IR 05000282/1997008
| ML20217R346 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/27/1997 |
| From: | Grobe J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | Wadley M NORTHERN STATES POWER CO. |
| References | |
| 50-282-97-08, 50-282-97-8, 50-306-97-08, 50-306-97-8, EA-97-290, NUDOCS 9709050079 | |
| Download: ML20217R346 (36) | |
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l August 27, 1997 EA 97 200 Mr. M. Vice President, Nuclear Generation Northern States Power Company 414 Nicollet Mall Minneapolis, MN 66401 SUBJECT: NRC PREDECISIONAL ENFORCEMENT CONFERENCE
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Dear Mr. Wadley:
This refers to a Predecisional Enforcement Conference conducted by Mr. A. Bill Beach, Regional Administrator, and other members of the Region lli and Headquarters staff on August 8,1997. The subject of this conference was the apparent violations at Prairie Island identified in NRC Inspection Report 50 282(306)/97008(DRS). The purpose of this conference was to discuss the violations, root causes, contributing factors, and corrective actions.
You wPI be notified by separate correspondence of our decision regarding the enforcement action based on the information presented and discussed at the Predecisional Enforcement Conference. No response is required until you are notified of the proposed enforcement action.
j During the Predecisional Enforcement Conference, you acknowledged the test control and j corrective actions violations related to test acceptance criteria. You also acknowledged
< that the USAR had not been updated in a timely manner with respect to the auxiliary feedwater system flowrotes under a main feedline break, and presented information clarifying the auxiliary feedwater system design and licensing basis. Also presented was a 7f synopsis of the causes, safety significance, and corrective actions taken for each potential violation.
9709050079 970827 hkflflfllflf[jfllf PDR ADOCK 05000282 e
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M. August 27, 1997 The presentation also included a detailed review of your corrective action program status including actions related to design control calculation verification and information on your USAR Upgrade project. A copy of the handouts provided by Northern States Power and the NRC are enclosed.
Sincerely,
/s/ J. M. Jacobson (for)
John A. Grobe, Acting Director Division of Reactor Safety Decket Nos. 50 282:50 306 Licenses No. DPR 42: DPR-60 Enclosures: 1. List of Attendees 2. Licensee's Handout 3. NRC Handout cc w/encts: Plant Manager, Prairie Island
John W. Ferman, Ph.D.
Nuclear Engineer, MPCA State Liaison Officer, State of Minnesote State Liaison Officer, State of Wisconsin Tribal Council, Prairie Island Dakota Community Distribution:
Docket File w/ encl SRis, Prairie Island, TSS w/enct
[PUBLIC Nw/ enol Monticello w/enci J. Lieberman, OE w/ encl OU/LFDCB w/enci LPM, NR:1 w/enci J. Goldberg, OGC w/enci DRP w/enci-- A. B. Et> '- Rill w/ encl R. Zimmerman, NRR w/ encl DRS w/enci J. L, Caldovil, Rlll w/enci Rlli PRR w/ encl Rlli Enf. Coordinator w/enct SEE ATTACHED CONCURRENCES DOCUMENT NAME: G:\PRA082 7.DRS ~
Ta receive a copy of this document, Indicate ln the box: "C" o Copy w/o attachment / enclosure "E" = Copy with attachment / enclosure *N" * No copy IOFFICE Rlll:DRS lRlil:DRS l Rlll:DRP Rlli:EICS l Rlli:DRS $ N '
NAME Guzman/kJc Ring McCormick Barger Clayton Jacobso@ Whe)
DATE 08/ /97 08/ /97 08/ /97 08/ /97 08Q /97 2 W'
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OFFICIAL RECORD COPY
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ls M. The presentation also included a detailed review of your corrective action program status including actions related to design control calculation verification and Information on your USAR Upgrade project. A copy of the handouts provided by Northern States Power and the NRC are enclosed.
Sincerely, John A. Grobe, Acting Director Division of Reactor Safety Docket Nos. 50 282: 50 306 Licenses No. DPR 42; DPR 60 Enclosures: 1. List of Attendees 2. Licensee's Handout 3. NRC Handout cc w/encis: Plant Manager, Prairie Island John W. Ferman, Ph.D.
Nuclear Engineer, MPCA State Liaison Officer, State of Minnesota St9te Liaison Officer, State ,
of Wisconsin Tribal Council, Prairle Island Dakota Community 5-Distribution:
Docket File w/enci - Rlli PRR w/encI J. L. Caldwell, Rlil w/enci
PUBLIC IE-01 w/enct SRis, Pralrle Island, Rlli Enf. Coordinator w/enci OC/LFDCB w/enci Monticello w/enci TSS w/enci DRP w/enci LPM, NRR w/enci DRS w/ encl A. B. Beach, Rlli w/ encl DOCUMENT NAME: G:\PRA082 7.DRS ~
To receive a copy of _this document, Indicate ln the bort *Ca e Copy w/o attachment /endoeure "E" = Copy win attachnent/endosure *N's No copy OFFICE Rill:DitS l C- Rill:DRS c, Rill:DRP IV Rlli:EICS _ Rill:DRS l NAME Guzman/kje d Ring n in r MR. McCormick Bargern Clayton f4 Jacobson/Grobe DATE 08n$ /97 / 08/as/97 08/M/97 MIOJ 08/21/97 08/ /97 OFFICIAL RECORD COPY '
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LIST OF ATTENDEES Northern States Power E. Watzl, President, Nuclear Generation M. Wadley, Vice President, Nuclear Generation J. Sorensen, Plant Manager T. Amundsen, General Superintendent Engineering -
K. Albrecht, General Superintendent Engineering 9. Heideman, Superintendent Mechanical Systems Engineering M. Heller, AFW System Engineer NBC A. Beach, Regional Admini:trator J. Grobe, Acting Director, Division of Reactor Safety (DRS)
. M. Depas, Acting Deputy Director, Division of Reactor Projects (DRP)
M. Ring, Chief, Lead Engineers Branch, DRS -
J. McCormick Barger, Chief, Reactor Projects Branch 7, DRP
' J. Guzman, Reactor Engineer, DRS P. Krohn, Resident inspector, DRP G. O'Dwyer, Reactor Engineer, DRS J. Heller, Enf orcement Coordinator D. Diec, NRR Project Manager
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ENCLOSURE LICENSEE HANDO'JT
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PREDECISIONAL ENFORCEMENT CONFERENCE AUGUST 8.1997
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AGENDA
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AFW SYSTEM FLOW REOUIREMENTS AFW LICENSING DOCUMENTS TIMELINE
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EXTENT / SHORT TERM MRRECTIVE ACTIONS e TEST CONTROL AND UNTIMELY CORRECTIVE ACTION
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APNTEST ACCEPTANCE CRITERIA
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EXTENT OF TEST CONTROL ISSUE
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UNTIMELY CORRECTIVE ACTIONS ON AFW, RH; CS
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REVIEW OF OTHER ASME SECTION XI PUMPS
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SHORT TERM CORRECTIVE ACTIONS l
e CORRECTIVE ACTION PROGRAM STATUS
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STRENGTHS
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WEAKNESSES
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CHANGES e DESIGN CONTROL CALCULATIONS
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CORRECTIONS & SIGNIFICANCE
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CALCULATION VERIFICATION MEASURES
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CORRECTIVE ACTIONS '
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Prairie Island AFW System
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Flow Requirements
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Loss of Cooldown Main Main LOCA Feedwater Feedline Steamline (station . Break -Break blackout)
FSAR 200 'gpm 200 gpm Not Not 200 gpm Required Required (analysis used 200 gpm)
USAR 200 gpm variable 400 gpm 200 gpm 200 gpm (200 gpm AOR)
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AFW Licensing Documents
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Timeline
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Date Description Comments 09/69 Westinghouse letter PlW-P-519 Description of generic AFW
requirements for 2 Loop plants with supporting analysis, uses !
400 gpm for MFLB 10/69 Westinghouse letter PlW-P-540 Description of specific AFW requirements for PI and supporting analysis (AOR),200 gpm used for MFLB 10/69 Pioneer letter PIP-N-353 Description of Pi AFW requirements and conservatism analysis 08/73
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FSAR Approved MFLB analysis not included
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AFW Licensing Documents
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Timeline (cont'd?
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Date Description l
_ Comments i 02/81 NSP Letter to Director N REG 0737 submittal, j NRR, AFW System Westinghouse
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information contracted to provida i
- AFW requirements, ,
states MFLB analysis
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! references PlW-P-540 06/82 USAR Rev. O Submitted incorporates 0737 '
information :
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Design Basis Document Operability and program notes re' portability assessed inconsistencv
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06/93 Nuclear Anaiysis No operability or Department requested reportability. Item was !
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to perform MFLB given lower priority analysis verification !
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. J AFW Licensing Documents l Timeline (cont'd)
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Date Description Comments
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09/93 USAR Update Special update to incorporate SBO mods i
06/94 USAR Update NAD analysis verification not complete 12/95 USAR Update NAD analysis verification not complete 05/97 NAD analysis verification Verification by aitemate completed calculation (analysis)
l 07/97 Safety Evaluation for Safety evaluation verifies MFLB USAR Update 10/92 assessments.
completed 08/97 Correction to 02/81 submittal 09/97 USAR Update Submittal
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e Prairie Island's Configuration Management Efforts
- 34 Design Bases Documents (DBDs)- allissued and verified
- 910 Follow-On-Items (FOls)- all assessed
- 1311 Action items (135 remain open)
- 92 USAR Actions (22 remain open: 8 to be submitted 9/97, for 14 remaining items all SE's to be approved by 12/98)
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TEST CONTROL e AFW TEST ACCEPTANCE CRITERIA
- Non conservative ASME Section XI acceptance criteria considering instrument uncertainties, pump recirculation flow, and allowances to the Steam Generator safety valve setpoint.
- All month ly and outage test procedures for all AFW pumps have been updated with appropriate acceptance criteria.
- The revised acceptance criteria have always been met by the AFW pumps except when one pump was declared inoperable for repairs. Three original pumps from 1971 are stillin service and operable. Current rronthly test procedures have some margin to the lower acceptance value. No degradation over previous cycle.
- AFW Pump current status:
Unit 2 system benchmarking preliminary shows the calculation modelis accurate and gives conservative results. Unit 1 system benchmarking data will be collected during 11/97 refueling outage.
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e EXTENT OF TEST CONTROL ISSUE
- AFW Pump trend is minor since plant preoperational tests (i.e. 21 AFWP). No degradation at 220 gpm on any pump since previous outage testing.
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- AFW margin recovery efforts.
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More precise uncertainty calculations 2 GPM Recirculation Flow instrumentation 1 GPM
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Replace MD AFWP MOVs to reduce dp 2 GPM
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Reanalyze Design Basis Flow 40 GPM
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Increase Driver Speed 20 GPM-
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Replace pump intemals (Not Degraded) 0 GPlVi
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Reduce MS Safety Tolerance from 3% to 2% 3 GPM
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Reduce recirculation flow 10 GPM :
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New Pumps and/or Drivers ?
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TEST CONTROL e EXTENT OF TEST CONTROL ISSUES
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NSP initiatives to review other ASME Section XI areas:
Began a review in 1996 of all valve stroke maximum times. Review completed July 1997. All valves operable. All AFW system valve stroke maximum times unchanged. A few changes made to other valve stroke maximum times. Test procedures will be revised before next use.
Check valve full flow values were captured in H10.1 during PI 3rd 10 year.
QA audit of Safety injection (SI) system conducted at same time as SSOPl.
Generally found design criteria properly implemented in pump and valve l test procedures. Found Si pump test acceptance criteria not conservative in comparison with ASME Section XI (plus or minus) 10% at high flow end of pump curve. Si pumps meet design criteria. Si pumps have met the more restrictive ASME criteria. Test procedures will be revised before next use. .
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TEST CONTROL UNTIMELY CORRECTIVE ACTION
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e UNTIMELY CORRECTIVE ACTIONS ON AFW, RH, CS
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- An Operational Experience Assessment corrective action in 1991 recommended review of Auxiliary Feedwater (AFW), Residual Heat Removal (RH) and Containment Spray (CS), for proper acceptance criteria. The corrective action was untimely.
- AFW Pump criteria reviewed and revised during SSOPl.
- RH Pump criteria found acceptable. Currently at least 5.6% margin between ASME Section XI and design criteria considering instrument uncertainties.
- CS Pump criteria found acceptable. Currently at least 9.2% margin between ASME Section XI and design criteria considering instrument
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uncertainties.
- S1 Pump test acceptance criteria modified per QA audit.
/2. - LAR on RH will be submitted to address actual design flow requirements l
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TEST CONTROL UNTIMELY CORRECTIVE ACTION e
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Review of other ASME Section XI Pumps
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Cooling Water pump criteria previously found conservative to OSAR.
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Component Cooling (CC) & Control Room Chilled Water (ZH) test instrument uncertainties will be calculated irf 12h ~
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CC & ZH Pump capacities exceed DBA load listed in USAR. Normal operation flows can exceed DBA flow requirements. Extra capacity ensures test acceptance criteria exceeds DBA requirements.
CC ZH DBA Flow (gpm) 2980 283 CC Design Flow (gpm) 4000 ZH Test Flow (gpm) 400 Pump Curves Plot Flows to (gpm) 5500 550 Pump Curve vs Pump Performance 2 psid (2%) 4 psid (4.7%)
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Pumps and valve design criteria will be incorporated into the ASME Section XI program contro!
document M10.1. Future changes will have easy reference to the design criteria.
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Engineering Sur -ort Personnel will receive training on changes to H10.1. I B 1 i
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TEST CONTROL UNTIMELY CORRECTIVE ACTION
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Management review of old OEA recommendations for regulatory significance, priority and schedule. Management communicated expectations for completion.
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NSP QA Finding on OEA recommendations lead to process revision 4/97 providing more management control on due dates.
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Corrective Action Program Status
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Stren_gths .
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Numerous problems self-identified (several generic issues identified)
. Staff empowered to identify problems and initiate corrective actions
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. System engineer! rig ownership results in high degree of system operability and reliability
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Operability and reportability determination documentation improving
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Corrective Action Program Status (cont'd)
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Weaknesses
. Many separate corrective action vehicles
. Lack of management involvement in prioritization I
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. Status of some corrective action vehicles not reported to upper management or other affected personnel I
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Corrective Action Program Status (cont'd)
_ Changes / Actions Completed
. Procedures revised to require additional management involvement in OEA prioritization
. Open OEA recommendations reprioritized
. Sample of open FOls reviewed for proper operability and reportability assessment
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I i Corrective Action Program Status (cont'd)
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Near Term Changes- i i
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. i New condition reporting system to be introduced by September 1997 (consolidates some corrective action vehicles)
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Corrective action process owner designated by September 1997 i
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Employee Observation Report system revised by September 1997
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Resource authorizations approved and vacancies being filled i
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Longer Term Changes
. Institute a management oversight committee by December,1997
. Improve corrective action status reporting functions by December,1997
. Perform industry review of effective corrective action programs by December,190" ,
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. Institute additional corrective action program improvements in 1998
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DESIGN CONTROL e WEAKNESSES IDENTIFIED DURING THE INSPECTION
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Errors in five mechanical / hydraulic calculations.
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All Discrepancies were minor.
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No Operability Concerns.
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No Changes to Original Conclusions.
e CAUSE - FAILURE TO FOLLOW SITE PROCESS l
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Training. !
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Communication of Management Expectations Regarding Rigorous .
Approach to Design Calculations.
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Calculations Revised.
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Heightened Awareness of Engineering Staff. !
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Added management staff.
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Training. ,
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Sampling of Calculations (focus on mechanical / hydraulic design calculations).
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Review of Calculations which support USAR.
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CONCLUSIONS
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- ALL SYSTEMS REMAIN OPERABLE o WEAKNESSES ACKNOWLEDGED IN THE FOLLOWING AREAS:
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Test Control.
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Conective Action Program.
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USAR Update.
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Design Control Calculation rigor.
e IMMEDIATE SHORT TERM CORRECTIVE ACTIONS HAVE BEEN TAKEN TO ADDRESS WEAKNESSES IDENTIFIED
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Revised acceptance criteria and procedures for AFW.
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Review of other ASME Section XI pumps for appropriate acceptance criteria.
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Completion of va!ve stroke maximum times.
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Review and prioritization of old OEA Recommendations.
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Completed Safety Evaluation to update USAR AFW Flow requirements.
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Revision to calcula; ions with errors.
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Height?ned Awareness of Engineering Staff on rigor ' wociated with calculations. i a/
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CONCLUSIONS (cont'd)
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o ON-GOING ACTIONS TO PREVENT RECURRENCE
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Unit 1 AFW System benchmark testing.
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AFW System Margin Recovery.
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Upgrade to n3ME Section XI Program Document H10.1.
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New Condition Reporting System to be introduced.
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Increased management oversight of Corrective Action Program.
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USAR Upgrade Project. l
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Engineering Resources being added. i
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Communicate management expectations via training. l
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Hold engineers accountable to expectations.
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USAR UPGRADE PROJECT
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e USAR UPGRADE PROJECT INITIATED
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Project Manager assigned.
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Multi-Discipline Review Team formed.
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Project and implementation Plans written.
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Management Oversight Committee to be formed.
I e OBJECTIVES OF USAR UPGRADE PROJECT
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Assure USAR is current, accurate and reflects the current ;
cesign bases and operation of the Plant. ;
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Administrative processes are in place to properly maintain i its status as a Living Document.
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Assure that plant personnel are fully aware of the !
significance of the USAR.
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Assure integration with the improved Technical Specifications Project.
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'I USAR UPGRADE PROJECT
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SCOPE .
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THREE SEPARATE PHASES .
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USAR DOCUMENT REVIEW
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Thorough review of the USAR on both a chapter and l
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Discrepancies identified will be tracked.
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Discrepancies affecting operability or reportability will be dispositioned as they are identified.
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Other discrepancies will be prioritized for later resolution. '
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USAR UPGRADE PROJECT SCOPE (cont'd)
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USAR VERIFICATION
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USAR will be compared to the physical plant and current plant procedures.
Ensure physical plant and USAR description are consistent.
Plant is operated in accordance with the USAR.
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Discrepancies affecting operability or reportability will be dispositioned as they are identified.
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This phase will involve a significant portion of the plant Staff in the review (i.e. system engineers, operations, radiation prclection, etc.).
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USAR UPGRADE PROJECT SCOPE (cont'd) I e PHASJJ -
USAR RECONSTITUTION
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Not specifically defined, but will concentrate on the area of calculations and analyses.
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For calculations that are retrievable assumptions and inputs will be verified and referenced.
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For calculations that are not found, new calculations will be performad, as necessary, to support design bases.
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USAR UPGRADE PROJECT APPROACH AND METHODOLOGY
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e COMPLETION OF USAR UPGRADE CHECKLIST FOR DOCUMENTS REVIEWED
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Scope Document will be written summarizing the revisions that are made to cach section of the USAR.
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Documents to be reviewed include:
AEC SER - FSAR Comparison FSAR / USAR - Comparison USAR Revisions - correct and complete information NUREG 0800 - guidance for review of USAR i R.G.1.70 - guidance for format and content of USAR FSAR Appendix E - AEC Questions and Answers USAR-Technical Accuracy Pending USAR Revisions
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USAR UPGRADE PROJECT SCHEDULE
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USAR DOCUMENT REVIEW
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Currently in Progress with Target completion of 6/98.
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USAR VI RIFICATION
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Target completion of 12/98.
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USAR RECONSTITUTION
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Target completion of 12/01.
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Q ENCLOSURE NRC HANDOUT
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The apparent violations descuss9d in the predecisional enforcement conference are subject to further review and are subject to change prior to any resulting enforcement action.
PRAIRIE ISLAND SOPlINSPECTION ENFORCEMENT, EA 97 290:
1. 10 CFR Part 50, Appendix B, Criterion XI, Test Control, requires,in part, that a test program shall be established to assure that all testing required to demonstrate that systems will perform satisfactorily in service is identified and
]- performed in accordance with written test procedures which incorporate the requirements and acceptance limits contained in applicable design documents.
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I Contrary to the above, since at least February 19,1992 until April 1997, the Unit 1 and Unit 2 monthly and refueling outage surveillance tests procedures of the Auxiliary Feedwater Pumps did not incorporate the requirements and acceptance limits contained in applicable design documents such as the USAR.
Consequently these test procedures would have permitted an AFW pump to degrade below the minimum performance requirements that would have been required during a design basis accident.
2. 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, requires,in part, that conditions adverse to quality are promptly identified and corrected; and in the case of significant conditions adverse to quality, the cause of the condition 1 shall be documented, appropriately reported to levels of management, and corrective action taken to preclude repetition.
Contrary to the above, as of May 16,1997, for a significant condition adverse to quality identified in 1991, corrective action to preclude repetition did not occur.
Specifically, in 1991, the Safety injection Pump surveillance test acceptance criteria was identified to be inadequate. However, the 1991 proposed corrective actions which included a review of other pumps' acceptance criteria were inadequate to prevent repetition as demonstrated by the Auxiliary Feedwa*er Pump surveillance tests acceptance criteria being inadequate until April 1997.
3, 10 CFR 50.71(e) and 10 CFR 50.9(a) require that each licensee periodically update the final safety analysis report (FSAR) to assure that the information included in the FSAR contains the latest material developed and assure that information is complete and accurate in all material aspects.10CFR50.71(M(4) requires, in part, that revisions be filed such that the intervals between successive updates to the USAR do not exceed 24 months. It further states that the revisions must reflect all changes up to a maximum of 6 months prior to the date of filing.
- 10 CFR 50.59, " Changes, Tests and Experiments," permits the licensee, in part, to make changes to the facility as described in the safety analysis report without prior
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Commission approval provided the change does not involve an unreviewed safety question. It requires, in part, that the licenst;e maintain records of <:hanges in the facility and that these records include a written safety evaluation which provides the bases for the determination that the change does not involve an unreviewed safety question.
, 10 CFR Part 50, Appendix B, Criterion XVI, " Corrective Action," requires, in part, that measures be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencFes, deviations, defective n,aterial and equipment,
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The apparent violations discussed in the predecisional enforcement conference are subject to further review and are subject to change prior to any resulting enforcement action.
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l0 * Th3 apparent viot:tions discussed in tha prIdicitiond enforcement conference tre subject to further review and tra subj!ct to changs prior to any resultin0 enforc;; ment action.
I I and nonconformances are promptly identified and corrected.
10 CFR 50.73(2)(ii)(B) requires, in part, that the licensee report any event or condition that resulted in the nuclear power plant being in a condition that was outside the design basis of the plant.
The applicable section of the FSAR, now titled the Updated Safety Analysis Report (USAR), is Section 11.9.3, " Performance Analysis (Condensate, Feedwater, and Auxiliary Feedwater Systems]"
a. Contrary to the above, as of May 16,1997, lnformation provided to the Commission was inaccurate in that it li:;ted 400 gpm vs. 200 gpm available and had not been corrected, in accordance with S0.71(e), although two opportunities to correct it were available. The inaccuracy was materialin that in December 1992, the licensee identified that the 400 gpm flow rate specified in USAR Section 11.9.3 under the Main Feedwater Line Rupture accident scenario disagreed with actual Auxiliary Feedwater pump performance capabilities, in June 1993, the licensee determined that the pumps were operable and the USAR value was incorrect. Since June 1993, the licensee had two opportunities to update the USAR under 10 CFR 50.71(c) in December 1993 and again in December 1995, yet no action was taken to address this issue.
b. Contrary to the above, as of May 16,1997, the licensee had not taken prompt corrective at. tion to the above described significant conditions adverse to quality. Specifically, in December 1992, during the licensee's design basis reconstittttion program, licensee engineers identified that the auxiliary feedwater pumps could not meet the 400 gpm flow rate specified in USAR Section 11.9.3 under the Main Feedwater Pipe Rupture accident scanario. In 1995, the licensee cietermined tha* the Main Feedwater Line Rupture Analysis used the 400 gpm value. Although a preliminary revision to the analysis was performed by May 16,1997, the analysis had not been reviewed and approved, approximately four and a half years after the issue was discovered, c. Contrary to the above, as of May 16,1997, the licensee had not reported that the plant was out:,ide its design basis and had nc,t performed a safety evaluation to make permanent this change to the facility as described in the USAR and to verify that no unreviewed safety question existed. Specifically, in June 1995, the licensee identifieo that the Main Feedwater Line Rupture
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Analysis used an Auxiliary Feedwater flow rate of 400 gallons per minute (gpm) to the steem generators while the Auxiliary Feedwater pumps were capable of only supplying slightly over 200 gpm. This was not reported.
Additionally, a safety evaluation was not performed to justify accepting the 200 gpm as a permanent change to the plant as described h the USAR and design basis.
The apparent violations discussed in ine predecisional enforcement conference are subject to fur'her review and are subject to change prior to any resulting enforcement action.
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