ML19242D143

Responds to Request for Addl Info Re IE Bulletin 79-08. Design of BWR Containment Is Such That All Lines Penetrating Containment Have Valves Which May Be Isolated

ML19242D143
Person / Time
Site:	Cooper
Issue date:	08/07/1979
From:	Pilant J NEBRASKA PUBLIC POWER DISTRICT
To:	Ippolito T Office of Nuclear Reactor Regulation
References
TAC-13206, NUDOCS 7908140549
Download: ML19242D143 (9)

Text

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- - COOPER NUCLEAR STATtoN P.o. Bo X 98, BROWNVILLE, NE BR ASKA 68 321 Nebraska Publ.ic Power D. is tr.

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.- 2-August 7, 1979 Mr. Thomas A. Ippolito, Chief Operating Reactors Branch No. 3 Division of Operating Reactors U.S. Nuclear Regulatory Commission Washington, D.C. 20555

Subject:

IE Bulletin 79-08, Request for Additional Information

Reference:

Ippolito to Pilant letter dated July 20, 1979-

Dear Sir:

This letter is written in response to the referenced letter requesting additional information concerning our response to IE Bulletin 79-08 dated April 25, 1979. The following items address the items as listed in the referenced letter.

Question Teen 30, i

1. Confirm that the review of item 1 of IEB 79-08 by all licensed operators and plant management and supervisors with operational responsibilities has been documented in your plant records.

Response

1. The review of item 1 of IE Bulletin 79-08 by all licensed operators and plant management and supervisors with operational responsi-bilities has been documented in plant records.

Question It_am No. 2

1. Your response indicates that ycu reviewed containment isolation of all valves whereas the Bulletin refers to all lines. Confirm that your review considered isolation of all lines penetratirg containment.

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'r. Thomas A. Ippolito August 7, 1979 Page 2.

p.sponse

1. The design of the B'4R contain=ent is such thar all lines pene-trating containment have valves which may be isolated as required.

Our previous response stating our review of all containment iso-lation valves i= plied we reviewed isolation of all lines pene-trating containment.

Question Item No. 2

1. For the manual actions related to restart and continued RCIC op-eration, and for any other manual actions required, specify whether these actions are addressed by written procedures.

Response

1. The manual actions related to restart and continued RCI.' operation are addressed in written procedures.

Question Item No. 4

1. Your response is incomplete. Describe all uses and types of vessel level indication for both autonctic and manual initiation of saf aty systems.

2. Your response is incomplete. In addition, describe other instru-mentation which the operator might have to determine changes in reactor coolant inventory, e.g. , drywell high pressure, radio-activity levels, suppression pool high temperature, containment sump pump operation, etc.

3. Clarify your response to indicate whether operators have beer.

instructed to utilize other available information to initiate safety systems. Provide your schedule for completion of this action.

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Mr. Thomas A. Ippolito August 7, 1979 Page 3.

Respense

1. Reactor Water Level Instrumentation

a. LIS-57 A&B (Yarway)

Location: Rack 25-5 Range: -150" to +60" (366.75" to 576.75" vessel height)

Function: Trip RR Pumps, Close MSIV, and Closes Reactor Water Sample Valves (>-37")

b. LIS-58 A&B (Yarway)

Location: Rack 25-6 Range and Function same as LIS-57 A&B

c. LIS-72 A,B,C,D (Yarway)

Location: Racks 25-5 & 25-6 Range: ~150" to +60" Function: Initiate RCIC & HPCI at (>-37")

RHR, CS, DG & ADS (>-145.5")

d. LIS-83 A&B (Yarway)

Location: Racks 25-5 & 25-6 Range: 0 to 60" (516.75 to 570.75" vessel heighc)

Function: ADS Permissive (>12.5")

e. LITS-59 A&B (Yarway)

Location: Racks 25-5 & 25-6 Range: -150" to +60" Function: Feeds LI-85 A&B on Panel 9-5, Control Room

f. LITS-73 A&B (Yarway)

Location: Racts 25-51 and 25-52 Range: -100" to +200" (252.56" to 552.56" vessel height)

Function: RHR 2/3 core height interlock (1-39")

Feeds LI-91 ASB, Panel 9-3

g. L1-70 (Gemac)

Location: Racks 25-52 Range: 0" to 100" (516.75" to o16.75" vessel height)

Function: Feeds LR-98, Panel 9-5, Control Room L ,S g ,[s?.w"

Mr. Thomas A. Ippolito August 7, 1979 Page 4.

h. LT-52 A,B,C (Gemac)

Location: Racks 25-5 & 25-6 Rande: 0" to 60" Function: Feedwater Control, Main TG & RFFT Hi level trips Feeds LI-94 A,B,C on Control Room Panel 9-5 LR-97 on Control Room Panel 9-5

1. LT-61 (Gemac)

Location: Local Rack 25-5 Range: 0" to 400" (516.75" to 916.75" vessel height)

Function: Feeds LI-86 on Control Room Panel 9-4

j. LIS-101 A,B,C,D (Gemac)

Location: Racks 25-5 & 25-6 Range: 0" to 60" Function: Reactor Scram, Primary Containment Isolation at >12.5". Trips ROIC & HPCI at <58.5"

2. The following instrumentation could assist the operator to deter-mine changes in reactor coolant inventory:

a. Drywell equipment and floor draic cump flow recorders are provided in control room.

b. Drywell equipment su=p te=perature indicator is provided in the control room.

c. Mismatch between reactor feedwater flow and steam flow re-corders and indicators.

d. The suppression pool water level is detected by level switches, indicators and transmitters. Three suppression pool water indicators and one recorder are supplied in the control room on Panci 9-3. The recorder is a dual channel recorder, PC-LR-11 torus level, and PC-PR-512A drywell pressure. LI-10 pro-vides torus indication over a range of -4.4 ft. to +6 ft. LI-12 provides narrow range indication over a rance of -10 in. to

+10 in. LI-13 provides narrow range indication over a range of -5 in, to +10 in. Should suppression chamber level reach a level of +5 inches two alarms will be annunciated. One from LT-12 " suppression chamber level high/ low" which is activated at +2 in. (high) and -1 in. (low). The other from LT-10

" suppression chamber level high/ low" which is activated at +5" (high) and -5" (low).

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Mr. Thomas A. Ippolito August 7, 1979 Page 5.

e. Three primary containment and one wetwell pressure indications are provided in the control room.

f. Primary containment internal temperature is detected by 38 temperature elements of which four are used for wetwell pool temperature.

g. Drywell process radiation monitor which monitors particulate, gaseous, and iodine activities, plus provides the capability of a grab sample,

h. Main steam line high radiation and main steam line high flow alarms.

1. Rene.*.or water cleanup high flow alart J. High area temperatures (steam leak detection) alarms.

3. Operators have been instructed to utilize all available infor-mation to initimte safety systems. This is documented in plant training records.

Question Item No. 6

1. It is not clear from your response that safety related "alve po-sitioning requirements were reviewed to ensure proper operation of engineered safety features. Please supplement your response to provide a commitment to conduct this review and a schedule for completion.

2. Please augment your response to indicate the extent to which po-sition and locking device checks are performed for locked safety system valves.

3. 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 information.

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Mr. Thomas A. Ippolito August 7, 1979 Page 6.

Response

Item No. 6

1. Prior to our April 25, 1979 response to IEB 79-08, our safety related valve positioning requirements were reviewed to ensure proper operation of engineered sateguard feature systems. An I&E inspection team also independently inspected and performed a de-tailed review of the engineered safeguard feature systems on May 1-4 and 14-15, 1979. This inspection is described in IE Inspection Report 50-298/79-10. Two minor discrepancies were noted and have since been corrected.

2. The following excerpt is taken from our administrative procedure concerning valve seals:

"The Valve Seals Log shall contain a list by system of the normally locked open/ closed valves. These valves shall be sealed in such a manner as to prevent their being operated without destroying the seal. The seal will consist of a lead-wire, numerically numbered seal.

The Shift Supervisor will maintain the Valve Seals Log.

Whenever a seal is broken for any reason, the Shift Supervisor shall be notified and the seal returned to him. He will then cross out the seal number as listed in the log and record the valve with the missing seal until the seal is replaced in the Shift Supervisor's log as a red arrow item.

Upon return of the valve to its normal locked open or closed position, the Shift Supervisor shall be notified. He will then issue a new seal with the next higher number for that sarticular system and list it in the Valve Seal Log. Each system is issued a block of seals in numerical order."

The valve seals broken during surveillance testing or maintenance are replaced and a specific sign off is required. All seals are checked as part of the valve lineups performed after an extended outage or at the discretion of the Operations Supervisor.

3. .ollowing our April 1979 outage which was in progress at the time of our initial response to IEB 79-08, all safety related valve positions were verified to be in the proper position. As described above, this position verification was independently checked by an ISE inspection team.

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Mr. Thomas A. Ippolito August 7, 1979 Page 7.

Question Item No. 7

1. In your discussion regarding systems designed to transfer radio-active gases and liquids, no explicit discussion is presented regarding valve action or resetting safety features instrumen-tation. Provided assurance that inadvertent transfer of radio-active gases or liquids out of containment will not occur or re-setting safety feature instrumentation.

Response

1. At th. ;; resent ti=e our procedures caution the operator to ensure that the problem that caused the containment isolation has been rectified, prior to resetting the instrumentation. We will review our procedures and update them as necessary to provide additional guidance concerning parameters which should be checked prior to resetting isolation instrumentation. This review and update will be completed by September 15, 1979.

Question Item No. 8

1. We understand from your response that operability is verified for redundant safety related systems prior to removal of aa> safety related system from service. Since you may be relying on prior operability verification within the cr rent technical specification surveillance interval, operability should be further verified by at least a visual check of the system status to the extent practic-able, prior to removing the redundant equipment f rom service.

Please supple =ent your response to provide a cocmitment that you will revise your maintenance and test procedures to adopt this position.

2. It is not clear from your response that all involved reactor op-erational personnel in the oncoming shif t are explicitly notified about the status of systems removed from or returned to service.

Please indicate how this information is transferred at shif t turn-over.

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Mr. Thomas A. Ippolito August 7, 1979 Page 8.

Response

1. The Technical Specifications at Cooper Nuclear Station require that redundant systems be tested ic=ediately and periodically thereafter when a system is made or found inoperable. When a system is to be made inoperable for maintenance, station procedures require that the redundant systems be verified operable uith an actual system test prior to removal of that system. This generally involves a visual. inapection of the equipment being tested, however the con-trol room operator doing the testing has enough information im-mediately available to him to determine system operability.

2. The following items are required by our procedures as a minimum turnover for all licensed operators:

a. Review status of plant and operating procedures in progress.

b. Review safety system status panel.

c. Review respective log back to their last working shift.

d. Review surveillance tests in progress.

e. Significant changes in routine operation which have occurred during the last two shif ts.

f. Review any abnormal circumstances.

The shif t supervisor must also review all maintenance work orders in progress, review any new special orders issued since the last working shift, and review the Night Orders Log.

Events such as removing from or returning to service of systems are typical entries required by procedures to be made in the Control Room and Shift Supervisor's Logs. In addition, the safety system status panel is provided on the reactor operator's panel. This panel is checked at each shift turnover and it indicates whether an engineered safety feature system or loop of that system is in service, in test, or out of service. Our test procedures have specific steps in them through which the status of the panel is updated.

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Mr. Thomas A. Ippolito August 7, 1979 Page 9.

Question Item No. 9

1. Amend your response to provide assurance that your procedures stipulate that NRC will be notified any time the reactor is not in a controlled or expected condition of operation.

Response

1. Our station procedures stipulate that NRC will be notified any time the reactor is not in a controlled or expected condition of operation.

Sincerely,

. . Pilant Director of Licensing and Quality Assurance JMP:PJB:cg

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