ML19322D926
| ML19322D926 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 12/23/1978 |
| From: | Grimes B Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19322D920 | List: |
| References | |
| NUDOCS 8003110218 | |
| Download: ML19322D926 (2) | |
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. Conccraing the Steam Generator Dual level Set Point Modificatic (December 23, 1978)
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aeference n 12/23/78 Conference consisting of Brian Grimo, co coraan, von Kirkpatrick, Symore Weiss, Mort Fairtile, G. S. Vissing concerning TEco letter dated December 22, 1978 - Steam Generator Level Set Points Questions which were developed:
l.
What happens shen operator fills SG and pressurizer goes solid?
2.
What is the effect of over cooling transient - when SG becomes filled with cold HPI? Do we exceed TS limits on cooling?
3.
What is the ' basic reason for the 35" level setpoint?
Is it for loss of
. off. site power or for loss of feedwater?
, Concern for operator going to 35" during a LOCA.
4.
5.
Concer'n for Loss of offsite power in which operator goes to 35" set point and then we have a small break.
Does 10 min. operator action effect ECCS analysis?
12/23/78 Call to Region III, Tom Tambling, John Streeter, Richard Knop with the people previously mentioned.
Discussed above.
12/23/73 Conference cill to TECo, IE & D0R IE.III - Tambling, Streeter, Knop 1
IE Hdqr. - Ed Jordan, Don Kirkpatrick
'TEco - C. Domeck, E. Novak, Plant Rep.
00R - Grimes, Weiss, Fairtile, Vissing 1.
Would system go solid on HPI?
Answer HPI pump shut off head is 16004/d" Would not go solid.
j 2.
Consequences of overcooling transient.
Answer - Plant is designed to sustain 40 loss of power ~to coolant pumps.
This is similar to overcooling transient.
3.
What is reason for controlling SG at 35"? Operation is at 40".
Have had loss of feedwater 0 40% power and controlled at 35" by operator action.
Ope ~rator can quickly take this kind of action.
4.
Concerning implicakions on small break LOCA. Does natural' circulation test show better heat transfer capability?
Are they still in compliance with the regulations with regards to LOCA?
TEco says that natural circulation flow test showed greater flow than anti-cipated.
S&W says there is no change with analyses for small break.
How can TECo be sure that operator is can go into 35" level set point when there is a small break? TECo says operator. has good indication, alarms, etc.
.TECo says,that B&W tells them that -with RC pump's running the 35" level set; point
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av-is OK for small break.
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What happens when the plant has a loss of off site power and then a small break? 4 TEcb could not answer this.
They said that it would be cleaner to change to auto set poiht at 35" but this would take a long time and money with new.ECCS analysis.
They think the dual level is quickest method.
Brian Grimes said that the design should consider loss of offsite power' and then a LOCA. TECo said they have not analyzed that event.
We asked if this analysis is required under 50.45.
We asked them to talk to B&W to check and make sure if this analysis should be in the record.
Conclusions (1) We asked TECo to look at theloss of offsite power and then a small break -
make sure they are in conformance to 50.46 (ECCS)
(2) We requested within a commitment to have a (1) a design within 60 days of penmanent modifications, (2) 50.59 determination within 60 days for the mod with copy of safety analysis, (3) schedule (as early as possible - before next refueling) of the modifications.
We concluded interim fix OK and TECo has no restraints.
Also let us know outcome with talk to B&W concerning compliance with ECCS 50.46.
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Reference 15 TOLEDO
%s EDISON L77-312 October 7,1977 FILE: RR.2 (NP-32-77-16)
Docket No. 50-346 License No. NPP-3 4GEA
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Mr. James G. Keppler Regional Director, Region III f S/7 C.
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Of fice of InoPection and Enforcenent cQ4 M
U. S. Nuclear Regulatory Com:sission 799 Roosevcit Road Q
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-/w-Glen Ellyn, Illinois 60137 7A
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Dear Mr. Keppler:
Reportable Occurrence NP-32-77-1 Davis-Besse Nuclear Power Station Unit 1 Date of Occurrence:
Scotember 24, 1977 Enclosed find three copies of Licensee Event Report NP-32-77-16 with a supple-mental infore.ation sheet, which is being submitted in accordance with Technical Specification 6.9 to provide 14 day written notification of the subject occury rence.
Yours truly, J Th /%
Jack Evans Station Superintendent Davis-Besse Nuclear Power Station C[
JGE/JRL/1jk ffe yh Enclosures
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Dr. Ernst Volgenau, Director Office of Inspection and Enforcement Enc 1: 40 copies Licensee Event Report 40 copies supplemental Information Sheet Mr. Wiliam G. Mcdonald, Director Office of Manage =ent OC 121977 i
Information and Program Control Enc 1: 3 copies Licensee Event Report 3 copies Supplenental Infor=ation Sheet 7729401M 2 copies Telecopied Report THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLECO. CHIO 43S52 1
. p,a p CCu 3rd U. S. NUCLE AD CECULA720Y COtt,pagg;3gg t.lCENSEE EVENT REPORT EXHIBIT A CO*.7 AoL SLOCat: l l l l 1 l lh (PLEAst PRINT CR TYPE ALL REOUI AED INFoRt.tATION) i s
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[o i a j f Half trip of the Steam and Feedvater Rupture Control System (SFRCS) r ioi3i i causing a rise in Reactor Coolant System (RCS) temperature and pressure. ;
gogii This resulted in the pressurizer power relief valve to open and this l
Io tsj i valve failed to close, causing a reduction in RCS pressure.
Limiting i
Conditions For Operation vere exceeded 'for five Technical Specifications,,
13, 1
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- 3. 4.1, 3.' 4. 5, 3. 4. 6. 2, 3. 6.1. 4 and 3. 7.1. 2.
(NP-32-77-16)
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3 copies Supplement
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1.
SDLva75 On Septe=ber 24, 1977, a series of events occurred at the Davis-Besse Unit 1 which resulted in depressurization of the pri=ary system from a normal operating pressure of 2150 psi to 900 psi in approWmtely 8 ninutes, and the release of apprcri=ately 11,000 gallons of water in the for= cf steam within the containment through the pressurizer quench rz:h rupture disc.,
On the m.fternoon of Saturday, Septe=ber 24, 1977, the =ain turbine was shut do a to repai a leak in a pressure sensing connection on a steam 16 from the turbine governing valves to the turbine inlet.
The re-actor was being held critiez1 at apprezi=ately 9% ther=al power.
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1.t 2134 heurs, a spurious half trip occurred in the Steam Feedvater Rupture Centrol System (SFRCS).
This caused the startup feedvater valve on the No. 2 sream generator (which is the normal feed path at thie pcue: level) to close.
Closure of this valve resulted in a lov Noa 2 steam generator level, which then resulted in a no: mal full trip of the S71CS for this condition and initiation of the S?RCS.
STFIS initiation closes both nain steam isolation valves and initiates feed-vater flov to both steam generators fron their individual steam-driven au:rf 14 ary fee dpt._-.pi..
The bl # trip and resulting full trip of the STRCS caused a reduction in heat renoval fica the primary systen and a corresponding tenperature/
pressure -ise in the pri=ary syste=.
The pressure rise in the primary syste= caused the pressurizer power relief valve to lif t.
This valve s then rapidly escillated closed-to-open approxinately nine tices and re mired in the full open position.
y The terperature rise in the pr#~=ry system caused an increa.;e in the pressurizer level, and the operator manually tripped the reactor on high pressuri:er level approxinately two minutes after the half trip on the S?lCS cecurred.
The pressurize pcver relief valve, in the full open position, rapidly reduced the pri=ary systen pressure, and a Safety Features Actuation Systen (STAS) trip occurred at the 1600 psi setpoint of the primary.
systen.
The power relief valve discharge goes to the pressuriser quench tank, which became overloaded and overpressurized, and ipproxicately 4% --4 utes af ter reactor trip the rupture disc in this tank relieved due to overpressure, venting the stean into the contain=ent.
Apprc:cim tely 20 =inutes af ter reactor. trip, the operators diagnosed the reason for the p:#-*:7 systen depressurization as being the power relief valve, and from -de control room closed the motorized block valve ahead of the power relief valve, terminating tha loss of primary coolant into the courain=ent.
I Subsequent operator action using =akeup pu=ps and high pressure injection i
, pu=ps stabilized the pri=ar'/ systen pressure and pressurizer level and a centrolled shutdon to cold shutdown conditions followed.
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Tne =ajor physical. da: age fron the incident was to the reflective netal insulation on the lover part of the No. 2 stead generator, which re-ceived the jet of stean cocing from the pressurizer quench tank. A ventilating duct in the area of.the quench tank was di= pled and required straightening. Twenty-three panels of reflective =etal insulation required replace =ent.
Entry into the containment was made at 0550 Sunday, September 25, 1977, for cleanup operations.
-Another event occurred 14 the' course of this incident that did not contribute =aterially to the above events, but did result in the No. 2 stean generator going dry. Tais was the failure of the No. 2 am414 ary feedpu=p to cece up to full speed following the SFRCS trip. This feedpu=p ca=e up to approri=ately 2600 rp= and stayed at this level with no flow to the steam generator until approxicately 12 cinutes af ter reactor trip, when the operators placed its control in canual and brought it up to full speed (co==encing feedvater flow to the stean generator).
The depressurization of the pricary systen resulted in stea= for=ation in the primary systen, but evaluation has shown there was no appreciable boiling in the core. Tne pressure /te=perature transients in the primary systa= co:ponents including the stean generator, reactor coolant pt. ps and fuel vere severe, but analysis and subsequent p'ucp testing has shown that these transients are within the de. sign allovables and' that no detrinental effects are to be expected on the pri=ary system, pu=ps or fuel.
Systen/conponent aloperation or failure occur, red in three areas:
SFRCS (half-trip initiation), pressuriter power relief valve (oscillation and failieg in the open position)' and auxiliary feedpt._.p (failure to ce=e up to full speed).
The causes of these maloperation/ failures have been investigated and corrective action taken to prevent recurrence.
Additional syste=/ equip:ent modifications have been co=pleted or initiated, and additional training has been initiated to strengthen the systens intelligence available.to the operators and facilitate operator action.
At no time during the sequence of events was there any jeopardy to the health and safety of the public or plant operators, and there was no l
relea'se of radioactivity to the envi'roncent.
Activity levels within the conta4-'t at no tice i=peded containment access.
All safety systc=s perforced their design functions in the proper Operator action was ti=ely a' d proper throughout the sequence n
canner.
of events.
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2.
EvzNr DESCRI? TION At the tice this incident occurred, the reactireter data logging systen was in service which recorded at high speed a nu=ber of
. system parameters that would not have been available on such a time base through nor=al station instrumentation and records.
Tnis in-for ation, together with the ce=puter ala:n logging, has per=itted a very detailed plotting of the transient conditions in the pri=ary and secondary systems ke'yed to the system, co=ponent and operator actions. This data is plotted on four Figures in hMhit 3.
Figure 1 is an 11 inute plot of prinary systen parameters from one (1) minute prior to event initiation (SFRCS half trip).
Figure 2 is a 130-ninute plot of three pricary systen pars =eters.
71gures 3 and,4 are 95-cinute plots of pressure and ce=perature for steam generators No.1 and No. 2 respectively.
The event started at ti=a 21:34:20 (T = 0) on September 24, 1977.
The plant was in Mode 1 vith Power (MWT) = 263.
The turbine had been shutdown earlier in the evening to repair a leak in the mHn stean line at an instru= ant connection betvoen the turbine stop valves and the.high pressure turbine.
At this time a half trip of the Steam and Feedvater ?.upture Control System (SFRCS) was initiated by an unknown cause.
The trip closed the startup feedvater valve to No. 2 steam generator and stopped all feedvater.to this 2,enerator (at this lev power level the ain feedvater block valve is closed, isolating the
=ain feedvater control valve).
The lov level alarm was reached in No. 2 stean generator at T = 24 sec. Before the operator could identify and correct. the problem, this lov level in No. 2 steam f
generator correctly produced a full trip of the STRCS.
This trip closed the main stern isolation valves and feedvater isolation valves in both stean generators (T = 58 sec.). SFRCS initiation also started both auxiliary feedvater pu=ps.
The nu=ber one pt..p perfor=ed as intended, hevever, nu=ber two auxiliary feeduater pu=p only came up to 2600 rps, insuf ficient to feed its steam generator (No. 2).
The loss of feedvater, first to one and then both steam generators, caused. an increase in reactor coolant temperature, which resulted in an increase in pressuriter level and reactor coolant system pressure.
At 2255 PSIG the pressurizer.electronatic relief valve received an.
open signal. During the next 40 se'conds, it received open and close
. sig al s, cycled close-to-open nine times and then re=ained open.
This provided a continuous vent path fron the pressurizer to the cuench tank.
Enen pressurizer level rose to 290", the operator manually tripped the reactor (T = 1 cin. 47 sec.).
Energy escaping through the electro =atic relief valve and main steam relief valves caused a rapid cooldown and depressurization of the reactor coolant systas.
Reactor coolant' systen pressure dropped to 1600 PSIG (T = 2 nin.
51 sec.) initiating the Safety Features Actuation Systes (STAS).
This started the high pressure injection pu=ps and closed certain contain=ent isolation valves.
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With the electronatic relief valve still open, the quench tank rupture disc ruptured (T = 6 min.), relieving steam into the container.nt.
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Ynen the reactor coolant system pressure decayed to approrfr ately 1500 psig full high pressure injection flow was established and started to raise pressurizer level. At T = 6 =in. 14 see. the operator stopped the high pressure injection pe=ps.
(The operators had been heavily involved before this tine in regaining seal injec-tion flow to the reactor coolant pu=ps which had been stopped by the SFAS actuation. By T = 5 =in. 20 sec. the appropriate SFAS signals had been overridden and nor=al flows restored to the seals of the pu=ps). Reactor coolant system pressure continued to decrease until saturation pressure vaa racched and steam began to form in the reactor
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coolant systen (approxicately T = 8 nin).
This caused an insurge of vater into the pressurizer and the pressurizer level vent off scale I
at 320 inches.
Dur1ng this level increase the operator, seeing average reactor coolant system temperature and pressurizer level increasing, stopped onc reactor coolant pu=p in each loop (T = 9 nin.)
to reduce the heat input into the reactor coolant systen.
Due to d2 creasing presse=t in No. 2 steam generator, the SFRCS system gave a low pressure block persit signal at T = 14 nin.13 sec..This alerted the operator to the low level and feed condition of No. 2 stean generator. He ble:ked the low pressure trip. (T = 15 min.
18 sec.), took canual centrol' of the speed of No. 2 auxiliary feed-vater pe=p, which con =enced full feedwater flow to No. 2 steam generator (T = 16 cdn.). The operator saw the rapid addition of cold feedwater into No. 2 steam generator was dropping the reactor coolant systen temperature and reduced the feedvater addition to this generator.
At approxicately 'T = 21 nin., it was deter =ined. that the power relief valve was re=aining open and the block valve uas closed, isolating the power relief valve on the-pressurizer and stopping the venting' of the, reactor coolant system to the quench tank.
At T = 31 min.,
pressurizer level cane back on scale.
At T = 41 min. the, operator started a second cakeup pe=p to try and stop the pressurizer level decrease.
This additional. cold water started the reactor coolant '
system on a slow decreasing ta=perature transient. At T = 43 min.,
pressurizer level rasched the lov level interlock and cut off the pressurizar heaters. At T = 49 cin. the operator started a high pressure injection pu=p to try end stop the decreasing pressurizer level.
The level and pressure in No. 2 steam generator again decreased to the point where the SFRCS gave a low pressure block pernit signal.
The operator again blocked the trip and, through canual speed control of its auxiliary feedvater pe=p, restored level and pressure in No. 2 steam generator (T = 31 =in.)
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With pressurizer level vell en its way to recovering, the operator stopped the high pressure injection pe=p (T = 53 nin. 24 sec.).
At T = 57 a#. he restored reacto'r coolant =aMeup flow to nor=al.
This stopped the slow decreasing reactor coolane temperature transient
'which started at T = 41 nin. All plant para =eters were nov fully under centrcl and the plant was brought to a steady state condition, and a nor:21 plant cooldown started.
3.
SYSTEM-IOUTIME2.r EAL?UNCIION A.
General Tnere vere three system. /co:ponents where calope' ration or f=4'ure occurred during the event.
Tnese are:
1.
Stean Teedvater Rupture Control Systen - STRCS (half-trip initiation) 2.
Pcuer Relief Valve (oscillation and failing in the open
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position) 3.
Auxiliary Feedpu:p (failure to come up to full speed)
Tne SFRCS is a safety systen designed to provide feedvater to the steam generator /s for renoval of decay heat fro = the pri=ary systa= under a variety of hypothesized plant operating conditions,.
Tcese hypothesized conditions include loss of nor=al feedvater flo, stessi line breaks and feedvater line breaks.
The co=po-nes:s of this systen include sensing systens, logic and initiation systecs, cain steam isolation valves, stean turbine-driven auxiliary feedvater pu=ps, feedvater isolation valves, auxiliary stes= and feedvater supply valves and cross connect valves.
A description of this systen is contained in Exhibit C of this re,pe rt.
A half trip of the SFRCS initiated this event by closing the sta cup feed.ater valve. to the.No. 2 steam generator, which resulted in a full trip due to low stean generator level.. Tais spurious or inadvertent half trip, and possible reasons for it occu_ing, are discussed in core detail below.
rne pressurizer power relief valve is a 2h" pilot-actuated relief valve connected to the top of the pressurizer s.-ith a noter-operated isolation or block valve located in the line i
4-adia:ely ahead of the relief valve.
Tne purpose of this power relief valve is to provide a :'eans of relieving pressur-iner pressure _vithout requiring ~ operation of the spring-loaded ASME Code relief valves.
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During this event, the power-operated relief valve opened, oscillated closed-to-open cud then failed to close and re= ined in the open position.
Operator action from the control room closed the isolation valve ahead of the power relief valve about 20 cinutes after reactor trip.
The reasons for the oscillatio'os and the failure of the power relief valve to close.are discussed in = ore detail below.
The steas turbine-driven auxilir.ry feedwater punps are a part of the STRCS.
Upon initiation of the SFF.CS, the auxiliary steam supply valve to the feedvater pump turbine opened cs called for. The.No. 2 auxiliary feedvate; pump turbine ca=c up to 2600. rpm and reMned at this speed rather than con-tinuing up to 3600 rpm, which is the design speed.
Operator action at 14 minutes after reactor trip brought this pu=p up to design speed by placing the control (in the control room) in :.inual.
Failure of this pu=p to come up to speed did not racerially contribute to this event, but did result in the No. 2 stea= generator boiling dry, which added to the transient
. condition in the pr4" y system.
The rearons for this feedvater pe=p turbine to come up to speed are disc 2.ssed in detail below.
3.
STRCS The initiating event was a Steam and Feedvater Rupture Control e
System (SFRCS) Channel 2 co=entary one-half trip from an u:0cnown
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cause that went back to normal before the station co=puter could record the source.
This one-half trip caused the following events:
1.
The startup feedvater control valve (S?7A) on steam generator No. 2 closed.
This caused a loss of feedvate,r incident on stean generator No. 2.
2.
A one-half trip on Channel 2 sealed in on both =ain stes= line isolation valves, (MSIV).
This on~e-half trip deenergized at least one solenoid valve on each MSIV, and resulted in a "Mn Sem Iso 1 (2) Trbl" alarn on th'e station co=puter for both MSIV's.
This comentary one-half trip could have been caused by a spurious contact opening or a loose connection in a vire in,a S?RCS input signal from a steam generator low pressure switch, a steam generator low level bistable or a cain feedwater high pressure differential switch. "The coment'ary one-half trip eduld-also have been caused by
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trouble internal to the SFRCS cabinets.
All possible causes were investigated.
As a result of this investigation, it was dete==ined that an input buffer card had failed.
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Auriliarv Teedeu=3 Turbine Covernor The auxiliary feedpe=p No. 2 f ailed to accelerate to the nor:21 speed of 3600 rps.
The stean isolation valve opened preperly and the pe=p case up to about 2600 rpm.
The governor, a Woodvard Type P.-?L vith a speed changer motor driving the canuti speed settb.g knob, was calling for a higher speed (the speed cha'nger nate.: was turning in the " increase" direc-tien). As required, the governor was lef t in accordance with precedures with the speed adjust =ent at the " full speed" position when the pe=p was shutdown. ITnen the pu=p was called on to auto-start, with steam generator level belov setpoint, the speed changer notor continued to drive, through a slip clutch, in the " increase" direction.
However, the speed setting cethanism was already at its mechanical high speed step a; plying a binding torque to the "T" bar, a portion of the " feed back" Lickage, not allowing it to drop dovn and allev :he piston red to cove down in the increase speed di:ecticn. The undesired binding in the feedback linkage Save the severnor a f alse signal that the turbine was at the desired speed. Once the torque was recoved. by operator remote nanual action, from the "T" bar, the "T" bar dropped dovn and the auxiliary feed pu=7 turbine proceeded to the high speed step ( 3600 ryn)..
D.
Pressurizer Pover Eelief Valve
- When the reaeter coolan: systen pressure reached the setpoint for the power relief valve, 2255 psig, the valve opened properly.
However, there is a seal-in relay which then keeps the valve open until pressure is reduced to a lower " reset" pressure -(2203 psig).
This seal-in relay that controls the l
closing of the valve was nissing fron the circuit.
Without the re*ay, the valve reclosed as soon as pressure decreased below the "open" setpoint. The result was open-close cycles as pressure went above and below the "open" setpoint pressure instead of one or two longer blows to relieve the high pressure down to the " reset" pressure.
Af ter apprcxicately nine open-close cycles the power relief valve ra-afned in the cpen position.
When the valve was dis-asse= bled it was found that the pilot valve was stuck in the open posit on causing the =ain valve to stay open.
The pilot i
Yalve was stuck in the open position due to unknown foreign
._aterial binding the ste= in the guide area of the pilot valve no :le.
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SYSTEM TRANS1:.h n AND AKAI.TSIS A.
Transients During this rapid depressurization event (see section 2 above and Exhibit 3 Figures 7-1 through 7-4), the reactor coolant system pressure d:cpped fres about 2300 psig to about 930 psig in 7h cinutes and gradually recovered to 1800 psig in two hours (see Figure 4-1).
During this 7 ninutes the t aaetor coolant outlet te=perature dropped at varying rates from about 580 F to about 533 F.
Approxicately 30 minutes af ter this initial te=perature change, a second ever and s= aller te=perature change from 540 ? to 505 F occurred over a 21-=inute period.
Following this second terperature decrease, the, temperature gradually.-increased over a 2-hour period to 528 F.
The reactor coolant inlet temperature changes and, durations were similar to those of the reactor coolant outlet te perature (see Figure 4-2).
The secondary side pressure in stean generator No.1 reached a
-n 4 us of 1050 psig and decreased to about 860 psig within 15 cinutes (see Figure 4-3).
The secondary side pressure in stean seaerator No. 2 reached a =aximum of 980 psig, decreased to 610 psig in 14
<-utes, and returned to 860 psig in 2 =Lautes.
Twenty cinutes later Che pressure in steam gen.erator No. 2 again decreased to 610 psig and graduilly recovered over a 2-hour period (see Figure 4-4).
3.
Analysis of the Reactor Coolant Systen 3&W has co=pleted its evalu ~ ion of the September 24 transients and has found no harnful short or long tern effects on the reactor
- coolant system ce=pesents.
For this evaluation it was conserva-tively assu=ed that the total temperature decrease occurred at the initial rate.
This results in a 43
? decrease in the reactor coolant outlet tarperature over a 6-uinute period.
The design specification for Davis-3 esse Unit 1 recuired the evhluation of 40 cycles of a rapid depressurization event, which included a decrease in the reactor coolant pressure fro = 2200 psig to 800 psig, a change in the reactor coolant system average temperature from 563 Y to 500 F in 15. cinutes, and a decrease in secondary system pressure from 1050 psig to 640 psig.
The =ajor difference between the actual transient and the design transient is the rate of the ta=perature change in the reactor co.olant system. The actual rate of te=perature ch'ange was t rice the rate of the design transient, but the total te=perature change was only 7S" of that of the design transient.
The net result is that the fatigue usage of this one rapid depressurization is about the same as that predicted for one cycle of the design transient.
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-2 As a nore direct conparison, the transient event idertified was m,m17:ed for the renetor vessel shell and compared to the design transient.
The results were that the range in thermal ramal gradient stress for the actual transient was 5400 psi, and the range of ther=al radial gradient stress for the design transient was 6600 psi. This co=parison vould be representative of other thicknesses throughout the reactor coolant systen pressure bocadary.
The conclusions of the analysis are:
(1) Stresses in the pressure boundary did not exceed those already calculated on a design basis.
This is verified by the actual pressure not exceeding 2500 psig and the thernal transient being less severe than a conbination of design transients for a ' rapid depressurization and a reactor trip.
(2) Tatigue life of the reactor coolant co=ponents is not
.1 affected if one cycle of the reactor trip design transient and two cycles of a rapid depressurization design transient are considered to be used for this transient.
Two cycles of the rapid depressurization transient are necessary be-cause the E?I systen was actuated twice during the event and two cycles are necessary to reflect the ther=al transient in the high pressure injection nozzle.
The effect of the entire event on the fatigue life of the sr steza generators can be accounted for by using one cycle of the design transient for rapid depressurisction and one cycle of the design transient for loss of feedvater to one generato'r.
(3)
The eff.ect of the change in water level on the pressurizer has a vezy ninor effect on the press'urizer shell stresses.
The pressurizer has been previously analyzed for the thernal effect cf water-stean interface, and the change of level
- does not affecc that analysis.
(4) No significant teer=al shock should occur to the heaters, -
because the hentets were deactivated due to a low water level sensor and not reactivated until the level recovered.
(5) No dynanic effects were caused by the rapid pressure decrease.
No specific analysis was done, but a dynanic response of the shells vould require a large pressure change in the order of
- 114 seconds, and the actual change was on the scale of ninutes.
4 6
+
3 The reduced feedvater flow to steam generator No. 2 was not sufficient to caintain a water level during the first five cinutes of the event and this steam sendra:or boiled dry. The pri=ary concern with a dry generator is the tube to shell te=-
perature differene2.
Is this event a water level was established before the systen cocidown was started, and acceptable tube to shell te:perature differences were =aintained.
This condition is sbnar to th'e loss. of feedvater design transient, followed by restart of a dry pressurized generator using the aas:111ary feedvater systen.
The burs: rupture disc on the pressuri:cr quench tnk resulted in a stream of s:can and water i= pinging on stea= generator No. 2.
Tais strean re=oved a see:1ca of insulation 10' high and 20' vide from the lower shell of the generator and impinged directly on the generator shell. The te=perature' of the ispinging water was assured to be 2120 7.
A conserva:1ve evaluation of the rapid temperature change in this local region of th'e vessel shell was perfor=ed.
Tne results of this evalua ion indicate that this one event used less than 1::: of the to:a1 fatigue life of the vessel.
Tne predicted fatigue usage factor for the 40-year design life of the vessel in this area was less than 0.10.
This jet i=pingenent did not significantly reduce the fatigue life of the stean generator.
The reae:or coolcat pe=ps (2C?) experienced the following conditions during the Septecher 24 trarsient.
All four RC pa=ps were subj ected to the following:
0:00 Reactor trip 1:10 STAS trip 1:12 Seal return valves shut for 1:16 1:13 Seal inj ec: ion valves shut for 1:52 All fcur pu=ps operated for 1:15 vith no seal injection and.no seal return flow during :he RCS de-pressurization 2:25 Seal retu:n valves open 3:05 Seal inj ectica valves open 6:00 Steam.forcation Pressure oscillating near PSAT f0r 30 to 45 cinutes 36:07 Tota.1 seal injection flov lov alarm Pu=p 1-1:
7:04
?c=p tripped 7:45 Shaf t stopped 36:07-About one =inu:e of lov ' seal inj ection flow (near 2 gp=)
1
?lov # alance starved seal injection 36:30 Seal return valve shut 1:12:55 S andpipe level high i
1:17:07 Standpipe level nor:al j
j
~..
v S
Pump 2-2:
4:20 Eigh vibration 7:04 Pu=p tripped 36:07 Lost seal injection for _about one minute
~
36:22 Seal return valve shut for about 40 seconds Checkout of the. reactor coolant p_~ps was initiated to assess vhether maintenance and/or repait was required as a result of the transient.
Operational checks were required to de=onstrate that no signifi-cant da age had occurred to the pump bearings, shaft and senis.
The first series' of tests were performed in Mode. 5 due to operational restrictions.
Later operational checks were perfor=ed in P. ode 3.
Each pu=p was to be opera.ted individually for a duration not to exceed ten (10) =inutes, providing all defined para:erers renained vithin established li=its.
The operational sequence was as follows:
' l.
Lif t pu=ps were started and pu=p shaf ts rotated by hand.
Torque values were not to exceed 200 ft-lbs. A stethoscope vas provided to detect any unusual =echanical noises in seal housing area.
(This was satisfactorily ce=pleted on 10/3/77).
2.
Mode 5 testing at 225 psig.
2.1 Instru=entation Required:
a.
Upper and lower cavity pressures - all four pe=ps.
b.
Both horizontal vibration probes - all four pu=ps.
c.
Systen pressure or suction pressure.
d.
Vertical probe on 2-2 pe=p.
Standpipe leakage was collected and ceasu' red during e.-
the tes't.
2.2 Co puter Data -
htout NSS special su=aary trend for r -ning RCP cvery 15 seconds.
2.3 Tne folleving limits were not to be exceeded:
a.
Shaf t vibration - 15 mils peak.co peak.
4 mmph-
- B-b.
Total standpipe leakage (upper seal leakage) plus seal return should not exceed 0.6 gpm.
If, during the test this limit is exc'eeded, the possibility exists of an open seal.
In no case will total seal leakage be allowed to exceed 1.5 gpm.
If this limit is exceeded, maintenance vill be required before further pe=p operation.
All oth'er no'r=al plant licits and precautions prevail.
c.
2.4 Sequence of Operation:
a.
Secure standpipe flush.
b.
Establish seal injection in accordance with plant operating procedure.
c.
Measure and record standpipe leakage and return flow.
Confir= that total leakage limits are not exceeded.
t d.
Assure ~ coc=unication between control roon and personnel stationed at RCP standpipe leakage drain line.
e.
Countdo n from 10 to 0 Start strip chart recorders at high speed; Start Reactor Coolant Pu=p 2-2 in accordance with plant operating procedure.-
/'
After approxi=ately 11 sec., reduce strip chart speed.
~
f.
Run pe=p for two (2) ninutes unless any above limits ~
are exceeded.
g.
Data assesscent by 35W and Byron-Jackson representatives.
~
h.
Following assessment of data, pump =ay be run for an additional five (5) minutes to allow for venting procedure requirements.
Follow kbove seque' ce on 2-1,1-2 and 1-1.
1.
a
- j. Assessment of this data vill determine whether any maintenance is required before high pressure operati'on is allowed.
3.
54 diar, tests were repeated with system pressure at greater than 1300 psig before a final determination on the condition of the pu=ps was =ade.
S e
e
sa -
All four reactor coolant pu=ps were run on 10/5/77 with the following results:
RC? 2-2 10/5/77 Run (2 ein.):
Stean pressure 225 psig 3rd Seal leakage 2nd Seal cavity pressure 165 psig plus seal return flow 4.4gp=
3rd Seal cavity, pressure 123.9 psig Eorizontal vibration 5 - 7.5 mils Vertical vibration.25 mils After the 2-ninute run, the pu=p was run for 10 minutes for syste:: venting. About 30 seconds before the pump was shutdown, there.vas a step increase in vertical vibration to 2.5 mils.
Tne pu=p was run again on 10/6/77 for 10 minutes to check out this phenenenon. Tne ve'rtical vibration was again.25 rd.ls until about 5 seconds before shutdown, when it increased to 2.5 tils.
To allow a longer run tice, 2-1 and 2-2 pu=ps vare run together for 10 minutes, then 2-2 was run alone for 10 ninutes.
The vertical vibration stayed at.25 =ils for the entire run.
This was conitored during pu=p runs during plant beat up.
It should be noted that the step increase in vertical vibration was later assessed to be spurious instrument.
. noise as a result of a loose connector on 'an instru=ent line.
Af ter the connec:or was tightened, vertical vibration recained less than.25 mils peak-to-peak a:plitude.
EC? 2-1 Stea: pressure 225 p~ ig 3rd Seal leakage s
g.4 EEm
~
2nd Seal cavity pressure 132 psig plus return flow 3rd Seal cavity pressure 70 psig Eori:enal vibration 5 - 7.5 =ils 2C? 1-2 Systen pressure 225 psig 3rd Seal leakage
< *4 82 2nd -Seal cavity pressure 40.29 psig plus return flow 3rd Seal cavity pressure 81.3 psig Eori:catal vibration 5 - 7.5 mils RC? 1-1 Systen pressure 225 psig 3rd Seal leakage
<..4 gp=
2nd Seal cavity pressure 77.98 psig plu= caer.m flew 3rd Seal cavity pressure 89.27 psig Horizontal vibration 5 - 7.5 mils t
O O
+
9 e
. a, o The apparent discrepancy on seal cavity pressures on 1-1 and 1-2 vas checked on 10/6/77 by installing pressure gauges, at the pressure transmitters.
The gauges read as follows:
1-1:
2nd Seal Cavity Pressure 184 psig 111 psig 3rd Seal Cavity Pressure 1-2:
2nd Seal Cavity Pressure 184 psig 112 psig 3rd Seal Cavity Pressure The readings indiente the seals are staging properly.
Based on the abo've performance, En' saw no concern which would justify =aintenance at the ti=e.
By 10/13/77 all four reactor coolant pu=ps had been run at a-system pressure greater than 1300 psig.
RC Pu=ps 2-1 and 2-2 have continued to run from the initial cold pu=p starts.
Below is a typical line of data from each pu=p.
RC? 2-1 1650 psig f
System ?ressure 2nd Seal Cavity ?ressure 1034 psig 3rd Seal Cavity Pressure 500 psig Eorizontal Vibration 3 mils RCP 2-2 Systec ? essure
- 1650 psig 2nd Seal Cavity Pressure 1075 psig 588 psig 3rd Seal Cavity Pressure Eorizontal Vibration 3.5 nils RC? 1-1 1650 psig Steza ?: essure 1056 psig 2nd Seal Cavity ?ressure 540 psig 3rd Seal Caviry ?ressure 4 mils Ecrizontal Tibration RC? l-2 1650 psig System Pressure 920 psig 2nd Seal Cavity Pressure 520 psig 3rd Seal Cavity Pressure Horirontal Vibration 3 milss 4
A-
3 1.s -
e e
o 3ased on the above data, B&W felt that all.four pu=ps were in gcod op'erating condition and require nothing core at this, ti=e than periodic cenitoring.
e 357 has reviewed the results of the operational checks and has concluded that no detectable damage has occurred to the
- t
- :p cc=ponents. 35*J considers the reactor coolant pt=ps to be serviceable for sustained full operational conditions vith no require =ents -for maintenance.
A = ore detailed malysis was co=pleted to assess the core ther=al conditions during the September 24 depressurization event at Davis-3 esse Unit 1.
Core conditions were analyred to (1) deter =ine if steam was produced in the core, (2) deterni.ne the cart =un internal fuel rod pressure during the transient, and (3) dete: ine if naxi=um lif e force exceeded the li=it.
71gure 4-5 shows transient ther=al conditions as monitored by the reacti=eter. The systen pressure is =easured at the pressure cap, which is approximately 65 feet above the top of the, core.
The RC pressure at the top of the core is approxi-ately 50 psi higher than the censured pressure because of tarecoverable and elevation, pressure losses.
As shown in Figure 4-6, the predicted core coolant temperature is slightly higher than the sini un saturation te=perature (based upon taasured pressure); however, there is some uncertainty in both the reasurecent and the prediction.
Therefore, it is possibly that se=e vapor bubble for:ation (stean bubbles in vater) co61d have occurred t-ithin the core. An exa=ination of the rea.ctimeter data (Figure 4-7) indicates that the RCS pressere level was near the saturation pressure for less than one hour and that during this time period the pressure oscillated with a variation.of i 50 psi.
Therefore, the =aximum time period during which the core could have been subjected to bubbly flow vas less than one hour.
If bubbles were formed during this period', the for:ation
' vould be in the liquid as well as on the surface, as opposed to forr.ition from a hot surf ace. With the temperatures, ti=e duration, and type of for=ation, no significant.cffect on the co=penents would be predicted.
~
Prior to the depressurizaticn event the reactor had been operating at 15% power for approximately one ueek.
I= mediately prior to reactor trip the power level was 9% of rated power.
The core buruup vas 1 :.ceD, therefore no significant fission gas prodccrion had occurred and none was released.
During the 60-=inute cine period in which the indicated RCS pressure was es izated to vary from 900 to 1000 psia at the top of' the core, the average coolant te=perature was less than 540 F and no sign'.ficant heat generation' occurred in the fuel.
An initial e
4 e
p
~
evaluation had predicted tensile stresses in the cladding based upon a W-u:n pressure differential across the cladding of 200 to 300 psi.
This evaluation had been based cpon a 30L TAFT analysis with an arbitrary safety factoi.-
added to ensure that actual conditions vould be bounded by -
the prediction.
A more recent analysis, again using TAFY, has resulted in a predicted ca.v.icun internal fuel rod pressure of 1000 psia. Tnis analysis considered as-built fuel properties and hot, near zero power conditions at a coolant average te=perature of 5400 F.
On the basis of this analysis it is
~
concluded that the' fuel rod cladding was not subjected to any significang level of tensile stress during the subject depressurination event.
Because the cladding was not subjected to a large, long ters tensile stress, no significant long ter:s effects on the cladding resulted. Tne tensile stresses which could have occurred would have little effect en the cladding due to the snall stress level.
and the short duration of the censile stress.
6 lb/=in Assu=ing a coolant te=perature of 537 F and 150 x 10 systen flov (per Figures 4-8 and 4-9), the net lif t force vill be less than 375 lb. Tne maximu:s allevable lif t force is 472 lb.
Therefore, we conclude that fuel asse=bly lift-off did not occur.
.0 6
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IOUT? VENT DAMCE, CLEASTP.GD REPAIR A.
Entrv and Cle2nuo
?rior to entering containment, air sa=ples were collected at 2E5030 (contain=ent air conitor) for radioactive noble gases, particulstes, iodines, and tritium; no airborne radioactive raterials were detected. knen contain=ent was first entered
- at 0550 on Septe=ber 25, 1977, to deter =1ne the levels of cont *-5=tica, dirt was found on the valkvays on elevation
' 565' and 585' in the e20: side of containment, and on 545' elevation the floer was coepletely covered with dirt which was vashed do a duri:g the period when stea= vas being released f ca the quench bank and condensing on contain=ent structures.
The dirt was contaminated with activation products of Cr-51, WIS7, Co-58, Zn-97, and Na-24 which were present in the reactor coolant syste=.
Smears of the dirt indicated levels 2
up to 40,000 dp:/100cm,
Decosta=ination was accomplished by shoveling gross amounts
,of dirt into dru=s, and vacuum sweeping the re=ainder.
The level of contaation in valkways was reduced to neet clean area 1 N ts.
Lir sa ples collected during the decentamination vork verified that contamination did not become airborne.
Tne outer surf ace of stes: generator 1-2 was inspected in the reglen where the metal reflective insulation was blown off.
3ori. acid stains were cbserved on the outer surface of steam f
generator 1-2; however, these minute quantities do not present any concern since the te=perature of these surfaces are on th.e order of 5000 F.
3.
Icui ent Danage Tne pressuri er quench tank rupture disc ruptured from high pressure in :he quench tank.
The steam from the pressuri:er quench tank vent damaged metal reflective insulatica on the
. lever part of No. 2 steam generator. A ventilating du,ct above the quench tank was bent, and a ventilation louver had to be replaced.
Several pressurizer heater cables vere da pened fro: the coisture, causing low insulation resistance, and had to be dried out.
Tour cabics were also found shorted to ground, but it is not knot.n if the failures were a direct result of the incident. Two light fi):tures and a ccebustion detector sensor in the quench tank area vere also da aged.
Treaty-thre'e (23) panels of reflective insulation were deforced, loosened or detached from the lover e>:terior of the steam gen-erator. The panels, fabricated from thin stainless steel sheets vith air spaces between the=, are apprc:cinately 36" x 30" x 4".
e a
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_ 27 -
t Tne panels are formed to the contour of the steam generator and attached to the exterior oc. a frane to support rhe weight.
3uckles and clips fasten the panels together, panels blovn fron the stean generator fell to the floor, piping and ventilation duct in the i==ediate vicinity.
Some panels were repaired and reused; others had to be replaced.
The da= aged panels were intact but were bent.
C.
Recairs
. All damaged equipnent was repaired or replaced.
Instrumentation and equipnent in the area uns checked or tested for possible da= age from the steam and water.
Trenty-three (23) panels of reflective insulation vere replaced.
The other effected panels were straightened, reposirioned and reinstalled on the steam gen'rator.
e All essential and autonatically-controlled pressurizer heaters were returned to service.
The vet pressurizer heater cables
' vere baked, heated or air dried to restore insulation resistance to vendor reco== ended values.
Only two of the four cables shorted to ground were replaced '-ith spares. The other two are on order.
A new rupture disc was installed on the pressurizer quench tank.
rne deforned ventilation duet was straightened and a new louver,
vas installed in the duct.
Tne damaged light fixture and conbustion detector vere replaced.
9 e
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Q.
q
6.
SYSTEMS /EOUIPMENT MODIFICATION AND TESTING.
A.
STRCS The Davis-Besse Instru=ent and Control (I&C) Group has tested logie channels 2 and 4 (channel 2) of the SFRCS, since it was indicated that the closure of SP7A (start-up feedvater valves) led to the secuence of events on 9/24/77.
Logic channels 2 and 4 are the only SFRCS c,hannels that accuate S?7A.
"On 9/26/77, Maintenance Work Order (WO) IC-622-77 was written
.to check the main steam line pressure switches PS 3687A through PS 3687H. A calibration check was coupleted on 9/27/77. All pressure switches acenated within 12 psig of the 612 psig set-points.
I6C personnel had nothing to report fron the visual inspection.
On 9/27/77, rio IC-636-77 was written to investigate the ra-aining inputs to the SFRCS.
Pressure differential switches 2686C, 2686D, 2685A and 26853 were tested per ST 5031.14, Section 6.3.
The setpoint of the pressure diff erential switches tested ranged fren 176 psig to 187 psig, the setpoint being
'177 120 psig.
m The stean generator level inputs to the STRCS were tested per j
ST 5031.14, Section 6.4.
Againi logic channels 2 and 4 were t es.t ed.
All bistables tripped at the desired setpoints.
The desired trip setting is.509 i.013 volts and the range of voltages for the.bistables tested were fron.5054 volts to
.509 volts.
In addition, the level transnitter calibration was checked per ST 5031.16.
I&C tested for any non-Sinearities between transnitter input and output, especially at the lower ranges.
LT-SP9AS, LT-S?9A9, LT-S?936, and LT-S?937 vere well within the acceptable linits as specified by ST 5031.16 and no non-linearities were observed.
The inputs to the SFRCS fron the loss of 4 reactor ' coolant pe=ps vote not tested since this input actuates auxiliary feddvater only.
This input does not affect the feedwater valves o'r cain steam isolation valves.
In addition to testing all the input devices, I&C checked C5792.
This is. the cabinet for logic channels 2 and 4.
All.
inputs and outputs were nor=al for existing plant conditions.
I&C checked nechanical connections on the input and output b.uffers, and induced nechanical vibration on the input buffers, n logic panels, and output relays without output buff ers, s e any systen effect.
The nain logie panels were heated slightly with a beat la=p and slowly cooled ' to check for ther=al varia-tions, but this had no effect on the systen.
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.0a Stpte=5er 29, I&C complated thnir chtek of SFRCS ter-inntions. Tha following are the results of that chtek:
1.
Screvs on T337 (yellow & blue) were tightened 1/2 of a turn.
This is as input to the sorenson 15 volt logic supply for C5792.
2.
En C5721 (Feedvater Panel) one loose screv was found on 21T311 Ter * ' 17 (lef t side of TB).
This screw required little =ovement to thoroughly tighten.' This is a cain stea:t pressure switch input to logic channel 2.
- 3..
In C5721, 21T327. had 3 loo'se screws. Terninal 17 (right side of Ter=inal Board) had to be tightened 1 full turn.
This is a min stean pressure switch input to logic channel 3.
Ter=inal 18 (right side,of Ter=inal Board) had to be tightened 1/2 a turn. This is a pressure differential switch input to the SFRCS.
Logic channel 3 Te,rrtinal 4 (lef t side of T3) had to be tightened slightly. This is a =ain staan pressure switch input to the SFRCS.
On Septe:Ser 30, BIS 4870 A and C vere tightened to their counting.
These are stacked s -itches.
The switch units the=selves were secure, but the estire package was loose en the =ounting.
This switch unit being loose would probably not have af fected systen operatien.
Temporary jumpe s were installed to prevent an inadvertent main steam isolation valve closure during STRCS checkout.
On Oc:ober 6, 1977, the Stean Generator level instrumentation was checkec' '
out.
ISC was specifically looking for noise spikes that could have
.r caused an erroaccus trip. All analog inputs and outputs only had a 20 MV
,, typical) AC n oise.
DC signals appeared " clean".
(
Da Oc:ober 8,1977, eight 6 cha:inel chart recorders were patched into the systen for centinuous =onitoring.
The recorders were connected per the attached sheets. The system was then checked out for operability.
Pressure dif f erential and Steam Generator low level trips were tested.
Since the S?RCS was blocked due to lov steam pressure, pressure switch trips were -initiated and the input to the logic was verified by a voltneter reading. These pressure switch inputs vill be further tested during the STRCS conthly, ST 5031.14, Section 6.2, at a later date.
Connecting the recorders has indicated no effect on system operability.
e
LOGIC CHANNEL 1 SPRCS TEST CONNECTIONS CONNECT SIGNAL COFr.~I COMNON TO LEAD TO '
iPUT 3br::2 TEST POINT Bur:m TEST POINT P2 CORDER
' CHANNEL
' 33689A 1-1 TP4 1-1 TPS 9.1.48 1
336893 2-1 TP2 1-1 TP7 9.1.48 2
53689C 2-2 TP2 1-1 TP9 9.1.46 9
33689D 1-2 TP4 1-2 TPS 9.1.48 4
>2686A 1-3 I?4 1-3 TP7
- 9.1.48 5
32685C
~ 2-3 TP2 1-3
'TP9 9.1.48 6
- SP938 1-4 TP4 1-4 TPS 9.1.46 7
- S?9A6 2-4 TP2 1-4 TP7 9.1.46 8
i V. Power 2pply Output 1-5 TP2 1-5 TP10 9.1.48 3
581 2-7 TP2 2-7 TP10 9.1.46 10 571 2-6 TP2 2-7 TF16 9.1.46 11
.c 386 1-6 TP2 1-6 TP10 9.1.49 12 1
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0 LOGIC CHANNEL 2 SPRCS TEST CONNECTIONS
~
CONNECT SIGNAL
~
CONNECI' COEON TO LEAD TO N?UT Burr u.
TEST ?OINT BurrER TEST POINT RECORDER CE NMEL S3687A 1-1 TP4 1-1 TP5 9.1.44 1
S36873 2-1 '
TP2 f-1 TP7 9.1.44 2
S36870 2-2' TP2 1-1 TP9-9.1.41 9
S3637D-1-2 TP4 1-2 TPS 9.1.44 4
DS2635A 1-3 TP4 1-3 TP7 9.1.44 5
DS2686C 2-3 TP2 1-3 TP9 9.1.44 6
T S?936 1-4 T?4 1-4 TP5 9.1.41
- 7 T S?9AS 2 TP2 1-4 TP7 9.1.41 8
.5 V. Po.:er upply output 1-5 TP2 1-5 TP10 9.~1.44 3
$680 2-7 T?2
'27 TP10 9.1.41 10 1
2-6 T?2 2-7 TF16 9.1.41 IT 896 1-6 TP2 1-6 TP10 9.1.41 12 k'
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~
LOGIC CHANNEL 3 SFRCS TEST CONNECTICNS
~
CONNECT SIGNAL CONNECT CO)".50N TO IZAD TO FUT Bu::.x TEST SOIh7 Bus :&
TEST POINT RECORDER C3ANh'EL 3689E 1-10 TP4 1-10 TPS 9.1.47 13 3689F-_
2 TP2 1-10 TP7 9.1.47 14 3689G 2-11' TP2 1-10 TP9 9.1.45 21 3689H 1-11 TP4 1-11 TPS 9.1.47 16 S26863 1-12 T?4 1-12 TP7
'9.1.47 17 S2685D 2-12 I?2 1-12 TP9 9.1.47 18 SP939 1-14 T?4 1-13 TPS 9.1.45 19 SP9A7 2-13 TP2 1-13 TP7 9.1.45 20
- v. Power pply Output 1-14
'TP2 1-14 TP10 9.1.47 15 S1 2-15 TP2 2-16 TP10 9.1.45 22 71 2-16 I?2 2-16 TP16 9.1.45 23-TP2 1-15 TF10 9.1.45
, 24 86 1-15 9
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LOGIC CHAIO EL 4 SFRCS TE3T CO?GECTIONS CONNECT SICIUl.
.CO!CIECI C0YS.ON TO LEAD TO
?UT E u:::.r.
TEST ?OI'C 3D::u TEST POIIE RECORDER CEMNEL 3637E 1-10 TP4 1-10 TPS 9.1.43 13 e
3687F 2 e TP2 1-10 TP7 9.1.43 14 e
3687G 2-11' TP2 1-10 TP9 060404 21 3687E 1-11 TP4 1-11 TPS 9.1.43 16 S2685B 1-12 TP4 1-12 TP7 9.1.43 17 S2686D 2-12 TP2 1-12 TP9 9.1.43 18 SP937 1-13 TP4 1-13 TPS 060404 19 SP9A9 2 TP2 1-13 TP7 060404 20 e
V. Power pply Output 1-14 TP2 1-14 TP10 9.1.43 15 80 2-15 TP2 2-16 TF10 060404 22 2
2-13 TP2 2-16 TP16 060404 23" TP2 1-15 TP10 060404 24 96 1-15 O
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34
'On 10/23/77, thh SFRCS again trippad from a epurious signal. The Startup Feedvater Valve on stean generator No. 2 vent closed.
This ulHmtely resulted in a valid Ste' m Generator lov icvel trip input" to the SFRCS a
and the syste= functioned as intended.
This was the first spurious trip received since the chart recorders had been donnected to the SFRCS. All information en the chdrts could be exp1n4nad except for a problem on SFRCS logic Channel 4 co=puter alarm,-
T~ is particular channel on the recorder was internittently failing, i
P680.
n l
giving spurious t-ip indications.. Of the 48 total chart recorder ch5-nels, this was the caly one that had failed.
~
I&C Tee 4c4 m " checked out" the bad recorder channel for operation.
They found that the cha:inel was sensitive to any nechanical vibration, it did respond t'o a given ' input, and that the pens were slightly misaligned.
From =11 of the infor=ation gathered it was concluded that the indication
.on the bad recorder channel was an input from the SFRCS.
The logic point under question then was the conputer point ("p680" lov Main Stean Pressure Trip).
E:<=-ining other charts indicated no change in the input :o SIRCS logic Channel 4.
Thus it was concluded the problem was internal to the syste=,
In er=- 4ning the logic control diagran, it was deceinined 3 IC " chips", 2 input buf fers and associated viring could have caused the fault.
ISC personnel replaced all of the above equipment, with the e tception of the interconnecting v4. ring.
The viring and buffer connections were visually inspected, and no faults were observed.
A functional logic :est was perforned and the systen responded satisfactorily.
Power Engineering had contacted Consolidated Controls Corporation, the
=annfacturer, dad their representative was on site the norning of 10/26/77.
The nanufacturer also 'reco=nended. changing the same equipment that TECo
-I&C personnel had changed.
Tne --=-efacturer perforned a response time check on both input buffers La question.
The response tine test shoved no defects.
TECo I&C personnel continued to =nitor one of the two input buffers in a test set.
Failure.
of one input buffer did occur on the test sei, which indicates that this was the cause cf the half trip.
The canufacturer's representative also took a look at the logic system with an oscilloscope.. He was looking for any erratie, noisy points, but i
everything tes:ed appeared to be trouble free.
The two input buf fers vill re4" with TICo' for further test and evaluation, while the 3 IC chips were returned to the canufacturer for evaluation.
rne canufacturer's representative on 10/27/77 conciled a list of addi-tional points they want =onitored. TECo I&C personnel are assisting to connect cp the recorders.
A S
J
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)
~ 'Aftcr th2 10/23/77 cvent, a study w:s also conducted to sne if any single 120 VAC or 125 V DC f ault induced voltage dip could have cacsed the one-half trip on bo'th P.SIV's and closed the SG-2 SU b.ontrol valve.
This study revealed that no single fault on these power supplies could have caused this problem.
The following changes have been rade to the design of the SnCS since the Septe ber 24, 1977 incident:
1.
. Annunciator vindevs have be'en added where computer n'ams presently exist for:
a.
Stean Generator Level Half / Full Trip for both Cunnels 1 & -2 b.
hin 7eedvater D? Half / Full Trip for both Channels 1
&2 c.
Loss of 4 Reactor Coolant Pump Trip-2.
A new annunciator and conputer alar = has been added for a SFRCS Full Trip.
3.
The reshteing of all SFRCS related alarm vill be delayed long enough to allow the computer to, record 'the event.
These changes will be made as soon as possible.
Y
~
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36 -
3.
Auxiliarv"Feedou=n Turbine Governor Before describing the modifications cade to the auxiliary feedpu=p turbine (AT?T) governor, the governor action which resulted in the binding will be described. Figure 6-1 is c.
drawing of the Woodward Governor PG-PL speed se.cting nachanis=,
showing the. governor in the bound up condition.
The sequence of events creating this condition is as follows:
l.
When the 3'odine =otor was at a mini =un speed setting, the speed setting shaf t nut was fully to the left.
The link raised the collar, contacting the base speed setting nut, i
i raising it and the "T"-bar to an idle condition. The pivot l
be'aring would be contacting the floating lever.
2.
Because the governor is not rotating, the speed setting servo renains in a fLxed position at idle (as shown).
It cannot nove until oil pressure is available.
s 3.
The thu=bscrew is contacting the low speed stop pin..
4.
hs the 3cdine speed setting notor is rotated toward high speed, the following events occur:
4.1 The speed setting shaft nut noves towards the high speed stop pin.
4. 7.
The link allows the collar to =ove downward.
4.3 The collar moving downward, allows the base speed setting nut and "T"-bar assenbly to nove downward.
4.4 The floating lever is fixed at.the speed setting servo piston end.
4.5 The low speed stop pin end of the link pushes down
~
on the thu=bscrew, which pushes down on the speed setting pilot valve until the dashpot land contacts the dashpot plug.
4.6 3ecause the floating lever is now fixed on both ends it stops coving.
4.7 The "T"-bar continues downward, following the collar.
The pivot bearing leaves the floating lever.
The "T"-bar continues dtwnward until the retainer screw contacts the floating lever.
~
4.8 The collar separates fro = the base speed setting nut and continues downward until the stop pin in the speed shaft contacts the stop pin in the speed setting shaft nut.
O e
=
b S
,er
4.9 Because the Bodine motor continues to rotate the manual speed setting knob, slipping the clutch, a torque ir placed on the speed setting shaf t nut, link and collar.
This torque against the "T"-bar causes friction that locks the "T"-bar in place.
5.
When the turbine is started, the speed setting servo piston coves downward with increasing oil flow, increasing the speed setting of the governor.
When the floating level.
contacts the pivot, bearing, the speed setting pilot valve j
begins to raise.
- 6.. When the pilot valve control land covers the metering port, the speed setting servo piston stops coving.
7.
Because the torque is still p' resent on the speed setting shaft, the "I"-bar is bound up, and the governor is at 2200r2600 rp=.
8.
When the Bodine speed setting cotor is backed off fron the stop, the "T"-bar falls down to its high speed stop, dropping the pivot bearing.
The pilot valve coves downward, increasing oil flov to the speed setting servo until the high. speed condition is reached.
9.
Any changes in speed setting shaf t position are nos nor= ally followed by the "T"-bar, pivot bearing, pilot valve, and speed setting servo piston.
I When the AFFT governors arrived at the Woodward Governor Co=pany factory, one of the governors was placed on the test stand.
While observing the operdtion of the speed setting linkage, it beca=e evident that a si=ple link from the speed setting pilot valve (plunger) to the ficating lever vould allow removal of the bellows, coupling spring, low speed pin, "C" link and dashpot plug in the speed setting pilot valve sleeve (see Figure 2).
This would allov the speed setting pilot valve to overtravel when the =otor vas' set in a high speed condition with the speed s'etting servo j
at the mini =um position (see Figure 6-3).
j 1
I The required parts were canufactured, the unneeded parts removed and thd governors were reasse= bled.
The governors were tested
)
at the Woodward factory and the tests confir=ed that the codifications did recove all possibility of the undesired bind-ing of the governors.
Surveillance testing at the station has j
also co fir =ed that the auxiliary feedpu=p turbine governors function properly.
I l
I l
1 l
1
FEN 41137 "T" Sar x
i f'T lgase Speed Setting. Nut
,e --
f ow Speed Stop Pin F1.cating L.ever3 l
Retainer Speed Setting j
Se m i lf r
Servo Pisten (g
l y
1 3
Collar
___5.Eh.!&
- . a..,
I
,of<f // ///A 3 d j s.
4y Link
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frg Th0=bscre-g$1 EA Manual Scaed N
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' Set-ing knob i
i ar-
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(motor input) l*
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To Soeed A
Sett'ing Servo ^
V
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(peed Setting Pilot Valve i%
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s g 96 h
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Da'shpot Area
. ?
Dashpot Plug
/sY l'
FIGURE 6-I e
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39 MiGH S'!10 sfC8 --- s
. - Fi!.) Ace.: Tint.G Nur.LtA'auAto
- AOJi/s Ts*.C sE75CAE3 \\
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e SHUTCCV.N NUTS 0 h -h.
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stat?
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PVA:P GEARS CHEOst VALVE RO TA!iNG *lLOT-J sOPENs VALvs SusHiNC
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FIGURE 6-2
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FEN 41137 y
Y j
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IT 1
Floating 1.ever\\
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k I kr Pilot Valve
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Rilot vaIve can overtravel -
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alicwing pivot bearing to
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k g
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) L_Jb FIGURE 6-3 e
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C.
Pressurizer Power Relief Valve On September 28, 1977, the valve was completely disassembled.
The cain valve was found to be clean.
The seats on the no::le and cain valve disc vere lapped.
The pilot valve was found stuck in the open position and it was thought that the pilot stem was bent so the pilot stem was replaced and the no::le guide area was clamd up to rc=ove the = arks from the galling of-the foreign =aterial. The valve was reasse= bled and on October 12,1977,.the valve was stroked six (6) times with a pressurizer pressure of approximately 600 psi.
During this tasting the pilot valve again stuck and the isolation valve had to be closed.
The valve was again disasse: bled and under closer observation it was found that the pilot valve' stem was moving too far (3/8" vs 1/8" desired).
It was also found that the clearances between the pilot sten and the tozzle guide were too s=all
(.0005" vs desired cini=en of.001").
The clearances were opened up and the stroke of the pilot was shortened by adjustment of solenoid position. The valve was tested again successfully by stroking it twelve (12) times on October 15, 1977, at a pressurizer pressure of approxi=ately 900 psi and one time at a pressure of 2200 psi.
D. ~ lelar/ Fuse /Wirin; Checks l
Eecause of the tissing relay in the pressurizer electro =stic~
relief valve control circuit, an extensive review program of j
i checking all other relay cabinets was perfor=ed.
All relay cabinets in the plant were inspected for missing plug-in relays and fuses.
A detailed review of drawings.vas =ade to deter ine the service of each =issing item and its effect cn plant operations. The one additional relay and ten fuses found missing vere replaced.
There vere no essential functions l
af fected by the additional missin; relay and fuses.
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' tissing fuses and relay were for generator iso pha,se bus control, alarm and indications; relay cabinets power supply and heater supply circuits; tain feed pu=p turbine lube oil tank level indicatica; and reactor coolant pu=p co=ponent cooling vater return valve control.
Neither the =issing relay nor the fuses were controlled under the station ju=per and lif ted vire control procedure.
This
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indicates the fuses and relay were removed by unknown persons after checkout and testing.
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Other Actions Follo -ing this incident a. training progra= vas developed and' presented.
This program was approxirately eight (8) hours of instruction and discussion covering the events of this incident.
including a detailed coverage of the transient and the actions taken by the operators, and a refresher training session cover-ing the operation of the steam and feedwater rupture control systen.
~
The tr.'d,f ng was presented to all in the operating shif t crevs, the anage=ent and staff level engineers and the QA/QC staff.
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.B.. D a e Variables Plots.
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10 Cil Part 21 Letter on Auxiliary Feedpu:2 Turbine Governor v
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7A Tisant Chrtnology 21:34:20 Secrtup Teidvatar valva o OTSO #2 vent closed on a '% trip" of the Steam and Teodwater Rupture Control Systen (SFRCS).
21:35:18 Received a complete S7RCS trip due to low level in OTSG #2.
2'l:35:23, Main Stean Isolation valves vent closed.
21:35:26-Pressurizer ?over Relief Valve cycled 9 times before sticking open.
49 21:36:04 ~ Auxiliary Feed ?u=p (AFP) #1 was feeding #1 Stean Generator (SG).
A7? !2 did not cone up to full speed (3600 rpm), and the discharge pressure was not, sufficient to feed #2 SG.
21:36:07 Operator tripped t'le reactor.
21:37:17 Safe-- Features Actuation Syste= Incident Levels 1 and 2 were ini' ated due to reactor coolant system pressure less than 1600 psi.
21:37:33 -High Pressure Injec'tiion (EFI) Pu=p 1-2 was on and had nor=al flev.
21:37:49 EPI Pump 1-1 was on and had acrnal flow.
~
21:38:13 Re-established Reactor Coolant Makeup flow.
21:40:22 Contain=ent Nor. a1 Sunp Pu=p case on indicating the Quench Tank Rupture Disk had blown.
21:40:36 EFI Pu=ps were shutdown.
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21:43:16 Auxiliary 3:iler Systen vas' started and at normal conditiens.
21:43:41 Tripped Reactor coolant ?u=ps (RC?'s) 1-1 and 2-2.
21:44':05 Re-established Reactor Coolant letdown flow.
21:49:57 Put A7? #2 in hand and ran it up to speed (3600 rp=) and then lowered the speed.
21:58:00 Closed block valve to ?:essurizer Power Relief Valve.
1
- 22:15:22-Started second Reactor Coolant Makeup Pump.
22:22:57 Started #2 EPI Pu=p.
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22:27:24 Brought f2 Ma% Feed Punp back on with Auxiliary Boiler stean.
22:27:44 Shutdown #2 EPI Pump.
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22:33:23 Shutdown el Reactor Coolant Makeup Pump.
22:43:54 Shutdown #1 and #2 Ar?'s.'
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C.
Syntes descriotion Steam and Feedvater Ruoture Control System 1.
General The stems and feedvater rupture control syste= (SFRCS) is an aut,o=atic systes designed to protect against the following incidents:
a.
Main steam line. rupture, either upstream or downstream of r-Mn steam isolation valve (MSIV).
This condition, if allowed to
" proceed, could rapidly blow down both steam generators, resulting in a rapid RCS cool down and therefore a rapid reactivity insertion under certain core conditions.
b.
Main feedwater line rupture.
If on the stesa generator side i
~
of the feedwater check valve, this is approxi=ately the same i
1
- ~
accident as the steam line rupture; on the feedwater side of the feedwater check valve this results in a total loss of f eedvater.
c.
Loss of all feedwater. This (as well as the above incidents) could result in boiling both steam generators dry.
If thi s happens, there vould be no steam available for rtmning aM 1hry feedvater pc=ps to remove decay heat.
d.
IEss of 4 reactor coolant pu=ps (RCP).
This results in loss of reactor coolant flow and therefore auxiliary feedwater is j
needed to establish reactor coolant natural circulation flow.
-r The SFRCS, 'upon indication of conditions a, b and c above vill isolate'.
both stean generators (close the main feedwater valves and =ain steam line valves and trip the turbine) and start the auxiliary feedwater systen.
Auxiliary feedwater is initiated to keep steam available for the aM16' y. feed punp turbines and to re=ove decay heat from the reactor coolant systen.
Once this is accomplished, the operator will have time to begin a cool down in an orderly canner..
2.
Desien Criteria The design criteria for the SFRCS and the auxilisry feedvater system. are as.follows:
a.
The systes =ust perform its safety function after a single
, active failure has occurred. This means that the single fM1m e of any power supply, pu=p, turbine, instrument or control system logic channel will not prevent the system from re=oving decay heat from the reactor coolant system.
b.
A =ain steam line break upstream of the MSIV or a main feedwater break do mtream of the main feedwater isolation valve vill disable one steam generatbr.
Af ter this event both auxiliary
-feed pc=ps and turbines will be aligned to the re=aining intact 1
steam generator. This re=aining steam generator has adequate capacity to,re=ove the decay-heat from the reactor coolant-system.
6 1
s
,, + -,,, - -
j 3.
Functional Descriptien (Refsr to Inclosuras I cud 2)
The SFRCS is divided for redundancy, diversity, and testability into four logic channels. Logic chnnels 1 and 3 form ch=nel 1, and logic. channels 2 and 4 form channel 2.
In one cabinet one logic chmnel has an AC power supply, the other a DC supply:
. Logic Chn,el Cabinet Power Suoply 1
C5762A Y1 (120V AC)
~
2 C5792
.Y2 (120V AC) 3 C5762A Dl? (125V DC) 4-C5792 D2P (125V DC) -
Each logic chmnel receives the following inputs which will cause it to trip:
Six pressure switches, 'two on each main steam line set a.
at 600 psig decreasing and one on each nain -steam line set at 650 psig decreasing.
b.
Two cain feedvater pressure differential ss-itches, one fron each nain feedvater line (see Enclosure 1 for sensing points) set at 177 psid steam generator pressure higher than nain feedwater line pressure.
c.
Two level transmitters with bistables, one on each steam generator ' set at 17" decreasing level on the startup s
range.
d.
A contact from R?S pump power sensing circuit; cont.act opens on loss of all four RCP's.
The SFRCS cabinets consist basically of an AC and a DC power supply, j
input buffers, logic =odules, and output relays. The output relays de-energize to actuate their associated equipment.
They also turn out a light on the cabinet when in the tripped state.
Each input to STRCS has a test switch and light so that a trip
.of that input can be initiated for testing purposes.
~
The outputs from the 57RCS are contacts from the output relays.
These contacts are in the control circuits for the SFRCS actuated equipment. Most components rec.uire two SFRCS logic channels to,
trip to actuate.
See Enclosure-2 for a listing of actuated equipment.
There is a block feature associated with the lov steam pressure trip. To prevent the systen from actuating on cooldown, each logic chmanel has a " block" pushbetton on C5721 and on the SFRCS cabinet.
When steam pressure goes'belou 650 psig a block permissive light is received on C5721 along with annunciator and computer alar as.
When.the block button is pushed, the channel will not trip on low steam pressure _ and a "W. 5" LOW PRESS TRI? BLKD" light is actrated
.on C5721 as well'as annunciator and computer alar =s.
On a beatup the block signal is. auto =ntically removed when the steam generator pressure exceeds.650 psig..
t
+
8 Thera is anoth r bicek which is uM14rtd.cn cooldown.
If th2 d: cay heat cystem cuction valvas from tha recetor ctolant sy tem (DH11 and 12) are open, this block will prevent the, opening of the steam inlet valves to the auxiliary feed pu=p turbines.
This prevents the SFECS from ; starting the auxiliary feed pu.ps when all reactor coolant pumps are secured on shutdown. This " block" is auto =atically. re=cved when the decay heat system is shut down on startup.
4.
Systen T,oeic a.
The response of the actuated components depends on the type of trip:
(refer to Inclosure 2) 1.
On icw stiean pressure on one main steam line, both stean generators are isolated.
In ad,dition, both auxiliary feed pu=ps are aligned to the steam generator which is
~
above 600 psig.
If both steam generators go belev 600 psig, both steam generators are isolated and no, auxiliary feedvarer is initiated.
If, any other trip (such as lov steam generator level) acco=panies a low steam pressure trip, the valves will align per icv stea= pressure trip logic.
2..
On high feedwater pressure differential or low steam generator level an one stean generator, both steam generators are
_ and each auxiliary feedwater s
pu=p is alignt s feed its respective steam generator
~
(1 to 1 and 2 to 2).
~
3.
On loss of all four reactor coelant purns, each am-411ary f eedwater pe=p is aligned to its respective steam generator.
~
The stean generators are not isolated.
4.
On all of the above events, the turbine is tripped by the SFRCS.
b.
The auxiliary feedwater pu=p governor control switch in the control roon bus has 3 positionsi Auto Issential (SFRCS) r ICS Manual In the auto-essential position, the auxiliary feedwater pu=p is in auto-essential level centrol.
In the ICS position, the amr*14mry feedvarer pu=p is on level control froa the ICS; via the Eand-Auto station.
In =anual, the anviliary feed-vater pump is controlled by the operator with the Raise-lover switch.
..t 9
p
.-a.-
t Thd STRCS' starting cf th2 cuxiliary facdvetar pumpa vill e,uto-c.
coticany re. set enco th2 trip condition on tha input is removad.
l None of the valves, bovaver, will* return to their origN1 position until operated individually frca the control room or-l a new trip condition occurs..
5.
Systen coeration 4
In order to understand the operation of the STRCS system, it is best to follow the various system actions under several accident i
conditious. The fonowing <.ases s-ill be considered:
a.
Steam Line Rupture j-b.
Feedwater Line Rupture l
c.
Loss of Feedvater Pumps d.
Los's of Four Reactor Coolant Punps Enclosures 1 and 2 s~nould be used as an aid to understanding the i
description. All discussions assume 100% ?? operation at start.
I Some non-SFRCS actions are considered to aid in. understanding the j
)'
(1)
Steza 1.ine Ruoture - Assume stean line 1 shears downstream of r. 7 Steam pressure vill rapidly drop.
When either stea. generator reaches, 600 psig, all four logic channels, j
vill crip, isolating both steam generators.
(See Enclosure 2 for specific valves.)
The MSIV takes five seconds to shut, ch'e main feedwater isolation valve 15 seconds.
Both steam j
lines vill probably drop below 600 psig, therefore, auxiliary i
feedvater vill not start until one stean generator recovers to above 600 psig. Auxiliary feedvater pumps vill align as i
described in Section 3 above to feed the stea= generator that first recovers to. 600 psig, s-ith both auxiliary feed pu=ps.
The SFRCS will trip the turbine. The ranctor will
]
trip on low pressure.
1 hen both steam generators are above 600 psig, the trip condi-l tion auto =atically clears and the atsospheris vent valves nay be used for pressure control cooldown if required and i
provided no other trips are present.
(2)
Feedvater Ruoture Line - Assume feedvater line 1 shears up-4 stream of.the leedwater line check valve. Feedvater pressure vill rapid.ly drop. When either feedvater heater drops.co 177 psig less than steam generator pressure, the SFRCS will 4
isolate both steam generators and align the auxiliary feed
' pumps to their respective steam generator (1 to 1; 2 to 2).
4 The reactor vill trip on. high pressure and the STRCS will trip the turbine.
(3) Loss of Four Reactor Coolant Pumos - If all four reactor coolant.
pu=ps trip, the turbine will be, tripped by the SFRCS and the reactor protection system t-ill trip the reactor. The STRCS will initiate auxiliary feedwater.
The :. team generators will i
not be isolated.
.db 20-23 s
ENtt.03URE 1 - SFRCS ACTUATED E0.UlfMENT (FEEDWATEF,)
.(foKSTEAM YAi.V5 SEE tif.XT PA
~
MN FTEDWATER NN FEEDWATCR CJNTCL VALVF.
ICOL VALVE FW *lfo F'C.P:.5 FW f.lZ.
5 Mi FEED v
A r
~
enseva l PDS 2L1'.A-D l SG-1 MN FIIDY.% TEA SAL CoNTPa VALY*..
c custs y. w
& 2r70 AF L02 c.1X fjED PUMPi
,f h
h, a
L3LL sess a,.-
2 31b 9N AF 3571 c
i Ar 2 n z.
AF Sfl 6
AUXFEED PUM? 2 g
g,-
g x
y v
SG - 2 y
lF05 2!.35A-Dl PW FTT OWATE R C O N 7 Rot. W LV E F# 771 FOP t.A FW Lol B
MilN FEED 4
~
s-A n
1-i 8.rvs774 IRL
)
- sf A ML VM.Fa.
i, j
.y 1
j i
~
ENtLO5URE 1 SFRCS ACTUATED EQUIPMENT (57EAM) i g ATM. VENT VALYE W'
MN STri 150L V.~.LV E py gg gg 8 tqs to uay
/
- PS w
2 t,. it A O
ORAIN TO
- 9;ap.i-we C 0ND.
$Q l
'dai 1;
-STFAIS 3 L 7 7 r.
tis '01-f
. r1S 374 3Li'i n 35? ?:
_ff.
h his !Clo 3:,rrr:
- <5 % I C.
K 3L770 Q. MS 10L O.-
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X
- 0Y<
1-1 k
.g.
CM.
P X.-
F2
-I K P3 b* N 3RI
~E h
N t
5 L.*/ */ *.
.h51070.
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3 GTl t.
4 G !. 't.
2 L I ?.'
MS107 U5??T"l MSICO.
~
2G 3' Y7 t:
3657ti
- W V
X(
O MS 275
% ORP f N TO NS SW-t C.OND b
ATM PY-ICI flA
.j VENT VALYE g
^
g g
I'llCI.05URE 2 STEAll-FEEullATEft ItUPTUltE CutTTR0!. SYSTEtt ACTUATION CllAllflEI. I HS 101 113 100 HS101-1 18 394 ICS lin FW G12 FW 780 SP 7a (c 576fA)
NOTE 3 Il0TE 3 NOTP. 3 NOTE 3 CilAll!!El. 2 115 101 MS 100 HS100-1 tlS 375 ICS IIA FW 601 FW-799 sr FA NOTE 3 (C 5792)
NOTE 3 HOTE 3 liOTE 3 l.oti PitESSURE IIAlli Stuna f.lun 1 (</>00f)
SilUT SIIUT SilUT E!!UT SIIUT Sl!UT S!!UT SIIUT S!!UT SilUT SilUT
' S!!UT SilUT SilUT,
-1.0'i PRESSilitE HAIN Stena 1.line 2 ((600f)
SilUT SituT "
SiluT SIIUT Si1UT SituT SilliT S11UT SilUT SilUT SilUT SilVT SilUT Silut S1tl. - FW A P SG 1 lilull (>l77 PSID)
SilUT S11UT S!!UT SilUT S!!UT S!!UT Sl!UT SilUT SilUT SilUT Sl!UT Sl!UT Sl!llT SilllT SlH, - I'll A P SO 2 111011 ()l77 PSID)
Sl{UT SilUT.
SilUT SilUT Sl!UT SilUT SillTP SilUT SilUT SilUT S!!UT SilUT SilUT SilUT l.UU !.EVEl. SG I
(<l7" SUR)
SilUT SilVT SilUT SilUT SiluT SIIUT SilUT SituT SilUT 5110T Bl!UT SillTP SLIUT 5110 7 l.oW !.Evt:!. Sa 2
. -(Cl ?" SilR)
SMUT Sl!UT SituT SituT Silut S!!UT SituT Sil0T S!!UT Sl!UT Silut SilUT SilUT SilUT 1.05S Ol* 4,FC PilHPS Cil Atlt!EI. "1 SP 6A HS 106 HS 10'6A AF3870 AF 3869 l AF 608 IIAIN (C 5762A)
TURRINE CllAtEIEl. 2 SP 65 HS 107 HS107A AF3872 AF3671 AF 599 (C 5792) 1.0W PRESSURE IIAIN OPEN OPEN OPEtt OPEN
$1EAll I.INE.1 (<600f)
SilUT SilUT SilUT NOTE 1 NOTE I SilUT SilUT NOTE I NOTE 1 SilVT Sl!UT OPEN TRIP LOW PRESSilRE !!AIN OPEN OPEN OPEN OPEN STEAH 1.INE 2 (<600f)
SilUT Sl!UT NOTE I SituT SilUT tiOTE I NOTE 1 SilUT SilUT
!!OTE 1 OPEN SilUT TRIP STil-FW 4 P SC 1 lilGil 7177 PSin SilUT S!!UT OPEN OPEN SilUT SllUT OPDI OPEN SIIUT SilUT OPEN OPEN TRIP Sin-FWA Y Sc 2 111 011 >l77 PSID SilUT SilUT OPEN OPEN SilUT SillTP OPEN OPEN SilUT SilUT OPEN OPEN TRIP 1.011 I.EVEl. SG 1
((17" SUR)
SilUT S11UT OPEN UPEN SilUT SilUT UPEli OPEN SilUT SilUT OPEN OPEN TRIP I.011 1.EVEl. SG 2
((17" SUR)
S!!UT
' SilUT OPEN OPEN S!!UT SilUT OPEli OPEN SilUT S!!UT OPEN OPEN TRIP OPEN OPEN 1.055 0F 4 RC PliHPS NOTE 2 NOTE 2 SiluT Silltf OPEN OPEN SilUT S!!UT OPEN OPEN TRIP Il0TES:
1.
It linth minin etcani lines aro <600f, thono valves ahut.
2.
Thusq valvos will not open if Dil 11 and Dil 12 (Dil Suction from RCS) are open.
'3.
Thene valves are closed on a is channel trip.
y
pg c
- w.,u
.)
L
.)rcEc Tile: 0017,0486, N W
,'~~ L*U~.o..,eit.pl.sk October 11, 1977 LowEn.E. ROE ww Serial No. 391
' ~ o-sem nSun Docket No. 50-346 Mr.- Jacies C. hppler
~
Regional Direc or, Region III
~
Of fice of Inspection & Enforcement U.S. Nuclear Regulato:y Co:= ission 799 Roosevelt Road Cicri Ellyn, Illinois 60137
Dear Mr. Keppler:
)
This le'tter supersedes my letter.to you on this subj ect dated October 5, 1977.
In accordan c.<ith 10 CFR Par: 21.21(b), this is a report of a defect in a co poncat installed in the Davis-ncsse Nuclear Powcr Sta:ica Unic No.
The component involved is the governor on the auxiliary facd pu::ps.
"1,.
.The auxilidry feed pumps were supplied by Syron fackson'Pung Division'.
The secam driven pu=p turbine was supplied by Ierry Corporation to Byron Jnckson.
In turn, the tu:bine governor w:s supplied to Terry Corporation by Woodward Covernor Co=peny.
The turbine governor is identified as a type l'C-PL, which has a 'servonotor control employing a Bodine Electric Company motor.
The defect involves a potential for the governor to bind under cettain conditions and preventing the tur,bine fron coning up to design speed.
The operating procedures for :his equipment called for the governor to be placed in the high speed stop position prior :o shu: ting down the turbine.
Investiga' tion has shown that with the Bodine servo =otor driving against the high speed stop, a misalignment force is applied to the T-bar of the governor linkage. This misalignment force treates a potential for the governor to bind at a speed position less ;ne.n duisi.qued upon a turbine startup. This misalignment force does not always cause the governor to bind and this misalignment force can be re=oved by driving the Bodine servocotor away from the high speed stop.
The safc y ha:ard which could be created is the potential for both auxil-inry feed pumps to fail to come up to design speed upon startup. This could result in a substantial loss of auxiliary f eedwater flow to the
, stean ;tencrators when such flow was required. This in : urn could caese significant renctor coolant sys:ca pressure /te=peraturc transients, and.
' - I-
-significant boiling in the reactor coolan: systen if substantial decay
' heat were presen: in the reactor core.
1:41s% rest) # W> *<; t e:* r's.w CO::;7;; p t.7t 7,C3 MA;'.::;rv;.vTretJf-TCt ( C3. 01::0. T '.? -
~
_g,.
(
The evaluation zad identification of this defect'vas provided to me on Septc=ber 30, 1977, and ves' discussed with Mr. T.11arpster of your of fice en Septe=ber 30, 1977.
There are two identical auxiliary feed pumps with the turbine governors, described above", installed in the Davis-Sesee Nuc1 car Power Station Unit '
No. 1.
The corrective actica taken was to.odify the governor including the recoval of "portiens of the pneu=stic speed-setting mechanism to assure that the governor will properly respond to speed dc=and signals. The pneumatic speed-setting =echanisa was never an integral part of the functioning of the governor, because the governor c= ployed servo =otor
~
~
control. This =odification was acccaplished at the Woodward Governor Company facilitics. Subsequent t'esting at these facilities has proved the proper functioni g of the sovernor.
The codifications were co=pleted.
priot to the current unit starcup. The governors have been tested for proper
- functioning c: auxiliary steam, and the surveillance cist will be completed during Eode 3 of the current startup.
Yours ved truly, n.
y?'
s 4 tA p towell E. Roe ~
Vice Presidcat Facilities Develop =c:t db b/9-10 bec:
P. M. S= art,'Esq.
C.
Charnof f, Is q.
D. II. Ilauser, Esq.
W. A. Johnson t
J. S. Grant I'.
E. C. Novak
- C. R. Doneck/
J. D.Lenardson J. C. Evcas
~
R.
R. o. senthal
. P. P. Anas A. H. Laza::-,.. -
e e
~2
59 -
_z_m13IT E
(
a E.
Historical Log of Station Ooerstions (September 24 - October 28)
Sept. 24 1Lsactor critical at 15% po'ver, generator on the line at 110-140 nu, performing controls tuning 1700 - Discovered steam leak on steam lead between No. 2 Turbine Control Valve and high pressure turbine 1830 - Turbine-generator taken off the line to repair steam leak. Reactor critical at about 9%
'powe r.
2135 - Received Stean and Feedwater Rupture Control System Actuation, resulting in Reactor Trip, and Safety Features Actuation 2345 - Plant stable at 1800 psig, Tay,+ 525 F Sept. 25 0415 - Started Plant Cooldova 0645 - Co=pleted initial survey of Containment.
Sept. 26 Cleanup and repairs begun Sept. 30 co=pleted repair and replacenent of mirror insulation on No. 2 Steam Generator x
Oct. 3 Auxiliary Teedpt. p Governors r.e=oved and sent' to Woodward Governor Factory Quench Tank Rupture Disc replaced' Oct. 5
. Vented Reactor Coolant Systen and run Reactor Coolant
?u=ps to get data to evaluate status of pt. ps and seals.
Oct. 6 Started Feedvater Cleanup in preparation for Reactor Coolant Syste= heatup.
1
~
Oct. 7 1830 - Received NRC approval to proceed with plant startup.
Oct. 8 1530 - Checkout of Auxiliary Feedpe=ps (using.
Auxiliary Stean) conpleted Oct. 11 Atte=pted to test pressuri:ar power relief valve.
Unsuccessful due to electrical circuit proble=s.
1
~
Oct. 12 Pressurizer power relief valve control circuit working, stroke'd valve and it stuck open again e
e e
e
.o e
c Oct. 13 Crosby can on site again working on Power Relief Yalve.
Increased RCS pressure to co=plete testing of Reactor Coolant Pu=ps.
Oct. 15 Completed repairs to Power Relief Valve and tested
.it successfully 0ct. 16 Co=pleted testing of Auxiliary Feedpu=ps.
Governors had been codified by Woodward to prevent sticking 1306 - Reactor critical 2316 - Rolled the turbine
~
Oct. 17 0307 - Generator synchronized 1135 - Generator off the line for overspeed tests
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1805 - Co=menced Reactor Shutdown for 'Unic Power Shutdown Test.
We are nov back in Power Escalation Sequet.ae Oct. 18 0200 - Coepleted Unit Power Shutdown Test 0300 - Co=menced Reactor Startup 1036 - Generator Synchronized e
Oct. 19 1124 - Generator off line to repair steam leak betwe6n No. 1 Control Valve end HP Tc:bine Oct. 20 2241 - Cenerator synchronized Oct. 23 1009 - A half trip of STRCS caused lov stea= generator level, resulting in full SFRCS trip, Reactor trip and STAS Oc t '. 27 2330 - Reactor Critical Oct. 28 1151 - Generator Synchronized s
e s,
m Reference 16 0FFICE OF NUCLEAR REAC' (Week ending Occcmber 29, 19/s)
Seabrook, Units 1 & 2 Amendment llo.1 to the Construction Permits which changed owners and ownership percentages was issued on December 27, 1978.
8 Zimmer 1 The issuance of the Zimmer Safety Evaluation Report will be delayed about 1 month to permit further review and subsequent rework of the report.
Trojan SLB on interim operation A Partial Initial Decision was issued by the of Trojan as related to the Control Building seismic non-conformance.
The decision, dated December 21, 1978, permits interim operation of Trojan pending modifications to the Control Building structure, with These certain conditions on operation during this interim period.
(1) no modifications which may weaken the shear walls may be made without prior NRC approval, (2) the plant must shutdown for are:
inspection for any earthquake exceeding 0.08g (design OBE is 0.15g),
and (3) some additional pipe supports must be added prior to reactor A
These conditions were recommended by the 00R staff.
operation.
conforming license amendment was issued by DDR on December 2?,,1978.
Davis Besse 1 Test data at Davis Besse Unit I has shean that because of the high capacity of the auxiliary feed pumps it is necessary to use operator control of steam generator level in the event of actuation of the Auxiliary Fe'edwater System during transients such as loss of offsite power to limit the cooldown rate and prevent the pressurizer level from decreasing below the level at which it could be monitored.
After the regional inspector questioned the licensee's determination that no unrev'iewed safety question was involved, 00R requested a copy of the licensee's evaluation and, subsequently, additional information.
22, 1978.
After The additional information was submitted on DecemberDDR concluded t review on December 23, 1978, would occur if the operator actions were not takco, no unreviewed safety question was involved and no licensing action was required.
The licensee agreed to submit a description of a design change and an implementation schedule w':Lhin 60 days which would avoid continued reliance on operator action to maintain indication of level during This matter is being pursued to determine whether transient events.
the problem exis.ts at other B&'d plants.
ENCLOSURE B
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