IR 05000333/1979009
| ML19210E193 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 09/19/1979 |
| From: | Foley T, Mccabe E, Stetka T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML19210E183 | List: |
| References | |
| 50-333-79-09, 50-333-79-9, NUDOCS 7911300680 | |
| Download: ML19210E193 (22) | |
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U. S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT
REGION I
Report No.
79-09 Docket Nc 50-333 License No.
DPR-59 Priority Category C
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Licensee:
Power Authority of the State of New York 10 Columbus Circle New York, New York 10019 Facility Name:
James A. FitzPatrick Nuclear Power Plant Inspection At:
Scriba, New York Inspection Conducted:
June 25-29, 1979 C,
.L
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Inspectors:
/M 9/7/79 T.
. St'etka Reactor Inspector
/dath
$b WM/77 T. Foley R6 actor Ilfspector date
.f $diobv h /79 3. P. Ourd,iReactor Inspector
'
d6te R. M. Knoll, Reactor Inspector (Co-oo)
date Approved by:
fON i}ld'7T E. C. McCabe, Jr., Chief, Reactor date Projects Section No. 2, RO&NS Branch Inspection Summary:
Inspection on June 25-29, 1979 (Report No. 50-333/79-09)
Areas Inspected:
Special, announced inspection of the pipe support verifica-tion program and the licensee's action taken in response to IE Bulletins 79-02 and 79-08.
Plant tours were conducted.
The inspector involved 135.5 inspection hours onsite by three NRC regional based inspectors and one NRC regional based inspector co-op.
This report also documents NRC:RI in office review of the Cycle III Startup Test Report.
Results:
Two items of noncompliance were identified (Infraction-failure to frisk properly when exiting a contaminated area; Infraction-failure to post and barricade high radiation areas).
1439 332 Region I Form 12 (Rev. April 1977)
7911300 g[j
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DETAILS 1.
Persons Contacted Power Authority of the State of New York
- E. Abbott, Operations Superintendent
- R. Converse, Assistant Operations Superintendent
- M. Cosgrove, QA Site Engineer
- W. Fernandez, Assistant to Maintenance Superintendent
- J. Hoddy, Plant Reliability & Performance Supervisor
- L. Johnston, Plant Engineer (Civil)
- H. Keith, Instrument & Control Superintendent J. Leonard, Resident Manager R. Liseno, Plant Engineer 0. Mallori, Principle Civil Engineer L. Milesi, Security / Safety Supervisor
- E. Mulcahey, Radiation Protection & Radio-Chemistry Supervisor
- R. Pasternack, Superintendent of Power Target Technology J. Dainora, Engineer Other personnel contacted included operations, engineering, maintenance, and health physics personnel.
- Present at the exit interview.
2.
Plant Tours During the course of this inspection numerous plant tours were conducted.
These tours included the reactor building, turbine building, diesel gene-rator rooms, intake structure, and the relay room.
The plant is presently in a cold, shutdown condition as a result of an NRC order issued on March 13, 1979 (see NRC:IE Report 79-02, paragraph 3).
As a result of this order, the licensee has been re-analyzing their ori-ginal seismic stress analyses for piping systems.
In addition, the licensee is also involved in testing concrete expansion anchor bolts used on pipe support based plate in accordance with IE Bulletin 79-02 (see paragraph 9 of this report).
Due to the plant's outage status, the plant tours placed particular emphasis on radiation area controls and plant housekeeping conditions.
These tours identified the following:
1439 333
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(1) Plant housekeeping was observed to be in need of improvement.
Licensee representatives stated that due to the anchor bolt testing and other maintenance activities occurring during the outage, it was difficult to maintain proper housekeeping.
TMse representatives also stated that general plant cleanup would be accomplished prior to plant startup.
(2) Radiation control zones were observed to verify proper identifica-tion and implementation.
These observations included review of step-off pad conditions, disposal of anti-contamination clothing and area posting.
These observations identified two items of noncompliance.
a.
On June 25 at approximately 1530 hours0.0177 days <br />0.425 hours <br />0.00253 weeks <br />5.82165e-4 months <br />, two personnel were observed frisking only the bottom of their feet while exiting the contamin-ation control point.
The frisking was done quickly (within about two seconds) and did not include any other part of their body.
These frisking practices were discussed with Health Physics (HP)
personnel at the control point, who stated that they would be alert for poor frisking techniques.
On June 26 at approximately 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> another individual was observed performing a " quick" frisk on the bottom of his feet with no frisking of his body or hands.
During this event a number of HP perscanel were present and no attempt was made to correct the frisking technique.
This individual was stopped and made to re-monitor himself after the inspector alerted the HP personnel of the situation.
These,~ risking techniques are contrary to the Radiation Protec-tion Operating Procedures and Technical Specifications 6.11 and is an item of noncompliance (333/79-09-01).
The licensee's management was apprised of these violations and it was noted that an HP technician was stationed at the control point to assure proper frisking.
b.
On June 26 at approximately 0930 hours0.0108 days <br />0.258 hours <br />0.00154 weeks <br />3.53865e-4 months <br />, the Reactor Water Cleanup (RWCU) Room was observed to be unbarricaded and not posted as a high radiation area.
Review of radiation work permits (RWP)
and area survey sheets indicated area dose rates to be from 75 to 100 MR/HR.
On the same date at apprcximately 0945 hours0.0109 days <br />0.263 hours <br />0.00156 weeks <br />3.595725e-4 months <br />, the East Pipe Tunnel was found to be unbarricaded and not posted as a high radiation area.
Review of the RWP and area survey sheets indicated area dose rates to be from 15 to 200 MR/HR.
Technical Specificatien 6.11(A)1 requires that areas where dose rates are greater than or equal to 100 MR/HR be barricaded and conspicuously posted as a high radiation area.
Failure to comply with this Technical Specification is an item of noncompliance (333/79-09-02).
}439 334
3.
Review of Operator Training A review of training records and discussions with licenseu operators on each shift were conducted to verify the adequacy of licensee administered operator training.
The review and discussions verified the following:
Tnat operators are aware of the specific details of the THI-2 inci-
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dent to the extent available and have received training on any pro-cedure changes initiated as a result of Julletin 79-08; That operators have been instructed on the specific measures which
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provide assurance that engineered safety features would be available if required, in particular, measures for returning such systems to operable status following maintenance and testing; That operators hsve been instructed on the specific and detailed measures
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to assure that automatic actuations of emergency safety features are not overridden; That operators have reviewed plant automatic actions initiated by
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reset of engineered safety features, that could effect the control of radioactive liquids and gases; and,
' hat plant operators and supervisory personnel have been instructed
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in the provisions and directives for early NRC notification of serious events.
All operators interviewed appeared knowledgeable of the items listed above.
In addition to performing this training, the licensee has written and issued Operation Department Standing Order No. 6, Procedure to Insure ECCS Equipment is Operable After Maintenance, dated April 18, 1979, and hcs written and is in the process of issuing special procedure F-E0P-32, Recovering From An Isolation.
Procedure F-EOP-32 provides precautions that require an evaluation as to whether primary containment sampling should be performed prior to unisolating and in one instance makes such sampling mandatory prior to ur. isolating the systems.
The licensee revised their Fmargency Plan and Procedures on April 25, 1979 to specify notification of the NRC and establish continuous communications.
The inspector noted tFat the change requires calling the original tele-phone number and does not recognize the new direct contact system now in effect.
The licensee will revise the plan to coincide with the new sys-tem prior to plant startup.
This item is unresolved (333/79-09-03).
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4.
Review of Engineering Safety Features (ESF)
A.
A detailed review of the ESF was conducted to verify by independent exam-ination of records and procedures that ESF are operable according to T.S.
requirements and that the licensee's procedures and administrative con-trols provide adequate assurance of continued operability.
Valve / breaker /
switch lineups were reviewed for the following systems using system pro-cedures noted below.
These procedures were compared to current system diagrams to verify adequacy of the lineups.
F-0P-1, Main Steam System, Rev. 5;
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F-0P-5, Condensate Transfer System, Rev. 2;
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F-0P-6, Demineralized Water Storage and Transfer System, Rev. 1;
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F-0P-13, Residual Heat Removal System, Rev. 4;
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F-0P-14, Core Spray System, Rev. 2;
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F-0P-15, High Pressure Coolant Injection System, Rev. 3;
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F-0P-17, Standby Liquid Control System, Rev. 2;
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F-0P-19, Reactor Core Isolation Cooling System, Rev. 3;
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F-0P-20, Standby Gas Treatment System, Rev. 3;
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F-0P-21, Emergency Service Water System, Rev. 0;
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F-0P-22, Diesel Generator Emergency Power, Rev. 3;
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F-0P-40, Reactor Building Closed Loop Cooling System, Rev. 1;
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F-ST-40G, Locked Valve Surveillance, Rev. 1;
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F-0P-43A, 125 V D-C Power System, Rev. 3; F-0P-46A, 4160 V and 600 Y Normal A-C Power Distribution System, Rev. 3;
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and,
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F-0P-46B, 120 V A-C Power System, Rev. O.
Review of these valve / breaker / switch lineups revealed a number of inade-cuacies.
Some of these inadequacies follow:
(1) The system valve lineups and the Locked Valve Surveillance kst (F-ST-40G)
are inconsistent in that some valves loded on one lineup are not locked on the other lineup.
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(2) The valve lineup for the Main Steam System (F-0P-1) has no required position indicated for 29 MOV-2048.
(3) Figure OP-13-1, Revision 1, shows RHR-4038, Suppression Pool Pump Suction line ILRT Connection, as RHR-4308.
(4) Vent and Drain valves RCIC-408 and SLC-701 are not shown on their respective system drawings.
(5) Lube Oil Circulating Pump Discharge valves EDG-53A, B, C, and D appear on the system valve lineup (F-0P-22) but are not on the system drawings.
(6) Motor Driven Injector Pump Isolation valve EDG-101 is not on the system valve lineup (F-0P-22).
The valve was not correctly located on the system drawing and was changed, via pen and ink, to another incorrect location on the system drawing.
(7) Figure OP-43-A-1, and the respective one line diagram illustrate Battery
"A" with an output breaker.
The battery is actually directly wired to the bus and no such breaker exists.
(8) Figure OP-46-B-3 and the respective one-line diagram show the cir-cuit breaker supplying power to the standby liquid control system pipe heat tracing as radwaste piping heat tracing.
(9) Valves 10-MOV-151A and B (Suppression Pool Discharge) are shown on figura OP-46A as being connected to a circuit breaker.
In actuality these valves have been physically disconnected from their power supplies.
(10) Table II of procedure F-0P-21 shows the breakers for emergency service water pumps A & B as "RI" (Racked-In) with no normal position for these breakers being designated.
(11) Table II of procedure F-0P-17 designates valves EV-14A and EV-148 (Explosive Valve Firing Power) as separate breakers.
Figure OP-46A shows that these valves share the same circuit breaker the standby liquid control pumps use.
The licensee will review and revise valve / breaker / switch lineups and sys-tam diagrams to be consistent with actuai system cnnfigurations.
These revisions will be completed by August 31, 1979.
This item is unresolved (333/79-09-04).
1439
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B.
During the review conducted in paragraph A, the inspectors noted inade-quate tagging of instrument root valves and rack mounted valves.
The li-censee performs an instrument valve lineup check in accordance with procedure F-ISP-73, During Refueling Prior to Startup Instrument Line Valve Check Off List.
Due to the lack of instrument valve labeling, the effectivenes; of this check-off list is questionable.
In addition, the system lineups conducted by the inspectors indicate that F-ISP-73 is not complete.
The licensee will tag all safety-system instrument valves, revise F-ISP-73 to insure a complete check off list, and implement the new valve check off list prior to plant startup.
In addition, the licensee will perform independent verification of instrument valve positions until these changes have been made.
This item is unresolved (333/79-09-05).
C.
Due to the plants cold shutdown condition, the verification of actual valve / breaker / switch positions for power operation could not be parformed.
This verification will be performed during a subsequent inspection after the plant returns to power operation.
5.
Surveillance Test / Maintenance Procedure Review A.
The inspectors reviewed the following Surveillance Test Procedures, Instrument and Control Surveillance Procedures, Lub"ication, and Maintenance Procedures to assure that these systems and components are returned to an operational configuration.
SURVEILLANCE TEST PROCEDURES (OPERATIONAL)
F-ST REV.
NO.
TITLE NO.
DATE
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1B MSIV Fast Closure Test (ISI)
12/1/78 IC Primary Containment
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Isolation Valve Exercise
7/30/76 1G Main Steam Line High
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Radiation Functional Test
6/4/76
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II Main Steam Isolation Valves Limit Switch Instrument Func-tional Test
8/10/76
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1M MSLCS Valve Exercise
10/11/78
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2A RHR Pump Flow Rate Test (ISI)
12/1/78 2B RHR Pump Operability Test (ISI)
12/1/78
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2C RHR M0V Operability Test
8/20/76 1439 338
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9/74 Headers and Nozzle Air Test (ISI)
2E RHR Service Water Pump
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Flow Rate Test (ISI)
12/1/78 2G RHR Isolation Valve Control
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Logic System Functional Test
1/23/75 2H LPCI SuLsystem Logic System
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Functioral Test
6/04/76 2M ECCS Trip Systems Bus Power
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Mor.itors Functional Test
8/10/76
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2R RHR Service Water Pump and MOV Operability Test (ISI)
12/1/78
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2S Valve Testing - Residual Heat Removal (ISI)
12/1/78
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3A Core Spray Pump Flow Rate
12/1/78 3C Core Spray Pump Operability
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Test (ISI)
12/1/78
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3D Core Spray MOV Operability Test
6/15/76
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3G Core Spray Subsystem and LPCI Subsystem Discharge Piping Vented
9/74
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3J Core Spray Subsystem Logic System Functional Test
2/27/79
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3L Core Spray and Containment Spray Keep Full Switch Functional Test
8/3/78
12/1/78
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4C HPCI Pump Operability Test (ISI)
12/1/78 1439 339
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SURVEILLANCE TEST PROCEDURES (OPERATIONAL)
(Continued)
F-ST REV.
NO.
TITLE NO.
DATE 4D HPCI MOV Operability Test
6/15/76
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4E HPCI Subsystem Logic System
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Functional Test
3/22/77
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4E HPCI Subsystem Automatic Isolation Logic System Functional Test
2/18/77
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5B APRM Instrument Functional Test
1/18/78
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SQ Flow Bias Functional Test
1/5/79 SR RBM Upscale and Downscale
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Instrument Functional Check
12/9/77
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6A Standby Liquid Control Pump Functional Test (ISI)
12/1/78
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Logic System Functional Test
9/21/77
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ESW Pump Operability Test (ISI)
12/1/78
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6/15/76
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8D ESW Pump Flow Rate Test (ISI)
12/1/78
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1/31/79
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15A Pressore Suppression Chamber -
Drywell Vacuum Breaker Open-ing and Closing Test
8/74 15C Pressure Suppression Chamber -
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Reactor Building Vacuum Breakers Operability Test
2/18/77 15G Pressure Suppression Chamber -
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Reactor Building Vacuum Breaker Differential Simulated Automa-tic Actuation and Set Point Test
10/75 1439 340
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SURVEILLANCE TEST PROCEDURES (OPERATIONAL)
(Continued)
F-ST REV.
NO.
TITLE NO.
DATE 15H Primary Containment Integrity,
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Manual Isolation Valves Position Verification
9/14/76 16A 125 VDC Battery Weekly Tests
3/21/78
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16B 125 VDC Battery Monthly Tests
7/16/76 16C 125 VDC Battery, Specific
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Gravity, Temperature and Electrolyte Level Checks
3/21/78
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16E LPCI Independent Power Supply Weekly Battery Test
1/25/79 16F LPCI Independent Power Supply
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Three Month Battery Check
1/25/79 16G LPCI Battery Monthly Test
11/22/78
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22C ADS Logic System Functional
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Test
3/12/76 24A RCIC Pump Operability Test
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(ISI)
12/1/78
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11/15/77 24C RCIC Flow Rate Test (ISI)
12/1/78
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24D RCIC Subsystem Automatic
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Isolation Logic System Functional Test
11/15/77
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34A Reactor Building Isolation Logic System Functional Test
9/21/76 34B Reactor Building Isolation
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Logic System Functional Test
1/10/75 35A Containment Cooling subsystem
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Logic System Functional Test
6/04/76
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SURVEILLANCE TEST PROCEDURES (OPERATIONAL)
(Continued)
F-ST REV.
NO.
TITLE NO.
DATE 40G Locked Valve Surveillance
12/1/78
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41B Valve Testing - RBCLC Water (ISI)
12/1/78
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[439 342
INSTRUMENTATION AND CONTROL SURVEILLANCE PROCEDURES ISP NO.
TITLE REV. NO.
DATE
Inst. Line Flow Check Op.
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Test
9/77
Reactor Water Level
6/77
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3-1 Reactor Lo/Lo Water Level
6/79
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3-2 Reactor Lo-Lo/Lo-Lo-Lo Water
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Level
3/79 3-3 Reactor Low Water Level (ADS)
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Instrument
4/79 3-4 Reactor Level Instrument
3/78
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Drywell High Pressure
4/78
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4-1 Drywell High Pressure (HPCI, LPCI, RHR, SBGT, EDG, Core Spray Instrument
4/76 4-2 Drywell High Pressure Inst.
4/78
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4-3 Drywell High Pressure (Cont.
Spray Permissive)
4/78
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4-4 Drywell High Pressure Inst.
5/79
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4-5 Drywell Supp. Chamber Diff.
Press
12/78
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Reactor High Pressure
11/78 5-2 Reactor Low Pressure (Break
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Detection)
2/76 5-3 Reactor Low Pressure (LPCI,
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Core Spray Permissive)
6/78 5-4 Reactor Pressure Instrument
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Calibration
6/78 5-5 Reactor High Pressure
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Transmitter
12/78 1439 343
6 Pump Discharge Pressure
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6/78 6-1 Pump Discharge Pressure
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Interlock (Core Spray)
1/79
Core Spray Sparger Diff.
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Prtssure
4/78
Steam Line Flow--High HPCI/RCIC
3/78
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9-1 RCIC Steam Line High Flow
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Instrument
3/78
HPCI/RCIC Steam Line High
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Temperature
12/78
-- 10-1 RCIC Steam Line Temperature Instrument
12/78
HPCI Steam Line Low Pressure
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Instrument
6/78
-- 12-1 RCIC Steam Line Low Pressure
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Instrument
5/78
HPCI Turbine High Exhaust
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Pressure Instrument
8/78
-- 13-1 RCIC Turbine High Exhaust Pressure Instrument
6/79
HPCI Low Pump Suction Press
8/78
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-- 14-1 RCIC Low Pump Suction Press
8/78
APRM-Downscale, Upscale
' 12/78
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HPCI Turbine Exhaust Diaphram
4/79
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-- 22-1 RCIC Turbine Exhaust Diaphram
4/79
Emergency Service Water
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Lockout Matrix
8/78
-- 23-1 Emergency Service Water System Instrument
8/78
Rod Block Monitor
6/78
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27 Drywell Temperature
1/79
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Suppression Chamber High Water
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Level
11/77
-- 29-1 Suppression Chamber Water Level Instrument Cal.
12/77
-- 63-1 APRM Flow Bias Signal
2/77
-- 64-1 Main Steam Line High Radiation
10/78
-- 65-1 Main Steam Line Isolation Valve Closure (RPS)
3/70
-- 66-1 Scram Discharge Volume High Water Level Instrument
12/77
-- 70 Reactor High Pressure Permissive
3/79
During Refueling Prior to
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Startup Instrument Line Valve Check Off List
12/74
Condensate Storage Tank Low
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Water Level
4/78
-- 77 Reactor Shroud Level
12/78
LPCI Discharge Piping Press 9/77
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-- 80 C. S. Discharge Piping Press 9/77
Standby Liquid Control Level /
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Temperature
9/77
-- 82-1 Standby Liquid Control System Temperature
1/79
Condenser Low Vaccuum
4/79
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Standby Gas Treatment
4/76
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Suppression Chamber /RX BVID
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Vaccuum Breaker Isolatior Valve
5/79 1439 345
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LUBRICATION PROCEDURES LP #
COMPONENT REV. N0.
DATE
-- 002 Motor, Inductic-
9/78
-- 003 Pump, Positive Displacement Triplex
9/78
-- 005 HPCI Turbine
9/78
-- 007 Couplings
9/78
-- 009 Air Compressor
9/78
-- 011 Diesel Engine
9/78
-- 016 Motor Generator Set
9/78
-- 021 Centrifugal Pump
9/78
-- 025 RCIC Turbine 13-TV
9/78
-- 030 Pump Gear Reducer Drive Mechanism
9/78
-- 033 Hydraulic Scissor Lift
9/78
-- 036 Pump, Metering
9/78 1439 346
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MAINTENANCE PROCEDURES PROCEDURE NMPC PASNY NO.
TITLE DATE REV.
DATE REV.
M1.3 Main Steam Isolation Valves 7/77
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-- M1.4 Reactor Vent and Drain Valves Maintenance 6/76
-- M3.1 Control Rod Drive Unit Repair 6/77
M3.2 Hydraulic Control Unit
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Repair 6/77
-- M7.1 Containment Control Valve and Vacuum Relief Valve Maintenance 7/76
-- M8.1 SGTS Fan and Motor Maintenance 7/76
-- M8.3 SGTS Filter Replacement 7/76
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M8.4 SGTS Heating Coil
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Maintenance 7/76
-- M10.1 RHR Pump Maintenance 7/76
M10.2 RHR Heat Exchanger
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Maintenance 7/76
M10.3 RHR Valve Maintenance 6/77
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-- M11.1 SLC Pump Maintenance 9/77
M11.2 SLC Explosive Valve Replace-
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ment and Bench Testing of Trigger Assembly 7/76
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W /
1439 347
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MAINTENANCE PROCEDURES (Continued)
TITLE DATE REV.
DAJE REV.
M11.4 SLC Accumulator
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Repair 7/76
-- M13.1 RCIC Turbine Maintenance 9/76
M13.2 RCIC Valve
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Maintenance 9/76
M13.3 RCIC Pump Maintenance 9/76
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-- M14.1 Core Spray Pump Maintenance 9/76
M14.2 Core Spray Valve
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Maintenance 9/76
-- M23.1 HPCI Turbine Maintenance 9/76
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M23.2 HPCI Pump Maintenance 9/76
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-- M23.3 HPCI Valve Maintenance 9/76
-- M23.4 HPCI Gland Condenser Maintenance 9/76
-- M23.6 HPCI Booster Pump Maintenance 9/76
-- M33.1 Emergency Service Water P.,' Maintenance 9/76
M33.2 Emergency Service Water
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Valve Maintenance 9/76
M52.1 Diesel Generator
--
Maintenance 9/76
e 14'39 348
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MAINTENANCE PROCEDURES (Continued)
TITLE DATE REV.
DATE REV.
M52.2 Diesei Engine and
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Auxiliaries Maintenance 9/76
M52.3 Fus! Oil Transfer Pump 9/76
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M54.1 4 KV Switchgear
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Maintenance 9/76
-- M55.1 600V Emergency Load Center Maintenance 9/76
M56.1 600V Emergency Motor Con-
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trol Center Maintenance 9/76
M57.1 125V DC Equipment
--
Maintenance 9/76
M57.3 Replacement of a Station
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Battery Defective Cell
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10/77
M57.4 LPCI Independent Power
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Supply Charger / Inverter L/78
M58.2 UPS Motor Generator
--
Maintenance 9/76
-- M58.4 Reactor Protection System M-G Set Maintenance 9/76
M59.1 Bolted Bonnet Valves
--
Maintenance 9/76
-- M59.2 Pressure Seal Valves Maintenance 9/76 h
f439 349
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PREVENTIVE MAINTENANCE PROCEDURES PROCEDURE NO.
TITLE REV.
DATE EP001 Circuit Breaker
8/78
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EP003 Valve Operator
5/79
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EP004 Motor
8/78
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EP005 Motor
8/78
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IP010 Transformer
8/78
--
--
IP013 Switchgear and Distribu-
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tion Panels
8/78
--
IP016 Disconnect Switch
8/78
--
EP017 Transformer and Breaker Insulating Oil
8/78 B.
The inspectors identified the following discrepancies with respect to the return-to-normal criteria during the review of the preceding procedures:
(1)
F-ISP-5-4, Reactor Pressure Instrumentation Calibration, Rev. 5, refers to " Main and Auxiliary Feedwater Systems" in paragraphs 3.5, 5.2 and 5.3.
This facility does not contain an Auxiliary Feedwater System.
This procedure will be revised to reflect the actual plant conditions.
This item is unresolved (333/79-09-06).
(2) A number of procedures including F-ISP-3-4, F-ISP-4, F-ISP-4-1, F-ISP-4-2, F-ISP-4-3, F-ISP-9-1, F-ISP-13, F-ISP-13-1, F-ISP-14-1, F-ISP-22-1, and others leave the test connection drain valve open at the completion of the procedure.
This action would leave the instrument inoperable and allow spillage of fluids via the drain line.
Additional steps are required in the procecure to assure proper return to service.
The licensee will revise ail affected procedures by September 1, 1979.
This item is unresolved (333/79-09-07).
(3) Procedure F-ST-2C, RHR MOV Operability Test, Rev.
3., does not specify how to return the system to a standby mode because the valve lineup reference is left blank.
This procedure will be revised to incor-porate the appropriate valve lineup to return the system to a stand-by mode by July 16, 1979.
This item is unresolved (333/79-09-08).
1439 350
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(4) Procedure F-ST-25, Valve Testing - Residual Heat Removal (ISI), Rev. O, specifies that a return to normal valve lineup is not required, however, a preceding step requires the re positioning of certain valves and closes valve 10 MOV-278 without reopening the valve.
An open position appears to be the normal configuration for this valve.
This procedure will be reviewed and revised as requi*ed to ensure the system is returned to normal by July 16, 1979.
This item is unresolved (333/79'09-09).
(5) Preventative Maintenance (PM) procedures EP001 through EP017 do not specify to initiate a work request to perform the maintenance and only require performance of a system tag-out.
Though the licensee's practice had been to require a work request for such maintenance, it should be stated in each PM procedure to assure that the work request controls for a proper return to normal are in effect.
The licensee will revise the PM procedures to require -issuance of a work request.
This item is unresolved (333/79-09-10).
6.
Review of Administrative Controls for System Tagging and Return to Service Administrative controls imposed by the licensee 'or maintenance and test activities were reviewed to verify the following:
ESF components are properly returned to service;
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ESF syste$ns are returned to operability at the conclusion of extended
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outages; and, Tagging practices on control panels do not obscure status indicators.
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The following procedures and orders were reviewed:
AP 3.1, Procedure for Maintenance Procedures, Rev. 1;
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Standing Order No. 5, Valve / Electrical 'ineup Checkoff List Review,
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Rev. 0; Standing Order No. 6, Procedure to Insure iCCS Equipment Is Operable
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After Maintenance, Rev. 0;
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Work Activity Control Procedure (WACP) 10.1.1, Procedure For Control of Maintenance, Rev. 1, and, Work Activity Control Procedure (WACP) 10.1.:, Equipment and Person-
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nel Protective Tagging, Rev. 1.
The licensee has established the use of " mini-tags" in accordance with procedure WACP 10.1.2.
These " mini-tags" are utilized on control panels to prevent obstruction of status indicators and were judged by the inspector to be effective in their purpose.
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.
Review of procedure WACP 10.1.1 indicated that the licensee is using a slightly different Work Request / Event / Deficiency / Form than is described in the ;'acedure.
This discrepancy was identified during Quality Assur-ance Audit 295 conducted to revise WACP 10.1.1 prior to plant startup.
This item (333/79-09-11) is unresolved.
7.
Reactor Vessel Water Level Control The inspector queried licensee representatives with reprd to the manual actuation of RCIC or HPCI to assist in level control of the reactor during a routine operational transient.
The representatives answered that they do not require manual actuation of these systems.
The HPCI and RCIC sys-tems initiate automatically when reactor vessel level falls to -38 inches.
During a routine operational transient (e.g., a reactor scram), it is expected that reactor vessel water level will fall to -38 inches before an operator could react to manually start the HPCI/RCIC systems.
If this automatic actuation does not occur, procedures direct the operators to start the systems manually.
The inspector had no further quastions on this item at this time.
8.
Independent Verification of Valve / Breaker / Switch Lineups
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The inspector queried the licensee representatives concerning their use of independent verification of valve / breaker / switch lineups.
Licensee representatives stated that they use independent verification for instru-ment root valves but not for other valve lineups and do not have proce-dures that require these verifications.
The inspector had no further questions on this item at this time.
9.
Pipe Support Verification Program Reference:
IE Inspection Report 50-333/79-06 The purpose of this inspection was to verify the accuracy of the licens-ee's "as-built" pipe support drawing information used to support the pipe stress analysis.
The inspection included 23 pipe supports selected from the list of drawings (MSK's) and systems referenced in attachment 1 to the letter to NRR, dated April 30, 1979, No. JPN-79-24.
The drawings were compared to the piping and pipe support installation for the accu-racy of the relative locations and configuration.
Selected dimensions of the pipes and pipe supports were measured and verified to the drawings.
The following listed pipe supports were examined:
ii-10A-988 H-29-146 H-46-177A H-13-3 H-PC-185 H-46-178A H-13-3A H-46-169 H-46-178 (439 352
.
.
H-13-19A r.46-170 H-46-280 H-13-49 H-46-176 H-46-282 H 14-73 H-46-177 H-46-284 H-46-334 PFSK-1120 H-46-347 PFSK-840 H-46-348 The following deficiencies were noted:
H-46-284 The welds between the 3/4" x 2 1/2" x 8" plates are cracked H-46-282
" clearance between the sliding pad and the base plate H-46-347 Undersize anchor bolts H-46-176A Undersize anchor bolts H-46-348 Undersize anchor bolts The licensee issued Work Requests for the foregoing deficiencies which will evaluate and correct the condition.
The foregoing items are con-sidered unresolved pending completion of corrective actions and review by the NRC (333/79-09-12).
The inspector reviewed the licensee's procedure No. MPT-15, Revision 4, for technical content and adequacy ralative to Bulletin 79-02.
No dis-crepancies were identified.
10.
NRC in Office Review NRC:RI in office review of the Cycle III Startup Test Report for the James A.
Fitzpatrick Nuclear Power Plant has been completed.
The report contains test results consistent with design predictions.
No unacceptable conditions were identified.
11.
Unresolved Items Unresolved items are those items for which further information is required to determine whether they are acceptable or items of noncompliance.
Un-resolved items are contained in Paragraphs 3, 4.A and B, 5.B, 6 and 9 of this report.
12.
Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1)
at the conclusion of the inspection on June 29, 1979, and summarized the scope and findings of the inspection as they are detailed in this report.
During this meeting, the unresolved items and items of noncompliance were identified.
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