater nozzle inner diameter. In the previous saarger/thennal sleeve arrangement leakage occurred. The cooler feedwater leakage mixed with much hotter do ncomer flow creating turbulent eddies. The result of this unanticipated mixing was that the nozzle blend radius region, in particular, was alternately wetted by hot coolant then by cooler feedwater at high frequency. High cycle thermal fatigue initiated cracks through the i cladding. Normal thermal duty, viz, start-up/ shut-down cycles and
; SCR#tS, propagated the high cycle initiated cracks into the base metal.
In the discussion to follow it is assumed that the intended performance of the piston ring seal thermal sleeve is adequate to prevent feedwater leakage, and in so doing, remove concern about an on-going crack initiation phase. The pumose of this evaluation is to address OC GORB recomendation l' OC -9 8-1. The OC GORB has recomended that an ultrasonic inspection of the RPV FW nozzle be conducted during the present (1983) outage i based on a concern that small defects, possibly lef t after nozzle clad and defect removal, may have propagated to unacceptable size since the 1977 repairs. The purpose of this evaluation is to demcnstrate that a 0.0625" (1/16") deep nozzle inside cover region crack would not propagate to a size requiring repair, viz 0.5". The AStE reporting criteria governing performance of the dye penetrant examination did not require that a linear indication of 0.0625", or less, be documented. I This evaluation addresses the predicted propagation of a defect which !' may have been undetected during the 1977 repairs. THe ASME reporting criteria governing performance of the dye penetrant examination did .,
- not recuire that a linear indication of 0.0625",or less, be W #
documented. This evaluation is to demonstrate that a 0.005" (1/y') long crack, with an assumed semi-circular profile,, located in the , insidecorner region, would not propagate to a depm greater than N ,/ f.HS*[f, sh 0.5", the conservatively established limit for repair. [ d.0.$It{&l
$fnn.i l
TOR No. 497 Rev. 0 Page 3 of 8 II. Methods A. Analytical Strategy The methods employed in this evaluation are precisely those identified in Reference 1. Crack propagation is determined using
; linear elastic fracture mechanics (LEFM) relationships which i consider the interaction of presumed cracks with the immediate mechanical and thermal stress field tending to exercise the crack. Crack growth results are benchmarked against domestic nuclear industry experience and experiments conducted by the
', Japanese Atomic Energy Research Institute (JAERI). i The basis for the mechanical and thermal stresses used in this *
! evaluation is Ref. I for the unclad condition with the exception ! that an adjustment is made to take benefit of the thermal j performance of the thermal sleeve installed at Oyster Creek The methods of Ref. I were extended in that ASME Section XI crate growth rules (CGR) available only since the publishing of Ref. I were used. Another extension is that in addition to the generic
, duty cycle used previously a duty cycle specific to Oyster C 't
- operating practice was developed and used.
1 The LEFM code used in this evaluation was the same as that used in Ref.1. It was obtained from one of the authors (P. C. Riccardella) and put-up in the GPUN computer system. Individual problem solutions were executed by GPUN. B. Mechanical and Thermal Stresses To benchnark the method, in-house pressure and thermal stresses J' in the RPV feedwater nozzle region were taken from Fig.1 (Ref. 2). It is important to recognize that these stresses originate from an axisymmetric analysis. The feedwater nozzle intersection with the RPV is not axisymmetric, there being one radius to describe the long axis of the cylinder and a second to describe , tha circular cross-section. To account for this an adjustment is ' made to the model of the intersection to provide that stresses are equal to or greater than stresses predicted from simple fo rmulas. A fictitiously large radius is used to accomplish ! this. A check on the accuracy of the radius chosen is that the
- stress concentration factor (KSCF) of the inner corner radius of j the nozzle is not greater than 2.6, as detennined by JAERI '
(Reference 3). i. l l
i i t 4 TDR No. 497 Rev. 0 Page 4 of 8 l The thermal stress analysis captures the effect of the high cycle thermal mixing condition resulting from feedwater leakage as well j as the presence of the stainless steel cladding. The presence of
- the cladding increases stresses because of the thermal expansion difference and thermal conductivity difference between cladding and base metal.
- The LEFM model and-governing equation for stress intensity of a l feedwater nozzle blend radius crack is shown in Figure 2.
The integration of the crack growth equation that is a best bit I to ASME Section XI,1977, data: da = 0.3795 x 10-9(4LK)3.726 g l 1 ON I . a, crack radius N, number of cycles is accomplished by executing GECRACK (Ref. 2) over a generic duty i cycle _(discussed below). Starting with an initial crack size of O.25", the initiation phase crack arrest depth as well as i cladding depth, yields the results shown in Figure 3. The GPUN
! results track through the data from industry experience. This is : )
a key benchmark in verifying the method. 4 In addition, another bench' mark case is attempted. The LEFM model is executed against JAERI results (Ref. 3). The Japanese notched , a model nozzle forging and cycled pressure as crack propagation - was measured. Figure 4 shows that GPUN results followed the
- observed crack extention reasonably well.
] fhe above cases serve to verify the accuracy of the LEFM method l l and demonstrate the reliability with which GPUN can execute 4 problem solutions of this kind. l 1 i C. CRACK GROWTH RULE 4 I Eq.1, above, is a besc estimate of experimental data for crack propagation for both BWR and PWR materials available in 1977. I The upper bound Sect. XI CGR for 1977 is more conservative in 4 that the majority of experimental data points lies below the curve. Use of the upper bound CGR yields different results than those obtained using the best estimate. The upper bound results i suggest crack propagation at a much f aster rate, a rate not { verified by industry experience (Figure 3). ! i l
TDR No. 497 Rev. 0 : Page 5 of 8 ; i
- In 1980 a revised ASME, Section XI CGR became available (Fig. 5, ;
Ref. 4) better representing crack extension at high stress intensities. It is used below. D. Thermal Duty Cycle l Two thennal duty cycles are considered: the GE generic dut cycle and a GPUN derived Oyster Creek specific duty cycle. y E. G.E. Generic In Ref.1, G.E. developed a thermal duty cycle description representing the most severe start-up/ shut-down and SCRAM 1 frequencies and profiles af ter considering all BWR experience. j To briefly summarize key features of the GE generic thermal duty cycle, Fig. 6, feedwater flow cycling creates nozzle region j temperature extremes of 4330F to 1000F,.at 6 cycles per hour, for high pressure, hot sta7dby conditions during both heat-up and i shut-down with additional but lower temperature range cycling occurring during low pressure, hot standby during shut-down. i For each start-up/ shut-down cycle, three SCRAMS will occur whose i thermal and pressure profiles are on Fig. 6a. Each SCRAM is associated with a 1200 psig pressure spike and is divided into a high pressure and low pressure recovery phase. During each ! feedwater cycling is taken to occur at 6 cycles per hour with 4300F to 1000F temperature extremes during feedwater on/off 4
- cycles at high pressure and 3500F to 1500 temperature
{ extremes during low pressure hot standby. When feedwater is
, secured, hot downcomer water constitutes the boundary condition at the nozzle; when feedwater is restarted, cooler feedwater conditions dictate boundary conditions. , ; F. Oyster Creek Specific Thermal Duty Cycle ! Exception to the G.E. generic duty cycle can readi.ly be taken.
1 With the assistance of the plant (Ref. 5) and with the detailed 1 review of control room logs, control room instrument strip charts, and plant procedures provided by the GPUN Mechanical . System groups (Ref. 6 and Attachment 1) and with reference to + plant SCRAM logs (Ref. 7) an Oyster Creek specific duty cycle was developed (Fig. 7). The specific Oyster Creek duty cycle is an
; estimate approximating actual practice.
r l, I
I TOR No. 497 Rev. O Page 6 of 8 Key features of the Oyster Creek specific duty cycle are, firstly, that while feedwater pump on/off cycling has been a necessary practice to control reactor water level, it has not occurred at the frequency or regularity suggested by the G.E. generic cycle. Nine feedwater pump on/of f cycles is a reasonable upper limit dJring start-up and shut-down maneuvers. Second, there have been no, nor can there be, pressure spikes to 1200 psig during SCRAM's. EMRV set-points prevent this as does automatic isolation condenser initiation at 1060 psig. Finally, SCRAM to a low pressure, hot standby condition is not procedurally recognized at Oyster Creek. On the other hand, the 1/3 relationship of start-up/ shut-downs to SCRAM's is the same as that used in the GE generic assessment. There have been three SCRAMS to each start-up/ shut-down at Oyster Creek. III. Results Ref. I provides pressure and thermal stresses for an unciad nozzle region (Fig. 9). The thermal stresses for the film coefficient case of U=100 BTV/hr - f t2 . oF were adjusted (upward) to include the effect of the thermal sleeve as installed at Oyster Creek. The thermal performance assessment of the installed thermal sleeve was made by its designer, MPR Associates (Ref. 8). MPR recommends using a film coefficient of U'150 ST U/h r - f t2 . of to represent the installed thennal sleeve. Fig. 9, the thermal stress profile of the nozzle using as-built conditions, was developed in this manner. Interpolation between the G.E. generated cases of U=500 & U=100 made possible results f or U=150. The stress distribution was . input to GECRACK for execution of the crack propagation prediction. The initial crack size was 0.0625", the largest non-reportable dye penetrant Indication permitted by the ASME Code (Ref. 9) To be conservative, the hypothetical crack depth was taken equal to its surface exter.t. 0.0625". Crack growth was integrated over the 1977 best estimate relationship (EQ.1) and the ASFE, Section XI,1980, bi-Itnear CGR . The model was executed for both the G.E. generic duty cycle and the Oyster Creek specific duty cycle developed by GPUN. The results are indicated in Fig.10 and summarized in Table 1.
TOR No. 497 Rev. O Page 7 of 8 Results obtained by using a method demonstrated valid by duplication of published results indicate that for an initial 0.0625" deep feedwateg nozzle corner crack interacting with an adjusted GE develow' stress field for pressure and thermal conditions propagates to not more than 0.070" deep, but no less than 0.064" deep in five years. The most conservative case, that being the more extreme thermal duty cycle (GE generic) provides no alarming crack growth predication for the period since repairs in 1977. IV. Ccnclusions
- 1. It is very unlikely that small defects lef t af ter the 1977 feedwater nozzle repairs would propagate to a depth requiring repair at this time.
- 2. The ASFE Code threshold for raportability is demonstrated as appropriate. The apparent disconnect between Section XI &
Section III is reconciled.
- 3. The design basis for the RPV, in terms of anticipated SCRAM cycles, is being exceeded by rate of occurrence, but not by total count.
V. Recommendation
- 1. It is recommended, as a conservative step, to inspect the nozzle region during the next refueling outage.
E I 1 TDR No. 497 ! Rev. O i i Page 8 of 8 ! 1 .. VI. References I
- 1. G.E. NED0-21821-02, Boiling Water Reactor Feedwater Nozzle,
- ) Sparger, Final Report (Supplement 2), by H. Watanabe, P.C. i Ricardella, et. al, August 1979. l j 2. GECRACK Users Manual, P.C. Riccardella,1/19/77.
j 3. Boiling Water Reactor Feedwater Nozzle (Sparger Interim Program, 1 Supplement 2), GE-NE00-21480-2, by P. C. Riccardella and S. A. Delvin, November 1977. l ] 4. ASME, BPVC, Section XI,1983. l 5. Mr. N. Boulware, Group Shift Supervisor, OCNGS, private ;
- connunication to S. D. Leshnoff.
I l 6. IOM, B. Elam to S. D. Leshnoff, OCNGS Feedwater Nozzle : j Inspections, MSS-83-868, 11/4/83. i 7. OCNGS, Plant SCRAM Log. i
- 8. MPR, Report #783, OCNGS, Evaluation of Low Flow Feedwater Control System, 8/83.
i j 9. ASFE, 8PVC, Section 111, NS-5 350,19 74. l { i l ) ! i I i L 1 i 1 i l i 1 - - _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ - -
TABIE 1 Intial Size WR Stresses Duty Cycle Final Depth (pressure & 1hermal) I load block = 1 SU/SD atter 5 yrs.
+ 3 SGWG 0.0625 6 77 IE G.E. adjugtal to U=150 GE .064 InU/hr f t hr 'F, &
Ppeak = 1200 psig 0.0625 M 80 ASME GE 0.070 (bi-lineer) q 0.0625 677 BE Pp k = 1050 T 18 h eycles during SU/SD; 0.064 no SGAM to IP/IIS 0.0625 s;80 ASME 9.07o DC
BWB FEEDWA TER NOZZLE i 240 g I 31o N
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( l FEEDWATER NOZZLE REPRESENTATION l I I : I I I I . I I I I l l l l , I l l l p A, RA01US OF CR ACK l I A s, i j (VESSEL i J l i i i I l
's I x
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N N FUN 11. SIMULATED 3.D NOZZLE CORNER CRACK Kg = fra (0.706 Ao + 0.537 (2a/n) At + 0.448 (a 2/2) A2 + 0.393 (4a3/3n) A3 ] l Where A o, Aj, & A3 ARE CONSTRAINTS IN A POLYNOMIAL REPHESENTATION OF THE STRESS DISTRIBUTION.
. FIG. 2
. _,._.-_-_ __.__ _ _ .- _ . _ . _ _ _ . . _ _ _ _ _ , _ . . . _ _ _ _ _ _ _ . _ _ _ _ -._______.,_ .___ _ -_ _ _ - ~ _ . _ _ _ _ _
COMPARISON OF ANALYTICALLY DERIVED CRACK CR3WTH CURVES AND MAXIMUM CRACK DEPTHS OBSERVED IN OPERATING REACTOR FEEDWATER NOZZLES (GE VS GPUN RESULTS) l'I OBSERVED CRACKED DEPTHS O BLEND RADIUS 1.6 - A PLANTS WITH PARTIAL FIXES (WELDED SPARGER, INTER. FERENCE FIT SPARGERS. OR 1.5 - HIGH FERRITE CLADDING) X GPUN RESULTS 1.4 - ; 1.3 - i rUPPER BOUND
- / ANALYTICAL
! CURVE (BLEND RADIUS) 1.2 -
l 1.1 - I I 1.0 - I 0.9 - I 0.8 - HUM 80 LOT B AY 0.7 - BEST ESTIMATE O
, j f ANALYTICAL CURVE f (BLENO RADIUS) OYSTER CREEK 0.6 -
l 0
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O O 0 0RESDEN 2
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/ CITIES 2 OQUA0 CITIES 1
,/ PEACH BOTTOM 2 O 0 0RESDEN 3
- oO VERMONT 0.3 ,.*** YANKEE 0.2 - COOPER OVERSEAS 6 RS -
BROWNS FERRY 1 OVERSEAS 0.1 - PEACH O BRS OM 3 BROWNS FERRY 2 1 HATCH 1 0 ' '^ ' ^ I I ^l OA I I l 0 V 20 40 60 80 100 120 NUMBER OF STARTUP/ SHUTDOWN CYCLES FIG. 3
JAERI RESULTS 2.0 A
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, X GPUN RESULTS g O
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20 30 NUMBER OF CYCLES (thousands) FIG. 4
1983 Edition APPENDIX A - NONNIANDATORY Fig. A43001 1000 .
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6o0 - FULL POWE R m y OPE R ATION
~
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l FULL POWEn OPE R A-FULL POWEli TION OPE RA TION 600 - -4 EIo 425 43o 42,5 Ii 3 o 400 - l
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FULL PCWER CPERA pyLLPCwER TCN CPE A AT CN 4 600 - ON/0FF FW FLCW CVCLING AT Six [ 500 - cvctes PE R mwwr 4M 4;g z 425 _.
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.1
OYSTER CREEK PLANT SPECIFIC SCRAM CYCLES 800 700 - EC w- 2 600 - b eh 12 CYCLES {8 500 [ 425 430 d ga 425
),
w y 400 -
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200 -
, 100 -
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- f 1400 -
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[ A1045 1045 g 1000 "
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=
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; - FIG. 7 (a)
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OYSTER CREEK STARTUP/ SHUTDOWN CYCLES 700 - 600 - w-
$9 500 - 485 9 CYCLES 9 CYCLES
; 430 ,f 425 430 d ##
g* 400 - - I ' N g2 300 - d 200 - 100 too - 100 TIME (HR) 1400 - IP - 3 1200 w 1045 1045 ff h1000 - 930 [ \ 930 e E 800 - 8 y 600 = u 400 - 200 - TIME (HR) FIG. 7 (b)
n I NOM 4N AL CLAD DEPTH 180
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200 200 A. U = 2000 8tu/he.ft2.eP /, S. U = 1000 Stufhe ft2 .o, ,. . 180 - * * , 180 -
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-20
' ' ' I I ' ' ' ' ' ' ' I I ' '
- 20 1 2 3 4 5 6 7 8 1 2 3 4 5 6 7 8 INSIDE DISTANCE ALONG OUTSIDE INSIDE - DISTANCE ALONG OUTSIDE SURFACE NOZZLE SECTION 8 linches) SURFACE SURFACE NOZZLE SECTION 8 (Inches) SURFACE l
l l Unclad Nozzle Thermal Stress Profiles f16. 6
THERMAL STRESS AT BLEND RADIUS REG 70 - i 60 4 6
\.
\.
50 - \ ,
\.
. \
, .m h = 500 Btu /hr - 12 . .p 40 = * * , .
MnteWabon) s \ 3 u i. m
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.. c\.
h = 500 Btu /hr-12. .p
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+
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h = 100 Bru/hr- 2 t . .p 4
,,..* *.g.
., .\ , s l l t .**.*.*.\ I f 2 4 6 10 W 8 -
-10 -
-20 FIG. 9
OCNGS FW NOZZLE BLEND RADIUS - CRACK PROPAGATION ,
.70
'80 1980 ASME CGR
}
.60 - BE 1977 BEST ESTIMATE CGR ,!
/
GE GENERIC DUTY CYCLE *
/
OC SPECIFIC j
.50 -
./
/
GE *80 . 40 -
./
./
,/ OC 80
.30 - -
j
./
/
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l I l l 100 5 YRS 200 ' 300 20 YRS 400 CYCLES 10 YRS 30 YRS 500 FIG.10
PAGE 2 of 2 5 868 November 4, 1983 If you have any questicns regarding this subject, please ccntact D. Wong on extensicn 2034. FJ.am Mechanical Systems N/BDE:mcr cc: G. R. Capodarno, Fluid Systems Director D. K. Crcneberger, Director, Derg. & Design T. M. Dempsey, Secondary Plant Manager A. P. Pcchino, Manager, Ehgineering Mechanics D. Wcog, Ehgineer, Mechanical Systems EDECC Files - OC 10300 and 5050 1
- P;.GZ 1 of 2 Intor4ffl30 Momsrandum MSS-83-868 Dan N:nreser 4,1983 Sueject Oyster Creek Ebedwater Nazzle
'Ihermal Cycling Nuclear To S. Isshnoff Parsippany-C B Project hgineer Location
'Ihis is to supplement GPUN Mercrandum MSS-83-851, B. D. Elam to S. Ieshnoff dated 10/26/83, same subject.
In review of the Oyster Creek Specific Startup/Shutdcwn and Scram Cycles, the use of these cycles to approximate a representative duty cycle for Oyster Creek reactor feedwater nozzles appears supported by the review of the strip chart recordings and control room logs used by MPR in generating their report.
'Ibe firdings frce the strip chart recordings and control room logs which are in support of the specific Oyster Creek duty cycles, as stated previously in GPCN memorandum MSS-83-851, are:
- 1. In 1982, the nu.nber of times the reactor feed pumps were turned cn and off was equivalent to produce 9 feedwater cycles for the duration of the plant startup. 'Ihe turning on and off of the reacter feedwater pu::ps is a definite indicaticn of feedwater cycling at the nozzles. 'Ihe occurrence of this is greatest in 1982.
It appears that the centrol of the feedwater to the reactor was most difficult in 1982, as indicated in the total number of reactor level fluctuaticms per hour in cceparison to the other previous years. It could therefore be crnservative to use the 9 cycles experienced in 1982 as the worst case and apply it to the other previous years.
- 2. 'Ihe fluctuaticns of reactor water level occur during the ramp up to and at hot standby and not cnly at hot standby.
- 3. 'Ihere are no indicaticns that Oyster Creek normally proceeds or remains in a low pressure hot standby ccndition for any extended period of time.
It appears that the reactor is suberitical if the correcticn of the scram requires a long period of time.
- 4. 'Ihe presence of a 1200 psig pressure spike during each and every scram is not justified for Oyster Creek. A 1200 psig pressure spike can occur, but cnly during a rare transient conditicn, i.e., a turbine trip with bypasses closed. Please note that pressure spikes over 1100 psig can occasionally occur at Oyster Creek.
'Ihe records of Oyster Creek reactor and feedwater parameters are not conclusive and ccxnprehensive enough to justify the replacerent of the General Electric (GE) duty cycle; however, they do provide enough justificaticn to produce a duty cycle which can be used to approximate the Oyster Creek actual situaticn for use in a ecx::parison with the GE typical cycles.
A0000844 {
A 4 4 i ATTACHMENT 3 h f e k I I i 1 1 st 1 1 4 4 4 i I 4
lNf YW TIC 023 NCC : #3-4 ef MPR ASSOCIATES. INC. FDIu.: a B'l *fA01 __ DISTSSED WI"'1T 4
. e- >
OYSTER CREEK NUCLEAR GENERATING STATION EVALUATION OF LOW FLOW
'- FEEDWATER CONTROL SYSTEM MPR-783 J
P a F' d E u 7 T l l Prepared for: GPU Nuclear Corporation August 1983 1050 CONNECTICUT AVENUE. N.W. WAsMINGToN. D.C. 20036 202 659 2320
M P R ASSOCIATES. INC. TABLE OF CONTENTS
- 1. _
INTRODUCTION 1.1 PURPOSE OF EVALUATION 1.2 SCOPE OF EVALUATION
~
1.3 BACKGROUND
- BWR FEEDWATER NOZZLE CRACKING PROBLEM 1.4 PREVIOUS ACTIONS TAKEF AT OYSTER CREEK TO REDUCE - POTENTIAL FOR FEEDWATER NOZZLE CRACKING <1
- 2.
SUMMARY
AND RECOMMENDATIONS 2.1 RESULTS OF EVALUATION 2.2 RECOMMENDATIONS 2.2.1 Short Term O 2.2.2 Long Term l 3. EVALUATION tJ _ . . . . . 3.1 PRESENT LOW FLOW FEEDWATER CONTROL SYSTEM 3.2 OPERATING EXPERIENCE - 1978 TO PRESENT i a 3.3 STRESS AND FRACTURE MECHANICS EVALUATION OF FEEDWATER
,J NOZZLE
', 3 3.3.1 Feedwater Nozzle Pressure and Thermal Stress Analysis
,_ ,High Cycle Fatigue ss 3.3.3 Low Cycle Fatigue v-MT Crack Growth Rate i
M PR ASSOCIATES. INC. TABLE OF CONTENTS (cont'd)
- 4. REFERENCES
- 5. APPENDICES y Ig FEEDWATER NOZZLE PRESSURE AND THERMAL STRESS ANALYSIS L B. HIGH CYCLE FATIGUE EVALUATION f g-Q LOW CYCLE FATIGUE EVALUATION d D. CRACK GROWTH RATE EVALUATION -
8 8 r1 d - 1 a t V P s. l I ii j
l 4 proportional to reactor power at those pressures ( 1 percent at the lower p'ressures, 2 to 3 percent at the higher pres-sures). Hold periods observed extended up to 15 hours. 4 I For the purposes of evaluating the effect of flow variations I on nozzle fatigue usage, a typical startup/ hot standby cycle i is required. GE has described such a reference duty cycle for BWRs in Reference 1. (See Section 3.3.3 of this report j( for a description of the reference duty cycle.) Although I there is a wide variation in.the actual path followed in each of the actual startups reviewed at Oyster Creek, the GE 3 reference duty cycle, in terms of elapsed periods, is con-jd sidered applicable on the average. The specific charater-
} istics of the reference duty cycle as described by GE in g Reference 1, such.as number of feedwater flow cycles per hour and the potential for variation in the fluid tempera- ; y ture at the nozzles, vary somewhat from that observed in the
] [ actual ~startup data. Tne effects of variation in these parameters were evaluated in assessing fatigue usage of the
- U feedwater nozzles.
\
, W 3.3 STRESS AND FRACTURE MECHANICS EVALUATION OF FEEDWATER NOZZLE 3
! Ul Stress and f racture mechanics evaluations of the feedwater j nozzle were performed to determine: 1 1 - } a. The fatigue. usage on the feedwater nozzle due to high i L frequency thermal cycling and low frequency pressure jI and thermal cycling associated with normal startups, I shutdowns, and hot standby conditions. i i i 9 3.7 i ne-- + - , , - - , - , - - , - - -.e-,m-,, - ,-, ,---.n,--,-m -
, -s--- - - - , . , - - , - -w-,--- -,-. ~a , . - -,-,-w,-. --,e.-
- b. The crack growth rates for any existing cracks in the feedwater nozzles (i.e., cracks that may have been
' missed' during the 1977 inspections) or any new thermal fatigue cracks initiated by high or low fre-quency thermal cycling. (Note: Based on analyses presented in Appendices B and C of this report, initiation of any new thermal fatigue cracks is considered unlikely for the expected remaining life of the plant.)
v f l
- These analyses were performed for conditions considered
- typical of recent Oyster Creek operating experience, as well w as for postulated system improvements. Results of these i
y stress and fracture mechanics evaluations are described in
'J the follosing sections of this report.
I El { 3.3.1 Feedwater Nozzle Pressure and Thermal Stress Analysis Results of a pressure and thermal stress analysis of the feedwater nozzle are presented in Appendix A of this report. Analyses were performed using the finite element computer program ANSYS. The results of the feedwater nozzle pressure and thermal stress analyses were used as inputs to fatigue and crack growth rate analyses which are discussed in the following sections of this report. 1 Pressure Stress Analysis The finite element mesh for the pressure stress analysis of the feedwater nozzle is shown in Figure 3-2. The model in-cludes the feedwater nozzle and a sufficient length of the 1 3.8
reactor vessel shell and attached feedwater pipe such that the stresses at the end of the model are reduced to clean shell values. The 3D nozzle-to-cylinder configuration was represented ! using a 2D axisymmetric model of a nozzle in a sphere. The equivalent spherical radius was chosen to be 3.2 times the
, cylinder radius. Results of 2D axisymmetric and 3D finite
- f. element of analyses reported in Reference 4 indicate that
( this will result in peak hoop-stresses and stress intensi-ties in the nozzle that are approximately equal to the r [ef stresses that would be obtained from a more exact 3D model. 5 As shown in Figure 3-2, the finite element model includes O the final machined configuration of the feedwater nozzle due F to removal of stainless steel cladding and a thin layer of ( base metal during the 1977 modifications. {;} g A unit internal pressure load of 1000 psi was applied to the inside surface of the finite element model. An axial stress p of 2820 psi was applied at the attached pipe end of the finite alement model to represent the pressure load from the at-tached f'eedwater pipe. A zero rotation boundary condition was specified at the attached pipe and reactor vessel shell f ends of the finite element model. Radial displacement was b permitted at these ends. The maximum hoop stress and stress intensity were found to occur on the inside surface in the blend radius region of the feedwater nozzle. The hoop stress is the predominant streas so that the maximum stress intensity is approximately equal to the hoop stress. The distribution of hoop stress through the nozzle wall thickness for an internal pressure 3.9
l of.1000 psi is shown in Figure 3-3 for the worst case blend radius and bore sections. Thermal Stress Analysis The temperature distribution in the feedwater nozzle was determined for an assumed 550*F to 100*F step change in feedwater temperature. This transient represents the low frequency thermal shock due to on-off operation of feedwater flow during low flow conditions. When feedwater flow is r-off, the temperature.of the water in the feedwater sparger i thernal sleeve and feedwater nozzle will approach the tem-V perature of the reactor recirculation water (assumed to be 550*F at high pressure hot standby). Initiation of feed-O water flow to maintain reactor water level will result in a rapid decrease in the feedwater temperature flowing through
, ]i the feedwater sparger thermal sleeve (assumed to be 100*F i , when feedwater heaters are out-of-service).
The same finite element mesh that was used for the pressure stress analysis was used for the thermal transient analysis. The values of heat transfer coefficient and bulk temperature used in the thermal transient analysis are shown in Figure 3-4.
}O A conservative heat transfer coefficient of 150 Btu /hr ft27
} with a bulk temperature of 100*F was used for the surface of d the feedwater nozzle insulated by the thermal sleeve (Region I).
This value is based on calculations given in Appendix A of t, , this report for bypass leakage rates up to 5 gpm. The sur-face of the feedwater nozzle in contact with the reactor recirculation water (Region III) was assigned a heat trans-fer coefficient of 1000 Btu /hr ft2 .F with a bulk temper-ature of 550'F. The region between the double flow baffles , (Region II) was assigned values of heat transfer coefficient i 3.10
and bulk temperature equal to the average of the Region I and Region III values. The outside surface of the feedwater nozzle was assumed to be insulated (i.e., h = 0). Finally, the ends of the model were assigned zero heat transfer coef-ficient to simulate no axial (meridional) hea't flow at these locations. The temperature distribution (i.e., isotherms) in the feed-4 water nozzle at 5, 10, 30, and 310 seconds after initiation
' I.
of the tr:nsient and at steady state conditions are given in i. Figures 3-3 through 3-9. Thermal stresses were determined for each temperature distribution shown in Figures 3-5 through 3-9. The hoop {*L stress and stress intensity in the blend radius were found to be a maximum at steady state. The hoop stress and stress C4 l intensity in the bore were found to be a maximum at
, 310 secondu. The distribution of hoop stress through.the nozzle wall thickness due to the 550*F to 100*F thermal v
shock is shown in Figures 3-10 and 3-11 for worst case blend bf radius and bore sections, reapectively. d5 3.3.2 High Cycle Fatigue L , Results of a high cycle fatigue evaluation of the feeddater k' nozzle for the rapid temperature fluctuations associated r with mixing of hot reactor reciruelation water with cooler L feedwater bypass leakage are presented in Appendix B. The
, method of calculation was in accordance with Sections 4.7.2.1 through 4.7.2.4 of Reference 1.
3.11 1
The high cycle fatigue evaluation was based on the random temperature cycling trace shown in Figure 4-125 of Refer-ence 1 using the Ordered Overall Range (OOR) cycle counting approach. (Note: The referenced figures from Reference 1 are included in the calculations given in Appendix B.) The OOR load spectra and envelope spectrum for the high cycle fatigue evaluation are given in Figure 4-127 and Table 4-25 of Reference 1, respectively. The design fatigue curve used in the evaluation is shown in Figure 4-128 of Reference 1. This curve is the same as the design fatigue curve given Section III of the ASME Code up to 10 6 cycles but has been extended by GE out to 10 11 cycles. J High cycle fatigue usage factors were calculated for the [! b metal temperature cycling at the nozzle blend radius shown in Figure 4-33 and the feedwater time / temperature / flow map f shown in Table 4-29.a of Reference 1, respectively. Calcu-lations were performed for zero gap and 0.02" gap at the periphery of the flow baffle. From Figure 4-33 of Refer-e ence 1, the metal temperature cycling as a precentage of the Eh fluid temperature difference (i.e. (TRRW-Tpg) is 12 and 17 percent for flow baffle gaps of zero and 0.02", respec-tively, for bypass leakage rates up to 15 gpm. Further-9 more, the metal temperature cycling is essentially 7 independent of the bypass leakage in the 0-15 gpm range. v
- r. In the replacement feedwater sparger design, the flow
[, baffles are preloaded against the conical surface of the machined feedwater nozzle with a significant preload. In addition, since two flow baffles are provided, with leakoff holes between them, there will be essentially no leakage I 3.12 I
l around the periphery of the flow baffles. For these rea-sons, the replacement feedwater sparger flow baffle arrange-ment is considered equivalent to the tested single flow baffle with zero gap. The calculated high cycle fatigue usage factors are summarized below. l.. {~ High Cycle Fatigue Usage Factor y Gap at Periphery of Flow Baffle y Per Year In 30 Years In 40 Years [ Zero 5.18E-5 0.00155 0.00207
, 0.02" 4.55E-3 0.137 0.182 n The calculated fatigue usage factors indicate that initia-A tion of fatigue cracks due to high frequency thermal cycling is not expected.
n - - - - - . . . . . . . . .- .- . - .,. . .. __. . . . n . a i- d.2l ; .. asu..: '. i . '.2.-- -
- -: L: ~ 4
. ._.u- --- . ~-
3/.314llT 5 mg h g yr s , r'z,.e% L t. *;A;.tt xe <
=R
3.3.4 Crack Growth Rate Results of crack growth rate analyses are presented in Appendix D of this report. The method of calculation was in accordance with Section 4.7.3 of Reference 1. The pressure and thermal stress distribution through the nozzle wall thickness were obtained from the feedwater nozzle pressure k and thermal stress analysis given in Appendix A of this report. The stress intensity factors as a function of the flaw size and stress were calculated using the correlations given in Figure 4-136 of Reference 1. (Note: The referenced figures from Reference 1 are included in the calculations in Appendix D.) F [' Crack growth rates were calculated for the same reference duty cycle and variations used in the low cycle fatigue evaluation discussed in the previous section of this report. However, a simplification in the duty cycle was introduced in the crack growth rate calculations. Since the fatigue damage associated with scram to high pressure hot standby I (62 cycl'es per Figure 4-145) is obviously less than the 0 fatigue damage associated with scram to low pressure hot standby (349 cycles per Figure 4-144), the two types of d scrams were combined into a single scram to low pressure hot standby per Figure 4-144 which occurs 411 times in 40 years t- (62 + 349). The crack growth rate in the feedwater nozzle was then evaluated for 130 occurences of a sequence con-sisting of one startup/ shutdown cycle per Figure 4-143, followed by three scrams to low pressure hot standby per Figure 4-144, for a total of 130 startup/ shutdown cycles and 390 scram cycles in 40 years. (The non-conservatism of i l 3.16 l . _ _ , . .-- - - - - .
eliminating 21 scram cycles (411 vs. 390) by the above method is considered to be compensated for by the conser-vatism in defining all scrams as scrams to low pressure hot standby.) Calculations were performed for two initial flaw sizes; O.062" and 0.250". The 0.062" initial flaw size represents ( the maximum flaw that may have been left in the feedwater nozzle during the 1977 inspections. The basis for the f. {- 0.062" flaw size is given in GPU Nuclear Letter, MSS-83-537, to W. R. Schmidt (MPR), dated June 16, 1983. The 0.250" I initial flaw size represents the maximum flaw that could be D initiated by the high cycle fatigue mechanism. Although the replacement feedwater sparger is expected to eliminate any new crack initiation by the high cycle fatigue mechanism (see Section 1.4), the 0.250" initial flaw size is evaluated () in order to determine conservative inspection intervals for
, , the feedwater nozzles.
k Calculations were also performed for two crack growth laws;
~~' an Upper Bound Curve from Section III of the ASME Code and a Best Estimate Curve from Figure 4-137 of Reference 1. The crack growth rate in the blend radius is controlling since the stress is a maximum at that location.
n 1 v Results of crack growth rate calculations are presented in
- Figures 3-12 through 3-15. The estimated number of years for an assumed 0.062" and 0.25" initial flaw to grow to 1 inch are summarized on the following page.
3.17
3./ V l NO. YEARS FOR INITIAL FLAW TO GROW TO 1 INCH Initial Flaw 0.062" Crack Growth Law
' Case Description Best Estimate Upper Bound G) 1 Reference Duty Cycle 6 cycles /hr, 100% AT 16.2 2.7 L ?
- I
(, _____________ _______..._________ _ _ _ _ _ _ _ _ _ _ _ . . _ _ .._________ ___ p, . . Initial Flaw 0.25" Crack Growth Law
.. Case Description Best Estimate
(' Upper Bound d e> 1 Reference Duty Cycle - e- 6 cycles /hr, 100% AT 4.3 0.8 f-w ___ .___________________________ ______________________________ ,P _ . j ____..____._______________ ,______ ___________~____.._____________ l 5 < . .I . , P
- r. . -
.. l l ' Notes:
- 1. Case 1 represents current Oyster Creek operation.
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Iil l l M I llll ::: I y l l , , I l ,i I q . 1 L . , - III -. II q L . FEEDWATER NOZZLE HEAT TRANSFER COEFFICIENTS FIGURE 3-4
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+
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Stress ... . . - . . .. .... .. ...- .. . . . . . (Kst) ....-
. ._.- - - - - . ~ ~ _ t., .
.r , ,s. .; . . -
........i
..... . . . .. . .. . 6. p % . . .%. .,. .. .. ;. ..
....... . .. . . . . . ~ . . .. 3. ., .w, .k. .
40 - .. ... ....... ....... ........,.7 .. . . . F
....... . . . . . . . . .... 1 ..7gp :... ..g.......
r
..,. ..... . ..... . . . . . . . o.........
T
.. ..... ... ... ....s..... . . . . . . . . . . . .
6
...A6-j< 4.. . p .;. ....Q...
30 Sec . . . . .. ..... ... . . . . .. 1 4. . .
# 20 - - ' ~
q
- -----;.- g310.Sec..l.. .... . . - . . . . .
g . .
.- 10.Sec .. . . . .... . . .; . : .SS...
- t. .
5.Sec .
-. ; . .. . L . .
0 ....... ....... . x s l 6....... ..s......
.....G.....
..y
.........L..
....!..,. .,.y...
_. .u
.}........
- r. ....... . . . . . . . . . . . . . . . . . . . . . . . . . ...... ...
f -20 . . i 0 0.2 0.4 0.6 0.8 1.0
- e
. Thickness U o
u (tx/t) e e u 44 54 54 3
. m . m m
e e
.e.t
.C.4 a e a a a
.M O HOOP STRESS PROFILE FOR 550 - 1 OO OF THERMAL SHOCK BLEND RADIUS REGION FIGURE 3-1 O t
120 :.: _ Section Thickness B-B 4.345" B 100 . .n C. , . a. 3 .c - s W g-
- r. . %
O - 80 = . . _ . . . . ._ _ 1 9g q......... . . . - . . 60 - )..... . . ..-- -
. . ~ - ... . - . .. .
5 tress . ...... . . . - . . ,. -.. . - . . - - :.. (Ksi) . . .....
. - . . . _ . . . . . . . ~ . . .
_ ..A.........Q........... r-
, . .... g ...... . ...~... ..; ...a... .. .. .
~
Sec.j. g*.** g, . 30 . ' ' . . ~ * . . . . . . . . . . . ~..~ ~". (.\ .. .. .. ... .. .. .. . .. .'. ~ . . . ~ ~ . . ~ ~ . p . .. . . . . .. . . ......... ......... 20 - .\. . . % i - N.s..... .... 40
- Sec . A. .. .. '
. . . * * *..' N. . g' ' ' . . . . " . .. ...' .."...".. .
\. .
.N . .
. 3e c . q . . . . .._ .. . . .
... .... . .g. .
....s....
.. . . . . . ..g...... .. .g.
. . . g. .N,..s.
O ~% =~. N.. eC.
. ......'"'"*=__nN
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F'.
.. . . . . . . . . . . . .... ..... ... .... ........~.
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6 0 0.2 0.4 0.6 0.8 1.0 e Thickness e u o (tx/t) e e u M M be 'J
= =
m e e e .e
< m a
e o
>4 O HOOP STRESS PROFILE FOR 550 -1000 F THERMAL SHOCK -
BORE RADIUS REGION FIGURE 3-1 l
, t ;
.._.._._.4 a__
. t .
.! 'j:- >
- - :{ '
1.75 -
- ,i ". . ....
l . _ _
.i
,. l. .8
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.l .
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...........4.. ...j - I .._-.a W. -
,w
.i. . .
I j . . j - L . . s 1.00 Crack
.i . . .! .
Depth - - - - - ' : t -- ; ~ ~- ; - - - - -
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(Inch) . . ,
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---- --. ,-- C a s e i j -- ,
w- ! , t ,
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~
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g 1
..l..
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- ~ ~
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r' t
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g . - . . . . 0.062 *
. ' ' ' ! 40 F 10 ..
- 2,0
' 30 Years e- e 0 . . . . .
(. 0 100 200 300 400 500 600
, Cycles (Every Fourth Cycle = Startup/ Shutdown; '
other Three = Scrata To LP Hot Standby)
. CRACK GROWTH RATE O.OG2" IfflTIAL FLAW BEST ESTIMATE CRACK GROWTH LAW .
FIGURE 3-12 l
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Title:
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Title:
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Title:
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Title:
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Title:
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Title:
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Title:
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Title:
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Title:
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Title:
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Title:
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Title:
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10-3 ASME - SECT XI - BEST FIT CURVE UPPER SOUND CURVE '
, da/dN
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1050 Connecticut Ave., NW - Washington DC 20036
Title:
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1050 Connccticut Ave., NW - Washington, DC 20036
Title:
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MPR ASSOCIATES, INC. 1050 Connccticut Ave., NW - Washington DC 20036 i-
Title:
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- 1050 Connecticut Ave., NW - Washington. DC 20036
Title:
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Title:
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n [ MPR ASSOCIATES, INC. ! 1050 Connecticut Ave., NW - Washington, DC 20036 !
Title:
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1050 Connecticut Ave., NW - Washington DC 20036
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- 1050 Connecticut Ave., NW - Washington, DC 20036 l c '-
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1050 Connecticut Ave., NW . Washington DC 20036
'n'l?_ Date: H1'f lVi k
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Title:
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1 1 } MPR ASSOCIATES, INC. 1050 Connecticut Ave., NW - Washington, DC 20036 1
Title:
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MPR ASSOCIATES, INC. / 1050 Connecticut Ave., NW - Washington. DC 20036 g
Title:
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/W C' Date: '
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I i MPR ASSOCIATES, INC. 1050 Connecticut Ave., NW - Washington, DC 20036
Title:
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Title:
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1050 Connecticut Ave., NW - Washington, DC 20036 i
Title:
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1050 Connecticut Ave., NW - Washington, DC 20036 I;
Title:
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I: MPR ASSOCIATES, INC. 1050 Connecticut Ave., NW - Washington, DC 20036
Title:
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E MPR ASSOCIATES, INC. 1050 Connecticut Ave., NW - Washington, DC 20036
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Title:
tl '-A T %VG FL (d Eli' fr edM Calculated by: * /C Date:GM 2 /4 ,
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l MPR ASSOCIATES, INC. 1050 Connecticut Ave., NW - Washington, DC 20036
Title:
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;. 1050 Connecticut Ave., NW - Washington, DC 20036
Title:
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1050 Connecticut Ave., NW - Washington DC 20036 ri
Title:
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! MPR ASSOCIATES, INC.
1050 Connecticut Ave., NW - Washington. DC 20036
Title:
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T) MPR ASSOCIATES, INC. lt 1050 Connecticut Ave., NW - Washington. DC 20036 ,il
Title:
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-' MPR ASSOCIATES, INC. :
1050 Connecticut Ave., NW - Was'hington, DC 20036 l Pk" I'4L Plo fd
- C 'A42 - Date: d2 d73 s
Title:
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Title:
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1050 Connecticut Ave., NW - Washington. OC 20036
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Title:
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MPR ASSOCIATES. INC. 1050 Connecticut Ave., NW - Washington, DC 20036 D
Title:
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MPR ASSOCIATES, INC. 1050 Connecticut Ave., NW - Washington, DC 20036 h
Title:
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L fi .. MPR ASSOCIATES, INC. I, , 1050 Connecticut Ave., NW - Washington, DC 20036
Title:
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MPR ASSOCIATES, INC. 1050 Connecticut Ave., NW - Washington DC 20036
Title:
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Title:
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MPR ASSOCIATES, INC. 1050 Connecticut Ave., NW - Washington. DC 20036 1 Date: 5/.~ / r/ ~, N
Title:
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- Project: l T'< 'i l 2 93-61 Page II of /2_
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MPR ASSOCIATES, INC. I 1050 Connecticut Ave., NW - Washington, DC 20036 s
Title:
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1 3 MPR ASSOCIATES, INC. 1050 Connecticut Ave., NW - Washington, DC 20036
Title:
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y M [ k MPR ASSOClATES, INC. 1050 Connecticut Ave., NW - Washington, DC 20036 1 L'/Ar k G3 wik Loc 4 m t
Title:
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" MPR ASSOCIATES, INC.
1050 Connecticut Ave., NW - Washington, DC 20036 f
Title:
hit U4 ** K km WN*'s Calculated by: + ?E Date: 7/7/F3 _' No Po k sk'A C NW A +7 Checked by: T-
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i J MPR ASSOCIATES, INC.
; 1050 Connecticut Ave., NW - Wcshington, DC 20036 l
Title:
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m j MPR ASSOCIATES, INC. I Il 1050 Connecticut Ave., NW - Washington, DC 2003 6 1
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Title:
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1050 Connecticut Ave., NW - Washington, DC 20036
Title:
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f MPR ASSOCIATES, INC. [ 1050 Connecticut Ave., NW - Washington, DC 2003G
Title:
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