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Discusses Allegation Received from Former Welder Concerning Quality of Welds Performed on Main Steam Bypass Line Replacement.Related Info,Including 860117 Briefing Package for V Stello,Insp Plan & Deviation Event Rept,Also Encl
	ML20210U435
Person / Time
Site:	Fermi, 05000000
Issue date:	12/09/1985
From:	Nicole Parker; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:	Weil C; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML20210T193	List: ML20210T187; ML20210T243; ML20210T439; ML20210T483; ML20210T532; ML20210T569; ML20210U435;
References
	FOIA-87-50 NUDOCS 8702180481
	Download: ML20210U435 (91)
	v • d • e

{{#Wiki_filter:.. 1 .4 Detenbre 9, 1985 MEMOIANDUM FOR: C. H. Wc21, Investigotion end Compliance Specialist TH:<U: F'. t;. Byron, Seni or Rctident Ir:t. pert or, Fe tri :? FRDM: M. E. F eri:c: Resicient Inspcc. tor, Fermi ? EUI/ JECT: A'.LE6/0 I D:J FERMI, WELDE DlJ MAIN STEAM LYPASS LINE O. December 9, 1905. ot E: 4 5 c. tv.., I rc:ca ved a call f rcim a f o: mer wc2 dor e c.. t. r i n g t l i:- cr.sr 12 t y o# the we)cc per#crmed on ths Mrin 5'. w a r. I y; oc.n Li ne r rr1 ccoment. Tt.e i nc.i vi dual wr u upc.c i f a r.ol l y e c., c t

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'2/2-MmN M iqg6 SCHEDULED DATE(s): ; INSPECTOR (s): -E b, g,,-. s~._g gogg m. N WODULE NO. % COMP,[ INSPECTION FOCUS ASSIGNED TO 9 21cl [ REJ 6vJ oF' ucester 's TTeceu / resma cuasr wces *+*v=/ l t oH tcH bc / jot ~ / GLEcrN ~~~ A rJ V / [ k 5 OSw.(C 00ALI FI G ATI 0.'s) [ J \\ exne~s rac-sas.s9-J \\ AV625 -rw E Pt'Te D. - # ~ ~v f A 5 D / 9'Z.~7DI REv t EM RE9t AEmnaT-T.w esuJ ) oG' Nod S A FC'T Y - (2 e.~.LA70 / so" maw celemm / \\ (by Pass Ei Ptro6 4Mb. / RGn A9 IUGR RENLT/ \\ Y n ^ \\ nn h u..nn f vA.VJ/M M ^""1 T 'GerwfR 5037 L M nbeck nNW O v 1 PLAN PREPARED BY:W GO F'u A c^J DATE: / 7//86 3 PLAN _ APPROVED BY: (Wbd O/b-DATE: /// 2 < //." M ) PLAN REVIEWED BY: DATE: (( m PROJECT SECTION CHIEF mrAantwe 9275MMu3ser

_ : __.___.n ' ' ~ ~ " ~' DAILY CEP3GT A DATE: 09/19/85 i ~ ~~ ITN OR EVENT ~ ~~ REGIONAL ACTION w /*\\ A FCCELIYY7LI 5 SEE' NOTIFICATION (4p 0FFICE OF THE RESIONAL ADMINISTRATOR pm. JAMES 9. REPPLER WILL ADORESS THE CHICA00 INFORMATION .SECTION/AMERICAN NUCLEAR SOCIETY AT.A DINNER.MELIING IN CHICAGO. OIVISION OF REACTOR PROJECTS SRI L REGIONat THF LICENSEE WAS OPERATING THE UNIT ON 9/IT/OS AT 4.It InPECT.OR FERMI 2 SRI FAX power.AND.650 PSIS WITH THE STEAM SYPASS. VALVES PARTIALLY OPEN WHEN A LEAK WAS 08 SERVE 0 IN THE EAST FOLLOWINS ife m**r's Act10NR l. _ _ CTEAM WYPASS LINE. WHILE ATTEMPTING.TO.lSOLATE_INF EAST RYPASS STEAM LINE ANO OPEN THE WEST SYPAss VALVE, j A_TRANSILNI PDWER SPI AE WAS 08SERVEDe__ POWER _Pramra SEVERAL SECON05. INVESTIGATION SY THE OPERATORS S.75 FOR REVEALEO SUSPECT.SNUSSER DAMAGE.AT_.THE.SNUSSER_AN0_PIPr IN THE EAST BYPASS STEAM LINE. LONGITUDINAL CRACRS clamp -.- WERE. OklSERVED WHERE THE. SNURSER. ORACKET IS. RESTR AINED_BY LUGS ON THE PIPE AFTER THE INSULATION WAS REMOVED. THE I AFFECTED.SECTION OF PIPING.1S_IN_THE.MORIZONT*!_ *L* DOWNSTREAM OF THE EAST STEAM SYPASS VALVE. THE LICENSEE ALSO OWSERVED INSTRUMENT. ROOT.VALVLDAMAGE-THE LICENSEE 08 SERVE 0 INDICATIONS.0F.A_ LEAK _.IN_IME_ NEST BYPAS$ STEAM LINE. INSULATION IS BEING REMOVED TO INVEST! SATE. THE EXTENT..._.NO CAUSE._HASJEEN.DEIERMIAlan Fnu THE FAILU9E AND THE LICENSEE IS CONTINUING THEIR . _. _. _. _.. INVESTIGATION. . AT 10t!5 AM. CST 04 9/IS/8Se _WHILL OPERATING.AT. 905_ RESIDEftf INSPECIDR ..DUANE ARNOLO. ____..TELECON FRON.R1 ANO PRESENT DURINe HOIDO ON 9/19/05 POWER THE RCIC SYSTEM FAILED ITS SPECIAL WEEKLY INIIt _ n_ = a SURVEILLANCE TEST WHEN_THE_. MOT 0A_0PERATED_VALVF HETWEEN THE RCIC TURBINE AND THE REACTOR VESSEL FAILED FOLLOWUP TESTING 70 OPEN..THE THERMAL OVERLOA05 ON THE MOTOR _ TRIPPED AND IT WAS THEN DETERMINED THAT THE CURRENT SEING DRAWN WAS HIGMER THAN NORMAL.._THE HPCI SYSTEN_WAS IMME0!ATELY TESTED ANO VERIFIED To SE OPERABLE. THE. UNIT. POWER WAS. REDUCED TO.405. TO..PERMLT._ENJRY INTO THE STEAM TUNNEL'WHERE THE RCIC VALVE IS . LOCATED. TESTINS VERIFIED THAT THE. VALVE.TOROUE SWITCH HAD FAILED ANO WAS DRAWI46 EXCESSIVE CURRENT. THE 10 ROUE Sw!TCH WAS REPLACED AND THE VALVE.. SATISFACTORILY TESTED. UNIT POWER HAS 8EEN RETURNED i.. _.... _ _ _.. ....T0_905. 4 4.._._._____..____......__.._.. p .. gem q. _ew e... a

6 FERMI 2 BRIEFING PACKAGE FOR VICTOR STELLO. ACTING EDO Friday January 17, 1986-l ( I,/

? Fermi History Low Power License March 20, 1985 i Premature Criticality July 1, 1985 Event Comission Briefing July 10, 1985 Full Power License July 15, 1985 l Unit Shutdown for Planned October 11, 1985 Outage l Projected Availability Mid to late February,1986 l for Startup i ... - - - - -. = - - -

/ Technical Issues Diesel Generators January,19P3 Failure Lack of pre-lubrication New Failures Lack of proper engine break in* Foreign material in lube oil

Mis-alignment during repair
Turbine Bypass Lines Cracking - large diameter thin wall piping plus high frequency acoustic vibration - piping replaced, RIII inspected.

South Reactor Feedpump Damaged bearings and pedestal Turbine i4RC contractor to review (Jan.1986) TIP Purge Line Penetration isolation does not meet GDC-56. l l Temporary modification being made and T.S. exemption requested. Unofficial Licensee desigr.ated root cause(s). l t 1

Technical issues (continued) Condensate Storage Tank Tank weld ruptured apparently due to excessive fill rate. Released low level activity water to pit /birm. No plans to repair during present outage. Remote Shutdown (3-L) Cable separation / backup power supply Panel issues. Licensee's proposed actions need to be reviewed and accepted by NRR. Environmental Qualifications / - Allegation on lack of review. RIII Review inspected - No problems identified. Seismic Review Allegation on lack of review. RIII continuing to inspect. Resolution expected by 1/17/86. Initial review-l no pr.-blems with hardware. Concrete Embedment As-built loads greater than those assumed during initial design. Licensee has evaluated - no problems requiring i j modifications. RIII inspection in progress. l l

(. Technical Issues (continued) Security Issues Fifteen violations of security' plan-Enforcement Conference scheduled for 1/17/86. Escalated enforcement action anticipated but not yet finalized. Premature Criticality Special Report 50-341/85040 on technical (7/1/85) issues issued. Enforcement Package to Headquarters (12/13/85). 01 & OIA reports not yet issued. l l l l

Regulatory Actions Confirmatory Action Letter Issued July 16, 1985. Licensee has completed all items. 5% power limitation still in place. j Escalated Enforcement Packages Premature criticality event, LCO violations, License condition violation. Special report issued 1/7/86. Enforcement Package forwarded to Headquarters 12/13/85 covering technical issues, but not potential wrongdoing issue. Security items - Enforcement-Conference scheduled 1/17/86. 10CFR 50.54(f) Letter Issued 12/24/85. Decision to issue concurred in during November meeting among W. Dircks, J. Taylor, H. Denton, J. Keppler and G. Cunningham, l s l I

+ Regulatory Action (continued) 10CFR 50.54(f) Letter (cont) Anticipate Improvements Improved Management awareness Impraved Management support. Improved Training at all Levels Assurance that unit can support each increase in power for testing Improved Regulatory and Operational performance

C / Performance Issues Excessive Personnel Errors Security Problems Failure to Disposition known Problems Lack of Candor Ineffective Communications Weak Management Chain i l \\

. Ih Personnel Errors Review of 78 Total LERs March 20 to November 25, 1985: 41 of 78 LERs caused by Personnel Errors July 10 to September 10, 1985 9 of 25 LERs caused by Personnel Errors + l September 10 to November 23, 1985 l i 8 of 17 LERs caused by Personnel Errors l l l

4 Management bleaknesses Failure to Reduce Personnel Errors Failure to Disposition known Problems Operations Engineering Security Licensing Quality Assurance i l l

. <4 Other Issues 01 Investigation: 1 OIA Investigation: External Interest: 4 l l l l l l

6 Other Issues O! Investigation: The Office of Investigation is reviewing the licensee's actions surrounding the premature criticality event of July 1, 1985. Current schedule is to present their final findings and recommendations to the Commission on January 27, 1986. DIA Investigation: The Office of Investigation and Audits is reviewing the NRC's actions surrounding the premature criticality event of July 1, 1985. Monroe County Board The Monroe County Board of Connissioners has taken of Commissioners: an active interest in the activities at Fermi 2. Both the Region III Administrator and Deputy Division Director, Division of Reactor Projects have met with and briefed the Commissioners. A copy of Special Report 50-341/85040 was sent to County Board. Additional briefing to be conducted before plant exceeds 5% power. I t

Y Otherissues(continued) Congressman Congressman Dingell, in a letter to John D. Otogell: Chairman Palladino, requested an investigation into the July 1, 1985 premature criticality event. A copy of the Region III Inspection Report No. 50-341/85040 was sent directly to Congressman Dingell upon release. The OI &OIA Reports must be sent when available. j e +w7-- y,7,-- y-y- -y s,e 4. ,p,-w- ,-,---yw-- -,y-r

A i' T. J. O'KEEFE g MAR 6 BB6 l4 Edison ~ L l t I Date: F cruary 28, 1986 NL-fts-86-0066 hs E. Preston, Jr. l Operation Engineer

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Frcm: J. R. Gree @ Supervisor Systems Engineer

Subject:

man Stearn Bvoass Lines Limitations A new concern has risen in regard to the Main Stem Bypass lines that will limit their operation. As you are aware, the Main Stee Bypass lines (PIS N3017) were recently replaced (EDP 4410). S e new 1" thick wall pipe was designed with a fatigue limit greater than maximum stresses in the been a cone rn. pipe. Fatigue failure was not supposed to ever have Howwer, a new analysis by Bopper and Associates demonstrates the vibration will cause cracks to grow. Using fracture medanics and information obtained last stener while testing the original lines, i Bopper predicted the time required to failure as a function of ficw for the new installation. Se results are startling! For example, --~ing the new installation reduces noise by 1/2, Bopper showed that at flows equivalent to 20 percent reactor power the lines would fail in seven days. I'there is a threshold below which no growth occurs; even with no imprwement in noise level the lines (2) can carry 5 percent reactor power indefinitely). Flows at 20 percent reactor power will not be that unommon; for exanple, during startup that power level will typically be realized before the turbfra r,ererator is synchronized and loaded. Attached is Hopper's curve for "J/2 noise". Engineering is reviewing Bopper's analysis and will potentially have to issue design changes; but that, at the earliest, would be at the 1st Refueling outage. In the interim, Operations will have to log the flow / time for the bypass lines. [ Flow can be inferred fran valve position,100 percent OPEN (per valve) equates to 121/2 percent t reactor flow). The effects of flow / time are accunulativer so the l total postulated crack growth should be logged. When the crack length approaches the limit (0.7867"), the lines can not be used. In order to minimize crack growth, i.e. extend the life of the lines, flows equivalent to greater than 5 percent reactor power should be avoided. Wrbine-generator synchronization and loading should be expedited; if pr&lems arise during this tine, reactor power should be reduced to 5 percent. nese actions will caplimte plant startups N

-4 T E. Preston, Jr. February 28, 1986 NENIS-86-0066 Page 2 and be an extra burden on Operations: Engineering is proceeding to alleviate the problem. Written by s A. D. Peluso f(6v systems Engineer anG/en Attactment cc: F. E. Agosti W. F. Colbert E. P. Griffi J. W. Bankala k!h R. S. Imnart S. B. Noetzel T. J. O'Keefe / L. E. Schuerman A. A. Shoudy L. J. Sisekin D. Spiers L. F. Wooden i I l l

. -........ ~.......... Atta::tunent HOPPER AND ASSOCIATES NE-NS-86-0066 ENGWEERS i 100 .7867 mi mi !!! W

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F E Q.- 4i . Enclosure 5 Main Steam Bypass Lines .On September 17, 1985, the licensee identified a leak in the east steam bypass line..It was determined through. investigation that the 30 inch diameter Steam Bypass Lines have experienced the development of a number of large cracks as well as damage to lugs and pipe restraints. Although this line is not safety-related, due to its relative importance to safe plant operations, a review of the licensee's actions concerning this problem was performed. Items reviewed during this inspection included: Operational history of the Steam Bypass Lines Licensee's evaluation of the cause of the cracking Results of rea' dings from the instrumentation installed on the pipe Techniques used to repair the cracks Results of metallurgical tests conducted on a sample removed from an area of the cracked pipe L The NRC inspectors concluded after this review that the Steam Bypass Lines were experiencing very acute fatigue induced failure. This is unusual in that the lines have a very short (approximately 2 months) operational history and that the unit has been limited to 5% of rated power. It appears that the fatigue problem is related to two factors; (1) the large diameter thin wall p configuration of the piping (30 inch diameter, 0.375 inch wall); (2) the high frequency acoustic vibrations induced by the steam bypass-valves. A permanent modification to address the fatigue problem must consider both of these factors. l Additionally, during the NRC inspection the licensee discovered a new crack j which had developed since the last repair. The NRC inspectors concluded that L further failure of the lines was imminent and that continued operation represented a risk to continues operation of the unit. The licensee agreed to shut the unit down and initiate replacement of the damaged piping. The above is documented in Inspection Report No. 50-341/85045 (Enclosure SA). The licensee has replaced the steam bypas.s lines from the bypass valves to where they. penetrate the main condenser. Region III has inspected the replacement ' activities and has identified no problems with the modification work. The l-results of the followup inspection are documented in Inspection Report No. 85049 l (Enclosure SB). 4v.uffa a w lls & } J &.. % %y :,ssfn E u/ Un. s e.e.* S .ll* t <. & N-

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.? Dettolt Edison Date: March 4, 1986 NE-PJ-86-0103 File: 0801.05 To: N. Shimshock Dire: tor - Pcwer Systems - ERD From: L. C. FronfCYM Supervising Engineer - P&PE Nelear Engineering

Subject:

Service Life er Ferri II Turbine Steam Bvness Pinine On February 28, 1986 Hopper & Associates presented a fra:ture mechanics analysis of the bypass lines to kclear Enginetring. This presentation concluded that the lines, under full flow conditions would last between 100 minutes and seven days. The analysis assumes a maxinum allowable flaw, as defined in API SL Grase B material specifications, is present. The cyclic stresses observed on the original 3/8" thick wall pipe are superinposed on the 1" thick well pipe at its' base stress level. The base stress level being the pipe wall stress due to internal presstre retention. Eclear Engineering is requesting Engineering.Research - Power Systems Group to review the Hopper report, evaluate the failure mechanics evaluation against the material intique life, and clarify to Eclear Engineering the validity and context within which this report is to be A written response is due to Eclear Engineering by 3/31/86. used. Work order 910BB281J150 may be used for this task. In the event that l you have questions contact J. Contoni - Lead Mechanical Engineer on l 165-1757. A Written by: J. Conto 3 8NM M Approved by: D. Spiers / 3*W l General Supervisor - P&PE i /dbg I cc: A. Peluso l ( I)

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HEPPER AND ASSOCIATES suomesas i. MXIMUM ALLOWABLE INITIAL FLAW SIZE - 0.0625 INCH (API) 3 TEMPERATURE - ROOM TEMPERATURE MEAN AND ALTERNATING STRESSES - LINEARLY a TO FASS FLOW RATE (M) THRESH 0LO STRESS INTENSITY-FATIGUE - 2.5 KSI/IN STRESS CONCENTRATION FACTOR - 2.78 CRITICAL CRACK LENGTH - 0.7867 INCH FOR 1 INCH WALL PIPE HEAN STRESS - 7.1 KSI CRACK GROWTH RATE h=1.02x10-2 g 3g.95 5 STRESS INTENSITY FACTOR l AK = Y AS /i = (2.78) + AS /a

WHERE, AS = (MAXIMUM STRESS - HINIMUM STRESS)

I CRACK LENGTH a = l l ASSUMPTIONS FOR FATIGUE CRACK PROPAGATION CALCULATIONS l

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A (* HOPPER AND ASSOCIATES ENGINEERS g. 210 AVENUE I, SUITE E 1 REDONDO BEACH CA 90277 213-316 4835 May 15,1986 HA-5/86-494 Contract #274884 Suborder #D59113 Task #B7 PIS #N30-17 Detroit Edison Company Enrico Fermi 2 Job Site 6400 North Dixie Highway Newport, MI 48166 Attention: Mr. Dave Spiers, General Engineering Supervisor

Subject:

Summary of Procedure for Fatigue Crack Growth Analysis of Steam Bypass Pipeline Gentlemen: We have prepared the subject document and hereby transmit it for your use. Please adivse us if you would like for us to transmit a copy of this document to others. If questions arise, please contact the undersi ned. Ver truly y D' avid M. Hopper Professional Engineer cc: Mr. John Contoni NOC Approval 3 YY

HOPPER AND AssocturES HA-5/86-494 ENGINEERS ,j SUPNARY OF PROCEDURE FOR FATIGUE CRACK GROWTH ANALYSIS OF STEAM BYPASS PIPELINE I Prepared for: Detroit Edison Company Enrico Fe"mi 2 Job Site 6400 North Dixie Highway Newport, Michigan 48166 Prepared by: Hopper and Associates 210 Avenue I, Suite E Redondo Beach, CA 90277 May 15, 1986

HOPPER AND Assocxxrrs HA-5/86-494 = ENGINEERS TABLE OF CONTENTS PAGE OBJECTIVE 1 BACKGROUND 1 J CALCULATION PROCEDURE 2 Stress (Strain)LoadingCycles 2 Mean Stresses 6 Noise Power Levels 6 Crack Length at Failure 6 Stress Intensity 8 Fatigue Crack Growth 10 RESULTS 12 COMMENTS 12 REFERENCES 19 APPENDICES A - Strain Gage Location 21 B - Partial Strain Trace 22 C - Vibration Mode Shapes and Frequencies 23 j l D - Mode Shapes and Frequencies 24 E - FCP Curve (Typical) 25 i F - FCP Cal:ulation Assumption Summary 26 G - FORTRAN Program - FCP Calculation 27 H - Pipeline Service History 28 I - Crack Growth 3/8" Pipewall - Continuous Crack 29 J - Crack Growth 3/8" Pipewall - Semi-elliptical Crack 30

.;e. HOPPER AND ASSOCIATES HA-5/86-494 ENGINEERS OBJECTIVE The main objective of this fatigue crack propagation analysis is to approxi-mate the fatigue life of 3/8" pipewall by theoretical fracture mechanics methods and utilize these methods to derive a procedure to approximate the life expectancy for the 1" pipewall. BACKGROUND After approximately 22 days of startup operations at Detroit Edison's Enrico Fermi 2 nuclear power plant, the steam turbine bypass pipelines were observed to hav2 cracks and resultant steam loss. These cracks were located in the 30" diameter 3/8" pipewall near welded mounting lugs. Failure analysis determined that the cracks were due to fatigue (1), and structural analyses showed that valve-generated noise was inducing structural vibrations (2) (shell wall flutter converging to primarily breathing mode vibrations at a frequency of approxi-mately 2000 Hz). A decision was made to replace the 3/8" pipewall containing welded lugs by 1" pipewall containing no welded mounting lugs but with baffle I plates inserted to restrict flow and dampen noise. i 1 i i i.. - -. -

y .c

" ~

HOPPE21 AND ASSOCIATES HA-5/86-494 ENGINEERS ' CALCULATION PROCEDURE The' fracture mechanics approach to the detemination of fatigue life requires the calculation.'and-application of several parameters. These parameters include:- -o Stress Loading Cycles o Mean Stresses o ' Noise Levels o Final Crack Length o - Stress Intensity (Flaw shape, size, and location) o Fatigue Crack Growth {- The calculations and basic assumptions for each of these parameters are expressed. in the following sections. . Stress (Strain) Loading C.ycles A strain gage rosette was located at location G13 on the west main stream bypass pipeline. Strain traces were recorded for this location on the original 3/8" pipewall during operation at 31% valve opening, and the corresponding hoop strain j trace was analyzed for a 0.27 second duration recording (see Appendices A and B). The representative hoop strains were measured and the corresponding hoop stresses were calculated: (oH = Pr/t). From the recorded strain trace, the maximum and minimum peak stresses were neasured and the number of occurrences were counted c for each value (Table 1). These values were plotted and found to comply to normal distributions (Figure 1), which indicates that the number of occurrences A for each stress is predictable and can be expressed mathematically. The frequency of the alternating stresses is determined empirically to be 2000 Hz, this matches the calculated convergant structural vibration frequency. The loading block diagram for the pipe was determined by utilizing a simplified rainflow count for repeating load histories (3). For valve openings (mass flow rates) from 0 to 31%, the maximum stress ranges were measured to be a linear function of percentage valve opening (4). Therefore, assuming the linearity continues to 100% valve opening, the full-flow loading is HOPPER AND ASSOCIATES HA-5/86-494 ENGINEERS S,x Ncount N (ksi) (#/.27 sec) 5 4 3 4.5 12 8. Mean = 1.747 35 33 Standard Deviation = 1.258 25 2 Distribution: 15 (-(S - 1.747)2 ) 1 71 72 0.5 50 53 max N = 270 expl 0 30 33 t 3.164 -) -0.5 17 17 \\ -1 4 8 -1.5 2 3 S N min count curve fit 1.5 4 4 1 9 11 0.5 29 24 Mean = -1.455 0 49 43 -0.5 66 66 Standard Deviation = 1.200 -1 81 83 -1.5 82 91 Distribution: -2 81 81 I-min + 1.4555)2 ) N = 270 exp'( (S -2.5 61 62 I -3 40 39 2.880 / -3.5 24 21 -4 8 10 -4.5 5 3 -5 1 1 1 Ntotal = 540 f = 2000 Hz (ttotal = 0.27 sec j OCCURRENCES OF PEAK FATIGUE LOADS - MEASURED AND CALCULATED TABLE 1..

a a1 a r ..s 9 ? HOPPER AND ASSOCIATES HA-5/86-494 ENGINEERS -

E ASURED Sg
"^5""E" S

~ NUMBER OF OCCURRENCES MIN - CALCutATED s,. snin (no./.27 sEC) c (. 1.45 100 ( = 1.75 v. j SMin I ~ l 4 50 l 40" I l j 20 i -5 -4 -3 -2 -1 0 -1 2 3 4 5 S. STRESS (K$1) i' PEAK MAXIMUM AND MINIMUM STRESSES MEASURED FROM STRAIN GAGE #8 TRACE (3/8" WALL PIPE) FIGURE 1 ._,...,..._,_.___._.._.--.___..__..-.m.,__

4 HOPPER AND ASSOCIATES - HA-5/86-494 ENGINEERS proportional to the 31% flow loading. For the 1" pipewall, the alternating thoop stresses are decreased by an additional factor of 0.375 due to the thick-ness effects. The loading block diagrams are given as occurrences of stress ranges, AS, in Table 2, and they are determined for a one-second time interval and full valve opening. AS AS N N N 3/8" FULL FLOW I" FULL FLOW FULL NOISE 2/3 NOISE 1/2 NOISE l (KSI) (KSI) (#/SEC) (#/SEC) (#/SEC) 35.48 13.31 1 32.26 12.10 1 29.03 10.89 4 1 25.81 9.68 15 2 22.58 8.47 44 8-l 19.35 7.26 109 31 16.13 6.05 220 112 1 12.90 4.84 361 275 23 l 9.68 3.63 474 500 194 6.45 2.42 471 622 737 3.23 1.21 299 449 - 1045 0CCURRENCES OF LOAD RAN3ES - PER SECOND TABLE 2 L

HOPPER AND AssocirrES HA-5/86-494 ENGINEERS - Mean Stresses The mean stresses assume.the value of the hoop stresses due to static steam pressures in the pipeline (these mean stresses are also linearly dependent on mass flow rate) and are given as: 10.0 ksi for the 3/3" pipewall at full flow, and 7.1. ksi for the 1" pipewall at full flow. S,(l") = S,(3/8") x wall thickness factor x baffle plate factor I = S,(3/8") x (3/8) x (1.9) Noise Power Levels Assuming the stress / strain generating noise power is equivalent to the area under a stress range versus count curve (Rayleigh distribution-fit curve), any noise power mitigations will be reflected by a corresponding decrease in the area'under the distribution curve (Figure 2). Taking noise power j levels of 2/3 and 1/2 of the original unmitigated noise power level, re-spective distribution curves can be generated for full flow conditions. l Crack Length at Failure The pipe materials are ASTM A-155 and API K55, which are low carbon steels + with UTS 60 ksi and YS 35 ksi. Therefore, due to high fracture toughness (ductility), the pipe will fail by ductile yielding not by fast fracture. Assuming a " worst case" crack, namely a continuous edge crack, failure will occur when the effective cross-section is reduced to a point where the maxi-mum (instantaneous) applied tensile stress exceeds the UTS. The " critical" crack length is dependent on the valve opening (steam mass flow rate, A), and is given by: 1 f =.375 .1734 A (in) for 3/8" wall thickness: a 1" wall thickness: a = 1 .2292 A (in) f

l. l

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HOPPER AND AssocrrrES HA-5/86-494 ' ENGINEERS 9 1100 - S 1000 0 g 900 - g 800 - Wy 700 - Ey _600 - o 500 !5 g 400 - 1/2NOISELEVEL(PREDICTED) = 300 - 2/3 NOISE LEVEL (PREDICTED) FULL NOISE (MEASURED) 200 100 - 2 4 6 8 l'0 3/8"(.31 ) AS = 25, (KSI) I"(1.0k) 2 4 6 8 10 12 ALTERNATING STRESSES AND OCCURRENCES PER SECOND FOR DIFFERENT NOISE INTENSITY LEVELS ' FIGURE 2,

HOPPER AND AssocirrES HA-5/86-494 ENGINEERS For calculation simplicity, the crack length at failure for the 3/8" pipe-wall was taken as constant values: a = 0.295 in., and for the 1" pipewall, c a =.7867 in., for all flow rate valves. c For elliptical or semi-elliptical cracks, the critical crack lengths, a ' c are equal to the wall thickness (leak before break failure). Stress Intensity Stress intensity was evaluated for the 3/8" and 1" pipewall as (5): K=Y o 6

where, K E stress intensity Y E stress intensity factor a E applied stress a E crack length Stress intensity factors (or stress concentration factors) for the near crack region have been estimated based upon:

1. shape of flaw 2. size of flaw 3. location of flaw. The shape of the flaw was assumed to be for " worst case" conditions as an edge continuous crack, and for " realistic case" conditions as a semi-elliptical surface flaw (see Figure 3). HOPPER AND ASSOCIATES HA-5/86-494 ' ENGINEERS SEMI-ELLIPTICAL CONTINU0US -a .- a I 'i i y- ~ 2c .1. L - _- _a T l .i: K=Ya6 K=Ya6 Y = 1.12 6//6 Y = 1.12 6 Q = 1.65 (for a/2c =.38) SEMI-ELLIPTICAL AND CONTINUOUS EDGE CRACK CONFIGURATIONS FIGURE 3 i .g.

HOPPER ann ASSOCIATES HA-5/86-494 ENGINEERS In the determination of stress intensity factors, the initial allowable flaw size of 0.0625" was used, even though crack growth occurs resulting in new factor values. The location of the flaw was assumed to be the " worst case" wherein the flaw is located in a weld zone. The total stress intensity factor is the product of all factors: YT = Y) Y2*Y3

where, Y) = f1aw shape = 1.55 semi-elliptical (a/2c =.38) crack (6)

=1.99continuouscrack(7) 2 = flaw size factor - neglected (assumed constant) Y 3 = flaw location factor = 1.4 unground weld (8) Y = 1.9 welded lug (9) In the 3/8" wall: Y = 3.77 for continuous crack = 2.945 for semi-elliptical crack In the 1" wall: Y E 2.78 for continuous crack E 2.17 for semi-elliptical crack These factors were used in the crack growth evaluation as constants with no additional considerations for correction factors due to geometry, plasticity, or environment. Fatigue Crack Growth Fatigue crack propagation (FCP) is commonly expressed as the Paris crack growth rate equation (5,10,15): 4+

e> +
86 IMAGE EVALUATION I

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l HOPPER AND ASSOCIA*TES HA-5/86-494 ENGINEERS h=AAK" The crack growth rate equation for pressure boundary steels in LWR environ-ments has been given as (11,12,13): h = 1.02 x 10-12 q] gg.95(in/cy) for AK < x 5 1.01 x 10-7 Q AK * (in/cy) for AK > x = 2

where, AK = AS Y6 R=Smin/Smax"(S ~ 3 )/(S +3) m a

m a Q) = 26.9 R - 5.725 Q2 = 3.75 R + 0.625 x = 17.739 Q /0 2 1 b AK = Y AS a This equation is based on tests conducted with compact fracture specimens, therefore, growth is calculated based upon one dimensional (single axis) propagation. The minimum allowable crack growth per cycle is given-byJ,he rger's vector K.95, 5 for steel, b = 1 x 10-8 inch. Thereerf,whenda/dN=1.1x'10~ 1 x 10-8, the threshold stress int nsity, AK = 3.1 kri / inch. This,value th was assumed to remain constant and o crack growth will occur for tress in-tensities below this value. HOPPER AND ASSOCIATES HA-5/86-494 ENGINEERS RESULTS A FORTRAN program was compiled to calculate the fatigue crack propagation (14) utilizing the crack growth equation and the loading block diagrams. The pro-gram encompasses the Wheeler model (15) for crack growth retardation due to high tensile overloads assuming the crack tip plastic zone is in plane strain. The resulting cumulative damage is expressed as: aa (the percentage of total crack growth = x 100c D a - "o to crack growth at failure) c The program was used to calculate crack growth for both a continuous flaw and semi-elliptical flaw utilizing previously stated assumptions. The calcu-lations for the semi-elliptical flaw assume that radius "a" grows in accor-dance to the crack growth equation and the ellipse remains cons' tant in shape (a/2c = constant). The corresponding graphs of damage versus time have been plotted for the 3/8" and 1" pipewalls at various flow rates. Since it is not yet known what the effects of the baffle plates are on noise power mitigation, sets of damage curves have been generated for the 1" pipewall at full, 2/3 and 1/2 noise power levels (Figures 4 - 8). COMENTS I The first evidence of the 3/8" pipewall failure (ie., steam leakage) was observed afterapproximatgly 22 Anys of operation at various flow rates. Utilizing a histogram of valve position versus time for the 3/8" pipewall, an approximate service cycle was extracted, and from this service cycle, the progression of crack propagation was estimated by using the damage curves. Assuming a continuous flaw, the life was detennined to be 16 days from the damage curves, while assuming a semi-elliptical flaw, the damage was only 8 3% after 22 days. These are both excellent approximations since they give a range for theoretical fatigue life by which the actual observed fatigue I life is bound. I ! i L

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into which the actual life will be bound; this is the approach that has been taken in these calculations and has been shown to yield excellent results.

l l l l.

HOPPER AND AssocirrES HA-5/86-494 ENGINEERS REFERENCES 1. Detroit Edison Company, Technical Interchange, October 25, 1985. 2. Blevins, R.D., Formulas for Natural Frequency and Mode Shape, Krieger Publishing, Malabar, Florida, 1984. 3. " Standard Practice for Cycle Counting in Fatigue Analysis," ASTM Standard E 1049-85, Annual Book of ASTM Standards, Volume 03.03, 1985. 4. Detroit Edison Company Data, " Dynamic Strain Measurements", TI-877, Technical Interchange, November 11, 1985. 5. Hertzberg, R.W., Deformation and Fracture of Engineering Materials, Second Edition, John Wiley & Sons, New York, 1983. 6. Newman, J.C., Jr., "A Review of Assessment of the Stress-Intensity Factors for Surface Cracks," Part-Through Crack Fatigue Life Prediction, ASTM STP 687, J.B. Chang, Editor, ASTM,1979. 7. Sih, G.C.M., Handbook of Stress Intensity Factors, Lehigh V'niversity, Lehigh, Pennsylvania, 1973. 8. Blodgett, Omer.W., Design of Welded Structures, The Lincoln ARC Welding Founding, Cleveland, Ohio, 1966. 9. Faupel, J.H., Fisher, F.E., Engineering Design, John Wiley & Sons, New York, 1981.

10. Nelson, Drew, V., " Review of Fatigue Crack-Growth Prediction Methods,"

Experimental Mechanics, February 1977.

11. Bamford, W.H., " Technical Basis for Revised Reference Crack-Growth Rate Curves for Pressure Boundary Steels in LWR Environment," Journal of Pressure Vessel Technology, Volume 102, November 1980.

12. Anderson, W.F., et al., " Preliminary Analysis of the Effect of Fatigue Loading and Crack Propagation on Crack Acceptance Criteria for Nuclear l

Power Plant Components," NUREG-0726 U.S. Nuclear Regulatory Commission, 1981. )

13. Simeonan, F.A., Goodrich, C.W., " Parametric Calculations of Fatigue Crack-Growth in Piping," NUREG/CR-3059, PNL-4537, U.S. Nuclear Regulatory Commission, 1983.

14. Peterson, D.E., Vroman, G.A., " Computer-Aided Fracture Mechanics Life Prediction Analysis," Part-Through Crack Fatigue Life Prediction, ASTM STP 687, J.B. Chang, Editor, ASTM, 1979.

15. Fuchs, H.0., Stephens, R.I., Metal Fatigue in Engineering, John Wiley & Sons, New York,1980.

HOPPER AND ASSOCIATES HA-5/86-494 ENGINEERS APPENDICES 1 I t ~20-l

HOPPER AND ASSOCIATES ENGINEEM HA-5/86-494 VALVE HOOP STRAIN GAGE NO 8 G 13 ROSETTE C 9 HOOP STRAIN GAGE N0 4 ROSETTE G 18 CONDENSER WEST MAIN STEAM BYPASS DUMP HOOP STRAIN GAGE LOCATION APPENDIX A _

HOPPER AND ASSOCIATES HA-5/86-494 .c ENGINEERS +n A 1 AA 1. /./ N f V'l\\vW\\\\14(y)k,1 a. i m \\ T1Me TIME = 0.01 SECOND = o l STRA1H GAGE #8 TRACE - HOOP STRAIN (TAKEN FROM 3/8" WALL PIPE) l APPENDIX B I HOPPER MD AJSOCIATES HA-5/86-494 RING FLEXURAL MODES 3/8" WALL n(1-n)2 2 2 n=2 f= 41 cps 'N 1_ gI g' f, _2R 2 \\ / y AR4 1+n n=3 f = 115 cps BEAM FLEXURAL MODES 3/8" WALL 2 L = 6' f= 630 cps EII I f. [En AY L=4' f = 1400 cps s ' ~ ~ _ _ ',,, 4 a LONGITUDINAL VIBRATION MODES WALL INDEPENDENT L = 65' f= 800 cps ,.mya L Y L = 45' f = 1150 cps o a,- - - - - - - g RING EXPANSION MODE WALL INDEPENDENT ,.;g(i.nbn.O < = 2u0 cp, o 2R7 / MODE SHAPES & FREQUENCIES APPENDIX C HOPPER AND AssocrxrES HA-5/86-494 ENGINEERS COUPLED SHELL MODES ~~ \\ ~~ I l'l l I h I! il l I Q, ) l l 's ~-- i=0 l=2 l=3 @$[) %M i=4 CIRCUMFERENTIAL N00AL PATTERN m A p"s A ~ ' ~ ~ %#'w sa v a ~ ~ -, _ ,,- s p ,-% n,%, h V % /%,,,e V l*l l=2 l=3 AXIAL NODAL PATTERN / 2N NODAL ARRANDEMENT ' ':,s' f \\ p' l FOR i= 3,l = 4 \\ ,s ' ',~m \\ / k' I CIRCUMFERENTIAL ,, e ' N00E g \\s' l! / AXIAL N00E WALL INDEPENDENT A Eg l ij " 2IIR g),y) 2 FLUTTER CONVERGENCE f = 2186 cps MODE SHAPES AND FREQUENCIES j APPENDIX D_ HOPPER AND ASSOCIATES HA-5/86-494 ENGINEER $ I l I i l I gl 41% T fi El l-I \\$ w u 3

I I$

I l w'" l l i I I w g b _ NO.i. CRACK 5.: FRACTURE GROWTH GROWTil STRESS INTENSIFICATION FACTOR 109 AK CRACK GR614H AND FRACTURE PHENOMENA APPENDIX E HOPPER M gSOCIATES HA-5/86-494 MAXIMUM ALLOWABLE INITIAL FLAW SIZE - 0.0625 INCH (API) TEMPERATURE - ROOM TEMPERATURE MEAN AND ALTERNATING STRESSES - LINEARLY a TO MASS FLOW RATE (M) THRESHOLD STRESS INTENSITY-FATIGU - 3.1 KSI4fi STRESS CONCENTRATION FACTOR - 2.78 CRITICAL CRACK LENGTH - 0.7867 INCH FOR 1 INCH WALL PIPE MEAN STRESS - 7.1 KSI FOR 1 INCH WALL PIPE CRACK GROWTH RATE h=1,02x10-2 9 gg.95 5 STRESS INTENSITY FACTOR AK = Y AS 6 = (2.78). AS. 6

WHERE, AS = (MAXIMUM STRESS - MINIMUM STRESS) a = CRACK LENGTH ASSUMPTIONS FOR FAT!GUE CRACK PROPAGATION CALCULATIONS APPENDIX F HOPPER AND ASSOCIATES HA-5/86-494 ENGINEERS C FATICUE CRACC PROPACATION ANALYSIS FOR STEAM PIPELINE C UTILIZING A$nE CODE-SECTION XI REVISED CRACC CROWTH RATE CURVES C IMITIALI2E UALUES AND READ LOAD SPECTRUM DATA D!nEN510N DELS(15)

DIRENSION NUM(153 REAL KTH DOUILE PRECIS!0H 7,TT,70TN PI*ACOS(-1.) KTHe3.1 YS*35 READ (5,101AC.A4,Y,$MS 16 FORMAT (4F7.4) READ (5,30)N RE AD( 5,20 )( DELS (! ), NUM (! ),!*1,M ) 29 FORRAT(F8.4,I5) 34 FORMAT (13) C PERCENT RASS FLOU RATE LOOP DO 3094 K+1,10 Pa(11-C)/10. T*S. TOTNat. $MesMes? A*At ATH+(KTH/(DELS(1)*PSV))*s2 IF (A.LT.ATH.AND.ATH.LT.AC) A*ATH+0.0001 20*A+(($M+DELS(11sP/213Y3A339.5)**2/(63Pl*YS232) DELARee WRITE (6,200 )P. A att FORMAT (*1*,*FRACT. RASS FLOW RATE *,FS.2,5X,*INIT. FLAU S!ZE',F3.5) URITE(8.300 ) 340 FORMAT (11X,' TIME-MIM.',21X,' CRACK $12E=IM.',5X,'fERCENT DARACE') C RAIM PROCRAM 30DY - FCP CALCULATIONS + 490 DC 1000 !*1,H C CALCULATE STRESS INTEMSITY FACTOR - DEtt i DELK*DELS(!)tPSY* Asst.5 IF(DELE.LT.CTH3C0 To lite C CALCULATE STRESS RATIO R Ar(D CRACC CROUTH putTIPLIERS 01 AND 02 R * ( $ R-DELS ( I ) 3 P/23 /( $ R+DELS ( ! )2P/2 ) IF(R-4.25)419,419,420 419 R*0.25 l 420 IF(R-e.851444,434,430 438 Rec.65 440 01*26.98R-5.725 82+3.758R+4.es25 X*17.739s(02/01)sse.25 C CALCULATE CAACC CROUTH RATE-LINEAR (FROM ASMC CODE.5ECTION XI) IF( DELE.CC.X) DADNet.41E-7:02sDELKast.95 IF(DELE.LT.X) DADNet.02[-12:013 DELE **5.95 C CALCULATE RETARDATION FACTOR RP I * ( ( $M* DEL $ ( 1 ) $P/2 )IYS A339. 51 s s 2/ ( E s PI sYs s st ) ZaA+RPI IF(2.CT.20) Ze*Z r l C CALCULATE CRACK CROWTH RATE-RETARDED (WHEELER MODEL) DADNR*(RP!/(24-A13881.3*DADN DELAR*DADNRrNUM(Ilsse l C+20-2+DELAR IF(C.CE.9) C0 70 500 M*(20-23/DADNR C CALCULATE CRACE CROUTH-RETARDED DELAR*(MSDADNR+(NUM(!)-M)*DADN)t80 C NEU CRACE LEMCTH 500 A*A*DELAR IF(A.CE.AC)C0 TO 1550 1940 CONTINUE lite IF(DELAR31700,1700,1200 i C WRITE CRACE LINCTH AND CROUTH 70 FAILURE AT TIME (EVERY 1983H MINUTES) 1240 TOTM*T0TN+1. TT*DL0010(TOTN) IF(TT-T)1520,1400,1400 1444 PCTF*190.*( A-A0 3/( AC-AG) URITEt$ 1564)T0TH.A.PCTF l 1540 FORMAT (1X,714,E14.3,T49,F11 7,TT4,E14.5) l T*T+4.1 itte CD TO 400 1554 WRITEts,1600) TOTH [ 1690 FORMAT (1M,' FAILURE TIRE-MIN *,F14.1) C0 70 2900 l 1764 WRITE (Seitet) i 1964 FORMAT (* INFINITE LIFE-KCETH FOR ENTIRE LOAD SPECTRUM *) 2000 CONTIPJC m ConTI,,ut \\ nW END APPENDIX G

1

HOPPER AND ASSOCIATES HA-5/86-494 ENGINEERS P 20-ACTUAL g APPR0XIMATED }4 r u. s 5' 7 AJ~\\f LA_/ f V E 10-Wd> 0 0 10 20 DAYS i APPROXIMATED CYCLE A .19 .11 .12 .15 .18 .13 t 60 7850 7850 3500 1750 7000 (MIN) ACTUAL AND APPROXIMATED SERVICE CYCLES APPENDIX H -.

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References:

7 - p.m -17 HH. itiator Org./Section Time lS l l'2 5'7 Date T/ ~2 72/, T/b 7n. Alfr/Adsts /ft V$f Aff/W l 22-I. Initwor's Supy, Time Date J. OSL No. T J, -hPM// l 225 l/.25? pr l } PART 2 OPEllASILITY/REPORTABILITY l ) A.10CFR50.72 [ ] 1 Hr. (TNo NRC Contact's Name: []4 s. Date Contacted: Time: Determined By M [ Ab Date: /- 2 "O '7 B.10CFR50.73 f ( K] No LER No. Determined EV Date: 8 /r /F) LER Due Date C. Other: ~ LER No. Determined By Date: .LER Due Date 1 PART3 CARB REVIEW l i A. Evaluation Assignment Organization: Section: 1 RadChem ()NQA IJNOS WMaintenance Technical ( ) NE M NP Operations Other C. CARB Approv Date S. Evaluation Dpe Date $1 V& W I l-S

C l -

PART 4 EVALUATION l A. Describe the i.ot cause o' the deviation Determined By Org/Section Date S. Remarks rkn DFf2 L b be cLoiro hu DGP # 7 O i l N

1. 5-M '1 T5&E21 Reportability Review Requested?

l l yes h4 No i PART5 WORK COMPLETION l @. Approved By Org/Section Date I l l PART 8 NOA COMPLETION REVIEW l %eviewed by Section Date l l

Send to CAC lll

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&-stacAwr meneros m.s, FMa so whou e sa A//-. Pha Jan,erswAA-ruix/MrAn614 anC" zp 4ewAm+ tawrammr /sai,- e. 51~ E-Yn Assv1 M < AaTNA7s*A (h1hfJ ~ fainoMe& ) G M terer.:s s lHgltintor Org/ tion Time Da NfI /oWO IN N5l-favorola f/F&> n-f's d' L T*mt De II,W$Daf1f X M>hvM W Astr77a?/-97/1~/ l l.33S' llO J. O$. tic I /e lPAF1 2 OPERABILITY /REPORTABILITY A 10M R!,0.72 []1 r. 1)Q room IriRC Contact s hame. IDate Contacted: Tim e - 4ff Dete* mined BV ,e 4 ete /-71-87 E. 10C,i RZ 73 Ye p(Nc, LER No. , ; _1 Datt //s /B*7 LER Due Date Determined Bv v LER No C. Otner Dete rrr.ine d Be Date LER Due Date l l PART 3 CARB REVIEW /, 6 se'urtio i Assignment Organization Section. [ } RadChem ) T echnics' ) f. Q t-l ) NOS N Maintenence i i ) f.E bd Nr . Operations, ) Ott.e r E_i.aivat:on Due C.CA I EVALUATION l FART 4 A Otscrib6 thE root cause o1 the Ctviation fMvlnWgg AWAsAAto/T' s/s. _/~ se'di,n/.S 723r>et Ls./d* AAAbh/a11 aW.s 13'Mwf~ LJJL' SHILG1td"1 =/di7"andL ANb'aleM.edL AnnGn00&Af1*

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De* ermine: By Org'Se: tion Date k RemetLt w g y Ojc b, knj pineon f,u tf6 ha,qp O,y J.3-.y 9 ~ / 1CiCFR21 hepo tability Review Requested) l l Yes FAINo l l PART 5 WORK COMPLETION Date G Approv6d by Org/5ection i I I j l' A R T (, NOA COMPLETION REVIEW ha glev.ed B y Section Date l I 3Ie nt to C AC \\ v

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w....... -. - -............ -... POM 12 000 052. Page 2 of 2 070186 CORRECTIVE ACTION SHEET Page 2 of I PART 4 PROPOSED CORRECTIVE ACTIONS PART 5 COMPLETED CORRECTIVE ACTIONS I C. Remedial Corrective Action A. Remedial Action Taken: /. f /J'TAL L CA/ J44 VEL 4 47-1. Assigned Org: 4 .5it(4).S$AAr WS74vosatt fAA Sec1 ion: Jac Ar,s.rs.

2. Reg'd Due Date:

3 CARB Approval

2. PDife' t r+1 ADfE7dAv7' 1E3 7" of Date.

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Send to CAC

Armuutavr ds. / %ls7 n >.r z Ma i, S m L,a c. Aw oo s w.Tmwca r be. Ruun' Inerin En cia <cieinc, Asstss n r eo /?ccome<om.wr or ria was va n-e usiwn. mes. g,4-Visv.) e-in.h of A fa il eci sects.-s revealed n o o6veova Ja.re were..no rk ,,eIv no,s, d a f i e e-, e, e > % wJd i-p o r is i ). 7h..s,taiWea: ..n o Ia e tc o f.fa.ria,._an ct n o ,sfeef we/d au mal. witi sta mlers welcl deli +o ca 4a - mdud. Lca.ne, at th, %,e., it app eac.s w3 tG. .. E Iv r c. can k c. cdtriloutect to v,1 cotio -in daccc/ fob ve.. y ~ A rea,n a f th. sy e e,p>emii-> c onditio-> shoo d tk. d pie u y u ntuned p r<.asure. ( tv< b s - e in l<f p ressur c) tc> be a y y toximie y%s.ps > unckr.W... recs,t.(100 MWs ) Jto ..=- opvatie7 conAi h <. fece, _1h. _p reu vec. c4 ce g a ccro.u; th. YS V 's wo u l c\\ he. a p p co z o mc.is]p 8.f0 p *. e.sso a sc<ft) Nc. o re esiI Sc, e n ) o at wi14. piping vib re tion s h ca paeit ..Las th L a c ca.s t'ic ener3 ge.,va.1ect hw h19 eas vedves).y. ec eive, y va lves : ic. f L EP2 Eyp. ....... rusur e. ~ fo - e t%. so u rce. o th. po,:>in vi L ce fia., is L . TSv'.'.s.,. -.... ~ ~in kn wiit A c.bo-c i, it M fc4 i Ju rc a ssex.ro,3 cJ, ~ ~ .En3m e ering reco i,,ve ds TLn. fo IIou u,9 (in acic\\,1oc, +o %- n work da % NP 10 re,sfetc. L sya1em +:>.D dl<.sinfus.).

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11.000 53 Revision 1 Mikgp s / o F 3 Page 20 ATTACHMENT 3: SAFETY EVALUATION CHECKLIST (SEC) SEC NUMBER PlS NUMBER A/3'O - // Technical Specifications change? l lYes NNo 1. IDENTIFICATION: Document Number A/. A Rowlsion Description Rasouvrioa or Hs F.m to mm La,oe wirw.furprermo.CI.o - Linueto Jrnaov S rarr Vinna rma oc or brian sna>r sur DV'.r var nn Ti>Amion FSAR or Technical Specifications Sections Affected: A/ou s. 2. SAFETY EVALUATION llYes SNo a. Does the proposed change or test increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety previously evaluated in the FSAR? Basis: R ucA To rac.s. v ee n. t x eta u -ri.s l lYes $No b. Does the proposed change or test create the possibility of an accident or malfunction of a type different from any evaluated previously in the FSAR7 ) Essis: No AlfU Act o Dw T SK AllA IO.r ed MALfdAlt f/*J/ Adt CHATED E4 'Th c n ea.st s vs rw Vn sAA-rrass M f Y.s scout. rs. l l lYes $No c. Does the proposed change or test reduce the margin of safety as defined I 11 the basis for any of the Technical Specifications? Easis' ht viktAyra er nia tiF1at:.s v.rTr>< ts Asar DIMc71.N OR frJDIALL-rL~r \\ A Nw ssto us =nJE RASU kB Amr Y *TXe NN/tkt.3PL C/Cf CA~flDA) 3. If the answer to any of the questions in Part 2 is "yes." then the proposalis considered to present an unreviewed safety question. l lYes ENo UNREVIEWED SAFEV OUESTION? If an unreviewed safety question is involved or a change in the Technical Specifications is required. review by the NSRG and approval by the NRC is necessary before the proposed action may be implemented. 4. PREPARED BY: 177.*184N/ 2 / St /a" ci.,e r4 8/E Date: /-% R7 Title REVIEWED BY: / Date: Title APPROVED BY: / Date: Title 5. OSRO CONCURRENCE: / Date: Title (Enclose original SEC and any attachments with proposed change package or procedure working file. Provide copy of OSRO-approved SEC and any attachments to the OSRO clerk.]

PRELIMINARY EVALUATION CNICKLIST ATTACHMENT 1: 1. DENTIFICATION: Desument Number A/.A. Revlelen REstnArriano eYh $w e 'T*>earce u>ra Susrecres s,,m M rua ae rnor emat hw ="s e rrv *r Desertption A e > Tanueen Srr m ra ruc 7aamiue 2 CLASSIFICATION Does the proposed change lavolve QA Level I Equipment? llVes De

3. PRELIMINARY EVALUATION Would this modify plant therectoristice of procedurel steps described in

[ ffe. De e. FSAR? If yes. Identify sections: Could this adversely effect the ebility of egulpment or structures to pe ((Yes De b. their eefety-reisted funettons? Does this treste e new test not described in the FSAR tha llYes SNo c. plant esfety? For design thenges only. would this change modify essumptions us occident snelyses described in FSAR Chapter 157 If yes. identify sec l[Yes SNo d. For design changes only.could this edversely effect the function of or components required for compliance with the Umiting Condition llyes pio e. Operation in the Technical Speelfscetions? i In the judgment of the evoluetor. la e Safety Evolustion required? lNo f. RVes t If the answer to any questions la part 3 is *yes.' th br u a rwren-rin he wearuuse Tnt nessenn er r et" occontrace MA ni.s *! tar 4 PIPt. M UP7dAC A CC/ DFA 7 sY A SAFETY EVALUATION REQUIRED? Stes lMe 4. If a thenge to the Technical Specifications is involved, prior NRC a Note: required (see NOIP 11.000.118). /- 4' - 8'A frT LI42MC/ fefue,aree. NFE Date ~ S. PREPARED BY: ~ Title / . Dete-APPROVED BY: Title itnclose original PEC and ettschments with change package or p bk' l C ~ ~ " * ' - - -

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.5ee. [d) Akb sA% SERM LIES ACTION PIAN Y IMDIATE ACTION RESPONSIBLE /. Plug and cap six (6) instr taps (permanent fix)

Maint.

By 1/4/87 4-4" taps already capped @ 1600 2. Grind out weld at the 2 utilized instr taps and reweld

Maint.

3 Restore 12 hangers and 5 snubbers

Maint.

g. Inspect loop drain lines - hangers & welds

Maint.

f, Engineering Research to evaluate failures

Eng Res.

1-3-87 3 existing failures Q. n> +.. C< e- . b r 3 additional valves yet to be cutout -s f.'. 3. v. awI* Maint. 1-4-87 6. Evaluation of DER through root cause and remedial action CORRTTIVE ACTION /ANYSIS 7 Vibration monitoring program for instrunent lines Eng/ISI l-4-87 8, Snubber evaluation analysis 6 %.'.. t u M., r- ' l,. .,c-ISI l-3-87 J LI ,G.. % j e. t. s.~r y 9, Engineering evaluation of weld failures and hangers way Eng/CA j dictate additional inspections or analysis of other lines 1

Must be conplete prior to placing lines in service.

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___..~. WAeuteAves Uy. nea f%% Uru -fse aAO % a bad del is' Wa.'DOJb4D t'. U1 FOR: ED GREENMAN FROM: RALPH SYLVIA Date January 6, 1987 1-7-87/mb To: B. R. Sylvia Group Vice President From: J. R. On1houn, Outirman Independent overview Committee subject: Test Candition 2 'the IOC has been requested to advise tether Permi 2 is ready to proceed to Test Condition 2, whid allows operation between 20 percent and 45 percent power. 'the IOC considers Fermi 2 will be ready to proceed with Test condition 2 after operating for five to seven days at appromisstely 20 percent electrical power, while consistently unintaining acomptable water chemistry, as a demonstration that planned repairs and modifications have been effective. In making this roccamendation, the IOC gave recoytition to the fact that the Permi 2 reactor has been unintained at power from December 19, 1986 to the present. In addition, we expect that you will proceed with your plans to j further verify the <=i=== of the broken instrument lines en the main steam line; assure that corrective actions to date are satisfactory; and collect additional steam line vibration data during the 20 percent l power run. In looking ahead, we will continue to look for improvements in the time required to identify root causes of equipment failures or salfunctions and to implement appropriate corrective action. We also believe that the Independent Safety Engineering Group report rveillance testing warrants special unnagement attention to ire that the questions raised in that report are thoroughly heed r ate corrective ation taken. f..~,_,- JRyp1r ocs W. J. McCarthy i l 1 i TOTAL P * ..}}

ML20210U435

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