IR 05000302/1998006
| ML20236S649 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 07/20/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236S648 | List: |
| References | |
| 50-302-98-06, 50-302-98-6, NUDOCS 9807270146 | |
| Download: ML20236S649 (42) | |
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, U.S. NUCLEAR REGULATORY COMMISSION
REGION II
Docket No: 50-302 License No: DPR-72 Report No: 50-302/98-06 t
Licensee: Florida Power Corporation Facility: Crystal River 3 Nuclear Station Location: 15760 West Power Line Street Crystal River. FL 34428-6708 Dates: May 10 through June 20, 1998 Inspectors: S. Cahill. Senior Resident Inspector S. Sanchez. Resident Inspector
, J. Bartley Resident Inspector. Farley. Section 08.1 i
P. Fillion. Reactor' Inspector. Sections E E8.6-E8.13 M. Miller. Reactor Inspector.' Sections E8.1-E S. Ninh, Project Engineer. Sections 08.3-0 Salyers. Emergency Preparedness Specialis Sections R8.1-R8.5 l
l- Approved by: R. Musser. Acting Chief. Projects Branch 3
. Division of Reactor Projects
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- 'i POR ADOCK 05000302 F-8 PDR (- t
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EXECUTIVE SUMMARY Crystal River 3 Nuclear Station NRC Inspection Report 50-302/98-06 l
This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a six-week period of resident ins)ection; in addition, it includes the results of announced inspections Jy regional inspectors, including a project engineer and an emergency preparedness specialist. as well as a visiting resident inspecto Operations:
. Operations responded aggressively to several performance problems this period. However these problems were indicative of ongoing problems with unclear radiological work practice expectations, poor use of procedures, and human performance lapses (Section 01.1).
. The opening of a breaker cubicle door in a condition that clearly indicated an abnormal condition was an example of extremely poor electrical safety practices by Operations. Operations management also did not initially recognize or consider these electrical safety practices to be a significant problem (Section 01.2).
. Personnel associated with the evolution to deenergize Engineered Safeguards Motor Control Center MCC-3Al at power to remove a faulted breaker worked with a good sense of urgency. Overall, the evolution was executed well and controlled in a timely manner commensurate with safety
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(Section 01.2).
. During the evolution to deenergize Engineered Safeguards Motor Control Center MCC-3A1. 0)erations was uncertain of several equipment impacts which could have Jeen resolved using resources outside of Operations to determine the impacts. Most of the unknown load effects.had minimal operational consequences. While the evolution received appropriate management oversight, the potential existed for this oversight to be bypassed due to the informal action plan and lack of any process to capture it (Section 01.2).
. After discussions with Operations personnel and management following a delayed operability decision. Operations understanding of timely operability determinations was identified as a weakness (Section 01.3).
. Accurate and timely Operations operability determinations was hindered by acceptance criteria in surveillance )rocedures (SP) that were poorly referenced to a source requirement and lad no bearing on safety function and operability. Consequently the operability impacts of failing to i , meet SP acceptance criteria were not readily apparent to operators (Section 01.3).
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. The use of a wordprocessors macro for limiting condition for operation action verification was questionable because of the possibility to miss a required action if the Nuclear Shift Manager does not verify the actions against the Improved Technical Specifications (Section 01.3).
. The licensee's initial deportability determination for a makeup pump lube oil pump failure was timely but the content of the information sent to the NRC was poor. Their explanation did not stand alone and did not I a) pear to be well understood by the operators documenting and reporting tie event (Section 01.4).
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. The licensee equipment label upgrade project was an ambitious and comprehensive effort, which also corrected and upgraded configuration control databases. Good oversight and ownership of the project were displayed by the operator assigned to manage it. An action plan to resolve discrepancies was thorough and planned in advance between Operations and Engineering. A large number of missing labels and flow print discrepancies have been found during system walkdowns which have 1 validated the need for the project (Section 02.1).
. The licensee's Nuclear General Review Committee was meeting the commitments as described in Final Safety Analysis Report Section 12. Nuclear General Review Committee. Each member of the Engineering and Technical Support Subcommittee, as well as the members at the full committee meeting, consistently contributed to discussions on the various subject matter presented (Section 07.1).
. A revision to the corrective action program requirements that did not *
consider the impact on existing items in the system was identified as an example of poor change managemen Two examples of active and open ,
nonconformance Deficiency Reports were identified with closed Precursor !
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Cards tracking the issues in the corrective action program (Section 07.2).
Maintenance:
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. A non-licensed operator consistently' implemented the S.T.A.R. (Stop, l Think. Act. Review) 3rocess during the conduct of a surveillance procedure on the tur)ine-driven emergency feedwater pump and valves (Section M1.2).
. Surveillance procedure SP-3498 for the turbine-driven emergency feedwater pump and valves contained steps that were not tr, ded the same by two non-licensed o)erators performing the procedure on different i occasions; however, tais resulted in minimal significance because the l operator only had to stroke a valve for timing purposes (Section M1.2).
. The use of a temporary calibrated gauge to verify if a permanent gauge was malfunctioning was considered to be a proactive and appropriate
, action during a turbine-driven emergency feedwater pump and valve surveillance (Section M1.2).
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Engineering:
! . The ir.spector concluded the licensee responded appropriately to a minor
! fire in the diesel exhaust header flange. They displayed excellent
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concern for operability of the opposite train equipment and performed a ,
thorough preliminary investigation including significant industry I experience research (Section E2.1). '
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. During an open item closure review. several discrepancies were note Administrative tracking of precursor card deficiency reports was poor.
, A design issue with the control complex habitability envelope chiller i ability to automatically start at elevated service water temperatures had not been adequately reported to the NR ,
l Changes had been made by l l the licensee to engineering design assumptions in their modification
! package after a license change request had been submitted. The licensee did not then revise the license change request to notify the NRC reviewers. The licensee control of modifications associated with licensing actions was informal (Section E8.14).
l . An inspector identified two discrepancies with the licensee's implementation of reactor vendor owners group guidance.for reactor l coolant pump restart following a boron dilution event. The i discrepancies could have resulted in incorrect operator action and i reference to incorrect procedure source requirements (Section E8.15).
Plant Sucoort:
. During review of the licensee's corrective actions for a violation of contamination control procedures, the inspector noted that " lessons learned ~ from the problem were not self-critical. The licensee's Health Physics organization did not acce)t responsibility for the problem and a
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root cause was not identified. Tlese were considered poor corrective actions (Section R8.2).
. The licensee's emergency plan drill on May 27. 1998, was successful in accomplishing it's training objectiv However. conversations took l place without the Emergency Coordinator's knowledge or awareness that could have resulted in the loss of relevant information to help in decision makin Also, suspension of safeguards controls was not well understood by drill participants outside of the Security organization j (Section Pl.1).
. The temporary relocation of the protected area boundary to support I removal of the Technical Support Center for a construction project was !
l verified to be done well. Compensatory measures were established and l
! effective to ensure no decrease in security effectiveness per 10 CFR l 50.54(p). The controls for accessing individuals in an emergency were l observed during an Emergency Plan drill through a temporary gate and 1 were considered appropriate and sufficient (Sections Pl.1 and S2.1).
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! Reoort Details l-Summary of Plant Status
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The plant began the inspection period in Mode 1 at 100% rated thermal power but reduced it to 85% power for several hours on May 10, 1998, to remove a condenser water box from service for repairs. The plant returned to full power operation later the same day and remained at full power operation for the remainder of the report perio I. Doerations 01 Conduct of Operations 01.1 General Comments (71707)
Using Inspection Procedure 71707. the inspectors performed routine t reviews of plant o emergent problems,perations which included, shift turnovers, response to and log review Several problems occurred this period that resulted in Operations performing prompt investigations per Operations Instruction (01)-1 Investigation of Abnormal Events. The )roblems included: an auxiliary building operator who contaminated his lands while working in the overhead beyond the scope of his radiological work plan (RWP): a containment isolation air-operated valve found gagged closed with no record of the gagging in configuration control documents: a failure to'
follow procedure steps to realign chill water when switching control l
complex chillers; and a failure to reposition a condenser off-gas
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radiation monitor valve following log readings. The inspector reviewed I the completed 01-12 investigations and the details of each event. The ins)ector noted that Operations management responded aggressively to eac1 of these problems by initiating the investigation and implementing immediate corrective actions. Each problem was entered into the corrective action system for further investigation and identification of recurrence controls. The inspector did not identify any safety-significant concerns with the results of each problem. However the problems were further examples of ongoing Operations deficiencies with unclear radiological work practice expectations, poor use of procedures, and human performance problems. Although responded to aggressively by Operations management at each occurrence, the problems continue to occur. Operations management has continued to focus significant effort on reducing these types of problem )
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l 2 l 01.2 Breaker Failure and Subsecuent Removal of ES MCC-3Al From Service While l at Power l Insoection Scope (71707)
l The inspectors reviewed the sequence of events that resulted in the necessary deenergization of an Engineered Safeguards Motor Control Center (ES MCC-3A1). The inspectors also observed the plant activities to deenergize ES MCC-3Al and remove air handling fan (AHF-20A) breaker cubicle while the plant remained near full powe ] Observations and Findinas On May 24, 1998, while performing a monthly surveillance on the control room emergency ventilation system. AHF-20A was restarted and immediately tripped, as indicated by the lights on the main control room boar It was later determined that an electrical short had taken place inside AHF-20A breaker cubicle located on ES MCC-3A Precursor Card (PC) 98-2647 was written to document the event. After the loss of control room indication for AHF-20A and the failure of the fan to run, an operator )
was dispatched to the breaker at ES MCC-3A1. The breaker cubicle had I clearly visible soot marks around the sides of the door but the operator i elected to o)en the door anyway. Opening the cubicle door also required opening the areaker, which had not tripped. The operator did not take any electrical safety precautions while doing these actions. The Operations Nuclear Shift Manager (NSM) subsequently also opened the cubicle without taking electrical safety precautions. A licensee electrical maintenance e.upervisor became aware of this and initiated PC 98-2671 for the poor practice, j
Precursor Card 98-2671 was originally screened a D level PC by the l
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licensee's screening committee. Operations management originally thought this appropriate and considered specific guidance needed to be developed for operators regarding electrical safety. However, the inspector noted that general employee electrical safety training and guidance in Administrative Instruction (AI) 1800. Safety Management, and AI-1802. Personal Protective Equipment Policy, was sufficient to allow the operators to recognize the potential consequences of their action After discussion at the morning management meeting the next day the licensee decided to upgrade the PC to a C level, thus requiring an apparent cause evaluation. The breaker cubicle was potentially deranged equipment due to the loss of breaker indication and the soot on the cubicle. Opening a breaker cubicle door in this condition without electrical safety gear is an example of extremely poor electrical safety practices by Operations, which was not initially fully recognized by licensee management.
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l During the inspector's review of the licensee's plans to remove the breaker cubicle, several problems were observed with uncertain equipment
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impacts from deenergizing MCC-3A1. Initially the emergency diesel (- generator (EDG) was to be considered inoperable due to the loss of power r to a voltage regulator circuit. After further review it was determined
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this did not impact EDG operability. Other observations included uncertain main turbine throttle valve and governor valve position indications and an unknown impact on radiologically controlled area (RCA) access computers. Based on interviews with operators who planned the MCC removal from service, the impacts on several loads were unknown due.to the limited scope of electrical schematic prints for vendor supplied equipment. Some of the aforementioned components had several
)ower su) plies leading to them and the loss of the single supply from iCC-3Al lad uncertain results. Given the several days of preparations
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for the MCC action plan the. inspectors considered other resources !
outside of Operations. such as instrument technician references, could '
have been utilized to determine the impacts. However, the inspectors recognized that most of the unknown load effects had minimal operational consequence The inspectors also noted that the licensee's action plan was an informal document. Per procedure Administrative Instruction (AI)-55 Infrequently Performed Tests or Evolutions (IPTE), the evolution was consicered an IPTE and consequently received enhanced management
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oversight. The Plant Review Committee (PRC) also reviewed the evolution due to the required change to Operating Procedure (0P)-703. Plant Distribution System, to remove the MCC while the plant was operatin Nevertheless, most of the ) reparatory activities for the MCC removal
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! checklist developed by the single-)oint-of-contact operator. The PRC
! chairman appropriately requested tie entire evolution be reviewed, even
- though only the scope of the OP was required to be reviewed. A formal
! action plan procedure would have ensured a full PRC review of the evolution. The licensee pointed out that much of the action plan i l sequenced entry into already existing procedures. So, while the
- evolution received appropriate management oversight and the inspectors ]
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had no major concerns with its conduct. the potential existed for this
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oversight to be bypassed due to the informal action plan and lack of any licensee process to capture it.
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The inspectors observed the activities associated with the deenergization of ES MCC-3Al and subsequent removal of the AHF-20A i breaker cubicle. These activities included observations in the main control room, accompaniment during hanging of the clearance tags, and l actual breaker cubicle removal. No problems or concerns were note j l Personnel associated with the evolution worked with a good sense of l urgency, which was appropriate for the eight hour limiting condition for ,
l operation (LCO). All work to remove the breaker cubicle, restore power '
l to the MCC. and exit the LCO. was accomplished in three hour i c. Conclusions The inspectors concluded that the opening of a breaker cubicle door in a ,
condition that clearly indicated an abnormal condition was an example of
, extremely poor electrical safety practices by Operations. Operations management did not initially recognize or consider these electrical safety practices to be a significant problem. Several problems were
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l observed with uncertain equi) ment impacts from deenergizing the MC .
l The inspectors considered otler resources outside of Operations could I l have been utilized to determine the impacts. The inspectors also recognized that most of the unknown load effects had minimal operational consequences. While the evolution to deenergize the MCC received appropriate management oversight, the potential existed for this oversight to be bypassed due to the informal action plan and lack of any licensee process to capture it. Personnel associated with the evolution {
worked with a-good sense of urgency. The inspectors concluded that, overall, the evolution was executed and controlled in a manner commensurate with safet .3 Steam Driven Emeroency Feedwater Pumo Surveillance i Insoection ScoDe (61726. 71707)
On May 22, 1998, the inspector observed the performance of Surveillance Procedure SP-3498. EFP-2 (Turbine Driven Emergency Feedwater Pump) and Valve Surveillance. The general discussion on the performance of SP-349B is contained in the Conduct of Maintenance section of this report, however the discussion of the 0)erations-related issues identified s during the performance of SP-3493 are discussed her I Observations and Findinas During the performance of SP-349B. a check valve (EFV-35)~on the recirculation line from the motor driven emergency feedwater pump (EFP-1) to the emergency feedwater tank failed to fully seat. This valve had two functior.s: "open" to permit minimum recirculation flow, and "close" to prevent backflow when the steam driven emergency feedwater pump (EFP-2) is operating. The inspector's review of the Nuclear Shift Manager's
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(NSM) logs identified that the log entry at 14:43 failed to classify the o)erability of the surveilled emergency feedwater pump (EFP-2) after ErV-35 failed its acceptance criteria. A later log entry at 18:00 indicated that the safety function of EFV-35 was not met. and thus required entry into Improved Technical Specifications (ITS) Action i 3.7.5. Condition D for an inoperable emergency feedwater ; rain. The ;
time this action was taken was at 17:15 on May 22. 1998. The inspectors l questioned the operator's delay in determining operability because l according to the arocedure, the acceptance criteria for EFV-35 was !
clearly not met w1en the valve disk did not perform the function '
described within the procedur .
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Subsequent discussions with Operations personnel and management resulted in. differing views on when an operability determination should be mad The NSM that failed to make the initial determination indicated that ;
more information was needed in order to make an operability l l determination on EFP-2. Per Compliance Procedure (CP) 150. Identifying '
i and Processing Operability Concerns, the NSM has to make an operability
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determination when a question of operability arises, and make the j determination based on the information available at the time. At a ;
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! go back and rescind or confirm the initial operability determinatio In this instance. not making the prompt operability determination, but instead waiting on an Engineering evaluation. resulted in a moot point
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l The inspectors determined that accurate and timely Operations l operability determinations was hindered by acceptance criteria in i surveillance procedures (SP) that were poorly referenced to a source l requirement and had no bearing on safety function and operability. The licensee's SPs often contain a mixture of regulatory required testing for American Society of Mechanical EngineersSection XI pump and valve testing or ITS requirements, as well as licensee internal commitment Consequently, the operability impacts of failing to meet SP acceptance j criteia are not readily apparent to operator The inspectors also identified that when the Limiting Condition for Operation (LCO) was entered for EFP-2 being inoperable. the NSM log entry did not contain a statement ' indicating that the "B" train of the Nuclear Services and Decay Heat Seawater (RW) was verified to be operable. - This was one of the many required actions to be taken upon entering the LCO. After the inspector brought this to the attention of Operations management, it was determined that a computer word processing program macro was used to generate the electronic log entry. The use of j a macro for LCO action verification was questionable because of the '
possibility to miss a required action if the NSM does not verify the actions against the ITS. The missed entry for the RW system being verified operable was of little consequence because the system was
- operable and available at all times during the surveillanc Conclusions
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The inspectors concluded after discussions with Operations personnel and management, that Operations understanding of timely operability determinations was a weakness. The inspectors determined that accurate and timely Operations operability determinations was hindered by acceptance criteria in surveillance procedures (SP) that were poorly referenced to a source requirement and had no bearing on safety function I and operability. Consequently, the operability impacts of failing to i meet SP acceptance criteria are not readily apparent to operators. The !
inspectors consider the use of a macro for LC0 action verification i questionable because of the possibility to miss a required action if the i
' hSM does not verify the actions against the IT .4 Evaluation of Failure Modes Effects Analysis for a Motor Control Center i
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' Insoection Stone (71707)
l ., The inspectors reviewed the details of a licensee-identified problem ,
that could potentially render the 1B makeup and purification pump (MUP) 3 p
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I Observations and Findinas On April 30, 1998. PC 98-2230 was presented to the Operations Nuclear y Shift Manager (NSM) and determined to be reportable under 10 CFR l 50.72(b)(2)(iii)(D) as a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> phone notification for a single event l that could have impacted the ability of an accident mitigation syste PC 98-2330 described the results of a Failure Modes and Effects Analysis
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(FMEA) on the swing motor control center (MCC) 3A The FMEA determined that a single failure of the MCC could render the IB MUP inoperable by l
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removing power from the AC lube oil pump and could also create an inability to isolate normal makeup and reactor coolant pump seal injection. This in turn could lower the amount of high pressure injection flow going to the core in an emergency. The inspe tor reviewed the text of the licensee's initial notification and considered it poor. It omitted necessary details and didn't stand alone'. It illustrated a potential lack of complete and independent understcodina of the issue by Operations who compiled the text from the PC and FMEA results. The text indicated the issue placed the licensee outside their design basis which would have been reportable as a one hour report per 50.72(b)(1)(ii)(B). After further analysis, the licensee retracted the deportability' determination on May 15. 199 This raised questions regarding the applicability of the original Operations * deportability determination for an accident mitigation system. The licensee's retraction was based on the ability of a backu supply lobe oil flow upon loss of the AC pum TheDC lube was DC pump oil pump not to safety-related but the licensee processed an operability determination that classified it as operable but not fully cualified. They were pursuing dedicating the pump as safety relatec. The licensee initiated PC 98-2523 to address the discrepancy with the reporting requirement interpretation. The inspector review of the basis for the licensee *s deportability retraction did not identify any technical discrepancie . Conclusions The licensee's initial deportability determination for a makeup pump lube oil pump failure was timely but the content of the information sent to the NRC was poor. Their explanation did not stand alone and did not a) pear to be well understood by the operators documenting and reporting tie event. A subsequent retraction of the deportability determination indicated further misunderstanding of outside design basis deportability criteri Operational Status of Facilities and Equipment 02.1 Labelina Uoorade Procram
, Insocction Scooe (71707)
t The inspectors reviewed the status of the licensee labeling upgrade
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previously discussed in NRC Inspection Report (IR) 50-302/97-1 l
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! 7 Observations and Findinas Considering that the licensee had originally committed to the plant equipment label upgrade 3rogram in August 1997, the inspectors had considered progress on t1e project slow. As of May 1998, no new labels had yet been installed. However, this review revealed that the scope of the project had evolved beyond the original plan and was progressing well. The licensee determined that it was appropriate to address many known weaknesses in their configuration management information system (CMIS) and electronic clearance system databases and the engineering schematics in conjunction with the relabel effort. The labeling group in Operations coordinated with the System and Design Engineering departments to develop a comprehensive action plan for dispositioning discrepancies they ex)ected to find in the plant. They developed this prior to commencing tie walkdowns to ensure the findings would be correctly dispositioned. The CMIS database field for unique component identification was expanded from six to 50 characters to provide more flexibility and better labeling capacity. The inspector determined the project was an ambitious and comprehensive effort to fix the licensee's database The licensee hired seven new potential non-licensed plant operators to perform the walkdowns and an instrument technician was supplied by the Maintenance department on a rotating basis. System walkdowns by these individuals were being done against the flow prints and instrument drawings and commenced in April 1998. They have identified numerous discrepancies. primarily with missing labels but also with the flow prints. The inspector verified many of these discrepancies and noted that temporary labels were hung as the problems were found and the print problems were transferred to engineering for resolution per the )re-arranged action plan. Seventeen systems had been completed at t1e time
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of the review and discrepancies were found with over 20% of the components checked. The inspector considered that this validated the value of the licensee's project to reduce operator and configuration control challenge The inspector interviewed the operator assigned full-time management of the projec He was cognizant of the status of all assects of the project and the findings from the completed system wal(downs. The inspector noted the operator exhibited good ownership of the projec Several aspects of the project for pipe labels, room markings, and system train designators were being integrated with an in-progress plant Jainting t]furbishment. A standard process to label instrument valves lad been defelo)ed which fully resolved the concerns previously ,
identified in 11 50-302/97-13. The operator displayed samples of the l new labels which are planned to start being hung in the plant in October 1998. The inspector observed that they were of very good quality and ,
incorporated color-coding and icons to delineate features such as normal status, failure mode, and various program applicability such as
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l containment isolation valves. The licensee had considered unique l requirements for labels in the reactor building and had developed separate labels and attachment devices for that application. The
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licensee's project was scheduled to complete in May 199 Conclusions The inspector concluded the licensee label upgrade project was an j ambitious and comprehensive effort to correct and upgrade configuration control databases. Good oversight and ownership of the project was displayed by the operator assigned to manage it. An action plan to resolve discrepancies was thorough and planned in advance between ,
Operations and Engineering. A large number of missing labels and flow l print discrepancies has been found during system walkdowns which has validated the need for the label projec )
07 Quality Assurance in Operations 07.1 Nuclear General Review Committee (40500) <
On June 6.1998, the Nuclear General Review Committee (NGRC)
subcommittees gathered. with a full committee meeting on June 7.199 The inspector attended the Engineering and Technical Support (E&TS)
subcommittee and the full committee meetings and determined that the NGRC was meeting the commitments as described in Final Safety Analysis Report (FSAR) Sectiors 12.8.2. Nuclear General Review Committee. The inspector observed that each member of the E&TS Subcommittee, as well as the members at the full committee meeting, consistently contributed to discussions on the various subject matter presented. No concerns or problems were note :
07,2 Corrective Action Proaram (40500)
While reviewing open items for closure, the inspector identified several discrepancies in the licensee's corrective action program. Specific problems were found when reviewing items involving restrictions on the ultimate heat sink temperature. These problems were indicative of poor integration of the nonconforming condition Deficiency Report (DR)
process in the corrective action program, previously discussed in IR 50-302/97-04. Although the licensee has recognized the scope of the I problem and initiated a project to separate the DR process from the i corrective action program (CAP). some short term corrective actions were i not implemented well. Compliance Procedure (CP)-111. Processing of Precursor Cards in the Corrective Action System, had recently been ,
changed to require the precursor card tracking an issue to remain open l in the corrective action program as long as a DR evaluation remained in l effect to disposition a nonconformance. However, the two examples the
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inspector reviewed both had open DRs. but the PCs were fully closed in l f the corrective action program. The failure to address the change in the :
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CP-111 requirements on already existing DRs and PCs was considered poor change management. DRs continued to be tracked and retained by the design engineering organization, but their process was not formal and
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not part of the CA Closure or resolution of a DR was also
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consequently difficult to ascertain due to the lack of integration in the CAP process.
L 08 Miscellaneous Operations Issues l'
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08.1 (Closed) IFI 50-302/98-02-04: Radiological Mission Dose Consequences (92901)
l This item was opened to review the licensee's evaluation of Emergency Operating Procedures in plant action steps which were to be performed
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"if accessible." The evaluations were to determine the consequences of not performing the step and if additional guidance at the step was necessary. The inspector reviewed the licensee's evaluations and determined that the items were adequately addressed. The inspector also verified that the recommended procedural enhancements were either already incorporated or entered into the licensee's Nuclear Operations
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Procedure Observations and Suggestions Tracking (NUPOST) procedure l
comment syste This Inspector Followup Item (IFI) is closed 08.2 Soent Fuel Pool Safety Issues Resolution (92901)
In the fall of 1996, the NRC determined that a safety issue associated
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with the Crystal River Unit 3 (CR-3) spent fuel pools (SFP) existed.
l The issue was that during refueling periods when the blank flange on the l containment side of the transfer tube is removed. improper operation of l the spent fuel transfer system or the SFP cooling and cleanup system E could lead to a loss of coolant from the SFP through the transfer tube to the refueling cavity inside containment. The licensee voluntarily committed to enhance procedures to assure the integrity of.SFP inventory. The procedures affected were: Operating Procedure OP-406
-' Spent Fuel Cooling System; and. Maintenance Procedure MP-125. Fuel Transfer Tube Covers and Drain Line Cover. The' inspectors verified that the procedure enhancements implemented were adequate to assure that the
~ fuel. transfer canal and fuel transfer tube drain valves would be closed and locked prior to removal of the fuel transfer covers. In the Summary section of a letter from Florida Power Corporation to the NRC dated November 15, 1996, a typographical error was identified. The error was a component identification number (SFV-182). The correct com)onent should have been valve SFV-180. as already identified in the )ody of the l
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letter. The licensee indicated that the error would be corrected in the next annual commitment update lette .3 (Closed) URI 50-302/97-14-09: NRC Evaluation of' Acceptability of Makeuo System Trains Crosstied Without Ability to Remotely Isolated Trains
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i L This unresolved item (URI) was opened to gather additional information from the licensee for NRC review to address two issues: 1) Should the
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High Pressure Injection (HPI) system discharge header cross-connect valves MUV-3 and 9 have continuous power available for operation from the control room in order to provide for safety system train separation
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l and 2) Should these valves be included in the licensee's in service i testing (IST) program.
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By letters to NRC dated November 10, 1997. January 8. and 23, 1998 Florida Power Corporation (FPC) submitted additional information to clarify the development of the design and licensing basis for the makeup pump discharge header cross-connect operation. FPC stated that the normal operating position for the common discharge header cross-tie valves MUV-3 and 9 are maintained in the open position with their respective circuit breakers locked in the off position (i.e.. power removed). The licensee committed to place valves MUV-3 and 9 for periodic stroke testing in the closed direction as part of the Crystal River 3 IST progra Based on its review. NRC concluded that valves MUV-3 and 9 should be in the open position and deenergized during normal operation, and they would be appropriate to include in the licensee's IST program. The inspector verified that the licensee had included valves MUV-3 and 9 in the ASME Section XI IST program for quarterly stroke time-testing in the closed direction. Therefore. URI 50-302/97-14-09 was closed and no violation of requirements was identified.
i 08.4 (Closed) VIO 50-302/97-07-02: Inadeauate Plannino and Controlling of Hydrostatic Testina (92901)
The inspector reviewed the violation dated July 7.1997. and the licensee's response in a letter dated August 5,1997. The inspector determined that a stand down meeting with the involved maintenance crew was conducte Maintenance Procedure MP-137. System Hydrostatic Pressure Testing, was revised to include additional guidance for the arotection of systems adjacent to the hydrostatic test pressure
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Joundary. An operations department night order addressing pre-job briefings and expectations was issued. Field engineering planners were l briefed on procedure revisions. The inspector verified that all of the required corrective actions had been completed. Therefore, VIO 50-302/97-07-02 was close ,
08.5 (Closed) VIO 50-302/97-11-02: Inadeouate Procedural Guidance for Quality-Related Work (92901)
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The inspector reviewed the violation dated September 12. 1997, and the licensee's response in a lethr dated October 8.1997. The inspector determined that Health Physics personnel assigned responsibility for providing ventilation for the once-through steam generator (OTSG) work were counseled for failing to secure the temporary purge fans prior to manway installation. Procedure OP-301. Operation of the Reactor Coolant System, was revised to include a caution statement that Reactor Coolant l level indication perturbation may be caused during the manway
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installation. Health Physics Block Training regarding Nuclear Safety
, was conducte Health Physics Coverage during OTSG work was proceduralized. Maintenance Procedure MP-110 A. Steam Generator Primary Side Maintenance, was revised to add steps requiring either one manway
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is still open or the inspection cover purge fans are secured prior to manway closure. The inspector verified that all of the required corrective actions had been completed. Therefore. VIO 50-302/97-11-02 was close .6 (Closed) VIO 50-302/98-01-02: Failure to Post Documents as Reauired by 10 CFR 19.11 (92901)
The inspector reviewed the violation dated March 4.1998, and the licensee's response in a letter dated March 31. 1998. The inspector determined that Nuclear Compliance Procedure NC-02. Posting Requirements, was issued on February 5.1998. to specifically address the postings of notices to workers. Also. Nuclear Compliance
" Regulatory Action Items" tracking list was revised to include a column for posting requirements. This will identify those violations and responses that need to be posted and preclude further omissions in postings. The inspector reviewed selected postings boards in the plant and found no discrepancie Therefore. VIO 50-302/98-01-02 was close .7 (Closed) LER 50-302/98-03-00: Loss of Power to Intearated Control System (ICS) Caused a Trio of the Reactor (92901)
The inspectors reviewed the licensee's final corrective actions for this item. As previously discussed in IR 98-03, although the licensee was unable to implement their corrective action to inspect other circuit boards in the plant with ICS in operation, they had also only committed to do spare boards in storage by the next major outage and not if replacing them in the plant. The licensee has subsequently implemented a corrective action under PC 98-0972 to inspect spare boards in the warehouse and trained instrument technicians to inspect boards being installed in the plant. This resolved the remaining inspector concern so this item is-close II. Maintenance M1 Conduct of Maintenance M1.1 General Comments (62707. 61726)
Using Inspection Procedures 62707 and 61726. the inspectors observed all or portions of several work requests (WR) and surveillance including WR 353987, for investigation and repair efforts on the IB emergency diesel exhaust gasket fire and WR 354189. AHF-20A Feeder Breaker Replacement -
ES MCC 3Al Cubicle 8B. These items are discussed further in Sections E2.2 and 01.2 res)ectivel The inspectors noted that all work observed was performed wit 1 the work packages present and in active us Technicians were experienced and knowledgeable of their assigned task The inspectors frequently observed supervisors and system engineers
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monitoring job progress. The inspectors did not identify any notable deficiencies.
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i M1.2 Turbine Driven Emeroency Feedwater Pumo Surveillance Inspection Scope (61726)
On May 22, 1998. the inspector observed the performance of Surveillance Procedure SP-349B. EFP-2 [ Turbine Driven Emergency Feedwater Pump] and Valve Surveillance.
l Observations and Findinas Prior to the start of the surveillance the inspector reviewed the procedure for conformance to Technical Specification requirements and general adecuacy. No concerns were noted. The inspector questioned the non-licensec operator (NLO) assigned to )erform the surveillance and determined that this was the first time le was to perform this particular surveillance. In addition, the inspector determined that the NLO had previously been in a licensed operator training program. The o)erator's supervisor also questioned him on previous performance of j tais surveillance and concluded that the NLO was knowledgeable and experienced enough to perform this surveillance. However, the 3rimary plant operator was directed by the Nuclear Shift Supervisor to )e available if neede The ins)ector attended the pre-job briefing and considered it to be thoroug1 and properly conducted. During the performance of the surveillance, the inspector observed the NLO consistently implement the S.T.A.R. (Stop. Think. Act. Review) process. The use of a second-party verification was utilized when called for and all test instrumentation was within its calibration due date and appeared to be functioning properly. At step 4.1.3.6 in the procedure. the NLO was directed to record the reading on the EFP-1 (motor driven emergency feedwater pump)
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discharge pressure gauge. A pressure of less than or equal to 200 l pounds per square inch gauge (psig) would indicate that emergency 1 feedwater valve EFV-35, which is a check valve in the recirculation line to the emergency feedwater tank, was closed. However, the pressure gauge indicated approximately 500 psig. Operations and In service Inspection (ISI) personnel decided to replace the gauge with a temporary calibrated gauge to assure themselves that the indicated pressure on the permanent gauge was reading correctly. The temporarily installed gauge indicated the same pressure as the permanently installed gauge. The ISI personnel re-installed the permanent gauge and the NLO continued on with the surveillance. No other problems or concerns were identified with the completion of the surveillanc Several 0)erations related issues were identified during the performance
- of SP-3493. These are discussed in Section 01.3 of this report. One of l these issues resulted in EFP-2 being declared inoperable due to excessive leakage past EFV-35. After the licensee flushed EFV-35 as
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part of the troubleshooting effort and determined that the valve was
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functioning properly, portions of SP-3498 were reperformed. The inspector reviewed the documentation for the reperformance and discovered several steps that ask if the surveillance was being
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performed to satisfy American Society of Mechanical Engineers (ASME)
Section XI quarterly requirements. If the answer was ~NO" the operator was instructed to go to another step in the procedure. A different NLO conducted the reperformance of SP-349B and both NL0s treated these particular steps differently. One NLO marked these steps "not applicable." The consequence of this was minimal because the proceeding steps instructed the NLO to aerform some valve stroke timing. The licensee acknowledged that t1ese ste a procedure revision would be done. Meanwhile, ps were confusing and SP-349A/B indicated were that placed on administrative hold for this reson and because engineering was going to perform a review of the accepance criteria for the different components in the procedures. This review would determine if the acceptance !
criteria was reasonable and necessary for all the components listed in the procedur Conclusions The inspector concluded that the NLO consistently implemented the S.T.A.R. process during the conduct of SP-349B. The use of a temporary calibrated gauge to aid in determining whether the permanent gauge was i functioning properly was considered to be a proactive and appropriate action. The inspector determined that procedure SP-349B contained steps that were not treated the same by different NL0s performing the l l procedure, however, this resulted in minimal significance because the
! operator only had to stroke a valve for timing purpose III. Enaineerina El Conduct of Engineering l l
EL1 Comoonent Failure Analysis for Fault inside Motor Control Center (MCC) '
, Insoection Scooe (92903)
l The inspector reviewed a preliminary failure analysis report, which covered a failure of MCC internal wiring. Inspection activity included
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examination of the failed equipment and examination of photographs showing the undisturbed post-failure condition of the equipmen Observations and Findinas
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On May 25, 1998, a failure occurred inside 480 V motor control center No.3A1, compartment 88, which controls air handling fan AHF-20A. This is a non-safety-related control complex ventilation system recirculation i fan. The failure occurred in the internal wiring which runs between the MCC removable compartment bus stabs and the compartment circuit breaker The equipment experiencing the failure is identical to safety-related equipment in type and applicatio _ _ _ _ - - _ - _ - _ . _ _ _ - - - . . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
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A preliminary failure analysis report had been completed. Shortly before this NRC inspection activity (June 15 and 16,1998L. the failure report was ] resented to a special licensee review committee, which made certain ver3al recommendations and comments for further investigatio A root cause analysis team had been formed having a charter to issue a final report by July 15. 1998. The licensee plans to incorporate the component failure report into the root cause analysis report. The licensee determined the event was not reportable under 10 CFR 50.7 The component failure analysis team developed an explanation of the failure as follows: one or more of the ring tongue connectors at the breaker line side lugs had a high resistance connection at the barrel to wire crim). This high resistance connection generated sufficient heat to melt t1e wire and connector barrel. At the same time, wire insulation material was charring and off-gassing. Energized parts are insulated from ground at this point so there was not yet any ground fault. When the connection burned open, an arc was drawn which caused gasses to become ionized. The ionized gasses caused a high resistance short-circuit between all three phases at the breaker line side lug All three connections burned open, thus clearing the fault. Fault current levels remained below the set point of the short time element of the trip device at the upstream circuit breaker. The failure took place
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within about one minute of energizing the load. The failure was ,
accompanied by a small explosion inside the MCC starter compartment '
l great enough to slightly deform the MCC starter compartment doo Other relevant facts are as follows; there was virtually no damage l electrically down stream of the line side lugs. Post-failure testing showed the internal insulation system of the MCC circuit breaker was l good. The upstream circuit breaker was working properly. There was no evidence of a pre-existing phase to ground or phase to phase leakage
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path. The failed wiring was original wiring. All lugs were the proper-size for the applicatio What was not explained, and may never be definitely determined. is what
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caused the high resistance connection at the lug to wire crimp. The
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licensee will attempt to answer this question by examining similar connections throughout the MCCs. The examinations will be made using thermography. If another hot connection is found, that may reveal a i plausible answer. In addition the licensee plans to obtain input from i
MCC manufacturers and the wire iug manufacturer. One possible cause of the high resistance connection could be an improperly made crim). Other i crimps in the failed compartment appear to be proper crimps. T11s !
statement was based on examination of the crimps and appropriate crimp L tool and a cross section cut made by the licensee. Some examples will j be sent to the lug manufacturer to obtain another opinion. A second.
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and equally important, reason for examination of safety-related motor i control centers is to determine whether any deterioration of corresponding wires or lugs can be observed. Inspection of the mo'.or
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control centers, as planned by the licensee, to look for any anomalies or deterioration of the stab to breaker wiring is an important part of ;
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After visual examination of the physical evidence and the licensee's analysis. the inspector concluded that the licensee had developed an explanation of the May 25. 1998. MCC equipment failure that fits the evidence and is supported by sound theory. The investigation is continuing. and may uncover new informatio l
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E2 Engineering Support of Facilities and Equipment j
E2.1 Emeroency Diesel Generator 1B Exhaust Gasket Fire Inspection Scone (37551) (93702)
The inspectors observed licensee corrective maintenance and reviewed the licensee's investigation following a small fire on the 1B emergency diesel generator (EDG) engine exhaust piping during routine testing on May 11. 199 Observations and Findinas Routine Surveillance Procedure (SP) 354B. Monthly Functional Test on IB EDG. was being performed when an operator detected excessive smoke t during initial loading of the engine. An immediate shutdown of the engine was performed from the control roo The local operator had discovered a small fire at one of the two exhaust header transition pieces which he extinguished with a small amount of dry chemical from a fire extinguisher. The control room entered the abnormal procedure for a plant fire but the event did not cause any actuations of automatic fire suppression equipment and no further actions were require The inspector observed the damaged area and verified it originated from the transition joint flange on the control side of the engine, just prior to the turbocharger. The inspector noted that the damage appeared minor and was limited to a discolored insulation blanket that was over the joint and a singed insulation jacket wrap on wiring to the EDG governor. The inspector verified the licensee inspected this wiring and determined the damage was superficial. limited to the wrap. and the wire and it's insulation were not damaged. The licensee disassembled the sus)ect flange and observed that a several inch section of the graphoil gascet was missing and had ap)arently been forcefully ejected. The licensee also discovered the Jolts holding the flange together were loose and that leakage was present on the EDG's opposite side exhaust header flange. The inspector also visually inspected the gasket and flange and confirmed the licensee's observations. To verify a common mode failure mechanism did not exist, the licensee immediately inspected the 1A EDG exhaust flanges. They observed that no leakage had previously occurred on that EDG due to the absence of any oil on the insulating blanket and any exhaust soot marks on the flange. They
, therefore considered the flange bolts tight and the 1A EDG operabl .
The inspector also inspected the 1A EDG and did not identify any problems with the licensee's determination. The licensee confirmed the
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1A EDG bolts were adequately torqued during the next scheduled surveillance outage of the 1A EDG. The inspector considered the licensee displayed prompt and appropriate consideration of the opposite train operabilit The licensee initiated a root cause investigation per PC 98-243 Although not finalized at the end of the report period, the inspector reviewed the licensee *s assessment and conclusions which were not expected to change. The licensee attributed the failure to the inadequate torquing of the flange bolts. They determined the leaking
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oil was lube oil from the engine pre-lubrication cycle that had not been l burned off because the engine was still lightly loaded. They found much Operating Experience (0E) information from other plants ~that indicated exhaust header leakage was an inherent problem with opposed piston diesels due to leakage into the cylinders from the pre-lubrication cycle
during lightly loaded operation. Other utilities had also seen problems l with loose exhaust header bolts and had instituted periodic torquing i requirement The licensee was unable to conclusively determine why the i
bolts were loose but discovered that their vendor manual and maintenance
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procedure did not specify a torque value for the exhaust flange bolt The bolts were therefore last torqued to a standard vendor bolting table l torque value. The licensee identified that the EDG had undergone an
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elevated number of thermal cycles since the flange was last torqued due
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to the extensive EDG radiator upgrade modification testing which may have contributed to the bolts loosening. The licensee was working with i the EDG vendor to determine if any higher torque value was desired and was planning periodic re-torquing of the flanges. The IB EDG flanges
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were repaired and the EDG was returned to service on May 1 c. Conclusions l
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The inspector concluded the licensee responded appropriately to the fire L
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in the diesel exhaust header flange. They displayed excellent concern for operability of the opposite train equipment and performed a thorough i
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preliminary investigation including significant industry experience research.
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E Miscellaneous Engineering Issues
'E f00en) VIO 50-302/97-14-13: Failure to Take Corrective Actions to
, :dentify Des;an Weakness for Past 10 CFR 50.59 Reviews for Positioning
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of DHV-34 and DHV-35 (92903)
Violation 50-302/97-14-13 was examined by the inspector to determine its .
L current statu )
Initially the violation was cited since the licensee failed to implement adequate corrective actions to identify and correct design weaknesses associated with the past 10 CFR 50.59 review for positioning of decay
, heat valves (DHV)-34 and DHV-35. The licensee identified the problem in September 1996. The proposed safety evaluation to support the position of DHV-34 and DHV-35 was not completed by the scheduled date of March _ - _ _ _ _ _ _ .
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1997, and still had not been performed by November 6. 1997. At this time. the violation was cited because DHV-34 and DHV-35 were maintained in the " closed" position without an adequate safety evaluation.
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The licensee issued PC 97-7755 to address the violatio In December 1997 they performed a Safety Assessment, an Unreviewed Safety Question Determination (USOD), an Operability Determination and Justification for Continued Operation. These items were discussed in detail in IR 50-302/97-19. No safety concern was identifie During this inspection, the inspector reviewed Florida Power letter i 3F0298-11 dated February 6, 1998, to the NRC. This letter stated that l the due date for the commitment to submit documentation to resolve the t
USOD for DHV-34 and DHV-35 was being extended to October 16. 1998. The licensee had made very limited progress in this area since the February 6. 1998 letter to the NR The inspector concluded that there were no safety concerns with the j documentation due date being extended to October 16, 199 E8.2 (Closed) LER 50-302/97-07-00 & 01: Buildina Temperature Variations Larcer Than Assumed. Resultina in Unknown Instrument Uncertainties
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LER 50-302/97-07-00 and -01 stated that on February 13, 1997, the l licensee discovered that the temperatures in the plant were not being
- maintained in accordance with the Environmental & Seismic Program Manual. The NRC cited the licensee with Violation 50-302/97-01-07 for
- this deficiency. This violation was closed in NRC Inspection Report 50-l 302/97-12.
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The inspector verified that the licensee completed implementing the necessary corrective actions to meet their commitments for closing this LER. The following corrective actions implemented were: 1) engineering ( completed an evaluation to determine operability of temperature variations in the various buildings: 2) engineering instituted
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requirements for temperature calculations and instrumentation l calibrations: 3) engineering established ambient temperature requirements for the instrumentation calibrations in all buildings: 4)
l management established controls for maintaining the required l temperatures in all buildings: 5) Instrumentation & Control (I&C)
l calibrated all Technical Specification *s temperature instrumentation:
i and 6) engineering revised Nuclear Engineering Procedure (NEP)-213 to l- include requirements for all engineering calculations including temperature. The inspector concluded that appropriate corrective
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actions were implemented by the licensee and this LER was close f
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E8.3 (Closed) VIO 50-302/97-14-05: Failure to Provide Adeauate Procedure for l Calibration of Reactor Vessel Level Instrumentation Reduced Inventory Doeration (92903)
This violation resulted from an inadequate procedure for calibrating the reactor vessel level instrumentation. The level instrumentation was calibrated on February 21, 1997. On February 25. 1997, the level instruments were readjusted nine inches to match the vessel level
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I reading provided by the tygon tubing. The nine-inch readjustment resulted in the reactor vessel level instrumentation being outside the calibration acceptance criteria in Surveillance Procedure (SP)-195.
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Calibration procedure SP-195. Remote Reactor Vessel Level Instrument Calibration. directed the technician to readjust the level instrumentation to match the tygon tubing reading following completion of the normal calibratio The inspector verified the licensee had implemented corrective action by i
revising procedure SP-195 and instrumentation calculation procedure I-90-1017. Reactor Vessel Shutdown Level Instrument Loop Tolerance Scaling, and Setpoints. The calculation was revised to include compensation for density effects from temperature and boration conditions. The results of revised calculation I-90-1017 were incorporated in procedure SP-195. .The inspector concluded that i
appropriate corrective actions were implemented and this violation was j closed.
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E8.4 (Closed) VIO 50-302/96-01-05: Two Examoles of Failure to Uodate FSAR as Reauired by 10 CFR 50.71(e) (92903)
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1) In 1986 the licensee made a modification to the make-up system regarding an interlock installed to open the borated water tank
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isolation valves, makeup valves (MUV)-58 and MUV-73. on a low tank level l
and locking open MUV-64 to satisfy 10 CFR 50. Appendix R requirement No revision was made to the FSAR to address the interlocks for the valves.
l 2) The design basis of the spent fuel pool system as revised by License l l
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Amendment 134 issued on April 16, 1991, was not incorporated into the FSAR as follows: 1) 1180 fuel assemblies were used instead of 1357; 2) ,
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FSAR Table 9-6 incorrectly states 16 refuelings instead of 191/3: 3) l 140 degrees F was used instead of 157 degrees F: and 4) it incorrectly stated that leakage from spent fuel pool through the leak chase is !
monitored daily.
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The inspector verified the leakage was monitored on each shift daily as i required by Surveillance Procedure SP-301. Shutdown Daily Surveillance Log. The inspector also verified that the Safety Evaluation Report (SER) for Amendment 134 issued by the NRC stated that the spent fuel coolant temperatures of 157 degrees F complied with the guideline
, temperature of less than 212 degrees F and was therefore acce) tabl !
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The licensee revised the tem)erature in the FSAR to be less tlan 160 degrees F. which was accepta)l l l
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' The inspector verified that the FSAR was revised to reflect the low level interlock for the isolation valves and the items in License Amendment 134 were also incorporated. This violation was closed.
l E8.5 (Closed) LER 50-302/97-34-00: Nonconservative Assumotion in Breaker I Setooint Calculations Could Prevent Startino of Motor Ooerated Eauioment l (92903)
- The licensee determined that set)oint calculations for safety-related
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molded case circuit breakers wit 1 adjustable instantaneous trip settings did not provide adequate allowance for breaker tolerances for greater than nominal voltages at the motor terminal. The trip settings were set at a notch between the low value of 1.6 times locked rotor current and the high value of 13 times locked rotor current. The instantaneous trip setting adjustment for the breaker consisted of a screw that could be moved to several fixed notched positions such as 0, 25. 50. 75. and 100%. The concern was breaker tolerances at the notched position could be higher than stated by the vendor The licensee revised the electrical calculations in procedure EDC Electrical Design Criteria - Molded Case Circuit Breaker Trip Setting, by raising the lower trip valu Plant Modification Approval Record (MAR) 97-10-11-01. MCC Motor Circuit Protector Trip Setpoint Evaluation, was implemented to reset the trip settings. The breakers available for on-line maintenance were reset. The licensee committed to reset the remaining breakers during the next refueling outage. A safety assessment and justification for continued operation were satisfactorily performe Work orders for resetting the remaining breakers were issued. The licensee was tracking the completion of the MAR to ensure it was completed as scheduled through the Nuclear Operation Tracking &
Expediting System (NOTES). No. 2719 The ins)ector examined the
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operating conditions for electrical power in tie plant. The plant did not have a history of these breakers false tripping on high voltage conditions. The switchyard was being maintained within the 238 - 242kV range. The main concern with the switchyard voltage was not high voltage; but a degraded voltage condition. The insnactor concluded there was no safety concern and the licensee had su'fficiently addressed this item. This LER was close Although this item was a noncompliance with regulatory requirements, for reasons discussed in Inspection Report 50-302/97-21. the licensee met the criteria for enforcement discretion per Section VII.B.2 of the NRC Enforcement Policy as described in NUREG-1600. Consequently this item was closed and was identified as another example of Non-cited Violation .
NCV 50-302/97-21-01. * Examples of Noncomallance in Design Control. 50.59 Evaluations. Procedure Adequacy. fleportaaility, and Corrective Actions l That are Subject to Enforcement Discretion."
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l 20 E8.6 (ClosedTLER 50-302/97-11-00: Pressure Lockina in Emeroency Feedwater l
Valves EFV-32 and 33 (92903)
l The licensee identified in a readiness review project conducted on the Emergency Feedwater System that isolation valves EFV-32 and EFV-33
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could have a safety function to open for certain scenarios where they
- close due to a single failure then must later open to fulfill their safety function. The previous analysis excluded these valves from consideration of pressure locking based on there being no safety function to open. This new information meant that the pressure locking should be considered for these valves. potential fortwo There were corrective actions stated in the LER. First, the valves would be modified by drilling a hole in the disk to preclude ' pressure locking
, and second, the entire analysis for pressure locking / thermal binding of
- - valves would be re-reviewed in light of the new informatio The inspector reviewed Field Change Notice No.3 to MAR 96-10-10, and confirmed that the hole-drilling modification was. implemented. In addition, the inspector confirmed that administrative controls were in
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place to hel) ensure that this modification would be maintained in the event that t1ese valves would be reworked in the future. The control was to place a note in the equipment list against these valves ( indicating they had a hole drilled in the disk. The equipment list
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should always be checked by planners when preparing work orders thus
3 helping to ensure that the modification would be maintained in the
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l l The inspector reviewed Calculation M94-0003. Pressure Locking and Thermal Binding Evaluation. Revision 3. dated' December 16. 1997, and l confirmed that the re-review committed to in the LER had been performed.
l The re-review did not identify any problems similar to that described in
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the LER, and the inspector agreed with this conclusio .In addition, the inspector reviewed the following two documents:
. Letter FPC to NRC, dated September 14, 1996, on subject of l " Response to Request for Additional Information on Generic Letter i 95-07, Pressure Locking and Thermal Binding of Safety-Related l Power-Operated Gate Valves".
. Letter. NRC to FPC, dated October 16, 1997, on subject of " Crystal River Unit 3 - Safety Evaluation of Licensee Response to Generic l Letter (GL) 95-07. Pressure Locking etc...."
Based on the above described inspection activity, the inspector concluded that the issue described in LER 97-11-00 was resolve Although this item is a noncompliance with regulatory requirements, for the reasons-discussed in Inspection Report 50-302/97-21. the licensee
~, ' meets the criteria for enforcement discretion per Section VII.B.2 of the NRC Enforcement Policy as described in NUREG-160 Consequently this item is closed and is identified as another example of Non-cited
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Violation NCV 50-302/97-21-01. Examples of Noncompliance in Design ,
Control. 50.59 Evaluations. Procedure Adequacy. Deportability, and '
Corrective Actions That Are Subject to Enforcement Discretio E8.7 (Closed) VIO 50-302/97-11-07: Deletion of Water Quality Requirements o from the FSAR (92903)
i The corrective actions for this violation were reviewed during a
]revious inspection, and those results were documented in NRC Inspection i Report 50-302/98-04. This inspection focused on the extent of condition reviews conducted by the licensee as part of the corrective actions, and focused specifically on FSAR related issues. The extent of condition reviews consisted of looking at text taken from the standard technical L specification and placed in the FSAR and plant procedures at the time of conversion to-the improved technical specifications. A matrix or road
! map was already available to help expedite these reviews. Any difference between the original text and the current text was highlighted. These differences would not necessarily mean there was a problem.- because text transferred from the original technical specification to the FSAR or a procedure may be changed as long as documented safety evaluations are performe The inspectors reviewed most of the identified cases where text transferred to the FSAR differed from the original standard specification. The inspectors agreed that the changes to the ;
L transferred text were properly handled. The situation of the violation.
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i e. deletion of water quality standards from the FSAR. was a unique case where the subject text was moved to a procedure after moving to the FSAR. Thus, the final corrective action to perform an extent of condition review was adequately completed, and the violation was close ' E8.8 ~ (Closed) VIO 50-302/97-17-01: Failure to Conduct an Adecuate Unreviewed L Safety Ouestion Evaluation for a Modification Functional Test (92903)
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l The cause of the problem described in this violation was lack of L procedural guidance for develo] ment of test procedures. Accordingly the-
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corrective actions stated in tie revised response to the violation.
l . dated March 12, 1998, specified certain specific procedure changes. The
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inspector confirmed that Compliance Procedure (CP)-134.. Revision 2 Preparation and Approval of MAR Functional Test Procedures, effective l April 15.-1998. Steps 3.2.2.1. e, h. and g, contained additional
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guidance over previous revisions to help preclude the problem that occurred. These steps had been revised or added in accordance with the response to the violation. The inspector also confirmed through review L of.the following procedures they had been revised and/or were consistent l= with each other as stated in the response to the violation:
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- Administrative Procedure AI-550. Infrequently Performed Tests or Evolution ?
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Administrative Procedure AI-400C. New Procedures and Procedure Change Process. Revision 24. dated April 9.199 .
Compliance Procedure CP-213. Preparation of a Safety Assessment l and Unreviewed Safety Question Determination (10 CFR 50.59 Safety l Evaluation). Step 3. 'In summary, the inspectors confirmed that the corrective actions stated in the response to the violation were implemented, and the violation was close E8.9 (Closed) IFI 50-302/97-12-03: Enclosure 17/18 Interaction (92903)
The inspector reviewed the flow diagram for the Control Complex Ventilation System. PC 3-97-8564 (closed status) documented the resolution of the concern expressed in the IFI. The resolution was to perform a calculation to support development of procedural o)erator guidance for realigning the chilled water system to supply t1e running air handling heat exchanger (AHHE-5A or 5B). After the calculation was done. it was routed to operations aersonnel who identified four procedures that could be affected Jy the conclusion of the calculatio The calculation also identified that the position of certain valves (and corresponding flow rate) in the chilled water loop could not be definitely ascertained by document review. A PC and work request were generated to make flow measurements on the system so that all design inputs to the calculation would be verified inputs, and this work was don The develo)ed operator guidance was to close down a specified number of turns on t1e chilled water outlet valve of the previously running heat exchanger. then sequentially open a specified number of turns on the corresponding valve at the running heat exchanger. shut off flow to the
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shut down heat exchanger, then fully open the valve to the running heat exchanger. This 3rocedure would prevent tripping of the chiller on low flow as well as cliller pump motor overloa The inspector reviewed all the procedures and the calculation results, and concluded that the issue was resolved. The inspector noted that the PC evaluation stated that there was an error in the Enhanced Design Basis Document (EDBD) concerning the position of certain chiller inlet valves, but there was no corresponding corrective action in the PC. The licensee initiated PC 3-C98-3007 to correct the EDBD and to document incomplete corrective action on the first P E8.10 (Closed) VIO 50-302/98-02-02: Failure to Provide Adeouate Instructions for Installation of LR-82-FE/FI and LR-83-FE/FI (92903)
Originally. Emergency Plan Procedure (EM)-225A. Post Accident Reactor j
Building Hydrogen Control. did not give adequate instructions for
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installing flow elements to be used for purging the reactor building to maintain safe post-accident hydrogen concentrations. Corrective actions included creation of Maintenance Procedure MP-815. Installation of Post-
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Accident Hydrogen Purge Flow Instruments, which became effective February 17. 1998. Secondly. Administrative Instruction AI-400F New Procedures and Procedure Change Process for Emergency Operating Procedures (EOP). Abnormal Procedures (APs) and Supporting Documents, was revised with Revision 5. effective May 27. 1998, to specify full review of supporting procedures, which should help prec'lude the problem of the violatio Similarly. AI-402C. AP and E0P Verification and Validation Plan. added ac'ditional guidance. An additional corrective action was to review the adequacy of other supporting procedure Evidence of this review was provided in Interoffice Correspondence (IOC)
OP98-0016 dated March 27. 1998. This IOC lists the specific supporting actions reviewed. Review of these actions and the guidance provided identified no examples where there is inadequate guidance for critical actions required for accident condition E8.11 (Closed) VIO 50-302/97-16-04: Failure to Follow Procedure CP-111 by not Performina a 10 CFR 50.59 Safety Evaluation Within 90 Days After Identification of a Non-conformina Condition Which Conflicted with the FSAR Description (92903)
The corrective actions for this violation were reviewed in previous inspections, and the results were documented in irs 50-302/98-03 and 50-302/98-04. This inspection focused on the extent of condition review which the licensee conducted as part of the corrective actions. The objective of the extent of condition review was to identify cases where a PC was initiated for a problem that could represent a non-conformance, and the evaluation of that PC was not completed and the 90 days for such evaluations had elapsed. The scope of the review was open PCs initiated prior to October 1. 1997. and there were 610 PCs in this category. The review went beyond the scope of the violation in that it was not limited to FSAR discrepancies. There were 74 PCs in the set of 610 classified
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as non-conformances. Evaluations by the licensee determined that 32 of these non-conformances had been corrected. A summary report of the remaining 42 PCs in the set was generated. Licensee evaluators determined that 28 of the set of 42 would never result in a Deficiency Report through evaluation of the summary statement itself. The remaining 14 PCs were evaluated by the respective responsible engineers, and at least six of these did not result in a Deficiency Report. The inspector evaluated the specifics of the PCs on the summary list, and agreed with the licensee's dispositio One conclusion made by the licensee as a result of the extent of condition review was that the problem of not meeting the 90-day requirement for evaluating PCs representing potential equipment deficiencies was extensive. The licensee took at least two corrective actions to prevent this problem. First. the PC form itself was revised to provide a field for indicating whether a Deficiency Report evaluation is required and who is responsible for the evaluation. Second the licensee initiated daily tracking of the required evaluations to ensure
, the 90-day requirement is met. The inspector confirmed these corrective actions through review of recent PCs and the daily tracking shee _ _ - _ - _ _ _ _ _ _
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24 E8.12 (Closed) VIO 50-302/97-13-03: Failure to Follow Procedure for Control 11na Breakers Removed from Switchaear Cubicles (92903)
This violation involved a situation where an NRC ins)ector, in a
, walkdown inspection, identified that 4160 V circuit areakers had been removed from the switchgear and were standing unrestrained adjacent to safety-related equipment. This situation was not consistent with seismic design. The inspector confirmed that Preventive Maintenance Procedure PM-101. 4.16 and 6.9 kV Switchgear Preventive Maintenance Procedure. Revision 26. dated May 29. 1998, Step 3.2.4. was an instruction to place.any removed circuit breaker in the approved storage racks or remove the circuit breaker from the Engineered Safeguards switchgear room. The inspector went to the switchgear rooms (both trains) and verified that seismically designed circuit breaker restraining racks had been installed. There were two wall mounted restraining racks in both the Train A and Train B switchgear rooms. The inspector also walked down the remote storage location referred to in
- the PM procedure and observed that it was a safe. storage locatio E8.13 (Closed) VIO 50-302/98-01-07: 500 kV Backfeed Not a Qualified Source of Offsite Power (92903)
(Closed) VIO 50-302/98-01-08: Inadeouate Procedures for Use of 500 kV Backfeed (92903)
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(Closed) LER 50-302/98-02-00: Unqualified Power Source Used for Safety-Related Eauioment (92903)
An NRC inspection, conducted'in 1996. questioned the adequacy of a power source known as the 500 kV backfeed. This power source, which would be used during plant shutdown modes, consists of isolating the main
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generator and allowing power to flow from the 500 kV buses, through the generator step-up unit and the unit auxiliary transformer to plant auxiliary buses including the safety-related buses. The NRC originally questioned the adequacy of this source because they were aware of a l problem report describing a case where an inverter switched from normal AC source to backup DC source upon starting a motor while aligned to the 500 kV backfee It was later determined that the inverter switch over never actually occurred, but there was an incorrect set point for the alarm which indicated switch over. The NRC had also-identified that the 500 kV backfeed had not been analyzed by any formal calculations. When
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calculations were performed to support use of the 500 kV backfeed, the-
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calculations indicated that the source had limited capacity as compared to the other offsite power sources. The calculation. E96-0004 Revision 0. 500 kV Backfeed Voltage Drop. Load Flow. Motor Starting. Short Circuit and Parallel Operation Analysis, dated August 21, 1997, recommended that administrative controls be implemented to allow use of this sourc .
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25 Engineering and Operations developed the following. very conservative, restrictions on use of the 500 kV source:
. Only one train of safety-related equipment would be powered from the 500 kV backfeed at any one tim .- The source would only be used in Modes 5 and 6.
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l . Not more than 726 A load on the unit auxiliary transformer 4160 V l winding.
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.. 500 kV bus voltage at 515 - 530 k . When backfeed is in use, the Energy Control Center shall enable the under/over voltage alarm, which is set at 517/528 kV.
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The following procedures were reviewed and revised appropriately to ensure the above stated restrictions are adhered to:
L e Operations Procedure OP-209. Plant Cooldow . Operations Procedure OP-703A. Establishing. Maintaining and Removing 500 kV Electrical Power Backfee . Administrative Instruction AI-504. Guidelines for Mode 5 Outages and Reduced RCS Inventory Operation . Surveillance Procedure SP-301. Shutdown Daily Surveillance Log.
, . Surveillance Procedure SP-321. Power Distribution Breaker Alignment and Power Availability Verification.
l- The inspector reviewed these procedures and confirmed that the guidance was sufficient to maintain the self imposed restrictions on use of the 500 kV backfeed which was well within the limits determined in the analysis of this sourc E8.14 (Ocen) LER 50-302/97-25-00 and 01: Service Water Raw Water Temperature Calculation Contains Non-conservative Assumptions-
,a. Insoection Scooe (92903)
As previously discussed in IR 50-302/98-01. the licensee analyses performed in 1987 to increase design basis seawater ultimate heat sink (UHS)-temperature from 85 to 95 F were deficient in considering limiting combinations of running service water (SW) and raw water (RW) pumps and reactor building (RB) fan fouling factors. An inspector previously '
reviewed a Justification for Continued Operation (JCO) under Deficiency Report PC 97-5265, which administratively limited the maximum UHS -
, temperature to 79.9 degrees F. A modification involving a change to the starting-logic for Reactor Building (RB) Air Handling Fans (AHF) 1A and
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IB to ensure only one fan starts automatically on an Engineered
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Safeguards (ES) signal was identified as an Unresolved Safety Question per.10 CFR 50.59 and therefore required NRC review and approval prior to implementing. The licensee submitted License Amendment Request 224 to the NRC on December 5. 1997 which was then issued by the staff as License Amendment 165 on March 9. 199 Following receipt of this, the licensee installed the modification under Modification Approval Record (MAR) 97-09-05-01 on March 19 for AHF-1A and March 21 for AHF-18. The inspector reviewed the scope of the modification and verified the j installation training and procedure changes were complete and that the temporary operational limits imposed in the JCOs had been modifie Observations and Findinas As discussed in section 07.2 the inspector identified that the two precursor cards-(97-8080 and 97-5265) covering these UHS issues were closed but the associated Deficiency Reports (DR) containing the active JCOs remained open, contrary to administrative. procedural guidanc This created an administrative tracking problem but did not affect the adequacy of the JCO. The inspector reviewed each of the DR and JC0 conclusions and noted that the RB Fan problems in PC 97-5265 were corrected by the MAR installation and the DR and JC0 were appropriately closed. However, the inspector noted that an administrative UHS limit of 89 degrees F remained in effect after the MAR due to. separate problems with control complex habitability envelope (CCHE) chiller
. starting problems at elevated SW temperatures which were delineated in PC 97-8080. This PC and the CCHE chiller problems were initially screened by the licensee as not reportable when first discovered in December 1997. The licensee based this conclusion on the assumption
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that the issue was similar to the RB Fan problems and was encompassed in
- .LER 50-302/97-25 as another problem with the analysis to upgrade the UHS-temperature from 85 to 95 degrees F. This assumption was incorrect
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because the ins)ector determined the CCHE chiller problem was an independent pro)lem from the RB fan issue and was not discussed in either revision 00 or 01 of the LER. All written correspondence to the NRC on the RB Fan issue, including the License Amendment Request (LAR).
implied that the UHS limit would be restored to the full 95 degree L Technical Specification limit when the LAR was approved and the MAR installed. The DR for PC 97-8080 had been revised several times since March to raise the UHS limit as a result of ongoing engineering and vendor analysis efforts. The ins)ector concluded that the CCHE chiller L problems described in PC 97-8080 lad not been reported. The licensee L was processing revision 02 to LER 50-302/97-25 to correct the oversigh !
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The inspector reviewed the final field change notice (FCN) 5-to MAR 97-09-05-01 which implemented the last phase of the installation following receipt of the license amendment. The inspector noted that FCN 5 also changed a time delay relay setting from six. down to three seconds, i This relay controlled the time delay to start the second RB fan if the first were to fail to start and was described in detail in the LAR
' submitted to the NRC. The inspector determined that the licensee had l not updated their LAR to reflect this change for the NRC reviewer's i consideration. The inspector verified the Safety Evaluation Report for ;
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Licensee Amendment 165 did not specify a value for the time delay and l confirmed with the reviewer it was not a determinant in approving the change. However, the inspector considered this indicated a flaw in the licensee's process for ensuring license changes a) proved by the NRC were implemented per the requirements of the SER and t1e assumptions the approval was based upon. The inspector questioned the level of control on the modification process for MARS associated with license actions and determined they were informal. The licensee considered the failure to update the LAR an oversight and a minor degree of detail change.
l However, they were reviewing their licensing process to ensure the problem would not recu The inspector verified the licensee revised operations procedure OP-103B. Plant Operating Curves, and associated Curve 15 as well as surveillance procedure SP-300. Daily Logs, to reflect the changes in the
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UHS limits. The inspector reviewed the return to service package for MAR 97-09-05-01 and identified some concerns with the timing and adequacy of training for the operators. These concerns had not been resolved by the end of the report perio c. Conclusions During the review for closure, the inspector identified several discrepancies associated with this item had yet to be resolved so this item was not closed. Administrative tracking of precursor card deficiency reports was poor. A design issue with the CCHE chiller ability to automatically start at elevated SW temperatures had not been adequately reported to the NRC and required a supplement to the LE Changes had been made by the licensen t o engineering design assumptions in their modification package after o hcense change request was submitte The licensee did not then revise the change request to
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notify the NRC reviewers. The inspector concluded the licensee controls of modifications associated with licensing actions were informa E8.15 (Closed) IFI 50-302/96-15-03: Actions Taken to Resolve !
Recriticality Concerns Due to Localized Boron Dilution (92903)
This item was originally opened to track the licensee's resolution of a Babcock and Wilcox Owners Group (BWOG) Preliminary Safety Concern on the ,
potential for reactor recriticality due to localized boron dilution and a reactor coolant pump restart following loss of subcooling. The licensee had initiated Problem Report (PR) 96-0437 to track resolution of this problem by the BWOG. The BWOG issued final Emergency Operating Procedure (EOP) guidance in a letter to the licensee from the BWOG Operator Support Committee dated March 18, 1997. Resolution was determined to be emergency operating procedure (EOP) changes to limit allowed reactor coolant pump (RCP) restart following a loss of subcooling. The inspector reviewed the guidance and verified it was :
implemented by the licensee in Enclosure 16. RCP Recovery, of revision 2
, of E0P 14. E0P Enclosures. issued on January 14. 1998. However, the inspector identified that the E0P-14 enclosure guidance for restart of a RCP if natural circulation conditions existed in only one loop stated
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.that it was preferred" to start the RCP in the loop with the natural circulation. The final BWOG resolution stated that the RCP in the loop with natural circulation conditions "must" be the one restarted. The-licensee initiated PC 98-2916 to investigate the discrepancy and issued Operations Short Term Instruction (STI)98-026 as a short term corrective action to ensure the operators started the correct RCP. The inspector verified this had been completed. The licensee's investigation determined that the correct "must" terminology had originally been accurately incorporated into a draft of E0P-14 but that it had been subsequently revised again in a subsequent draft to the
" preferred" terminology and was issued with that terminology in revision 2. The licensee was still. investigating-the cause and the safety significance of the error at the end of the report perio The corrective actions.for PR 96-0437 also included actions to develop programmatic controls for future core reloads to ensure adequate review was performed to assess any resultant impacts on the RCP restart criteria. The inspector verified the licensee implemented these controls via revision 8 of Nuclear Engineering Procedure (NEP)-28 Reactor Core Design. The inspector noted that the BWOG Preliminary Safety Concern resolution was in the reference section of NEP-281. but it referenced letter ESC 96-560 dated October 15. 1996. This letter was an earlier draft of proposed BWOG guidance for the problem. The final BWOG resolution was issued in letter INS 97-1040 issued March 18, 199 The licensee revised the NEP to reflect the correct lette Based on the inspector's reviews. the licensee has implemented the BWOG guidance adequately to address the original IFI concern, so this item is close However, the ins)ector identified two discrepancies with the licensee's implementation tlat could have resulted in incorrect operator action and reference to incorrect. procedure source requirements. These were under-investigation by the licensee at the end of'the report perio . .
IV. Plant Support R8 Miscellaneous RP&C Issues R8.1 (Closed) VIO 50-302/97-20-02: Failure to Follow Radiation Protection
, ,Proaram Procedures for Documenting Personnel Contamination Events n192904 The licensee committed in their response to the violation to:
e= Make Health Physics Supervisors and aersonnel aware of this finding during a stand down on Decem)er 10 and 11.199 * Review all forms being maintained in the Health Physics functional area, and verify that the forms were the current revisio ,
e Promulgate Management's expectations, including expectations to follow procedures. The expectations were included in the HP
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. Add Section 5.0 to the Follow-Up Actions of Health Physics Procedure. HPP-110. Radiation Protection Forms. This section was directed at ensuring obsolete HP forms are not available for us . Emphasize during training, that management expected completed documents to be reviewed for accuracy. legibility, and completenes The inspector reviewed and verified: changes to procedure HPP-11 Section 5.0. were adequate corrective actions, current forms were in place in the Health Physics functional area, and completed contamination forms in the Health Physics functional area were properly complete The inspector reviewed attendance sheets and documentation of the December 10 and 11. 1997. Health Physics area stand down informing Health Physics personnel of the findin The licensee corrective actions for the violation were appropriate and complet R8.2 (00en) VIO 50-302/97-20-04: Failure of Radiation Workers to Follow Contamination Control Procedures (92904)
The inspector reviewed the licensee ~s proposed implementation of their response to the violation. The inspector's review concluded that the corrective actions were not effectiv The licensee had committed to incorporate the violation as a " lessons learned" into contractor in-process training. The inspector reviewed the ~ lessons learned" associated with the violation in General Employee Training manual and concluded that there was no discussion as to ap]ropriate corrective actions or expected action the worker should have tacen. The ~ lessons learned" was not self critical. The licensee ,
Health Physics department did not share the responsibility for the j violation by the contract workers nor was a " root cause~ identified and I discusse The inspector did not close the violation. The licensee had not completed their commitments in response to VIO 50-302/97-20-0 R8.3 (Closed) VIO 50-302/98-03-02: Failure to Label a Container of Radwaste in Accordance with 10 CFR 20.1904(a) Requirements (92904)
The licensee's response to the violation contained the following corrective actions:
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. Place another radioactive material label on the subject Outside Storage Shield Cask (OSSC) and adjacent OSSC. The remaining OSSCs were checked for visible labels. Other containers in the Radiation
, Controlled Area and outside the radioactive material storage warehouse were checked for proper labeling.
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e L Write a Health Physics Supervisor information notice regarding l
visibility of container label .
L . Cover the violation and corrective actions taken in continuing l training.
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. Modify Health Physics Procedure HPP-202A Radiological Surveys and Inspections.-survey routine forms to designate the areas and l- container checking requirements.
l . Develop a document standard covering area posting and container l labeling practice The insoector toured the OSSC storage area and the Radiation Controlled Area, and verified that the posting committed to in the licensee's L response had been performed. The inspector observed that labels on other
- radiological containers in the area were clearly visible and contained
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the appropriate radiological information. The inspector verified that:
the committed training was completed, procedure HPP-202A, Radiological s Surveys and Inspections, survey routine forms were modified to designate
~ the areas and container checking requirements, and that a documented
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practice The licensee corrective actions for the violation were appropriate and complete R8.4 100en) IFI 50-302/98-03-03- Review Licensee Maintenance on the RM-Al l Sample Line and Licensee Evaluation of the Adecuacy of Rubber Tubino in )
Iodine Samolina Systems (92904)
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The licensee's resolution to the question of iodine adsor) tion into
! rubber. hoses used in RM-Al sample lines was to contact tie present vendor. Nuclear Research Cor oration. and perform an Internet solicitation of approximately fifty Radiation Monitoring system engineer Nuclear Researc did not provide any technical response but L
L the licensee received eleven responses to their Internet solicitatio Ten of the responders stated that they used all stainless steel. One of the responders used some rubber hose in their ventilation radiation monitors. One responder commented that iodine absorption may be a valid concer The ins)ector verified by drawing and actual.. system observation that rubber loses were used extensively in the licensee's radiation monitoring systems. The ins)ector concluded that the licensee's response did not adequately answer tie. question of iodine absorption in rubber hoses. The licensee plans to perform additional containment air sampling i to determine the impact of using rubber hoses. The IFI was left open, pending the results of this containment air samplin l l
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R (Ocen) VIO 50-302/98-03-04: Failure to imolement licensino commitments to meet TMI Action item II.F.1-3. calibration of all ranaes of CHRMs The inspectors did not close VIO 50-302/98-03-04. The licensee had not yet completed all of their commitments in response to the Violation at the time of the inspectio P1 Cenduct of EP Activities Pl.1 Emeraency Plannino Drill Insoection Scooe (71750)
On May 27, 1998, the licensee conducted an Emergency Planning (EP) J drill . The inspector observed the drill from the Technical Support Center / Operational Support Center (TSC/OSC). Observations and Findinas One of-the purposes for the drill was to test- the adequacy and
' effectiveness of the TSC/OSC functions subsequent to the building being removed from the protected area boundary to support the construction of additional floors. Ingress and egress to the plant protected area was accomplished, and would be accomplished in the case of a real emergenc through temporary gates that were erected along the northeast berm are Security personnel were stationed at these gates to assist OSC 3ersonnel entering the plant. This was accomplished by the use of a log )ook and picture identification for each individual obtaining access through the temporary gate The recent renovations of the TSC/0SC in 1997 were successful in
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providing an area that will better serve its purpose during use. The inspector verified that all NRC equipment was in sufficient working
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order and that the areas provided for NRC personnel were adequate for ,
E their purpose. Communications during the drill were frequent and I informative, however some communications between various coordinators '
i- did not always include the Emergency Coordinator (EC). At times. the activities in the OSC appeared uncontrolled and hurried, but the l resultant activities to respond to plant conditions as directed were not :
hampered. There was some confusion when the simulated accident !
! progressed to a point where site evacuation was required and the l L security coordinator made a recommendation to suspend safeguards. The ,
confusion came when the security coordinator based the recommendation to !
suspend safeguards _on 10 CFR 50.54(x) and (y). Apparently in past l licensee drills this had not been done. Suspension of safeguards would '
have delayed response teams entering vital areas because with the TSC/0SC outside the protected area (PA), Security personnel would have )
to telephone the Central Alarm Station and provide the badge numbers of persoi.nel entering the PA. If there were no security personnel at the
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temporary entrance to the PA. then the response teams would not be logged into the security card reading system and therefore would be unable to card into vital areas. However the control room maintains vital area keys and would provide them to response team personnel as neede c. Conclusions The inspectors consider the method of logging and positively identifying individuals with the use of photographs upon entering the protected area I from the TSC/OSC appropriate and sufficient. Some communications amongst the various coordinators that could lead to an emergency response were observed to circumvent the Emergency Coordinator depriving him of relevant information to help in his decision making. Suspension of safeguards controls was not well understood by drill participants outside of the Security organization. The inspector concluded that, in general, the EP drill was successful in accomplishing it's objective S2 Status of Security Facilities and Equipment S2.1 Temocrary Relocation of Protected Area Fence (71750)
On May 21. 1998, the protected area (PA) boundary was relocated to allow the Technical Support Center / Operational Support Center (TSC/OSC) to ;e outside the PA. This action was taken to su) port a construction ,,,oject to add three floors to the existing TSC/OSC auilding. Compensatory measures have been established to ensure that there will be no decrease ;
in security effectiveness per 10 CFR 50.54(p). These compensatory measures also ensure that the temporary modification does not decrease the effectiveness of the Crystal River Unit 3 Radiological Emergency l Response Plan per 10 CFR 50.54(q). The inspectors reviewed the
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compensatory measures and inspected the temporary fence: no concerns I were note I S8 Miscellaneous Security and Safeguards Issues S8.1 (Closed) VIO 50-302/97-07-06: Inadeouate Corrective Action For Use of an Uninterruotible Power Suoolv Unit in the Security Central Alarm Station (92904)
The inspector reviewed the corrective actions described in the licensee *s response letter, dated August 5. 1997, and determined them to be reasonable and complete. The inspector interviewed several alarm station o>erators to verify their knowledge of the backup power supply
, . syste Vo further problems were identified; therefore this item is l considered closed.
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V. Manaaement Meetinas
- X1 Exit Meeting Summary L The inspection scope and findings were summarized on June 23, 1998.
! Proprietary information is not contained in this report. Dissenting l comments'were not received from the license PARTIAL LIST.0F PERSONS CONTACTED Licensees j l
R. Anderson. Senior Vice President. Energy Supply-J. Baumstark. Acting Director, Nuclear Engineering & Projects S.-Bernhoft. Manager. Nuclear Licensing J. Cowan. Vice President. Nuclear Operations R. Davis. Assistant Plant Director. Operations and Chemistry R. Grazio. Director. Nuclear Regulatory Affairs
.G. Halnon. Acting Director. Nuclear Quality Programs B. Hickle. Acting Director. Nuclear Operations Training J. Holden. Director. Site Nuclear Operations M. Marano. Director. Nuclear Site & Business Support
- C. Pardee. Director. Nuclear Plant Operations W. Pike. Manager. Nuclear Regulatory Compliance 1 H. Schiavoni . Assistant Plant' Director. Maintenance NRC'
l J. Bartley, Resident Inspector. Farley (June 18. 1998)
l P. Fillion. Reactor Inspector. Region II (June 1-5.. June 25-19. 1998)
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F. Hebdon. Director. Directorate 11-3. NRR (May 14. 1998)
M. Miller. Reactor Inspector. Region II (June 1-5'. 1998)
S. Ninh. Project Engineer. Region II (June 15-18. 1998)
- G. Salyers. Emergency Preparedness Specialist. Region II (June 1-5. 1998)
L. Wiens. Senior Project Manager..NRR (May 12-14. 1998)
INSPECTION PROCEDURES USED IP 37551: Onsite Engineering .
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IP-40500: Effectiveness of Licensee Controls in Identifying. Resolving and L Preventing Problems
, IP 61726: Surveillance Observations
~IP 62707: Conduct of Maintenanc IP 71707: Plant Operations .
l IP.71750: Plant Support Activities l IP 92901: Followup - Operations IP 92903:. Followup - Engineering IP 92904: Followup - Plant Support J
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. IP 93702: Prompt Onsite Response to Events at Operating '
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f-l ITEMS OPENED. CLOSED'. AND DISCUSSED
~ Opened lyng Item Number Status Description and Reference None
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Closed f
Tygg Item Number Status Description and Refere~nce L .IFI 50-302/98-02-04 Closed Radiological Mission Dose Consequence (Section 08.1)
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URI 50-302/97-14-09 Closed NRC Evaluation of Acceptability of Makeup System Trains Crosstied without Ability to j Remotely Isolate Trains. (Section 08.3)
L VIO 50-302/97-07-02 Closed Inadequate Planning and Control of Hydrostatic Testing. (Section 08.4)-
VI0' 50-302/97-11-02 Closed Inadequate Procedural Guidance for-Quality-Related Work (Control of RCS Draindown). (Section 08.5)
VIO 50-302/98-01-02 Closed Failure to Post Documents as required by 10 CFR 19.11. (Section 08.6)
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.-LER 50-302/98-03-00 Closed L'oss of Power to the ICS Caused a Reactor Trip. (Section 08.7)
LER 50-302/97-07-00- Closed Building Temperature Variations 50-302/97-07-01 Larger Than Assumed. Resulting in Unknown r Instrument Uncertainties. (Section E8.2)
VIO 50-302/97-14-05 Closed- Failure to Provide Adequate Procedure for Calibration of Reactor Vessel Level Instrumentation Reduced Inventory Operation. (Section E8.3)
, : VIO 50-302/96-01-05 Closed Two Examples of Failure to Update FSAR as
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Required by 10 CFR 50.71(e). (Section-E8.4)
F LER 50-302/97-34-00 Closed Nonconservative Assumption in Breaker
,' Setpoint Calculations Could Prevent
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Starting of Motor Operated Equipmen (Section E8.5)
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LER 50-302/97-11-00 Closed Pressure Locking of EFW Isolation Valves <
(EFV-32/33). (Section E8.6) J VIO 50-302/97-11-07 Closed Deletion of Water Quality Requirements from the FSAR. (Section E8.7)
VIO 50-302/97-17-01 Closed Failure to Conduct an Adequate US0 Evaluation for a Modification Functional Test. (Section E8.8)
IFI 50-302/97-12-03 Closed E0P-14. Enclosure 17/18 Interaction (Control Complex Chiller). (Section E8.9)
VIO 50-302/98-02-02 Closed Failure to Provide Adequate Instructions i for Installation of LR-82 and 83-F '
(Section E8.10)
VIO 50-302/97-16-04 Closed Failure to Follow CP-111 by not Performing a 10 CFR 50.59 Safety Evaluation Within 90 Days. (Section E8.11) !
VIO 50-302/97-13-03 Closed Failure to Follow Procedure for I Controlling Breakers Removed from '
Switchgear Cubicles. (Section E8.12)
LER 50-302/98-02-00 Closed Use of 500kV Backfeed while not a i Qualified Source of Offsite Powe (Section E8.13)
VIO 50-302/98-01-07 Closed 500kv Backfeed not a Qualified Source of
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Offsite Power. (Section E8.13)
VIO 50-302/98-01-08 Closed Inadequate Procedures for use of 500kV Backfeed. (Section E8.13) l IFI 50-302/96-15-03 Closed Actions Taken to Resolve Post-Accident Recriticality Concerns due to Localized Boron Dilution. (Section E8.15)
VIO 50-302/97-20-02 Closed Failure to Follow Radiation Protection Program Procedures for Documenting 4 Personnel Contamination Events. (Section !
R8.1) )
VIO 50-302/98-03-02 Closed Failure to Label a Container of Radwaste
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in Accordance with 10 CFR 20.1904(a) i Requirements. (Section R8.3) I I
VIO 50-302/97-07-06 Closed Inadequate Corrective Action for UPS in j
, Security Alarm Station. (Section S8.1) i
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Discussed Tvoe Item Number Status Description and Reference VIO 50-302/97-14-13 Open Failure to Take Corrective Actions to Identify Design Weakness for Past 10 CFR 50.59 Reviews for Positioning of DHV-34 and DHV-35. (Section E8.1)
VIO 50-302/97-01-07 Closed Instrument Loop Uncertainty Setpoint Calculation Assumptions not Translated Into Procedures. (Section E8.2)
NCV 50-302/97-21-01 Closed Noncompliance in Design Control. 50.59 Evaluations. Procedural Adequacy / Adherence. Deportability and Corrective Actions. (Sections E8.5. E8.6)
LER 50-302/97-25-00 Open Service Water Raw Water Temperature 50-302/97-25-01 Calculation contains Non-conservative Assumptions (Section E8.14).
VIO 50-302/97-20-04 Open Failure of Radiation Workers to Follow Contamination Control Procedures. (Section R8.2)
IFI 50-302/98-03-03 Open Review Licensee Maintenance on the RM-Al Sample Line and Licensee Evaluation of the Adequacy of Rubber Tubing in Iodine Sampling Systems. (Section R8.4)
VIO 50-302/98-03-04 Open Failure to implement licensing commitments to meet TMI Action Item II.F.1- calibration of all ranges of CHRMs (Section R8.5)
LIST OF ACRONYMS USED AHF - Air Handling Fan AHHE - Air Handling Heat Exchanger AI - Administrative Instruction AP - Abnormal Procedures BWOG - Babcock and Wilcox Owners Group CAP - Corrective Action Program
. CCHE - Control Complex Habitability Envelope
- Code of Federal Regulations
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CFR
- CHRM - Containment High Range Monitor i CMIS - Configuration Management Information System l CP - Compliance Procedure l ,
CR3 - Crystal River Unit 3 DH - Decay Heat DHV - Decay Heat Valve
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J i' 37 DR - Deficiency Report i EC - Emergency Coordinator !
EDBD - Enhanced Design Basis Document !
EDG - Emergency Diesel Generator i EFV- - Emergency Feedwater Valve EM - Emergency Plan Procedure-E0P - Emergency Operating Procedure EP - Emergency Planning ES - Engineered Safeguards E&TS - Engineering and Technical Support l FCN - Field Change Notice i FMEA - Failure Modes and Effects Analysis i
- Florida Power Corporation
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FSAR - Final Safety Analysis Report j GL - Generic Letter ;
HP - Health Physics HPI - High Pressure Injection )
ilPP - Health Physics Procedure :
ICS - Integrated Control System j I&C - Instrumentation and Control i
~IFI - Inspection Followup Item IOC - Interoffice. Correspondence ]
. IPTE - Infrequently Performed Tests or Evolutions l IR - NRC Inspection Report ISI - Inservice' Inspection IST - Inservice Testing
.ITS - Improved Technical Specifications JC0 - Justification for Continued Operation kV - Kilovolt '
kW - Kilowatts LAR - License Amendment Request LC0 - Limiting Condition for Operation LER - Licensee Event Report ;
MAR - Modification Approval Record MCC - Motor Control Center MP - Maintenance Procedure MUP - Make-up Pump MUV - Make-up Valve NC - Nuclear Compliance !
NEP- --Nuclear Engineering Procedure NGRC - Nuclear General Review Committee NLO . - Non-licensed Operator .
NOTES - Nuclear Operations Tracking and Expediting System NRC - Nuclear Regulatory Commission NRR- - Office of Nuclear Reactor Regulation ;
NSM - Nuclear Shift Manager '
NUPOST - Nuclear Operations Procedure Observations and Suggestions Tracking 01 - Operations Instruction
- 0P - Operating Procedure
) , OSC - Operational Support Center i OSSC --Outside Storage Shield Cask j OTSG - Once-Through Steam Generator
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PA - Protected Area
- Precursor Card
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PC PRC - Plant Review Committee
.RCA - Radiologically Controlled Area
'RP&C - Radiological Protection and Chemistry .
RW - Decay Heat Seawater RWP - Radiological Work Plan SER - NRC Safety Evaluation Report SFP - Spent Fuel Pool SP - Surveillance Procedure j STI - Short Term Instruction TS - Technical Specification TSC' - Technical Support Center URI - Unresolved Item .
l USQD - Unreviewed Safety Question Determination VIO - Violation-WR - Work Request ,
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