IR 05000285/1987029

Insp Rept 50-285/87-29 on 871101-30.Violations Noted. Major Areas Inspected:Followup on Previously Identified Items,Ler Followup,Operational Safety Verification,Plant tours,safety-related Sys Walkdowns & Monthly Maint

ML20148S935
Person / Time
Site:	Fort Calhoun
Issue date:	01/24/1988
From:	Harrell P, Reis T, Westerman T NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:
Shared Package
ML20148S861	List: IR 05000285/1987029
References
RTR-NUREG-0737, TASK-2.E.1.1, TASK-2.F.2, TASK-TM 50-285-87-29, IEB-87-002, NUDOCS 8802030178
Download: ML20148S935 (19)

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APPENDIX B U. S. NUCLEAR REGULATORY C01411SS10N

REGION IV

NRC Inspection Report: 50-285/87-29 License: OPR-40 Docket: 50-285 Licensee: Omaha Public Power District 1623 Harney Street Omaha, Nebraska 68102 Facility Name: Fort Calhoun Station Inspection At: Fort Calhoun Station, Blair, Nebraska Inspection Conducted: November 1-30, 1987

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Inspector: .

_M.RI I2*T-67 P>q.j%rreQSenior H Resident Reactor Date IT1spector w vu A?- 7-87 T. Reis, Reslident Reactor Inspector Date AI Approved: 9 d /h;v 6 T. F. Westeman, Chief Project J:M -EB Date

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Section B. Division of Reactor Projects 0802030170 000127 PDR ADOCK 05000205 O PDR

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Inspection Summary ,

Inspection Conducted November 1-30, 1987 (Report 50-285/87-29)

Areas Inspected: Routine, unannounced inspection including followup on j previously identified items, licensee event report followup, operational safety t verification, plant tours, safety-related system walkdowns, monthly maintenance observations, monthly surveillance observations, security observations, '

radiological protection observations, in-office review of periodic and special ;

reports, followup on NRC Compliance Bulletin 87-02, and followup on l NUREG-0737 (THI) Items II.E.1.1.1, II.F.2.1, II.F.2.3.8, and II.F. i

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Results: Within the 12 areas inspected, 2 violations (failure to provide postmaintenance testing instructions for operability verification of CQE (safety-related), limited-CQE and fire protection equipment, paragraph 7; and :

4 failure to issue and implement a procedure for surveillance testing of the core t

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exit and heat junction thermocouple systems, paragraph 13) were identified.

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OETAILS

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> Persons Contacted

• Gates, Plant Manager C. Brunnert, Supervisor, Operations Quality Assurance

• M. Core, Supervisor, Maintenance T. Dexter, Supervisor, Security G. Fleischmann, Quality Assurance Inspector

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J. Foley, Supervisnr, I&C and Clectrical Field Haintenance

"L. Gundrum, Plant Licensing Engineer J. Kecy, Acting Reactor Engineer

J. Lechner, Plant Engineer

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• J. Mattice, Plant Health Physicist

• T. Patterson, Supervisor, Technical

• A. Richard, Manager, Quality Assurance G. Roach, Supervisor, Chemical and Radiation Protection 1 *F. Scofield, Supervisor, Outage Projects

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, *J. Spilker, Senior Nuclear Production Engineer D. Trausch, Supervisor, Operations

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• S. Willrett, Supervi'or, Administrative Services and Security

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• Denotes attendance at the monthly exit interview.

. The NRC inspectors also contacted other plant personnel, including operators, technicians, and administrative personnel.

, Followup on Previously Identified Items (0 pen) Severity Level IV Violation II.F.1.m (Deficiency 2.1-9) of NRC Inspection Report 50 285/85-22: Incorrect Information provided in the system design descriptions.

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This item involved the failure of the licensee to maintain system

! design descriptions (SDD) in an up-to-date condition to refle:t

changes made in systems since the S005 were issue To ensure that personnel do not use the inaccurate information contained in the SDDs, the licensee labeled the SDDs as "for information only." The labeling of the 500s indicates to all licensee personnel that the manuals contain inaccurate information and shall not be used for the basis of design change The NRC inspector reviewed a selected sa

nple of SODS to verify that the manuals had been labeled as "for information only. No cases j were noted where the manuals had not been properly labeled. The NRC l inspector also discussed with various licensee personnel, the i

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g significance of labeling the SDDs. In each case, the licensee individual was aware that the SDDs may contain inaccurate informatio This item remains open pending the updating and reissuance of the SDDs as formally controlled documents. The reissuance will be .

performed when the licensee completes the current effort of design o '

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basis reconstitution for all plant safety-related systems. It is a currently anticipated that the design basis reconstitution effort

/. Wil,1 be completed in 199 f (Clo' sed) Severity Level V Violation II.E.2 (Deficiency 3.2-6) of N Inspeetion Report 50-285/85-22: Inadequate instructions for purchase s of critical quality element (CQE) component

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This item was related to the failure of the licensee to provide g specific procedures or guidance for purchaca of the steam generator r.czzle dams. The purchase order for the nozzle dams failed to

,,7 include any requirements related to seismic criteri *

W licensee completed a seismic analysis for the nozzle dams. The

+ results of the analysis indicated that the nozzle dams were capable of withstanding the plant design seismic event. The analysis was i completed prior to the initial installation of nozzle dam To establish a orogrem for providing specific guidance for the purehase of CQE items, the lictnsee issued Procedure N-TSAP-14,

' Determination and Procurement of CQE and Limited-CQE Items and Services." This procedure provided instructinns and guidance for engineers to determir.e whether or not items or services are CQE or limited-CQE and to provide guidelines for implementation of special purchasing requirements associated with CQE and limited-CQE item The licensee performed a review of purchase orders that had been

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issued since 1982 to verify that the appropriate CQE designations had been made, as appropriate. The results of the licensee's review indicated that all purchase orders properly implemented the

• requirements, except for three All three of the purchase orders were issued to Combustion Engineering (CE). CE subsequently supplied documentation for each purchase order to indicate that the items obtained by the purchase orders complied with the licensee's established quality assurance requirement The NRC inspector reviewed the actions taken by the licensee, as discussed above. Based on this review, it appeared that the licensee 8.ad taken actions to correct the identified deficiency and had take7 actions to prevent recurrence of this problem, (Closed) Severity Level IV Violation II.G 3 (Deficiency 2.3-2) of NRC Inspection Report 50-285/85-29: Pen-and-ink changes were made to design installation documents *.;ithout approved field change _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _

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This item is related to licensee personnel making pen-and-ink changes

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to design documentation without processing a formal document chang The basis of this item was that the licensee had not provided a procedure that clearly identified when pen-and-ink changes could be mad This item identified Modification Instruction Packages MR-84-96, MR-83-158, and MR-84-61 as examples of design change documents containing unapproved pen-and-ink changes. Subsequent to the inspection, the three modifications were reviewed by quality control, quality assurance, and the systems acceptance committee. The modifications were found to be adequate and no changes were made by the review group To provide clearer guidance for the performance of field changes, the i

licensee issued a revision to Procedure 50-G-30, "Setpoint/ Procedure i Changes." The procedure revision specifically addressed the details of the mechanism to be used in making the field change and obtaining approval for all controlled documents, including construction installation instructions and drawings. The licensee also revised

Procedure 50-G-21 ~,tation Modification Control," to clarify the circumstances under which pen-and-ink changes could be used in the field during the installation of plant system mcdification The NRC inspector reviewed Modification Instruction Packages MR-84-96, MR-83-158, and MR-84-61 to verify that the pen-and-ink changes made during installation activities were properly

'eviewed and accepted by quality control, quality assurance, and the systems acceptane.e committee. The NRC inspector also reviewed the context of the pen-and-ink changes mide to the packages to verify

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that the cnanges were acceptable. No problems were noted with the revie A review of the changes made to Procedures 50-G-21 and S0-G-30 was performed by the NRC inspector to verify that the procedures adequately addressed the control of field changes. It appeared that the licensee had established an adequate program for field changes made to controlled documentatio d. (Closed) Unresolved Item 285/8710-03: Availability of data from a Technical Specification (TS) amendmen This item involved the failure of the licensee to use the data from a TS amendment during the performance of a test to verify that the trisodium phosphate dodecahydrate (TSP) located in containment was capable of performing its intended safety functio This item is closed based on the issuance of a violation in paragraph 13 of this inspection report. The violation is related to the basis of this unresolved item in that the licensee failed to

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l 8 adequately implement changes made to the operating license through

! issuance of a TS amendmen Closeout of the violation will include a review of concerns identified by this unresolved item, j 3. Licensee Event Report (LER) Followup Through direct observation, discussions with licensee personnet, and .

review of records, the following event reports were reviewed to determine that reporteteility requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had .

been accomplished in accordance with T The LERs listed below are closed:

87-003 Setpoint of tne main steam safety valves were not within tolerance 87-008 Loss of .ffsite power

,87-012 Partial actuation of engineered safeguards features (ESF)

equipment due to the loss of Inverter A A discussion of the review of each LER is provided below

LER 87-003 provided information regarding the as-found setpoints of i t

three of the main steam safety valves (MSSV). The licensee  :

discovered, while perforniing Surveillance Test ST-MSSV-1 during a

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scheduled shutdown of the unit for refueling, that 3 of the 10 MSSVs ,

, were not within the TS required band of 1 percent of their setpoint ,

i values. In each case, the setpoint had drifted higher. The licensee

performed an analysi<, Operations Support Analysis Report (0SAR) 87-17, and determired that the existing

out-of-tolerance setpoints did not adversely affect the loss-of-load i analysis performed for the plant. The resulting pressure transients  !

were below the design basi acceptance criteria provided in i Section 14.9 of the Updated Safety Analysis Report (USAR). ,

The three valves were recalibrated and retested satisfactoril The  ;

licensee attributed the out-of-specification condition to normal i

drift of the safety valve lift setting over an operating cycl To l

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ensure that the setpoint drift was minimized during the next  !

operating cycle, the licensee revised the procedures for retesting of ,

the safety valves after refurbishment. The procedure revisions '

required that the laboratory environment more accurately simulate i actual plant conditions. Specifically, the procedure changes stated l that tle valve inlet neck temperature be within 50*F, instead of the j previous 200*F, of saturation temperature of the inlet steam, and i

holding the temperature condition for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, instead of the previous

specified 15 minutas, prior to the initiation of testin The  !

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licensee contended that these changes in testing conditions are ,

expected to improve future perfe-mance, i

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The NRC inspector reviewed OSAR 87-17 in conjunction with Section 14.9 of the USA Based on this review, it appeared that the licensee's conclusion that the event posed no significant safety hazard was correc b. LER 87-008 reported an event where offsite power was lost during housekeeping activities. The oil-fault pressure relay for the 161-kV transformer was inadvertently tripped causing the loss of the offsite power. The loss of the offsite power feeder caused the loss of the 4160- and 480-volt vital buses for approximately 5 minutes, at which time an emergency diesel generator was placed on the 4160-volt bu At the time of the event, the plant was in a refueling shutdown and the shutdown cooling system was in operation. The loss of pcwer did not cause any core temperature limits to be exceeded due to thc temporary loss of shutdown cooling. Confusion existed in attempts by operations personnel to restore of fsite power in that operations personnel were not aware that housekeeping was being performed on the transformer. In addition, operations personnel did not know how to reset the oil-fault relay due to deficiencies in the operating procedur The relay was reset by an electrician and offsite power was restored in approximately 22 minute The licensee took corrective actions to ensure that this event did not recur. The actions taken by the licensee are discussed below:

. Signs were installed on all transformer boxes warning that power can be tripped by switches inside the transformer bo . The appropriate sections of Operating Procedure OP-10

"Annunciator Response Procedure," were revi n d ai,8 reissued to include adequate instructions on how to respord to '.ne M1-fault relay annunciators for the 345- and 161 kV transformer . Operations and electrical maintenance personnel reviewed the details of this even . Preventive maintenance logs for the transformers were revised to require notification of the shift supervisor prior to performing housekeeping activities on all offsite power transformer The NRC inspector reviewed the actions taken by the licensee in response to this event. It appeared, based on the review, that the licensee had taken actions to correct the problem and had taken actions to prevent recurrenc c. LER 87-012 detailed an event related to the loss of Inverter A during a refueling shutdown. The loss of the inverter initiated actuation of the ventilation isolation actuation signal (VIAS), an ESF syste All systems associated with the VIAS functioned normally. Other ESF

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systems were also actuated by the loss of the inverter; however, the equipment was in a pull-to-lock condition and did not operate. This condition for the ESF equipment was allowed by the T ;

At the time of the loss of Inverter A, the bypass transformer was not ;

available for automatic loading. A check of the inverter by operations personnel indicated that the static switch had transferred !

from the normal supply to backup supply provided by the bypass transformer. When the power was lost, technicians were pulling the power fuses for a steam generator pressure indicating controller. No other maintenance that affected the' inverter output was in process at the time,

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A review was performed by the licensee to determina if the pulling of the fuses caused the inverter to switch to the bypass transformer by i reperforming the steps of the maintenance procedure. When the  :

procedure was reperformed, no inverter problems were initiated. Based ,

on this review, the licensee concluded that the cause of the loss of l the inverter was indete minat t l

The NRC inspector reviewed Maintenance Order (MO) 871595 to verify  ;

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that the instructions provided were clear and concise and did not l

] contribute to this event. Based on the review, it appeared that the [

procedure was adequat A review performed at the time of the event i

. indicated that no other maintenance activities affecting the inverter j

were in progress. Assuming that the maintenance was being performed in accordance with the procedure at the time the inverter output was i lost, it appeared that the reason for the loss could not be  !

determined. The NRC inspectors will continue to monitor the ,

licensee's activities related to the performance of maintenance that !

affect the output of the inverter (

No violations or deviations were identifie . Operational Safety Verification The NRC inspectors conducted reviews and observations of selected

activities to verify that facility operations were performed in conformance with the requirements established under 10 CFR, administrative procedures, and the TS. The NRC inspectors made several control room observations to verify the following

. Proper shift staffing

. Operator adherence to approved procedures and TS requirements

. Operability of reactor protective system and engineered safeguards equipment

. Logs, records, recorder traces, annunciators, panel indications, and switch positions complied with the appropriate requirements

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. Proper return to service of components

. Maintenance orders (MO) initiated for equipment in need of ,

maintenance  !

. Appropriate conduct of control room and other licensed operators

. Management personnel toured the control room on a regular basis -

e On November 11, 1987, the licensee experienced an actuation of the auxiliary feedwater actuation system during the performance of a surveillance test. The actuation was Mttributed to personnel error in that a switch was incorrectly operate The licensee notified the NRC Emergency Operations Center via the

! emergency notification system as specified by 10 CFR Part 50.72. However, i licensee personnel failed to notify either of the NRC resident inspectors ,

of the event. Subsequent to the event, the plant manager issued a memo to all shift supervisor This memo stated that the shift supervisors shall ensure that the NRC resident inspectors are notified of the occurrence of .

all events at the plan Based on the issuance of the memo, it is l l anticipated that one of the NRC resident inspectors will be notified of i any event occurring at the plant.

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No violations or deviations were identifie !

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The NRC inspectors conducted plant tours at various times to assess plant  ;-

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and equipment conditions. The following items were observed during the i j

tours.

. General plant conditions, including operability of standby equipment, ,

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. Equipment was being maintained in proper sondition, withaut fluid leaks and excessive vibration, f

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. Plant housekeeping and cleanliness practices were observed, including I no fire hazards and the control of combustible materia l (

, Performance of work activities was in accordance with approved (

p rocedure t t

. Portable gas cylinders were properly stored to prevent possible i 4 missile hazard ]

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. Tag out of equipment was performed properly, t

. Management personnel toured the operating spaces or regular basis.

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. The auxiliary feedwater pumps were not steam boun No violations or deviations were identifie . Safety-Related System Walkdowns The NRC inspector walked down accessible portions of the following safety-related systems to verify system operability. Operability was determined by verification of selected valve and switch positions. The systems were walked down using the drawings and procedures note . 480-volt electrical distribution system (Procedure 01-EE-2, Checklist A, Revision 11; and Figure 8.1-1 of the USAR, Revision 31)

. Raw water system (Procedure OI-RW-1. Checklist A, Revision 17; and Drawing 11405-H-100, Revision 31)

During the walkdowns, the NRC inspector noted minor discrepancies of an editorial nature between the drawings, procedures, and plant as-built conditions for the selected areas checked. None of the conditions noted affected the operability or safe operation of the systems. Licensee personnel stated that the noted minor discrepancies would be correcte No violations or deviations were identifie . Monthly Maintenance Observations The NRC inspectors reviewed and/or observed 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 and/or observations:

. The TS limiting conditions for operation were met while systems or components were removed from servic . Approvals were obtained prior to initiating the wor . Activities were accomplished using approved MOs and were inspected, as applicabl . Functional testing and/or calibrations were performed prior to returning components or systems tn 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 _ ____ _ _ _ _ - _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _

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I The NRC inspectors reviewed and/or observed the following maintenance activities:

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. Inspection of Valve HCV-808B for water (HO 874599)  ;

. Inspection of Valve HCV-8078 for water (H0 874600) '

. Inspection of Valve HCV-2805A for water (MO 875175)

. Inspection of Valve PCV-514A for water (MO 875166)

. Inspection of Valve FCV-532A for water (M0 875167)

. Inspection of Valve LCV-1173 for water (M0 875183)

The NRC inspector reviewed M0s 875167, 875183, and 875166 associated with disassembly of valves for cxamination for water intrusion from the  :

instrument air system. During this review, it was found that the MOs did  ;

not specify postmaintenance operability tests prior to returning the  ;

j affected CQE (safety-related) components to servic ;

Section 5.8.1 of the TS states, in part, that written procedures and ,

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administrative policies shall be established, implemented, and maintained  !

that meet or exceed the minimum requirements of Section 5.3 of ANSI 18.7-1972. Section 5.3.5(3) of ANSI 18.7-1972, which is applicable j to all safety-related equipment, requires that maintenance procedures  :

i provide instructions for postmaintenance checkout and return to servic Section 6.2 of the licensee's Quality Assurance Plan (QAP) states that l l  !

l postmaintenance testing shall be performed on CQE, limited-CQE, and  !

fire protection equipmen i r

Procedure 50-G-17, "Maintenance Orders", was issued to implement the above  ;

requirements. Paragraph 4.3 of Procedure 50-G-17 states that when -

maintenance is performed on safety-related equipment that is required to  ;

l function during accident conditions, it shall be tested, if deemed t

] necessary by the technical review supervisor (TRS), to verify operability l l Defore it is declared fully operational. This procedure requirement  !

i restricted the applicable scope of postmaintenance testing to equipment  ;

required for accident conditions and did not include all CQE, limited-CQE,  ;

j and fire protection equipment, as required by the QA t

! Contrary to the above, the TRS failed to provide postmaintenance l

instructions for testing of CQE equipment in that no postmaintenance F

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operability testing was specified on M0s 875166, 875167, and 876183 which were issued for repair of safety-related Valves PCV-514A, FCV-532A,

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and tCV-1173, respectively. This is an apparent violation. (285/8729-01)  ;

t l Upon notification by the NRC inspector, the licensee agreed to change the l l requirements of 50-G-17 to conform with the requirements of the QAP. The j

! TRS will be required to specify postmaintenance testing instructions for  ;

I CQE, limited-CQE, and fire protection equipment. At the end of this  !

inspection period, the licensee had not completed the change to l

Procedure 50-G-1 !

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I Monthly Surveillance Observations

~ t t The NRC inspectors observed selected portions of the performance of and/or reviewed completed documentation for the TS-required surveillance testing 4 on safety-related systems and components. The NRC inspectors verified I the following items during the testing:

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. Testing was perfor:1ed by qualified personnel using approved  !

procedure .

Test instrumentation was calibrated,

. The TS limiting conditions for operation were me .

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. Removal and restoration of the affected syste;n and/or component were i

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accomplishe ,

. Test results conformed with TS and procedure requirements.

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. Test results were revievid by persennel other than the individual j directing the tes ;

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. Deficiencies identified during the testing were properly reviewed and l l resolved by appropriate management personn The NRC inspectors observed and/or reviewed the documentation for the  !

I following surveillance test activitie The procedures used for the test i

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activities are noted in parenthesi ,

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. Monthly test of the high containment pressure channels (ST-RPS-8) {

. Monthly test of the low pressurizer pressure signal (ST-ESF-1)

! . Monthly test of the safety injection tank low level signal (ST-ESF-7) I

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. Monthly test of the pressurizer level signal (ST-PL-1)  :

, . Verification of reactivity balance (ST-RA-1) [

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I While witnessing Surveillance Test ST-PL-1, performed to verify

operability of the pressurized level signal, the NRC inspector observed an l electrical anomaly that requires further licensee investigatio The  !

l performance of Steps F.2(20) and F.2(22)(b) appeared to have caused the l l actuation of the emergency feedwater storage tank (EFWST) low level  ;

annunciator. Immediate investigation determined the tank level was not  :

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low. After removing the test equipment, operators and instrument and  ;

control technicians were unsuccessful in their attempts to duplicate the  ;

event. This item remains open pending a review of licensee efforts to I determine and correct the cause of the annunciator actuation for the EFWST  !

low level. (285/8729-02) [

In addition to normally scheduled surveillance testing, the NRC inspector j observed selected portions of the performance of Special Procedure  !

SP-STR0KE-1, "Inservice Testing of Air-0perated CQE Valves." This  !

procedure was performed to test selected safety-related valve operators,  !

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that are not part of the inservice inspection program, for possible

. effects resulting from the intrusion of water into the instrument air '

syste The 38 valve operators that were found to have been wetted during ,

water intrusion into the instrument air system on July 6, 1987, were !

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stroked and timed. Additionally, selected valve operators from each instrument air riser were stroked. Both diaphragm- and piston-type i

. actuators were tested on each riser. None of the valves stroked exhibited ,

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evidence of moisture in the operators. Several exhibited questionable r characteristics during operation such as noise and chattering. These I i characteristics will be reviewed by plant engineering. None of the valves i j stroked failed to operate on demand. The NRC inspectors will perform a i

, followup, during an upcoming inspection period, to review the engineering i evaluation of the data obtained during the performance of (

, Procedure SP-STROKE-1. This item remains open pending the completion of -

the review by the NRC inspectors of the evaluation done by plant r

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engineering. (285/8729-03) l l No violations or deviations were identifie f 9. Security Observations [

] i The NRC inspectors verified the physical security plan was being l implemented by selected observation of the following items: (

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, The security organization was properly manne !

. Personnel within the protected area (PA) displayed their j identification badges, j

. Vehicles were properly authorized, searched, and escorted or I controlled within the P l i

. Persons and packages were properly cleared and checked before entry

into the PA was permitte . The effectiveness of the security program was maintained when security equipment failure or impairment required compensatory {

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. The PA barrier was maintained and the isolation zone kept free of l transient material,

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. The vital area barriers were maintained and not compromised by i breaches or weaknesse !

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. Security monitors at the secondary and central alarm stations were functioning properly for assessment of possible intrusion No viciations or deviations were identifie . Radiological Protection Observations The NRC 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 requirement The activities 'isted below were observed and/or reviewed:

. Health physics (HP) supervisory personnel conducted plant tours to check on activities in progres . Radiation work permits contained the appropriate information to ensure work was performed in a safe and controlled manne . Personnel in radiation controlled areas (RCA) were wearing the required personnel monitoring equipment and protective clothin . Radiation and/or contaminated areas were properly posted and controlled based on the activity levels within the are . Personnel properly frisked prior to exiting an RC No violations or deviations were identifie . In-office Review of Periodic and Special Reports In-of fice review of periodic and special reports was performed by the NRC resident inspectors and/or the Fort Calhoun project inspector to verify the following, as appropriate:

. Reports included the information required by appropriate NRC requirements.

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. Test results and supporting information were consistent with design j

predictions and specificationt.,

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. Determination that planned corrective actions were adequate for resolution of identified problems.

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i . Determination as to whether any information contained in the report should be classified as an abnormal occurrenc The NRC inspectors reviewed the following:

! . October monthly operating report, dated November 13, 1987

. Monthly operations report for October, undated l

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. NRC questions on the Fort Calhoun IST Program, dated October 30, 1987 No violations or deviations were identifie . Followup on NRC Compliance Bulletin 87-02 NRC Compliance Bulletin (CB) 87-02 was issued to require licensees to verify that the fasteners used in the plant meet the purchase specifications under which they were obtained. To verify that the fasteners met the applicable requirements, CB 87-02 required the licensee to select a sample of ten safety-related and ten nonsafety-related fasteners for mechanical and chemical testin A representative from the licensee's quality assurance department and the NRC resident inspector selected a sampling of fasteners and nuts from the licensee's warehouse. The number and types of fasteners and nuts were selected based on plant usage. After selection of the fasteners and nuts, each item was labeled and prepared for shipment to a testing laboratory. A listing of all the fasteners and bolts was compiled for comparison with the testing result The testi'ig results will be formally submitted to the NRC by the licensee in January 198 During a future inspection, a review of the licensee's program for receipt inspection, storage, and issue of fasteners and nuts will be performe C8 87-02 remains open pending completion of this review and a review of the fastener and nut data supplied by the license No violations or deviations were identifie . Followup on NUREG-1737 (THI) Items II.E.1.1.1, II.F.2.1, !!.F.2.3.B, and II.F. A followup was performed to verify that the licensee had coapleted a reliability analysis of the auxiliary feedwater (AFW) system, as required by THI Item !!.E.1.1.1, and to verify that the NRC had reviewed and approved the analysi On November 8, 1985, a meeting was held between the licensee and the NRC staff to discuss the reliability of the AFV system. During the meeting, the licensee presented the preliminary results of the AFW reliability analysis to the NRC staf Based on the NRC staff comments at the meeting, the licensee was directed to submit the final reliability analysi In a letter, dated December 23, 1986, the licensee transmitted the j final reliability analysis to the NR In this letter, the licensee I concluded that their study indicated that the AFW system met the i established unreliability criteria as specified in NUREG-0611

and NUREG-0635.

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i

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. .

In a letter, dated April 20, 1987, the NRC staff requested additional information be provided by the licensee. The licensee provided the additional information in a letter dated May 18, 198 In July 1987, an onsite inspection of the AFW system was performed by NRC headquarters personne The onsite inspection was performed to review the as-built configuration of the AFW system to gain first-hand knowledge of the reliability analysis data that had been provided by the licensee. This onsite inspection was performed besause the FCS is one of seven plants that currently have only two pumps installed in the AFW syste The NRC staff is currently reviewing the reliability analysis submitted by the licensee. This item remains open pending the completion of this review and formal acceptance of the analysis by NRC headquarters personnel, b. A review was performed to verify that the licensee had installed inadequate core cooling instrumentation (ICCI) in accordance with the requirements of TMI Items II.F.2.1, II.F.2.3.B. and II.F.2.4. The NRC inspector reviewed the following elements:

. The ICCI system, comprised of three subsystems, met all the requirements stated in the safety evaluation report (SER) that was issued by the NRC on February 14, 1987. The three subsystems that make up the ICCI system are the saturation margin monitor (SMM), core exit thermocouples (CET), and the heat-junction thermocouples (HJTC) that are used in the reactor vessel level monitoring syste . A TS amendment implementing operability and surveillance requirements for the ICCI subsystems was submitted by the licensee for approval and was subseqicntly approved by the NR . The licensee implemented the approved TS amendment operability and surveillance requirement . Licensed-operator personnel reviewed the TS amendment prior to implementation of the requirement . Operations personnel and shift technical advisors (STA) have been trained in the operation of the safety parameter display system (SPDS) for obtaining data from the ICCI subsystem . The appropriate emergency operating procedures (EOP) were upgraded to include the requirements for review of SMM, CET, and HJTC parameters as an indication of plant status during abnormal or emergency event The results of the review performed by the NRC inspector are discussed below:

_ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ - _

. _ _ _ _ .

• ,

. The installation of the HJTC, SMM, and CET instrumentation complied with the requirements of the TMI items and the SER issued by the NRC. The requirements included display of the appropriate parameters on the SPDS, the appropriate number of channels were provided, the range of indication of each ,

subsystem parameter was appropriate, the subsystem alarm setpoints were correctly established, and the status of each t ICCI subsystem was recorded on the control room alarm monito . On August 31, 1987, the NRC issued Amendment 110 to the TS to include the operability and surveillance test requirements for the HJTC, SHM, and CET subsystems of the ICCI system. The

.

amendment was issued based on an application submitted by the licensee on August 5, 1986, as supplemented by additional information supplied by the licensee on November 24, 1986, and April 15, June 22, and July 1, 198 . Licensee implementation of the operability and surveillance requirements for the ICCI subsystems specified in TS Amendment 110 were found to be inadequate. When Amendment 110 was issued by the NRC, an implementation period of 60 days was granted before the operability and surveillance requirements became effective. With a 60-day implementation period, the requirements of the amendment did not become effective until October 30, 198 On November 10, 1987, the NRC inspector reviewed the status of the ICCI subsystems and associated surveillance tests. The NRC inspector noted that the licensee had not written, issued, or performed any surveillance testing on the HJTC or CET subsystem Section 5.8.1 of the Technical Specifications states, in part, that procedures shall be established, implemented, and maintained that meet or exceed the requirements of Appendix A to Regulatory Guide 1.3 Section 8.b of Appendix A to Regulatory Guide 1.33 states, in part, that specific procedures for surveillance tests should be written for each surveillance test listed in the Technical Specification Contrary to the above, the licensee failed to issue and implement procedures for surveillance testing of the CET and HJTC subsystems on October 30, 1987, the date that Amendment 110 became effective. This is an apparent violation. (285/8729-04)

On November 24, 1987, the licensee issued Procedures ST-CET-1 and ST-HJTC-1 to provide instructions for testing the CET and HJTC subsystem channels. The procedures were successfully completed on November 24, 1987, to establish that the CET and

,

HJTC subsystem were operable.

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.. .

A surveillance test of the SM subsystem channels was performed by the licensee prior to the implementation date of TS Amendment 110 using Procedure ST-SM- Although not officially declared operational until the TS amendment was issued, the SM subsystem has been installed and used since 1984. For this reason, the licensee complied with the operability requirements as each SM subsystem channel was established as operational by the performance of the surveillance tes The licensee has experienced difficulties in the past in implementing requirements established by TS amendment In NRC Inspection Report 50-285/87-10, the NRC inspector identified a problem where the licensee had failed to properly perform a chemical analysis of the TSP stored in containment because the information in TS amendment had not been implemented in the test procedure. To resolve the problem, the plant manager was added to the distribution list for all TS amendments and each amendment submitted by the licensee for approval would request a period of time for implementation of the new amendment requirements. It is expected that the implementation period requested by the licensee will be used to change all appropriate documentation such that the new TS requirements can be implemented at the time the amendment becomes effectiv In this case, the licensee took no action, until notified by the NRC inspector, to implement th6 surveillance testing requirements of TS Amendment 110. After discussions with the NRC inspector, the licensee performed a review of other 75 amendments issued recently. During the review, the licensee identified another case where the requirements of a TS amendment had not been properly implemente TS Amendment 111 changed the requirement for operating the emergency diesel generators fully loaded during the monthly surveillance test from 15 minutes to I hour. Subsequent to the effective date of the amendment, the licensee performed the monthly surveillance test and ran the diesel for only 15 minutes fully loaded instead of the required I hour. The licensee is subnitting an LER, in accordance with 10 CFR Part 50.73, to document the details of the proble The NRC inspector also noted on November 10, 1987, that Channel B of the NTC subsystem was inoperable. The channel had been inoperable since the operability requirement became effective on October 30, 1987. The limiting condition for operation , as specified in Table 2-10 of TS 2.21, requires that the inoperable channel be restored to an operable status within 7 days, or a special report be submitted to the Commission pursuant to TS 5.9.3, within 30 days of discovery of the inoperability. In accordance with the TS requirements, the licensee submitted a special report for the inoperability of Channel B of the NTC subsystem on November 30, 1987. On November 16, 1987, Channel B of the HJTC subsystem was ressored to operabilit ,

<* r

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. On November 10, 1987, the NRC inspector interviewed onshift ;

licensed operators to verify that they had reviewed TS i Amendment 110 prior to its implementation and understood the .

ICCI subsystem operability requirements. Based on the !

interview, it appeared that the onshift operators had not read i

'

or had been given instructions on the new requirements. At the time the interviews were conducted, the operators were not aware that LCOs were in effect for the ICCI subsystem ;

In discussions with licensee personnel, the NRC inspector noted ;

that the licensee had not established a formal internal i distribution system for daumentation received from outside !

sources. For this reason the licensee shall include a i discussion in the response to Violation 285/8729-04, documented i above, what actions will be taken to ensure that documentation !

received from outside sources is promptly distributed internally to the appropriate individual [

!

. The NRC inspector verified that licensed operators and STAS had- l been trained in the use of the SPOS for obtaining data from the l ICCI subsystems. In each case, the operators and STAS l demonstrated that the data could be retrieved and that they 1 understcud the significance of the data with respect to !

post-accident plant status, j

. The E0ps were reviewed to verify that they had been revised to !

include consideration of the ICCI subsystem parameters during !

accident condition. It appeared, based on the review of selected E0Ps, that the licensee had appropriately included ,

requirements for monitoring the ICCI subsystem parameters during ;

accident condition ;

!

In July 1986 a review was performed by NRC inspectors to verify that !

the ICCI system complied with the requirements of Revision 2 of !

Regulatory Guide 1.97. The details of the review are provided in NRC [

Inspection Report 50-265/86-20. In September 1987 an NRC team from I headquarters performed an onsite review of the SPDS displays. No j problems were noted with the displays related to the ICCI system nor were any human engineering deficiencies identified, f

Based on the reviews performed by NRC inspectors, it appeared that t the licensee had installed an ICCI system that complied with the f requirements of THI Items II.F.2.1, II.F.2.3.8, and 11.F.2.4- I therefore, these items are considered close I I

14. Exit Interview  ?

[

The NRC inspectors met with Mr. W. G. Gates (Plant Manager) and other I members of the licensee staff at the end of this inspectio At this f meeting, the NRC inspectors summarized the scope of tne inspection and the ,

findings.