ML18219D963

Letter Implementation Option Selected for Installation of Acceptable Long-Term Improvements to Mitigate Over-Pressurization Transients

ML18219D963
Person / Time
Site:	Cook
Issue date:	03/03/1977
From:	Maloney G Indiana Michigan Power Co, (Formerly Indiana & Michigan Power Co)
To:	Rusche B Office of Nuclear Reactor Regulation
References
Download: ML18219D963 (51)

Text

NRC FORM 196 I2.78)

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U.S. NUCLEAR REGULATORYCI" MISSION cW NBC DISTRIBUTION FoR PART 50 DOCKET MATERIAL DOCKET NUMBER

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FILE NUMBER TO: Mr B C Rusche LETTER gIoRIGINAL Dcopv gNOTORIZED 2 UN C LASS I F I E D PROP INPUT FORM FROM:

Indiana Bc Michigan Pwr Co New York,'NY.

G P Maloney DATE QF $0c3U PjvT DATE RECEIVED 3 Q 77 NUMBER OF COPIES RECEIVED one signed DESCRIPTION Ltr notarized 3-3-77....re our 1-10-77 ltr

....intrans the fallowing:

ENCLOSURE Info concerning implementation option re ins ta llation of acceptable long-term 'improvements t mitigate overpressuriiation transients........

2p REACTOR VESSEL OVERPRESSURIZATION DISTRIBUTION PER G, EECH 10-21-76 PLANT NAME: Cook 81.

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INDIANA II MICHIGAN POWER COMPANY P. O. BOX 18 BO WL IN G G R E EN STAT ION NEW YORK, N. Y. 10004 Donald C.

Cook Nuclear Plant Unit No.

1 Docket No. 50-315 DPR No. 58 March 3, 1977 Mr. Benard C. Rusche, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.

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Dear Mr. Rusche:

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i/i This letter is in response to Mr. Dennis L. Ziemann's letter dated January 10, 1977 which requested that we identify the implementation option we have selected for the installation of acceptable long-term improvements to mitigate overpressurization transients.

There were also staff positions and requests for additional information listed in enclosure 1 to Mr. Ziemann's letter.

Our responses to the staff positions and requests for information are provided in Attachment l.

Attachment 2 presents two sketches which represent the control function - logic diagram for the circuits to be installed at the Cook Nuclear Plant to mitigate the effects of an overpressurization transient.

This control scheme will be duplicated on both existing RCS wide range pressure transmitters with each train actuating one Pressurizer Power Oper ated Relief Valve.

This control scheme will be the long-term solution for RCS overpressure protection.

The selection and ordering of hardware (e.g.-

Block-Unblock switch) is currently underway.

The estimated delivery for the hardware is late September 1977.

The remaining equipment that is needed is already available or can be fabricated on site.

The mitigating system we have designed incorporates controls for two (2) of our existing Pressurizer Power Operated Relief Valves and the control scheme of Attachment 2.

The mitigating system will represent good engineering practice and will not adversely affect plant safety or introduce potential comon mode failures that could both cause a pressure transie isable the protection system.

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March 3, 1977 We will have all the necessary equipment available and be pre-pared to do the installation work by December 31, 1977.

We will install the equipment at the first system depressurization subsequent to having all of the equipment available on site.

We believe this to be the most logical approach since there is no need for the mitigating system until the reactor coolant system is depressurized from operating conditions.

Your August 13, 1976 letter required us to provide "an analysis of the RCS response to pressure transients."

We expect that the analysis, when completed by Westinghouse, will show that our existing relief capacity is adequate to mitigate all the anticipated pressure transients.

We will advise you as to the results of the analysis when they become available.

Very truly yours, GPM:mam Attachment

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Mal ne Vice Pr ide t Sworn and subscribed to before me on this G'~day of March 1977 in New York County, New York Notary Pu lac KATHIiE BARRY QOTARY I UBl.i, State ol How York No. 41-4604'i9Z Qua!iliad in Queens County g@ificate lifed in rabbi York County'prnn>>ssion Expires March 30, ]flub'7 cc:

G. Charnoff R. J. Vollen R.

C, Callen P.

W. Steketee R. Walsh R, S, Hunter R.

W. Jurgensen

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Attachment 1

Res onse to Xtem 1:

Discussion of PWR overpressurization transients will be conducted as a part of the Licensed Operator Requalification Program.

The program is conducted on a one shift per week basis.

The required personnel attendance is normally all RO and SRO licensed operators on shift.

For the purpose of overpressurization discussions, attendance will also be required of-shift. auxiliary equipment operators and CSX Technicians.

The requalification

program, year three, will start in March, 1977.

a)

Xt is estimated that all plant operators will have participated in formal discussions of overpressure transients by May 15, 1977.

b)

A formal discussion of overpressure transients at PWR's will be conducted as a part of the licensed operator requalification program.

The discussions will cover in detail the following subjects:

1.

The causes of past pressure transients that, have occurred at other PWR facilities.

2.

Plant conditions at the time of the pressure transients.

3.

The mitigating action that could have been or was taken.

4.

The preventive measures that could have been taken to avoid the event.

5.

The steps that have been taken to prevent similar occurrences.

6.

Cook Nuclear Plant similarities and differences with other PWR facilities and how these relate to over-pressure events which occurred at these other facilities.

c)

We have been able to collect details of eleven overpressure events at PWR facilities.

These have been analyzed as to their possibility of occurrence at the Cook Nucleax Plant and the results are summarized in the following table.

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Table of Over ressure Events (Item 1(c)

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Donald C. Cook Nuclear Plant Plant Name Cause of Event Event Nature Mater-Solid 0 eration*

Steam-Bubble 0 eration**

Indian Point 3 Component Failure RHR-Letdown Isolation Credible Not-Credible Indian Point 2

(2 events)

Component Failure Loss of Instrument Air Not-Credible (Redundant Dryer Trains)

Not-Credible St. Lucie 1 Defective Procedure RCP Start Credible Not-Credible D. C. Cook 1 Defective Procedure Letdown Isolation Credible Not-Credible Beaver Valley 1 76-13/3L Personnel Error SIS plus Insufficient Information to Letdown Isolation draw a conclusion.***

Point Beach 2

Defective Procedure Letdown Isolation Credible Deliberate Operator Action Not-Credible Beaver Valley 1 76-11/3L Personnel Error Letdown Isolation Insufficient Information Electronic Instr.

to draw a conclusion.**+

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• This column reflects the Cook Nuclear Plant prior to the revision of the three major operating procedures described in Item 2 response.

• • This column reflects the Cook Nuclear Plant with its current operating procedures which provide for a steam cushion in the pressurizer.

• • • The description of the event in the available reports. did not contain enough information to be able to determine the event credibility at the Cook Nuclear plant.

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Donald. C. Cook Nuclear Plant Plant Name Cause of Event Event Natu1e=-

Water-Solid J!%"

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'yTest Insufficient Information to draw a conalusion.***

Zion l Operator Error

~Letdown Isolation Credible Not-Credible TurkeyyPoint 3

Insufficient Information RHR-Letdown Isolation Credible Not-Credible Indian Point 2

Defective Procedure RCP Start Credible Not-Credible

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d)

To reduce the likelihood of occurrence of the events which are listed as credible in the preceeding

table, we have revised three major operating procedures to include non-water-solid operation of the reactor coolant system (RCS) and have installed an "operator alert" alarm.

The description of the details of the procedural changes and the alarm function are provided in the response to Items 2 and 5 below.

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3 Res onse to Xtem 2:

Three major operating procedures in which the plant would experience a water solid condition have been re-written.

These procedures cover plant heatup,

cooldown, and. reactor coolant system filland vent.

As a result of the revisions to these operating procedures described below, it is not anticipated that the Cook Nuclear Plant willbe operated or maintained in a water solid condition.

Therefore the possibility of an overpressurization during these plant conditions has been greatly reduced.

a)

Revised Plant 0 eratin Procedures:

1.

Fillin and Ventin Reactor Coolant S stem-OHP 4021.002.001 Prior to revising this procedure the pressurizer (PZR) and coolant loops were filled to a condition in which four air pockets remained in the steam generator tubes.

The RCS pressure was then increased to about 400 psig using the charging pumps.

awhile maintaining this pressure with the charging pumps and letdown through RHR, each reactor coolant pump (RCP) was bumped-and the RCS was vented.

After this venting the RCS was in a water solid condition using charging and letdown to control pressure and RHR system to control temper-ature.

After the first vent the reactor coolant pumps were run again for 1 minute and the system vented.

This cycle was repeated two (2) more times for 5 and 10 minute runs on the reactor coolant pumps.

After completing the ten minute run/vent cycle the fill and vent procedure was. considered complete and the plant remained in this condition until plant heatup, at which time a pressurizer bubble was drawn during heatup.

The revised, procedure fills the RCS using the charging system to the condition where four air pockets remain in the steam generator tubes.

At this point instead of pressurizing the system, bumping reactor coolant pumps and venting to a water solid condition, the pressurizer heaters are energized and a steam bubble is formed in the pressurizer.

Pressurizer temperature is then increased to bring system pressure up to about, 325 psig for reactor coolant pump operation.

The

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pressurizer steam space is put on sample recirculation to the volume control tank to remove air and M2 from pressurizer.

The reactor coolant pump run/vent cycle is now performed using steam bubble for pressure control.

After the last RCP is run, the reactor coolant system pressure is reduced to 50-100 psig for final vent, to remove more N2 from solution at lower pressure.

After venting at 100 psig the fill and vent. procedure is complete and the system will remain in this condition until heatup.

Using this revised filland vent method, which makes use of the air pockets in the four steam generator

tubes, the RCS will never be in a water solid condition.

Pre-vious experience has shown that this procedure pro-vides a "soft" system while drawing the steam bubble.

Plant Heatu from Cold Shutdown to Hot Standb

-OHP 4021.001".001 This procedure was revised so that initial conditions for the heatup are that the RCS pressure is being main-tained by a pressurizer steam bubble.

The steam bubble has already been established during the filland vent, therefore, the heatup procedure is not involved with water solid operation under any conditions.

Plant Cooldown from Hot Standb to Cold Shutdown-OHP 4021.001.004 Prior to revising the cooldown procedure the RCS was placed in a water solid condition during the last phase of plant cooldown.

The first phase of plant cooldown remains the same, since water solid condition is not a concern during this part of the cooldown.

The RCS temperature is reduced to about 150 F in the reactor vessel and loops, and 425'F in the pressurizer (which maintains system pressure at about 325 psig), by dumping steam to the condenser and placing RHR system in service at 425 psig/350'F.

During the first phase of cooldown the following pertinent steps are taken at. the indicated system conditions.

i)

System Pressure 4 1000 psig:

Accumulator isolation valves IMO-ll0, 120,

130, 140 are closed and breakers are racked out and tagged.

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System temperature g 350'F:

SI pump breakers are xacked out and tagged.

Note:

A revision is now being made to add the following valves to condition (ii), (system temp.

4 3504 F).

Close the following valves, rack out and tag supply breakers for each:

SI Pump discharge Valves ICN-260 Im-265 Cold Leg SI Valves IMO-51 IMO-52 X@0-53 IMO-54 The steps listed above will completely disable the SIS system.

A major revision was made to the last. phase of the cooldown procedure to cool the pressurizer down and depressurize the system using Auxiliary spray with a maximum level in pressurizer of 85%, instead of filling the pressurizer to water solid condition for cooldown.

The revised procedure completes the plant cooldown as follows:

Plant Conditions:

Loop and vessel temperature 1504F Pressurizer temperature 420'F Pressurizer pressure 325 psig 1 reactor coolant pump running, RHR in service.

First the running RCP is secured and the PZR heaters are de-energized.

Using Auxiliary spray the PZR level is slowly increased, cooling the PZR and depressurizing

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The PZR level is allowed to increase to 85% maximum level and recirculated using Auxiliary and letdown through RHR until a steam space temper-ature of 250 F to 275 F is reached.

When this condition is established, the PZR level is reduced by increasing letdown through RHR system and the charging system is secured.

Draining of the RCS to necessary level for maintenance/and/or refueling can be staxted at this point and the cooldown is considered complete.

For all other conditions which do not require opening of the RCS a low pressure steam bubble will be maintained during the cold shutdown condition.

b)

With the revision of the above three procedures, it is not anticipated that the RCS will be operated or main-tained in a water solid condition at any time during further operations.

c)

Xt is not anticipated that the need will exist, to operate the RCS in a water solid condition.

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Res onse to Item 3:

The safety injection system (SIS) pumps willbe de-energized during cold shutdown operation as required.

However, oux current Technical Specifications

(3.5.2 S 3.5.3) require that surveillance testing of the SXS be performed during cold shutdown and hence the SXS pumps must be operated during this mode.

We will not be able to de-energize the SXS pumps during this testing period but they will be de-energized at all other times during cold shutdown.

a)

Figures 1 and, 2 (attached) show the SXS flow paths into the RCS.

b)

Figures 3 and 4 show the head-flow characteristics of both (North 6 South)

SIS Pumps.

c)

The valves and pumps to be closed and de-energized are shown on Figure 1 by being circled with a dotted line.

Xtems are identified as North and South

("N" and "S")

pumps, valves V-9 (XCM-265) 6 V-10 (ICM-260) and valves XMO-51, IM0-52, IMO-S3, IMO-54.

d)

The present revision of the plant cooldown procedure requires the accumulator valves to be closed and the breakers for each to be racked out and tagged.

The SXS pumps are also de-energized and the supply breakers racked out. and tagged.

A temporary sheet has added the following valves to the cooldown procedure:

Cold leg injection valves IMO-51, 52, 53, and 54 SI pump discharge valves ICM-265 and 260 (V-9 &

V-10 xespectively)

The change requires these valves to be closed and the breakexs for each xacked out and tagged.

e)

All equipment identified in 3(c) will remain closed and/or de-energized during cold shutdown operation, except when technical specification surveillance testing is being performed.

It is felt that by..closing and de-energizing

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the cold leg injection valves IMO-51, 52, 53, 54 that SIS injection would be prevented in the event of an inadvertent operation of the SIS during cold shutdown condition.

f)

Safety Injection Pump "S" is supplied from circuit breaker TllA1 located in the Bus A and Bus B 4kV switchgear room.

Safety injection pump "8" is supplied from circuit breaker TllD5 located in the Bus C and Bus D 4kV switchgear room.

Both switchgear rooms are on elevation 609'.

These circuit breakers are normally operated from the control room.

Disconnecting these circuit breakers, which makes them inoperable, is done at the 4 kV switchgear rooms.

The north safety injection pump shutoff valve, ICM-260, is supplied from MCC~1-ABV-D and the south safety injection pump shutoff valve, ICM-265, is supplied from MCC-1-ABV-A.

Both motor control centers (MCC) are located on elevation 587'.

The valve motor operators are normally controlled from the control room.

Opening the circuit breakers to the valve controller, which will disable the valve operator, is done at the motor control

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center.

g)

Disconnecting the circuit breakers for the safety injection pump motors will turn off both the red and green breaker position indicating lights and the white breaker closing status light.

The status monitor lights are still oper-able and will indicate a failure of the safety injection pump motors to start if a safety injection is initiated.

The red and green valve position indicating lights will both go out on both valves.

The status monitoring lights will indicate the valves closed if a safety injection is initiated.

h)

The plant cooldown procedure will provide the administrative control necessary to insure that each designated valve and pump is de-energized and/or closed at the correct time during cooldown.

When the equipment. is placed in the indicated position the step in the procedure willbe signed off.

The equipment will also be tagged out. in

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accordance with the plant tagging procedure, which pro-vides another administrative control.

When the equipment is tagged, it can not be energized or its position changed without first obtaining authorization from the shift operating engineer.

When the plant is returned to power, proper equipment alignmerit and operability is administratively controlled by the technical specification surveillance signoff sheets.

These data sheets are attached to the plant heatup and startup procedures and require that all equipment associated with a mode change are OPERABLE prior to that mode change.

All equipment listed above willbe checked and verified OPERABLE in this manner.

Each shift operating engineer is responsible for plant operation and control of plant equipment on his shift.

Xt is the shift operating engineer's responsibility to insure that designated equipment is closed, de-energized and tagged out in accoxdance with plant procedures.

i)

The accumulator isolation valves willbe closed when the RCS pressure is at 1000 psig or less.

The supply bxeakers for the motor operated accumulator isolation valves are located on Motor Control Centers 1-EZC-A,B,C, and D.

These MCC's axe located in the electrical bay mezzanine.

The switches for the isolation valves are located on the control room panel SXS.,

The supply breakers are opened and closed locally at the MCC's.

Disabling the valve motor operators will cause the red and green valve position lights to go out.

The annunciator "accumulator valve not fully open" willbe opexable when operating above Pll interlock and the status monitor lamps willbe operable.

j)

The overall impact on plant operation, to lower accumulator nitrogen pressure for each cooldown, would be the possible increase in time to either the cooldown or heatup cycle due to depressurization or pressurization of accumulators.

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Also the increased usage and availability of N2 could effect plant operation.

There is also a remote chance that the N2 would have to be vented to the gas system due to high gaseous activity which would overload the waste gas system.

We believe that closing the isolation valves, de-energizing and tagging the breakers supply power to the motor operated valves is sufficient protection.

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11 Res onse to Item 4:

a)

The charging system and the safety injection system are routinely tested while in cold shutdown.

b)

For extra measures taken see the response to Item 3.

Res onse to Item 5:

A high pressure alarm ("operator alert,") used during low RCS temperature operation is presently installed at the Cook Nuclear Plant.

a)

The "operatohn,alert" alarm was installed prior to the December 23, 1976 refueling outage.

b)

The system modifications are shown in Figure 5 (attached).

c)

The alarm setpoint, mode of annunciation and sensor are shown in Figure 5.

d)

Et is not anticipated that the Cook Nuclear plant willbe operated in a water-solid condition.

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12 Res onse to Item 6:

a)

RHR Design pressure is 600 psig.

b)

Figure 2 identifies the RHR isolation valves relevant to overpressure incidents by means of two hexagons.

The valves are Copes-Volcan, double disk, motor operated gate valves.

c)

RHR Xnterlocks, setpoints, alarms:

RHR pump suction valves:

IMO-128 and ICM-129.

IMO-128 is interlocked with RCS wide range pressure channel NPS-122.

When pressure drops below 425 psi, a permissive signal allows the operator to~anually open this valve. If the pressure increases above 600 psi, this valve is closed automatically.

An RCS high pressure alarm in the control room is energized whenever pressure is above 600 psi, as sensed on either wide range pressure

channel, NPS-121 or NPS-122.

An RCS low pressure alarm is energized whenever both pressure channels are below 425 psi.

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ICM-129 is interlocked with RCS wide range pressure channel NPS-121.

Operation and setpoints are as above.

d)

The nominal stroke time for these valves is 35 seconds.

e)

The setpoint and capacity of the relevant RHR safety valve is 450 psig and 900 gpm, respectively.

This safety valve is designated on Figure 2 by an asterisk(*).

f)

The other alarms and setpoints associated with RHR not discussed in (c) above are as follows:

Pressure alarms,"';

XPA-310, XPA-320 RHR pump discharge pressure high alarms annunciate in the control room above 590 psi.

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We feel that adequate provisions have beenpprovided by means of the revised procedures described in the response to Item 2.

a)

Using our revised procedures for RCS filland vent, plant cooldown and heatup, the RCS will no longer be in a water solid condition during any of these operations.

During RCS filland vent, air bubbles are maintained in the steam generators while a pressurizer steam bubble is established.

The reactor coolant pumps (RCP's) are not operated prior to PZR steam bubble formation.

System pressure is increased for RCP operation by increasing PZR temperature to about 440 F.

Pump run and vent cycles are conducted at these conditions.

With this procedure the RCP's are not operated in a water solid condition.

The RCS is not placed in a water solid condition at anytime during cooldown.

The PZR is cooled down and RCS depressurized using Auxiliary spray and slowly filling the PZR to a maximum level of 85% and recir-culating until conditions are met for draining the PZR/RCS.

Either air or N2 blanket is placed on system at start of drain down. If the RCS is not to be opened during a unit shutdown, a low pressure (i.e. 50-75 psig) bubble willbe maintained in the PZR during shutdown precluding a water solid condition.

Using these revised procedures, RCP operation during water solid conditions is no longer a question.

b)

A PZR steam bubble is established for all conditions listed in Item 7(a) prior to the start of the first reactor coolant pump.

c)

Listed below are the limits associated with system temperature before the first RCP can be started.

This question is not applicable to the Cook Plant due to the response of (b) above.

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RCS below 150 F without seal water or with RCP seal leakage rate at 5 gpm or less and about 35 gpm of component cooling water at inlet temperature of about 80'F is flowing through thermal barrier.

2.

A pressure differential of at least 275 psi is available across the Nl controlled leakage seal.

This is met if reactor coolant pressure is above 375 psig.,

and temperature of 350'F or less.

d)

RCS temperature profile instruments:

1.

Individual hot and cold leg reactor coolant temperature are recorded over the range of 0-7004F.

2.

RHR pump discharge temperature is recorded over a range of 0-400'F.

e) with steam bubble operation it is not. necessary to establish isothermal conditions in the RCS for RCP operation.

f) pressure spikes are not a problem on RCP starts when a PZR steam bubble has been established.

Several uses of the revised procedures as described in the response to Item 2 have shown this to be the case.

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