IR 05000285/1989028

Insp Rept 50-285/89-28 on 890701-31.No Violations or Deviations Noted.Major Areas Inspected:Review of Previously Identified Items,Ler Followup,Operational Safety Verification,Plant Tours & Monthly Maint Observations

ML20247B442
Person / Time
Site:	Fort Calhoun
Issue date:	08/23/1989
From:	Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20247B425	List: IR 05000285/1989028 ML20247B419
References
50-285-89-28, GL-89-06, GL-89-6, NUDOCS 8909130035
Download: ML20247B442 (22)

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APPENDIX U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

.NRC Inspection Report: 50-285/89-28 Licensee: DPR-40 Docket: 50-285

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Licensee: Omaha Public Power District (OPPD)

-444 South 16th Street Mall Omaha, Nebraska 68102-2247 Facility Name: Fort Calhoun Station (FCS)

. Inspection At: FCS. Blair, Nebraska Inspection Conducted: July 3-31, 1989 Inspectors: P. H. Harrell, Senior Resident Inspector T. Reis, Resident Inspector Approved: b -

1. 4 7-97 T. F. Westerman, Chief, Project Section B Date Division of Reactor Projects Inspection Summary '

Inspection Conducted July 1-31, 1989 (Report 50-285/89-28)

-Areas Inspected: Routine, unannounced inspection including review of previously identified items; licensee event report followup; operational safety verification; plant tours; monthly maintenance observations; monthly surveillance observations; security observations; radiological protection observations; in-office review of periodic, special, and nonroutine event reports; emergency preparedaets; and the Systematic Assessment of Licensee Performance public meetin Results: During this inspection period, the inspectors reviewed the areas L discussed below. The discussion provides an overall evaluation of each are The inspectors reviewed the actions taken by the licensee in response to previously identified items and licensee event reports. Based on reviews of the actions taken by the licensee it appeared that the licensee had l appropriately implemented both the short- and long-tena actions to prevent

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recurrence of the identified problem During observations of activities and evolutions perfonaed by the operations staff, the inspectors noted no problems with the performance of the staff. It 8909130035 890s29  ;

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appeared that the licensee's operations staff performed their duties in a

. professional manner to ensure safe plant operatio The-inspectors performed numerous tours of the plant during this inspection period. During the tours, no problems were noted. The inspectors did note that the licensee is continuing the ongoing efforts to upgrade plant appearance and housekeepin Maintenance and surveillance activities were observed by the inspectors during

- this insptction period. During observation of these activities, the inspectors noted that the activities were performed in a professional manner. During surveillance observations, problems were noted with system performance. In l testing of the emergency diesel generators, a problem related to elevated

' jacket cooling water temperature caused the licensee to question the validity

.. of a 10 CFR Part 50.59 evaluation relating to a diesel generator modification ,

completed during the last refueling outage. This is an unresolved ite !

During observations of the activities and tasks performed by security and health physics personnel, the inspectors noted that these personnel performed '

their duties in a professional manner.- No concerns were -identified with these activities during this inspection perio I

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-3-DETAILS l Persons Contacted l l

• K. Morris, Division Manager, Nuclear Operations

• G. Peterson, Manager, Fort Calhoun Station A. Richard, Assistant Manager Fort Calhoun Station J. Tills, Assistant Manager, Fort Calhoun Station J. Adams, Reactor Engineer

• J. Bobba, Supervisor, Radiation Protection

• J. Fisicaro, Manager, Nuclear Licensing and Industry Affairs S. Gambhir, Division Manager, Production Engineering Division

• J. Gasper, Manager, Training

• R. Jaworski, Manager, Station Engineering  ;

J. Kecy, Supervisor, Systems Engineering D. Lieber, Supervisor, Security Operations

• G. Manoran Shift Supervisor

• T. Mathews, Station Licensing Engineer

• D. Matthews, Supervisor, Station Licensing K. Miller, Supervisor, Maintenance

• H. Sefick, Manager, Security Services

• C. Simmons, Station Licensing Engineer

• F. Smith, Plant Chemist

• S. Swearngin Engineer, Nuclear Safety Review Group

• D. Trausch, Supervisor, Operations

• S. Willrett, Supervisor, Administrative Services

• Denotes attendance at the monthly exit intervie The inspectors also contacted other plant personnel, including operators, technicians, and administrative personne . Plant Status During this inspection period, the plant operated continuously at 100 percent power. No plant perturbations or demands for safeguerds equipment were experience . Review of Previously Identified Items (92701 and 92702) (Closed)OpenItem 285/8846-06: Inadvertent rise in the reactor coolant system water leve This open item was generated to track licensee's corrective actions subsequent to an unplanned rise in the reactor coolant system (RCS)

water level which occurred on December 5,1988. The plant was in a refueling shutdown at the time with no fuel in the reactor vesse The inspector witnessed licensed operators perform a surveillance of all feasible sources and verify them secured. After approximately

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l-4-L 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, it was determined that the inlet source was demineralized water caused by open manual Valve CH-363 which is used to flush Charging Pump CH-1 The licensee's immediate action was to secure Valve CH-363. The following day, Incident Report (IR) 880480 was generated. As corrective action, the licensee verified that, had there been fuel in the reactor vessel, a boron dilution incident of this type was bounded by the Updated Safety Analysis Report (USAR). Further, the IR stated that to prevent recurrence, all demineralized water inlets to the charging pumps had been changed from " closed" to " lock closed" per Procedure OI-CH-1, " Normal Operation of the Charging System."

Also a comprehensive review of the locked-component list was performed by engineering. Engineering performed a review of all operating instructions to ensure that all locked components in the plant are included in the comprehensive locked-component list. A field check was performed and then used to verify that all applicable operating instructions require a component to be locked if it is designated on the locked-component list. The results of the audit and walkdowns found nine valves which were required to be locked and not presently procedurally controlled. A revision to Procedure 50-0-44,

" Administrative Controls for the Locking of Components," was issued on April 4,1989, to include the previously omitted valve During the aucit, engineering became aware that the 40-month system hydrostatic tests were vulnerable to a boron dilution event. In soms instances, a gate valve is the only barrier preventing demineralized water from entering the RCS. If a valve should leak, dilution would result. The Supervisor, Systems Engineering directed the Procedure Upgrade Program to highlight this condition in the test procedure The procedures have not yet been revise Based on the above, it appeared that the licensee had taken appropriate action to analyze its plant for dilution vulnerabilitie The inspector has reviewed the IR, procedure changes, licensee's safety analysis, and applicable USAR sections and found no discrepancies. Based on the review, this open item is considered closed, b. (Closod)OpenItem 285/8819-03: Inadequacy of new fuel receipt procedur This open item was generated from weaknesses identified during the performance of Procedure SP-NRF-1, " Fuel Receipt Procedures," during June 1988. During the inspection, the inspector observed, on numerous occasions, the uncrating, receipt inspection, and storage of 1 new fuel bundles. The licensee committed to review and take actions, l as appropriate, to the following inspector observations: )

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The procedure in use by the health physics (HP) technicians radiologically surveying the new bundles was not a controlled documen *

Persons were performing the receipt inspection of new fuel in street clothe *

Fuel bundles appeared vulnerable to dataege from the aluminum triped ladder being used for inspectio *

The procedure in use for transferring the new fuel bundles from the receipt inspection location to the new fuel iacks did not require the operator to monitor the drag 1orce between the bundle and the new fuel rac In response to this item, the licensee revised Radiation Protection Procedure VII-9-40, " Radiation Contamination Surveys," to incorporate guidance on the survey of new fuel bundles. Procedure VII-9-40 is a standing procedure applicable to all contamination surveys. Also, the licensee has revised Engineering Procedure SP-NFR-1 to incorporate clothing requirements, proper procedure for movement of the aluminum tripod ladder, and the measurement of any drag force on the new fuel bundles. The inspector reviewed the revised procedures and, based on the reviews, it appeared the licensee has taken appropriate action to address the weaknesses identified by the inspector. This item is considered close c. (Closed) Severity Level IV Violation 285/8822-01: Thermal margin / low pressure trip setpoint error This violation cited a potentially significant error found by the licensee in calculations used to formulate the thermal margin / low pressure trip setpoints of the reactor protection system. The error resulted in setpoints that reduced the margin of safety; however, i other conservation in the calculations offset the nonconservative calculational error The root cause of the error was inadequate design control. Prior to Cycle 11, the licensee performed its setpoint analyses manually using an Office of Nuclear Reactor Regulation (NRR) approved methodolog The results were verified internally within the licensee's i organization. Beginning with Cycle 11, the licensee employed a computer-aided technique of performing the setpoint analyses. The computer program was supplied by Combustion Engineering (CE). CE j provided training to the OPPD staff on the use of the progra Apparently due to a misunderstanding during training, the licensee  ;

wrongly assumed that the required overpower margin (ROPM) tenn and 3 the total integrated radial peaking or penalty factor had been ]

incorporated into the computer program algorithm. However, the CE j instructor assumed that the licensee would it corporate these factor I

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-6-The licensee committed to the following corrective steps to avoid further violations:

Implement procedural requirements to provide qualified personnel to conduct independent reviews of analyse Develop and implement program requirements that ensure training programs match incumbent skills with job requirements for setpoint analyses wor *

Implement a verification and validation program for all safety-related softwar The licensee revised Production Engineering Division Quality Procedure 5 (PED QP-5), " Engineering Analysis Preparation, Review and Approval," to delineate qualification requirements for preparing and providing analyses verification of reload analyses. Procedure PED QP-5 requires that individuals performing independent verification of reload analyses have verified and validated training evaluations on file with the PED training coordinator. Additionally, the procedure requires that reviewers also have previously and successfully completed a similar analysis. The inspector verified that the requirements stated above were incorporated into Procedure PED QP-5, and the licensee has implemented procedural requirements to provide qualified personnel to conduct independent reviews of the analysi Procedure PED QP-5 was also revised to require that the Supervisor, Reactor Physics or the Supervisor, Reactor Performance verify that the individuals assigned reload analysis duties meet the training requirements for that task. Additionally, Procedure PED QP-16,

" Nuclear Production Division Technical Services Administrative Procedure 11," has been revised to require an annual review of the department's activities affecting safety of the FCS and the qualification of department personnel to perform these activitie The annual review of qualifications specifically includes an assessment of the need for training in setpoint analyses work and all other areas affecting safety. Also, Section 3.0 of the PED Training Program Master Plan includes in the manager's duties, scheduling of training to meet individual qualification needs. The plan also states that PED personnel are responsible for ensuring that their qualifications are current for activities in which they are engaged.

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The inspector verified the incorporation of the above into the

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appropriate documents. The licensee has developed and implemented I program requirements that ensure training programs match incumbent skills with job requirements for setpoint analyses wor In April 1988 the licensee's Quality Assurance Division conducted its first audit of the company's computer sof tware control program. The l  ;

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-7-audit was conducted prior to the discovery of the error with the implementation of the program for determining the reactor trip setpoints. While the audit did not find the specific problem noted with respect to the program for determining reactor trip setpoints, it did find significant problems associated with the implementation of the quality assurance program for control of computer software within the Nuclear Production Divisio The audit found numerous deficiencies with respect to software control in each division which made use of safety-related softwar Since the quality assurance plan requirements had been newly implemented, the audit team management decided not to write deficiency reports on the findings but to notify each respective manager, in writing, of the deficiencies in order that action could be taken to achieve complianc A followup audit was conducted by quality assurance in February 198 This internal audit found, in summary, that although seven deficiencies were identified, OPPD had developed an effective software control program in its quality assurance plan and associated divisional procedures. The audit also found that while most of the requirements of the software control program were being effectively implemented by personnel, the licensee noted that more emphasis should be placed upon ensuring that personnel are aware of, and complying with, the requirements of the software control progra Based on the implementation of the software control program via incorporation into the quality assurance plan, the audits performed by the quality assurance department, and the corrective actions taken with respect to audit findings by the licensee's internal divisions, it appeared that the licensee was satisfactorily implementing a verification and validation program for safety-related softwar d. (Closed) Open Item 285/8823-01: U assemblies for the raw water (RW) pgrade pump of thevalve discharge air accumulator In NRC Inspection Report 50-285/89-09, the inspector reviewed and discussed the licensee's modification to upgrade the air accumulator assembly rn RW Block Valves HCV-2850, HCV-2851, HCV-2852, and HCV-2853 to a safety-related status. This was done to providr assurance that the valves would prevent RW backflow through an idle loop since the check valves designed for that purpose were found to be defective. The item remained open in the referenced report due to i apparent deficiencies identified by the inspector. Specifically, the inspector noted that the installed air line for Valve HCV-2851 exhibited excessive vibration and the RW block valves had not been l

included in the inservice inspection program with respect to back leakage.

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On April 12, 1989, the inspector was provided with verification that the subject air line was seismically qualified despite the vibration, l

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On June 30, 1989, the licensee issued a revision to

,1 Procedure ST-ISI-RW-1, " Raw Water Valves Inservice Testing," which

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indirectly verifies that back leakage through RW Block l

Valves HCV-2850, HCV-2851, HCV-2852, and HCV-2853 is acceptable by verifying that each individual RW pump discharge pressure is acceptable when the remaining pumps are idle.

l The inspector reviewed the procedure revision and considers it an L acceptable method of verifying the integrity of the RW pump block valves. Based on the licensee's actions and the inspector's review,

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this item is considered close . (Closed)UnresolvedItem 285/8903-05: Installation o+ Modification Request (MR) FC-88-01 This item involved' discrepancies noted during a review, by the inspector, of the installation of MR-FC-88-011. " Penetration M-73 Upgrade, Instrument Air System." The MR was issued by the licensee to modify the instrument air (IA) containment penetration pipin During review of this modification, the inspector identified a number of discrepancies' The discrepancies identified by this unresolved

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item were discussed with the licensee and conclusions as to the acceptability were reached. Each discrepancy identified is discussed below: Torque wrench information was not recorded in the installation procedure to provide for equipment traceabilit The licensee reviewed quality control records and was able to locate the appropriate torque wrench information. The data was entered into the completed modification documentatio Based on the actions taken by the licensee to retrieve the appropriate data, it appeared that this concern has been properly addresse . TheboundarybetweentheCQE(safety-related)andnon-CQE portions of the IA system was apparently incorrectly establishe The licensee reviewed this item and established that a personnel error occurred when the piping classification designation was added to Drawing II405-M-264. The licensee reissued Drawing M-264 to correct the erro The inspector reviewed the drawing and verified that the error had been corrected. It appeared that the licensee had acequately addressed this concern.

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The specific details of this. concern were reviewed with licensee i'~ personnel. In each case, it was determined that an approved

. < change had been made to the MR to reflect the actual l" # installation of the modificatio _

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Based'on the review of the documentation, it appeared that the

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licensee had taken the actions to ensure that the modification had beenfinstalled in accordance with approved instruction '

Based onithe reviews performed by the ins)ector, it appeared that the licensee's actions adequately addressed t1e identified concern During an ' additional review of selected portions of MR-FC-88-011, no

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additional problems were notad. This item is considered close ' No violations or deviations were 1dentifie .l Licensee Event Report (LER) Followup (92700)

Through direct observation, discussions with licensee personnel, and review of records, the following event report was reviewed to determine

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that deportability requirements were fulfilled, immediate corrective-action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with the Technical Specifications (TS).

The LER listed below is closed:

88-023 Failure of Main Steam Safety Valves to Lift LER 88-023 reported an event where three main steam safety valves (MSSV)

failed to lift within the tolerance specified by TS 2.1.6(3); therefore, the valves were inoperable. The TS specifies that the MSSVs shall lift within a tolerance of plus or minus 1 percent of the nominal nameplate setpoint values. The licensee discovered the failure of the three MSSVs to properly lift during the 1988 refueling outag TS 2.1.6(3) states that eight of the ten MSSVs shall be operable or the plant shall be placed in Mode 3 (hot shutdown) within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> from the time of discovery. The plant was in Mode 3 at the time of discovery of the problem. To address the,inoperability of the MSSVs, the licensee generated Operations Support Analysis Report (0SAR) 88-57, " Safety Valve As-Found Setpoint Variance Analysis." The results of OSAR 88-S7 indicated

. that, eveJ With only seven of ten MSSVs operable, the design basis criteria for the limiting pressures in the reactor coolant and secondary systems could be met. Based on the analysis, the licensee concluded that plant operation was safe with the as-found condition of the MSSVs.

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-10-During the testing of the MSSVs performed in the past refueling outages, problems-were identified with MSSVs not lifting within the TS-specified tolerance. The licensee is reviewing the current methodology used for testing the MSSVs to verify that the method is appropriat Based on the actions taken by the licensee to address inoperability of three of the~ ten MSSVs and the actions in process to review the testing method, this LER is considered closed. The three MSSVs were subsequently refurbished and satisfactorily +?sted by Wylie Laboratory. The inspectors will review the performance of the testing of the MSSVs during the next refueling outag Based on the reviews perfonned by the inspectors, as described above, it appears that the licensee took appropriate actions in response to the identified event to provide tinely corrective actions and implementation of controls to prevent recurrence of the even No violations or deviations were identifie . Operational Safety Verification (71707)

The inspectors conducted reviews and observations of selected activities to. verify that facility operations were performed in conformance with the requirements established under 10 CFR, the licensee's administrative procedures, and the TS. The inspectors made several control room observations to verify the following:

Proper shift staffing was maintained and conduct of control room personnel was appropriat *

Operator adherence to approved procedures and TS requirements was eviden *

Operability of reactor protective system, engineered safeguards equipment, and the safety parameter display system was maintaine If not, the appropriate TS limiting condition for operation (LCO) was

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Logs, records, recorder traces, annunciators, panel indications, and l

switch positions complied with the appropriate requirement *

Proper return to service of components was performe * Maintenance work orders (MWO) were initiated for equipment in need of maintenanc *

Management personnel toured the control room on a regular basi *

Control room access was properly controlle * Control room annunciator status was reviewed to verify operator awareness of plant condition = _ _=

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Mechanical and electrical temporary modification logs were properly maintaine *

Engineered safeguards systems were properly aligned for the specific plant conditio During this inspection period, the following items were reviewed: During this inspection period, the Missouri River, which is the reactor plant's ultimate heat sink, approached a temperature of 85* The USAR indicated that, to ensure adequate cooling of component cooling water (CCW) during normal operation and direct cooling of containment during a design basis accident, the peak river temperature was assumed to be 85 F. To this date, the assuned limit has not been exceeded, but becouse of persistent heat and drought during the previous year, the licensee began to analyze for an elevated river temperature in May 198 On July 7, 1989, the licensee issued Safety Analysis for Operability (SAO)89-012. " Elevated Component Cooling Water Temperature." The analysis demonstrated that a CCW temperature as high as 110*F did not present an unreviewed safety question and is thereby bound by the current USAR accident analysis. The analysis stated that when two RW pumps and three CCW heat exchangers are operating, the allowable river temperature may increase to 92* The analysis was reviewed by the NRC's Region IV Office, and the Office of Nuclear Reactor Regulation (NRR). No significant preblems were noted during the revie However, the licensee determined that the flow rates assumed in SA0 89-012 were higher than measured flow rates. The analysis assumed a combined flow rate of 7200 gpm through two pumps and three heat exchangers; whereas, testing demonstrated that for two pumps on a comon bus, flow could be as low as 5000 gpm, depending on the condition of the heat exchanger Since the river temperature has rever exceeded 85*F and SA0 89-012 has never been implemented, the plant never operated outside its USAR analysis. The condition of RW flow versus river temperature is further discussed in paragraph 8 of this inspection report, During monthly testing of Emergency Diesel Generator (EDG) 2 on July 12, 1989, operators received an alarm indicating high jacket cooling water temperature. Investigation found the local indication .

to be 197*F, although this was a noncalibrated gage. The alarm  !

setpoint is 190*F plus or minus 9.5*F. EDG 2 was declared inoperable  !

pending an investigation. The investigation found no anomalies in either the generator output or the cooling system. EDG 2 was tested

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declared operable after successful completion of Procedure ST-ESF-6,

" Monthly Testing cf the Errergency Diesel Generator."

The licensee discussed the situation with the inspectors and indicated their intent to perform some additional troubleshooting and retest EDG 2 when outside air temperatures increased. The licensee believed that the problem may be attributed to elevated ambient temperatures compounded by heat input from an uninsulated diesel exhaust line. During the additional troubleshooting, the licensee found that the thermostatic control valves to be operating properly, the jacket water alarm setpoint to be within calibration, the coolant level to be within specifications, and the radiator to be unobstructe Procedure ST-ESF-6 was performed again on July 20, 1989. The ambient air temperature was 76*F. Jacket cooling water temperature reached 180*F and no alarm was received. Prior to this run, engineering vented the cooling system.

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On July 26, 1989, the monthly testing for EDG 1 came due and was

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performed. Jacket water temperatures came up quickly to 190 F, the cooling system was vented, but temperatures continued to climb and

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settled out at 200"F. No alarm was received. EDG 1 satisfactorily met the requirements of the surveillance test but the licensee opted to conservatively declare EDG 1 inoperable pending an investigatio As with EDG 2, the instrumentation was found to be within tolerance and no cooling system anomalies coulo be identified. By this time, the licensee had enough data to indicate that a near linear correlation existed between ambient air temperature and diesel room temperature. Diesel room temperature was approximately 20 F above ambient temperatur It appeared that the uninsulated diesel exhaust line was increasing diesel room temperature more +han had been analyzed for when the modification was aproved. The 10 CFR Part 50.59 c. iuatiori for the u s .sulated exhaust line indics.ed a rise in room temperature of 9' Further data gcthering by the licensee and evaluation of the 50.59 analysis by the inspector is necessary prior to assessing the safety implication,s of this analysis; therefore, it will be carried as an unresolved ite (285/8928-01)

During the month, the licensee had been keeping the vendor, Morrison Knudsen Company, appraised of the situation. Enough data was now available to ascertain that there were no performance problems with the cooling system and that the elevated temperatures could be attributed to environmental (diesel room) temperatures. The vendor performed an analysis and determined that the diesel generator trip and alarm setpoints could be raised to 215 F and 208 F, respectivel EDG 1 setpoints have been changed and EDG 2 setpoints will be changed just prior to its next surveillance tes _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ - - _ _ _

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-13-TS 3.7.1.c.iii requires verification that station emergency loads do not exceed the 2000-hour kilowatt (kW) rating of the engine. It can be conservatively implied, therefore, that the diesel must be able to maintain its 2000-hour kW capacity. The licensee has established, by i calculation, that EDG 1 and EDG 2 carry station loads of 2374 kW and 2465 kW, respectivel From the manufacturers curves, che licensee determined that EDG 2 could not meet its station loading requirements, 2465 kW, while maintaining a 2000 hour0.0231 days <br />0.556 hours <br />0.00331 weeks <br />7.61e-4 months <br /> ratino if air intake temperature reached 117'F, which would correspond to an ambient air temperature of 97* Therefore, the licensee declared EDG 2 administratively inoperable when outside air temperature reaches 97*F, until a rmmdy can be achieved. The inspector discussed this approach with NRC Region IV and NRR and all concurred that this was an acceptable approac The licensee considers it inappropriate, however, to prove operability of the alternate diesel when temperatures reach 97 F and inappropriate to reperform Procedure ST-ESF-6 for EDG 2 to prove operability when the temperature falls below 97 F since the status, mechanically, had not changed from the point in time prior to temperature exceeding 97 F. EDG 2 will be logged administratively inoperable but in an emergency, will still receive its start signa EDG 1 is being started to demonstrate operability until this issue is resolve The inspector is monitoring the licensee actions. At this point, the actions are considered aggressive and conservativ c. As a result of a recent discovery of unqualified electrical cable splices located within condulets at the Cooper Nuclear Station, a concern arose that other reactor plants may have similar deficiencies. The inspector performed a suryny to determine controls the licensee has and has had over splices in conduit and condelet It was found that, during 01ignal construction, r,plicing of cables was prohibited by the construction contract. The inspector verifieo the clause in the contrac At this time, no written policy on splicing within conduit or condulets is in place. The licensee states that, in general, the original construction contract is followed, with the exception being where instrument pigtails have been spliced within a conduit in lieu of a junction box to avoid seismic mounting problems. The licensee provides controls to ensure only qualified splices are installed by directing splices in modification packages to be performed in accordance with Procedure ETS-11, " Conductor Splice Installation Specification."

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L .-14-The licensee's two-wire splices are made with Burndy uninsulated, in-line, butt-type. splices covered by properly sized Raychem heat shrinkable sleeves. The three-wire bolted lug splices are made using Burndy uninsulated, compression-type, ring terminal lugs. These splices are then covered by a properly fitted Raychem heat-shrinkable sleeve or V-type kit. Both splices are controlled under Procedure ETS-11. This procedure is employed in both CQE and non-CQE applications with the only difference being that in some cases the splice materials used for non-CQE splices are not CQE and will not retain any traceabilit No violations or deviations were identifie j 6.- P1 ant Tours (71707)

The inspectors conducted plant tours at various times to assess plant and equipment conditions. The following items were observed during the tours:

General plant conditions, including operability of standby equipment, were satisfactor *

Equipment was being maintained in proper condition, without fluid leaks and excessive vibratio *-

Valves and/or switches for safety-related systems vere in the proper positio *-

Plant housekeeping and cleanliness practices were acceptable, including no fire hazards and the coi+.rol of combustible materia *

Performance of work activities was in accordance with approved procedure *

Portable gas cylinders were properly stored to prevent possible missile hazard *

Tag-out of equipment was performed properl *

Management personnel toured the operating spaces on a regular basi No violations or deviations were identifie . Monthly Maintenance Observations (62703)

The inspectors reviewed selected station maintenance activities on safety-related systems and components to verify the ; maintenance was conducted in accordance with approved procedures, regulatory requirements, and the TS. The following items were considered during the reviews:

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L The TS LCOs were met while systems or components were removed from-

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Approvals.were obtained prior to initiating the wor *

Activities'were accomplished using approved MW0s and were inspected, as applicabl *'

Functional testing and/or calibrations were performed prior to returning components or systems to servic *

Quality control records were maintaine Activities were accomplished by qualified personne *

Parts and materials used were properly certifie *

Radiological and fire prevention controls were implemente The inspectors reviewed the following maintenance activities:

Halon indication, west switchgear room, cover light missing and bulb burnt out (HWO 885160)

Chlorine gas toxic gas monitor flow light malfunction (MWO 891014)

125-Yde Battery No. I acid corrosion in spots on th'e battery pack (MWO891887)

• - Raw Water Pump Motor D oil level low both upper and lower bearings

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(MWO892302)

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Testing of EDG 2 power supplies to the speed switches (Ml40 893036)

_ CH-1A Charging Pump A, oil addition to crankcase (MWO 893254)

Although scheduling limited the observation of maintenance activities during this inspection period; completed NW0s 885160, 891014, 891887, 892302, 893036, and 893254 were audited by the inspector to ensure compliance with the licensee's recently implemented improved Maintenance

Procedure 50-M-101, " Conduct of Maintenance." In each case, compliance was verified-in'accordance with the criteria listed in the introduction of

, , ' this paragrap In NRC Inspection Report 50-285/89-26, the inspector noted instances where the NW0s were not filled out in their entirety, which was attributed to the crafts unfamiliarity with the new system. The licensee stated that

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-16-training was in progress. It is assessed that, during this inspection period, documentation of maintenance activities had improve No violations or deviations were identifie . Monthly Surveillance Observations (61726)

The inspectors observed selected portions of the performance of, and/or reviewed, completed documentation for the TS-required surveillance testing on safety-related systems and components. The inspectors verified the following itec- during the testing:

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Testing was performed by qualified personnel using approved procedure *

Test instrumentation was calibrate *

The TS LCOs were me Removal and restoration of the affected system and/or component were accomplishe Test results conformed with TS and procedure requirement Test results were reviewed by perscnnel other than the individual directing the tes Deficiencies identified during the testing were properly reviewed and resolved by appropriate management personne * The test was performed on schedule and it complied with the TS-required frequenc The inspectors observed and/or reviewed the documentation for the

following surveillance test activities. The procedures used for the test l activities are noted in parenthesis:

Monthly check of the station batteries (ST-DC-1)

Monthly testing of the EDGs (ST-ESF-6)

uarterly testing of air accumulator assemblies for RW valves ST-ISI-RW-1)

RW pumps baseline performance test (SP-RW-3)

RW system performance test (SP-RW-4)

A discussion of each surveillance observed is provided below; i

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-17-a. On July 6,1989, the inspector observed technicians performing the monthly test on the station batteries. The testing was conducted in accordance with Procedure ST-DC-1, " Station Battery Checks."

Perform. ace of this evolution included measurement of specific gravity, temperature, and level in each battery cel '

i The inspector noted that the technicians completed this surveillance test in a professional manner. The test equipment used was in i calibration; the test was performed using the procedure, as written; and the technicians observed the appropriate personnel safety precautions while working with battery acid. No problems were identified during observation of this testin b. In a submittal dated December 16, 1987, the licensee requested relief from measuring inlet pressure, differential pressure, and flow on RW Pumps AC-10A, AC-108, AC-100, and AC-10D in accordance with the requirements of ASME Section XI, paragraph IWP-3100, and proposed to monitor pump motor amperage versus pump discharge pressure. In a response dated December 22, 1988, NRR denied the relief request stating the licensee had not demonstrated the impracticality of testing the pumps in accordance with the requirements of Section X During this inspection period, the licensee performed Special Procedures SP-RW-3, " Raw Water Pumps AC-10A, B, C and D Baseline Performance Test," and SP-RW-4, " Raw Water Flow Performance Test," on July 24 and 25, respectively. Procedure SP-RW-3 was performed to acquire baseline reference values for flow and developed head for each pump. Procedure SP-RW-4 was performed to determine flow rates for different combinations of two pumps flowing through three heat exchanger As indicated in paragraph 5 of this inspection report, the baseline data obtained from these test procedures determined that RW flow, from two pumps on a common emergency bus flowing through three RW/ component cooling water heat exchangers, was not as high as was assumed in the licensee's Safety Analysis for Operability (SA0) 89-01 In this analysis, the licensee determined that Missouri River temperature could reach 92*F and still provide adequate cooling to meetdesignbasisaccident(DBA) conditions. The analysis assumed RW flow through three heat exchangers would be at least 7200 gpm; whereas, actual testing revealed flow could be as low as 5000 gpm depending on the lineup of the heat exchangers used and the condition of the heat exchangers at the time. Therefore, it appears that the SAO, as well as cperating procedures, require revision in order that administrative controls exist for all ranges of river temperature to ensure DBA cooling capabilities are maintained. The revision of SA0 89-012 and the incorporation of administrative controls into operational procedures for the condition of elevated river temperatures is considered en inspector followup item. Additional discussion of this matter is provided in paragraph 8.c below.

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(285/8928-02)

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i-18-The inspector reviewed Procedures SP-RW-3 and SP-RW-4 and discussed the results.of t'.e test procedures with the system engineer. No problems were'r.%ed with respect to the methodology and execution of the test c. Procedure Sl'-ISI-RW-3, " Raw Water Pump Inservice Inspection," was -

revised on July 26, 1989, to incorporate acceptance criteria derived from SA0 89-012. The revised procedure required demonstration that various combinations of two RW pumps could provide flow of 7200 gpm through three heat exchangers. The satisfactory performance would ensure that DBA cooling could be aaintained at elevated river temperatures of up to 92 F. . Based on the results of Procedure SP-RW-4, it was doubtful-this could be achieve Procedure SP-RW-4 was completed on July 25, 1989. Between July 25 and the performance of ST-ISI-RW-3 on July 27, 1989, Heat Exchangers AC-10A and AC-10B were sparged for extended periods and the RW flow lineup was changed from Heat s.xchangers AC-10A, AC-108,

and AC-10D to AC-10A, AC-10C, and AC-10D. It is normal procedure to rotate heat exchangers on a daily basis. With this lineup, the minimum flow from two pumps on the same emergency bus was 6700 gpm, which did not meet the acceptance criteria of 7200 gpm. After obtaining the results, an emergency session of the Plant Review Committee was convened to address operability of the RW ' syste It was determined that based on current river temperature of 79 F, the system was certainly. operable, but the conclusion reached in SA0 89-012 that river temperature could rise to 92'F will need to be reevaluate The inspector has reviewed the design basis and, for the above lineup, concurs that the system met operability requirement However, the inspector indicated to the licensee that there are no procedural controls to alert operators to keep RW flow high during periods of elevated river temperature. This topic will be further pursued during the August inspection period by the inspector as part of inspector follow Item 285/8928-0 d. On July 25, 1989, the insp"ector witnessed portions of the performance of Procedure ST-ISI-RW-1, Instrument Air Accumulators Check Valve Operability Test," for RW pump discharge Valves HCV-2850, HCV-2851 HCV-23E2, and HCV-2853. The instrumentation and control technician encountered no difficulties in completing the test, e. On several occasions during the inspection period, the inspector witnessed selected portions of the performance of Procedure ST-ESF-6,

" Monthly Testing of the Emergency Diesel Generator," for both EDG 1 and EDG The inspector had been tracking these tests to follow t, on the problem of elevated jacket water temperature as discussed in

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-19-paragraph 5 of this inspection repart. There were no problems noted with the execution of the test as writte No violations or deviations were identifie . Security Observations (717M)

The inspectors verified that the physical security plan was being implemented by selected observation of the following items:

The security organization was properly manne *

Personnel within the protected area (PA) displayed their identification badge *

Vehicles were properly authorized, searched, and escorted or controlled within the P "

Persons and packages were properly cleared and checked before entry into the PA was permitte * lhe effectiveness of the security program was maintained when security equipment failure or impairment required compensatory measures to be employe * The PA barrier was maintained and the isolation zone kept free of transient materia The vital area barriers were maintained and not compromised by breaches or weaknesse * Illumination in the PA was adequate to observe the appropriate areas at nigh "

Security monitors at the secondary and central alarm stations were functioning properly for assessment of possible intrusion No violations or deviations were identifie . Radiological Protection Observations (71707)

The inspectors verified that selected activities of the licensee's radiological protection program were implemented in conformance with the facility policies and procedures and in compliance with regulatory requirements. The activities listed below were observed and/or reviewed:

Health physics (HP) supervisory personnel conducted plant tours to check on activities in progres HP technicians were using calibrated instrumentatio _ _ _ _ _ _ _ _ _ _ _ -

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Radiation work permits contained the appropriate information to ensure that work was performed in a safe and controlled manne *

Personnel in radiation controlled areas (RCA) were wearing the required personnel monitoring equipment and protective clothing and were properly frisked prior to exiting an RC *

Radiation and/or contaminated areas were properly posted and controlled based on the activity levels within the are During this inspection period, the following items were reviewed: On July 28, 1989, the inspector was notified by the Supervisor, Radiological Protection that material with loose surface contamination reading 1500 dpm was found in a noncontrolled area of discarded material. The equipment, a pressure transmitter, had been stored in a trailer within the owner controlled area but outside the protected area. The licensee has numerous such holding bins of excess material which was supposedly surveyed and released prior to staging ther The licensee is in the process of ascertaining exactly what material and equipment is out in the yard in order to discard what is not needed. As they inventory each trailer, bin, or other storage area, they are radiologically surveying each article prior to ultimate dispositio The inspector surveyed the yard with the Supervisor, Radiological Protection and observed several teams of health physics technicians and supervisors surveying equipment in different locations. The personnel were appropriately protected with anticontamination clothing for their tasks. The inspector relayed'his observations to health physics specialists in the Region IV offic Licensee respiratory protection personnel inquired of the inspector any information regarding an industry report indicating that the NRC-had lifted the ban on wearing contact lenses with full-face respirators. The inspector consulted with specialists from the Region IV office who provided the appropriate correspondence. In a memorandum dated June 5,1989, the Director, Division of Radiation Protection and Emergency Preparedness, NRR, informed Regional Directors.that the NRC was following OSHA's lead in lifting the ban on the use of contact lenses with respirators, and until such time as 10 CFR 20 could be revised, citations of violations were not to be issued. The inspector provided a copy of this nemorandum to the licensee's Supervisor, Radiation Protectio No violations or deviations were identifie I L -

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1 In-Office Review of Periodic Special, and Honroutine Event Reports

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(90/12 and 90/13)

L In-office review of perirt.c, special, and nonroutine event reports was performed by the NRC '.nspectors to verify the following, as appropriate:

Correspondence included the information required by appropriate NRC

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Test results and supporting information were consistent with design

. predictions and specification *

. Planned corrective actions were adequate for resolution of identified-problem , ,

Whether or not any information contained-in the correspondence report should be classified as an abnormal occurrence or additional reactive

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Correspondence did not contain incorrect, inadequate, or incomplete informatio The inspectors reviewed.the following correspondence:

Inservice Testing Program for Pumps and Valves, Program Clarification and Schedule, dated July 3, 1989

.IntroductionofUnauthorizedWeaponintoProtectedArea(LER89-504),

dated July 10, 1989

Special. Report on Inoperability of Fire Barriers, dated July 10, 1989

• Certification of the fort. Calhoun Unit No.1 Safety Parameter Display System to Requirements'of NUREG-0737, Supplement 1 as Required by Generic Letter 89-06, dated July 11, 1989 Monthly Operations Report for June 1989 undated June Monthly Operating Report, dated July 14, 1989 ,

No violations or deviations were identifie . Emergency Preparedness (51332)

.On July 19, 1989, the ins)ector participated in the licenste's emergency preparedness exercise. T1e results of the exercise will be discussed in NRC Inspection Report 50-285/89-2 e n

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o  : 13. .Public SALP Merting (94703)

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., On July 13,-1989, a public meeting was held at the licensee's emergency operations' facility in Omaha, Nebraska. The meeting was held to discuss a the Systematic Assessment of Licensee Performance (SALP) report issued on June 30, 198 ,

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14. ExiiInterview

' The inspectors met with Mr. K. J. Morris, Division Manager, fluclear

- Operations and other members of the licensee staff on August 1,1989. The

- meeting. attendees are listed in paragraph 1 of this inspection report. At this meeting, the inspectors suur.arized the scope of the inspection and the findings. The licensee did not identify eny proprietary information

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