IR 05000313/1993026

Insp Repts 50-313/93-26 & 50-368/93-26 on 930907-1008. Violations Noted.Major Areas Inspected:Licensee Incore Fuel Loading & Fuel Storage Configurations,Core Component Performance & Outage Work Controls

ML20059E217
Person / Time
Site:	Arkansas Nuclear
Issue date:	10/20/1993
From:	Powers D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20059E190	List: IR 05000313/1993026 ML20059E186 ML20059E208
References
50-313-93-26, 50-368-93-26, NUDOCS 9311030130
Download: ML20059E217 (37)

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APPENDIX B U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Inspection Report: 50-313/93-26; 50-368/93-26 Operating Licenses: DRP-51; NPF-6 Licensee: Entergy Operations, In .

Route 3, Box 137G Russellville, Arkansas 72801  ;

Facility Name: Arkansas Nuclear One, Units 1 and 2 Inspection At: Russellville, Arkansas Inspection Conducted: September 7-24, 1993, with in-office inspection of records until October 8, 1993 Inspectors: H. Bundy, Reactor Inspector, Plant Support Section Division of Reactor Safety L. Ellershaw, Reactor Inspector, Maintenance Section Division of Reactor Safety L. Gilbert, Reactor Inspector, Maintenance Section Division of Reactor Safety C. Johnson, Reactor Inspector, Maintenance Section Division of Reactor Safety R. Stewart, Reactor inspector, Maintenance Section Division of Reactor Safety APPROVED:

Dr. Dale A. Powers, Chief, Maintenance Section Date Division of Reactor Safety Inspection Summary Areas Inspected (Unit 1): Special, announced inspection of the licensee's in-core fuel loading and fuel storage configurations, core component performance, outage work controls and critical path scheduling, potential for fuel-talated problems identified at other facilities, fuel handling procedures and practices, and disposition of degraded core components. In addition, a routine, announced inspection of the licensee's inservice inspection program and related nondestructive examination activities was performe PDR ADOCK 05000313 ..

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Areas Inspected (Unit 2): The inspectors examined the Unit 2 spent fuel pool area while performing inspections of the Unit 1 spent fuel pool are Results (Unit 1):

• The licensee was responsive and consistently employed conservatism in the evaluations of Information Notices associated with fuel handling '

problems (Section 2.1).

• The licensee performed comprehensive and in depth surveillance audits at its fuel vendor's fuel fabricating facilities (Section 2.3).

• A weakness was identified in that there did not appear to be any established requirement to cause review and/or approval of fuel vendor identified manufacturing deviations (Section 2.3).

• The licensee had appropriate concern for nuclear safety considerations during the scheduling of refueling outage work. Performance of .

computerized probabilistic shutdown safety assessments each shift was l considered a strength (Section 2.6).

• The licensee's use of a computer program for maintaining core component status was considered a strength (Section 2.7).

• The licensee's failure to ascertain the compatibility of control components with fuel assemblies in advance of conducting fuel movements ;

resulted in the unnecessary movement.of irradiated core components '

(Section 2.7).

• An effective training and qualification program for fuel handling personnel had been established (Section 2.8).

• The licensee had made reasonable provisions for improving fuel; performance in Unit 1 (Section 2.9).

• The licensee has become more proactive during the last three cycles of !

operation in its efforts to determine fuel failure mechanisms . ,

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(Section 2.14).

• Work controls were inadequate in keeping foreign materials from the j Unit I reactor vessel and spent fuel pool Level I housekeeping areas, i and a violation was identified (Section 2.15). i e A weaknesses was identified in that refueling task job orders did not clearly indicate what preventive maintenance activities were to.be performed on the fuel hantiling equipment (Section 2.16).

• Certain members of the licensee's management did not fully appreciate the safety significance or the need to further evaluate the polar crane

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-3-l slippage that occurred while moving the reactor vessel head (Section 2.18).

• The licensee was appropriately pursuing a means for increasing its onsite storage capacity for spent fuel (Section 2.19).

• The inservice inspection program was well documented and effectively '

implemented, with performance of work activities and nondestructive examinations observed to be good, and the licensee's oversight activities of the inservice inspection program was good (Section 3.2).

Summar_y of Inspection Findings:

• Violation 313/9326-01 was opened (Section 2.15).

• Inspection Followup Item 313/9326-02 was opened (Section 2.18).

Results (Unit 2):

• Work controls were inadequate in keeping foreign materials from the Unit 2 spent fuel pool Level I housekeeping area, and a violation was identified (Section 2.15).

Summary of Inspection Findings:

- * Violation 368/9326-01 was opened (Section 2.15).

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Attachments:

• Attachment 1 - Persons Contacted and Exit Meeting

• Attachment 2 - Documents Reviewed

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-4-DETAILS 1 PLANT STATUS During this inspection period, the plant transitioned from a hot shutdown '

condition (initiated at 12:01 a.m. on September 7, 1993) to Refueling Outage 1Rl FUEL INTEGRITY AND REACTOR SUBCRITICALITY (60705/60710/62700/86700/

92701/92702)

The objectives of a Fuel Integrity and Reactor Subcriticality (FIRS)

inspection are to review, inspect, and determine the adequacy of the licensee's activities related to the protection of reactor fuel. Attachment'2 to this inspection report is a tabulation of documents that were reviewed by the inspectors during the inspection and provided some of the basis for the findings documented in this report. Other licensee documents that discussed fuel-related activities and associated equipment designs and operational characteristics were made available to the inspectors and were examined in much less detail. In general, the reviews of procedures and records were not detailed in nature, but rather were broad overviews to determine that essential issues were addressed in reasonable fashion. Many of the findings !

in this inspection report were the results of inspectors' observations of licensee activities in progres Information on several aspects of the i licensee's activities were based on interview statements taken from licensee staff members and, a sampling of those statements were verified by review of Technical Specifications (TSs) or the licensee's procedures and record Emphasis, however, was given to reviewing the following areas:

• In-core fuel loading and fuel storage geometrical controls to preclude configurations that have not been specifically approved by NRC in safety >

evaluation reports and that conceivably could result in situations involving inadequate shutdown margin or inadvertent criticality; ,

• Operational work control practices, communications, procedures, physical systems and equipment, and training that preclude unsafe fuel' movements from occurring; ,

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• Licensee evaluations and corrective actions that were performed ,

subsequent to any self-identified problems that were indicative of- >

accident sequence precursors or that had the potential to lead to fuel damage; and -

• The susceptibility of the licensce's operations, procedures, and equipment to fuel-related problems that have occurred.at other nuclear i power plant ;

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. i-5-NRC Inspection Manual Procedures 60705, " Preparation for Refueling"; 60710, ,

" Refueling Activities"; 62700, " Maintenance Program Implementation"; 86700,

" Spent Fuel Pool Activities"; 92701, " Followup"; and 92702, " Followup on -

Corrective actions for Violations and Deviations"; provided guidance for this inspection effor .1 Fuel-Related Incidents at Other Facilities ,

2. Discussion The inspectors discussed with the licensee several fuel-related events that have occurred at commercial nuclear power plants. Specifically, the incidents that were discussed are described in NRC Information Notices (ins) and Bulletins that were issued during the past decade. Attachment 2 contains a listing of those ins and Bulletins, which the inspectors reviewed and discussed with the licensee for this assessment. Certain ins were viewed by the inspectors not to be directly applicable to the licensee's operation This ncn-applicability was because, for instance, the licensee's designs, practices, or procedures at the time of IN issuances should have precluded i such incidents from occurring. These nonapplicable ins are not discussed in this report. A portion of one IN did not appear to be adequately addressed by the licensee and is discussed belo .1. IN 88-65 IN 88-65, " Inadvertent Drainages Of Spent Fuel Pools," informed licensees of

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incidents in which the level of spent fuel pools were inadvertently lowered as .

a result of the failure to realign a valve in the spent fuel pool cleanup system, and an inadequate procedure for lowering the level in conjunction with '

a plugged anti-siphon device. The IN also noted that the spent fuel pool level indicator and low-level alarm were inoperabl ,

The licensee evaluated the IN and revised the applicable procedures to include

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prerequisites and precautions to prevent inadvertent lowering of the spent fuel pool levels. The Unit I spent fuel pool gates utilized solid rubber seals, which did not require the maintenance of a pneumatic supply. In addition, half-inch siphon holes were drilled in spent fuel circulation pump suction piping at a depth of 1 1/2 feet below the normal water level to assure that no depletion of water level due to siphoning after a pipe break could occur. The inspectors noted that the licensee's summary of actions contained in an internal response dated August 31, 1989, with respect to level indicators and/or alarms was basically a recommendation to perform surveillances, without indicating that the recommendation would be adopted and ,

incorporated into the maintenance progra i Upon questioning by the inspectors, the licensee provided information to show that -the spent fuel pool level instrumentation had been calibrated and was on a scheduled calibration frequency. The initial calibration was performed as 1 corrective maintenance on the spent fuel pool . level indicator and related loop-instrumentation on Job Order 825861. This job was shown to have been

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completed on October 18, 199 Subsequently, an NRC-inspection identified that certain instruments used during surveillance tests were not themselves calibrated. The licensee initiated Condition Report 2-91-0395 on June 8, 1991, which addressed gages and instrumentation that had not been calibrated within an 18-month period due to the restructuring of the preventive maintenance program. The licensee performed an evaluation of all safety-related instrumentation (both units) to identify those instrument Calibration activities were initiated, and as procedures were established and, instruments calibrated, they were incorporated into the preventive maintenance program with specified calibration frequencies. Certain instruments require calibration, but have not yet been placed into the preventive maintenance .

program. To assure that they do get calibrated, they are listed in the condition report which remained open and under the responsibility of the Unit 1 Instrumentation & Controls Supervisor. The inspectors noted that the spent fuel pool instrumentation had not been placed into the preventive maintenance program; however, it was listed in the condition report, and was now on an 18-month calibration frequency (with the last calibration performed on November 5, 1992, using Corrective Maintenance Job Order 880720).

During this review, the inspectors examined drawings of the Unit 1 spent fuel pool syste The licensee had previously completed upgrading piping and *

instrumentation drawings. The inspectors noted a problem with Drawing No. M-235, " Spent Fuel Coolant System," Revision 48. The drawing was inaccurate with respect to its symbology for tell-tail drains as opposed to drain lines on the spent fuel pool and cask loading pit liner. As a comparison, the inspectors determined that the symbology for tell-tail drains on the refueling cavity liner drawing were proper. When the inspectors brought this matter to the licensee representative's attention, they were informed that the licensee had previously identified this problem when system engineers had been looking to find alternate means of draining'the cask loading pit cavit The inspectors concluded that for thoroughness, the licensee should have initiated a drawing design change at the time of their discover .1.2 Conclusions Regarding IN 88-65, the inspectors considered the actions taken by the licensee to be appropriate to the circumstances. The inspectors, after review l and verification of licensee evaluations associated with other ins, concluded I that a meaningful IN evaluation program had been established and that i conservatism was consistently employe l

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2.2 Shutdown Marcin and Premature Criticality I l

2. Discussion The licensee informed the inspectors that the following two issues had been previously identified regarding inadequate shutdown margin l l

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-7-On February 1, 1989, the licensee submitted Licensee Event Report No. 50-313/89-005, that concerned a personnel error resulting in an inadequate procedure which caused calculated reactor shutdown margin to be less conservative than assumed in the plant's design basis. The NRC subsequently ,

evaluated the licensee's actions taken to address the issue and considered them to be acceptable. The Licensee Event Report was closed in NRC Inspection Report 50-313/90-02; 50-368/90-0 ,

On February 23, 1989, the licensee initiated Condition Report 1-89-0109, which addressed a potential for inadequate subcritical. margin that might exis ,

during heatup or cooldown conditions witn high steam generator levels. The condition report resulted from the receipt of a Babcock & Wilcox letter dated February 22, 1989, that notified the licensee of potential conditions to be aware of in order to ensure that the reactor would be sufficiently subcritical during certain shutdown conditions. For followup purposes, the NRC identified this condition as Inspection Followup Item 313/8910-03. The NRC subsequently evaluated the actions taken by the licensee and determined that their approach was conservative and adequately bounded the issue. The item was closed in NRC Inspection Report 50-313/90-15; 50-368/90-1 ;

Regarding current refueling activities, the inspectors reviewed the licensee' shutdown margin calculations for September 7 and 8,1993, and confirmed that -

they were appropriat 'i 2.2.2 Conclusions The licensee, after identifying adverse conditions, took responsible and '

proper actions to correct the conditions. Current shutdown margin .

calculations were appropriat ,

2.3 Procurement and Receipt Inspection 2.3.1 Discussion

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The inspectors reviewed the licensee's procurement and inspection activities regarding the ANO-1 Batch 14 fuel assemblies and burnable poison rod assemblies (bPRAs) delivered to ANO-1 July 28 through August 11, 199 The inspectors confirmed that the licensee's fuel supplier has. always been a .

domestic source; consequently, 10 CFR 74.15 (b) receipt inspection

requirements were not applicable. The inspectors observed an informative j training video that the licensee's staff had prepared on the ANO-1 fuel '

receipt proces ;

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Fuel Contract 660-067/068/069/070 dated December 21, 1973, through .

Amendment 12 dated March 1, 1991, represented the procurement agreement for :

reload Batches 7.through 16, with Babcock & Wilcox Company, whose nuclear fuel manufacturing unit subsequently became known as B&W Fuel Company, and the licensee. The technical and quality assurance information provisions of th *

contract were rather general in nature; however, for evaluation purposes, the t

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purchaser had access to technical, manufacturing, and inspection informatio This also included access to required drawings, specifications, procedures, and the witnessing of manufacturing and-inspection operation The inspectors reviewed the procedures (identified in Attachment 2) that dealt I with supplier qualification and the performance of fuel vendor audits. The inspectors verified that B&W Fuel Company was maintained on the Qualified ,

Suppliers List based on the performance of an audit dated October 21, 1991, and annual evaluations, the last of which was performed on August 3,199 In addition, Quality Plan, "AN0-1 Batch 14, Cycle 12 Reload Nuclear Fuel Design and Fabrication Surveillance Audit Program at Babcock & Wilcox Fuel-Company," was established to plan and conduct five surveillance audits during the fabrication of the Batch 14 nuclear fuel. Review of the Quality Plan, which was Revision 0 and transmitted under cover letter dated February 18, 1993, revealed it to be a comprehensive and in-depth document that provided general and specific attributes to be observed and verified during each of the scheduled audits. The inspectors reviewed the five audit reports (identified in Attachment 2) and the applicable checklists which provided the details in support of the results summarized in the audit report The inspectors reviewed the procedures (identified in Attachment 2) and documentation associated with receipt inspection activities conducted by the .

licensee regarding Batch 14 nuclear fuel assemblies. _ Receipt inspection was limited to assembly serial number identity verification, observation of the fuel assembly containers for physical appearance / condition (e.g., obvious damage which could have affected the assemblies), and review of B&W Fuel Company's supplied documentation (i.e., shipping notices, certificates of conformance, B&W radioactive survey information, B&W final visual inspection reports, and Form 741, " Nuclear Material Transaction Report"). ,

The inspectors reviewed, "B&W Fuel Company's Q.A. Data Package," which contained certificates of conformance related to the 60 fuel assemblies and 52 BPRAs that constituted Batch 14 nuclear fuel that was delivered to ANO-1 July 28 through August 11, 1993. The certificates.of conformance provided assembly identification, pertinent element weights, and attestations with respect to meeting the listed specifications and procurement documents. In addition, the certificates identified "CVARs" and "DRs" (contract variation approval requests and deviation reports, respectively), which were described ;

as exceptions that had been reported to and approved by the Fuel Engineering Section of the B&W Fuel Company. The contract variation approval requests and deviation reports were identified by number only, with no verbal descriptio The inspectors asked licensee representatives what the procedure was pertaining to their review and approval process of vendor identified variations and deviations. The inspectors were informed that a specific i procedure or instruction did not exist that would cause a review and approval process to occur, but that the contract would provide guidance in this regar It was noted that Amendment 12 to the contract, dated May 1, 1991, addressed design changes. It provided for the purchaser to be informed of any significant design changes, which were defined as changes that could affect

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the design / licensing basis or safety limits of the fuel assembly, or AN0-1 TSs. However, the inspectors did not identify any guidance with respect to review and approval of vendor identified manufacturing deviations. The licensee was not able to provide any information as to what the identified contract variation approval requests or deviation reports pertained to. This was considered by the inspectors to be a programmatic weaknes "

2. Conclusions Procurement of nuclear fuel assemblies is currently based on a 1973 contract that established minimal quality requirements. Rather than performing an attribute verification upon receipt of fuel assemblies, the licensee has chosen to rely upon the performance of surveillance audits at the nuclear fuel vendor's facility. The licensee developed a comprehensive and in-depth quality plan that provided the attributes / characteristics to be evaluated and verified at the time the surveillance audits were performed. The surveillance audit reports and their integral checklists reflected implementation of the quality plan. A programmatic weakness was identified in that there did not appear to be any established requirement to cause a review and/or approval of vendor identified manufacturing deviation .4 Commitments on Fuel-Handlino Activities 2. Discussion The licensee committed to Regulatory Guide 1.33, " Quality Assurance Program Requirements (Operations)," November 1972, in TS 6.8. The guide endorsed ANS 3.2-1972, " Administrative Controls and Quality Assurance for the Operational Phase of Nuclear Power Plants." This standard required written procedures for core alteration, accountability of fuel, and partial or complete refueling operations. Specific procedures (listed in Appendix A of Regulatory Guide 1.33) were also required for each refueling outage and for receipt and shipment of fue . Conclusions The inspectors verified by review, that the licensee had established the procedures (identified in Attachment 2) required by Regulatory Guide 1.33 and ANS 3.2-197 .5 Service Information on Fuel-Handlino Eauipment 2. Discussion The licensee informed the inspectors that the process for handling vendor service information on fuel-handling equipment was described in Procedure 5510.203, ' Vendor Technical Manual Review & Update," Revision To determine the effectiveness of the licensee's implementation of the procedure, the inspectors selected for review a major modification which involved vendor provided post-installation service information. The modification involved the ,

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-10-main fuel handling bridge, transfer system load cells, and limit switches, which involved vendor provided post-installation service information. Design Change DCP 86-1069 was partially initiated in 1990 and completed during the current outage. The new equipent was provided by the Stearns-Roger Division of UEC, the designer and installer. While the equipment and main fuel handling bridge were classified as non-Q (i.e., nonsafety-related), the main fuel handling bridge was classified as Quality Assurance Category Q for '

Seismic Class 1 considerations, and, therefore, considered as equipment important to safety. The design change consisted of modifying existing equipment and the installation of new equipment, including:

• New control consoles which provided all existing control and indication functions for the bridge, trolley, and hois The primary control unit of the console was a programmable logic controller, which was removable (for storage outside of containment) by quick disconnect connection ,

• A new motor control center which provided power to the bridge, t olley, and hoist motor. The motor control center was removable (for st rage outside of containment) by quick disconnect connections that coupled it to the control console or console tester / simulato * A control rod mast modification to accommodate the differently designed MK B-4 and MK B-5 control rod '

• A fuel mast modification from hydraulic to pneumatic grapple contro * All necessary equipment furnished to convert the bridge grapple actuation system from a hydraulic operation / control system to a pneumatic syste * A new electronic load sensing system provided for the fuel and control rod masts (included new load cells).

• Programmed mast limit switches with remote. digital readout (both fuel and control rod masts).

• A console tester / simulator which simulated inputs and loads for a complete functional check-out of the control console and motor control cente * A new two-camera indexing system to replace the existing system used for positioning the main fuel bridg The inspectors conducted a review of design change package DCP 86-1069'. The package included a 10 CFR 50.50 evaluation that documented that the new and modified fuel handling equipment did not impact the TS nor a fuel handling accident as analyzed and discussed in Chapter 14 of the Safety Analysis Report. In addition, detailed installation and test procedures were contained in the design change package. The inspectors also noted that prior to each ,

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-11-reinstallation of the control console and motor control center, they could be

" pre-installation" tested by using the simulator console that facilitated all inputs and loads to the control console and motor control center for a complete function check-out. The simulator also enabled the out-of-containment training of operators and maintenance personne During this portion of the inspection, the inspectors reviewed the " pre- .

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installation" test conducted on September 3, 1993, in accordance with the Procedure SR-PP-1106, " Simulator Interconnection and Operating Procedure,"

Revision 2. The procedure appeared to be well detailed, encompassing interlocks, overloads, fuel and control rod masts, and complete functions of the bridges, trolley, and hois In addition, on September 15, 1993, the .

inspectors witnessed the calibration of the new load cells installed on the-main fuel handling bridge cabinet. The load cells were an integral part of the new electronic load sensing system and the calibration was performed in accordance with Procedure SR-PP-998 under Job Order CWP 86-1069/830171-1 All calibration values were left'within the acceptance criteria delineated.in the procedure. Subsequent to witnessing the load cell calibrations, the inspectors made a visual examination of the accessible components and equipment newly installed on the main fuel handling bridge as part of DCP 86-1069. The inspectors observed that the quality of workmanship appeared goo In addition, the new control console panel controls and instruments'were well arranged with human factors considerations apparen The inspectors noted that all design modifications were complete, including post-installation testing. However, prior to fuel movement, a final preoperational check out of the main fuel handling bridge over the open reactor vessel was to be conducted. This check out was not observed by the inspectors. Later, however, the inspectors observed operators experiencing minor difficulties in utilizing the control rod mas These difficulties appeared to be related to load cell tripping. The tripping was preliminarily attributed to a small increase in system friction loading when control rods ,

were raised to a certain elevation in conjunction with the small allowable load margin preset in the load trip circuitry. The inspectors understood that the licensee subsequently resolved this proble .5.2 Conclusions The inspectors considered the design modification to the main fuel handling bridge to be appropriate. The procedures and observed activities were well'

planned and implemente .6 Outage Work Controls Responsibilities. Delegations, and Critical Path Scheduling 2. Discussion The current refueling outage (IRll) had been preplanned in accordance with the

"ANO Unit 1 1Rll Outage Planning Handbook," and scheduled in accordance with Procedure 1001.002, " Outage Scheduling and Management." Nuclear safety and

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-12-l shutdown risk considerations were appropriately addressed in accordance with ,

the "ANO Unit 1 Shutdown Operations Protection Plan," which was responsive to current NRC and industry guidanc The refueling crews were staffed to provide sufficient coverage of refuelin activities with periodic relief from continuous dutie There were four individuals qualified as senior reactor operators in charge of refueling '

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assigned to each of two, 12-hour shifts. There were also eight refueling equipment operators assigned to each shift. The senior reactor operator in charge of refueling was expected to direct the refueling crew no more than .

.three continuous hours. It was also planned to rotate the duties for refueling equipment operators approximately every 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. The inspectors

. observed that this was occurring. The refueling crews were scheduled to work ,

the TS limit of six continuous 12-hour days prior to a day of res Management stated that no exemptions to this limit, as allowed by TS, were expected to be grante Outage reports were generated for each 12-hour shift. These reports were discussed during the management outage meetings that were held twice a day (i.e., 6:30 a.m. and 6:30 p.m.). Most of the 6:30 a.m. meetings during this inspection period were attended by an inspector. The reports addressed the status of significant plant parameters including the estimated time to steam .

release in the event of loss of decay heat remova Expected work for the

next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the status of key safety functions, and any safety concerns '

were discussed. The following key safety functions were considered: decay heat removal capability, inventory control, electrical. power availability, reactivity control, containment closure capability, and instrumentation and i instrument air status. These key safety functions were analyzed with the ORAM-TIP computer progra Probabilistic shutdown safety assessments.were !

performed each shift and the results were analyzed by a nuclear safety engineer. The inspectors observed performance of one shutdown safety '!

assessment and concluded from the brief review that the results appeared comprehensive. Performance of the probabilistic shutdown risk assessment, !

which was performed each shift, was considered a strengt The inspectors reviewed various outage work schedules to ascertain any adverse impact of out-of-service equipment and systems on nuclear safety. It appeared :

that the impact of out-of-service equipment and systems was being considered each shift, and that.probabilistic shutdown risk assessments indicated that core damage was acceptably lo ;

The shutdown operations protection plan required performance of a detailed safety analysis of the outage schedule by the shutdown risk assessment task-force for all planned outages. The shutdown risk assessment task force.was comprised of representatives from operations, safety analysis, system engineering, industry events analysis, licensing operations training, and outage scheduling. The inspectors learned through interviews that the safety-analysis of the schedule for Outage IRll had been completed approximately 2 months prior to the outage and comments had been incorporated. Also, a safety review was required prior to implementing any safety significant

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changes to the outage schedule after the initial revie Requirements'for the :

seven allowed shutdown conditions were described in the protection plan and .

discussed in each shift report. Any deviations from the guidelines required  !

formal approva The protection plan listed numerous equipment requirements to minimize the l risk of releasing radioactive materials. Among these were containment and '

spent fuel handling area ventilation requirements and controlling protected systems and their power supplies by physical barriers with signs that required plant personnel to cont &ct the control room prior to entr ;

2.6.2 Conclusions The licensee had scheduled refueling outage work with appropriate concern for nuclear safety. Performance of computerized probabilistic shutdown safety ,

assessments each shift was considered a strength. The refueling crews were suitably staffed to avoid excessive worker fatigue and overtime was properly controlled. Areas around protected systems and their power supplies were ,

controlled by physical barrier ; Fuel-Handlina Controls 2. Discussion ,

Suitable instructions for fuel reloading and storage of spent fuel and control  ;

components were provided by the following documents:

• Procedure 1502.004, " Control of Unit 1 Refueling,"

• Procedure 1506.001, " Fuel and Control Component Handling,"

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• Procedure 1409.492, " Refueling Shuffle,"

• Procedure 1203.042, " Refueling Abnormal Operations," and

• Memorandum ANO-93-02479, "1Rll Refueling Shutdown Margin."

Suitable refueling administrative controls were established in the ' procedures l

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and referenced documents. Instructions for communications and documentation of fuel and control component m vements were sufficient. Criteria for stopping fuel handling were appropriate. Procedure 1409.492 provided the specific sequence of fuel and control component movements. It included a  ;

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10 CFR 50.59 evaluation and provided appropriate instructions for review and approval of sequence changes. Appropriate instructions for interim parking of fuel assemblies were included. The shutdown margin calculation had been

. performed by the fuel contractor and Memorandum ANO-93-02479 provided bounding 1 conditions for this calculation. This memorandum was referenced by Procedure 1203.042 with regard to sequence changes. Instructions for verification, communication, and logging of fuel assembly and control component movements l

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-14-were appropriate. The post-fuel load core verification inspection was comprehensive. Procedure 1506.001 provided suitable instructions for dealing with fuel assembly interferences. Refueling crew personnel interviewed had not experienced problems with fuel rod twisting or bowing. There were appropriate precautions for preventing accidental dropping or damaging of fuel assemblies. Procedure 1203.042 provided suitable pre-established actions for responding to a fuel handling accident. It further'provided for safely ,

shutting down fuel handling in the event of a tornado warnin The SHUFFLEWORKS computer program for core status was supportive of the fuel reloading procedures and its use was considered a strength. It was observed to be very useful when interim in-core parking of control components was required. The inspectors observed that sequence step changes were correctly performed and the required safety reviews were complete The licensee made some unnecessary control component moves in trying to place a spent BPRA in the fuel assembly located in position H8. After unsuccessful attempts to place two different BPRAs " the fuel assembly the licensee theorized that since this was the oldest fuel assembly in the core (a Mark B4 design), it might not be compatible with the current BPRA. design (Mark B5).

The BPRA did not seat at as low of an elevation as had been anticipated. This appeared to have been caused by a design difference in the fuel assembly holddown spring retainer plugs. The plug is not present in the newer fuel assembly design. Failure to integrate core component design modifications into a fuel and control component compatibility revie,f prior to establishing the fuel movement plan caused additional, unnecessary handling of irradiated core component The refueling crew was observed to be adequately staffed during movement of control components. Communications between the main fuel handling bridge and the main control room were clear. Control component engagements, movements, and disengagements were independently verified by a certified fuel handler equipped with binoculars. The inspectors, without binoculars, had no difficulty discerning which components were being handle . Conclusions Suitable administrative controls and fuel handling instructions had'been implemented to accomplish safe fuel reloading and discharging of spent fue Communications between the fuel handling crew and the control room were clea The use of a computer program for maintaining core component status was considered a strength. Failure to ascertain the compatibility of control components with fuel assemblies in advance of conducting fuel movements resulted in the unnecessary movements of irradiated core component ..

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-15-2.8 Fuel Handlina 0"glification and Trainina Proaram 2. Discussion The inspectors reviewed the licensee's lesson plans and qualification forms for fuel handling and fuel handling equipment operators. The lesson plans covered training for senior reactor operators, reactor operators, and ,

contractor personnel. Review of contractor personnel qualifications indicated that most of the personnel had prior experience in fuel handling operations; however, contractors were still required to complete the licensee's screening ,

and fuel handling qualification training. Review of licensee records indicated that most of the personnel had completed the classroom training, but some equipment operators had not completed the practical portion. The inspectors reviewed 17 qualification cards for fuel handling operators and 6 qualification cards for senior reactor operators. No discrepancies were identified during this review. The scope of training was determined to be appropriat .8.2 Conclusion

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It appeared that the licensee had established an effective training and qualification program for fuel handling personnel, and was properly screening contractor personnel for prior fuel handling experience. The inspectors determined that the scope of operator training appeared appropriat .9 Fuel and Core Component Performance

2. Discussion The licensee presented the inspectors with an overview of its fuel and core component performance at ANO-1. The licensee explained that it has a Company Directive (C7.910, Revision 1) that set forth a goal of zero leakage fuel through fuel design, chemistry controls, power condition practices, and reactor coolant system cleanliness. The licensee had a failed fuel action plan (based upon a fuel reliability indicator), however, the inspectors did not review this plan during the inspectio The original fuel decign loaded into AN0-1 was the Babcock & Wilcox designated Mark B4 15 x 15 standard design. During the 1980s, the licensee participated in an extended burnup lead test assembly program. The lead test assembly program was sponsored by Babcock & Wilcox, Department of Energy, and Arkansas Power & Light. The extended burnup program included four lead test assemblies that prior to discharge achieved maximum burnups approaching 60,000 MWD /MTU.- ,

The latest fuel design for Cycle 12 loading is designated Mark B8ZL, and the fuel rod design is designated Mark B In general, core performance problems at ANO-1 can be categorized into two areas: broken fuel assembly holddown springs and fuel rod fretting perforation In regard to the former problem, the licensee routinely replaced broken holddown s,3;ngs during refueling outage In regard to the

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latter problem, the licensee had experienced a number of fuel failures in each cycle of operation, except the last cycle (Cycle 11). The licensee's representative expressed the viewpoint that .it believed that during Cycle 11 operation all Babcock & Wilcox designed-plants, except ANO-1,:had fue failures in their cores. The table below lists the number of failures for each cycle of ANO-1 operatio , ,

TABLE 1 ANO-1 FUEL FAILURES

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CYCLE OF OPERATION NUMBER OF FUEL' FAILURES

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1 3 2 9 3 2 4 43 5 11 ,

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6 12 7 7 8 4-9 6 10 11 11 0-l In describing its role to obtain better performing fuel, the licensee's  !

representative presented generic information to the inspectors on the industry's experience with Babcock & Wilcox fuel. The greater majority of fuel failures that have been determined in Babcock & Wilcox Inconel-grid fuel,  : were mostly attributed to fretting failures that occurred at locations where grid strap springs contacted the fuel cladding. The.second most prevalent fuel failure mechanism was attributed to fretting failures due to debris in the reactor coolant system. In response to this awareness, the fuel vendor provided the first all Zircaloy-4 gridded fuel for loading during the IR8 outag In addition, since the IR9 outage the licensee has been loading fuel with a deb ris-resistant design. This design provides a solid lower-end plu in the area below the first grid where debris often becomes trapped and frets fuel claddin l

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-17-The licensee's representative stated that the licensee and the B&W Owners' i Group urged its fuel vendor to design and supply reconsitutable fue Subsequently, the licensee loaded its first reconstitutable fuel during the IR9 outage, and has continued with this improved fuel design. The Cycle 12 ,

core will have only nine non-reconstitutable fuel assemblies remainin j 2. Conclusions ,

i The inspectors concluded that the licensee had made reasonable provisions to '

improve fuel performance in ANO- ,

2.10 Fuel Handling Area Ventilation System 2.10.1 Discussion -

The inspectors reviewed surveillance test reco"ds of the fuel handling area ventilation system to determine if the licensee had performed the surveillance test required by TS 4.17. The TS required the licensee to verify an acceptable level of efficiency and operability of the fuel handling area !

ventilation system. Operability of the fuel handling ventilation must be verified before handling of irradiated fuel. A pressure drop test across the HEPA filters and charcoal absorber banks is required every 18 months. The '

inspectors were provided Job Orders 895000 and 897317, which showed that the licensee had performed the required operability test of the fuel handling '

ventilation system. The inspectors noted that the licensee had established a procedural requirement for operability test performance every 720 hours0.00833 days <br />0.2 hours <br />0.00119 weeks <br />2.7396e-4 months <br /> of operation, which is more conservative than the TS requirement of 18 month '

2.10.2 Conclusion From review of records and discussion with the responsible engineer, the inspectors concluded that the licensee was performing the required testing on the fuel handling ventilation system prior to movement of irradiated fuel as required by T .11 Boron Concentrations 2.11.1 Discussion The licensee had established procedures to control boron concentrations during '

refueling operations. The inspectors reviewed daily chemistry logs and determined that the primary water chemistry was being maintained above the minimum 1600 ppm required by TS 3.8.18, and above the minimum 1800 ppm administratively required by Procedures 1502.004 and 1502.01 '

2.11.2 Conclusion Adequate primary water boron concentrations were being maintaine .

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The inspectors reviewed the licensee's control for primary system water clarity. Visual examinations by the inspectors verified that spent fuel pool

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and refueling cavity water clarity were good. Operations personnel who were , ;

interviewed indicated that primary water clarity usually decreased in the

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later part of the fuel handling operation This loss of clarity was attributed to the release of corrosion products from core components that were moved during the outage. The inspectors noted that there was no refueling ,

cavity skimmer system to remove dust and other minute floating debris. There was, however, a spent fuel pool skimmer syste ;

The licensee determined spent fuel pool water clarity monthly by sampling the .

water chemistry to determine the amount of suspended solids. The inspectors were informed by the licensee that when suspended solids exceeded 100 ppb, the water would be recirculated through demineralizer tanks to lower the suspended solid The inspectors were briefed on a lighting upgrade modification that was implemented during this refueling outage. The modification consisted of three additional underwater lights of high intensity. That, in conjunction with the use of a B&W Fuel Company underwater video camera, provided good visibility to enable fuel handling operators ease of fuel assembly identificatio Observations by the inspectors verified the use of this equipmen .12.2 Conclusions  :

The inspectors concluded that the licensee had established good primary water clarit .13 Reload Analysis 2.13.1 Discussion The inspectors reviewed the licensee's reload analysis for Cycle 12. (ANO-1, !

Cycle 12 - Revised Final Fuel Cycle Design Report). The inspectors' review .

was focussed on parameters (burnup, mechanical design) that had changed relative to previous Cycle 11 parameter During the review, the licensee's representatives from Jackson, Mississippi',

and its fuel vendor answered questions. Specific questions included those on the adequacy of the Cycle 12 fuel assembly shoulder gap. This issue was selected for review because certain parameters that the reload analysis discussed as having been changed were changed in ways that should result in the need for an increased shoulder gap, yet the design shoulder gap had not been increased. The inspectors' concern centered around three parameters that together act to enhance fuel rod axial growth. These parameters were an

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-19-increased design burnup (i.e., effective full power days), a reduction in fuel rod prepressurization, and an increase in fuel pellet diamete !

t The inspectors were informed that these parameters had been explicitly modeled in the Cycle 12 analysis. The inspectors did not audit the actual data input ;

to the calculations, which were not available on site, but believed that the '

licensee had presented reasonable evidence that the data input was acceptable, for the purpose of this inspection. Among the information provided to the

inspectors was the proprietary growth correlations for fuel rods and fuel

assembly guide tubes. These were the correlations given in the NRC-approved j version of BAW-10179, " Safety Criteria and Methodology for Acceptable Cycle Reload Analyses." The inspectors noted that the Cycle 12 analysis (a)  !

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utilized the fuel rod thermal-mechanical code TAC 03, which is approved fo ANO-1 application by NRC, and (b) determined that the Mark B9 fuel rod would ,

continue to meet all mechanical design criteri j

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2.13.2 Conclusions During the discussions and.the brief review, the inspectors determined the licensee's reload analysis properly accounted for fuel assembly shoulder gap !

provision !

2.14 Fuel Assembly Post-irradiation Examination and Reconstitution +

2.14.1 Discussion ,

In regard to broken fuel assembly holddown springs, the licensee had conducted- t a visual inspection of holddown springs prior to beginning fuel and core .i

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component moves during each refueling outage. These inspections were performed with an underwater camera obtained from its fuel vendor. The inspectors witnessed the quality of the reception on the camera and found it

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to provide a good quality, color picture. The camera inspection did not necessarily reveal incipient cracks in holddown springs, but should have revealed catastrophic failures of holddown spring During an earlier outage, the licensee recaged one fuel assembly that contained a failed fuel rod. This assembly was of the earlier design and

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because the design was non-reconstitutable, the activity was considerably more involved than a reconstitution would be with the more improved desig The inspectors questioned the licensee on what post-irradiation examinations had been performed other than holddown inspections. It appeared that the '

licensee was becoming more proactive in this area over the past three cycles of operation. Up until the IR9 outage, the licensee had performed limited i

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visual (of peripheral rods) and ultrasonic examinations of fuel assemblies suspected of having failed fuel. However, during the IR4 outage, a sipping campaign was performed in response to the large number of fuel failure Consequently, the post-irradiation examinations performed up until the IR9 outage were basically focused on which rods had failed and not on. establishing .

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the failure mechanism. During the IR9 and IR10 outages, the licensee

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-20-i conducted full core ultrasonic examinations and performed rod lifts on peripheral rods that had ultrasonic indications of water logging. These 1 inspections-enchled the licensee to confirm fretting at grid spring contact- !

points as-an operative failure mechanis '

There were no scheduled post-irradiation examinations of fuel assemblies in IR11, because chemistry measurements indicated the absence of certain

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radioisotopes in the reactor coolant during Cycle 1 .14.2 Conclusions The inspectors concluded that the licensee had become more proactive over the last three cycles of operation in its efforts to-determine fuel failure '

mechanism .15 Loose Parts and Foreign Material Exclusion 2.15.1 Discussion i

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Debris fretting by foreign material is one of the leading causes of fuel l failures in the nuclear industr Foreign material which might enter the ;

spent fuel pool may eventually enter the reactor with fuel insertion during:

refueling. Also, foreign material may enter the reactor core when the reactor- J vessel head is remove .

The inspectors examined the licensee's controls for general cleanlines:, loose l parts, and housekeeping activities, for Units 1 and 2 spent fuel pool .

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refueling areas. The inspectors also examined Unit 1 containment refueling

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areas. The spent fuel pool and containment areas were designated as Housekeeping Level 1 areas, the highest level of housekeepin ;

The licensee had established Procedure 1000.018, " Housekeeping,"' Revision 19, which was developed to prevent foreign material from entering the housekeepin Level 1 areas (e.g., within 10 feet of the spent fuel pool areas), and ' i Procedure 1025.019, " Maintenance of Fluid and Air' System Cleanliness Control," '

which provided controls to prevent foreign material from being introduced into .

systems, structures or components during maintenance and modification-Procedure 1000.018 specified that clear plastic is prohibited in Levei- 1 areas "

and required that tools be secured with lanyards. Procedure 1025.019 *

specified that loose items be secured with tape or safety strap '

During the inspection of the spent fuel pool Level 1 areas, the inspector identified the following:

. On September 8, 1993, clear plastic inside the designated Level 1 area of the Unit 1 spent fuel pool are I b

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• On September 8, 1993, a screwdriver and a weight used to secure stanchions were unsecured in the designated Level-1 area of the Unit 2 spent fuel pool are * On September 10, 1993, a small wad of electrical duct tape floating in the tilt pit area of the Unit 2 spent fuel pool are * On September 20, 1993, two pair of anti-contamination garments lying on'

the floor inside the designated Level 1 area of the Unit l' spent fuel pool, and a piece of electrical duct tape (approximately 1 inch)

floating in the tilt pit area of the Unit 1 spent fuel poo * On September 22, 1993, a grey paint chip (approximately 1-1/2 inch by 2-1/2 inch) floating in the Unit I spent fuel pool, and an 8-inch piece of electrical duct tape in the spent fuel pool attached to the spent-fuel pool cooling pump suction straine During the inspection period, the licensee identified the following examples of foreign material inside designated Level 1 areas:

• On September 8, 1993, a mop bucket in the Unit I spent fuel area. :The licensee informed the inspectors that the bucket was in the area for cleanup.of rain water from a roof leak. The licensee stated that the bucket was 10 feet from the spent fuel pool area, however, it was still within the designated Level 1 are * On September 20, 1993, an object on the horizontal section of the reactor vessel former plate adjacent to core grid locatinns F-15 and G-15. This area is just below where the upper core barrel bolts are installed. The' licensee determined that this object was most likely a tool (modified thread chaser) used during 1R6 for the upper barrel bolt replacement work. The tool was subsequently removed during this outag Apparently, the tool had been held in place by the upper internals package during five cycles of operation. When the inspectors reviewed, with the licensee's representative, a previous post-core load verification video, the tool was vaguely apparent. 'It is believed that the use of the improved reactor vessel lighting and a new und.erwater camera resulted in the recent identification of the too * On September 28, 1993, an object resembling a piece of wire or nail on the reactor core plate. This object may have created the need for unnecessary movements during this outage. When operators had attempted to remove the fuel assembly from the G-9-location, it would' not unseat from the core plate. Consequently, all surrounding fuel assemblies were

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removed to see if any interference was evident. This being unsuccessful, the licensee arranged with its fuel vendor to increar:e the lifting force on the fuel assembl The fuel assembly was then unseated. The licensee's investigation into the role of the foreign

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object related to the difficulty in unseating the fuel assembly was not -

complete during this inspection perio The examples listed above are violations of Unit I and 2 TS 6. (313/9326-01; 368/9326-01).

A noncited violation identified in NRC Inspection Report 50-313/92-26; ,

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50-368/92-26 was similar to the above violation. This ' earlier problem was ,

associated with the accountability of loose parts during Unit 2 steam '

generator repairs. The licensee informed the inspectors that corrective actions regarding the noncited violation had been completed; however, the final closure package had not been assembled. The inspectors did not review the documented corrective actions because the closure package was not availabl Management was cognizant of the problems associated with the Level I housekeeping areas in the spent fuel pool areas and was attempting to resolve ;

these issues. The licensee had initiated condition reports to track and document corrective actions. The General Manager issued a memorandum to all personnel emphasizing the importance of Level I housekeeping requirement The heusekeeping procedure (1000.018) was revised.(Procedure Change 4) to provide clearer guidance and implement a more common sense approach to Level I housekeep ng concepts. Also, a Level I housekeeping monitor watch wa established in the spent fuel pool area with monitors provided by all departments cn site, with participation pro-rated on the number of employees ,

in the deparcmen While perfarming inspection activities in the Unit I spent fuel pool area on September 8,1993, the inspectors noted that the roof over the. spent fuel pool and tilt pit area was leaking (a thunderstorm was occurring at that time).

The inspectors questioned the licensee regarding the possibility of. impurities (e.g., chlorides) being introduced into the spent fuel pool water, thus, changing the water chemistry. The TS required that the chemistry of the spent fuel pool water be controlled and maintained within the established concentration levels for boron, chlorides, and fluorides. The licensee initiated -a condition report in order to evaluate and determine if corrective .

actions were warranted. In addition, the licensee repaired the leaking roo (No leaks were observed by the inspectors in the roof during a subsequent thunderstorm several days later.)

In response to the inspectors' questions, the licensee obtained water samples from the spent fuel pool and performed a chemical analysis. The results of . .

the analysis showed the water chemistry to be within the TS requirements. The inspectors, however, questioned the licensee's chemistry supervisor about the results to ensure that the sample had been taken from the tilt pit cavit The inspectors learned that the samples had been obtained according to established procedure, and the procedural specified sampling had not resulted in the sampling of the tilt pit. This was because the tilt pit was not serviced by the spent fuel pool cooling system and its cavity inventory was separated by a gate from the spent fuel pool. The sampling of the wrong

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-23-cavity was an example of poor interdepartmental communication. Upon bringing l this to the attention of the licensee's representative, a chemistry analysis L of the tilt pit pool was performed. The results were found to be acceptabl During periodic visits to the Unit I containment building, the inspectors observed Level I housekeeping areas to be maintained in good conditio ,

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2.15.2 Conclusion Several observations of foreign materials were made during the inspection of ,

the Unit I reactor vessel and the Unit 1 and 2 spent fuel pool housekeeping Level 1 areas. These observations included examples of a violatio ,

2.16 Maintenance of Fuel Handlina Equipment 2.16.1 Discussion The inspectors requested preventive maintenance (PM) records for the main fuel-bridge, spent fuel bridge, and spent fuel transfer equipment for review. The -

licensee had established preventive maintenance activities for certain fuel ;

handling equipment in Preventive Maintenance Engineering Evaluations (PMEE)

No. 148, " Fuel Handling Cranes," Revision 2. The PMEE listed required preventive maintenance, which included vendor recommendations, for certain fuel handling equipmen !

Discussions with the licensee indicated that the PMEE had not been implemented because of equipment modifications. The inspectors were also informed that preventive maintenance was a combination of typical preventive maintenance activities and pre-refueling operational checkouts. The inspectors requested preventive maintenance job orders for this refueling outage and the previous refueling outage. The inspectors reviewed one Corrective Maintenance Work Order No. 827214 (fuel transfer mechanism) and two Refueling Task Job Orders -

Nos. 859130 and 863618 (nuclear fuel transfer and main fuel handling bridge, ,

respectively). Review of both refueling task orders did not clearly indicate that required preventive maintenance activities were actually performe ,

Further discussions with the licensee revealed that prior to 1Rll, contractor personnel, using ANO procedures and supported by Entergy maintenance "

personnel, performed the preventive maintenance while doing checkouts of the fuel handling equipment. The inspectors requested supporting documentation to verify that preventive maintenance activities were performed. The licensee t could not provide the inspectors with documentation that verified' preventive maintenance activities had been performed on the fuel handling equipment. The-licensee did present the contractor's work plan, which covered some preventive j maintenance requirements; however, there was no documented evidence available -

that the work was performed. Review of Procedure 1502.003, " Refueling Equipment And Operator Checkouts," Revision 12, indicated that operational checks of the equipment were performed using this procedure. However, the procedure also indicated that operational checks did not cover or include all required preventive maintenance identified in the PME i

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The licensee informed the inspectors that the fuel handling equipment was non-  !

safety related; however, it was designed to Seismic Category 1 requirement Regulatory Guide 1.29, which the licensee committed to, required that nuclear power plant structures, systems, and components important to safety be ,

designed to withstand the effects of earthquakes without loss of capability to '

perform their safety function. 10 CFR Part 50, Appendix B, " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," established ,

quality assurance requirements for the design, construction, and operation of nuclear power plant structures, systems, and components that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the publi ,

Subsequent to the inspection, continuing in-office review was performed in .

order to establish the safety-related aspects of the fuel handling equipment '

and the applicability of 10 CFR 50, Appendix Additional information (facsimile received October 8,1993) from the licensee was obtained and discussions with appropriate personnel from the Office of Nuclear Reactor Regulation were conducte It was established that the only

safety-related function of the fuel handling equipment was to remain structurally intact to prevent damage due to overall equipment failure. Other considerations that could require the applicability of 10 CFR Part 50, -

Appendix B to fuel handling equipment, such as calculated offsite dose greater than 0.5 Rem (whole body or thyroid), did not appl .16.2 Conclusion s The licensee had performed operational checks of the fuel handling equipment prior to refueling activities. However, a weakness was identified in that -

refueling task job orders did not clearly indicate what preventive maintenance activities were to be performe .17 Primary Makeup Pumps

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2.17.1 Discussion The inspectors visually examined Unit 1 Primary Pump Rooms P36B and P36 Survey maps of the rooms were posted on the door entrances. The rooms appeared to be fairly clean, radiation and contamination areas were marked clearly. The inspectors questioned the licensee on why these pump rooms were roped off as contaminated. The licensee's representative stated that when the inboard or outboard seals leaked, contamination occurred. Decontamination efforts caused radiation exposure; therefore, it was feasible to rope off the rooms as contaminated to prevent further contamination of other personne The inspectors selected maintenance records for review of Unit 1 Primary Makeup Pump P36A. The inspectors reviewed selected corrective and preventive maintenance job orders to determine if required maintenance activities were performed as required. The inspectors also reviewed associated TS surveillance test records. Review of work packages indicated that maintenance

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activities were performed as required, and review of access records into the radiological control areas supported that determination. Relevant Work Order 1 Nos. 869131, 882944, 872923, and 859094 were completed and properly i documente .17.2 Conclusion

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Maintenance activities on the primary makeup pumps appeared to have been performed in accordance with requirement .18 Polar Crane Problem On September 17, 1993, the inspectors observed the licensee commence reactor vessel head removal operations using the main hoist of the polar cran Pursuant to procedure, the head was lifted off the reactor vessel a few inches and suspended for observation of (a) any suspended equipment under the head .

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and (b) level indication. The head was subsequently lowered back onto the reactor vessel for turnbuckle adjustment to more accurately level the head during the subsequent movements. The head was then raised in one lift sufficiently high enough to clear the alignment pins. Next, the head was tro11 eyed horizontally away from the reactor vessel to the refueling cavity location where the succeeding vertical lift was.to occur. Upon initiating the-lift movement, personnel noticed that the head started to slip downwar Over several ensuing hours into September 18, 1993, eight attempts were made to lift the reactor vessel head, but those attempts were stopped each time as a result of slippage. Licensee personnel estimated that individual slips ranged from 1-6 inches. During this time period, the licensee was in telephonic contact with its crane vendor for technical assistance. On each occasion, the crane's motor brake operated properly. Subsequently, the licensee placed the head onto cribbing, which was placed on the floor of the refueling cavity. After further evaluation, the licensee postulated that the likely cause of slippage was due to a lower than typical bus voltage. The licensee assessed that the crane's induction motor could not generate enough starting torque without initial slippage occurrin Following this determination, the licensee raised (in one movement) the head out of the refueling cavity and placed it on the stan Purchased in 1970 through the Bechtel Corporation, the polar crane was ,

manufactured by the P&H Harnischfeger Corporation. The crane has an electro- '

mechanical motor brake and a magnetorque load brake. The magnetorque load brake is coupled to the motor shaft and produces a retarding torque for the motor when motion is underway. The magnetorque load brake is designed to maintain excitation current so a load will lower at a safe speed if the motor brake fails to hold. The crane has stepless controllers, and was rated for 150 tons; however, the crane vendor was reported to have stated that for short i durations the crane would lift as much as 180 tons. The crane has an auxiliary hoist with a 25-ton capacit ,

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This 1Rll outage was the first occasion for the license.to backfeed Unit 1 through the auxiliary transformer, thus providing a redundant means for bringing offsite power on site. (During some Unit 2 outages, the licensee had backfed that unit through its auxiliary transformer. The Unit 2 polar, crane ,

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motor is not an induction motor like the Unit 1. polar crane motor.) The licensee representatives stated that prior to using the crane, safety bus voltages had been adjusted, but the B7 Load Center that supplies the-crane was,.

found to be operating at 450 volts without load and 440 volts under loa During previous outages, the licensee has backfed Unit 1 through either Startup Transformer 1 or 2. With backfeed through startup transformers, the licensee typically had encountered higher bus voltage than desired. Licensee personnel stated the bus voltage was about 40 volts higher than with auxiliary transformer backfeed. (As a result of concerns over safety equipment being supplied by higher than desired voltage, the licensee had a modification that was underway to reduce bus voltages when backfeeding through a startup transformer.)

t Before beginning reactor vessel head lift operations, the licensee had determined that the lower B7 Load Center voltage obtained by backfeeding through the auxiliary transformer would not place any serviced equipment ou of specification. After this event occurred, the crane vendor representative was reported as having stated that crane motor operation would be permissible with a voltage that was substantially below the observed supply voltage (i.e.,

450 volts). Also prior to placing the polar crane into service, the licensee determined that it passed system preoperational check The reactor vessel head along with the lifting structure and lead shielding (that had not been removed for the lift operations) was reported by the licensee as having weighed 147 tons. . Consequently, during the lift operations, the licensee was attempting to lift a load near the maximum rated load, while supplying the crane motor with a voltage below the desired nominal value. During previous occasions when the reactor vessel head was lifted no a slippage was notice On those occasions, there may not have seen as much lead shielding carried on the reactor vessel hea On September 21, 1993, a meeting was held with the licensee and the inspectors. Also, an inspector in the Region IV office participated in the meeting telephonically. .At that time, the licensee believed that the root cause of the crane slippage had been established and no further evaluation was planned. During this meeting, the inspectors questioned the licensee representatives on whether the crane motor was running hotter than expected, whether the crane motor brake was dragging, or whether there were any loose electrical connections on the crane motor or in the logic circuitr ,

On September 22, 1993, the licensee informed the inspectors that an action plan would be developed prior to reusing the crane to replace the reac_ tor vessel head. Also, the licensee conducted a search of the INPO Nuclear Network data base but did not identify a similar crane problem. The licensee-placed a question on the Nuclear Network to see if other licensees had observed a problem of this nature. As of October 5, 1993, the licensee had ,

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not received any feedback of a similar occurrence. Also, the' licensee prepared a condition report (CR-1-93-0349) for administrative tracking of the- +

crane problem and corrective action On September 23, 1993, the inspectors met with licensee management representatives to review the action plan. The inspectors identified no ,

problems with the plan. During the meeting, the licensee's management representative stated that the crane problem was not considered a safety '

issue. The inspectors differed with the licensee's viewpoint, and were concerned about the lack of significance given to the crane problem. The licensee'; management held the opinion that there was no safety' issue involved, inasmuch as the polar crane was not safety related, the crane brake functioned properly, no TSs were violated, equipment was not damaged, and personnel were not injured. During the meeting, the inspectors expressed the i viewpoint that the crane problem was a safety issue because a heavy safety-related component (reactor vessel head) was being moved when an unplanned ,

evolution (slippage) occurred. Although no damage to the component-or other safety-related equipment or personnel had occurred, the potential was presen The crane problem was later discussed with the plant manager on September 24,  :

1993. The plant manager expressed a conservative view and stated that-he '

would not have allowed the reactor vessel head to have been replaced until further evaluation of the crane had taken place.

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The action plan specified that prior to attempting to replace the reactor vessel head, the 87 Load Center would be energized via one of the startup.-

transformers, test equipment would be placed on the crane hoist and control '

circuitry,. lift verifications would be conducted that would include lifting the reactor vessel head from a suspended-position, and certain other actions would be accomplishe The action plan addressed all inspector concerns that were outstanding. During a conference call on October 1,1993, the ~1icensee's representatives stated that it suspected that not only the reduced voltage contributed to the crane slippage, but also potential wear degradation of the i motor brush contacts may have contributed to reducing the voltage supplied to

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the motor. Consequently, the vendor had identified' additional " fine tuning" of the crane motor that could be performed. This " fine tuning" was basically to adjust mag-amp During the week of October 4, 1993, the licensee lifted the reactor' vessel head from its stand. For this lifting, the licensee powered the_ polar crane 87 Load Center from the Unit 1 Startup Transformer Number 2, which provided_a bus voltage of approximately 500 volts. The licensee's personnel again observed that the reactor vessel head slipped when lifted from a suspended I position. The reactor vessel head was then replaced on the reactor vesse i The licensee's representative indicated that they suspect the crane problem is 1 attributable to the crane's logic' circuitry (perhaps the anti-hunt logic). j The NRC followup of this polar crane problem is an inspection followup item '

(313/9326-02).

2.18.2 Conclusion  ;

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-28-The licensee's investigation into the cause(s) of the polar crane problem is incomplete. The inspectors noted that certain members of the licensee's

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management did not fully appreciate the safety significance or the need to further investigate the polar crane-slippage that occurred while moving the reactor vessel head, a safety-related heavy loa .19 Dry Fuel Storage ,

2.19.1 Discussion The licensee presented the inspectors with an overview of their plans to store spent fuel on site. The licensee representative stated that they have reracked the spent fuel pool for additional storage' three times, and currently i anticipated that_they would run out of storage room in 3 to 5 year Consequently, a plan was being actively pursued to increase onsite fuel storage capacity in a dry-storage facilit The storage facility will be built adjacent to the switchyard in an expanded ,

part of the protected area. The licensee has contracted for major component fabrication, which began in July 1993. The major component consists of a ventilated storage cask, which will be a sealed _ steel vessel within a concrete cask. Each cask will have a 24 fuel assembly capacity. The storage facility is designed to be expandable as need for additional cask storage arise The licensee indicated that the project has been discussed with personnel from the Office of Nuclear Reactor Regulation, and anticipated the.need for a minor .

amendment to the TSs (for overhead crane use) and a 10 CFR 50.59 evaluation prior to putting the facility into use. Site construction for the project was projected for October 1993, with concrete placement for the pad in February'

1994. The project was scheduled for completion in August _1994. The 3 inspectors indicated that this information was useful and would be factored into future NRC inspection consideration .

2.19.2 Conclusion From a brief review, the inspectors concluded that the licensee was appropriately pursuing a means for increasing its onsite storage capacity for )

spent fue j 2.20 Remote Reactor Shutdown I

2.20.1 Discussion  !

i The inspectors spot checked the licensee's Procedure 1203.29 that was !

available for operator use in conducting shutdown of the reactor at a remote -)

location (turbine building). The available prrcedure was determined to be the !

current licensee-approved version (May 1992) and that it contained the latest change (April 2, 1993) that had been issue ,

2.20.2 Conclusion l

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-29-The licensee was maintaining a current version of the remote reactor shutdown procedure at an operator station in the turbine buildin INSERVICE INSPECTION - OBSERVATION OF WORK AND WORK ACTIVITIES (73753)

The objectives of this inspection were to ascertain whether performance of inservice examinations, repair, and replacement of Class 1, 2, and 3 pressure retaining components were performed in accordance with the TS, the applicable'

ASME Code, correspondence between the Office of Nuclear Reactor Regulation and the licensee concerning relief requests, and requirements imposed by NRC and industry initiative .1 Discussion During the week of September 13-17, 1993, the inspectors observed inservice inspection activities pertaining to the ultrasonic examination of a Class 1 piping weld in the reactor coolant system that was identified as ISI Exam 12-002, and the liquid penetrant examination of four Class 1 piping welds in the high pressure coolant injection system, ISI Exams23-055, 23-057, 23- ,

060, and 23-062. The ultrasonic examination was performed using Pr ocedure 1415.022, " Ultrasonic Examination of Dissimilar Metal Welds,"

Revision 4. The liquid penetrant examination was performed using Procedure 1415.004, " Liquid Penetrant Examination ASME Section XI,"

Revision For these examinations, the inspectors verified that: the examinations were performed by contract personnel certified as Level II examiners for the applicable method, the equipment used was calibrated, the system calibration was performed prior to using the transducer for the ultrasonic examination, the temperature of the components and calibration block were within the limits specified in the applicable procedure, the penetrant materials were appropriately certified, and the applicable examination techniques were consistent with the procedure used for the examination. The examinations were performed satisfactorily and no indications requiring evaluation were observed or reported for these weld The inspectors also observed the inservice inspection visual examin_ation, ISI Exam 31-050, of a pipe support for a 36-inch diameter main steam line. The visual examination was performed using Procedure 5120.241, " Inservice Inspection (ISI) Visual Examinations (VT-1 and VT-3)," Revision 0. The examination was performed satisfactorily by a Quality Control Level II examiner certified for performing the VT-3 examination. The quality control examiner identified three construction type discrepancies between the existing pipe support and the support drawing, EBB-3/MS-Ill, Sheet 1, Revision 1. The discrepancies were documented on engineering action requests for engineering to evaluate. No service related degradation of the pipe support was observed or reporte The inspectors reviewed the qualifications and certifications of all nondestructive examination personnel involved in the performance, evaluation,

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-30-and supervision of the inservice inspection activities observed during the inspection. The inspectors were provided satisfactory personnel-certifications for the appropriate level of qualification for each method of nondestructive examination. The certifications also documented a satisfactory annual visual acuity and color test for each of the nondestructive examination personnel. The certifications were found to be current and consistent with the requirements of Procedure QC0-10, " Qualification, Certification, and "

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Training of NDE Personnel," Revision The inspectors reviewed the inservice inspection program plan, "ISI Inspection Plan," Revision 29, for Arkansas Nuclear One Unit 1 Inspection Interval - 2, and verified that the observed examinations were accomplished as scheduled for the current inspection period which is Period 3. The inservice inspection

examinations scheduled for the current outage were documented in the Outage Plan, "1Rll ISI Scope," Revision The inspectors noted that issuance of the inservice inspection plan and outage scope documents, including changes thereto, were satisfactorily controlled in accordance with Procedure 5120.201,

" Control of Inservice Inspection Program Documents," Revision 0, and Change The inspectors also observed the steam generator tube data analyses of three existing explosive plugs, which were examined with the rotating pancake coil eddy current technique. The plug examinations observed were identified as plugs at the upper tubesheet primary surface in Row 87 Tube 81, Row 97 Tube 125, and Row 116 Tube 111. The eddy current data on each plug was ,

independently evaluated by a primary analyst and a secondary analyst who were encloyed by different companies and had their own work stations and computer The data evaluations were consistent with the analysis guidelines in Engineering Standard HES-27, "ANO-1 Steam Generator ECT Data Analysis-Guidelines," Revision 0, for performing .the analysis and characterization of ,

flaw and permeability indications. The inspectors verified that documentation

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was available regarding ' current, satisfactory certification of qualification, ,

physical ability, indoctrination training, and performance demonstration test $ng of the primary and secondary eddy current data analyst The inservice inspection engineer informed the inspectors that based on the initial examination of five plugs that resulted in the identification of flaws in 4 out of the 5 plugs examined, a repair list was generated to repair all Alloy 600 explosive plugs with welded Alloy 690 plugs. The inspectors verified that the approved procedure, identified as Proc / Work Plan 1409.493,

" Installation of Welded Plugs in the ANO-1 OTSG's by BWNS," Revision 0, was available for performing the plug repairs. The inspectors were informed that -

the plug examinations were the only nondestructive examination activities ongoing that pertained to the Section XI repair and replacement progra The inspectors reviewed the licensee's oversight of the contractors performing inservice inspection examinations. The inspectors were informed that the oversight included auditing, surveillance, and overview by the site and corporate nondestructive examination specialists. The inspectors reviewed the ,

most recently completed Quality Assurance Audit QAP-21-93, "ASME B&PV Code Section XI - Inservice Inspection (ISI)," that was performed during the period i

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-July 7 through August 4, 1993. In addition to the audit, the licensee documented quality assurance oversight of inservice inspection activitie during the last 18 months in six surveillance reports. The report numbers were: 92-040,92-063, 92-102,92-106, 93-020, and 93-026. The inspectors concluded that the licensee's oversight-of the inservice inspection activities was goo '

3.2 Conclusions The inservice inspection program was effectively implemented, with performance of work activities and nondestructive examinations observed to be good. The nondestructive examinations were performed by contract personnel that were well qualified for the proces The licensee's oversight activities of the inservice inspection program was goo !

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ATTACHMENT 1

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1 PERSONS CONTACTED 1.1 Enterqy Personnel

• S. Boncheff, Licensing Specialist

1. +05 Bennett, Acting Supervisor, Licensing

E. Burns, Inservice Inspection Er.gineer .

• M. Converse, Supervisor, Engineering Programs

• 1. M. Cooper, Licensing Specialist

1. B. Day, Manager, Unit 1 System Engineering J. Dobbs, Supervisor, Design Engineering

1. R. Douet, Manager, Unit 1 Maintenance

• B. Gordon, Supervisor,-Maintenance Engineering

1. M. Harris, Manager, Unit 2 Maintenance

1. L. Humphrey, Director, Quality Assurance

• 1. + R. King, Acting Director, Licensing

• R. Lane, Director, Design Engineering

• D. Lomax, Manager, Engineering Programs -

D. McKenney, Supervisor, System Engineering

1. M. McKinnen, Reactor Engineer

1. T. Mitchell, Supervisor, System Engineering -

1. + N. Mosher, Licensing Engineer

• K. Panther, Specialist, Non-Destructive Examinations J. Priore, Reactor Engineer

1. S. Pyle, Licensing Engineer

1. T. Reichert, Superintendent, Reactor Engineering i

1. D. Roderick, Assistant Unit 1 Outage Manager

1. E. Rogers, Superintendent, Maintenance Engineering

• J. Ray, Supervisor, Non-Destructive Examinations

• M. Sellman, General Manager F. Smith, Supervisor, Core Design

• M. Spinelli, Design Engineer

1. A. Taylor, Supervisor, Maintenance Engineering .

• 1. J. Vandergrift, Plant Manager, Unit 1 R. Wardlaw, Supervisor, Drafting B. Wilkins, Supervisor, Fuel Fabrication

• 1. J. Yelverton, Vice President, Operations 1.2 NRC Personnel

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• S. Campbell, Resident Inspector

• P. Goldberg, Reactor Inspector

• C. Paulk, Reactor Inspector

1. L. Smith, Senior Resident Inspector In addition to the personnel listed above, the inspectors contacted other personnel during this inspection perio t

• Denotes personnel that attended the exit meeting on September 17, 1993. !

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1. Denotes personnel that attended the preliminary exit meeting on September 24, 199 + Denotes personnel that participated in the telephonic exit conducted on September 30, 199 Denotes personnel that participated in the telephonic exit conducted on <

October 8, 199 '

2 EXIT MEETING An exit meeting pertaining to the inservice inspection was conducted on September 17, 1993, and a preliminary exit meeting regarding the FIRS inspection was conducted on September 24, 1993. During these meetings, the '

inspectors summarized the scope and findings of the inspections. The licensee did not express a position on the inspection findings documented in this ,

report. The inspectors acknowledged that during the inspection, the licensee had provided certain information which was considered proprietary. The inspectors informed the licensee that this information would be destroyed after subsequent in-office inspection of records. Additional telephonic exits were conducted on September 30 and October 8, 1993

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ATTACHMENT 2 I

DOCUMENTATION REVIEWED

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PROCEDURES 1025.019, " System Cleanliness Controls During Modification And Maintenance,"

Revision 4 >

1000.018, " Housekeeping," Revision 19 ,

1409.278, " Fuel Handling Bridge And Transfer System Modification Testing,"

Revision 0 1502.003, " Refueling Equipment And Operator Checkouts," Revision 12 1102.015, " Filling And Draining The Fuel Transfer Canal," Revision 13 1502.010, " Control Of Fuel And Control Rod Movement In Unit 1 Spent Fuel Area," Revision 0 l 1506.001, " Fuel And Control Component Handling," Revision 13 l 1502.004, " Control Of Unit 1 Refueling," Revision 24 1042.003, " Radiochemistry Routine Surveillance Schedule And Technical 1 Specification Reporting," Revision 10  !

l 000.144, " Training And Qualification Of Contractor Personnel," Revision 0 l l

1000.115, " Preventive Maintenance Program," Revision 3 1001.002, " Outage Scheduling and Management," Revision 7 1022.012, " Storage, Control, and Accountability of Special Nuclear Material,"

Revision 16  ;

1506.001, " Fuel and Control Component Handling," Revision 13 1502.004, " Control of Unit 1 Refueling," Revision 24 1022.013, " Preparation and Conduct of Refueling Activities," Revision 5 1203.042, " Refueling Abnormal Operations," Revision 0 1502.003, " Refueling Equipment and Operator Checkouts," Revision 12 1409.492, " Refueling Shuffle," Revision 0 1503.002, " Fresh fuel Receipt Unloading Inspection Report," Revision 6 1503.003, " Control Component inspection," Revision 5

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1010.008, " Industry Events Analysis Program," Revision 5 1015.001, " Conduct of Operation," Revision 46  ;

1504.007, "RX Vessel Closure Head Removal and Storage," Revision 5 5010.020, " Drawing Revision hotices," Revision 3 ,

1504.019 " Reactor Plenum Removal and Replacement (Dry)," Revision 3 Station Directive A2.201, " Overtime," March 31, 1992 -

SQAP-04, " Supplier Qualification / Maintenance of Qualification," Revision 0 QAP 18.04, " Performance of Fuel Vendor Audits," Revision 0

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other Documents

"ANO Unit 1 1Rll Outage Planning Handbook," January 6, 1993 PMEE 148, Fuel Handling Cranes," Revision -2 Section 3.8, ANO Unit' l Technical Specifications, " Fuel loading and Refueling" ,

Memorandum AN0-93-02479, "1Rll Refueling Shutdown Margin," January 16, 1993

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PC 93-7031, " Spent Fuel Pool Lighting Upgrade," Revision 1

"ANO Unit 1 Shutdown Operations Protection Plan," Revision 0 Radioiodine Penetration And Retention Test Report Nos. 7324 and 7251 Fuel Handling Qualification Forms And Lesson Plans Schedule f

"ANO Unit 1 Refueling Outage Master Management Summary Bar Chart," August 14,

1993 Schedule, "ANO Unit 1 Refueling Outage Master Critical Path, Major Milestones, I Shutdown, Refueling Sequence, and Startup," August 23, 1993 q Report, "lRll Shift Outage Meeting," September 24, 1993 Nuclear Fuel Audit Reports Nuclear Fuel Quality Assurance Surveillance Audit Report F93-1, April 16, 1993 ,

Nuclear Fuel Quality Assurance Surveillance Audit Report F93-2, June 4,1993 Nuclear Fuel Quality Assurance Surveillance Audit Report F93-3, June 25,1993 )

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o-3-Nuclear fuel Quality Assurance Surveillance Audit Report F93-4, July 22,1993 Nuclear Fuel Quality Assurance Surveillance Audit Report F93-5, September 3, 1993 Job Orders

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897317 895000 827214 863618 859130 882944 889377 869131 Information Notices and Bulletins IN 81-23 IN 84-93 IN 85-12 IN 86-06 IN 86-58 IN 87-19 IN 88-65 IN 88-92 and Supplement 1 IN 89-31 IN 89-51 IN 90-77 and Supplement 1 IN 91-26 IN 92-21 IN 92-25 Bulletin 84-03 Bulletin 89-03