ML20209J232

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Proposed Tech Specs 3/4.7.D Replacing Limit for Any One Msli Valve of Less than or Equal 11.5 Sfch with Aggregate Value of Less than or Equal 46 Scfh for All MSIVs
ML20209J232
Person / Time
Site: Quad Cities  Constellation icon.png
Issue date: 07/16/1999
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20209J213 List:
References
NUDOCS 9907210152
Download: ML20209J232 (7)
v • d • e


Text

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. ATTACHMENT D, Pr posed Ch:nge ta TechnicIl Specific;ti:n f r Qusd Citico Nuclear Power Station Units 1 and 2, Page 1 of 2 REVISED PAGES 3/4.7-7 B 3/4.7-2 9907210152 990716 l PDR ADOCK 05000254  !

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CONTAINMENT SYSTEMS

. PCIVs 3/4.7.D 3.7 - LIMITING CONDITIONS FOR OPERATION 4.7 - SURVEILLANCE REQUIREMENTS

2. With one or more reactor '
a. At least once per 31 days by instrumentation line excess flow check verifying the continuity of the valves inoperable, operation may continue and the provisions of explosive charge.

Specification 3.0.C are not applicable, b.

provided that within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either: At least once per 18 months by removing at least one explosive

a. The inoperable valve is restored to squib from an explosive valve such OPERABLE status, or that each explosive squib will be tested at least once per 90 months,
b. and initiating the removed The instrument line is isolated and explosive squib (s). The the associated instrument is declared inoperable. replacement charge for the exploded squib (s) shall be from the Otherwise, be in at least HOT same manufactured batch as the SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> one fired or from another batch and in COLD SHUTDOWN within the which has been certified by having 4

following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. at least one of that b'atch successfully fired. No squib shall remain in use beyond the expiration of its shelf-life or operating life, as applicable.

6.[A quency specified by the Primary Containment Leakage Rate Testing Program, verify leakage for any one main steam line isolation valve when tested at P, (25 psig) is s11.5 scfh.

I M5e c E QUAD CITIES - UNITS 1 & 2 3/4.7-7 Amendment Nos. 171 a 167

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CONTAINMENT SYSTEMS B 3/4.?

BASES leakage tests). The acceptance criteria were established i during init al air lock and primary containment OPERABILITY testing. The periodic testing requirements verify that air lock leak does not exceed the allowed fraction of the overall primary containment leakage rate. The Frequency is required by the Primary Containment Leakage Rate Testing Program. The surveillance requirements have been annotated such that an inoperable air lock door does not

-invalidate the previous successful performance of the overall air lock leakage test. This is considered reasonable since either air lock door is capable of providing a fission product barrier in the event of a DBA. Additional annotation is provided to require the results of air lock leak tests being evaluated against the acceptance criteria applicable to the surveillance requirements .

This ensures that the air lock leakage is properly accounted for in determining the com B and Type C primary containment leakage.

3/4.7.D Primary Conts:r.rr.;r.: laa!= tion Valves The OPERABILITY of the primary containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radicactive material to the containment atmosphere or pressurization of the containment. Containnut isolation within the time limits specified for those isolation valves designed to close auto ensures that assumptions used the in therelease of aradioactive analyses for LOCA. material to the environment will be consistent The containment is also penetrated by a large number of small diameter instrument lines which contact the primary coolant system. A program for periodic testing and examination of the flow check valves in these lines is performed by blowing down the instrument line during an inservice leak or hydrostatic test and observing conditions which verify that the flow check valve is operable, that e.g., a to quickly reduces distinctive a slight trickle. 'cuck' when the poppet valve seats, or an instrumentation high flo The main steam line isolation valves are tested at lower pressures, per an approved exempt ,

but the leakage rate is included in the Type B and C test totals. The surveillance testing for measuring leakage rates is consistent with the requirements of Appendix J of 10CFR Part 50 with the exception of approved e .emptions. (Ref: Exemption Request Approval, Mr. D. B.

Vassallo (NRC) to Mr. D. L. Farrar (CECO) dated12,1984.) June 3/4.7.E Snaaremaion Charr.her - Drv ::ll Vaennm Breakers M'I The function of the suppression chamber to drywell vacuum breakers is to relieve vacuum in t drywell. These internal vacuum breakers allow air and steam flow from the suppression chamb to the drywell when the drywell is at a negative pressure with respect to the suppression chamber. Each vacuum breaker is a self actuating valve, similar to a check valve.

The safety analysis assumes that the intemel vacuum breakers are closed initially and open at a differential pressure of 0.5 psid. Additionally, three of thue internal vacuum breakers QUAD UNITS 1 & 2 By NRC Letter dated October 5,1998 B 3/4.7 2

, ATTACHMENT O, Proposed Change t3 Technic:1 Specift:ati:n for Quad Cities Nuclear Power Station Units 1 and 2, Page 2 of 2 INSERT IN PLACE OF THE CURREN14.7.D.6:

6. In accordance with the methods and at the frequency specified by the Primary Containment Leakage Rate Testing Program, verify total maximum pathway leakage for all main steam isolation valves (MSIVs) is s 46 scfh when tested at Pi (25 psig).

INSERT AT THE END OF BASES SECTION 3/4.7.D:

The individual main steam isolation valve (MSIV) leakage limit has been replaced by the aggregate leakage limit of 5 46 scfh for all MSIVs. The leakage will be determ;ned for the maximum pathway leakage in accordance with the Primary Containment Leakage Rate Testing Program. This is a very conservative total for MSIV leakage because it takes the MSIV with the maximum leakage in each steam line and sums the leakage for each of those valves to determine the maximum pathway leakage.

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ATTACHMENT C, Proposed Change t3 Technic l Specificati:n f r Quad Citi;s Nuclear Power Station Units 1 and 2, Page 1 of 2 Comed has evaluated thic proposed amendment for Quad Cities Nuclear Power Station, Units 1 and 2 and 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 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.

Comed proposes to amend Appendix A, Technical Specification, of Facility Operating Licenses DRP-29, DPR-30. The determination that the criteria set forth in 10 CFR 50.92 are met for this amendment request is indicated below:

Does the change involve a significant increase in the probability or consequenc3s of an accident previously evaluated?

The proposed changes to the Technical Specifications, Appendix A, modifies the allowed leakage limit to an aggregate value with no change to the total allowed leakage rate. This change does not affect either the automatic or manual features that would close the MSIVs.

Thare are no physical changes to the plant and plant operations remain unchanged.

Therefore, this proposed amendment does not involve a significant increase in the probability or consequences of an accident previo'isly evaluated.

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

The safety function of the MSIVs is to provide a timely steam line isolation to mitigate the release of radioactive steam and limit reactor inventory loss under certain accident and transient conditions. The MSIVs are designed to automatically close whenever plant conditions warrant main steam line isolation. Changing the leakage limits to include an

! aggregate value does not affect the isolation function. No new equipment will be installed or utilized, and no new operating conditions will be initiated as a result of this change.

Therefore, the proposed change does not create the possibility of a new or different kind of  ;

accident from any previously evaluated.

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, ATTACHMENT C, Proposed Ch:nge 13 Technical Specific tion far Quad Citi;3 Nuclear Power Station Units 1 and 2, Page 2 of 2 Does the change involve a significant reduction in a margin of safety?

The total allowed leakage rate for all MSIVs remains unchanged at 46 scfh. Therefore, there will be no change in the types or significant increase in the amounts of any effluents released offsite, and, thus, the radiological analyses remain unchanged and within the guidelines of 10 CFR 100 and General Design Criteria 19. Therefore, these changes do not involve a significant reduction in the margin of safety.

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

, ATTACHMENT D, Proposed Change 13 Technicil Specifl:ati:n f :r Qu:d Citi;s Nuclear Power Station Units 1 and 2, Page 1 of 1 Comed has evaluated this proposed operating license amendment request against the criteria for identification of licensing and regulatory actions requiring environmental assessment in 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 significant increase in the amounts, and no significant change in the types, of any effluents released offsite.

(iii) There is no significs nt increase in individual or cumulative occupational radiation exposure.

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 increase in individual or cumulative occupational radiation exposure resulting from this change.

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