Text

Monthly Operating Repts for Mar 1986
	ML20141E426
Person / Time
Site:	Quad Cities
Issue date:	03/31/1986
From:	Kronich C, Tamlyn T; COMMONWEALTH EDISON CO.
To:	; NRC OFFICE OF ADMINISTRATION (ADM), NRC OFFICE OF INSPECTION & ENFORCEMENT (IE)
References
	TKT-86-30, NUDOCS 8604210439
	Download: ML20141E426 (42)
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i QUAD-CITIES NUCLEAR POWER STATION UNITS 1 AND 2 MONTHLY PERFORMANCE REPORT MARCH 1986 COMMONWEALTH EDISON COMPANY AND IONA-ILLIN0IS GAS & ELECTRIC COMPANY NRC~ DOCKET NOS. 50-254 AND 50-265 LICENSE NOS. DPR-29 AND DPR-30 L

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o TABLE OF CONTENTS I.

Introduction II.

Summary of Operating Experience A.

Unit One B.

Unit Two III.

Plant or Procedure Changes, Tests. Experiments, and Safety Related Maintenance A.

Amendments to Facility. License or Technical Specifications B.

Facility or Procedure Changes Requiring NRC Approval C.

Tests and Experiments Requiring NRC Approval D.

Corrective Maintenance of Safety Related Equipment IV.

Licensee Event Reports V.

Data Tabulations A. -Operating Data Report B.

Average Daily Unit Power Level C.

Unit Shutdowns and Power Reductions VI.

Unique Reporting Requirements i

A.

Main Steam Relief Valve Operations B.

Control Rod Drive Scram Timing Data

-VII.

Refueling Information

- VIII.

Glossary t

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INTRODUCTION l

Quad-Cities Nuclear Power Station is composed of two Boiling Water i

Reactors, each with a Maximum Dependable Capacity of 769 MWe Net, located in Cordova, Illinois.

The Station.ls jointly owned by Commonwealth Edison Company and Iowa-Illinois Gas'& Electric Company. The Nuclear Steam Supply Systems are Generel Electric Company Bolling Water Re ctors.

The Architect / Engineer'was Sargent &.Lundy, Incorporated, and the primary

. construction contractor was United Engineers & Constructors. The Mississippi River is the condenser cooling water source. The plant is subject to Itcense numbers OPR-29 and DPR-30, issued October 1, 1971, and March' 21, 1972, respectively; pursuant to Docket Numbers 50-254 and 50-265. The date of initial Reactor criticalities for Units One and Two, respectively were October t

18, 1971, and April 26, 1972.

Commercial generation of power began on February 18, 1973 for Unit One and March 10, 1973 for Unit Two.

This report was compiled by Becky Brown and Carol Kronich, telephone number 309-654-2241, extensions 2240 and 2157.

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r II.

SUMMARY

OF OPERATING EXPERIENCE A.

UNIT ONE March 1-31 Unit One remains shutdown for the End of Cycle Eight Refueling and Maintenance Outage for the entire month of March.

B.

UNIT TWO March 1-18 Unit Two began the month of March shutdown for a short scheduled Maintenance Outage and Battery Discharge Tests.

On March 2, at 0855 hours0.0099 days 0.238 hours 0.00141 weeks 3.253275e-4 months , a normal unit startup commenced. The unit was critical at 1100 hours0.0127 days 0.306 hours 0.00182 weeks 4.1855e-4 months and was placed on line at 50 MWe at 2225 hours0.0258 days 0.618 hours 0.00368 weeks 8.466125e-4 months . On March 3, at 0205 hours0.00237 days 0.0569 hours 3.38955e-4 weeks 7.80025e-5 months , load was held at 205 MWe for Turbine surveillances.

At 0310 hours0.00359 days 0.0861 hours 5.125661e-4 weeks 1.17955e-4 months HPCI was declared inoperable. A GSEP Unusual Event was declared at 0430 hours0.00498 days 0.119 hours 7.109788e-4 weeks 1.63615e-4 months . The unit was taken off line at 0800 hours0.00926 days 0.222 hours 0.00132 weeks 3.044e-4 months , and at 0810 hours0.00938 days 0.225 hours 0.00134 weeks 3.08205e-4 months the unit was m:nually scrammed. At 2242 hours0.0259 days 0.623 hours 0.00371 weeks 8.53081e-4 months a normal unit startup commenced.

On March 4, at 0126 hours0.00146 days 0.035 hours 2.083333e-4 weeks 4.7943e-5 months , the unit was critical, and at 1145 hours0.0133 days 0.318 hours 0.00189 weeks 4.356725e-4 months the unit was placed on line.

Full power was reached at 1800 hours0.0208 days 0.5 hours 0.00298 weeks 6.849e-4 months . At 2200 hours0.0255 days 0.611 hours 0.00364 weeks 8.371e-4 months load dropped to approximately 700 MWe due to a xenon trans-ient. On March 5, at 0410 hours0.00475 days 0.114 hours 6.779101e-4 weeks 1.56005e-4 months , load began an increase to full power.

Eight hundred twenty-five MWe was reached at 0615 hours0.00712 days 0.171 hours 0.00102 weeks 2.340075e-4 months and held steady until March 11, when load was dropped below 800 MWe, and placed on Economic Generation Control (EGC) at 1347 hours0.0156 days 0.374 hours 0.00223 weeks 5.125335e-4 months .

At 1545 hours0.0179 days 0.429 hours 0.00255 weeks 5.878725e-4 months the unit was taken off of ECC and load was increased to full power.

On March 13, at 0230 hours0.00266 days 0.0639 hours 3.80291e-4 weeks 8.7515e-5 months , load was dropped below 800 MWe for surveillances. At 0345 hours0.00399 days 0.0958 hours 5.704365e-4 weeks 1.312725e-4 months load was increased to full power.

On March 14, at 0000 hours0 days 0 hours 0 weeks 0 months , load was dropped below 800 MWe, and the unit was placed on EGC at 0040 hours4.62963e-4 days 0.0111 hours 6.613757e-5 weeks 1.522e-5 months .

EGC was tripped on March 17 at 0020 hours2.314815e-4 days 0.00556 hours 3.306878e-5 weeks 7.61e-6 months , and load was dropped to 750 MWe for surveillances. The unit returned to EGC at 0125 hours0.00145 days 0.0347 hours 2.066799e-4 weeks 4.75625e-5 months . At 0830 hours0.00961 days 0.231 hours 0.00137 weeks 3.15815e-4 months EGC was tripped and load was increased to full power. On March 18, at 0010 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months , load was dropped below 800 MWe and the unit was placed on ECC.

I' B.

UNIT TWO (cont.)

March 19-31 On March 19, at 0120 hours0.00139 days 0.0333 hours 1.984127e-4 weeks 4.566e-5 months , EGC was tripped, and load decreased to 675

.MWe for surveillances. At 0535 hours0.00619 days 0.149 hours 8.845899e-4 weeks 2.035675e-4 months full power was reached after a normal load increase, and the unit was returned.to EGC at 0930 hours0.0108 days 0.258 hours 0.00154 weeks 3.53865e-4 months .

On March 20, at 0650 hours0.00752 days 0.181 hours 0.00107 weeks 2.47325e-4 months , the unit was taken off ECC and load was

-increased to full power,'and held until March 21, at 0235 hours0.00272 days 0.0653 hours 3.885582e-4 weeks 8.94175e-5 months , when the unit was placed on EGC.

On March 22 at 0044 hours5.092593e-4 days 0.0122 hours 7.275132e-5 weeks 1.6742e-5 months , the unit was taken off of EGC for surveillances. The unit returned to EGC operation at 0143 hours0.00166 days 0.0397 hours 2.364418e-4 weeks 5.44115e-5 months .

On March'24, at.0720 hours0.00833 days 0.2 hours 0.00119 weeks 2.7396e-4 months , the unit was taken off of EGC and load was increased to full power. The unit was returned to.EGC at 0333 hours0.00385 days 0.0925 hours 5.505952e-4 weeks 1.267065e-4 months , on March 25.

At 1053 hours0.0122 days 0.293 hours 0.00174 weeks 4.006665e-4 months the unit was taken off of EGC and load increased to full power. At 1327 hours0.0154 days 0.369 hours 0.00219 weeks 5.049235e-4 months the unit returned to EGC operation. On March 27, at 2026 hours0.0234 days 0.563 hours 0.00335 weeks 7.70893e-4 months , the unit was taken off of EGC and power was increased to full power. EGC operation was returned to.

on March 28, at 0057 hours6.597222e-4 days 0.0158 hours 9.424603e-5 weeks 2.16885e-5 months . At 0107 hours0.00124 days 0.0297 hours 1.76918e-4 weeks 4.07135e-5 months , load was increased to maximum power.

On March 30, at 1410 hours0.0163 days 0.392 hours 0.00233 weeks 5.36505e-4 months , load was dropped to 800 MWe for a Condensate Demineralizer differential pressure problem. On March 31, at.0340 hours0.00394 days 0.0944 hours 5.621693e-4 weeks 1.2937e-4 months load was increased to maximum power and was held for the remainder of March.

- III. PLANT OR' PROCEDURE CHANGES, TESTS, EXPERIMENTS, AND SAFETY RELATED MAINTENANCE-A.

Amendments to Facility License or Technical Specifications On February 7,1986, the NRC issued Amendment 92 to License DPR-29 and Amendment 89 to License DPR-30. These Amendments allow data to be obtained for scram insertion times during a-unit scram, and establish clarification on the requirements for the scram insertion times of control rods.

B.

Facility or Procedure Changes Requiring NRC Approval The attached information is in response to Quad-Cities Unit 1 Technical Specifications, section 6.10: MAJOR CllANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS. The 'B' train of the Unit 1 Off-Gas System has been modified to eliminate off-gas recombination problems. This modification was completed during the spring 1986' Unit 1 Refuel Outage.

Response to Technical Specification items are included as follows:

6.10.A.l' a.

Reference 2 b.

Reference 3 c'. Reference 3 d.

Reference 4 e.

Reference 4 f.

Reference 5 g.

Reference 6 L.

Tests and Experiments Requiring NRC Approval There were no tests or experiments requiring NRC approval for the reporting period.

r D.

Corrective Maintenance of Safety Related Equipment i

'The following represents a tabular summary of the major safety l

related maintenance performed on Units 1 and 2 during the reporting period. This summary includes the following: Work Request l

Numbers, Licensee Event Report Numbers, Components, Cause of Mal-functions, Results and Effects on Safe Operation, and Action Taken j

to Prevent Repetition.

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Table of Contents 1.

Quad-Cities, Unit'1, Technical Specification, Section 6.10 (copy).

2.

10 CFR 50.59. Format for Safety Evaluation (copy).

3.

Letter from J. E. Hausn.an to N. J. Kalivianakis, dated January 31, l

1986; subject: Off-Gas Modifica lon; Approval Letter (copy).

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Evaluation of change in release of radioactive materials in gaseous effluents.

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Evaluation of exposure to plant operatlng personnel.

6.- Modification Approval Sheet.(copy).

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QUAD-CITIES DPR-29 6.10 MAJOR CHANGES TO RADIOACTIVE WASTE TREATMENT SYSTEMS (LIQUID, GASEOUS, SOLID)

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Licensee initiated major changes to the radioactive waste systems may be made provided:

1.

The' change is reported in the Monthly Operating Report for the period in which the evaluation was reviewed by the onsite review function. The discussion of each change shall contain:

A summary of the evaluation that led to the determination a.

that the change could be made in accordance with 10 CFR 50.59; b.

Sufficient detailed information to support the reason for the change; c.

A detailed description of the equipment components, and process involved and the interfaces with other plant systems; d.

An evaluation of the change which shows the predicted releases of radioactive materials in liquid and gaseous effluents and (or quantity of solid waste that differ from those previously predicted in the license application and amendments);

A comparison of the predicted releases of radioactive e.

materials in 11guld and gaseous effluents and in solid waste to the actual releases for the period in which the changes were made; f.

An estimate of the exposure to plant operating personnel as a result of the change; and Documentation of the fact that the change was reviewed and 9

found acceptable by the onsite review function.

2.

The change shall become effective upon review and acceptance by onsite review function.

10328 6.10-1 Amendment No.

1 Revision 11 C

SNED Procedure Q.6 Exhibit C Commeswealth Edissa 10CFR50.59 FORMAT FOR SAFETY EVALUAT1

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STATICN Quad Cities Station UNIT 1

SYSTEM off-Gas System MOOlFICATION No. w-1 85 58 ECUIPMENT NAME Steam Jet Air Ejectors EQUIPMENT No.

N/A DESCRIPTION OF MODIFICATION:

This modification will increase the concentrations of steam in the off-gas system where existing off-gas concentrations are combustible.

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SAFETY EVAL.UATICN: Answer the following questions with a "yes" sr "no",

and provide specific reasons justifying the decision:

1 Is the probability of an occurrence or the consequence of en occident,. or malfunc equipment important to safety as previously evoivated in the Final Safety Analys increased?

Yes I __ No, Because:

probability of off-gas fires without effecting equipment important to safety as previously evaluated in the FSAR.

2.

Is the possibility for an accident er malfunction of a different ype then any p evaluated in the Final Sciety Analysis Report createo?

_Yes - I

_ No, Because:

This modification does not interface with any safety related equipment and any failure of equipment installed for this modification will not fall outside the boundaries of any single failure event of design basis accident already analyzed in the FSAR.

3.

Is the mergin of safety, as defined in the basis for any Technical Specifiestion, re Yes.

No, Because:

The modification will not change the design intent of the technical specification requirements of the off-gas system.

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Aporoved By Date2/#N0 0 6 (13)

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January 31, 1986

Subject:

Off-Gas Modification; Approval Letter

-Quad Cities Station, Unit 1 M4-1-85-58 Revision 0 Reference 1:- Letter from N.J Kaltvianakis to J.S. Abel dated October 11, 1985.

Reference 2:

Letter from J.E. Hausman to li.J. Kaltvianakis dated November 18, 1985 concerning the off-gas system code of construction.

Mr. N.J. Kalivianakis:

Modification Review The Station Nuclear Engineering Department has_ reviewed the request for engineering and approval for Modification Number M4-1-85-58.

(Reference 1).

Modification Approval The Station Nuclear Engineering Department approves of the nodification to'the off-gas system that will eliminate off-gas fires.

Reason for Change Over the past two years, off-gas fires have been experienced-at both the Dresden and Quad Cities Stations.

The Station Nuclear Engineering Department R&DE Group has been investigating these fires.

R&DE has determined that the source of off-gas ignition.is catalyst fines that.have migrated upstream of the off-gas recombiners.

The existing off-gas mixture in the presence of sufficient

~ catalyst can auto ignite.

Once the off-gas ignites, the flame propagates back to the region of the after condenser where a stable flame is established.

Continuous burning produces nitrates because of the'high flame. temperatures.

These compounds adversely affect condensate water chemistry.

Therefore, it is necessary to extinguish-these off-gas fires when they occur.

To extinguish these fires, it is required that plant operators follow a special procedure which may result in a plant derating.

Since fires have been occuring with some regularity, the potential for unit unavailability increases.

In addition to the chemistry problems and possible deratings, the NRC has recently taken an interest in the offgas fire situation with respect to its effect on the hydrogen water chemistry For these reasons it is desirable to modify the system to

. program.

preclude these fires.

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. Proposed Solutions Three methods of dealing with this problem were evaluated; l) removal of the catalyst by chemically cleaning the piping, 2) replacing the contaminated pipe with new pipe, and 3) eliminating the flammable region of the off-gas system.

The first two approaches remove the source of ignition while the last approach renders the off-gas mixture non-combustible.

Although the first two approaches were found to be technically feasible, these actions would not necessarily result in a permanent solution.

Since the mechanism for migration of catalyst fines is not wall understood, the clean or new pipe may recontaminate with time.

Therefore, the third option provides a higher degree of confidence that the problem will not reoccur.

J Dasign Change The proposed modification will render the flammable regions of the off-gas system non-flammable by making the co T.entration of steam in this region greater than 60% by volume.

In order to increase the concentration of steam in the flammable region, the after condenser will be by-passed.

This is accomplished by routing the outlet of the second stage SJAE 's directly to the loop seal.

This change will result in off-gas flow with a steam concentration of greater than 60% by volume.

The increase in off-gas flow would choke the booster SJ AE 's, therefore, the booster jets are being replaced by spool pieces.

Due to the loss in off-gas motive force previously provided by the booster SJAEs, this modification design includes replacing the 2nd i

Stage SJAEs with larger jets.

I If the after condenser is bypassed and the piping from the 2nd stage SJAEs to the preheater is insulated, the flow of condensed off-gas steam through the loop seal to the condenser will be significantly reduced.

The after condenser loop seal level control will be modified to accommodate the changes to the system.

The instrument measuring flow downstream of the after condenser loop seal consists of a restricting orifice and a differential

-pressure sensor.

The restricting orifice must be removed to accommodate the higher flow conditions.

This change will result in the loss of this flow indication.

The larger 2nd Stage SJAEs and the non-condensing nature of new system will result in a reduced steam dilution requirement.

The restricting orifice used to control dilution steam flow will be resized to accommodate the system change.

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. In its present configuration, the SJAE cannot be used until reactor steam pressure reaches approximately 700 lbs.

This limitation is a function of an interlock requiring at least 125 lbs steam be available at the dilution line.

Due to the reduced dilution steam requirement and the change in the restricting orifice size, the dilution line pressure will reach 125 lbs when reactor steam pressure reaches 200 lbs.

change will allow the SJAEs to be available hours earlier during startup.

This It is anticipated that these changes will reduce the time required to evacuate the condenser, and therefore, reduce startup time.

off-gas system shall be modified at this time.It has been determined that only the "B" train of the Unit 1

_ Summary of Design Evaluation The Station Nuclear Engineering Department has reviewed the proposed design as delineated above and the design documents developed for this modification and approves of the modification to eliminate off-gas fires that have affected the operation of Unit Station.

I at Quad Cities Verbal Approval Verbal approval was not requested or provided for this modification.

Classification of Modification The off-gas system is non-safety related.

Environmental Qualification This system is not required to function after an accident.

Therefore, environmental qualification documentation is not required for this equipment.

supplied equipment manuals andThe equipment should be maintained in accordance with maintenance practices.

in accordance with normal station Quality Requirements The professional engineering needed for this modification will Division Procedures of NUS.be performed in accordance with the appropriate sections standard non-safety related engineering practices.This work is performed in accordance w i

modification will be procuredThe materials procured for the ANSI B31.1 portion of this ASTM specifications.

in accordance ANSI B31.1 requirements and The materials procured for the Section XI portion of this modification shall be procured with additional documentation requirements.

in NUS installation specification 7957-N-201.The procurement documentation requireme h

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- l Installution labor and NDE requirements shall follow the l

requirements of NUS installation specification 7957-M-201.

This document specifies standard power plant practices.

SNED Procedure Q.6 requires the following paragraph to be included into the modification approval letter: "The final documentation checklist for this modification should include dimensional verification documentation.

This documentation should be checked to ensure that it includes evidence that ANSI N45.2.6 qualified inspectors have verified the installed dimensions of load carrying safety related structures are within design tolerances."

There are no safety related structures associated with this modification.

Fire Protection Requirements SNED has reviewed the Fire Hazards Analysis Report as part of the modification review process.

Based on figure 3.3-21 of this report SNED has determined that the off gas wall penetrations will not be considered penetrations in fire barrier walls and therefore require no special fire protection consideration.

These penetration must be sealed in accordance with normal station procedures.

Non-combustible seal material must be used.

Code Requirements Except as noted below, the code of construction for this modification is non-safety related ANSI 831.1 1967 Edition.

The code of construction for that portion of this modification which replaces the booster jet is ASME Section XI 1980 Edition, Winter of 1980 Addendum.

Section XI allows reference to the original code of construction which is ANSI B31.1 1967 Edition. (Reference 2)

Test Criteria 1.

Lnop seal drain valve -

The new loop seal drain valve shall be tested prior to start-up testing.

The valve shall operate from closed to full l

open within 20 seconds.

The orientation of the valve shall be l

varified to be correct.

2.

Flow indicators and pressure regulators -

The existing flow indicators and pressure regulators shall be inspected and calibrated prior system start-up in accordance with standard mainte an; practices.

e 3.

During system startup, the he is

.ak detection equipment shall be connected to an off-gas sample line of the active train.

Helium shall be directed at all the modified piping and equipment.

The helium leak detection equipment must show no leakage.

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4.

Start-up testing -

Modification testing to verify operability shall be performed by an inservice test.

During system start-up the station shall verify that the modified operating procedures I

are used.

The station shall monitor the dilution steam supply pressure, recombiner temperature and condenser vacuum.

Additionally, the station will visually inspect the off-gas condenser, the new SJAEs and the new insulation for excessive vibration.

5.

Following the completion of the equipment installation the I

SJAE area radiation monitors shall be recalibrated.

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These testing requirements have been reviewed against the i

testing Checklist (Exhibit G) of SNED procedure Q.6.

Heat Rate This modification reduces the steam requirement of the off-gas system.

This reduction of steam may effect heat rate calculations.

Technical Specifications The Off-gas system which handles radioactive gaseous effluent l

is mentioned in several sections of the technical specifications.

These sections are as follows:

1.

Tech. Spec. 3.8/4.2-4 Radioactive Gaseous Effluent Instrumentation 2.

Tech. Spec. 3.8/4.8-1 Radioactive Gaseous Effluent Release Limits 3.

Tech. Spec. 6.8-1 Off-Site Dose Calculation Manual 4.

Tech. Spec. 6.10-1 Major Changes to Radioactive Waste Treatment Systems Although this modification does not change the design intent of the system, it may be considered as a major change to the radioactive waste treatment systam.

The station shall evaluate the need to include a description of this change in the Monthly Operating Report following the completion of the modification in accordance with the requirements of Tech. Spec. 6.10-1.

No technical specification changes are required for this modification.

e

. Final Safety Analysis Report As a result of a major modification to the off-gas system in the early seventies, the original and updated FSARs have two different descriptions of the off-gas systems.

the design and operation of the off-gas system.This modification will also change The Station Nuclear Engineering Department recommends that the Station change the system description in the updated FSAR to incorporate the changes resulting from this modification.

The design basis of the off-gas system in the original and updated FSAR is identical.

located in Section 9.2.

The FSAR off-gas system description is Outage Related Portions The off-gas modification must be installed during a unit outage.

Therefore, this modification is outage related.

Drawings NUS modification drawings will be issued for construction by February 12, 1986.

changes shall be made using an FCR or DCR as appropriate.After these draw ALARA The off-gas system rooms are high radiation areas during normal operations.

During outages, the off-gas system is no longer a high radiation area.

Therefore, during the outa exposure to workers will be less than 1 mange it is anticipated that the rem.

Therefore, the corporate ALARA Manual does not require a cost-benefit comparison or an ALARA Action Review.

Equipment Procurement The Station Nuclear Engineering Department will procure the 2nd Stage Steam-Jet Air Ejectors.

February 24, 1986.

These SJAE 's are scheduled for delivery by The piping, supports, spool pieces and miscellaneous materials will be procured by-Station Construction in accordance with a bill of materials delineated on NUS construction drawings.

These drawings were provided on January 24, 1986.

Craft Labor the on-site labor contractor, Morrison Construction. Station Construction will provid

_.. Budget The Station Nuclear Engineering Annual Budget 810000 will be used for this project.

All labor and material procurement costs shall be-charged to Work Order fl07832A.

All engineering costs will be charged to Work Order fl05257A.

If there are further questions concerning this modification please contact me on x-8321.

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Written by:

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Approved by:

f-2sr#t d U<E. Hausman EJR/dg/68780 cc:

L.D. Butterfield (TSN) w/att.

E.J. Rowley w/att.

F.A. Palmer (NS) w/att.

D.L. Sanderson-(SNED) w/att.

G.

Carney (Quad Cities Station) w/att.

L.

Petrie (SC) w/o att.

D.

Gibson (QA) w/o att.

M.R. Mychajliw (NUS) w/o att.

'W.J.

Shewski (QA) w/att.

R.

Zentner (Dresden) w/o att.

J.

Achterberg (Dresden) w/o att.

K.

Hill w/o att.

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4. - Evaluation of change in release of radioactive materials in gaseous effluents.

The release of radioactive materials in gaseous effluents as a result of this modification is not expected to change. There are several reasons why this is true.

First, the mass flow rate of gases from the condenser is not changed. The new secondary jets are designed to operate at the same condenser vacuum, steam pressure, and off-gas flow rate ~as the previous secondary jets.

Second, the mass flov rate of steam to the recombiner is not changed. More steam exists in the mixture being discharged from the secondary jets (due to more motive steam and the noncondensing operation of the jets) but this is off-set by reducing the amount of dilution steam added downstream. Finally, the major component of the off-gas system.is the 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months holdup pipe. This modification does not affect the hold-up pipe.or the holdup time.

Actual releases in gaseous effluents can be found in the monthly operating reports of this Station.

'5.

Evaluation of exposure to plant operating personnel.

This modification is not expected to result in any additional exposure to plant operating personnel. The affected off-gas l

piping is surrounded by biological shielding. Also, all valves which require to be operated have remote manual controls.

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UNIT 1

MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETIT10N Q37512 1/2-6601 Diesel Suspected that the Without an air supply, The solenoid of the Generator Air solenoid valve the air start motors solenoid valve on the air Start System bound up, &

could not spin. The start line was replaced.

Solenoid Valve stopped the air Diesel could not be flow to the air star.ed.

starting motors.

Q37514 1-6601 Diesel Suspected that the Without an air supply, The solenoid of the Generator Air solenoid valve the air start motors solenoid valve on the air Start System bound up, &

could not spin. The start line was replaced.

Solenoid Valve stopped the air Diesel could not be flow to the air started.

starting motors.

Q43490 1B-1002 RHR An oil seal of the The motor tripped while A replacement motor was Pump Motor motor upper being run. The other RHR installed.

bearing failed.

systems, Core Spray This allowed oil systems, & Diesel to cover the pump, Generators were avail-motor, & motor able for emergency use, windings. This if needed.

effectively ruined the motor.

Q43601 1B-1002 RHR An oil seal of the The motor tripped while A replacement motot was Pump hator motor upper being run. The other RHR installed.

bearing failed, systems, Core Spray This allowed oil systems, & Diesel to cover the pump, Generators were avail-motor, & motor able for emergency use, windings. This if needed.

effectively ruined the motor.

0027H/0061Z

y-UNIT 1

MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q43602 1B-1002 RHR An oil seal of the The motor tripped while.

A replacement motor was Pump Motor motor upper being run. The other RHR installed.

bearing failed.

systems, Core Spray This allowed oil systems, & Diesel to cover the Jump, Generators were avail-motor, & motor able for emergency use, windings. This if needed.

effectively ruined the motor.

Q43687 85-05 CA 18101 1B Previous cable The noise in the cable The cable was rerouted Fuel Pool route was near would cause a spike (high to avoid the suspected Radiation various equipment.

or low) of the radiation noise inducing equipment.

Monitor Cable The operating monitor. This started the equipment induced logic to start various electrical inter-safety related -

tilation ference (noise) systems.

in the cable.

Q47066 86-01 1-220-58B Worn pins &

Failed Local Leak Rate Installed new pins &

Inboard Feed-busings allowed

Test, busings, water Check enough play in Valve valve to prevent good seating of the disc.

Q47219 l-302-109D Personnel error The amplifier failure Immediate action was to Scram Dis-while removing caused a

'B' SCRAM signal.

replace the amplifier board.

charge wiring contacts The

'A' SCRAM signal was A modification is in Volume Level allowed one not received. Unit I was progress to install new Transmitter contact to touch shutdown at the time of circuitry to eliminate the another. This the

'B' SCRAM signal, need to lift the contacts.

caused a short in the amplifier board circuitry.

0027H/0061Z

4 UNIT 1

MAINTENANCE

SUMMARY

CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NUEER NUMBER COMPONENT MALFUNCT10N SAFE OPERAT10N PREVENT REPETIT10N Q47260 Bus 14-1, Personnel error The energized contact Immediate action was to Cubicles 6 &

allowed an in-simulated an under-clear the Diesel alarm &

9 - Security correct panel voltage trip, which start the Diesel. A Diesel contact to be opened the Security procedure has been Emergency energized while Transformer breaker, changed to allow a cool Generator a modification Gatehouse lost power down period for the Diesel.

test was in when the Diesel did not Personnel were reminded progress. A start due to its high to double-check wiring, procedural error temperature alarm, shut off the Security Diesel Genertor before it had cooled down.

Q48039 86-01 A0 1-2001-16 Possible packing Failed Local Leak Rate Disassembled valve, Drywell Equip-leak.

Test.

inspected, reassembled ment Drain with new packing.

Sump Isolation Valve.

Q48235 86-01 A0 1-1601-60 Valve disc was Failed Local Leak Rate The valve disc was adjusted Torus Vent not seating Test.

for proper seating.

Valve properly.

Q48293 86-01 Penetration Penetration seal Failed Local Leak Rate Replaced Penetration link MK-489 - RHR material had Test.

seal and recaulked.

Service Water deteriorated.

Pump Vault Penetration 0027H/0061Z

UNIT 1

MAINTENANCE SUhWARY I

i CAUSE RESULTS & EFFECTS f

W.R.

LER OF ON ACTION TAKEN TO NUREER NUREER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q48294 86-01 Penetration Penetration seal Failed Local Leak Rate Replaced link seal and MK-479 - RHR material had Test.

recaulked around Service Water deteriorated.

Penetration.

Pump Vault Penetration Q48427 86-01 1-3999-516B Valve internals Failed Local Leak Rate Valve internals were RHR Service were dirty.

Test.

cleaned. Valve bonnet Water Vault Valve bonnet was was honed to remove Sump Pump corroded.

corrosion.

Check Valve Q48567 86-01 AO 1-2001-3 Unknown.

Failed Local Leak Rate Disassembled, inspected, Drywell Floor Test.

and reassembled. No Drain Sump reason for leakage found.

Isolation Valve 0027H/0061Z

D"7 UNIT 2

MAINTENANCE

SUMMARY

a CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NUMBER NUMBER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q46993 2-2001-16 An accumulation The valve failed to open The valve was manually Drywell Equip-of sediment in when the pump was needed opened. The valve will be ment Drain the valve to drain the sump. The disassesbled during next Sump Line internals pre-valve's safety isolation outage of sufficient Isolation vented valve position is closed, so duration.

Valve opening.

isolation was still intact.

Q48058 2-1705D Main No cause found.

The 'D' radiation monitor The 'D' radiation monitor Steam Line This is an reads lower than the other was tested fully. No reason Log. Radiation apparent charac-three monitors. However, for low reading. Apparently Monitor teristic of this all four monitors are just a characteristic of individual within limits.

this monitor, monitor. It is within all limits, just on the low side.

Q48095 86-03 2-2303 HPCI Piping reroute The wet side (at higher The incorrect piping was Gland Seal design error by pressure) forced water removed and drain line fed Condenser Operations & Tech up the drain line into to floor drain. Modifica-Exhauster Staff personnel the gland seal. Water tion in progress to route Drain Line routed the line to was forced out around the drain line into correct the wet side of seal. HPCI was declared portion of the gland seal the condenser.

inoperable to remove condenser.

incorrect line.

0027H/0061Z

UNIT 2

MAINTENANCE SUhMARY CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NLABER NLAEER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q48097 86-03 2-2372-3/8" HPCI Piping reroute The wet side (at higher The incorrect piping was Cland Seal design error by pressure) forced water removed and drain line fed Condenser Operations & Tech up the drain line into to floor drain. Modifica-Exhauster Drain Staff personnel the gland seal. Water tion in progress to route Line routed the line to was forced out around the drain line into correct the wet side of seal. HPCI was declared portion of the gland seal the condenser, inoperable to remove condenser.

incorrect line.

Q48139 2-733-2 TIP The valve The valve would not close, The manual isolation valve Valve internals had but isolation was avail-was closed. A modification corroded, and the able using the manual is in progress to use friction caused valve & the explosive nitrogen for the TIP drive by this corrosion shear valve used for purge system. Nitrogen will prevented the primary isolation, not have moisture in it.

valve from closing.

The moisture was from the instrument air system which supplies the TIP drive purge system.

Q48175 2-733-4 TIP The valve The valve would not close The manual isolation valve Valve internals had within the required 5 was closed. A modification corroded, & the seconds. Isolation was is in progress to use friction caused by available using either nitrogen for the TIP drive this corrosion the manual isolation purge system. The nitrogen prevented the valve or the explosive will not have moisture in it.

valve from closing shear valves, within 5 seconds.

The moisture was from the instrument air line which supplies the TIP drive purge system.

0027H/0061Z

UNIT 2

MAINTENANCE SUWARY CAUSE RESULTS & EFFECTS W.R.

LER OF ON ACTION TAKEN TO NLABER NUWER COMPONENT MALFUNCTION SAFE OPERATION PREVENT REPETITION Q48267 2-2301-49 IIPCI Electrical The Unit 2 valve was not The wiring was hooked back-Cooling Water Maintenance operable. It is a test up to the valve, & it was Test Return personnel in-valve, which is normally functionally tested. Several Valve correctly dis-closed & is not required procedures have been connected the during emergency llPCI created to double-check that wiring from the operation.

the correct equipment is Unit 2 '49' valve, being worked on, instead of the intended Unit 1

'49' valve.

e 0027H/0061Z

IV.

LICENSEE EVENT REPORTS The following is a tabular summary of all licensee event reports for Quad-Cities Units One and Two occurring during the reporting period, pursuant to the reportable occurrence reporting requirements as set forth in sections 6.6.8.1. and 6.6.B.2. of the Technical Specifications.

UNIT ONE Licensee Event Report Number Date Title of Occurrence 86-13 3-10-86 Excess Flow Check Valve Surveillance Causes ATWS Initiation 86-14 3-12-86 Loss of November 1985 Gaseous Particulate Composite Samples 86-15 3-12-86 Reactor Scram From Scram Discharge Volume High Level While Switching Reactor Protection System Bus Feed 86-16 3-17-86 Unit 1 Diesel Generator Auto Start After Core Spray Logic Test 86-17 3-18-86 Reactor Vessel Instruments Isolated 86-18 3-20-86 Reactor Scram While Per-forming 4KV Undervoltage Test of Bus 13-1 86-19 3-17-86 Spurious ATWS Trip 86-20 3-28-86 Group I Isolations Due to Bumping Rack 86-08 3-25-86 1/2 Diesel Generator Auto Start Due to Inadvertent Relay Actuation

F-

.v

.,33:

IV.

LICENSEE EVENT REPORTS (cont.)

UNIT TWO Licensee Event Report Number Date.

Title of Occurrence

86-03 3-3 Unit 2 HPCI Declared Inoperable Due to HPCI Gland Exhaust Piping Error 86-04 3-14-86 HPCI Turbine Trip Due-to Turbine Exhaust Pressure Switch l

u a

V.

DATA TABULATIONS The following data tabulations are presented in this report:

A.

Operating Data Report B.

Average Daily Unit Power Level C.

Unit Shutdowns and Power Reductions

e a

f3 v g.:

OPERATING-DATA REPORT DOCKET NO.

50-254-UNIT ONE-DATEAPRIL 3 1986 COMPLETED BYCAROL L KRONICH TELEPHONE (309)654-2241 OPERAT'ING STATUS-i L

0000-030186'

i. Reporting period 2400 033186 Gross hours in reporting period 744

2. Currently authorized power level (MWt):~2511 Max. Depend capacity (MWe-Net): 769* Design electrical rating (MWe-Net): 789-

[-.

3. Power level to which restricted (if any)(MWe-Net): NA l

4. Reasons for restriction (if any):

This Month Yr.to Date Cumulative l

L5.: Number of hours reactor was critical 0.0 120.0 96781.4 l

6. Reactor reserve shutdown hours 0.0 0.0 3421.9

7. Hours generator on line 0.0 120.0 93399.5 l

8. Unit reserve shutdown hours.

0.0 0.0 909.2

9. Gross thermal energy generated (MWH) 0 240485 195205642

10. Gross electrical energy generated (MWH) 0 80622 63199993

11. Het electrical energy generated (MWH)

-2985 69225 59095758

12. Reactor service factor 0.0 5.6 79.5 13.-Reactor avo11ob111ty factor 0.0 5.6 02.3

14. Unit service factor 0.0 5,6 76.7_

l

15. Unit availability factor 0.0 5.6 77.5

16. Unit capacity factor (Using MDC)

.5 4.2 63.1

17. Unit copocity factor (Using Des.MWe)

.5 4.1 61.5

18. Unit forced outage rate 0.0 0.0 5.9

19. Shutdowns scheduled over next 6 Months (Type,Date,and Duration of each):

20. If shutdown at end of report per i od, e st inat ed da te o f s tar t up ___4-3_fL6 ____

L GMFFICIAL COWANT NUNIERS ARE USED IN THIS REPORT

OPERATING DATA REPORT 3

DOCKET NO.

50-265 UNIT TWO DATEAPRIL 3 1986 COMPLETED BYCAROL L KRONICH TELEPHONE (309)654-2241 OPERATING STATUS 0000 030186

1. Reporting period 2400 033186 Gross hours in reporting period 744 2.. Currently authorized power level (MWt): 2511 Max. Depend copocity (MWe-Net)- 769* Design electrical rating (HWe-Net): 789 3.

Power level to which restricted (if any)(HWe-Net): NA 4.

Reasons for restriction (if any):

This Month Yr.to Date Cumulative

5. Number of hours reactor was critical 691.7 1934.3 93202.0 l

6.

Reactor reserve shutdown hours 0.0 0.0 2985.8 i

7. Hours generator on line 669.8 1909.0 90206.6

8. Unit reserve shutdown hours.

0.0 0.0 702.9.

I

9. Gross thernal energy generated (MWH) 1599870 4542126 190651149 l '

10. Gross electrical energy generated (MWH) 520410 1497469 60914251 i

11. Het electrical energy generated (MWH) 504779 1430873 57305132

12. Reactor service factor 93.0 89.5 77.1

13. Reactor avo11obility factor 93.0 89.5 79.6

14. Unit service factor 90.0 88.4 74.7

15. Unit avo11ob111ty factor 90.0 88.4 75.2 i <6. Unit capacity factor (Using MDC) 88.2 86.1 61.7_

17. Unit capacity factor (Using Des.MWe) 86.0 84.0 60.1 18._ Unit forced outage rate 2.5 1.3 0.0

19. Shutdowns scheduled.over next 6 months (Type,Date,ond Duration of each):

20. If shutdown at end of report period,estincted date of startup NA SUMEFICIAL COMPANT NUMERS ARE USED IN THIS REPORT

APPENDIX B AVERAGE DAILY UNIT POWER LEVEL e

DOCKET NO.

50-254 UNIT ONE DATEAPRIL 3 1986 COMPLETED BYCAROL L KRONICH TELEPHONE (309)654-2241 MONTH Horch 1986 DAY AVERAGE DAILY POWER LEVEL DAY AVER AGE DAILY POWER LEVEL (HWe-Net)

(MWe-Net) 1.

-3.5 17.

-5.6 2.

-3.0 10.

-5.0 3.

-3.3 19.

-5.8 4.

-3.5 20.

-5.5 5.

-2.0 21.

-5.8 6.

-3.3 22.

-5.1 7.

-5.0 23.

-4.0 8.

-5.5 24.

-5.4 9.

-5.5 25.

-5.2 10.

-5.0 26.

-5.3 11.

-5.3 27.

-5.4 12.

-5.5 20.

-5.1 13.

-6.9 29.

-E.3 14.

-7.5 73.

-7 o 15.

-7.7 31.

-6.9 16.

-7.7 INSTRUCTIONS On this forn, list the overage daily snit pour level in 1%ie-Net for each day in the reporting nenth.Ceppste to the nearest whole negewett.

These flgeres will be esed to plot a graph for each reporting eenth. Note that when naninen dependable capacity is used for the net electrical rating of the snit be occasions when the daily everage powr level exceeds the ill! line (or the restricted pour level line),there not.In such cases,the overage delly snit pou r estpet sheet sheeld b festnoted to explain the apparent onently

APPENDIX B AVERAGE DAILY UNIT POWER LEVEL j-

1. g DOCKET NO.

50-265

-UNIT TWO DATEAPRIL 3 1986 COMPLETED BYCAROL L KRONICH 1ELEPHONE(309)654-2241 MONTH March 1986 DAY AVERAGE DAILY POWER LEVEL- -

DAY AVERAGE DAILY POWER LEVEL:

(MWe-Net)

(MWe-Net) 1.-

-7.2 17.

834.8 2.

-6.8 18, 664.0 3.

40.5 19.

728.2 4.

468.i 20.

779.0 5.

757.4 21, 740.0 6.

792.8 22.

750.0 7.

'918.8 23.

'731.6 8.

781.4 24.

7G7.0 9.

790.7 25.

742.4

~ 10 '.

798.4 26.

723.9

.11..

806.5 27.

747.0

12..

774.5 28.

803.5 13, 789.3 29, 768.7-14, 750.8 30.

783.9 15, 751.8 31, 789.0 16.

760.5 INSTRUCTIONS en Cis f list the enrega delly wit peer level in late-ht for each der in the reporting noth.Cupote to the neerest I negemett.

These figures will k esed to plot a graph for each reporting nenth.16ete that when neelnen dependeble cepecite is osed for ik set electrical reting of the salt there ner k occasions when the dellt eserege power level exceeds the issa line (or the restricted power level line).,In suh cases,the enrega deilt enit pwer setpet sheet shoold be feetnoted to esplein tk apperent onenely a..

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c VI. UNIQUE REPORTING REQUIREMENTS The following items are included in this report based on prior commitments to the commission:

A.

MAIN STEAM RELIEF VALVE OPERATIONS There were no Main Steam Relief Valve Operations for the reporting period.

B.

CONTROL RCD DRIVE SCRAM TIMING DATA FOR UNITS ONE AND TWO There was no Control Rod Drive Scram Timing Data for Units One and Two for the reporting period.

I

o 1

o VII.

REFUELING INFORMATION The following information about future reloads at Quad-Citle-Station was requested in a January 26, 1978, licensing memorandum (78-2% from D. E.

O'Brien to C. Reed, et al., titled "Dresden Quad-Cities, and Zion Station--NRC Request for Refueling Information", dated January 18, 1978.

l l

f

QTP 300-S32 Revision 1 QUAD-CITIES REFUEL. LNG March 1978 INFORMATION REQUEST 1.

Unit:

Q1 Reload:

7 Cycle:

8 9-14-87 2.

Scheduled date for next refueling shutdown:

3 Scheduled date for restart following refueling:

4-3-86 4.

Will refueling or resumption of operation thereaf ter require a technical specification change or other license amendment: YES. A ROUTINE MAPLHCR AMENDMENT HAS BEEN SUBMITTED AS A PREPARATORY CHANGE TO ALLOW A 10 CFR 50.59 REVIEW OF THE RELOAD. TECHNICAL SPECIFICATION CHANGE TO ALLOW HAFNIUM ABSORBER MATERIAL HAS BEEN SUBMITTED.

5 Scheduled date(s) for submitting proposed licensing action and supporting information:

6.

Important licensing considerations associated with refueling, e.g., new or

' different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:

A) USE OF 8 ASEA-ATOM CONTROL BLADES CONTAINING HAFNIUM.

B) CONTINUED USE OF TWO BARRIER LEAD TEST ASSEMBLIES.

7 The number of fuel assemblies, 724 a.

Number of assemblies in core:

b.

Number of assemblies in spent fuel pool:

1894 8.

The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is planned in number of fuel assemblies:

3657 Licensed storage capacity for spent fuel:

a.

0 b.

Planned increase in licensed storage:

9.

The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2003 WPPROVED APR 2 01978 C3.c:.c).s.Ft.

I QTP 300-S32 Revision 1 QUAD-CITIES REFUELING March 1978 INFORMATION REQUEST 1.

Unit:

Q2 Reload:

7 Cycle:

8 2.

Scheduled date for next refueling shutdown:

10-13-86 3

Scheduled date for restart following refueling:

1-19-87 4.

Will refueling or resumption of operation thereaf ter require a technical specification change or other license amendment:

NOT AS YET DETERMINED.

5.

Scheduled date(s) for submitting proposed licensing action and supporting Informa tion:

SEPTEMBER 19, 1986, IF REQUIRED.

6.

Important IIcensing considerations associated sith refueling, e.g., new or

different fuel design or supplier, unreviewed design or performance analysis methods, significant changes in fuel design, new operating procedures:

NONE PLANNED AT PRESENT TIME.

7 The number of fuel assemblies.

a.

Number of assemblies in core:

724 b.

Number of assemblies in spent fuel pool:

838 8.

The present licensed spent fuel pool storage capacity and the size of any increase in licensed storage capacity that has been requested or is planned in number of fuel assemblies:

a.

Licensed storage capacity for spent fuel:

3897 b.

Planned increase in licensed storage:

0 9.

The projected date of the last refueling that can be discharged to the spent fuel pool assuming the present licensed capacity: 2003 APPROVED APR 2 01978 Q.c.o.S.R.

O VIII. GLOSSARY l

.The following abbreviations which may have been used in the Monthly Report, are defined below:

ACAD/ CAM -

Atmospheric Containment Atmospheric Dilution / Containment Atmospheric Monitoring ANSI American National Standards Institute APRM Average Power Range Monitor ATHS-Anticipated Transient Without Scram BHR Bolling Water Reactor CRD Control Rod Drive EHC Electro-Hydraulic Control System EOF Emergency Operations facility GSEP Generating Stations Emergency Plan HEPA High-Efficiency Particulate Filter HPCI-High Pressure Coolant Injection System HRSS High Radiation Sampling System IPCLRT Integrated Primary Containment Leak Rate Test

.IRM Intermediate Range Monitor ISI Inservice Inspection LER Licensee Event Report LLRT Local Leak Rate Test LPCI Low Pressure Coolant Injection Mode of RHRS LPRM Local Power Range Monttor MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MFLCPR Maximum Fraction Limiting Critical Power Ratio MPC Maximum Permissible Concentration MSIV Main Steam Isolation Valve NIOSH National Institute for Occupational Safety and Health PCI Primary Containment Isolation PCIOMR Preconditioning Interim Operating Management Recommendations RBCCW Reactor Building Closed Cooling Water System RBM Rod Block Monitor RCIC Reactor Core Isolation Cooling System Residual Heat Removal System RHRS

.RPS Reactor Protection System RWM Rod Worth Minimizer SBGTS Standby Gas Treatment System SBLC Standby Liquid Control SDC Shutdown Cooling Mode of RHRS Scram Discharge Volume SDV SRM Source Range Monitor TBCCW Turbine Building Closed Cooling Water System TIP Traversing Incore Probe TSC Technical Support Center

I I

j Commonwealth Edison 4

ouad Cities Nuclear Power Station 22710 206 Avenue North Corcova, tilinois 61242 Telephone 309/654-2241 TKT-86-30 April 1, 1986 Director, Office of Inspection & Enforcement United States Nuclear Regulatory Commission Washington, D. C.

20555 Attention: Document Control Desk Enclosed for your information is the Monthly Performance Report covering the operation of Quad-Cities Nuclear Power Station, Units One and Two, during the month of March, 1986.

Raspectfully, COMMONHEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION V

T. K. Tamlyn Services Superintendent bb Enclosure l

4 9

6

.

ML20141E426

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