ML19322C294

Summary of 780610 Meeting W/Util at Rancho Seco to Discuss 780320 Cooldown Incident

ML19322C294
Person / Time
Site:	Rancho Seco, Crane
Issue date:	07/31/1978
From:	Lobel R Office of Nuclear Reactor Regulation
To:	Check P Office of Nuclear Reactor Regulation
References
TASK-TF, TASK-TMR NUDOCS 8001160817
Download: ML19322C294 (13)

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NUCLEAR RE UL T RY cOMT.'.lsSION NgQQg4 p,Q, jQ')$

WASHINGTON. D. C. 2055s y

Jul 312 I

MEMORANDUM FOR:

Paul S. Check, Chief, Reactor Safety 3 ranch, 00R FROM:

Richard Lobel, Reactor Safety Branch, 00R

SUMMARY

OFMEEJINGHELDATMNCNO[SN@NUCLEARPOWER PLANT ON JUNE }l0,1978 TO DISCUSS A RECENT C00LDOWN EVENT 4

On June 20, 1978 I attended a meeting at the Rancho Seco_ nuclear power plant to discuss an abnormal event which occurred od Mardh220,1978:

This event resulted in a cooldown of the reactor vessel iTsic~eis~'of't}iat allowed by the Rancho Seco Technical Specifications, specifically, Section 3.1.2.

An attendance list is given in Attachment 1.

An agenda is given in Attach-ment 2. is a disucssion of the cooldown event and a sunnary of the meeting. Attachment 4 is a SMUD internal report on the transient with proposed actions to prevent the reoccurrence of such an event. contains my conclusions reached from studying this'e-nt.

f'= m LM LMA Richard Lobel Reactor Safety Branch Division of Operating R.eactors

Enclosure:

As stated G. Zwetzigi x W S cc:

M. Chiramal

0. Tondi T. Marsh R. Woods P. Norian 9

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ATTACHMENT 1 LIST OF ATTENDEES MEETING ON RANCHO SECO C00LDOWN TRANSIENT SMUD_

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Ron Colombo

~

S. l. Anderson 1

Lloyd Stephenson John D. Dunn Norm Brock Bob Dieterich Ron Rodriguez Donald C. 81achly John V. McColligan Pierre Dubre NRC Phil Johnson-IE:V M. Chiramal John Anderson-Oak Ridge Nat'1. Lab Richard Lobel Dam Tondi gy czggg-~~-@

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t B&W Joel T. Janis 9

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RTTRCHMEHT 2 AGENDA.

March 1978 Reactor Transient as installed (1)

A brief description of the Integrated Control Syste:::

at Rancho Seco, presented by Nor::: Brock, Senior Instrument & Control-

, Engineer.

(2)

Ceneral descript' ion of sequence of events prior to, during and following cooldown and a sec:ary of the valid and acavalid instrument indications available to Operators during the transient, presented

, by, Don 31achly, Associatie Mechanical Engineer.

(3)

Review of electrical drawings pertinent to event, presented by John Dunn, Supervising Electrical Engineer.

(4)

Corrective action (icplemented and planned) including a discussion of adequacy, presented by John Dunn, Supervising Electrical Engineer.

(5)

Reasons for providing SFAS automatic start of auxiliary feedpur:ps and consequences if S?AS automatic start la deleted, presented by Stan Anderson, Associate Nuclear Engineer.

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ATTACHMENT 3 MEETING SUPPARY The meeting began with a description by Nonn Brock of SMUD of the Integrated tontrol System (I.CS) used at Rancho Seco.

The Integrated Control System coordinates the reactor, steam generator feedwater control and the turbine under all operating conditions.

The ICS maintains the required balance between the reactor, the steam generato;and_.,the._ turbine f.or coordinated control. operation under all conditions. (The;ICS isi..

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pa pa bliIsf Tfollisiif tWisa~d ~dE' and T6v~eFtnEl Md Ta~ngeif :r m

attached - Un'it 1.oad Demand, Integrated Master, Reactor Control, Steam Generator /Feedwater Control and Turbine Control.

There are four modes of operation of the ICS.

These are:

S ta tt.u p,,_l o,w (Joad, control _(5% to,15",), ful) Icad, range _(15Lto 1,00Qan

.o pe ra ti oni( r u n_ba c k;o ril i mi t foa d Lo ni c e r ta i ni ma l fu nc t i_o_ng ).

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In.,the,ful1, load frange_the_ basi _c r equirement of the ICS is to 7natch;the, 59enerated&wi.thitheXdemand.

The ICS does this by coordinating the flow of steam to the turbine and the rate of steam production.

The flow of steam is controlled by the turbine throttle valves.

The rat ~e of steam generation is controlled by varying the total amount of feedwater and reactor power and also maintaining a proper rati.o between feedwater flow The feedwater and reactor power so that the proper steam conditions exist.

flow is controlled by the feedwater valves and pumps and the reactor power is controlled by the control rods.

The primary flow in the steam generator is constant for all load points'so that temperature of the reactor outlet fluid (T ) is a measure of the Btu's available.

The react'or inlet temperature

.h (Tc) compared against the outlet temperature would produce an index of the With a fixed primary flow amount of energy transferred to the feedwater.

for any load, Tc is affected by the feedwater flow.

If more and a given Th feedwater flows throagh the steam generator at any given load the Tc will tend to be lower. The values of Th and.Tc vary with load 'and the average of these two temperatures (Tave) is used as an index of the balance between feedwater flow and the heat available in the primary fluid or r.eactor power.

Tave is controlled at a constant value from 15% to 100% reactor ower.

There are a number of variables that determine the amount of ene. ;y available from the steam generator.

These are:

reactor coolant flow, reactor outlet temperature, feedwater temperature, steam gene,rator pressure j

and feedwater flow.

If an attempt is made to remove more energy from the steam generator than there is available, a reduction in final steam temperature i

would occur.

To prevent this from happening the demand for each loop feed-water flow is limited by the conditions in that loop.

The effect of the 3

l limiting values are expressed by:

e d

Stu limit = (TL + Fwt, + Psg - 200) Rcf.

where Ti, = Reactor outlet temperature limit

.~at = Feedwa+ar tenperature limit Psg -

..erator pressure limit J

R Rcf = Reactor Coolant Flow limit The feedwater control subsystem (See Figure 2) is designed to maintain a total feedwater flow equal to the feedwater flow demand.

Feedwater flow i

is con' trolled through the use of feedwater control valves while the pressure difference across these valves is maintained constant through the use of the turbine driven feed pumps.

The pressure drop across the feedwater valves is monitored and compared to an adjustable set point and i

the error signal controls the pumps.

In order to anticipate the pressure difference change due to a load change a feed forward signal frcm feed-water flow demand is added to the pressure difference signal.

The feedwater flow to each steam generator is compared to the loop feed-B wa ter demand and the difference, feedwater error, establ.ishes the position i

of the feedwater valves.

In order to provide a uniform sensitivity over the full range of flow control the two feedwater valves (main a,nd start-up)

H are used.

To prevent the feedwater flow control from reducing the level in the steam generator below a predetermined minimum, a low range level meter output is compared to an adjustable set point and the difference is auctioneered against the feedwater flow error.

If the feedwater flow error tries to reduce feedwater to a point where the level will drop b'elow its minimum value, the level error will take control of the feedwater valves. A similar control is imposed a t high loads where the level can rise to a j

maximum value.

iIf mainjeedwater'puinps' fail _ the inir imum steaml~ generator 1eviliTs~ maihtai6ed (by; the emergencyj;feedwaterypumpf through _the: emergency feedwater ~ valve, ~ ' ""

The sequence of events during the cooldown transient was next described by Don

'~

Blachly of SMUO.

The reactor was initially operating at 72% of rated power at a pressure of 2155 psi.

The plant was generating 680 Mw(e).

The average reactpr coolant f s

temperature was 582 F and the inlet (cold leg) temperature was 520)F.

The electromatic relief valve was out of service.

been leaking, it was gagged so that it could not open).

One HPI pump was in opera tien as a make-up pump which is the normal configuration at Rancho Seco.

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. The plant has two hot legs and four cold legs with two once-through steam generators (OTSG's).

The ICS monitors steam pressure from either loop A or loop B.

The selection is made manually.

Before and during this transient, loop B was selected.

The pressurizer is attached to the B loop.

At 4:25 a.m. an operator attempted to change a light bulb in a control panel causing a short to ground which caused the loss of a majority of the indication and control parameters.

Upon loss of. one of the two power supplies (' designated NNI-Y) the hot This made the term in the first leg temperature indication went to zero.

parentheses of the Stu limit equation less than zero and therefore set the feedwater demand equal to zero.

This resulted in essentially a tennination of main feedwater flow to both steam generators by running both feedwater pumps down to minimum speed and shutting the main and startup feedwater valves.

Upon loss of main feedwater flow (as measured b'y feedwater pump However, due di,scharge pressure) the auxilliary feedwater pumps started.

to loss of function of the ICS, the auxilliary feedwater valves did not This was because the steam generator water level indications to the open.

ICS were also rendered incorrect.

The indicated water level in the A steam generator began drifting downward while the indicated water level in the B steam generator began drifting upward.

The auxilliary feedwater valves did not open until the indicated water level in the A steam generator reached the low level set;aint (20 inches).

Thus, there was no main or auxilliary feedwater flow to either steam generator at this time.

This loss of feedwater flow caused the reactor coolant tpaperature to increase, causing an insurge to the pressurizer which caused the reactor coolant system pressure to rise sharply.

One of the two pressure safety valves opened-below its setpoint of 2500 psi p oviding some pressure relief.

At 2355 psi the reactor was tripped by a The peak pressure reached was 2425 psi.

It is assumed high pressure trip.*

that the safety valve was alternating open-to-closed-to-open (simmering)

'during the remainder of the transient.

Shortly after trip, the reactor operator initiated flow from a second HPI pump (the makeup pump was still running). This pump uses the Sorated Water Storage Tank as a water source.

As

'The backup trip would have been high hot leg temperature at 619 F.

seen from Figure 3 the hot leg temperature was never above apprcximately, 602 F.

I F

, The reactor coolant system pressure then dropped due to opening of the main steam safety valves in response to increased pressure in the steam This provided some cooling to the generators due to loss of feedwater.

The pressure then remained fairly constant for approximately primary side.

7 minutes.

Both steam generators dried out at approximately 1 minute after initiation of th_ e event.

Tile steam generator level signals are shown in Figure 3.

During this time the reactor operator was trying to detennine what instru-mentation was available to provide him with reliable indications of the plant's condition.

He chose pressurizer level and pressurizer pressure (at Rancho Se'co there are no valves.between the core and the pressurizer so -the pressure transducers are mounted on the pressurizer).

At approxi-2 mately 7 minutes, the water level indication in the A ~ steam generator fell below the icw level set point (See Figure 4).

This caused the ICS to ser.d a demand signal to the main feedwater pump control valves to provide water to the A steam generator.

These valves opened.

However; the main feedwater pump was at low flow due to the false hot leg temperature indication and could not respond to the new demand (since the hot leg signal is given priority in tne ICS over the level signal).

The ICS then called upon the aufilliary feedwater valves to supply water to the A steam generator.

The auxilliary feedwater pumps were already operating at this time due to the loss of main feedwater flow signal.

Thus, flow from the auxilliary feed-water pumps was sent to the A steam generator.

The reactor operator observed that there was a 100", demand from the. A steam generator but 0". demand from the B steam generator.

He therefore removed the trip signal from the main feedwater pumps, leaving the main feedwater control valves in automatic.

Operation of the main feedwater pumps caused the closure of the auxilliary feedwater valves.

The reactor operator was then manually controlling the main feedwater flow to the A steam generator.

Due to the cooldown from the feedwater flow to the A steam generator, the reactor pressure fell to the set pobit for actuation of t'he Safety Features Actuation System (SFAS).

SFAS consists of two HpI pumps and the auxilliary feedwater pumps.

The two HPI pumps and the auxilliary feedwater pumps were now delivering 100". flow (the third HPI pump was still delivering makeup flow).

The water source for the HPI pumps was the Borated Water Storage Tank and the water source for the auxilliary feedwater pumps was the Condensate Storage Tank.

Soon after SFAS initiation, the operator took over manual coeration of the H?I flow and reduced it from the full flow called for by SFAS.

The steam driven auxilliary feedwater pump continued operation, su? plying water to

-toth steam generators after SFAS actuation.

The notor driven auxilliary j

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feedwater pump was recoved (to shed electrical lead upon SFAS initiation).

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-S-r The reactor coolant system pressure and temperature were decreasing at 3

this point.

The pressure reached a mininum of 1490 psi, After approximately one hour, the NNI-Y power supply was restored and the operator became aware that the reactor coolant temperature was only 3S F

and he was therefore in violation of Section 3.1.2 of the Technical Specifications.

[,

The operator then took the following actions to bring the reactor into compliance with the Technical Specifications:

(1) spraying the pressurizer to reduce the reactor coolant system pressure, (2) keeping three reactor coolant pumps operating (operating procedure calls for only three pumps in operation when the Q

coolant temperature is less than 520*F).

(3) shutting off the auxilliary feedwater heaters,.and (4) draining the steam generators.

]

.g Aft'er this description of the transient John Dunn, Supervising Electrical 4

Engineer, presented a review of the electrical drawings pertinent to the y

event.

I was not involved in this discussion.

g After a lunch break John Dunn presented an oral s'mmary of corrective VJ u

actions, both implemented and planned.

These actions are disctissed in an

.s internal SUMD document which is Attachemnt 4 of this meeting report, i

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1 Il Among the actions proposed by SMUD is the preparation of a precedure which M

tells the operator what instrumentation he can rely upon if he loses the Non-Nuclear Instrumentation power supplies.

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Stan. Anderson.of SbjuD gave5 reasons, for providing,automa!;ic start of:

t-auxilliary feedwater pumpsialongzwith the HPIf actuation safety. signal.

There are three aEcidents in which credit is taken for the auxilliary feed-T water actuation.

These are the Steam Line Break, and the large Break and W'

fSmall B'reak LOCAf.

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y SMUD is asking Babc'ock and Wilcox (B and W) to evaluate the necessity for I

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, automatic actuation of the..au.xillia.ry_ feedwater after HPI actuat on.. ?The.

1 effect on,the:Large -Break and Small Break LOCAs ~is ~ thought to be: small.

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'The' B&W evaluation is expecte'd within'two weeks ~of 'the medtini hate.

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Q l NRC raised the cuestion of preservation of shutdown margin during the cooldo, a

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transient.

SMUD, presented calculations which showed that a li shutdown Ii margin was maintained during the transient.

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h ihe final item on thef agenda was a discussion of..o.ther :possible~mecha'nisms, n,;for causing.la severe cooldown transient..Depressurizationidue to f a faultyi, electromatic: relief val.ve orf.safeti. valve was the onlyL possib.ilityJdiscussed.

- 3 The meeting adjourned at approximately 4:30 p.m. after which the NRC representatives were given a tour of the control room to get a better idea 7",

of the location of the equipment available to the operator during the tra.ns i e n t.

This included the RPS panels which were in a separate room adjacent to the control room.

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