ML20236L345
| ML20236L345 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 07/30/1987 |
| From: | Chan T Office of Nuclear Reactor Regulation |
| To: | Office of Nuclear Reactor Regulation |
| References | |
| TAC-65471, TAC-65472, TAC-65473, NUDOCS 8708100220 | |
| Download: ML20236L345 (55) | |
Text
, _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _
o July 30, 1987 Docket No.:
50-344 DISTRIBUTION
~ Docket file /EJordan LICENSEE:
Portland General Electric Company (PGE)
NRC & Local PDRs/JPartlow PD5 RDF/ACRS (10)
FACILITY:
Trojan Nuclear Plant GKnighton/HBClayton JLee/TChan/T0 Martin (ED0)
SUBJECT:
MEETING
SUMMARY
REGARDING THE JUNE 16, 1987 OGC-Beth MEETING ON TECHNICAL ISSUES RELATED TO RESTART On June 16, 1987, the staff met with representatives of, and consultants for PGE to discuss the results of analyses related to the Accumulator Fill Line Failure, Main Feedwater Line Restraint Failure and the Main Steam Line Thin Wall issues at Trojan. A list of attendees is contained in Enclosure 1.
Main Feedwater Line Restraint Failure During the 1987 refueling outage, concrete in and around a seismic restraint (SR-8) on the Loop B feedwater piping inside containment was found to have failed. The support itself appeared to be undamaged. The concrete expansion anchor shells did not pull out of the concrete.
PGE performed analyses to determine 1) the load required to cause the observed failure; 2) the effects of a 1986 Integrated Leak Rate Test (ILRT) on the l
restraint; 3) potential causes of the restraint failure; and 4) the cause and l
effects of a " cold sprung" line on the failed restraint's mirror imaged counterpart (SR-4) on the "A" feedwater line. These analyses were submitted to the staff by letter dated June 12, 1987, and formed the bases for discussion.
PGE's presentation handouts are included as Enclosure 2.
PGE stated that the equivalent static load which caused the restraint's failure at the baseplate was not less than 40 kips, assuming 5000 psi compressive strength concrete, and that such a load would most likely have been generated by steam condensation-induced waterhammer. PGE also discussed their analytic methodology for evaluating thermal stratification and water hammer effects.
The staff found that this issue appeared to have been thoroughly investigated by PGE. However, it was pointed out that concrete with compressive strengths less than that assumed in PGE's equivalent static load calculation could significantly alter the minimum expected applied load, and as such, determined that further verification of concrete adequacy was warranted. PGE agreed to perform such additional verification, and has since determined from pour card records, that the assumed compressive strength of the concrete was valid. The licensee's assessment that a small waterhammer caused the failure of the re-l straint, and that the waterhammer was an isolated event, appears to be reason-l able, but will be evaluated further.
8708100220 B70730 PDR ADDCK 05000344 S
l Accumulator Fill Line Failure While transferring water from the "A" accumulator to the "D" accumulator through the fill lines, a rupture of the "A" fill line at the accumulator nozzle-to-pipe weld occurred. A metallurgical evaluation revealed that a low-cycle fatigue failure had occurred.
The piping was subsequently repaired and satisfactorily hydrostatically tested.
While repeating the water transfer operation, a loud banging noise was heard, and the operation was stopped. A second attempt was made and the loud banging noise continued.
Following a valve line-up check, the operation was started again, and the fill line nozzle weld for the "A" accumulator failed again.
The metallurgical evaluation for the second failure also indicated that the failure was low cycle fatigue.
It was demonstrated that the failure was caused by a cyclic generation of pressure pulses resulting from valve chatter from a packless diaphragm valve subjected to backflow under high differential pressures.
PGE evaluated nozzle stresses for three different flow conditions, and Bechtel Corporation evaluated the dynamic effects. These analyses were submitted to the staff by letter dated June 12, 1987, and provided the bases for discussion.
PGE's presentation handouts are included as Enclosure 3.
PGE conducted a backflow test through a packless globe valve similar to the one in the accumulator fill line, at their Boardman Coal Plant. The test resulted in a pipe failure from water hammer which occurred at about 56 gpm.
This test was captured on video tape; the tape was viewed during the meeting.
Results of analytic modelling and evaluation predicted piping stress levels to be consistent with the observed low cycle fatigue failure, and concluded that the accumulator fill lines could have permitted flow rates to exceed that which caused the piping failure at the Boardman Coal Plant. Stress calcuutions for the accumulator concluded that there was an insignificant loss of fatigue life for the accumulator and for the fill line nozzle.
The staff concluded that the root cause of the fill line failure appeared to have been properly demonstrated and that this issue appeared to have been throughly investigated and evaluated. The staff will evaluate the analyses, calculations, and the mock-up backflow test to ensure that they are technically sound.
Main Steam Line Below Required Min. Wall l
While perfoming ultrasonic wall thickness measurements on the "B" main steam line as a part of PGE's erosion / corrosion inspection program, the licensee j
discovered an area at the outlet of the steam flow venturi that was below the specified minimum wall thickness. Further examination revealed the same condition exists at the venturi outlet for the "A", "C", and "D" lines. The cause of the wall thickness discrepancy was attributed machining of the pipe spool in preparation for installation of the venturi during original construction.
In order to demonstrate the acceptability of continued use of the affected piping, fatigue and finite element analyses were perfonned by the licensee in
1
- l accordance with Subsection NB of Section III of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. These analyses were submitted to the staff by letter dated June 11, 1987, which provided the bases for discussion. PGE's presentation handouts are included as Enclosure 4.
Trojan's Code of Record for this portion of main steam piping is B31.7 " Nuclear Power Piping," Class 2 (1969 with Addenda to Summer 1971). The pressure design of this piping, according to B31.7, Chapter 2-II, "shall be in accordance with Division 104 of USAS B31.1.0," with noted exceptions.
PGE stated that, for the "as is" condition, Equations 11, 12 and 13 of B31.1,
" Power Piping" (1973) regarding longitudinal and thermal stress could be met; however, Equation 3 or 3a regarding hoop stress could not be met.
As a result, a more rigorous analysis was performed to determine general membrane stress and primary stress intensity. The pipe stress and fatigue evaluation was performed in accordance with ASME Section III, NB3600 (1983).
The results of these analyses showed that the maximum general membrane stress and the primary stress intensity were below the ASME Code allowables. The analyses also determined that the cumulative usage factor was acceptable, t
The staff did not agree with PGE's assertion that the use of analytic techniques and stress allowables permitted by ASME III Subsection NB rules and requirements were, by themselves, appropriate in demonstrating equivalency to USAS B31.7 Class 2 rules and requirements. The staff recognizes that although the use of more rigorous analytic methods may be one facet of the design / construction j
process which may justify the use of higher stress allowables, we also feel that the contribution of material " quality" toward added justification of permitting higher stress allowables, that is provided under the categories of fabrication, inspection and testing cannot be ignored. The staff expressed that selective use of Subsection NB provisions cannot be employed to demonstrate compliance with USAS B31.7 Class 2 requirements.
Resolution of this matter will require additional discussion, and will be addressed in future correspondence.
Original signed by Terence L. Chan, Project Manager Project Directorate V Division of Reactor Projects - III, IV, V and Special Projects cc w/ enclosures:
As stated A'
DRS /'hton DRSP/PD
- PD5 C#
TChan:cd 7/q/87 7/g/87
,,,e.
(
4 Mr. David W. Cockfield Portland General Electric Company Trojan Nuclear Plant CC:
Senior Resident Inspector U.S. Nuclear Regulatory Commission Trojan Nuclear Plant Post Office Box 0 Rainier, Oragon 97048 Michael J. Sykes, Chairman Board of County Commissioners Columbia County St. Helens, Oregon 97501 Mr. David Kish Oregon Department of Energy Labor and Industries Building Room 111 Salem, Oregon 97310 Regional Administrator, Region V U.S. Nuclear Regulatory Commission 1450 Maria Lane, Suite 210 Walnut Creek,' California 94596 e
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LIST OF ATTENDEES NRC/PGE MEETING RECORDING START-UP ISSUES l
JUNE 16, 1987 l
NRC PGE BECHTEL l
T. Chan T. Walt T. Bostrom G. Knighton A. Roller R. Fosse S. Hou R. Wehage A. Arastu L. Marsh T. Bushnell K. Wichman B. Elliot A. Serkiz IMPELL P. Wu l
J. Ma W. McLeod l
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1 MAIN FEEDWATER PIPING INSIDE CONTAINMENT I
RESTRAINT FAILURE EVALUATION i
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BACKGROUND I
1 SEISMIC RESTRAINT EBB-3-I-SR8 ON LOOP B WAS FO o
HAVE A FAILED STRUCTURAL ATTACHMENT TO THE C THE FEEDWATER PIPING IS 14 INCH (NPS) SCHEDULE 60 o
A-106B PIPING.
THE PIPING RUNS FROM THE CONTAINMENT PENETRATION TO THE STEAM GENERATOR.
MOST OF THE PIPING j
IS HORIZONTAL, WITH VERTICAL RUNS NEAR THE CONTAINMENT {
PENETRATION AND THE STEAM GENERATOR.
SR-8 IS THE ONLY f
f RIGID RESTRAINT ON THE PIPING SYSTEM.
i 1
ANALYSES WERE UNDERTAKEN TO IDENTIFY CAUSES ND AFFECTS o
THE DAMAGE TO THE SYSTEM IS LIMITED TO FAILURE.0 o
CONCRETE EXPANSION ANCHOR BOLT CONNECTIONS TO ISOLATED RESTRAINT THAT IS THE LIMITING COMPONEN j
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ANALYSES PERFORMED
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s THERMAL EXPANSION LOADINGS o
RESTRAINT / SNUBBER LOCKING o
l (INTEGRATED LEAK RATE TESTING (ILRT) o i
o EXCESSIVE PRELOAD 1
2 THERMAL STRATIFICATIONIB0 WING o
o.WATERHAKMER l
1 ISOLATION VALVE CLOSURE o
o CHECK VALVE CLOSURE o' CONTROL VALVE CLOSURE STEAM CONDENSATION-INDUCED o
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RESULTS OF ANALYSES NORMAL THERMAL LOADS COULD NOT CAUSE FAILURE o
l SNUBBER LOCKINGlRESTRAINT BINDING COULD NOT CA 0
FAILURE ILRT AFFECTS COULD NOT CAUSE FAILURE o
EXCESSIVE PRELOAD DID NOT EXIST o
THERMAL STRATIFICATION COULD RESULT IN SIGNIFICA o
LOADINGS ON SR-8 WATERHAMMER FROM THE VARIOUS VALVE CLOSINGS WERE o
E. VALUATED AND SHOWN TO NOT RESULT IN S LOADINGS STEAM CONDENSATION-INDUCED WATERHAMMING UNDERG j
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SYSTEM WAS OPERABLE FOR DESIGN LOADINGS WITH 0
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. I
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ANALYSIS FOR HYPOTHETICAL THERMAL B0 WING SINCE SR-8 IS THE ONLY RIGID RESTRAINT IN THE SYSTEM, o
.AND ITS CLOSE PROXIMITY TO THE CONTAINMENT PENETRATIO COULD RESULT IN A MOMENT COUPLE TO RESTRAINT ANY THERMAL BOWING, AN ANALYSIS FOR AN ASSUME THERMAL STRATIFIED FLOW WAS UNDERTAKEN HYPOTHETICAL TOP TO BOTTOM THERMAL GRADIENT OF l00F o
ASSUMED.
o ANSYS STIF 20 (PLASTIC STRAIGHT PIPE) AND STIF 60 (PLASTIC CURVED PIPE) ELEMENTS USED l
o ANALYSIS PERFORMED WITH AND WITHOUT SR-8 ACTIVE RESULTS p
o SIGNIFICANT THERMAL BOWING MAY OCCUR j
o SR-8 MAY ACT AS COUPLE THERMAL RESTRAINT WITH THE
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CONTAINMENT RESULTING IN SIGNIFICANT THERMAL BOWING LOADS ON SR-8 l
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LOADING ON SR-8 WAS 5.399 KIPS o
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PARAMETRIC ANALYSIS FOR HYP0THETICAL WATERHAMMER i
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o HYP0THETICAL WATERHAMMER ASSUMED 100 PSI oAP o
TWO ASSUMED RISE-TIMES o
.001 SEC o
.01 SEC LINEAR-ELASTIC TIME HISTORY ANALYSIS PERFORMED o
o WITH SR-8 o
WITHOUT SR-8 I
RESULTS l
LOW FREQUENCY OF SYSTEM RESULTS IN SIGNIFICANT DYNAMIC 1
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ISOLATION - MAX PIPE STRESS LOW l
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PIPING RESPONSE WITH AND WITHOUT SR-8 ACTIVE IS SIMILA o
0 LOADING ON SR-8 IS 6.4 KIPS l
LOADINGS ON OTHER RESTRAINTS SUCH THAT SR-8 IS 0
LIMITING COMPONENT l
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>7 FEFDWATER LINE ANALYSIS LOAD AT SR-8 DUE:70 CONTAlyMENT ILRT
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PURPOSE DURING THE INTEGRATED LEAK RATE TEST FOR THE j
CONTAINMENT BUILDING, THE STEEL LINER EXPANDS I
DETERMINE THE LOAD ON SR-8,
l OUTWARD UNDER LOAD.
DUE TO THIS OUTWARD MOVEMENT AT THE CONT AINMENT PENETRATION.
I CALCULATION J
MODEL WAS DEVELOPED USING PIPE CONFIGURATION FROM THE AS-BUILT ISOMETRIC.
PIPE WAS MODELED FROM THE
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CONT AINMENT PENETRATION TO THE N0ZZLE ON THE STEAM GENERATOR.
NUP~1PE STRESS ANALYSIS PROGRAM WAS USED.
THE CONTAINMENT DEFLECTION OF 0.140 INCH WAS INPUTED AS AN ANCHOR MOVEMENT AT THE PENETRATION.
RESULTS l
LOAD AT SR-8 IS 1626 POUNDS FORCE 1
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RADIAL DISPLACEMENT AT AZIMUTH 270'
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.f FEEDWATER LINE ANALYSIS LOAD AT SR-8 DUE TO CONTAINMENT ILRT l
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SAFET1 INJECTION ACCUMULATOR FILL LINE EVALUATION i
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BACKGROUND THE SAFETY INJECTION FILL LINE TO THE A ACCUMULATOR o
EXPERIENCED A FAILURE TWICE TO THE PIPING TO ACCUMULATOR A N0ZZLE SOCKET WELD HEAT AFFECTED ZON THE PIPING IS 1 INCH (NPS)
SCHEDULE 40S, A-312 TP 304 o
BOTH FAILURES OCCURRED DURING SLUICING OPERATIONS o
1 ACCUMULATOR A TO ACCUMULATOR D BACKFLOW THROUGH THE Y-PATTERN PACKLESS METAL DIAPH l
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-(PMD) VALVE WAS SUSPECTED AS CAUSE OF THE FAILURE ANALYTICAL EVALUATIONS WERE PERFORMED TO DETERMINE i
o THE REVERSE FLOW COULD HAVE RESULTED IN THE OBSERVED FAILURES i
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EVALUATION i
I HYDRAULIC FORCING FUNCTIONS WERE DEVELOPED BASED Ul o
PMD VALVE CLOSURE IN.D05 SECONDS AT 56 GPM BACKFL 0
l SIGNIFICANT TRANSIENT WAS PRDDUCED l
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LINEAR-ELASTIC TIME HISTORY ANALYSIS PERFORMED BASED 0
f UPON ANALYTICALLY DERIVED FORCING FUNCTIONS i
RESULTS PREDICTED STRESS LEVELS CONSISTENT WITH OBSERVED LOW o
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ACCUMULATOR FILL LINE ANALYSIS FLOW ANALYSIS PURPOSE TO DETERMINE ANALYTICALLY WHETHER FLOW RATES HIGH ENOUGH TO CAUSE DIAPHRAGM VALVE DISC CHATTER COULD HAVE OCCURRED.
CRITERIA FOR SELECTION OF FLOW CONDITIONS FOR WHICH ANAtYSIS WOULD BE PERFORMED 1.
ANY CONDITION WHICH RESULTS IN REVERSE FLOW THROUGH l
2.
CO.NCENTRATE ON THOSE CONDITIONS WHERE UNUSUALLY.HIGH FLOW RATES COULD OCCUR.
CONDITIONS SELECTED FOR ANALYSIS 1.
WATER TRANSFER BETWEEN TWO ACCUMULATORS USING THE l
INTERCONNECTED FILL LINES.
2.
WATER TRANSFER BETWEEN TWO ACCUMULATORS OSING THE INTERCONNECTED SAMPLE LINES.
3.
WATER TRANSFER FROM AN ACCUMULATOR TANK TO AN f'
OUT-OF-CONTAINMENT CHEMICAL AND VOLUME CONTROL
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SYSTEM HOLDUP TANK.
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ACCUMUL ATOR FILL LINE ANALYSIS
/
FLOW ANALYSIS ASSUMPTIONS CONDITIONS NO. 1 AND No. 2 WATER IS TRANSFERRED FROM THE "A" ACCUMULATOR TO THE "D" ACCUMULATOR CONDITION No. 3 WATER IS TRANSFERRED FROM THE "A" ACCUMULATOR TO THE "A" CHEMICAL AND VOLUME CONTROL SYSTEM (CVCS) HOLDUP TANK SOURCE IANK (ACCUMULATOR "A") IS AT 600 POUNDS PER SOUARE INCH GAGE RECEIVING TANK (EITHER ACCUMULATOR "A" OR CVCS HUT "A")
IS AT ATMOSPHERIC PRESSURE WATER IS AT AMBIENT TEMPERATURE
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FLOW IS STEADY-STATE, NON-CHOKED
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i ACCUMULATOR FILL LINE ANALYSIS FLOW ANALYSIS CONDITION No. 1 f
WATER TRANSFER BETWEEN TWO ACCUMULATORS USING INTERCO i
FILL LINES MODEL INPUTS 30 FEET OF SCH. 40, 1 INCH PIPE 2 90* ELBOWS, 1 INCH 2 45' ELB0WS, 1 INCH 1 TEE, FLOW THRU BRANCH, 1 INCH 1 PIPE ENTRANCE, 1 INCH 1 PIPE EXIT, 1 INCH l
2 GLOBE, Y PATTERN, DIAPHRAGM VALVES, 1 INCH i
ONE WITH FORWARD FLOW, CV = 7.1 ONE WITH REVERSE FLOW, CV = 6.5 CONTROL VALVES, 1 INCH 2 GLOBE, STRAIGHT PATTERN, CV = 15 (SAME FOR BOTH FORWARD AND REVERSE FLOW)
CALCULATION METHOD
/
IN-HOUSE COMPUTER PROGRAM."L10-PIPE" f
USES TWO K METHOD (ADJUSTS X VALUE AS A FUNCTION OF REYNOLDS NUMBER)
RESUlTS FLOW RATE - 94.68 GALLONS PER MINUTE
ACCUMULATOR Fill LINE ANALYSIS FLOW ANALYSIS i
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t ACCUMut_ATOR Fill IINE ANALYSIS l
FLOW ANALYSIS CONDITION No. 2 WATER TRANSFER BETWEEN TWO ACCUMULATORS USING THE INT SAMPLE LINES MODEt INPUTS 106 FEET OF SCH. 40, 3/4 INCH PIPE 14 90 ELB0WS, 3/4 INCH l
4 45 ELBOWS, 3/4 INCH i
5 TEES, FLOW THRU BRANCH, 3/4 INCH 1 PIPE ENTRANCE, 3/4 INCH 1 PIPE EXIT, 3/4 INCH 2 GLOBE, Y PATTERN, DIAPHRAGM VALVES, 3/4 INCH ONE WITH FORWARD FLOW, CV = 5.6 ONE WITH REVERSE FLOW, CV = 4 2 GLOBE, STRAIGHT PATTERN, MOTOR-0PERATED VALVES, 3/4 INCH CV = 6 (SAME FOR BOTH FORWARD AND REVERSE FLOWS) l l
l CALCULATION METHOD IN-HauSG COMPUTER PROGRAM " LIQ-PIPE" h
k USES TWO K METHOD (ADJUSTS K VALUE AS A FUNCTION OF REYNOLDS NUMBER)
RESULTS FLOW RATE - 44.74 GALLONS PER MINUTE
i-i ACCUMULATOR Fil1 LINE ANALYSIS FLOW ANALYSIS l
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FJLL LINE TO S!Y l
g-TO B & C TEST LINE l
r-ACCUMULATORS l
CONDITION No. 2 FLOW PATH ACCUMULATOR "A" 10 ACCUMULATOR "D" VIA THE SAMPLE LINES
t ACCUMULATOR FILL LINE' ANALYSIS FLOW ANALYSIS CONDITION NO. 3 WATER TRANSFER FROM AN ACCUMULATOR TANK TO AN OUT-OF-CONTAINMENT CHEMICAL AND VOLUME CONTROL SYSTEM HOLDUP TANK MODEL INPUTS PIPE AND PIPE ELEMENTS GREATER THAN 1 INCH WERE IGNORED 1 INCH PIPE AND FITTINGS 104 FEET OF SCH. 40 AND SCH. 80 PIPE 5 90 ELB0WS 3 45 ELBOWS 1 ENTRANCE 4 TEES, FLOW THRU THE BRANCH 1 GLOBE, Y-PATTERN, DIAPHRAGM VALVE CV = 6.5 (REVERSE FLOW) 1 GLOBE, STRAIGHT PATTERN, CONTROL VALVE CV = 15 ORIFICE (0.304 INCH) 3/4 INCH PIPE AND FITTINGS 25 FEET OF SCH. 40 AND SCH. 160 PIPE 12 90 ELBOWS 4.45 ELBOWS 4 1EES, FLOW THRU THE RUN 6 TEES, FLOW THRU THE BRANCH 1 GLOBE, Y-PATTERN, DIAPHRAGM VALVE CV = 5.6 (FORWARD FLOW) 2 GLOBE, STRAIGHT PATTERN, CONTROL VALVE j
k CV = 10.6 CALCULATION METHOD IN-HOUSE COMPUTER PROGRAM " LIQ-PIPE" USES TWO K METHOD HEAD LOSS THRU ORIFICE DETERMINED USING IN-HOUSE COMPUTER PROGRAM "0RIFICE" USES METHODS OUTLINED IN ASME PAPER MFC-3M-1984 RESULTS FLOW RATE - 26.28 GALLONS PER MINUTE J
l
4 s
ACCUMULATOR FILL LINE A_ALYSIS N
ACCUMULATOR N0ZZLE STRESS CALCULATION PURPOSE TO EVALUATE THE CONDITION OF THE N0ZZLE ON THE ACCUMULATOR TANK FOR THE 1 INCH FILL LINE AS A RESULT OF TWO LINE I
SPECIFICALLY, THE EFFECT ON FATIGUE LIFE WAS FAILURES.
DETERMINED.
APPROACH CALCULATE THE MOMENT RANGE REQUIRED TO PRODUCE A PART I BENDING STRESS EQUAL TO TWICE THE YIELD STRESS PART II USING THESE MOMENTS, CALCULATE THE STRESS INTENSITIES AT THE N0ZZLE AND AT THE VESSEL PART III USING THESE STRESS INTENSITIES, DETERMINE THE FATIGUE FACTORS i
e e
____.m_.--__-__m_.-_._._..___n_____-mm__._.___m.
m
_ _ _.=_
i
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ACCUMUL ATOR Fill LINE ANALYSIS ACCUMULATOR N0ZZLE STRESS CALCULATION
/
T205A l
Location Location (2)
(1) l l
I THIS END CORRESPONDS l
TO A 1" 3000' 500KET VELD COUPLING l
3 l
g I
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I I
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i A
B i
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_._.___._._._._._._._._.;_.___._ r_._._._.
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AS-MEASURED DIMENSIONS:
A s 1.760" B
1.350" C s.457" D s.833" E s 1.338" r s.ve" "A" ACCUMULATOR FILL LINE N0ZZLE
e H
i ACCUMULATOR FILL LINE ANALYSIS ACCUMULATOR N0ZZLE STRESS CALCULATION CALCULATION - PART 1 l
1 APPROACH CALCULATE THE MOMENT RANGE REQUIRED TO PRODUCE A BENDING STRESS EQUAL TO TWICE THE YIELD STRESS RESULT $
BASED ON A YIELD STRESS OF 42 KSI FOR TYPE 304 STAINLESS STEEL, A MOMENT LOADING OF 4979 INCH POUNDS FORCE WOULD BE REQUIRED TO CAUSE A STRESS EQUAL TO TWI.CE THE YIELD STRENGTH AT THE PIPE-TO-N0ZZLE WELD.
A MORE CONSERVATIVE MOMENT WILL BE USED BY APPLYING THE STRESS INDEX AT THE SOCKET WELD OF 2.1.
RESULTANT MOMENT - 10,456 INCHES POUNDS FORCE.
USING A 12 INCH MOMENT ARM, THE LOAD CORRESPONDING TO THIS,M0 MENT IS 871 POUNDS FORCE-APPLIED AT THE VERTICAL RISER PIPE.
THE MOMENT AT THE N0ZZLE-TO-VESSEL INTERFACE 13 INCHES FROM THE RISER PIPE IS 11,328 INCHES POUNDS FORCE.
h b
ACCUMULATOR FILL LINE ANALYSIS ACCUMULATOR N0ZZLE STRESS CALCULATION CALCULATION - PART 11 J
l APPROACH USING THE MOMENTS CALCULATED IN PART I, CALCULATE THE COMBINED MEMBRANE AND BENDING STRESS INTENSITIES AT THE OUTER SURFACE OF THE N0ZZLE AND THE VESSEL.
RESULTS USE THE METHODS IN WELDING RESEARCH COUNCIL BULLETIN 297.
COMBINED MEMBRANE AND BENDING STRESS INTENSITY AT THE J
VESSEL WALL:
1,484 POUNDS PER SOUARE INCH (PSI).
COMBINED MEMBRANE AND BENDING STRESS INTENSITY AT THE N0ZZLE:
31,303 PSI.
VESSEL STRESS INTENSITY IS WELL BELOW THE FATIGUE LIMIT OF 14,000 PSI.
THEREFORE, NO LOSS IN FATIGUE LIFE HAS OCCURRED.
N0ZZLE STRESS INTENSITY IS ABOVE THE FATIGUE LIMIT OF Uk FA bRW B
PkRORMED j
l
O i
ACCUMULATOR FILL LINE ANALYSIS ACCUMULATOR N0ZZLE STRESS CALCULATION CALCULATION - PART III APPROACH b
USING THE STRESS INTENSITIES CALCULATED IN PART. 'll FOR l
THE N0ZZLE, DETERMINE THE FATIGUE USAGE FACTORS.
l RESULTs f
FATIGUE USAGE FACTORS.WERE CALCULATED.AT TWO LOCATIONS:
1)
AT THE BOTTOM 0F.THE N0ZZLE SOCKET; l
AND 2)
AT THE VESSEL-TO-N0ZZLE INTERFACE.
l CALCULATE THE PEAK STRESS INTENSITY RANGE USING ASME III, NB-3653.2 FOR LOCATION (1) 87 KSI FOR LOCATION (2) 53 KSI AT THE BREAK POINT - 173 KSI SINCE THE FATIGUE USAGE FACTOR AT THE BREAK POINT IS, BY DEFINITION, EQUAL TO 1.0, WE ARE ABLE 10 DETERMINE THE NUMBER OF CYCLES FROM THE S-N CURVE. ' TOTAL CYCLES IS 2500 FOR EACH PIPE FAILURE.
TWO FAILURES - USE 5000 CYCLES.
l APPLYING 5000 CYCLES, CALCULATE THE FATIGUE USAGE f
I
[
FACTORS:
~
FOR LOCATION (1) - 0.083 FOR LOCATION (2) - 0.004
ACCUMul_ ATOR FILL i INE ANAL _YSIS ACCUMULATOR N0ZZLE STRESS CALCULATIONS i
CONCLUSIONS i
1 NO LOSS OF FATIGUE LIFE FOR THE VESSEL i
i
(
l INSIGNIFICANT EFFECT ON FATIGUE LIFE FOR THE FILL l
LINE N0ZZLE I
1 4
1 r
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k MAIN STEAM WALL THICKNESS EVALUATION 6
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I.
BACKGROUND AREAS WITH-REDUCED WALL THICKNESS IDENTIFIED IN o
BETWEEN FLOW ELEMENT AND ELBOW.THE REDUCED WALL WAS AS A RESULT OF MACHINING THE ID OF THE PIPING FOR INSTALLATION OF THE FLOW ELEMENT PIPING DESIGNED PER USAS B31.7. 1969 EDITION WITH l
0 ADDENDA THROUGH SUMMER 1971 PIPING IS CLASSIFIED AS CODE CLASS 2 0
PIPING IS 28" A-155 KC 70. CLASS I SEAM WELDED, MIN o
WALL OF.858" PURPOSE OF THE EVALUATION IS TO DEMONSTRATE THA o
PIPING IS ACCEPTABLE FOR RETURN TO SERVICE l
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II.
EVALUATION APPROACH-PERFORM DETAILED ANALYSIS TO DEMONSTRATE PIPING IS o
ACCEPTABLE FOR SERVICE FROM THE INTRODUCTION OF USAS B31.7 THE SPECIFIC DESIGN REQUIREMENTS OF THE CODE USUALL REVOLVE AROUND A SIMPLIFIED ENGINEERING APPROACH TO SUBJECT.
IT IS INTENDED THAT A DESIGNER CAPABLE OF APPLYING MORE COMPLETE AND RIGOROUS ANALYSIS TO S UNUSUAL PROBLEMS SHALL HAVE LATITUDE IN THE DEVEL SUCH DESIGNS AND THE EVALUATION OF COMPLEX OR COMB STRESSES.
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III.
EVALUATION CRITERIA l
DEMONSTRATE PRIMARY GENERAL MEMBRANE STRESSES ARE LES 0
I THAN SM DEMONSTRATE LOCAL MEMBRANL STRESSES ARE LESS THAN o
SM DEMONSTRATE PRIMARY STRESS INTENSITY IS LESS THAN o
o 1.5SM FOR NORMALlUPSET o
2.25SM FOR EMERGENCY o
3.0SM FOR FAULTED DEMONSTRATE THAT THE CUMULATIVE USAGE FACTOR (CUF) IS o
l LESS THAN l.0 i
DEMONSTRATE THAT ANSI B31.I EQUATIONS II, I2. AND 13 i
o ARE ACCEPTABLE.
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l IV.
STRESS EVALUATION i
A.
PRIMARY GENERAL MEMBRANE STRESS PERFORMED 5-D FINITE ELEMENT ANALYSIS USING AN o
ANSYS STIF 45 (3-D ISOPANAMETRIC SOLID ELEMENT) o THIN WALL REGIONS BELOW MIN WALL MODELED BA 1" x I" GRID UT DATA MINUS.010".
THIN WALL REGIONS ABOVE MIN WALL MODELED AS MIN WALL (
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1 RESULTS-1 MAX PRIMARY GENERAL MEMBRANE STRESS INTENSIT o
3-D MODEL WAS 17.649 KSI (SM - 19.456) 1 O
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~ROJAN PROJECT - 29" PIPE MINIMUM WA a
1 3-D FINITE ELEMENT MDDEL n
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IV.
STRESS EVALUATION (CONT'D) 1 13.
PRIMARY STRESS INTENSITY AND FATIGUE ANALYSIS i
I
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t 0
ANALYSIS' PERFORtiED PG HB-3500
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SECTION MODULUS FOR PRESSURt AND MOMENT L0fDS ASSUMED T 0.8bO" 1
l AND3Tg[DASEDUPONTl=1.000" o AT I
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IV.
STRESS EVAL'.lATION (c0NT'D) i THREE SETS OF STRESS INDICES USED o
t 4
LONGITUDINAL BUTT WELD AND GIRTH BUTT WELD o
o ELB0W l
0 TAPERED TRANSITION RESULTS PRIMARY STRESS INTENSITY IS ACCEPTABLE i
o DESIGNINORMALlUPSET STRESS INTENSITY WAS o
i 18.01KSI429.18KSI'(i.5Sg)
-l o
FAULTED STRESS INTENSITY WAS i
20.13 KSI(-58.37 KSI (3S )
M I
FATIGUE (USAGE FACTOR) IS ACCEPTABLE o
]
i f
CUMULATIVE USAGE FACTOR (CUF) IS.04(1.0 o
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,s IV.
STRESS EVALUATION (CONT'D)
C.
ANSI B31.1 /J'ALYSIS PRESSURE AND MOMENTS LOADINGS ASSUME T.780" o
EVALUATED EQUATIONS II (SUSTAINED LOADINGS), 12 o
(OCCASIONAL LOADINGS) AND 13 (THERMAL EXPANSION)
RESULTS o
ALL EQUATIONS SATISIFIED l
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y 7
0 55 0
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6 7
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B 0
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7 10 s
3 1
s 7
r 3
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A B
C A
8 D
E 5
7 P
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tee i
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ss tE u ss a as 1
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5.
t V.
CONCLUSIONS l
PIPING SATISFIES EVALUATION CRITERIA 0
PIPING SUITABLE FOR RETURN TO SERVICE o
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