IR 05000341/1989200
| ML20248G375 | |
| Person / Time | |
|---|---|
| Site: | Fermi  |
| Issue date: | 09/28/1989 |
| From: | Gramm R, Imbro E, Jeffrey Jacobson, Lanning W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20248G364 | List: |
| References | |
| 50-341-89-200, NUDOCS 8910100216 | |
| Download: ML20248G375 (24) | |
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U.' S. NUCLEAR REGULATORY COMMISSION-~
OFFICE OF NUCLEt.R REACTOR REGULATION Division of Reactor Inspection.and Safeguards Report'No.: 50-341/89-200 Docket'No.: 2 50-341 Licensee: . Detroit Edison Electric Company racility:' Enrico Fermi Atomic Power Plant, Unit 2 t
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Inspection Conductet. July 17 through '21 and July 31 through'tagust 4,1989
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Inspection Team bers:
w h R6bert Grd..., Team Leader .
El N Date Signed Special Inspection. Branch, NRR h h Jeffrey Jacobson, Assistant Team Leader 9 fg,(
Da".e Signed H
Special. Inspection Branch, NRR Electrical Power / Instrumentation and Control: S. V. Athavale, NRR, RSIB 0. Mazzoni, Consultant-Mechanical Cor..ponents: J..Lucena, Consultant
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Mechanical Systems: J. Rugieri, Consultant Motor-0perated Valves: J. Jacobson, NRR L. Zerr, NRR Team Member: *J. Stang, NRR Reviewed By: b 7. b b
.E. V. Imbro, Chief
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Date Srgned
t Team Inspection Section 2B Special Inspection Branch, DRIS, NRR Approved By: /lh ws W; D.~/ Lann~iTig, Chief f/28[ff Ddte figped Special Inspection Branch, S, NRR
- Part time
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SUMMARY INSPECTION REPORT 50-341/89-200 l DETROIT EDISON COMPANY-(DECO)
ENRICO FERMI ATOMIC POWER PLANT, UNIT 2 (FERMI-2)
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The' design phase of the safety systems outage modifications inspection (SSOMI)
was conducted at the Fermi-2 nuclear power plant during the weeks of July 17 and July 31, 1989. The prinary purpose of the design inspection was to examine the detailed design and engineering work products required to support the modifications planned for the first refueling outage. The design inspection
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was expanded to include a vertical slice review of the emergency diesel genera--
tors and associated auxiliary system ,
The inspection team concluded that the recently issued engineering design-packages for the first refueling outage were generally satisfactory. However, several weaknesses were identified by the inspection team. The following weaknesses were identified as being the most significant:
The lack of a program or procedures for the receipt and evaluation of vendor service bulletins. As a result, the last seven service bulletins for the emergency diesel generator had not been received or evaluate *
Post-modification test requirements were not properly delineated in the associated engineering design package *
Proper acceptance criteria were not contained in the procedures written to perform MOVATS testing on motor-operated valve Several discrepancies were noted between the piping design and the plant as-built conditio * Inconsistencies and unverified assumptions were made in calculations relating to the emergency diesel generator and the residual ~ heat removal system overcurrent relay *
A setpoint calculation was performed using an assumed drywell temperature which had been previously shown to be incorrec The inspection team also identified several strengths with the design control program. These included the performance of thorough safety evaluations, the utilization of an efficient computerized design document database that provided excellent document retrievability, well-maintained emergency diesel generator areas, and the conduct of a strong self-conitoring program for technical and quality oversight of the engineering and design activities.
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l 1.0 INTRODUCTION The huclear Regulatory Commission (NRC) initiated the Safety Systems Outage Modification Inspection (SSOMI) program in 1985 as a means to ensure that the licensir,g bases of the plant are not compromised during the process of making changes to the plant. The SSOMI usually consists of two inspections: (1)a design inspection to evaluate the planned outage design changes and modifica-
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tions with regard to regulatory requirements and licensing commitments and (2) an installation and test inspection to evaluate the implementation and closecut of the modification packages. Only the design phase of the SSOMI program was accomplished during this inspectio The NRC inspection team, with contractor assistance, conducted the design-related portion of the SSOMI at Enrico Fermi Atomic Power Plant, Unit 2 (Fermi-2), during the weeks of July 17 through 21 and July 31 through August 4, 1989. An exit meeting was held on August 4, 1989. This report describes the activities and findings generated by this design inspection. Some of the findings may result in potential enforcement items. Region III will initiate and execute any required enforcement action that results from this inspectio The inspection team examined a sample of the detailed design and engineering required to support the planned modifications for the first refueling outage to determine the technical adequacy of the modifications, compliance with regula-tory requirements and licensing commitments, and the effectiveness of the design controls for the modification process. During the cour e of the inspec-tion, the team considered:
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validity of design inputs and assumptions
proper component safety classification
adequacy of the safety evaluations for the design change
proper performance of design verification
consistency between the as-built plant and the design
compliance with regulatory requirements and design bases implementation of programmatic controls for the design change process l
The team examined several engineering areas including mechanical systems, !
mechanical components, electrical power, instrumentation and control, and motor-operated valves. To gain a broader perspective of the design process, the team also did a limited vertical-slice review of the emergency diesel generators and the associated auxiliary system Sections 2.1 and 2.2 provide the details of the inspection activities and the associated findings of the selected engineering design packages for the modifi-cation and the review of the diesel generator and its auxiliary system Section 3 discusses the observed strengths and weaknesses of the design control process. Appendix A provides a list of the engineering design packages for the modifications that were reviewed during the inspection. Appendix B lists the licensee personnel that were contacted through the course of the inspection as ,
well as indicates those present curing the exit meeting. Appendix C documents l the deficiencies that were found by the inspection team. Within the main body ,
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'of the report, NRC open item numbers have been assigned to the more significant deficiencies which correlate with the associated deficiency description in Appendix .0 INSPECTION DETAILS 2.1 Design and Engineering 2.1.1 Mechanical Components The inspection team reviewed six engineering and design packages (EDPs) and ;
associated calculations to confirm that the effects of plant modifications on I
the design of pressure retaining components and equipment supports were properly accounted for and that the as-modified systems structures and components were in accordance with regulatory requirements and licensing commitments. The following EDP's were reviewed: EDP 1971 " Replacement of Control Complex HVAC Ductwork," EDP 4940 "TIP Nitrogen Purge Line Isolation,"
EDP 5299 " Elimination of Cross Tie Flood Control Valves," EDP 8044 -
" Replacement of Butterfly Valve Carbon Steel Components," EDP 1049A -
" Installation of Flanges on Drywell Cooler Piping," and EDP 8383A " Reroute of Condensate Storage Tank Vent Line."
Except as noted below, the team generally found the documentation packages associated with the modifications to be complete and detailed. Specifically, the stress analysis and pipe support calculations were clear, complete, techni-cally adequate, and well documente Hcwever, during the review of the design documentation, the team found several discrepancies between the design documents and the as-built plant condition One installed pipe support for penetration X-35G was not shown in the design drawings and calculations for this pipe support and the associated piping were not available. Adriitionally, four small-bore pipe supports for the core spray pump casing vent, seal leak-off, and casing drain lines had not been installed although they were shown on design documents. The licensee reanalyzed the stress levels for the piping and determined they were acceptabl Because the inspection sample size was small, the team was concerned whether other pipe supports had been properly installed. The licensee was similarly concerned and agreed to perform additional sample walkdown verifications to determine whether the required supports are installed. This item is open (341/89-200-01).
2.1.2 Mechanical Systems The team reviewed the following five EDPs to determine whether technical and procedural requirements had been met: EDP 4940 "TIP Nitrogen Purge Line Isolation," EDP 1971 " Replacement of Control Complex HVAC Ductwork,"
" Installation of Double Grapple Hooks," EDP 8044 " Replacement of
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EDP 10093 Butterfly Valve Carbon Steel Components," and EDP 7042 "HPCI Booster Pump -
l Impeller Replacement."
With the exception of a concern related to post modification test criteria for a modification to the nitrogen supply piping, the design activities were found to be well controlle ]
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Hitrogen Parge to the Traversing In-Core Probe (TIP) System The TIP nitrogen purge system provides dry nitrogen to the TIP guide tubes to prevent corrosion of the steel drive cables and associated mechanisms. The licensee issued EDP 4940 to install a new TIP purge control panel inside the drywell and to modify the source of nitrogen to the panel so that it would be
- supplied by the Division I primary containment pneumatic system (PCPS) header inside the drywell. The existing nitrogen supply piping, control panel outside the drywell, and nitrogen supply line drywell penetration were eliminated by this modification. The drywell penetration assembly was modified to be used as a spar The PCPS had been designed and installed to provide makeup nitrogen to the automatic depressurization system (ADS) accumulators. The TIP purge system was classified as nonsafety-related (Quality Level III) while the PCPS was safety-related (Quality Level I). The design modification provided a safety-related restriction orifice in the supply line to the TIP purge control panel to limit the flow to 20 standard cubic feet per minute (SCFM) in the event of a loss of pressure boundary integrity in the TIP purge syste The licensee had performed an evaluation of the PCPS design calculations and had determined that a loss of 20 SCFM through the restriction orifice would not decrease the pressure in the PCPS header below that which is required for makeup to the ADS valves accumulators. The team found that the post-modification test requirements specified that the flow through the restriction orifice be measured to verify that the flow does not exceed 20 SCFM while the PCPS header is pressurized; however, the requirements did not specify the
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header pressure to be maintained during the test. The licensee issued a clarifying memorandum No. HE-PJ-89-0539, dated August 3, 1989 to specifically require the header pressure to be maintained at normal operating values during the test. The team considers this to be acceptable. This item is close .1.3 Electrical Power and Instrumentation and Control The team reviewed the following 10 EDPs and relateo design documents:
EDP 1022 " Automatic Depressurization System Logic Modification," EDP 7838 -
"MSIV Last Command Interlock," EDP 9094 "Rosemount 1152 Transmitter Replacement," EDP 6740 " Modification of Instrument Racks," EDP 8239 -
" Replacement of Chlorine Detectors," EDP 7577 " Secondary Containment High Temperature Alarm Window," EDP 1511 " Modification of Station Air Compressor Cooler Drain Line," EDP 8355 "EDG Start Logic Deficiency," EDP 8315 -
" Reactor Coolant System Setpoint Change," and EDP 9348 " Motor-0perated Valve Electrical Power." In addition, a review was conducted of the engineering specification and procedures applicable to the EDP process. From this review, the team concluded that detailed and prescriptive procedures were in place and were being implemented for the engineering and design work currently in process. Hewever, the following deficiencies were noted by the team in the electrical and instrumentation and control are Setpoint Calculations
, Review of the calculations related to EDP 9094 revealed that the effects of installing new instrument transmitters with a longer response time had not been considered. These instrument loops were associated with post-accident monitor-ing. The licensee personnel indicated that response time changes had been-3-
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reviewed informally and found to not be a critical attribute. The licensee l agreed to add a design verification check for assessment of response time change effects.
h~ During review of Temporary Modification Number 87-0069, written to silence the I
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for the high-pressure reactor scram setting were based on an ambient drywell !
temperature' equal to 135'F whereas the actual temperature was found to be over i 190*F.- Although the overall effect of the higher temperatures on the instru- l ments accuracy appeared to be negligible, the licensee agreed to revise the l cWulations as necessary to include the appropriate temperature inputs. This !
item is open pending revision of the calculations (341/89-200-02).
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_ Installation Requirements and Post-Modification Testing .
l For the electrical EDP's reviewed, the team fcund that the description of the I installation requirements and the post-modification testing (PMT) was too !
general. The description was mainly related to a listing of applicable sur- j ve111ance procedures. As an example, in EDP 9094 for the Rosemount transmitter I replacement, Section 9 of.the EDP only referenced surveillance procedure nunbers and did not include specific PMT acceptance criteria. The actual .
detailed description of the requirements was left to the maintenance group, who I
. would take full responsibility for the proper interpretation of the EDP ttst requirements. There was no requirement for an engineering group review of the maintenance group procedures for installation and testing. Thus, while the responsibility for origination of the EDP and its design implementation fell on the design engineering group, the responsibility for testing and installation fell on a different grou The team reviewed the guiding procedure for EDPs, FIP-CM1-12, and found that it contained explicit requirements for inclusion of PMT instructions and accep-tance. criteria within the EDPs. As a result, it was evident to the team that the procedure had not been fully adhered to during the generation of the EDP The team noted that a previous DECO quality assurance audit had found that the EDPs were lacking details for PMT requirements. Several hundred EDPs were scheduled to be reviewed by engineering to ensure that the PMT was adequat In response to the team's concern, the licensee agreed to have the current refueling outage PMT requirements reviewed for adequacy as part of the engi-neering EDP ownership program. An engineering directive was issued to reiter-ate the EDP requirements for PMT criteria. This item remains open pending completion of the review of the PMT requirements (341/89-200-03).
Voltage Rating of Relay Coils The team reviewed EDP 1022 and EDP 7838 for an automatic depressurization system logic modification and an MSIV last command interlock modification, respectively. During this review the team found that relays were procured without specifying the applicable minimum and maximum voltage range in which the relay coils are requireo to function. The expected range of the source voltage at the relay coils is 110 138 volts dc. The licensee determined that these specific relays will meet the voltage requirements. The licensee agreed to verify the adequacy of previous relay insta11aticns with respect to coil voltage capabilities and to revise the design verification checklist to require-4-
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~ an explicit check of this attribute for future design changes. This item remains.open(341/89-200-04).
2.1.4 Motor-Operated Yalves Motor-Operated Valve (MOV) Program The inspection team reviewed DEC0's MOV program as it related to the design and implementation of MOV modifications. Specifically EDP 8931 was reviewed for the upgrade of the spring packs on four safety-related MOVs. These spring packs were upgraded as a result of testing performed in accordance with NRC Bulletin 85-03. The testing indicated marginal performance of these MOVs and
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upgraded spring packs would increase the thrust available from the subject actuators. The replacement spring packs were procured from Limitorque, Inc.,
under safety-related Purchase Order NP-18831 The EDP contained recomended torque switch settings that Limitorque had derived from standard industry equations. The EDP referenced Preliminary Design Change (PDC) 8505 for obtaining thrust criteria. The team reviewed PDC 8505 and found that it contained both Limitorque calculated thrust require-ments and thrust requirements provided by the M0 VATS database. The team could not determine which number was to be designated for use as the minimum thrust requirement. Licensee personnel told the team that DEC0's policy is to try to meet both numbers; however, they considered the governing requirement to be the Limitorque calculated number. The team found this methodology inadequate because the thrust requirements for the particular valve must be determined at design conditions to ensure a properly sized valve operator. This will be evaluated by DECO as part of their response to Generic Letter 89-1 MOV Test Instructions The team's review of the installation and post-modification testing instruc-tions contained in EDP 8931 showed procedures NPP 35. LIM.008 and 35. LIM.005 referenced as the applicable installation and disassembly procedures. These procedures were found to contain spring pack drawings but no specific instruc-tions on spring pack installation or removal. Procedures NPP 35.306.003 for periodic MOV inspection, NPP 35.306.006 for MOV electrical testing, and NPP 35.306.010 for M0 VATS testing were also referenced in the EDP. The MOVATS testing procedure NPP 35.306.0?0 and the M0 VATS evaluation procedure NPP 47.306.001 were found to be delicient because they do not delineate where to obtain thrust acceptance values required to be documented in the procedur When the team discussed this concern with the licensee, it was discovered that these numbers apparently come from PDC 8505, which contains both Limitorque calculated and M0 VATS database numbers. The NRC has generally accepted the use of information obtained during differential pressure testing of similar valves, such as that contained in the MOVATS database; however, the ARC has not accepted the use of only calculated numbers. As a result of this finding, the licensee has agreed to changing the appropriate procedures, as necessary, to indicate where thrust acceptance values are obtained and to add the minimum thrust and torque switch setting criteria to the central equipment databas This item remains open (341/89-200-05).
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MOV Purchase Order l
The team reviewed Purchase Order NS-347484 to M0 VATS Inc. for the procurement I of equipment, services, and engineering data to be used in the testing and I evaluation of MOVs. Although the purchase order appeared to be safety-related, I it was unclear as to what items and services were being provided as safety- (
related and which ones as commercial grcde. When the team discussed this concern with the licensec, they were told that the MOVATS thrust data was not considered safety-related data, although it is used as a reference in validat- j ing torque switch settings on safety-related MOVs and is contained in PDC 850 The team expressed concern that, although this data may only have been used as a reference, it is still being used as part of engineering evaluations being
. performed on safety-related MOVs and therefore should be controlled, reliable, l and verifie MOV Material Control ]
The team reviewed Work Request 0226890319/01W, which involved the replacement of shaft key material on the declutch lever of the MOVs. This item was of particular interest because it related to a previously identified NRC concern involving the use of nonsafety-related material on safety-related operator The team reviewed the material traceability, document control, and engineering evaluation process involved with the shaft key replacement. The documents 1 reviewed include quality control receipt inspection documents, the inspection checklist, certificates of compliance for the shaft key material, certificates of tests, the purchase requisition, sales order, and packing order. The documents were found technically satisfactor .2 Inspection of Diesel Generator and Auxiliary Systems The team performed a limiteo review of the design and installation of the emergency diesel generators (EDGs) and auxiliaries on a system basis to gain a broader perspective of the design process at Fermi-2. The team also reviewed design documentation and performed a limited plant walkdown of EDGs 11 and 1 .2.1 Piping and Supports The team reviewed Temporary Modification 87-0094, which was initiated to replace the inlet damper actuator for EDG 11. In addition, representative piping stress analysis and pipe support calculations from the EDG service water and fuel oil systems were reviewed to determine the adequacy of analysis, correctness of design information, and agreement with the as-built condition The seismic analysis of the piping for the 8-inch EDG service water return line was found to be inconsistent with the as-built plant configuration. Specifi-cally, the U-bolt clearances depicted on a design sketch for hanger R30-2181-G13 were not in the same configuration as the installed support. The piping analysis assumes that hanger R30-2181-G08 carries a dead-weight load of 690 pounds ulthough, in actuality, a gap exists in the field installation so the support does not carry the assumed load. The licensee agreed to reassess the effects of gaps on the piping seismic analysis. Review of the licensee's reassessment of the piping seismic analysis is an open item (341/89-200-06).
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f 2.2.2 Auxiliary Systems j The' team inspected the following mechanical systems that support the EDGs:
diesel generator service water
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fuel oil fill'and transfer
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. residual heat removal complex heating and ventilation system
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residual heat removal complex fire protection
scavenging air ~
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The team performed a walkdown of the mechanical systems associated with EDG 11 and found the installation in agreement with the design documsnt The team found the associated design documents to be complete and in agreement with the diesel generator manufacturer's requirements and with FSAR commitment The team investigated two deviatina event reports (DER)' associated with a
. problem involving lube oil leaks onto the exhaust manifold insulation that caused fires during the diesel generator testing. The licensee indicated that gaskets were planned'to be replaced on all four EDGs at identified leak points as a preventive measur .2.3 Electrical and Instrumentation and Control Aspects Capability of Power Circuit Breakers Exceeded The team reviewed calculations DC-2575 and DC-2566 for the 4160 volt switchgear short circuit interrupting capability. The team found that the interrupting and short-circuit ratings of the 4160-volt circuit breakers could be exceeded if the system voltage exceeds 104 percent of the nominal value. Subsequent-review of the system voltage calculation DC-835, page 3, indicated that the system voltage could go up to 106 percent ano that the breaker ratings could be exceeded. Further discussion with the licensee revealed that this issue had been previously identified during a self-initiated SSFI in February 1989 and as a result, calculation revisions were made which showed that the actual loads postulated on the 4160 volt buses were lower than originally assumed. The team examined the proposed calculation revisions and found them adequate. However, the formal calculations had still not been revised at the time of the inspectio Unsupervised Fuses in Protective Relaying Circuits The team reviewed one-line diagram SD-2500-08, which indicated that protective relay device 67 (reverse power protection) and device 81 (underfrequency protection) had unmonitored fuses in their potential circuits. If a fuse were to blow, it could remain undetected. These relays do not fail safe, as such, they will fail to trip upon the loss of their potential supply. Although the relays are only active during the test mode of the EDG, the EDG could be seriously damaged if while in the test mode either a loss of offsite power or a loss of motive power (the engine) should occur. The EDGs were not provided
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with backup overload protection, which makes the underfrequency protection even more important. While this situation does not impair the ability of the diesel ;
-to perform its design safety function,'it appears that equipment protection "
considerations have not been fully examined. The licensee agreed to review the acceptability of the unmonitored relays, a i
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Ambient Temperature Alarms J I
During a field walkdown'of the EDG.and switchgear room, the team observed that I the ambient temperature alarms were set above the maximum design ambient
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temperatures.for the rooms as follows:
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maximum design temperature alarm setpoint, i Switchgear. room .104* F 110' F Engine room 122' F 130' F Although the alarms are not safety-related, the licensee agreed to evaluate the basis for. the alarm setpoints and to adjust them as necessar Monorail Hook Storage-Position
'During the field walkdown of the EDG switchgear room, the' team observed that-the monorail hook was in a lowered position and located between two safety-related switchgear panels. Under a postulated seismic event, the hook could swing and impact both switchgear panels. The licensee found that engi-neering had established proper guidelines (PDC 8048) for the hook storage; however, these guidelines were not incorporated into the proper prucedure The licensee indicated that the procedures would be revised to incorporate the necessary guidelines and that personnel would be instructed accordingl Lube Oil and Jacket Water Cooling Heaters The team reviewed the logic diagram for the EDG auxiliaries, the EDG manufac-turer recommendations, and the actual plant setpoints of the heaters provided for the lube oil and jacket cooling water. Inconsistencies were identified and the plant setpoints were unconservative with regard to the design documenta-tion. The licensee issued a DER to reconcile the inconsistencie Emergency Diesel Generator Calculations The team reviewed sizing calculation 5003 for the EDG, logic for the load-shed circuits of the emergency bus, and logic for sequential connection of engi-neered. safety features (ESF) loads to this bus. The team noted that the size of EDG appeared adequate based on the load tables included in the calculatio However, the validity of the load values listed in the calculation could not be verified because of the unavailability of the related documentation. Specifi-cally, in calculation DC 2116 full load currents were not always used and when full load currents were used, adequate justifications for the assumed values were not provided. The team was informed that DECO is currently revising these calculations and that in the future only full load currents will be used. This item will remain open until the appropriate calculations are revised (341/89-200-07).
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Residual Heat Removal (RHR) Motor Overcurrent Eclays The team also reviewed the trip settings of the motor overcurrent relays for the RHR and core spray pumps. From the data proviced by the DEC0 engineers, it appeared that the RHR pumps would trip before they would attain the required speed to mitigate an accident condition. DECO engineers evaluated this situation and informed the team that the initial data provided to the team by DECO was not correct. By performing a preliminary calculation using the correct data, DECO engineers showed the team that a premature trip of RHR and core spray pump would not occur. The licensee issued a DER to document the concern of using erroneous data and committed to formalize the calculation for the RHR overcurrent relay and to investigate similar core spray pump motor protective relays. Thisitemremainsopen(341/89-200-08).
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Emergency Diesel Generator Physical Condition
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During the i.e. walidown, the team noted a large loop of loose cable hanging by a small rope over instrument tubing lines above HVAC panel H21-P351. If the rope broke, the cable could damage the instrument sensing lines required for EDG 12 HVAC control and monitoring. DECO investigated this further and informed the team that the cable in question was an abandoned cable that should have been cut immediately after abandoning per existing station procedure DECO initiated corrective action to remove the abandoned cable. The EDG area was excellent with regard to housekeeping considerations: all identification and caution tags were readily visible, illumination levels were adequate, and instrumentation was accessible for calibration and readout . 2.2.4 Vendor Documentation The team reviewed DECO's receipt and evaluation of the service information letters (SILs) sent out periodically by the diesel manufacturer, Colt Industries. These SILs contain maintenance recommendations, material deficiencies, operating information, and other pertinent data relevant to the reliable operation of the diesel generators. During the review, it was discovered that DEC0 had either not received or had received and not evaluatea the last seven SILs issued by Colt, dating from June 22, 1987 to the presen Additionally, further review indicateo that there was no program or procedure in place at Fermi-2 for the receipt and evaluation of this type of vendor information. Item 2.2.2 of !!RC Generic Letter 83-28 specifies that licensees should establish, implement, and maintain a continuing program to ensure that vendor inforroation for safety-related components is complete, current, controlled, and appropriately referenced or incorporated in plant instructions and procedures. This letter also specifies that vendors of safety-related equipment be contacted to establish a coununication interfac As a rcsult of this finding, DEC0 has agreed to the following: (1) to estab-lish a list of key safety-related components and then to expeditiously contact pro those receivedcomponent vendors and evaluated and by site ensure and personnel that all ap(2) priate to review theinformation adequacy ofhas the been existing site procedures that govern the control of vendor documentation. This item remains open (341/89-200-09).
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- The~ team noted that previous- Region' III; inspections (50-341/88-24,and 89-03)
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ihad identified' deficiencies with'the use of uncontrolled vendor technical manuals and the untimely implementation of. vendor recomended lubrication requirements. The licensee has also comitted in DEC0 letter, NRC-89-0139, to c revise the applicable site-procedures for handling vendor informatio .0, INSPECTION FINDINGS INDICATIVE OF LICENSEE STRENGTHS AND WEAKNESSES
- 3.1' Strengths
. Safety Evaluations
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The team found that the safety evaluations were thoroughly performed and contained sufficient detail to assess the validity of the evaluations and conclusions.- For' example, the safety evaluation for the relocation of.the-TIP purge panel land nitrogen supply header to inside the drywell contained a detailed analysis of the consequences of tapping into the existing primary-containment pneumatic supply system to support the conclusion that no
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unreviewed safety concern would exist as a result of the modificatio The team _found all of the evaluations.and conclusions to be acceptabl : Design Document Control The team found during the review'of the engineering design packages that the'
111censee maintained a design document index to identify the types of document that contain current as-designed or as-built information. The effectiveness of-the document control-was proved by the various drawings, specification, and calculations reviewed by the team. The team also found the licensee's document retrievability and computerized component database to be'very efficient. The team noted that the design drawings were neat, legible, and contained the i required cross reference . Quality Oversight of Engineering Activities l The team found DECO vigilant about safety and noted DECC was involved continu-
'ously in a self monitoring program for quality and technical adequacy of its engineering and design activities. Examples of such activities were the conduct of an SSFI by the in-house staff using help from independent consul-tants, upgrading of all engineering and design procedures, performance of technical quality assurance audits and quality assurance monitoring of the
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I design' process implementatio EDG Physical Condition Plant areas for the EDG were well maintained with proper equipment layouts to facilitate operation and maintenance activitics. Component tagging was quite
. goo .2 Weaknesses Calculational Errors Review of the calculations associated with the EDG sizing were found to contain assumptions regarding the loads which could not be verified by the team. Some of the loads were improperly derived from equipment brake horse power values.
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While'the available margin of the EDG appears adequate, thesesuncertainties could not be' reconciled for the team.
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Control of Vendor'Information
. Review of the program for receipt, distribution, and evaluation of vendor information identified that several service information letters had not been received for the emergency diesel generators. No clear responsibility had been
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established for handling of the vendor documentation. Concerns over the control of vendor technical manuals and implementation of recommended preventative maintenance had been identified previously by Region III inspector *
Definition of Post-Modification Test Requirements ,
The design changes issued in the electrical and instrumentation areas were found to lack definitive acceptance criteria and post-mootfication test requirements. While the guiding engineering procedures specified that the design change must delineate the test criteria, in practice only a reference was made to an existing surveillance procedure. After issuance of the design change, it was the responsibility of the plant staff to determine the test requirements and acceptance criteri Consistency of Design Documents with the As-Built Plant In several cases the plant configuration was not found consistent with the associated design documents. These involved pipe stress calculation assump-tions that did not reflect the hanger configuration, four cases where small-bore pipe supports were not installed as depicted on the piping isomet-ric, and a case where a support was installed but had not been analyzed. These situations raise concern with respect to the performance of the as-built walkdown verifications to ensure that the design has been properly translated into the-plant hardware. The licensee agreed to perform additional sample walkdowns to ascertain the scope of this proble .0 MANAGEMENT EXIT MEETING The inspection team conducted an exit meeting on August 4, 1989, with licensee personnel as identified in Appendix B to this report. During the meeting, the team members presented the inspection findings and provided clarification in response to licensee questions regarding the finding Other NRC personnel at the exit meeting included: Mr. John Zwolinski, Mr. Eugene Imbro, Mr. Jeffrey Harold, and Ms. Patricia Eng from NRR and Mr. Thomas Martin, Mr. Mark Ring and Mr. Stanley Stasek from Region II . _ -. -. _ _ _ _ _ - _ - _ _ _ _ _-
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APPENDIX A 4 ENGINEERING DESIGN PACKAGES REVIEWED BY THE INSPECTION TEAM Number Title 1022 Automatic Depressurization System Logic Modification
, 1971- Replacement of Control Colaplex HVAC Ductwork
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4940 TIP Nitrogen Purge Line Isolation 5299 Elimination of Cross-Tie Flood Control Valves l 6740 . Modification of Instrument Racks {
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7042 HPCI Booster Pump Impeller Replacement 7838 MSIV Last Command Interlock 8044 Replacement of Butterfly Valve Carbon Steel Components 8239 . Replacement of Chlorine Detectors 8505 IE Bulletin 85-05 Criteria-8931- Motor-0perated Valve Springpack Upgrades 9094 Rosemount 1152 Transmitter Replacement 10093 lasta11ation of Double Grapple Hooks 1049A Inst 611ation of Flanges on Drywell Cooler Piping 8383A Reroute of Condensate Storage Tank Vent Line 7577 Secondary Containment High-Temperature Alarm Window 1512 Modification of Station Air Compressor Cooler Drain Line 8355 EDG Start Logic Deficiency 8315 Reactor Coolant System Setpoint Change 1023B Vent fans and Filters 9348 Motor-0perated Valve Electrical Power ;
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APPENDIX B PERSONNEL CONTACTED R. Anderson Electric Supervisor
- P. Anthony Licensing Engineer
- R. Ballis Supervisory Instrumentation and Control Engineer T. Banek Nuclear Instrument and Control Maintenance Specialist D. Bergmooser System Engineer L. Borr Nuclear Engineering D. Bramlett Lead Engineer Procedures
- S. Catola Vice President Nuclear Engineering and Services
- L. Collins Lead Electrical Engineer .
- G. Cranston General Director Nuclear Engineering 1. Di Magio Engineer-in-Charge Maintenance Procedures
- T. Dong Plant Safety
- J. Dudlets Supervisory Design Engineer L. Ferguson Lead Instrument and Control Engineer R. Filipek Lead Instrument and Control J. Fix Electrical Supervisor R. Godnek Electrical Engineer
- Goodman Director Nuclear Licensing J. Green Electrical Supervisor A. Hassoun Nuclear Engineering M. Hoffman Senior Operator D. Jax Senior Engineer H. Kantrowitz Assistant Lead Mechanical / Structural Engineer
- A. Keltos Engineering Assurance Supervisor, SWEC S. Larry Lead Mechanical / Structural Engineer A. Lim Nuclear Engineering
- V. Manta Senior Quality Assurance Specialist
- R. Mathews Acting Superintendent of Maintenance
- R. McKeon Supervisor of Operations W. McNeil Nuclear Engineering
- T. Riley Supervisor of Licensing H. Sahiner Seismic Engineer
- A. Settles Supervisor of Technical Engineering G. Sharma Lead System Engineer
- G. Trahey Director of Special Projects
- E. Wilds Lead Mechanical Engineer M. Williams Senior Engineer
- Zoma Electrical Engineer
- Denotes licensee personnel attending the exit meeting on August 4, 1989. ,
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APPENDIX C m
DEFICIENCIES
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DEFICIENCY 341/89-200-01: Discrepancies Between the Piping Design and As-Built Condition Discussiun: During the review of the nitrogen purge to TIP system modifications associated with the EDP-4940, the team discovered that the pipe support
, that presently holds the seismic Category 1 pipe from the drywell liner up to the connectiun to'the tubing in the TIP drive room at. penetration X-35G
'is not shown in the baseline design drawing 61721-2837-9. The existing pipe support restrains-the 1-1/2" pipe in the lateral horizontal direction l and in the downwards direction. However, the support does not restrain the motion of the pipe in the upwards direction. Furthermore, the licensee was unable to show design calculations qualifying the present piping.and pipe support arrangement The licensee issued a DER that evaluated the present piping'and pipe support configurations and found them acceptabic. The DER recommended that the piping system and primary containtrent-. remain operable until a documented design basis is provided for the existing installation or when EDP-4940 is installe Modifications planned for the coming outage (EDP-4940) will modify this
. specific piping and pipe suppor . The team also reviewed the modification for the elimination of the cross-tie flood control valves from the reactor building sub-basement sumps associated with EDP-5299. The team discovered that four small bore pipe. supports had not been installed. The missing pipe supports, shown in the analytical isometric 6WM-E21-5110-3 and drawing 6WM-E21-5086-1 affect the Division 1, core spray pump casing vent, seal leak-off and casing drain line Two of the missing supports (G02 and G03) were designed to be located in the 3/4" casing drain line in the seismic /nonseismic interface area, downstream of the safety-related isolation valve V8-2052. Support G02 is a guide and support, G03 is a deadweight hange The other two missing supports (G06 and G08) were designed to be located in the 3/4" casing vent line and 3/4" seal leak-off line, respectively. Both were deadweight hanger The licensee issued a DER with evaluations that showed that the three affected pipe lines meet all the design code requirements in the as-installed conditio The DER recommends that all the affected piping and the core spray pump remain operable until the design basis calculations are updated to reflect the installed condition or until EDP-5299 is installed. Modifications planned for the coming outage (EDP-5299) will modify the routing and supports of these three lines. The licensee also agreed to verify, on a sampling basis, the actual installation of small bore safety-related pipe supports against the representative small bore stress isometrics. This item is open (341/89-200-01).
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Regulatory Basis:-
10 CFR 50, Appendix B, . Criterion V requires'that activities affecting qualit be accomplished in accordance with appropriate drawings'or procedures. Con-trary to the above, pipe supports were not installed as indicated on the applicable drawing References:
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Dwgs.'6WM-E21'-5086-1, . Rev. B and 6WM-E21-5110-3, Rev. B and 61721-2873-9,
. Rev. 0 .
Calculation DC-1938
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DCP-5116-IO1 (sheet 20 of 60)
EDP-4940,."TIP Nitrogen Purge Line Isolation" .
EDP-5299, " Elimination of Cross-Tie Flood Control Valves" ASME Section III, 1971 Edition lfFSAR sections 3.7.3.13 and 3.7.3.14 DEFICIENCY 341/89-200-02: Setpoint Calculations Discussion:
During this inspection the team reviewed Calculation 4522. " Instrument Setpoint
' Validation." The' team noted that the setpoint calculation for the steam dome
'high-pressure scram was performed on June 17,1989 using 135*F as a drywell
' ambient. temperature even though certain areas in the drywell were known to be at temperatures.substantially higher than this number. Although engineering
personnel were aware of higher drywell temperatures for two years, they failed to use the actual ambient temperature value for density compensation for the scram setpoint calculation. Increased temperature lowers the density of water resulting in a lower value of static head. As a result, the'high-pressure scram signal may not be generated at the designated trip setting. Although the overall of fect of the' higher temperature on the instrument accuracy appeared to be negligible, the licensee agreed to revise the calculations as necessary to include theap repriate temperature inputs. This deficiency is considered open (341/89-200/02 .'
Fegulatory Basis:
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10 CFR 50, Appendix B, Criterion III requires that measures be established for the identification and control of design interfaces. Contrary to the above,
,
DEC0 did not use the proper drywell temperature in setpoint calculations for the steam dome high pressure instrumen References:
DECO Calculation 4522, Rev. B " Instrument Setpoint Yalidation" DEFICIENCY 341/89-200-03: Failure to Specify Post-Modification Test Requirements
'The team reviewed EDP 9094, which was issued to replace the existing QA level 1, Seismic Category I, Rosemount Model 1152 series pressure and level transmitters on the primary containment accident monitoring (PCAM) system with
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Rosemount Model 1153 series transmitters to comply with current environmental qualification requirements and to reduce future environmental qualification maintenance requirement The team noted that the EDP did not specify requirements for the post-modification test (PMT) acceptance criteria; instead, in Section 9 of the EDP, a few surveillance procedure numbers were included in lieu of specific accep-tance criteria for the PMT. The team was informed by licensee engineers that the plant maintena:.ce staff establishes the test acceptance criteria on the basis of the desigi changes in the EDP and performs the test to the test acceptance criterie established by the plant maintenance staff. The site procedures require that PMT acceptance criteria should be specified by the
. design engineer involved in the modification and should not be left for the interpretation of instrumentation test technicians who may not be familiar with the design objectives of the modification. The team was informed that the PMT results were not reviewed by the design enginee The team was also concerned that a proper design verification of the PMT acceptability could not be done when the EDP is issued because the referenced PMT procedures may be later revised without reverification of the PMT require-ments. Although Procedure FIP-CM1-12 was issued in March 1989 and provides explicit requirements for inclusion of PMT instructions and acceptance criteria within the EDP, it was evident to the team that this procedure had not been fully adhered to during the generation of the EDPs. The licensee agreed to ensuring that the engineering personnel would review the adequacy of the EDP PMT requirements as part cf the new EDP ownership progra The team also reviewed a DECD quality assurance audit report 88-252 that had previously identified that PMT requirements were not adequately defined within the EDPs. The engineering organization bsd initiated a review of several hundred EDPs for FMT adequacy as part c e e corrective action. This deficiency is open (341/89-200-03).
Regulatory Basis:
10 CFR 50, Appendix B, Criterion V states that activities affecting quality should be accomplished in accordance with the governing procedures and instruction ANSI N18.7-1976, Section 5.2.19.2, specifies that 1) the design organization should participate in the definition of the testing requiremer,ts before placing a nuclear plant system into operation and 2) the design organization should determine that the testing is edequate and that the test results are acceptabl ANSI N45.2.11-1974, Section 6.3.1, states that the design verification should L address whether acequate preoperational and subsequent periodic test require-ments have been appropriately specifie Contrary to the above, the licensee did not adhere to its procedures and the engineering design organization did not adequately specify PMT and the accep-
, tance criteria.
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i DEC0 procedures FIP-CMI-12, Rev. O, " Engineering Design Packages" and' -
FIP-CM1-13, Revision 0, " Design Verification" ,
EDP 1022, Revision 0, " Automatic Depressurization System Logic Modification" i EDP 9094, Revision 0, "Rosemount 1152 Transmitter Replacement" l DEFICIENCY 341/89-200-04: Relay Coil Voltage Discussion:
Review of EDPs 1022 and 7838 identified that. replacement relays had been specified by engineering. No documentation was found to substantiate that a review had been performed to assess the voltage rating of theirelay coils with respect to the anticipated' system source voltages. The licensee evaluated the rated pickup voltage and-the maximum permissible continuous coil voltage of the replacement relays versus the anticipated maximum voltage during battery equalization and the lowest battery voltages. The analysis found the replace-ment relays were satisfactory for the intended servic The licensee agreed to revise the process to incorporate an explicit check of this aspect for future design changes. The licensee additionally agreed to analyze istics. This theitem existing relay341/89-200-04).
is open (installations with respect to relay coil character-Regulatory Basis:
10 CFR 50, Appendix B, Criterion IV requires that measures be established to ensure that design basis requirements are suitably referenced in procurement documents. Contrary to the above, minimum and maximum coil operating voltages were not suitably referenced in procurement documents for the above relay References:
EDPs 1022, Revision 0, and 7838, Revision 0 IEEE 279, " Criteria for Protection Systems for Nuclear Power Generating Stations" DEFICIENCY 341/89-200-05: M0 VATS Test Procedures
l' Discussion:
During the review of the DEC0 procedures for performing MOVATS testing of MOVs, the team identified that procedures NPP 35.306.010 and NPP 47.306.001 did not l
delineate where the thrust acceptance values are obtained. While PDC 8505 was
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indicated to be the source, the PDC contains both thrust values that are calculated by Limitorque, Inc., and thrust values from the H0 VATS, Inc.,
database. No formal mechanism was in place to ensure the engineering defined acceptance criteria were utilized during the performance of site MOVATS L testing. This item is open (341/89-200-05).
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kegulatory Basis:
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10 CFR 50, Appendix B, Criterion XI requires measures be established to ensure j
.that procedures are written which include appropriate acceptance criteria l contained in design documents. Contrary to the above, appropriate acceptance (
criteria were not contained in the procedures used for M0 VATS testing and j evaluatio i i
References- !
10 CFR 50 Appendix B DECO NPP 35.306.010, "MOVATS Testing", Revision 21 )
. DECO NPP 47.306.001, " Signature Analysis of Motor Operated Valves", Revision 20 !
DEFICIENCY 341/89-200-06: DiscrepanciesBetweenAs-BuiltPipingandthePiping '
Analysis Discussion:
During the review of piping calculations and plant walkdowns for the EDG service water system, the team discovered several discrepancies between the piping analysis and the as-built conditions. Calculation DC-2924 provides the piping stress analysis for the 8" EDG service water return line from the EDG 11 (R3001S001) up to the tie-in with the 24" residual heat removal service water (RHRSW)line. The latest revision to this calculation, done on November 28, 1984, is designated as Addendum (C) to the piping stress report for subsystem SX-10. This revision was done to reconcile the as-built gaps of U-bolt type of supports with the piping analysis. Sketches depicting the actual U-bolt gaps are shown in the Addendum C to the calculation, pages 15 through 20. The team discovered that one of these sketches does not agree with the actual as-built condition. The sketch for hanger R30-2181-G13 shows the U-bolt with the strong axis located horizontally and with gaps all around the pipe. In the actual support configuration as confirmed by the team's walkdown, the U-bolt is located with the strong axis in the vertical direction and with no noticeable gaps in the vertical direction. The orientation of U-bolts and the location and magnitude of the gaps influence the modeling of restraints in the piping analysis and the generation of pipe support load The team's walkdown also confirmed that the U-bolt for hanger R30-2181-G08 is installed with a gap at the bottom of the pipe. This U-bolt does not presently I touch the pipe and consequently, does not carry any deadweight load. The piping analysis, however, assumes this hanger to be a deadweight support and to carry a deadweight load of 690 pounds. This assumption is not in agreement with the as-built condition. This particular piping subsystem appears to be conserva-tively supported and the effect of these discrepancies does not appear to be ,
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significant. The licensee agreed to evaluate and reconcile the effect of thesc discrepancies in the piping analysis of record. This item is open (341/89-200-06).
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Regulatory Basis:
10 CFR 50, Appendix B, Criterion XVI requires measures.be established to ensure that conditions adverse to quality are promptly identified and correcte Contrary to the above, DECO haa not identified discrepancies between installed pipe supports and pipe. support calculation References:
Calculation DC-2924, Vol. 1A, Revision B Hanger Mark Nos. R30-2181-G13, R30-2176-G32, and R30-2181-G08 ASME Section III, 1971 Edition
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NRC Bulletin 79-14 " Seismic Analysis for As-Built Safety-Related Piping System" .
DEFICIENCY 341/89-200-07: EDG Sizing Calculation
!
Calculations 5003 and DC 2116 pertaining to the loading of the EDG were reviewed by the team. The team noted that the size of EDG appeared adequate
' based on the load tables included in the calculations. .However, the validity of the load values listed in the calculations could not be verified because of the unavailability.of the related documentation. Specifically, in calculation DC 2116 full load currents were not always used and when full load current were used, adequate justifications for the assumed values were not provide Additionally, in Section 72B-2A of Calculation 5003, a 45 brake horse power (bhp) load was converted to 33.3 kilowatts (KW) rather than assuming a 90 percent efficiency and conversion factor of 1 HP=746 watts which would yield 37.3 K The team was informed that DECO is currently revising these calculations and that in the future only full load currents will be used. This item will remain open until the appropriate calculations are revised (341/89-200-07).
Regulatory Basis:
ANSI N45.2.11-1974, Section 4.2, Paragraph I requires design analyses be performed in a planned, controlled and correct manner., Contrary to the above, calculations performed on the EDG loading were done in an inconsistent manner using unverified assumption References:
DECO Calculation 5003, Revision 0, "EDG Loads" DECO Calculation 2116, Revision C, " Bus Loading" l DEFICIENCY 341/89-200-08: Residual Heat Removal (RHR) Motor Overcurrent Relay
.
l Discussion:
The team reviewed the zero block loads of the EDG load secuencer including the l
trip settings of the RHR pump overcurrent relays XZ51-IAC66B type. This relay was set at a tap setting of 3.5 amps. The data reviewed indicated the trip time of this relay is equal to 212 cycles at 15.2 amps, 320 cycles at 9.5 amps, and that it will trip instantaneously at 28 amps. The current transformer l ratio was 500/5 and the locked rotor current and the full load current at ,
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4160 volts was 1360 amps and 245 amps, respectively. Previous test data and FSAR section 8.3.8 showed that during the worst case start of the ESF loads, the acceleration time for the RHR pump was 4.5 seconds. It appeared that the relay would. trip in approximately 240-255 cycles, i.e. , during pump startu No specific calculation had been performed to show the pump start versus relay characteristic Subsequently, the team was informed by the licensee engineers that earlier data given to the team was incorrect. The team was then shown a preliminary draf t calculation using the correct data that showed the margin between the relay curve and the acceleration time curve was sufficient to preclude premature tripping of the RHR pump. The licensee informed the team that a final calcula-would be
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tion and analysis performed of theand for the RHR pump coreacceleration spray pumps.and overcurrent This relay (341/89-200-08).
item is open Regulatory Basis: ,
10 CFR 50, Appendix B, Criterion III requires measures be established for ensuring the design bases are correctly translated into specifications and drawings. Contrary to the above, no basis existed for the overcurrent relay tap setting at the time of the inspectio References:
FSAR Section 8. DECO Data Sheets, "RHR Pump Overcurrent Relays XZ-51, IAC66B Type" DEFICIENCY 341/89-200-09: Control of Vendor Information Discussion:
The receipt and evaluation of the diesel generator manufacturer service infor-mation letters was reviewed. These letters contain pertinent information regarding maintenance recorrnendations, material deficiencies, operating infor-mation, and other information germane to the reliable operation of the diesel generators. The last seven information letters had not been evaluated for applicability. Additionally, upon further review the team found that there is no program or procedures in place at Fermi-2 for the receipt and evaluation of this type of vendor informatio The licensee acknowledged this concern and agreed to establish a list of critical components, to set a time frame to contact the associated vendors, and to revise the site procedures as necessary to ensure that vendor information is received and evaluated. As a result of the inspection, DECO committed in letter NRC-89-0139 to revise procedure FIP-DCl-02 to require that selected safety-related vendors be contacted on a two-year cycle to ensure that vendor notices have been received, and to revise procedure FMD-AD3 to oesignate that Huclear Engineering is responsible for the receipt and disposition of vendor technical information. This item is open (341/89-200-09).
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Regulatory' Basis: .
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10 CFR-50 Appendix B, Criterion V requires activities affecting quality be -
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Contrary to the' above, DECO did not have procedures for th'e' receipt and evalua-
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- tion of vendor service bulletins, nor had a connunication interface been
establishe :l
' References:
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. DECO FIP-DCI-03 and FIP-DCI-02, " Vendor Manuals," Revision 1
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DECO NE-001, " Correspondence Control," Revision 0- (
DECO FND-AD3, Revision 1 ..
NRC Generic. Letter 83-28 DECO Letter VP-85-0134 dated July 5,1985
DECO Letter NRC-89-0139 s
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