IR 05000424/1989020
| ML20248D201 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 07/21/1989 |
| From: | Jape F, Casey Smith NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20248D156 | List: |
| References | |
| 50-424-89-20, 50-425-89-24, NUDOCS 8908100289 | |
| Download: ML20248D201 (11) | |
Text
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'\\ , . 50-424/89-20 and 50-425/89-24 .
, ? Report Nos.: . , -Licens'ee: Georgia Power Co...pany E P. O. Box 1295 _ ( Birmingham, AL. 35201~ ] " , . ,.*; , .
, ~ Docket: Nos. :. - 50-424 and 50-425 ' License Nos.: 'NPF-68 and'NPF-81 , Fadility Name: Vogtle 1 and 2 - j . - i Inspection Conducted: June 26-30, 1989 '! " . Inspector: & h[1 9 4#ac_
pSmith,TeamLeader
- V Date Signed j
L Team, Members: M. Thomas ! E. Lea
- j S. Ninh l
, '7b# [[7 ' ' ' Approved'by:' A (4M- /. ~~Date Signed . ' F. Jape, Section Chief gf ' , . , Quality Performance Section Operations Branch Division of Reactor' Safety !
- StNMARY ! +
-Scope:. , ! This routine, unannounced. inspection was conducted in the areas of design, , design ' changes, and. plant modification.
l ' Results: ] . .
. Independent design reviews of completed DCPs were performed to. determine their i
technical adequacy.
The Safety Evaluations performed in connection with the ,4 J plant -' modifications were technically adequate; post-modification -test a; requirements and test acceptance criteria were specified;~and-the design output l ' documents were technically correct, (para 2.0).1 Additionally, the licensee's ! 1 temporary modification process appeared to.be adequately. planned, executed, and
documented, (para 3.0).
. , One violation was, identified in~ connection 'with the disposition of REA-7604, p.
dated October.14, 1987, and is discussed in paragraph 4.0. - ! l ' > l.
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+ . ] . ... .. . ,. ,' REPORT DETAILS j 1.- Persons Contacted.
Licensee Employees -*J. Aufer Kampe, Manager, Technical. Support
- C. Cross, Nuclear Production Supervisor, Technical Support
- G. Frederick, QA Site Manager, Nuclear Operations-
- W. Gabbard, Senior Regulatory Specialist. Technical Support
- C. Garrett, Nuclear Operations Engineer
- H. Handfinger, Maintenance Manager
- M. Horton, Engineering Support Manager
- W. Kitchens, ^ Assistant General Manager
- G. McCarley, 'ISEG Supervisor
- R. Odom Plant Engineering Supervisor
- J. Swartzwelder, Manager - Operations Other licensee. employees contacted during this inspection included j
engineers, operators, and administrative personnel.
. Other Organizations' W. Cherault, Engineer, American Engineering and Technical Associates
- A. Gillette, Engineering Manager, Enterprise Engine Services, Division of Cooper Industries NRC Resident Inspectors J. Rogge. Senior Resident Inspector R. Aiello, Resident inspector
- Attended exit interview
- Contacted via telecon on June 29, 1989 2.
Design Control Program (37700) a.
DCP No. 37-VIE 0311-01, Revision 0 The above DCP was developed and implemented to improve system operation and performance in connection with the MFW 1 solation Bypass valves.- These valves require 3-5 minutes to go from the closed position to the fully open position.
'If the operatcr should inadvertently release the hand switch the valves fail close. Nuclear Operations department stated in a DCR that the slow opening and fast closure of these components provides an unwanted distraction when the control room operators are in the process of changing steam-generator makeup from the AFW system to the MFW regulator valves.
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The stated design objective was to provide a seal-in contact to reduce operator action in maintaining the hand switch in the open position.
The design scope involved installing stem mounted limit switches and associated cables for MFW 1 solation Bypass valves IHV-15196(B) and 1HV-15198(B).
Additionally, the control circuits for va'ives 1HV-15196,1HV-15197,1HV-15198 and 1HV-15199 would be modified to incorporate a logic input from the ZSC, indicative of the valve not being in the closed position.
The safety evaluation performed in accordance with the requirements of 10 CFR 50.59 was reviewed by the inspectors to assess the technical adequacy of the evaluation.
No deficiencies were identified during this review. The following design output documents were reviewed by the inspectors to verify that the DCP was consistent with the design scope.
Drawing No. B2-J-87-VIE 0311-100, Control Logic Diagram MFW System, Revision A.
Drawing No. B2-E-87 VIE 0311-101 Condensate and FW Sys.1HV-15196 A, B Sheets 3-8.
Drawing No. 82-E-87 VIE 0311-102, Elem. Diag-Condensate and FW Sys. l'iY-15197 A and B Revision A.
Drawing No. B2-E-87 VIE 0311-103, Elem. Diag. Condensate and FW System 1HY-15198 A and 'B, Revision A Drawing' No. B2-E-87 VIE 0311-104, Elem. Diag. Condensate and FW System,1HY-15199 A and B, Revision A.
These drawings represent the electrical schematics required for installation of the DCP.
Additional wiring diagrams included in the DCP were selectively reviewed to verify conformance with the plant modification design scope.
Based on the above independent design review, the inspectors determined that the hardware changes shown on the above drawings were consistent with the design scope as described in the Narrative Design Summary.
Requirements for procurement of EQ limit switches were specified in the DCP.
The inspectors verified, by review of Purchase Order No. G-04197, and NAMCO Controls Certificate of Compliance, that the limit switches procured met the requirements of 10 CFR 50.49.
Additionally, functional test requirements and test acceptance criteria specified 'in the DCP were reviewed by the inspectors and verified to be adequate to demonstrate achievement of the design objective.
I Within this area no violation or deviation was identified.
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DCP No. 87-VIE 0183, Revision 0 - _ _ - _ _ _ _ - _ _ _ _ _ - _ _ _ _ _. _
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i, DCP No. 87-VIE 0183,125V DC System Cables IND3A0lLA, IND3A0lLB and IND3A0 LLC, was developed and installed to address a concern related , to the charging of 125V DC flon-class IE battery IND 3AB.
The
licensee's consultant determined that following a battery discharge, ! the battery would be recharged at the maximum output of the chargers (105%), i. e.1159.9 amperes, which exceeds the emergency overload cable ampacity of 1083 amperes.
This situation arises when all loads of the 125V DC switchgear IND3A01 are disconnected except for the battery.
The consultant also stated that the above conditions for ' recharging battery IND3AB is outside the desigr basis of the system, and provided information concerning this issue in the following letters: Bechtel Western Power Corporation letter Log: BG-35153 dated March 11, 1987 Bechtel Western Power Corporation letter Log: BG 34689 dated July 22, 1986 The recommended design change to facilitate recharging 125V DC Non-class IE battery IND3AB under the above conditions was to add a 500 MCM cable to each set of cables from switchgear IND3A01 to battery IND3AB. This plant modification was intended to increase the ampacity of the feeder cables with a resulting reduction of the cable temperature rise to less than 130"C.
The installation would thereby meet the guidelines of IEEE 242-1986 under a complete recharge condition.
The inspectors reviewed design basis calculations X3CF01, Battery Sizing (Normal D.C. System); X3CR08, D.C. Power Cable Sizing; and X3BC09, D.C. Breaker Sizing, to verify the technical adequacy of the design inputs.
Based on review of the calculations and discussions with licensee engineering personnel the inspectors determined the root cause for the plant modification to be the design requirements for static inverters IND312 and 1ND313.
For a Non-class 1E system, upon restoration of the Non-class IE 480 VAC power supply, the inverters are aligned for operation from this system. This reduction in the 125 VDC switchgear normal load therefore resulted in the load sharing battery chargers output current exceeding the ampacity rating of the feeder cables to battery IND3AB, when the battery is fully discharged.
DCP 87-VIE 0183 was implemented to currect this design deficiency in a non-safety related system.
Discrepancies in listed values for (1) Inverter Load; (2) Nonnal Operating Load; and (3) Rating Used for Charger Sizing were identified between the design basis calculations and the above referenced letters.
These discrepancies were resolved during discussions with licensee engineering personnel.
Based on additinnal reviews of the Safety Evaluation and the design-output drawings, the inspectors verified that the plant modification was technically adequate and was consistent with the design scope.
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Within this area no violations or deviations were identified.
c.
DCP No. 888 VIN 0078, Revision 0.
i The above DCP was developed and implemented to provide redundant thermal protection for containment penetration assemblies conductors.
The design scope included replacing existing magnetic only circuit breakers with 15 Amperes thermal magnetic circuit breakers in the following MCC cubicles,1ABC13,1ABC15, 1ABC32, 1ABC33, IBBC15, IABD44, IBBD44, 1 ABE10, 1 AB E11, 1 ABE13, 1 ABE16, 1 ABE24 1BB E15, IBBE24, IBBE25, and 1BBE37.
The inspectors performed a partial review of the above DCP and determined that the plant modification was intended to address a concern related to Class 1E MOVs which had their then6al overload l relay contact bypassed per Regulatory Guide 1.106.
The ' magnetic circuit breakers provided short circuit protection only. Because of this, they could not be set such that they would trip below the containment penetration continuous rating and still allow starting the MOVs without tripping the MOVs on starting inrush current.
The 15 amperes thermal magnetic circuit breakers were installed to provide redundant thermal overload protection for the containment penetration conductors.
Design basis calculation MX3CM01, Electrical Penetration Short Circuit Currents, Revision A1, was not available to the inspectors for review.
However, the inspectors reviewed the Nuclear Safety Evaluation, functional testino requirements, and selected drawings and verified that the DCP was consistent with the design scope.
Based on the above review of DCP No. 88 VIN 0078, no violations or deviations were identified.
d.
DCP No. 87-VIE 0262, Revision 0 DCP No. 87 VIE 0262, Rev. O, Unit 1. " Containment Pressure Indicator PI-10945, war developed and implemented to provide a new non-safety instrumentation loop for monitoring containment narrow-range pressure of -2 to +4 psig.
This Loop provides backup instrumentation to ERF computer point P-987 in accordance with T.S. requirement 4.6.1.4.
Additionally, it provides narrow-range pressure indication in the main control room.
This installation was modified in response to a single failure of either (1) ERF computer (2) pressure transmitter (PT-10944) or (3) the interconnecting cable, which could eliminate any method of accurately determining containment pressure in the normal operating ranges.
The inspectors reviewed the DCP package and determined that engineering safety evaluation was thorough, addressing the potent 12i effect on the FSAR and technical specification as well as unreviewed safety questions.
Functional calibration test results were reviewed
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< and found to be acceptable.
Training was performed as a result of this modification, the affected procedure, Technical Specification, drawings and FSAR were updated.to s * ect the change.
The' DCP package was completed in accordance with the procedure ' requirements and found to be adequate.
In addition, the inspectors independently verified that the PT-10945 indication was installed as-specified in the DCP.
No violations or deviations were identified in this area.
'e.
DCP;No. 88-V1N0049, Revision 0 This DCP involved limiting the travel of the three-way thermostatic valves 1-TCV-19096 and 1-TCV-19097 installed at the inlet to the shell side of the jacket water heat exchangers-for Unit 1 EDGs IA and.
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The original design specified jacket water flow through the heat-exchangers to vary between 0-1800 gpm according to the temperature of the water, and as controlled by the three-way thermostatic valves.
However, it was determined that the shell side of the heat exchangers ~ was. only designed for 750 gpm instead of the 1800 gpm specified in the original design.
This design deficiency was reported to the NRC on January 19,1988, under 10 CFR 21 by ' the EDG vendor (formerly Enterprise Engine Division of IMO Delaval Inc., but now Enterprise Engineer Services-Division of Cooper Industries).
Prior to the determination that there was a design deficiency, _the licensee had reported the item under 10 CFR 50.55(e) on September 17, 1986, stating that the EDG's jacket water heat exchanger was damaged , during flushing on Unit 1 (CDR 424/86-109 and 425/86-109). The heat exchanger was repaired and installed in Unit 2, but was damaged again during testing' for Unit 2. The subsequent-investigation resulted in the discovery of the design deficiency which led to the 10 CFR 21-report by the vendor.
It appears that the damage to the heat exchanger was caused during flushing when the original design flow of 1800 gpm was used.
The licensee submitted a revised 10 CFR 50.55(e) report for Units 1 and 2 on April 14, 1988.
The modification was installed in Unit I during the 1988 refueling outage under MW0s 18806163 and 18806186. Unit 2 valves, 2-TCV-19096 and-2-TCV-19097, were modified under MW0s 28801578 ano 28801577 respectively, prior to Unit 2 receiving an operating license.
The , ' vendor' performed additional analyses and calculations for the EDG design bases operating condition and determined that the maximum - , L jacket water flow to the shell side of the heat exchangers should be limited to '<36.9 gpm. No damage would be caused to the heat exchangers at this flow rate.
The inspectors reviewed the post modification test results where it was demonstrated that the three-way thermostatic valves had been properly modified to limit jacket water flow to less than 786.9 gpm.
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In addition, the inspectors also reviewed test results for the Unit I surveillance tests and the Unit 2 preoperational test listed below to verify that EDG jacket water temperatures were maintained within design requirements during operation of the EDGs under design basis loading conditions. The following tests were reviewed: 54055-1 Train A Diesel Generator and ESF System Actuation Test 54065-1, Train B Diesel Generator and ESF System Actuation Test 2-300-01, Integrated Safeguards and Load Sequencing Test, Revision 1 No violations of deviations were identified, f.
DCP No. 88-VIN 0108, Revision 0 This design change involved modifying the ESF chiller circuitry in order to prevent the ESF chillers frem tripping (if already running) when a safety injection signal is initiated. This DCP was completed during the 1988 Unit I refueling outage under MW0s 18807440 and 18807514. The inspectors reviewed the completed DCP package and MW0s for accuracy and adequacy. The inspector also reviewed the completed post modification test which was performed under temporary procedure T-ENG-88-018, ESF Chiller Sequence Block Test.
The inspectors verified that the testing satisfied the design requirements and I demonstrated that the design change had been properly implemented.
No violations or deviations were identified.
3.
Temporary Modifications As of June 26, 1989, there were 55 active temporary modifications for both units.
Five TMS were randomly selected to determine the effectiveness of design control and documentation of TMS.
The following five TMS were reviewed: TM Number Title 1-87-395 Hotwell Level Indicator - LI-4415.1305 1-87-437 To Protect the Normal Chilled Water System from Overpressurization Due to Thermal Expansion 1-87-481 Remove Blades in Backdraft Dampers 1-88-035 Remove Rad Drain Pipe Support Attached to Auxiliary / Containment Building 1-88-082 Remove Fire Detector from Zone 390 i ___ ... . .
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, The documentation. reviewed was completed as. per p1 ant: procedure Number 00307-C, Temporary Modifications.
However, the inspectors identified one deficiency-involving the 10 CFR 50.59 safety evaluation prepared for TM No. 1-88-082, " Removal of. Fire Detectors. from Zone 39D, Unit 1".
The documented design bases for removal of the fire detectors from Zone 39D was provided by a memorandum prepared by licensee consultant for removal of similar Unit 2 detectors. Licensee management was informed that this memorandum did not provide adequate design bases for removal of ^,1 Unit 1 detectors. Additionally, this issue was identified as an Unresolved ' Item at the exit interview, pending receipt of additional information from the' licensee that was not available at the time of the inspection.
The follcwing documents were subsequently provided by the licensee; Southern Company Services letters No. X4BJ01 Log: SG8137, dated June 6, 1989;: and. Number X4BJ01 Log: NPFSG-03735, dated April 12, 1989. Pursuant to review of attached calculation No. 'X4C2301515, the inspectors verified that-the combustible loading for fire zone 39D is zero.
Based on this review, the issue is ' closed.
Overall,. the licensee's temporary modification process and control appeared to be adequately planned, executed, and documented.
It was also noted that the licensee periodically review all active TMS to maintain awareness of the effect of all TMS on plant status.
~ No violations or deviations were identif sed in this area.
4.- Engineering Support
The inspectors reviewed RERs and REAs and interviewed licensee personnel to determine how engineering respond to concerns identified in the plant.
The initiating document for both RERs and REAs-is plant procedure 00400-C, Revision 9, dated September 30, 1988; Plant Design Control.
i A total of 20 RERs and 20 REAs were reviewed.
In general adequate attention-was given to the RERs reviewed.
Several of the RERs reviewed , ' resulted in DCRs/DCPs being issued. RER 88-0014, dated December 29, 1987, identified a problem with Unit 2's diesel generator jacket cooling water heat exchanger.
Tube failure occurred during flushing, Information pertaining to the tube failure was assembled and evaluated by Southern Company Services.
As a result of the evaluation DCR 88-VIN 0049 was initiated.
The scope and results of the implementation of this DCR is , addressed in paragraph 2.e of this report.
RER 88-0735, dated October 10, 1988, identified a problem with Essential Chilled Water System.
During ESFAS testing the ESF "A" chiller was noted to have both an amber and green light indication following the initiation of Train "A" Safety Injection Signal.
Engineering generated DCR 88-VIN 0108 to eliminate the double indication.
The scope and results of , this DCR is also addressed in paragraph 2.f of this report.
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8 While reviewing REAs files the inspector identified an REA (REA-7604) initiated October 14, 1987.
At the time of the inspection no evaluation had been performed on the request.
The inspectors were informed by the licensee that the REA had not left the department at this time.
No departmental approval had been granted to have the technical support staff review the con:: erns addressed on the REA.
The REA addressed an apparent prcblem regarding the acceptability of 14 oversized penetration seals.
The penetration seals were installed without subdividers as required by desitri, therefore no design test data.
The design test data is required , I to show evidence that penetration seals provide a minimum fire resistance rating of three hours. The three hour minimum fire rating is addressed in NUREG-0800, Standard Review Plan, Section 5., and in the licensee's FSAR.
Section C.5.
At the time of the inspection the licensee was unable to provide technical data which demonstrate the following oversize penetrations seals are capable of providing the minimum three hour fire rating as required by design bases Seal Tag No.
026/2-1-08-288-0 053-1-08-816-C 053-1-08-817-C 067-1-08-312-B 067-1-08-320-B 068-1-08-329-B 068-1-08-330-B 069-1-08-357 '8 070-1-08-371-B 071-1-08-409-B 125-1-08-617-A 125-1-08-618-A 126-1-08-634-A 133/8-1-08-904-A Penetration problems were first identified during an NRC inspection conducted February 24-28, 1986.
The items were identified as Inspector Followup Item (424/86-13-03 and 424/86-64-06).
Both followup items addressed oversize penetrations and the absence of records to verify that the penetration seals met the design requirements for a three hour fire resistant rating.
Both IFI were closed during an NRC inspection conducted January 12-16, 1987.
The licensee provided information to the NRC/NRR on a letter dated October 24, 1986 and during a meeting on November 4, 1986.
The information provided was used to justify the acceptability of the oversize penetration.
The letter dated October 10, 1986, specified that the subdividers/ hot board were installed in oversize penetration.
It also stated that special test were performed and proved that the oversize penetration seals with subdividers/ hot board installed c!id provide effective fire barriers as required by design.
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Vogtle SSER 4, section 2.14 also indicates that all oversize penetrations were constructed with subdividers. Acceptance of the oversize penetration seals were based on. test data provided in " Bisco Test Report 798-134, May 14, 1984."
In a letter from the licensee to the NRC, dated December 14, 1988, there is further indication that oversized penetrations at Plant Vogtle Unit I were installed with subdividers or a material called hot board.
This information is used as a method of justifying acceptability of Unit 2's oversize penetration seals.
Several other RERs/REAs reviewed resulted in action being taken (DCRs, procedure changes, etc...). The failure of the licensee to take adequate corrective to determine if existing penetration seals are capable of providing the three hour minimum fire rating as required by design is identified as violation 50-424/89-20-01, 50-425/89-01, failure to provide timely and adequate corrective action on a REA.
5.
Exit Interview The inspection scope and results were summarized or June 30, 1989, with those persons indicated in paragraph 1.
The inspectors described the areas insper,ted and discussed in detail the inspection results listed below.
Proprietary information is not contained in this report.
) Dissenting comments were not received from the licensee.
Untimely corrective action related to the preparation, review and disposition of REA No. 7604, dated October 14, 1987, was identified as a violation; 50-424/89-20-01, 50-425/89-24-01.
6.
Acronyms and Initialisms AFW Auxiliary Feedwater DC Direct Current DCP Design Change Package DCR Design change Request ERF Emergency Response Facility FSAR Final Safety Analysis Report MCC Motor Control Center MFW Main Feed Water PT Pressure Transmitter TMS Temporary Modifications V Volt ZSC Closed Position Limit Switch REA Request for Engineering Assistance RER Request for Engineering Review IFI Inspector Followup Item CDR Construction Deficiency Report CFR Code of Federal Regulations _ _ _ _ _ _ _ _ _ i
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EDG Emergency Diesel Generator 'ESF Engineered Safety Features GPM Gallons Per Minute l-MWO.
-Maintenance Work Order TCV . Temperature Control Valve -! I l - _ _ _ - _- }}