IR 05000220/1999004
| ML20196G025 | |
| Person / Time | |
|---|---|
| Site: | Nine Mile Point  |
| Issue date: | 06/21/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17059C694 | List: |
| References | |
| 50-220-99-04, 50-220-99-4, 50-410-99-04, 50-410-99-4, NUDOCS 9907010035 | |
| Download: ML20196G025 (42) | |
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U.S. NUCLEAR R.EGULATORY COMMISSION
REGION I
Docket / Report Nos.: 50-220/99-04 50-410/99-04 License Nos.: DPR-63 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: March 28,1999 - May 8,1999 Inspectors: G. K. Hunegs, Senior Resident inspector R. A. Fernandes, Resident inspector R. A. Skokowski, Resident inspector F. J. Arner, Ill, Reactor Engineer A. L. Della Greca, Senior Reactor Engineer K. S. Kolaczyk, Operations Engineer G. W. Morris, Senior Reactor Engineer -
R. C. Ragland, Radiation Specialist Approved by: Michele G. Evans, Chief Projects Branch 1 Division of Reactor Projects !
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[DR ADOCK 05000220 PDR I
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i EXECUTIVE SUMMARY
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Nine Mile Point Units 1 and 2 50-220/99-04 & 50-410/99-04 March 28,1999 - May 8,1999 This inspection report included aspects of licensee operations, engineering, maintenance, and plant support. The report covered a six-week period of resident inspection. The results of an occupational radiation exposure and radweste management and transportation inspection from April 7 - 16, an inservice inspection program review from April 19 - 23, and an engineering inspection from April 12 - 16 were also included in this inspection repor Operations The April 24 Unit 2 automatic reacto, i hutdown from 100 percent power was characterized by the licensee and NRC staffs as a risk algnificant transient. The cause was determined to be a generator protection circuit relay failure which also resulted in a residual (slower) transfer to off-site power. The slow transfer caused large motor loads such as reactor feedwater pumps, i
- reactor recirculation pumps, and condensate booster pumps to trip. Operator performance with
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respect to procedure use, communications, and control of plant equipment was good. Senior management oversight of scram recovery efforts was appropriate. . Major equipment failures .
included the reactor core isolation cooling system and a partial loss of the uninterruptible power supply system. These equipment failures and other minor equipment problems did not significantly impact recovery efforts. (01.2)
The Unit 1 outage shutdown risk program was wellimplemented. The communication of plant protected equipment and safety system status was good. (01.3)
Overall, NMPC's approach to identifying and resolving equipment performance problems following the April 24 Unit 2 reactor scram was acceptable. Positive aspects of NMPC's post-scram evaluation process included the establishment of multi-discipline teams to review equipment performance, the conduct of periodic status briefs, and the use of vendor service . Senior management effectively challenged their staffs post-scram analysis which contributed to i a more rigorous evaluation and the re-creation of the event using the plant simulator. However, i a few performance shortcomings related to the scram evaluation process were apparen i Although the overall process was thorough, equipment troubleshooting and failure analysis were not methodical. NMPC management recognized these shortcomings and was developing methods to improve its staff's problem solving skills. (07.1)
Between March 5 and March 12,1999, Unit 2 experienced two events where the automatic .
depressurization system nitrogen storage tanks had excessive leakage. NMPC failed to recognize that the leakage exceeded the allowed limit, and therefore, did not take the required limiting condition of operation actions. This was a non-cited violation of Unit 2 Technical Specification 3.5.1. (08.2)
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l Executive Summary (cont'd) '
Maintenance j
Ths Unit 1 fuel off-load was well controlled. Communications between the operators on
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,. the cefuel bridge, as well as between the refuel bridge and the control room were observed to be good. - (M1.2)
i The installation of the emergency core cooling system torus suction strainers was well controlled. The work environment was clean, organized and good foreign material exclusion controls were in place. (M1.3)
During the Unit 2 scram, the reactor core isolation cooling (RCIC) system failed to operate as .
required and was manually tripped. This RCIC system failure was attributed to an inadequate maintenance procedure and the licensee's over-reliance on vendor support for a 1998 RCIC turbine trip throttle valve rebuild. The failure to ensure an adequate maintenance procedure was prepared and used to perform work on the RCIC system was a non-cited violation. Based on recent operating history, the RCIC system has exceeded its Maintenance Rule performance
- criteria. (M2.1) .
On April 24, a Unit 2 generator protection circuit relay failed which caused a reactor scra NMPC effectively evaluated the cause and consequences of the relay failure and implemented-acceptable corrective action. (M2.2)
Non-destructive examination personnel were qualified, and adhered to procedures while )
e performing examinations.; The core shroud and reactor vessel weld inspection plans were in
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accordance with the requisite NRC safety evaluation. Deficiencies identified during inspection activities were properly documented. A new surveillance program provided enhanced oversight of vendor activities. (M3.1)
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Enaneenog During the Unit 2 reactor scram transient, one of the two reactor protection systen, uninterruptible power supplies (UPS) failed. Excessive currents, which caused the inverter DC power supply fuse to blow, were the result of a UPS design deficiency involving a newly installed maintenance byp*Ss switch. _ NMPC identified that this vendor supplied UPS design change received an inadequate engineering design review. (E1.1)
Unit i design changes that were reviewed, correctly addressed the concems for which the
. modifications and been developed.' Typically, the analyses accurately described the purpose of the modification and the intended results; the calculation and safety evaluations satisfactorily supported the design changes; and the design change process was acceptably implemente (E1.2)
in the case of the Emergency Core Cooling System (ECCS) strainer modification, the technical
. , scope of the design change was comprehensive, but the licensee's original review of an ECCS pump air ingestion calculation lacked thoroughness regarding a small break loss of coolant lii
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Executive Summary (cont'd)
-. accident (LOCA) scenario and required a more detailed analysis and a revision of the supporting calculation. Additionally, the safety evaluation required revision to provide stronger bases for the conclusions contained therein regarding a large break LOCA. The lic qsee's review of air ingestion phenomena associated with the large break LOCA resulted in he j conservative decision to declare inoperable, in the future, any ECCS pump placed in operation i for surveillance testing or torus cooling. (E2.1)
l From March 20 to March 23,1999, Unit 1 operated with a maximum average planar heat generation rate (APLHGR) exceeding the limits specified by the technical specifications. This technical specification violation was non-cited. NMPC determined that the cause was the inadvertent processing of traverse in-core probe (TIP) data, due to inadequate computer system security on the 3D-Monicore system. Specifically, TIP data could be processed without authorization or operator knowledge from uncontrolled locations. Additionally, the oversight by station personnel with regards to reactivity management and core performance monitoring was poor, in that this discrepancy was not recognized for three days. (E4.1)
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The failure to conduct the required ASME Code inservice inspections of the reactor recirculation
. pump seal housing bolts and flange surfaces during the first and second ten-year inspection
- intervals was non-cited. (E8.1)
NMPC self-identified and promptly corrected a condition which could have adversely affected the ability of the unit to achieve safe-shutdown, involving the Unit 2 service water intake de-icing ,
heater control circuits which were not protected against a control room fire. This violation of !
License Condition 2.G was non-cited. (E8.3)
Plant Suo.pg_r1 Radioactive material / waste managcment and transportation programs wra effectively implemented as evidenced by use of up-to-date regulations and facility unses, appropriately trained personnel, proper procedural guidance and adequate maintenance of procedures, ;
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appropriate use of scaling factors to estimate isotopic content of radioactive material / waste l packages, and proper shipping records. (R1.1) i Radiological controts for the Unit i refuel outage were effectively planned and implemented and focused on jobs with elevated exposure estimates, high dose rates, and radiologically cor ) plex work. (R1.2)
Radiological posting practices for access to radiation areas, high radiation areas, and airborne ,
radioactivity areas were effective as evidenced by well defined boundaries and clear radiological
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postings. Some opportunities to enhance informational postings on the refuel floor that required
" health physics notification prior to entry" beneath the drywell dome and reactor head insulation were identified. (R1.2)-.
s Contamination monitoring requirements for access to the Turbine Building 305' Green Area
. (clean area within the radiologically controlled area [RCA)) did not include an entire whole body IV E
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Executive Summary (cont'd) l frisk similar to the requirements for RCA exit. However, they were adequate to minimize the risk for the spread and ingestion of s!;nificant amounts of radioactive contamination based on use of detailed procedures, restrictions 6n personnel that could use the facility, and close health physics oversight. (R1.2)
Effective high radiation area controls were implemented as evidenced by clear radiological I postings, use of locked doors when required, use of " Alarming" dosimetry, use of radiation work
- permits (RWPs), use of remote door alarms, requirements for a minimum available exposure for l access, and increased health physics oversight and monitoring for high radiation area entr ,
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Material conditions were good and housekeeping practices were effective as evidenced by clear aisles and walkways, neatly stored tools and equipment, and painted floor and wall surface (R2.1)
l Self-assessments, audits, and the deficiency / event reporting system were effectively used to L identify, evaluate, and resolve radiological control issues as evidenced by the conduct of multiple self-assessments and audits to satisfy the radiation protection program review l requirements in 10CFR20.1101(c) and use of the DER system to implement appropriate
! corrective actions and controls to pre /ent unplanned exposures. (R7.1) ,
One non-cited violation was ioentified associated with the failure to maintain access restrictions in the upper elevation of the drywell during movement of an irradiated core component on March 15,1997. (R7.1) .
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TABLE OF CONTENTS page EXECUTIVE SUMMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . li TABLE OF CONTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .vi i
Summary of Plant Status ~. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1
. l . Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01 Conduct of 0perations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 General Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.2 ' Automatic Reactor Shutdown (Unit 2) . . . . . . . .. . . . . . . . . . . . . . . . . . . 1 l 01.3 5 Outage Shutdown Safety (Unit 1) . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 f 07.1 Assessment of Post-Scram Troubleshooting Efforts (Unit 2) . . . . . . . . 4 .q 08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 t 08.1 (Closed) Licensee Event Report (LER) 50-410/98-26 . . . . . . . . . . . . . 5 08.2 (Closed) LER 50-410/99-03 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 l l . Mainten ance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 M1.1 General Comments (61726, 62707) . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 M1.2 Fuel Off-load Activities (Unit 1) . . . . . . . . . . .. . . . . . . . . . . . . . . . .. . . . 8 )
- M1.3 Installation of Core and Containment Spray Strainer Assemblies (Unit 1) i
........................................................ 8 M2 - Maintenance and Material Condition of Facilities and Equipment . . . . . . . . . . 9 M2.1 Reactor Core isolation Cooling (RCIC) System Failure During Reactor Scram Transient (Unit 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 M2.2 Generator Protection Relay Failure Resulted in Scram (Unit 2) . . . . . 10 M3 Maintenance Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . 11 M3.1 - Inservice Inspection (Unit 1) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 M8 - Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 q M8.1 (Closed) LER 50-410/99-02 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 j M8.2 (Closed)VIO 50-410/98-05-02 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 l M8.3 (Closed) VIO 50-220/98-02-05 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 Ill . Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 E1.1 Uninterruptible Power Supply (UPS) Failure During Reactor Scram Transient (Unit 2) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 l E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . 15 l E2.1 Design Modifications (Unit 1) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 E4 Engineering Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . 17 i
_ E4.1 .(Closed) LER 50-220/99-03 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 I'
E8 Miscellaneous Engineering lasues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 vi
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Table of Contents (cont'd)
l E (Closed) LER 50-220/98-19 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 l
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E (Closed) VIO 50-410/98-05-03 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 E8.3 (Closed) LER 50-410/99-04 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 E8.4 (Closed) IFl 50-410/98-19-03: Leakage of contaminated water l following a scram reset . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 E8.5 (Closed) Violation 50-220/98-16-02 . . . . . . . . . . . . . . . . . . . . . . . . . . 22 IV. Pla nt Support . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . 2 3 R1 Radiological Protection & Chemistry Controls . . . . . . . . . . . . . . . . . . . . . . . . . 23 R1.1 Unit 1 and Unit 2 Solid Radioactive Waste Management and Transportation of Radioactive Materials . . . . . . . . . . . . . . . . . . . . . . 23 I l R1.2 Refuel Outage Radiological Controls (Unit 1) . . . . . . . . . . . . . . . . . . 24 L R2 Status of RP&C Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 l R2.1 Radiological Housekeeping (Units 1 and 2) . . . . . . . . . . . . . . . . . . . . 27 R7 Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28 R Radiological Control Program Corrective Actions (Units 1 and 2) . . . 28 1 V. Mane mment Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30 Xi Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 30
' ATTACHMENTS Attachment 1- Partial List of NMPC Persons Contacted-Inspection Procedures Used l - Items Opened, Closed, and Updated
- List of Acronyms Used
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Report Details Summarv of Plant Status Nine Mile Point Unit 1 (Unit 1) began the inspection period at 100 percent reactor power. On April 11; Unit-1 was shutdown to begin a scheduled refueling outage (RFO15) and remained shutdown through the end of the inspection period. Significant outage activities included inspection of core shroud vertical welds and reactor vessel longitudinal welds, and installation of new emergency core cooling system suction strainer Nine Mile Point . Unit 2 (Unit 2) began the inspection period at 100 percent reactor power. Unit 2 automatically shutdown on April 24 due to the malfunction of an electrical relay associated with the generator protection circuit. Maintenance activities completed during the shutdown included:
reactor core isolation cooling system troubleshooting and repair; uninterruptible power supply system modifications; and recirculation system flow control valve (FCV) maintenance. Following the forced maintenance outagte, Unit 2 was retumed to service on May 4. Due to recirculation system flow control valve fluctuations, Unit 2 was placed in single-loop operation on May After FCV adjustments were made, Unit 2 was retumed to two-loop operation on May 9 and was
- retumed to 100 percent power on May 1 I. Operations 01 Conduct of Operations 2 01.1 GeneralComments (7170Z)
' 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 ano shift staffing, verification that the units were operated in conformance with Technical Specifications (TSs), and verification that logs -
and records accurately identified equipment status or deficiencies. In genaral, the i conduct of operations was professional and safety-consciou .2 Automatic Reactor Sh'utdown (Unit 2) Insoection Scooe (71707)
On April 24, at 4:19 a.m., Unit 2 experienced an automatic reactor shutdown (scram) i from 100 percent power. The inspectors responded to the site and observed portions of j the scram recovery process. The inspectors also reviewed the operator logs, post-scram review documentation, and the sequence of events. Additionally, the event was
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discussed with Unit 2 operations and management personne i l
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' Topical headings such as o1, M8, etc., are used in accordance with the NRC standardized reactor inspection report outlin Individual reports are not expected to address all outline topics. The NRC inspection manual procedure or temporary instruction that was used as Inspection guidance is listed for each applicable report sectio . .
2 Observations and Findinas The cause of the reactor shutdown was the failure of a relay in the generator protection circuitry (see section M2.2). The relay failure caused a turbine trip and subsequent automatic reactor shutdown. Because of the particular relay failure and design of the generator protection circuit, instead of a fast transfer of electrical loads to off-site power, a residual (slower) transfer occurred which caused all feedwater, condensate booster, and recirculation pumps to trip. All control rods fully inserted on the automatic shutdown and vessel level control was maintained by automatic initiation and injection of the high pressure core spray (HPCS) system. The reactor core isolation cooling (RCIC) system also initiated, but failed to come up to speed and was tripped by the control room operators (see section M2.1). All five turbine bypass valves opened to control reactor pressure below the main steam safety relief valve setpoint. Plant cool-down was commenced on natural circulation using the turbine bypass valves. By late evening on April 24, the plant was in cold shutdow ,
Coincident with the reactor shutdown, there was a trip of the uninterruptible power supply (UPS) which provides a portion of the power to the reactor protection system (see section E1.1). The partial loss of UPS resulted in several primary containment isolation valve group isolations. The partial loss of UPS had negligible impact on scram recovery effort The Updated Safety Analysis Report (USAR) classifies this event as an incident of moderate frequency. Notwithstanding the equipment problems, plant response was in accordance with the USAR. Because of the residual transfer of power, and notable equipment failures, the NRC staff performed an initiating event risk assessment. The NRC staffs assessment showed that the event was risk significant, in that, lhe total conditional core damage probability had increased and exceeded the accident sequence precursor threshold value used by the NRC staff for assessing significance. The licensee's risk assessment of this event was consistent with the NRC staffs assessmen The inspectors evaluated operator performance with respect to emergency operating procedure use, emergency plan use, communications and control of the plant. Based on interviews of operators and operations management and review of operator logs and plant sequence of event information, the inspectors determined that operators responded appropriately to the event. Reactor v6ssel level and pressure were well controlle Subsequent re-creation of the event on the simulator also showed that operator response was appropriate. The inspectors noted that Niagara Mohawk Power Corporation (NMPC) senior managers responded to the site and provided appropriate oversight. Additional operators were made available to provide assistance, as necessary. The inspectors observed good communications and good procedure use by operators during the post-scram recovery perioo. With the exception of the RCIC and UPS system failures, other major equipment operated as designed. A few minor system
~. m discrepancies were appropriately documented in the corrective action program for follow-up evaluation and repai . ..
3 Conclusions The April 24 Unit 2 automatic reactor shutdown from 100 percent power was
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, characterized by the licensee and NRC staffs as a risk significant transient. The cause was determined to be a generator protection circuit relay failure which 6.Do resulted in a residual (slower) transfer to off-site power. The slow transfer caused large motor loads such as reactor feedwater pumps, reactor recirculation pumps, and condensate booster pumps to trip. Operator performance with respect to procedure use, communications, and control of plant equipment was good. Senior management oversight of scram recovery efforts was appropriate. Major equipment failures included the reactor core isolation cooling system and a partialloss of the uninterruptible power supply syste These equipment failures and other minor equipment problems did not significantly impact recovery effort ,
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O1.3 Outaae Shutdown Safety (Unit 1) I 1 Inspection Scope (71707)
The inspectors reviewed the methods used by Unit 1 personnel to monitor shutdown safety as outlined in station procedure Observations and Findinas Procedure N1-ODG-11 " Shutdown Operations Protection Guideline," is used to monitor plant status during shutdown conditions. NMPC uses an attachment to the procedure for tracking the status of plant equipment important to shutdown safety associated with decay heat removal, inventory control, electrical power availability, secor.dary containment, and reactivity control. The attachment is updated each shift by control room operators and is used to brief station personnel at various meetings throughout the day. The inspector observed several briefings and noted good communication from station personnel with respect to emphasizing safety system status and protected components. The inspector noted that visual aides were used on control panels and in equipment rooms to warn personnel of the protected status of safety significant equipmen Conclusions The Unit'1 outage shutdown risk program was wellimplemented. The communication of plant protected equipment and safety system status was goo i
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07 Quality Assurance in Operations 07.1 Assessment of Post-Scram Troubleshootina Efforts (Unit 2) Insoection Scooe (71707)
NMPC appointed a post-scram review team to investigate the cause of the Unit 2 scram (see section 01.2). The inspectors attended post-scram review tsam and site operations review committee (SORC) meetings, observed NMPC troubleshooting efforts and discussed the scram evaluation activities with several members of the Unit 2 management staff. Additionally, the inspectors reviewed the post-scram review procedure. The inspectors assessed NMPC's overall performance in identifying the causes of the equipment deficiencie Observations and Findinas
.. A post-scram review team was appointed to investigate the cause of the scram and to determine corrective action. The technical services department had the lead on identifying the causes and equipment performance issues and were assisted by design engineering. Several teams were formed to focus on individual equipment performance'
problems including the RCIC, UPS, and generator protection relay failure. The teams used Procedure N2-REP-6, " Post-Scram Review," which provided an overall approach to evaluate the causes of a reactor scram and to review plant equipment performance. The i -
- teams held periodic briefs and formal SORC reviews were conducted; Senior management challenged the preliminary and apparent causes of equipment malfunctions during the review process The inspector noted that the operators that were involved in the transient prepared !
critique sheets listing their recollection of the event and actions taken, but a formal method to evaluate operator performance was not evident. Station practice has been to
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' sely on the self-evident nature of operator errors to determine if a more rigorous review is
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wanPnted.- In the case of this transient, the operations manager determined that it would be benefidal to re-create the event on the simulator to more thoroughly evaluete operator performance and learn from the event. As a result, some minor simulator fidelity issues were identified and additional operator performance incights were gaine l l The evaluation of the cause of the scram and equipment performance was particularly
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I challenging because of the cor1plexity of the transient. Additionally, data was not
. . available from the transient analysis recorder following the event, since it had not been !
properly aligned to automatically trigger the recording of data because of an earlier l operator error. : NMPC documented this problem in' deviation / event report (DER) 2-gg-
' 1260. Based on the inspectors' observations, NMPC's approach to the troubleshootirg )
and analysis of the transient and equipment problems was not methodical. For example, )
+ _ . the troubleshooting efforts to identify the cause of the RCIC trip were extensive and i drawn out. The eventual identification that tolerances for the trip and throttle valves were !
incorrect was the result of additional trip throttle valve agitation, late in the !
troubleshooting process, rather than the conduct of a formal root cause analysis. NMPC
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management attributed their staff's problem resolution weaknesses to the absence of necessary skills and training. The inspectors leamed that NMPC was developing methods and training to improve performance in this area, including increased formal root cause analysis trainin Conclusions !
Overall, NMPC's approach to identifying and resolving equipment performance problems following the April 24 Unit 2 reactor scram was acceptable. Positive aspects of NMPC's post-scram evaluation process included the establishment of multi-discipline teams to review equipment performance, the conduct of periodic status briefs, and the use of vendor services. Senior management effectively challenged their staff's post-scram analysis which contributed to a more rigorous evaluation and the re-creation of the event using the plant simulator. However, a few performance shortcomings related to the scram evaluation process were apparent. Although the overall process was thorough, equipment troubleshooting and failure analysis were not methodical. NMPC management recognized these shortcomings and was developing methods to improve its staff's probier' solving skill Miscellaneous Operations issues (92700)
O8.1 (Closed) Licensee Event Report (LER) 50-410/98-26: Seismic Monitor inoperable for More Than Thirty Days and Special Report Not Submitted. On October 9,1998, NMPC -
discovered that a reactor building triaxial response spectrum recorder (TRSR) was not properly oriented which rendered the seismic monitor inoperable. The seismic monitoring instrumentation is installed to monitor and record data in the event of an earthquake. This data would be used following an earthquake to verify that the event ;
was bounded by the analytical model provided in the USAR. NMPC determined that the monitor had been inoperable since at least May 1997, when the equipment was last i tested, and was potentially inoperable since initial installation. On October 23,1998, i NMPC corrected the orientation of the TRSR and verified proper orientation of the other accessible seismic monitoring equipmen i i
NMPC concluded that station personnel who developed and revised the seismic monitor surveillance procedures were not aware of the importance and precise tolerances required for the orientation of the seismic monitor instruments. Consequently, the instruments were improperly positioned and rendered inoperable for greater than thirty days. NMPC's failure to maintain the seismic monitors in an operable status and submit a Special Report, as required by TS 3.3.7.2, constitutes a violation of minor significance and is not subject to formal enforcemen* actio The inspectors completed an on-site review of the LER and verified that it was completed in accordance with the requirements of 10CFR50.73. Specifically, the
__ _ _ description and analysis of the event, as contained in the LER, were consistent with the inspectors' understanding of the event The root cause and corrective and preventive actions, as described in the LER, were reasonable. This LER is close , ,
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08.2 (Closed) LER 50-410/99-03: Automatic Depressurization System (ADS) Nitrogen Leakage in Excess of Unit 2 Technical Specifications Surveillance Limits Lnx.@ tion Scope (92700) ,
Between March 5 and March 12,1999, Unit 2 experienced two events where the leakage from the ADS nitrogen storage tanks exceeded the design basis leak rate. The inspMhrs reviewed the associated DERs, attended pertinent SORC meetings, reviewed the sGsequent LER, and discussed related issues with NMPC personne Observations and Findinas The nitrogen system supplies high pressure nitrogen to the ADS valves. The system consists of high pressure storage tanks located outside the reactor building, which supply nitrogen to enk Nos. 4 and 5, located within the reactor building, which in tum supply nitrogen " > - ADS accumulators. Tank No. 4 supplies three ADS valves, while !
t- , tank No. 5 supp..e , sur ADS valves. The tanks are normally isolated from the outside l storage tanks and are periodically re-pressurized to make-up for normal leakag * i Event 1 l l
On March 5, Unit 2 completed a surveillance test on the ADS nitrogen system. The test i exercised various valves within the system and required operators to remove the blank j
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. flange on the nitrogen emergency fill connection and attach a test assembly. Following i the test, operators removed the test assembly, re-installed the flange and verified ;
acceptable tank pressures. Over the next day, operators re-pressurized tank No. 5 several times. Subsequently, operators identified a leak at the blank flange. The gasket was replaced and the pressure in the tank was stabilized. DER 2-1999-0682 was written to evaluate the event, and the subsequent review showed that the leak rate from tank No. 5 exceeded the TS allowed valu Event 2 On March 9,1999, operators responded to a low nitrogen pressure alarm on tank No. 4 and manually re-pressurized the tank. From March 9 to March 12, operators re-pressurized the tank five more times while searching for leaks. Initially, the station shift supenrisor (SSS) considered the leakage to be of a similar magnitude as past leaks and concluded that the TS leakage limit was not exceeded. However, on March 12, the SSS determined that Unit 2 may have exceeded the TS limit and initiated DER 2-1999-074 Later that day, NMPC identified and repaired a few small nitrogen system leaks. The leak rate decreased to below the TS limit. Subsequently, NMPC found the normally closed valve 2GSNV73A slightly open, and difficult to operate. NMPC concluded that since the first re-pressurization of tank No. 4, on March 9, that valve 2GSNV73A had permitted leakag . .
in addition to the two events described in the LER, NMPC believes that excessive nitrogen leakage occurred at other times and it was not recognized that the TS limit or the design basis was exceede As described in the LER, NMPC determined that, based on the maximum observed leakage during these two events, approximately 1.68 and 1.86 days of nitrogen for tank Nos. 4 and 5 would have been available. Although, this was less than the design basis of five days, NMPC concluded that 1.5 days was sufficient to allow for a nitrogen truck to arrive on site to resupply the nitrogen tanks. The inspectors considered this to be reasonable. Nonetheless, the failure to take the actions required by TS 3.5.1.e.2 during the period when the ADS tanks leakage rate exceeded the TS allowed limits is a violation. This Severity Level IV violation is being treated as a Non-Cited Violation, l
consistent with Appendix C of the NRC Enforcement Policy (NCV 50-410/99-04-01).
This violation is in the licensee's corrective action program as DERs 2-1999-0682 and 2-1999-074 .The inspectors completed an on-site review of the LER and verified that it was completed in accordance with the requirements of 10CFR50.73. Specifically, the description and analysis of the event, as contained in the LER, were consistent with the
- inspectors' understanding of the event. The root cause and corrective and preventive actions as described in the LER were reasonable. This LER is close Conclusion Between March 5 and March 12,1999, Unit 2 experienced two events where the automatic depressurization system nitrogen storage tanks had excessive leakag NMPC failed to recognize that the leakage exceeded the allowed limit, and therefore, did not take the required limiting condition of operation actions. This was a non-cited violation of Unit 2 Technical Specification 3. . Maintenance
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M1 Conduct of Maintenance M1.1 General Comments (61726. 62707)
Using NRC inspection Procedures 61726 and 62707, the resident inspectors periodically observed various maintenance activities and surveillance tests. As part of the observations, the inspectors evaluated the activities with respect to the requirements of
- the Maintenance Rule, as detdad 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 approprit'e focus on safet , Speedic activities and noteworthy observations are detailed in the inspection report. The inspectors reviewed procedures and observed all or portions of the followirg j maintenance / surveillance activities:
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e WO 99-08109, Hydraulic control unit post maintenance testin . . WO 99-06450, Uninterruptible power supply inverter cleaning and inspectio e N2-OSP-ICS-R002, Reactor core isolation coolin e WO 98-03424, Feedwater heater replacemen e RFMSHRD30, Electric discharge machine shroud weld V9 and V10 activitie M1.2 Fuel Off-load Activities (Unit 1)
i Insoection Scone (60710)
The inspectors observed portions of Unit 1 fuel off-load activities using the guidance provided in NRC Inspection Procedure 60710, " Refueling Activities." Observations and Findinas The inspectors observed fuel off-load activities from the control roorn and from the refueling bridge. .The off-load was performed in accordance with approved procedures, and was well controlled. The inspectors considered the communications by the operators on the refuel bridge and between the refuel bridge and the control room operators to have been good. The inspectors independently verified installation of the refueling interlock jumper and that a sample of the fuel moves were correc Conclusions :
The Unit 1 fuel off-load was well controlled. Communications between the operators on the refuel bridge, as well as between the refuel bridge and the control room were observed to be goo M1.3 Installation of Core and Containment Sorav Strainer Assemblies (Unit 1) Insoection Scope The inspector reviewed work order packages for installing the emergency core cooling system (ECCS) torus strainers to ensure the installation was being conducted in accordance with station drawings and work instruction Observations and Findinas The inspector utilized work order packages 98-03314-05,06,07, and 08 to verify that the core spray and containment spray systems' strainer assemblies were properly installe The peckeges included quality assurance hold points, foreign material exclusion signatures, as well as signatures for verification of component fit-up. The inspector verified that welders were utilizing approved welding procedures and welding materia Additionally, the inspector verified the welder's qualifications were current and compatible for the weld procedure and process being utilized on the modificatio W
4 4
The inspector toured the torus and work site and observed that the work areas were clean and organized and that waste materials were kept to a minimum to reduce the challenges to foreign material exclusion controls. In addition, the inspector noted that
- quality assurance personnel were assigned to the project and were conducting routine surveillance sctivitie ' Conclusions The installation of the emergency core cooling system torus suction strainers was well controlled. The work environment was clean and organized and good foreign material exclusion controls were in effec M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Reactor Core Isolation Coolina (RCIC) System Failure Durina Reactor Scram Transient (Unit 2) Insoection Scooe (62707. 37551)
. During the April 24 reactor scram transient, the RCIC system failed (see section 01.2).
The inspectors reviewed applicable sections of the USAR, RCIC operating procedures, and the DER disposition. The inspectors walked down portions of the system, observed system troubleshooting efforts, and interviewed the operator who was responsible for the operation of the RCIC system at the time of its failur Observations and Findinas During the reactor scram, reactor low water level was reached and the RCIC system ,
received an automatic start signal. The RCIC injection valve opened and the trip throttle !
valve indicated that it was open. However, the maximum RCIC turbine speed observed I by operators was 200 rpm with zero discharge flow indicated. Based on these control l
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room indications, the control room operator manually secured the RCIC turbine. Based on the observed system operating parameters, the inspector concluded the operator's .
action to trip the RCIC turbine was appropriat l
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Subsequent troubleshooting showed that the RCIC system had received a valid initiation signal and that the steam admission and outboard injection valves had opened. Data -
recorders confirmed that the RCIC turbine speed had increased to 200 rpm, at which point the turbine trip valve was trippe NMPC conducted extensive troubleshooting and determined that the latching mechanism for the trip throttle valve was not sufficiently engaged. NMPC determined that the set-up of the overspeed trip linkage and associated valve components was not )
- a correct.~The inspectors determined that the RCIC turbine trip throttle valve had been disassembled and rebuilt during the 1998 outage. NMPC obtained vendor assistance to complete the work and had relied upon the vendor's expertise. Licensee review determined that the overspeed trip linkage tolerances were not described in the work
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package and consequently the linkage was re-assembled with incorrect tolerances. The failure to provide an adequate work procedure is a violation of 10 CFR 50 Appendix B Criterion V, " Instruction, Procedures, and Drawings." This severity level IV violation is l
. being treated as a Non-Cited Violation, consistent with Appendix C of the NRC 1 Enforcement Policy (NCV 50 410/99-04-02). This procedural adequacy violation is in the licensee's corrective action program as Deficiency Event Report (DER) 1099-125 Corrective actions included revising the maintenance procedure and providing additional training for maintenance personnel. The licensee also determined that additional industry operating experience was available, but not used, conceming RCIC system trip throttle valve maintenanc The Maintenance Rule performance criteria for the RCIC system is two functional failures over a two-year period. This failure was classified as a maintenance preventable ,
functional failure and actual performance shows three functional failures during the previous two-year period. At the end of the inspection period, NMPC was evaluating the RCIC system for classification in Maintenance Rule category (a)(1). I Conclusions During the Unit 2 scram, the reactor core isolation cooling (RCIC) system failed to operate as required and was manually tripped. This RCIC system failure was attributed to an inadequate maintenance procedure and the licensee's over-reliance on vendor support for a 1998 RCIC turbine trip throttle valve rebuild. The failure to ensure an 6- adequate maintenance procedure was prepared and used to perform work on the RCIC system was a non-cited violation. Based on recent operating history, the RCIC system has exceeded its Maintenance Rule performance criteri M2.2 Generator Protection Relav Failure Resulted in Scram (Unit 2) . i l Inspection Scoos i The Unit 2 reactor scram was caused by the failure of a generator protection circuit rela The inspector observed and reviewed NMPC's troubleshooting and evaluation methods i used to determine the cause of the relay failur l l Observations and Findinas i NMPC troubleshooting effort showed that the volts / hertz relay associated with the generator protection circuit had failed. The circuit design was such that the relay failure i caused the turbine trip (and reactor scram) arid caused a residual (slower) transfer of ,
electrical loads to offsite power source !
No apparent cause for the relay failure was identified. Inspection of the relay did not
, .. . - reveal any physical characteristics for the failure mode. Bench testing showed that the i relay was defective and that the malfunction would provide a spurious trip signal with an !
outcome the same as the event that was experienced. To obtain more specific information concoming the failure mode, the failed relay was shipped to an independent I l
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laboratory for failure analysis. The relay was replaced and calibrated satisfactoril Other similar relays in use were recalibrated and tested satisfactoril The inspector reviewed the failed volt / hertz relay work history which showed that this relay was replaced during outages in 1990 and 1998. The failed relay and similar relays used in the generator protection circuit fall under the Unit 2 preventive maintenance program and are calibrated every refuel outage. These relays were included within the scope of the Maintenance Rule and this event was classified by NMPC as a functional failur Conclusions On April 24, a Unit 2 generator protection circuit relay failed which caused a reactor scram. NMPC effectively evaluated the cause and consequences of the relay failure and .
implemented acceptable corrective actio I M3 Maintenance Procedures and Documentation l
l M3.1 Inservice inspection (Unit 1) ]
! Inspection Scope (73753) l I
The inspectors reviewed the inservice inspection (ISI) activities that were part of i Refueling Outage (RFO) 15. The review involved performing a walkdown of portions of the core spray system piping and venfying piping welds were reflected in the ISI program manual and system isometric drawings. Non-destructive examination (NDE) activities were observed, and the qualifications of NDE personnel verified. Additionally, the inspectors assessed NMPC's oversight of contractor NDE activitie Observations and Findinas ISI Proaram Manual No deficiencies were noted in the ISI program manual during the field walkdown of the core spray system. The list and location of core spray system piping welds contained in the manual, matched the as-built system configuration. However, during the field walkdown the inspectors identified errors in the core spray system weld map isometric drawing F-45183-C. Specifically, the ISI progtam manual indicated welds 81-WD-128 and 81-WD-183-A were located downstream of core j spray pumps 11 and 12. The inspectors confirmed the welds were located in the l correct location on the core spray piping. However, they were not shown on the !
corresponding weld map drawing. NMPC documented this drawing error in DER 1- l 99-122 Both we!ds were located in ASME Code Class 2 piping and were not among the population of welds that NMPC had selected for NDE activities. This approach was in accordance with ASME Code requirements, which indicate only 25% of the l
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applicable welds in ASME Code Class 2 piping need be examined over the 10-year inspection interval. Accordingly, the weld map error did not result in an ASME Code violation and this ISI Program administration oversight was of minor safety consequence not subject to formal enforcement actio Observation of NDE Activities The inspectors witnessed several NDE field inspections, including an ultrasonic (UT)
examination performed by NMPC personnel on a recirculation system piping weld, and a visual examination of a service water system piping hanger conducted by a contractor. The individuals who performed the examinations met the training and experience requirements outlined in procedure SNT-TC-1 A " Recommended Practice, Personnel Qualification and Certification of Non-Destructive Testing."
While observing the UT examination, the inspectors verified the UT test equipment was calibrated in accordance with industry and NMPC standards. Further, the inspectors verified that deficiencies uncovered during the visual and UT examinations were documented as required in DER Core Shroud and Beltline Weld Insoection Activities NMPC had made arrangements with two vendors, General Electric and Framatone, to perform NDE activities on the horizontal welds in the reactor vessel and core shroud, respectively. By review of vendor inspection plans and interviews with NDE personnel,-
the inspectors verified the inspection scope for the shroud and reactor vessel welds were in accordance with the NRC approved inspection plans described in NRC correspondence to NMPC, dated March 24 and April 7,1999, respectively. To minimize the possibility that relevant indications would be overlooked, both vendors had at least two individuals who independently review the NDE dat Oversiaht of NDE Activities During this outage, NMPC changed its philosophy regarding oversight of contracted NDE activities. Prior to the change, NMPC NDE personnel provided little formal oversight of contracted NDE activities. Instead, oversight was provided on an informal basis, whereby NMPC personnel would observe contractor activities on a time-available basi Formal oversight was limited to yearly audits of the ISI progra During this outage, NMPC developed a formalized ISI surveillance schedule that outlined q which NDE activities would be monitored. Most monitoring was conducted by NMPC J NDE personnel. However, contract personnel where scheduled to oversee some NDE activities where NMPC did not have the necessary in-house experience to adequately observe and evaluate. Surveillance plan observations were to be documented and
. forwarded to management for revie .
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The inspectors did not have the opportunity to review any completed surveillance reports I or observe performance of surveillance in the field, so it was not possible to comment on l
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the quality of the effort. However, the new surveillance program does provide NMPC additional assurance that contracted activities will be properly conducte Conclusions Non-destructive examination personnel were qualified, and adhered to procedures while performing examinations. The core shroud and reactor vessel weld inspection plans were in accordance with the requisite NRC safety evaluation. Deficiencies identified I during inspection activities were properly documented. A new surveillance program provided enhanced oversight of vendor activitie M8 Miscellaneous Maintenance issues (37551, 92700, 90712, 92903)
l M8.1 (Closed) LER 50-410/99-02: Missed Technical Specification Channel Functional Test of the Recirculation Flow Upscale Rod Block. The technical details associated with this LER were discussed in NRC Inspection Report (IR) 50-410/99-03, Section M1.2. The l Inspectors completed an in-office review of the LER and verified that it was completed in accordance with the requirements of 10CFR50.73. Specifically, the description and analysis of the event, as contained in the LER, were consistent with the inspectors'
l understanding of the event. The root cause and corrective and preventive actions as-described in the LER were reasonable. This LER is close M8.2 (Closed) VIO 50-410/98-05-02: Failure to conduct surveillance test on batteries.
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Specifically, during Refueling Outages (RFOs) 4 and 5, credit was inappropriately taken I for the battery performance test, in lieu of the battery service test for the Division i 125 l
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volt battery. Subsequently, NMPC issued LER 50-410/98-09 " Missed Battery TSSR
[ technical specification surveillance report] Due to Inappropriate Interpretation." This LER was reviewed and closed in NRC IR 50-410/98-05. Based on the review of the LER, which provided the root cause and corrective actions regarding the event, NMPC was not required to provide a separate response to the violation. The inspectors verified l= implementation of the corrective actions associated with this event. Violation 50-410/98-05-02 is closed.
L M8.3 (Closed) VIO 50-220/98-02-05: Inadequate plant impact in Work Order (WO) packag Specifically, during the development of a troubleshooting WO associated with a control l room chilled water temperature control valve, the impact of removing two lesds was not l
adequately evaluated. As a result, removing these leads caused an unanticipated L opening of the control room ventilation outside air and return air dampers. The inspectors confirmed the completion of the corrective actions associated with the event l ma described in NMPC's June 28,1998, response to the violation. Violation 50-220/98-l 02-05 is closed.
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111. Enaineerina E1 Conduct of Engineering E Uninterruotible Power Suoolv (UPS) Failure Durina Reactor Scram Transient (Unit 2) Inspection Scope (37551)
During the reactor scram transient, one of the two reactor protection system uninterruptible power supplies (2VBB-UPS38), failed. The inspector reviewed applicable
' sections of the USAR, UPS operating procedures, and the DER dispositio Observations and Findinas The power supply for the reactor protection system (RPS) consists of two UPS system . Each UPS has three power sources: the preferred alternating current (AC) source, the direct current (DC) source, and the maintenance AC source. Upon loss of the preferred AC source, the UPS automatically switches to the DC source. Each UPS is connected to
. its RPS through two redundant electrical protection assemblies (EPAs). The EPAs are .
designed to protect the RPS circuits from voltage or frequency deviation During the event, power was lost to the preferred UPS power supply causing the UPS to transfer to the battery. Upon re-energizing the AC switchgear, the UPS DC source fuse o e blow JAlso, the voltage had lowered sufficiently enough to cause the EPAs to trip on -
undervoltage, resulting in a loss of power to the RPS. The impact of the loss was limited because the plant was already shutdow NMPC performed a formal event and causal factor analysis to determine the root cause !
of the UPS failure, it was determined that, excess current caused the DC power supply !
fuse to blow. The excess current was the result of a design deficiency with the control l circuit board for the maintenance bypass switch. The specific design deficiency was 1 incorrect grounding of a control circuit board. . NMPC determined that the design - -]
deficiency ws.: created during the installation of the UPS maintenance bypass switc The maintenance bypass switch for the failed system was a vendor supplied modification which was installed during the 1998 refueling outage and for the other train of UPS, in 1996. During the modification review process, NMPC did not recognize that the grounding circuit for the maintenance bypass switch was incorrect. Contributing to NMPC's oversight was the absence of appropriate vendor supplied design change drawings. As documented in DER 2-1999-1707, corrective actions included a review of the engineering design change process. In addition, NMPC's interim corrective action included the installation of a temporary modification to remove the motor-operated feature of the maintenance bypass switch, effectively removing the circuit card design deficienc i NMPC classified the UPS failure as a Maintenance Rule functional failure. The inspector determined that this was the only functional failure for the UPS system during the previous two-year perio . .
' Conclusions During the Unit 2 reactor scram transient, one of the two reactor protection system uninterruptible power supplies (UPS) failed. Excessive currents, which caused the inverter DC power supply fuse to blow, were the result of a UPS design deficiency involving a newly installed maintenance bypass switch. NMPC identified that this vendor 1 supplied UPS design change received an inadequate engineering design revie I E2 Engineering Support of Facilities and Equipment ,
E Desian Modificetions (Unit 1)
l a insoection Scope (37700)
The inspectors reviewed selected Unit 1 design change packages (DCPs) to assess the quality of engineering analyr 's and to verify that the design change process complied with plant administrative procedures and regulatory requirements. The inspection
- addressed temporary and permanent design changes and included a review of the background information, applicable analyses, calculations, safety evaluations, internal review process, and post-modification testing activities. The inspectors also conducted walkdowns of selected installations to verify their conformance with applicable document Observations and Findinos The inspectors determined that the reviewed system and component changes correctly addressed the concems for which the design modifications had been prepared and that
- the applicable analyses accurately described the purpose of the modification and the intended results. Except as described below, the analyses, calculations, and safety evaluations were detailed, supported the design changes, and had been appropriately reviewed. The inspectors identified no concems with the installed equipment or post-modification testing performe ECCS Suction Strainer Replacement This modification pertained to the installation of new horizontal stacked disc strainers in the torus and was initiated to address strainer plugging concems raised by NRC Bulletin 96-03, " Potential Plugging of Emergency Core Cooling Suction Strainers by Debris in I Boiling-Water Reactors." Besides the installation of the strainers, the modification package addressed other needed changes, including: (1) the addition of new spectacle flanges and a strainer between the condensate storage tanks and the core spray pumps; and (2) the removal of retired-in-place hydrogen-oxygen monitoring tubing in the torus to eliminate a potential direct debris source. The inspectors found the technical scope of the design changes to be comprehensive and the licensee's review of the procedures requiring revision due to the design changes to be thoroug I j
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in vendor calculation No. S14STRAINERM002, "ECCS System Strainer Air / Steam Ingestion Analysis," the licensee evaluated whether the air bubbles that formed in the Unit 1 torus during c loss of coolant accident (LOCA) or a safety-relief valve (SRV)
s discharge presented a challenge to the operation of the core spray and containment spray systems. The inspectors found that, although the licensee had performed the required independent review of the calculation, tha review lacked thoroughness in certain areas. For instance, based on the results of the calculation, the licensee had concluded that an SRV actuation while the containment spray pumps were operating, such as in response to a small break LOCA, would result in air being ingested into the containment spray pump strainers and cause degradation of the pumps. However, the licensee did not address this potential condition in the unreviewed safety question determination section of the safety evaluation. As, a result of the inspectors' questions in this area, the licensee asked the vendor to perform a more detailed analysis of this specific scenario and subsequently determined that air ingestion and, hence, pump degradation would not j occu i
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- During a large break LOCA, the vendor calculation indicted that the time lapse between the onset of the LOCA and the start of the pumps would prevent air ingestion into the i new ECCS suction strainers and would not challenge the operability of the pumps. In the calculation, the vendor concluded that a pump, which was already running at the onset of a large break LOCA may be momentarily degraded due to air ingestion; however, the pumps would still be able to achieve the flow rates assumed in the accident analysi The licensee accepted the calculation results, but conservatively decided that, whenever l a core or containment spray pump was placed in operation for testing or torus cooling, J they would declare that loop of the system not operational and follow the TS j requirements regarding the limiting condition for operation (LCO) of that system. As in !
the case of the small break LOCA, the licensee had not specifically addressed in the safety evaluation the potential degraded condition of the pump for which the LCO was necessary. The licensee explained that the potential for air ingestion by the pumps was a pre-existing unrecognized condition that was being alleviated by the new strainers and that the LCO was a conservative measure to assure the reliability of the affected syste The inspectors determined that the two conditions described above should have been included in the licensee's unreviewed safety question review. Upon further review of the 4 issue, the licensee determined that the safety evaluation should be revised to provide stronger bases for the conclusions reached. Accordingly, the licensee initiated Deviation / Event Report (DER) 1-1999-148 Conclusions Unit i design changes that were reviewed, correctly addressed the concems for which the modifications had been developed. Typically, the analyses accurately described the purpose of the modification and the intended results; the calculation and safety evaluations satisfactorily supported the design changes; and the design change process l
.was acceptablyi mpemente l l In the case of the Emergency Core Cooling System (ECCS) strainer modification, the technical scope of the design change was comprehensive, but the licensee's original
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review of an ECCS pump air ingestion cc!culation lacked thoroughness regarding a small break loss of coolant accident (LOCA) scenario and required a more detailed analysis and a revision of the supporting calculation. Additionally, the safety evaluation required revision to provide stronger bases for the conclusions contained therein regarding a large break LOCA. The licensee's review of air ingestion phenomena associated with the large break LOCA resulted in the conservative decision to declare inoperable, in the future, any ECCS pump placed in operation for surveillance testing or torus coolin E4 Engineering Staff Knowledge and Performance E4.1 (Closed) LER 50-220/99-03: NMP1 Thermal Limit Exceeded the Requirements of Technical Specifications Inspection Scope (37551. 92700)
On February 19,1999, the Unit 13D-Monicore system was mistakenly updated with a traverse in-core probe (TIP) power distribution from February 4,1999. This action resulted in the plant being operated with the average planar heat generation rate (APLHGR) exceeding the limits specified in the Technical Specifications. The inspector reviewed the LER and the documentation associated with NMPC's corrective action progra Observations and Findinas 3D-Monicore is a system of computer programs designed to monitor and predict important core parameters. The programs criculate reactor power, moderator void and flow distributions in the core. This information is used to determine other core parameters such as margins to thermal limits, and fuel exposure. The programs are designed to track current reactor parameters automatically (usually once per hour) or on demand. The computer program accuracy is enhanced by making use of in-core neutron flux measurements. NMPC utilizes the TIP system for in-core flux measurements and inputs this data to the 3D-Monicore. On March 23, NMPC completed a routine TIP data collection run and subsequently transferred the new data to the 3D-Monicore system. The new TIP data correctly updated the computer's calculated core power distribution and following the printout of the core parameters the operators determined that the APLHGR value for one area of the core was 2.2 percent above TS l limits. Operators immediately reduced power to restore the APLHGR within TS limit NMPC's investigation into the event determined that an inadvertent " Process TIPS" command was entered into the 3D-Monicore system. This action essentially put old in-core flux measurements into the program. During the investigation, NMPC determined that the action had a discemable effect on the computer printout of the core parameters, but was not identified by the operators or the reactor engineering group at the time of the error.~ NMPC's investigation also determined that the error did not cause any adverse effects until after a control rod pattem adjustment was made on March 20,1999. Had the proper TIP case been in the computer program at that time, there is a high probability that the particular rod adjustment would not have been done. Evaluation by NMPC l
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determined that an APLHGR of 2.2 % above the limit would not have exceeded any of the 10 CFR 50.46 licensing criteria, and therefore, had minimal safety consequence " NMPC determined that the cause ofinadvertent processing of TIP data was inadequate computer system security on the 3D-Monicore system. The system was not protected, in that the design allowed TIP data to be processed without authorization and without waming from uncontrolled loc &tions. The inspectors determined that additional information and follow-up was required in the area of computer security (IFI-50-220/99-04-03). NMPC determined that the reactor engineering group failed to recognize the corruption of the system due to insufficient analysis of daily 3D-Monicore data, in that the discrepancy was not recognized for three days. Corrective actions in the LER included disabling the ability to process TIPS from uncontrolled locations. The NMPC root cause evaluation identified several other corrective acticas including developing tools to aid in monitoring the accuracy of 3D-Monicore and tracking key core thermal limit parameters for trending and analysis purpose As discussed above, NMPC determined that the actual impact of exceeding the thermal limit was small. Nonetheless, the failure to maintain core thermal limits as required by TS 3.1.7.a is a violation. This severity level IV violation is being treated as a Non-Cited Violation (NCV), consistent with Appendix C of the NRC Enforcement Policy (NOV 50-220/99 0444). This violation is in the licensee's corrective action program as DER 1-1999-0837.
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The inspectors completed an on-site review of the LER and verified that it was completed in accordance with the requirements of 10CFR50.73. Specifically, the description and analysis of the event, as contained in the LER, were consistent with the inspectors' understanding of the event. The root cause and corrective and preventive actions as described in the LER were reasonable. This LER is close Conclusiong
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From March 20 to March 23,1999, Unit 1 operated with a maximum average planar heat j generation rate (APLHGR) exceeding the limits specified by the technical specification I I
This technical specification violation was non-cited. NMPC determined that the cause was the inadvertent processing of traverse in-core probe (TIP) data, due to inadequate computer system security on the 3D-Monicore system. Specifically, TIP data could be processed without authorization or operator knowledge from uncontrolled location Additionally, the oversight by station personnel with regards to reactivity management !
and core performance monitoring was poor, in that this discrepancy was not recognized for three day l
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E8 Miscellaneous Engineering issues (92700)
E (Closed) LER 50-220/98-19 Missed ASME Section XI Inservice inspection Due to Cognitive Error, Inspection Scope On November 13,1998, while reviewing an issue associated with a contingency work package, NMPC discovered certain visual inspections had not been performed on plant
components as required by the ASME Code. The missed inspections concemed the <
l failure to perform visual examinations of the reactor recirculation pump seal housing bolts and flange surfaces, during the first and second ten year inspection interval Observations and Findinas -
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NMPC determined the inspections were missed since the reactor recirculation pumps i
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bolts had not been classified as pressure retaining components in design document The recirculation pump flanges were not examined, because ISI personnel
- overlooked an ASME Code requirement that stated, if any of the five recirculation
' pumps are disassembled during an interval, one pump flange surface inspection must be performed. During the first and second ten year intervals, several recirculation pumps were disassembled.
P -
Niagara Mohawk corrective action included revising the ISI inspection plan to incorporate the required inspections. A review of the ISI plan was conducted to ensure ASME Code and regulatory requirements were identified in the plan. No other missed inspections were identified. Finally, an inspection of the recirculation pump seal housing belts was scheduled for completion during refuel outage (RFO) :
1 The inspectors reviewed the ISI program plan and self assessment reports, and
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I verified the plan had been modif'ad. The failure to conduct the required ASME code inspections is a severity level IV violation and is being treated as a Non-Cited Violation, censistent with Appendix C of the NRC Enforcement Policy (NCV 50- l
. 220/99-04-05). This violation is in the licensee's corrective action program as LER '
'50-220/98-1 ,
i The inspectors completed an on-site review of the LER and venfied that it was !
completed in accordance with the requirements of 10CFR50.73. Specifically, the 1 description and analysis of the event, as contained in the LER, were consistent with the inspectors' understanding of the event. The root cause and corrective and preventive ,
actions as described in the LER were reasonable. This LER is close i i
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20 Conclusion The failure to conduct the required ASME Code inservice inspections of the reactor s < - recirculation pump seal housing bolts and flange surfaces during the first and second ten-year inspection intentals was non-cite E8.2 (Closed) VIO 50-410/98-05-03: Failure to perform adequate design for emergency diesel i generator (EDG) modification on fuel line. Specifically, the installation of a 1993 l modification to the EDG fuel lines failed to include a rubber grommet at the piping l support to compensate for system vibration. As a result, vibration of the fuel line pipe caused fretting of the pipe at the location of the pipe support and on April 14,1998, during surveillance testing of the Division 11 EDG, a fuel leak developed. NMPC issued DER 2-98-0891 to address this issue. The root cause and corrective actions associated with this DER were reviewed in NRC IR 50-410/98-05. Based on this review, NMPC was not required to provide a separate response to the violation. The inspectors verified completion of the corrective actions associated with this event. Violation 50-410/98-05-03 is close E8.3 (Closed) LER 50-410/99-04: NMP2 Service Water Intake De-Icing Heater Control Circuits do not Meet Fire Protection Program Requirements insoection Scooe (92700)
o .On March 18,' 1999, during a review of the Safe Shutdown Analysis for a control room l fire, Un;l 2 personnel determined that the service water intake de-icing heater control circuits were not included in the analysis. The inspectors reviewed the associated DER, attended pertinent SORC meetings, reviewed the subsequent LER, and discussed related issues with NMPC personne Observations and Findinas The Unit 2 service water intake structure openings are equipped with bar rack heaters to eliminate the potential for frazil ice adhesion. Frazil ice formation can occar when the
- intake structure temperature drops near freezing. Therefore, TS require the heaters to ba operable whenever the intake tunnel water temperature is below 39 degrees F. As part of the corrective actions for LER 50-410/99-01, "NMP2 Outside the Design Basis l Due to Safe Shutdown Service Water Pump Bay Unit Coolers Being Out-of-Service,"
Unit 2 determined that a control room / relay room fire that renders the service water intake de-icing heaters inoperable, coincident with service water temperaturet that approach freezing, could lead to a complete loss of service water. This condition was not in accordance with the Safe Shutdown Analysis as described in the Unit 2 UFSAR, and this condition has existed since the initial operation of the plan Upon identification, NMPC established a fire watch for the control room / relay room fire l area, which will be in place until a design change to correct the deficiency is implemented. NMPC expects to complete the design change by November 30,1999. In L
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. addition, NMPC is continuing their review of the safe-shutdown capability as described in LER 50-410/99-01. The inspectors considered this actions to be appropriat NMPC Licensee Condition 2.G requires implementation of the fire protection program described in the UFSAR. The fire protection program includes an analysis of the ability to achieve safe-shutdown of the unit in the case of a control room fire. The failure to ensure that the service water intake de-icing capability is available during a control room fire, could impair the ability to achieve safe-shutdown, and is a violation of this licensee condition. This Severity Level IV violation is bein,q treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (NCV 50-410/99-04 06).
This violation is in the licensee's corrective action program as LER 50-410/99-0 The inspectors completed an on-site review of the LER and verified that it was completed in accordance with the requirements of 10CFR50.73. Specifically, the description and analysis of the event, as contained in the LER, were consistent with the inspectors' understanding of the event. The root cause and corrective and preventive actions as described in the LER were reasonable. This LER is close Conclusion NMPC self-identified and promptly corrected a condition which could have adversely affected the ability of the unit to achieve safe-rhutdown, involving the Unit 2 service water intake de-icing heater control circuits which were not protected against a control
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-room fire. This violation of License Condition 2.G was non-cite E (Closed) IFl 50-410/98-19-03: Leakage of contaminated water following a scram rese Draining evolutions of the scram discharge volume (SDV) have resulted in the contamination of a small area of the reactor building floor. The contamination is caused by steam condensing into und leaking from a ventilation duct to which the equipment drain cooler and the SDV vent piping are connected. NMPC compensated for this by ensuring that applicable procedures incorporate a plant announcement to stand clear of the area during SDV venting and draining. The NRC originally became aware of this issue in 1997 (NRC inspection report 97-11). Later, the NRC opened the IFl to condu::t further review of the issue.- The purposs of this inspection was to review the system design and to evaluate the safety implication The potential for the leakage was created in the 1994-1995 time frame, after the licensee implemented a design change that relocated the reactor building equipment drain header piping. Before the implementation of this change, the hot, pressurized drain lines from the reactor core isolation cooling and the residual heat removal steam condensing systems shared a common header with the cool, gravity equipment drain lines. This common header provided a path for the fluid from the pressurized sources to flash into steam, through the gravity drains, into the reactor building. The relocation of the piping to separate the high pressure drain header from the gravity drain header was successful in eliminating steam from the gravity drains. However, an associated change that tied the SDV vent piping into the equipment drain cooler vent line, effectively created a path
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of less resistance to a connected ventilation duct allowing leaks from the duct onto the floor during scram rese . The inspectors' review of the design implications determined that the small amount of contaminated water emitted from the ventilation duct was controlled and contained in accordance with approved site radiation protection procedures. The inspectors also determined that within the affected area there was no safety equipment which could be adversely impacted by the water spillage. Based on this review, the inspectors did not consider the small amount of water spillage to be safety significant but noted that, although the licensee had implemented several design changes in the equipment drain system during the last five years, they had not implemented an acceptable piping configuration that resolved the original system interaction and the venting contamination concems. The water spillage also constituted an operator work-around in that it required additional operator actions during the drain evolution of the SD NMPC had proposed and was evaluating attematives to eliminating the inadvertent contamination experienced during SDV draining. The proposed actions included the rerouting of the SDV vent line and the use of a check valve. Based on the licensee actively pursuing the resolution of this low safety significant issue, this item is close E (Closed) Violation 50-220/98-16-02: Failure to identify and promptly correct a condition adverse to quality. On August 20,1998, NMPC determined that they had failed to
- recognize, in 1996 and 1997, a low cooling water flow to the motor bearing of core spray pump No.122. The licensee reported the finding in LER 98-16. At the time of the finding, the NRC reviewed the issue and was satisfied that the licensee had developed a comprehensive corrective action plan to resolve the discrepancy. Nonetheless, the licensee's failure to evaluate and correct their finding was a violation of the Appendix B, corrective action progra During the current follow up review, the inspectors confirmed that the actions described in the LER and the subsequent engineering evaluation had been satisfactorily completed. Specifically, the inspectors verified that: (1) the repair of the affected pump had been satisfactorily completed; (2) the extent of condition had been addressed; (3)
the applicable maintenance procedures had been revised; and (4) the lesson leamed from the event had been discussed with responsible personnel from engineering, operations, and maintenance. This item is close I
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IV. Plant Support R1 Radiological Protection & Chemistry Controls R Unit i and Unit 2 Solid Radioactive Waste Manaaement and Transoortation of Radcactive Materials Inspection Scone (86750)
A selective review was performed to evaluate NMPC's basis for certifying that radioactive materials and wastes intended for disposal were properly classifed, described, packaged, marked and labeled and that radioactive material / waste shipments were made in accordance with applicable shipping regulations. Information was gathered by a review of the following: possession or access to applicable federal regulations; maintenance of licenses for facilities that radioactive materials or wastes were shipped; training records; procedural guidance and procedural maintenance; use of scaling factors to infer the concentration of difficult-to-measure radio nuclides; methods used to classify radioactive wastes; shipping records; and through interviews with cognizant personnel and tours through the plant, b.- Observations and Findinos A review of records verified that NMPC had ready access to up-to-date copies of federal o -
regulations including 4g CFR Parts 100-179 and 10 CFR Parts 20,61, and 71. A reviw of selected records verified that up-to.< late copies of licenses were maintained for facilities to which radioactive materials and wastes were shipped. A selected review of
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training reco.Vs showed that members of the Radwsste organization responsible for preparing radioactive waste shipments had received current training on regulations and computer programs used to preparc shipments. An interview with a training specialist ,
revealed that initial training and periodic training were provided on equipment and I processes that generate radioactive waste. A review of selected procedures verified that ;
detailed instructions were provided for personnel involved with the transfer, packaging j
- and transport of radioactive wastes. Procedural guidance was adequately maintained i and several procedures were in the process of being revised to incorporate recent changes to the regulations and to make needed improvement The Radweste group primarily used computer programs to c;assify and prepare radioactive material and waste shipments. A selected rev'mw of shipping records '
showed that waste classifications were accurate and no discrepancies were identife Shipping records were prepared in accordance with procedural guidance and included appropriate information such as radiation and contamination surveys, emergency ,
response information, and shippers certification of the adequacy of the shipmen ! Conclusions Radioactive material / waste management and transportation programs were effectively implemented as evidenced by use of up-to-date regulations and facility licenses,
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appropriately trained personnel, proper procedurd guidance and adequate maintenance of procedures, appropriate use of scaling factors to e. stimate isotopic content of radioactive material / waste packages, and proper shipping records.
R1.2 Refuel Outaae Radioloaical Controls (Unit 1) Insoection Scooe (83750)
A review was performed of radiological controls implemented for outage wor Information was gathered by a review of radiation exposure goals, selected licensee
. Initiatives to maintain radiation exposures as low as is reasonably achievable (ALARA),
through discussions with cognizant personnel, through a review of radiological posting practices, by a review of administrative controls for the Unit 1 Turbine Bu'Iding " Green Area," a review of high radiation area access controls, and a review of the following documents:
1. ALARA Review 9912, "Drywell in-Service inspections (ISI), Erosion / Corrosion (E/C)
Exams and Support Work" 2. ALARA Review 99-10, "Dewater and Desludge Torus, Replace ECCS Suction Strainers" 3. ALARA Review 99-06, " Replace Thirty-Seven Control Rod Drives (CRDs)"
4. ALARA Review 99-11, " Disassembly / Reassembly of Reactor Vessel, Fuel Movements and Decontamination" 5. ALARA Review 99-03, "Drywell - Repack Valves - All Elevations" 6. ALARA Review 99-17. "Drywell Floor Drain Sump and Associated Work" 7. GAP-RPP-08, Rev 5., " Control of High, Locked High, and Very High Radiation Areas" 8. S-RAP-RPP-0801, Rev. 8, "High Radiation Area Monitoring and Control" !
9. S-RAP-RPP-0103," Posting Radiological Areas" i 10.10CFR61.55 data for NMP1 dry active waste (DAW) ,
11. Safety Evaluation 96-102, " Safety Evaluation for Turbine Building Green Area"
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12. Deviation / Event Report No. 1-1999-1106, " Unlocked Source Storage Locker - Turbine i Building 261'- Condenser Bay," j 13.N1-RSP-1Q, " Accountability of Calibration and Check Sources at NMP Unit 1, 1 Rev. 0." Observations and Findinas An ALARA exposure goal of 280 person rem was set for RFO15. The majority of dose (more than 200 person-rom) was estimated to be received from drywell work. Significant dose jobs included 70 person-rem for drywell ISI,25 person-rem for torus desludging and ECCS suction strainer replacement,23 person-rem for CRD exchanges,20 person-rem for refuel floor activities,18 person-rem for drywell valve repacks, and 17 person-rem for miscellaneous drywell inspections. Interviews with cognizant personnel and
. reviews of documentation rovealed that the radiological controls organization was staffed with trained and qualified personnel and there was early involi'ement in planning forjobs with elevated exposure estimates, high dose rates, and radiobgically complex wor =
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ALARA reviews were thorough and provided details of applicable industry event Examples of radiological controls and ALARA/ initiatives included the following:
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Significant amounts of temporary shielding were installed in the Unit 1 drywell to reduce general area and job specific dose rates; a
Floor and equipment drain lines were flushed to reduce general area dose rates in the radwaste building, reactor building, and drywell and allowed the drywell to be down posted from a locked high radiation area to a high radiation area;
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Detailed component and equipment locations were included on drywell briefing maps;
- The torus was desludged with a back-flushable filter skid which reduced the need for filter handling and disposal; and
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Cameras were stationed in key work areas for remote monitoring of wor RFO15 Handbook An outage handbook was distributed to plant personnel to communicate important outage information such as responsible contacts and telephone numbers, meeting times, and plant maps. However, the inspector noted that the handbook included information regarding the expected response to an alarming dosimeter that was inconsistent with p procedural guidance and general employee training. The handbook stated that if an
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individual received a dose rate al arm on their alarming dosimetry they should move to a lower dose rate area. Procedural guidance and general employee training instructed personnel that if an electronic dosimeter alarm occurs, personnel shall leave the work area and report to radiological protection. NMPC issued a DER and distributed a correction to the RFO15 Handbook. No known examples of improper response to alarming dosimetry occurred as a result of this temporary incondstency and no violations of NRC requirements were identified.-
Radioloaical Boundaries Overall, radiological boundaries were clearly defined and posted. However, several ,
opportunities for improving radiological postings were identified on the Unit i refueling floor. For example, during reactor vessel disassembly, the drywell dome and the reactor I vessel head insulation were stored on the refueling floor. General contamination beneath these components was approximate!y 10,000 - 60,000 dpm/100 cm2 and areas on the bottora of the insulation package had levels up to 24 mrad /hr/100 cm2. Access to
' these components was restricted witn several signs indicating " Contact Health Physics Prior to Entry." During a tour of the refuel floor, severallocations that provided access to
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areas beneath the drywell domq and reactor head insulation did not have readily-mm , L_.m observable radiological postings. Upon notification by the inspector, health physics staff ;
members immediately posted the identified areas. The improvements in radiological c posting were considered an enhancement to existing postings and no violations of NRC requirements were identifie ,, n
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Administrative Controls For the Unit 1 Turbine Buildina " Green Area" A clean area (" Green Area") was setup within the radiologically controlled area (RCA) on Unit 1 Turbine Building 305 foot elevation to allow turbine building workers to eat, drink, and use senitation facilities. A safety evaluation was used as the basis for establishing the area and administrative controls for setup, use, and health physics oversight of the facility were included in procedure S-RAP-RPP-0103, Rev. 9 " Posting Radiological Area." Controls for" Green Area" access included continuous health physics oversight; requirements and provisions for contamination monitoring of hands, feet, and face prior to entry; frequent contamination and radiation surveys; and special provisions for the transport of food and other clean items within the RC The radiation protection manager acknowledged that personal contamination surveys for
" Green Area" access did not include a whole-body frisk. However, he pointed out that the risk for the spread of contamination and potential ingestion of significant amounts of contamination were minimized by limiting use of the facility to individuals that were briefed on use of the facility and who worked in areas with relatively low contamination levels and lower risk of hot particles (i.e., turbine building workers); contamination monitoring was performed for body parts with the highest probability of contamination (bands, feet, and face); and the health physics staff maintained close oversight and monitoring of the facility. Additionally, all RCA workers were required to receive a whole-body frisk prior to exiting the RCA and experience gained during the previous outage damonstrated that radiological controls implemented for the " Green Area" were effective in controlling contaminatio Hiah Radiation Area Access Controls Controls for high radiation area access included detailed procedural guidance; radiological postings; frequent use of locked doors and required use of locked access controls for areas that could result in an individual receiving a dose equivalent in excess of 1000 mrom per hour at 30 centimeters; radiation work permit (RWP) controls; use of
" alarming" dosimetry; requirements for a minimum available exposure for access; use of remote door alarms; and increased health physics oversight and monitoring. Two health I physics technicians and two health physics supervisors selected for interviews demonstrated thorough knowledge of high radiation area controls. Tours of the plant confirmed that high radiation and locked high radiation areas were appropriately posted and doors that were required to be locked were found locked or appropriately controlled by health physics staff. A review of DERs for t'ne last year showed no negative trends in high radiation area postings or high radiation area access control : DER 1-1999-1106 dated April 15,1999, was written to document and investigate the (discovery of an unlocked radioactive source cabinet by a health physics supervisor
' during a routine tour. The source cabinet was used to store radioactive sources used for m .. instrument source checks and was posted as a high radiation area. The maximum dose rate found in the cabinet was 80 mrem per hour at 30 cm from a source. Upon
- identification, a source inventory was conducted which showed that all sources were accounted for and the cabinet was locked. Actions taken to identify and correct the t
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deficient condition and to place the issue into the corrective action system were appropriate. The failure to maintain the posted high radiation area storage cabinet locked was contrary to radioactive source storage requirements in N1-RSP-1Q, J
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' Accountability of Calibration and Check Sources at NMP Unit 1," Rev. O. This violation is minor in nature and is not subject to formal enforcement action, Conclusions Radiological controls for the Unit i refuel outage were effectively planned and l implemented and focused on jobs with elevated exposure estimates, high dose rates, -
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and radiologically cornplex wor Radiological posting practices for access to radiation areas, high radiation areas, and >
airborne radioactivity areas were effective as evidenced by well defined boundaries and clear radiological postings. Some opportunities to enhance informational postings on the refuel floor that required " health physics notification prior to entry" beneath the drywell
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dome and reactor head insulation were identifie l Contamination monitoring requirements for access to the turbine building " Green Area" (clean area within the radiologically controlled area {RCA)) did not include an entire whole body frisk similar to the requirements for RCA exit. However, they were acceptable to minimize the risk for the spread and ingestion of significant amounts of radioactive contamination based on use of detailed procedures, restrictions on the personnel that could use the facility, and close health physics oversigh Effective high radiation area controls were implemented as evidenced by clear radiological postings, use of locked doors when required, use of " Alarming" dosimetry, use of radiation work permits, use of remote door alarms, requirements for a minimum available exposure for access, and increased health physics oversight and monitoring for high radiation area entr R2 Status of RP&C Facilities and Equipment R Radiolooical Housekeeoino (Units 1 and 2) Insoection Scooe (83750 and 86750)
Plant tours were conducted to evaluate housekeeping and cleanliness and material conditions. Information was gathered through tours of Unit 1 and Unit 2 radwaste buildings and through the Unit i drywell, reactor and turbine building, Observations and Findings Housekeeping practices were effective as evidenced by c'sar isles and walkways, neatly stored tools and equipment, and well illuminated work areas. Material condition for the reactor and radwaste buildings were generally very good with painted floor and wall surface .
28 Conclusion Material conditions were good and housekeeping practices were effective as evidenced by clear aisles and walkways, neatly stored tools and equipment, and painted floor and wall surface R7 Quality Assurance in RP&C Activities R7.1 Radioloalcal Control Proaram Corrective Actions (Units 1 and 2) Insoection Scooe (88750)
A review was performed of the use of audits, appraisals, and DERs for the identification and resolution of deficiencies in the area of radiological controls and radwaste management and transportation. Information was gathered through discussions with cognizant personnel and selected reviews of quarterly self-assessments, audits, and DER I Observations:and Findinas A combination of self-assessments and quality assurance audits were performed to meet the requirements for an annual review of radiation protection program content and implementation as required by 10CFR20.1101(c). Quarterly self-assessments were _
performed at each unit to identify trends in program areas. Self-assessments included a review of radiation exposure, radiological safety indicators, and radiation worker ,
performance. Semi-annual self-assessments in radiological controls were performed to review common elements of Unit 1 and Unit 2 radiological controls programs. Quality assurance audits of radiological controls and radwaste management were often performed with the assistance of industry peers. Deficiencies arising from self-assessments and audits were addressed through the DER system and opportunities for
- improvement were maintained on an " Action item List" for evaluation and revie Two significant deficiencies were selected to evaluate the effecu uness of corrective actions:
DER 1-97-0762 dated March 16,1997, was written during the previous refuel outage at l NMP1 (RFO14) after an irradiated core component (tie-rod mid-support piece) was placed on the reactor cavity seal plate which resulted in elevated dose rates and an l audible indication of increased dose rates on the upper drywell radiation monitor syste l Two individuals who were or may have been in the upper drywell at the time of the event were alerted by the audible indication of increased dose rates (increased chirping rate)
and exited the drywell without receiving a significant exposure. Dose rates on the mid- ,
support piece were subsequently measured to be 200 rem per hour on contac !
- . . . . Immediate corrective actions included restriction of personnel access to the upper l elevations of the drywell; relocation of the mid-support piece to the equipment pit; and 1 issuance of a stop work order on the fuel floor. A root cause analysis identified multiple barrier failures which resulted in the event including the failure to recognize and
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communicate that the mid-support piece was an irradiated core componen Preventative actions included revision of multiple procedures to include appropriate communications and radiological controls for restriction of personal access during movement of the mid-support piece and other irradiated core component Radiological control barriers that were in place at the drywell included 1) use of electronic alarming dosimetry,2) use of a remote radietion monitoring system with detector probes in the upper drywell, and 3) the health physics staff had a policy to evacuate the upper drywellif elevated radiation readings / alarms were received by the upper drywell radiation monitoring system. Total dose to the individuals electronic dosimetry for the drywell entry was 10.7 mrem and 29.5 mrem and the maximum dose rates measured by the worker's electronic dosimetry was 145.9 mrem per hour and 174.9 mrem per hour. These readings were typical for routine drywell entries. Maximum dose rates measures by the radiation monitoring system averaged 159.3 mrem per hou Conservative exposure calculations estimated that the maximum dose rate to the head of an individual standing in the upper drywell would have been 978 mrem per hou A selected review of procedures verified that appropriate revisions had been made and
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interviews with cognizant personnel confirmed that personnel were knowledgeable of the event and actions to take for the movement of irradiated core component Technical Specification 6.11, " Radiation Protection Program" states that " procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 CFR 20 and shall be approved, maintained and adhered to for all operations involving personnelradiation exposure." Procedure S-RAP-RPP-0801,"High Radiation Area Monitoring and Centrol" step 3.8.1 required the drywell 259 foot elevation ladder to be in place and locked to control access to the upper e:evations of the drywell during the movement of an irradiated coro component. Contrary to this requirement access restrictions to the upper elevations of the drywell were not maintained on March 15,1997 during movement of a tie rod mid-support piece. This Severity LevelIV violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy (NCV 50-220199-04-07). This violation was in the licensee's corrective action program as DER 1-97-076 DER 1-97-1346, dated April 29,1997, was also written during the previous refuel outage at Unit 1 after irradiated fuel was placed in the 198 cell rack adjacent to the spent fuel pool gate, resulting in unexpected elevated dose rates after drain down of 70 Rem per hour at contact with the bottom of the gate and 700 mrem per hour immediately above the spent fuel pool gates. Access to areas with elevated dose rates were appropriately controlled prior to and after the discovery of the elevated dose rates, no uncontrolled exposures resulted from the event, corrective and preventative actions were determined to be appropriate and no violations of NRC requirements were identified. An apparent cause evaluation determined that this event occurred because of a mis-communication
.or understanding between engineering and reactor engineering personnel regarding precautions for moving spent fuel within six feet of the spent fuel gate. Additionally, there were no procedures which specifically prevented the storage of fresh spent fuel adjacent to the spent fuel gate. Immediate corrective actions included restriction of access to the
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communicate that the mid-support piece was an irradiated core componen Preventative actions included revision of multiple procedures to include appropriate communications and radiological controls for restriction of personal access during
' movement of the mid-support piece and other irradiated core component Radiological control barriers that were in place at the drywell included 1) use of electronic a! arming dosimetry,2) use of a remote radiation monitoring system with detector probes in the upper drywell, and 3) the health physics staff had a policy to evacuate the upper drywell if elevated radiation readings / alarms were received by the upper drywell radiation monitoring system. Total dose to the individuals electronic dosimetry for the drywell entry was 10.7 mrem and 29.5 mrem and the maximum dose rates measured by the worker's electronic dosimetry was 145.9 mrem per hour and 174.9 mrem per hour. These readings were typical for routine drywell entries. Maximum dose rates measures by the radiation monitoring system averaged 159.3 mrem per hou Conservative exposure calculations estimated that the maximum dose rate to the head of an individual standing in the upper drywell would have been 978 mrem per hour.
I A selected review of procedures verified that appropriate revisions had been made and j interviews with cognizant personnel confirmed that personnel were knowledgeable of the event and actions to take for the movement of irradiated core component Technical Specification 6.11, " Radiation Protection Program" states that " procedures for personnel radiation protection shall be prepared consistent with the requirements of 10 ~
,> CFR 20 and shall be approved, maintained and adhered to for all operations involving personnel radiation exposure." Procedure S-RAP-RPP-0801, "High Radiation Area Monitoring and Control" step 3.8.1 required the drywell 259 foot elevation ladder to be in 1
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place and locked to control access to the upper elevations of the drywell during the movement of an irradiated core component. Contrary to this requirement access restrictions to the upper elevations of the drywell were not maintained on March 15,1997 during movement of a tie rod mid-support piece. This Severity Level IV violation is being i treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement j Policy (NCV 60-220/99-04-07). This violation was in the licensee's corrective action j program as DER 1-97-076 l l
DER 1-97-1346, dated April 29,1997, was also written during the previous refuel outage at Unit 1 after irradiated fuel was placed in the 198 cell rack adjacent to the spent fuel pool gate, resulting in unexpected elevated dose rates after drain down of 70 Rem per hour at contact with the bottom of the gate and 700 mrem per hour immediately above the spent fuel pool gates. Access to areas with elevated dose rates were appropriately controlled prior to and after the discovery of the elevated dose rates, no uncontrolled exposures resulted from the event, corrective and preventative actions were determined to be appropriate and no violations of NRC requirements were identified. An apparent cause evaluation determined that this event occurred because of a mis-communication
.or understanding between engineering and reactor engineering personnel regarding precautions for moving spent fuel within six feet of the spent fuel gate. Additionally, there were no procedures which specifically prevented the storage of fresh spent fuel adjacent to the spent fuel gate. Immediate corrective actions included restriction of access to the
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area, establishing the area above the sper* fuel ste as a high radiation area, and I flooding the area between the spent fuel pt. a .ss which dropped dose rates to 7 mrem per hour. Preventive actions included revisic. : of procedures to require health physics notification prior to movement of the lower spent fuel gate shield plug (No. 9); to require shield plug No. 9 to be in-place for cavity work or to control the area as a locked high radiation area when the shield plug was not in-place; and a revision of fuel handling i procedures to limit the storage of fuel in the 198 cell to only new fuel or spent fuel greater )
than one year old. A selected review of procedures during the inspection verified that '
appropriate revisions had been made and interviews confirmed that health physics personnel assigned to the refuel floor were knowledgeable of the event and actions to take when the lower spent fuel gate was move Conclusions Self-assessments, auoits, and the deficiency / event reporting system were effectively used to identify, evaluate, and resolve radiological control issues as evidenced by the j conduct of multiple self-assessments and audits to satisfy the radiation protection ;
program review requirements in 10CFR20.1101(c). Appropriate corrective actions and controls to prevent unplanned exposures were implemented as a result of previous deficiencie One non-cited violation was identified associated with the failure to maintain access restrictions to the upper elevation of the drywell during movement of an irradiated core
. component on March 15,199 V. Manaaement Meetinas X1 Exit Meeting Summary
! The inspectors presented the inspection results to members of the licensee management at the conclusion of the inspection on May 27,1999. The licensee acknowledged the findings presente ( . s-ATTACHMENT 1 PARTIAL LIST OF PERSONS CONTACTED Niagara Mohawk Power Corporation D. Bosnic ' Manager, Operations, Unit Two S. Doty Manager, Maintenance, Unit One
- N. Paleologos - Plant Manager, Unit Two -
F. Fox Acting Manager, Maintenance, Unit Two R. Smith : Plat Manager, Unit One N. Rademacher Manager, Quality Assurance D. Topley - Manager, Operations, Unit One ,
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INSPECTIOMRQCEDUJtES USED
. IP 37550 Engineering IP 37551 . On-Site Engineering .
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IP 61726 Surveillance Observations iP 62707 Maintenance Observations IP 71707 Plant Operations IP 71750 Plant Support '
IP 73753 - Inservice inspection IP 83750 Occupational Radiation Exposure _
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IP 86750 . Solid Radwaste Management and Transportation of Materials
- 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'92904 Followup - Plant Support
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ITEMS OPENED. CLOSED. AND UPDATED j OPENED l
.50-410/99-04-01 NCV Automatic Depressurizaton System Nitrogen Leakage in Excess l of NMP2 Technical Specifications Surveillance Limit ]
50-410/99-04-02 NCV Unit 2 Reactor Core isolation Cooling System Failure During Reactor Scram Transien ,50-220/99-04-03 IFl Computer Security for 3D Monicor /99-04-04 NCV ' Failure to Maintain Core Thermal Limits as Required by Technical SpecWication . .
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Attachment 1 2
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50-220/99-04-05 NCV Failure to Conduct the Required ASME Code inspection /99-04-06 'NCV Failure to Ensure that the Service Water intake De-Icing Capability is Available During a Control Room Fir /99-04-07 'NCV Radiological Control Program Correction Action CLOSED 50-410/99-04-01 NCV Automatic Depressurization System Nitrogen Leakage in Excess of NMP2 Technical Specifications Surveillance Limit /99-04-02 NCV Unit 2 Reactor Core Isolation Cooling System Failure Dudng Reactor Scram Transien /98-05-02 VIO Failure to Conduct Surveillance Test on Batterie /98-02-05 VIO Inadequate Plant Impact in Work Order Packag /99-04-04 NCV Failure to Maintain Core Thermal Limits as Required by Technical Specification /99-04-05 NCV Failure to Conduct the Required ASME Code Inspection /98-05-03 VIO Failure to Perform Adequate Design for Emergency Diese!
Generator Modification on Fuel Lin /99-04-06 NCV Failure to Ensure that the Service Water intake De-icing Capability is Available During a Control Room Fir /98-19-03 IFl Leakage of Contaminated Water Following a Scram Rose /98-16-02 VIO Failure to identify and Promptly Correct a Condition Adverse to Qualit /99-04-07 NCV Radiological Control Program Correction Action '50-410/98-26 LER Seismic Monitor Inoperable for More than Thirty Days and Special Reports Not Submitte !99-03 LER ADS Nitrogen Leakage in Excess of Unit 2 Technical Speedications Surveillance Limit /99-03- LER . NMP1 Thermal Limit Exceeded the Requirements of Technical Specification P.- , ,
Attachment 1 3 50-220/98-19 LER Missed ASME Section XI Inservice Inspection Due to Cognitive Erro /99-02 LER Missed Technical Specification Channel Functional Test of the Recirculation Flow Upscale Rod Bloc {
50410/99-04 LER NMP2 Service Water intake De-Icing Heater Control Circuits do Not Meet Fire Protection Program Requirement ;
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Attachment 1 4 LIST OF ACRONYMS USED AC Attemating Current ADS Automatic Depressurization System ALARA As Low As is Reasonably Achievable APLHGR Average Planar Heat Generation Rate APRM Average Power Range Monitor ASME American Society of Mechanical Engineers BWRVIP Boiling Water Reactor Vessel and Internals Project CDF Core Damage Frequency CFR Code of Federal Regulations CRD Control Rod Drives DAW Dry Active Wastes DC Direct Current I DCP Design Change Package l DER Deviation / Event Report DOT Department of Transportation l dpm Disintegration Per Minute l EC Emergency Condenser
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E/C Erosion / Corrosion ,
ECCS Emergency Core Cooling System i ECS Emergency Cooling System EDG Emergency DieselGenerators EP Emergency Preparedness EPA Electric Protection Assemblies ESA Engineering Supporting Analysis ESF Engineered Safeguards Feature ESL Equipment Status Log FCV Flow Control Valve FWBP Feedwater Booster Pump GAP Generation Administration Procedure HPCI High Pressure Core injection HPCS High Pressure Core Spray IFl Inspector Followup Item IGSCC Intergranular Stress Corrosion Cracking IR inspection Report ISEG Independent Safety Engineering Group ISI in-Service Inspection LC Limiting Condition for Operation LER Licensee Event Report j LOCA Loss of Coolant Accident LPRM Local Power Range Monitor LSFT Logic System Functional Test i
'APRM . Local Power Range NCV Non Cited Violation i NDE Nondestructive Examination NMPC Nine Mile Point Corporation
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Attachment 1 5
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NRC Nuclear Regulatory Commission PRA Probability Risk Analysis
'PRNM Power Range Neutron Monitor QA Quality Assurance RCA Radiological Controllec' Area RCIC . Reactor Core isolation Cooling RFO15 Refueling Outage RPS- Reactor Protection System RWP Radiation work Permit-SDV Scram Discharge Valve SORC Station Operating Review Committee SRV Safety Relief Valve SSS Station Shift Supervisor TIP Traverse Incore Probe TS Technical Specification USAR Updated Safety Analysis Report
. Unit 1 - Nine Mile Point Unit 1 Unit 2 Nine Mile Point Unit 2 j UPS Uninterruptible Power Supply
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