IR 05000220/1998014
| ML20155B918 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 10/26/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20155B914 | List: |
| References | |
| 50-220-98-14, 50-410-98-14, NUDOCS 9810300340 | |
| Download: ML20155B918 (29) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
Docket / Report Nos.:
50-220/98-14 50-410/98 14 License Nos.:
DPR 63
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NPF-69 Licensee:
Niagara Mohawk Power Corporation P. O. Box 63 Lycoming, NY 13093 Facility:
Nine Mile Point, Units 1 and 2 Location:
Scriba, New York Dates:
August 16 - September 26,1998 Inspectors:
B. S. Norris, Senior Resident inspector R. A. Fernandes, Resident inspector, FitzPatrick G. K. Hunegs, Senior Resident Inspector, FitzPatrick R. A. Skokowski, Resident Inspector 9810300340 981026
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PDR ADOCK 05000220
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j EXECUTIVE SUMMARY l
l Nine Mile Point Units 1 and 2 50-220/98-14 & 50 410/98-14 August 16 September 26,1998 This NRC inspection report includes reviews of licensee activities in the functional areas of operations, engineering, maintenance, and plant support. The report covers a six-week period of inspections and reviews by the Nine Mile Point and FitzPatrick resident staffs.
OPERATIONS l
In general, the conduct of operations was professional and safety-conscious. During the l
period, the inspectors noted improved attention-to-detail on the part of Unit 1 operators, l
especially the licensed control room operators, in the areas of shift briefings, routine j
communications, and the use of procedures.
l MAINTENANCE / SURVEILLANCE
l NMPC effectively modified the Unit 1 average power range monitors to account for thermal hydraulic instabilities, as required by NRC Generic Letter 94-02. The inspectors noted that l
the modification was completed, as designed, the work package and safety evaluation
were thorough, and the training provided to the control room operators was acceptable.
l During a Unit 2 surveillance test of the Division I standby liquid control system (SLS),
l operators discovered that the Division li pump suction manualisolation valve was locked closed vice locked open. This rer'ted in both divisions of SLS being declared inoperable.
This issue remains open pending inspector review of NMPC's completed root cause l
analysis and determination of corrective actiom> to prevent recurrence. (eel 50-410/98-15-01)
During planned maintenance on the Unit 1 containment spray system, NMPC discovered l
that primary containment integrity had been breached due to an inadequate boundary valve markup. Subsequent review by NMPC identified that in 1994 a similar condition existed, but was also not recognized as a breach of primary containment integrity. In both cases, NMPC's f ailure to maintain primary containment integrity resulted in leakage in excess of l
that allowed by the Unit 1 Technical Specifications, Section 3.3.3.a. These licensee identified and corrected Technical Specification non-compliances are being treated as a non-cited violation. (NCV 50-220/98 14-02)
l ENGINEERING NMPC determined that several Unit 2 pipe welds had not been examined within the appropriate time interval between the first and third refueling outages, as required by Generic Letter 88-01 and the Technical Specifications. This licensee identified and
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corrected violation was of minor significance and not subject to formal enforcement action.
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Executive Summary (cont'd)
Since initial plant startup, the Unit 2 offgas pre-treatment radiation monitors had been set non-conservatively because the associated procedure improperly reduced the conversion factor to eliminate the effect of short-lived isotopes. Although the offgas system would not have isolated at the Technical Specification value, several related alarms would have provided the operators sufficient warning and allowed for timely operator action to effectively mitigate the consequence of high activity in the offgas system. Upon identification, NMPC took prompt and appropriate corrective actions. This licensee identified and corrected offgas radiation monitor Technical Specification non-compliance is being treated as a non-cited vjolation. (NCV 50-220/9814-03)
PLANT SUPPORT in g'eneral, the performance in the area of plant support was professional and safety
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conscious.
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TABLE OF CONTENTS i
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E X EC UTIV E S U M M A R Y.............................................. ii
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TA B L E O F C O NT E N T S.............................................. iv S U M M ARY O F ACTIVITIES........................................... 1 Niagara Mohawk Power Corporation (NMPC) Activities.................. 1 Nuclear Regulatory Commission (NRC) Staff Activities................... 1 1 OPERATIONS
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Conduct of Operations.................................... 1 01.1 G e n eral Com m e nts.................................. 1
Miscellaneous Operations issues............................. 2 08.1 (Closed) URI 50-220 & 50-410/97-04-06: SORC Review of TS Viol a tio n s........................................ 2 08.2 (Closed) VIO 50-410/97-02-01: Failure to implement Unit 2 Control Room Deficiency Program...................... 2 11. M A I N T E N A N C E................................................. 3 M1 Conduct of Maintenance.................................. 3 M 1.1 G e neral Com m ents........,......................... 3 M1.2 Unit 1 - Installation of the Thermal Hydraulic Stability Modification
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per GL 94-02
....................................4 M1.3 Unit 2 Standby Liquid Control System inoperable Due to a Valve Inadvertently Locked Closed........................... 4 M8 Miscellaneous Maintenance issues............................ 5 M8.1 (Closed) eel 50-220/98-09-01: Inadequate Markup Resulted in a Breach of the Unit 1 Primary Containment Integrity........... 5 M8.2 (Closed) LER 50-220/98-15: Breach of Primary Containment Due to Personnel Error
..................................6 M8.3 (Closed) LER 50-220/98-17: Breach of Primary Containment Due to Personnel Error in 1994
............................6 Ill. E N G I N E E R I N G.................................................. 6 El Conduct of Engineering...,................................ 6 E1.1 General Comments (375 51 )........................... 6 E8 Miscellaneous Engineering issues............................. 6 E8.1 (Closed) LER 50-410/98-21: Missed inservice Inspections Required by TS Caused by inadequate Change Management..... 6 E8.2 (Closed) VIO 50-410/97-03-09: Failure to Install Eight Hour Battery-Pack Emergency Lighting in the Vicinity of Appendix R Remote Shutdown Equipment.......................... 8 E8.3 (Closed) LER 50-410/98-22: Radioactive Gaseous Effluent
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Monitoring Instrumentation Set Non-Conservative............ 8
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Table of Contents (cont'd)
I V. PLA NT S U PPO RT..............................................
F2 Status of Fire Protection Facilities and Equipment................
F2.1 Off-site Fire Department Causes Automatic Start of Fire Pumps.
FS Miscellaneous Fire Protection Issues
.........................11 F8.1 (Closed) URI 50 410/97-11-08: Breach Permit Greater than 3 Ye a r s O l d......................................
1 1 V. M AN AG EM ENT MEETING S........................................
1 1 X1 Exit M e eting Sum m ai y...................................
1 1 X3 Management Meeting Summary.............................
X3.1 NRC/NMPC Meeting Related to Extension Request for Inspection of the Unit 1 Core Shroud, Followed by Public Question / Answer Pe rio d..........................................
1 2 ATTACHMENTS Attachment 1
- Partial List of NMPC Persons Contacted
- Inspection Procedures Used-Items Opened, Closed, and Updated
- List of Acronyms Used
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Attachment 2
- Handouts from September 24,199E, Meeting with NMPC Concerning the Unit 1 Core Shroud i
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REPORT DETAILS Nine Mile Point Units 1 and 2 50-220/98-14 & 50-410/98-14
August 16 - September 26,1998
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SUMMARY OF ACTIVITIES
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Niagara Mohawk Power Corporation (NMPC) Activities l
Both units maintained essentially full power throughout the inspection period.
Nuclear Regulatory Commission (NRC) Staff Activities Inspection Activities The NRC inspectors conducted inspection activities during normal, backshif t, and deep backshift hours.
Uodated Final Safety Analysis Report Reviews While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the Updated Final Safety Analysis Report (UFSAR). The inspectors verified that the UFSAR descriptions were consistent with the observed plant practices, procedures, and/or parameters.
1. OPERATIONS
Conduct of Operations 01.1 General Comments (71707)'
Using NRC Inspection Procedure 71707,the resident inspectors conducted frequent reviews of ongoing plant operations. The reviews included tours of accessible areas of both units, verification of engineered safeguards features (ESF) system operability, verification of adequate control room and shift staffing, verification that the units were operated in conformance with Technical Specifications (TS), and verification that logs and records accurately identified equipment status or deficiencies. In general, the conduct of operations was professional and safety-conscious. During the period, the inspectors noted improved attention-to-detail on the part of Unit 1 operators, especially the licensed control room operators, in the areas of shift briefings, routine communications, and the use of procedures.
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' Topical headings such as o1, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline.
Individuat reports are not expected to address all outline topics. The NRC inspection manual procedare or temporary instruction I
that was used as inspection guidance is listed for each applicable report sectio __ ----.- -
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08 Miscellaneous Operations issues 08.1 (Closed) URI 50-220 & 50-410/97-04-06: SORC Review of TS Violations (92901)
In June 1997, the NRC identified that the Unit 1 and Unit 2 Station Operating Review Committees (SORCs) were not reviewing all TS violations, as required by the TS 6.5.1.6. At that time, the inspectors questioned whether this included all procedural non-compliances which were cited as violations of TS 6.8.1. This was left as an unresolved item pending further review by the NRC.
This issue was discussed between the regional staff and the NRC Headquarters Technical Specification Branch, Quality Assurance (QA) Branch, and the Office of Enforcement staffs. The NRC staff agreed that a literal reading of the TS would imply that all procedure violations needed SORC review. However, it was concluded that this was not consistent with the general intent of TS 6.5.1.6. The requirements of the on-site review committee are listed in American National Standards Institute (ANSI) N18.7, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants," which is endorsed by the NRC in Regulatory Guide (RG) 1.33, " Quality Assurance Program Requirements (Operation)." ANSI N18.7 states that on-site review committees should review violations that have safety significance.
Based upon the inspectors' broad review of the deviation / event report (DER)
database and detailed examination of selected DERs, no significant TS violations were identified by the inspectors that were not reviewed by the SORCs.
Consequently, there was no violation of NRC requirements. This unresolved item is closed.
08.2 LQlosed) VIO 50-410/97-02-01: Failure to Imolement Unit 2 Control Room Deficiency Proaram (92901)
In May 1997, the inspectors identified several noncompliances with the Nine Mile Point Unit 2 (Unit 2) control room deficiency program, as described in Procedure
N2-ODP-OPS-0001," Conduct of Operations." The discrepancies were classified as violations of TS 6.8.1, regarding procedure adherence. NMPC's response, dated May 16,1997, provided the root cause and corrective actions for this violation.
The inspectors reviewed this letter, conducted an on-site review of the current control room deficiency program, and concluded that the root cause determination and corrective actions were appropriate. The inspectors verificJ that the work control computer software was revised to prevent personnel outside of the operations department from changing the tracking code that identifies the deficient condition as a control room deficiency. In addition, the inspectors reviewed the current control room deficiencies and found the list to be accurate and complete.
This violation is closed, s
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3 11. MAINTENANCE 2 i
j M1 Conduct of Maintenance
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M 1.1 General Comments (61726,62707)
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j Using NRC Inspection Procedures 61726 and 62707, the resident inspectors i
periodically observed various maintenance activities and surveillance tests. As part
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of the observations, the inspectors evaluated the activities with respect to the i
requirements of the Maintenance Rule, as detailed in 10CFR50.65. In general,
maintenance and surveillance activities were conducted professionally, with the
work orders (WOs) and necessary procedures in use at the work site, and with the appropriate focus on safety. Specific activities and noteworthy observations are
j detailed in the inspection report. The inspectors reviewed procedures and observed all or portions of the following maintenance / surveillance activities:
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l WO 95-4157-03 Thermal Hydraulic Stability Modification - #11 APRM i
WO 98-6269 Repair of Electrical Penetration RW-67 N1-CSP-A326 OGESMS Stack Detector #112-07 Calibration -
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N1 IPM-081-005 Core Spray System Flow Instruments N1-ISP-092-321 APRM #11 Instrument Channel Calibration / Test
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N1-PM-Q11 Primary Containment Manual Valve Check i
N1 REP-8 Core Thermal Power N1 RSP 6Q Control Room Ventilation Radiation Monitor Instrument j
Channel Test j
N1-RSP-13Q Stack Radiation Monitor Quarterly Calibration Check &
Channel Test
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N1-ST-Q6A Containment Spray Loop #111 Operability Test i
N1-ST-Q15 Condensate Transfer System Operability Test N2-OSP-EGS-M@002 Diesel Generator & Diesel Air Start Valve Operability i
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Test, Division lil N2-OSP-SLS-QOO1 Standby Liquid Control Pump, Check Valve, Relief Valve Operability & 40 Month Functional Test N2-OSP-SLS-QOO2 Standby Liquid Control Valve Operability Test N2 RSP-RMS-R102 Channel Calibration Test of the Main Control Room Area Radiation Monitor
- N2-RSP-RMS-R111 Channel Calibration Test of the Drywell Atmosphere Offline Gas & Particulate Process Radiation Monitors
surveillance activities are included under " Maintenance." For example, a section involving surveillance observations might be included as a separate sub-topic under M1, " Conduct of Maintenance."
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4 M1.2 Unit 1 - Installation of the Thermal Hydraulic Stability Modification ner GL 94-02 l
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Inspection Scone (61726,62707)
The inspectors reviewed the modification to the Nine Mile Point Unit 1 (Unit 1)
average power range monitors (APRMs) to account for thermal-hydraulic l
instabilities. The review included observations of some in-field work and the pre-
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evolution brief, and review of the safety evaluation, work package, post-maintenance acceptance work, and the operations training material, b.
Observations and Findings l
During the inspection period, NMPC implemented a modification to the Unit 1 APRM circuitry to incorporate the requirements of NRC Generic Letter (GL) 94-02,"Long-
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Term Solutions and Upgrade of Interim Operating Recommendations for Thermal-Hydraulic Instabilities in Boiling Water Reactors." NMPC treated the modification as a special evolution, in accordance with NMPC Procedure GAP-SAT-03, " Control of
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l Special Evolutions," due to the potential for a significant plant transient (e.g.,
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reactor scram). Special evolutions require direct upper management involvement.
Accordingly, the Unit 1 Technical Support Manager was designated the Senior Manager for this modification.
The inspectors observed one of the pre-evolution briefs by the Senior Manager, portions of the actual modification, and portions of the post-modification test (PMT).
The pre-evolution briefing was thorough and detailed. The modification work and the PMT were appropriately detailed in the work package and completed, as
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designed. The oversight by instrumentation and Controls (l&C) supervision and the system engineer was good. The inspectors reviewed the safety evaluation and the operations training material and found them to be consistent and complete.
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Conclusion NMPC effectively modified the Unit 1 average power range monitors to account for thermal hydraulic instabilities, as required by NRC Generic Letter 94-02. The inspectors verified that the modification was properly completed, as designed, the work package and safety evaluation were thorough, and the training provided to the control room operators was acceptable.
M1.3 Unit 2 Standbv Liouid Control System Inocerable Due to_a Valve Inadvortentiv Locked Closed (61726,71707)
On September 11,1998, during a Unit 2 surveillance of the Division I standby liquid control system (SLS), operators discovered that the Division ll pump suction manual isolation valve (2SLS*V46) was locked closed vice locked open. With Division I already inoperable because of the surveillance, and with 2SLS*V46 locked closed, both divisions of SLS were inoperable. The Station Shift Supervisor (SSS) directed
that 2SLS*V46 be locked open and independently verified locked open.
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Subsequently, NMPC wrote DER 2-98-2730 to document the evant and to initiate a formal root cause analysis.
Additionst corrective action included a valve line up verification of select safety i
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systems at both units; no other mispositionings were identified nor was there any
l indication of tampering, initial review by NMPC indicated that the valve was last l
manipulated in August 1998 during a routine surveillance, and that the valve was l
locked closed since that surveillance. This issue remains open pending completion of the DER disposition, issuance of the associated Licensee Event Report (LER), and
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l review by the inspectors of the root cause and corrective actions.
l (eel 50-410/98-14-01)
i M8 Miscellaneous Maintenance issues l
l M8.1 ] Closed) eel 50 220/98-09-01: inadeouate Markuo Resulted in a Breach of the Unit 1 Primary Containment Intenrity (90712. 92902)
On August 4,1998, during planned maintenance on the Unit 1 containment spray system, NMPC discovered that primary containment integrity had been breached due to an inadequate boundary valve markup. This event was initially discussed in l
NRC IR 98-09, Section M1.2, and a tracking item assigned pending issuance of the l
associated LER and inspector review.
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I As documented in LER 98-15, NMPC's investigation determined that the root cause was complacency on the part of the licensed operators who developed, reviewed, and approved the markup for isolation of the containment spray system. The j
operators relied on experience and other operators, rather than reviewing the l
system drawings and the plant impact statements in the work packages. Corrective
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actions included a briefing for each shift as to the requirements and expectations for markups and enhanced training to include lessons learned from this event. NMPC evaluated the increased containment leakage and noted that the resultant dose to
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l the control room from a design basis accident (DBA) would remain within the limits of 10CFR100. During their review, NMPC identified that in 1994, a similar condition existed, but was likewise not recognized as a breach of primary containment integrity. The NMPC analyses for each event concluded that neither posed a significant safety threat to the public or plant personnel. eel 50-220/98-09-01 is closed.
The failures to maintain primary containment integrity resulted in leakage in excess of that allowed by the Unit 1 Technical Specifications, Section 3.3.3.a. These licensee identified and corrected violations are being treated as a Non-Cited Violation (NCV), consistent with Section Vll.B.1 of the NRC Enforcement Policy.
(NCV 50 220/98-14-02)
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M8.2 LQlosed) LER 50-220/98-15: Breach of Primary Containment Due to Personnel Error (90712)
The inspectors previously reviewed the technicalissues associated with this event as documented in NRC IR 98-09. Additional inspector observations are discussed in Section M8.1 of this inspection report. The in:cet. tors completed an in-office review cf the LER and considered the root causs cnd corrective actions to be reasonable. The description and analysis of the event, as contained in the LER, wera cons; stent with the inspectors' understanding of the event. The LER is closed.
M8.3 (Closed) LER 50-220/98-17: Breach of Primary Containment Due to Personnel Error in 1994 (90712)
The technical issues associated with this LER were described in Section M8.1 of this inspection report. The inspectors completed an in-office review of the LER and considered the root cause and corrective actions to be reasonable. The description and analysis of the event, as contained in the LER, were consistent with the inspectors' understanding of the event. The LER is closed.
Ill. ENGINEERING E1 Conduct of Engineering E1.1 General Comments (37551)
Using NRC Inspection Procedure 37551, the inspectors reviewej design and system engineering activities and the support by the engineering organ:zations to plant activities throughout the inspection period.
E8 Miscellaneous Engineering issues
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E8.1 (Closed) LER 50-410/98-21: Missed Inservice inspections Reauired by TS Caused i
by inadeauate Chanae Manaaement
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Inspection Scope (90712,92700)
in June 1998, NMPC determined that some Unit 2 inservice inspection (ISI) pipe weld examinations had not been performed, as required. The inspectors discussed
the missed ISI examinations with NMPC management, including the engineering manager responsible for ISI, and the Quality Assurance Manager. In addition, the i
inspectors performed an on-site review of the LER, associated DER for both units, portions of the Unit 2 ISI Program, and the appropriate sections of the ASME
[American Society of Mechanical Engineers] Code.
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Observations and Findinas in June 1998, NMPC determined that ISI examinations had not been completed as required by Unit 2 TC Surveillance Requirement (TSSR) 4.0.5.f, which states that the ISI Program for piping shall be performed in accordance with the schedule requirement of Generic L atter (GL) 88-01. Specifically, GL 88-01 requires all designated Category D p.pe welds shall be inspected every two refueling outages.
NMPC identified that the examinations of some reactor vessel Category D welds were not conducted during the 18-month operating cycles between Refueling Outage One (RFO1) and RF03. This was initially identified by NMPC in 1994 during the preparation of the post-RFO3 ISI Summary Report. At that time, NMPC resolved this discrepancy using the rationale that the new 24-month operating cycle time interval could be applied, and thus they were complying with the current Unit 2 i
TSs. During the recent refueling outage, RFO6, completed June 1998, NMPC identified that this 1994 assumption was incorrect.
All of the Category D weld examinations were completed during RFO4 (1995).
Since then,50 percent of the welds were examined during RFOS and the other 50 percent were inspected during RF06. NMPC stated in the LER that all of the weld examination results were acceptable. The failure to complete the required examinations between RFO1 and RFO3 is contrary to TSSR 4.0.5.f. However, this failure constitutes a violation of minor significance and is not subject to formal enforcement action.
The inspectors verified that the LER was completed in accordance with the I
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requirements of 10CFR50.73. Specifically, the description ann analysis of the event, as contained in the LER, were consistent with the inspectors' understanding of the event, based on the inspection observations discussed above. The root cause and corrective and preventive actions described in the LER were reasonable and appropriate. This LER is closed.
During review of the above event, the inspectors noted that there were several DERs addressing ASME Code required examinations that may not have been completed. Specifically, visual inspections of several piping system structural supports were not performed in the manner prescribed by the ASME Code. The inspectors determined that NMPC's initialinterpretation of the ASME Code was that
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insulation on structural supports could remain installed during visual examinations.
NMPC sent a letter to the ASME Code Committee requesting a formalinterpretation.
The ASME Code Committee ruled that NMPC's interpretation was incorrect, and that the insulation needed to be removed. Subsequently, NMPC issued DERs 198-2593 and 2-98-2594(Units 1 and 2 respectively) to initiate a review of the missed inspections.
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For Unit 1, NMPC determined that the required examinations were completed for the 1" and 2"d periods of the current Ten-Year Interval, and that the remaining examinations were planned for the 3'd period (during RFO15, scheduled for Spring 1999). For Unit 2, NMPC determined that some of the inspections had not been performed prior to the end of the First Ten-Year Interval (April 1998). However, the
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l ASME Code, Section IWA-2430, provides for extending the interval up to 1 year.
l The inspector observed that NMPC planned to exercise this option and complete the
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required examinations within the next 12 months. The inspectors found this approach consistent with the ASME Code.
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Conclusion I
l NMPC determined that several Unit 2 pipe welds had not been examined within the appropriate time interval between the first and third refueling outages, as required
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by Generic Letter 88-01 and the Technical Specifications. This licensee identified l
and corrected violation was of minor significance and not subject to formal enforcement action.
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E8.2 (Closed) VIO 50-410/97-03-09-Failure to install Eiaht-Hour 8atterv-Pack
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Emeraency Liahtina in the Vicinity of Accendix R Remote Shutdown Eouloment l
(92903)
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l In May 1997, the inspectors identified that eight-hour battery pack emergency l
lighting was not provided in the vicinity of the RHR minimum flow valves.
l Emergency lighting is required because local operation of these valves is necessary
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for safe shutdown of the plant, in the event of a control room fire requiring
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evacuation. The inspectors completed an on-site review of the root cause and corrective actions documented in NMPC's August 11,1997, violation reply letter,
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l and determined the actions to be adequate. NMPC installed the required emergency l
lighting and completed a broad review of their Appendix R program. The inspectors
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verified proper installation of the new emergency lighting, in addition, the inspectors reviewed the findings of the licensee's Appendix R program review and l
confirmed that the findings were being appropriately addressed through NMPC's
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corrective action program. This violation is closed.
E8.3 (Closed) LER 50-410/98-22: Radioactive Gaseous Effluent Monitorina Instrumentation Set Non-Conservative a.
Inspection Scope (90712,92700)
On June 23,1998, while Unit 2 was shutdown for RFO6, NMPC personnel l
determined that the Unit 2 offgas pre-treatment radiation monitor setpoint had been l
set non-conservatively since initial plant startup. The inspectors completed an on-i site review of the issues associated with this LER. Particularly, the inspectors assessed the licensee's root cause analysis and corrective actions as described in the LER, including a review of the TS, UFSAR, applicable licensee procedures, and
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Offsite Dose Calculation Manual (ODCM). The inspectors also discussed the issue with the responsible Unit 2 Radiation Protection Supervisor. In addition, the inspectors verified the completion of the LER in accordance with 10CFR50.73.
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Observations and Findinas During a review of the Unit 2 emergency operating procedures, NMPC personnel determined that the offgas pre-treatment radiation monitors had been set non-conservatively since initial plant startup. As required by TS 3.3.7.10, the alarm / trip setpoints for the offgas pre-treatment radiation monitors are determined and adjusted in accordance with the methodology and parameters provided in the ODCM. In 1986, NMPC Chemistry and Radiation Protection personnel developed proceduros to adjust the offgas pre-treatment radiation monitor setpoints. Review l
of this procedure in June 1998 determined that the setpoint conversion factor used by the radiation monitors was improperly reduced to eliminate the effect of short-lived isotopes and background radiation contributions, which resulted in a non-conservative setpoint. Upon identification, NMPC determined an appropriate conversion factor consistent with the methodology provided in the ODCM and reset
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the offgas pre-treatment radiation monitor alarm / trip setpoints, j
The offgas pre-treatment radiation monitors isolate offgas flow to the main stack in i
the event of a high radiation condition. In addition, the monitors provide a control room alarm to alert the operators of higher than normal offgas activity. NMPC analyzed the consequence of the event and determined that although the offgas system would not have isolated at the value required by TS, the alarm would have provided the operators sufficient warning and allowed for timely operator action to reduce a monitored radioactive release, in accordance with the alarm response procedure (ARP). Furthermore, changes in offgas activity would have been detected by the main stack gaseous effluent monitor, and other plant radiation monitors, including the main steam line radiation monitors and particular area radiation monitors. The alarms associated with these monitors would have prompted operators to investigate and take action to reduce a monitored radioactive release. Based upon the alarm indications available from the offgas monitors and other radiation monitors, NMPC concluded that timely operator action would have been taken to effectively mitigate the consequence of high activity in the offgas system.
l The inspectors reviewed plant procedures and discussed the issue with the i
responsible Unit 2 Radiation Protection Supervisor and several Unit 2 senior reactor operators iSROs), and determined that NMPC's conclusion was technically sound.
Nonetheless, the failure to determine and adjust the offgas pre-treatment radiation
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l monitors in accordance with the methodology and parameters provided in the ODCM is a violation of TS 3.3.7.10. This non-repetitive, licensee identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-410/98-14-03)
The inspectors verified thet the LER was completed in accordance with the requirements of 10CFR50.73. Specifically, the description and analysis of the i
event, as contained in tne LER, were consistent with the inspectors' understanding l
of the event. The rcot cause and corrective and preventiva actions as described in the LER were reasonable. This LER is closed.
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Conclusion l
l Since initial plant startup, the Unit 2 offgas pre-treatment radiation monitors had been set non-conservatively because the associated procedure improperly reduced the conversion factor to eliminate the effect of short lived isotopes. Although the offgas system would not have isolated at the Technical Specification value, several related alarms would have provided the operators sufficient warning and allowed for timely operator action to effectively mitigate the consequence of high activity in the
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offgas system. Upon identification, NMPC took prompt and appropriate corrective actions. This licensee identified and corrected offgas radiation monitor Technical Specification non-compliance is being treated as a non-cited violation. (NCV 50-220/98-14-03)
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IV. PLANT SUPPORT Using NRC Inspection Procedure 71750, the resident inspectors routinely monitored l
the performance of activities related to the areas of radiological controls, chemistry,
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emergency preparedness, security, and fire protection. Minor deficiencies were discussed with the appropriate management, significant observations are detailed below.
l F2 Status of Fire Protection Facilities and Equipment l
l F2.1
.Qff-site Fire Department Causes Automatic Start of Fire Pumos (71750)
l On August 22,1998, with the fire suppression water headers cross-connected, the Scriba Fire Department used a fire hydrant located outside of the protected area to l
fill the department's tanker truck. This action resulted in the automatic start of the j
diesel and electric driven fire pumps to maintain normal fire suppression water
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system pressure. A verbal agreement between the local fire department and NMPC allowed use of the hydrant, provided prior permission was received. In this instance, no permission was granted and the fire pump starts were unanticipated.
The inspectors discussed this event with the Nine Mile Fire Department and were j
informed that the planned corrective action was limited to a call to the Scriba Fire l
' Department to remind them of the need to request permission prior to using the hydrant. The inspectors were concerned that NMPC's corrective actions were weak, in that, no positive means were instituted to prevent an individual outside of the protected area from inadvertently impacting site fire equipment located inside the protected area. Further discussions with the Supervisor of the Nine Mile Fire Department resulted in more definitive action (i.e., the hydrant was locked closed and a letter was sent to the Scriba Fire Department informing of the actions and the i
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need for prior permission to operate the hydrant). The inspectors considered these subsequent actions to be appropriate.
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F8 Miscellaneous Fire Protection Issues F8.1 (Closed) URI 50-410/97-11-08: Breach Permit Greater than 3 Years Old (92904)
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In October 1997, the NRC identified that a fire-door in the radioactive-waste building was removed from its hinges. The associated breach permit was dated September 9,1994, and stated that the door was removed to allow hoses to pass through the doorway for a temporary modification. At that time, the inspectors questioned: (1) whether the door being removed for over three years was considered in the fire hazards analysis, and (2) whether the excessive time was in essence a permanent modification without the requisite safety evaluation.
During NMPC's disposition of the associated DER, they determined the cause to be due to a misapplication of design inputs during the installation of a plant
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modification. Specifically, the modification did not include an evaluation of the breached door. A contributing cause was the failure of management to monitor fire protection activities, in that the breach was allowed to remain open for an extended
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period of time. Corrective actions included a design change to re-route the hoses, the door was re-hung, and the breach permit was closed. Also, fire protection personnel reviewed all existing breach permits at both units, and identified others that were greater than 90 days old. The conditions requiring the permit were either corrected, or the breach was re-evaluated for extension, as allowed by the procedure. Finally, the controlling NMPC Procedure (GAP-FPP-03, " Breach Permit")
was revised to include a requirement for a DER to be initiated for any breach permit greater than six months old that was not also part of a temporary modification.
The inspectors performed an on-site review of the associated DER and toured the facility for similar conditions with no additional concerns identified. The inspectors concluded that although the temporary modification and associated breach permit were poorly controlled, no violation of NRC requirements occurred. This unresolved item is closed.
V. MANAGEMENT MEETINGS X1 Exit Meeting Summary At periodic intervals, and at the conclusion of the inspection period, meetings were held with senior station mana0ement to discuss the scope and findings of this inspection. The final exit meeting occurred on October 9,1998. During this meeting, the resident inspector findings were presented. NMPC did not dispute any of the inspectors findings or conclusions. Based on the NRC Region I review of this report, and discussions with NMPC representatives, it was determined that this report does not contain safeguards or proprietary informatio...
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i X3 Management Meeting Summary
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X3.1 NRC/NMPC Meetina Related to Extension Reauest for Inspection of the Unit 1 Core Shroud. Followed by Public Question / Answer Period On September 24,1998, the NRC met with NMPC to discuss their request to extend the inspection interval of the Unit 1 core shroud from 10,400 hot operating hours to 14,500 hot operating hours. This equates to an extension from late November 1998, until the next refueling outage scheduled for May 1999. This meeting was conducted at the Oswego Campus of the State University of New York (SUNY)-Oswego and was open for public observation. Following the meeting between the NRC and NMPC, a second meeting was held to receive public comments regarding NMPC's request. The handouts from the NRC - NMPC meeting is included as Attachment 2 to this report.
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ATTACHMENT 1
l PARTIAL LIST OF PERSONS CONTACTED Niaaara Mohawk Power Corooration l
R. Abbott Vice President, Nuclear Engineering
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D. Barcomb Manager, Unit 2 Radiation Protection
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D. Bosnic Manager, Unit 2 Operations J. Burton Manager, Training
H. Christensen Manager, Security
J.Conway Vice President, Nuclear Generation W. Davey Manager, Unit 1 Work Control (acting)
R. Dean Manager, Unit 2 Engineering i
S. Doty Manager, Unit 1 Maintenance G. Helker Manager, Unit 2 Work Control A. Julka Director, ISEG C. Merritt Manager, Unit 2 Chemistry P. Mezzafero Manager, Unit 1 Technical Support N. Paleologos Plant Manager, Unit 2 L. Pisano Manager, Unit 2 Maintenance N. Rademacher Manager, Quality Assurance l
R. Randall Manager, Unit 1 Engineering V. Schuman Manager, Unit 1 Radiation Protection
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l C. Senska Manager, Unit 1 Chemistry R. Smith Plant Manager, Unit 1 C. Terry Vice President, Nuclear Safety Assessment & Support D. Topley Manager, Unit 1 Operatians K. Ward Manager, Unit 2 Techn: cal Support D. Wolniak Manager, Licensing INSPECTION PROCEDURES USED IP 37551 On-Site Engineering IP 61726 Surveillance Observations IP 62707 Maintenance Observations IP 71707 Plant Operations IP 71750 Plant Support IP 90712 In-Office Review of Written Reports of Non-Routine Events at Power Reactor Facilities IP 92700 Onsite Follow-up of Written Reports of Non-Routine Events at Power Reactor Facilities IP 92901 Follow-up - Operations l
1P 92902 Follow-up - Maintenance l
IP 92903 Follow-up - Engineering IP 92904 Follow-up - Plant Support A1 l
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Attachment 1
ITEMS OPENED, CLOSED, AND UPDATED OPENED 50-410/98-14-01 eel Standby liquid controlinoperable due to a valve inadvertently locked closed 50-220/98-14-02 NCV Failure to maintain primary containment integrity resulted in leakage in excess of TS allowable 50-410/98-14-03 NCV Failure to adjust offgas rad monitors per the ODCM CLOSED 50-220/98-09-01 eel Inadequate markup resulted in a breach of primary containment integrity 50-220 &
URI SORC reviews of TS violations 50-410/97-04-06 50-410/97-02-01 VIO Failure to implement the Unit 2 CR deficiency program 50-220/98-15 LER Breach of primary containment due to personnel error 50-220/98-17 LER Breach of primary containment due to personnel error in 1994 50-410/97-03-09 VIO Failure to install eight-hour battery-pack emergency lighting in the vicinity of Appendix R remote shutdown
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equipment
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50-410/98-22 LER Radioactive gaseous effluent monitoring instrumentation set non-conservative 50-220/98-14-02 NCV Failure to maintain primary containment integrity resulted in leakage in excess of TS allowable 50-410/98-14-03 NCV Failure to adjust offgas rad monitors per the ODCM 50-410/97-11-08 URI Breach Permits Greater than 3 Years Old 50-410/98-21 LER Missed inservice inspections required by TS caused by inadequate change management UPDATED none LIST OF ACRONYMS USED l
ASME American Society of Mechanical Engineers l
APRMs Average Power Range Monitors
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ARP Alarm Response Procedure CFR Code of Federal Regulations DBA Design Basis Accident DER Deviation / Event Report l
eel Escalated Enforcement item l
ESF Engineered Safeguards Feature GL Generic Letter IR inspection Report
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Attachment 1 I&C instrumentation and Controls ISI Inservice Inspection LER Licensee Event Report NCV Non Cited Violation l
NMPC Nine Mile Point Corporation NRC Nuclear Regulatory Commission ODCM Offsite Dose Calculation Manual QA Quality Assurance l
l RFO Refueling Outage
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RG Regulatory Guide RHR Residual Heat Removal SLS Standby Liquid Control System
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SORC Station Operating Review Committee SRO Senior Reactor Operator SSS Station Shift Supervisor l
SUNY State University of New York l
TS Technical Specification TSSR Technical Specification Surveillance Requirement
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UFSAR Updated Final Safety Analysis Report
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Unit 1 Nine Mile Point Unit 1 l
Unit 2 Nine Mile Point Unit 2 VIO Violation l
WO Work Order
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i ATTACHMENT 2
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HANDOUTS FROM i
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SEPTEMBER 24,1998
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MEETING WITH NMPC
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CONCERNING THE UNIT 1 CORE SHROUD
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AGENDA
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l September 24,1998 Meeting Regarding Inspection of Core Shroud Vertical Welds at l
Nine Mile Point Nu. clear Station Unit 1 1. NRC SESSION WITH NIAGARA MOHAWK POWER CORPORATION (NMPC)
l 5:00 NRC Opening Remarks Darl Hood Purpose introduction of Participants 5:05 Background Robert Hermann 5:10 NMPC's Review of Request to Richard Abbott Extend Core Shroud et al.
Inspection Interval
Introduction Core Shroud Boat Sample Tests and Evaluations Application of BWRVIP-14 to Unit 1 Core Shroud Weld Cracks Conclusions 6:30 NRC Questions / Comments 7:00 Break 11. NRC SESSION WITH PUBLIC 7:30 NRC Opening Statements Darl Hood 7:35 Questions / Comments from Audience 9:30 NRC Closing Remarks Singh Bajwa l
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Enclosure 1
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NRC ATTENDEES
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Office of Nuclear Reactor Regulation, Rockville, MD:
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l Singh S. Bajwa Director Project Directorate.l-1 l
Darl S. Hood Senior Project Manager
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Project Directorate 1-1 Robert A. Hermann Senior Level Advisor-Materials Science Materials and Chemical Engineering Branch Division of Engineering William H. Koo Senior Materials Engineer i
Materials and Chemical Engineering Branch Division of Engineering l
Ralph Caruso Section Chief l
Reactor Systems Branch Division of Systems Safety and Analysis l
Kerri A. Kavanagh Reactor Systems Engineer l
Reactor Systems Branch j
Division of Systems Safety and Analysis
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Dr. Lambros Lois Senior Reactor Systems Engineer Reactor Systems Branch Division of Engineering l
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Region I, King of Prussia, PA:
Lawrence T. Doerflein Chief, Project Branch 1 l
Division of Reactor Projects
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Barry S. Norris Senior Resident Inspector Nine Mile Point Nuclear Station Neil A. Sheehan Senior Public Affairs Officer Public Affairs Staff NRC Contractor:
Dr. William J. Shack Associate Division Director of the
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Energy Technology Division l
Argonne National Laboratory i
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O MACARA M0tlAWK POWE R CORPORA TION s
y M Ntht At1Lt POINTNUCLLAR STArlCN UNrf I ngCHUA Openmg Renurks
.J H Muelkr introductions.
. R B. Abbott NRC/NMPC Purpose.
. u Adeoti Background.
..C. D. Terry Nine Mile Point Um,t 1 B, suits of E.aiuahon..
..canai R Horn Core Shroud Meeting ut Manamn Results of Structural Margin Anessment.
..C. Inch Corslusion.
. Il B. Abbott M
Meeting Purpose M
Background e The BWRVIP developed industry standardized w Present supplementalinformation applied as shroud inspection, evaluation and repair criteria basis for extending shroud reinspection which were approved by the NRC
- NMP1 shroud metallurgical, fluence, and e Unit 1 shroud horizontal welds preemptively crack growth assessment submitted repaired in 1995 February,1998 e All vertical welds inspected in 1997 consistent
- NMP1 supplemental shroud structural with BWRVIP cntena for repaired shrouds margin analysis submitted April,1998 e Cracks were observed and boat samples removed
- Neutron transport analysis - September, for rnetallurgical evaluation 1998 e Applicability of BWRVIP 14
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M Background M Basis of the Vertical Weld 10,600 Hour Inspection Interval e Aptd 1997, NMPC provided jushfuatum, consistent with e 100% insptuon of all accessble vertre and hansontal welds BWRYlP41 guidelanes, for 10400 hours of har operauon conastent with BWRYlP41 and 8WRVIP47 e May 8.1997 NRC assued an SER allowmg operauen for 10400
, ywww elemens Lawar Elasex Fracture Mechanus (LEPM)
hours pner to remspecuan of ttw vertual welds analysis of V9 and VIO part ihtough wall cracks based on e February 27, 1996, the NMPC subnuttal requested to entend frut.are ioughness of (130 ks:(m)consissent with 8WRVIP41 operauen irarn 104C0 hours to 14300 hours, based upon evaluation gualelmes rnetallurgwal evaluaten and reassessment of crack growth rates e Lanut Load Analysis for VL Vi$,and VM tar welds V9 and V10 e Opraung mwrval was defmed based on CGR of 10 a le' m/hr e Aptd 30,1990. NMPC subavited result.s of supplemerdal e Ne credit ice honzental weld inwgnty structural margm assessment of welds VL V9 and VIO, consistent e Part hgh wdicrzkmg ash ohnens where1/r wuh BWRVIP41 guidance, to further support operauon for 14300 idenaAed uncre.*ed hgament hours e Opraw wittun EPRI wawr chenustry guurknes e C Nne s.1996, the NRC issued an SER on BWRVIP.14 wtuch ts
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d nly apptrable to the NMP) cratkmg
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M Actions Since April 1997 M Basis of the Vertical Weld 14,500 Inspection and Evaluation Hour inspection Interval e NRC approved tre NMPC firute el.nwne fruture enecharucs e Metallurgical and fluence evaluanons justify 14,500 and lun t load analysis of the venn al wetas and etw safety hours based upon lower CCR.
ancierrent of itw vertaal a, eld crackmg
- PLIDGE analvsis CCR cordtrms 2.2 x 10-5ut/hr e NMPC has operated well twlow the EPRI water ctwnustry with sagruficant margin guadelme comnutment (conductivny <.3 S/cm, sulfate <5 ppb,cNoruse <5 ppb)
- Cracking conftrined as ICSCC, consistent with basis of BWRVIP-14
. avg conductmty 0.076.5/cm
- avg sulfaw 2,01 ppb, avg chlande <0.5 ppb
. Analysis satisfies the BWRVIP-14 SER condations e NMPC complewd detailed metallurgical evaluatwns of the
. Fluence wdi remam below 5 m 10a'n,'cm3 verucal weld boat samples e Supplemental structural analysis which satisfies e Addinonal structural margm analysis completed BWRVIP-01 analysis guidelmes justthes greater than e The NRC usued DWRVIP 14 SER wtuch supports lower CCR
! t.500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br /> at the assumed 5 x 105 in/hr CCR I-M Vertical Weld Boat Sample M
NRC SER Crack Growth Evaluations Assessment e
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Role ofIrradiation Effects on Assessment Summary NMP1 Shroud
. Evaknatmas bened en both CE PLIDGE endeland ihr BWEVIP-l4 carvensaan e Ivaneseems car.nlet a5 ihe inequrs.kd aHoct pawnesi cach pro.d. me
. Venerai.eid seasheal and Inten aaon speesen (9wRVIP-14, NMP1 ansiva.1
. NMP) operstmg chemssary (Plant Ostal
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e GE presentation (Dr. R. Hom)
. Meerrul Auemy (Annivtis and East templel
. Meenal emnessaan (Euse easyer does,CE daea. BWRvtP-14)
e Conclusions
. PLIDCI piedr in CCR m er hvione e 42 e 10* an/hr
. Der of 2J e IO$ m/hr boundo poedu.d CCA 1einaw et1)
. Appiameson of 2J u 10 m/hr suppere.4 nde geesert han 24 months
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EEE Effects ofIrradiation on Shroud EEE Comparison of Boat Sample Data Cracking e thgh fluence can contribute to the suscepubdity of tlw fn=
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- Sensitization can be found outside of the weld HAZ spicani esa.n rama Vu h
. Crackmg willextubit additional features:
cract Dancrwe Yes No
- Sigruncant Eraan fallout 5 onifmans maanano vos
- Sipuficant crack branchmg in higlwr fluence regions e irradiation will also produce sipu6 cant hardening of the base snatenal
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Summary E8E NMP1 Shroud Neutron U"9dM%$
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. m.mn st oud. oat sampw.oi ed a comp _.ita ttw eartwr evaluation, performed on a boat sample from another shroud, arradated to lugher Buence
- Lacauans of sensamanon
- Crackmg snarphology
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- Base suaanalcharaciensucs
. umi n aa r wi non indican s noirradianon effecu e MPM Technologies,Inc.
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.d ie, sis.f n.se ma n.i marden Presentation
- No sigmfrans gram fallour
- No sainc= =a nr= h==
(Dr. M.P. Manahan, Sr.)
. Crmkmg correlated with regions of weld educed senatuauan
- Fluence was below tevois where eradiauan effecu are
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EE Analysis of Boat Sample EgE Analysis f Boat Sample Dosimetry Data (continued)
Dosimetry Data e Analysis by Framatome in January,1998 using e Two boat samples were cut from the shroud cycle 7 transport data showed a discrepancy at the end of cycle 12 between the Fe and Ni dosimeters
- ID surface of V9 26.4 inches above midplane e Analysisof the210degreesurveillancecapsule (peak ID measured fluence = 3.49 x 10%/cm2)
dosimetry in May,1998 by MPM using a mid-cycle 12 transport analysis showed a similar
- OD surface of V108.3 inches below midplane (peak OD measured fluencr= = 1.42 x 10%/cm2)
discrepancy e In May,1998 MPM suggested that a large flux e Dosimetry data taken at three depths within dr P through cycle 12 would explain the each boat sample discrepancy i
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e Through cycle analysis has resulted in close Analyses include:
agreement between Fe and Ni dosimeters e R-0, R-Z, and R calculations for 5 'cle e Average ratio of the fluxes from Ni to those from Fe are 0.991 with a standard deviation of 3.3%
12 representative power profiles (15 e Calculations at the boat sample locations have transport calculations)
been shown to be conservative by comparison e Uncertainty Analysis with the measured fluxes
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NMP-1 R-6 Geometry EM Calculated Fast Fluence to Welds V9 and V10 at End ofCycle 13
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Neutron Transport Results gg Calculated Fast Fluence to Weld ggy H4 at End ofCycle 13 5 ' **"..d W'Id'
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EM urnmary and Conclusions S
l EMNeutron Transport ResultsforV9/V10 at End ofCycle 13 e Through cycle transport calculations for cycle IJ j
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M Supplemental Structural Margin EE Concluding Remarks Analysis eThere is substantialbasis for
. The euow.up suppiementai rractu,e mecha,ocs anairs demonstraks that the reqmred ASME cme reqmrd reduced crack growth rate margms are mamtamed. tor inore than 14.500 hours0.00579 days <br />0.139 hours <br />8.267196e-4 weeks <br />1.9025e-4 months <br />, even g p}uence e{fects are not significant
,
assunung a N a 5 m 105an/hr l
. Analysis consstent with BWRVIP-01 guidelmes l
. cr=w e er uncracw locanons coairma by eStructural analysis demonstrates mapecnons <svr.iuor v, aa vio inspection interval of 14,500 hrs is both votumetne mapections (UT) and voual
.
.
j
. c,mu e sor nar me d,=cnon capaduity a ur a justified without reducing CGR
,
quahfied by BWRVIP-a) for V4 weld
. V4, V9 and V10 haut load evaluations show sigruf. : ant l
marge l
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EVALUATED CRACK GEOMETRIES ORIGINAL ANALYSIS SUPPLEMENTAL ANALYSIS
- -- 0.9 7'
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6.31 6.00'
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b s
n 9.73'
.
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91.00'
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h 38.03'
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