ML20207N439

10CFR50.59 Annual Rept 1988

ML20207N439
Person / Time
Site:	Braidwood
Issue date:	12/31/1988
From:	Hunsader S COMMONWEALTH EDISON CO.
To:	Murley T NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM), Office of Nuclear Reactor Regulation
References
5229K, NUDOCS 8810190149
Download: ML20207N439 (9)

Text

. .

[D Commonwealth Edison

, ) One First Nabonal Ptata. Chicago, Ilknois C 7 Addrcss R: ply to: Post Offics Bog 767 Chicago. tilinois 60690 October 12, 1988 Mr. Thomas E. Murley, Director Office of Nuclear Reactor Rer,ulation U.S. Nuclear Regulatory Commission Washington, DC. 20555 Attn: Document Control Desk

Subject:

Braidwood Station Units 1 and 2 10 CFR 50.59 Annual Report NRC_Dacket Nos. 50-456 and 50-457

Dear Mr. Murley:

Pursuant to 10 CFR 50.59(b)(2), Commonwealth Edison Edicon is providing the required annual report for Braidwcod Station (Facility Operating License Nos. NPF-72 and NPF-77). The annual requirement is based on the Unit i fuel load license (NPF-59) issuance date of October 17, 1986.

The report consists of descriptions and safety evaluations for changes to the facility as described in the safety analysis report. The report only contains changes made to the facility; no tests or experiments governed by paragraph (a) of 10 CFR 50.59 were executed.

Please direct any questions regarding this matter to this Very truly yours,

/J c. L +

S. C. Hunsader Nuclear Licensing Administrator

/kij att.

cc: S. Sands (NRR)

J. Hinds (RIII)

Braidwood Resident Inspector 5229K j i

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b Braidwood Nuclear "ower Station 10CFR50.59 Annual Report 1988 NRC Docket No. 50-456 and 50-457 License No. NPF-72 and NPF-77 l

2314m(092988)/1

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e Modification: M20-0-87-003

Description:

This modification provides a process radiation monitor (OPR14J) to monitor the radiation level in the condensate polisher Hi" and "Lo" conductivity sump.

This modification isolates the sump by tripping pumps OCPO4P and OCP05P, initiates a shutdown of the condensate polishing system, and provides an alarm on the condensate polisher control panel in the event of high radiation or monitor failure.

Safety Evaluation Summary:

1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because this modification provides an additional process radiation monitoring point for affluent -

discharge.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because this modification provides an additional process radiation monitoring point for effluent discharge.

3. The margin of salsty, as defined in the basis for any Technical Specification, is not reduced because radiation monitoring and control of effluent is improved by this modification.

2314m(092985)/2

Modification: M20-0-87-021

Description:

This modification installs an additional isolation valve (0FP791) in the Lake Screenhouse ring header so that the ring header will not have to be out of service if maintenance is required on the existing valve. No other piping or support changes are needed.

Safety Evaluation Summary:

1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because the additional valve increases the availability of the fire protection system.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because except for additional isolation capabilities, the function of the fire protection system remains unchanged.

3. The margin of safety, as defined in the basis for any Technical Specification, is not reduced because the additional isolation valve increases the availability and the margin of safety of the fire protection system.

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2314m(092988)/3 i .

Modification: H20-0-87-022

Description:

This modification adds an isolation valve (0FP898, 14 in., gate valve) in series with valva 0FP579 to improve the maintainability of the fire protection system.

Safety Evaluation Sumary:

1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because the additional isolation valve increases the availability of the fire protection system.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because the function of the fire protection system is not changed. Valve OTP898 is added to provide additional isolation to the fire protsetion system for maintenance.

3. The margin of safety, as defined in the basis for any Technical Specification. is not reduced because tha fire protection system availability is increased by this modification.

1314m(092988)/4 4

Modification: M20-0-87-030

Description:

This modification adds valves 0FP901 and 0FP902 to replace the functions of valves 0FP515A and 0FP515B. This modification relocates the isolation valves for the manual deluge of the charcoal filters, outside the filtse room.

Safety Evaluation Summary:

1

1. The probability of an occurrence or the consequence of an accident, or malfunction mf equipment important to safety as previously evaluated in the Final Safety (salysis Report is not increased because the function si the fire protection system is not changed by the additional valves and the relocation.

2. The possibility for an accident or malfunction of a differenN type than any previously evaluated in the Final Safety Analysis Report is not :reated because the function at the fire protection system is not changed by th- additional valves and the relocation.

3. The margin of safety, as defined in the basis fc* any Technical Specification, is not reduced because the modification only provides relocated additional valves.

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Modification: M20-0-87-037 i

! Descriptions s

Thas modification revises the internal wiring of area radiation monitors ORT-AR056, ORT-AR055, 1RT-AR011, and 1RT-AR012 so that actuation occurs from "High Radiation" and from "instrument failure" signals, j

Safety Evaluation Summary:

, 1. The probability of an occurrence or the consequence of an accident, or

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malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not increased because the internal wiring change increases the reliability of the area radiation monitors.

2. The possibility for an accident or malfunction of a different type than any 4 previously evaluated in the Final Safety Analysis Report is not created because the reliability of the area radiation monitors is increased and the function of the system is not changed.

3. The margin of safety, as defined in the basis for any Technical d Specification, is not reduced beca.se the operation of the system is not i changed.

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j 2314m(092988)/6 i

4 Modification: M20-1-87-053

Description:

This modification provides a level indication for reactor cavity local level during refueling by painting a scale on the refueling cavity wall.

Safety Evaluation Su. wary:

1. The probability of an occurrence or the consequence of an accident, or malfunction of equipment important; to safety as previously evaluated in the Final Safety Analysis Report is not increased because the scale allows local verification of refueling cavity level. It has no effect on system operation as described in FSAR Chapter 15.

2. The possibility for an accident or malfanction of a different type than any previously evaluated in the Final Safety Analysis Report is not created because the scale does not affect plant operation in any way. It has no affect on any PSAR Chapter 15 safety analysis.

3. The margin of safety, as defined in the basis for any Technical Specification, is not reduced because the scale will aid in compliance with Tech Spec 3.9.10 which states that a minimum of 23 feet of water will be maintained above the reactor vessel flange.

2314m(092988)/7 l

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Modification: M20-1-87-092

Description:

This modification involves rewiring the reactor protection circuitry to change the reactor trip on turbine trip permissive from P7 to PS. Previously, a turbine trip caused the reactor to trip if reactor power was above 10% (P7).

The modification results in a turbine trip causing a reactor trip only if reactor power is above 30% (pS). The modification also revises steam dump control actuation from turbine trip to reactor trip, so that on turbine tt'ip below PS steam dumps are controlled by the load reject controller. The modification eliminates reactor trips which occur between 10% and 30% power yielding savings in tine by avoiding reactor restart. The elimination of these unneeded coactor trips will enhance plant performinco ratings.

Safety Evaluation Summary:

1. The probability of an occurrence or the consequence of an accident, or malfunction of equiprner.t important to safety as previously evaluated in the Final Safety Analysis Report is not increased because turbine trip initiation of reactor trip is not credited in the safety analysis.

2. The possibility for an accident or malfunction of a different type than any previously evaluated in the Final Safety Anal' isis Report is not created beca'se safety systems remain unchanged.

3. The margin of safety, as defined in the basis for any Technical Specification, is not reduced because safety systems remain unchanged.

2314m(092988)/S l