IR 05000528/1994004
| ML17310B132 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 03/10/1994 |
| From: | Ang W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION V) |
| To: | |
| Shared Package | |
| ML17310B130 | List: |
| References | |
| 50-528-94-04, 50-528-94-4, 50-529-94-04, 50-529-94-4, 50-530-94-04, 50-530-94-4, NUDOCS 9403290077 | |
| Download: ML17310B132 (44) | |
Text
U.'.
NUCLEAR REGULATORY COMMISSION
REGION V
II tl.:
Docket No.:
IJ tl Licensee:
~lit II
~lt d Ins ect'o date:
- Lnne~cto:
50-528/94-04, 50-529/94-04, and 50-530/94-04 50-528, 50-529, and 50-530 NPF-41, NPF-51, and NPF-74 Arizona Public Service Company Palo Verde Nuclear Generating Station Units 1, 2, and
Palo Verde Nuclear Generating Station, Wintersburg, Arizona January 24 through February 11, 1994 C. Clark, Reactor Inspector
~dd tt W. P. Ang, Ch ef Engineering Branch Q-lb-g Date Signed Ins ectio Summar :,
Ins ection dur the eriod Jan ar 24 throu h
ebruar ll 1994 e ort os 50-528 94-04 50-529 94-04 and 50-530 94-04 d
.'his announced routine inspection, reviewed the licensee implementation of maintenance activities and the licensee's actions for open items identified in NRC inspection reports.
NRC inspection procedures 62700, 92701, and 92702 were used as guidance for this inspection.
Sa et Iss an e ent S st S
MS Ite
None jlesults:
eneral Co clus o s and S ecific indin s:
The inspector concluded that the licensee's maintenance activities observed were performed in an effective manner.
Adequate engineering support was i
provided for the observed maintenance activities.
94032'70077 I9403li
'DR ADQCK 05000528
I
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I Si
'cant Safet atte s:
None.
Summar of Vio ation or Deviations:
No violations or deviations of NRC requirements were identified during this inspection.
0 en Items Summar
The inspector closed seven open items.
The status of four additional items was reviewed and the results are identified in Section 5 of the repor I l
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'getail s 1.
Persons Contacted izon Pub ic Service Com an
- R. Adney, Plant Manager, Unit 3
'J.
Bergstedt, Senior guality Monitor, guality Assurance and Monitoring
- R. Bernier, Supervisor, Nuclear Regulatory Affairs
- M. Berryman, Supervisor, Unit 3 Mechanical Maintenance
- P. Brandies, Manager, Maintenance Support T. Cannon, Supervisor, Plant Inservice Inspection and Testing
- F. Floyd, Supervisor, Unit 1 Mechanical Maintenance D. Garchow, Manager, Site Mechanical Engineering Department
- D. Kanitz, Senior Engineer, Nuclear Regulatory Affairs W. Lehman, Engineer (Check Valve), Inservice Inspection and Testing D. Oakes, Primary Discipline Engineer, Inservice Inspection and Testing D. Phillips, Manager, Maintenance Standards
- R. Prabhakar, Manager, Independent Safety /guality Engineering
- R. Russo, Manager, guality Control
- T. Taylor, Supervisor, Maintenance Standards M. Woloszyn, Senior Technical Specialist, guality Assur ance Audits W. Wong, Primary Discipline Engineer, Inservice Inspection and Testing
- R. Younger, Manager/Staff, Site Maintenance Others e
"J. Draper, Site Representative, Southern California Edison Company
- F. Gowers, Site Representative, El Paso Electric U. S. Nucle r Re u ato Commis
'on K. Johnston, Senior Resident 'Inspector The inspector also held discussions with other licensee and contractor personnel during the course of the inspection.
- Denotes those attending the exit meeting on February 11, 1994.
2.
a tenance m lementatio 62700
a e PiLrrose The purpose of this inspection was to:
Verify that maintenance activities for safety related systems and components were being conducted in a manner that results in the reliable and safe operation of the plant.
Evaluate the adequacy of maintenance activities and associated engineering support by observing ongoing maintenance activitie f l
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b.
~ck rou d
During this inspection the licensee was conducting the Unit 2 mid-cycle outage.
This outage was scheduled by the licensee to perform steam generator tube inspections.
c.
Ins ector's etio s
Du in Present Ins ection The inspector reviewed the following:
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The Unit 2 Mid-Cycle Outage Baseline Schedule dated January 6,
1994.
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Mechanical Maintenance Department daily work schedules for the dates the inspector was on site.
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guality Assur ance Audit Report 93-001,
"Special Process,"
dated April 13, 1993.
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guality Assurance Audit Report 93-013,
"Maintenance,"
dated December 3, 1993.
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Approximately 108 guality Control Reports issued by the licensee for observation of maintenance activities performed in Units 1, 2, and 3.
The reports were dated from September 9,
1992, to January 21, 1994.
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Ten guality Assurance (gA) Monitoring Reports issued by the licensee for observation of maintenance activities.and documentation associated with work performed in Units 1, 2, and 3.
The reports pertained to observations performed from May 4, 1993 to September 9,
1993.
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Work Order No. 00644464, issued to disassemble, inspect, and reassemble Unit 2 turbine driven auxiliary feedwater pump (AFA-POl).
d.
Obse vation of ainte ance Acti ities Back round The licensee identified that the Unit 2 turbine driven auxiliary feedwater pump (AFA-POl) vibration levels were trending upward toward unacceptable levels.
The licensee initiated an investigation to identify the cause for this increase in pump vibration.
P ocedures The inspector reviewed the licensee's work packagegmaintenance procedures issued for the maintenance activities observed by the inspector and concluded that the procedures conformed to the
licensee's administrative requirements for format, approval, and control.
ork Observations The licensee performed an alignment check of the assembled pump with laser equipment and did not identify any unacceptable alignment conditions.
On January 26, 1994, the inspector observed the disassembly and initial inspection of the pump and associated piping, performed with a pump vendor (Sulzer Bingham) representative observing the activities.
This work was performed in accordance with instructions in Work Order No. 00644464 and Work Request 863641.
The initial inspection of the pump components, performed during and after disassembly, did not appear to identify any abnormal conditions that would have caused increased pump vibration by themselves.
The pump rotating element was sent to the manufacturer for "as found" balancing, disassembly, inspection, reconditioning, reassembly, and "as left" balancing.
After the vendor disassembled the rotating element, the vendor initially notified the licensee that it appeared the rotating element's shaft had an "as found" maximum total indicated runout (TIR) of 0.007 inch over the length of the shaft.
The licensee indicated that this "as found" condition of the shaft may have been the cause of increased pump vibration.
The rotating element had not been returned to the licensee by the
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end of this inspection.
The licensee is initiating corrective actions to identify and resolve the problem.
Personnel involved in the pump inspection activities were knowledgeable and performed their assigned functions in a competent manner.
Conc us'o s
The inspector concluded that the maintenance documents reviewed and
~ the maintenance activities observed, appeared to indicate the
'icensee maintenance activities were being implemented, monitored and performed in accordance with the guidance of the licensee's Haintenance Program.
The licensee appeared to be taking appropriate actions to identify the cause of the increase in pump vibration.
Adequate engineering support was provided for the work observed.
No safety significant problems were identified.
No violations o} deviations of NRC requirements were identifie l
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3.
o ow-u of Corrective Actions for Previous de t fied tems 9 70 a.
C osed Follow U Item 50-528 89-08-01:
Emer enc
erat'rocedures U
rade ro ect 0 i na NRC ol ow U t
NRC inspectors identified that the licensee's initial emergency operating procedures (EOPs)
presented significant usage problems to the operators.
Those problems were identified as human factors related and included:
Inconsistencies in structure and format.
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An overly complex structure.
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Ill defined, excessive, and buried transitions within the optimum and functional recovery procedures.
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Heavy reliance on operator judgement and discretion.
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The lack of a well defined philosophy of use.
NRC inspection report 90-02 noted on February 17, 1990, that licensee implementation of corrective actions for these problems
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would be delayed from July 1990 to July 1991, and that the licensee was in the process of identifying these delays to the NRC.
This report left this item open until the noted delays where identified to the NRC by the licensee and the licensee issued a revised schedule for when the revised EOPs would be issued.
. Licensee's Act ons in Res o se to th s te In a letter dated March 21, 1990, the licensee notified the NRC of the status of their actions to upgrade the EOPs and provided a
revised schedule for completion of the EOP upgrade project.
I s ecto 's Act o s Dur the Prese t I s ect o
The inspector reviewed the following:
A licensee letter dated Narch 21, 1990, which provided the status of the EOP upgrade project and a revised schedule showing that Revision 0.01 of the EOPs would be effective September 30, 1990.
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A licensee letter dated August 19, 1992, which identified that Revision 0.02 of the new EOPs became effective on August 17, 199 I
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NRC inspection report 93-33, dated September 3,
1993, which identified that the licensee had initiated corrective actions to improve the useability of EOPs based'n lessons learned from the recent Unit 2 steam generator tube rupture event and the latest NRC requalification examinations.
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NRC meeting report 93-50, dated December 2, 1993, which identified a licensee proposed projected timeline showing completion of rewriting all EOPs and training by August 1995.
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Various available licensee documents and records identifying problems and-actions taken for the EOP upgrade project.
The inspector also discussed this item and the information reviewed, in the documents identified above, with various licensee personnel.
Discussion a
d Conclusion The inspector determined that the licensee had notified the NRC on March 21, 1990, of delays in the original EOP upgrade project initial schedule and provided a new schedule for issue of revised EOPs.
The inspector noted that the NRC and the licensee were still tracking problems associated with the EOPs and that the NRC had tentatively schedule future EOP inspections in 1994 and 1995.
Based on this review, the inspector concluded that the licensee's actions adequately resolved the original follow up item.
This item is closed.
C osed ice see vent Re o t em 50-5 8 9 -0 1-00:
Postu ted Loss of En 'ered Safet Features E ui e t Wit a F're i the
~Ct 1 t Ori inal Licensee Event Re ort On October 29, 1991, the licensee determined that a design basis
)0
,CFR 50 Appendix R fire in the Control Room could result in the loss of one Train "B" essential air handling unit (AHU).
The Train
"B'ssential AHU provides cooling to Train "B" Engineered Safety Features (ESF) equipment, and the Train "B" DC battery rooms.
The Train "B" equipment is necessary for the safe shutdown of the plant if there was a fire in the Control Room.
The cause of this postulated event was a failure of the original
CFR 50 Appendix R evaluation to recognize the potential for a fire in the Control Room affecting the control circuits for one essential AHU.
Licensee's Actions n Res onse to th s Item Upon discovery of this potential event, appropriate compensatory measures were established in accordance with the licensee Fire'
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Protection Program.
The licensee also initiated actions to:
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Revise the procedure,
"Shutdown Outside the Control Room Due to Fire and/or Smoke," to return the applicable essential AHU to service in the event of a fire in the Control Room.
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To issue a plant change request (PCR) 91-13-HJ-003, to provide a transfer switch to isolate the Control Room circuits and install a local control switch for the affected essential AHU.
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To issue PCR 91-13-ZJ-002 to provide door hold open devices which trip closed on a fire for the doors between the "8" and
"D" DC equipment rooms and the Train "B" ESF switchgear room.
Ins ector's Actions Durin the Present Ins ection The inspector reviewed the following:
Condition report/disposition request (CRDR) 9-1-0225, dated October 21, 1991, which identified that a design basis
CFR 50 Appendix R fire in the Control Room could result in the loss of one Train "B" essential air handling unit (AHU).
A licensee conversation memorandum (File No. 91-003-726)
dated October 21, 1991, with the PVNGS Fire Department, which established interim compensatory measures for fire protection.
In response to CRDR 9-1-0225 and this memorandum the PVNGS Fire Department established an hourly fire watch for the Control Room for all units.
The inspector determined that the licensee was performing an hourly fire watch in the affected area of the Control Room for all units.
Licensee Event Report 91-011-00 dated November 27, 1991.
Procedures 41A0-1ZZ44, Revision 4; 42A0-2ZZ44, Revision 3; and 43A0-3ZZ44, Revision 2;
" Control Room Fire."
These procedures provided the current licensee instructions to return the applicable essential AHU to ser vice in the event of a fire in the Control Room.
The inspector determined that these instructions appeared adequate to perform the required actions in the event of a fire in the control room.
A licensee letter (392-00023-FDG/DRD)
dated March 11, 1992, which identified that Plant Change Request 91-13-ZJ-.002 to install "hold-open" devices on the applicable doors would be cancelled.
This letter stated that revised calculations
[Calculation 13-MC-HJ-261 (R/I)] demonstrated that DC equipment rooms would not overheat for 45 minutes with a loss of essential air and the doors closed.
Consequently, adequate time existed for operator action to open the doors during a
Control Room fire scenario.
Based on the information noted in
(,
this letter, PCR-13-ZJ-002, was cancelled by Fire Protection Engineering.
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A licensee letter (102A-01152-RAB/JNI) dated October 25, 1993, which requested licensee actions to identify specific due dates on the cycle milestone schedule for each unit, for implementation of plant modifications addressed under the
CFR 50 Appendix R Reconstitution Project.
This letter noted that a meeting held with the NRC on July 20, 1993, to discuss the licensee's
CFR 50 Appendix R Reevaluation Effort, at PVNGS, included a schedule for implementation of plant modifications.
This schedule was included in an NRC letter, dated July 27, 1993, summarizing the July 20, 1993, meeting.
The inspector determined that the internal letter requested that PCR 91-13-HJ-003 be tentatively scheduled to be performed during the fifth refueling outages for Units 1, 2, and 3, and that this schedule agr eed with the schedule included in the July 27, 1993, NRC letter noted above.
The inspector determined that the licensee requested actions in this letter appeared adequate.
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A licensee letter (322-00515-REG/PJM)
dated December 29, 1993, that contained the Plant Modification Committee (PMC) meeting minutes for a meeting held December 14, 1993.
These minutes identified th'at PCR 91-13-HJ-003 was tentatively scheduled to be performed in 1995 during the fifth refueling outages for Units 1, 2, and 3.
The inspector determined that the tentative schedule for performance of PCR 91-13-HJ-003 noted in this letter appeared adequate.
The inspector also discussed this item and the information reviewed, in the documents identified above, with various licensee personnel.
isc ssio d
o us on The inspector determined that the licensee had implemented appropriate compensatory measures in accordance with the licensee Fire Protection Program.
Resolution of this item and implementation of PCR 91-13-HJ-003 was being tracked under the licensee 10 CFR 50 Appendix R Reevaluation Effort as part of the
CFR 50 Appendix R
Reconstitution Project.
The licensee initial corrective actions and those that were tentatively scheduled, appeared adequate to resolve the original item.
This item is closed.
c.
C osed ollow U Ite 50-528 529 530 9 -036-02:
Chec V lve o ra es Or nal NRC Follow U Ite NRC inspectors identified that licensee procedure 73AC-OXI03,
r i
Revision 01.02,
"Preventive Haintenance of Check Valves," did not accurately reflect the check valves being actually worked as part of the check valve preventive maintenance program and that inspection scope changes were being done informally.
Licensee's Actions in Res onse to this Ite The licensee agreed with the inspector's findings.
The licensee committed to review this item and revise procedure 73AC-OXI03 by January 4,
1993.
Revision 01.03 of procedure 73AC-OXI03 was issued December 31, 1992.
ns ecto 'etio s Dur n the Present Ins ectio The inspector reviewed the following:
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Revisions 01.03 and 01.04 of procedure 73AC-OXI03.
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Revision 4 of Bechtel Power Corporation Study No. 13-HS-A24,
"Check Valve Evaluation Program for PVNGS."
The inspector also discussed this item and the information reviewed, in the documents identified above, with various licensee personnel.
iscussion nd Conclusio The inspector determined that Revision 01.03 of procedure 73AC-OXI03 had added information to clarify how the valve inspection scope changes should be handled.
However, neither revisions 01.03 or 01.04 of procedure 73AC-OXI03 had revised the procedure to include all the check valves identified for inspection in Study No. 13-HS-A24 (see Tables 5-9 and 5-13 of the study), or those check valves being actually worked as part of the check valve preventive maintenance program.
The licensee acknowledged this procedure discrepancy.
The licensee stated that as part of the initial long term corrective action for this item they were preparing a Revision 01.05 for procedure 73AC-OXI03, which would accurately reflect the applicable check valves identified for inspection in Study No. 13-HS-A24, or those cheep valves being actually worked as par t of the check valve preventive maintenance program.
The licensee identified that proposed Revision 01.05 to procedure 73AC-OXI03 would divide the'applicable check valves into four groups of valves instead of the ten valve categories now used.
The licensee commitment to issue Revision 01.05 to procedure 73AC-OXI03 to accurately reflect the applicable check valves identified for inspection in Study No. 13-HS-A24 and those check valves selected (based on actual check valve inspection results) for inspection as part of the check valve preventive maintenance program provides adequate action to resolve this item.
This item is close I
Closed Follow U Item 50-528 529 530 93-017-01:
Licensee to Include an 0 crab'1't Review for 11 the eat Exchan ers
'n Conditio a1 Re o t D's osition e uest No.
93053 r'n R
ol ow U Ite NRC inspectors identified that Generic Letter (GL) 89-13,
"Service Water System Problems Affecting Safety-Related Equipment,"
recommendations for heat exchanger thermal performance tests, although being implemented, were not verifying actual heat transfer capability of all safety related heat exchangers because of inadequate testing methods.
GL 89-13, Action 2, recommended that licensees conduct periodic test programs to verify heat transfer capability of all safety related heat exchangers cooled by service (spray pond) water and of heat exchangers in closed cycle systems
[Essential Cooling Water (ECW)] cooled by service water systems.
The letter further recommended monitoring flow, inlet and outlet temperature, and functional testing with adjustments to design conditions to verify adequate heat removal.
The inspectors noted that there were no installed flow instruments to measure flow for most of the important heat exchangers; e.g. the ECW, the shutdown cooling water, the diesel gener ator jacket water, air after cooler, lube oil, and fuel oil heat exchangers.
Without measurements, the system engineer assumed system flows'were split in accordance with design and then used a proportionate amount of total flow in the calculation.
The licensee committed to include an operability review for all the heat exchangers in Condition Report/Disposition Request (CRDR) No.
930532.
L'censee's Actions i Res onse to this Ite CRDR No. 930532 was issued August 25, 1993, which included a
licensee operability review for heat exchangers identified in the CRDR.
Ins ecto 's etio s Durin the P ese t ns ectio The inspector reviewed the following:
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n ee in issue a
E P
GL-00 3-2 dated Janua
99 This document detailed licensee corrective actions for the licensee's GL 89-13 Service Mater Reliability Program.
En inee in 'al t Re uest E
93-SP-005 dated Se te be
993 This document stated that "the diesel generators are run at
I
10 "'.
full load conditions at every outage when the integrated safeguards test is run.
This subjects them to 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> at IOOX power and an additional 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> at llOX power.
Data sheets from these full load runs show that there has never been a
tendency for any of the heat exchangers to run hot or close.to the high temperature alarm setpoint.
In Unit 3 at the first refueling, the, integrated safeguards test was run in July when the spray pond bulk water temperature was approaching its maximum.
Even with the higher cooling water inlet temperature, there was no overheating tendency.
A review of the data for successive tests shows that the temperatures experienced are consistent with only a slight upward trend [I to 2 degrees]
noted.
This slight trend provides ample time for maintenance cleaning well before the heat transfer capability is challenged, and validates the inspection results by demonstrating that the heat transfer capability has not been compromised."
licensee res o se lette 02-02626-MFC B DLK to t e NRC dated Se tember
1993 issued in es onse to Notice oF eviat on 50-528 529 530 93-
-02 This letter stated,
"An operability. review was performed on all open and closed-cycle service water system heat exchangers under CRDR 9-3-0532.
Based on the satisfactory performance of the heat exchangers during Integrated Safeguards Testing, Performance Engineering system temperature trend results, high quality system chemistry control, and the results of visual.
heat exchanger inspections conducted during refueling outages, engineering determined that all open and closed-cycle heat exchangers are capable of performing their intended safety function and are therefore capable of meeting their design basis requirements."
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CRD o
930 32 d ed Au ust
1993 This CRDR stated that the heat exchanger operability was reviewed and noted the following:
(I)
Starting with the first refueling outage beginning after April 18, 1990, inspections of the tube side (open cycle side) of all heat exchangers cooled by the open cycle service water (Spray Pond)
system have been performed on alternate trains, during alternate refueling outages.
These inspections looked at the channel heads, tubesheets and tubes ends, and piping spools adjacent to the channel heads.
(2)
C c e S d f
e t xc an ers - None of the regular or special inspections, had ever shown biofouling, corrosion products or silt of any significance in the open
cycle side of any safety related heat exchanger cooled by open cycle service water.
Further, metallurgical examinations of tubes pulled from the Unit 2 ECW heat exchangers (both trains)
showed no deposits of this kind in the tubes themselves.
On the basis of this evidence, there is no degraded condition due to flow blockage or fouling which would prevent the open cycle service water system heat exchangers from performing their safety function.
Further, monitoring and trending of the diesel generator operating temperatures, under load to demonstrate heat exchanger functional performance, has shown no tendency for the diesel to run hot or overheat.
(3)
C ose C
le Side o
Heat cha ers The closed cycle service water system has an existing high quality chemistry control program including weekly sampling for biological activity.
Inspection through pulled tube holes of the closed cooling water side of the EC'W heat exchangers (both trains) confirmed the effectiveness of the chemistry treatment program, in that no fouling or deposits of any kind were observed.
It was assumed that these conditions were representative of the conditions in the ECW side of the Essential Chiller condense and the shutdown cooling heat exchanger.
Thus, operability of the closed cycle service water system heat exchangers was assured by the inspections which confirmed the results of the chemistry treatment and sampling program.
Po t'ons of com leted rocedu e 73 EW01 Revis'o
" e t Bal nc'n of sse ti l Coolin Water S stem
"
erfo ed n
U t
o Na c 9- 0 986 This procedure test was performed to verify the heat balance and rejection capability of'the Essential Cooling Water System (ECWS), in conjunction with the Shutdown Cooling System (SCS)
and the Essential Spray Pond System (ESPS).
This test determined that the total heat rejection capability of the ECWS, SCS, and ESPS was acceptable under as tested and design conditions.
This test was performed during a reactor shutdown with one diesel generator train started and loaded.
However, the inspector noted that this test did not determine the heat rejection capability/performance of individual heat exchangers in the systems tested.
Various inspection work order data for visual inspections performed on heat exchangers.
The inspector also discussed-this item and the information reviewed, in the documents identified above, with various licensee personne i 1,
Dis ss o
a d Co c us An operability review of a heat exchanger is normally completed by gathering measured and calculated performance parameters for each heat exchanger and then performing engineering calculations to verify the thermal performance of each heat exchanger under design conditions.
For this item the licensee stated that since the original inspection did not identify an immediate operability concern based on the periodic full load tests of the diesel generators and the satisfactory periodic visual inspections of the heat exchangers, they elected to perform a review of historical test and visual inspection data for the operability review of the applicable heat exchangers.
The licensee review of historical test and visual inspection data did not identify any safety related heat exchanger operability concerns.
The licensee identified that they were tentatively scheduled to implement a revised heat exchanger thermal performance and heat exchanger capability testing program by March 31, 1994.
Therefore, since the historical data review did not identify any previous heat exchanger operability concerns, the licensee determined that a detailed operability review with engineering calculations for all the heat exchangers identified in CRDR No. 930532 was not required.
Based on this review, the inspector concluded that the licensee's cur rent and scheduled actions appeared adequate to resolve the inspector concerns.
The licensee is tracking the performance of GL 89-13 testing.
After the licensee issues the new heat exchanger test procedures and performs the associated testing, the licensee would perform an operability review of all heat exchangers based on the new test data.
This item is closed.
Closed Follow U Item 50-529 93-021-01:
Ade uac of Steam G ne ato Edd Current est'
ami at'on lan and Techn ues 0 i inal NRC Fo low U Item An NRC inspector noted that during eddy current examinations of the Unit 2 steam generators tubes, following the recent tube rupture event, the licensee made extensive use of rotating pancake coil (RPC) probes to examine suspect tube areas.
The inspector encouraged the licensee to evaluate eddy current testing (ECT)
techniques such as rotating field probes, multi-pancake coil probes and pilger-compensating data analysis software to improve the reliability of its ECT flaw detection capability.
The inspector identified that the adequacy of the licensee's current steam generator tube ECT examination plan and technique was being evaluated by NRR.
ce see's ctions es onse to t s Ite The licensee submitted their Unit 2 Steam Generator Tube Rupture
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Analysis Report (dated July 1993) to NRR for review.
Ins ector's Actions 0 r n the Present Ins ection The inspector reviewed the following:
An NRC Confirmatory Action Letter to the licensee dated June 4, 1993, to confirm commitments made by the licensee to the NRC.
A licensee letter to the NRC dated July 18, 1993, which submitted the licensee Unit 2 Steam Generator Tube Rupture Analysis Report (dated July 1993) to NRR for review.
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An NRC letter to the licensee dated August 19, 1993, which included an NRR Safety Evaluation (SE) related to the Startup and operation of Unit 2 following the steam generator tube rupture of March 14, 1993.
This letter also noted the licensee commitments to the NRC up to that date, regarding NRC concurrence with restart of the facility and other matters, had been satisfied.
The inspector also discussed this item and the information reviewed, in the documents identified above, with various licensee personnel.
iscussion and Co elusion Based on review of the above information the inspector determined that the adequacy of the licensee ECT examination plan and techniques in use as of the date of this inspection, had been evaluated acceptable by NRR.
The licensee, NRC, and NRR were continually reviewing ECT examination plans and techniques to improve ECT examinations of steam generator tubes.
This item is closed.
f.
Closed U resp ed Item 50-528 93-04 -Ol: Fai r
o s
Fo d
oc eak ate est of Check Valve IPSB 533 Ori nal RC ol ow U
te An NRC inspector noted that while performing an as found local leak rate test (LLRT) on penetration 67 (long term recirculation) safety injection check valve 1PSIBV533, the valve disk seat leakage was so large that test pressure could not be obtained.
After testing of adjacent valves in the same line a subsequent test of the check valve measured an acceptable
'LLRT leakage rate of 411 SCCA.
However, no corrective action was taken, other than a retest of the check valve.
The valve was not opened, nor was an examination performed of the valve internals.
The licensee issued Condition Report / Disposition Request (CRDR) 1-3-0584 dated October 18, 1993, which identified the following:
,
l l~(
t
arent Cause of the Cond tion The potential causes for the excessive valve seat leakage exhibited by check valve 1PSIBY533 were as follows:
(1)
Debris existing between the valve disk and the seat.
(2)
An accumulation of deposits between the ball and the race creating a stiffness that prevented the valve disk from seating properly.
(3)
The disk coming into contact with the valve body causing it to hang-up as identified in a Borg Marner 10 CFR Part 21 notification.
The licensee considered that debris existing between the seat and the disk, or some stiffness between the ball and race, or a combination of the two, would be the more credible of the three identified potential causes.
A definitive inspection of the valve internals was not performed.
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Back rou d - Borg Marner issued a
CFR Part 21 notification (CFRN-9302) that identified a concern with disc to body contact in various models of Borg Warner check valves which could result in a disc becoming stuck open.
The area of contact was the junction between the horizontal and vertical bores in the valve body.
A CRDR 9-3-0149 identified containment isolation check valve 1PSIBVV533 as a potential candidate that this concern could apply to.
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Corrective Actions - CRDR 9-3-0149 dated October 28, 1993, addressed the concern of the Borg Marner 10 CFR Part
notification and the failure of the LLRT of valve 1PSIV533 identified in CRDR 1-3-0584.
The corrective action of CRDR 9-
.3-0149.was to inspect valve 1PSIBV533 at the next available opportunity with continued inspections as required by the check valve program.
No additional actions were taken.
The
CFR Part 21 report identified similar conditions that could result in a stuck open valve.
The inspector considered that the licensee actions in response to CRDR 1-3-0584 may have not sufficiently taken into account the recommendation of. the
CFR Part 21 report, including examination of the valve internals.
The inspector recommended additional NRC review of the licensee's basis for the action it took.
ce ee's ct'o s es o se t
s te The licensee implemented action to investigate the applicability of the Borg Warner
CFR Part 21 notification CFRN-9302 to valve 1PSIBY533 and the basis for their original valve corrective action Ins ector's Actions Dubbin t e res nt Ins ect on The inspector reviewed the following:
~
A licensee letter (315-00242-TNM) dated September 17, 1992, which discussed the findings of a licensee inspection of two BM check valves obtained from the licensee warehouse stock.
These inspections were performed prior to the date of issue of 10 CFR Part 21 Notification CFRN-9302.
~
CRDR 9-3-0149 dated Dctober 28, 1993, which requested evaluation of 10 CFR Part 21 Notification CFRN-9302 for various models of Borg Warner (BW) check valves.
~
CRDR 1-3-0584 dated November 12, 1993, which identified valve 1PSIBV533 as a model 77700 BM valve that would be inspected per CRDR 9-3-0149.
~
A BM/IP letter dated January 11, 1994, which discussed the differences between the original
CFR Part 21 Notification CFRN-9302 valve and a model 77700 BM valve.
~
A licensee letter (315-00420-TCC)
dated January 31, 1994, which identified the following:
{1)
The original
CFR Part 21 Notification CFRN-9302 dated February 18, 1993, identified a concern regarding interference between the disc/swing arm assembly and the valve body on BW bolted bonnet swing check valves.
{2)
A BM model 77700 was not a bolted bonnet design valve.
However, a subsequent BW/IP letter dated April 21, 1993, identified that 10 CFR Part 21 Notification CFRN-9302 could apply to a BM model 77700 valve, in regards to the potential for disc to body contact.
This was contrary to the original
CFR Part 21 Notification.
(3)
Per a BW/IP letter dated January 11, 1994, the clearance between the disc and the valve body on a model 77700 valve was much greater (0.200 vs. 0.0625)
than that of the
.
particular model valve that was found stuck open per
CFR Part 21 Notification CFRN-9302.
Therefore the model 77700 valve should not have any interference between the disc and the valve body.
The inspector also discussed this item and the information reviewed, in the documents identified above, with various licensee personnel.
iscussion nd Conclus o
Valve 1PSIBV533 failed its as found i.LRT toward the end of the
I
Unit 1 fourth refueling outage.
The latest recommended corrective action for valve 1PSIBV533 was to examine the valve internals during the next available opportunity, tentatively scheduled for the next Unit 1 refueling outage.
Since the original NRC inspection, the licensee had contacted BW and obtained additional information on how the problems discussed in 10 CFR Part 21 Notification CFRN-9302 could or could not have been related to valve 1PSIBV533.
Based on this review, the inspector determined that while the original licensee actions in response to CRDR 1-3-0584 did not appear to sufficiently take into account the recommendation of 10 CFR Part
Notification CFRN-9302 to examine valve internals, the new BW information (contained in BW/IP letter dated January 31, 1994, noted above)
appeared to reduce the concerns for immediate examination of the valve internals.
Based on the licensee's intention to inspect the valve internals during the next refueling outage, this item is closed.
4.
Fo low 92702 Corrective Actions o
P ev ou Ide ti ied V'o a 'ons Closed V'olation 50-530-36-03 F 'l e to o rec Oocu ent Identif ed Unacce table Conditions Noted in a Check Valve a
d a Mec anical Snubber.
Ori i al NRC Violation NRC inspectors identified two examples where the licensee had observed'hat appeared to be unacceptable component conditions and failed to correctly document these conditions.
QLaaml~e On April 11, 1991, an examination of non-essential auxiliary feedwater (AFM) pump AFN-P01 discharge check valve 3PAFN-V012 was performed'and the examination results were documented in check valve examination report number 91-354.
This report identified that examination checklist item 4.f (hinge arm / bracket connection locking devices)
and item 6 (additional exams)
were rejected.
This report identified that the internal hinge bracket cap screws did not have their locking device tabs bent up to lock the cap screws in place and that the existing external valve bonnet cap screws were not made from the design specified material.
An engineering evaluation request (EER) was not initiated following licensee identification of,. the unacceptable conditions.
An engineering evaluation and corrective action were annotated on the check valve examination report.
However, this process did not provide sufficient documentation of the evaluations, the supervisory review, the feedback for possible generic consideration nor possible operability considerations.
/~am
~le On October 7, 1992, an examination of mechanical snubber support 3-SI-
i I
il l
241-H007 was performed and the examination results were documented on visual examination report number 92-3258.
The report identified that the snubber evaluation was rejected, a nut and bolt were found loose, and a
note on the report stated
"design load on snubber reads 6000 lbs., load on drawing reads 8308 lbs.."
A footnote added by a licensee reviewer to the original examiner note stated in part that, "condition verified acceptable per conversation with...
The footnote did not document the basis for the snubber engineer acceptance of the 6000 lbs. snubber.
HNCR No. 92-SI-3118 was issued to correct the loose nut and bolt.
Neither a material non-conformance report (HNCR) nor an EER was initiated to evaluate an identified potential unacceptable condition with the identified loads.
icensee's Actions I Res o se To he V'olatio Exam le of t e Violat on The licensee, in their April 23, 1993, response to example 1 of the violation, identified the following:
~
The violation occurred because the Check Valve Engineer believed at the time of the inspection, that since the conditions did not affect the operation of the check'valve, documenting and correcting them using the APS Work Control Program, was adequate.
~
The visual examination procedure, 73TI-9ZZ19, and the administrative control procedure, 73AC-OXI03, have been revised to require a
Haterial Non-Conformance Report, or an EER for Non-guality Related items, be generated for all rejected inspection blocks on the check valve examination checklist.
am e
2 of the Violat'on The licensee, in their December 31, 1992, response to example 2 of the violation, identified the following:
The evaluation block was marked "reject" on the Inservice Inspection (ISI) Report 92-3258.
The reason for the rejection was a loose nut and bolt.
This condition was documented on HNCR 92-SI-3118 and subsequently reworked and accepted on ISI Report 92-3262, dated October 10, 1992.
The initial ISI Report (92-3258) contained a note, which read,
"design load on snubber reads 6000 lbs.,load on drawing reads 8308 lbs."
This apparent discrepancy was clarified by the APS Snubber Engineer and documented on ISI Report 92-3258.
Stress calculation 13-HC-SI-501 identifies the maximum calculated stress loads under various accident conditions.
The worst case condition, Level D, is 8308 lbs.
The design drawing, 13SI-241-H007, Revision 5, reflects this worst case maximum calculated design stress'oad (8308 lbs.).
As documented in APS specification 13-PN-209 (the subject snubber's
l
procurement specification),
snubber 3SI-241-H007 has a maximum load capacity of 10385 lbs.
The qualified load rating for 3SI-241-H007 under accident conditions A and B is 6000 lbs., which is the nominal design load rating.
The nominal design load rating (6000 lbs.) is stamped on the snubber nameplate.
Since no deviation exists between the snubber nameplate, the design drawing, or the stress calculation, no HNCR or EER was warranted.
~
APS considers the action taken to resolve the nonconformance condition and the questions that arose during the snubber inspection appropriate.
~
Based on the facts surrounding example 2 of the violation, APS denied example 2 of the violation.
The NRC response to this denial of example 2 of the violation stated the NRC would review the information regarding snubber sizing during a followup inspection for this violation.
Ins ector's Aetio s Durin T e Present Ins ect on
~am le l The inspector reviewed the following:
~
Licensee visual examination procedure 73TI-9ZZl9, Revision 06.02,
"Visual Examination of Pump and Valve Internal Surfaces."
Step 8.4.2 of the procedure had been revised to require a Material Non-Conformance Report, or an EER for Non-guality Related items, be generated for all rejected inspection blocks.
~
Licensee administrative control procedure 73AC-OXI03, Revision 01.04,
"Preventive Maintenance of Check Valves."
Step 3.2.6 of the procedure had been revised to state,
"EERs are required for Non-guality Related (NgR) components for all rejected inspection blocks.".
The inspector also discussed this item and the information reviewed, in the documents identified above, with various'icensee personnel.
~amide The inspector reviewed the following:
~
Licensee CRDR 3-2-0572, dated December 21, 1992, which stated that ISI Visual Examination Report 92-3258 evaluation block was marked
"reject" only because of the loose nut and bolt.
This CRDR also stated the difference between the "Design Load" information stamped on the snubber nameplate and the "Design Load" information provide in the support drawing was resolved
"on the Spot" by the cognizant (snubber)
engineer.
The licensee identified that they did not
f I
consider the difference between these loads as abnormal, and therefore this difference was not respectable.
Licensee procedure 60AC-Ogg01, Revision 3, "Control of Nonconforming Items," which was the licensee MNCR procedure in effect at the time of the snubber inspection.
~
Stress calculation No. 13-MC-SI-501, sheet for Problem No. 13-MC-SI-5018, support drawing No. 13-SI-241-4-007, Rev. 5, which identified the maximum calculated stress loads under various accident conditions for snubber 3SI-241-007.
~
Licensee EER 92-SN-017, dated November 13, 1992, which determined that the design load had not exceeded the load carrying capacity of the snubber.
Grinnel Corporation Load Capacity Data Sheets for Figure 306N and 307N Mechanical Snubbers, which identified Maximum Loads for snubber 3SI-241-007.
The inspector also discussed this item and the information reviewed, in the documents identified above, with various licensee personnel.
scussion and conclusio
~xam le
The inspector concluded that the licensee's corrective actions adequately resolved this example of the violation.
g)gmm'~e The inspector concluded that the initia'l ISI snubber inspection was performed by a contract examiner, who documented a concern with the difference design loads identified on the snubber and snubber drawing.
The licensee considered the contractor's concern a statement of a discrepancy between the two documents, not an identification of a rejectable condition.
While the licensee's use of a footnote (with a reference to a conversation)
to determine the contractor's concern to be acceptable appeared to be a weak method for documenting an engineering evaluation, it did not appear to be a violation of their inspection procedure.
After reviewing the available information for Example 2 of
'his violation, the inspector concluded the concern with the difference in design loads had been adequately resolved.
This item is closed.
5.
Ot e 0 e Items
0 a d 9 70 The inspector checked on completion of other open items.
The licensee stated that their actions on the following items were not yet complete.
LER 93-002-LO
- Loss Train "B" EDG Due to Fire in Control
II
)
I l
LER 93-005-03 Deviation 93-017-02 Unresolved 93-049-02 6.
~Ed tl
Room
- Loss of Redundant Trains of SSE Due to Fire
- Failure to meet GL 89-13 Commitment
- Local Leak Rate Testing of Air Lock Door Equalizing Valves The inspector conducted an exit meeting on February 11, 1994, with members of the licensee staff as indicated in Section 1 of this report.
During this meeting, the inspector summarized the scope of the inspection activities and reviewed the inspection findings as described in this report.
The inspector identified that he had requested some additional information that would be reviewed in the region.
This additional information was reviewed by the inspector and included in sections 3 and 4 of the report.
The licensee acknowledged the inspector findings as identified in this report.
The licensee did not identify as proprietary any of the information provided to, or reviewed by, the inspector during this inspection.