ML20207C952

Rev 1 to Monthly Operating Rept for Mar 1986

ML20207C952
Person / Time
Site:	Davis Besse
Issue date:	03/31/1986
From:	Storz L TOLEDO EDISON CO.
To:	Haller N NRC OFFICE OF RESOURCE MANAGEMENT (ORM)
References
KB86-0510, KB86-510, NUDOCS 8607210461
Download: ML20207C952 (10)

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. . s Rev. 1 - 5/26/86

, OPERATIONAL

SUMMARY

MARCH, 1986 i

The unit remained shutdown the entire month of March following the reactor trip on June 9, 1985. Investigation of the causes of the event and corrective actions continues. See NUREG 1154 for further details.  !

Below are some of the major activities performed during this month:

1) Continued testing as part of the System Review and Test Program.

2) Continued MOVATS testing.

t 3) Completed rewiring of SFAS Channel 4.

4) The ultrasonic testing for all four Reactor Coolant Pump (RCP) shafts revealed significant areas of interest. This inspection was conducted as a result of the Crystal River Power Plant inspection of the RCPs which revealed a failed shaft.

All four RCP seals have been removed, and the RCP 2-1 motor and impeller also have been removed from the pump.

Plans are to replace RCP 2-1 shaft with a spare rotating assembly and conduct hot functional tests. However, the other pump shafts will be replaced after the hot functional tests.

The RCP shaft replacements will extend the outage longer than

expected, k .

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, Rsv. 1 - 5/16/86 COMPLETED FACILITY ^ CHANGE' REQUEST

i FCR No.: 77-433 SYSTEM: RC/SP COMPONENT: Instrumentation CHANGED. TEST OR EXPERIMENT: This FCR improved instrument transmitter &

RTD's inside the containment vessel to promote hermeticity, by replacement and installation of conax wire leads and seals. This FCR also promoted the hermeticity of wire pigtail terminations inside existing Crouse Hinds outlet boxes.

Work was completed May 13, 1985. ,

REASON FOR CHANGE: To promote hermeticity of seals and wire terminations and instrument logic to withstand C.V. design environment.

SAIETY EVALUATION

SUMMARY

The proposed design change promoted and improved the hermeticity of instrumentation inside the containment vessel. This change will not adversely affect the operation of the instrumentation or plant, nor the function of the safety protection system.

The proposed change does not involve a technical specification, nor an unreviewed safety question.

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, Rev. 1 - 5/16/86 COMPLETED FACILITY ~ CHANGE REQUEST FCR NO.: 79-400 SYSTEM: Reactor Coolant System COMPONENT: TE-RC3A3, TE-RC3Al and TE-RC3B1 Inconel Mounting Bosses CHANGE, TEST OR EXPERIMENT: This FCR replaced the resistance temperature detector mounting bosses, for TE-RC3A3, TE-RC3Al and TE-RC3B1 in accordance with B&W field change (FC-138).

This work was completed May 13, 1985.

REASON FOR CHANGE: The change was necessary due to damage on the Inconel Mounting Bosses. Loop #1 had developed galled threads on TE-RC3Bl. Loop #2 (TE-RC3A3 and TE-RC3A1) bosses were bent and also had galled threads.

SAFETY EVALUATION

SUMMARY

The repairs meet the requirements of the applicable specifications for the reactor coolant piping. Field changes 138 and'139 include a verified statement that an appropriate analysis was made and the repairs would be acceptable. Included in field changes 138 and 139 are the required changes to the equipment specification for reactor coolant piping to allow the reapirs, with applicable codes and standards. Original design of this piping was 1968 draft of ANSI B31-7.

The repaired bosses have the same dimensions as the original bosses.

This is not an unreviewed safety question.

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Rev. 1 - 5/16/86 COMPLETED' FACILITY' CHANGE ~ REQUEST FCR No.: 82-018 SYSTEM: Containment Vessel COMPONENT: Technical Specifications CHANGE, TEST OR EXPERIMENT: This FCR revised tables 3.3.10-and 4.3.10 of the Technical Specification.

This work was completed March 4, 1986.

REASON FOR CHANGE: The revision to Technical Specification Tables 3.3.10 and 4.3.10 was required per NRC requirement NUREG 0737 Item IIF-1-4 containment wide range pressure (FCR-79-425), IIF-1-5 containment normal sump level (FCR-79-408), and containment wide range level (FCR-79-409).

SAFETY EVALUATION

SUMMARY

The safety function of the containment pressure normal sump level and wide range level is to monitor the containment pressure and containment level and to inform the operator of post accident condition.

Technical Specification Tables 3.3.10 for post accident monitoring instrumen-tation and Technical Specification Surveillance 4.3.10 requirements are adequate to verify that these systems are available during post accident conditions, and the operability is maintained in the applicable mode. It is concluded .that the proposed Technical Specification changes do not involve an unreviewed safety question.

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Rev. 1 - 5/16/86' i

COMPLETED' FACILITY ~ CHANGE REQUEST 1

3 FCR NO: 82-127 i

) SYSTEM: 480V Essential MCC i

j COMPONENT: Breakers BF 1223 and BF 1217 i

CHANGE. TEST OR EXPERIMENT: This FCR will replace existing 200 amp trip

units on essential heaters BF 1223 and BF 1217 with 250 amp trip units.

i i This work was completed January 14, 1983.

! REASON FOR CHANGE: Breakers were causing spurious losses of pressurizer heaters. Current through these breakers is 160 amps, which is normal for a full heater bank. FCR 78-430 replaced the trip units on the non-essential breakers with 250 amp units. These breakers supply 126 KV loads also.  ;

SAFETY EVALUATION

SUMMARY

This FCR will not compromise the integrity of the existing pressurizer essential heater banks or their power supplies.

All modifications are internal to the breaker units and will not prevent

{ the safe shutdown of the plant. Increasing the current rating of the j thermal magnetic trip units from 200 amps to 250 amps will improve the

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reliability of the power supply to the pressurizer essential heater banks j by decreasing the susceptibility of these two breakers'to current surges,

! which had been resulting in loss of pressurizer heaters. I

! Therefore, the work authorized by this FCR does not create any new adverse j environment and does not constitute an unreviewed safety question.

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Rev. 1 - 5/16/86 COMPLETED FACILITY CHANGE REQUEST FCR No: 83-067 SYSTEM: Various COMPONENTS: MVO-5990, MVO-06010, MVO-6080, MV0-6120, MV-15170, MV-15180, MVRC-020 CHANGES, TEST OR EXPERDfENT: This FCR replaced the motor and brake with identical motor and qualified brakes, complete with special radiation resistant coils for the following motor operated valves:

MV-15170 MV-5990 4

MV-15180 MV-6080 MV-06010 MV-020 MV-6120 This work was completed February 19, 1986.

REASON FOR CHANGE: NRC IE Bulletin 79-01B required a review of environmental qualification of Class IE Electrical Equipment. The above equipment must be replaced to meet the requirements of NRC IE Bulletin 79-01B.

SAFETY EVALUATION

SUMMARY

The above equipment is being replaced because it does not meet environmental qualification requirements.

Since this FCR calls for replacing the motor with identical motor and brake and coil with qualified brake and coil, the operators on these valves will still perform their intended safety function. This replacement will increase reliability of these actuators during accident conditions. The work authorized by this FCR does-not create any new adverse environment and does not constitute an unreviewed safety question.

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Rev. 1 - 5/16/86 COMPLETED ~ FACILITY' CHANGE REQUEST FCR No.: 83-075 SYSTEM: Decay Heat COMPONENT: DH-13A

CHANGE, TEST OR EXPERIMENT

This FCR repaired and replaced valve'DH-13A to its original condition. (Refer to FCR 82-130 for present pinned condition).

This work was completed February 21, 1986.

REASON FOR CHANGE: This valve (DH-13A) had been placed in a pinned position.

This FCR will repair and replace valve DH-13A to its original condition so that it can be operated.

SAFETY EVALUATION

SUMMARY

The safety function of valve DH-13A is to maintain the pressure boundary of the decay heat system and to fail close upon SFAS actuation to prevent 'ypass of the decay heat cooler.

This FCR will restore valve DH-13A to its original design condition, which will enable it to perform its safety function.

i This FCR based on the above analysis does not constitute an unreviewed safety question.

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Rev. 1 - 5/16/86 COMPLETED' FACILITY CHANGE' REQUEST FCR NO.: 84-160 SYSTEM: Main Feedwater COMPONENT: FW-601 and FW-612 4

CHANGE, TEST OR EXPERIMENT: FCR 84-160 changed the open torque switch settings for the Main Feedwater Isolation Valves, FW-601 and FW-612, from .

1.5 to 1.0. Work was completed Januarv 9, 1985.

REASON FOR CHANGES: The new torque switch settings are to improve valve reliability. The basis for the new settings is by recommendation from the Torrey Pines Technology Report on Limitorque motor-operated valves.

SAFETY EVALUATION

SUMMARY

The safety function of the torque switch settings is to close the valve tight enough to prevent any leakage and to break the circuit in case of high mechanical force to prevent overtraveling of the

valve stem. The new torque switch settings do not affect the safety function of the torque switch. Therefore, an unreviewed safety question does not exist.

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Rev. 1 - 5/16/86 COMPLETED FACILITY CHANGE REQUEST FCR No.: 84-201 SYSTEM: HPI/MU Systems COMPONENT: HPI and MU valves (HP-2A, B, C, & D; HP-22 & 23; MU-169, 196 and 197)

CHANGE, TEST OR EXPERIMENT: This FCR leak tested the HPI and MU valves using both liquid measuring devices (LMD) at pipe drains or vents, and acoustic emission (AE) inspection equipment for locating any leaks.

This work was completed February 21, 1986.

REASON FOR CHANGE: This information will be used as I.S.I. documentation for ASME Section XI alternative inspection requirements.

SAFETY EVALUATION

SUMMARY

The change to both the HPI and MU systems is a temporary test committed to the NRC that will be conducted by providing a measurable leakage path through drain valves in various locations during Modes 5 and 6. As long as a flowpath from the BWST via a decay heat removal pump to RCS is available, the Makeup System does not need to be operable.

The test pressure of 1900 psig will not exceed either the preservice hydro of 2500 psig, nor the required mill test of 2500 psig for each length of pipe per ASTM-1530.

The conduct of this test will not affect the~ integrity of the system tested; therefore, the testing proposed does not constitute an unreviewed safety '

question.

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TOLEDO

%s EDISON May 16, 1986 Log No. KB86-0510 File: RR 2 (P-6-86-03)

Docket No. 50-346 License No. NPF-3 Mr. Norman Haller, Director Office of Management and Program Analysis U. S. Nuclear Regulatory Commission Washington, D.C. 20555

Dear Mr. Haller:

Enclosed are ten copies of Revision 1 to the Monthly Operating Report for Davis-Besse Nuclear Power Station Unit 1 for the month of March 1986.

Please remove and destroy the Operational Summary and the Completed FCR Summaries for FCRs77-433, 79-400,82-018, 82-127,83-067, 83-075,84-160, and 84-201, and replace with the attached revised sheets.

If you have any questions, please feel free to contact Morteza Khazrai at (419) 249-5000, Extension 7290.

Yours truly, W

p> WeO 7 ,! i Louis F. Storz Plant Manager

• Davis-Besse Nuclear Power Station LFS/MK/lj k Enclosures cc: Mr. James G. Keppler, w/l Regional Administrator, Region III Mr. James M. Taylor, Director, w/2 Office of Inspection and Enforcement Mr. Paul Byron, w/1 i NRC Resident Inspector CiiAlit2A:!-:iLC D JK/002 M

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THE TOLEDO EDISON COMPANY EDISON PLAZA 300 MADISON AVENUE TOLEDO, OHIO 43652