ML19261E525

Responds to 790720 Request for Addl Info Re IE Bulletin 79-08.Procedures Revised by 790620 to Assure Isolation of RHR to Fuel Pool Deminimizing Line & Torus Makeup Line in Event of Containment Isolation Signal

ML19261E525
Person / Time
Site:	Pilgrim
Issue date:	08/21/1979
From:	Andognini G BOSTON EDISON CO.
To:	Ippolito T Office of Nuclear Reactor Regulation
References
79-165, NUDOCS 7908280592
Download: ML19261E525 (11)

Text

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• t ' ' s BOSTON EDISON COMPANY oENENAL OFFICES Boo BOYLSfCN STmEET 80sTON MAESACNuSETTE o2199

• G. CAmL ANDOGNINO supEmINTENCENT NUCLEAm QPEmATIONE DEPARTMENT August 21, 197?

BEco. Ltr. #79-165 Mr. Thomas A. Ippolito, Chief Operating Ranctors Branch #3 Division of Operating Reactors Office of Nuclear Reactor Regulation -

U.S. Nuclear Regulatory Co= mission Washington, D. C. 20555 License No. DPR-35 Docket No. 50-293 Supplementary Information to IE Bulletin 79-08 Ref. a.) Boston Edison Company wer to Boyce H. Grier titled, " Response to IE Sulletin 79-08", dated April 25, 1979 b.) Nuclear Regulatory Commission letter to G. C.

Andognini titled " Request for Additional Information on IE Bulletin 79-08" dated July 20, 1979

Dear Sir:

Reference b.) above, requested additional information to our response to IE Bulletin 79-08 Ref. a. ) . The following is our response to your items of concern:

Item No. 1 Provide the date on which you completed the actions required by Item 1 of IE Bulletin 79-08.

Response

The review was completed on May 18, 1979.

Item No. 2

1. Your response indicated that procedures would be developed to isciate the Residual Heat Removal to Spent Fuel Fool Demineralizer line and ne Torus Makeup line in the event either was in service on receipt of a containment isolation signal. Provide your schedule for completion of this action.

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BOSTON E* DIS O N COMPANY Mr. Thomas A. Ippolito "

August 21, 1979 Page 2

Response

The following station operating procedures were revised to assure that in th2 event of a containment isolation signal, both the RHR to Fuel Pool Demin I.ine and the Torus Makeup I.ine would be isolated.

2.2.30 Reactor Building Closed Cooling Water 2.2.35 Condensate Storage and Transfer System 2.2.36 Instrument Air System 2.2.85 Fuel Pool Cooling and Filtering System 2.2.86 Residual Heat Removal These revised procedures were in place by June 20, 1979.

2. The Main Steam Drain lines and Reactor Water Sample lines ire not isolated on high drywell pressure. Isolation of these lines on high drywell pressure would not appear to result in degradation of needed safety features or cooling capability. Therefore provide justification for not isolating these lines or prepare and implement the necessary changes to isolate these lines as required by the Bulletin.

Response

A high drywell pressure isslation function will be added to the Main Steam Drain Lines and the Reactor Water Sample Lines. A Plant Design Change Request has been initiated to accomplish these modifications.

This will be completed during the scheduled 1980 refueling outage.

_ Item No. 4

1. Describe other redundant instrumentation which the operator might have to determine changes in reactor coolant inventory, e.g., drywell high pressure, radioactivity levels, suppression pool high temperature, containment su=p pump operation, etc.

Resnonse The attached table is a complete listing of all instrumentatica, other than vessel level instrumentation, which an operator might have to determine changes in reactor coolant inventory.

2. Clarify your response to indicate whether operators have been instructed to utilize other available information to initiate safety systems. Pievide your schedule for completion of this action.

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BOSTON EDISON COMPANY Mr. Thomas A. Ippolito August 21, 1979 Page 3

Response

The Training Department is preparing an outline which details all information available that an operator may utilize to initiate safety systems. This out-line will be distributed to all operations personnel and on shif t lectures will be held by the Watch Engineers.

This training will be completed by October 9,1979.

Item No. 5

1. Verify that the redundant and diverse instrumentation referred to in your response to item 5b includes instrunentation other than water level instru-mentation, e.g., drywell high pressures drywell high radioactivity levels, suppression pool high temperature, etc.

2. Provide a schedule for any actions on Item 5 that have not yet been completed.

Response

Station procedures will be revised to emphasize the necessity of verifying con-ditions utilizing all available instrumentation. This revision will emphasize both level and instrumentation delineated in the response to Item No.4.

These procedure changes will be completed by October 1,1979.

Item No. 6

1. It is not clear from your response that safety related valve positioning requirements were reviewed to ensure proper operation of engineered safety features. Please supplement your response to provide a commitment to con-duct this review and a schedule for completion.

Response

A review is being conducted of all station operating procedures which include safety related valve positioning requirements. These valve positioning re-quirements are being compared to those specified in the Final Safety Analysis Report so that the valves remain positioned in a manner which uill ensure the proper operation of engineered safety features.

This review will be completed by October 1, 1979.

2. Your response did not clearly indicate that all accessible safety-related valves had been inspected to verify proper position. Nor was a schedule for performing the position verification for all safety-related valves provided. Please supplement your response to provide this infornation.

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BOSTON EDISON COMPANY Mr. Thomas A. Ippolito August 21, 1979 Page 4 Respc'se All accessible safety related valves have been inspectad to verify proper position. These positions are verified daily by operations personnel. All safety related valves located in the quadrants are inspected bi-weekly and since the issuance of IE Bulletin 79-08 have been locked in the proper position.

All inaccessible safety related valves in the drywell are locked in the proper positica and the position indication verified daily.

Item No. 7

1. Your response regarding inadvertent transfer of radioactive liquids by operation of the containment sump pump on resetting engineered safety features instrumentation is unclear. Verify that your present procedures recuire the switch for the sump isolation valves be taken out of the " auto" position prior to reset in order to prevent inadvertent transfer or radio-active liquids out of containment. In addition provide your schedule for removing the " auto" function f rom the switch.

Response

In our initial response to Item No. 7, we stated that the control switch governing the containment sump isolation valves would have its auto open function perman-ently removed in order to preclude the possibility of inadvertently transferring radioactive liquids from containment. However, during the review of the planned modifications for this commitment it was determined that a more practical and conservative modification could be implemented on the sump pump control logic as follows.

Thi; modification would remove the currently existing " Auto Pump Start" feature but would still retain the " Auto Pump Trip" on low level feature when operated in the " Auto" mcde. Upon implementation of the modification, the pumps when operated in the " Auto" mode would require manual starting by means of a " Start" push button. Pumping would continue until either the sump reached the low level trip setting or the circuit timer timed out. The intent of the circuit timer which is adjustable, is to back-up the low level trip setting by limiting the total pumping time to a marimum of 30 minutes for each sump pump. With this

" start" push button feature, transfer of liquids requires the administrative interlock of contacting the Main Control Room to open containment isolation valves, as well as manual sump pump actuation, thereby eliminating the possibil-ity of undersired pumping. These containment isolation valves receive a closure signal on high drywell pressure which causes the valves to close. This modification is currently installed and in operation.

Also, an evaluation is being made of additional design modifications such as a gross radiation monitor downstream of the isolation valves, that in operation would generate a signal to close these valves on high radiation.

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BOSTON E" DIS O N COMPANY Mr. Thomas A. Ippolito August 21, 1979 Page 5 Item No. 8 We understand from your response that operability tests are performed on redundant safety-related systems prior to removal of any safety-related system from service. Since you may be relying on prior operability verifi-cation within the current Technical Specification surveillance interval, operability should be further verified by at least a visual check of the system status to the extent practicable, prior to removing the redundant equipment f rom service. Please supplement your response to provide a commit-ment that you will revise your maintenance and test procedures to adopt this position.

Response

Pilgrim Station procedures require that operability tests be performed on redundant safety related systems immediately prior to removal of any safety related systems from service. Reliance on prior operability verification within the current Technical Specification surveillance interval would there-fore be a violaticn of station procedure. No further revisions to maintenance and test procedures are necessary.

Item No. 9 Your response is unclear as to the extent of commitment to the requirements of Item 9 of IE Bulletin 79-08. We require a rewording of your response to ensure an explicit commitment to the requirements of Item 9 of IE Bulletin 79-08.

Response

All station reporting procedures for NRC notification have been reviewed and revised as necessary, to assure that the :RC is notified within one hour of the time the reactor is not in a controlled or expected condition of operation and at that time an open continuous communication channel shall be established and maintained with the NRC.

Additionally these notification requirements have been included in the station Emergency Procedures. This procedural review and revision was completed by August 15, 1979.

Item No. 10 You stated in your response that you have reviewed only one procedure in connec-tion with this item. Please provide a summary description of the operating modes and procedures which you reviewed for adequacy to deal with significant amounts of hydrogen for both transient and accident conditions.

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BOSTON EDISON COMPANY Mr. Thomas A. Ippolito, Chief August 21, 1979 Page 6 Response _

Those transient and/or accident conditions, which have potential for generating significant amounts of hydrogen gas, have been reviewed. The procedure detailed in our initial response, Post Accident Venting, has been determined to be ad-equate for dealing with hydrogen within the Primary Containment regardless of the source, A new procedure, Loss of Coolant with No Pipe Breaks, has been developed to deal with those situations which result in fuel damage and hydrogen generation which remains within the primary coolant system.

This procedure details utilization of the Augmented Offgas System's hydrogen recombiners and charcoal absorber vaults.

We trust that you will find this information satisfactory; however, should you desire additional information, please contact us.

Very truly yours,

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Atta chments cc: Mr. Boyce H. Grier, Direct.cr Office of Inspection and Enforcement Region I U.S. Nuclear Regulatory Ccmmission 631 Park Avenue King of Prussia, PA. 19406 2022 048

ATTACIIMEliT Page 1 of 6 Instrument Loop ,

Use Type Location PI 5067A Indicates drywell Indicator Control Room pressure Panel 904 PI 5067B Indicates torus Indicator Control Room pressure Panel 904 DPI 5021 Indicates drywell/ torus Indicator Control Room

• differential pressure Panel 904 i I l PT 5060 Monitors drywell 59- Computer to Control Room

torus pressure & provides Recorder or Panel 905

input for computer printout printout i -

- Indicating MR-5051 Records relative humidity Recorder Reactor Building at (2) drywell and (3) - Computer Panel C-85 torus points Printout N

C) TR 50508 Records temperture Indicating rs) Reactor Building at (6) drywell and Recorder Panel C-85 IN3 ~

at (3) torus points CD 42- '

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TR 5050A Reco rds temperature at Indicating Reactor Building (12) drywell points Recorder Panel C-85

Page 2 of 6 Instrument Loop Use Type Location TI 9018 Indicates Torus Air Temp- Indicator Panel 903 erature t

i TI 5047 Indicates Torus Water Indicator control Room Temperature Panel C7 I

t i

TI 5048 Indicates Torus Water Control Room l Indicator j Temperature Panel C7 I

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i TRU 9045 Records drywell Temperatures Indicating I Control Room .

Recorder Panel 903 i

I TRU 9045 Records drywell pressure Indicating Control Room Recorder Panel 903

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Instrument Loop Use Type hacation Pall 9046 Annunciates high drywell Alarm Unit Control Room pressure Panel C6 PI 9046A Indicates drywell pressure Indicator Control Room Panel C6 i

P1 9046B Indicates drywell pressure Indicator Control Room Panel 903 I

TRU 9044 Records Drywell Temperature Indicating Control Room Recorder Panel 903 6

I

TRU 9044 Records Drywell Pressure Indicatin8 Control Room 1

Recorder Panel 903 N

O N

, ps3 LR 5038 -

Records Torus Water Level

, Indicating Control Room Recorder Panel 903 O ,

tm LR 5049 Records Torus Water Level Indicating Control Room Recorder Panel C7

Page 4 of 6' -

Instrument Loop ,

Use Type Location LS 5066 Annunciates Low Torus Water Level Switch Torus Compartment Level Alarm at Control Room Panel 904 LS 5037 Annunciates Low Torus Water Level Switch Torus Compartment Level Alarm at Control Room Panel 903 Level Switch Torus Coppartment >,

LS 2351A Annunciates liigh Torua Water Level Alarm at Control Rocx Panel 903 i

i i LS 2351B Annunciates liigh Torus Water Level Switch Torus Compartment i Level Alarm at Control Room Panel 903 l

PS 512 A+D Initiate Scrams on drywell Pressure Switdi Reactor Building high pressure Alarm at Control Rooa i)rs Panci 905 C:3 -

N N

PS 1001 90A*D '

Initiate Safeguards on Pressure Switch Reactor Building 123 drywell high pressure Alarm at Contiol Roon Panel 903

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Permissive for Drywell Pressure Switch Reactor Building PS 1001 83A*D Alarm at Control Room

, Spray !! ode of RIIR on high drywell pressure Panel 903

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natrument Loop y,, Type Location vS 1001 89A*D Initiating signal on ADS Pressure Switch Reactor Building system on high drywell , Alarm at Control

. Pressuta Room Panel 903 C-19 Annunciates and records Continuous Portable Within

, drywell radiation conditions Sampling by Reactor Building Pumps

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C-41 Records 02 Level in Drywell Continuously rotating Panel C-41 samples of (4) drywell and (3) torus Points and (1) h Bldg. atmosphere Ilydrogen Analyzer "A" Records llydrogen Concentra- Continuous Sampling Panel C-118 tion in drywell Recorder on Panci 902 1

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Ilydrogen Analyzer "B" Records llydrogen Concentra- Continuous Sampling Pan-1 C-119

.p) tion in drywell Recorder on Panel 902

' CD N

rs) , ,

CE? TR 302-94 Records CRD temperatures and Indicating Reactor Building LJ7 -

annunciates on high tempera- Multipoint Annunciator on Panel L/4 tures Recorder 905 in Control Room TR 260-20 Records relief and safety Indicating Control Foom valve discharge termperature Multipoint Panel 921 Recorder e

Page 6 of 6 -

9 Instrument Loop usa Type Eocation

, 1 i.

TR 260-151 Records suction temperature Indicating Control Room

. for both recirc loopa Recorder Panel 904 j

I -

e TI 260-151A Indicates recirc suction Indicator Control Room temperature - A loop Panel 904 8

1 i

TI 260-151-B Indicates recirc suction Indicator Control Rcom temperature - B loop Panel 904 t

! TR 263-104 Vessel temperatures Indicating Control Room I Recorder Panel 921 i - -

TR 263-105 Recorda (2) points on Indicating Control Room vessel for temperature Recorder Panel 904 i

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