ML20205J991

Proposed Tech Specs Allowing Alternative Methodology for Quantifying RCS Leakage When Normal RCS Leakage Detection Sys Is Inoperable

ML20205J991
Person / Time
Site:	Quad Cities
Issue date:	03/30/1999
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20205J990	List: ML20205J991 SVP-99-038, Application for Amends to Licenses DPR-29 & DPR-30,allowing Alternate Methodology for Quantifying RCS Leakage When Normal RCS Leakage Detection Sys Is Inoperable
References
NUDOCS 9904120285
Download: ML20205J991 (6)

Text

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ATTACHMENT B Marked Up Technical Specifications Pages 1 SVP-99-038 (Page1of1) 3/4.6-10 B 3/4.6-4 J

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PRIMARY SYSTEM BOUNDARY Lcakage Detection 3/4.6.G 3.6 - LIMITING CONDITIONS FOR OPERATION 4.6 - SURVEILLANCE REQUIREMENTS G. Leak' age Detection Systems G. Leakage Detection Systems The following reactor coolant system The reactor coolant system leakage leakage detection systems shall be detection systems shall be demonstrated OPERABLE: OPERABLE by:

1. The primary containment atmosphere 1. Performing the leakage determinations porticulate radioactivity sampling of Specification 4.6.H.

system, and

2. Performing a CHANNEL CAllBRATION

2. The drywell floor drain sump d e APPLICABILITY: '

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OPERATIONAL MODE (s) 1, 2 and 3. h Icad Astre e W$on 4 yee er 18 t*oN4

ACTION: _

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1. With the primary containment atmosphere particulate radioactivity sampling system inoperable, restore the inoperable leak detection radioactivity sampling system to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />; otherwise, be in HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the l following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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2. With the drywell floor drain sumpJ ~

s stem inoperable,(rsst a -

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rjrnpih sufntVsvit tGsfwithin o [j hourct,sttherwise, be in at least HO SHUTDOWN within the

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next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> nd in COLD / gUj gg,,, argin 5,q SHUTDOW within the following . w, S ( I s r be. y?U mo44orsrg

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QUAD CITIES - UNITS 1 & 2 3/4.6-10 Amendment Nos.

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F PRIMARY o. TEM BOUNDARY B 3/4.6  ;

w-- l BASES hhe' Leakage from the reactor coolant pressure boundary insid the drywell is detected by at least one or two independently monitored variables, such as sum evel changes and drywell atmosphere i radioactivity levels. The means of quantifying leakage in the drywell is the _drywell floor dr '

sump pumps. With the drvwell floor drain sumo .ystem inoperable 9/pf 7gMfprgi(

(Mo#itArAg'ghp dg(4d fy(g(u/idlWidf6stnMdsfdhlbdeGd/ondsfi/ containment atmannhere l sampling for radioactivity can provide indication of channes in leakage rates.MaWeen.Q men

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3/4.U.H unerational Leakaos The allowable leakage rates from the reactor coolant system have been based on the predicted and experimentally observed behavior of cracks in pipes. The normally expected background leakage due to equipment design and the detection capability of the instrumentation for determining system leakage was also considered. The evidence obtained from experiments suggests that for leakage somewhat greater than that specified for UNIDENTIFIED LEAKAGE the probability is small that the imperfection or crack associated with such leakage would grow rapidly. However, in all cases, if the leakage rates exceed the values specified or the leakage is located and known 'to be PRESSURE BOUNDARY LEAKAGE, the reactor will be shutdown to sidow further investigation and corrective action.

An UNIDENTIFIED LEAKAGE increase of more than 2 gpm within a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is an indication of a potential flaw in the reactor coolant pressure boundary and must be quickly evaluated.

Although the increase does not necessarily violate the absolute UNIDENTIFIED LEAKAGE limit, IGSCC susceptible components must be determi~ui not to be the source of the leakage within the required completion time.

l 3/4.6.1 Chemistry The water chemistry limits of the reactor coolant system are established to prevent damage to the reactor materials in contact with the coolant. Chloride limits are specified to prevent stress corrosion cracking of the stainless steel. The effect of chloride is not as great when the oxygen 4 concentration in the coolant is low, thus the 0.2 ppm limit on chlorides is permitted dunne POWER OPERATION.

Conductivity measurements are required on a continuous basis since changes in this parameter are an indication of abnormal conditions. When the conductivity is within limits, the pH, chlorides and other impurities affecting conductivity must also be within their acceptable limits. With the conductivity meter inoperable, additional samples must be analyzed to ensure that the chloridos are not exceeding the limits.

Action 1 permits temporary operation with chemistry limits outside of the limits required in OPERATIONAL MODE 1 without requiring Commission notification. The surveillance requirements provide adequate assurance that cnneentrations in excess of the limits will be detected in sufficient time to take corrective action.

QUAD CITIES - UNITS 1 & 2 B 3/4.6 4 Amendment Nos. ma

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ATTACHMENT C Information Supporting a Finding of No Significant Hazards SVP-99-038 (Page 1 of 2)

Comed has evaluated this proposed amendment for Quad Cities Station, Units 1 and 2, and has determined that it involves no significant hazards consideration. According to 10 CFR 50.92(c),

a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

Involve a significant increase in the probability of occurrence or consequences of an accident l previously evaluated; Create the possibility of a new or different kind of accident from any previously analyzed; or Involve a significant reduction in a margin of safety.

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Pursuant to 10 CFR 50.90, Comed proposes to change Appendix A, TS Section 3/4.6.0, Leakage Detection Systems, for Facility Operating Licenses DPR-29 and DPR-30. The purpose of this proposed change is to allow an alternate methodology for quantifying reactor coolant system (RCS) leakage when the normal monitoring system is inoperable. This change is consistent with NRC guidance provided in letter L. Olshan (USNRC) to T. Kovach (Comed) dated August 21,1990, concerning Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping."

The determination that the criteria set forth in 10 CFR 50.92 are met for this amendment request is indicated below: l Does the change involve a significant increase in the probability or consequences of an l accident previously evaluated?

The current Technical Specifications require a periodic measurement of RCS leakage. The ,

normal method for quantifying RCS leakage is to use the DWFDS and DWEDS flow totalizers.  !

The proposed TS change would allow an attemate method for quantifying RCS leakage when a flow tota:izeris not available. The proposed change has no impact on the frequency for monitoring RCS leakage and would only be used for a maximum of 30 days while the normal leakage monitoring system is being restored to an operable condition. The alternate methodology for quantifying leakage has a measurement sensitivity that is consistent with the normal method. The proposed change does not impact any system structure or component used I to mitigate the consequences of an accident and there will be no change in the types or j significant merease in the amounts of any effluents released offsite.

Therefore this proposed amendment does not involve a significant increase in the probability or

. consequences of an accident previously evaluated.

ATTACllMENT C Information Supporting a Finding of No Significant flazards

( SVP-99-038 (Page 2 of 2)

Does the change create the possibility of a new or difrerent kind of accident from any accident previously evaluated?  !

The proposed change involves no physical modifications to any system, stmeture or component used to mitigate the consequences of an accident. The operation of the DWEDS and DWFDS are not being altered in any way that could affect their ability to function during an accident condition.

l Therefore, the proposed changes do not create the possibility of a new or different kind of l accident from any previously evaluated.

l Does the change involve a significant reduction in a margin of safety?

The current TS's require a periodic measurement of RCS leakage. The normal method for l quantifying RCS leakage is to use the DWFDS and DWEDS flow totalizers. The proposed technical specifications change would allow an a' ternate method for quantifying RCS leakage when a flow totalizer is inoperable. The proposed change has no impact on the frequency for

! monitoring RCS leakage and would only be used for a maximum of 30-days while the normal leakage monitoring system is being restored to an operable condition. The proposed alternate methodology for quantifying leakage has a measurement sensitivity that is consistent with the

normal method.

l Therefore, these changes do not involve a significant reduction in the margin of safety.

l Therefore, based upon the above evaluation, Comed has concluded that these changes involve no significant hazards consideration.

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l ATTACHMENT D Information Supporting an Environmental Assessment SVP-99-038 l (Page 1 of 1) l l

Comed has evaluated this proposed operating license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in l accordance with 10 CFR 51.21, Comed has determined that this proposed license amendment request meets the criteria for a categorical exclusion set forth in 10 CFR 51.22(c)(9) and as such, has determined that no irreversible consequences exist in accordance with 10 CFR 50.92(b).

This determination is based on the fact that this change is being proposed as an amendment to a license issued pursuant to 10 CFR 50 that changes a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, or that changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:

(i) the amendment involves no significant hazards consideration.

As demonstrated in Attachment C, this proposed amendment does not involve any significant hazards consideration.

(ii) there is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

As documented in Attachment C, there will be no change in the types or significant increase in the amounts of any effluents released offsite.

(iii) there is no significant increase in individual or cumulative occupational radiation exposure.

The proposed changes will not result in changes in the operation or configuration che facility. There will be no change in the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the -

proposal result in any change in the normal radiation levels within the plant. Therefore, there will be no significant increase in individual or cumulative occupational radiation exposure resulting from this change.

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