IR 05000266/1989012

Insp Repts 50-266/89-12 & 50-301/89-11 on 890425-0512.No Violations Noted.Major Areas Inspected:Corrective Actions Initiated in self-initiated SSFI of Emergency Diesel Generator Sys

ML20244C172
Person / Time
Site:	Point Beach
Issue date:	06/06/1989
From:	Phillips M, Yin I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20244C170	List: IR 05000266/1989012 ML20244C167
References
50-266-89-12, 50-301-89-11, NUDOCS 8906140151
Download: ML20244C172 (7)

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U.S. NUCLEAR REGULATORY COMMISSION REGION III  ;

Reports No. 50-266/89012(DRS); 50-301/89011(DRS)

Docket Nos. 50-266; 50-301 Licenses No. DPR-24; DPR-27 Licensee: Wisconsin Electric Power Company 231 West Michigan Street - P379 Milwaukee, WI 53201 Facility Name: Point Beach Nuclear Plant, Units 1 and 2 Inspection At: Two Rivers, WI 54241 Inspection Conducted: April 25-28, and May 10-12, 1989 Inspector: in / ~

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$ $ $9/ Date Approved By: M. P. Phillips, Chief 6[ f Operational Programs Section Date Inspection Summary Inspection on April 25-28, and May 10-12, 1989 (Reports No. 50-266/89012(DRS);

No. 50-301/89011(DRS))

Areas Inspected: Routine, announced inspection of licensee corrective actions initiated for the issues identified in its self-initiated Safety Systems Functional Inspection (SSFI) of the Emergency Diesel Generator (EDG)

system. The inspection was performed based on selected portions of NRC Inspection Procedure 90713 and 3070 Results: The inspection concluded that the licensee had not provided adequate nor iTi5ely corrective actions for most of the issues selected by the inspector for review, such as electrical protective device coordination and compliance with 10 CFR 50.59 safety evaluations for plant modifications.

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DETAILS Persons Contacted Wisconsin Electric Power Company (WE)

• C. W. Fay, Vice President, Nuclear Power Department G. Frieling, Superintendent, Systems Engineering

+* Rinzel, QA Engineer P. J. Katens, Senior Project Engineer, Electrical

+*R. Heiden, Superintendent, Nuclear QA

'+*G. Krieser, General Superintendent, QA

• R. Newton, General Superintendent, Nuclear Systems Engineering and Analysis Section S. Schellin, Superintendent, Reactor Engineer D. Bell, Project Engineer B. Lunde, Specialist, Mechanical

+ J. E. Knorr, Regulatory Engineer

+F A. Flentje, Administrative Specialist

+ W. B. Fromm, Modification Engineer U.S. Nuclear Regulatory Commission

+ R. J. Leemon, Resident Inspector i

• Indicates those attending the exit meeting in the WE corporate office on

April'28, 1989.

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+ Indicates those attending the exit meeting at the site on May 12, 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 Emergency Diesel Generator (EDG) system was selected to be the first system receiving the S$FI. The SSFI team was composed of nine Westec staff and three supporting WE engineer The inspection was conducted on January 4 through February 5,1988, with the final report being issued on March 15, 198 The NRC inspector reviewed the Westec report in the WE corporate office on March 27-30, 1989, and concluded that the SSFI review scope was extensive, and the SSFI was conducted in a professional manner.

l The purpose of the NRC inspection conducted on April 25 through May 12, 1989, was to evaluate the adequacy of licensee corrective actions for the issues identified during the SSFI, and assess the effectiveness of the corrective actions take i

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. Review Sample Selection The inspector considered the following Inspection Observations (10s)

contained in the SSFI report to be technically significant issues:

No. DCP-5 Inattention to modification design details involving fuel oil supply flow resistance, flow rate, gravity flow scheme, and friction factor selectio No. DCP-6 Safety evaluations not performed, or inadequate for a number of modification No. DCP-7 Safety relief valves incorrectly installed on EDG starting air accumulators. Deficiencies identified in Spare Parts Equivalency Evaluation Determination (SPEED) progra No. GM-1 Lack of formal battery sizing calculation for the DOS and D06 batterie No. GM-3 DC distribution bus short circuit breaker exceeded main breaker UL ratin No. GM-8 Inadequate breaker coordination study per 10 CFR Part 50, Appendix R to ensure power supply for plant safe shutdow No. GJ0-1 Lack of flow balance calculation for service water to determine the effects of the addition of the fourth heat exchanger onto the syste No. RB-2 The four undervoltage channels on the 4.16 kV buses not adequately tested to ensure EDG auto start. LER 88-003-00 was forwarded to NRC on March 8, 1988, and updated on September 6, 1988, and on December 19, 198 No. RB-3 Inadequate testing of EDG start 1 and 2 auto control circui No. WCS-1 Some EDG preventive maintenance procedures not in compliance with manufactures' recommendation No. WCS-7 Inadequate root cause analysis to determine failure modes, and trend equipment failure No. WGD-1 EDG having potential of being overloaded during injection and recirculation phase of operation The inspector selected 10 Nos. DCP-5, DCP-6, GM-1, GM-8, and WGD-1 for review and assessmen ___________ ____ _ ___-________-_-__ _ _ _

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l Licensee Corrective Action Weaknesses j The inspector. identified some potential weaknesses within the licensee's corrective action program / action l' (10-No. DCP-5): In the area of piping seismic analysis, the inspector identified that a calculation, dated July 18, 1985, for a proposed-piping configuration, that had since been revised, was not voided, and-no new calculation was performed for the actual piping that had been installed in December 1985. QA closed this 10 on January 13, 1989, without verification of the seismic design adequac (10 No. DCP-6): The 10 stated that safety evaluations (SEs) were not consistently performed for all modifications to determine if unreviewed safety questions exist. Specifically, the following five modifications-were.without.any SEs because they were not nuclear. safety-related:

M-605 Wet pipe sprinkler systems were installed in safety-related equipment area Buried fuel oil transfer line was re-routed to allow construction of the addition to the gate hous >

.M-515 An inspection port was added to the engine exhaust-manifol Drain flaps were installed.to prevent possible backflow flooding in the EDG room in the event of flooding elsewhere in the turbine buildin PT.R-5 Battery DOS test acceptance final voltage was changed from 105 volts to 103.25 volt The WE disposition, documented in a Management Supervisory Staff Meeting (MSSM) meeting minutes No. 88-13, dated June 20, 1988, concluded that an SE should have been performed for M-605, but not for the rest. The inspector reviewed the four where WE disagreed with the SSFI finding, and concluded:

Agreement with VE disposition for M-515 and 83-11 *

(82-51) A 50.59 evaluation which included the seismic effects regarding the re-routing of the Seismic Category I fuel oil transfer line should have been conducte (PT.R-5) The 105 volts is based on 60 cells at 1.75 volts per cell (per design); and the 103.25 volts is based on 59 cells at 1.75 volts per cell (as-built). There was no SE L performed on design modification from 60 cells to 59 cell ,

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. (10 No.'DCP-6): During a discussion with the site Modification Engineer on whether a seismic evaluation should be performed for the re routing of the Seismic Class 1 fuel line, he insisted that since the re routed piping has the same pipe size, same material, only one 90 elbow was added, and installation was performed based on the same specification, no SE to address seismic effects on piping was required. This is incorrect. An SE is required if a change to the facility described in the FSAR is made. Therefore, the original finding was valid, and an SE was required, since the Emergency Diesel Generator's fuel supply was discussed in the FSAR. See additional discussion on the matter in Paragraph d. (IO No. DCP-6): In a followup of this 10, WE contracted Westec to conduct an evaluation of all 241 modifications that were performed in 1987. The Westec evaluation was completed on April 8, 1988, and concluded that 88 of the 241 modifications should have received 10 CFR 50.59 SEs, but had no To-date, no documentation was available to show how WE resolved this finding, nor were plans in place to look previous to 198 (IO No. DCP-6): The inspector reviewed WE procedure QP 3-3,

" Authorization of Changes, Test, and Experiments (10 CFR 50.59)

Reviews," Revision 2, dated October 3, 1988. QP 3-3 allows onsite personnel, through discussions such as MSSMs, to decide whether an SE is required for a given plant modification without Offsite Review Committee concurrenc Although QA performs periodic checks on modifications without SEs being performed, its past focus has been consistently placed on programmatic compliance. The present WE review process is contrary to ANSI N 18.7-1976, Section requirements which WE committed to in the Point Beach FSAR Section (10 No. GM-1): The inspector's findings on batteries were incorporated into another Region III Inspection Report No. 50-266/89016; No. 50-301/8901 (10 No. GM-8): The issue of inadequate breaker coordination study was identified by the SSFI team on March 15, 1988; and there has not been any licensee actions taken to address the specific safety concerns identified in the SSFI findings to-dat Impe11 has been contracted to complete the re-run of 10 CFR Part 50, Appendix R portions of the protective devices coordination study by December 31, l 1989, and complete studies for the rest of the safety-related circuits by June 1, 1990. In review of the WE design specification for Impell in PB-347, " Requirements for Contractor Services, Electrical Analysis / Design Projects," Revision 0, dated July 20, 1988, the inspector determined that coordination acceptance criteria for match or mix of relays, circuit breakers, and fuses had not been specified in the document.

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5. Acceptable Licensee Corrective Actions (10 No. DCP-5): The inspector reviewed the updated WE Calculation No. N-88-036," Diesel Generator Day tank Gravity Fill," dated June 21, 1988, and had no adverse comments. The calculation was verified by actual flow tes (10 No. DCP6): In conjunction with Paragraph 4.d. Westec selected 25% of the 153 1987 modifications that had received SEs for review, and concluded five SEs (for Modifications87-002, 085, 089, 098, and 066) were inadequat WE followup review concluded these five SEs were performed adequately. The inspector reviewed the following WE onsite MSSM meeting minutes:

87-18, dated September 8, 1987 87-21, dated October 12, 1987 88-22, dated October 18, 1988 89-04, dated March 16, 1989 The inspector concurred with the licensee's review method and conclusio (IO No. WGD-1):

(1) The SSFI team recommended that some non-safety-related loads, such as the control room filter fan and air conditioning unit, should be-powered from a safeguard bus. WE plans to commence a study regarding this issue on June 1, 1989, and to complete it before September 1, 198 (2) Various loads powered by the safeguard MCCs have not been accounted for in accordance with the FSAR Section 8 analysi WE plans to perform a safeguard bus load study that is based on worst case loading involving LOCA occurring in one unit, shutting down of the second unit, and loss of offsite power at the same time. The study will be completed by December 31, 1989.

t- (3) A more detailed load analysis model will be established based on actual expected electric loads under most potential adverse conditions. This analysis will be a part of the station blackout studies recently required by the NRC. The scheduled completion date ic December 31, 1990. The licensee installed a gas turbine driven electric generator at the site that can tie into the auxiliary AC power system as a backu _ _ _ - _ _ _

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6, Conclusion- The licensee corrective actions for the SSFI findings were not in all cases considered timely to address specific safety concerns; for example, the breaker coordination problems identified in March 1988 will not be resolved until June 1990 (Paragraph 4.g). The licensee control of 10 CFR 50.59 safety evaluations has been improved as discussed in Region III Inspection Reports L No. 50-266/89005; No. 50-301/89005; however, there are still a number of specific issues that should be addressed by the license (1) Resolution of the 88 1987 modifications that were without safety evaluations (Paragraph 4.d).

(2) Based on the results of the 88 1987 modifications evaluation, a determination should be made by WE on extending its evaluations to 1986 modifications and beyon (3) The licensee's past determinations, that a deifications on non-safety-related systems or components required no safety evaluations, may not be in conformance with 10 CFR 50.59 requirements. The old dispositions should be reviewed for correct 50.59 applicability, and an SE performed if the change involved FSAR described item (4) The licensa's present practice to let onsite personnel determine whether a modification shculd receive a safety evaluation without any offsite committee overview was contrary to ANSI N 18.7, and was identified to be deficient by the SSFI team findings (Paragraph 4.b), Westec follewup review (Paragraph 4.d), and the inspector's observation (Paragraph 4.c and 4.e), The WE followup for the station battery design capacity and test issues identified by the SSFI team was inadequate. (See Inspection Reports No. 50-266/89016; No. 50-301/89015). There was a lack of followup of seismic design deficiencies identified by the SSFI team. Furthermore, the QA group cloted the finding apparently without technical review of the specific issues (Paragraph 4.a). The licensee's corrective action to ensure EDG fuel oil gravity flow in accordance with the plant design condition (Paragraph 5.a), and plans to ensure adequate power supply for all safeguard load demands during worst case design conditions including station blackout (Paragraph 5.b) was considered to be adequat . Exit Interview The inspector met with licensee representatives (denoted in Paragraph 1)

on April 28, 1989, in the licensee's corporate office and on May 12, 1989, at the Point Beach Nuclear Plant site and summarized the purpose, '

scope, and findings of the inspection. The licensee stated that the inspector had no access to proprietary informatio