ML19259C050

Ack Receipt of NRC 790524 Comments Re Util Actions in Response to IE Bulletin 79-06B.Clarifies Actions Re Personnel Training,Void Formation,Containment Isolation, Reactor Coolant Pump Operation & Sys Operability

ML19259C050
Person / Time
Site:	Calvert Cliffs
Issue date:	06/05/1979
From:	Lundvall A BALTIMORE GAS & ELECTRIC CO.
To:	Reid R Office of Nuclear Reactor Regulation
References
NUDOCS 7906120265
Download: ML19259C050 (6)

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BALTIMORE G AS AND ELECTRIC COMPANY

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B A LTI M O R t. . A R YL AN D 21203 $

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June 5, 1979 _

Office of Nuclear Reactor Regulation

• U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Mr. Robert W. Reid, Chief Operating Reactors Branch #4 Divisicn of Operating Reactors

Subject:

Calvert Cliffs Nuclear Power Plant Unit Nos.1 and 2, Docket Ne3. 50-317 and 50-318 IE Bulletin 79-06B

References:

(1) A. E. Lundvall, Jr. letter to Boyce H. Grier dated April 26, 1979, same subject (2) A. E. Lundvall, Jr. letter to Boyce H. Grier dated May 8,1979,. same subject (3) Combustion Engineering, Inc. report on Small Break LOCA to Mr. Ashok Thadant (USNRC) of May 15,1979 Gentlemen:

This letter responds to comments we received by telecopier on May 24, 1979 from your Mr. E. L. Conner, Jr. (DOR) regarding References (1) and (2). These comments were clarified by a phone conversation with Mr.

Dixter on May 29, 1979. The item numbers below are the same as those used in the telecopied corments we received and generally refer to the item numbers of the subject bulletir..

ITEM 1 - TRAINING OF LICENSED PERSONNEL In addition to the training described by Reference (1) given by the "NRC Briefing Team", the following training has been given, or is -

planned, to cover subjects related to the TMI-2 incident:

7906120 6 2284 038 ool 5 sl

IE BULLETIN 79-06B Page 2

1. All licensed operations personael have been trained in the procedure changes made as a result of THI-2 which were described in References -

(1) and (2). This training was verified by interviewing operators during the last inspection visit of your Mr. D. F. Johnson.

2. The regular " retraining" sessions have been modified to schedule specific training on the subjects required by item 1 of the bulletin, 0 as well as other aspects o' the TMI-2. incident is it relates to the Calvert Clif4 plants. T! A retraining has been attended by approxi-mately two-thirds of N 10:1: sed personnel to date. As more infoma-tion becomes availa'. . , d. as the result of investigations and design reviews. Qis information is added to the requalification training lesta plans.

3. The Company has reserved simula <

• time at the Combustion Engineering simulator July 9 - July 28,19i.' ir the retraining of licensed personnel . The time spent at the simulator will be totally devoted to accident scenarios and will include small-break simulations. It should be noted that this simulator is a replica of the Calvert Cliffs Unit 1 Control Room.

ITEM 2 - VOID FORMATION As stated by Reference (2), the nomal olant operating procedures and non-routine plant operating procedures ensure that a subcooled condition is maintained and verified by the observation of RCS pressure (Pressurizer pressure), cold leg and hot leg temperatures and Pressurizer temperature.

During nomal operation, at least a 500F margin of subcooling is maintained.

While shutdown and recovering from a solid water condition, approximately a 1000F margin is provided by procedural control while the Pressurizer bubble is being fomed. When the transition between shutdown cooling and RCP operation is being made, the in-core thennocouples are used to monitor temperature in the core.

For transient and accident conditions, the emeroency procedure for a loss of reactor coolant was reviewed and revised as described under item 6 of Reference (2); these changes addressed the operation of HPSI pumos, Reactor Coolant Pumps and the use of pressurizer level instrumentation.

The coments pmvided by Mr. Conner request more infomation reoarding the recognition of voiding, avoidance of voiding, and core cooling enhance-ment in the event of voidino. The review and revisions to the loss of reactor coolant prccedure have provided for these considerations as follows:

1. The recognition of voiding due to steam famation can be accomplished by the detemination of saturation conditions in the RCS; these conditions can be recognized by the use of the RCS temperature instru-ments, in-core themocouples and Pressurizer pressure instruments.

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IE BULLETIN 79-06B Page 3

2. When possible, in a small break situation, the prevention of-unnecessary voiding can be accomplished by the continued operation ,

of the HPSI pumos to provide make-up water and a source of system pressure. As stated in Item 6 of Reference (2), the criteria for inhibiting HPSI required by the subject bulletin has been incorporated into the procedure (a classification of " endorsed by our NSSS supplier" is provided in 6.2 below). Additionally, the maintenance of an-  ;

adequate heat sink to provide for heat removal (in addition to th,at removed by the postulated break) is important. Procedural provisions for the assurance of such cooling is discussed in 3 below.

3. Depending upon the size of the break and upon the availability of an off-site power source, three basic modes of cooling the core are provided fo by the procedure.

a. Forced circulation by RCP's - as discussed in Item 6 of Reference (2), the continued 6peration of one RCP in each 1000 is required by the procedure whenever off-site power is available and as long as forced flow is provided. The NSSS supplier recomends continued RCP operaticn under these conditions until it can be established that the core outlet is 200F subcooled. Based on your coment 6.3, the 200F is "not acceptable" - this criterion has been removed from our pmcedure,

b. Natural circulation - should offsite power not be available or should the operating RCP's incur damage causing them to be inoperable, RCS natural circulation may be available (in the absence of core boiling /S. G. condensation). The subject procedure has been revised to include instructions for the verification of natural circulation by the observation of temper-ature rise acmss the core (expected to be less than 200F for subcooled circulation), the stabilization and decrease of core outlet temperature and the stabilization and/or decrease of cold leg temperature. The system should be maintained at least 200F subcooled during natural circulation.

c. As discussed in Reference (3), for a certain size range of small breaks, the core will partially uncover, and heat removal, in addition to that removed by the break, is removed by core boiling-steam generator condensation. This condition would be evidenced by the observation of saturation conditions in the RCS and an empty pressurizer.

In all of the above described modes of heat removal, it is imoortant to establish and maintain approximately a nonnal water level in the steam _

generators; this action is set forth in the subject pmcedure as well as a caution to control feed rate to prevent the excessive cooldown/ depress-urization of the RCS.

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IE BULLETIN 79-06B Page 4 The instrumentation observed by the operators to monitor the core and RCS '

r conditions in the above situations includes:

l. RCS Tem 6650 F ) peratures - cold leg (range 0-6000F) and hot leg (range 515-In-core Themocouples (range 50-17620F) 2.

3. Pressurizer Level (range 0-360 in.)

4. Pressurizer Pressure (range 0-2500 psia)

5. Steam Generator Level (-100 to +60 in.)

Parameters which can be monitored from the Control Room to indicate the status of the RCP's include

1. Seal Pressures (range 0-2500 psia)

2. Lower Seal Temperature (range 0-3000F)

3. Thrust Bearing Temperatures (range 50-2500F) 4 V'bration

5. Motor Amperes As an aid to the operator, curves displaying a saturation line and a 50 0F subcooling line plotted versus temperature and pressure has been added to the subject procedure.

ITEM 3 - CONTAINMENT ISOLATION The list of penetrations to be isolated from the Control Room upon receipt of automatic safety injection set forth in item (3) of Reference (1) includes all lines except component cooling (to the RCP's), instrument air and RCP seal bleedoff. These three systems remain unisolated to support RCP operation and valve (CV) operations.

ITEM 4 No further infomation required (per phone conversation with Mr. E. L.

Conner, Jr.). _

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IE BULLETIN 79-06B Page 5 ITEM 5 - PORY INDICATIONS The following indications are available to detemine the status of the PORV; the loss of reactor coolant procedure includes this list of indications:

1. Quench tank level, pressure and temperature . 2

2. PORY position indication

3. PORV discharge line tenperature ITEM 6 - RCP/HPSI OPERATION 6.1 Upon receipt of a Containment Isolation Signal (CIS), the Component Cooling water for the RCP's is isolated. Should this system remain isolated, eventual failure of the seals will occur. Since the failure would include deterioration of the seal "0-rings", such failure would occur whether or not the pump is running. More importantly, the lack of cooling to the oil system would result in the eventual failure of the thrust bearing, probably causing the misalignment and seizure of the pump. No time estimate is available until failure of the bearing occurs; however, manufacturer's tests indicate that the maximtun recommended temperature (1950F) would be exceeded in approximately 10 minutes. As indicated in item 6 of Reference (2), it is important to reset CIS and re-establish cooling flow to the RCP's if containment isolation has occurred.

A design study has been initiated to detemine the feasibility of system modification to retain component cooling flow to the RCP's in all post-accident situations or to provide for the rapid restoration of the system following a non-seismic accident scenario.

6.2 All of the criteria set forth by IE Bulletin 79-06B, item 6.b have been incorporated into the appropriate procedure; these requirements were not modified by the recommendations of the NSSS supplier.

6.3 The 200F subceoling criteria for stooping the RCP's, if an alam parameter is exceeded, has been deleted from our procedure based on your comment that you do not recognize this as an acceptable criteria.

6.4 No further information requested. -

ITEMS 7 and 8 2284 042 No further information requested.

IE BULLETIN 79-06B Page 6 ITEM 9 - VERIFICATION OF SYSTEM OPERABILITY (1) As used in the response set forth by item 9 of Heierence (1), the tem " equipment" was intended to mean system or cnponent as applicable. In the event tha non-operability of a single component would cause a system or sub-system to be inoperable, then the system or sub-system would be classified as such under the requirements. O set forth by the plant Technical Specifications.

(2) No further infomation reouested (per phone conversation with Mr.

E. L. Conner, Jr.).

(3) Whenever a safety related system or component is taken out of service, the appropriate Technical Specification " action statement" is logged in the Control Room and Shift Supervisor's logs. Each succeeding shift must then enter a summary of these action state-ments as the first entry of each shift; this is carried on until the action statement is cleared. This mechanism ensures that the personnel are continually aware of the outage status of safety related equipment.

ITEMS 10 and 11

~

No further infomation required.

ERRATA Corrections to Reference (1) should be made as follows:

1. Page 6, under " Item 8", the last sentence of the paragraoh numbered "2" should read, " Isolated by SIASor CVCS Isolation Signal".

2. Page 6, a new oaragraph should be added after "7" to read, " Steam Generator Blowdown Isolated by CIS".

Should you have further questions regarding these matters, we would be pleased to discuss them with you.

Very. truly yours,. f'

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/ N , n d --- L /

C ~A. E. Lundvall, Jr. T N 2284 Qf}

Vice President - Supply AEL/ RED /dds Copies To: Mr. E. L. Conner, Jr. - NRC, D0R Mr. Boyce H. Grier - NRC, I&E