ML19274F103

Forwards Response to IE Bulletin 79-06A.Procedural Changes & Training Planned for Completion Prior to Facility Startup

ML19274F103
Person / Time
Site:	Beaver Valley
Issue date:	04/30/1979
From:	Woolever E DUQUESNE LIGHT CO.
To:	Grier B NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
References
IEB-79-06A, IEB-79-6A, TAC-11685, NUDOCS 7906130134
Download: ML19274F103 (10)

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(412) 471-4300 435 S.ath Avenue Pittsburgh, Pennsylvan~.a

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April 30, 1979 United States Nuclear Regulatory Com ission Attention: Boyce H. Grier, Director Region I 631 Park Avenue King of Prussia, Pennsylvania 19406

Reference:

Beaver Valley Power Station, Unit No. 1 Docket No. 50-334 License No. DPR-66 Response To IE Bulletin 79-06A Gentlemen:

We have reviewed your IE Bulletin No.79-06A which identifies actions to be taken by the licensees of Westinghouse Pressurized Water Power Reactors as a result of the Three Mile Island Unit 2 incident. Attached are the responses to Items 1 to 12. All procedural changes and training identified are expected to be completed prior to BVPS Unit 1 startup following resolution of the pipe hanger support problem.

If you have any questions regarding this response, please contact my office.

Very truly yours, E. . Woo ever Vice President, Engineering and Construction Attachment cc: United States Nuclear Regulatory Commission Office of Inspection and Enforcement Division of Reactor Operations Inspection Washington, D. C. 20555 790613013'

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DUQUESNE LIGHT COMPAh"I Beaver Valley Power Station Unit No. 1 RESPONSE TO IE BULLETIN No.79-06A Letter Dated April 14, 1979

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Item 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 auxiliary feedwater trains at the Three Mile Island Unit 2 plant and other actions taken during the early phases of the accident; (2) the noparent operational errors which led to the eventual core damage; (3) that the potential exists, under certain accident or transient conditions, to have a water level in the pressurizer simultaneously with the reactor vessel not full of water; and (4) the necessity to systematically 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 will result in unsafe plant conditions (see Section 7a.); and (2) not make operational decisions based solely on a single plant parameter indication when one or more 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 documented in plant records.

Response 1 A program has been initiated to assure that all licensed operators, plant management, and supervisors with operational responsibilities will be familiarized with the Three Mile Island incident. Participation by the necessary personnel will be documented. The program will consist of, at a minimum, the following informa tion:

a. Description of the TMI event including the sequence of events as described in IE Bulletin 79-05A.

b. The necessity to analyze more than one plant parameter to determine the overall status of a system.

c. Guidelines to assure that ESF operation will not be overridden unless the plant is in a stable, safe condition or unless unsafe plant conditions will result from its cantinued operation.

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Beaver Valley Power Station, Unit No. 1 Response To IE Bulletin No.79-06A NRC Letter Dated April 14, 1979 ,

page 2 Response 1 (continued)

A training program addressing the above items is presently being conducted however total completion of item C is dependent on operator familiarization of additional procedures presently being formalized.

Item 2 Review the actions required by your operating procedures for coping with transients and accidents, with particular attention to:

a. Recognition of the possibility of forming voids in the primary coolant system large enough to compromise the core cooling capability, especially natural circulation capability,

b. Operator action required to prevent the formation of such voids.

c. Operator action required to enhance core cooling in the event such voids are formed. (e.g., remote venting)

Response 2 Operating procedures for coping with transients and accidents that could result in void formation were reviewed. As a result of this review, procedures dealing with depressurization will be revised and upgraded from abnormal procedures to Emergency Operating Procedures.

Item 3 For your facilities that use pressurizer water level coincident pressurizer press"ee for automatic initiation of safety injection into the reactor coolant system. trip the low pressurizer level setpoint bistables such that, when the pressurizer pressure reaches the low setpoint, safety injection would be initiated regardless of the pressurizer level. The pressurizer level bistables may be returned to their normal operating positions during the pressurizer pressure channel functional surveillance tests. In addition, instruct operators to manually initiate safety injection when the pressurizer pressure indication reaches the actuation setpoint whether or not the level indication has dropped to the actuation setpoint.

Response 3 Site personnel are looking into revising safety injection logic to provide for 2/3 pressurizer pressure safety injection logic. In the interim, the pressurizer level coincident bistables will be placed in the trip mode whenever the RCS pressure is

> 2000 psig. Procedures are being revised to equire manual initiation of safety injection at the pressurizer pressure setpoint of 1845 psig.

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Beaver Valley Power Station, Unit No. 1 Response To IE Bulletin No.79-06A NRC Letter Dated April 14, 1979 Page 3 Item 4 Review the containment isolation initiation design and procedures, and prepare and implement all changes necessary to permit containment isolation whether manual or automatic, of all lines whose isolation does not degrade needed safety features or cooling capability, upon automatic initiation of safety injection.

Response 4 Present station design provides a containment isolation Phase A on any initiation of safety injection which isolates all lines not required for an orderly and safe hot shutdown of the reactor. This includes the flow paths for the containment sump, primary drains transfer tank in containment, and the containment vacuum pumps. Resetting CIA will not automatically reopen these flow paths. Manual operator action is required to open the flow paths.

Item 5 For facilities for which the auxiliary feedwater system is not automatically initiated, prepare and implement immediately procedures which require the stationing of an individual (with no other assigned concurrent duties and in direct and continuous communication with th; control room) to promptly initiate adequate auxiliary feedwater to the steam generator (s) for those transients or accidents the consequences of which can be limited by such action.

Response 5 Present station design provides for automatic initiation of auxiliary feedwater flow upon receipt of either loss of heat sink indication or safety injection actuation.

Item 6 For your facilities, prepare and implement immediately procedures which:

a. Identify those plant indications (such as valve discharge piping temperature, valve position indication, or valve discharge relief tank temperature or pressure indication) which plant operators may utilize to determine that pressurizer power operated relief valve (s) are open, and

b. Direct the plant operators to manually close the power operated relief block valve (s) when reactor coolant system pressure is reduced to below the set point for normal automatic closure of the power operated relief valve (s) and the valve (s) remain stuck open.

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Beaver Valley Power Station, Unit No. 1 Response To IE Bulletin No.79-06A NRC Letter Dated April 14, 1979 Page 4 Response 6 Present procedures and proposed revisions identify the following indicators which are in the direct view of the operator:

a. Position indicating lights for power operated relief valves.

b. Temperature indicator for power operated relief valves line.

c. Temperature indicators for each of the safety valve relief lines.

d. Pressure, level and temperature indication for the pressurizer relief tank.

The procedures will direct the operator to utilize the aboea indications to determine that pressurizer power operated relief valve (s) tre open and direct the operators to manually close the power operated relief alock valve (s) when reactor coolant system pressure is reduced below the setprint for normal automatic closure of the power operated relief valve (s) atd the valve (s) remain stuck open.

Item 7 Review the action directed by the operating procedures and training instructions to ensure that:

a. Operators do not override automatic actions of engineered safety features, unless continued operation of engineered safety features will result in unsafe plant conditions. For example, if continued operation of engineered safety features would threaten reactor vessel integrity, then the HPI should be secured (as noted in b(2) below).

b. Operating procedures currently, or are revised to, specify that if the high pressure injection (HPI) system has been automatically actuated because of low pressure condition, it must remain in operation until either:

(1) Both low pressure injection (LPI) pumps are in operation and flowing for 20 minutes or longer; at a rate which would assure stable plant behavior; or

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Beaver Valley Power Station, Unit No. 1 Response To IE Bulletin No.79-06A NRC Letter Dated April 14, 1979 Page 5 Item 7 (continued)

(2) The HPI system has been in operation for 20 minutes and all hot and cold leg temperatures are at least 50 degrees below the saturation temperature for the existing RCS pressure. If 50 degrees subcooling cannot be maintained after HPI cutoff, the HPI shall be reactivated. The degree of subcooling beyond 50 degrees F and the length of time HPI is in operation shall be limited by the pressure / temperature considerations for the vessel integrity.

c. Operating procedures currently, or are revised to, specify that in the event of HPI initiation with reactor coolant pumps (RCP) operating, at least one RCP shall remain operating for two loop plants and at least two RCPs shall remain operating for 3 or 4 loop plants as long as the pump (s) is providing forced flow.

d. Operators are provided additional information and instructions to not rely upon pressurizer level indication alone, but to also examine pressurizer pressure and other plant parameter indications in evaluating plant conditions, e.g. , water, inventory in the reactor primary system.

Response 7

a. A checklist is being developed to provide the operators guidance in resetting and overriding safety injection. It must be realized that resetting the safety injection signal neither terminates nor changes the state of any automatic actions, but allows the operator to realign his system during a recovery. The guidelines will assure that ESF operation will not be overridden unless the plant is in a stable, safe condition or unless unsafe plant conditions will result from its continued operation, as before.

b. Procedures are being revised to assure that the plant is in a safe condition prior to terminating high head safety injection flow following actuation because of a low pressure condition.

1) For those LOCA conditions where both the high head and low head safety injection systems would be operating and delivering water to the primary system, current procedures call for continued operation of both systems.

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Beaver Valley Power Station, Unit No. 1 Response To IE Bulletin No.79-06A NRC Letter Dated April 14, 1979 Page 6 Response 7 (continued)

2) For those LOCA conditions where only the high head safety injection system is delivering water to the primary system, the criteria permitting termination of high head safety injection reflect upon conditions which are at least 50* subcooled although no specific subcooling criterion is included. Rather, procedures are being based upon parametric indication available to the operator and the parametric values have been chosen to be. consistent with a highly subcooled state in the primary system. These parametric values will include:

a) Wide range RCS pressure > 2000 psig and increasing, b) Wide range level indication in at least two steam generators and increasing, c) Pressurizer level at normal program range and rising, and d) Wide range RCS hot leg temperatures < 580F.

These provide for a subcooled primary side, stable primary side conditions, and a heat sink via the steam generators.

It is felt th>r these criteria are adequate and that a 20 minute minimum running criterion for the high head safety injection system is hazardous and could lead to overpressurization or unnecessary cycling of the power t,erated relief valves (which in turn could result in an accident condition) for the following reasons:

1) Each HHSI pump is capable of delivering 150 gpm and two could deliver 6000 gallons over a 20 minute period. Normal operating water volume in the pressurizer is at least 5200 gallons and the pressurizer capacity is 10,500 gallons. Mathematically it is possible to fill the pressurizer solid and vent off the steam bubble in less than 20 minutes.

2) With two HHSI pumps running and initial pressurizer pressure and level being 1800 psia and 5%, respectively, mathematically it could take less than 5 minutes to require the lifting of the power operated relief valves to avoid excessive pressure buildup in the pressurizer.

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Beaver Valley Power Station, Unit No. 1 Response To IE Bulletin No.79-06A NRC Letter Dated April 14, 1979 Page 7 Response 7 (continued)

3) D'u ring the transient experienced at BVPS Unit 1 on July 28, 1978 as a result of the main transformer failure and the resulting station blackout for 20 minutes, with only one diesel generator providing power and only one HHSI pump delivering flow, it took less than seven minutes to reach the power operated relief valve setpoint starting from an initial pressurizer pressure of 1970 psig and level of 21.5% following the reactor trip.

Item 8 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. Also review rela' 2d procedures, such as those for maintenance, testing, plant and system startup, and supervisory periodic (e.g., 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 8 Station personnel are reviewing the positioning requirements of all ESF valves for normal operation, plant or system startup, or return to service after testing or maintenance, to assure that they are positioned in a manner to ensure the proper operation of the engineered safety features. Station personnel are also taking the positive approach of requiring all manual valves in the direct flow path of these systems which do not have indication in the control room to be locked in position to assure the availability of the flow path.

Item 9 Review your operating modes and proc' Wres for all systems designed to transfer potentially radioactive gases and iquids out of the primary containment to assure that undesired pumping, venting or other release of radioactive liquids and gases will not occur inadvertently.

In particular, ensure that such an occurrence would not be caused by the resetting

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of engineered safety features instrumentation. List all such systems and indicate:

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

b. Whether such systems are isolated by the containment isolation signal.

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

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Beaver Valley Power Station, Unit No. 1 Response To IE Bulletin 79-06A NRC Letter Dated April 14, 1979 Page 8 Response 9

a. The systems which transfer radioactive gas or liquid outside of the containment do not have interlocks with high radiation levels. However, other parameters are monitored which more reliably detect accident situations and provide for automatic initiation of containment isolation.

b. These systems are isolated by a containment isolation signal as described in the response to Item 4.

c. Operability of the above features is accomplished by periodic testing in accordance with Technical Specification requirements.

Item 10 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 systems prior to the removal of any safety-related system from service.

b. Verification of the operability of all safety-related systems when they are returned to service following maintenance or testing.

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

Response 10

a. Station administrative procedures require that a licensed senior operator verify that redundant safety related equipment or systems will not be affected by the removal of any safety related system from service.

It is felt that additional testing beyond the established periodic surveillance and testing program is not necessary. Periodic surveillance and testing provides a sufficient level of confidence that the system will function as designed when necessary, without requiring additional testing. Testing intervals are selected in recognition of the fact that certain equipment may be unavailable on a temporary basis due to maintenance, and restrictions are imposed on the length of time such equipment can be removed from service.

b. Present procedures require the verification of the operability of safety related systems when they are returned to service following maintenance.

The surveillance testing procedures are conducted to verify operability and the acceptance criteria are set up such that adherence to the acceptance criteria demonstrates operability of the systen.

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Beaver Valley Power Station, Unit No. 1 Response To IE Bulletin 79-06A NRC Letter Dated April 14, 1979 Page 9 Response 10 (continued)

c. Present procedures require notification of reactor operational personnel prior to removing from or returning to service any safety related system.

Item 11 Review your prompt 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. Further, at that time an open continuous communication channel shall be established tnd maintained with the NRC.

Response 11 Changes are being made to the station administrative procedures to require one hour notificatio:. to the NRC in the event that the reactor is not in a controlled or expected condition of operation. Notification will be given directly to the Region I office and they will be asked to return the call to confirm the occurrence of the event. The return call line will be left open and will be considered the continuous communication channel.

Item 12 Review operating modes and procedures to deal with significant amounts of hydrogen gas that may be generated during a transient or other accident that would either remain inside the primary system or be released to the containment.

Response 12 Operating modes to deal with significant amounts of hydrogen gas that may be generated during a transient or other accident that would either remain incide the primary system or be released to containment will be reviewed and genei t guidance will be provided in the operating manual to deal with the problem.