IR 05000373/1986044

Insp Repts 50-373/86-44 & 50-374/86-44 on 861118-1229.No Violations Noted.Problems W/Verifications Identified. Observation of Switch Activities Continuing.Major Areas Inspected:Maint,Training Lers,Unit Trips & Investigations

ML20209F607
Person / Time
Site:	LaSalle
Issue date:	01/28/1987
From:	Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20209F593	List: IR 05000373/1986044 ML20209F589
References
50-373-86-44, 50-374-86-44, NUDOCS 8702050149
Download: ML20209F607 (15)

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NUCLEAR REGULATORY COMMISSION

REGION III

Reports No: 50-373/86044(DRP); 50-374/86044(DRP)

Docket Nos: 50-373; 50-374 Licenses No. NPF-11; NPF-18 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: LaSalle County Station, Units 1 and 2 Inspection At: LaSalle Site, Marseilles, Illinois Inspection Conducted: November 18 through December 29, 1986 Inspectors: M. J. Jordan R. Kopriva J. Mueller R. Paul lie Approved By: -(f e Wg Chief

/ Reactor Projects Section 2C

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Inspection Summary Inspection on November 18 through December 29, 1986 (Reports No. 50-373/86044(DRP); No. 50-374/86044(DRP))

Areas Inspected: Routine, unannounced inspection conducted by resident inspectors of licensee actions on previous inspection findings; operational safety; surveillance; maintenance; maintenance program; training; Licensee Event Reports; unit trips; onsite followup of events; regional request; and investigation Results: Although there were no violations issued, the licensee continues to have problems with second verifications during surveillance or operational activities. Also, observation of licensee's activities on the SOR differential pressure switches continues, especially in light of the recent SOR switch diaphragm failure $50b$$73 PDR

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DETAILS 1. Persons Contacted

• J. Diederich, Manager, LaSalle Station R. D. Bishop, Services Superintendent J. C. Renwick, Production Superintendent D. Berkman, Assistant Superintendent, Work Planning W. Huntington, Assistant Superintendent, Operations

• P. Manning, Assistant Superintendent, Technical Services T. Hammerich, Assistant Technical Staff Supervisor W. Sheldon, Assistant Superintendent, Maintenance J. Atchley, Operating Engineer

• R. W. Stobert, Quality Assurance Supervisor

• D. Enright, Quality Assurance Engineer

• Denotes personnel attending the exit interview on December 29, 198 . Licensee Action on Previous Inspection Findings (92701)

(0 pen) Open Items (373/85012-01; 374/85012-01): The licensee was to determine how much sediment in station batteries should be considered excessive. A Thermal Group Procedure (TGP) is currently being written by the LaSalle Technical Staff to establish a specific procedure to inspect station batteries. This TGP is to be implemented by February 198 The inspector will revien the TGP upon its issuance and determine whether it satisfies his concern regarding excessive battery sediment. These items remain ope (Closed) Open Items (373/8100-137; 374/8100-40): These items tracked a Unit 1 open item and Unit 2 License Condition 2.C(10).

The licensee was required to resolve the deficiencies described in Appendix 0 to Supplemental Safety Evaluation Report (SSER) No. 7 regarding separation between redundant cables prior to startup after the first refueling outag For both units, the NRC Office of Nuclear Reactor Regulation (NRR) reviewed the licensee's response and issued a Safety Evaluation Report for Unit 1, dated May 15, 1986, and for Unit 2, dated December 2, 1986. NRR concluded that the licensee's analysis and addition of zipper tubing to the cable for the Automatic Depressurization System (ADS) valves satisfied the requirements for these items. The licensee completed the installation of zipper tubing on Unit 1 by Work Request (WR)29-969 and as an original installation prior to startup on Unit These items are considered close (Closed) Open Item (373/8100-39): This item tracked Unit 1 License Condition No. 2.c.21(a) which required that the licensee install a heavy duty turbocharger drive gear assembly on each of the diesel generators prior to the first refueling outage to improve their

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reliability and availability. The inspector confirmed that the licensee has installed heavy duty turbocharger gear drive assemblies, which were approved in NUREG-0519, Safety Evaluation Report, Section No. 8.3.1.1(1), on the 0, 1A, 1B, 2A and 2B diesel generator This item is considered close (Closed) Open Item (373/8100-151): This open item tracked Unit 1 License Condition No. 2.c(13) which required a response to IE Bulletin No. 79-26, Revision 1, " Boron Loss from BWR Control Blades," dated August 29, 1980. This bulletin required operating BWRs to perform various actions including (1) determining boron depletion for individual control blades; (2) limiting the maximum extent of boron depletion in individual control blades under certain conditions, and (3) conducting shutdown-margin tests and modifying shutdown margin requirements to accomodate the boron loss phenomeno NRR evaluated the licensee's response and issued a Safety Evaluation Report dated December 4, 1986, which concluded that the response was acceptable and satisfied the license condition. This item is considered close (Closed) Unresolved Item (373/83014-04): This unresolved item tracked the inspector concerns regarding response time testing of Turbine Stop Valve Closure (TSVC), reactor scram, and End-0f-Cycle Recirculation Pump Trip (E0C-RPT) functions. A review of the surveillance procedure for measuring the reactor protection system TSV fast closure scram response time, LIS-RP-04, identified that the TSV limit switches which initiate the scram signal and supply an E0C-RPT signal were not included in the response time measuremen NRR, in a memorandum from R. Bernero to C. E. Norelius dated November 14, 1986, reviewed the installation of a test switch in parallel with the limit switch to simulate the limit switch function and response time. NRR concluded that the licensee's current method of performing the response time verification test for TSVC is reasonable and will provide adequate assurance that scram response time is within the Technical Specification limit. This item is considered close (Closed) Open Item (373/84005-01): This open item tracked a licensee commitment to modify or eliminate the following valves prior to startup after the first Unit I refueling outage:

1E21-F333 Low Pressure Core Spray (LPCS) inboard testable check isolation valve bypass line isolation valve 1E22-F354 High Pressure Core Spray (HPCS) inboard testable check isolation valve bypass line isolation valve 1E51-F354 Reactor Core Injection Coolant (RCIC) outboard testable check isolation valve bypass line isolation valve 1E51-F355 RCIC inboard testable check isolation valve bypass line isolation valve 1E12-F099A Residual Heat Removal (RHR) inboard testable check isolation valve bypass line isolation valve

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1E12-F099B RHR inboard testable check isolation valve bypass line isolation valve These six automatic containment isolation valves would reopen upon reset of an isolation signal without deliberate operator actio This was not consistent with the requirements of NUREG-0578, "TMI Lessons Learned Task Force and Short Term Recommendations,"

Appendix A, Section 2.1.4, Position 4, which required deliberate operator action to reopen automatic containment isolation valve The licensee chose to eliminate the four LPCS, HPCS, and RCIC valves listed above. This design change was approved by NRR by Amendment 31 to the Unit 1 operating license. The licensee chose to modify the control circuitry of the two RHR valves such that, following a primary containment isolation, resetting of the isolation signal would not result in automatic change of position of the valves. The inspector has reviewed the modification packages for these changes and has no additional concerns. This item is considered close (Closed) Violation (374/83055-01): This item concerned the licensee's failure to perform an adequate engineering review in the area of Emergency Safety Feature (ESF) reset controls. The action remaining on this item was to track the modification of Unit 2 Valves No. 2E12-F099A and No. 2E12-F0998 as committed to in a letter dated February 4, 1984, from C. W. Schroeder to H. R. Denton. This modification was completed and verified by the inspector during the inspection period documented in Inspection Report No. 374/8400 This item is considered close (Closed) Unresolved Item (374/8100-50): This unresolved item tracked Unit 2 License Condition No. 2.c(15)(k) and documented

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inspector concerns that control cabling for the fuel oil transfer pumps for the "0" and "2A" diesel generators passes through the same fire zone, creating the possibility that a single fire could render both fuel oil transfer pumps inoperable and ultimately render both diesel generators inoperable. The licensee has completed a modification which rerouted two control cables for diesel generator "0" fuel oil transfer pump from fire zone 3G-1 to fire zone 2H1-1. This assures that a fire in fire zone 3G-1 will not affect the fuel oil transfer pumps for both diesel generators "0" and "2A". The inspector has reviewed the modification package and verified that the cables have been proper y rerouted. No additional concerns were identified. This item is considered close (Closed) Open Item (373/85024-01): This item tracked the inspector's concern regarding the test procedure for diesel generator air start motors. During the test, one set of two air start motors is isolated by closing an isolation valve while the other set of two air start motors starts the diesel. The concern was whether the volume of pressurized air between the isolation valve and the air start motors not being tested could provide significant starting force to the diesel. Under these conditions, the capability of one set of air

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start motors to start the diesel was not fully tested. The licensee reduced the volume of pressurized air assisting the start of the diesel via the isolated air start motors by changing the isolation valve to a valve which is about one tenth the distance from the isolated air start motors. The licensee changed Procedures No. LOS-DG-M1, -M2, and -M3, " Diesel Generator Operability Test,"

for all five diesel generators to specify the new isolation valve The inspector reviewed these procedures and confirmed that these changes have been incorporated. The inspector observed the replacement and testing of the air start motors on the 1A diesel generator and found the test technically adequate. This item is considered close (Closed) Violation (374/86008-01): On March 2, 1986, the licensee failed to adhere to Procedure No. LOP-RT-05, " Reactor Water Cleanup System Filter /Demineralizer Backwash," for the Unit 2 Reactor Water Cleanup (RWCU) System which subsequently caused the system to isolate. The licensee has taken several corrective actions which include counseling the operators involved, revising procedures to identify critical steps, and repairing the remote mechanical indicators on the valves involved. The licensee's actions appear to be sufficient. This violation is close (Closed) Violation (373/86007-03): On February 26, 1986, the licensee's maintenance procedure for replacing the Unit 1 Standby Liquid Control System (SBLC) explosive valves failed to assure the system piping was properly filled and vented which resulted in a perturbation in the instrument piping causing an unnecessary fast start of the "1B" Emergency Diesel Generator. The licensee walked down the system checking for any degradation. The maintenance procedure, LMP-SC-01, " Standby Liquid Control Explosive Valve,"

has been revised to clarify the filling and venting instruction The operating procedure, LOP-SC-01, has been revised to include instruction on how to fill the SBLC injection line. These actions appear to be sufficient. This violation is close (Closed) Violation (373/86007-04): On February 18, 1986, the licensee failed to obtain a reactor coolant sample which is required once every twenty-four (24) hours when the continuous conductivity monitor is inoperable. The licensee's immediate corrective actions were: Reactor coolant samples were obtained and analyzed for conductivity within the specified sample frequency since 2:00 p.m. (CST) on February 19, 198 The Laboratory Foreman was reprimanded upon his return to the station on March 3, 198 The Shift Foreman was counseled upon his return to the station on March 5, 198 .

. Tailgate meetings on this event were held with the Operating Department on April 24, 1986, and the Radiation Chemistry Department on March 11, 198 This event was discussed at a March 5, 1986, meeting with all station department head The licensee's long term corrective actions were: The Radiation Chemistry Technicians have been retrained to understand the sampling requirements of the Technical Specifications and the proper completion of the attachments to LAP-1800-4, " Rad Chem Surveillance." This training will be completed by June 6, 198 The Radiation Protection Foremen have been retrained to understand the sampling requirements of the Technical Specifications and the requirements for review of the attachments to LAP-1800-4. The Radiation Protection Foreman have initiated review of these attachments (sampling requirements) shiftl Chemistry management has been retrained to understand the sampling requirements of the Technical Specifications, verify that they have been met, and provide a detailed turnover of non-routine sampling requirements to the offshift Radiation Protection Foreman should unusual conditions exis LOP-RT-03, " Reactor Water Cleanup System (RWCU) - Shutdown,"

was revised to include a note to inform the Radiation Chemistry Department when the Reactor Water Cleanup System is shutdow LOS-AA-D1, " Daily Surveillance," was revised to accurately indicate that the surveillance for an inoperable continuous conductivity monitor is applicable at all time LAP-1800-4, " Rad Chem Surveillance," has been revised to accurately indicate that the surveillance for an inoperable continuous conductivity monitor is applicable at all time When chemists are on-site, a chemist will review the surveillance schedule daily, along with the Laboratory Forema LAP-1800-6, " Rad Chem Foreman Shift Turnover," has been revised to require the Rad Chem Shift Foreman to review the Technical Specification sample sheet in the laboratory area on a shiftly basi These actions appear to be sufficient. This violation is close (Closed) Open Item (373/8210-09): This item tracked a generic concern regarding the methodology for tests that measure times used to verify safety-related parameters which are of relatively short

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duration, such as Main Steam Isolation Valve (MSIV) closure. The concern was that the methodology allowed factors such as human error to influence the test data. Recognizing that the question of short time measurement methodology applied, not only to LaSalle, but to other nuclear generating stations, the inspector forwarded this item to NRR for resolution. NRR referenced ANSI /ANS 58.4-1979, "American National Standard Criteria for Technical Specifications for Nuclear Power Plants," Paragraph 5.1.(6), which discusses error assumption Since the NRC Technical Specification basis did not provide a discussion of measurement error for the error associated with the use of stopwatches, NRR concluded that 200 milliseconds was an acceptable assumption. NRR's complete response was forwarded to Region III by memorandum dated January 12, 1983, from D. G. Eisenhut (NRR) to R. Spessard. This item is considered close . Operational Safety Verification (71707)

The inspector observed control room operations, reviewed applicable logs, and conducted discussions with control room operators during the inspection period. The inspector verified the operability of selected emergency systems, reviewed tagout records, and verified proper return to service of affected components. Tours of Unit I and 2 reactor buildings and turbine buildings were conducted to observe plant equipment conditions, including potential fire hazards, fluid leaks, and excessive vibrations and to verify that maintenance requests had been initiated for equipment in need of maintenance. The inspector, by observation and direct interview, verified that the physical security plan was being implemented in accordance with the station security pla The inspector observed plant L,usekeeping/ cleanliness conditions and verified implementation of radiation protection control During the month of December 1986, the inspector walked dcwn the accessible portions of the following systems to verify operability:

Unit 1 Low Pressure Core Spray Unit 1 and 2 Diesel Fire Pump System Unit 1 and 2 High Pressure Core Spray On November 7, 1986, while performing LOS-AA-W1, " Technical Specification Weekly Surveillance," for Control Rod Drive (CRD)

exercises on Unit 2, CRD No. 42-47, when given an insert signal, continued to drift into the full in position. The CRD should have only moved one notch (six inches). The operators verified the CRD to be full in and disarmed the CRD, taking it out of servic Upon inspection of the CRD Hydraulic Control Unit (HCU) insertion Valve (No. 123), small metallic chips were found. These chips were lodged under the valve seat, allowing flow through the valve and inserting the CRD. The licensee has had the metallic chips analyzed and preliminary indication is that the chips are part of the 123 insertion valve. The valve was replaced and the CRD placed back

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into service. The licensee is finalizing the analysis of the metallic chips and is reviewing the concern that this event may effect other CRDs. The resident inspector is following this item which will be considered an Open Item (374/86044-01).

On November 25, 1986, at 1:15 p.m. (CST), while Unit I was operating at approximately 77% power and ramping to 100% power, an Instrument Maintenance (IM) technician was performing monthly instrument surveillance LIS-RD-301, " Scram Discharge Volume Level Alarm,. Rod Block and Scram Functional Test," on Unit 1. When work was started on the IC11-N013F switch, which provides the Scram Discharge Volume (SDV) Hi-level alarm, the switch's inlet (IC11-F361) and vent (1C11-F350) valves were discovered closed. These valves are normally open during operation. This alarm switch performs no reactor safety functions. The IM technician realigned the valves to their proper position and notified the Shift Engineer. The Shift Engineer immediately had LOS-RD-W1 run on Unit 1 to verify the SDV switches were valved in properly, notified the Operating Engineer, and then had the Unit 2 SDV switches verified to be valved in correctl The licensee's investigation found that no work had been performed on the SDV since November 20, 1986, at 11:55 p.m. (CST), when surveillance LOS-RD-W1, " Scram Discharge Volume (SDV) Water Test,"

was completed. On the LOS-RD-W1 checklist, the valves in question had been independently verified to have been left in the correct positio The licensee determined that the root cause of the valves being left in the wrong position was personnel error due to inattention to detail. Since no other work or surveillances had been done on the switch in question, the valves are assumed to have been left in the closed position from the time of the November 20, 1986, surveillanc Also determined to be a contributing cause was.the format of the

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surveillunca, which required double verification on closing the valves in order to do the surveillance, and double verification on i

opening the valves following the surveillance. Since two operators verified each other's work, the " apart-in-time" feature required by

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the independent verification was los The inspector reviewed the licensee's short-term and long-term l corrective actions and found them adequate.

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On November 26, 1986, Unit 1 experienced a Reactor Water Cleanup (RWCU) system isolation. The system isolated from a high differential temperature signal. The isolation caused both the "A" and "B" RWCU pumps to trip. The cause of the high differential temperature signal appeared to have been a spurious electrical spike in the high differential temperature instrumentation. No other causes were found. The RWCU pumps were restarted and the system returned to normal. The licensee performed a complete review of

this event on December 1, 1986.

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On December 3, 1986, at 2:15 p.m. (CST), the inspector observed the reactor building air sampler on the 820' level in the Unit I reactor building alarming on high radiation. The strip chart indicated that it had been alarming for approximately six hour The inspector immediately contacted the Radiation Protection Department. Radiation Protection personnel stated that the air sampler cycles between sampling the IA, IB, and 1C Reactor Water Cleanup (RWCU) demineralizer areas and the air sampler had alarmed on the IB and 1C RWCU demineralizers. The Rad Protection personnel stated that the alarm was thought to be spurious since no Area Rad Monitors (ARMS) had alarmed in the IC RWCU demineralizer roo The inspector contacted the Radiation Chemistry (Rad Chem)

Supervisor who arranged for Rad Chem to take air samples from the area. The samples indicated no increased radiation levels in these areas. The problem was determined to be a faulty power supply for the air sampler. The power supply was replaced on December 18, 1986, and the sampler returned to servic . Monthly Surveillance Observation (61726)

The inspector observed Technical Specification required surveillance testing and verified for actual activities observed that testing was performed in accordance with adequate procedures, that test instrumentation was calibrated, that limiting conditions for operation were met, that removal and restoration of the affected components were accomplished, that test results conformed with Technical Specification and procedure requirements and were reviewed by personnel other than the individual directing the test, and that any deficiencies identified during the testing were properly reviewed and resolved by appropriate management personnel.

l The inspector witnessed portions of the following test activities:

l LOP-RT-08 Unit 2 Reactor Water Cleanup System Strainer Backwash LOS-DG-M2 2A Diesel Generator Operability Test

! On November 22, 1986, during the weekly surveillance of Control Rod l Drive (CRD) exercising (LOS-AA-W1) on Unit 1, CRD 10-47 became uncoupled (loss of Position 48 indication and receipt of the overtravel alarm). CRD 10-47 was inserted to Position 44 in an attempt to recouple the CR Upon withdrawal, the drive failed to reach Position 48 (full out). Subsequent troubleshooting revealed l the following about the status of CRD 10-47:

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the blade was following the drive, (verified by neutron monitoring instrumentation)

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the drive could not be moved to Position 48 or overtravel, but only to a position between 47 and 48 (the overtravel alarm was never received again following the initial uncoupling event, the Position 46 indication lit up, but neither the Position 48 indicator nor the rod fully withdrawn indicator illuminated),

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coupling verification could not be performed by the normal method (with the drive at Position 48, by demanding a single notch withdrawal and verifying that the drive does not go to the overtravel position).

CRD 10-47 was inserted (along with its three symmetric control rod drives) to Position 00 (full in). This complies with Technical Specification 3.1.3.6. The drive at core location 10-47 was replaced during the recent Unit I refuel outage. CRD 10-47, which experienced numerous cycles (insert and withdrawals) and coupling verifications since April 1986 (fuel load), had performed normally prior to this event. That is, no overtravel alarms were received until November 22, 198 The licensee is contemplating the operation of Unit 1 with CRD 10-47 withdrawn. Since CRD 10-47 will not be at Position 48 for the remainder of the cycle, an operating license amendment has been requested to modify Technical Specifications 3/4.1.3.2,3/4.1.3.3, 3/4.1.3.4, and 3/4.1.3.6 to allow operation with the inability to verify couplin The basis for such operation is the high probability that coupling does exist. However, if for some unforeseen reason inadequate coupling exists, there is assurance that there would be no degradation of scram performance and actions could be taken to minimize Rod Drop Accident (RDA) concerns:

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CR0 10-47 will be fully inserted and disarmed when less than or equal to 20% powe This negates RDA concerns at low power and satisfies the basis of Technical Specification 3/4.1.3.6.

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During withdrawal of CRD 10-47 to its target position (46),

l neutron instrumentation (Low Power Range Monitor (LPRM) or Traversing Incore Probe (TIP)) will be monitored to verify that i the control blade is following the drive. This satisfies the basis of Technical Specification 3/4.1. If neutron instrumentation response does not verify that the rod is following the drive, CRD 10-47 will be inserted to Position 00.

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Should surveillance scram testing of the drive be required

! during the course of the cycle, an additional time increment (the scram time between notch Positions 03 and 01, located in the drive's buffer region) will be added to the scram times from l

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positions 45, 39, 25, and 05. This additional time increment will conservatively compensate for using Position 46, instead of Position 48, as the fully withdrawn position. For Technical Specifications 3/4.1.3.2, 3/4.1.3.3, and 3/4.1.3.4 the fully withdrawn position for CRD 10-47 is Position 4 CRD 10-47 will be exercised weekly (Technical Specification 3/4.1.3.1).

General Electric Company (GE) has postulated a scenario as to how CRD 10-47 beccme uncoupled and to why CRD 10-47 cannot be withdrawn to Position 48. GE also indicated that, given LaSalle's potential problem, removal of the CRD from below the reactor vessel is rarely successfu The inspectors will continue to monitor the licensee's action This is considered as an Open Item (373/86044-01).

On November 24, 1986, while performing the calibration of the reactor water low level and Automatic Depressurization System (ADS)

permissive Static-0-Ring (SOR) differential pressure switches, the 2B21-N038A switch (ADS permissive) was found outside of its reject limit. The differential pressure calibration setting was between 58.4 psi and 59.4 psi with a reject limit of less than 56.0 psi or greater than 61.8 psi. The calibration found the switch at 63.4 psi, which was equivalent to a reactor vessel water level of 13 inche The Technical Specification limit is 11 inches. The switch was declared inoperable and a seven day Limiting Condition for Operation was entered. The licensee replaced the switch within the seven day On December 9,1986, during the surveillance on the Unit 2 Reactor Core Isolation Cooling (RCIC) high steam flow isolation logic, SOR differential pressure switch 2E31-N013BB failed the surveillance test. Upon failing the surveillance, the licensee took the actions required by Technical Specifications, took the RCIC system out of

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service, and replaced the 50R differential pressure switch. The i switch failed due to a bad diaphragm.

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On December 15, 1986, while performing surveillance LIS-MS-102,

" Unit 1 Main Steam Line High Flow MSIV Isolation Calibration,"

the main steam line high flow S0R differential pressure switch (1E31-N0080) was found to be in the action range. Normal calibration range for this switch is 96.6 to 98.6 psid. Switch i No. IE31-N0080 was found to trip at 107.6 psid with the reject limit being 108 psid. The licensee replaced the switch and sent the bad switch to the manufacturer for inspection. The licensee reported that the manufacturer could not repeat the problem through testin Upon disassembly of the switch, there were no apparent causes as to the behavior of the switch when installed at the plant. There was a burr found on the switch adjusting screw, but it did not appear to be the problem.

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The licensee's surveillance program seems to be adequate in testing the SOR differential pressure (dp) switches. There is a new concern pertaining to S0R dp switch diaphragm failures. The licensee had identified approximately six diaphragm failures over the last year and a half. The licensee, along with the manufacturer, is following this potential generic proble The resident inspectors continue to follow the licensee's actions due to the history of the SOR dp switche . Monthly Maintenance Observation (62703)

On December 17, 1986, the Maintenance Department had finished replacing a seal on the Unit 2 Motor Driven Reactor Feed Pump (MDRFP). While the MDRFP was being started to check operability of the pump, the seal failed, causing a leak of about 200 gallons per minute (gpm). Approximately 3,000 to 6,000 gallons of water and steam sprayed into the pump room in the turbine buildin The licensee had some difficulty in isolating the leak because of steam and water in the pump room. All of the water drained to the radioactive waste water treatment system, but there was some residual radioactive contamination in the area of the MDRF Decontamination activities of the affected areas are complet There were no personnel injuries or contaminations. A regional based inspector has reviewed the incident at the site and is following the licensee's actions pertaining to the water spill and cleanup effort. The resident inspectors are following the licensee's actions on the MDRFP proble . Maintenance Program (62702)

On December 24, 1986, the inspector attended a meeting with the station Technical Staff, Maintenance Department, and Station Nuclear Engineering Department. The meeting was a planning session for the upcoming Unit 2 feedwater sparger inspection and potential contingencies if indications / cracks are found on the sparger General Electric Company, supplier of the spargers, has informed the

! licensee that the Unit 2 sparger nozzles are susceptible to i cracking.

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The meeting covered the inspection techniques to be used, scheduling i of the inspection, indication / crack evaluation, and temporary and permanent fixes if problems are found.

l There is to be a second meeting planned for January 9,1987.

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Regional inspector followup on this item is anticipate . Training (41400)

The inspector, through discussions with personnel and a review of l training records, evaluated the licensee's training program for operations and maintenance personnel to determine whether the

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general knowledge of the individuals was sufficient for their assigned tasks. Specific areas reviewed are identified in Paragraphs 3 and 4 No items of concern were identifie . Licensee Event Reports (92700)

Through direct observations, discussions with licensee personnel, and review of records, the following Licensee Event Reports (LER's)

were reviewed to determine that reportability requirements were fulfilled, immediate corrective action was accomplished, and corrective action to prevent recurrence had been accomplished in accordance with Technical Specification (Closed) 373/86036-00 - During the first Unit I refueling outage, 104 safety related snubbers failed to meet functional test acceptance limits during a Technical Specification surveillanc Corrective actions have been implemented for all failures including some support mcdifications to prevent failure recurrenc (Closed) 373/86040-00 "A" control room HVAC system ammonia detector tripped due to insufficient flow caused by sample pump degradation. The pump was replaced and an additional detector will be added to each train. The trip logic will be modified to institute a two out of three trip logi (Closed) 374/86011-01 - This revision presents additional testing performed in response to a Unit 2 feedwater transient which caused vessel level to drop below Level 3 trip setpoint. Followup is documented in Inspection Report No. 374/86023. (See Bulletin No. 86-02).

(Closed) 373/85032-02 - This revision provides additional clarification to the cause of a Local Leak Rate Test (LLRT) failure of the 1821-F010B, Inboard Feedwater Check Valv Long term corrective action will be evaluated at the first refueling outage in January 1987.

l (Closed) 373/86039-00 - With Unit 1 at 90% power, the Residual Heat Removal (RHR) high reactor pressure shutdown cooling isolation switch instrument stop Valve, No. 1833-N018B, was found lockwired j closed instead of open. A Notice of Violation was issued in l conjunction with Inspection Report No. 373/86040.

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(Closed) 374/86015-01 - Unit 2 Reactor Water Cleanup (RWCU) System isolated on high differential flow due to flow losses through a l stuck open "B" heat exchanger shell side relief valve. This revision describes a modification to replace the relief valves with valves that have slower opening times and to analyze relief line piping and supports to possibly reduce thermal expansion loadings.

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(, Closed) 373/84062-01 - This revision stated that the followup investigation of a stuck closed shutdown cooling suction inboard isolation valve could not conclusively determine the cause of its failure to open remotely or manually. The valve has since been operated with no recurrence of this problem. See Inspection Report No. 373/84026 for details of the even . Unit Trips (93702)

On December 4, 1986, Unit 1 experienced a reactor scram from 94%

power. The licensee was performing a routine surveillance (LIS-NR-303) of the Average Power Range Monitor (APRM), Channe:

"E", which initiated a half scram signal on the Reactor Protection System (RPS) channel "A". At the same time, the unit received a spurious main steam line "B" high radiation signal. This signal generates a half scram signal to channel "B" of RPS which caused the unit to scram. All systems functioned as expected. The reactor vessel water level appeared to have reached a level below the required scram set point for reactor water level and, upon investigation, there was evidence that the Static-0-Ring switches, used for the low reactor water level scram, actuated within their set point rang The unit was restarted on December 6, 198 . Onsite Followup of Events (93702)

On December 17, 1986, an inboard seal on the motor driven feedwater pump (732' level) failed and caused a leak which resulted in the loss of approximately 3000 gallons of hot (350 F), slightly contaminated main condenser water onto the floor. The leak apparently occurred as a result of an improperly adjusted seal. With the exception of small amounts of

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liquid found on the floor of the 710', 694', and 680' levels, all the water drained into the radwaste system. No employees were injured or contaminated during the steam / water leak and the licensee isolated and controlled the contaminated areas imediately after the event occurre The contamination levels ranged from 100 to 800 dpm/100 cm . The contamination was mainly cobalt-60 and manganese-54. Less than 100 microcuries of radioactive material was spilled. No contamination l was identified on the ductwork or other areas as a result of the

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contaminated water which had migrated to lower floor levels. Air samples taken in the area after the release indicated no airborne activity was presen The contaminated areas were decontaminated and released for

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uncontrolled use by noon on December 18, 1986. To prevent recurrence, the licensee has modified the maintenance procedure for changing out seals by j strengthening the inspection / identification program of replaced seal . Regional Requests (92705)

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A memorandum from C. E. Norelius dated October 28, 1986, which referred j

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to Temporary Instruction (TI) 2500/17, " Inspection Guidance for Heat Shrinkable Tubing," requested the inspectors to determine the extent of deficient splices involving Heat Shrinkable Tubing (HST) at LaSalle County Statio l

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This concern was previously reviewed and documented in Inspection Reports No. 373/86034; No.-374/86035 and Licensee Event Reports (LERs)

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No. 374/86012, No. 374/86013, and No. 374/86014.

1 Investigations (99014)

(0 pen)OpenItem(373/86004-02;374/86004-02): Regional based inspectors reviewed an allegation (RIII-85-A-205) received from an individual

outside.of the licensee's organization pertaining to the adequacy of

~ training provided to fire brigade members and qualifications of certain licensee staff members responsible for the fire protection program at LaSalle County Statio The review determined that while no explicit regulatory requirements were violated, the concerns expressed by the alleger were substantiated and,

therefore, the allegation is considered closed. However, pending NRC review of the results of the licensee's re-evaluation and implementation of corrective actions, this portion of the item remains open,

, 13. Open Items Open items are matters which have been discussed with the licensee, which will be reviewed further by the inspector, and which involve some action

! on the part of the NRC or licensee or both. Open items disclosed during l

the inspection are discussed in Paragraphs 3 and I

14. Exit Interview (30703)

l i The inspectors met with licensee representatives (denoted in Paragraph 1)

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throughout the month and at the conclusion of the inspection period and summarized the scope and findings of the inspection activities. The

, licensee acknowledged these findings. The inspector also discussed the i

likely informational content of the inspection report with regard to '

i documents or processes reviewed by the inspector during the inspection.

. The licensee did not identify any such documents or processes as

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proprietar ,

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