IR 05000341/1989016

Insp Rept 50-341/89-16 on 890531-0602 & 26-31.No Violations Noted.Major Areas inspected:self-initiated SSFI of HPCI Sys. Present Design Control Measures Showed Significant Improvement

ML20247J169
Person / Time
Site:	Fermi
Issue date:	07/19/1989
From:	Phillips M, Yin I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20247J142	List: IR 05000341/1989016 ML20247J137
References
50-341-89-16, IEIN-77-05, IEIN-77-5, IEIN-88-071, IEIN-88-072, IEIN-88-71, IEIN-88-72, IEIN-89-011, IEIN-89-11, NUDOCS 8907310253
Download: ML20247J169 (8)

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U. S.. NUCLEAR REGULATORY COMMISSION

REGION III

> Report No; . 50-341/89016(DRS)

Docket No. 50-341 License No. NPF-43 Licensee: 'The Detroit Edison Company 6400 North Dixie Highway Newport, MI 41866 Facility.Name: Fermi'2' Nuclear Power Station

Inspection At:. Newport, Michigan

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f FInspection Ati May 31-June 2,'and June 26-19, 1989

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" Inspector:

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.I. T. Yin /

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Approved B . Monte . Phillips, Chief

' 7[ I/87 Operational Programs Section Date Inspection Summary Inspection on May 31-June 2, and June 26-31, 1989 (Report No. 50-341/89016(DRS))

Areas--Inspected: . Routine, announced inspection of licensee corrective actions g 4 initiated for the issues identified'in its self-initiated Safety Systems Functional Inspection (SSFI) of the High Pressure Coolant Injection (HPCI):

system. .The. inspection was performed based on selected portions of NRC Inspection Procedures 90713 and 3070 Results: Licensee effort in conducting the'SSFI for the HPCI system, and followup on the issues identified was good. Based on this review and evaluation, the inspector determined the following:

• - The SSFI review scope was extensiv *. . The' licensee had made commitments which were very responsive to the issues raised in the SSFI; although in some cases, the licensee had not implemented these commitment * The licensee's present design control measures have shown significant-improvement; however, there could be better referencing of design. basis documentation within a calculatio *

8907320253 890724 PDR

'O ADOCK 05000341 PDC

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DETAILS f

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Detroit Edison Company (DECO)

B. R. Sylvia, Senior Vice President

• +J. Pendergast, Licensing Engineer

• +J. Contoni, Supervisor, Mechanical Engineering

• +J. Dudlets, Principal Engineer

• W. S. Orser, Vice President, Nuclear Operations

• J. R. Green, Supervisor, Electrical Engineering i *+L. E. Schuerman, General Supervisor

• P. Zoma, Electrical Principal Engineer

• + Cranston, General Director, Nuclear Engineering

• +L. Goodman, Director, Nuclear Licensing

+R. B. Stafford, Director, Nuclear QA and Plant Safety

+S. G. Catola, Vice President, Nuclear Engineering and Service

+J. G. Walker, General Supervisor, Plant Engineering

+ McKeon, Superintendent, Operations A. Banek, I&C Specialist

.S. Williams, Senior Engineering Technician, Electrical

'R. L. Raisanen, Supervisor, Engineering Design and Support F. J. Svetkovich, Assistant to Plant Manager B. J. Sheffel, ISI/ Performance Evaluation Manager R. Matthews, Acting Superintendent, Maintenance and Modifications E. Wilds, Lead Engineer, Mechanical J. Melito, Senior Engineer D. Thomas, Mechanical Engineer D. Jax, Mechanical Engineer L. B. Collins, Lead, Electrical Engineer T. Dong, Supervisor, Plant Safety O. S. Nuclear Regulatory Commission W. Rogers, Senior Resident Inspector

+ Stasek, Resident Inspector

• Indicates those attending the exit meeting at the site on June 29, 198 + Indicates those attending the exit meeting at the site on June 2, 198 . Introduction The licensee contracted Westec, a Division of ERC International Corporation, to develop a detailed engineering oriented Safety System Functional Inspection (SSFI). Subsequently, the High Pressure Coolant Injection (HPCI) system was selected to be the first system receiving the SSFI. The SSFI team was composed of eight Westec staff and six supporting DECO engineer The inspection was conducted on November 16 through December 18, 1987, with the final report being issued on February 5, 198 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _- _ _ . _ _ _ _ _ _ - _________-__ -_ _-__- -

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.The purpose of the NRC inspection conducted on May 31 through June 29, 1989, was to evaluate the effectiveness of the corrective actions taken in response to the issues identified during the SSFI, and to assess the scope of this self-initiated quality oversight activit . Review of SSFI Report The inspector reviewed the Westec report and the licensee responses and concluded that the scope of the SSFI was extensive and that the SSFI was conducted in a professional manner. The licensee was very responsive to the issues raised in the SSFI; however, weaknesses were observed in the depth of the evaluation of the SSFI findings, and the implementation of corrective actions to the SSFI findings (Paragraphs 7c, d, and e). Site Review Samp'm Selection The inspector considered the following Inspection Observations (10s)

contained in the SSFI report to be significant design issues in the areas of mechanical, electrical, and instrumentation and control, and performed reviews on these issues:

No..MS-2 The determination of motor operated valve (MOV) maximum differential pressures in accordance with NRC Bulletin No. 85-03 was based on certain assumptions which may not be conservativ No. MS-9 Emergency equipment cooling water (EECW) make-up tank outlet MOV operators were not mounted in accordance with the manufacturer recommendatio No. MS-11 The reactor building closed cooling water (RBCCW) system may not be capable of dissipating design basis accident heat loads when EECW is not automatically initiate No. EP-2 Inadequate battery sizing calculation for Division II Class 1E safety related batterie No. EP-4 Inadequate battery cell electrolyte temperature surveillanc No. EP-5 Potential thermal overload relay degradation due to prolonged testing at 600% equipment current ratin No. IC-4 The as-found set point calibration data for a safety-related instrument / loop was not trended to determine if any time period adjustment is needed for the next calibratio . Licensee Corrective Action Strengths The inspector observed the following licensee corrective action activities which can be interpreted to be strengths within its program:

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a. (I0'No.'MS-2) A specific concern was raised in the 10 that the MOV maximum differential pressure (dp) calculation was not based on low trip pressure stated in the Technical Specification (TS). A licensee Deviation Event Report (DER) 88-0148 was issued to document and resolve the problem. The corrected dp calculation showed less than 2% operator thrust load increase. No hardware adjustment or modification was required. In responding to recent NRC Region III findings on MOV settings (see Inspection Report No. 50-341/88025),

and in anticipation of NRC issuance of a generic letter on MOV design calculation and dynamic test verification, Deco assigned a Nuclear Engineering MOV Issues Task Lead on November 4, 1988, and established a detailed calculation guidelines based on GE NEDC-31322,

" Valve Opening / Closing Differential Pressure Due to Steam / Fluid Acceleration / Deceleration," in December 1988. Staffing and plans for the establishment of permanent controlled files for all MOV data and calculation were in place. The target completion date for the 176 safety-related MOVs is December 1989, and for the 486 non-safety-related MOVs, November 199 b. (IO No. MS-2) In conjunction with the dp issue identified in this 10, the inspector randomly selected four other " generic" design problems and evaluated whether DECO was aware of the issues and what actions had been take (1) Pressure Locking The issue was addressed in NRC IN 88-71, NRC IE Circular 77-05, and INPO Significant Operating Experience Report 84- According to records reviewed by the inspector, DECO evaluated all Q Level I valves and installed pressure relief taps and connections on all affected valves in 198 (2) Thermal Binding The issue was addressed in NRC IN 88-7 The DECO solution is twofold:

• Installation of double disc parallel seat gate valves, and utilization of compensatory spring packs in the MOV operator to absorb initial valve closure forc * Operation procedures, such as NPP-23.138.01 for the recirculation loop, and the Nuclear Operations Training Lessons Guide No. 19, require operators to partially open and close the concerned valves several times when system cooldown takes plac _ - _ _ _ _ _ _ _ _ _ _ _ _ - . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ - _ _ - _ _ - _ -

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, l(3) 'D'egraded c'c Bus' Voltage .

D The issue was addressed in NRC ins 89-11, and 88-72. Deco issued DERs 89-0266, and 88-1717 to evaluate the problem. DECO

determined that the Fermi 2 MOV cable sizing calculations were based on locked rotor. currents, degraded battery voltage,.and-

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conservative conductor resistance values, and concluded the overall-design was adequat (4) Hot Motor Operating Environment 1 The issue was eddressed in NRC IN 88-72 and the Limitorque K Corporation Part 21 notification, dated November 3, 1988. The

, concern was that, in some cases, MOV motors may not develo full' rated starting torque in an elevated temperature environment. DECO issued DER 89-0158 to evaluate the situation. DECO calculation DE-4943,:" Verification of D.C. M0P Operability with Reduced Voltage and Elevated Temperatures,"

dated March 21, 1989, concluded the existing installations to be acceptabl .

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(IO No. EP-2) The ihspector reviewed the revised Deco calculation

DC-0213 " Sizing of 150/260 Volts Batteries," Revision E, dated January 25, 1989, with review focus being placed on the establishment of a de load profile. The inspector concluded that the calculatio was comprehensive; for example, the loading compilation included equipment running liphts and energization of relay coils, and the duration of MOV under locked rotor current was also conservativ The calculation was done in accordance with IEEE Standard 485 - 197 ' Acceptable Licensee Corrective Actions (IO fo. MS-9) The two EECW make-up tank outlet MOV operators were mounted with the limit switches oriented downward. The installation was not in compliance with vendor recommendation and plant design, and the deficiency was not observed / reported by QC during constructio The licensee evaluated the situation and concluded that no specific or generic corrective action was required. The conclusion was based

on the proactive~ maintenance activity requiring checking of the MOV operator switch shaft penetration for seal leakage, and the acceptance of the checking method by the MOV manufacturer. The inspector reviewed Fermi 2 maintenance procedure NPP-35.306.003, "Limitorque Motor Operator - Periodic Inspection," Revision 24, dated April 12, 1989, regarding the issue, and had no adverse coment (10 No. MS-11) RBCCW would not be able to remove LOCA heat loads if

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EECW was not activated. Auto-initiation control circuits of EECW were changed to initiate actions upon high drywell pressure in addition to the existing loss of offsite power supply, and low dp between supply and return headers initiation signal ___ _ _ _ - - _ - _ _ _ =

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. (10 No. MS-11) The )iping including the check valve and the relief j valve installed on t1e N, supply to the make-up tank for pressurizing.

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the EECW system was not seismically qualified. During a postulated line break event coincident with check valve failure to seal; and/or relief valve stuck open, depressurization of the EECW system wculd occur, and piping located at high elevation could experience flashing, which could cause pump cavitation and/or line vibration. The inspector concurred with the DECO plan, to be implemented during the first refueling outage (September 1989), to measure )ressure at the predicted flashing points based on analysis, then su) tract the N, pressure (loss of N, supply). If the resulting pressure is less than 1.27 psia (the vapor pressure of water at 110*F, design  !

temperature when leaving heat exchangers), a system modification may be necessar (IO No. EP-5) Potential thermal ocerload relay (TOR) degradation due to prolonged testing at 600% full load amperage (FLA). Some documented TOR damage was attributed to improper testings done in the past. Test procedure deficiencies identified by DECO, included (1) incorrect utilization of a 1.25 multiplication factcr in

. calculating heater pick-up current value for the 600% FLA tests, and (2) insufficient cooldown periods in between tests. Some TOR damage could be attributed to excessive heat buildup caused by failure to extent periods between tests as TOR ambient temperature increase The problems were corrected in DECO surveillance procedure NPP-42.000.02, " Thermal Overload Relay Calibration," Revision 21, dated March 6, 1989. Deco stated that there had not been any recurrence of the same proble . Licensee SSFI Program and Corrective Action Wecknesses The inspector identified some potential weaknesses in the licensee's SSFI program / actions: (10 No. EP-2) Retrieval of design basis documents for the revised battery calculation was difficult because not all essential equipment activation sequence tables, one-line electrical flow diagrams, equipment schematics, and vendor technical data were referenced in the calculation. The present DECO design control measures have been improved, but additional efforts should be applied to obtaining design basis documentation and associated verificatio (IO No. EP-2 and EP-4) The revised battery sizing calculation showed that the present Division II batteries will be adequate for a j design service life.of 20 years provided that the electrolyte '

temperature is maintained at approximately 77*F. The present weekly and quarterly surveillance (POM 42.309.01(SR), and 42.309.02(SR))

require checking and maintaining the average electrolyte temperature i

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to be above 60 This surveillance requirement was questionable for the following reasons:

• The 60 F TS based temperature is without technical basis. The battery sizing calculation for a 20 year operation was based on 77 F design temperatur * There is no quantitative upper bound acceptance criteria to prevent operating in a high temperature environment, which would shorten battery life, (10 No. EP-4) The SSFI response indicated that a TS change should be made in conjunction with Potential Design Change 8244, completed in July 1988, which require divisional battery rooms to be maintained at between 75 F and 90 F. This rqsponse had not been irrplemented by the license < (10 No. EP-4) The SSFI respons'_ stated that weekly divisional battery room temperature surveillance procedure P0M 42.309.01(SR)

should include acceptance criteria. The procedure was revised during the course of this inspectio (10 No. IC-4) Setpoint calibration data was not trended for adjustment to calibration cycle. A new instrument setpoint trending procedure NPP-CT1-05 was issued for use on February 1, 198 The inspector reviewed the content, and had no adverse comment For implementation assessment, the inspector visited the site instrument maintenance shop, and observed the following problems:

• Some setpoint trending was performed based on informal guidanc Not all as-found and as-left instrument loop surveillance data was recorded in the history / trend cards, such as Surveillance No. 44.010.166, " Quarterly Flow Unit C Loop Test on Reactor Water Recirculation," completed on March 4, 198 * As of June 28, 1989, five months after the issuance of the new procedure, it had not been implemented at the work statio * The responsible nuclear I&C foreman was not aware of the existence of the new procedur . Conclusion DECO program for conducting MOV design evaluations and dynamic functional tests is comprehensive.

l DECO re-calculation for sizing the Class 1E batteries is considered to be extensive; however, not all design basis documents were included or referenced in the ne calculatio _- _

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. a Deco effort in the areas such as monitoring the divisional battery rooms temperature, and instrument set point trending was not as comprehensive or timely as was expecte . Exit Meeting The inspector met with licensee representatives (denoted in Paragraph 1)

on June 29, 1989, at Fermi 2 Nuclear Power Station and summarized the purpose, scope, and conclusions of the inspection. The licensee stated that the inspector had no access to proprietary information.

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