ML20006E172

Forwards Listing of Changes,Tests & Experiments Completed During Jan 1990,including Items Completed in 1989. Interlocks Installed on Refuel Bridge Fuel Handling Machine to Prevent Raising Hoist While Hoist Loaded

ML20006E172
Person / Time
Site:	Quad Cities
Issue date:	02/02/1990
From:	Robey R COMMONWEALTH EDISON CO.
To:	Office of Nuclear Reactor Regulation
References
RAR-90-11, NUDOCS 9002220223
Download: ML20006E172 (12)

Text

g Commonwealth Edison s

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'*a-A' Ouad Cites Nuclear Power Statm -

._22710 206 Avenue North Cordova, Illinois 61242-9740 Telephone 309'654 2241 RAR-90,

February 2, 1990 Director of Nuclear Reactor Regulations U.S. Nuclear Regulatory Commission

. Mail Station P1-137 Washington, D.C.-

20555 3.

Enclosed please find a listing of those changes, tests, and experiments L.

completed during the month of January 1990, for Quad-Cities Station Units-1 and 2, DPR-29 and DPR-30.

In addition, several items completed R

-during the previous year are being submitted. This'is in response to a.

request by the previous NNR Project Manager to conduct a review of 1989 safety evaluations. A summary of the safety evaluations are being

reported in compliance with 10CFR50.59 and 10CFR50.71(e).

Thirty-n?"e copies are provided for your use.

Respectfully,

. COMMONWEALTH EDISON COMPANY QUAD-CITIES NUCLEAR POWER STATION 6s

. R.A..Robey v Technical Superintendent

'RAR/LFD/vmk Enclosures i

cc:

R. Stols T. Watts /J. Galligan 9002220223 900202 l' {

'PDR ADOCK 05000254 R-PNV

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Special' Test #1-121 4

Description Special-- Test - No'. 1-121 was completed on April 6 -'1989.' :The purpose of i

this-test was to determine the Reactor Core Isolation Cooling System Steam _Line

'High' Flow Setpoint.. Per' Technical Specifications.Tablef3.2-1,the RUIC-system 1s designed to isolate at less'than 300 percent of' rated steam flow. Steam line flow is measuredhas a differential pressure acrossoa flow elbow installed l

infthe steam supply line.

The instrumentation, which is used to isolate'RCIC Lon:high differential pressure, is located locally in the RCIC cornerEroom on-

-instrument rack 2201-58.- This test measured 1the differential pressure developed

.under test conditions and then determined a value corresponding 1to 300 percent-of rated-l steam flow using a_ formula provided by General Electric.-

I Evaluation It was determined that no 10CFR50.59 Safety Evaluation was required for.

this special test.

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Special Test #2-89 v-

-Description 5 Special, Test No.I2-89 was completed-on January 16, 1989. The purpose of this; test-was to determine the Reactor _ Core Isolation Cooling System Steam _Line

lligh' Flow Setpoint. Per~ Technical Specifications Table 3.2-1, the RCIC system is' designed;to isolate at-less than_300 percent of rated steam flow.. Steam lline flow is measured:as a differential pressure across a flow elbow installed in_the steam supply line.

The instrumentation, which is used to' isolate RCIC on high differential pressure, is located-locally in the-RCIC corner room on

'instrumentLrack 2201-58.- This test measured the differential pressure developed under test conditions and then datermined a value corresponding to 300 percent

-of rated steam flow using-a formula provided by General Electric.

Evaluation:

It was determined that no 10CFR50.59 Safety Evaluation was required for

-this special test.-

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Procedure Change QAP 300-2, Revision 24.

Conduct of Shift Operations The change is to the calculation used to determine the radwaste discharge f

rate.'.A requirement.was added to have the fish pumps off for the duration of

the discharge. This is'done to prevent pumping the discharge to the spray canal.

l The lift pumps were removed from the flow rate calculation since they are no longer.used.-

1.

The~ probability of an occurrence or the consequence of an accident, or malfunction of equipment important to safety as previously evaluated

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in the Final Safety Analysis Report is not increased because this change had no impact;on the possibility of an accident.

It only reflects

' current operating conditions,

The. possibility for=an accident or malfunction of a different type -

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g than any previously evaluated in.the Final Safety Analysis Report is.

not created because requiring the fish pumps to be off during.a discharge only affects the fish hatchery in the spray canal which.has no consequences 1

vx on the possibility of an accident..

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The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because this change reflected the fact that j

the lift pumps are not used and that it is preferrable not to pump

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the. discharge to the spray canal using the fish pumps..

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-Procedure-Change QO'S'1300-1, Revision 11 RCIC Monthly ~and Quarterly Test 4

This change clarified certain steps-involving quarterly flow rate testing.

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.-To change component names to match labelling in control room, change test para-meters for quarterly testing of Unit One RCIC and to delete steps involving a

.RHR/ Core Spray fill system. alarm which should not.ina related to.RCIC system operation.

"1 1.

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

e or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report is not-increased because changes L

clarified steps and provided more details, therefore the probability of an occurrence should be reduced.

2.

.The. possibility for-an accident or malfunction'of'a different type than any previously evaluated-in the Final Safety Analysis Report is l

h not created because: basic testing method remains unchanged, therefore 1

no new possibility for an accident is created.

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3.

The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because revised test procedure still meets all Tech Spec requirements for system operability.

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a Procedure Change QTP 1100-1, Revision 8:

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Core Verification.

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, 'This change-included a' step-to ensure distribution of completed core, veri-g fication'was performed in accordance with' CECO nuclear. procedures. Grammatical

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changes were also made.:-

i 1.--The probability of'an._ occurrence or the consequence of an: accident,

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or-malfunction ofiequipment important-to safety as previously evaluated

'1n_the Final Safety. Analysis Report is not-increased.because core veri-

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' fication.is still completed' as before.

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The possibility for an accident or malfunction _of a different type-Jj

than any previously evaluated in the-Final Safety Analysis Report isi O

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.not' created because-the core is still verified prior =to reactor operation.

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The margin of safety, as defined in the basis for any Technical Spect-

-fication, is'not reduced because core verification is unchanged.

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o Safety Evaluation #89-217

-Incomplete Modification M-4-1-86-017 Insta11' Light Activated-Sensor on Roll-0-Matic Filters-0 Mod'ification M-4-1-86-017 was incompleted. Three control systems. consisting of light activated sensors for filter, movement were installed, tested and operation t

- authorized.

The four remaining filters will not have the-new systemLinstalled.

1.

The prob' ability of an occurrence or the consequence of_an_ accident -

~or malfunction'of equipment important to safety as previously evaluated

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'in the Final Safety Analysis _ Report is not-increased because the FSAR does not, discuss the control system for_the advancement of.the filters.

Both control systems, however, will advance the filter media as designed.-

2. -The possibilit'y for an accident or malfunction of a different type than any previously evaluated in-the Final Safety Analysis Report is

-not created because the control system for the filters is not discussed in the FSAR.

3.

The margin of-safety, as defined in the basis for any Technical Speci-fication, is-not reduced because the Tech Specs do not discuss the-control system for the filters.

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. Safety Evaluation #89-519 w"

Modification M-4-2-89-152 Interlocks Installed on Refuel Bridge Fuci Handling Machine:

This modification was installed:as a corrective action per' potential signifi-icant event report PSE-89-006 titled "New fuel bundles drop while in fuel pool".

The PSE occurred.on 9-21-89 at Quad Cities Unit'l (see PSE-89-006 for details).

This modification installed an additional electrical interlock that prevented raising the hoist on the fuel moving machine while the hoist is loaded unless

the grapple is fully closed and in the engage position..This._ modification is contained in the G.E. fuel moving panel located on the refuel bridge,

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1. 3The probabilityfof 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 modi-

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fication will add an additional feature to the interlock system to enhance the safe movement of fuel.

-2.

The possibility for an accident or malfunction of a different type than any previously evaluated in the_ Final 1 Safety Analysis Report is.

not created because-this modification added an additional interlock protection;to an evaluated condition.

3.

The' margin of safety, as defined in the basis for any Technica1'Speci-fication :1s'not: reduced because this modification increased the margin' of safety while. moving fuel.

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c3 Safety Evaluation'#89-617 IPCLRT Sof tware: Changes L

The-IPCLRT software was_ adapted-to allow compatibility with the new j

' Revision 21 Eof-PRIMOS. The software was modified to collectcpoint history data l

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2thereby bypassing the Honeywell computer.

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1. _The probability of an occurrence or the consequence of an accident,c

-or malfunction of equipment important-to safety-as previously evaluated; l

c in the Final--Safety: Analysis Report is not increased because software

-is'not evaluated or described in the FSAR.

2.

The possibil'ity for an accident or malfunction of a different type-

-than any,previously evaluated in the Final Safety AnalysisLReportiis-n,

not created.because the accuracy of the software will be qualified and since it only processes test data, it should not result in an

accident-or malfunction.

3.--The margin of safety, as_ defined in the basis for any Technical Speci--

h fication, is not-reduced because software is not described in-the '

Technical Specifications.

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g; tra'4 Safety Evaluations #90-24 and #90-25 Reactor Recirculation and Reactor Water Cleanup System Decontamination i

During the Unit 2 Refuel Outage decontamination of piping associated with the Reactor vessel was performed, the Reactor Water Cleanup Piping was performed with fuel in1the vessel and the vessel head removed. The decontamination chemicals did noti enter the v(ssel during this process.

'The Recirculation Pump Suction and Discharge Piping was also decontaminated.

This was done with the fuel removed from the vessel. The vessel head was in place but.not tensioned. Water level in the vessel was maintained below the core area of the vessel. The decontamination chemicals were flushed from the vessel prior to reloading fuel.

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 since. metallurgy effects are minimal because the solvent corrosion rates are less than the original allowances.

304 stainless steel coupons were placed in the decontamination flow path and analyzed upon completion of the project for assurance of the actual corrosion rates. Water purity effects are minimal because the reactor coolant were returned to a conductivity and a TOC level that is acceptable to station chemistry.

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 effects of residual solvent in the system was determined'to be negligible.

Reactor Coolant is cleaned and returned to a conductivity.and-a TOC level which is acceptable to the station chemistry staff.

Station radiation protection procedures were followed throughout the decontamination.

During resin transfer to the solidifi-cation truck, the affected areas of the reactor building was evacuated.

Access into the drywell during the proca.ss was strictly controlled by station. health physicists. The level of'the solvent in the recircu-latiun system risers and annulus was continuously monitored.

Since SMAD has reviewed the material / solvent interface for materials within.

the core and has accepted the solvent for use, the consequences of a failure in the level controls causing a spill into the core are negligible.

3.

The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because the decontamination project was performed in accordance with the existing Technical Specifications. The reactor 1

was maintained in the shutdown or refuel mode with all interlocks in the shutdown or refuel position.

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Safety Evaluation #90-27 Unit One and Unit Two Technical Specification Combination This safety evaluation combined Unit One and Two Technical Specifications.

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 proposed Tech Spec change does not deal with accident precursors or equipment-malfunctions and, therefore, cannot impact safety consequences as previously evaluated in the FSAR. This change is not a technical change, but is strictly administrative.

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 proposed Tech Spec change, being administra-tive in nature, does not create the possibility of an accident or malfunction of a different type because no technical change is proposed.

No equipment or system modification is involved.

3.

The margin of safety, as defined in the basis for any Technical Speci-fication, is not reduced because no technical change is proposed.

This proposal only' involves combining two nearly identical. documents into one to minimize the potential for confusion and to enhance control of the documents.

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n., %y :9l,'g hb Safety Evaluation #90-48 Core Monitoring Code SPR Fix Installation

A review'of thelthermal limits code and core monitoring code _was installed-to correct.the' cold-predictor and_ backup functions.

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The probability of an' occurrence or the consequence of'an accident,

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,o or malfunction of equipment important to safety as previouslyJevaluated

in the Final Safety Analysis Report is not increased because this b

~ = revision involves.a computer _ program which has:no functional control over any equipment and, therefore...cannot increase the probability p

of_an occurrence or the consequences of an_ accident.

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2.-~ The possibility lfor an-accident or malfunction of a different type-

than any previously evaluated in the Final Safety Analysis Report is L

_not created because the functions being changed provide no information b

to the' operator and the computer has no functional control _over any:

equipment.

1 3.

The' margin of= safety, as defined in the basis for any Technical Speci-

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~fication, is not reduced because this code does calculate MFLCPR which-gy:

is used to maintain'the proper margin to the MCPR safety limit, however, the-functions being changed have no impact on that calculation.

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