ML19259C176

Forwards Response to IE Bulletin 79-08.Reviews Procedures for Dealing w/TMI-2 Incident Circumstances

ML19259C176
Person / Time
Site:	Big Rock Point File:Consumers Energy icon.png
Issue date:	05/04/1979
From:	Bixel D CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:	James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
References
IEB-79-08, IEB-79-8, NUDOCS 7906140246
Download: ML19259C176 (13)

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..o May 4, 1979 Mr James G Keppler Office of Inspection and Enforcement Region II' US Nuclear Regulatory Co==ission T99 Roccevelt Road Glen Ellyn, IL 60137 DOC:G2 50-155 - LICE'iSE DPR BI3 ROCK FOINT PL*drI - RESFCIISE TO IE EULLETI:179-03:

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Consumers Power Company's response to the subject Eu.lletin is fervarded as an attachment to this letter.

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xert Lavid F. Sixel

'uclear Licensing Ad inistrater CC Director, Office of Iiuclear Peactor Regulation Director, Office of Inspection and Enforcement t

1979 2278 ;07 7906140 ~246:4 !

RESPONSE TO IE BULLETIH 79-08 Question:

1.

Review the description of circumstances described in Enclosure 1 of IE Bulletin 79-05 and the preliminary chronology of the TMI-2 3/28/79 accident included in Enclosure 1 to IE Bulletin 79-05A.

a.

This review should be directed toward understanding:

(1) the extreme seriousness and consequences of the simultaneous blocking of both trains of a safety syste= at the Three Mile Island Unit 2 plant and other actions taken during t,n early phases of the accident; (2) the apparent operational errors which led to the eventual core damage; and (3) the necessity to syste=atically analyze plant conditions and parameters and take appropriate corrective action.

b.

Operational personnel should be instructed to (1) not override automatic action of engineered safety features unless continued operation of engineered safety features vill result in unsafe plant conditions (see Section 5a of this bulletin); and (2) not make operational decisions based solely on a single plant part=eter inlication when one or = ore confirmatory indications are available, c.

All licensed operators and plant management and supervisors with operational responsibilities shall participate in this review and such participation shall be docu=ented in plant records.

Restense:

This review will be conducted and documented in training sessions and vill ce completed prior to plant start-up.

Question:

2.

Review the containment isolation initiation design and procedures, and i=ple=ent all changes necessary to initiate containment isolation, whether manual or auto =atic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon autc=atic initiation of safety inj ection.

Resronse:

The Big Rock Point Plant design provides contain=ent isolation (excluding e=ergency core cooling, post-incident spray and makeup syste=s). The isola-tion occurs upon reactor vessel low-vater level or containment building high pressure. The low reactor water level set point used for this purpose is the same as that used to initiate e=ergency core cooling. Isolation capability is also provided =anually via a control console operated hand switch. The isolation valves vill re=ain closed and cannot be opened as long as either the low reactor water level or containment building high-pressure signals remain in effect.

If both sig als are absent, pneu=atically operated valves can return to the open position; the main stea= isolation 2278 308

2 valve which is d-c motor operated requires cperator action to open. If isolation is manually initiated via the hand switch, the valves remain in a closed position. Contain=ent isolation capability is periodically tested in accordance with the Plant Technical Specifications to assure continued operability.

Lines containing isolation valves are divided into four categories. See Drawing No M-539 (attached):

Type A - Lines which are or may be open to the interior mf the contain-ment shell have two valves in series, at least one of which closes automatically to prevent outward flow.

Except for check valves, both valves can be closed by manual initiation from either tne control roo= or another place that would be tenable after an accident.

(Items 1, 2, 3, h, 5, 6, 7, 8, 13, 14, 15 and 17)

Type B - Lines which are open to the reactor or any portion of the reactor recirculating loop are treated in the canner described in "A" above, with the added require =ent that the two valves are on opposite sides of the containment shell.

(Items 9, 10,11,12 and 16)

Type D - Lines that are ncrmally clo==1 have only a single valve. A lock, interlock, or opers/.ng prourdures and/or checklists protect this valve fro = being opened during reactor operation or conditions other that cold shutdown (Ite 30).

Type E - Ceri; n lines enter and leave the containment building without any openings to the con *ain=ent interior. Others leave and return without any open;ngs to the atmosphere (instrument tsps,etc). Such lines do not require isolation valves, pro-vided the lines are not in danger of being broken as a result of reactor system rupture. These lines are routinely checked to insure leak tightner s (Items 18 t hrough 29).

Normally, open lines which carry flui.ds out of the contain=ent building are closed automatically on the Isolatio 2 Signal, and/ cr power failure or upon manual trip from the control console hand switch.

dor = ally, open lines which carry fiv".ds into the containment building are each equipped with a check valve to preveat backflow upon loss of invard propellant force. In addition, operating perst nnel can secure these lines by manuall,,

operated gate valves or by air-operated control valves. The latter close on air or power failure, with exception of the supply line to the control rod drive hydraulic system. Control valves in this line fail open to insure continuous water supply and backup isolation is provided by integral spring-seated valves on the control rod drive pumps.

The two 2k-inch ventilation openings (supply and exhaust) are closed within six seconds after any scra= signal. Manual opening of these valves is per-mitted (by interlocks) only after the Reactor Protection Syste can be reset 2278 309

3 and high radiation levels (> 10 =R/h) are not present within the contain=ent building.

Review of Off-Nermal Procedure ONP 2.31 indicates that steps are addressed to insure isolation of all lines whose isolation does not degrade needed safety features or cooling capability. However, one procedural step, OnP 2.3.1.3 (f),

which is worded as follows, "If scra caused by hi-enclosure pressure or lcw reactor water level, check all automatic isolation valves closed and cIose nonauto=atic isolation valves not necessary to control scram incident" will be changed to include the following state =ent:

" Place Hand Switch S-5 in the ' ISOLATE' position to preclude valve opening upon loss of trip signal."

This action vill prevent the pneu=atically operated valves (Ite=s L, 5, IL and 16) fro: automatically opening if the sensors (low reactor water level and high contain=ent building pressure) returns to normal during the scra:

incident.

Question:

3.

Descrite the actions, both automatic and manual, necessary for proper functioning of the auxiliary heat removal syste=s (eg, RCIC, Reactor Core Isolation Cooling) that are used when the main feed-water syste=

is not operable. For any manual action necessary, describe in su==ary for= the procedure by which this action is taken in a timely sense.

Response

The RCIC syste= at Big Rock Point Plant is the Emergency Condenser (ECS).

It is co=pesed of a tank with two tube bundles, mini =um capacity per bundle of 16 x 10b Stu/h, a L-hour reserve of cooling water on the shell side of the tank, two makeup syste=s to the shell side, one of which is automatic with a level control using de=ineralized water and a backup =anually supplied fire vattriine. The syste= is designed to be in full operation 30 seconds after receipt of the opening signal for the d-c operated valves.

It is operable and ready for service during all " power" operations.

The syste: vill automatically come into service at approximately 1h35 psig upon initiation of PS REO 7 A L D for MO 7063 and PS REO 7 3 & C for MC 7052 Outlet Valves MO 7053 and 7063 open to place the syste= in service. The inlet valves are nonautomatic but are maintained open during power operation.

The ECS also is automatically placed in service upon loss of a-c output frc=

the motor generator sets to the safety syste Channels I and II.

It can also be manually placed in service fro: the control room console.

Makeup water to the shell side of the ECS is auto =atically supplied by a 2-inch de=ineralized waterline through an air-operated control Valve CV LO25 (fails closed on loss of air or a-c pover ).

This valve maintains the EC3 vater level approximately 28 inches above the tube bundles. As a backup to the normal fill line, there is a manually operated 1-inch fire water supply line inside contain=ent which can be opened to maintain level in the e=ergency condenser upon failure of the de=ineralized water syste=.

There is a L-hour 2278 310

k inventory of water fer decay heat re= oval in the emergency condenser before the level decreases to a point of uncovering the tube bundles. Operatic n beyond four hours =ay require manual starting of the de=in water pu=p or initiation of the fire water supply.

The ECS shell side also has a low-level alar = set at approximately 12 inches above the tube bundle which sounds in the control room.

Two radiation detectors of the scintillation type are counted in the ECS vent to the at=osphere to warn of possible tube leaks.

Question:

h.

Describe all uses and type of vessel level indication for both auto =atic and =anual ir tiation of safety syste=s. Describe other redundant instru-

=entation S-.ch the operator night have to give the same information regarding p. ant status. Instruct operators to utilize other available infor=ation to initiate safety syste=s.

Resoonse:

Auto Initiation:

- LS REO9 A-D - ECCS actuation on RPS isolation trip.

- LS RE09 E-H - Redundant ECCS actuation.

- LT 3180-3183 - RES actuation.

Manual Initiation:

- LI 3380-3383 - RDS Rx level.

There is no other redundant instrumentation available to the operator on vessel level that is LOCA qualified.

Question:

5 Review the action directed by the operating procedures and training instructions to ensure that:

c.

Operators do not override e.uto=atic actions of engineered safety features unless continued operation of engineered safety features vill result in unsafe plant conditions (eg, vessel integrity).

b.

Operators are provided additional information and instructions to not rely upon vessel indication alone for manual actions, but to also examine other plant parameter indications in evaluating plant conditions.

2278 511

5

Response

1.

The operations group and plant manage =ent involved in plant operation vill be provided special sessions dealing with the safety-related syste=

operations at Big Rock Point. Special emphasis will be placed on the ite=s covered by this bulletin. These sessions will be completed prior to plant start-up.

Question:

6.

Part 1:

Review all safety-related valve positions, positioning requirements and positive controls to assure that valves remain positioned (open or closed) in a manner to ensure the proper operation of engineered safety features.

Restonse:

1.

Before start-up from cold condition, valve checklists are perfor=ed on all syste=s which have had work perfor=ed on the=.

Systems which have re=ained in service have been excluded from this check. The checklists include all valves, automatic and manual, in the system.

2.

The checklist shows observation of the position of the valve for operating conditions, as well as locked open or locked closed condition and includes the initial of the individual who checked the valve. Sign-offs include

e. signature by the operator who completed the check sheet, a signature by 2

the control operator who reviewed the check sheet, and a second review and signature on the check sheet by a Shift Supervisor.

Part 2:

Review related procedures, such as those for maintenance, testing, plant and syste= start-up and supervisory periodic (eg, daily / shift checks) surveillance to ensure that such valves are returned to their correct positions following necessary manipulations and are maintained in their proper positions during all operational modes.

Response

1.

There are several checks performed to ensure that equipment (valves) is stored to the operational positf on following the =aintenance activity.

First, the maintenance order requests the valve or equip =ent to be tagged and/or released for the maintenance to be done. The tagging order is written by a Control Operator and reviewed by the Shift Supervisor to re=ove the equipment (valve) from service so that it or the system can be worked on.

This is a step-by-step order to perform proper tagging of equipment. Upon completion of the maintenance, a tagging order is written in a step-by-step sequence to re=ove the tags and restore the equip =ent to operable status.

2278 312

6 2.

Work performed on safety-related equipment is done either by a procedure or by use of an Equipment Outage Request (EOR).

In either case, the precautions are listed such as length of ti=e it can be out of service, effect on other systems, testing to be done before declaring it operatie, and a Shift Supervisor's signature declaring it operable when returned to service. The testing and declaring operability of a system or piece of equipment indicates that the valve align =ent is proper to allow the equipment or syste to perform its intended function.

3 Surveillance testing is completed by use of a step-by-step procedure which designates the movement of valves and also the restoring of valves to their proper position. Each major step has a sign-off for the individ-ual performing the step. The test is completed by a sign-off of the operator conducting the test followed by reviews and signature of the Control Operator and the Shift Supervisor.

k.

Plant and system start-up is conducted according to the Master Checklist.

This checklist designates which valve checklists have to be perforced and which, if any, surveillance tests are required to be completed during start-up.

Question:

7 Review your operating modes and procedures for all systems designed to transfer potentially radioactive gases and, liquids out of primary contain-ment to assure that undesired pu= ring, venting or other release of radio-active liquids and gases will not occur inadvertently.

In particular, ensure that such an occurrence would not be caused by the resetting of engineered safety features instrumentation. List all systens and indicate:

a.

Whether interlocks exist to prevent transfer when high radiation indication exists, and b.

Whether such systess are isolated by the containment isolation signal.

The basis on which continued operability of the above features is c.

assured.

Response

(a)

(b)

High Radiation Isolat'e Systen Ccemodity Ir.terlock On Signal 1.

Ventilation Air Yes Yes Supply Valve (1) 2278 313

7 (a)

(b)

High Radiation Isolate Syste=

Co==odity Interlock On Signal 2.

Ventilation Air Yes Yes Exhaust Valve (1) 3 Clean Su=p Water No Yes Punp Dis-charge Valves (2) k.

Dirty Su=p Water No Yes Pu=p Dis-charge Valves (2) 5 Resin Sluice Water &

No Yes Valves (3)

Resins 6.

Reactor and Water No Yes Fuel Pit Drain (2)

Continued operability of these features is verified by test at least annually in accordance with Technical Specifications requirements.

Question:

8.

Review and modify as necessary your maintenance and test procedures to ensure that they require:

a.

Verification, by test or inspection, of the operability of redundant safety-related syste=s prior to the re= oval of any safety-related system fro service.

b.

Verification of the operability of all safety-related syste=s when they are returned to service following maintenance or testing.

c.

Explicit notificatien of involved reactor operational personnel whenever a safety-related system is removed from and returned to service.

Response

Maintenance and test procedures were reviewed and the appropriate changes vill be made prior to plant start-up.

2278 314

6 Question:

9 Review your pro =pt reporting procedures for NRC notification to assure that NRC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation.

Purther, at that time, an open continuous co==unication channel shall be established and maintained with NBC.

Response

A procedure change to require these actions vill be Lsplemented prior to plant start-up.

Question:

10.

Review operating modes and procedures to deal with significant a=ounts of hydrogen gas that may be generated during a transient or other accident that would either re=ain inside the pri-n y syste= or be released to the containment.

Resnonse:

No procedures presently exist for dealing with significant atounts of hydrogen gas which =ay be generated during an accident. Hydrogen gas generation is not considered to be a significant problem at Big Rock Point. The primary system design is such that any hydrogen gas generated in the reactor vescel is im=ediately vented to the primary steam drum. The line providing this vent path cannot be isolated. The stea= drum is more than 20 feet above the reactor vessel head.

Hydrogen gas generated during nor=al operation (due to radiolytic deco =posi-tion of water) is carried into the turbine and main condenser with nor=al steam flow.

It is then vented to the atmosphere via the condenser air ejec-tors. During full power operation, approximately 6 scfm of hydrogen gas is continually vented from the condenser. The hydrogen off-gas rate drops es-sentia11y to zero once the plant has been shut down.

The only potential mechanism for the generation of significant a=ounts of hydrogen gas is a loss of coolant accident (LOCA). In a LOCA, hydrogen gas would be generated pri=arily as a result of the chemical reaction between steam and the "ircaloy fuel rod cladding. No chemical additives are used in the e=ergency core cooling or contain=ent spray water; hence, there vould be no significant corrosion of other metals in the contain=ent or reactor vessel. Further, hydrogen generation due to the radiolytic deco = position of water would be insignificant compared to that produced by the zirconium steam reaction.

In a LOCA resulting from a small pipe break, the reactor depressurization syste= (RDS) would automatically actuate and any hydrogen generated would, therefore, be vented to containment via the RDS relief lines which are connee-ted to the top of the steam drum. Only the relatively small primary coolant system volume within the emergency condenser tube bundles is at a higher 2278 315

F 9

elevation than the RLS line connections. At any subsequer.t time when reactor syste= pressure is greater than containment atmosphere pressure by the a=ount necessary to operate the RDS relief valves (50 psi), hydrogen which had col-lected in the stea= drum could be vented via this same path. This venting could be perfor:ed from the control roo=.

In a LOCA resulting from a pipe break cf sufficient size to preclude ROS operation, there would be no means of venting hydrogen which had collected in the primary syste=, except via the break.

In this case, any hydrogen gas produced could collect in the steam drum.

(It should be noted, however, that Big Rock Point core spray syste=s would provide adequate core coolinC even in a hydrogen environ =ent.)

Consu=ers Power Co=pany does not consider this to be of safety significance since no potentia 2 ignition source is present to initiate an explosion.

Once any hydrogen gas has been vented to the contain=ent, the potential for a hydrogen gas concentration exceeding the explosive concentration exists.

Based on Regulatory Guide 17, " Control of Combustible Gas Concentrations in Contain=ent Following a Loss of Coolant Accident," a four volu=e percent (k v/0) concentration of hydrogen gas in stea= and air is potentially fla==able.

The Guide further states that a hydrogen concentration of 6 v/0 would not be expected to result in adverse consequences to containment syste=s.

It has been calculated that, if all the zirconiu= in the Zircaloy cladding (cladding OD =.LL9", thickness =.03L") in the fueled zone (70") of all of the fuel rods (8L bundles with 117 full rods per bundle for G-3 fuel) were to react with stea=, 7.178 x 10k = oles of hydrogen gas would be produced. Conservative calculations of the =axi=u= amount of H2 which could be di coolant at the time of an accident indicate that 195 x 10gsolved in the primary additional coles of H2 cculd be released to the containment as a result of depressurizing the pri-

=ary coolant system. The number of moles of air in contain=ent during nor=al 6

operation is 1.267 x 10.

(This corresponds to a contain=ent free volu=e of 9L0,000 ft3.) Thus, if all of the cladding in the active zone on all of the fuel rods were to react with steam, and if this were added to the =axi=u:

amount of dissolved H2 which could be released from the coolant, a hydrogen concentration in the contain=ent of less than 5.7 v/$ vould result. Approxi-mately 46% of the zirconium in the active zone on all of the fuel rods vould have to react with steam to produce an average contain=ent hydrogen gas con-centration of 4 v/$ (using the sa=e conservative assu=ptions).

In conclusion, explosive mixtures of hydrogen gas in the containment should not occur following a LOCA and, therefore, a =eans for dealing with hydrogen gas in the contain=ent are not necessary.

2278 316

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