ML18005A306
| ML18005A306 | |
| Person / Time | |
|---|---|
| Site: | Harris  |
| Issue date: | 02/26/1988 |
| From: | Watson R CAROLINA POWER & LIGHT CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| CON-NRC-594 HO-880062-(O), NUDOCS 8802290266 | |
| Download: ML18005A306 (14) | |
Text
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
CCESSION NBR: 8802290266 DOC. DATE: 88/02/26 NOTARIZED:
NO FACIL: 50-400 Shearon Harris Nuclear PoUjer Plant.
Unit i. Carolina AUTH. NAME AUTHOR AFFILIATION WATSON> R. A.
Carolina PoUjer 5 Light Co.
RECIP. NAME RECIPIENT AFFILIATION.
Document Control Branch (Document Control Desk>
R
SUBJECT:
Responds to NRC 871110 ltr re violations. noted in Insp Rept 50-400/87-37 0 50-400/87-40. Corrective actions: operations supervisor 8c manager operations counseled re need to be sensitive to significant events.
DOCKET ¹ 05000400 DISTRIBUTION CODE:
IEOID COPIES RECEIVED: LTR ENCL I
SIZE; TITLE: General (50 Dkt)-Insp Rept/Notice of Violation Response NOTES:Application for permit reneeal filed.
05000400 RECIPIENT ID CODE/NAME PD2-1 PD COPIES RECIPIENT LTTR ENCL ID CODE/NAME 1
1 BUCKLEY'S B COPIES LTTR ENCL 2
2 INTERNAL: ACRS DEDRO NRR/DLPG/PEB118 NRR/DOEA DIR11E NRR/DREP/RPB10*
NRR /P MAS/ILRB 12 OGC 1 5-B-18 RES/DRPS DIR 2
2 1
1 1
1 1
2 2
1 1
1 1
1 1
AEOD NRR MORISSEAUz D NRR/DLPG/GAB10A NRR/DREP/EPB10D NRR/DRIS DIRPA2 HEMG~Y.LS RQN2 FILE 01 1
1 1
1 1
1 1
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1 EXTERNAL:
LPDR NSIC 1
1 NRC PDR 1
1 1
1 TOTAL NUMBER OF COPIES REQUIRED:
LTTR 24 ENCL 24
Carolina Power & Light Company HARRIS NUCLEAR PROJECT P.
O.
Box 165 New Hill, North Carolina 27562 XEB 26 198g File Number.'SHF/10-13510E Letter Number'HO-880062 (0)
NRC-594 Document Control Desk United States Nuclear Regulatory Commission Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT DOCKET NO. 50-400 LICENSE NO. NPF-63 REPLY TO A NOTICE OF VIOLATION Gentlemen:
In reference to your letter of November 10, 1987 and Notice of Violation dated January 29, 1988, referring to I.E. Reports RII:
50-400/87-37 and 50-400/87-40, Attachments 1
and 2 is Carolina Power
& Light Company's reply to the violations identified as "A"
and "B" in Enclosure 1.
Attachment 3 provides the additional information requested by your January 29, 1988 letter concerning operability determinations.
It is considered
. that the corrective actions taken are satisfactory for resolution of the item.
Thank you for your consideration in this matter.
Very truly yours, 88022q02bb 5000000 PDR hooch 0 PDR G
R. A. Watson Vice President Harris Nuclear Project MGW'ddl Attachment cc'.
Messrs.
B. C. Buckley (NRC)
G. Maxwell (NRC-SHNPP)
Dr. J. Nelson Grace (NRC)
+go MEM/HO-8800620/1/OS1
ATTACHMENT 1
Response
to NRC I.E. Report RII:
50-400/87-37 Violation "A" Re orted Violation'.
Technical Specification 6.8.la requires that written procedures be implemented covering the procedures outlined in Appendix A of Regulatory Guide 1.33, Rev.
2, February 1978.
Administrative Procedures are identified in Appendix A of the Regulatory Guide.
An Administrative Procedure titled "Operations Conduct of Operations" OMM-l, Rev.
3, Step 3.2.3.4, requires the shift foreman to ensure the safe operation of the plant-at all times.
Contrary to the
- above, On October 9,
1987 OMM-1 was not implemented, in that the shift foreman authorized control room operators to manipulate reactor coolant vent valves RC-900, RC-901, and RC-902 with the knowledge that operating these valves probably would cause the uncontrolled opening of RC-904 and RC905.
The vent valves were individually operated and in each instance RC-904 and RC-905
- opened, thus allowing a vent path for reactor coolant to fl'ow into the containment atmosphere and into the pressurizer relief tank.
This is a Severity Level IV violation (Supplement I).
Denial or Admission and Reason for The Violation.'he violation is correct as stated.
On October 9,
- 1987, the plant was in Mode 1
at 91K power.
Operations Surveillance Test (OST)
- 1043, Reactor Coolant System Vent Path Operability, was scheduled to be performed.
This test procedure, required by the In-Service Inspection (ISI) Program to be performed once per 92 days, strokes and times the valves in the Reactor CooLant System (RCS)
Head Vent System.
The system flow drawing at the time is shown in Figure A.
On October 9,
- 1987, testing was being conducted on some of the valves more frequently than each 92 days as required by the ISI Program due to increases in the valve stroke times in previous testing.
Testing commenced at approximately
- 0500, when valve LRC-904, (Manufacturer.'Target
- Rock, Model No.'79Q-017) vent path to containment atmosphere, was satisfactorily cycled.
The next valve tested was LRC-900, one of two vent valves from the reactor vessel head.
When this valve was
- opened, valve 1RC-904 was observed to spuriously open.
This created an open path from the RCS to the containment atmosphere, so the operator immediately closed 1RC-900.
Valve 1RC-904 reclosed immediately when 1RC-900 was closed.
RCS pressure was observed to decrease slightly, by one or two psig, but no change in pressurizer level was noted.
The Shift Foreman was not consulted about the opening of 1RC-904.
MEM/HO-8800620/1/OS1
0
Due to the unexpected opening of 1RC-904 and the prompt closing of 1RC-900, a
response time for 1RC-900 was not obtained.
Valve 1RC-900 was reopened at 0503 to obtain a closure time',
on this
- attempt, 1RC-904, opened
When the operator attempted to reclose 1RC-900, it did not respond.
Prior to this point the Shift Foreman was unaware of the vent valve malfunctions.
All three valves remained open until their control switches were simultaneously placed in PULL TO LOCK position, interrupting power to the solenoid; after several
- seconds, 1RC-900
During this evolution, a
leakage path existed from the RCS vessel head into the PRT and to containment atmosphere.
Pressurizer level decreased approximately 3X and pressure dropped to 2210 psig.
Shift personnel discussed the event to determine if it was safe to continue testing.
Throughout the discussion, the Shift Foreman was unaware of the results from the first cycle of 1RC-900, which occurred at approximately 0500.
Shift, personnel were unsure as to whether the observed problem was specific to valve 1RC-900 or was generic to all the valves in the head vent system.
It was believed that the valves could be relied on to reclose when their control switches were placed in PULL TO LOCK.
- Also, a scheduled outage would commence the next day, and there would be no further opportunity to obtain data prior to shutdown.
Since such data would be needed to support any repairs, and since the valves could be reclosed if the event were to reoccur, limiting RCS leakage to a
few
- seconds, the decision was made by the Shift Foreman to resume testing to obtain this data.
Three additional openings of the valves were performed.
Valve 1RC-900 was opened at
- 0516, valve 1RC-901 at
- 0519, and valve 1RC-902 at 0521.
In each
The causes of the RCS vent valve malfunctions and corrective actions are discussed further in Shearon Harris Nuclear Plant Licensee Event Report 87-058-01.
Corrective Ste s Taken and Results Achieved:
An estimate of RCS leakage concluded that the limits of Technical Specification 3.4.6.2 (Identified leakage less than 10 gpm) were
- exceeded, with the majority of the leakage occurring at 0503, when 1RC-900 was opened the second time.
A total of 198 gallons of coolant was lost, with 120 gallons discharging into the
- PRT, and the remainder entering the containment.
At 0610, an unusual event was declared in accordance with the plant emergency plan due to exceeding the Technical Specification limits.
When the valves were successfully
- closed, RCS identified leakage returned to less than that allowed by Technical Specification 3.4.6.2.
MEM/HO-8800620/2/OS1
Corrective Ste s Taken to Avoid Further Noncom liance:
1.
Personnel have been briefed on the design of Target Rock
- valves, and the significance of continuing an evolution which raises a safety concern.
This briefing was conducted prior to personnel assuming licensed duties for their shift.
The Shift Foreman who was on duty during the event was counseled by plant management regarding the seriousness of the event.
In addition, the Shift Foreman participated in the subsequent investigation of the event.
2.
The Operations Supervisor and Manager-Operations were counseled by the Plant General Manager regarding the need to be sensitive to significant events and their implications to plant safety.
3.
The requirement to test these valves at normal operating pressure and temperature has been changed.
On February 12, 1988 the NRC issued Amendment No.
4 to Harris Operating License NPF-63.
The amendment deletes Surveillance Requirement 4.4.11.1, which requires quarterly testing of the reactor coolant system (RCS) vent path block valves and modifies Surveillance Requirement 4.4.11.2b to include the testing of the block valves at least once every 18 month interval.
On December 22, 1987 CP6L revised the ISI Program to require testing of these valves only at cold shutdown.
The NRC was informed of this change by a letter dated December 4, 1987.
Date When Full Com liance Was Achieved:
Full compliance was achieved on February 12,
- 1988, upon completion of the actions as stated above.
MEM/HO-8800620/3/OS1
Document Control Desk Page 4
Figure A
REACTOR VESSEL HEAD VENT SYSTEM NRC-594 SAMPLINGLlt+
1RC-902 CONTAINMENTATt10SPHERE IRC-905 S
IRC-QH PRZR 9s IRC-900 1RC-00S 1RC&01 S
PRZR RELIEF TANK RY HEAD Ref. Dug.'CPL - 2165-S-1301 MEM/HO-8800620/4/Osl
ATTACHMENT 2
Response
to NRC I.E. Report RII:
50-400/87-40 Violation "B" Re orted Violation:
- 10CFR50, Appendix B, Criterion XVI, FSAR Section 17.2.16 and CPGL Corporate Quality Assurance
- Program, Section 15.2.5 collectively require that the licensee take prompt corrective action on conditions adverse to quality.
Contrary to the
- above, prompt corrective action on conditions adverse to quality was not taken in that, in June
- 1985, and February
- 1987, the licensee experienced spurious opening of Reactor Coolant Vent System valves during testing.
Both these problems are conditions adverse to quality and were not corrected until October
- 1987, which resulted in a loss of reactor coolant inventory on October 9, 1987.
This is a Severity Level IV violations (Supplement I).
Denial or Admission and Reason for The Violation:
The violation is correct as stated.
On October 9,
1987 the plant was in Mode 1 at 91/ power.
While performing Operations Surveillance Test (OST)
- 1043, Reactor Coolant System Vent Path Operability, an unexpected release of reactor coolant to the Containment atmosphere and to the Pressurizer Relief Tank (PRT) occurred.
The release was the result of a
malfunction of the block valve (1RC-904) to containment atmosphere and the block valve (1RC-905) to the PRT.
During performance of OST-1043, on October 9, 1987 when either of two Reactor Vessel Head Vent Valves (1RC-900, 1RC-901) or a
Pressurizer Vent Valve (1RC-902) were cycled open, the downstream block valves 1RC-904 and 1RC-905 would spuriously open creating an open path from the Reactor Coolant System (RCS) to the containment atmosphere and the PRT.
During investigation of the above incident the following similar incidents and related information were revealed.
In February
- 1987, at SHNPP a
similar event occurred during performance of OST-1043.
In this event, the downstream valves were able to be immediately reclosed, and no indications of RCS inventory loss were noted.
Position indication for one of the valves was lost during the test, so testing was suspended.
Upon repairs to the indication circuit, testing was satisfactorily completed.
MEM/HO-8800620/5/OS1
In resolving the February 1987 event, no specific problem with the valves or electric control circuits could be found by the maintenance workers.
A maintenance Feedback Request was generated to investigate the situation',
- however, since the work request did not identify the safety significance of the
- problem, a
low priority was assigned to the item.
The test procedure (OST) was satisfactorily completed on May 25, 1987 and August 25, 1987, with the plant at power, and no problems with spurious valve operation were reported.
Previous industry experience with this phenomena was known to the vendor (Target Rock).
This information was not supplied to the Shearon" Harris Project in 1984 when the project requested verification and validation of the Valve Technical Manual.
Again in April 1987, the project requested verification and accuracy of'urrently available information.
Target Rock replied in June 1987 that the information was current and accurate.
Information pertaining to the valve unseating problems was not included.
CP&L's H.
B.
Robinson Plant experienced similar problems with their Target Rock RCS vent valves in 1980.
The problem was subsequently corrected by inverting the valves as recommended by the vendor (Target Rock).
During this time period CP6L had no effective program for disseminating information of this type between its nuclear plants.
Therefore, the problem and correction at H.
B.
Robinson was not incorporated into the Shearon Harris design.
A related valve response was observed at SHNPP on December 22,
- 1985, during preoperational testing during fill and vent of the RCS for hot functional testing.
Since the event occurring during preop testing was of concern only during depressurized conditions, it was resolved by incorporating-a precaution into the RCS filland vent procedure (GP-008) regarding the tendency of these valves to open when subjected to the back pressure normally present in the PRT while the RCS is depressurized.
An evaluation of this event concluded that there was no deficiency since the valves were designed to normally be pressurized on the top of the valve disc tending to keep the valves shut.
Corrective Ste s Taken and Results Achieved:
A Target Rock vendor representative was brought on-site following this event (October 9,
1987) to assist in the investigation and corrective action.
Target Rock advised that the unseating phenomena will not occur when the space above the valve disc is maintained full of liquid, or when the pressure transient is sufficiently slow to allow the pressure above the valve disc to equalize with the inlet pressure.
As a result, test procedures for these 'alves were revised to change the sequence by which the valves are tested to help ensure the valves remain filled with fluid.
MEM/HO-8800620/6/OS1
The design of the vent system was changed to reorient the valves per the vendor recommendation (e.g.,
block valves inverted such that the disc cavity remains full at all times).
A check valve was added to correct the problem observed during the December 22, 1985 event.
See Figure B.
Subsequent to repairs, all valves were tested satisfactorily at Mode 5, Cold Shutdown, conditions.
Corrective Ste s Taken to Avoid Further Noncom liance'.
A thorough investigation of the event was conducted to determine why previous similar experiences with these valves were not considered in the plant design, and appropriate design.
A review of outstanding work items was conducted to determine if other items existed which represented a challenge to safe plant operations, but were not given proper priority.
No similar items were found.
The technical manual for these valves has been appropriately updated.
A review of other applications of Target Rock valves in the plant has been conducted.
While similar behavior is possible in the other applications, the consequences are acceptable, and have been determined not to present a safety concern.
To ensure that important deficiencies get identified and appropriately pursued work lists are screened at a
greater frequency, each normal work
- day, a
list of work tickets (deficiencies) identified in the past 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is provided to a
number of management personnel for review.
This includes the System Engineers, Maintenance
- Manager, Operations
- Manager, Technical Support
- Manager, and the Plant General Manager.
Additionally, System Engineers frequently review backlogged work for their system to identify items of importance.
Subsequent to the 1980 event at the H. B. Robinson plant which was not made known to
- SHNPP, the following actions have taken place which should preclude similar oversites in the future'.
a ~
An engineering Central Design Organization (CDO) has been formed in the corporate office with responsibility for configuration control for all CPSL nuclear plants.
- Thus, information from one project with applicability to another project will be screened by the same CDO group.
b.
Each nuclear site has an active operational experience feedback (OEF) system which reviews events from a
variety of sources such as the
- NRC, INPO OEF
- System, Nuclear
- Network, Plant Incident
- Reports, other plant LERs, etc.
MEM/HO"8800620/7/OS1
c.
Significant LERs and Incident Reports from CP&L nuclear plants are distributed to the other nuclear plants for possible applicability.
Date When Full Com liance Was Achieved:
Full compliance was achieved on January 14, 1988 upon completion of the actions as stated above.'EM/HO-8800620/8/OS1
Document Control Desk
'age.
9 Figure B
REACTOR VESSEL HEAD VENT SYSTEM NRC-594 AFTER MODIFICATION SAMPLING Lith 1RC-902 CONTAINt1ENT ATt10SPHERE PORV 1RC-903 1RC 304 BLOCK VAI.VES IHVERl cD Qs I
1RC&00 1RC&05 CHECK VALVEADDED TO PREVENT BACKPRESSURE ON 1RC&05 1RC&01 PRZR RELIEF TAIIK RV H-"AD Ref.
Dwg:
CPL-2165-S-1301 MEM/H0-8800620/9/OS1
0 0
Document Control Desk Page 10 ATTACHMENT 3 NRC-594 In the cover letter transmitting the Notice of Violation, the NRC requested an evaluation of CP&L's process for making operability determinations.
The primary means used to determine operability is the applicable surveillance test which identifies specific acceptance criteria.
This means is direct, specific, and does not rely on judgement.
- However, as stated in Technical Specification, a component may be capable of satisfying surveillance requirements and still not be capable of satisfying its intended safety function.
This could be caused by inadequate maintenance of equipment qualification, failure or removal of supports or
- snubbers, or the determination that some other defect exists.
In the event with the Reactor Vessel Head Vent Valves, only those valves which had failed their stroke response time were declared inoperable.
The valves that were pressurized open could be manually closed within the required response time and for this reason were not declared inoperable.
Even though CP&L subsequently found the opening to be predictable, the opening produced an unanticipated, undesired leakage from the Reactor Coolant System.
This should have also resulted in declaring these valves inoperable consistent with the position taken by the NRC.
In response to the NRC's position, CP&L will provide additional guidance to applicable personnel.
The guidance is that for equipment to be considered operable it must:
1.
Satisfy all applicable Technical Specification Surveillance Requirements.
2.
Satisfy applicable design and installation requirements.
3.
Respond as designed to control and protection signals.
MEM/HO-8800620/10/Osl