IR 05000409/1985022

Insp Rept 50-409/85-22 on 851217-860214.No Violations Noted. Major Areas Inspected:Licensee Actions on Previous Insp Findings,Operational Safety Verification,Monthly Maint Observation,Ler Followups & TMI Action Items

ML20153G140
Person / Time
Site:	La Crosse File:Dairyland Power Cooperative icon.png
Issue date:	02/21/1986
From:	Boyd D NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20153G128	List: IR 05000409/1985022 ML20153G124
References
TASK-2.B.3, TASK-TM 50-409-85-22, NUDOCS 8602280041
Download: ML20153G140 (15)

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U. S. NUCLEAR REGULATORY COMMISSION

REGION III

Report No. 50-409/85022(DRP)

Docket No. 50-409 License No. DPR-45 Licensee: Dairyland Power Cooperative 2615 East Avenue - South La Crosse, WI 54601 Facility Name: La Crosse Boiling Water Reactor Inspection At: La Crosse Site, Genoa, WI ,

Inspection Conducted: Decembe- 17, 1985 through February 14, 1986 Inspectors: I. Villalva J. Wiebe

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Approved By: D. C. Bo'yd, Cfief 2-2 /~ N Reactor Projects Section 2D Date Inspection Summary Inspection from December 17, 1985 through February 14, 1986 (Report No. 50-409/85022(DRP))

Areas Inspected: Routine, unannounced inspection by the resident inspectors of Licensee Actions on Previous Inspection Findings; Operational Safety

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Verification; Monthly Maintenance Observation; Licensee Event Reports Followup; Systematic Evaluation Program Action Items; TMI Action Items; Regulatory Improvement Program; Plant Trips; Preparation for Refueling; Organization and Administration; and Onsite Review Committee. die inspection involved a total of 140 inspector-hours onsite by two NRC_ inspectors including a total of 20 inspector-hours during back shifts.

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Results: No violations of NRC requirements were note PDR O ADOCK 05000409

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DETAILS 1. Persons Contacted

• J. Parkyn, Plant Superintendent

• G. Boyd, Operations. Supervisor

• L. Kelley, Assistant to Operations Supervisor

• L._ Nelson, Health and Safety Supervisor

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R. Wery, Quality Assurance Supervisor ,

M. Polsean, Shift Supervisor L. Goodman, Operations Engineer R. Brimer, Electrical Engineer The inspectors also interviewed other licensee personnel during the course of the inspectio ,

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• Denotes those attending exit interviews during the inspection perio . Licensee Action on Previous Inspection Findings (0 pen) Open Item (409/84009-04(DRP)): Minor Differences Between First Half of 1984 Radioactive Effluent Report and Regulatory Guide 1.2 The La Crosse Boiling Water Reactor Technical Specifications, paragraph 6.9.3.a., requires in part that information concerning liquid and gaseous effluents be submitted in accordance with Appendix B of Regulatory

Guide 1.21 (Revision 1), dated June 1974. Regulatory Guide 1.21 (Revision 1), Appendix B, states che following

Section A summary description should be provided of the methods used for estimating overall errors associated with radioactivity measurement Section Estimates of the total error associated with

! total values should be provided as shown in Table 1 Section Estimates of the total error associated with certain total values should be provided as shown in Table 2 Section Estimates of the total error associated with certain total values should be provided as shown in Table Section Include the bases for the tritium limit in the supplemental report information, Introduction The reporting method includes the use of uniform notation (external floating point form).

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In Inspection Report No. 50-409/84009, the inspector noted that the first half of.the 1984 Radioactive Effluent Report did not include the above items. The licensee agreed to take action to correct the repor During the present inspection period, the inspector noted that the licensee's followup system identified the action on this item as complete. A review of the first half of the 1985 Radioactive Effluent Report shows that a summary description of the methods used for estimating overall errors associated with radioactivity measurement is still not provide This item remains open pending licensee resolution and NRC revie . Operational Safety Verification The inspector observed control room operations, reviewed applicable logs and conducted discussions with control room operators during the period of this report. The inspector verified the operability of selected emergency systems, reviewed tagout records and verified proper return to service of affected components. Tours of the crib building, reactor building and turbine building 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 housekeeping cleanliness conditions and verified implementation of radiation protection controls. The inspector walked down the accessible portions of the shutdown condenser and alternate core spray systems to verify operability. The inspector also witnessed portions of the radioactive waste system controls associated with radwaste shipment and barrelin These reviews and observations were conducted to verify that facility operations were in conformance with the requirements established under technical specifications,10 CFR, and administrative procedure . Monthly Maintenance Observation Station maintenance activities of the following safety related systems and components were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with technical ;

specifications:

j Control Rod Drive No. 2 Alternate Core Spray Pump Diesel 1A Lower Control Rod Drive Mechanisms

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The following items were considered during this review: the limiting

conditions for operation were met while components or systems were

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removed from service; approvals were obtained prior to initiating the work; activities were accomplished using approved procedures and were inspected as applicable; functional testing and/or calibrations were nerformed prior to returning components or systems to service; quality control records were maintained; activities were accomplished by

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qualified personnel; parts and materials used were properly certified; and radiological controls were implemente Work requests were reviewed to determine status of outstanding jobs and and to assure that priority is assigned to safety related equipment maintenance which may affect system performanc Following completion of maintenance on the above listed equipment, the inspector verified that these systems had been returned to service properl . Licensee Event Reports Followup

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Through direct observations, discussions with licensee personnel, and

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review of records, the following event reports 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 (0 pen) LER 85-004: Type C Leakage Test Failures - Retention Tank and Demineralized Water Valves. In a previous inspection report (50-409/85012(DRP)), it was stated that the most recent failure of valve 54-25-006 (the retention tank pump discharge valve) had led

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DPC to reevaluate the need to improve this valve. Accordingly, the event was left open pending the findings of DPC's evaluation. DPC has completed its evaluation of this event and has approved a facility change proposal to replace the suspect valve with a General Twin Seal Valve which can be repaired on-line, if necessary. The

licensee plans to install the-new valve during the forthcoming i refueling outage. Consequently, this item is being held open pending acceptable installation of the valve and its passing the acceptance criteria for Type C leakage testin (0 pen) LER 85-009

Seal Inject System Leak With Scram During Shutdow In a previous inspection report (50-409/85012(DRP)), it was stated that this event was being held open pending the results of DPC's evaluation of the design of the line between the orifice and transmitter. DPC has completed its evaluation of the failed line and plans to replace the stainless steel flexible hoses to the seal inject control rod drive flow orifice during the forthcoming refueling outage. DPC estimates that the dose associated with this task is about 300 mR. This item is being held open pending the

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c. (0 pen) LER 85-017: Reactor Scram Due to Loss of Load Due to Maintenance on 2NB11 and Loss of Offsite Power. The most salient feature of this event, from a regulatory viewpoint, was the loss of offsite power. Said loss of offsite power occurred while an electrical technician was winterizing a 69 kV oil circuit breaker (2NB11) in the main switchyard. Winterizing the breakers is accomplished by adding methyl alcohol to each breaker's air receiver, thereby preventing ice from forming in the air receive Prior to adding the alcohol, the receiver is depressurized by opening a relief cock and then removing a plug on one end of the receiver for the adding of alcoho In this instance, difficulty was experienced in removing.the plug; therefore, the technician used a " persuader" to help remove the plug. While tugging on the persuader, the technician bumped the breaker's manual trip lever with his shoulder, causing the breaker to trip, thereby resulting in the loss of load for LACBWR and, in turn, causing a reactor trip and loss of offsite powe This event was left open in a previous inspection report (50-409/

85018(DRP)), pending the licensee's implementation of acceptable corrective actions to prevent the recurrence of a similar even Corrective actions previously taken by the licensee included removing the manual trip lever on Breaker 2NB11 to prevent a similar event, and serving the air compressor that charges the air receiver from a power source independent from the affected breaker. Since a tripped breaker cannot be reclosed with a depre:surized air receiver, this latter change will reduce the time required to reclose a breaker with a depressurized air receiver. Due to other safety considera-

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, reinstalled and a protective bolt was placed thru the trip lever to prevent the accidental tripping of Breaker 2NB1 In brief, the corrective actions taken to date by the licensee have not been well coordinated nor carefully thought out and do not warrant the closing of this even Since LACBWR has only one source of offsite power, extensive and intensive systematic efforts should be expended to strengthen any weak link in the offsite power syste Said efforts should consider and evaluate not only hardware modifications but also software changes such as administrative controls and procedural changes. The results of such integrated efforts would, among other features, ensure that the breakers adjacent to the LACBWR plant (e.g., Breaker 25NB1) are provided the same corrective actions as those being implemented for the main switchyard breakers. Accordingly, this event is being held open pending the implementation of acceptable corrective actions by the license d. (Closed) LER 85-19: Reactor Partial Scram - Short Due to Leakage From Control Rod Drive Mechanism. A previous inspection report (50-409/85018(DRP)) stated that this event was being held open pending the results of certain licensee considerations and maintenance

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actions to be taken during the forthcoming refueling outage. In the

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interim, the leakage from the affected control rod drive mechanism increased. On January 4th, in an attempt to reduce the seal water leakage, a shift supervisor started tapping the seal water supply line to the affected control rod drive mechanism. While the line was being tapped, a Swagelock fitting on the seal water supply line separated, spraying high pressure seal water (@l400 psig) all over the immediate vicinity and drenching the shift supervisor. The shift supervisor contacted the control room operator and requested that the reactor be brought to a rapid shutdown by the all rod insert method; however, because of an apparent misunderstanding, the shutdown was terminated when reactor power had been reduced to 30%

rated power rather than to zer Later, the Swagelock fitting was replaced, and seal water to the affected control rod drive mechanism -

was restored. Ultimately, on January 5th, about fifteen hours after the Swagelock fitting separated, the plant was shutdown to repair the leaking seal. The maintenance actions eliminated the seal leakage; therefore this event is close e. (Closed) LER 85-20: 1A High Pressure Service Water Diesel Failur At 1053 on December 31, 1985 the 1A High Pressure Service Water (HPSW) diesel driven pump was started following the addition of a

fuel conditioner to the 1A and 1B HPSW diesel fuel tanks. The conditioner was being added to improve the fuel's pour point and cetane number and to reduce the possibility of fuel oil waxing. The HPSW pumps serve both the HPSW system (a fire suppression water system) and the Alternate Core Spray System (a low pressure emergency core cooling system). At 1100, on December 31, 1985, the 1A HPSW diesel driven pump stoppe Based on initial diagnoses of the failure, the fuel oil filters were replace In addition, a check valve which had been the cause of a previous failure was replaced. However, neither of these actions were successful. Therefore, at 1343, the IB HPSW diesel was tested to demonstrate its operability and to confirm that a common mode failure related to the fuel additive did not exist. The IB HPSW diesel performed satisfactorily, thereby demonstrating that a common mode failure mechanism did not exis Further troubleshooting on the 1A HPSW diesel determined that fuel oil was not reaching the. injectors. The fuel injection pump (a Roosa Master pump) was replaced, and a successful surveillance test was completed on the diesel at 1810. (The 1A HPSW diesel was also run on January 1, 2, 7, and 15, 1986, for approximately 30 minutes each time to further demonstrate that the corrective maintenance was successful).

The failed fuel injection pump was removed from the diesel, disassembled, examined and overhauled by a local Roosa Master repair agen Examination of the parts indicated that a polyurethane flexible ring in the fuel injection pump had embrittled and broken with age. The function of the flexible retaining ring is to dampen i

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normal engine operating shock and vibration from the governor. The shock and vibration to which the retaining ring is normally subjected eventually caused the retaining ring to fractur Loose pieces which were caught in the fuel passages of the pump caused the diesel to sto The preventive maintenance program will provide for periodic replacement or overhaul of the 1A HPSW diesel fuel injection pum The fuel injection pump on the IB HPSW diesel is a different make and design of pump and is not subject to this type of failur Based on the actions taken by the licensee, this event is close f. (Closed) LER 86-01: Type C Leakage Test Failure - Reactor Building Main Steam Isolation Valve (RBMSIV) and its Bypass. During a maintenance outage on January 6, 1986, a Type C test was performed on the and RBMSIV and its bypass. The test was required by the Inservice Testing Program because the leakage measured during the spring 1985 refueling outage had reduced the margin between and measured leakage rate and maximum permissible rate by more than 50%.

The leakage measured during the test varied from approximately 40 SCFH to greater than 48 SCFH, the top end of the rotameter scal Technical Specification Section 5.2.1.2.c requires that the combined leakage for all Type B and C tests shall be less than 60% of the Type A allowable test leakage rate. This requirement corresponds to less than 30 SCFH. The combined leakage rate of all other Type B and C tests was approximately 4.65 SCF The 10-inch RBMSIV and its 1-1/2 inch bypass are located in parallel sections of the main steam line and cannot be isolated from each other. Subsequent to the failed test, the bypass valve was disassembled and examined. It appeared that the valve plug may have been resting on one side of the seat more that the other. A new seat, inner valve, seat gasket, pin, stem, bonnet gasket and packing were installed. Afterward, the RBMSIV was cycled and greased, and its seating was adjusted to compensate for normal wea '

On January 9,1986, after the above adjustments and maintenance actions were taken, the measured leakage through the RBMSIV and its bypass was 14 SCFH. The cumulative Type B and C tests leakage rate was 18.65 SCFH, well within the acceptance criteria of less than 30 SCF Based on the results of the January 9,1986 test of the RBMSIV and l

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its bypass valve, this event is close g. (0 pen) LER 86-02: Momentary Loss of Power to 120V Noninterruptible Bus 18. On January 10, 1986, while performing a facility change during a plant cold shutdown, the 120V Noninterruptible Bus 1B was being powered from its reserve feed. Following completion of the l work, an operator was sent to clear the tags on the breakers which

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had been tagged out for the work, and to place the IB Noninterruptible Bu: back on its normal power supply, the IB Static Inverter. However, the operator did not close the breakers from which he removed the tags because he believed the procedure for placing the IB Static Inverter in service would include the closing of the breakers for correct electrical alignment. Therefore, when the reserve feed contactor was opened, Noninterruptible Bus 1B was momentarily de-energized until the reserve feed contactor automatically reclosed to supply Noninterruptible Bus 1B. When the operator reported to the control room and described his actions, he was given instructions for reclosing the open breaker on Noninterruptible Bus IB. The IB Static Inverter was then placed in. service on the Noninterruptible Bus IB without further inciden When Noninterruptible Bus 1B was momentarily de-energized (1420 on January 10,1986) both High Pressure Core Spray (HPCS) Pumps, both Emergency Diesel Generators (EDG) and the 1A High Pressure Service Water / Alternate Core Spray (HPSW/ACS) Diesel Pump started and the containment building isolated. (The 1B HPSW/ACS Pump was running at the time of the event because the internal fire suppression header had been depressurized for unrelated maintenance work and the IB HPSW/

ACS Pump was being used to supply water to the external fire suppression system.) The HPCS pumps and the 1A HPSW/ACS pump were secured after approximately 1 minute, and the EDGs, which are designed to start but not necessarily to load on a low water level signal, were secured within the next minut The ACS Valves may have started to open during the brief period tne Noninterruptible Bus IB was deenergized, but since the internal HPSW header was depressurized, no water would have been injected if they had momentarily opene This event was considered to have been caused by personnel error on the part of the operator and his supervisor (both of whom are licensed Senior Reactor Operators) and procedural inadequacy. Contributing factors included inconsistent nomenclature in the procedure and a continuous audible alarm due to troubleshooting being conducted on the control room annunciator system. Better preparation could have prevented this incident. To prevent recurrence of similar events, the procedure for placing the 1B Static Inverter in service on the Noninterruptible Bus 1B will be revised and the importance of proper performance of duties has been emphasized in operations department trainin Although the training associated with this event has been completed, the procedure has not, as yet, been revised. This event, therefore, is being followed as an open item (50-409/85022-01) pending the revision of the procedur ..

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h. (0 pen) LER 86-03: Reactor Scram - Fuse Blown While Adjusting Seal Inject Flow to Control Rod Drive 2. On' January 12, 1986, while the reactor was in Operating Condition 2 (startup and heatup) the effluent temperature from Control Rod Drive (CRD) No. 2 was hotter than from the other CRDs. Maintenance had been recently performed on CRD No. 2, and upon investigation, it was determined that a valve on the seal inject supply line to CRD No. 2 had been left closed. At 1107, while the closed seal supply valve at CRD No. 2 was being opened, the seal inject tubing was bumped into and shorted the terminal block mounted on the side of the CRD, causing fuses (FU) 29/2 and 55/2 to blo The blown fuses, in turn, caused the reactor to scram, the high pressure core spray (HPCS) pumps to start, containment building to isolate and the emergency diesel generators (EDG) to start. The equipment that is automatically actuated when Fuse 55/2 blows are those that would actuate on a low reactor water level signal, and all such systems responded properl The HPCS pumps were secured for approximately two hours. Between 1108 and 1300, when FU 55/2 was being replaced, the HPCS pumps would not have started on an automatic signal. By Technical Specifications, this placed the plant in Section 3.0.3, requiring hot shutdown within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and cold shutdown within an additional 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />. The plant was in hot shutdown at the tim At 1121, the diesels' shorting plugs were removed to permit securing of the diesels while FU 55/2 was blown. Removing the plugs eliminates the automatic start of the EDGs on low reactor water level but does not affect their ability to start and load automatically on an undervoltage condition. At 1122, the 18 EDG was secured and returned to auto. At 1123, the 1A EDG was secured and returned to aut At about 1300, new fuses were installed and the diesel shorting plugs were replaced, and at about 1302 the containment building isolation signals were rese Information on this incident has been disseminated to all members of the operations department. In addition, the importance of proper performance of duties and of post-maintenance verifications has been emphasized during operator training session Although the actions taken for this specific event warrant its closing, it is being followed as an open item (50-409/85022-02)

pending the licensee's evaluation of the generic aspects of the event, (e.g., the plants susceptibility to partial scrams because of the exposed terminals in the immediate vicinity of the CRDs).

Therefore, this event is being held open pending the licensee's evaluation of corrective measures and the implementation of any actions deemed necessary by the license .

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• (Closed) LER 86-04: Monitored Discharge of Unsampled Waste Water With Analyzed Waste Water Tank. At 1640, on January 13, 1986, while the 3000 gallon waste tank in the turbine building was being discharged to the river, an auxiliary operator noticed that the level in the 4500 gallon waste tank (GWT) had decreased from the 1300 reading of 23% to 14%. Since the turbine building drain system was lined up to the 4500 GWT, its level should have increased rather than decreased. Consequently, the discharge to the river was secure It was later determined that the suction valve from the 4500 GWT to

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IA Waste Water Pump had failed in such a manner that it was open,

regardless of the handwheel position. Since the 1A Waste Water Pump was in use at the time, it was drawing water from both waste water i

tank LACBWR Technical Specifications require that liquid wastes be batch-sampled and analyzed prior to release. Although the contents of the 4500 GWT were not sampled prior to release, the total liquid waste was monitored during the release. Since the indicated activity did not exceed the alarm setpoint, the activity of the water was within allowable limits. Additionally, the contents of the 4500 GWT were sampled following this incident and determined to be releasable at the dilution rate used. Drain system water had entered the 4500 GWT prior to the sample, but since it was from the same sources as would have been present in the tank during the release, the sample was considered representative. Approximately 500 gallons were estimated to have been discharged from the 4500 GW The failed valve, a 2-inch Dow lined diaphragm valve was repaired and returned to service. The failed valve's compressor had separated from the diaphragm and the compressor set pin was sheare The actions taken by the licensee prior to, during and after the

, event were acceptable. Therefore, this event is close . Systematic Evaluation Program Action Items (Closed) Integrated Plant Safety Assessment Report - NUREG-0827 (IPSAR) Reference 4.1 (409/8400S 05): Incorporate Technical Specification Change to Inform NRC of Any Changes in Occupancy of i Privately Owned Land. The licensee submitted the proposed Technical Specification change by letter LAC-9480 dated December 14, 1983. By letter from John A. Zwolinski to Frank Linder dated May 28, 1985, the NRC issued Amendment No. 41 to Provisional Operating License N DPR-45. This amendment includes the requirement to inform the NRC of any changes in occupancy of the exclusion area which lead to

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residential uses. The inspector verified that this amendment was entered in the Technical Specifications. This item, therefore, is considered close i

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t b. (Closed) IPSAR References 4.18, 4.20.1, and 4.20.2 (409/84009-06):

Revise Technical Specification Chloride and pH Limits to Conform With Regulatory Guide 1.56 and Reestablish Conductivity Limits and Sampling Frequency Following Review of System Capability. The 1 license submitted the proposed Technical Specification change by letter LAC-9480. By letter from John A. Zwolinski to Frank Linder *

dated May 28, 1985, the NRC issued Amendment No. 41 to Provisional i Operating License No. DPR-45. This amendment includes the proposed

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changes. Although some of the conductivity limits exceed the Regulatory Guide 1.56 recommendations, the NRC concluded that this is dictated by existing cleanup system capability and the licensee has taken adequate steps to minimize stress corrosion cracking. The inspector verified that this amendment was entered in the Technical Specifications; therefore, this item is close , c. (Closed) IPSAR Reference 4.4 (409/84009-07): Review Possible Loss i of Cooling Water to Plant, and if Needed Develop Procedures to

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Identify Alternate Sources of Water. The licensee submitted a i proposed Technical Specification by letter LAC-9725 dated March 21, 1984. On February 22, 1985, the licensee submitted additional information required to complete NRC's review. By letter from John A. Zwolinski to James W. Taylor dated October 8,1985, the NRC issued Amendment No 44 to Provisional Operating License No. DPR-45. This amendment includes the proposed Technical Specifi-cation change. The inspector verified that the amendment was entered

[ in the Technical Specifications. With regard to determining river water levels, the water level is readily available from the Army Corp

, of Engineers which operate Lock and Dam No. 8 just upstream of the plant. This item is, therefore, close "

d. (0 pen) IPSAR Reference 4.21.3.2 (409/84009-14): Install Remotely Operated Solenoid Valves Outside Containment in Shutdown Condenser Vent to Offgas Line. By letter LAC-10683 dated April 8,1985, the licensee requested to revise the completion date from the February 1985 refueling outage to the 1986 refueling outage. The revision was considered necessary because an upgrading of the entire line and penetration is required. By letter from John A. Zwolinski to Frank Linder dated April 26, 1985, the NRC accepted the revised completion date and requested that the licensee submit the proposed design and

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i related radiation exposure / cost analysis for analysis for NRC j revie i By letter LAC-11095, the licensee submitted the proposed design and

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i analysis to the NRC. The letter also reported that the delivery date for the solenoid valve would not be in time for the 1986 i refueling outage. As a minimum, the licensee intends to upgrade the piping system and install the electrical service for the solenoid valve during the 1986 outage. The licensee intends to install the solenoid valve during a subsequent outage. This item remains open pending licensee installation of the modification and subsequent NRC review.

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l (0 pen) IPSAR Reference 4.26.4 (409/84009-18): Provide Separate Power Supplies for the Full Scram Channels. This item was scheduled for the February 1985 refueling outage. By letter LAC-10491, dated January 3,1985, the licensee requested that one power-to-flow safety channel modification be delayed until the 1986 refueling outage. The reason for the delay was to prevent the existing 1A static inverter from being overloaded. The licensee plans to replace the existing 1A static inverter with a larger unit capable of handling the power-to-flow channel and of providing for future uninterruptable loads during the 1986 refueling outage. By letter from John A. Zwolinski to Frank Linder, dated May 16, 1985, the NRC concluded that the schedule was acceptabl By letter LAC-11243, dated November 14, 1985, the licensee informed the NRC that replacement nuclear instrumentation was being procure The estimated delivery date for the nuclear instrumentation was such that it would not be possible to install it during the 1986 refueling outage. Because of changes .necessary to the power-to-flow channels as a result of the replacement nuclear instruments, the licensee requested that the separation of the power-to-flow power supplies be delayed until proper coordination with the nuclear instrumentation modification can be achieved. Completion of these modifications are expected during the 1987 refueling outage. By letter from John A. Zwolinski to James W. Taylor, dated December 23, 1985, the NRC concluded that the revised schedule was acceptabl This item remains open pending licensee completion and subsequent NRC reviews of the above listed action . TMI Action Items (0 pen) II.B.3.2.B. - Post-Accident Sampling - Modify. By letter LAC-11396, dated February 4, 1986, the licensee informed the NRC

, that the standard test matrix to test the chloride and pH analysis

techniques (409-84009-01) does not apply to the La Crosse Boiling Water Reactor. The licensee also stated that-semi-annual training

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will be held for the health physics personnel who would be involved

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with the collection and handling of a coolant sample following an accident (409/84009-02). The training includes the handling of the sample but not the analysis since these personnel perform all the required analytical procedures in the normal sampling program. This item remains open pending NRC's review and acceptence of the above ,

action . Regulatory Improvement Program (Closed) RIP Task (409/RP-00A-1): Evaluation of Organizational Responsibilities, Management Controls, and Staffing Levels. The

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NRC's Systematic Assessment of Licensee Performance (SALP 5) for the period of July 1,1983 through December 31, 1984, identified

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improvement in 6 of the 9 rated areas. The other 3 areas remained J

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the same with one of these 3 areas having a Category 1 ratin During 1985, no significant decline had been noted. As a result, the inspector determined that the licensee had demonstrated adequate management ability to allow closure of this item. Management will be reviewed during future SALP reviews to ensure this area does not degrade. This item is considered close b. (Closed) RIP Task (409/RP-00C-1): Periodic Review of Outstanding Special Information Tags for Proper Placement, Applicability, and Readability. The inspector reviewed the tag log and hanging tag The log indicates that shift supervisors are reviewing the tags semi-annually as required by procedure. The tags were readable, properly placed and appeared to be applicable. The inspector determined that the semi-annual review is effective and adequat This item is considered close c. (Closed) RIP Task (409/RP-00E-1): Implement Scheduling and Coordination of Maintenance Activities with other Department Activities. Although this area still needs improvement, adequate improvements have been made to ensure timely and safe maintenanc The SALP 5 report identified maintenance as improving, and no significant degradation has been noted in this area. This . item is considered close d. (Closed) RIP Tast (409/RP-00G-3): Improve Scheduling and Coordination of Outages. See item (409/RP-00E-1), above. No significant scheduling and coordination problems are being noted during outages. This item is considered closed, e. (Closed) RIP Task (409/RP-001-1): Develop and Implement Systems

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Training for Mechanical Maintenance Personnel. The training program implementation has been completed. This item is considered close f. (Closed) RIP Task (409/RP-00R-1): Establish an Improved Records Management Control and Commitment Response System. The system is functional and appears to be effective in ensuring commitments are met. This item is considered close g. (Closed) RIP Task (409/RP-00S-1): Improve and Expand the Technical Support to the LACBWR Operation and Engineering Activitie Although offsite company engineering staff is only lightly utilized, this appears to be a result of the offsite expertise needed. Use of contractor engineering expertise is being utilized as necessary and is effective in providing the operations, maintenance, and design expertise required. This item is considered closed, b. (Closed) RIP Task (409/RP-00V-1): Improve Communications Among Members of the Plant Staff. Also see item (409/RP-00A-1) abov Although some communication among the plant staff are still strained, this appears to be caused by personality conflicts and is not adversely affecting the safe operation of the facility. This item is considered close .

. Plant Trips Following the plant trip on January 12, 1986, while the plant was in a heat-up mode of operation, the inspector ascertained the status of the reactor and safety systems by observation on control room indicators and discussions with licensee personnel concerning plant parameters, emergency system status and reactor coolant chemistry. The inspector verified the establishment of proper communications and reviewed the corrective actions taken by the license All systems responded as expected, and the plant was returned to operation on January 13, 198 . Preparation for Refueling The inspector verified that technically adequate procedures were approved for the new control rods to be used in the forthcoming refueling outag The inspector verified that the licensee had submitted a proposed technical specification change to NRR for the new control rods and that the licensee's 10 CFR 50.59 safety evaluation of the reload core showed that prior NRR review is not require . Organization and Administration The inspector verified that changes in the organizational structure and assignments had been reported to the NRC through the licensee's QA program and verified that persons assigned to new or different positions in the licensee's organization since the last inspection of this area satisfy qualifications identified in the technical specifications, the licensee's QA program, and applicable national standard . Onsite Review Committee The inspector attended onsite review meetings conducted during the period of this report and examined onsite review functions to verify conformance with technical specifications and other regulatory requirements. This review included: changes since the previous inspection in the charter and/or administrative procedure governing review group activities; review group membership and qualifications; review group meeting frequency and quorum; and, activities reviewed including proposed technical specification changes, violations and corrective actions,-proposed facility and procedure changes and proposed tests and experiments conducted per 10 CFR 50.59, and others required by technical specification . 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 bot New open items are described in paragraph 5 g. and 5 .

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14. Exit Interview The inspector met with licensee representatives (denoted in paragraph 1)

throughout the inspection period and at the conclusion of the inspection i and summarized the scope and findings of the inspection activities. The licensee representative acknowledged tre findings as. reported herein and did not identify such documents or processes as proprietar .

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Appendix A NUS Corporation's Responses to U.S. NRC Inspection Report, Docket No. 99900516/85-01 1. Notice of Nonconforn:ance; U.S. NRC Reference - Appendix A, Item A:

Criterion 111 of Appendix B to 10 CFR PAR 50 states, in part, " Design changes, including field changes shall be subject to design control measures commensurate with those applied to the original design and be approved by the organization that performed the original design unless the applicant designates another respon-sible organization."

NRC Finding:

Contrary to the above, design control document ASD-5492-3 was changed in the field by Tracer Technology personnel without the benefit of proper NUS revie NUS Response Corrective Action Cognizant NUS/ESD staff reviewed and approved the subcontractor field change as Revision I to Design Control Document ASD-5492-3, in accord-l ance with the NUS/ESD Quality Assurance Manual. Revision I to the l

Design Control Document was reissued as a controlled document on Decem-ber 12,198 . Measures to Preclude Recurrence Responsible NUS/ESD personnel and the subcontractor were readvised of pertinent program requirements in the event future field activities warrant modifications to NUS/ESD Design Control Document . Effective Date of Corrective Action Completion

. The stated corrective action and preventative measures were completed on December 11,198 . Notice of Nonconformance; U.S. NRC Reference, Appendix A, Item B:

Criterion V of Appendix B to 10 CFR Part 50 states, " Activities affecting quality shall be prescribed by documented instructions, procedures, or drawings of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings. Instructions, procedures, or drawings shall include appropriate quantitative or qualitative acceptance criteria for determining that important activities have been satisfactorily accomplished."

Chapter 5 of the NUS Corporate QA Policy manual requires a "Semlannual report to the Corporate Executives for Division Operations on the status of Operating Unit programs and procedures and the effectiveness thereof. These reports will include summaries of audit findings and associated corrective action, an analysis of any significant unsatisfactory trends with recommended resolutions,

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and a' review of the extent and effectiveness of program monitoring by Operating Unit Quality Assurance Program Managers or Representatives."

l i NRC Finding:

I

[ Contrary to the abo've, only a single QA' status report for Operating Divisions;

was prepared in 1982 and 1984, none was prepared in 1983, and the single 1985 *

l report was in the process of being prepared at the time of the inspection.

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l NUS Response:

I Due to the classification of this finding as a nonconformance to the requirements of 10CFR50, Appendix B, it is considered appropriate to provide a clear explana -

tion of the NUS Corporate QA Policy and the functional responsibilities of the

,

Corporate QA organization, j The nonconformance is apparently based on the assumption that NUS Corporate i

. QA policy and the Corporate QA staff function provide direct control over safety related activities which are governed by federal regulatory requirement '

This is not a correct assumption. The NUS Corporate, QA Policy is an internal- ,

i administrative document which provides minimum requirements to all NUS ,

Operating Units for the control of quality for all work undertaken by the Corpora-

,

tion, both safety related and non-safety related alike. With regard to safety j related work, the NUS Corporate QA policy delegates all responsibility and j authority to the Operating Units. The Operating Units are individually responsible !

to establish QA programs which respond to the applicable regulatory requirements *

] and to implement, monitor and enforce those programs independent of the Cor-4 porate QA function. The NUS Corporate QA staff provide an administrative

]

overview, through audits of all operating units for NUS Corporate Administration, ;

i- strictly as a management tool to monitor the implementation of Corporate l Policy. This overview is not a safety related function, Corporate QA records

<

are not classified as safety related records, and there are no line responsibilities

{ delegated to the Corporate QA staff for safety related wor :

l NUS Operating Units with established QA programs which respond to regulatory j- requirements for safety related work, each have their own QA manager or repre-j sentative. These QA staff report directly to the management of the Operating

Unit, and in accordance with their individual safety related QA programs, have

) the direct line responsibility for evaluating and reporting program effectiveness

! and implementation adequacy through their own audit and surveillance procedure : In the event that Corporate QA staff are called on to support Operating Unit i f projects or activities which are classified as safety related, Corporate QA staff j operate as an extension of the Operating Unit QA organization and in accordance j with Operating Unit QA programs. In those instances, Corporate QA staff report i directly to the Operating Unit management. Other than interface and informa-

{ tional arrangements for annual Corporate QA administrative audits, there are j no reporting lines from Operating Unit QA representatives to the Corporate QA organization.

, It is NUS' position, therefore, that the -NRC nonconformance which was issued ,

j against -the requirements of the NUS Corporate Policy is invalid on the basis '

t of its nonapplicability to safety related functions or activities.

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To preclude future misunderstandings, the NUS Corporate QA Policy manual revisions, presently scheduled for the second quarter of 1986, will more clearly identify the purpose of the Corporate policy manual as an administrative docu-ment and the functions of the Corporate QA staff as an internal management overview activity with no line responsibilities for safety related wor We apologize for not more clearly explaining the foregoing points to the NRC inspectors at the time of their visit, and for any resulting inconvenienc III. Other Findings and Comments; U.S. NRC Reference, Appendix A, Item C:

Failure to Properly Pass Down Part 21 Requirements On July 13, 1984, NUS signed a contract with Control Data for Control Data to provide information concerning errors discovered in safety related computer programs supplied for use by NUS. The term of this contract was for six months and has since expired. Control Data is therefore not contractually obligated to supply NUS information concerning errors discovered in computer programs that may have been used in analyzing safety related systems. NUS acknowledged this discrepancy and indicated it would rewrite their contract with Control Data. Documentation was provided showing that although not contractually required, error reports were still being received from Control Dat NUS Response: Corrective Action As noted in the NRC findings, NUS provided the NRC Inspectors with documentation at the time of the inspection, demonstrating that service bureau error reporting provisions were still in effect irrespective of the expiration of financial arrangement The referenced documentation is shown as Exhibit A-1 of this respons . Measures to Preclude Recurrence The NUS contract with Control Data was rewritten and sent to Control Data as Amendment Five to CDC Contract No. EKD315. The modified contract is shown as Exhibit A-2 of this respons . Effective Date of Corrective Action Completio Written confirmation of the continuation of error reporting by CDC was provided on November 19,198 The NUS/CDC contract was rewritten and approved by NUS on December 2,198 !

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. EXilll31T A-1

@@GONTROL DATA 8100 34th Avenuo South Maihng Address Box 0 Minneapoks. Minnesota $5440 November 19, 1985 Mr. Ray Sacramo NUS Corporation 910 Clopper Road Gaithersburg, MN 20878 Control Data considers that the Quality Assurance Agreement

between NUS Corporation and Control Data Corporation dated March 15, 1984 and accepted on July 13, 1984 by Control Data, is still in effect. TI,e six month term was originally intended to allow for price changes to occur as our program evolve ,

Please disregard the six month interim stipulation. The terms and conditions are valid and not restricted to a specific time fram

Sincerely, CONTROL DATA CORPORATION O ,

Y , ~4 K Blaine H. Patrick, Manager CYBERNET Quality Assurance cc: Fraser CDC/ ROC 215 t

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@h IIXillBIT A-2

'. cot 4TRfX. DATA .

CORf0 RATION CONTROL DATA CORPORATION .

8100 34th Asenue South P O Ilos 0 Niinneapohs. Niinnesota $5440 AMENDMENT N Five TO CDC CONTRACT N EKD315 CUSIONtER NAhtE NUS CORPORATION STREET ADDRESS 910 CLOPPER ROAD CITY STAIE ZIP GAITHERSBURG MARYLAND 20878 Customer and Control Data agree that the abose referenced Agreement is amended as follows:

This amendment replaces and supercedes Amendment #2, and incorporates the attached Quality Assurance statemen Escept as prosided abose, all terms and conditions of the abose referenced AF reement will remain in full force and in effec AG R E ED 1 ACCEPIED HY:

NUS CORPORATION CONIROI. DATA CORPORAllON

[i , f -'

BT(A 4 (i d, na' re HY ( Authorized Signature)

• I% _

N A M E ( l >'pe or'Pri[tl ' N A%1E (I3 pe or Print)

P.D. Artovsmith il T LE I Type or Pnnt) ll TEE Ii>pe or Pnnt Vice President DAIE DAIL December 2, 1985 CDC Contract No AAT387 3,82 COC PRINTED IN U S A

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QUALITY ASSURANCE: 'The computer programs listed below and

designated as " Safety-Related Applications" are. currently

, available as part of Control Data's CYBERNET Services.

! Control Data understands that customer may use one or more of

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the projects, and that customer may be required to comply with u

various laws and reg'lations regarding quality assurance,

records retention and disclosure of errors and defects. To j assist Customer in complying with these laws and regulations, i Control Data has implemented a Quality Assurance program which j it believes is consistent with the pertinent provisions of

! " Quality Assurance Criteria for Nuclear Power Plants and Fuel Reprocessing Plants," 10CRF50 Appendix B, and " Quality Assurance Program Requirement for Nuclear Facilitics," ANSI standard

! N45.2. A copy of Control Data's CYBERNET Application Quality Assurance Program Policies and Procedures, as revised from time i

to time, is available for examination at Customer.'s request on a confidential basis.

i j In addition, in accordance with the requirements of " Reporting i of Defects and Noncompliance," 10CFR21, Control Data-will

!

provide Customer during the term of this Agreement with I information concerning errors discovered in the Safety Related

!

Applications programs listed below. Customer will be respon-sible for determining whether a particular error constitutes

. a defect (as defined in 10CFR21) as a result of Customer's j use of the progra SAFETY-RELATED APPLICATIONS:

ADLPIPE ANSYS APADS BASEPLATE-II DIS

! GTSTRUDL ,

NUPIPE

.PIPESD j STARDYNE UNIPLOT Send 10 CFR Reporting Correspondence to:

NUS Corporation

910 Clopper Road l Gaithersburg, MD 20878 ATTN: Stephen Smith

!l 301/258-6078 i

!

$

$

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