ML19320B537
| ML19320B537 | |
| Person / Time | |
|---|---|
| Issue date: | 06/13/1980 |
| From: | Pappas H NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Wright G ILLINOIS, STATE OF |
| References | |
| NUDOCS 8007140421 | |
| Download: ML19320B537 (2) | |
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NUCLEAR REGULATORY COMMISE!ON 3
aE REGloNlil b[
799 ROOSEVELT ROAD Q
GLEN ELLYN, ILLINOIS 6o137 o
JUN 131980 State of Illinois
-Department of Public Health ATTN:
Mr. Gary N. Wright, Chief Division of Nuclear Safety 535 West Jefferson Street Springfield, IL 62761 Gentlemen:
The enclosed IE Bulletin No. 80-14 titled " Degradation of Scram Discharge Volume Capability" was sent to the following licensees on June 12, 1980.
Cincinnati Gas & Electric Company Zimmer (50-358)
Cleveland Electric Ill. Company Perry 1, 2 (50-440, 50-441)
Commonwealth Edison Company Dresden 1, 2, 3 (50-237, 50-249)
LaSalle 1, 2 (50-373, 50-374)
Quad-Cities 1, 2 (50-254, 50-265)
Consumers Power Company Big Rock Point (50-155)
Illinois Power Company Clinton 1, 2 (50-461, 50-462)
Iowa Electric Light & Power Company Duane Arnold (50-331)
Northern Indiana Public Service Company Bailly (50-367) 8007140 M
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State of' Illinois Northern States Power Company Monticello (50-263)
Sincerely, b
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. Helen Pappas, Chief Administrative Branch
Enclosure:
IE Bulletin No. 80-14 cc w/ encl:
Mr. D. W. Kane, Sargent & Lundy_
~ Central Files
' Reproduction Unit hTC 20b Local PDR NSIC TIC i.
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UNITED STATES 8005050056 NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT mn0.N(( 7 WASHINGTON, D.C.
20555 a d L% M b L _G June 12, 1980 IE Bulletin No. 80-14 DEGRADATION OF BWR SCRAM DISCHARGE VOLUME CAPABILITY During our review of BWR operating experience, two events have raised concern on operations related to the control rod drive system gcram discharge volume (SDV).
Description of Circumstances:
At Hatch Unit 1, on June 13, 1979, while performing surveillance to func-tionally test SDV high level switches, two switches (C11-N013A, B) were found to be inoperable.
Redundant switches (C11-N013 C, D) were operable.
The reactor was in the refuel mode and these switches had been modified prior to this occurrence.
Inspection of the inoperable level switches revealed that the float rod was bent and binding against the side of the float chamber on both switches.
The licensee believes that the float rods were bent during or prior to initial installation and that metal particles from the modification caused binding of the float.
(LER 79-038)
-Brunswick 'Jnit 1 reported that slow closure of the SDV drain valve during a reactor scaam on October 19, 1979 apparently caused a water hammer event which damaged several pipe supports on the SDV drain line.
Drain valve closure time was approximately five minutes due to a faulty solenoid controlling air sup.~j to the valve.
The damaged pipe supports were repired but repair parts for the faulty solenoid were not available.
To prevent possible damage from a scram, the unit started up with the SDV vent and drain valves closed except for periodic draining.
During this mode of operation the reactor scrammed from high level in the SDV, without prior actuation of either the high level alarm or rod block switch.
Subsequent inspection revealed that the float ball on the rod block switch was crushed and the float ball stem on the high level alarm switch was bent such that the switches would not operate.
The water hammer event discussed above was the reported cause of failure of these two switch assemblies.
(LER 79-74)
As a result of these events and related anticipated transients without scram (ATWS) studies, concern arises that the SDV function may be degraded by the undetected presence of fluid in the SDV.
The second event is significant in that it indicates the potential for a common cause failure (faulty solenoid) to result in operation of the SDV in a manner which could defeat both the level switch function and the SDV draining function.
The ATWS generic studies (NUREG 0460) have led the staff to propose, among other requirements, improve-ments in the SDV' designs to reduce susceptibility to common cause failures.
By separate correspondence, the staff will provide example Technical Specifica-tions related to the action items discussed below.
-g.
IE Bulletin No. 80-14 June 12, 1980 Fage 2 of 2 A.
GE BWR's With an Operating License The following actions are to be taken by licensees of GE designed EWR facilities with'an operating license:
1.
Review plant records for instances of degradation of any SDV level switch
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which was or may have been caused by a damaged or bent float assembly.
Identify the cause and corrective action for each instance.
2.
Review plant records for instances of degradation of SDV vent and drein valve operability.
Provide the closure times required and typically observed for these valves and the basis for the required closing times.
Identify the cause and corrective action for each instance of degradation.
3.
-By procedures, rcauire that the SDV vent and drain valves be normally operable, open and periodically tested.
If these valves are not operable or are closed for more than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period during operation, the reason shall be logged and the NRC notified within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> (Prompt Notification).
4.
Review instances in which water hammar or damage which may have been caused by water hammer has occurred in SDV related piping.
Identify the cause and corrective action for each instance.
5.
Review surveillance procedures to ensure that degradation of any SDV level switch due to a damaged float or other cause would be detected and that inoperability from any cause would be reported to the NRC.
6.
If no functional test or inspection which would detect degradation of each SDV level switch has been performed during the past 3 months, make provisions to perform an inspection and functional test of all SDV level-switch assemblies at the next reactor shutdown of greater than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> duration.
B.
Reporting Requirements The action taken in response to the items in Part A shall Le completed and a written report on the results submitted to the NRC 'ithin 45 days from the date of this Bulletin.
This report should be submitted to the Director of the appropriate NRC Regional Office and a copy should be forwarded to the NRC Office of Inspection and
-Enforcement, Division of Reactor Operations Inspection, Washington, D.C.
20555.
Approved by GAO, B180225 (R0072); clearance expires 7-31-80.
Approval was given under a blanket clearance specifically for identified generic problems.
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- D IE Bulletin No. 80-14 Enclosure June 12, 1980 RECENTLY ISSUED IE BULLETINS Bulletin Subject Date Issued Issued To No.
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80-13 Cracking In Core Spray 5/12/80 All BWR's with an Spargers OL 80-12 Decay He'at Removal System 5/9/80 Each PWR with an OL Operability 80-11 Masonry Wall Design
~5/8/80 All power reactor facilities with an OL, except Trojan 80-10 Contamination of 5/6/80 All power reactor i
Nonradioactive System and facilities with an Resulting Potential for OL or CP U.1 monitored, Uncontrolled Release tc Environment 80-09 Hydramotor Actuator 4/17/80 All power reactor Deficiencies operating facilities and holders cf power reactor construction permits 80-08 Examination of Containment 4/7/80 All power reactors with
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Liner Penetration Welds a CP and/or OL no later than April 7, 1980 80-07 BWR Jet Pump Assembly 4/4/80 All GE BWR-3 and Failure BWR-4 facilities with an OL 80-06 Engineered Safety Feature 3/13/80 All power reactor (ESF) Reset Controls facilities with an OL 1
80-05 Vacuum Condition Resulting 3/10/80 All PWR power reactor In Damage To Chemical Volume facilities holding Control System (CVCS) Holdup OLs and to those with Tanks a CP l
79-01B Environmental Qualification 2/29/80 All power reactor of Class IE Equipment facilities with an OL 80-04 Analysis of a PWR Main 2/8/80 All PWR reactor facilitie Steam Line Break With holding OLs and to those Continued Feedwater nearing licensing Addition