IR 05000282/1989010
| ML20244C221 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/06/1989 |
| From: | Phillips M, Yin I NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20244C217 | List: |
| References | |
| 50-282-89-10, 50-306-89-11, NUDOCS 8906140166 | |
| Download: ML20244C221 (8) | |
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l U. S. NUCLEAR REGULATORY COMMISSION
REGION III
Reports No. 50-282/89010(DRS); 50-306/89011(DRS)
Docket Nos. 50-282; 50-306 Licenses No. DPR-42; DPR-60 Licensee: Northern States Power Company 414 Nicollet Mall Minneapolis, MN 55401 Facility Name: Prairie Island Nuclear Generating Plants, Units 1 and 2 Inspection At: Prairie Island Site, Red Wing, Minnesota Inspection Conducted: April 11-13, 1989 h/, f/
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Inspector:
1. T. Yin
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Date hf
!d,7 Approved By: Monte P.
illips, Chief Operational Programs Section Date Inspection Summary Inspection cn April 11-13, 1989 (Reports No. 50-282/89010(DRS);
No. 50-306/890ll(DRS))
Areas Inspected: Routine, announced inspection of licensee plant design changes and modification program improvement in light of issues identified by NRC inspections at Monticello. This inspection was based on selected portions of NRC Inspection Procedures 37700 and 30703.
Results: The present permanent modification, and temporary functional bypass programs lacked design oriented engineering input in the modification control processes. Some of the corrective measures being initiated by NSP were not considered to be timely. The modification program deficiencies identified at Monticello in 1988, and the bypass program deficiencies identified at Prairie Island in 1985 had not resulted in associated program upgrades as of the inspection date. However, the inspector concluded that there were strengths within NSP's nuclear audits in responding to the NRC concerns. NSP's site engineering department's awareness of NRC's Safety System Functional Inspection (SSFI) findings and the actions taken to address similar problems at Prairie Island, were also strengths. The inspection did not reveal any safety system operability concerns.
8906140166 890606 PDR ADOCK 05000282 O
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DETAILS 1.
Persons Contacted Northern States Power Company (NSP)
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- E. Watz1, Plant Manager
- A. Smith, General Superintendent, Planning and Services
- K. Albrecht, Director, Power Supply Quality Assurance
- G. Miller, Superintendent, Operations Engineering
- T. Silverberg, Lead Production Engineer
- A. Hunstad, Staff Engineer
- E. Eckholt, Senice Engineer
- J. Goldsmith, General Superintendent, Nuclear Technical Services
- B. Stephens, Superintendent, Design Standards
- C. Hoglin, QA Specialist
- D. Mendele, Plant Superintendent, Engineering and Radiation Protection
- J. Ruether, Lead Engineer E. Burke, Senior Engineer C. Mundt, I&C Engineer P. Hellen, Electrical Engineer D. Krech, Nuclear Operations Quality Supervisor K. Beadell, Superintendent, Technical Engineering J. Schuelke, System Engineer G. Thoraldson, System Engineer U. S. Nuclear Regulatory Commission (NRC)
- J. Hard, Senior Resident Inspector
- Indicates those attending the exit meeting at. Prairie Island Nuclear Generating Plant on April 13, 1989.
Other licensee personnel were contacted as a matter of routine during the inspection.
2.
Introduction NRC Region III performed a design modification inspection at the Monticello nuclear power plant in September 1988 (Report No. 50-263/88017),and identified control process weakness and procedure implementation deficiencies.
In responding to the findings, NSP presented a " Plant Design and Modification Control Process Current and Future Action Program," to NRC Region III management on November 17, 1988. The issues were followed up during an NRC inspection in December 1988 (Report No. 50-263/88027),
which concluded that improvements had been made at Monticello. NSP subsequently presented a "Monticello Modification Audit Status Report,"
to NRC Region III management on January 11, 1989, The purpose of this inspection was to review and evaluate NSP actions i
taken to correct similar problems, which the licensee had determined
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existed at Prairie Island (PI).
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3.
Documents Reviewed NSP Corporate Nuclear Administrative Control Directive (ACD)
N1 Administrative Work Instructions (AWIs) for PI, Revision 2, dated December 31, 1987.
5.1.1, Nuclear Plant Modification General Instructions, and Advanced Change, dated January 29, 1988.
5.1.3, Design Inputs.
5.1.4, Internal Design and Review.
5.1.9, Safety Evaluation.
5.1.13, Installation and Construction Test Procedures.
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5.1.14, Pre-operational and Operational Test Procedures.
5.1.17, Plant Design Change Close-Out.
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5.1.19, Alteration Process.
PI 5ACD 3.9, Bypass Control, Revision 8, August 1988.
SACD 6.5, Temporary Modification, Revision 0, draft.
Monticello Audit Report No. AG 88-54-12, "Monticello Modifications,"
January 27, 1988.
PI Audit Report No. AG 87-43-3, on process senors response time deficiency event followup.
PI Audit Report No. AG 89-3-12 "IEB 79-14 Modifications,"
February 17, 1989.
PI Audit Report No. AG B3-39-12, "PI Delta - I Instrument Incident of July 25, 1988," September 2, 1988.
Modification No. 88A 0059 to modify Foxbcro nuclecr flux tilt 62-H-2E style C controlierc requested on July 27, 1988, completed on July 28, 1988.
Modification ho. 85L867 to improve vibration measurements on auxiliary feedwater pumps, requested on July 70, 1968, not yet completed.
Active Bypasses (temporary modifications not yet closed out):
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Request Nos.
Date 7356 December 27, 1988 7129 February 13, 1988 6758 December 26,.1986 5901 August 28, 1983 PI LER l-88-04, Unit Shutdown Required Due' to Degraded Delta - I Reactor Trip Channels," Revision 1, dated October 12, 1988.
4.
Licensee Modification Program Strengths The inspector observed the following licensee actions relative to plant system modifications which can be interpreted to be strengths within its overall program.
a.-
The inspe'ctor reviewed four NSP audits (refer to Paragraph 3)
performed at Menticello and PI, and had the following comments:
(1) AG 88-54-12, performed on September 27, 1988, through January 9, 1989, was comprehensive, and revealed a substantial number of problems concerning Monticello plant modifications; however, specific safety system functional operability was not compromised.
(2) AG 87-4-43-3, presently tracked under NRC Region III inspection unresolved items No. 282/88012-01; and No. 306/88012-02, was an effective technical audit, which identified a safety issue involving inadequate measurements of protective channel process sensors-(transmitters) response time in line.with the station accident analysis. The audit followup involved overview of improvement made in testing efforts.
'(3) AG 89-3-12, performed on January 9 through February 2,1989, identified deficiencies in piping modifications, and was being expanded to more generic issues, such as Engineering Change Request process and modification processes control.
(4) AG 88-39-12, performed in July and August 1988, was very responsive to the PI Notification of Unusual Event on the Foxboro Delta - I controller inoperability problem.
b.
The inspector randomly selected three commonly seen system design deficiencies identified by the NRC, and/or other licensee self-initiated SSFIs and evaluated whether NSP was aware of the issues and what action they had taken. Although NSP has not performed any SSFI, the PI site engineering staff was familiar with the issues, and stated that the following actions have been taken at PI:
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(1) Protective Devices Coordination H
NSP contracted with Impell to perform en electrical coordination study in 1986 and 1987. The scope included:
station auxiliary transformer down to 120 Vac circuits, battery chargers to
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distribution panels, and instrument inverters to distribution panels. Changes.in breaker types and/or settings were'made as a result of the study.
a-Any modifications to the station's electrical system now
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require'the following reviews:
(a) NSP corporate staff conducts relay setting analysis on all circuit breakers at the 480 Vac bus level and above.
(b)' PI staff conducts circuit coordination analysis on all modifications below the 480 Vac bus level. This is done-using a computer program obtained from Impe11.
(2) Electric Load Management NSP contracted with Fluor to conduct a study, in conjunction with the station blackout project, to assess present bus
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loading and to analyze effects of reduced voltage.
Modifications to existing bus loads are now evaluated and concurred in by the PI staff using_the F.luor study as a reference. The evaluation covers'the effects on loading, cable sizing, cable routing, reduced voltage, and 10 CFR Part 50, Appendix R criteria applicability.
(3) Battery Loadings In 1982, NSP contracted with Fluor for a battery (125 Vdc)
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load study to determine if the batteries could meet.the most limiting scenario defined in the USAR. The study identified that one major non-vital load, i.e., the turbine dc' lube oil p>tmp, needed to be removed from the safeguards batteries.
In 1985, two new inu rters were added to the de system.
Evaluation of this additional load showed that the batteries were still adequately sized. The licensee is having the battery load profile updated in preparation for a calculation that will be based on IEEE 485 methodology.
Additional battery loads are handled in the following manner:
(a) PI staff analyzes and must concur on minor load additions.
(b) Fluor engineers are responsible for concurrence on any request for major load additions.
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5.
. Licensee Modification Program Weakness The inspector identified some potential weaknesses within the licensee's
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modification' program / actions as follows:
. a.
Administrative Control Weakness
'The' inspector reviewed the AWIs listed in Paragrapn 3 and had the following comments:
(1) There'is a lack of design engineering participation to determine original design basis or' criteria and commitments to the NRC during the 10 CFR 50.59 evaluation preparation.
(2) There is a lack of design engineering participation in the determination of appropriate material' certification, inspection, installation and fabrication procedures, and observation of acceptance or qualification tests.
(3) -The present control for the Bypasses (temporary modifications)'
does not provide for marking revisions on essential drawings, testing after installation and removal of bypasses, or verification of installation.
Some of'these problems were identified as far back as 1985.
The inspector's review of:
some of the past onsite Operations Committee meeting minutes indicated that past efforts to reduce the number of old (more
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than five years) Bypasses was minimal.
It was noted that efforts have improved; however, there are still a number of old Bypasses which remain active (1-1982;,4-1983; 4-1986; and 5-1988).
Any aspects of the modification process, including evaluation, requirement establishment, verification, procurement, and testing, could be enhanced with design engineering input and involvement.
Discussion can be found in the following NRC publications:
NRC NUREG/CR-5147, Fundamental Attributes of a Practical
Configuration Management Program for Nuclear Plant Design Control.
NRC-NRR Ger.eric Letter 89-02, Actions to Improve the Detection of Counterfeit and Fraudulently Marketed Products, dated March 21, 1589.
b.
Modification Action Weaknessy (1) (Modification No. 88A-0059) On July 26, 1988, a Xenon build-up following a Unit 1 load reduction caused an axial flux difference (AFD) sufficient to adjust the OTAT setpoints.
One of the four OTAT setpoint channels was found unresponsive to the AFD input.
The licensee investigation showed that the affected flux tilt controller was a Foxboro 62-H-2E Style C, while the other three controllers, which responded properly, were Style B.
All four.
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controllers in Unit 2 were Style C..
Both units were shutdown,
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and all Style Cs were modified by reversal of input signal.
polarity. Both units were returned to power following testing:
to verify-the operability of the modified controllers.
The inspector's review of the modification package and the LER (see Paragraph 3) determined the following:
The root cause/ accountability was not addressed by.NSP.
- The root cause of the deficiency was attributed to inadequate original design control by Westinghouse in safety instrument evaluation and selection, and the past site system surveillance test not being able to reveal this deficient design condition. PI staff stated that documentation will be revised to address.the issue.
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Any effects of polarity reversal on the Foxboro 62-H-2E'
Style C controller, including possible alteration of-performance characteristics, had not been evaluated by.NSP in accordance with 10 CFR 50.59 requirements. The failure to perform a safety evaluation is a violation of 10 CFR 50.59 requirements. No Notice of Violation will be issued due to NSP's prior notification to NRC that they had-identified the problems, and had initiated measures to correct these problems. PI staff stated that additional inquiry on the issue will be forwarded to Westinghouse and Foxboro.
(2)
(Modification No. 85L867) The purpose of this modification was to improve _the vibration monitoring and trending capabilities for both Units' auxiliary feedwater pumps (AFWPs). The scope included installation of a velocity transducer at each AFWP bearings, and a third transducer to be installed on the thrust end of the pump for monitoring axial vibration. The inspector observed the partial transducer installations on April 13,.1989, and had no adverse comments..
The inspector rev!ewed the modification package, and commented
.that axial vibration criteria should be added to the routine test procedure. The present test acceptance criteria for both motor driven and turbine driven AFWPs were documented'in surveillance l
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test procedures, such as SP 1100 and SP 1102. The reference, acceptable, alert, and action pump vibration ranges within these
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procedure were based on PI ISI/IST Second Ten Year Program, Section 1.5-4 in accordance with ASME Section XI IWP-4510, Vibration Amplitude. The inspector pointed out that IWP 4510 addressed only vibration measurements perpendicular to the rotating shaft. The acceptance criteria for axial vibration, which could indicate possible bearing misalignments, should be established by NSP engineers.
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Conclusion a.
The recent audits performed by NSP Nuclear Operations QA Department were considered to be effective in responding to NRC concerns and plant significant events.
b.
The plant technical staff was familiar with findings identified in the past SSFIs conducted by HRC and other nuclear licensees.- Actions.have been initiated to resolve similar issues at PI.
c.
The weaknesses observed and discussed in Paragraph 5 could be attributed to the lack of sufficient design engineering input and overview. For example:
The procedures for modifications and bypasses could be
strengthened by ensuring that design basis documents are reviewed or established prior to initiation of design revisions.
There was no safety evaluation on the possible change of performance characteristics resulting from the reversal of Foxboro component circuit polarity.
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There was a lack of acceptance criteria for pump axial vibration testing.
7.
Exit Meeting The inspector met with licensee representatives (denoted in Paragraph 1)
on April 13, 1989, at Prairie Island Nuclear Generating Plant and summarized the purpose, scope, and findings of the inspection. The inspector discussed the likely informational content of the inspection report with regard to documents reviewed by the inspector during the inspection. The licensee did not identify any such documents as proprietary.
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