ML20236W218
| ML20236W218 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 12/01/1987 |
| From: | Grace J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236W201 | List: |
| References | |
| 50-413-87-30-01, 50-413-87-30-1, NUDOCS 8712070334 | |
| Download: ML20236W218 (31) | |
Text
_ - _ _ _ _ -
A 2
ENCLOSURE 1 NOTICE OF VIOLATION Duke Power Company Docket No. 50-413 Catawba Unit 1 License No. NPF-35 During the Nuclear Regulatory Commission (NRC) inspection conducted on August 26, 1987, through September 25, 1987, a violation of NRC requirements was identified.
In accordance with the " General Statement of Policy and Procedure for NRC Enforcement Actions," 10 CFR Part 2, Appendix C (1987), the violation is listed below:
Technical Specification 3.7.1.2 requires at least three independent steam generator auxiliary feedwater pumps and associated flow paths to be operable.
With one auxiliary feedwater pump inoperable, restore the required auxiliary feedwater pump to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least hot standby within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in hot shutdown within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
Technical Specification 6.8.1 requires tnat written procedures shall be established, implemented, and maintained covering the applicable procedures recommended in Appendix A to Regulatory Guide 1.33, Revision 2.
Catawba Nuclear Station Directive 3.1.15, Activities Affecting Station Operations or Operating Instructions, requires that permission be obtained from the Shift Supervisor or other Supervisor with operational control prior to removing from service an instrument or component that may affect unit operation.
Contrary to the above, pressure switch ICAPSS131 was removed from service at some point between July 17, 1986, and July 6, 1987, by shutting its isolation valve without obtaining permission frcm the Shift Supervisor or other Supervisor with operational control. This caused the flow path from Auxiliary Feedwater Pump 1B to the IC Steam Generator to be inoperable as it would have isolated under certain conditions and the licensee did not maintain the unit in an Operational Mode in which Technical Specification 3.7.1.2 did not apply.
This is a Severity Level IV Violation (Supplement I) applicable to Unit 1 only.
Pursuant to the provisions of 10 CFR 2.201, Duke Power Company is hereby required to submit a written statement or explanation to the Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555, with a copy to the Regional Administrator, Region II, and a copy to the NRC Resident Inspector, Catawba, within 30 days of the date of the letter transmitting this 8712070334 B71210 gDR ADOCK 05000413 PDR
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' Duke Power Company 2
Docket No. 50-413 l
Catawba Unit 1 License No. NPF-35 Notice.
This reply should be clearly marked as a " Reply to a Notice of Violation" and should include: (1) admission or denial of the violation, (2) the reason for the violation if admitted, (3) the corrective steps which have been taken and the results achieved, (4) the corrective steps which will be taken to avoid further violations, and (5) the date when full compliance will be achieved.
Where good cause is shown, consideration will be given to extending the response time.
If an adequate reply is not received within the time s,pecified in this Notice, an order may be issued to show cause why the license should not be modified, suspended, or revoked or why such other action as may be proper should not be taken.
FOR THE NUCLEAR REGULATORY COMMISSION S
J. Nelson Grace Regional Administrator Dated at Atlanta, Georgia this 4/
day of dec.1987 l
l 1
t ENCLOSURE 2 ENFORCEMENT CONFERENCE
SUMMARY
On November 6,1987, representatives of the Duke Power Company (DPC) met with the NRC at the NRC's request in the Region II office in Atlanta, Georgia. The conference was held to discuss the unauthorized isolation of an Auxiliary Feedwater (AFW) system pressure switch which rendered the AFW system unable to function as designed under certain cotditions. A list of conference attendees is contained in Attachment 1.
Following opening remarks given by M. L. Ernst, NRC, Deputy Regional Administrator, DPC gave a presentation addressing the NRC concerns raised by the isolation of the pressure switch.
An outline of the DPC presentation is contained herein as Attachment 2.
The DPC presentation showed that while the AFW system would have operated in an off-norma'l configuration, the isolation of the pressure switch would not have prevented the system from perform ng its intended function under any postulated i
conditions.
This meeting served to enhance Region II's understanding of the issue and DPC's plan to prevent recurrence of similar problems.
The NRC enforcement action concerning this issue is discussed in Enclosure 1.
Attachments:
1.
List of Attendees 2.
DPC Presentation Summary
i ATTACHMENT 1 LIST OF ATTENDEES U.S. Nuclear Regulatory Commission M. L. Ernst, Deputy Regional Administrator T. A. Peebles, Chief, Reactor Projects Section 2A, Division of Reactor Projects (DRP)
W. T. Orders, Senior Resident Inspector, McGuire C. W. Hehl, Deputy Director, DPR E. W. Merschoff, Deputy Director, Division of Reactor Safety (DRS)
G. R. Jenkins, Director, Enforcement Investigation Coordination Staff (EICS)
B. Uryc, Enforcement Coordinator B. Bonser, Project Engineer M. S. Lesser, Resident Inspector D. Hood, Project Manager, NRR Duke Power Company D. Rains, Superintendent of Maintenance, McGuire J. W. Hampton, Manager, Catawba Nuclear Station (CNS)
H. B. Barron, Superintendent of Operations, CNS G. Smith, Superintendent of Maintenance, CNS F. P. Schif fley, II, Licensing Engineer, CNS G. E. Swindlehurst, Superintendent Design Engineer E. O. McCraw, Compliance Engineer, McGuire H. B. Tucker, Vice President, Nuclear Production N, A. Rutherford, System Engineer, Licensing i
1 i
i
._______ __ - _ _ _ _ _ _ _ _ _ _ _ - _ - _ a
is ATTACHMENT 2 DPC PRESENTATION
SUMMARY
DUKE POWER COMPANY /NRC ENFORCEMENT CONFERENCE i
NOVEMBER 6, 1987 1:00 PM 9
CATAWBA EVENT PROBLEM DESCRIPTION J. W. HAMPTON SYSTEM DESCRIPTION H. B. BARR0N SAFETY CONSEQUENCES G. B. SWINDLEHURST ROOT CAUSE/ CORRECTIVE ACTIONS G. T. SMITH
SUMMARY
STATEMENT J. W. HAMPTON
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C MP SURE.1CAPT5130 PROVIDES A SIGNAL TO REMOTE INDICATOR SEE LINE REV. 7 5130 PROVIDES LOCAL INDICATION. ICAPS 5130 PROVIDES A INPUT FOR RESPONSE TI E TESTING. IF AN AUTOMATIC AUXILIARY SIGNAL IS RECEIVED AND LOW DISCHARGE PRESSURE IS INDICATED, 9 D DDED VENT,.
NITI ATE A 30 SEC. TIM DELAY. IF AFTER THE DELAY LOW PRESSURE A lRFv PER D(
E9, A SETPolNT MONITOR CLOSES VALVE 1CAMB. ICAPT5150 MONITORS UXILi ARY FEEDWATER PUMP BlSCHARGE PRESSURE AND PROVlDES SIGNALS 7 lADDED NOTE TORS ICAP 5150 AND ICAP 5153 A SETP0009T MONITOR PROVIDES COMPUTER f,l(%
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AUXILIARY FEEDWATER SYSTEM Design Basis:
Assure sufficient feedwater supply to the steam l
generators to remove energy stored in the core and primary coolant in the event of a loss of normal feedwater.
Design Requirements:
Provide 491 gpm auxiliary feedwater flow to the S/G under the following conditions:
- Failure of any pump to start.
- No credit for flow to a faulted S/G.
- No operator action for thirty minutes
- S/G pressure at first safety I
setpoint plus 3% (1210 psig).
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Postulated Conditions
- 1 CAPS 5131 Isolated.
- Feedwater line rupture at A S/G.
- Turbine Driven pump start and successful run for thirty seconds, then failure.
- Loss of normal feedwater.
Actual Capacity
- Immediate operator action available.
- 433 gpm supplied to 1D S/G at 1210 psig.
- Flows in excess of 491 supplied at lower l
S/G pressures.
- Capability to remove core and primary system heat without operator action to unisolate intact S/G.
CONCLUSION:
While the auxiliary feedwater system would have been operating in a configuration different from that described in the FSAR, i.e. supplying only one S/G and flowrate lower than assumed at 1210 psig, the isolation of pressure switch 1CA5131 would not have prevented the system from performing its intended function under any postulated conditions.
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H ANALYSIS OF THE-IMPACT OF A i
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ON~THE LIMITIflG FSAR TRANSIENT e
BACKGROUND
.e WORST CASE SCENARIO e
ANALYSIS METHODOLOGY e 'RESULTS e
CONCLUSIONS i
- 4 BACKGROUND e -FSAR STATES THAT AT LEAST TWO STEAM GENERATORS CAN BE SUPPLIED WITH A MINIMUM 0F 491 GPM OF AUXILIARY FEEDWATER
- e MINIMUM FLOW WITH A NORMAL ALIGNMENT IS 499 GPM AT 1210.PSIG, BASED ON ONE MOTOR-DRIVEN PUMP SUPPLYING TWO STEAM GENERATORS e
DEGRADED SITUATION PROVIDES A MINIMUM 0F ONE MOTOR-DRIVEN PUMP SUPPLYING ONE STEAM GENERATOR SG PRES (PSIG)
FLOW (GPM) 1293 400 1210 433 (86% OF 499) i 1168 450 1015 500 864 550 685 600 277 700 o
LESS AUXILIARY FEEDWATER FLOW THAN ASSUMED IN THE FSAR CHAPTER 15 TRANSIENT AND ACCIDENT ANALYSES l
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WORST CASE SCENARIO e
AS ' IDENTIFIED IN NRC INSPECTION REPORT 50-413(414)/87-25, THE LIMITING SCENARIO IS A MAIN FEEDWATER LINE BREAK ALL MAIN FEEDWATER LOST AT TIME ZERO ALL FOUR STEAM GENERATORS BLOW DOWN UNTIL MSIV CLOSURE ON LOW STEAM LINE PRESSURE ONE MOTOR-DRIVEN CA PUMP SUPPLIES ONE INTACT STEAM GENERATOR e
NO OPERATOR ACTION FOR 30 MINUTES REGARDLESS OF CLEAR INDICATIONS OF SYMPT 0MS AND EXISTING PROCEDURAL GUIDANCE ISOLATE AUXILIARY FEEDNATER TO RUPTURED STEAM GENERATOR MSIV CLOSURE ON RUPTURED-STEAM GENERATOR REALIGN AUXILIARY FEEDWATER TO INTACT STEAM GENERATOR SAFETY INJECTION TERMINATION
s ANALYSIS METHODOLOGY PLANT. SPECIFIC SIMULATION OF-THE WORST CASE SCENARIO USING A e
THREE-LOOP CATAWBA UNIT 1 RETRAN-02 MODEL e
BOUNDARY CONDITIONS MAIN FEEDWATER LINE BREAK AT STEAM GENERATOR A MINIMUM AUXILIARY FEEDWATER DELIVERED TO STEAM GENERATOR D DECAY HEAT CORRESPONDING TO CONTINUOUS OPERATION OF CYCLE 2 AT FULL POWER UNTIL JULY 6, 1987 (195 EFPD).
NO CREDIT FOR HEAT TRANSFER DUE T0-AUXILIARY FEEDWATER SUPPLIED TO THE RUPTURED STEAM GENERATOR (SIGNIFICANT CONSERVATISM)
SCENARIO SIMULATED FOR 30 MINUTES AT WHICH TIME OPERATOR INTERVENTION CAN BE CREDITED
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SEQUENCE OF EVENTS TIME (SEC)
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173 SAFETY INJECTION ON LOW PZR PRESSURE 183 NV AND NI PUMPS STARTED 310 MINIMUMRCSPRESS0RE=1797PSIG 344 STEAM LINE ISOLATION SIGNAL ON LOW SG PRESSURE 350 MINIMUM PZR LEVEL = 6%
370 MINIMUM T-AVE = 517 F I
0 559 PZR HEATERS OFF 571 PZR SPRAY ON 1
1615 PZR WATER SOLID 1624 PZR PORVs BEGIN CYCLING 1800 END OF SIMULATION L
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UNIT STATUS,AT'30 MINUTES IS NOT, SERIOUSLY DEGRADED e
RCS IS WATER SOLID'WITH CHARGING AND SAFETY INJECTION PUMPS ON:
0 0
RCS^T-AVE IS 565 F (557 F IS NORMAL POST-0 TRIP) AND IS INCREASING ~1.5 F/ MIN AUX FDW REQUIREMENT TO MATCH DECAY HEAT AND REACTOR COOLANT PUMP HEAT IS 484 GPM e
OPERATOR INTERVENTION AT 30 MINUTES REALIGN AUX FDW MINIMUM' AUX FDW FLOW AVAILABLE IS 499 GPM TERMINATE SAFETY INJECTION e
UlllT. RECOVERY AND STABl).ZATION WITH OPERATOR ACTION AT 30 MINUTES WAS ASSURED l
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INSTRUMENT ISOLATIONS'
., a :,
rll NUMllER OF CASES:
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.. JALVE TYPES:
DRAGON INSTRUMENT ISOLATION VALvsi 5
DRAGON INSTRUMENT. MANIFOLD VALVE:
1 6
ROOT CAUSES:
OTHER:
4
' INSTALLATION DEFICIENCY:
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PERSONNEL ERROR:
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NUMBERREPORTA$LE:
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INSTRUMENT ISOLATIONS f
- DATE:
12-6-86 VALVE TYPE:
DRAGON INSTRUMENT AFFECTED:
INSPT5050, CONTAINMENT PRESSURE NARROW RANGE METHOD OF DISCOVERY:
OPERATIONS NOTED ABNORMAL PRESSURE READING ON PRESSURE GAUGE.
IAE DISCOVERED' INSTRUMENT' ISOLATION VALVE CLOSED WHILE INVESTIGATING PRESSURE READING.
ROOT CAUSE:
OTHER REPORTABLE:
YES CORRECTIVE ACTION:
RETURNED TO OPERABLE STATUS
_____-____--__a
b INSTRUMENT ISOLATIONS DATE:
4-24-87 VALVE TYPE:
DRAGON INSTRUMENT AFFECTED:
INSPT5040, CONTAINMENT PRESSURE NARROW RANGE METHOD OF DISCOVERY:
OPERATIONS NOTED ABNORMAL PRESSURE READING ON PRESSURE GAUGE.
IAE DISCOVERED INSTRUMENT ISOLITION' CLOSED AND VALVE HANDLE LOOSE (SLIPPING) ON VALVE STEM.
2 v
ROOT CAUSE:
INSTALLATION DEFICI.ENCY-REPORTABLE:
YES
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4 CORRECTIVE ACTION:
ALL CONTAINMENT. PRESSURE (BOTH UNITS) ISOLATION VAVLE HANDLES WERE INSPECTED (3 0F = 30 WERE LOOSE) AND CORRECTED.
THE CALIBRATION PROCEDUR$S WERE REVISED TO POSITIVELY VERIFY RETURN TO SERVICE.
ALL IAE AND PERFORMANCE TECHNICIANS WERE TRAINED ON THIS INCIDENT AND THE DRAGON VALVE HANDLE PROBLEM.
- ~
INSTRUMENT ISOLATIONS DATE:
7-7-87 VALVE TYPE:
DRAGON
. INSTRUMENT AFFECTED:
ICAPS 5131 PRESSURE SWITCH TO VALVE OPERATOR 1CA46B METHOD OF DISCOVERY:
AFTER A MANUAL REACTOR TRIP, TURBINE DRIVEN AUXILIARY FEEDPUMP AUTO STARTED.
VALVE 1CA46B AUTO CLOSED UNEXPECTEDLY.
OPS INITIATED A WORK-REQUEST T0' INVESTIGATE / REPAIR.
IAE FOUND 1 CAPS 5131 ISOLATION VALVE CLOSED.
ROOT CAUSE:
0THER
-REPORTABLE:
YES CALIBRATION CHECK OF 1 CAPS 5131/ RETURNED TO SERVICE CORRECTIVE ACTION:
' ADD THIS AND OTHER SIMILAR ISOLATION VALVES TO THE INSTRUMENT STARTUP CHECKLIST FOR THE AUXILI ARY FEEDWATER SYSTEM.
(BOTH UNITS)
REVIEWED ALL CA PRESSURE SWITCH CALIBRATION I
DOCUMENTATION l
aA INSTRUMENT ISOLATIONS DATE:
9-11-87 VALVE TYPE:
DRAGON MANIFOLD INSTRUMENT AFFECTED:
ICAFT5090 AUXILIARY FEEDWATER FLOW TRANSMITTER ROOT CAUSE:
PERSONNEL ERROR REPORTABLE:
NO CORRECTIVE ACTION:
THE ERROR WAS IDENTIFIED BY IAE ENGINEERING STAFF.
THE PERSONNEL INVOLVED WERE INTERVIEWED AND DISCIPLINARY NOTICES WERE ISSUED TO TWO IAE TECHNICIANS FOR FAILURE TO FOLLOW PROCEDURE I
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__._______.___a
- I INSTRUMENT ISOLATIONS DATE:
9-15-87 VALVE TYPE:
DRAGON INSTRUMENT AFFECTED:
2NSPT5380 CONTAINMENT PRESSURE WIDE RANGE METHOD OF DISCOVERY:
IAE CONDUCTING SURVEILLANCE PROCEDURE PER SCHEDULE FOUND VALVE CLOSED.
ROOT CAUSE:
OTHER REPORTABLE:
YES CORRECTIVE ACTION:
IAE ISSUED WORK REQUESTS TO VERIFY INSTRUMENT VALVE POSITIONS.
UNIT 11 PROCEDURES WERE STARTED ON 9-16-87 AND COMPLETED ON 9-30-87 ON THE FOLLOWING SYSTEMS:
CA, CF, FW, KC, NC, ND, N1, NS, NV, NW, RN, AND SM, UNIT I SYSTEMS WERE STARTED PRIOR TO THE REFUELING OUTAGE WITH CA AND NS COMPLETED ON 9-30-87.
ALL UNIT I ABOVE LISTED SYSTEMS WILL BE VERIFIED PRIOR TO RESTART AT OUTAGE COMPLETION.
,y_
3.,
INSTRUMENT'ISOLATIONS DATE:
9-21-87 VALVE. TYPE:
DRAGON INSTRUMENT AFFECTED:
2KCFT5810 METHOD OF DISCOVERY:
IAE CONDUCTING SYSTEM STATUS REVIEW ROOT CAUSE:
0THER' REPORTABLE:
NO - NON SAFETY RELATED; NO SAFETY FUNCTION CORRECTIVE ACTION:
NONE e
CATAWBA NUCLEAR STATION Corrective Actions for inadvertent Instrument Valve Mispositioning
- 1. Prevention Of inadvertent Miscositioning Of instrument Valves
- Training of IAE personnel on instrument valve operation.
Completed Sept. 24,1987.
- Training of Performance personnel on instrument valve operation.
Completed Nov. 4,1987.
II. Efforts implemented To Detect Mispositioned Instrument Valves
- Upgrade instrument valve checklists to verify proper valve position after an outage.
Complete Dec.1,1987
- Use of performance test procedures to identify questionable as-found instrument valve position.
Completed Nov.1,1987
- Procedure upgrade for returning instruments to service.
Completed Nov. 2,1987
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Ill.
Programmatic Efforts Underway To Ensure Surveillance / Calibration Of Technical Specification / Safety-Related Instruments
- Calibration program review and maintenance procedure revisions.
Complete Dec. 31,1987
- Nuclear Station Modification review process.
Complete April 1988 i
k.
__._____-_-__-_ - _