ML18098A035

Saint Lucie Unit 1 - Letter Regarding a Reportable Incident Under 10 CFR 50.55(e) Relating to a Split Flange on Intake Cooling Water Line and Requesting Postponement of This Report Until October 6, 1975

ML18098A035
Person / Time
Site:	Saint Lucie
Issue date:	09/09/1975
From:	Uhrig R E Florida Power & Light Co
To:	Moseley N C NRC/RGN-II
References
L-75-426, L-75-434
Download: ML18098A035 (9)

Text

g4 F~~f~~-C'-J f I+Ej FLORIDA POWER 5 LIGHT COIilPANY September 9, 1975 Q L-75-434.7 V Mr.Norman.C.Moseley', Director-Office of Inspection and Enforcement, Region II U.S.Nuclear Regulatory Commission 230 Peachtree Street, N.W., Suite 818 Atlanta, Georgia 30303

Dear Mr.Moseley:

Re: Construction Incident Report, St.Lucie 1 S lit Fl'an e on Intake Coolin Water'ine On August 11, 1975, Florida Power 6 Light Company notified your office of a reportable incident under 10 CFR 50.55(e), in which.a bolted flange on Intake Cooling Water Line I-3-CtI-70 had split.The report on this incident has not yet been completed and is due to be submitted on September 10, 1975.Accordingly, FPL requests postponement of this report until October', 1975.This extension was discussed with and noted by Mr.Foster of your office on Sep-tember 9, 1975.Yours very truly,)Qkj~~Ac~i.t>tfW+Robert E.Uhrig Vice President REU:nch cc:'ack R.Newman, Esquire

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~=I-'"""IL~m@~'LORIOA POWER 5 LIGHT COMPANY September 4, 1975 I,-75-426 Mr.Norman C.Moseley, Director Office of Inspection and Enforcement, Region II U.S.Nuclear Regulatory Commission 230 Peachtree Street, N.N., Suite 818 Atlanta, Georgia 30303

Dear Mr.Moseley:

Re: IE:II:MSK Reactor Coolant'System Overpressurization, St.Lucie Unit No.1 On August 12, 1975, letdown isolation valve gV2516 failed closed and permitted an 80 psi overpressurization of the, Reactor Coolant Syst: em of St.Lucie Unit No.1.The valve closure was caused by a broken electrical lug on HGA relay f94-3-159.

A report of this incident, stating the circumstances, the analytical results, and the corrective action taken, is attached to this letter.In view of the fact that no damage occurred.and that no extensive evalua-tion was required, FPL believes that this is not a reportable incident in accordance with'10 CFR 50.55 (e)(iii).~Yours very truly, pJ Robert E.Uhrig Vice President REU:nch Attachment cc:,.Jack R.Newman, Esq.

ST.LUCIE UNIT HO.1 REACTOR COOLANT SYSTEM OVERPRESSURIZATIOH SEPTEMBER, 1975 TABLE OF CONTENTS'

SUMMARY

1.Synopsis of Incident 2.Resolution of Damaged Equipment 3.Resolution of Occurrence V II DESCRIPTION OF INCIDENT IXI.CORRECTIVE ACTION 1.Response by Field Personnel 2.Damage Investigation and Results 3.Action Taken to Prevent Reoccurrence IV SAFETY IMPLICATIONS PAGE 1 1'2 2 5 I~Summar 1.S o sis of Incident ep On August 12, 1975;St.Lucie'nit No.1 was undergoing hot functional testing, when letdown isolation.

valve gV2516 failed closed.The resultant associated transient caused the RCS pressure limitations, as shown in Figure 3.4-2a of the standard technical specifications to be exceeded by 80 psia.The failure of valve jjV2516 was caused by a broken electrical lug on HGA relay,'f94'-3-15 9.2..Resolution of Dama ed E ui ment'The affected relay leads were relugged and tightened and the letdown isolation valve was returned to an operable status.No damage to the RgS was experienced.

3.Resolution of Occurrence An analysis of the overpressurization condition determined that the overpressure transient was acceptable, See Section III for details.It is believed that failure of the electrical conne-tion on the relay lead was an isolated occurrence, The other factory-wired relay panels have been checked to assure lug tightness.

II Descri tion of Incident The failure of letdown isolation valve f<V2516 on St.Lucie Unit No.1 occurred on August 12, 1975, at 10:45 A.H.as the unit'was under-going hot functional testing.At the time of occurrence the Reactor'oolant System was under the fo'llowing conditions:

1.Reactor coolant pressure at 210 psia as read on PI-1103 (scale 0-1600 psia)2.Reactor coolant temperature at 105 F as read on pressurizer water phase temperature indicator TI-1101 (scale 1-700 F)3;Charging and letdown system in operation with one (1)charging PP running.'.Control element drive mechanism venting in progress.The transient conditions following closure of V2516 were: 1.RCS pressure 600 psia as read on PI-1103.2.RCS pressure approximately 660 psia as read on Heise gauge installed at vessel head 3.RCS temperature at 105 F's read on TI-1101.

~.~Containment ambient tempeiature was 92 F as measured on recorder TR-25-1 (inlet temperature to the containment coolers).~p 4 The failure of V2516 occurred due to'broken lug'onnection on sealing relay f94-3-159, which is the solenoid relay associated

'with.V2516.

The plant's Instrumentation and Control group were in the process of checking lug tightness on the factory-wired relay panels on the RTGB, and in the process of removing, the cover from relay fj94-3-159, the letdown isolation valve (V2516)failed closed.Subsequent investigation-revealed that the two lead lugs were broken off at the head of the lugs, and the.relay, cover was apparently holding the leads in place.The transient, associated with V2516 failing closed resulted in RCS pressure exceeding the pressure/temperature limitations as shown on Figure 3.4-2a in the standard technical specifications.

Per Figure 3.4-2a the unit must be at greater than 156~F (RCS.temp-erature)before exceeding 520 psia RCS pressure.KX.Corrective Action 1.Res onse b Field Personnel Upon seeing the RCS pressure increase, the operator terminated the transient by securing charging to the RCS.The,RCS pressure remained at approximate3,y,600 psia for 2 to 3 minutes, at which time it was reduced to approximately 210 psia by venting via the'essel head vent.-2.The affected leads were relugged and,tightened,*

and the letdown isolation valve was returned to an operable status at 1:40 PM on August 12, 1975.Dama e Investi ation and Results a.Rela Wirin The relay.and wiring involved was factory-wired at General Electric where the panels were fabricated.

The lug used was an insulated x'ing tongue lug compressed onto a size ,'f16 Vulkene flexible wire.It is probable that during the start-up check outs, that this lug was flexed each time the wire identification number was'read and the lug finally broke at the junction of the ring and barrel when the relay cover was removed for inspection.

The relay and the broken lug were inspected by FPL Elec'trical OA personnel on August 12, 1975.It was their opinion based upon'the lack of any similar experience with the thousands of other similar factory-wired relays of this.identical type, that.

this occurrence was not a generic problem with either~'the lug or the relay, but an isolated instance.b..Reactor.Coolant S stem On August 13, 1975, Combustion-Engineering, the NSSS vendor, initiated an evaluation.

of the overpressurization condition.

The transient which occurred when V2516 failed imposed conditions on the RCS outside the limits specified in.Figure 3."4-2a of the technical specifications.

The particular.limit violated was that of the lowest service temperature of l60 F as determined by NB-2332(b) of ASHE Boiler 6 Pressure Vessel Code,Section XXI, 1971 Edition, Summer 1972 Addenda.The reference temperature, RT>DT, upon which this 160 F lowest service temperature x.s based, is 50 F.A comparison between the xeference (critical) intensity factor (K K)and tha calculated stress intensity.factors (KI and K was made using the procedures of ASME SectioŽn III, Appendix G, in order to verify the acceptability of the overpressure transient.

1.Determine KIR.RT~T=50 F Zemperature

=105 F., Assume-10 F instrument error and consider temperature relative to RT~T, (T-RT>DT), to be (95 F 50oF)45oF Enter Figure G-3110-1 at+45 F and determine KIR KSI~In.2.Determine KIm a Pr m=-.Indicated pres'sure on Heise gauge was 660 psia.With 1/2%accuracy this pressure could have been 663.psia.(a)30" D Reactor Coolant Piping r=15", t=2.5", a=38 KSX s=-=3980 psi=3.980 KSI.m t m.105,~t=1.58 From Figure G-2114.1, find M=1.88 and a=, 1.88 x 3.980=7.48=, K m Im

~83 (b)~42" Reactor Coolant Piping r=21", t=3.75", a=38 KSI.e=-=.713 KSI Pr.663'x 21".m t.~m-a ,v y From.098, I t=.1.94 3.713 Figure G-2114.1 find Q=<1.88 and a=<1.88 x 3.713=<6.98 K m'.Im (c)Steam Generator Plenum r=74.7", t=, 7.0",a=70 KSI 3.338 883 Pr'.663 x 74.7 m=3.538=0.50 From Figure 0-2114.1 find M=<<2.46 end M x a=<<2.46 x 3.538=<8.70 K m m'.'Im (d)Pressurizer r=48.250't=4.875", a=70 KSI a=.663 x 48.25=6.562KSI m'=6.562=.094, Wr=2.21 70 From Figure G-2114.1 find M=2.08 and m Q x a=2.08 x 6.562=13.65 KI (e)The reactor vessel temperature-pressure limits have been determined in accordance with Appendix G.and are shown in Figure 5.4-1A of the St.Lucie Unit No.1 FSAR.The pressure transient experienced did not violate the temperature-pressure restrictions as shown in this figure.3.Since the transient was a pressure transient only, with the insulated system at a steady 1050F and.the containment ambient at'20F, the temperature difference through the' walls (b T)is insignificant and therefore the calculated thermal stress intensity, KIt, is insignificant.

4.Allowable pressure.Using the criterion of Appendix G-2215, 2 KI+KIt<KI , and evaluating for the highest value oV K~as.tound for the pressurizer shell, we find thaC~E'IR=" 2 K=2 x 13.65=27.3 Im 27.2<50'herefore, it can be concluded from this analysis that no stress intensities,-

such as to possibly propagate the conservatively assumed defects of Appendix G, Section G-2120, were encountered.

It should be noted here that the value of RT<n~=50~F is'conservative and was used for this entire evaIQation.(expect for the vessel), not just the piping to which it is applicable.

3..Action Taken to Prevent Recurrence The other factory-wired relay panels have been checked to assure lug tightness.,V Safet Im lications There will be no effect on the safe operation of St.Lucie Unit No.l as a result of this incident.Failure of the letdown isolation valve by the mechanism described is considered to be an isolated event.Operator action terminated the transient immediately after observation and brought the RCS within pressure limits shortly thereafter.

An analysis of the overpressurization condition determined that the stress intensities experienced by the REC were acceptable in accordance with ASME Code,Section III, Appendix G.1