IR 05000313/1987037

Insp Repts 50-313/87-37 & 50-368/87-37 on 881021-23 & 27-30. No Violations or Deviations Noted.Major Areas Inspected: Justification for Continued Operation Commitment Items & Leakage of Lower Insp Handhole Cover

ML20147D157
Person / Time
Site:	Arkansas Nuclear
Issue date:	12/18/1987
From:	Barnes I, Haag R, Stewart R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20147D149	List: IR 05000313/1987037 ML20147D143
References
50-313-87-37, 50-368-87-37, NUDOCS 8801200077
Download: ML20147D157 (10)

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. APPENDIX U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

NRC Inspection Report: 50-313/87-37 Operating Licenses: 'NPF-6-

'50-368/87-37 DPR-51'

Dockets: 50-313 50-368 Licensee: Arkansas Power & Light Company (AP&L)

P.O. Box 551 Little Rock, Arkansas 72203 Facility Name: Arkansas Nuclear One (AN0),. Units 1 and 2 Inspection At: AND Site, Russellville, Arkansas Inspection Conducted: October 21-23 and 27-30, 1987 Inspector: # /

R. C. Stewart, Reactor Inspector, Materials & Dat'e

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Quality Programs Section, Engineering Branch

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gleR.C. Haag,ReactorInspector, Materials & Date Quality Programs Section, Engineering Branch

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Approved: / <t. red _

/a//,s'/y7 1. Barnes, Chief, Materials & Quality Programs Date Section, Engineering Branch Inspection Summary Inspection Conducted October 21-23 and 27-30, 1987 (Report 50-313/87-37)

Areas Inspected: Routine, unannounced inspection of the justification for continued operation commitment items related to the effects of the elevated reactor building temperatures and the leakage of the lower inspection handhole cover on A and B once-through steam generator Results: Within the areas inspected, no violations or deviations were identified. Near-term action items identified in the justification for continued operations commitments appear to have been complete '

8801200077 880114 PDR ADOCK 05000313 0 . .DCD

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Inspection Conducted October 21-23 and 27-30,1987 (Report 50-368/87-37)

Areas Inspected: No inspection of Unit E was conducte Results: Not applicabl !

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DETAILS Persons Contacted AP&L

• D. Lomax, Plant Licensing Supervisor

• 0. Howard, Manager, Special Projects

• P. Michalk, Plant Licensing Engineer H. Green, Quality Assurance Superintendent

• Quennoz, General Manager, Operations

• McGregor, Manager (Acting), Engineering

• Ewing, General Manager, Plant Support

• Huff, Mechanical Engineering Supervisor

• J. Taylor-Brown, Quality Control Superintendent T. McDonnell, Nuclear Services

• J. McWilliams, Manager, Maintenance

• B. Baker, Manager, Operations fiRC

• Harbuck, Resident Inspector

• Denotes those present at the exit meetin The NRC inspectors also contacted other plant' per;onnel, including operators, technicians, and administrative personne . Justification for Continued Operation Commitment Items Background As a result of an initial inquiry by the NRC Senior Resident Inspector on August 3, 1987, concerning unusually high temperatures in the Unit 1 Reactor Containment Building, an inspection was

, performed by an NRC Augmented Inspection Team during August 18-20, 1987. It was ascertained from this inspection that Unit 1 containment ~

temperatures were initially identified in 1974 to range between 103'F and 180 F during plant operations. The Arkansas Nuclear One, Unit 1, Safety Analysis Report, Section 6.3, indicates that the design basis temperature for the reactor containment building under normal operating conditions is 110 F. The licensee was requested to provide a justification for continued operation (JCO) under these high temperature condition The JC0 was provided by the licensee on August 28, 1987.

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b. Scope of Inspection The goal of this NRC inspection was to obtain a general overview of the licensee's JC0 and to verify that applicable JC0 commitments were satisfied during the midcycle outage. The licensee's main objective during the outage was.to complete _the "Near-Term Actions" identified-in Section IV of the JCO. Near-term actions were conducted to:

validate the JC0 evaluation inputs, provide a data base for future problem evaluation and/or resolution, replace or refurbish limited life components as necessary, and identify any limited temporary measures which could further reduce reactor building temperature The NRC inspectors focused their attention on the above items and the extent to which the licensee pursued these item c. Confirmatory Inspections Parformed by the Licensee These inspections were intended to validate assumptions used in the JC0 and to verify conclusions regarding equipment and structure conditions which have been exposed to the elevated reactor building temperature Selected portions of the following areas were included in this inspection:

reactor building liner and penetrations, internal structure concrete and steel (including the Swiss hammer impact tests on s91ected concrete stru;tures),

piping systems,

mechanical equipment, insulation, and instrumentation and electrical component i l

The NRC inspectors reviewed Special Work Plan (SWP) No. 1409.033, l

"ANO-1 Midcycle Outage JC0 Walkdown," Revision 0, dated October 15, 1 1987, which provides instructions to accomplish the inspections and I tests needed to fulfill JC0 commitments. For each inspection the SWP provides individual JC0 walkdown sheets which include the inspection requirements and acceptance criteria. The NRC inspectors verified that all JC0 near-term actions were included in the SWP or referenced to other work instruction l

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The NRC inspectors observed portions of the walkdown inspection of the reactor building firewater' piping and supports. This inspection, which was performed by an engineer and technician, was not limited in scope to only temperature-related deficiencies but also included as-built and general type deficiencies. During this inspection, no problems resulting from elevated temperatures were discovered; however, the licensee did identify several as-built deficiencies, Following the licensee's inspection of the service water cooling coil nozzles on the 2B reactor building cooler, the NRC inspector performed a followup inspection of the nozzles. The results of both inspections indicated no damage to the nozzles. The NRC inspectors also' verified that a location which had a low Swiss hammer impact test result did have surface air-voids. The licensee had accepted of the low impact test result in accordance with ASTM C 805-85 and satisfactory testing on adjacent area The licensee informed the NRC inspectors that the walkdown inspections revealed no damage which could be attributed to reactor building elevated temperatures. Of the discrepancies discovered by the licensee inspections, the NRC inspectors verified from review of engineering evaluations and inspection documentation that proper attention and resolution were given to these items. The NRC inspectors reviewed ?2 completed JC0 walkdown sheet I d. Developing the Data Base To obtain the data base necessary to accurately model reactor 1 building conditions and to support later efforts in determining long l term actions, the licensee performed certain tests, measurements, ,

inspections, and installed additional instrumentatio '

During plant cooldown, an infrared survey of reactor coolant system (RCS) insulation was taken to identify hot spots. A visual inspection of damaged or suspect insulation was also performed. Both inspections revealed numerous areas of deficient insulatio These items were evaluated as not being an immediate safety concer No l corrective action for insulation deficiencies was performed during l this outage. During reactor building inspections, the NRC inspectors ;

noticed similar areas of insulation degradation, particularly on the chilled water system. The condition of insulation is an issue that

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l must be addressed when dealing with the long term actions to reduce reactor building temperature The reactor building cooling system was inspected and tested to verify that performance met the minimum requirement and to provide design input for the long range action plan. Based on the results of the cooling system visual inspection and initial air flow results, the licensee repositioned several dampers and cleaned the cooling coils. Flow for the four chilled water cooling coils met the design requi ements with the exception of Coil VCC-1C, which was within 10 percent of the design flow. This condition only existed with one

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chilled water pump operating. With both chilled water pumps operating, flow for Coil VCC-1C met design requirements. The system design requires that both chilled water pumps run during normal operation The NRC' inspectors observed the reactor building cooling system air flow test which was coordinated by the licensee and actually performed by a contractor. The licensee had proper control of the test with AP&L engineers verifying measurements taken by the contracto Following cleaning of the cooling coils, air flows in the cooling system met the design requirement Additional instrumentation necessary for monitoring of reactor building and cooling system conditions was installed. Using the data gathered from the instrumentation, the licensee can evaluate the

"at power" performance of the reactor building cooling system and obtain a more detailed reactor ouilding temperature profile. The NRC inspectors observed portions of the instrumentation installation and performed a final walkdown inspection of the installed-instrument The NRC inspectors noted that, while the majority of the new instruments are designed for permanent installation, the remaining equipment (i.e., multiplexer, recorders) was temporarily installe The licensee indicated that the temporary equipment will be replaced with permanent equipment during the next outage. The NRC inspectors also reviewed Design Control Plan (DCP) 870-1095, "Containment Atmosphere and Local Component Temperature Monitoring," Revision 0, which installed the additional instrumentatio e. Component Replacement In Section II (Safety Analyses) of the JCO, the licensee identified one component with a qualified lifetime that may be exceeded before the next scheduled outage. The NRC inspectors verified that this component, a reactor building pressure transmitter, was replaced during the outage. No other components were identified by the previously discussed inspections as requiring replacement or refurbishment, l

f. Temporary Temperature Reduction Measures 1 In the area of temperature reduction measures accomplished during the outage, the licensee initiated one direct modification to allow future temperature reduction This modification involved the installation of valves and tie-in connections in the chilled water system. This will provide the capability to isolate the chilled ,

water flow to the reactor building from the chilled water flow to the j

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auxiliary and turbine building. The licensee now has the option of I adding a separate chiller unit to service the isolated auxiliary and l

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turbine building. Presently, the two existing chiller units will j continue supplying chilled water flow to all designed loads. The NRC ]

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inspectors observed portions of-the piping modification and also-reviewed DCP 87-1096, "Chilled Water System Tie-in for Future Chiller."

Based on the visual inspection of the reactor-building coolers and ( the initial airoflow test of the cooling system, the licensee cleaned the cooling coils. With the cooling system air flows now meeting design requirements and the increased cooling capacity of the cooling coils, the reactor building cooling system should now operate at a higher efficienc Quality Control Participation The-licensee's nuclear quality group performed a quality overview of various JC0 tasks accomplished during the outage. Quality assurance and quality control (QC) personnel were assigned to particular JC0 actions in the reactor building, however, while in the reactor l building they randomly reviewed 'other JC0 actions, lvhile observing j the cooling system air flow test, the NRC inspectors noted that a QC inspector assigned to the cooler nozzle inspection was also observing the air flow test. The QC engineering group also performed an initial and final review of the JC0 work pla No violations or deviations were identifie . Once-Through Steam Generator - Lower 5-inch Handhole - Flexitallic Gasket Leakage During the planned outage and a routine containment walkdown on October 22,1987, an ANO-1 operator observed evidence of a small boric acid (reactor coolant) leak on the lower, 5-inch handhole flange of the

"B" once-through steam generator (OTSG). Traces of boric acid leakage were also observed on the 5-inch handhole flange of the "A" OTS There was residual evidence of primary coolant leakage and slight l'

boric acid corrosion on the "A" handhole flange assembly. The "B" handhole flange assembly was found to have steam erosion partially through one stud i bolt and the outer area around the stud bolt. There was no evidence of damage to the stainless steel cladding, nor was there evidence of corrosion wastage. The leakage of primary coolant was estimated by the licensee to be less than .01 gpm. Repairs were made in accordance with Ap&L maintenance procedures, which included new threaded studs, new gaskets, and a new flange cover on the "B" 0TS )-

During the subsequent heatup phase of startup (week of November 8,1987),

both the "A" and "B" 5-inch handhole flange gaskets were again observed to show evidence of leakage. During the subsequent gasket replacement, the licensee employed the use of 5 millimeter thick "grafoil" tape on the flexitallic gaskets which appears to have corrected the leakage proble The licensee considers this particular repair as an interim fix until further review and analyses are mad ,

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During a conference call with the cognizant licensee personnel on November 10, 1987, the NRC inspector was. informed that the problem has '

been extensively discussed with both the OTSG designer, Babcock &

Wilcox (B&W), and Flexitallic Company, Both firms have concurred with the interim fi This matter is considered an open item until final corrective action is initiated (313/8737-01).

No violations or deviations were identifie . Review of Prior Reactor Coolant Pressure Boundary Leakage and Corrective Action Commitments In view of the current leakage problem with the OTSG 5-inch handhole '

flanges and the prior leakage corrosion damage found on the "A" high pressure injection (HPI) nozzle (October 1986), the NRC inspector reviewed prior corrective actions and commitments resulting from those events. In addition, the NRC inspector reviewed the licensee's response and commitments made relative to IE Bulletin No. 82-02, "Degradation of Threaded Fasteners in the Reactor Coolant Pressure Boundary of PWR Plants," dated June 2,198 Documents reviewed included the following:

AP&L letter, John R. Marshall to John T. Collins, NRC, dated August 2, 1982 (No. OCAN088201), "Initial Response to IEB 82-02" AP&L letter, John R. Marshall to John T. Collins, NRC, dated September 29, 1982 (No. OCAN098210), "Additional Information Concerning IEB 82-02" AP&L letter, John R. Marshall to John T. Collins, NRC, dated November 8, 1982 (No. OCAN118209), "Additional Information Concerning IEB 82-02" AP&L letter, John R. Marshall to John T. Collins, NRC, dated December 31, !

1982 (No. 2CAN128218), "Report Required by IEB 82-02, Action Item 4" l AP&L letter, John R. Marshall to W. C. Seidle, NRC, dated March 24, 1983 (No. 2CAN038312), "Quality Assurance Regarding Threaded Fasteners" AP&L letter, John R. Marshall to W. C. Seidle, NRC, dated April 21, 1983 (No. 2CAN048308), "Quality Assurance Regarding Threaded Fasteners" AP&L letter, John R. Marshall to John T. Collins, NRC, dated July 25, 1983 (No. ICAN078310), IEB 82-02, Action Item 4" AP&L inter C apany Correspondence, David Young to A. B. McGregor, dated ;

November 6, 1986 (No. ANO-86-14401) "Workplan for Concentrated Boric Acid l Attack on ANO, Unit 1 RCS" i l

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NRC IE Inspection Report 50-313/82-19, dated October 7, 1982  !

NRC IE Inspection Report 50-313/82-30, dated December 7, 1982 NRC IE Inspection Report 50-313/82-32, dated December 23, 1982

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NRC IE Inspection Report 50-313/83-10, dated June 15, 1983

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NRC IE Inspection Report 50-313/83-34, dated January 20, 1984

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NRC IE Inspection Report 50-313/86-36, dated December 23, 1986 In reviewing the historical documentation relative to threaded fasteners in the reactor coolant pressure boundary (RCPB), the NRC inspector noted

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that ANO Units 1 & 2 have had a limited number of primary coolant leaks occurring on bolted closures during their commercial operating historie It was noted~by the NRC inspector that during the 1982-83 period, in

] response to IEB 82-02 action items, the-licensee expanded the then existing administrative program controls and procedures (which include training and inspection requirements applicable to RCPB fasteners).

As a resu'it of discovering boric acid corrosion during an inservice ,

inspection (ISI) examination of the "A" HPI nozzle in October 1986, IE Inspection Report 50-313/86-36, dated December 23, 1986, the licensee initiated an inspection and evaluation program involving 68 locations in the RCS, which had the potential for leakage and to determine whether the corrosion mechanism was active in those area The licensee's workplan, dated November 6, 1986, included the following statement:

, "Acceptance criteria for preliminary concentrated boric acid attack inspections will be as follows:

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! There shall be no red rust on carbon steel surface . There shall be no visible residual boron deposit . There shall be no corrosion attack, pitting or grooving of carbon steel surfaces of dissimilar weld metal interface.

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Indications that fail to meet any of the three above criteria will be

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documented in Section E of the worksheet. These will be evaluated for j further action."

The NRC inspector observed that records of inspections performed on

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November 17 and 20, 1986, identified residual deposits of boric acid on j OTSG "A" and "B" handhole flange bolts, respectivel The evaluations

regarding those deposits state, "These crystals should be removed during j IR8 (Unit 1, refueling outage 8), and the bolts should be reinspected at that time."

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Since the primary coolant erosion.found on the lower handhole flange assembly of OSTG B discovered during this outage was of-principal concern ,

during this inspection, this matter was ey.tensively discussed during the exit meeting held on October 30, 198 The licensee representatives indicated that since.the boric acid crystals did not appear to threaten the structural integrity of the OTSG and because of ALARA considerations (150R at the surface), cleaning and reinspection was deferred. However,.in view of the present concern, the

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licensee committed to reviewing the acceptance' criteria established in the

methods of evaluating this type of. inspection / observations. This matter-is considered an open item pending completion of licensee review of acceptance criteria and NRC evaluation (313/8737-02).

No violations or deviations were identifie . Exit Interview i The NRC inspectors met with the licensee personnel denoted in paragraph 1

and the resident inspector on October 30, 1987, and summarized the scope and finding of the inspection.

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