ML18219C357

LER 1976-018-00 & LER 1976-019-00 for Donald C. Cook, Unit 1 One Accelerometer Out of Three Found Movable Mass Hard-Up Against Stops & Reactor Coolant System Pressure Increasing

ML18219C357
Person / Time
Site:	Cook
Issue date:	05/14/1976
From:	Jurgensen R American Electric Power Service Corp, Indiana Michigan Power Co, (Formerly Indiana & Michigan Power Co)
To:	James Keppler NRC/IE, NRC/RGN-III
References
LER 1976-018-00, LER 1976-019-00
Download: ML18219C357 (6)

Text

NRC FORM 195 U.S. NUCLEAA AEGULATORY COMMISSION DOCKET NUMIIER I2.78) 50-315 FILE NUMBER XNCIDENT REPORT Tp: ..G. Keppler,.~'RQM:Indiana & Michigan Power, Co. DATE OF DOCUMENT Bridgman, Michigan 5-14-76 R.W. Jurgensen DATE RECEIVED 5-20-76 QLETTER CINOTOAIZED PROP INPUT FOAM NUMSER OF COPIES RECEIVED 0ORIGINAL CYUNCLASSIFIED

@COPY ENCLOSURE 40'ESCRIPTION Licensee Event Report (ROO. 76-18) pn 4-19-7 Ltr. trass the following........ )n)

Concerning one A'cceleromhher out of'three foun movable mass .hztC-up against stops....

Licensee Event 'Report (R.O. 8 76-19) on 4-14-7 (oncern)igg Reactor Coolant System Pressure Increasing to 1040 PSIG (40 Carbon Cys. Received)

.Z)O @OT aEMOVF, ACKNOWLED<<D PLANT NAM.: Co>> > 1 IF. PERSONNEL

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EXPOSURE IS INVOLVED

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tray 14, 1976 6 0 1876 Mal Section Doc4t c4<k Nr. J.G. Keppler, Regional Director M(A{g@g

)A(41 ~egg Office of Inspection and Enforcenient United States Nuclear Regulatory Commiss 1 0 cr' Region III 799 Roosevelt Road Glen Ellyn, IL 60137 Operating License DPR-58 Docket No. 50-315

Dear ter. Keppler:

Pursuant to the requirements of Appendix A Technical Specifications and the United States Nuclear Regulatory Commission Regulatory Guide

l. 16, Revision 4, Section 2.b, the following reports are submitted:

RO 50-315/76-18 RO 50-315/76-19 Since ely, R.W. g sen Plant tlanager

/bab CC: R.S. Hunter J.E. Dolan G.E. Lien R. Kilburn R.J. Vollen BPI R.C. Callen tlPSC K.R. Baker RO: III P.lt. Steketee, Esq.

R. Walsh, Esq, G. Charnoff, Esq.

G. Olson J.H. Hennigan PNSRC RES. Keith Dir., IE (30 copies)

Dir., HIPC (3 copies)

IUCENSEE EVEtIT I~EPOAT ~

CONTROL~/LOCK: (PLEASE PAINT ALI. AEOUIAED INFOAIVIATION)

LLCCNSEC I ICftISE f.vfNI txAME LICENSE NULIOLA I YPE I Yl'E 0 0 0 0 0 0 0 o o Loa~f

~ 25 31;I2 1 ~ I 15 20 30 .

AfPO AT AfPAAI CATEOOAY Z YPE SUUACE FIOAT( I tIUMI IF A LVI t4I OATE IIIPOIIT IIAfC

~0'I CQf41 L ~L 0 5 0 0 3 1 5 [0 Il 1 9 7 '6 0 5 4 7 6

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FVENT DESCRIPTION

[oO~] Nhil'e in trode 5 Performance of Calibration Test on Seismic Peak Recording Accelerometer.

7 09 UO located in the s ent fuel pit area and reactor pit area found one AcceleroiIIeter out of 7 09 UO Jog three on each instrument with the movable IIIass hard-up against the stops. The units 7 0 9 Qg were re laced with functional units from more accessible areas (RO-50-315/7C-19) 7 0 9 00

~06 7 0 9 FAME 00 SYSTEM CAUSE COMPIINEN'f COMPONFNT COOE COOS COI IPONENT COOC SUPPUCA IJIANLIFACIU~A VIOLATION

~a~ ~XX G X X X X X X ~L T 1 0 0 7 0 9 10 11 12 17 43 CAUSE DESCRIPTION

[os nt deterioration. The surveillance schedule has been changed 7 0 9 00 toOj9 t r uire annual calibration of these instruments vice 18 Honthly.

00 Qoj 7 09 FACILITY MCTHOO OF S'f A'T US POWEA. OTHEA STATUS OlSCOVEAY OISCOVEAY OESCAlPTION 0 ~QO 0 IIA 8 Surveillance Testing 7 0 9 10 12 13 40 FOAM OF ACTIVITY COATENT AELCASEO OF AELEASE AMOUNT OF ACTIVITY LOCATION OF AELEASE Kg 9 Z Z NA NA 7 0 10 11 44 45 PERSONNEL EXPOSURES NUMOEA TYPF. OESCAIPTION

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/ 09 11 12 13 00 PERSONNEL INJURIES NUMOEA OESCAIPTION

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~00 0 IIA 7 11 12 7

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PROBABLE CONSEQUENCES LOSS OR DAMAGE TO FACILITY TYPE OESCAlPTION Q~g Z HA 7 09 10 00 PUOLICITY NA 7 0 9 00 ADDITIONAL FACTORS Q~g I HA

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7 09 NAF,IE G. Swan PHON-. 616-465-5901 (368)

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~ LICENSEE EVEIVT AEPonr r PLEASE PAINT ALL AKOUIAEO INFOAlVIATION)

LICENSEE I.ICI NSE EVENT NAME LLCllvSE NUMI'I'n I YPE TYPE Jog tl I D C C 1 O O 0 0 0 00 0 0 4 1 1 1 1 LOO3 )

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7 09 14 15 05 >6 30 31 30 AEponT nfponl nil>onl DAIr cATroonv -

Tvvf souncE OOCIiLT l'UMliln l'VI IIT UA'll.

Iocijcowc L ~L 0 5 0 0 3 I 5 0 O I II 7 6 0 5 ] 4 7 6 7 0 57 50 59 60 61 60 69 74 75 Uo EVENT OESCRIPTION Jog ftHILE IN tfODE 5, WITH REACTOR PROTECTION SYSTEtl RESPONSE TIt1E TESTING IN PROGRESS, AN 7 0 9 00 m09 7

It)ADVERTENT LET-DOWff ISOLATION WAS INITIATED WlfICH CAUSED REACTOR COOLAtfT SYSTEtf PRESSURl II0

~gq TO INCREASE TO 1040 PSIG WHILE REACTOR COOLANT SYSTEtl PRESSURE llAS 1100F EXCEEDING LItfIT.

7 8 9

[oOJ SET FORTH It< TECffNICAL SPECIFICATIONS PARAGRAPH 3.4.9.1.

7 0 9 00 tm (SEE SUP PLEf 1ENT ) RO-50-315/76- 18 7 0 9 PAME SYSTEM CAUSE COMPONENT COMPONENT CODE CODE COMPONENT CODE sUPPUEA MA> vr*cTunln VIOLATIOV IOOYI ~cA 0 z z z z z z ~z z z z z ~Y 7 0 9 10 11 12 17 43 44 47 40 CAUSE OESCRIPTION Jog PRERE(jUISITES FOR RESPONSE TIME TESTING INCLUDED PLACING BOTH PROTECTION SYSTEt1 TRAINS 7 0 9 Q~g It< TEST SIIIULTANEOUSLY AND REtfOVING TRAIN B OUTPUT FUSES. REMOVAL OF THESE FUSES 7 0 9 00

~10 DEEr<ERGIZED RELAYS GIVING LETDOWN ISOLATION AND RHR ISOLATION. (SEE SUPPLEtdENT) 7 09 fACILITY METHOD OF STATUS 5I POWEA OTHEA STATUS OISCOvEnv OISCOVEAY OESCAIP'AON G ~00 0 NA 44 45 OPERATIONAL EVENT 7 0 9 10 12 13 46 fOAM OF ACTIVITY COATENT AELEASED OF RELEASE AMOUNT OF ACTIVITY LOCATION OF AELEASE Q~g z ~z tTA 7 0 9 10 11 44 45 PERSONNEL EXPOSURES NUMOEA TYPE OESCAIPI'ION

~s ~oo o ~z w 7 09 11 12 13 80 PERSONNEL INJURIES NUMSEA OESCAIPTION Pg ~QO 0 IN 7 09 11 12 00 PROBABLE CONSEQUENCES Q~s] NA 7 89 00 LOSS OR DAMAGE TO FACILITY

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TYPE 10 OESCAIPTION NA PUBLICITY tQ 7 09 AOOITIONAL FACTORS Q~g NA 00 Q3gJ 7 89 OU NAME. G SWAtf i'tiovr::(~1~) I65-%01 +68)

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LICENSEE EVENT REPORT RO-50-315/76-18 SUPPLEt1EHT SUPPLEtlENT TO EVENT DESCRIPTION A fracture mechanics analysis of this overpressurization transient was performed by Westinghouse Electric Corporation. This analysis shows that the transient did not affect the structural integrity of the reactor vessel.

ASSUNPTIOHS OF THE ANALYSIS - Since irradiation effects the fracture toughness of the beltline region of the reactor vessel, and the pressure temperature curves are determined for this region, only the core beltline region was considered in this analysis. Ho there>al stress contribution was used in the analysis. The fracture mechanics analysis was based on the methods as defined in Appendix G to Section III of the ASME Code.

CONCLUSIONS OF THE ANALYSIS - As indicated by performed analysis, the stress intensity factor for a 1/4 thickness flaw in the beltline region is less than the fracture toughness by a factor of approximately 1.3. The 1/4 thickness flaw would not have become critical. Furthermore, the assumption of a 1/4 thickness flaw is extremely conservative as compared with any flaw that may be present in the pressure vessel. In addition, a fatigue evaluation was made which indicated that the contribution of the overpressuri zation transient to the total fatigue usage factor is negligible. It should be noted that the total cumulative fatigue usage factor due to all the transients specified to occur during the 40 year life of the plant is less than 0.0024. The results of the fracture mechanics analysis and the fatigue evaluation indicate that the integrity of the reactor vessel was not affected and that the reactor coolant system is acceptable for continued operation.

AEP Service Corporation engineers and the Plant Nuclear Safety Review Committee have evaluated the Westinghouse analysis and .oncur with the conclusions reached by this analysis.

SUPPLEMENT TO CAUSE DESCRIPTION A Temporary Change Sheet was written to the procedure to disable the isolations by lifting the lead on TB 148-12 in the Train B Auxiliary Relay Cabinet to prevent this problem from recurring.