IR 05000334/1999003
| ML20209D834 | |
| Person / Time | |
|---|---|
| Site: | Beaver Valley |
| Issue date: | 07/07/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20209D825 | List: |
| References | |
| 50-334-99-03, 50-334-99-3, 50-412-99-03, 50-412-99-3, NUDOCS 9907140050 | |
| Download: ML20209D834 (31) | |
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U. S. NUCLEAR REGULATORY COMMISSION REGION l (
License Nos.- DPR-66, NPF-73 Report No /99-03,50-412/99-03 Docket No ,50-412 l
Licensee: Duquesne Light Company Post Office Box 4 Shippingport, PA 15077 l
Facility: Beaver Valley Power Station, Units 1 and 2 l
Inspection Period: May 2,1999 through June 12,1999 l
Inspectors: D. Kern, Senior Resident inspector G. Dentel, Resident inspector G. Wertz, Resident inspector
,. J. Fuila, Senior Radiation Specialist J. Jang, Senior Radiation Specialist T. Hipschman, Resident inspector, Oyster Creek Approved by: N. Perry, Acting Chief Reactor Projects Branch 7 l
9907140050 990707
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EXECUTIVE SUMMARY Beaver Valley Power Station, Units 1 & 2 NRC Inspection Report 50-334/99-03 & 50-412/99-03 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 6-week period of resident inspection. In addition, it includes the results of announced inspections by regional radiation specialists and a Year 2000 readiness assessmen Operations The Onsite Safety Committee effectively recognized, reviewed, and evaluated plant changes ~affecting nuclear safety. The meetings were well organized as the meeting agenda and items under review were consistently distributed and re, viewed in advanc This allowed the committee members to review a large volume of items and focus on the items with greatest safety significance. (Section 07.1) The licensee properly evaluated and reported the January 1999 Unit 1 degraded condenser vacuum reactor trip event. Corrective actions were appropriate and engineers performed a detailed assessment which correctly elevated the circulating water system into Maintenance Rule category (a)(1) performance monitoring. (Section 08.1) Operators failed to properly evaluate a source range nuclear instrumentation surveillance test. Corrective actions were appropriate. (Section 08.2)
Maintenance Surveillance testing was performed safely, with appropriate supervieory attention, and in accordance with proper procedures. . System er'gineers and mainMnance person .el coordinated effectively to investigate on unexpected 2-4 vital bus invedtr trensfer to its attemate power supply. (Section M1.1). On-line maintenance was managed consistent with equipment availabihty assumptions containeo in the Beaver Valley Unit 1 & 2 Probabilistic Risk Assessments. The on-line maintenance procedure was comprehensive and well understood. Incorporation of maintenance ru!e insights was a strength. Work week managers actively tracked and !
communicated job status to the Nuclear shift Supenisor (NSS), which support scund !
decision making with respect to configuration control. Occasional performance j deficiencies, such as fJSS authorization of work without recognizing system operability ;
relationships were property addressed through the condition repoit system. (Section l
M1.2) 1 1 In response to an NRC viGation, the licensee took strong actions to reiterate station l
, policy for safe operation and maintenance of plant equipment. Training was effcetive ii
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Enclosure 1 and lessons leamed from the violation were stressed during pre-evolution briefings for subsequent auxiliary feedwater pump tests. (Section M8.2)
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In one instance, corrective maintenance for a degraded river water pump was not performed in a timely manner. The delay reflected communications and scheduling deficiencies for time sensitive recommended maintenance. (Section E1.1)
Enaineerina
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System engineers recently performed good trending of Unit 1 river water pump performance. Performance data was carefully reviewed and used to determine pump maintenance frequency. (Section E1.1)
Controls for the receipt, storage, and handling of safety-related equipment and material were being effectively maintained. A minor weakness in the ownership and control of materials being staged to wpport the 12-week schedule was identified and captured in Jhe comectivw action program. (Section E2.1)
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The nuclear engineering salf assessments rsviewed satisfied administrative requirements at.d confirmed expected results. However, the assessments did not provide substantiat recommendations or corrective actions. (Section E7.1)
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Engineers identified and corrected several nonconservative assumptions used in doss assessment calculations for design basis accidents (DBA). For two DBA types, the errcrs could have permitted control roorn operator radiological dose to exceed regulatory requirements. Extent of condition causal analysis, basis for continued operation evaluations, and 10 CFR 21 reporting were corrprehensive. (Section EB.1)
Plant Suooort
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A generally effective program for maintaining occupational exposures as low as is reasonably achievable (Al. ARA) has been established. Management involvement in the !
Al. ARA program, incorporation of ALARA principles !nto plant mod;fications, and
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establishment of exposure goals were appropriate. (Section R1.1)
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Radioactive liquu and gaseous effluent control programs were effective. The Offsite Dose Calculation Manuai contained sufficient specification and instruction to acceptably implement and maintain the radioactive liquid and gaseous e.Tuent control program (R1.2j .
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The licensee maintained an adequate program for effluent radiation monitoring system calibration. (R2.1)
.. ~ .The licensee established, implemented, and maintained an effective ventilation system surveillance program with respect to charcoal adsorption surveillance tests, h!gh lii i
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Enclosure 1 efficiency particulate and charcoal filters mechanical efficiency tests, and air flow rate tests. (Section R2.2)
. An effective training program for ALARA has been implemented for both radiological workers and radiation protection technicians. Specific training both as part of initial and continuing training has been established. (Section R5.1)
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The Quality Assurance program for the effluent control program was effectively implemented. Quality Control programs to validate radioactive liquid and gaseous effluent control program analytical results were effective. (Section R7.1)
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TABLE OF CONTENTS Page EXEC UTIVE SU MMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1i TAB LE OF CONTE NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v 1. Operation s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01 Conduct of 0perations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 General Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 02 Operational Status of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . . 1 O Engineered Safety Feature System Walkdowns . . . . . . . . . . . . . . . . . . 1 07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 0 Onsite Safety Committee . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 08 ~ Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2
- O8.1 (Closed) Licensee Event Report (LER) 50-334/99-01 . . . . . . . . . . . . 2 08.2 (Closed) LER 50-412/99-01 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 s 08.3 (Closed) LER 50-412/99-0 5 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 i i
l 11. M ainten a nce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 i M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 M1.1 Routine Surveillance and Maintenance Observations . . . . . . . . . . . . . 4 M1.2 On-Line M aintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 M8 Miscellaneous Maintenance issues . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 M8.1 (Closed) LER 50-334/98-15-01 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 M8.2 (Closed) Violation 50-334/98-04-01 ...........................7 lli. Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 E Unit 1 River Water Pump Lift Adjustments . .. . . . . . . . . . . . . . . . . . . . . 8 E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . 9 E Receipt, Garage and Handling d Safety Related Equipment and M ateria ls . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 E7 Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 E7.1 Nuclear Engineering Department Self-Assessments . . . . . . . . . . . . . 11
<scellaneous Engineering Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 (Closed) LER 50-334(412)/99-02 . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 E0.2 Year 2000 Computer System Readiness / NRC Temporary Instruction 2 515/141 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 IV. Plant S upport . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 R1.1_ Radiation Exposure Management . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 R1.2 Implementation of the Radioactive Liquid and Gaseous Effluent Control Prog ram s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16 R2 Status of RP&C Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 R2.1. Calibration of Effluent / Process / Area Radiation Monitoring Systems, Flow j
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Rate Measurement Devices, and Oxygen Monitors . . . . . . . . . . . . . . 18 v
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! R2.2 Air Cleaning Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 f R5 Staff Training & Qualifications in RP&C Activities . . . . . . . . . . . . . . . . . . . . . 20 R5.1 Radiation Exposure Contro: Training . . . . . . . . . . . . . . . . . . . . . . . . . . 20 R7 Quality Assurance in Radiological Protection and Chemistry Activities . . . . . 21 R Quality Controls for Radiological Effluent Monitoring . .... ..... 21 V. Management Meetings ..................................................21 X1 Exit Meeting Summa ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 X2 N RC Management Site Visit . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 ITEMS OPENED, CLOSED AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 LIST OF ACRONYMS U S ED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 24 l
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Report Details Summary of Plant Status i Unit 1 began this inspection period in Mode 3 (hot standby) . On May 7, the unit was
- synchronized to the offsite power transmission grid following a six day forced outage to recondition the main unit generator hydrogen seal oil system. The unit reached full power operation on May 1 Unit 2 began this inspection period at 100 percent power and remained at or near full power throughout the inspection perio l. Operations I
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01.1 General Comments (71707)
Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations. - In general, the conduct of operations was professional and safety-conscious; specific events and noteworthy observations are detailed in the sections belo Operational Status of Facilities and Equipment O2.1 Enaineered Safety Feature System Walkdowns (71707)
~The inspectors walked down accessible portions of selected systems to assess equipment operability, material condition, and housekeeping. Improvements wera noted in the material condition of the intake structure pump cubicles. Minor discrepancies were brought to Nuclear Shift Supervisor's (NSS) attention and corrected. No substantive concems were identified. The following systems were walked down:
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Unit 1 River Water !
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Unit 2 Service Water ]
07 Quality Assurance in Operations 0 Onsite Safety Committee
' Inspection Scooe (71707)
The inspectors observed the Onsite Safety Committee (OSC), focusing on the effectiveness of the group to recognize, review, and evaluate plant changes affecting nuclear safety. The inspectors reviewed the administrative requirements for the OSC and its membership, attended several OSC meetings, and reviewed over 50 items presented to the OS .
2 Observations and Findinas Guidance defining the organization, responsibilities, and administrative requirements for
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the OSC was clearly and effectively provided by Nuclear Power Division Administrative Procedure (NPDAP) 8.10, "Onsite Safety Committee," Rev. 6. Plant activities requiring OSC review, OSC member composition, and documentation of meeting records were listed in the OSC procedure. The OSC membership included representatives from various departments including Operations, Maintenance, Engineering, Health Physics, and Chemistry. The diverse membership provided a knowledgeable and balanced group review. The training records of the members were curren The meetings observed by the inspectors were well planned and organized. An electronic agenda was available in advance of the meeting. The meeting items were reviewed by the OSC members in advance. This enhanced the review process and allowed the OSC members to place more attention and discussion on items with greater nuclear safety implications. The group discussions were effectiva in resolving the issues raised. No schedule pressure to finish a review and move on to the next item was observed. The committee effectively recognized, reviewed, and evaluated plant changes affecting nuclear safet The procedure changes for OSC review were generally of good quality. In most cases, the procedure presenters were cognizant of the changes and effectively communicated them to the committee. One minor deficiency noted by the inspectors was that some of the procedure changes presented to the OSC consisted of poor editorial quality and were difficult to read and understand. The OSC properly rejected proposed procedure changes if the committee could not adequately determine its safety relevance due to the poor editorial qualit Conclusions The Onsite Safety Committee effectively recognized, reviewed, and evaluated plant changes affecting nuclear safety. The meetings were well organized as the meeting agenda and items under review were consistently distributed and reviewed in advanc This allowed the committee members to review a large volume of items and focus on the items with greatest safety significanc Miscellaneous Operations issues (90712)
08.1 (Closed) Licensee Event Report (LER) 50-334/99-01: Manual Reactor Trip Due to Continuing Degradation of Main Condenser Parameters Insoection Scope (62707. 92700)
The inspectors performed an on-site review of the LER and a maintenance rule assessment of equipment performance to determine whether the licensee had properly evaluated the even )
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3 Observations and Findinas This event was previously documented in detail in NRC Inm; rated Inspection Report Nos. 50-334(412)/98-11. The LER described the event a wrately, including ceusal assessment and planned corrective actions. The inspectors independently verified that the corrective actions were properly completed or scheduled for completion within a reasonable time fram The inspectors reviewed this event with respect to system pe~tformance monitoring required by 10 CFR 50.65. Degraded main condenser performance and improper maintenance had caused main condenser vacuum to decrease rapidly. Operators manually tripped the reactor in anticipation of an automatic turbine trip and reactor trip due to low condenser vacuum. Engineers categorized this event as a maintenance preventable functional failure (MPFF) of the circulating water (CW) system. The reactor trip and subsequent repair time caused plant level performance criteria to be exceeded for Unplanned Capability Loss Factor. The Maintenance Rule Steering Committee elevated the CW system to Maintenance Rule category (a)(1), which requires enhanced performance monitoring, establishment of specific performance goals, and appropriate actions to ensure the system's performance improves. The inspectors independently reviewed the MPFF evaluation form and category (a)(1) justification and determined that engineers had properly evaluated system performance during this even Conclusions The licensee properly evaluated and reported the January 1999 Unit 1 degraded condenser vacuum reactor trip event. Corrective actions were appropriate and engineers performed a detailed assessment, which correctly elevated the circulating water system into Maintenance Rule category (a)(1) performance monitoring.
08.2 (Closed) LER 50-412/99-01: Failure to Comply with Technical Specifications Due to Not Meeting the Acceptance Criteria for a Source Range Monitor During Surveillance Testing Insoection Scope (92700)
The inspectors performed an onsite review of the LER. The inspectors examined a sample of corrective actions and reviewed the description of the LE Observations and Findinas On February 27, Unit 2 was in Mode 3, beginning the seventh refueling outage. The reactor trip breakers were open, and all control rods were fully inserted. Operators performed testing on the source range nuclear instrumentation (SRNI). The testing was sequenced to ensure one instrument remained operable to comply with technical specifications (TS). Due to human performance errors by the reactor operator and the NSS and procedural weaknesses, the N-32 SRNI was incorrectly declared operable after testing. Operators continued the test on the N-31 SRNI. This resulted in both channels being inoperable at the same time. The safety significance of this event was negligibl The SRNls were being used for indication only and both channels, although inoperable, were functionally available and would have indicated a change in reactivity. Failure to take the actions required when both channels were inoperable was a violation of TS 3.3.1.1, Table 3.3-1, item 6.b. This failure constitutes a violation of minor significance and is not subject to formal enforcement action. Corrective actions to this event and other issues resulted in human performance improvement during the Unit 2 refueling outage (see NRC Integrated Inspection Report Nos. 50-334(412)/99-01 and 99-02). Conclusions Operators failed to properly evaluate a source range nuclear instrumentation surveillance test. Corrective actions were appropriate.
08.3 (Closed) LER 50-412/99-05: 4kV-2A Bus Trip on Ground Overcurrent Relay 51-VA207X The inspectors performed an in-office review of the LER. This event was evaluated and ,
documented in NRC Integrated inspection Report 50-334(412)/99-02. The LER adequately described the actual event. However, the LER was deficient in that it did not describe the additional problems associated with the recovery of the offsite power source. Specifically, the NSS made a poor configuration control decision when he secured the emergency diesel generator (EDG) prior to isolating the degraded 2-5 battery charger from the battery bus. Although permitted by TS, this configuration increased the likelihood of a repeat emergency 4 kilovolt bus trip and EDG start. These issues were properly documented and pursued through the licensee's corrective action progra II. Maintenance M1 - Conduct of Maintenance j i
M1.1 Routine Surveillance and Maintenance Observations (61726. 62707)
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The inspectors observed selected surveillance tests and maintenance task Operstional surveillance tests (OST) and work orders (WO) reviewed and observed by the inspectors are listed belo ST-3 " Reactor Plant River Water Pump 1 A Test," Rev.17
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2OST-2 " Main Steam Isolation Valve [2 MSS *AOV101 A] Partial Closure Test," Rev. 2
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10ST-6.2A " Reactor Coolant Water Inventory Balance," Rev.11
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WO 99-209207 2-4 Vital Bus Uninterruptible Power Supply
.The surveillance testing was performed safely and in accordance with proper procedures. The inspectors noted that an appropriate level of supervisory attention was given to the testing, depending on its risk significance. System engineers and
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maintenance personnel coordinated effectively to investigate an unexpected 2-4 vital bus inverter transfer to its alternate power suppl M1.2 On-Line Maintenance Inspection Scope (71707. 62707)
The inspectors reviewed procedures, scheduling, and performance of on-line maintenance activities to evaluate whether overall plant risk insights were properly considere Observations and Findinas The inspectors reviewed NPDAP 7.12, "Non-Outage Planning, Scheduling, and Risk Assessment," Rev. 6. The procedure was comprehensive and its incorporation of maintenance rule insights was a strength. Specific duties and coordination responsibilities between departments (Work Control, Nuclear Engineering Department, System and Performance Engineering, Maintenance Production Unit, and Operations)
were clearly identified. Separate attachments which provided system specific safety guidelines and highlighted non-technical specification and non-safety related equipment, used in emergency operating procedures, made the procedure easier to us Based on various interviews and field observations, the inspectors determined that station personnel were knowledgeably of NPDAP 7.12 requirements, including limitations for using the procedure's System Combination Matrix for emergent work scope risk assessment. Risk analysts made good use of plant specific probabilistic safety analysis (PSA) models to assess changes in core damage frequency (CDF) associated with planned and emergent work activities. In January 1999, station management established plant performance indicators to track overall monthly CDF associated with on-line maintenance activities. Based on these performance indicators and the method used to track actual equipment unavailability hours, the inspectors determined that on-line maintenance was being managed consistent with the equipment availability assumptions centained in the Beaver Valley Unit 1 & 2 Probabilistic Risk Assessment The inspectors reviewed several on-line maintenance activities performed during this inspection period. Risk assessment for planned maintenance was typically performed accurately and published in the Daily Schedule Report. The inspectors identified an exception early in the period, when the daily PSA was not updated to reflect Unit 1 "A" high head safety injection pump maintenance which had extended two days past the originally planned outage duration. The work week manager (WWM) initiated appropriate corrective actions to address this deficienc Operations and maintenance personnel used the Daily Schedule Report to discuss daily
. CDF and coordinate risk significant component work activities. . Work week managers actively tracked and communicated job status to the NSS throughout the day, which supported sound decision making with respect to configuration control. In general, NSSs effectively used the Daily Schedule Report and the NPDAP 7.12 program to manage on-
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line maintenance authorization in a safe manner. The inspectors noted five recent condition reports (CRs) initiated by NSSs and an electrical supervisor which identified work scheduling, authorization, or coordination deficiencies. In three cases, NSSs had authorized work without recognizing intersystem relationships which made an additional train or risk significant system inoperable. Subsequent PSA determined that overall plant risk had remained within allowable levels. The CRs clearly described the operability relationships in sufficient detail to support subsequent investigation and corrective actions. The remaining two CRs concerned work scheduling and coordination problems which extended TS limiting condition of operation durations for the recirculation spray and supplemental leak collection and release system Notwithstanding an overall safe on-line maintenance program, the inspectors identified several minor performance deficiencies which tended to extend equipment out of service times or limit NSS risk insights regarding ongoing maintenance activities. Pre-job parts i verifications and job walkdowns were not consistently performed prior to the scheduled J work day. Deficiencies associated with staging and chain of custody for parts used for i 12-week schedule maintenance are discussed in Section E2.1. Workers in the field did ;
not consistently contact the WWM when schedule deviations, which changed planned I equipment unavailability duration, occurred. Probabilistic safety analysis evaluations were not consistently updated when equipment unavailability exceeded the planned duration. The thresholds for initiating CRs to documem work schedule deviations was inconsistent. Station management acknowledged the inspectors' observations and initiated reviews of those issues not otherwise documented in recent CR ) Conclusions On-line maintenance was managed consistent with equipment availability assumptions contained in the Beaver Valley Unit 1 & 2 Probabilistic Risk Assessments. The on-line -
maintenance procedure was comprehensive and well understood. Incorporation of maintenance rule insights was a strength. Work week managers actively tracked and communicated job status to the Nuclear Shift Supervisor (NSS), which supported sound decision making with respect to configuration control. Occasional performance deficiencies, such as NSS authorization of work without recognizing system operability relationships were properly addressed through the condition report syste M8 Miscellaneous Maintenance issues (92700,92901)
M8.1 (Closed) LER 50-334/98-15-01: Inadequate Performance of Channel Functional Tests The inspectors performed an onsite review of the LER. The supplement provided clarification of a previous corrective action and did not change the previous NRC i assessment of the even I I
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M8.2 (Closed) Violation 50-334/98-04-01: Improper Response to Unit 1 Excessive Turbine Driven Auxiliary Feedwater Pump Packing Leakage On August 8,1998, the inspectors observed a maintenance supervisor manipulate a valve and perform maintenance on the auxiliary feedwater (AFW) pump without an -
approved work document as required by station procedures and TSs. This was done while the pump was being tested and delayed pump restoration by 22 hour2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> The inspectors reviewed various training records, interviewed station personnel, and observed preevolution briefings for subsequent AFW pump tests to assess the effectiveness of corrective actions and use of lessons learne Corrective actions included counseling the maintenance supervisor involved in the even The Vice President, Nuclear Operations and Plant Manager issued a site-wide memorandum to reiterate the policy for operation and maintenance of station equipmen The Operations, System and Performance Engineering (SPED) and Maintenance departments conducted training with their staffs. Each group reviewed the violation and their responsibilities for performing surveillance testing and maintenance. The inspectors interviewed station personnel associated with subsequent AFW pump test The inspectors concluded that, as a result of the training, station personnel knowledge of site requirements for equipment testing and maintenance was strengthene The inspectors observed the pre-evolution briefings for subsequent AFW pump tests for both Units 1 and 2. The tests were 1/20ST-24.9, "Overspeed Trip Test of Turbine Driven AFW Pump," (see NRC Integrated Inspection Report Nos. 50-334/99-02, 50-412/99-02 for revision numbers). The briefings were performed by assistant nuclear shift supervisors (ANSSs) with Maintenance and SPED personnel involved, and included a detailed discussion of the lessons leamed from the violation. The ANS$s stressed the - :
plant operator's responsibility for the control of plant equipment whi!e the test was in
. progress. The inspectors determined that the lessons leamed from the violation were effectively reviewed prior to the test The inspectors concluded that the licensee took strong actions to reiterate station policy for safe operation and maintenance of plant equipment. Effective training was performed
- with the appropriate work departments which strengthened worker knowledge of station policy for equipment testing and maintenance. Lessons leamed from the violation were stressed during pre-evolution briefings for subsequent AFW pump test ,
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111. Enoineerina E1 Conduct of Engineering E Unit 1 River Water Pumo Lift Adiustments Insoection Scooe (37551)
The inspectors reviewed trending data associated with testing and repairs to the Unit i rive water pumps. The inspectors interviewed the system engineer and performed field walkdowns on portions of the river water system. The river water system was reviewed based on the risk importance (ranked as the second most important system for Unit 1)
- and the indications of degraded pump performance ("B" river water pump was in the ASME alert range for low pressure differential during the last surveillance test). Observations and Findinos The Unit i river water system consisted of three 100 percent capacity pumps with two backup auxiliary river water pumps. The three river water pumps ("A", "B", and "C")
were overhauled in December 1993, November 1991, and October 1991, respectivel The inspectors noted that the system engineer had typically performed good trending of the pump performance data. The system engineer recommended overhauling the
, - pumps on a 10 year frequency, based on site experience and performance dat Between overhauls, pump lift adjustments were conducted to improve pump performance. Based on review of system engineering test data and operator logs, the inspectors determined that pump performance generally declined since the last lift
- adjustment. Since 1991, eight lift adjustments were made. Three occurred after a pump failed the quarterly surveillance test, and five were performed due to declining pump performance while the pump remained operable. The most recent pump lift adjustment ("C" river water pump in June 1999) was performed promptly following the quarterly surveillance test which indicated pump performance degradatio The inspectors identified weaknesses associated with pump performance evaluation and associated scheduling of one of the lift adjustments. On July 7,1998, the "A" river water pump quarterly surveillance was performed and the pump was just marginally operable (0.76 ft of head above the minimum operating point (MOP) curve). The system engineer had recognized the degraded pump performance earlier and had initiated a work request for performance of a pump lift adjustment on June 10. The lift adjustment was not made until September 21, Just prior to the next scheduled quarterly pump surveillance test. The inspectors determined that this delay was due to communications and scheduling deficiencies within the 12 week work planning process. The inspectors questioned the system engineer conceming pump operability from July 7 to September 21. After reviewing the declining performance data of the pump, the system engineer concluded that the pump probably would have dropped below the MOP curve before September 2 However, engineers also noted that due to reduced river water temperature in September, less pump flow would be necessary to remove heat during postulated design
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basis accidents. At the end of the inspection period, engineers continued to evaluate pump operability for the time prior to September 21, as documented in CR 99152 . The inspectors concluded that the conduct of the corrective maintenance in September 1998 was not timely. Performance of the lift adjustment at the end of the surveillance interval without evaluating or testing for operability was also a weakness. The system engineer stated that corrective actions to address these weaknesses would include implementing more timely maintenance on the pumps. Several appropriate corrective actions were in progress based on earlier condition reports and SPED self assessment findings initiated independent of the river water pump trending revie Conclusions System engineers recently performed good trending of Unit i river water pump performance. Performance data was carefully reviewed and used to determine pump maintenance frequency. In one instance, corrective maintenance for a degraded river water pump was not performed in a timely manner. The delay reflected communications and scheduling deficiencies for time sensitive recommended maintenanc E2 Engineering Suppod of Facilities and Equipment E Receiot. Storace and Handlina of Safety Related Eauioment and Materials l Inspection Scone (38702. 62707)
During a previous inspection period, inspectors observed trash and dirt in a material i staging area and questioned whether reasonable control of parts was being maintaine The inspectors reviewed the licensee's controls for the receipt, storage, and handling of safety related equipment and material. Specific areas reviewed included: receipt inspection; purchase order specification and documentation; material classification and
- traceability; material handling; environmental storage conditions; warehouse inventory maintenance; shelf life requirements and material issuance; staging; and return practices and control Observations and Findinas Material control procedures were well written and adequately described the requirements and implementation of the procurement program. Quality Assurance Procedure OP-6,
" Material Control." Rev.10, accurately described the requirements of the procurement program. Nuclear Procurement Department (NPD) procedure NPD 4.0, " Receiving inspection Function of Materials or Services," Rev.10 was referenced by the inspectors as they observed the quality control receipt inspectors process safety related items. The receipt inspectors verified that the purchase order specification requirements were met and that all required documentation was present. Material classification was performed by procurement engineering and specified on the purchase order. No problems were identifie I
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Material handling was observed and performed properly. Materials were staged in both an offsite warehouse and an in-plant storeroom in an orderly fashion. Both areas were clean and uncluttered. Combustibles such as oil and paint were stored separatel '
. Temperature and humidity requirements for safety related components were maintained and recorded. The warehouse supervisor reviewed the data monthly to ensure complianc l '
Warehouse inventory maintenance records were reviewed and found to be complete and current. Task sheets described the required stocked equipment maintenance and frequency. The inspectors noted one instance of overdue maintenance on a compressor which required verification of nitrogen gas pressure. The pressure verification could not be performed until a pressure gauge was installed. The compressor, however, was administratively prevented from being issued. Shelf life requirements were identified by tagging each affected stock item. Shelf life expirations were maintained on a computerized data base and tracked weekly. Several items were inspected and no discrepancies were note Normal material issuance was well controlled and in accordance with procedure NPD 12.0, " Issuing Function," Rev. 5. However, the inspectors identified a staging area used I by maintenance to obtain materials identified per the 12-week schedule that was not maintained in accordance with Operations Quality Assurance Procedure OP-6, " Material Control," Rev.10. Materials were issued, based on the requisitions identified from the 12-week schedule, placed in an open storage area, and subsequently left unattende ,
- Both the procurement and maintenance organizations identified each other as being l responsible for the material. Several discrepancies were noted. Candy wrappers were j found in bags containing safety related parts. Several issued requisition bags for work I orders were tom open with parts laying loose on the storage racks. Some parts had no identifying labels. The area was dirty. Some parts had been staged in this area for over 1 year. The inspectors expressed concem regarding material traceability and conditio . OP-6 requires that " control of material shall be maintained to prevent damage, loss, or deterioration." Failure to properly implement these controls in accordance with OP-6 is a violation of TS 6.8.1.a, which requires that , " written procedures shall be established, implemented, and maintained covering ... the applicable procedures recommended in Appendix "A" of Regulato:V Guide 1.33, Rev. 2, February 1978." This failure constitutes a violation of minor significance and is not subject to formal enforcement actio Condition Report (CR) 991390 was initiated to resolve the control and ownership discrepancies for the staging area. Station management informed the inspectors that all parts in this staging area would be returned to the store room until proper controls were established in the staging area. Items retumed to inventory were properly inspected and the required documentation was maintained. Items not having proper documentation were rejecte ) Conclusions Controls for the receipt, storage, and handling of safety related equipment and material
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were being effectively m. ..t? ined. Deficiencies in the ownership and control of materials being staged to support the 12-ween schedule were identified and captured in the corrective action program, j E7 Quality Assurance in Engineering Activities i
E Nuclear Enaineerina Department Self-Assessments Insoection Scooe (37551)
The inspectors reviewed Nuclear Engineering Department (NED) self-assessments to evaluate the quality of the self-assessment program. In 1998, the following self-assessments were complete *
Safety System Functional Evaluation (SSFE) Results and TS Compliance
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Unit 1 Twelfth Refueling Outage (1R12) Engineering Change Notice (ECN) and Field Change Notice (FCN)
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Nuclear Engineering Department Design Change Package (DCP) Review
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Administrative Controls for Non-conservative TSs , Observations and Findinas in 1998 and through May 1999, five self-assessments were conducted in the NED. The fire protection self-assessment was completed May 27,1999, and was not included as part of this inspection. Four additional scheduled self-assessments were postponed due to emergent work activities during the 1998 extended shutdowns and the Unit 2 refueling outage. The self-assessments performed were completed in accordance with station procedures. Recommendations were tracked and addresse The NED DCP review was completed in response to an NRC violation. The assessment l focused on the specific NRC concern associated with the scope of DCPs and 10 CFR <
50.59 reviews. The review addressed the concem. The Administrative Controls for Non-conservative TSs self-assessment verified that appropriate controls were in place to maintain equipment operable. The majority of the recommendations were for additional reviews of the procedures and processes used in controlling the nonconservative TS The 1R12 ECN and FCN self-assessment was an effectiveness review of changes made to engineering design documents during the Unit i refueling outage. Four of the five recommendations dealt with failures to follow NED procedures. The recommendations were to eliminate the procedure requirements. Failure to follow procedures for control of
. design change packages was a violation of 10 CFR 50 Appendix B, Criterion V,
" Instructions, Procedures, and Drawings." This failure constitutes a violation of minor significance and is not subject to formal enforcement action. The inspectors did not identify any safety issues with the failures to follow procedure. However, the common
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theme of procedure adherence was not addressed in the self-assessment. The reviewing engineer stated that during future reviews of ECNs and FCNs, additional emphasis would be placed on evaluating procedural compliance. The fifth recommendation was to improve DCP quality by emphasizing thorough reviews and walkdowns prior to DCP issuance. Prior corrective actions, such as institution of the installation and test plan process and the project manager concept, were already addressing this issue, therefore no additional action was take The SSFE Results and TS Compliance self-assessment addressed the immedir.te concern of TS compliance during the 1998 extended outages. The assessmert was a review of open items generated by the SSFE. The review found that generally the open items were properly prioritized, reviewed, and dispositioned. The review failed to identify inadequate administrative controls associated with a nonconservative TS. This issue l was later identified by the inspectors through review of the open items and resulted in a violation (see NRC Integrated Inspection Report 50-334(412)/98-01). The inspectors determined that with the exception of the fire protection assessment, the planned r ope of the NED self-assessment was narrow and in some cases, findings were incomplet In 1999, subsequent to the self-assessments discussed above, the Quality Services Unit initiated use of self-assessment checklists as a tool to improve self-assessment consistency and qualit Conclusions
The nuclear engineering self-assessments reviewed satisfied administrative )'
requirements and confirmed expected results. However, the assessments did not provido substantial recommendations or corrective action E8 Miscellaneous Engineering issues E8.1 (Closed) LER 50-334(412)/99-02: Nonconservative Concurrent lodine Spike Radiological Dose Calculation Methodology Insoection Scope (92700)
In February 1999, engineers determined that several non-conservative assumptions were used for design basis accident (DBA) dose assessments. As a result the calculated control room operator dose for the main steam line break DBA and small line break DBA could have exceeded the 10 CFR 50, Appendix A, Criterion 19 limit (interpreted as 30 Rem to the thyroid). The inspectors interviewed engineers, reviewed operability assessments and supporting records, and reviewed the LER and associated corrective actions to determine whether the deficiencies were properly addresse j Observations and Findinas in December 1997, engineers had identified that the Unit 2 control room emergency ventilation system was susceptible to single active failures. While evaluating the issue, radiological engineers determined that several related nonconservative assumptions had ;
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been used in DBA dose assessment calculation (see NRC Integrated Inspection Report Nos. 50-334(412)/97-11, NCV 50-334(412)/97-11-09, and LER 50-412/97-08-01). The revised (corrected) dose assessments for cont..)I room operators remained within regulatory limits. While performing extent of condition reviews in 1998, engineers identified that additional, unrelated nonconservative assumptions had been used in the DBA dose assessment calculations. The assumptions of concem had been provided by and used by the Nuclear Safety System Supplier to support original plant licensin These errors resulted in significant underestimation of the concurrent iodine (1-131) spike magnitude. The calculation errors affected both units. However, when corrected, only the Unit 1 DBA bounding calculations exceeded regulatory limits as described in the inspection scop The DBA concurrent lodine spike is a functior of iodine leakage from fuel and iodine presence in the reactor coolant system (RCS) during plant operation. As iodine removal from the RCS is increased, the release rate from the fuel may increase while the RCS coolant 1-131 activity remains within the TS limits. Therefore operating conditions which
. maximize RCS cleanup (and iodine removal) are the bounding conditions to be used for DBA dose assessment calculations. The four incorrectly used parameters are listed belo Non-Conservative Correct Limiting Pararneter Oriainal Value (Boundina) Value (1) RCS Letdown Flowrate
[ maximum operational value) 60 gallons per minute (gpm) 120 gpm (2) RCS letdown domineralizer efficiency 10 99
[ maximum nominal value)
(3) RCS Leakage
[ maximum allowed] Ogpm 12 gpm i i
(4) . RCS Volume
[ excluding pressurizer 1.91 E8 grams 1.763 E8 grams vapor space]
The LER accurately described the event, potential consequences (low), and corrective ,
actions including basis for continued operation (BCO) documents for each unit. The inspectors determined that the BCO documents were comprehensive, technically sound, and well written. The inspectors noted that BCO 1-98-009, Rev.1, specifically addressed Unit 1 Altemate Plugging Steam Generator Tube Plugging Criteria. The maximum allowed end-of-cycle analyzed faulted steam generator leakage for a main steam line break event was reduced to 6.5 gpm. The current projected value is 1.0 gp The LER also reported this event as a 10 CFR 21 issue. Specifically, engineers determined that the radiological calculation method used to calculate the DBA dose
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i assessments was a Basic Component which could create a Substantial Safety Hazar !
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The inspectors determined that the evaluation was thorough and technically sound. The licensee also reported the issue to the industry via the nuclear network. The complete review and revalidation of DBA dose calculations is scheduled for completion by August 30,1999. The inspectors conc!uded that engineers had properly evaluated the event to date and that remaining corrective actions were scheduled for completion in a timely i manne CFR 50, Appendix B, Criterion lil, " Design Control" states that " measures shall be established to assure that applicable regulatory requirements and the design basis are ...
correctly translated into specifications, drawings, procedures, and instructions." Failure to ensure that the design values used in DBA control room dose assessment calculations corresponded to actual plant conditions for more than a 12 year period was a violation of 10 CFR 50, Appendix B, Criterion Ill. Control room operator dose could have exceeded regulatory limits if a main steam line break DBA or small line break DBA had occurred while Unit 1 operated within licensed parameters. However, the likelihood of the unit operating at all of the pertinent parameter limits simultaneously was smal Appropriate corrective actions were implemented in a timely manner for this licensee identified, non-repetitive, and non-willful violation. This violation constitutes an additional example of NCV 50-334(412)/97-11-09 and is not being cited individuall Conclusions.
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' Engineers identified and corrected several nonconservative assumptions used in dose ,
assessment calculations for design basis accidents (DBA). For two DBA types, the i errors could have permitted control room operator radiological dose to exceed regulatory requirements. Extent of condition causal analysis, basis for continued operation evaluations, and 10 CFR 21 reporting were comprehensiv E8.2 Year 2000 Computer System Readiness / NRC Temoorary Instruction 2515/141
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The staff conducted an abbreviated review of Y2K activities and documentation using Temporary Instruction 2515/141, " Review of Year 2000 (Y2K) Readiness of Computer i Systems at Nuclear Power Plants." The review addressed aspects of Y2K management planning, documentation, implementation planning, initial assessment, detailed assessment, remediation activities, Y2K testing and validation, notification activities, and contingency planning. The reviewers used NEl/NUSMG 97-07," Nuclear Utility Year 2000 Readiness," and NEl/NUSMG 98-07, " Nuclear Utility Year 2000 Readiness Contingency Planning," as the primary references for this revie The results of this review will be combined with the results of other reviews in a summary report to be issued by July 31,199 .
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IV. Plant Support R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 Radiation Exposure Manaaement Insoection Scooe (71750. 83728)
A health physics inspection during routine operations was conducted. Areas of inspection focus were based on the following regulatory requirements from 10 CFR Part 20:
20.1101 Radiation protection program 20.1601 Control of access to high radiation areas 20.1602 Control of access to very high radiation areas 20.1902 Posting requirements 20.1904 Labeling containers 20.2103 Records of surveys Special focus during this inspection was on the program for maintaining occupational exposures as low as is reasonably achievable (ALARA). This included reviewing the processes for: (1) establishment and implementation of exposure goals, both site-wide and job specific; (2) incorporation of ALARA into engineering designs and modifications; (3) management of the ALARA program, and, (4) source term reduction, Observations and Findinas
Annual site, unit, and outage goals are developed by the health physics staff and presented to the Nuclear ALARA Review Committee (NARC) for discussion, review, and approval. Senior station managers make up the NARC membership, with the committee chaired by the Plant Manager. For 1999, a goal of 141 person-rem was established for the station, including 95 person-rem for the 2R07 refueling outage, and 10 person-rem l for the 1-S-99 mid-cycle maintenance outage, both of which have been completed. For the 2R07 outage, total exposure measured only 63 person-rem. Licensee personnel attribute the significantly lower than expected exposures to reduced staffing levels for the outage when compared to earlier outages; reduced exposure rates by the steam generator secondary side due to maintaining the secondary side flooded during the outape; and reduced work interferences. This last item was the result of having performed the primary side steam generator and reactor coolant pump seal inspections during 199 During the past year, a 12-week computer based planning cycle has been implemented for work control and planning during normal operaGons. Twelve system windows are created each quarter, and the work week plan is finalized two weeks prior to implementation. Representatives from each of the major work departments are part of the work control and planning group, including a representative from health physic Outago planning remains a separate work activity, due to use of a specific schedule
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16 l process. The health physics computer system, which is used to create the associated radiation work permits (RWPs) and to interface with the electronic dosimetry system for dose tracking, is not currently compatible with the work planning computer syste Accordingly there is complexity and some inefficiency in the work planning and control proces During their review of the recently completed mid-cycle maintenance outage at Unit 1 (1-S-99), the licensee recognized weaknesses in the work planning and control proces In this case, additional activities were allowed to be added to the scope of the previously i planned maintenance outage without estimating and planning for the additional -
accumulated exposure. Consequently, the outage was completed with 24.4 person-rem, ,
(i.e.,140% over the established outage goal). While no regulatory issue is involved, the ]
licenset ;1as initiated review of this matter to enhance work planning and control, and l improve ALARA efficiency and accountabilit !
l Plant design modifications are initiated by NED personnel, which by plant procedures, provide notification to the station ALARA coordinator during the conceptual design phase J of the modification process. Record reviews indicated that ALARA concepts were !
incorporated into plant modification Source term reduction initiatives have generally been focused around controlling primary chemistry in accordance with Electric Power Research Institute (EPRI) guideline Additionally, a hot spot tracking program has been in place for a number of years, which j identifies and recommends hot spot reduction efforts to plant management. Late in i 1998, a sump clean-out project was completed, which removed approximately seven curies of activity, and allowed for the reduction of several high radiation areas in both units, Conclusions
. A generally effective program for maintaining occupational exposures as low as is reasonably achievable (ALARA) has been established. Management involvement in the ALARA program, incorporation of ALARA principles into plant modifications, and )
establishment of exposure goals were appropriat i R1.2 Imolementation of the Radioactive Liauld and Gaseous Effluent Control Proarams Insoection S.ggse (84750)
The inspection consisted of:
(1) Physical walkdown of facilities and equipment, including radiation monitors; (2) Review of liquid and gaseous effluent release permits; (3) Review of effluent control procedures; (4) Review of the 1998 Annual Radioactive Effluent Report; (5) Review of the Offsite Dose Calculation Manual (ODCM);
(6) Review of Y2K issue; and
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(7) Review of the linplementation of TS/ODCM requirements,.
b. Observations and Findinas Effluent control procedures were detailed and easy to follow, and ODCM requirements were incorporated into the procedures. Radioactive liquid and gaseous eihuent release permits were properly completed. The licensee reviewed and resolved the potential Y2K issues for its computer software which is used to demonstrate compliance with projected dose calculations to the public from effluent release During the inspection, all effluent radiation monitors and air cleaning systems for both units were operable. Further, the primary auxiliary and containment buildings were maintained at a negative pressure, as require The 1998 Annual Radioactive Effluent Report properly reported: (1) the total released radioactivity through liquid and gaseous effluents; (2) effluent radiation monitoring !
systems not retumed to operable status within 30 days; (3) any unplanned radioactive liquid release; and, (4) the projected maximum doses to the public, as required. No reporting discrepancies were noted. Projected doses to the public were well below the TS/ODCM limit The ODCM provided descriptions of the sampling and analysis programs used to quantify radioactive liquid and gaseous effluents and used for calculating projected doses to the public. All necessary parameters, such as effluent radiation monitor setpoint calculation methodologies and site-specific dilution factors, were liste The licensee established an adequate program to implement the guidance in NRC Inspection and Enforcement Bulletin 80-10, ' Contamination of Nonradioactive Steam and Resulting Potential for Unmonitored, Uncontrolled Release of Radioactivity to ;
Environment," to detect and respond to potential unmonitored releases. In addition, programs were in place to implement guidance in 10 CFR 50.75(g) regarding documentation of residual contamination locations for purposes of decommissionin c. Conclusiong Radioactive liquid and gaseous effluent control programs were effective. The ODCM contained sufficient specification and instruction to acceptably implement and maintain the radioactive liquid and gaseous effluent control program :
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R2 Status of RP&C Facilities and Equipment R Calibration of Emuent/ Process / Area Radiation Monitorina S > stems. Flow Rate Measurement Devices. and Oxvoen Monitors Inspection Scope (84750)
The inspector reviewed the most recent calibration results for the following monitoring and measurement system Unit 1 RMS:
o RM-LW-104, Liquid Waste Effluent Monitor e RM-LW-116, . Liquid Waste Contaminated Drain Monitor o RM-DA-100, Auxiliary Feed Pump Bay Drain Monitor e RM-207, Fuel Storage Pool Area Monitor e RM-RIVER WATER-100, - Component Cooling-Recirculation Spray Heat Exchangers River Water Monitor e RM-GW-108 A&B, Gaseous Waste / Process Vent System Noble Gas Monitors e RM-VS-101 A&B, Auxiliary Building Ventilation System Noble Gas Monitors e- RM-VS-107 A&B, Reactor Building / Supplementary Leak Collection and Release System Noble Gas Monitors Unit 1 Flow Rate Measurement Devices:
e FR-LW-103 and FR-LW-104, Liquid Radwaste Emuent Lines e FT-CW-101 and FT-CW-101-1, Cooling Tower Blowdown Line e FR-GW-108 or RM-GW-109 Ch 10, Gaseous Waste / Process Vent System,
-e FR-VS-101 or RM-VS-109 CH 10,- Auxiliary Building Ventilation System, e FR-VS-112 or RM-VS-110 Ch 10, Reactor Building / Supplementary Leak Collection and Release System, Unit i Oxvoen Monitors:
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e On-AS-GW-110-1,2 Unit 2 RMS:
e 2SGC-RQ100, Liquid Waste Process Effluent Monitor e 2HVS-RQ101 A&B, Ventilation System Noble Gas Monitors e 2HVS-RQ109 A&B, Elevated Release Noble Gas Monitors
.e . 2RMQ-RQ301 A&B, Decontamination Building Vent Noble Gas Monitors e 2RMQ-RQ303 A&B, Waste Gas Storage Vault Noble Gas Monitors e 2HVL-RQ112 A&B, Condensate Polishing Building Vent Noble Gas Monitors e 2HVR-RQ104 A&B, Containment Purge Exhaust (Xe-133)
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e 2RMR-RQ303 B, Gaseous Activity (Xe-133) RCS Leakage Detection Unit 2 Flow Rate Measurement Devices:
- 2SGC-FS100, Liquid Radwaste Effluent
- 2CWS-FT101, Cooling Tower Blowdown Line Unit 2 Oxvaen Monitors:
e 2GWS-OA100 A&B Findinas and Observations All calibration results, including oxygen and hydrogen monitor calibration results, were within the licensee's acceptance criteria. The licensee performed radiation monitoring system (RMS) availability trending evaluations relative to its administrative goal for the RMS availability of greater than 95% for all RMS (i.e., effluent, process, area RMS). The 1999 availability was about 90% for both unitu. However, the availability of effluent RMS )
met the administrative goal. The radioactive liquid and gaseous effluent RMS 1 assessment results indicated that there was no Y2K issu Conclusions
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The licensee maintained an adequate program for effluent RMS calibratio j l
R2.2 Air Cleanina Systems I Inspection Scope (84750)
The inspector reviewed the licensee's most recent surveillance test results ((1) Visual Inspection, (2) In-F: ace high efficiency particulate (HEPA) Leak Tests, (3) In-Place Caarcoal Leak Tests, (4) Air Capacity Tests, (5) Pressure Drop Tests, and (6)
Laboratory Tests for the lodine Collection Efficienck s] to evaluate the implementation of ,
TS and UFSAR requirements for the following systems at both unit l (1) Control Room Emergency Habitability System:
(2) Supplemental Leak Collection and Release Systems; and (3) Containment Buildin Observations and Findinas All surveillance results were within the TS acceptance criteria. The licensee performed tracking and trending evaluations for the above surveillance items, including plant air balance The licensee used American Society for Testing and Materials (ASTM) D3803-1979 methodology to determine the iodine collection efficiency, as required by TS. The
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licensee also tested the efficiency using the ASTM D3803-1989 to satisfy a proposed NRC Generic Letter which has not yet been issued. The lodine collection efficiencies using both methods met the TS acceptance criteri g Conclusions i
The licensee established, implemented, and maintained an effective ventilation system I surveillance program with respect to charcoal adsorption surveillance tests, HEPA and charcoal filter mechanical efficiency tests, and air flow rate test ,
R5 _ Staff Training & Qualifications in RP&C Activities i
R5.1 ' Radiation Exposure Control Trainina j l Inspection Scope (83728) I
A review of training programs established for plant personnel covering maintaining occupational exposures ALARA was conducted. Training programs reviewed included i Initial and continuing radiological worker training and initial and continuing radiation protection technician training. The inspection was accomplished via a review of training l documents, instruction and course manuals, training facility tours, and interviews with cognizant plant personne Observations and Findinos
Radiological worker training is presented on an ahnual basis to all plant workers requiring access to the radiologically controlled area (RCA). Chiipter 5 of the training manual for this course covers the ALARA program at the station in depth. Additional ALARA concepts are reinforced during the practical factors training given at the conclusion of the classroom instruction, in a facility specifically dedicated for this purpos Radiation protection technician initial training covers the AI. ARA program in three distinct training modules. Additional ALARA training is provided periodically as part of the technician continuing training program. For 1999, one segment of radiation protection
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technician training was directed specifically at radiation exposure control,' including dose minimizatio Conclusions '
An effective training program for ALARA has been implemented for both radiological workers and radiation protection technicians. Specific training both as part of initial and continuing training has been establishe i
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R7 Quality Assurance in Radiological Protection and Chemistry Activities R7.1 Quality Controls for Radioloaical Effluent Monitorina Inspection Scooe (84750)
The inspection consisted of: l (1) Review of the 1998 Quality Assurance (QA) audit reports:
- QA Audit Report (BV-C-98-13), " Radiological Effluent and Environmental l Monitoring Programs", j
- QA Audit Report (BV-C-98-15), " Health Physics Program Audit" ; l (2) Review of inter-laboratory measurement comparisons; and (3) Review of the measurement laboratory quality control (QC) program for radioactive liquid and gaseous effluent sample ; Observations and Findinas l
The licensee audit team members identified a number of findings and strength '
Technical specialists were brought in to evaluate the radiation monitoring systems and other effluent areas. The scope and technical depth of these audits were sufficient to appropriately identify weaknesses in maintaining the radiation monitoring systems, and identified deficiencies were documented in condition reports and tracked until resolutio None of the identified deficiencies were significant with respect to actual effect on the public health and safety, and the environmen No discrepancies were evident from QC data for inter-laboratory comparisons. The QC l program consisted of measurements of spike / split samples through a vendor-supplied service. Quality control charts for the gamma spectrometry counting efficiency, gamma
- spectrometry full-width-half-maximum, and tritium counting efficiency were frequently reviewed by licensee staff and used as a mechanism to assess laboratory performanc ! Conc!usions The Quality Assurance program for the effluent control program was effectively implemented. Quality Control programs to validate radioactive liquid and gaseous effluent control program analytical results were effectiv V. Manaaement Meetinos X1 Exit Meeting Summary The inspectors presented the inspection results to members of licensee management on June 17,1999, after the conclusion of the inspection. The licensee acknowledged the findings presente i
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The licensee did not indicate that any of the information presented at the exit meeting was proprietary, i
X2 NRC Management Site Visit '
On May 27,1999, Mr. Hubert J. Miller, Regional Administrator, Region I and other NRC staff personnel visited the site, toured the facility, and discussed current plant performanc i i
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INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 38702 . Receipt, Storage, and Handling of Equipment and Materials Program IP 61726: Surveillance Observation IP 62707: Maintenance Observation IP 71707: Plant Operations IP 71750 Plant Support IP 83728 Maintaining Occupational Exposures ALARA IP 84750 Radioactive Waste Treatment, and Effluent and Environmental Monitoring l IP 90712 In-Office Review of Written Reports of Nonroutine Events at Power Reactor !
Facilities IP 92700: Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor - '
Facilities IP 92901: Follow-up - Operations I Tl 2515/141: Review of Year 2000 (Y2K) Readiness of Computer Systems at Nuclear Power Plants ITEMS OPENED, CLOSED AND DISCUSSED l Closed 50-334/99-01 LER Manual Reactor Trip Due to Continuing Degradation of Main Condenser Parameters (Section 08.1)
50-412/99-01 LER Failure to Comply with Technical Specifications Due to Not Meeting the Acceptance Criteria for a Source Range Monitor During Surveillance Testin (Section 08.2) {
50-412/99-05 LER 4kV-2A Bus Trip on Ground Overcurrent Relay 51-VA207X. (Section 08.3)
50-334/98-15-01 LER Inadequate Performance of Channel Functional Tests. (Section M8.1)
.50-334/98-04-01 VIO Improper Response to Unit 1 Excessive Turbine Driven Auxiliary Feedwater Pump Packing Leaking (Section M8.2)
50-334(412)/99-02 LER Nonconservative Concurrent lodine Spike )
Radiological Dose Calculation Methodology {
(Section E8.1)
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LIST OF ACRONYMS USED -
y "1R12 - Unit 1 Twelfth Refueling Outage ,
, AFW ' Auxiliary Feedwater - .
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ALAR . As Low As is Reasonably Achievable ANSS- Assistant Nuclear Shift Supervisor ..
. ASTM : American Society for Testing and Materials- . i BCO-- Basis for Continued Operation L ,
CDF' . Core Damage Frequency:
.CFR Code of Federal Regulations CR . Condition Report ' -,
CW Circulating Water '
DBA^ Design Basis Accident DCP- . Design Change Package l )
ECN Engineering Change Notice ,j ED Emergency Diesel Generator 1
.EPRI . Electric Power Research Institute j FCN Field Change Notice i
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gpm Gallons per Minut HEPA High Efficiency Particulate LE Licensee Event Report -
MOP 2 Minimum Operating Point
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MPFF Maintenance Preventable Functional Fai!ure i NARC Nuclear ALARA Review Committea l NCV ' Non-Cited Violation !
NED Nuclear Engineering Department NPD Nuclear Procurement Department NPDAP Nuclear Power Division Administrative Procedure NRC- Nuclear Regulatory Commission I NRR Nuclear Reactor Regulation NSS Nuclear Shift Supervisor ODCM Offsite Dose Calculation Manual 1 OSC Onsite Safety Committee i OST Operational Surveillance Test PDR Public Document Room PSA Probabilistic Safety Analysis Q Quality Assurance-QC Quality Control-RC Radiologically Controlied Area RCS Reactor Coolant System REMP Radiological Environmental Monitoring Program RETS Radiological Effluent Technical Specifications RMS Rac'istion Monitoring System RP&C , Radiological Protection and Chemistry RWP Radiation Work Permit SPED System and Performance Engineering Department SRNI Source Range Nuclear Instrumentation ,
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'25 SSFE Safety System Functional Evaluation-TS
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Technical Specification -
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UFSA Updated Final Safety Analysis I:- ViO :
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Violation WO Work Order WWM Workweek Manager Y2K Year 2000 1