ML20086U436

TS Change Request 95-09-0 to Licenses NPF-39 & NPF-8, Revising Ts,By Deleting Operability & SR Involving Secondary Containment Diffrential Pressure Instrumentation

ML20086U436
Person / Time
Site:	Limerick
Issue date:	07/28/1995
From:	Hunger G PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20086U437	List: ML20086U436 ML20086U455
References
NUDOCS 9508040103
Download: ML20086U436 (8)

Text

, Ctation support Department -

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" 10 CFR 50.90 v

< PECO ENERGY 3llJ3 "%

965 Chesterbrook Boulevard Wayne. PA 19087-5691 July 28,1995_-

Docket Nos. 50-352 '

50-353 License Nos. NPF-39 '-

NPF-85 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

Limerick Generating Station, Units 1 and 2 Technical Specifications Change Request No. 9549-0 Secondary Containment Differential Pressure instrumentation j Gentlemen PECO Energy Company is submitting Technical Specifications (TS) Change Request No. 95-09-0 in accordance with 10 CFR 50.90, requesting amendments to the TS (i.e., Appendix A) of j Operating License Nos. NPF-39 and NPF-85 for Limerick Generating Station, Units 1 and 2,' i respectively. The proposed TS Changes, which are consistent with the improved Standard ' )

Technical Specifications (NUREG-1433), delete the operability and surveillance requirements involving secondary containment differential pressure instrumentation. .

Information supporting this TS Change Request is contained in Attachment 1 to this letter, and j marked up pages showing the proposed changes to the LGS, Units 1 and 2, TS are contained in  !

Attachment 2. This information is being submitted under affirmation, and the required affidavit is  !

enclosed.

We request that, if approved, the amendments to the LGS, Units 1 and 2. TS be issued prior to January 26,1996 and become effective within 30 days of issuance.

If you have any questions, please do not hesitate to contact us.

Very truly yours, G.. A. d.fu . .

Hunger, Jr.,gir or Licensing Section-Attachments Enclosure cc: T. T. Martin, Administrator, Region I, USNRC (w/ attachments, enclosure) '

N. S. Perry, USNRC Senior Resident inspector, LGS (w/ attachments, enclosure)

R. R. Janati, Director, PA Bureau of Radiation Protection, (w/ attachments,~ enclosure) 0-10034 c

9500040103 950728 PDR .ADDCK 05000352 PDR g0hl P .

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COMMONWEALTH OF PENNSYLVANIA  : .

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COUNTY OF CHESTER  :

W. H. Smith, Ill, being first duly sworn, deposes and says:

That he is Vice President of PECO Energy Company, the Applicant herein; that he has read the ~

foregoing Technical Specifications Change Request No. 95-09-0 for Limerick Generating Station, Units 1 and 2, to delete the operabHity and surveHlance requirements invcMng secondary -I containment differential pressure instrumentation, and knows the contents thereof; and that the statements and matters set forth therein are true and correct to the best of his knowledge,-

Information, and belief. ,

i Vice President Subscribed and sw to before me thi day 1995.

Sa f ( 'Sh f I-Notary Public ._ [ ,

Notadal Seal Erica S. Rosendz. Not Pubre Trecyffrin Twp., Chester My Commission Expires July 10.1999

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l ATTACHMENT 1 UMERICK GENERATING STATION UNITS 1 AND 2 .

l Docket Nos. 50-352 50-353 l

Ucense Nos. NPF-39 l NPF-85 1 Technical Specifications Change Request  ;

1 No. 95-09-0 l

" Secondary Containment Differential Pressure instrumentation" Supporting Information for Changes - 4 pages I

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I Docket Nos. 50-352 l l

50-353 Ucense Nos. NPF-39 NPF-85 Discussion and Descriotion of the Prooosed Chanoes PECO Energy Company, Ucensee under Facility Operating Ucense Nos. NPF-39 and NPF-85 for Umerick Generating Station, Units 1 and 2 respectively, requests that the Technical Specifications (TS) contained in Appendix A to the Operating Ucenses be amended as proposed herein, to delete the following operability and surveillance requirements involving secondary containment differential pressure instrumentation for Units 1 and 2:

Table 3.3.2-1 (6f) (7e) (7f)

Table 3.3.2-2 (6f) (7e) (7f)

Table 3.3.2-3 (6f) (7e) (7f)

Table 4.3.2.1-1 (6f) (7e) (7f)

Table 3.6.3-1 Penetration 025 Isolation Signals U,T Table 3.6.3-1 Penetration 026 isolation Signals U,T Table 3.6.3-1 Penetration 201 A isolation Signals U,T Table 3.6.3-1 Penetration 202 Isolation Signals U,T Table 3.6.5.2.1-1 Zone I (Unit 1) Isolation Signals U,T Table 3.6.5.2.1 1 Zone il (Unit 2) Isolation Signals U.T Table 3.6.5.2.2-1 Zone lit isolation Signals U,T PECO Energy also requests that Table Notation (c) of Table 3.3.2-1 be revised to delete reference to isolation signals 'U' and 'T' in support of the changes identified above.

The proposed changes are indicated as mark-ups on the following TS pages (attached) for Units 1 and 2:

3/4 3-14 3/4 3-15 3/4 3-17 3/4 3-21 3/4 3-22 3/4 3-25 3/4 3-26 i l

3/4 3-30 3/4 3-31 3/4 6-22 3/4 6-26 l 3/4 6-49 I 3/4 6-51 l 3/4 6-51a )

The design and operation of the secondary containment automatic isolation instrumentation is not changed by this activity.

We request that, if approved, the TS changes proposed herein be issued by January 26,1996 and become effective within 30 days of issuance. l This TS Change Request provides a discussion and description of the proposed TS changes, a safety assessment of the proposed TS changes, information supporting a finding of No l Significant Hazards Consideration, and Information Supporting an Environmental Assessment. '

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E, 7 l Docket Nos. 50-352 50-353 License Nos. NPF-39 NPF-85 Safety Assessment The containment isolation system is designed to prevent or limit the release of radioactive mraterials that may result from postulated accidents. This is accomplished by providing isolation barriers in all fluid lines that penetrate primary containment. Secondary containment is designed j to control fission products that may leak into the secondary containment following a LOCA or fuel handling accident by maintaining a negative pressure relative to the outside atmospheric )

pressure and filtering the fission products prior to release. Selected plant parameters such as ,

reactor vessel water level, primary containment pressure, system flows, room temperatures, safeguard bus voltages, and radiation levels, are monitored to identify conditions symptomatic of a design basis accident. When these plant parameters exceed pre-determined values, appropriate actions are Initiated to isolate primary containment, and to initiate the Standby Gas i Treatment System (SGTS) and Reactor Enclosure Recirculation System (RERS).  ;

A limited primary containment isolation signal and a secondary containment isolation signal are initiated by the Outside Atmosphere To Reactor Enclosure Delta Pressure-Low trip function if reactor enclosure pressure is less than or equal to a negative 0.1 inches water gage with respect to outside atmospheric pressure. A secondary containment isolation signal is also initiated by the Outside Atmosphere To Refueling Area Delta Pressure-Low trip function if refueling area pressure is less than or equal to a negative 0.1 inches water gage with respect to outside i atmospheric pressure. These trip functions are initiated in anticipation of a potential design ,

basis accident. Consequently, these trip functions do not result from a design basis accident  ;

and no credit is taken for the operation of these trip functions in any design basis accidents evaluated in the SAR. .

Since no credit has been taken for the operation of the Outside Atmosphere To Reactor Enclosure Delta Pressure-Low and the Outside Atmosphere To Refueling Area Delta Pressure- j Low trip functions in any design basis accidents evaluated in the SAR, and sufficient plant parameters are monitored to detect a design basis accident and initiate the appropriate actions, ,

it is not necessary to include Technical Specifications requirements for operability and surveillance of these trip functions. All other Technical Specifications operablity and survegiance requirements associated with primary containment isolation and secondary containment remain ,

unchanged. The design and operation of the secondary containment differential pressure automatic isolation instrumentation remains unchanged.

There is no equipment important to safety that is adversely affected by this change. This activity does not change the design, function, or operation of any plant components or safety-related ,

systems including the secondary containment differential pressure automatic isolation instrumentation, primary containment, secondary containment, SGTS, RERS, Reactor Enclosure

, HVAC, or Refueling Area HVAC. There are no changes to the separation, redundancy, ,

qualification, quality assurance or fire protection requirements for the associated components and systems, nor are there any new fature modes created. This activity only removes -!

operabRity and survegiance requirements from the Technical Specifications for selected plant ,

components associated with secondary containment differential pressure trip functions. The i secondary containment differential pressure instrumentation wil continue to initiate the  :

appropriate actions if reactor buuding/ refueling area differential pressure cannot be maintained.

During a Loss-of-Offsite-Power, normal HVAC will be lost resulting in a loss of reactor enclosure / refueling area differential pressure. If differential pressure cannot be restored within ]

the associated time delay, the appropriate containment isolation and SGTS and RERS initiation actions WH1 stRI occur.

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e DockQ Nos. 50-352 50-353 Ucense Nos. NPF-39 NPF-85 The proposed changes have been made in accordance with the phHosophies and requirements of the improved Standard Technical Specifications (ISTS), NUREG-1433, issued September 28, 1992, and do not alter equipment configuration or operation in any way.

Information Suncortina a Findina of No Stanificant Ha7ards Consideration We have concluded that the proposed changes to the IJmerick Generating Station (LGS), Units 1 and 2 Technical Specifications (TS) involving secondary containment differential pressure instrumentation do not involve a Significant Hazards Consideration. In support of this determination, an evaluation of each of the three (3) standards set forth in 10 CFR 50.92 is provided below.

1. The orocosed Technical Soecifications (TS) chances do not involve a slanificant increase in the orobability or consecuences of an accident oreviousiv evaluated.

Deleting the operabuity and surveillance requirements for the secondary containment differential pressure instrumentation does not involve any changes to the design, function, or operation of any plant components or safety-related systems. There are no changes to the separation, redundancy, qualification, quality assurance or fire protection requirements for the associated components and systems, nor are there any new faHure modes created. This activity only removes operabHity and surveillance requirements from the Technical Specifications for selected plant components associated with the secondary containment differential pressure trip functions. No credit for operation of these trip functions is taken in any design basis accidents evaluated in the SAR.

These components will be maintained in accordance with the plant preventive maintenance program. The failure of any of these components does not result in the occurrence of an accident. Consequently, there is no increase in the probability of occurrence of an accident previously evaluated in the SAR.

The Outside Atmosphere to Reactor Enclosure Delta Pressure-Low and Outside Atmosphere To Refueling Area Delta Pressure-Low trip functions are not ,

symptomatic of a design basis accident. No credit for operation of the trip functions is taken in any design basis accidents evaluated in the SAR. Neither  ;

failure of the differential pressure components nor failure to generate the i associated trip functions affects the consequences of an accident previously evaluated in the SAR. The appropriate accident prevention and mitigation actions are generated from other plant parameters symptomatic of an accident.

Sufficient plant parameters symptomatic of a design basis accident are monitored to initiate the appropriate actions as evaluated in the SAR.  !

Furthermore, all safety-related systems wn! stHI be able to perform all of their j design basis safety-related functions. Consequently, there is no increase in the l consequences of an accident previously evaluated in the SAR. ,

Therdore, the proposed TS changes do not involve an increase in the  !

probab31ty or consequences of an accident previously evaluated. I

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'~s,; 1; DoclC3 Nos. 50-352 -i 50-353 i j

License Nos. NPF NPF-85 '

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2. The orocomed TS changes do not create the possibikv of a new or different kind i of accident from any accident oreviously evaluated.  !

The faNure of the differential pressure automatic isolation instrumentation - l components does not result in the occurrence of an accident. The faBure to .

generate the associated talo im:tions does not result in the occurrence of an  :

accident. This activity does not hvolve any changes to the design, function, or l operatim. of any plant components or s=fety-related systems. There are no - -

changes to the separation, redundancy, quellfication, quality assurance or fire protection requirements for the associated components and systems. These ,

components will be maintained in accordance with the plant preventative 1 maintenance program. Consequently, there is no possibuky of an accident of a different type than previously evaluated in the SAR. l Therefore, the proposed TS changes do not create the possibuity of a new or .

different kind of accident from any accident previously evaluated. -j

3. The oronosed TS chanoes do not involve a sionificant reduction in a maroin of safety.

r The ability of secondary containment to minimize any ground level release of -

radioactive material which may result from any accident is not affected. 3 i

Survellance and operabRity requirements for secondary containment SGTS and RERG are not changed by this activity. Draw down time, leakage factors, secondary containtnent system ratings, and secondary containment system  ;

response to a LOCA or refueling accident are not affected by this activity. SGTS and RERS initiation wul continue to occur when plant parameters symptomatic  ;

of a LOCA or refueling accident exceed predetermined values. There are no  !

changes to the inputs for the post-LOCA offsite dose analysis. ,

J Therefore, the proposed TS changes do not involve a significant reduction in a  ;

margin of safety.

Information Suooorting an Environmental Assessment An Environmental Assessment is not required for the changes proposed by this TS Change  !

Request because the requested changes to the LGS, Units 1 and 2. TS conform to the criteria for " actions eligible for categorical exclusion," as specified in 10 CFR 51.22(c)(9). The requested y changes wBl have no impact on the envirnnment. The proposed changes do not involve a' Significant Hazards Consideration as discussed in the preceding section. The proposed changes do not involve a significant change in the types or significant increase in the amounts of any effluents that may be released offsite.= In addition, the proposed changes do not involve a i significant increase in individual or cumulative occupational radiation exposure.  ?

Conclusion The Plant Operations Review' Committee and the Nuclear Review Board have reviewed these proposed changes to the LGS, Units 1 and 2, TS and have concluded that they do not involve an unreviewed safety question, and wHI not endanger the health and safety of the public.

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i ATTACHMENT 2 ,

i UMERICK GENERATING STATION UNITS 1 AND 2 Docket Nos. 50-352 50-353 -

Ucense Nos. NPF-39 ,

NPF-85 TECHNICAL SPECIFICATIONS CHANGE REQUEST No. 95-09-0 List of Affected Pages UnN1 UnN2 3/4314- 3/4 3-14 3/4 3-15 3/4 3-15 3/4 3-17 3/4317 3/4 3-21 3/43-21 i

3/4 3-22 3/4 3-22 ,

3/4 3-25 3/43-25 3/4 3-26 3/4 3-26 3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 6-22 3/4 6-22 3/4 6-26 3/4 6-26

'- 3/4 6-49 3/4 6-49 3/4 6-51 ' 3/4 6-51 3/4 6-51a 3/46-51a

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