Text

Forwards Response to RAIs Re DAEC Request to Convert TSs to Improved Tss,Per NUREG-1433
	ML20203F110
Person / Time
Site:	Duane Arnold
Issue date:	12/08/1997
From:	Franz J; IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:	; NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	RTR-NUREG-1433 NG-97-2097, NUDOCS 9712170288
	Download: ML20203F110 (250)
	v • d • e

IES Untros Inc.

200 Fust Stmet SE.

Ro. Bat 351 Cedar Rweds, IA 52406-0351 Telephone 319 398 8162 Fax 319 398 8192 UTILITIES ,,,,,,,,,

\4ce President, Nuclear l December 8,1997 NG 97-2097 i Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station PI-37 Washington, DC 20555-0001

- Sulject: Duane Arnold Energy Center (DAEC)

Docket No: 50-331 Op. License No: Di'R-49 Response to NRC Requests for Additional Information on the DAEC Improved Technical Specifications

References:

1) J. Franz (lES) to F. Miraglia (NRC)," Submittal of License Amendment Request to Convert the DAEC Technical Specifications to the improved Technical Specifications (NUREG-1433), (RTS-291)," NG-96-2322, October 30,1996.

2) Letter, G. Kelly (NRC) to L. Liu (IES), dated August 18,1997, " Request for Additional Information on the DAEC Improved Technical Specifications (TAC No. M97197)."

3) Letter, G. Kelly (NRC) to L. Liu (IES), dated September 8,1997,

" Request for Additional Information (RAl) on the DAEC Improved Technical Specifications (ITS) (TAC No. M97197)."

4) Letter, G. Kelly (NRC) to L. Liu (IES). dated October 2,1997, " Request for Additional Information on the DAEC Improved Technical Specifications (ITS)(TAC No M97197)."

5) Letter, K. Peveler (lES) to U.S. NRC," Partial Response to NRC Request for Additional InformaCon," NG-97-1598, September 17,1997.

I File: A-117, SPF-167 9712170298 971200 PDR ADOCK 05000331 P PDR

$ J An IES Industnes Concany

l De: ember 8,1997 NG 97 2097 Page 2 of 3 .

i Dear Sir (s).

1 In Reference 1, IES Utilities requested a conversion of the DAEC Technical Specifications to the Standard Technical Speci6 cations (NUREG-1433). In References 2, 3 and 4, the NRC transmitted Requests for Additional Infonnation (RAI) on selected Sections of the Reference 1 submittal. The l'rst enclosure to this letter contains our complete Response to the Questions on Section 3.6. This is not a complete response to the Referenced RAls, per our Reference 5 agreement.

Similar to our previous submittals ofImproved Technical Specification (ITS) RAI Responses, revisions to the Discussion of Changes (DOCS), Justification for Deviations (JFDs) and No Significant llazards Considerations (NSHC) have been annotated with revision bars and the NRC Question number. All of the DOC, JFD and NSilC pages in the affected Section have been replaced with new pages that are marked as Revision ti, to distinguish them from those made in the previous submittals.

In our previous ITS RAI Response submittals, we provided the Staff with a summary matrix showing the proposed final dertination of those items categorized as " Relocated Items" in the original submittal (Ref.1). As a number of our Responses in Enclosure 1 to this letter reference that matrix, we have enclosed another copy for the Staff s convenience (Enclosure 2). No changes to that matrix have been made.

The first enclosure to this letter provides the Response to all of the Staff's Questions in Sections 3.6. We are ready to discuss this Section with the Staff, at mutual convenience.

Responses to the Questions on the remaining Section 3.3 -Instrumentation, are under final development. Per the Staffs facsimile of December 4,1997, which transmitted the Staf"s priority issues on Section 3.3, we will respond to those issues first, with the remainder of our Responses to the StalTs Questions on 3.3 to follow in a subsequent transmittal. The next revision to the ITS and Bases (Rev. C) is also under development and will be submitted right after the first of the year.

Sincerely,

[

Kenneth E. Peveler hhmager, Regulatory Perfbrmance

Enclosures:

1) IES Responses to NRC Questions on the DAEC ITS Conversion

2) Relocated items Matrix lbr the DAEC Improved Technical Specifications

Decemocr 8,1997 NG-97-2097 Page 3 of 3 cc: T.13rowning/K. Putnam L. Root

J. Franz

D. Wilson

G. Kelly (NRC-NIUt)

A. B. Beach (Region Ill)

NRC Resident Office

w/o Enclosures l

l

Enclosure I to NG 97-2097 1 1

I i

IES Responses to NRC Questions on the DAEC ITS Conversion

DAEC ITS 3.6.1.1 PRIMARY CONTAINMENT ITEM - DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSE D

3.6.1.1-1 A.4 CTS 15.b CTS 15.b requires that at least one 7/8/97 Provide additional door in each airlock is closed and discussion and sealed. No change is associated with justification for this requirement, yet it is marked A.4. l the administrative A.4 changes " PRIMARY change CONTAINMENT INTEGRITY shall be associated with maintained" to " Primary containment the airbck docr.

shall be OPERABLE." This has nothing See item to do with the airlock door. See item Numbers 3.6.1.1-Numbers 3.6.1.1-2 and 3.6.1.2-1. ; 2 and 3.6.1.2-1.

DAEC RESPONSE: The CTS markup will be revised to associate CTS 15.b on airlocks with ITS 3.6.1.2 and the change will be reclassified Response to NRCas a Relocated Question detail to the Bases. The revised page will be included in the next ITS revision submittal. (

3.6.1.2-1).

1

DAEC ITS 3.6.1.1 PRIMARY CONTAINMENT ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS -

NO. JFD LCO OPENED CLOSE D

3.6.1.1-2 A.4 CTS 15. CTS 15 " PRIMARY CONTAINMENT 7/8/97 Revise the A.8 ITS B3.6.1.1 INTEGRITY" definition is divided markup to show R.1 Bases- into three parts in the CTS markup. that definition BACKGROU The first part is associated with ITS 1.0 CTS 15 ND and the char.ge is designated A.8 " PRIMARY which deletes the definition from TS. CONTAINMENT The second part is associated with INTEGRITY" is containment isolation valves (CTS 15.a) being relocated in and the change is designated R.1 which its entirety to the states tnat the majority of CTS 15.a is Bases of ITS relocated to the Bases. Since this R.1 B3.6.1.1.

is in the justifications for ITS 3.6.1.3 it Provide additional is assumed it is relocated to ITS discussion and B3.6.1.3, which it is not. The third justification for part is associated with airlock doors this Less-and manways (CTS 15b and c) and the Restrictive (LA) changes are designated A.4 and R.1. change. See item For A.4 see item Number 3.6.1.1-1. Numbers 3.6.1.2-R.1 relocated CTS 15.c to ITS B3.6.1.1 1 and 3.6.1.3-4.

Bases-BACKGROUND. The markup is incorrect. The entire definition for PRIMARY CONTAINMENT INTEGRITY is moved to ITS B3.6.1.1 Bases-BACKGROUND. in addition, the staff considers this a Less Restrictive change (LA) rather than a Relocation (R), which is reserved for the movement of whole specifications. See item Numbers 3.6.1.2-1 and 3.6.1.3-4.

2

DAEC ITS 3.6.1.1 PRIMARY CONTAINMENT.

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO_ OPENED' CLOSE-D '

DAEC RESPONSE: The information reiocated by R.1 to ITS 3.6.1.1 only refers to blind flanges and raceways and is -

contained in item c. in the Background section of Bases to 3.6.1.1. The information relocated by R.1 to ITS 3.6.1.3 can be ,

found in the second paragraph in the Background section of Bases to 3.6.1.3. As stated in our Response to Question J 3.6.1.1-1, the CTS mark-up will be revised to denote item 15.b is relocated to the Bases of 3.6.1.2. The relocation of this

information to the Bases is acceptable because the CTS definition of Primary Containment integrity contains OPERABILITY requirements fev the individual components that make up the primary containment boundary. This type of descriptive -

information d(es not belong in the Specification and can be relocated to the Bares, where it wi!! be controlled under the Bases Control Program of ITS Section 5.0. The CTS definition was marked up to denote where the individual items went i with respect to their corresponding LCOs. The existing markup is not incorrect and the proposed revision is merely another acceptable alternative and thus, will not be adopted. In addition, the DAEC ITS i bmittal denotes all relocated information with R-changes and does not use the LA-type designation. This has been discussed with the NRC and " Relocated items Matrix (attached) has been used to provide the proposed location of the relocated items.

3.6.1.1-3 LCT.2 CTS 4.7.E.4 CTS 4.7.E.4 requires a leak test of the 7/8/97 Delete this Bases ITS SR drywell to suppression chamber change.

P.8 3.6.1.1.2 structure every operating cycle (18 :

ITS B3.6.1.1 months). ITS SR 3.6.1.1.2 requires a l Bases- leak test every 24 months. This i SR extension of the surveillance frequency !

3.6.1.1.2 is considered by the staff as a beyond scope of review item for this !

conversion. i !

DAEC RESPONSE: Based upon our meeting with the NRC Staff on Se a. amber 10, 997, we understand that the technical !

resources have been applied to review our conversion from an 18 m . % to 24 month operating cycle in parallel with our [

ITS conversion. Consequently, this item is considered to be back witnia scopc for the purposes of our Final Safety

Evaluation on the ITS conversion. [

t f

l 3 ,

DAEC ITS 3.6.1.1 PRIMARY CONTAINMENT ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSE D

3.6.1.1-4 P.54 ITS SR CTS 4.7.A.1.aa requires leak rate 7/8/07 Licensee to Bases 3.6.1.1.1 testing in accordance with the Primary update submittal P.18 and Containment Leakage Rate Testing with regards to Ar,sociated Program. STS 3R 3.6.1.1.1 requires 11/2/95 letter Bases the visual examination and leakage rate and updated testing be performed in accordance TSTF 52 when with 10 CFR 50 Appendix J as OG provides modified by approved exemptions. ITS revision or SR 3.6.1.1.1 modifies STS SR provide additional 3.6.1.1.1 to conform to CTS justification for 4.7.A.1.A. The STS is based on deviations.

Appendix J Option A while the CTS /ITS are based on Appendix J, Option B.

Changes to the STS with regards to ,

Option A versus Option B are covered by a letter from Mr. Christopher I.

Grimes to Mr. David J. Medeen, NEl dated 11/2/95 and TSTF 32. The ITS changes are not in conformance with the letter or TSTF 52 as modified by staff comments.

DAEC RESPONSE: Option B to Appendix J was incorporated into the DAEC CTS by Amendment 219. The changes are in accordance with the letter from Grimes to Modeen (Nov. 2,1995) referencec in the NRC question. As committed in our original submittal letter (NG-96-2322), new CTS markup pages, based upon Amendment 219, will be provided in our next revision of the Specifications and Bases.

4

DAEC ITS 3.6.1.1 PRIMARY CONTAINMENT ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS -

NO. JFD LCO OPENED CLOSE D

3.6.1.1-5 Bases STS STS B3.6.1.1 Bases-SR 3.6.1.1.2 7/8/97 Delete this P.1 B3.6.1.1 states the following: "This SR change.

Bases- measures drywell to suppression SR chamber differential pressure...to 3.6.1.1.2 ensure that the leakage paths...are ITS 83.6.1.1 within allowable limits." iTS B3.6.1.1 Bases- Bases SR 3.6.1.1.2 changes this SR statement as follows: "This SR 3.6.1.1.2 maintains drywell to suppression chamber differential pressure...to measure and ensure that the leakage paths... The justification used (Bases P.1) is for plant specific nomenclature, system description, etc. This justification does not apply to the changes made. In addition, this SR does not maintain a differential pressure but measures the increase in pressure to determine the leakage rate.

Therefore, the change is unacceptable.

DAEC RESPONSE: As discussed in JFD P.6 to SR 3.6.1.1.2, the DAEC design does not permit the measurement of differential pressure between the suppression pool and drywell. Our SR will only i,easure the pressure on one side of the boundary and look for increases, which indicate leakage. The Bases were clarified to match the change in the SR. Either JFD P.1 (plant-specific information) or P.8 (changes to match changes in the Specifications) are acceptable.

1 5

DAEC ITS 3.6.1.1 PRIMARY CONTAINMENT ITEM DOC /- CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO ' OPENED CLOSE D

3.6.1.1-6 Bases STS STS B3.6.1.1 Basec-SR 3.6.1.1.1 7/8/97 See item Number P.8 B3.6.1.1 references a number of SRs in ITS 3.6.1.3 4.

Bases-SR 3.6.1.3. ITS B3.6.1.1 Bases SR 3.6.1.1.1 3.6.1.1.1 changes these numbers to ITS B3.6.1.1 reflect the deletion of a number of SRs Bases-SR in ITS 3.6.1.3. These changes are 3.6.1.1.1 dependent on resolution of item Number 3.6.1.3-4.

DAEC RESPONSE: See Reponse to Question 3.6.1.3-4. No change in numbering of the ITS is planned.

6

DAEC ITS 3.6.1.1 PRIMARY CONTAINMENT .

ITEM- ' DOC / CTS /STS DESCRIPTION OF ISSUE- DATE DATE. COMMENTS NO.' JFD LCO~ OPENED CLOSE D'

3.6.1.1 A.'4 CTS 15.b CTS 15.b requires that at least one 7/8/97 Provide additional -

door in each airlock is closed and discussion and -

sealed. No change is associated with justification for this requirement, yet it is marked A.4. the administrative -

A.4 changes " PRIMARY change.-

CONTAINMENT INTEGRITY shall be associated with maintained" to " Primary containment . the airlock door.

shall be OPERABLE." This has nothing See item to do with the air!ock door. See item Numbers 3.6.1.1-Numbers 3.6.1.1-2 and 3.6.1.2-1. 2 and 3.6.1.2-1.

DAEC RESPONSE: The CTS markup wC) be revised to associate CTS 15.b on airlocks with ITS 3.6.1.2 and the change will 'i

. be recirsified as a Relocated detail to the Bases. The revised page will be included in the next ITS revision submittal.- (See Response to NRC Question 3.6.1.2-1). '

t i

t t

1

DAEC ITS 3.6.1.1 PRIMARY CONTAINMENT ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE- COMMENTS NO. JFD LCO OPENED CLOSE l D

3.6.1.1-2 A.4 CTS '15. CTS 15 "PR' MARY CONTAINMENT 7/8/97 Revise the A.8 ITS B3.6.1.1 INTEGRITY" definition is divided markup to show R.1 Bases- into three parts in the CTS markup. that definition BACKGROU The first pstt is associated with ITS 1.0 CTS 15 ND and the change is designated A.8 " PRIMARY which deletes the definition frnm T3. CONTAINMENT The second part is associated with INTEGRITY" is containment isolation valves (CTS 15.a) being relocated in and the change is designated R.1 which its entirety to the states that the majority of CTS 15.a is Bases cf ITS relocated to the Bases. Since this R.1 B3.6.1.1.

is in the justifications for ITS 3.6.1.3 it Provide additional is assumed it is relocated to ITS discussion and B3.6.1.3, which it is not. The third justification for part is associated with airlock doors this Less and manways (CTS 15b and c) and the Restrictive (LA) changes are designated A.4 and R.1. change. See item For A.4 see item Number 3.6.1.1-1. Numbers 3.6.1.2-R.1 relocated CTS 15.c to ITS 83.6.1.1 1 and 3.6.1.3-4.

Bases-BACKGROUND. The markup is incorrect. The entire definition for P'IMARY CONTAINMENT INTEGRITY is moved to ITS B3.6.1.1 Bases-BACKGROUND. In addition, the staff considers this a Less Restrictive change (LA) rather than a Relocation (R), which is reserved for the movement of whole specifications. See item Numbers 3.6.1.2-1 and 3.6.1.3-4.

E

DAEC ITS 3.6.1.1 PRIWeY CONTAINMENT ITEM DOC / CTS lSTS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSE D

DAEC RESPONSE: The information relocated by R.1 to ITS 3.6.1.1 only refers to blind flanges and raceways and is contained in item c. in the Background section of Bases to 3.6.1.1. The information relocated by R.1 to ITS 3.6.1.3 can be ,

found in the second paragraph in the Background section of Bases to 3.6.1.3. As stated ir. Our 36sponse to Question 3.6.1.1-1, the CTS mark op will be revised to denote item 15.b is relocated to the Bases of 3.6.1.2. The relocation of this information to the Bases is acceptable because the CTS definition of Primary Containment integrity contains OPERABILITY requirements for the individual components that make up the primary containment boundary. This type of descriptive information does not belong in the Specification and can be relocated to the Bases, where it will be controlled under the Bases Control Program of ITS Section 5.0. The CTS definition was marked up to denote where the individual items went with respect to their corresponding LCOs. The existing markup is not incorrect and the proposed revision is merely another acceptable attemative and thus, will not be adopted. In addition, the DAEC ITS submittal denotes all relocated information with R-enanges and does not use the LA-type designation. This has been discussed with the NRC and " Relocated items Matrix (attached) has been used to provide the proposed location of the relocated items.

3.6.1.1-3 LCT.2 CTS 4.7.E.4 CTS 4.7.E.4 requires a leak test of the 7/8/97 Delete this Bases ITS SR drywell to suppression chamber change.

P.8 3.6.1.1.2 structure every operating cycle (18 ITS B3.6.1.1 months). ITS SR 3.6.1.1.2 requires a Bases- leak test every 24 months. This SR extension of the surveillance frequency 3.6.1.1.2 is considered by the staff as a beyond 4 scope of review item for this ',

convers an.

DAEC RESPONSE: Based upon our meeting with the NRC Staff on September 10,1997, we understand that the technical resources have been applied to review our conversion from an 18 month to 24 month operating cycle in parallel with our ITS conversion. Consequently, this item is considered to be back within scope for the purposes of our Final Safety Evaluation on the ITS conversion.

3

DAEC ITS 3.6.1.1 PRIMARY CONTAihMENT ITEM- DOC / : CTS /STS DESCRIPTION OF ISSUE DATE DATE: COMMENTS NO. JFD LCO- OPENED. CLOSE D_ ,

E 3.6.1.1-4 P.54 ITS SR CTS 4.7.A.1.aa requires leak rate - 7/8/97- Licensee to-Bases' 3.6.1.1.1 testing in accordance with tha Primary . update submittal P.18 and Containment Leakage Rate Testing with regards to'-

Associated Program. STS SR 3.6.1.1.1 requires 11/2/95 letter .

Bases the visual examination and leakage rate and updated '

testing be performed in accordance . TSTF 52 when" with 10 CFR 50 Appendix J as OG provides modified by approved exemptions. ITS revision or. -

SR 3.6.1.1.1 modifies STS SR provide additior.ai- ,

3.6.1.1.1 to conform to CTS justification for 4.7.A.1.A. The STS is based on deviations.

Appendix J Option A while the CTS /ITS are based on Appendix J, Option B.

Changes to the STS with regards to Option A versus Option B are covered ,

by a letter from Mr. Christopher I. L Grimes to Mr. David J. Modoen, NEl dated 11/2/95 and TSTF 52. The ITS changes are not in conformance with the letter or TSTF 52 as modified by staff comments. ,

DAEC RESPONSE: Option B to Appendix J was incorporated into the DAEC CTS by Amendment 219. The changes are in l accordance with the letter from Grimes to Modeen (Nov. 2,1995) referenced in the NRC question. As committed in our "

original submittal letter (NG-96-2322), new CTS markup pages, based upon Amendment 219, will be provided in our next revision of the Specifications and Bases.

4 l i

-- - - .- ,m . - __

DAEC ITS 3.6.1.1 PRIMARY CONTAINMENT I

ITEM. DOC / CTS /STS DESCRIPTION OF ISSUE DATE. DATE COMMENTS NO. JFD LCO OPENED CLOSE D

3.6.1.1-5' Bases STS STS B3.G.1.1 Bases-SR 3.6.1.1.2 7/8/97 Delete this P.1 B3.6.1.1 states the following: "This SR change.

Bases- measures drywell to suppression SR chamber differential pressure...to 3.6.1.1.2 ensure that the leakage paths...are ITS B3.6.1.1 within allowable limits." ITS B3.6.1.1 Bases- Bases SR 3.6.1.1.2 changes this SR statement as follows: "This SR 3.6.1.1.2 maintains drywell to suppression chamber differential pressure...to measure and ensure that the leakage paths... The justification used (Bases P.1) is for plant specific nomenclature, system description, etc. This justification does not apply to the changes made. In addition, this SR does not maintain a differential pressure but measures the increase in pressure to determine the leakage rate.

Therefore, the change is unacceptable. i DAEC RESPONSE: As discussed in JFD P.6 to SR 3.6.1.1.2, the DAEC design does not permit the measurement of differential pressure between the suppression pool and drywell. Our SR will only measure the pressure on one side of the boundary and look for increases, which indicate leakage. The Bases were clarified to match the change in the SR. Either JFD P.1 (plant-specific information) or P.8 (changes to match changes in the Specifications) are acceptable.

5

. . . . , . .. . _ ~ . . .- . . . - .. - . - _ . . .

DAEC ITS 3.6.1.1 PRIMARY CONTAINMENT:

ITEM- DOC / CTS /STS DESCRIPTION OF ISSUE- DATE. DATEJ bOMMENTS :

NO. JFD: LCO: OPENED CLOSE ,

D :i

. -3.6.1.1-6 Bases STS STS B3.6.1.1 Bases-SR 3.6.1.1.1 '7/8/97 See item Number -

P.8 B3.6.1.1 references a number of SRs in ITS 3.6.1.3-4.- .

Bases-SR 3.6.1.3. ITS B3.6.1.1 Bases SR 3.6.1.1.1 3.6.1.1.1 changes these riumbers to ,

ITS B3.6.1.1 reflect the deletion of a' number of SRs 1 Bases-SR in ITS 3.6.1.3. These changes are 3.6.1.1.1 dependent on resolution of item -

Number 3.6.1.3-4.

DAEC RESPONSE: See Reponse to Question 3.6.1.3-4. No change in numbering of the iTS is planned.

I I

l t 6

. _ . - _ = . . _. ._ . . . - .

e DAEC 'ITS 3.6.1.2 PRIMARY' CONTAINMENT AIRLOCK:

ITEM. DOC /.:- CTS /STS ' DESCRIPTION OF ISSOE - DATE. .DATE COMMENTS NO.' JFD' LCO' OPENED CLOSED >

3.6.1.2-1 A.4 CTS' 15.b CTS 15.b requires that at 7/8/97 ~ Revise tne CTS :

least one door in each ' markup to include airlock is closed and sealed. . CTS 15.b, and e CTS 15.b is definitely part provide additional ;

~

of the OPERABILITY discussion and ,

requirements for airlocks ' justification for the -

and as such should be changes made to -

included in the CTS markup CTS 15.b.

for ITS 3.6.1.2. See item Numbers 3.6.1.1-1 and 3.6.1.1-2.

DAEC RESPONSE: The CTS markup for ITS 3.6.1.2 will be revised to include CTS 15.b. The justification has been revised to a Relocated items and DOC A.4 has been revised accordingly (attached). (Reference Response to NRC Ouestion 3.6.1.1-1).

^

3.6.1.2-2 Bases STS SR STS SR 3.6.1.2.2 requires 7/8/97 Licensee to update P.1 3.6.1.2.2 verifying only one door in submittal to be in Bases ITSSR the airlock will open at a - accordance with P.8 3.6.1.2.2 and - time at six month intervals. TSTF,17 or provide Associated The interval is modifad in additional

] Bases the ITS from 6 months to justification for the '

24 months. This - deviations.

i nodificatien is in accordance with TSTF 17; however, the Bases changes are not in accordance with TSTF 17. ;

DAEC RESPONSE: The Bases will be revised to reflect TSTF 17, Revision 1, which 'was approved by the NRC on March 13,1997. A clerical error in the traveller Bases (NRC comment of September 18,1996 not fully incorporated) will also be 3 corrected. The change will be submitted with the next revision to the ITS Specifications and Bases. .

t i !

1

--- - + , _

, - -, - - ., . . . . v

DAEC ITS 3.6.1.2 PRIMARY CONTAINMENT AIRLOCK ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED _

3.6.1.2-3 Bases ITS B3.6.1.2 ITS B3.6.1.2 Bases RA A.1, 7/8/97 Delete this change.

P.2 Bases- A.2 and A.3 adds the RA A.1, A.2 following sentence: 'This 7 and A.3 day allowance is tracked from initial entry into condition A and is not intended to place a limitation on the use of Note 1 to the ACTIONS." The justification used (P.2) is for enhances clarity, typographical errors or Bases censistency. This change does not enhance clarity, changes the intent of the Note, anc' could possibly be generic. .

Therefore it is unacceptable. l l DAEC RESPONSE: Note 1 to Actions allows entry and exit to perform repai_rs with no limit on how long this may continue.

Note 2 to Required Actions A.1, A.2, A.3 allows entry and exit for only 7 days. As discussed in the Bases, this 7 day allowarece is for Surveillances, other Required Actions, and any other necessary activities, not fcr repsirs, which are covered by Note 1 to Actions. These two Notes are separate and distinct. The added sentence does not change the intent of either Note, it merely clarifies that they are separate, independent notes. Thus, the change is acequately addressed by JFD P.2. A TSTF will not be pursued as it is deemed to be "below threshold" for a generic change.

3.6.1.2-4 None ITS B3.6.1.2 See item Number 3.6.1.1 -4. 7/8/97 See item Number Bases 3.6.1.1-4.

yEC RESPONSE: See Response to Question 3.6.1.1-4.

2

'DAEC'ITS 3.6.1 3 PRIMARY CONTAINMENT ISOLATION. VALVES (PCIVs),

ITEM . DOC / : CTS /STS DESCRIPTION OF ISSUE' DATE DATE- COMMENTS ,

NO; JFD g LCO- OI5ENED CLOSED

+

3.6.1.3 A.4 CTS 15 See item Number 3.6.1.1-2 7/8/97 See item Number A.8 ITS 3.6.1.1 3.6.1.1 -2. :- ,

t R.1 ~ Bases BACKGROUM

. D DAEC RESPONSE: See Respone,e to Question 3.6.1.1-2.

.3.6.1.3-2 A.4 STS 3.6.1.3 Three new Notes are added to ITS 7/8/97 Delete this change.

P.3 ACTION 3.6.1.3 ACTIONS. Note 2 allows Bases Notes separate condition entry for each P.2 ITS 3.6.1.3 penetration flow path. Note 3 -

ACTION requires entering applicable -

Notes and Conditions and Required Actions ,

Associated for systems made inoperable by

, Bases. PCIVs. Note 4 requires entering Conditions and Required Actions of ITS 3.6.1.1 when MSIV or purge.

valve leakage rates exceed overall ,

containment leakage rate acceptance criteria. STS 3.6.1.3 Action Note 4 applies to all PCIVs

not to just MSIV.and Purge Valves.

Since leakage from a 1y valve could -

result in valve inoperability yet not result in ovt;rall containment leakage being exceeded, the Note has to apply to all PCi s. In addition, there is inadequate justification (P.3) to limiting Note 4 to just MSIVs and Purge Valves. .

The staff finds the change unacceptable, and potentially aeneric.

1

DAEC ITS 3.6.1.3 PRIKERY' C0tiTAINMEllT ISOLATI0fi VALVES (PCIVs)

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE. DATE COMMENTS -

JFD LCO OPENED CLOSED

_N O.

DAEC RESPONSE: ITS 3.6.1.3 does not contain Surveillance Requirements for leakage testing of any PCIVs other than MSIVs and Purge Valves. Therefore Note 4 is limited to only the MSIVs and Purge Valves to avoid confusico as a result of the other changes in STS 3.6.3.3. Additional leakage testing requirements for the remaining PCIVs are contained in ITS 3.6.1.1 and the Primary Containment Leakage Rate Testing Program (ITS 5.5.12). This change is plant-specific, as the DAEC Current l_icensing Basis does not include bypass leakage (see JFDs P.8) and JFD P.50 regarding leak testing of hydrostatically tested lines. Additional explanation will be added to JFD P.3.

3.6.1.3-3 M.4 STS 3.6.1.3 STS 3.6.1.3 Condition I defines the 7/8/97 Delete this generic P.37 ACT!ON I acronym OPDRVs in Condition I. change.

ITS 3.6.1.3 ITS 3.6.1.3 ACTION G removes ACTION G the phrase " Operation with a potential for draining the reactor vessel (OPDRVs) from Condition G and places it in RA G.1 in place of "OPDRVs." The justification (P.37) states that the only OPDRVs that need to be suspended are those associated with the RHR Shutdown Cooling System. The justification does not provide adequate justification as to why ITS 3.6.1.3 ACTION G should not apply to the other OPDRVs implied by the justification. Since the ras are connected by an "or" there is no ,

guaranty that RA G.1 will be used for when the RHR valves are inoperable rather than RA G.2.

The STS considers this condition into Condition I as unacceptable.

In addition, the staff has determined that thb. b a generic change which is beyond the scope of review for this conversion.

2

DAEC ITS 3.6.1.3 PRIMARY CONTAltiMEllT ISOLATION VALVES (PCIVs)

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED DAEC RESPONSE: The use of the term "or operations with the potential for dreining the reactor vessel (OPDRV's)" in Condition G is inconsistent with the Mode of Applicability specified for LCO 3.6.1.3. The DAEC Current Licensing Basis does not require any Primary Containment Isolation Valves (PCIV) Operable in Mode 4 or 5 (Reference DOC M.4 to CTS) nor does it require PCIVs to be Operable when performing OPDRV's. The only PCIV functions required Operable by the STS for the DAEC design would be the Shutdown Cooling isolation valves. Restoring Operability of the shutdown cooling valv.es will not mitigate any OPDRVs except those within the RHR system. The reviewer notes that either Required Action G.1 or G.2 may be applied. This is correct for both the ITS and the STS, and is not altered by change P.37. This is acceptable as either Action is a proper compensatory action to restore compliance with the LCO. JFD P 37 has been revised to better describe this deviation. Since this deviaticn preserves appropriate attributes of the Current Ucensing Basis, it is appropriately within the scope of the conversion.

3.6.1.3-4 R.1 CTS 15.a STS SR 3.6.1.3.3 and SR 7/8/97 Revise the CTS P.7 STS SR 3.6.1.3.4 verify that each manual markup and ITS Bases 3.6.1.3.3 PCIV and blind flange located 3.6.1.3 to include P.2 STS SR outside and inside containment - STS SR 3.6.1.3.3 Bases 3.6.1.3.4 and respectively that is required to be and SR 3.6.1.3.4 P.8 Associated closed is closed. ITS 3.6.1.3 does and its Associated Bases not include these STS SRs. The Bases. Also revise ITS B3.6.1.3 justification used (P.7) is incorrect. ITS B3.6.1.3 Bases. The STS SRs are required, based Bases-SR 3.6.1.3.1 SR 3.6.1.3.1 on the CTS requirement specified to reflect these in CTS 15.a.2). While it is correct changes. Provide that the majority of CTS 15.a is any additional relocated (R.1) to the Bases as discussion as background material for ITS necessary. See B3.6.1.1, the requirement specified item Number in CTS 15.a.2 needs to be 3.6.4.2-4.

specified as an SR just like the rest of the definition requirements.

CTS definitions are part of the TS requirements and can specify indirectly or directly SRs or OPERABILITY requirements that must be met, such as in this case.

See item Number 3.6.4.2-4.

3

.DAEC ITS 33 .1.3 PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) t l ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED DAEC RESPONSE: The reviewer contends that SR 3.6.1.3.3 and SR 3.6.1.3.4 are required by CTS definition 15.a. This is incorrect. While Definition 15.a for Primary Containment does indeed include requirements for these components to support Operability of the Primary Containtnent, by no means does this constitute a Surveillance Requirement to verify the-position of manual va!ves and blind flanges every 31 days or (inaccessible valves prior to Mode changes). JFD P.7 provides a detailed discussion of the Administrative Cc 7trols in place at the DAEC to ensure these components continue to effectively support Operability of the Primary Containment. The components affected by these STS Surveillances (manual, valves and blind flanges) are passive devices which do not change status without deliberate human action. The Administrative Controls in place at the DAEC ensure these activities do not occur without the cognizance of the operating crew and thus achieve an equivalent level of protection with less burden and less radiation exposure for the Operating staff.

As discussed in the Bases for SR 3.0.1 "Nothing in this Specification, however. is to be construed as implying that systems or components are OPERADLE when: a. The systems or components are known to be inoperable, although still meeting the SR's." Beth the STS and CTS recognize that there are attributes of Operability that are not periodically surveilled. The absence of these SR's does not obviate the requirement to have the equipment Operable, i.e., valves are in the inproper position and flanges are installed. The addition of a periodic surveillance, however, is not necessary provided there is reasonable assurance that the Operability of the function will be maintained with equal reliability. The existing Administrative Controls have been demonstrated to be equally effective to the STS periodic SR; fully implement the Current Licensing Basis; and, therefore, are acceptable as the mechanism for implementing this verification following conversion to the ITS.

4

~ . . . - . . , - . .

DAEC ITS 336.1.3 PRIMARY CONTAD#1ENT' ISOLATION VALVES .(PCIVs)

ITEM DOC / CTS /STS : . DESCRIPTION OF ISSUE ~ DATEi DATE COMMENTS -

' NO.' JFD LCO- ;

. OPENED l CLOSED 3.6.1.3-5' R.4 CTS 3.7.8.4 CTS 3.7.B.4.a requires that . 7/8/97 Correct this L Bases ITS B3.6.1.3 containrr :nt pu ge valves'not be discrepancy or (-

P.2 Bases- ' opened so as.to create a flow peth provide additional SR 3.6.1;3.4 from the primary containment. The discussion and specific valves listed in CTS justification for not !

i 3.7.B.4.a are moved to the ITS ' listing CV-4309 3.6.1.3 Bases SR 3.6.1.3.4. Two and CV-4310 in of the valves (CV-4309 and CV- ITS 3.6.1.3 Bases -

4310) listed in the CTS are not SR 3.6.1.3.4. -

identified in the ITS 3.6.1.3 Bases -

SR 3.6.1.3.4.

DAEC RESPONSE: The Specific SR (SRs 3.6.1.3) deals with purge valves with resilient seals. The two valves in questio'n -

are small bypass valves (2" valves) that do not have resilient seats and are not part of this SR. Thus, they ue not included

.in the Bases. -DOC R.4 also is used for lTS SR 3.6.1.1 (CTS 3.7.B.4.a) for verifying that the 18" purge valves are closed.

As stated above, the two valves in question are small bypass valves that are not subject to this SR either. DOC R.4 will be clarified to explain that the subject valves are relocated to the UFSAR with all the other power-cperated automatic PCIVs (attached).

I U

I l'

l

(

.L 5

l . __ __ . ~ . .

DAEC. ITS 3.6.1.3 PRIMARY CONTAINMENT' ISOLATION ' VALVES -(PCIVsl'.

.lTEM ~ DOC / CTS /STS DESCRIPTION OF ISSUE ' DATEi DATE- COMMENTS NO. JFD LCO OPENED CLOSED 4r 3.6.1.3-6 'L1- CTS 3.7.B.2 CTS 4.7.A.1.b specifies the MSIV 7/8/97 Delete this generic P.4 CTS leakage limits and remedial actions- change.

Bases' 4.7.A.1.b to take upon discovery of leakage P.3 STS 3.6.1.3 rates exceeding specified limits. ~

ACTION D CTS 3.7.B.2 provide additional.

and operability requirements, remedial Associated actions and associated times in Bases which to complete the repairs and ITS 3.6.1.3 retests associated with CTS ACTION D 4.7.A.1.b. The repair time per CTS and 3.7.B.2 is 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ITS 3.6.1.3 Associated Condition D changes STS 3.6.1.3 -

Bases Condition D from " Secondary containment bypass leakage rate not within limit to' "One or more penetration flow paths with one or more MSlVs not within leakage l j fimits." Based on STS B.3.6.1.3 Bases RA D.1 discussion, STS 3.6.1.3 Condition D includes both j secondary ccntainment and MSIV ;

leakage. Therefore, the proposed ,

i change to Condition D is acceptable. However, the change r of the Completion Time associated with RA D.1 and CTS 3./.B.2 from ;-

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to an ITS time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is not adequately justified.' The justification used is consistency i

with the Completion Time of RA ,

t 6 ,

DAEC ITS 3.6.1.3 PRIMARY CONTAINMENT IS0l> TION VALVES (PCIVs)

ITEM DOC / . CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS '

NO. ED LCO OPENED CLOSED A.1 and that generic change BWR 15 C.4 was not fully implemented.

The staff never approved BWR 15 C.4. The Completion Time associated with ITS 3.6.1.3 RA D.1 takes into account the safety significance of containment leakage versus valve inoperability.

Thus the STS Completion Time for leakage is less than the Completion Time for an inoperable MSIV. In addition, the staff finds this change to be generic and beyond the scope of review for a conversion.

DAEC RESPONSE: We disagree with t"e Staff's argument. An open MSIV that is not capable of being closed (Condition A) is not less significant to safety than a closeable MSIV that exceeds its leakage limit iCondition D). They are the same.

Consequently, the Completion Time for implementing the Required Action to restore compliance with the LCO should be the same. In addition, Industry records indicate that the NRC did approve BWR-15#C.4.

7

DAEC ITS 3.6.1.3 PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE - DATE DATE' COMMENTS NO. JFD- LCO OPENED CLOSED- 'l 3.6.1.3-7 L2_ CTS- CTS 4.7.A.1.c requires purge 7/8/97 Either modify ITS P.25 4.7.A.1.c system isolation valve laakage ' SR 3.6.1.3.4 and Bases ITS SR testing at least once every three its Associated P.8 ' 3.6.1.3.7 months. STS SR 3.6.1.3.7 Bases to conform and requires leakage testing every 184 to the frequency Associated days and once within 92 days after specified in CTS.

Bases opening the valve. ITS SR 4.7. A.1.c or to ITS SR 3.6.1.3.4 relaxes the CTS testing both the 3.6.1.3 4 and requirement to e'very 184 days and frequencies Associated deletes the STS testing specified in STS SR Bases requirement of once within 92 days 3.6.1.3.7 and its after opening the valve. STS Associated Bases.

B3.6.1.3 Bases SR 3.6.1.3.7 Provide additional states the staff position that the obcussion and -

92 day frequency was chosen justification for recognizing that cycling the valve these changes.

could introduce additional seal degradation, beyond that which occurs to a valve that has not been opened. Thus, decreasing the STS interval of 184 days is a prudent measure after a valve has been opened. The justification used (L.2) to delay the 3 month testing requirement to 184 days states that the vent / purge valves are cycled at least once per week to maintain consistent stroke times of the valves with Bettis Actuaturs.

Based on this and a commitment made to leak test the valves on a 8

DAEC ITS 3.6.1.3 PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 1

3 month frequency (stat'd in the SE for Amendment 219 Jated October 4,1996), the staff finds that the frequency changes made in ITS SR 3.6.1.3.4 when converting from CTS 4.7.A.1.c and STS SR 3.6.1.3.7 are unacceptable.

DAEC RESPONSE: DAEC committed to quarterly testing with Amendment 219 as part of removing T-ring seal replacement from TS. This Amendment was issued after the DAEC ITS was completed and was, thus, not incorporated. Because, adding this additional test does not create an additional burden on the DAEC and is consistent with the current licensing basis, the STS requirement will be, aciuded in the ITS. DOC L.2 and JFD P.25 have been revised accordingly (attached).

The ITS Bases will be revised to re-instate the STS wording in a future submittal. [ Note: the ITS page already includes the second frequencv (an oversight in the original submittal) and does not need to be revised}.

e 9

DAEC'ITS 3;6.113 PRIMARY CONTAU4 MENT ESOLATION VALVES.(PCIVs).-

ITEM - DOC / - CTS /STS - DESCRIPTION OF ISSUE _DATEf :DATE -l COMMENTS-l NO. ~JFD LCO OPENEDL CLOSED' 3.6.1.3-8 L.8 ITS SR A Note is added to 11S SR . 7/8/97 - Delete this change.

. P.31 '3.6.1.3-7 3.6.1.3.7 which allows not entering the appropriate system.

i Bases' and

- P.8 Associated Required Actions when performing Bases a Surveillanr e Requirement on ,

these systems. This Note is not

^

included in the CTS o- STS. The

! justification addresses previous communication to the NRC staff in ;

letter DAEC GL 89-10 Motor- ,

Operated Valve (MOV) Program (J.

Franz (lES) to W. Russell (NRC), , 1

, " Generic Letter 89-10 Program," ')

NG-94-4017, November 30.

1994). The staff's position is that if a surveillance test makes the system inoperable or puts the system into an alignment that is not a normal or emergency -

operational alignment, then the ,

system is considered inoperable . i and the appropriate ACTIONS shall be taken.

DAEC RESPONSE: Per our Response to the Strff's RAI on this Note (Ref. NG-97-1597) and our meeting with the Staff on .

September 9,1997, this change (DOC L.8) will be withdrawn. The associated DOC. (L.8), No Significant Hazards .

Consideration (NSHC) and JFD (P.31) have been revised (attached).

?

10

DAEC ITS 3.6.1.3 FRIMARY CONTAlfiMEfiT ISOLATION VALVES (PCIVs)

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED C1.OSED 3.6.1.3-9 LCY.2 CTS 4.7.8.1.a CTS 4.7.8.1.a require; performing 7/8/97 Delete this change.

CTS 4.7.C testing of the PCIVs at least once ITS SR per operating cycle. Under the 3.6.1.3.6 same circum 7tance ITS SR ITS SR 3.6.1.3.6 requires testing every 24 3.6.1.,1.7 months. CTS 4.T ' requires and ve..fying, once pe. merating cycle.

Associated the operability of the reactor Bases coolant system instrument line flow check valves. Under the same circumstances ITS GR 3.6.1.3.7 requires this verification ,

every 24 months. This extension of the Surveillance frequency is considered by the staff as a beyond scope of review item for this conversior .

DAEC RESPONSE: Based upon our meeting with the NRC Staff on September 10,1997, we understand that the technical resources have been applied to review our conversion from an 18 month to 24 month operating cycle in parallel with our ITS conversion. Coasequently, this item is considered to be back within scope for the purposes of our Final Safety l Evaluation oc the ITS conversion.

3.6.1.3-10 P.21 ITS 3.6.1.3 The renumbering of ITS 3.6.1.3 7/8/97 See item Numbers Bases and SRs and references to succeeding 3.6.1.3-4 and P.8 Associated specifications will depend on the S3.6.1.d-1.

Bases resolution of item Numbers 3.6.1.3-4 and S3.6.1.4-1.

i DAEC RESPONSE: See Response to 3.6.1.3-4 and S3.6.1.4-1. No renumbering of the ITS is planned.

I 11 l .

DAEC ITS 3.6.1.3 PRIMARY CONTAINMEffT ISOLATION VALVES (PCIVs)

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.1.3-11 P.52 STS SR STS SR 3.6.1.3.6 and Associated 7/8/97 Licensee to update Bases 3.6.1.3.6 and Bases has been modified by TSTF submittal with P.8 Associated 46 Rev 1. ITS SR 3.6.1.3.3 and regards to TSTF Bases its Associated Bases as stated in 46, Rev 1.

ITS SR P.52 has been modified to be 3.6.1.3.3 and consistent with TSTF-46. The Associated changes to ITS SR 3.6.1.3.3 and Bases its Associated Bases are not consistent with TSTF-46.

DAEC RESPONSE: ITS SR 3.6.1.3.3 SR & Bases will be revised to reflect recommended wording of TSTF-46, Rev 1 in the a next revision of the ITS.

12

DAEC ITS 3.6.1.3 PRIMARY CONTAlfdENT ISOLATION VALVES (PCIVs)

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED C1.OSED 3.6.1.3-12 Bases STS B3.6.1.3 STS B3.6.1.3 Bases-LCO states 7/8/97 Provide additional P.1 Bases-LCO that the valves covered by this discussion and ITS B3.6.1.3 LCO are listed with their associated justifications for Bases-LCO stroke times in the FSAR (Ref. 2). listing these PCIVs ITS B3.6.1.3 ITS B3.6.1.3 Bases-LCO states in Administrative Bases- that the list is in plant Procedures rather REFERENCES administrative control procedures than the FSAR, and implies, based on the changes include the made, that no stroke times are procedures or provided in those procedures. The FSAR location in specific procedures are not ITS B3.6.1.3 specified or listed in ITS B3.6.1.3 Bases-Bases--3FFERENCES Section nor is REFERENCES the procedure change control Section and the process defined. Also, the location procedure change -

of the PCIV stroke times is required control process.

for reference to the satisfactory Also provide performance of ITS SR 3.G.1.3.3. document that lists &

the PCIV stroke times as well as the document !

change control process and ,

reference it in ITS B3.6.1.3. Provide additional discussion and justification as ;

necessary.

13

, DAEC ITS 3.6.1.3 FRIMARY CONTAIMtENT ISOLATION VALVES (FCIVs)

ITEM - DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD i.C O OPENED CLOSED DAEC RESPONSE: The procedure referenced is an Administrative Control Procedure, " Guidelines for Primary Containment Valves and Penetrations". The content of this procedure is derived from various information contained ;n the UFSAR.

Changes to this procedure are evaluated under 10 CFR 50.59 relative to the UFSAR. Although, the information in question is contained in the UFSAR, no succinct listing which contains solely the applicable valves is available in the UFSAR for referencing in the ITS Bases. With respect to valve stroke times, the stroke times for the referenced va!ves are specified in the In Service Testing Program and implemented in associated Surveillance Test Procedures. The values are derived from

- information contained in a variety of locations in the UFSAR however, no succinct listing is available in the UFSAR for referencing the Bases for iTS 3.6.1.3. Changes to the procedures are evaluated under 10 CFR 50.59 against the applicable information in the UFSAR for the specific change. These controls are consistent with what was reviewed and approved by the NRC staff in the Safety Evaluation for Amendment 181 to the DAEC Ucense which removed tables listing primary containment valves and their stroke times from the Technical Specifications.

3.6.1.3-13 Bases STS B3.6.1.3 STS B1.6.1.3 Bases-SR 3.6.1.3.10 7/8/97 Provide additional P.1 Bases- states the following for excess discussion and SR 3.6.1.3.10 flow check valve : 'The 118] justification to ITS B3.6.1.3 month Frequency is based on the justify the sentence Bases need to perform this Surveillance ! deletion based on SR 3.6.1.3.7 under the conditions that apply '

current licensing during a plant outage and the basis, system potential for an unplanned transient design or if the Surveillance were performed cperational with the reactor at power." ITS constraints.

B3.6.1.3 Bases-SR 3.E.1.3.7 deletes this statement. The justification used (Bases P.1) is the general plant specific nomenclature, system dascription, etc., justification, which is

, inadequate to justify this deletion.

DAEC RESPONSE: Some EFCVs can be Wested on line because they affect instruments that provide indication only, or instruments that can be safely isolated on line without risk of an inadvertant equipment actuation. Hence, the reason the STS Bases were revised.

14

DAEC ITS 3.6.1.3 PRIMARY CONTAI?NENT ISOLATION VALVES (PCIVs)

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.1.3-14 Bases STS SR The Bases for STS SR 3.6.1.3.13 7/8/97 Add Note to ITS SR P.1 3.6.1.3.13 refers to a Note 1 while STS SR 3.6.1.3.13 and and 3.6.1.3.13 does not show a Note, retain Bases Associated Therefore, the Bases discussion on description of Note.

Bases the Note was deleted from the l'S Provide additional ITS SR SR 3.6.1.3.9. This is an error. justification and 3.6.1.3.9 and The Note should be added to ITS discussion to Associated SR 3.6.1.3.9 and the discussion sepport this Bases retained in the Bases. This Note change.

deals with leakage limit applicability and is associated with ITS 3.6.1.3 ACTIONS Note 4.

Also, BWR 16 C.5 corrected this error. This error has been corrected by TSB-13.

DAEC RESPONSE: The propo ed Note to the SR is not needed, as the LCO's Applicability for MSIVs is only Modes 1, 2 arrJ

3. Thus, per SR 3.0.1, the surveillance needs only to be met in Modes 1,2 and 3, so the Note would be redundant. This is analagous to changes made under JFD P.47 for the purge and vent valves.

3.6.1.3-15 Bases STS B3.6.1.3 STS B3.6.1.3 Bases-LCO states 7/8/97 Provide additional P.2 Bases-LCO that the passivo PCIVs or isolation discussion and ITS B3.6.1.3 devices covered by STS LCO justification for Bases-LCO 3.6.1.3 are listed in the FSAR (Ref listing these ITS B3.6.1.3 2). ITS B3.6.1.3 Bases LCO isolation devices in Bases- modifies this statement by adding more than one HEFERENCES after " Reference 2" "or in document rather applicable administrative than just the FSAR, procedures." The justification used include the (Bases P.2) is a general clarity procedures in ITS justification. This justification is B3.6.1.3 Bases-incorrect for this change. The REFE 'NCES specific procedures are not Section and provide specified or listed in ITS B3.6.1.3 the procedure Bases-REFERENCES Section n r is change control the procedure change control process.

process defined. r 15

DAEC ITS 3.6.1.3 FRIt%RY CONTAlfFtNT ISOLATION VALVES (PCIVs) z w . w , .. w. . ,

ITEM- DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE p i.'RNJiG d NO. JFD LCO OPENED CLOSED l- . _ .

m. .._kA DAEC RESPONSE: Neither the Current Technical Specifications nor the Updated Final Safety Analysis Repm Gw,ha a succint comprehensive listing solely of the passive isolation devices for the primary containment. Sir >ce the reWer h 'de STS is to a single list, this would be inaccurate for the DAEC as multiple documents will need to be consulted. #Mdican for Deviation P.2 correctly reflects this clarification. Also, see the Response to Question 3.6.1.3-12, above.

3.6.1.3-16 Bases ITS B3.6.1.3 ITS B3.6.1.3 Bases ACTIONS adds 7/8/97 Delete this generic P.2 Bases- additional statements to the change.

ACTIONS ACTIONS Note 1 description stating that Note 1 expands upon the allowance of LCO 3.0.5. The justification used (P.2) is the general clarity justification. The addition does not clarify Nots 1 and the change would be considered by the staff as a generic change which is a beyond the scope of review item for this conversion.

DAEC RESPONSE: LCO 3.0.5 allows equipment removed from service or declared inoperable to be returned to service under administrative control soleiv to perform testing required to demonstrate its Operability or the Operability of other equipment. It does not allow the affected penetrations to be opened for any other reason. Thus, the intent of Note 1 to the ACTIONS to provide this additional flexibility. The additional Bases wording merely explains what types of operations are allowed under the Note 1 exception to LCO 3.0.5. While this clarification may be applicable to other plants, it is telow the threshold for the generic TSTF process.

16

DAEC ITS 3.6.1.3 PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) i ITEM DOC /,. CTS /STS DESCRIPTION OF ISSUE . DATE- . DATE COMMENTS !

. NO. JFD LCO. OPENED CLOSED t

3.6.1.3-17 Bases STS B3.6.1.3 STS B3.6.1.3 Pases RA C.1 and 7/8,97 Delete this generic P.2 Bases- C.2 states the following: "In the change.

RA C.1 and event the affected penetration flow [

C.2 path is isolated in accordance with ;

iTS B3.6.1.3 Required Action C.1, the affected l penetration must be verified to be 4

Bases =

' RA C.1 and isolated on a periodic basis. This is .

C.2 necessary to ensure that primary I containment penetrations required ;

to be isolated followiag an accident

, are isolated." ITS B3.6.1.3 Bases RA C.1 and C.2 revised the wording of these sentences for l clarity (Bases P.21 and requires a system walkdown to verify i isolation. The modification does [

not clarify the STS wordir g and l i the addition of the system l walkdown imposes an additional l requirement that is not justified nor ;

imposed on Conditions A, B, or E.

The staff considers this change to j 4

be a generic change which is a beyond scope of review item for i i this conversion. t t

l DAEC RESPONSE: The purposed Bases change was done for consistency of wording with the same requirements under j j' Conditions A and E. The additional clarification of "by system walkdown" for how to perform the intended " verification

l

l. will be deleted for consistency with the other Bases sections. ' This change will be included in the next ITS revision j submittal.

I i

E i 17 [

_ . .. ~ - - . . ~ . - - - . . . . . . . ..

DAEC ITS 3.6.1.3 PRIfMRY CONTAlt.HENT ISOLATION VALVES (PCIVs)

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSko 3.6.1.3-18 Bases ITS B3.6.1.3 ITS B3.6.1.3 Bases-RA E.1, E.2 7/8/97 Delete this change.

P.2 Bases- and E.3 adds the following: "If the '

RA E.1, E.2 results of a combined leak rate or and E.3 pressure drop test indicate excessive leakage, credit can be taken for one of the purge valves to satisfy Required Action E.1, if it can be reasonably determined that only one of the purge valves is leaking excessively." This statement was added for clarity (Bases P.2); however, it does not clarify anything that is already written in the Bases and could be misinterpreted to allow the leaking valve to be used for the isolation requirement of RA E.1. The change is unacceptable.

DAEC RESPO.'c 2: This Bases change will be deleted. The revised Bases page will be included in the next ITS revision submittal.

3.6.1.3-19 None ITS B3.6.1.3 See item Number 3.6.1.1-4. 7/8/97 See item Number Bases- 3.6.1.1-4.

SR 3.6.1.3.4 DAEC RESPONSE: See Response to Question 3.6.1.1-4.

18

DAEC ITS 3.6.1.3 PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.1.3-20 None CTS 3.7.B.1 CTS 3.7.B.1 requires all primary 7/8/97 Provide a ITS LCO containment isolation valves (PCIV) discussion and 3.6.1.3 and instrument line flow check justification for this and valves OPERABLE except when a Less Restrictive Associated PCIV is inoperable and the affected change of the ITS Bases flow path is isolated. ITS 3.6.1.3 exception requires each PCIV, except reactor of the reactor building-to-suppression chamber building-to--

vacuum breakers, to be suppression OPERABLE. There is no discussion chamber vacuum or justification for excepting breakers CTS reactor building-to-suppression requirements.

chamber vacuum breakers from ITS LCO 3.6.1.3 when it is required by CTS 3.7.B.1.

DAEC RESPONSE: The CTS requirements for Reactor Building-to-Suppression Chamber vacuum breaker Operability are in CTS 3.7.D, not CTS 3.7.B.1. The corresponding ITS Section is 3.6.1.6. These valves are not classified as PCIV's ur. der CTS 3.7.B.1 at the DAEC. Thus, no justification is needed, as no change has been made between the CTS and !TS for these valves.

19

DAEC ITS 3.6.1.5 LOW-LOW SET (LLS) VALVES ITEM DOC / CTS /STS' DESCRIPTION OF DATE DATE COMMENTS NO. JFD LCO ' ISSUE OPENED CLOSED 3.6.1.5-1 A.3 CTS 4.6.D.1 CTS 4.6.D.1, CTS 7/8/97 Provide the A.4 CTS 4.6.D.2 4.6.D.2 and CTS 4.6.D.4 appropriate A.5 DTS 4.6.D.4 specify various discussions and R.1 surveillances,

justifications for the inspections, and tests to disposition of these be performed on the CTS requirements in safety relief valves. The the CTS /ITS markups CTS markup shows that of ITS 3.6.1.5.

these surveillance requirements are associated with ITS 3.4.3. The staff's review concludes that they also apply to ITS 3.6.1.5 as well, but the justifications A.3, A.4, A.5 and R.1 are not provided in ITS 3.6.1.5.

DAEC RESPONSE: See Response to Question 3.6.1.5-2, below, regarding an omission on the CTS markup of page 3.6-9 (page 23 of 47) for ITS LCO 3.6.1.5. The referenced DOCS only apply to ITS 3.4.3 (S/RV) channes.

1 i

DAEC ITS 3.6.1.5 LOA-LOW SET (LLS) VALVES ITEM DOC / CTS /STS DESCRIPTION OF DATE DATE COMMENTS NO. PD LCO ISSUE OPENED CLOSED 3 3.6.1.5-2 M.2 ITS LCO CTS 4.2.B.2.g and CTS 7/8/97 Provide a markup for 3.6.1.5 4.6.D.3 contain SRs for CTS 3.6.D.1 and CTS ITS 3.6.1.5 the LLS valves. M.2 3.6.D.2 and provide APPLICABILITY states that the CTS does the appropriate ITS 3.6.1.5 not contain specific LCO, Administrative (A),

ACTIONS APPLICABILITY or More Restrictive (M)

ACTIONS for the LLS or Less Restrictive (L) valves. This is incorrect. change discussions No M.2 is shown in the and justifications to markup of CTS 4.2.B.2.g convert these CTS or CTS 4.6.D.3 which requirements to the corresponds to this ITS requirements of change. The only M.2 ITS 3.6.1.5.

shown applies to ITS SR 3.3.8.1.5. However, CTS 3.6.D.1 and CTS 3.6.D.2 do contain the specific LCO, APPLICABILITY and ACTIONS for the LLS valves.

DAEC RESPONSE: DOC M.2 accurately reflects the change to the CTS. The Low-Low Set function was added to the DAEC design in 1983. In CTS Amendments 89 and 102, setpoints and Surveillance Requirements for this function were added to the DAEC Technical Specifications. However, no explicit or implicit limitations were placed on the modes of applicability or an operation with inoperable components. Change M.2 was inadvertantly omitted from the CTS mark-up of page 3.6-9 (page 23 of 47) and the NUREG mark-up for LCO 3.6.1.5. Mark-up pages will be revised. These changes will be included in the next ITS revision submittal.

i 2

l _ _ ___ _ _ _ - - - - _ _ _ _ _ _ _ _ . - _-

i

~!

DAEC ITS 3.6.1.5 LOW-LOW SET (LLS) VALVES a

f

?

ITEM DOC / ' CTS /STS DESCRIPTION OF DATE' DATE COMMENTS ,

j NO. JFD' LCO ISSUE' OPENED CLOSED 3

j 3.6.1.5-3 LCY.2 CTS 3.6.D.3 CTS 3.6.D.3 requires the 7/8/97 Delete this change. l l- Bases CTS 4.2.B.2.g manual opening of each ;

!' P.8 ITS SR 3.6.1.5.1 relief valve once per I

! ITS SR 3.6.1.5.2 operating cycle. CTS

! and Associated 4.2.B.2.g requires Logic ,

, Bases System Functional Tests -

(LSFT) once per operating

{'

3 cycle for the Low-Low l

i Set Function. Under the i l same conditbns the ITS i

!- SR 3.6.1.5.1 and ITS SR I l 3.6.1.5.2 have a

] Frequency of 24 months. f

This extension of the Surveillance Frequency is considered by the staff i l as a beyond scope of i

review item for this

. conversion. ,

4

.DAEC RESPONSE: Based upon our meeting with the NRC Staff on September 10,1997, we understand that the technical i resources have been applied to review our conversion from an 18 month to 24 month operating cycle in parallel with our

[ ITS conversion. Consequently, this item is considered to be back within scope for the purposes of our Final Safety j Evaluation on the ITS conversion.

l

, 3.6.1.5-4 P.21 ITS 3.6.1.5 See item Number 7/8/97 See item Number l i

Bases and Associated 3.6.1.4-1. 3.6.1.4-1. !

P.8 Bases ;

i...

DAEC RESPONSE: See Reponse to Question S3.6.1.4-1. No renumbering of the ITS is planned.

. i t

I i

3 l

l i

DAEC ITS 3.6.15 LOW-LOW SET (LLS) VA!NES ITEM DOC / CTS /STS DESCR!PTION OF DATE DATE COMMENTS NO. JFD LCO ISSUE OPENED CLOSED 3.6.1.5-5 Bases CTS 3.6.D.3 CTS 3.6.D.3 specifies 7/B/97 Provide a discussion P.1 ITS B3.6.1.5 that the LLS valves shall and justification for Bases- be manually tested when this More Restrictive SR 3.6.1.5.1 reactor pressure is 2100 change in test psig. ITS B3.6.1.5 pressure.

Bases-SR 3.6.1.5.1 states that this same test cannot be performed until the reactor pressure is 2 150 psig. No justification is provided for this More Restrictive change from 100 psig to 150 psig.

DAEC RESPONSE: CTS surveillance 4.6.D.3 requires that the safety relief valves, which includes the LLS valves, be tested at a reactor pressure equal to or greater than 100 psig with adequate turbine bypass flow available to detect that the safety relief valvas open and close. Reactor pressure and turbine bypass flow is controlled by the Electro Hydraulic Control (EHC) system. At DAEC, the EHC system is unable to control reactor pressure and turbine bypass flow be',w 150 ps.g reactor pressure. Consequently, the earliest time operability of the safety relief valves can be demonstrated is at a reactor pressure of 150 psig. DAEC currently performs the CTS surveillance 4.6.D.3 at approximately 150 psig with adequate steam flo y as defined by turbine bypass valve position. This satisfies the requirements of CTS surveillance 4.6.D.3 as neither an up,_ur limit on pressure nor time limit after reaching 100 psig is specified. To reflect the fact that the surveillance cannot be performed at reactor pressures less than 150 psig, the ITS Basis for SR 3.6.1.5.1 states, " adequate pressure at which this test is to be performed is approximately 150 psig which is the lowest pressure EHC can maintain". This is consistent with SRV and ADS specifications to avoid repeated cycles of the SRVs. This change has been characterized as More Restrictive l and is discussed under DOC M.1 to ITS 3.6.1.5.

4

DAEC ITS 3.6.1.6 REACTOR BUILDING-TO-SUPPRESSION CHAMBER VACUUM BREAKERS ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS -

NO. JFD *CO OPENE CLOSE D D 3.6.1.6-1 A.4 ITS SR Two Nctes are added to ITS SR 7/8/97 Provide 3.6.1.6.1 3.6.1.6.1 which state, that the additional SR is not required to be met for discussion and vacuum breaker assemblies justification for that are open for Surveillances the Less and not required to be met for Restrictive vacuum assembly valves open change.

for their intended function.

These Notes are not included in the CTS. There is inadequate discussion and justification for these Less Restrictive exceptions to the CTS SR.

DAEC RESPONSE: These notes are necessary to prevent unwarranted entry into the LCO Actions during a Surveillance, per SR 3.O.1. The DAEC CTS does not have SR 3.O.1. The addition of SR 3.0.1 is an essential part of the conversion process which is cor-sidered to be Administrative (Ref. DOC A.7 to ITS 3.0). By extension, the addition of these notes that are driven by SR 3.O.1 is also Administrative. In addition, the referenced change and supporting justification (DOC A.4) are identical to one previously found acceptable to the Staff for characterizing this change as Administrative (Ref. WNP-2).

3.6.1.6-2 A.4 ITS SR Two Notes are added to ITS SR 7/8/97 Correct A.4 to 3.6.1.6.1 3.6.1.6.1 which state that the delete the ITS SR SR is not required to be met for reference to ITS 3.6.1.6.3 certain conditions. A.4 states SR 3.6.1.6.3.

that both of these Notes apply to ITS SR 3.6.1.6.1 and SR 3.6.1.6.3. This is incorrect.

The Notes only apply to ITS SR 3.6.1.6.1.

DAEC RESPONSE: The Reviewer is mistaken. DOC A.4 does not say the Notes apply to SR 3.6.1.6.3. It merely references SR 3.6.1.6.3 as the Surveillance that creates the conflict with SR 3.6.1.6.1, necessitating the addition of the Notes. The DOC does not need correcting.

l l 1 l

DAEC ITS 3.6.1.6 REACTOR BUILDItJG-TO-SUPPRESSION CHAMBER VACUUM BREAKERS ITEM DOC / CTS /STS DESCR:PTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OFENE CLOSE D D 3.6.1.6.3 M.1 CTS 3.7.D.3 CTS 3.7.D.3 requ .n, when a 7/8/97 Provide a P.32 ITS 3.6.1.6 vacuum breaker assembly discussion and Bases P.1 ACTION A valve is open, that the other justification for and Associated valve in the assembly be the deletion of Bases verified closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . the CTS M.1 reduces the 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to 1 requirement to hour to be consistent with the verify the other time provided for when primar,e valve is closed containment is not maintained. within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ITS 3.6.1.0 ACTION A does not include this CTS ,

requirement, in either form. '

DAEC RESPONSE: The Reviewer is correct that CTS 3.7.D.3 is not included in ITS 3.6.1.6 Action A. The Action to verify the other valve closed in an assembly with one valve not closed is required by Condition B of ITS 3.6.1.6. (Sec CTS mark-up page 25 of 47 which shows that the " verification" part of CTS 3.7.D.3 is moved to Action B. It has not been deleted).

If both valves are open, Condition A and B would be entered simultaneously and Required Action B.1 would be taker'. If only one valve is open, then only Condition A is entered. This is fundamental to the way IT3 works and is an implicit part of the conversioni process (Ref. LCO 3.O.11. Action B.1 includes the 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Completion Time described in DOC M.1 to ITS 3.6.1.6 and is consistent with ITS 3.6.1.1 for primary containment restoration requirements. This is consistent with STS Section 3.6.1.7.

2

DAEC'ITS 3.6.1.6 RE#fTOR BUILDING-TO-SUPPRESSION CHM.BER VACUUM BRE#XERS e

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENE CLOSE D D 3.6.1.6-4 M.1 CTS 3.7.B.3 CTS 3.7.D does not contain 7/8/97 Provide P.32 CTS 3.7.D.3 ITS specific Required Action for discussion and Bases 3.6.1.6 ACTION B when one or two reactor justification for P.1 and Associated building-to-suppression the Less Bases chamber vacuum breaker Restrictive assemblies have both valves requirement not closed. One can assume when both that an immediate shutdown reactor building-would be required either by to-suppression CTS 3.7.B.3 or CTS 3.7.D.3. chamber vacuum ITS 3.6.1.6 RA B.1 requires breaker closing one open vacuum assemblies breaker assembly valve within valves are not 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months when one or two closed. See reactor building-to-soppression item Number chamber vacuum breaker 3.6.1.6-3.

assemblies have both .Was not closed. There is no discussion or justification for the new requirement when both reactor building-to-suppression chamber vacuum breaker assemblies valves are not closed. However, M.1 seems to apply here, but M.1 talks about when one valve is open. See item Number 3.6.1.6-3.

DAEC RESPONSE: See Response to NRC Ouestion 3.6.1.6.3.

i 3

l

DAEC ITS 3.6.1.6 REACTOR BUILDIfG-TO-SUPPRESSION CHAMBER VACUUM BREAKERS ITEM- DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENE CLOSE D D 3.6.1.6-5 L3 CTS 3.7.D.4 CTS 3.7.D.4 and 4.7.D.2 7/8/97 Provide CTS 4.7.D.2 provide ACTIONS and additional Surveillance Requirement for discussion and the Reactor Building-to- justification for Suppression Chamber Vacuum the procedure Breaker position indication and policy instrumentation. Position change control indication instrumentatien is process.

moved to plant procedures and policies. There is no discussion or justification of the specific proct. dure or policy and how changes to these procedures and policies are controlled.

This chcnge is considered by the staff as a Less Restrictive (LA) change.

DAEC RESPONSE: Per OGC direction, items that are ' relocated" from CTS to phnt procedures, which are not considered by the NRC to be under 10 CFR 50.59 control, mur t be considered to be " deleted" from Technical Specifications and characterized as a Less-Restrictive change, not a " relocated detail" under LA. DOC L3 states that this item is considered to be deleted from CTS and thus, is properly characterized per the NRC guidance.

4

DAEC ITS 3.6.1.6 REACTOR BUILDING-TO-SUPPRESSI0it CHAMBER VACL!UM BREAXERS ITEM DOC / CTS /STS DESCRIPTION OF ISSGd DATE DATE COMMENTS NO. JFD LCO OPENE CLOSE D D 3.6.1.6-6 L4 CTS 4.7.D.2 CTS 4.7.D.2 requires quarterly 7/8/97 Delete this P.42 STS SR demonstrations that the generic change.

Bases 3.6.1.7.2 and Reactor Building-to-Suppression P.8 Associated Bases Chamber Vacuum Breaker ITS SR 3.6.1.6.2 valve travels through one and Associated complete cycle of full travel.

Bases STS SR 3.6.1.7.2 requirer performing a functional test of !

each vacuum breaker every 1921 days. ITS Sri 3.6.1.6.2 cxtends this testir.g by ,

changing the frequency from "Once per quarter"/92 days to "In Accordance with the Inservice Testing Program" which allows up to six months depending on previous testing performance. This extension of the CTS /STS Surveillance Frequency from once per quarter to once per 6 months is considered by the staff to be a generic change which is a beyond scope of review item for this conversion.

DAEC RESPONSE: These components are in the current DAEC IST program. The change to the STS was made for consistency with other surveillances associated with the IS's program (see for example STS SRs 3.1.7.7. 3.4.3.1, 3.4.5.1, 3.5.1.7 and 3.5.2.5). The DAEC currently tests these valves quarterly (every 92 days), so no actual change in testing frequency is being made. Any future change in frequency will be evaluated under 10 CFR 50.55 (a) and 10 CFR 50.59.

The change was characterized as Less Restrictive for overall conservation, since the ASME Code would allow the frequency to be er. tended to six months. While this change is the DAEC's Current Licensing Basis thus, is plant-specific, a generic traveller has been initiated, based upon NRC request (Ref. DAEC Response to NRC Ouestion 3.1.8-3).

5

DAEC ITS 3.6.1.6 REACTOR BUILDING-TO-SUPPRESSION CH FoER VACUUM BREAKERS ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENE CLOSE D D 3.6.1.6-7 LIC.2 CTS 4.7.D.3 CTS 4.7.D.3 requires once per 7/8/97 Delete this Bases ITS SR 3.6.1.6.3 quarter demonstrating that the change.

P.14 and Associated setpoint of each vacuum Bases breaker is equivalent of s0.5 psid. Under the same circumstances ITS SR 3.6.1.6.3 requires every 12 months. This extension of the CTS Surveillance Frequency is based on Generic Letter 91-04 which only applies to increasing the SR frequency from 18 months to 24 months.

In addition, this change is considered by the staff to be a beyond scope of review item for this conversion.

DAEC RESPONSE: As discussed in iTS 3.6.1.6 DOC L.lC-2, this interval extension has been evaluated and found to be acceptable. This evaluation is based upon the NRC's methodology for extending surveillance intervals, as discussed in GL 91-04. The GL does not limit the use of this methodology to 18 to 24 month operating cycle extensions, but to any surveillance frequency change, especially for instrument calibration extensions such as this one. This change is part of our overall conversion to AI:owable Values for instrument setpoints, which the NRC has agreed is "within scope

for the DAEC.

3.6.1.6-8 P.21 ITS 3.6.1.6 and See item Number 3.6.1.4-1 7/8/97 See item Bases Associated Bases Number 3.6.1.4-P.8 1.

DAEC RESPONSE: See Response to Question S3.6.1.4-1. No change to the ITS is planned.

6

DAEC ITS 3.6.1.6 REACTOR BUILDING-TO-SUPPRESSION CHAPSER VACLUi BREAKERS

-- - n-TEM DOC / CTS /STS DESCRIPTION OF ISSUE !0 ATE DATE f.% MENTS NO, JFD LCO , OPCE CLOSE j

^C , D \

3.6.1.6-9 Bases STS B3.6. l.7 STS B3.6.1.7 Bases- k 7/8/97 Ste Otom P.2 Bases- APPLICABILITY states that vc Number APPLICABILITY vacuum breakers are reenM S3.6.2.4-1.

ITS B3.6.1.6 OPERABLE in MODES 1,2.ud Bases- 3 APPLICABILITY "...when the Suppresticn Ant Sr, ray System is requires u be C;'2RABLE." ITS B3.6.12 Bases-APPLICABILITY cnaeg.1 this to "...when Suppression '

Pool Spray System ope ation may be desirable." The acceptability of this change is dependent on the resolution of item Number S3.6.2.4-1. g DAEC RESPONSE: See Response to Question S3.6.2.4-1.

7

DAEC ITS 3.6.1.6 REACTOR BUILDING-TO-SUPPRESSION CHAM.BER VACUUM BREAKERS ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENE CLOSE D D 3.6.1.d-10 None CTS 3.7.D.2 CTS 3.7.D.2 identifies the 7/8/97 Provide ITS 3.6.1.6 required actions if one valve of additional ACTION C and a reactor building-to- disconsion and Associated Bases suppression chamber vacuum jusi;fication for breaker assembly is inoperable this for opening but known to be Administrative closed. ITS 3.6.1.6 ACTION C change of requires restoring the vacuum removing the breaker to OPERABLE status CTS requirement within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for one reactor that the vacuum liuilding-to-suppression breaker be chamber vacuum breaker known closed.

assembly with one or two valves inoperable for opening.

ITS 3.6.1.6 ACTION C does not assume knowing the inoperable vacuum breaker is c!osed nor does the ITS B3.6.1.6 Bases RA C.1. There is no discussion or justification for this Administrative change of removing the CTS requirement that the vacuum breaker be known closed.

DAEC RESPONSE: The Bases for Required Action C.1 does assume that the valve is clo::ed, as follows: "with one vacuum breaker assembly with one or more vacuum breaker assembly valves inoperable for openinci, the leak tightprimary containment boundary is intact. ~ (emphasis adc'ed). This obviously means that the valve (s) is closed, as the Bases for Required Action A.1 states: "with one or two vacuum breaker assemblies with one valve ng1 closed, the leak tight primary containment boundary may be threatened. ~ (emphasis added). Consequently, the CTS requirement of 3.7.D.2 is contained in ITS Action C. DOC L.1 to ITS 3.6.1.6 fully addresses this change.

8 L >

. . . ._ _ . _. ~ ._ . _. .. _ __

' DAEC ITS 3.6.1.7 SUPPRESSION CHAMBER-TO-DRYtdELL VACUUM BREAKERS

! TEM ~ DOC / CTS /STS _ DESCRIPTION OF ISSUE DATE: DATE COMMENTS NO. - JFD LCO OPENED CLOSED ,

, 3.6.1.7-1 A.2 ITS SR A Note is added to ITS SR 7/8/97. Provide additional 3.6.1.7.1 3.6.1.7.1 which states that the discussb. n and and Associated SR is not required to be met for justification for the .

i- Bases vacuum breakers that are open for Less Restrictive I Surveillances. This Note is not change. ,

included in the CTS. There is

inadequate discussion and justification for this Less

!- Restrictive exception to the Surveillance Requirement.

DAEC RESPONSE: See Response to NRC Question 3.6.1.6-1. In addition, the new Note could actually be characterized as an extension of the exsisting CTS footnote to TS 3.7.E.1 and 3.7.E.3, as opening for surveillance testing could be characterized as

performing their intended function." NRC approvat of the existing footnote is documented in the SER for DAEC Amendment 201.

l.

Nonetheless, this change is properly classified as Administrative.

3.6.1.7-2 A.2 ITS SR 3.6.1.7.1 A Note is added to ITS SR 7/8/97' Correct A.2 to delete and Associated 3.6.1.7.1 which states, that the the reference to ITS !

Bases SR is not required to be net for SR 3.6.1.7.3. i i vacuum breakers that are open ,

! g for Surveillances. A.2 states that -l i- the Note applies to ITS SR l

3.6.1.7.1 and ITS SR 3.6.1.7.3.

This is incorrect. The Note only i

applies to ITS SR 3.6.1.7.1.

DAEC RESPONSE: This is identical to the issue raised in Question 3.6.1.6-2. See Response to that Question. i i

1 i !

L t

1 l i

i-

DAEC ITS 3.6.1.7 SUPPRESSION CHAMBER-TO-DRYWELL VAClui BREAXERS i

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.1.7-3 L2 CTS 4.7.E.3 CTS 4.7.E.3 requires a visual 7/8/97 Provide discussion inspection of the drywell-to- and justification on suppression chamber vacuum how changes to the breakers. Visualinspection of the plant procedures and drywell-to-suppression chamber polkies wi!! be vacuum breakers is moved to cc ntrolled.

plant procedures and policies.

There is no discussion or justification of the specific procedure or policy and how changes to these procedures and policies are controlled.

DAEC RESPONSE: This issue is analogous to the one raised in NRC Question 3.6.1.6-5. That I>sponse is also applicable to this Question, as this CTS item is being " deleted" and thus, removed from regulatory contr-Is.

3.6.1.7-4 L3 CTS 3.7.E.4 CTS 3.7.E.4 and 4.7.E.2 provide 7/8/97 Provide discussion CTS 4.7.E.2 ACTIONS and a Survei!!ance and justification on flequirement for the suppression how changes to the chamber-to-drywell vacuum plant procedures and breaker position indication policies are instrumentation. Position controlled.

indication instrumentation is moved to plant procedures and policies. There is no discussion or justification of the specific proceduru or policy and how changes to these procedures and policies are controlled.

DAEC RESPONSE: This issue is analogous to the ones raised m NRC Questions 3.6.1.6-5 and 3.6.1.7-3. Those Respons. ', are also applicable to this Question, as this CTS item is beinq " deleted" and thus, removed from regulatory controls.

2

DAEC ITS 3.6.1.7 SUPPRESSION CHMBER-TO-DRYWELL VACUUM BRE#XERS ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.1.7-5 P.12 STS SR 3.6.1.8.1 STS SR 3.6.1.8.1 requires the 7/8/97 Delete this generic Bases ITS SR vacuum breakers be verified change.

P.8 3.6.1.7.1 and closed every 14 days and af ter Associated Bases any discharge of steam or any operation causing a vacuum breaker to open. ITS SR 3.6.1.7.1 deletes the second frequency (steam or operational opening). The justification (P.12) states that this frequency 's not needed since ITS SR 3.O.1 would not be met and appropriate actions taken. The justification also states that if conditions exist for the vacuum breakers to be potentially opened, control room operators would be alerted to the possibility and would ensure the vacuum breakers were closed at the completion of the evolution.

The SR frequency assures that this is done. The staff has determined based on the justification that this is a generic change which is beyond the scope of review for this conversion.

DAEC RESPONSE: The second surveillance frequency is not part of the CTS, i.e., current licensing basis (CLB). Existing control room indication of valve position should alert operators if a valve is incorrectly in an open position. Thus, per SR 3.0.1, the proper LCO Actions will 5e taken, as discussed in JFD P.12. These considerations, which were the basis for granting WNP-2 the same change (Ref NRC SER), are also valid at DAEC. Because this change is consistent with the DAEC's CLB, it is plant-specific, not withstanding the WNP-2 commitment to pursue a neneric traveller and should be approved for the conversion.

3

DAEC ITS 3.6.1.7 SUPPRESSION CHAMBER-TO-DRYWELL VACUUM BREAKERS ITEM DOCI CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO JFD LCO OPENED CLOSED 3.6.1.7-6 P.13 STS SR 3.G.1.8.2 STS SR 3.6.1.8.2 requires a 7/8/97 Delete this generic Bases ITS SR 3.6.1.7.2 functional test of the vacuum change.

P.8 and Associated breakers within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of any Bases discharge of steam into the suppression chamber and following any operation that causes the vacuum breaker to open. ITS SR 3.6.1.7.2 deletes these frequencies / conditions.

STS SR 3.6.1.8.2 covers a!!

aspects of valve opening including any other unexpected event which would open the vacuum breakers.

Thus, the staff considers the proposed change as a generic change and beyond the scope of review for this conversion.

DAEC RESPONSE: This issue is similar to the one raised in Question 3.6.1.7-5, above. (See also that Response). The surveillance frequencies in question are not part of the CTS, i.e., Current Licensing Basis (CLB). As d5 cussed in JFD P.13, the DAEC has installed "T-Quenchers" on the SRV discharge lines, that lessen the challenges to the Vaccum breakers efter SRV discharges. Consequently, this change is considered to be plant-specific and is consistent with the DAEC's CLB and should be approved.

3.6.1.7-7 P.2 ITS 3.6.1.7 and Ses item Number 3.6.1.4-1 7/8/97 See item Number Ba:;es Associated Bases 3.6.1.4-1.

P.8 DAEC RESPONSE: See Response to NRC Question S3.6.1.4-1. No channes to the ITS are planned.

4

.~

DAEC ITS 3.6.1.7 SUPPRESSION CHAMBER-TO-DRWELL VACUUM BREAKERS 1

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE OMMENTS NO. JFD LCO OPENED CLOSED 3.6.1.7-8 Bases STS B3.6.1.8 ITS B3.6.1.7 Bases- 7/8/97 See item Number P.1 Bases -- APPLICABILITY makes a number S3.6.2.4-1.

APPLICABILITY of changes to the STS B3.6.1.8 ITS B3.6.1.7 Bases-APPLICABILITY discussion.

Bases The changes deal with the APPLICABILITY operation of the Containment / Suppression Pool Spray System. The acceptability of these changes will depend on i the resolution of item Number S3.6.2.4-1.

DAEC RESPONSE: See Res mnse to Questions S3.6.2.4-1 and 3.6.1.6-9.

3.6.1.7-9 None STS SR STS SR 3.6.1.8.3 has a 7/8/97 Change the 3.6.1.8.3 Frequency of 118] months. ITS Frequency for ITS SR ITS SR SR 3.6.1.7.3 changes the 3.6.1.7.3 to 18 3.6.1.7.3 and Frequency to 24 months. months. Previde any Associated Bases Although this is a bracketed item additional discussion change, the 24 month period is and justification as i not in the current licensing btsis. necessary.

The current licensing basis Frequency for refueling outage frequencies is 18 months. The change in SR Frequencies from 18 months to 24 months has already been determined to be a beyond scope of review item for this conversion.

DAEC RESPONSE: Based upon our meeting with the NRC Staff on Septer iber 10.1997, we understand that the technical resources have been applied to review our conversion from ars 18 month to 24 month operating cycle in parallel with our ITS conversion.

Consequently, this item is considered to be back within scope for the purposes of our Final Safety Evaluation on the iTS conversion.

In addition, because the CTS does not currently contain this surveillance, its addition, at any frequency, is a More-Restrictive change (Ref. DOC M 1 to ITS 3.6.1.7). Th:s SR can only be performed with the plant shutdown and the containment de-inerted (Ref. .JFD P.20 to the Bases). Thus, the 24 month freque;'.cy is consistent with the changes to rest of the ITS for SRs that can only be performed during outages.

S

._. . . . . . _ _. _ . . _ _ _ _ . .- - _ . - _ .__ ._ m _-_ _ .. .

DAEC ITS 3.6.2.1 SUPPRESSION POOL AVERAGE TEMPERATURE 4 :

' DATE ITEM DOC / CTS /STS D3CRIPTION OF ISSUE DATE COMMENTS NO. JFD LCO OPENED CLOSED l j- 3.6.2.1-1 A.2 CTS CTS 3.7.G.2.b and 3.7.G.2.c requires the 7/8/97 Provide P.27 3.7.G.2.b plant brought to Hot Shutdown followed additionas j Bases CTS by Cold Shutdown if the suppression pool discussion and 1

P.8 3.7.G.2.c temperature is not reduced to s 95*F justification for '

A ITS 3.6.2.1 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . ITS 3.6.2.1 RA B.1 this Less

RA B.1 requires THERMAL POWER be reduced to Restrictive ;

s 1% RTP within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The ITS change. See ;

ACTION of reducing power to s 1% is Less item Numbers j i- Restrictive than the CTS ACTION of 3.6.2.1-2 and i i bringing the plant to COLD SHUTDOWN. 3.6.2.1-3. t l DAEC RESPONSE: Per ITS LCO 3.0.1, LCOs only have to be met in the Modes of Applicability or other specified conditions.

1 Once the defined Applicability is exited, the LCO is no longer required to be met including any Actions that extend beyond i the Applicability requirements. The Applicability of CTS 3.7.G.2.b is "during normal power operation." CTS Definition -

4 1.0.8 defines " power operation as "... above 1 % rated power. " Therefore, in order to exit the Applicability of CTS ,

! 3.7.G.2.b, the plant must only be taken to s 1.5 RTP and the 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowed to reach Hot Shutdown (the next lower l mode of operation) is also allowed to reach s 1% RTP. The CTS requirements to continue to Hot Shutdown and Cold Shutdown do not have to be met since the CTS LCO is no longer applicable. ITS 3.6.2.1 Action B implements the actual

, requirements'of the CTS by requiring the plant to reduce Thermal Power to s 1% RTP within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This change is '

I considered to be Administrative, consistent with our characterization of the addition of LCO 3.0.1 (DOC A.2 to Section 3.0) and the conversion of the CTS Definition of " Reactor Power Operation" to the ITS Definition of Modes 1 ano 2 as Administrative (DOC A.26 to Section 1.0). l 5

l

I

?

i

!I .

I 1 i

DAEC ITS 3.6.2.1 SUPPRESSION POOL AVERAGE TEMPERATURE ITEM DOC / CT, TS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.2.1-2 A.2 CTS CTS 3.7.G.2.a requires a maximum 7/8/97 Proside P.27 3.7.G.2.a average suppression chamber temperature additional Base CTS of 95'F during normal power operation. discussion and P.8 3.7.G.2.c CTS 7.G.2.c requires a maximum average justification for ITS LCO suppression chamber temperature of 105' this Less 3.6.2.1.a during testing which adds heat to the Restrictive ITS LCO suppression chamber. ITS LCO 3.6.2.1.a change. See 3.6.2.1.6 requires suppression pooi average item Numbers and temperature is s 95'F with THERMAL 3.6.2.1-1, Associated POWER > 1 % RTP and performing no 3.6.2.1-3, and Bases testing that adds heat to the suppression 3.6.2.1-5.

pool. ITS LCO 3.6.2.1.b requires suppression pool average temperature s 105'F with THERMAL POWER > 1% RTP and testing that adds heat to the suppression pool. Adding a specific THERMAL POWER level limits to these CTS LCOs is a Less Restrictive change and was inadequately discussed and justified.

See item Numbers 3.6.2.1 -1, 3.6.2.1 -3, and 3.6.2.1-5.

DAEC RESPONSE: The Applicability of CTS 3.7.G.2.a is "during normal power operation." CTS Definition 1.0.8 defines power operation as ". . above 1 % rated power." Therefore, the CTS 3.7.G.2.a requirement of "during normal power operation" is the same as ITS LCO 3.6.1.2.a requirement of "when THERMAL POWER > 1% RTP." While the phrase "during normal power operation" (i.e. > 1 % RTP) is not included in CTS 3.7.G.2.c. it is readily inferred from the CTS LCO, "At any time the nuclear system is pressurized above atmospheric, " (i.e., coolant temperature 2212'F), and the fact that the subject is " testing which adds heat to the suppression pool," which means that the reactor must be at a power level high enough to support this testing. Per JFD P.27 and the Bases to 3.6.2.1 the " point of adding heat" to the reactor coolant is roughly 1 % RTP. Thus, this change is a presentation preference (i.e. > 1% RTP) and is a direct conversion of the CTS to ITS.

2

DAEC ITS 3.6.2.1 SUPPRESSION POOL AVERAGE TEi!PERATURE ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED Ct.OSED 3.6.2.1-3 A.2 CTS CTS 3.7.G.2.a requires a maximum 7/8/97 Delete this P.27 3.7.G.2.a average suppression chamber temp rature generic change.

Bases CTS of 95'F during normal power operation. See item P.8 3.7.G.2.b CTS 7.G.2.c requires a maximum average Numbers CTS suppression chamber temperature of 105'F 3.6.2.1-1, 3.7.G.2.c during testing which adds heat to the 3.6.2.1-2, and STS LCO suppression chamber. STS LCO 3.6.2.1.a 3.6.2.1-5.

3.6.2.1 requires a s sppression pool average STS 3.6.2.1 temperature be s 95'F when any ACTIONS OPERABLE intermediate range monitor and (IRM) channelis 2 25/40 divisions of full Associated ' scale on Range 7, while STS LCOs Bases 3.6.2.1.b and c require a suppression pool ITS LCC average temperature be s 105"F when 3.6.2.1 any IRM channelis 2 25/40 divisions on ITS 3.6.2.1 Range 7 and s 110 F when allIRM Condition A channels are s 25/40 divisions on Range 7.

ITS 3.6.2.1 ITS 3.6.2.1 changes the IRM criteria to 1 % j RA B.1 RTP. Both STS B3.6.2.1 Bases-LCO and ITS 3.6.2.1 P.27 state that 1% RTP is not readily Condition C quantified with rnuch accuracy. However, and the Bases states that 25/40 divisions of Associated full scale on IRM Range 7 is a convenient Bases measure of when reactor is providing power essentially equivalent to 1 % RTP.

Since 1% RTP cannot be readily quantified with much accuracy the STS specifies an acceptable means to determine this.

Therefore, the staff finds the ITS change unacceptable and generic. See item Numbrs 3.6.2.1-1, 3.6.2.1-2 and 3.6.2.1-5.

3

DAEC ITS 3.6.2.1' SUPPrtESSION' POOL ' AVERAGE TEMPERATURE ITEM . DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS .

NO. JFD LCO OPENED CLOSED - ~ g - ,

DAEC RESPONSE: As stated in the Bases for LCO 3.6.2.1, the technical basis for the LCO limits are 1% RTP, per the -

Referenced analyses. The Reviewer's comment is accurate in that both the STS Bases and JFD P.27 recognize that 1 %

RTP is difficult to measure. However, the STS only proposes one acceptable method for determining when the LCO is entered, which may not be the best or most practical under all plant conditions (e.g., starting from cold conditions vs start up from hot standby conditions, early in core life vs late in core life, etc.). Hence, the DAEC chose to re-write the LCO based upon the technical (analy+ica!) limit and left the implementation of how to determine when the LCO~was entered to licensee-controls While this change could be considered to be " generic", it should be "within scope" of the DAEC-conversion, as it is also our current licensing basis, as discussed in our Response to Question 3.6.2.1-1 above.

f L

[

L b

t i

h g_

4 ,

DAEC ITS 3.6.2.1 SUPPRESSION POOL AVERAGE TEMPERATURE ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.2.1-4 L.1 CTS CTS 4.7.G.2.c requires an external visual 7/8/97 Provide 4.7.G.2.c inspection of the suppression chamber additional -

whenever there is indication of relief valve discussion and operation with the local suppression pool justifi-cation to temperature reaching 2OO'F or greater. show that L.2 states that ITS 3.6.2.1 does not retain NEDO-30832 this CTS requirement in accordance with has been NEDO-30832, " Elimination of Limit on reviewed and BWR Suppression Pool Temperature for approved by ,

SRV Discharge with Quenchers," dated the staff and December 1984. The discussion and its justification do not indicate if NEDO-30832 applicabilityto has been reviewed and approved by the DAEC.

staff. It also does not indicate its applicability to DAEC. This item may be considered a beyond scope of review item for this conversion.

l l

5

__ _. . . , _ . _- .._m _

i DAEC ITS 3.6.2.1. SUPPRESSION POOL AVERAGE TEMPERATURE f

1

- i

.DATE DATE I

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE COMMENTS '

NO. JFD OPENED CLOSED LC . .-.

DAEC RESPONSE: NRC W te+w and approval of NEDO-30832 is documented in the Safety Evaluation Report (SER) :

L issued on August 29,196. tw e ,a letter from Gary Holahan (NRC) to Robert Pinelli (BWROG), Transmittal of the Safety Evaluation of General Electric Co. Tooical Reoorts: NEDO-30832 entitled " Elimination of Limit on BWR Suporession Pool i Temperature for SRV Discharce with Quenchers" and NEDO-31695 entitled "BWR Suppression Pool Temoerature Technical Specification Limits"). This SER states, "The staff finds that the BWROG has presented sufficient data to demonstrate the capability of the "T" and

X" quenchers to reduce unstable CO loads to a negligible level when suppression pool temperature increases...The staff concluded that the local pool temperature limit may be eliminated provided that SRV discharges are delivered to the suppression pool through a "T" or "X" ques.:her device..."

From the CTS 3.7.G and 4.7.G BASES, "...As part of the program to reduce the loads on BWR containments, the NRC issued NUREG-0783, which limits local suppression pool temperatures during Safety Relief Valve (SRV) actuations.

Experimental data ir.dicate that excessive steam condensing loads can be avoided if the peak local temperature of the !

suppression poolis maintained below 200 degrees F during any period of relief valve operation. The requirement for an '

external visual examination following any event where potentially high loadings could occur provides assurance that no

' significant damage was encountered. ..."

, The CTS surveillance (4.7.G.2.C) relates to potentially high loadings as explained in the CTS BASES. At DAEC, each SRV discharge line tenninates in a T-quencher. NEDO-30832 provides the basis that high loadings will not occur when the SRV discharge is delivered th ugh a T-quencher. 't Therefore, the requirement for visualinspection is no longer required based for the DAEC on NEDO-30832 (which has been reviewed and approved by NRC staff) and T-quenchers being installed at DAEC.

3.6.2.1-5 None ITS LCO ITS 3.6.2.1.c requires suppression pool 7/8/97 Provide a 3.6.2.1.c average temperature s 110*F when discussion and !

and THERMAL POWER s1 % RTP. This More justification for Associated Restrictive requirement is not addressed in the More.

Bases the CTS. See item Numbers 3.6.2.1-2 and Restrictive 3.6.2.1-3. change. See.

Item Numbeis 3.6.2.1-2 and 3.6.2.1-3. .

- DAEC RESPONSE: The CTS markup of page 3.7-14 (page 31 of 47) will be revised to show that LCO 3.6.2.1.c comes from i CTS 3.7.G.2.d. New DOC A.3 has been written to describe this change (attached). The revised CTS mark'up page will be I submitted in the next revision to the ITS. j 6

i

f_ DAEC ITS 3.6.2.1 SUPPRESSI0tl POOL AVERAGE TEMPERATURE' 1

. ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS '

NO. JFD LCO OPENED CLOSED

, 3.6.2.1-6 None CTS CTS 3.7.G.2.e requires a depressurization 7/B/97 Provide a 3.7.G.2.e of the reactor to less than 200 psig within discussion and ITS 3.6.2.1 '12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the suppression pool justification for Condition E average temper .ture is 2120*F. ITS this Less and 3.6.2.1 Condition E requires the same Restrictive Associated required action when the suppression pool change.

Bases average temperature is > 120*F. No ;

justification is provided for the Less Restrictive change from "2120*F to >

120'F."

DAEC RESPONSE: The Reviewer is correct. The CTS markup of page 3.7-14 (page 31 of 47) will be revised to indicate this as a Less Restrictive change, similar to DOC L.2 for the 110'F limit. A new DOC (L.3) and No Significant Hazards Consideration have been included for this change (attached). The revised CTS markup will be submitted with the next ITS i revision.

1

[

3.6.2.1-7 None STS STS B3.6.2.153ases-BACKGROUND lists 7/8/97 Correct this

! B3.6.2.1 four technical concerns that lead to discrepancy. !

l Bases- development of the suppression pool

BACKGROU average temperature. ITS B3.6.2.1 Bases-i ND BACKGROUND only lists two concerns.

l ITS B3.6.2.1 The "and" between the end of concern a.

i Bases- and the beginning of concern b has BACKGROU inadvertently been deleted.

ND DAEC RESPONSE: This correction will be made in the next revision of the ITS.

i i '7 l

DAEC ITS 3.6.2.2 SUPPRESSION POOL WATER LEVEL ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.2.2-1 A.2 CTS CTS 3.7.G.1.a specifies the 7/9/97 Correct ITS LCO 3.6.2.2 P.43 3.7.G.1.a volume of the suppression to specify level in terms Bases STS LCO pool as cubic feet and of ft. Provide any P.8 3.6.2.2 percent. ITS LCO 3.6.2.2 additional justification and ITS LCO specifies the water level as discussion as necessary 3.6.2.2 a volume in cubic feet, ano for this change.

and the level is specified in the Associated Bases as volume (ft') and Bases level (f t). STS LCO 3.6.2.2 oNy specifies levelin ft.

The reason for this is that the control room indication of suppression pool water -

levelis given in ft. not volume (ft'). The STS was written this way as a convenience for the operator and to avoid confusion. Justification A.2 states that suppression pool water level indication in the control room is given in terms of level-ft, not volume. Therefore, the ITS change is unacceptable.

The Bases can specify the level as well as the corresponding volume.

DAEC RESPONSE: The suppression pool limits will be changed back to level from volume to be in accordatice with the NUREG. The new LCO limits are 210.11 ft and s 10.43 ft. These values correspond to the CTS values for suppression pool volume (see DAEC Bases pg. B 3.6-56). DOC A.2 and JFD P.43 have been revised to reflect this change (attached).

The CTS markup page, ITS and Bases revisions will be submitted in the next revision to the ITS.

1

DAEC ITS 3.6.2.2 SUPPRESSION POOL WATER LEVEL ITEM DOC / CTS /STS DESCRIPTION OF ISSUE ' DATE. DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.2.2-2 P 43 - STS LCO STS LCO 3.6.2.2 and its 7/8/97. - Correct this deviation Bases 3.6.2.2 and associated Bases specifies from the writers guide.

P.8 Associated the suppression pool water ~ See item Number 3.6.2.2-Bases level as "1" or "s" specified 1.

ITS LCO limits. ITS LCO 3.6.2.2 and 3.6.2.2 and - its associated Bases Associated changes "z" and "s" to Bases " greater than or equal to" and "less than or equal to",

respectively. This change is not in accordance with 4

the industry writer's guide ,

for TS conversions. See item Number 3.6.2.2-1. , ;

i DAEC RESPONSE: Per Response to Question 3.6.2.2-1 above, the above change will be incorporated.

i i

t L

6 2- .

_ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ r- _ . ~m + . - - . , ~ . . . . . . ,

1

DAEC ITS 3.6.2.3 RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING I

. ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS I

NO. JFD LCO OPENED CLOSED 3.6.2.3-1 P.16 STS SR STS SR 3.6.2.3.1 requires verifying each 7/8/97' Delete this Bases 3.6.2.3.1 RHR Suppression Pool Cooling System change.

P.8 ITS SR manual, . ower operated, and automatic valve 3.6.2.3.1 in the flow path that is not locked, sealed, or and otherwise secured in position is in the correct Associated position or can be aligned to the correct Bases position every 31 days. U1 der the same conditions ITS SR 3.6.2.3.1 requires the same except the manual 'alves are deleted.

The deletion is unacceptabh. The way the specification and associated .3ases is written would allow verification of the valves position through use of a paper verifict. tion rather than a system walkdown as is required 1

by the PWR STSs. Therefore, since this is a -

new specification for DAEC and a paper ;

verification is feasible based on the justific.ation (P.16), the staff requires that the manual valves be verified in the correct position.

+

i DAEC RESPONSE: CTS do not require this SR except as required to meet the definition of OPERABillTY. Walkdowns of systems to verify position of manual valves would add operator radiation dose and is not needed for the reasons stated in ,

JFD P.16. STS SR 3.6.2 3.1 does not specify that administrative means may be used. For this to be true, the specific wording of "by administrative means" would be included in NUREG SR 3.6.2.3.1, similar to the wording of Required Action A.1 to LCO 3.5.3; it is not. Hence, the NUREG does require this SR to be met by a system walkdown. This is consistent i with other SRs in the NUREG (e.g., SR 3.5.1.2, 3.7.1.1, etc.). Since a walkdown of manual valves is not in current -

licensing basis and because current administrative controls adequately control manual valves, this change will be retained for consistency with similar changes made in other sections of the DAEC ITS (e.g., 3.5.1, 3.7.1, and 3.7.2).

t 1

n r---- ,,-a = a_ . -_- - --

II DAEC.ITS3.6'2.3RESIDUALHEATREMOVAL_(RHR)SUPPRESSI0f(POOLCOOLING 4

ITEM - DOC / - CTS /STS DESCRIPTION OF ISSUE - DATE DATE COMMENTS .:

NO. JFD LCO' OPENED- CLOSED-3.6.2.3-2 P.21 ITS 3.6.2.3 The renumbering of ITS 3.6.2.3' ACTIONS 7/8/97. See item :-

( Bases ACTIONS ' will depend on the resolution of item Number ' Number ;

P.8 and 3.6.2.3-3. 3.6.2.3 Associated Bases ,

t DAEC RESPONSE: No renumbering of ITS is required. See Response to NRC Question 3.6.2.3-3.

3 i

f c

-i

-l t

i =l 1

+

i I

I t

2 .;

DAEC ITS 3.6.2.3 RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION P0OL COOLING ITEM DOC / CTS /STS DESCR!PTION OF ISSUE DATE' DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.2.3-3 P.34 STS 3.6.2.3 STS 3.6.2.3 ACTION B requires a shutdown 7/8/97 Delete this -

Bases ACTION B if the ras and associated Completion Times generic P.8 ITS 3.6.2.3 are not met and for two RHR Suppression change.

ACTIONS D Pool Cooling subsystems inoperable (loss of and E function). ITS 3.6.2.3 breaks STS 3.6.2.3 ACTION B up into two ACTIONS - ACTION D

- two subsystrms inoperable (loss of function) and ACTION E- ras and Competion Times not met. ACTION D instead of requiring a shutdown per the STS, requires the restoration of one RHR subsystem to OPERABLE status within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . The justification used (P.34) states that the time is consistent with time provided in NUREG-1433 when both RHR service water (RHRSW) subsystems are inoperable. This is not totally correct. The Bases for the RHRSW states that the time is consistent with both the RHR Suppression Pool Cooling and RHR Suppression Spray. However, only the spray system has the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months loss of function restoration based on the fact that there is alternate means of cooling containment. This is not true in this case. In this case, the staff finds that total loss of RHR Suppression Pool Cooling requires an immediate shutdown. Therefore, the change is unacceptable and is considered a generic change that is beyond the scope of review for this conversion.

3

DAEC ITS 3.6.2.3 RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED DAEC RESPONSE: The reviewer has misunderstood the justification provided by JFD P.34. The proposed change te allow up to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to restore one RHR Suppression Pool Cooling (SPC) subsystem to Operable status when both subsystems are inoperable being done for cons..tency with the similar condition in the RHR Se vice Water (RHRSW) LCO (Condition D to both STS and ITS LCO 3.7.1 - the Reviewer's assertion that only the RHR Sup,ression Pool Spray LCO has the 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months allowance for the loss-of-function Condition in the STS is incorrect). Because toe loss of RHRSW has the same impact or the loss of containment heat removal capability as does RHR-SPC, the Completion Times should be the same, as the impact on plant risk and safety are tha same. While the Reviewer's comment regarding this being a generic change is correct (Susquehana has initiated the traveller-BWROG-40), the proposed change should be approved for the DAEC ITS, as DAEC has voluntarily included the new specification ITS 3.6.2.3, Residual Heat Removal (RHR) Suppression Pool Cooling units ITS. This Spacification is not in the Current Technical Specifications. However, since DAEC has chosen to include this new specification, the DAEC believes that it should be able to obtain at least those allowances the Staff has granted in other similar ITS conversions. The Staff has approved ITS conversions for the Hatch and Peach Bottom plant that include a Completion Time of 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months with both RHR Suppression Pool Cooling Subsystems inoperable. The DAEC design is similar to these plants in this regard. The DAEC has provided adequate justification in JFD P.34, for this change.

3.6.2.3-4 Bases STS STS B3.6.2.3 Bases - RA A.1 states the 7/8/97 Correct this P.1 B3.6.2.3 following: "In this Condition, the remaining discrepancy.

Bases- RHR..." ITS B3.6.2.3 Bases- RA C.1 RA. A.1 decapitalizes the letter "C" in " Condition".

ITS B3.6.2.3 This is incorrect. The sentence is referring to Bases- Condition C; therefore, the "C ' in Condition" RA C.1 should be capitalized.

DAEC RESPONSE: This correction will be made in the next revision of the ITS.

4

DAEC ITS 3.6.2.3 RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION POOL COOLING ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.2.3-5 Bases ITS B3.6.2.3 ITS B3.6.2.3 Bases SR 3.6.2.3.1 is clarified 7/8/97 Delete this P.17 Bases- by a number of statements (Insert 2) which change.

SR allows the valves to be mispositioned for 3.6.2 3.1 surveillance testing and other approved operating system configurations. While the wording of the STS/ITS allows for the various system operating configurations, it does not allow for realignments due to surveillance tests. The staff's position is that if surveillance tests put the system into an alignrnent that is not a normal or emergency operational alignment then the system is considered inoperable and the appropriate ACTIONS shall be taken.

DAEC RESPONSE: Per our Response to the Staff's RAI on this Note (Ref. NG-97-1597) and our meeting with the Staff on September 9,1997, this change (Bases P.17) will be withdrawn. The associated JFD (Bases P.17) has been revised (attached).

5

..DAEC STS 3.6.2.4 RESIDUAL HEAT REMOVAL'(RHR): SUPPRESSION POOL SPRAYL ITEM ' DOC / ' CTS /STS DESCRIPTION OF ISSUE- DATE DATE- COMMENTS:

NO. 'JFD- LCO OPENED CLOSE.

D S3.6.2.4-1 P.17 STS 3.6.2.4 CTS 3/4.5.B " Containment 7/8/97 include CTS 3/4.5.B - ,

Bases and Spray Cooting Capability" in ITS 3.6. Provide P.8 Associated specifies the suppression pool additional' discussions B.ises and drywell spray MODES of and justifications for the RHR System OPERABILITY any changes made to '

requirements. STS 3.6.2.4 the CTS /STS.

specifies the OPERABILITY requirement for the RHR Suppression Pool Soray. ITS 3.6 does not include STS 3.6.2.4 based on the premise that it does not meet the Criterion specified in 10 CFR 50.36(c)(2)(ii). The staff has determined and stated in the ,

Bases of STS 83.6.2.4 that the RHR Suppression Pool Spray System does meet Criterion 3 of 10 CFR 50.36(c)(2)(ii).

E,ince this system was in the CTS and the staff

=

determination is that it meets '

Criterion 3, this specification a should be included in the ITS. *

, However, STS 3.6.2.4 of -

NUREG-1433 rray not be the appropriate TS in DAEC case, j

STS 3.6.1.7 "RHR l'

Containment Spray System" of NUREG-1434 (BWR-6) may be the more appropriate TS to a use.

1 s

.r , .i , .v

DAEC STS 3.6.2.4 RESIDUAL HEAT REMOVAL (RHR) SUPPRESSION' POOL. SPRAY-I TEM DOC / ~ CTS /STS DESCRIPTION OF ISSUE DATE DATE: COMMENTS NO. JFD LCO OPENED CLOSE 1

D.

DAEC RESPONSE: NUREG 1433 was developed based on a lead plant in the BWR/4 line, Hatch Unit 2. Many'of the specifications, parts of sprifications, and descriptions of these specifications in the Bases, do not represent th'e design or .

specific accident analysis assumptions used at the DAEC. Each plant choosing to convert to the improved Technical Specifications (ITS) must apply the four criteria in 10 CFR 50.36 (c) (2) (ii) to their Current Technicsi Specification (CTS)

LCOs in order to determine the composition of the plant specific ITS, not the generic application of the screening criteria to the STS. The application of these cliteria to the CTS LCOs is contained in the " Split Report" which was submitted to the staff in the ITS conversion package (volume 1). The DAEC plant specific accident analysis has been evaluated and this -

discussion is contained in the Split Report. The Split Report states in part, "...in the cmlysis of the bounding event of the -

containment analysis and the suppression pool pressurization due to bypass leakage, the drywell spray mode of RHR was not utilized for mitigation of the event." and "...the use of suppression pool sprays was not assumed in the analysis of the maximum containment bypass leakage, and is not relied upon to mitigate the event." The Split Report for each plant that converts to the ITS must be reviewed by the Staff in order to make a determination of which LCOs are to be included in a particular plant's ITS. Therefore, since none of the Criteria are met for inclusion in the DAEC ITS, neither the drywell sprays nor the suppression pool sprays are required to be included.

r I

2

o DAEC ITS 3.6.3.1 CONTAINMENT' ATMOSPHERE DILUTION (CAD) SYSTEM l

ITEM DOC / ' CTS /STS . DESCRIPTION OF ISSUE DATE DATE COMMENTS l NO. JFD LCO OPENED- C.LOSED i

3.6.3.1-1~ A.2 STS 3.6.3.4 STS 3.6.3.4 RA' A.1 specifies the 7/8/97; Delete this generic L P.33 RA A.1 required action to take if one CAD change.

Bases STS 3.6.3.4 subsystem is inoperable. STS i

P.8 ACTION B 3.6.3.4 ACTION B specifies the ITS 3.6.3.1 required actions to take if two CAD RA A.1 subsystems are inoperable (i.e., total' <

, and loss of function). ITS 3.6.3.1 deletes Associated STS 3.6.3.4 RA A.1 and modifies

Bases STS 3.6.3.4 ACTION B to apply to ' ;

total system inoperable (loss of ;

function). This modification adds a Note to ITS 3.6.3.1 RA A.1 to '

specify that LCO 3.0.4 is not applicable. While STS 3.6.3.4 RA -3 A.1 had this note, ACTION B did not because Condition A only applied to one subsystem inoperable, thus there was a redundant subsystem capable ;

of performing the safety function for ]

, the 30 days required to restore ,

system operability. With totalloss of function there is no backup, i redundant CAD system to rely on, therefore the exception to LCO 3.0.4 was not included in STS 3.6.3.4 I

ACTION B. The same reasoning

would also apply to the ITS changes to RA B.1. In addition the staff would consider this change as a !

aeneric chanoe.

i i

1 ,

1

DAEC ITS 3.6.3.1 CONTAINMENT ATMOSPHERE DILUTION (CAD) SYSTEM ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED DAEC RESPONSE: During our preparation of our Response to this Question, we discovered that the Applicability for DAEC ITS LCO 3.6.3.1 is inconsistent with the current design basis and CTS. The CAD system is sized in accordance with UFSAR section 6.2.5.3 to perform its function from an initial assumption that the containment is already inerted to 55%

oxygen by volume. Until the containment is inerted in accordance with LCO 3.6.3.2 the CAD system is not capable of performing its safety function and thus, is inoperable per LCO 3.0.1. The submittal will be revised to reflect the same LCO Applicability as ITS 3.6.3.2, which is consistent with the CTS and is supported by the current design and licensing basis.

As part of this correction, the LCO 3.0.4 note will no longer be required and will be deleted. DOCS A.2 and M.1, and JFD P.33 have been modified to reflect this change (attached). The CTS mark-up page (page 34 of 47) and ITS and Bases pages will be submitted in the next revision of the ITS.

3.6.3.1-2 P.20 STS SR STS SR 3.6.3.4.2 requires verifying 7/8/97 Delete this change.

Bases 3.6.3.4.2 each CAD subsystem manual, power P.8 iTS SR operated, and automatic valve in the 3.6.3.1.2 flow path that is not locked, sealed, and or otherwise secured in position is in Associated the correct position or can be aligned Bases. to the correct position every 31 days. Under the same conditions ITS SR 3.6.3.1.2 requires the same except the manual valves are j deleted. The deletion is unacceptable. The way the specification and associated Bases is written would allow verification of the valves position through use of a paper verification rather than a system walkdown as is required by the PWR STSs. Therefore since this is a new specification for DAEC and a paper verification is feasible based on the justification (P.20), the staff requires that the manual valves be verified in the correct position.

2

DAEC ITS 3.6.3.1 CONTAINMENT ATMOSPHERE DILUTION (CAD) SYSTEM i

ITEM DOC / CTS /STS DESCRIPTION OF ISSUE . DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED DAEC RESPONSE: The CTS does not currently require any periodic verification of manual valve positions. Valve positions have been controlled adequately in the past by the methods detailed in JFD P.20. This is the Current Licensing Basis and tha change will be retained. See also our Response to NRC Question 3.6.2.3-1 regarding verification of manual value position for further details.

3.6.3.1-3 P.21 ITS 3.6.3.1 The renumbering of ITS 3.6.3.1 and 7/8/97 See item Number Base and succeeding specification will depend S3.6.3.2-1.

P.8 Associated on the resolution of item Number Bases S3.6.3.2-1.

DA5C RESPONSE: See Response to NRC Question S3.6.3.2-1. No change to ITS numbering is planned.

3.6.3.1-4 Bases ITS B3.6.3.1 ITS B3.6.3.1 Bases SR 3.6.3.1.2 is 7/8/97 Delete this change.

P.17 Bases clarified by a number of statements SR (Insert 2) which allows the valves to 3.6.3.1.2 be mispositic,ned for surveillance testing and other approved operating system configurations. While the wording of the STS/ITS allows for the various system operating configurations it does not allow for realignments due to surveillance testing. The staff's position is that if surveillance tests put the system into an alignment that is not a normal or emergency operational alignment then the system is co .sidered inoperable and the appropriate ACTIONS shall be taken.

DAEC RESPONSE: Per our Response to the Staff's RAi on this Note (Ref. NG-97-1597) and our meeting with the Staff on Saptember 9,1997, this change (Bases P.17) will be withdra.vn. The associated JFD (Bases P.17) has been revised (attached).

3

s t

DAEC ITS 3.6.3.l CONTAINMENT ATMOSPHERE DILUTION'(CAD) SYSTEM

- v.

ITEM' DOC / CTS /STS . DESCRIPTION OF ISSUE ' DATE- DATE COMMENTS L NO. JFD LCO OPENED CLOSED 4

3.6.3.1 5- None STS SR STS SR 3.6.3.4.2 states that each 7/8/97 Delete this change, ,

3.6.3.4.2 CAD valve in the flow path is verified or provide. t ITS SR in the correct position. ITS SR additional 3.6.3.1.2 3.6.3.1.2 deletes the words "in the - discussion and -

and flow path." No justification is justification for this Associated provided for this' deletion. However, deletion. ;

Bases. M.2 which adds this SR to the DAEC ITS uses these words and the Associated Bases uses similar wording.

DAEC RESPONSE: SR 3.6.3.1.2 will be revised to state that the SR applies to valves in'the " required" flow path {s) to be consistent with the LCO requirement.' DOC M.2 and JFD P.33 have been modified accordingly (attached). 'Also, the -

"< DOC- M.3 >" in NUREG markup on p.3.6-46 should be < DOC-M.2>, there is no DOC M.3, for ITS 3.6.3.1. The revised ITS page was submitted with our most-recent revision to the ITS (NG-97-2OO8,11/21/97). -t i

4

.-++i- , -

g . . p g v.-, , w. & e-, -,

k DAEC'ITS 3.6.3.2 PRIMARY CONTAlte1ENT OXYGEN CONCENTRATION ITEM DOCi CTS /STS ' DESCRIPTION OF ISSUE' DATE DATE COMMENTS .i NO. JFD' LCO OPENED' CLOSED 3.6.3.2-1 P.21. ITS 3.6.3.2 See item Number 3.6.3.1- - -7/8/97 See item Number .

Bases and 3 3.6.3.1-3 P.8 . Associated Bases DAEC RESPONSE: See Response to NRC Question S3.6.3.2-1. No changes to the ITS numbering are planned.

3.6.3.2-2 Bases STS The changes made to STS 7/8/97 See item Number -

P.1 B3.6.3.3 B3.6.3.3 Bases- S3.6.3.2-1.

Bases- BACKGROUND to delete BACKGROU the references to STS ND 3.6.3.2 in ITS B3.6.3.2 ITS B3.6.3.2 will depend on resolution

Bases of item Number S3.6.3.2- i BACKGROU '1.

ND DAEC RESPONSE: See Response to NRC Question S3.6.3.2-1. No changes to the ITS numbering are planned.

I 4

r l

t

DAEC ITS 3.6.3.2 PRIMARY CONTAINMENT OXYGEN CONCENTRATION ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED S3.6.3.2-1 P.14 STS 3.6.3.2 STS 3.6.3.2 specifies the 7/8/97 Provide additional Bases and requirements and discussion and P.8 Associated surveillances for Drywell justification for this Bases Cooling System Fans. The deletion based on current '

lTS does not contain this lice.1cing bases, system specification. The design or operational justification (P.19) used constraints.

states DAEC does not assume Diywell Cooling System Fans are available to assure adequate mixing.

STS B3.6.3.2 Bases APPL.lCABLE SAFETY ANALYSES states that even though no credit for ;

mechanical mixing is assumed in the analysis, the system does meet Criterion 3 of 10 CFR 50.36(c) (2) Oi), for other reasons.

DAEC RESPONSE: The DAEC Drywell Cooling System fans do not satisfy any of the screening criteria of 10 CFR 50.36.

The DAEC system is non-safety related and is not relied upon to mitigate any transient or accidents and therefore does not satisfy Criterion 3 of 10 CFR 50.36 (c) (2) (ii). Requirements for Drywell Cooling fans are not contained in the Current Technical Specifications for the DAEC. While the Drywell Cooling fans do support maintaining the normal Drywell tempeectures at acceptable levels, LCO 3.6.1.4 adequately addresses this requirement. These fans may be of praater safety sigrdficance for other plant which utilize hydrogen recombiners; however, the DAEC design does not include hydrogen recombine.Es. Thus, this LCO is_ not r.eguired for the DAEC_lTS.

2

DAEC ITS 3.6.4.1 SECONDARY CONTAINMENT ITEM DOC / CTS /STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED 3.6.4.1-1 A.3 CTS 16 CTS 16 ' SECONDARY 7/8/97 Revise tho markup A.8 ITS B3.6.4.1 CONTAINMENT INTEGRITY" to show that the M.4 Bases definition is divided into three definition CTS 16 R.2 parts. The first part is associated " SECONDARY with ITS 1.0 and the change is CONTAINMENT designated A.8 which deletes the INTEGRITY" is being definition from TS. The second relocated to the part is associated with the access Bases of ITS openings and Standby Gas E3.6.4.1. Provide Treatment (SBGT) System (CTS audit:enal discussion 16.a and 16.b respectively) and and justification for the changes are designated M.4 this Less Restrictive and A.3 respectively. M.4 (LA) change. See modifies definition 16.a as iTS SR ltem Numbers 3.6.4.1.1 and SR 3.6A.1.2 while 3.6.4.2-4, 3.6.4.3.-

A.3 moves the definition for SBGT 1, and 3.6.4.3-2.

to ITS 3.6.4.3. The third part is associated with secondary 4 containment isolution valves / dampers (SCIV/D) and the -

char:ge is designated R.2 whici.

states that the majority of CTS 16.c is refor:ated to the Bases.

The markup is incorrect. The entire definition for SECONDARY CONTAINMENT INTEGRITY is moved to ITS B3.6.4.1 Bases in total or summarized in some part of ITS.

l B3.6.4.1 Bases. In 1

DAEC ITS 3.6.4.1. SECONDARY CONT'siNMENT ITEM DOC / CTS!STS DESCRIPTION OF ISSUE DATE DATE COMMENTS NO. JFD LCO OPENED CLOSED addition, the staff considers this relocation to be a Less Restrictive change (LA) rather than a Relocation (R) which is reserved for the movement of whole specifications.

See item Numbers 3.6.4.2-4, 3.6.4.3-1, and 3.6.4.3-2.

DAEC RESPONSE: We have reviewed the CTS markup and believe it to be correct. The information relocated by R.1 to ITS 3.6.4.2 can be found in the third paragraph the Background of ITS Bases 3.6.4.2. The CTS, definition was marked up to denote where the individual items went with their corresponding LCOs. The existing mark up is not incorrect and the proposed revision is merely another acceptable alternative and thus, will not be adopted. In addition, as noted in the onginal submittal cover letter (NG-96-2322) dated October 30,1996, the development of our conversion predated NEl 96-06 and therefore " Relocated Details" are categorized as " Relocated items" in our application. (See also our Response to Question 3.6.1.1-2).

3.6.4.1-2 M.4 CTS 16.a STS SR 3.6.4.1.3 verifies that the 7/8/97 Delete the TSTF 18 P.53 STS SR secondary containment access changes or provide Bases 3.6.4.1.3 doors are closed except when it is additional discussion P.2 ITS SR being used for entry or exit, then and justification for Bases 3.6.4.1.2 at least one door shall be opened. the deviations from P.8 and ITS SR 3.6.4.1.2 and its the STS.

Associated Associated Bases modifies STS Bases 3.6.4.1.3 and its Associated Bases based on plant specific design and TSTF 18. TSTF 18 has been rejected by the staff.

DAEC RESPONSE: Justification for Deviation P.53 has been revised to delete references to TSTF 18 (attached). This deviation is based on the current plant specific licensing basis which does not limit the duration of operation with an open Secondary Containment door provided the redundant Secondary Containment door is closed. Secondary Containment remains Operable when the access opening is closed by any combination of doors. Requiring a plant shutdown transient due to a single door being open for more than four hours when it redundant closure is fully capable of performing the intended function is not appropriate for the DAEC design.

2

DAEC:ITS 3.6.4.1 SECGNDARY CONTAINMaiT i

4- ITEM DOC / CTS /STS DESCRIPTION OF ISSUE' DATE DATE COMMENTS - .

e NO. JFD LCO OPENED CLOSED '

3.6.4.1-3 LCY-2 CTS CTS 4.7.J.1.a requires performing .7/8/97 Delete thIs change. :.

4.7.J.1.a testing of the secondary ITS SR containment capability at each ;

3.6.4.1.3 refueling outage (18 months).

and Under the same circumstance ITS Associated SR 3.6.4.1.3 requires every 24 Bases months. This extension of the Surveillance frequency is considered by the staff as a beyond scope of review item for this conversion. >

DAEC RESPONSE: Based upon our meeting with the NRC Staff on September 10,1997, we understand that the technical resources have been applied to review our conversion from an 18 month to 24 month operating cycle in parallel with our ITS conversion. Consequently, this item is considered to be back within scope for the purposes of our Final Safety :

Evaluation on the ITS conversion.

l 3.6.4.1-4 Bases STS STS B3.6.4.1 Bases- 7/8/97 Delete this generic l P.11 B3.6.

4.1 BACKGROUND

states the change.

Bases- following: "It is possible for the i

! BACKGROU pressure in the control volume to .i j NDITS rise relative to the environmental i

} B3.6.4.1 pressure (e.g., due to pump and t Bases- motor heat load additions)." ITS BACKGROU B3.6.4.1 B:ses-BACKGROUND i l ND deletes inis sentence. The l justification (Bases P.11) state that the sentence is confusing and did not add value to the discussion.

This is an inadequate justification 'i to determine why the sentence is j confusing. Furthermore, the j deletion based on the justification would be considered a generic !

, change which is beyond the scope ,

of review for this conversion. l l

.t i

3 -+

~ ., , , _ ,

. .~9.- ,

.DAEC'ITS 3.6 A 1:.SECONDARI. CONTAINMENT.

ITEM - l DOC / l CTS /STS DESCRIPTION OF ISSUE DATE~ D ATE . - COMMENTS .

NO.

JFD~ LCO' OPENED CLOSED .

DAEC RESPONSE: ' Following discussion with the NRC staff, such clarifications to the Bases are 'below threshold" for' processing generic change requests at this time and a traveller will not be pursued.. The Bases' paragraph, without the-deleted sentence, effectively communicates the required information for the DAEC design and as stated in JFD P.11 to the ;

Bases, is not needed. Thus, the proposed Bases change will be retained.

i I

l 4

J 3 F 1 .

p

'_E 1

1 t

4

.. #V"" 7i-hi+.. . _

DISCUSSION OF CliANGES ITS 3.6.1.1: PRIMARY CONTAINMENT ADMINISTRATIVE C11ANGES Ai All reformatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users.

The reformatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting)is made consistent with the NUREG. Dming NUREG development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or integretational) to the CTS. Additional information has also been added to more fully describe each subsection. This wording is consistent with the NUREG. Since the design is already approved by the NRC, adding more detail does not result in a technical change.

A: CTS 3.7.A.1 contains an exception such that Primary Containment Integrity does not have to be maintained while performing low power physics tests at power levels not to exceed 5 Mw(t). This exception is no longer needed since low power physics testing at DAEC was completed during the Startup Test Program. This change is considered to be administrative in nature.

A3 CTS 3.7.A.I.a contains an allowance that states that Primary Containment Integrity is maintained while in the actions for CTS 3.7.A.2.b,3.7.A.2.c, 3.7.A.2.d and 3.7.B.2. CTS 3.7.A.2.b,3.7.A.2.c and 3.7.A.2.d contain actions for when the primary containment airlock is inoperable and CTS 3.7.B.2 contains an action for when one or more primary containment power operated

. isolation valves are inoperable. The ITS does not contain, in individual TS, these types of provisions to prevent cascading from one LCO to another. ITS LCO 3.0.6 was added to direct the user of the TS to follow '.ndividual TS Actions, where provided, and to follow the supported system's Required Actions if so directed by the support system's TS provisions. Since the CTS provision is maintained by ITS LCO 3.0.6, this change is . onsidered administrative.

A., The definition of Primary Contaimnent Integrity has been deleted from the CTS.

In its place the requirement for primary containment is that it "shall be OPERABLE." This was done because of the confusion associated with these definitions compared to their use in the respective LCO. The change is editorial 1 in that all the requirements along with the remainder of the LCOs in the !

Containment Systems Primary Containment section (i.e., air locks, isolation valves, suppression pool, etc.) are maintained in the ITS and encompass the requirements of the delinition of Primary Containment integrity. Therefore, the j change is purely a presentation preference adopted by the NUREG. l l

DAEC 1 Revision E I i

- - . . .. . - - _ . . - - . - - . - - . - - . _ ~

t l

- DISCUSSION OF CIIANGES-

-ITS 3.6.1.1: PRIMARY CONTAINMENT:

TECliNICAL CIIANGES - MORE RESTRICTIVE D

- None a TECHNICAL CilANGES - RELOCATIONS

~

Ri The details in CTS Definition 15 that constitute Primary Containment Integrity l with respect to blind flanges and manways have been relocated to the Bases of-ITS 3.6.1.1," Primary Containment." These details are not necessary to ensure : ,

, . that the primary containment is maintained Operable. The requirements'ofITS LCO 3.6.1.1 and ITS'SR 3.6.1.1.l are adequate to ensure the blind flanges and manway covers are closed. Changes to the Bases will be controlled in accordance with the proposed Bases Control Program described in Chapter 5 of

= ITS.

4

-TECilNICAL CllANGES - LESS RESTRICTIVil Lcy 2 Generic Letter 91-04, Chances in Technical Snecification Surveillance Intervals to Accommodate a 24-month Fuel Cycle, describes NRC requirements for preparing such license amendment requests. The Generic

Letter indicates that the NRC staff has generically resiewed the extension of surveillance intervals from 18 to 24-months and found that "the effe't on safety is small because safety systems use redundant electrical and mechanical-components and because licensees perfomi other surveillances during plant operation that confirm that these systems and components can perforra their safety functions.~ Nevertheless, Licensees should evaluate the effect on safety .

of an increase in 18-month surveillance intervals to accommodate a 24-month fuel cycle. This evaluation should support a conclusion that the effect on safety is small."

- The Generic Letter specifies the following specific items for review:

Steam Generators Not applicable to DAEC -

Instrument Drift Addressed independent of this review by the

. DAEC Setpoint Control Program "Annendix J Exemotion TS Amendment No. 219 addressed DAEC adoption of Option B to Appendix J. No

- additional review is required in this

. evaluation.

DAEC 2 Revision E

. - .; .. . . - - - x

. DISCUSSION OF CilANGES ITS 3.6.1.1: PRIMARY CONTAINMENT TECHNICAL CIIANGES - LESS RESTRICTIVE (continued) ,

Lcy2 in addition, the Generic Letter indicates Licensee's should review the etTect (cont.) on safety of the extension of other surveillances to ensure that it is supported

. by historical maintenance and surveillance data.

Data was collected for a ten-year period from January 1986 to January 1996 of all deficiencies which occurred for the surveillances for which a frequency

, extension is being sought. The ten-year period was -'ected to ensure a broad oversiew oflong term performance and because a similar comprehensive review was performed in 1986 for preceding years to support changes from 12-month to 18-month intervals.

As a supplemental check, the database for 10CFR50.65 (Maintenance Rule) compliance was reviewed to confirm that equipment performance overall was compatible with a decreased surveillance frequency. The DAEC program for Maintenance Rule includes targets for safety system train availability and reliability compatible with assumptions in the DAEC Probabilistic Safety Analysis (data for the Maintenance Rule is limited to the period since 1991).

Data for the following surveillance tests were reviewed:

Description CTS Section ITS SR SBLC Squib Valve Firing 4.4.A.2.b 3.1.7.7 SBLC Flow Verification 4.4.A.2.c 3.1.7.8 SDV Vent and Drain Cycling- 4.3.B.3 3.1.8.3 Reactor Mode Switch Channel Functional 4.1. A. I 3.3.1.13 RPS Response Time 4.1. A .2 3.3.1.18/3.3.1.19 MSL Radiation Monito.* Logic System Functional 4.2.D.2.c 3.3.6.1.9 ATWS RPT Logic System Functional 4.2.G.2 3.3.4.2.4 RPT Breaker Response Time 4.2.G.3 3.3.4.1.3/3.3.4.1.5 SV Setpoint Verification 4.6.D.1 3.4.3.1 SRV Setpoint Verification 4.6.D.1 3.4.3.1 SRV Manual Opening 4.6.D.3 3.4.3.2-IIPCI Low Pressure Flow 4.5.D. I .e 3.3.1.6 CS Logic System Functional 4.2.B.2.a 3.3.5.1.9 RilR Logic System Functional 4.2.B.2.b 3.3.5.1.9 Containment Spray interlock Logic System 4.2.B.2.c 3.3.6.l.9 Functional llPCI Logic System Functional 4.2.B.2.d 3.3.5.1.9 ilPCl/RCIC Suction Transfer 4.5.D. I .f 3.5.1.7/3.5.3.5 (relocated)

ADS Logic System Functional 4.2.B.2.e 3.3.5.1.9 ADS Simulated Automatic Actuation 4.5.F.I a 3.5.1.8 ADS Valve Manual Opening 4.f..D.3 3.5.1.9 RCIC Low Pressure MM: 3.5.3.4 Drywell to Torus Leak Test

,4.7f4' 3.6.1.1.2 DAEC 3 Revision E

DISCUSSION OF CIIANGES ITS 3.6.1.1: PRIMARY CONTAINMENT.

TECTINICAL CIIANGES - LESS RESTRICTIVE (continued)

I{Y.2 (cont.) ,

Description CTS Section ITS SR PCIV Simulated Automatic Actuation 4.*i .B. I .a 3.6.1.3.6 (Groups 1 - 6,8. 9)

PCIV Logic System Functional Test (Groups 1-6) 4.2.A.2.a - g 3.3.6.1.9 EFCV isolation 4.7.B. I .c 3.6.1.3.7 LLS Valve Manual Opening 4.6.D.3 3.6.1.5.1 LLS Logic System Functional 4.2.B.2.g 3.3.6.3.6/3.6.1.5.2 Secondary Containment Integrity 4.7.J. l .a 3.6.4.1.3 SCIV/D Simulated Automatic Actuations 4.7.K.1 3.6.4.2.2 SBGT Simulated Automatic Actuation 4.7.L. I .d 3.6.4.3.3 River Water Supply Simulated Automatic 4.5.J.l.a 3.7.2.4 4 Actuation ESW Automatic Start w/ DG 4.8.E.1.a 3.7.3.2 SFU Simulated Automatic Actuation 4.10.A.3 3.7.4.3 Control Building Positive Pressure 4.10. A.3 3.7.4.4 LOOP /LOCA Test 4.8.A.2.b 3.8.1.13 Battery Service Discharge 4.8.B. I .c 3.8.4.7 in each of these tests, no tnin failures were identified by performance of the reference cyclic test during he ten-year period reviewed. In each case, the system performance was with r. targets established under the Maintenance Rule. This comLination of nc est failures and acceptable system performance is viewed as a strong indicator that interval extension is acceptable without more detailed review.

For six Surveillance Tests, more than one failure was identified during performance of the test during the ten year interval. These tests were singled out as requiring further review prior to extending the interval.

Diesel Generator and Emergency Service Water Automatic Actuation (ITS SR 3.7.3.2)

- IIPCI System Cycle Operability Test (ITS SR 3.5.1.6)

IIPCI Logic System Functional Test (ITS SR 3.3.5.1.8)

- Safety and Relief Valve Setpoint Verification imd Inspection Tests (3 tests)(ITS SR 3.4.3.1)

The majority of problems associated with failures of the Diesel Generator and l Emergency Service Water amatic actuation are related to personnel or :

DAEC 4 Revision E

DISCUSSION OF CilANGES ITS 3.6.1.1: PRIMARY CONTAINMENT TECIINICAl, CllANGES - 1 FSS RESTRICTIVI? (continued) j 1.cy.2 prowdural errors. The single exception was a failure in a diesel generator ;

(cont.) output breaker. The failures associated with the IIPCI logic system functional [

test include the failure of the turbine control valve to open due to the failure of ,

a newly installed relay, the failure of a pump suction motor-operated valve to i cycle (the valve is routinely cycled by the IST Program and would have been ;

detected at another time), and the failure of the tuibine stop valve to close due to a sticking limit switch. The failures associated with the llPCI System cycle 4 operability test were mainly associated with the inability to reach rated flow within the specified time of 30 seconds, in each case, the system responded within the analyzed 45 seconds. These and the other failures associated with this test would have been identified during the perfbrmance of similar i

quarterly testing. The failures associated with the SRV setpoint verification I and inspection tests include numerous instances of as fbund valves lifting more than 1% below the specified setpoint and a single failure of an SRV being above the 1% setpoint tolerance (see ITS change in setpoint tolerance from -l% to -3%).

For each of these tests, the nature of the failures, corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle tecting resulted in acceptable conditions for interval extension.

The equipment perfonnance supports interval extensions from 18 to 24 months, with a maximum proposed interval of 30 months in each case.

4 t

DAEC 5 Revision E

DISCUSSION OF CilANGES ITS 3.6.1.2: PRIMARY CONTAINMENT AIR LOCKS ADMINJSTRATIVE CIIANGES [

A All refonnatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. The reRimiatting, renumbering, and rewording process involves no technical changes to l the CTS.

Editorial rewording (either adding or deleting) is made consistent with the NUREG.

During NUREG development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the CTS. Additional information has also been added to more fully describe each subsection.11ds wording is consistent with the NUREO. Since the design is already approved by the NRC, adding more detail does not result in a technical change.

A2 A proposed Note has been added to CTS 3.7.A.2.b and 3.7.A.2.c. ITS Required Actions A and 13, Note 1: " Required Actions...are not applicable if... Condition C is entered," and proposed Condition C provide more explicit instructions for proper application of the Actions for Technical Specification compliance. In conjunction with the proposed Specifications 1.3," Completion Times," these Actions provide direction consistent with the intent of the existing Actions for one inoperable air lock door in the air lock. In the ITS Required Actions A and 11 Notes, there is a recognition that if both doors in the air lock are inoperable (Condition C entered),

then an " Operable" door does not exist to be closed (ITS Required Actions A.), A.2, A.3,13.1.11.2, and B.3 cannot be met).

TECilNICAl. CllANGES - MORE RESTRICTIVE None TECilNICAl, Cll ANGES - REl.OCNflONS Rr The details in CTS Definition 15 that constitute Primary Containment Integrity with respact to airlock doors have been relocated to the flases ofITS 3.6,1,2," Primary Containment Air Lock." These details are not necessary to ensure that the primary containment is maintained Operable. The requirements ofITS LCO 3.6.1.2 are adequate to ensure that at least one door in each airlock is closed and sealed when required. Changes to the llases will be controlled in accordance with the Bases Control Program described in Chapter 5 ofITS. {3.6.1.1-1 and 3.6.1.2-1)

DAEC 1 Revision E

DISCUSSION OF CilANGES ITS 3.6.1.2: Pillh1Al(Y CONTAINhiENT AIR I.OCKS TI:CilNICAl, Cil ANGES - 1 ESS ItESTRICTIVE 1.i The " equency for the air lock interlock test, CTS 4.7.A.2.b (ITS SR 3.6.1.2.2) is proposed to be changed from 6 months to 24 months. Typically, the interlock is installed aller each refueling outage, verified Operable with the Surveillance, and not disturbed until the next refueling outage, if the need for maintenance arises when the interlock is required, the performance of the interleck Surveillance would be required following the maintenance. In addition, when an air lock is opened during times the interlock is required, the operator first verifies that one door is completely shut before attempting to open the other door. Therefore, the interlock is not challenged except during actual testing of the interlock Consequently, it should be sullicient to ensure proper operation of the interlock by testing the interlock on a 24 month interval.

Testing of the air lock interk>ck mechanism is accomplished through having one door not completely engaged in the closed position, while attempting to open the second door. Failure of this Surveillance efTectively results in a loss of primary containment Operability. Procedures and training do not allow this interlock to be challenged for ingress and egress. One door is opened, all personnel and equipment as necessary are placeci into the air lock, and then the door is completely closed prior to attempting to open the second door. This Surveillance is contrary to processes and training of conservative operation, in that it requires an operator to challenge an interlock during a hiode when the interlock function is required. The door interlock mechanism cannot be readily bypassed; linkages must be removed, which are under the control of station processes such as temporary modifications, primary containment closure procec'ures, and out of service practices. Failure rate of this physical device is very low based on the design of the interlock.

llistorically, this interlock verification has had its Frequency chosen to coincide with the Frequency of the overall air lock leakage test. According to 10 CFR 50, Appendix J, Option A, this Frequency is once per 6 months. Ilowever, Appendix J.

Option 11, allows for an extension of the overall air lock leakage test Frequency to a maximum of 30 months.

Thereibre,it is proposed to change the required Frequency for this Surveillance to 24 mouths (and, with the allowance of SR 3.0.2, this provides a total of 30 months, which corresponds to the overall air lock leakage test Frequency). In this fashion, the interlock can be tested in a hiode where the interlock is not required. In addition, Note 8 has been deleted since it will not be needed once the Frequency is extended to 24 months.

DAEC 2 Revision E

. ~ - _ - . . - _ - .- -- -- - - _ _ _ - _ _ - -_ - --- .. - .-

l DISCUSSION OF CilANGES ITS 3.6.1.3: PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) :

ADMINISTRATIVE CilANGES Ai All refomiatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. The ;

refonnatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting) is made consistent with the NUREG. ,

During NUREG development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the CTS. Additional information has also been added to more fully describe each subsection. This wording is consistent with the NUREG. Since :

the design is already approved by the NRC, adding more detail does not result in a technical change.

In addition, the PCIV LCO (ITS 3.6.1.3) exempts the reactor building to-suppression chamber vacuum breakers since they are governed by another LCO.

Any changes to the requirements for these vacuum breakers are discussed in the new LCO Discussion of Changes.

A2 CTS 3.7.11.2.b contains Actions with one or more inoperable PCIVs.

The CTS stated Action is to isolate each affected penetration flow path, but the method to be used is not stated. The llases for CTS 3.7.11 and 4.7.B state that isolation barriers that are acceptable include a closed and de activated automatic PCIV, a closed manual valve, a blind flange, or a check valve inside primary containment with flow through the valve secured. The CTS llases methods for isolation match those specified in ITS 3.6.1.3.

A3 CTS 4.7.ll.l.b.2 contains a quarterly test to verify closure times for the MSIVs. The CTS Frequency has been changed in ITS SR 3.6.1.3.5 from " quarterly" to "in accordance with the Inser ice Testing Program." Since the IST frequency is quarterly, this change is administrative.

i i

DAEC 1 Revision E

T DISCUSSION OF CilANGES ITS 3.6.1.3: PRlh1ARY CONTAINhfENT ISOLATION VALVES (PCIVs)

ADhilNISTRATIVE CIIANGES (continued) i A4 Three new notes were added to ITS 3.6.1.3 Actions. The first note (Note 2) !

provides explicit instructions for proper application of the Actions for TS compliance, in conjunction with the proposed Specification 1.3 " Completion ;

Times," this Note provides direction consistent with the intent of the existing j Actions for inoperai'c isolation valves. 1 The second and third notes (Notes 3 and 4, respectively) facilitate the use and understanding of the intent to consider any system affected by inoperable isolation valves, which is to have its Actions also apply ifit is determined to be inoperable.

Note 4 clarifies that there " systems" include the primary containment. With the proposed LCO 3.0.6, this intent would not necessarily apply. This clarification is ,

consistent with the intent and intemretation of the existing TSs, and is therefore considered an administrative presentation preference.

TECIINICAL CHANGES h10RE RESTRICTIVE his CTS 3.7.13.1 requires Operability of the Primary Containment isolation Valves during reactor power operating conditions. The Applicability in ITS 3.6.1.3 has been changed for the PCIVs to also include hiodes 2 and 3 as well as hiodes 4 and 5 when associated instrumentation is required to be Operable per LCO 3.3.6.1 (which adds a hiode 4 and 5 requirement to the RilR Shutdown Cooling System isolation valves). This ensures that the PCIVs are Operable during times when the primary containment penetrations may need to be isolated in hiodes 1,2, and 3 a DilA could cause a release of radioactive material to primary containment. In hiodes 4 and 5 the probability and consequences of these events are reduced due to the pressure and temperature limitations of these hiodes. Therefore, most PCIVs are not required to be Operable. Only those PCIVs which isolate to prevent reactor vessel draindown are required in hiodes 4 and 5.

hi 2 This change proposes to add ITS Required Actions A.2 and C.2 to verify the penetrations that were isolated remain isolated every 31 days for isolation devices outside primary containment and prior to entering hiode 2 or 3 from hiode 4, if primary containment was de-inened while in hiode 4, if not performed within the previous 92 days, for isolation devices inside primary containment. The 31 days is reasonable because the valves are operated under administrative controls and the probabi;ity of their misalignment is low. The frequency for valves inside containment is consideied reasonable in view of the inaccessibility of the valves and other administrative controls ensuring that valve misalignment is an unlikely DAEC 2 Revision E

DISCUSSION OF CilANGES iTS 3.6.1.3: PRlh1ARY CONTAINhiENT ISOLATION VALVES (PCIVs)

~

TECl!NICAl CilANGl!S - hiORE RESTRICTIVE (continued) hi 2 possibility. These Actions are modified by a note that applies to valves and blind (cont.) fianges hicated in high radiation areas, and allows them to be verified by use of i administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these valves, once they have been verified to be in the proper position,is low.

h1 3 CTS 4.7.B.l.a contains a once per operating cycle test of the PCIVs for closure times. The CTS frequency has been changed in ITS SR 3.6.1.3.3 to "in accordance with the Insenice Testing Program." Since the IST frequency is quarterly, this change is more restrictive in the ITS. Ilowever, current operating practice at DAEC is to peribnn quarterly testing.

h1 4 ITS 3.6.1.3 Action G is a new requirement which was added in the event any Required Action and associated Completion Time cannot be met in hiodes 4 and 5.

The plant must be placed in a condition in which the LCO does not apply, in this case, flexible options are provided in Required Actions G.I. Initiate action to suspend Operations with a Potential for Draining the Reactor Vessel (OPDRVs) within the Residual lleat Removal (RilR) Shutdown Cooling System which is required to minimize the probability of a vessel draindown and subsequent potential fission product release. Suspending an OPDRV may require closing the RilR SDC isolation valves. Therefore, an alternative Required Action (G.2)is provided to immediately initiate action to restore the valve (s) to Operable status. This allows RilR to remain in senice while actions are being taken to restore the valve. This is ,

a new requirement and as such is an additional restriction on plant operation.

his Three SRs were added:

SR 3.6.1.3.1 - Verify every 31 days that each I8 inch primary containment purge valve is closed.

SR 3.6.1.3.2 - Verify continuity of the traversing incore probe (TIP) shear isolation valve explosive charge every 31 days.

SR 3.6.1.3.8 - Remove and test the explosive squib from each shear isolation valve of the TIP System in accordance with the Insenice Testing Program.

t DAEC 3 Revision E

\

l DISCUSSION OF CilANGES ITS 3.6.1.3: PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)

TECilNICAl,CllANGES MORE RESTRICTIVE (continued) !

M3 These SRs provide the means of ensuring that the PCIVs are Operable and able to (cont.) perform their safety function which is to provide primary containment isolation. ,

The addition of new SRs constitutes a more restrictive change.

TECilNICAl,CllANGES REl.OCATIONS Ri Tir :letails in CTS Definition 15 that constitute Primary Containment Integrity with resl~ct to PCIVs have been relocated to the llases. These details are not necessary to ensure that the primary containment isolation valves are maintained Operable.

The requirements ofITS 3.6.L3 Required Actions A.1,13.1, C.1 and E.1 and SR 3.6.1.3.1, SR 3.6.1.3.6 and SR 3.6.1.3.7 are adequate to ensure the valves will be in the proper poshion during accident conditions. Changes to the Ilases will be controlled in accordance with the proposed liases Control Program described in ITS Chapter 5.

R2 He details of the surveillance CTS 4.7.II.l.b.1 specifying that all normally open power operated isolation valves (except for the MSIVs, Well Water Supply / Return valves, Reactor fluilding closed Cooling Water Supply /Retum valves and the Containment Compressor SuetioWDischarge valves) shall be fully closed and reopened will be relocated to the plant procedures implementing the requirements of 4

the IST Program. These requirements may be relocated because they are duplicative ofIST requirements. The IST program, required by 10 CFR 50.55a, will satisfy this valve cycling requirement as allowed by plant conditions. Therefore, the cycling requirement is etTectively being met. The IST Program will be controlled in accordance with the proposed IST Program in Chapter 5 of the ITS. Changes to plant procedures implementing the IST Program will be evaluated in accordance with DAEC's 10 CFR 50.59 program. This change is consistent with the NUREG.

R3 The current requirement (CTS 4.7.ll.l.b.2) for power to be < 75% to perform MSIV isolation time testing will be relocated to the plant procedures implementing the requirements of the IST Program for MSIV stroke timing. This detail is not necessary to ensure that the MSIVs are maintained Operable. Limiting the initial reactor power to < 75% helps to prevent a reactor scram during test perfbmiance.

ITS 3.6.1.3 proposed SRs are adequate to ensure Operability of the MSIVs. Any changes to procedures implementing MSIV testing will be evaluated in accordance with the DAEC 10 CFR 50.59 program. This change is consistent with the NUREG.

DAEC 4 Revision E

,,, ,,, -g-- -- -- ,-, , , - .- - . , . . -m= -. ~ ~, - , -r , , n -

, . , . - m, ,-- ., - ,

DISCUSSION OF CilANGES ITS 3.6.1.3: PRIMARY CONTAINMENT ISOLATION val,VES (PCIVs)

Tl!CilNICAL CilANGES - RFl.OCATIONS (continued)

R4 CTS 3.7.ll.4.a contains a list of containment vent / purge valves that are requied to be ;

closed, except during allowed periods of time for inerting/de inerting, testing, etc.

This list of valves will be rekicated to the UFSAR, as this detail is not necessary to ensure that the subject valves are closed when required. Changes to there requirements will be evaluated in accordance with the DAEC 10 CFR 50.59 program.

CTS 4.7.A.l.c contains a list of containment vent / purge valves with resilient seals and groups they are leak tested in. The flow path valves identified as being subject to the primary containment vent / purge valve specification and groups they are tested in will be relocated to the liases. These details are not necessary to ensure the purge valves are maintained Operable. The requirements ofITS SR 3.6.1.3.4 are adequate to ensure the purge valves are leak tested. Changes to the flases will be controlled in accordance with the proposed Bases Control Program described in Chapter 5 of the ITS. This clumge is consistent with the NUREG. {3.6.1.3-5) l Tl! CLINICAL, Cll ANGES - i.ESS RESTRICTIVE ly This change relaxes the Completion Time in CTS 3.7,0.2 from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to isolate the afTected penetration (or restore to operable) if one main steam isolation valve (MSIV) in one or more penetrations is inoperable (due to leakage (ITS 3.6.1.3 Action D) or other reason (ITS 3.6.1.3 Action A)). This will allow a longer period of time to restore the MSIVs to Operable status in order to prevent the potential for 4

a plant shutdown by isolating the main steam line(s). During the additional time >

allowed, a limiting event would still be assumed to be within the bounds of the safety analysis, assuming no single active failure. Allowing this extended time to potentially avoid a plant shutdown, is reasonable and does not represent a significant decrease in safety.

la CTS 4.7.A.I .c requires purge system isolation valve leakage testing at least once every three months. ITS SR 3.6.1.3.4 relaxes this routine testing to every 184 days, but adds a second frequency of once within 92 days if the valve has been opened. The DAEC, currently cycles the vent / purge valves at least once a week in order to maintain consistent stroke times of the valves with llettis actuators.

DAEC 5 Revision E

i DISCUSSION OF CilANGES ITS 3.6.1.3: PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)

TECllNICAl, CilANGES - I.ESS RESTRICTlVE (continued) la While this would invoke the second frequency oilTS SR 3.6.1.3.4 routinely, (cont.) maintaining the CTS frequency, there is no requirement to continue to do so in the ,

future. Thus, the overall change to tk CTS is considered to be less restrictive.

There has only been one instance of a valve failing the quarterly pressure drop test since 1990, when the increased cycling was initiated. Therefore, current successful testing at a three month frequency (which includes the weekly valve cycling) has demonstrated that increasing the test interval up to every 184 days is acceptable.

{3.6.1.3 7}

lo in the event both PCIVs in an open penetration are inoperable, CTS 3.7.11.2, which requires maintaining one isolation valve Operable, would not be met and an immediate shutdown is required by CTS 3.7.11.3. ITS 3.6.1.3 Action Il provides I hour prior to commencing a required shutdown. This proposed I hour period is consistent with the ITS 3.6.1.1 time allowed fbr conditions when the primary containment is inoperable, The proposed change will provide consistency in actions for these various containment degradations.

la Currently, if containment purge valve leakage is not within limits, DAEC would enter CTS 3.7.A.l. Actions which only allows I hour to restore Primary Containment Integrity or enter CTS 3.7.11.2 for PCIVs which allows up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore inoperable valves or isolate the alTected penetration. ITS 3.6.1.3 Required Action E.1 with one or more penetration flow path with one or more containment purge valves not within purge valve leakage limits, allows 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to isolate the alTected flow path. This Completion Time extension is acceptable because of the low probability of an event requiring a containment isolation function during the time purge valve leakage is allowed to exceed the limit. Once the flow path is isolated, it will be verified to be isolated every 31 days to ensure the penetration is not inadvertently reopened. In addition, the valves used to isolate the penetration will be leak tested every 92 days, instead of the normal 184 day test of ITS SR 3.6.1.3.4. This aise ensures the isolated penetration maintains leakage limits.

L3 This change relaxes the Completion Time in CTS 3.7.11.2 from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months for penetrations in which the excess flow check valve is the only PCIV. The Completion Time in ITS 3.6.1.3 Required Action C.1 is reasonable considering it is a closed system and the small pipe diameter of the penetration. This Completion Time extension is considered acceptable because of the low probability of an event requiring a containment isolation function concurrent with a rupture of the piping in the closed system.

DAEC 6 Revision E

DISCUSSION OF CilANGES ITS 3.6.1.3: PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)

TEClINICAl. CllANGES - 1 ESS RESTRICTIVII (continued)

L6 'lhe reasons in CTS 3 ..7 lL4 a that the large primary containment purge and exhaust isolation valves may be opened are proposed to be expanded to also include ALARA or air quality considerations for personnel entry or for Surveillances that require the valves to be open in ITS SR 3.6.1.3.1. This is considered acceptable since these purge and exhaust valves are capable of closing in the environment following a LOCA. This change is consistent with the NUREG.

In CTS 4.7.II.l.a requirements for automatic actuation testing of the PCIVs stipulate a

" simulated" test be performed. The phrase " actual or," in reference to the automatic isolation signal, has been added to this ITS SR 3.6.1.3.6. This allows satisfactory automatic PCIV isolations for other than Surveillance purposes to be used to fulfill the SRs. Operability is adequately demonstrated in either case since the PCIV itself cannot discriminate between " actual" or " test" signals.

L Per our Response to the Staffs RAI on this Note (Ref. NG 97-1597) and our meeting with the StafTon September 9,1997, this change has been withdrawn.

(3.6.1.3 8}

Ley.2 Generic Letter 91-04, Chances in Technical Soccification Surveillance Intervals to Accommodate a 24 month Fuel Cycle, describes NRC requirements for preparing such license amendment requests. The Generic Letter indicates that the NRC staff has generically reviewed the extension of surveillance intervals from 18 to 24 months and ibund that "the effect on safety is small because safety systems use redundant electrical and mechanical components and because licensees perfbrm other surveillances during plant operation that confirm that these systems and components can perfonn their safety functions. Nevertheless, Licensees should evaluate the effect on safety of an increase in 18 month surveillance intervals to accommodate a 24 month fuel cycle. This evaluation should support a conclusion that the effect on safety is small."

i l

DAEC 7 Revision E

-w -

+.,--. ., . , . .y . , , , - . . . ,..gn,.---..-.,-,.-.m_... ~. ,.7+ -

, ,. . . , , , , ,._,~-...we,_. ,..,,,-n_.,..,w,- ...,,.,.,--r,-., n, w,,,--,

DISCUSSION OF CilANGES ITS 3.6.1.3: PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)

TECllNICAl, CllANGES - 1.ESS RESTRICTIVE (continued) lo.2 The Generic Letter specifies the following specific items for review:

(cont.)

Steam Generators Not applicable to DAEC

Jnstrument Drift Addressed independent of this review by the DAEC Setpoint Control Program

+ Annendix J Exemntion TS Amendment No. 219 addressed DAEC adoption of Option B to Appendix J. No additional review is required in this evaluation.

In addition, the Generic Letter indicates Licensee's should review the effect on safety of the extension of other surveillances to ensure that it is supported by historical maintenance and surveillance data.

Data was collected for a ten year period from January 1986 to January 1996 of all deficiencies which occurred for the surveillances for which a frequency extension is being sought. The ten year period was selected to ensure a broad overview of long tenu performance and because a similar comprehensive review was perfbrmed in 1986 for preceding years to support changes from 12 month to 18-month intervals.

As a supplemental check, the database for 10CFR50.65 (Maintenance Rule) compliance was reviewed to confinn that equipment peribnnance overall was compatible with a decreased sur'.eillance frequency. The DAEC program ihr Maintenance Rule includes targets for safety system train availability and reliability compatible with assumptions in the DAEC Probabilistic Safety Analysis (data for the Maintenance Rule is limited to the period since 1991).

DAEC 8 Revision E

DISCUSSION OF CilANGES ITS 3.6.1.3: PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs)

TECilNICAl, CllANGES - 1 ESS RESTRICTIVE (continued) l<v.2 Data for the following surveillance tests were reviewed:

(cont.)

Description CTS Section iIS SR SilLC Squib Valve Firing 4.4.A.2.b - 3.1.7.7 SilLC Flow Verification 4.4.A.2.c 3.1.7.8 SDV Vent and Drain Cycling 4.3.11.3 3.1.8.3 Reactor Mode Switch Channel functional 4.1.A.1 3.3.1.13

, RPS Response Time 4.1.A .2 3.3.1.18/3.3.1.19 MSL Radiation Monitor Logic System i unctional 4.2.D.2.c 3.3.6.1.9 Al WS RPT Logic System Functional 4.10.2 3.3.4.2.4 RPT lireaker Response Time 4.2.0.3 3.3.4.1.3/3.3.4.1.5 SV Setpoint Verification 4.6.D. I 3.4.3.1 SRV Setpoint Verification 4.6.D. I 3.4.3.1 SRV Manual Opening 4.6.D.3 3.4.3.2 11101 Low Pressure Flow 4.5.D. I .e 3.5.1.6 CS Logic System i unctional 4.2.ll.2.a 3.3.5.1.9 RilR Logic System Functional 4.2.ll.2.b 3.3.5.1.9 Containment Spray Interlock Logic System 4.2.ll.2.c 3.3.6.1.9 Functional llPCI Logic System Functional 4.2.ll.2.d 3.3.5.19 llPCl/RCIC Suction 1ransfer 4.5.D.I.f 3.5.1.7/3.5.3.5 (relocated)

ADS Logic System functional 4.2.ll.2 e 3.3.5.1.9 ADS Simulated Automatic Actuation 4.5.F. I .a 3.5.1.8 ADS Valve Manual Opening 4.6.D.3 3.5.l.9 RCIC Low Pressure 4.5.li. l .c 3.5.3.4 Dryw ell to Torus Leak Test 4.7.li.4 3.6.1.1.2 PCIV Simulated Automatic Actuation 4.7.ll. l .a 3.6.l.3.6 (Groups ! 6. 8. 9)

PCIV Logic System functional Test (Groups 16) 4.2.A.2.a - g 3.3.6.1.9 liFCV isolation 4.7.ll. l .c 3.6.1.3.7 LLS Valve Manual Opening 4 ').D.3 3.6.l.5.1 LLS Logic System Functional 4.2.ll.2.g 3.3.6.3.6/3.6.1.5.2 Secondary Containment integrity 4.7.J. l .a 3.6.4.1.3 SCIV/D Simulated Automatic Actuations 4.7.K.I 3.6.4.2.2 I

1 DAEC 9 Revision E

~ _ ._. , ._ . _ . _ _ . _.. ..

I i

i DISCUSSION OF CilANGES ITS 3.6.1.3: PRIMARY CONTAINMENT ISOLATION VALVES (PCIVs) l

! I TECIINICAL CllANGES - 1.ESS RESTRICTIVE (continued)

Lev.2 I (cont.)

Description . CIS Section iIS SR f SOOT Simulated Automatic Actuation 4.7.L. I .d 3.6.4.3.3 ,

River Water Supply Simulated Automatic Actuation 4.$.J.l.a 3.7.2.4 l LSW Automatic Start w/ IX3 4 8.li.l.a 3.7.3.2 !

SI'U Simulated Automatic Actuation 4.10.A.3 3.7.4.3 -i Controf fluilding Positive Pressure 4.10.A.3 3.7.4.4 i LOOP /LOCA lest 4.8.A.2.b 3.8,l.13 Itattery Service Discharge 4.8.II.l.c 3.8.4.7 In each of these tests, no train failures were identified by performance of the ;

reference cyclic test during the ten yw period reviewed. In each case, the system ,

performance was within targets established under the Maintenance Rule. This combination of no test failures and acceptable system performance is viewed as a strong indicator that interval extension is acceptable without more detailed ;

review.

For six Surveillance Tests, more than one failure was identified during performance of the test during the ten year interval. These tests were singled c ut ,

as requiring further review prior to extending the interval.

Diesel Generator and Emergency Service Water Automatic Actuation (ITS SR 3.7.3.2)

IIPCI System Cycle Operability Test (ITS SR 3.5.1.6) .

IIPCI Logic System Functional Test (ITS SR 3.3.5.1.8) !

4

'

Safety and Relief Valve Setpoint Verification and Inspection Tests (3 tests)(ITS SR 3.4.3.1) i The majority of problems associated with failures of the Diesel Generator and i Emergency Service Water automatic actuation are related to personnel or i procedural errors. The single exception was a failure in a diesel generator output ,

breaker. The fiillures associated with the llPCI logic system functional test l include the failure of the turbine control valve to open due to the failure of a j newly installed relay, the failure of a pump suction motor-operated valve to cycle i DAEC 10 Revision E l i

-m_ , ,. -, ,_s ,__..m_ _ _ . _ _ _ . _ , - .- , - _ , , . ~ . _ ., . . . _ _ _ - , . . . . - . - . . - _ . ,.

t DISCUSSION OF CilANGES ITS 3.6.1.3: PRIMARY CONTAINMENT ISOLATION VAINES (PCIVs)

I IEClINICAI. CilANGES - 1 ESS RESTRICTIVE (continued)

Ley.2 (the valvc is routinely cycled by the IST Program and would have been (cont.) detected at another tin.e), and the failure of the turbine stop valve to close due to a i sticking limit switch. The failures associated with the llPCI System cycle

operability test were mainly associated with the inability to reach rated flow

, within the sfecified time of 30 seconds. In each case, the system responded within the analyzed 45 seconds. These and the other failures associated with this test would have been identiFad during the performance of similar quarterly testing. The failures associated with the SRV setpoint verification and inspection tests include numerous instances of as found valves lifling more than 1% below the specified setpoint and a single failure of an SRV being above the 1% setpoint tolerance (see ITS change in setpoint tolerance from -l% to 3%).

For each of these tests, the nature of the failures. corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension.

The equipment perfcmnance supports interval extensions from 18 to 24 months, with a maximum proposed interval of 30 months in each case.

l I

l i

l DAEC 11 Revision E l

. , , . , , . ~ . . . . _ , , . _ , _ _ , . - _ , . , . . _ . _ _ . . , _ . . . . . . , , _ _ _ , _ , , _ _ , , , , _ , _ _ _ _ _ , , _ _ _ , . _ , _ . _ , _ _

DISCUSSION OF CilANGES ITS 3.6.1.4: DRYWELL AIR TEMPERATURE ADMINISTRATIVE CilANGES Ai All reformatting and renumbering is in accordance with the NUREO. As a result, the ITS should be more readable and more understandable by its users. The reformatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting)is made consistent with the NUREG.

During NUREG development ceitain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the CTS. Additional information has also been added to more fully describe each subsection. 'Ihis wording is consistent with the NUREG. Since the design is already approved by the NRC, adding more detail does not result in a technical change.

TFCilNICAl, CilANGl!S MORii RESTRICTIVE None TECIINICAl, Cil ANGES - REl.OCATIONfi None TEC1INICAl, CllANGES - I.ESS RESTRICTIVJi None I

DAEC 1 Revision E

DISCUSSION OF CilANGES ITS 3.6.1.5: LOW LOW SET (LLS) VALVES ADMINISTil6TIVE CilANGES Ai All refonnatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. The afonnatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting) is made consistent with the NUREG.

During NUREG development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or 4

interpretational) to the CTS. A dditional infonnation has e' . been added to more fully describe each subsection. This wording is consistent with the NUREG. Since the design is already approved by the NRC, adding more detail does not result in a 4 technical change.

A2 CTS 4.2.II.2.g contains Logic System Functional Tests (LSFT) once per operating cycle for the Low Low Set Function. Since the CTS definition of LSFT includes the actuated device, the CTS test includes the provisions ofITS SR 3.6.1.5.2 to verify the LLS System actuates on an actual or simulated automatic initiation signal. .

Iloth the CTS and ITS tests do not include valve actuation. Therefore, this change is administrative in nature.

TECllNICAl, Cil ANGES - MORE B ESTRICTIVE Mi CFS 4.6.D.3 requires each Low-Low Set valve to be verified to open when manually actuated with reactor pressure 2100 psig and tmbine bypass flow to the main condenser. The proposed change in ITS SR 3.6.1.5.1 will replace this requirement with a Note that states that the Surveillance is not required to 1 c perfonned until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months aller reactor steam pressure and flow are adequa.e to perfonn the test. This change applies a time limit for perfonnance of the Surveillance which constitutes a more restrictive change.

M2 CTS 4.2.ll.2.g and CTS 4.6.D.3 contain SurveiHance Requirements for the Low-Low Set (LLS) valves. Ilowever, the CTS does not contain a specific LCO, Applicability or Actions for LLS valves. The addition of these ITS 3.6.1.5 provisions is considered more restrictive.

DAEC 1 Revision E l

l DISCUSSION OF CilANGliS ;

ITS 3.6.1.5: LOW l.OW SET (LLS) VALVES Tl!CilNICAI,CilANGES RELOCATIONS ;

Ri CTS 4.6.D.3 contains specific details on how to verify that the Low-Low Set (LLS) valve has manually opened. These details are not necessary to ensure the Operability of the LLS valves. The requirements ofITS 3.6.1.5, Low Low Set valves, and the associated Surveillance Requirements are adequate to ensure the LLS valves are maintained Operable. These details on how to perfbrm Surveillances are being relocated to the 11ases and plant procedures. Any changes to the liases will be controlled in accordance with the TS 13ases Control Program in ITS Chapter 5 and changes to plant procedures will be evaluated in accordance with the DAEC 10 CFR 50.59 program. This change is consistent with the NUREG.

TECilNICAL CilANGES - 1 ESS RESTRICTIVE Ley.2 Generic Letter 91-04, Chances in Technical Soecification Surveillance Intervals to Accommodate a 24 month Fuel Cvels, describes NRC requirements ihr preparing such license amendment requests. The Generic Letter indicates that the NRC staff has generically reviewed the extension of surveillance intervals from 18 to 24 months and found that "the effect on safety is small because safety systems use redundant electrical and mechanical components and because licensees perform other surveillances during plant operation that confirm that these systems and components can perfbrm their safety functions. Nevertheless, Licensees should evaluate the efTect on safety of an increase in 18 month surveillance intervals to accommodate a 24 month fuel cycle. This evaluation should support a conclusion that the effect on safety is small."

The Generic Letter specifies the following specific items for review:

Steam Generators Not applicable to DAEC

Instrument Drift Addressed independent of this review by the DAEC Setpoint Control Program

. Aopendix J Exemotion TS Amendment No. 219 addressed DAEC adoption of Option 11 to Appendix J. No additional review is required in this evaluation.

DAEC 2 Revision E

DISCUSSION OF CilANGES ITS 3.6.1.5: LOW LOW SET (LLS) VALVES TECilNICAL CilANGES I.ESS RESTRICTIVE (continued)

Lem in additior, the Generic Letter indicates Licensee's shculd resiew the effect on (cont.) safety of the extension of other surveill uices to en.sure that it is supported by histoncal maintenance and surveillance data.

Data was collected for a ten year period from January 1986 to January 1996 of all deficiencies which occurred for the surveillances for which a frequency extension is being sought. The ten-year per iod was selected to ensure a broad overview of long tenn performance and because a similar comprehensive review was performed in 1986 for preceding years to support changes from 12-month to 18-month intervals.

As a supplemental check, the database for 10CFR50.65 (Maintenance Rule) compliance was reviewed to confirm that equipment performance overall was compatible with a decreased surveillance frequency. The DAEC program for Maintenance Rule includes targets for safety system train availability and reliability compatible with assumptions in the DAEC Probabilistic Safety Analysis (data for the Maintenetnce Rule is limited to the period since 1991).

Data for the ibliowing surveillance tests were reviewed:

Description ClS Section ! IIS SR SilLC Squib Valve I iting 4A.A.2.b 3.1.7.7 SilLC l' low Verification 4.4. A.2.c 3.1.7.8 SDV Vent t nd Drain Cycling 4 .3.11.3 3.1.8.3 Reactor Mode Switch Channel Functional 4.1.A.1 3.3.1.13 RPS Response ' lime 4.1.A.2 3.3.1.18/3.3.1.1 9

MSL Radiation Monitor Logic System functional 4.2.U.2.c 3.3.6.1.9 ATWh RPT Logic System functional 4.2.G 2 3.3.4.2.4 RPI'Ilreaker Response Time 4.2.G.3 3.3.4.1.3/3.3,4.1 SV Setpoint Verilication 4.6.D. I 3.4.3.1 SRV Setpoint Verilication 4.6.D.I 3.4.3.1 SRV Manual Opcnirg 4.6.D.3 3.4.3.2 IIPCI Low Pressure Flow 4.5.D. I .c 3.5.1.6 CS Logic System Functional 4.2.ll.2.a 3.3.5.1.9 RilR Logic System Functional 4.2.ll.2.b 3.3.5.1.9 Containment Spray Interlock Logic System 4.2.it2.c 3.3.6.1.9 Functional lilTI Logic S) stem Functional 4.2.3.2.d 3.3.5.1.9 DAEC 3 Revisien E

DISCUSSION OF CilANGES 4

ITS 3.6.1.5: LOW-1.0W SET (LLS) VALVES L

TECilNICAL CllANGES - LESS RESTRICTIVE (continued)

I Ley.2

~ (cont.) j i

Description CTS Section 11S SR lil'Cl/RCIC Suction 1ransfer 4.5.D. I .f 3.5.1.7/3.3.3.5 ,

(relocated) 4 ADS Logic System Functional 4.2.ll.2.e 3.3.5.1.9 l ADS Simulated Automatic Actuation 4.5.F. I .a 3.5.1.8 ,

ADS Valve Manual Opening 4.6.D.3 3.5.1.9 RCIC Low Pressure 4.5.E.1.e 3.5.3.4 Drywell to Torus Leak Test 4.7.E.4 3.6.1.1.2 PCIV Simulated Automatic Actuation 4.7.ll.l.a 3.6.1.3.6 (GroupsI.6.8.9) f PCIV Logic System Functionallest(Groups 1-6) 4.2. A.2.a . g 3.3.6.1.9 EFCV liolation 4.7.l!. i .c 3.6.1.3.7 LLS Valve Manual Opening 4.6.D.3 3.6.1.5.1 LLS Logic System Functional- 4.2.ll.2.g 3.3.6.3.6/3.6.1.5

.2 -!

Secondary Containment integrity 4.7.J. l .a 3.6.4.1.3 SCIV/D Simulated Automatic Actuations 4.7.K. I 3.6.4.2.2 SHOT Simule. ed Automatic Actuation 4.7.L.I.d 3.6.4.3.3 .

River Water apply Simulated Automatic Actuation 4.5.J. l .a 3.7.2.4 ESW Autoinat.c Start w/ DG 48 E. I .a 3.7.3.2 ?

SFU Simulated Automatic Actuation 4.10. A.3 3.7.4.3 Control lluilding Positive Pressure 4.10.A.3 3.7.4.4 LOOD/LOCA Test 4.8.A.2.b 3.8.1.13 Ilattery Service Discharge 4.8.H. I .c 3.8.4.7 In each of these tests, no train failures were identified by performance of the reference cyclic test during the ten-year period reviewed in each case, the system performance was within targets established under the Maintenance Rule. This combination of no test failures and acceptable system performance is viewed as a strong indicator that interval extension is acceptable without more detailed review.

For six Surveillance Tests, more than one failure was identif' icd during performance of the test during the ten year interval. These tests were singled out as requiring fur'her review prior to extecding the interval.

DAEC 4 Revision E ,

-mm., ,,-e- -,y,,. --, .--.m,. -y.-~.,-wm,-,.m._, .w w --..,, .-_,-

i DISCUSSION OF CllANGES 11S 3.6.1.5: LOW LOW SET (LLS) VALVES TECllNICAL CllANGES - 1 ESS RESTRICTIVE (continued)

Ley.2 - Diesel Generator and Emergency Senice Water Automatic Actuation (cont.) (ITS SR 3.7.3.2) a llPCI System Cycle Operability Test (ITS SR 3.5.1.5)

+ llPCI Logic System Functional Test (ITS SR 3.3.5.1.8)

+ Safety and Relief Valve Seipoint Verification and Inspection Tests (3 tests)(ITS SR 3.4.3.1)

The majority of problems associated with failures of the Diesel Generator and Emergency Service Water automatic actuation are related 10 personnel or procedural errors. The single exception was a failure in a diesel generator output breaker. The failures associated with the llPCI logic system functional test include the failure of the turbine contiol valve to open due to the failure of a newly installed relay, the failure of a pump suction motor-operated valve to cycle (the valve is routinely cycled by the IST Program and would heve been detected at another time), and the failure of the turbine stop valve to close due to a sticking limit switch. The failures associated with the llPCI System cycle operability test were mainly associated with the inability to reach rated flow within the specified time of 30 seconds. In each case, the system responded within the analyzed 45 seconds. These and the other failures asso:iated with this test would have been identified during the perfonnance of similar quarterly testing. The failures associated with the SRV setpoint verification and inspection tests include numerous instances of as-found valves lifting more than 1% below the specified setpoint and a single failure of an SRV being above the 1% setpoint tolerance (see ITS clumge in setpoint tolerance from -l% to -3%).

For each of these tests, the nature of the failures, corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension.

The equipment perfonnance supports interval extensions from 18 to 24 months, with a maximum proposed interval of 30 months in each case.

DAEC 5 Revision E

DISCUSSION OF CilANGES ,

ITS 3.6.1.6: REACTOR IlUILDING TO-SUPPRESSION CllAM13ER VACUUM 13REAKERS ADMINISTRATIVE CilANGES A All iefonnatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. He refonna' ting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting) is made consistent with the NUREG.

During NUREG development certvin wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the CTS. Additional information has also been added to more fully descrioe each subsection. This wording is consistent with the NUREG. Since the design is already approved by the NRC, adding more detail does not result in a technical change.

A2 CTS 3.7.D for Re.2 tor liailding to-Suppression Chamber Vacuum Ilreakers does not contain en Action if two vacuum breaker assemblies contain one or more valves that are inoperable for opening ITS 3.6.1.6 Action D. If these conditions exist, Primary Containment Integrity requirements would not be met and DAEC would default to CTS 3.7.A.2 which allows 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months for restoration. This is the same completion time allowed in ITS 3.6.1.6 Action D. There is no change to the intent of the CTS and thus th's change in presentation of material is considered administrative.

A3 CTS 3.7.D is being replaced by ITS LCO 3 6.1.6, Reactor lluilding to-Suppression Chamber Vacuum Breakers. The proposed LCO will contain a note stating that:

" Separate Condition Eritry is Allowed for each penetration branch line " This note clarifies that the Conditions r.nd Required Actions that follow may be applied to each of the two reactor building-to suppression chamber vacuum breaker assemblies without regard to the status of the other vacuum breaker assembly. Each vacuum breaker assembly contains two v-lves in series. This note prosides direction consistent with the intent of the CTS and ITS Required Actions. This change is consistent with the NUREG and considered to be administrative.

A4 Two notes have been added to ITS SR 3.6.1.6.1 to clarify that a vacuum breaker is allowed to be open (anc' i.ot fsiling SR 3.6.1.6.1) during Serveillance (e.g., when performing SR 3.6.1.6.2 and SR 3.6.1.6.3) and when performing their intended function. Since it is obvious that Operability is stin maintained, this addition is considered administrative.

DAEC 1 Revision E

-- - -_=~. _

i DISCUSSION OlkilANGES ITS 3.6.1.6: REACTOR BUILDING TO-SUPPRESSION CllAMBER VACUUM BREAKERS TECilNICAl, CilANGES - MORE RESTRICTIVE Mi When a vacuum breaker essembly vaJve is open, CTS 3.7.D.3 requires that the other l ,

valve in the assembly be verified closed within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . This time is being decreased to I hour, to be consistent with the time provided in CTS 3.7.A for when primary containment integrity is not maintained. {3.6.1.6-3) l l

TECilNICA1 CilANGES- REl,0 CATIONS Ri The CTS 3.7.D.1 details comprising wha' " Operable" means (e.g., closed) for the Reactor fluilding-to Suppression Chamtv r Vacuum Breaker assemblies and what a Reactor Huilding-to Suppression Chamber Vacuum Breaker assembly consists of(a vacuum breaker valve and a butterfly isolation valve) are proposed to be relocated to the Bases. The details for vacuum breaker assembly Operability are not necessary in the LCO. The definition of Operability suflices. In addition, the requirement that the vacuum breaker assemblies be closed is also explicitly required in proposed SR 3.6.1.6.1 and is not needed to be repeated in the LCO statement. Changes to the Bases will be controlled in accordance with the proposed Bases Control Program described in ITS Chapter 5.

TECliNICAl, CilANGES - 1.hSS RESTRICTIVE I. CTS 3.7.D.2 and CTS 3.7.D.3 identifies the current required actions if one valve of a Reactor Building to Suppression Chamber Vacuum Breaker assembly is inoperable. If more than one valve in a vacuum breaker assembly is inoperable, the existing specification assumes either containment integrity is lost or the ability to relieve negative pressure in the containment is lost. ITS 3.6.1.6 recognizes that there are two valves in series in each of two vacuum breaker assemblies between the reactor building and suppression chamber. If one vacuum breaker assembly valve will not open, the vacuum breaker assembly is inoperable to perform its relief function, thus the consequences af the second inoperable vacuum breaker assembly valve in the same assembly has no more efTect than the first inoperable vacuum breaker valve, i.e., the vacuum breaker assembly will not perform its relief function.

If two vacuum breaker valves in one vacuum breaker assembly are inoperable but closed (Condition C), containment integrity and venting capability are still maintained and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months is provided to restore the redundant vacuum breaker assembly. This change is consistent with the NUREG.

DAEC 2 Revision E

.- . _ _ _ _ . _ _ - _ - = _ _ - _ _ . _. - _ _ - - - .

DISCUSSION OF CilANGES ITS 3.6.1.6: REACTOR IlUILDING TO-SUPPRESSION CilAMBER VACUUM BREAKERS TECilNICAL CIIANGES - LESS RESTRICTIVE (continued)

I,2 CTS 4.7.D.1 requires that each Reactor Building to-Suppression Chamber Vacuum Breaker assembly be verified closed at least once per 7 days. ITS SR 3.6.1.6.1 relaxes this testing to once per 14 days. These vacuum breaker valves are normally closed during Modes 1,2 and 3 and the extension of the surveillance to 14 days is not a significant change in operating practice since the anticipated residt of each performance of the SR is to find the vacuum breaker valves in the closed position.

la CTS 3.7.D.4 and 4.7.D.2 provide Actions and Surveillance Requirement for the Reactor Building-to-Suppression Chamber Vacuum Breaker position indication instrumentation. Position indication does not necessarily relate directly to te respntive system Operability. The NUREG does not specify indication-only equipment to be Operablc to support Operability of a system or component. Control of the availability of, and necessary compensatory activities if not available, for position indication instrumentation is addressed by plant procedures and policies.

Vacuum breaker position is required to be known to be able to satisfy SR 3.6.1.6.1, SR 3.6.1.6.2 and SR 3.6.1.6.3. If position indication is not available and vacuum breaker position can not be determined, then the SRs can not be satisfied and the appropriate actions must be taken ihr inoperable vacuum breakers m accordance with the Actions ofITS 3.6.1.6. As a result, the requirements fbr the vacuum breaker position indication are adequately addressed by the requirements of Specification 3.6.1.6 and associated SRs and are proposed to be deleted from Technical Specifications.

19 CTS 4.7.D.2 requires quarterly demoustrations that the Reactor Building to-Suppression Chamber Vacuum Breaker valve will travel through one complete cycle of full travel. ITS SR 3.6.1.6.2 potentially relaxes this testing by changing the frequency from "once per quarter" to "In Accordance with the Inservice Testing Program" which allows up to six months depending on previous testing perfbrmance. Successful testing on a quarterly basis for these valves demonstrates that a potential extension to 6 months is acceptable. Any changes to the frequency will be evaluated in accordance with the DAEC 10 CFR 50.59 program.

1,ic2 Generic Letter 91-04, Chances in Technical Soecification Surveillance Intervals to Accommodate a 24 month Fuel Cvele, describes NRC requirements for preparing such license amendment requests. The Generic Letter indicates that the NRC staff has generically reviewed the extension of surveillance intervals from 18 to 24 months and found that "the effect on safety is small because safety DAEC 3 Revision E

DISCUSSION OF CIIANGES ITS 3.6.1.6: REACTOR BUILDING-TO-SUPPRESSION CilAh1BER VACUUh1 BREAKERS I ECllNICAl, CilANGES -I ESS RESTRICTIVE (continued)

L i c.: systems use redundant electrical and mechanical components and because (cont.) licensees perform other surveillances during plant operation that confirm that these systems and components can perform theh safety functions. Nevertheless, Licensees sheuld evaluate the effect on safety of an increase in 18 month !

surveillance intervals to accommodate a 24-month fuel cycle. This evaluation should support a conclusion that the efTect on safety is small."

The Generic Letter specifies the following specific items for review:

Steam Generators Not applicable to DAEC In_strument Drift Addressed independent of this review by the DAEC Setpoint Control Program

Annendix J Exemption TS Amendment No. 219 addressed DAEC adoption of Option B to Appendix J. No additional r: view is required in this evaluation.

In addition, the Generic Letter indicates Licensee's should review the effect on safety of the extension of other surveillances to ensure that it is supported by historical maintenance and surveillance data.

Data was collected for a ten year period from January 1986 to January 1996 of all deficiencies which occurred for the surveillances for which a frequency extension is being sought. The ten-year period was selected to ensure a broad overview of long term perfonnance and because a similar comprehensive review was performed in 1986 lbr preceding years to support changes from 12-month to 18-month intervals.

As a supplemental check, the database for 10CFR50.65 (hiaintenance Rule) compliance was reviewed to confirm that equipment performance overall was compatible with a decreased surveillance frequency. The DAEC program fbr hiaintenance Rule includes targets for safety system train availability and reliability compatible with assumptions in the DAEC Probabilistic Safety Analysis (data for the hiaintenance Rule is limited to the period since 1991).

DAEC 4 Revision E

l l

l DISCUSSION OF CIIANGES ITS 3.6.1.6: REACTOR BUILDING TO-SUPPRESSION CllAh1BER VACUUht BREAKERS TECilNICAL CilANGES - 1 ESS RESTRICTIVE (continued) 14c.2 Data for the following surveillance tests were reviewet -

, (cont.) ]

Description CTS Section ITS SR CS Simulated Auto Actuation 4.5. A. I .a 3.5.1.7 LPCI System Simulated 4.5.A.3.a 3.5.1.7 Automatic Actuation

^

llPCI Simulated Automatic 4.5.D. I .a 3.5.1.7-Actuation RCIC Simulated Automatic 4.5.E.1.a 3.5.3.5 Actuation ,

SRV Pressure Switch System Table 4.2-11 3.6.1.5.2 Functional Torus / Reactor fluilding Vacuum 4.7.D.1 3.6.1.6.3 lireaker SilGT Bypass Damper 4.7.L. l .e 3.6.4.3.4 in each of these tests, no train failures were identified by perfbrmance of trie reference cyclic test during the ten-year period reviewed. In each case, the system performance was within targets established under the hiaintenance Rule. This combination of no test failures and acceptable system performance is viewed as a strong indicator that interval extension is acceptable without more detai!ed review.

The equipment performance supports interval extensions from quarterly to 12 months, with a maximum proposed interval of 15 months.

Lav This change revises the Technical Specification opening setpoint Ibr the Reactor fluilding-to-Suppression Chamber Vacuum Breakers to reflect Allowable Values consistent with the philosophy of the NUREG These Allowable Values go be included in Technical Specifications) and the Trip Setpoints (to be included in plant proccJures) have been established by DAEC lustrument Setpoint hiethodology which is based on the General Electric (GE) Instrument Setpoint hiethodolegy; NEDC-31336," General Electric Instrumentation Setpoint hiethodology." The i

NRC approval of NEDC-31336 is documented in a Revision to the Safety Evaluation Report transmitte<l by letter from B. Boger (NRC) to R. Pinelli (IlWROG) dated November 6,1995. The setpoint evaluation used the

, DAliC 5 Revision E

DISCUSSION OF CilANGES t ITS 3.6.1.6: REACTOR llUll. DING-TO. SUPPRESSION CilAMBER VACUUM llREALERS TECilNICAl CllANGES -1 ESS RESTRICTIVE (continued)

Lay uncenainties associated with the DAEC instrumentation and actual DAEC physical (cont.) data and operating rectices to ensure the validity of the resulting Allowable Values and Trip Setpoints. I he methodologies used to derive the Allowable Values and Trip Setpoints are based on combining the uncertainties of the associated channels.

In the methodologies, the Trip Setpoints take into consideration calibration accuracies which were specifically assumed in the DAEC setpoint calculations.

Plant calibration procedures will ensure the assumptions regardinualibration accuracy are maintained. The proposed A!!owable Values and Trip Setpoints have been established from each design or safety analysis limit by accounting for instrument accuracy, calibration and drill uncertainties, as well as process measurement accuracy and primary element accu *acy using the DAEC Instrument Setpoint Methodology. The use of these methodologies for esublishing Allowable Values and Trip Setpoints ensures design or safety analysis limits are not exceeded in the event of transients or accidents. While the conversion of the existing instrument setpoint values in the CTS to Allowable Values in the ITS is technically an administrative change for the DAEC, as our current operating practice implements these Allowable Values by plant procedures, it has been characterized as a less restrictive change to the CTS sbr overall conservatism.

4 1

DAEC 6 Revision E

. - . .. . ~ _ - - . - - - _ _ - . .-- _-

DISCUSSION OF CilANGES ITS 3.6.1.7: SUPPRESSION CllAh1BER '"D DRYWELL VACUUhi BREAKERS ADhilNISTRATIVE CilANGES Ai All reformatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. 'Ihe reformatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting)is made consistent with the NUREG.

During NUREO development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the CTS. Additional information has also been added to more fully describe each subsection. This wording is consistent with the NUREO. Since the design is already approved by the NRC, adding more detail does not result in a technical change.

A2 A Note has been added to ITS SR 3.6.1.7.1 to clarify that a vacuum breaker is allowed to be open (and not failing SR 3.6.1.7.1) during Surveillances (e.g., when performing SR 3 6.1.7.2 ?.nd SR 3.6.1.7.3). Since it is obvious that Operability is still maintained, this addition is considered administrative.

TECilNICAL CilANGES - h10RE RESTRICTIVE hi i An additional SR, ITS SR 3.6.1.7.3, is being added. This verification of the Suppression Chamber-to-Drywell Vacuum Breaker opening setting is necessary to ensure that the safety analysis assumption regarding vacuum breaker full open ditTerential pressure oi0.5 psid is valid, h1 2 CTS 3.7.E.3 allows 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to close vacuum breders with one or more drywell to-suppression chamber vacuum breakers open. ITS 3.6.1.7 Action 11 only allows one vacuum breaker to be open. This change is more restrictive since only one vacuum breaker will be allowed to be open in the ITS with a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time for closure.

TECilNICAL CilANGES - RELOCATIONS None DAEC 1 Revision E

DISCUSSION OF CllANGES ITS 3.6.1.7: SUPPRESSION CilAMBER-TO DRYWELL VACUUM BREAKERS TECilNICAL CllANGES - I ESS RESTRICTIVE Li CTS 4.7.E.1 requires that each Suppression Chamber-to Drywell Vacuum Breaker be verified closed at least once per 7 days. ITS SR 3.6.1.7.1 relaxes this testing to once per 14 days. These vacuum breakers are normally closed during Modes 1J and 3 and the extension of this surveillance to 14 days i- 't a significant change in operating practice since the anticipated result of each c .,imance of the SR is to find the vacuum breaker in the closed position.

L2 CTS 4.7.E.3 requires a virual inspection of the dipvell to-suppression chamber vncuum breakers. This visual inspection does not necessarily relate directly to the respective system Operability. NUREO 1433 does not genera" r specify visual inspections and walkdowns of systems or components. Control of these activities are addressed by plant operational procedures ud policies. ITS Surveillance Requirements (SR 3.6.1.7.1, SR 3.6.1.7.2 and SR 3.6.1.7.3) must be satisfied '.,r the vacuum breakers to be Operable. In addition, SR 3.6.1.1.2 will ensure the vacuum breaker leakage is within limits. As a result, the requirements far the vacuum breaker visual inspection are adequately addressed by the requirements of Specification 3.6.1.7 and the associated SRs and are proposed to be deleted from Technical Specifications.

L3 CTS 3.7.E.4 and 4.7.E.2 provide Actions and Surveillance Requirement for the Suppression Chamber-to-Drywell Vacuum Breaker position indication instrumentation. Position indication does not necessarily relate directly to the respective system Operability. The NUREG does not specify indication-only equipment to be Operable to support Operability of a system or component. Control of the availability of, and necessary compensatory activities if not available, for position indication instrumentation ' addressed by plant procedures and policies.

Vacuum breaker position is required to be known to tw able to satisfy SR 3.6.1.7.1, SR 3.6.1.7.2 and SR 3.6.1.7.3. If position indication is not available and vacuum breaker position can not be ictennined, then the SRs can not be satisfied and the appropriate actions must be taken for inoperable vacuum breakers in accordance with the Actions ofITS 3.6.1.7. As a result, the requirements for the vacuum breaker position indication are adequately addressed by the requirements of Specification 3.6.1.7 and associated SRs and are proposed to be deleted from Technical Specifications.

DAEC 2 Revision E

DISCUSSION OF CilANGES ITS 3.6.2.1: SUPPRESSION POOL AVERAGE TEMPERATURE ADMINISTRNflVE C11ANGES Ai All reformatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. The reformatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting) is made consistent with tt' NUREG. r During NUREG development certain wording preferences or English tw tage conventions were adopted which resulted in no technical changes (either actual or interpretational) to the CTS. Additional information has also tren added to more fully describe each s 5 .ection. This wording is consistent with the NUREG. Since the design is already approved by the NRC, adding more detc.il do.., not result in a technical change.

A2 CTS 3.7.G.2.a requires the suppression pool temperature to be s 956 F during normal power operation (i.e., > 1% RTP). CTS 3.7.G.2.b and 3.7,G.2.c requires the plant to be brought to Ilot Shutdown followed by Cold Shutdown. Ilowever, once power is reduced to s 1% RT'), the LCO is not applicable. ITS 3.6.2.1 Action B requires Thermal Power to be reduced to s 1% RTP, consistent with the CTS Applicability. Therefore, this change is considered administrative.

A3 CTS 3.7.G.2.d requires the plant to be scrammed if suppression pool water temperatures e.vceeds 110 F. Thus, there is an implied CTS LCO limit on suppression pool temperature whenever the nuclear system is pressurized above atmospheric. In addition, CTS 3.7.G.2.b and 3.7.G.2.c have actions to verify that suppression pool temperatrie is less than 110 F when those conditions are entered, Thus, the addition of an explicit LCO limit in the ITC (LCO 3.6.2.1.c) is an administrative change. ' % 2.1-5)

WCIINICAI, Cil ANGyS - MORE RESTRICTIVE Mi CTS 3.7.G ge Mut. uppression pool temperature is applicable at any time the nuclear systen. is m shed above atmospheric and during nomial power operation (> 1% ki n'). IT S 3.6.2.1 Suppression Pool Average Temperature, is applicable in Modes 1,2, and 3. As a result, the proposed requirements for

, suppression pool tempemture are applicable when the reactor is critical or control rods a e being withdrawn when the reactor coolant temperature is s 212 F (i.e.,

depressurized)in addition to being applicable whenever the nuclear system is greater than 212 F Therefore, this change is more restrictive.

DAEC i Revision E

DISCUSS!ON OF CilANGES ITS 3.6.2.1: SUPPRESSION POOL AVERAGE TEMPERATURE TECIINICAL CllANGES - MORE RESTRICTIVE (continued)

M2 CTS 3.7.G.2.d and ITS 3.6.2.1 Condition D contain the required actions if suppression pool temperature is greater than 110 F. The proposed change adds an explicit requirement to verify that suppression pool temperature i: less than or equal to 120 F every 30 minutes whenever suppression pool tem serature a greater than 110 F and to place the reactor in Mode 4 within 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months . The exis'ing specification does not contain these explicit requirements foi w.i.oring temperature under these conditions or placing the reactor in a r on applicable Mode.

Therefore, this change is more restrictive. This change is connstent with the NUREG.

M3 CTS 3.7.G.2.e and ITS 3.6.2.1 Condition E rcquire that the reactor pressure vessel be reduced to less than 200 psig if suppression pool temperatt:re reaches or exceeds 120 F. ITS 3.6.2.1 Required Action E.2 requires that the cooldown continue until the reactor is in Mode 4. Therefore, the proposed change is more restrictive. This change is consistent with the NUREG.

TECilNICAl, CllANGES - REI OCATIONS None.

TliCilNICAl, Cll ANGES - 1.ESS RESTRlCTIVE Li CTS 4.7.G.2.c requires an external visual inspection of the suppression chamber whenever there is indication of relief valve operation with the local suppression pool temperature reaching 200 F or greater. This surveillance is being deleted in accordance with NEDO-30832, " Elimination of Limit on BWR Suppression Pool Tempemture for SRV Discharge with Quenchers," dated December 1984. The basis for deleting this surveillance is that testing has demonstrated that there are no undue loads on the suppression pool or its components at elevated temperatures and pressures when SRVs discharge through " quenchers"(spargers). DAEC UFSAR Section 6.2.1.6 states that each relief valve discharge line tenninates in a T-quencher (sparger). Therefore, the requirement for an external visual inspection of the suppression chamber is being deleted. This change is consistent with the NUREG.

DAEC 2 Revision E I

DISCUSSION OF CHANGES ITS 3.6.2.1: SUPPRESSION POOL AVERAGE TEMPERATURE TFCilNIC_AL Cil ANGES - LESS RESTRICTIVE (continued)

L2 The CTS 3.7.G.2.d suppression pool average water temperature limit of 2110' F for scramming the reactor has been slightly increased to > 110 F. This would allow suppression pool average water temperature to be equal te i 10' F and still be within the limit. The UFSAR assumes a 120* F initial suppressi 'n pool average water temperature prior to a LOCA blowdown of the RPV, which ensures average water temper-ture will not exceed 170' F. Due to the 10 F margin provided between 110' F and 120' F, the slight increase in the suppression pool average water te.nperature limit from 2110" F to > 110' F is insignificant.-

L3 The CTS 3.7 G.2.e suppression pool average water temperature limit of 2120 F for depressurizing the reactor has been slightly increased to > 120* F. This would allow suppression pool average water temperature to be equal to 120 F and still be within the limit. The UFSAR assumes the initial suppression pool average water temperature prior to a LOCA blowdown of the RPV is equal to 120 F, which ensures average water temperature u ill not exceed 170' F. Thus, allowing the temperature to equal 120 F before depressurizing the reactor remains within the UFSAR limits. Therfore, the slight increase in the suppression pool average water temperature limit from 2120 F to > 120 F is insignificant. {3.6 2.1-6)

L 1

l l

1 DAEC 3 Revision E

DISCUSSION OF CllANGES ITS 3.6.2.2: SUPPRESSION POOL WATER LEVEL ADMINISTRATIVE CilANGES A, All reformatting and renumbering is in accordance with the NUREG As a result, the ITS should be more readable and more understandable by its users. The reformatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting) is made consistent with the NUREG.

During NUREG development certain wording preferences or English language conventions were adopt:d which resulted in no technical changes (either actual or interpretational) to the CTS. Additional information has also been added to more fully describe cach subsection. Tnis wording is consistent with the NUREG, Since the design is already approved by the NRC, adding more detail does not result in a technical change.

A2 CTS 3.7.G.I.a specifies suppression pool volume in cubic feet and in percent. The suppression pool water limits in ITS 3.6.2.2 have been specified in terms of the corresponding level in feet, which is how level is indicated in the Control Room.

This change is administrative in nature since the level limits in the ITS correspond to the volume limits specified in the CTS. This is consistent with the NUREG.

Therefore, this change is also considered administrative, {3.6.2.2-1 }

4 TECIINICAL CilANGES - MORE RESTRICTIVE Mi CTS 3.7.G goveming suppression pool level is applicable at any time the nuclear system is pressurized above atmospheric. ITS 3.6.2.2, Suppression Pool Water Level, is applicable in Modes 1,2, and 3. As a result, the proposed requirements for suppression pool level are applicable when the reactor is critical or control rods are being withdrawn when the reactor coolant temperature is < 212 F (i.e.,

depressurized) in addition to being applicable whenever the nuclear system is pressurized (greater than 212 F). Therefore, this change is more restrictive.

Tl!CIINICAI. Cil ANGES - REl OCATIONS None DAEC 1 Revision E

- ...- . . - . . - - - - - - -..~ - - - -.- - . - . . . . - . . - - . . . ~ _ . . . . . . - .

l DISCUSSION OF CilANGES -

ITS 3.6.2.2: SUPPRESSION POOL WATER LEVEL TECIINICAL Cil ANGES - LESS RESTRICTIVE

, Li CTS 3.7.G.I.b allows I hour to restore suppression pool level to within limits. ITS i 3.6.2.2 Action A allows 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to restore level. An unanticipated change in the suppression pool level would require addressing the cause and aligning the

appropriate system to raise or lower the pool level. These activities require some time to accomplish. The Completion Time in the ITS is based on engineering judgment of the relative risks associated with: 1) the safety significance; 2) the probability of an event requiring the safety function of the system; and 3) the -

l i'

relative risks associated with the plant transient and the potential challenge of safety-systems experienced by requiring a plant shutdown. Upon further review and discussion with the NRC staff during the development of the NUREG, a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time was determined to be appropriate.

4 a

f 1

1 1.

I i-i i

i j

4

. DAEC 2 Revision E 4

-N N se--,w<,e,-n- ~n rwevv ,..w--r s-w,.-,,c, rr - - - - . - w -- .- - w- es-.-

. . - . . . _ . . . . _ .- __ . . . . ._ _ .. _ _ _ _ _ - . - _ . _ ~ _ _ ._.. _ . . _ . . _ . - . .

l DISCUSSION OF CilANGES : :

. ITS 3.6.2.3: RIIR SUPPRESSION POOL COOLING

_. ADMINISTRATIVE C11ANGES - !

' None -

TEClINIC^1/ CHANGES - MORE RESTRICTIVE l

- Mi - ITS 3.6.2.3, RHP. Suppression Pool Cooling is a new Specification. Appropriate [

- Actions and Surveillance Requirements have also been added. The addition of this '

, new TS is considered to be more restrictive.

[ ; TECliNICAL CHANGES - REl,0 CATIONS ,

1

- None

- TECilNICAL CHANGES - LESS RESTRICTIVE ;

None i

4 i-4

=DAEC- 1 Revision E.;

d 4

- ._ ._ - - . . ~ . _ . _ _ . _ - . . . _ _ - . _ - .-_ _ -.

1 DISCUSSION OF CliANGES ITS 3.6.3.1: CONTAINMENT ATMOSPIIERIC DILUTION SYSTEM ADMINISTRATIVE CilANGES Ai All reformatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. The reformatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting) is made consistent with the NUREG.

During NUREG development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretadonal) to the CTS. Additional information has also been added to more fully describe each subsection. This wording is consistent with the NUREG. Since the design is already approved by the NRC, adding more detail does not result in a technical change.

A2 Per our Response to NRC Question 3.6.3.1-1, the Note added to ITS 3.6.3.1 Required Action A.1 specifying ITS LCO 3.0.4 is not applicable has been withdrawn.

CTS 3.7.11.1 specifies the mode of applicability for the CAD System as the reactor being in reactor power operation (i.e.,21% RTP) and the primary containment atmosphere inerted (s5% Oxygen Concentration by volume) per CTS LCO 3.7.1.1 (ITS LCO 3.6.3.2).13ecause the CAD system cannot inert the containment by itself, it is only designed to maintain it inerted post-LOCA, (Reference UFS AR section 6.2.5.3), the LCO's Applicability is the same as CTS 3.7.1.1, i.e., Mode Switch in RUN Mode 1 not 21% RTP. [ Note: see DOC L.1 to ITS 3.6.3.2 for thejustification of change in Applicability from " Mode switch in RUN" to " Mode I with Thennal Power 215% RTP."] Ilowever, CTS 3.7.11.2 requires the CAD system to contain at least 50,000 SCF of nitrogen to be Operable (ITS SR 3.6.3.1.1) whenever the reacur is in " reactor power operation" only, which is not consistent with the overall system Operability requirement in CTS 3.7.lLI. Thus, in order to be consistent with SR 3.0.1 requirements, this requirement will be changed to agree with the LCO Applicability of CTS 3.7.IL1 (ITS 3.6.3.1). This change is considered to be administrative in nature, as it conforms to the NUREG requirements (SR 3.0.1) and corrects a inconsistency in the CTS. (3.6.3.1-1 }

DAEC 1 Revision E

DISCUSSION OF CilANGES ITS 3.6.3.1: CONTAINMENT ATMOSPilERIC DILUTION SYSTEM ADMINISTRATIVE CilANGES (continued)

A3 CTS 3.7.11.1 and CTS 3.7.11.2 contain Actions if CAD is inoperable after 7 days to be in at least ilot Shutdown within the next 12 hou s and in Cold Shutdown within the following 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , llowever, the CTS Applicability for CAD is only whenever the reactor is in power operation (i.e., > 1% RTP) and the primary containment is required to be inerted (i.e., oxygen concentration < 4%). Thus, once power has been reduced to < 1% RTP (conservatively, once Mode 3 has been entered), the LCO Applicability for CAD has been exited and it is not necessary to continue the shutdown to Mode 4 (Cold Shutdo un). Consequently, one can say that the CTS and ITS Actions are equivalent and this change can be considered to be administrative.

TECilNICAl, CilANGES - MORE RESTRICTIVE Mi Per our Response to NRC Question 3.6.3.1-1, the proposed addition of the LCO 3.0.4 Note has been withdrawn and the Mode of Applicability for LCO 3.6.3.1 has been revised to be consistent with the CTS and DAEC design basis (see DOC A.2 above). {3.6.3.1-1 }

M2 ITS SR 3.6.3.1.2 is a new SR perfomied every 31 days to verify each CAD System power operated and automatic valve in the required flow path (s) that is not locked, l sealed, or othenvise secured in position is in the correct poeition or can be aligned to the correct position. The addition of this new requirement is a more-restrictive change. Ilowever, it does not involve manipulation of plant equipment and does not place an undue burden on plant Operators. Thus, the addition of this new surveillance will not have an adverse effect on plant safety. (3.6.3.1-5)

TECilNICAl. CIIANGES - REI OCATIONS Ri The details in CTS 3.7.11.1, relating to Operability of the CAD System (i.e., it must be capable of supplying nitrogen to the containment for atmosphere dilution if required by post-LOCA conditions),is being relocated to the Bases. This infonnation, while important, belongs in the Bases. Changes to the Bases will be controlled in accordance with the proposed Bases Control Program described in ITS Chapter 5. This change is consistent with the philosophy of the NUREG which relocates these types of details to the Bases.

DAEC 2 Revision E

DISCUSSION OF CilANGES ITS 3.6.3.1: CONTAINMENT ATMOSPilERIC DILUTION SYSTEM TECliNICAl CliANGES - RELOCATIONE (continued)

R2 CTS 4.7.11.1 contains a requirement to functionally test the CAD System annually.

This Surveillance is being relocated to plant procedures. Any changes will be evaluated in accordance with the D AEC 10 CFR 50.59 program. This chang: is consistent with the NUREG.

R3 CTS 3.7.11.2 specifies the details of how to detennine that the CAD System contains a minimum of 50,000 scfofN (determined 2 by pressure and temperature measurements). These details are not necessary to ensure system Operability and are being relocated to plant procedures. Any changes to these procedures will be evaluated in accordance with the DAEC 10 CFR 50.59 program.

TEClINICAl, CIIANGfiS - LESS RESTRICTIVE Li CTS 4.7.11.2 contains a requirement to verify the volume in the N2 bank weekly.

The frequency of this Surveillance in ITS SR 3.6.3.1.1 has been extended to 31 days, similar to other surveillances for tank contents (e.g., diesel fuel oil). The N2 banks for the CAD System are dedicated fbr use in that system. The banks are filled with gaseous nitrogen and contml room indication is available for bank pressure and ihr outside air temperature which are used to determine N2 volume. Thus, the 31 day Frequency is considered appropriate.

DAEC 3 Revision E

DISCUSSION OF CilANGES ITS 3.6.3.2: PRIMARY CONTAINMENT OXYGEN CONCENTRATION AD_MINISTRATIVE CllANGES Ai All reformatting and remunbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. The reformatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting) is made consistent with the NUREG.

During NUREG development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the CTS. Additional information has also been added to more fully describe each subsection. This wording is consistent with the NUREG. Since the design is already approved by the NRC, adding more detail does not result in a technical change.

A2 CTS 4.7.1.1 requires the oxygen concentration to be verified to be within the limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after placing the reactor Mode Switch in Run and at least once every 7 days thereafter. The CTS frequency of"within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after placing the reactor mode switch in RUN"is not needed in ITS 3.6.3.2. This CTS frequency is part of the Applicability and by ITS SR 3.0.4, the SR must be performed prior to entering the Applicability of the LCO. This would require performance of the SR within the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period specified in CTS 4.7.1.1. Therefore, this change is administrative.

TECilNICAI. CilANGES - MORE RESTRICTIVE None TECliNICAl CHANGES - RELOCATIONS None TECHNICAL, Cll ANGES - 1 ESS RESTRICTIVE Li in order to establish Oxygen Concentration, CTS 3.7.1.1.a allows 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after placing the reactor Mode Switch in Run (approximately 5-10% RTP) following startup. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months time allowance on startup has been changed in ITS 3.6.3.2 to allow 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months aller exceeding 15% RTP. CTS 3.7.1.1.b allows de-inerting the drywell 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to taking the reactor Mode Switch out of Run prior to reactor DAEC 1 Revision E

DISCUSSION OF CllANGES ITS 3.6.3.2: PRIMARY CONTAINMENT OXYGEN CONCENTRATION TECilNICAL Cil ANGES - LESS RESTRICTIVE (continued)

Li shutdown. The ITS changes this allowance to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing Thermal (cont.) Power to <l5% RTP prior to reactor shutdown. If the oxygen concentration limits are not met, CTS 3.7.1.2 requires the reactor to be in at least Startup/llot Standby within the next 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months . ITS 3.6.3.2 Action B changes the shutdown requirement to reduce Thermal Power to s15% RTP within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The changes between the CTS r.nd ITS provisions are small and will add some time for inerting the drywell on a startup. On a shutdown, the ITS will allow de-inerting to begin slightly earlier than in the CTS. The ITS will also allow the reactor to stay at a slightly higher power level (s15% RTP) if the Actions of Condition B are l required to be met. These relaxations are minor, are within DAEC analysis assumptions, and provide consistency with the NUREG.

)

3 DAEC 2 Revision E

DISCUSSION OF CllaNGES ITS 3.6.4.1: SECONDARY CONT /.INMENT ADMINISTRATIVE CilANGES Ai All reformatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. The reformatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting) is made consistent with the NUREG.

During NUREG development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the CTS. Additional information has also been added to more fully describe each subsection. This wording is consistent with the NUREG. Since the design is already approved by the NRC, adding more detail does not result in a technical change.

A2 Per CTS 3.7.J.1, Secondary Containment Integrity is required "during all Modes of plant operation except when all of the following conditions are met (emphacic, added);

a. The reactor is suberitical and specification 3.3.A [ Shutdown Margin] is met.

b. The reactor water temperature is below 212 F and the reactor coolant system is vented.

c. No activity is being perfomied which can reduce the shutdown margin below that specified in specification 3.3.A.

d. The fuel cask or irradiated fuel is not being moved in the reactor building

[ secondary containment]."

Conditions a. and b. above constitute Cold Shutdown, which by the " exception" clause of 3.7.J.1 means Secondary Containment Integrity is required in Modes 1,2 and 3. Conditions c. and d. above constitute Core Alterations and moving irradiated fuel in Secondary Containment, as being activities requiring Secondary Containment integrity. Thus, the CTS Applicability is the same as the ITS for these Modes and Conditions except for the ITS addition for OPDRVs (see M-DOC to 3.6.4.1). Therefore, this is an administrative change.

DAEC 1 Revision E

DISCUSSION OI CIIANGES ITS 3.6.4.1: SECONDARY CONTAINMENT

~ ADMINISTRATIVE CTIANGES (continued)

A3 The definition of Secondary Contai'anent Integrity has been deleted from the CTS. In its place the requirement for secondary containment is that it "shall be Operable." This was dane because of the confusion associated with these definitions compared to its use in the respective LCO. The change is editorial in that all the requirements are specifically addressed in the ITS LCOs for the Secondary Containment, the Secondary Containment isolation Valves / Dampers and Standby Gas Treatment System Specifications. This change is consistent with the NUREG.

A4 ITS SR 3.6.4.1.3 requires the vacuum to be maintained greater than or equal to one quarter inch of water vacuum instead of requiring it to be maintained at one quarter inch, as in CTS 4.7.J.l.a. Maintaining at least one quarter inch of water vacuum during the performance of CTS 4.7.J.l.a is current operating practice at DAEC. The ITS provides clarification in that the vacuum does not need to be maintained at exactly one quarter inch vacuum. This change is a presentation preference and is therefore considered to be administrative.

As ITS 3.t,.4.1 Required Action C.1 has been modified by a Note stating,"LCO 3.0.3 is not applicable." If moving irradiated fuel assemblies while in Mode 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in Mode 1,2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement ofirradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. In addition, by adding an exception to LCO 3.0.3 for the suspension ofirradiated fuel movement in Mode 1, 2, or 3, the plant would still be required to shutdown aller 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months if secondary containment was not restored to Operable status per ITS Required Actions B.1 and B.2 and to suspend irradiated fuel movement per Required Action C.I. This is the intent of CTS 3.7.J.2. Therefore, this change is considered to be administrative in nature.

TECIINICAL Cll ANGES - MORE RESTRICTIVE Mi ITS 3.6.4.1 adds a new Applicability Requirement, a new portion of Condition C, and an appropriate Required Action for Condition C (Required Action C.3), for Operations with a Potential for Draining the Reactor Vessel (OPDRVs). Secondary containment is now required to be Operable during OPDRVs to provide mitigation if an inadvertent vessel draindown event occurs. The new Applicability and the DAEC 2 Revision E

l DISCUSSION OF C11ANGES ITS 3.6.4.1: SECONDARY CONTAINMENT TECIINICAL CilANGES - MORE RESTRICTIVE (continued)

Mi addition of the Condition and Required Action is an additional restriction to plant (cont.) operation and constitutes a more restrictive change.

M2 ITS 3.6.4.1 Action C adds a requirement that Core Alterations be "Immediately" suspended if secondary containment is inoperable. Although the CTS contains the requirement (CTS 3.J.2.a) to suspend movement ofirradiated fuel, it does not contain the requirement to suspend Core Alterations, which includes more than moving irradiated fuel in the reactor building, immediately suspending Core

. Alterations minimizes the probability of a fission product release if a reactivity event occurs while the secondary containment is inoperable imposing a new requirement to suspended Core Alterations is a more restrictive change.

M3 ITS SR 3.6.4.1.3 requires the secondary containment capability test ('i.e., to maintain 21/4" vacuum) using each tram of the standby gas treatment subsystem on a Staggered Test Basis. CTS 4.7.J.l.a requires secondary containment capability to be similnly demonstrated. Howeser, no requirement exists to preclude the same train from being used each time to conduct this test. Since this change adds a more prescriptive requirement, (i.e., on the Staggered Test Basis), it is classified as a more restrictive change.

M4 Two Surveillance Requirements were added (ITS SR 3.6.4.1.1 and 3.6.4.1.2):

ITS SR 3.6.4.1.1 verifies all secondary containment equipment hatches are closed every 31 days.

ITS SR 3.6.4.1.2 verities that either the outer door (s) or the inner door (s) in each sceandary containment access opening are closed.

CTS detinition 16 for Secondary Containment Integrity requires at least one door in each access opening be closed. ITS SR 3.6.4.1.2 requires that either the outer door (s) or the inner door (s) in each secondary containment access opening be verified closed. This requirement ensures that the infiltration of outside air of such a magnitude as to prevent maintaining the desired negative pressure does not occur.

Definition 16 has been slightly reworded from requiring "at least one door" to be closed to "either the outer door (s) or the inner door (s)" being closed. This is necessary since some secondary containment access openings at DAEC contain more than 2 daers (e.g.,2 outer doors and I inner door).

DAEC 3 Revision E

DISCUSSION OF CIIANGES ITS 3.6.4.1: SECONDARY CONTAINMENT TECilNICAL CilANGES - MORE RESTRICTIVE (continued)

M4 These tests help to ensure the integrity of the secondary containment boundary so (cont.) that it will perform as assumed in the safety analysis. Even though the CTS does not contain these specific SRs, this is current operating practice at DAEC. The addition of these SRs and the modification of Definition 16 constitutes a more ,

restrictive change.

TECllNICAL GIANGES - REl OCATIONS Ri CTS 3.7.J.l.d requires secondary containment be maintained if the fuel cask is being moved in the reactor building. CTS 3.7.J.2.a requires fuel cask movement be suspended in the reactor building if secondary containment cannot be met. The reference to fuel casks will be relocated to plant procedures goveming heavy loads.

The procedures goveming heavy loads provide assurance that an appropriate level of safety is provided and mitigation capability exists. Any changes to this requirement will be evaluated in accordance with the DAEC 10 CFR 50.59 program. This change is consistent with the NUREG.

R2 CTS 4.7.J.l.a provides a requirement that " Secondary containment capability to maintain 1/4 inch of water vacuum under calm wind conditions (<l5 mph) with a filter train . . ." The requirement of the wind conditions and filter train are being relocated to the Bases for ITS 3.6.4.3, " Standby Gas Treatment System." Any changes to this requirement will be made in accordance with the TS Bases Control program.

TEClINIC AI. Cil ANGES - I.ESS RESTRICTIVE ly CTS 3.7.J.2.b requires the restoration of secondary containment within one hour.

ITS 3.6.4.1, Action A extends the Completion Time to four hours to restore secondary containment to Operable status. The CTS requires the plant to begin shutting down when secondary containment is inoperable for more than I hour.

This change will allow a longer period of time to restore the secondary containment to Operable status in order to preclude an unnecessary plant shutdown. The ibur hours is commensurate with the importance of maintaining secondary containment during Modes 1,2, and 3. This time period also ensures that the probability of an accident (requiring secondary containment Operability) occurring during periods where secondary containment is inoperable is minimal. Allowing this extended time to potentially avoid a plant shutdown is reasonable and does not represent a significant decrease in safety.

DAEC 4 Revision E J

l I

l DISCUSSION OF CllANGES ITS 3.6.4.1: SECONDARY CONTAINMENT 1]? CLINICAL CllANGES - LESS RESTRICTIVE (continued)

L2 CTS 4.7.J.l.a requires that secondary containment capability shall be demonstrated

" prior to refueling,"(i.e., prior to Core Alterations). The ITS requirement to perform the secondary containment capability test with the standby gas treatment system subsystem (ITS SR 3.6.4.1.3) does not contain this requirement. Per SR 3.0.1, systems and components are assumed to be Operable when the associated SRs have been met (and not otherwise known to be inoperable), regardless of the clapsed time since the last successful performance of the SR. Therefore, it is not necessary to perform this SR " prior to refueling" as a precondition to Core Alterations, as long as the SR has been met within the specified Frequency (within the SR 3.0.2 allowance). In addition, the control of the plant conditions appropriate to perform required surveillances is best left to procedures and scheduling, and has been detemiined by the NRC staff to be unnecessary as a TS restriction (Reference GL 91. , Thus, this change, while less restrictive than the CTS,is consistent with the NUREG.

Lcy.2 Generic Letter 91-04, Chances in Technical Snecification Surveillance Intervals to Accommodate a 24-month Fuel Cycle, describes NRC requirements for preparing such license amendment requests. The Generic Letter indicates that the NRC staff has generically reviewed the extension of surveillance intervals from 18 to 24-months and found that "the effect on safety is small because safety systems use redundant electrical and mechanical components and because licensees perform other surveillances during plant operation that confirm that these systems and components can perform their safety functions. Ncvertheless, Licensees should evaluate the effect on safety of an increase in 18-month surveillance intervals to accommodate a 24-month fuel cycle. This evaluation should support a conclusion that the effect on safety is small."

The Generic Letter specifies the following specific items for review:

. Steam Generators Not applicable to DAEC Instrument Drill Addressed independent of this review by the DAEC Setpoint Control Program Annendix J Exemption TS Amendment No. 219 addressed DAEC adoption of Option B to Appendix J. No additional r: view is required in this evaluation.

DAEC 5 Revision E

DISCUSSION OF CilANGES ITS 3.6.4.1: SECONDARY CONTAINh1ENT TECllNICAL CilANGES -i ESS RESTRlCTIVE (continued)

Lm in addition, the Generic Letter indicates Licensee's should review the effect on (cont.) safety of the extension of other surveillances to ensure that it is supported by historical maintenance and surveillance data.

Data was collected for a ten-year period from January 1986 to January 1996 of all deficiencies which occurred for the surveillances for which a frequency extension is being sought. The ten year period was selected to ensure a broad overview of long term performance and because a similar comprehensive review was performed in 1986 for preceding years to support changes from 12-month to 18-month intervals.

As a supplemental check, the database for 10CFR50.65 (hiaintenance Rule) compliance was reviewed to confirm that equipment performance overall was compa:ible with a decreased surveillance frequency. The DAEC program for hiaintenance Rule includes targets for safety system train availability and reliability compatible with assumptions in the DAEC Probabilistic Safety Analysis (data for the hiaintenance Rule is limited to the period since 1991).

Data for the following surveillance tests were reviewed:

Description -

- Section iIS SR SHLC Squib Valve Firing 4 it.b 3.1.7.7 SilLC Flow Verification 42 . 3.1.7.8 SDV Vent and Drain Cycling 4.3.d.~.i ~ 3.1.8.3 Reactor Mode Switch Channel Functional 4.1. A. I 3.3.1.13 RPS Response Time 4.1. A.2 3.3.1.18/3.3.1.19 MSL Radiation Monitor Logic System functional 4.2.D.2.c 3.3.6.t.9 ATWS RPT Logic System Functional 4.2.G.2 3.3.4.2.4 RPT lireaker Response Time 4.2.G.3 3.3.4.1.3/3.3.4.1.5 SV Setpoint Verification 4.6.D. I 3.4.3.1 SRV Setpoint Verification 4.6.D.1 3.4.3.1 SRV Manual Opening 4.6.D.3 3.4.3.2 IIPCI Low Pressure Flow 4.5.D. I .c 3.5.l.6 CS Logic System Functional 4.2.H.2.a 3.3.5.1.9 RilR Logic System Functional 4.2.ll.2.b 3.3.5.l.9 Containment Spray interlock Logic System 4.2.B.2.c 3.3.6.l.9 Functional llPCI Logic System Functional 4.2.B.2.d 3.3.5.1.9 liPCl/RCIC Suction 'Iransl~er 4.5.D. I .f 3.5.1.7/3.5.3.5 (relocated)

DAEC 6 Revision E i

I

DISCUSSION OF CilANGES -

I ITS 3.6.4.1: SECONDARY CONTAINMENT

= TECilNICAL C11ANGES - LESS RESTRICTIVE (continued)

I4Y2 (cont.) ;

Description CTS Section ITS SR I ADS Logic System Functional 4..t.B.2.e 3.3.5.1.9 l ADS Simulated Automatic Actuation 4.5.F. I .a 3.5.1.8 ADS Valve Manual Opening 4.6.D.3 3.5.1.9 RCIC Low itessure 4.5.E.1.e 3.5.3.4 Drywell to Torus Leak Test 4.7.E.4 3.6.1.1.2 PCIV Simulated Automatic Actuation 4,7.B.I.a 3.6.1.3.6 (Groups I 6,8. 9)

PCIV Logic System Functional Test 4.2.A.2.a - g 3.3.6.1.9 (Groups 16)

EFCV isolation 4.7.11.1.c 3.6.1.3.7 LLS Valve Manual Opening 4.6.D.3 3.6.1.5.1 LLS Logic System Functional 4.2.B.2.g 3.3.6.3.6/3.6.l.5.2 Secondary Containment Integrity 4.7.J. l .a 3.64.1.3 SCIV/D Simulated Automat;c Actt ations 4.7.K. I 3.6.4.2.2 SilGT Simulated Automatic Actuation 4.7.L. I .d 3.6.4.3.3 River Water Supply Simulated Automatic 4.5.Jl.a 3.7.2.4 Actuation ESW Automatic Start w/ DG 4.8.E.1.a 3.7.3.2 SFU Simulated Automatic Actuation 4.10.A.3 3.7.4.3 Control lluilding Positive Pressure 4.10.A.3 3.7.4.4 LOOP /LOCA Test 4.8.A.2.b 3.8.1.13 Battery Service Discharge 4.8.B.I.c 3.8.4.7 In each of these tests, no train failures were identified by performance of the reference cyclic test during the ten-year period reviewed. In each case, the system performance was within targets established under the Maintenance Rule. This combination of no test failures and acceptable system perfonnance is viewed as a strong indicator that interval extension is acceptab!c without more detailed review.

For six Surveillance Tests, more than one failure was identified during performance of the test during the ten year interval. These tests were singled out as requiring further review prior to extending the interval.

_ Diesel Generator and Emergency Service Water Automatic Actuation (ITS SR 3.7.3.2)

llPCI System Cycle Operability Test (ITS SR 3.5.1.6)

DAEC- 7 Revision E

DISCUSSION OF C11ANGES ITS 3.6 4.1: SECONDARY CONTAINMENT TliCIINICAL CliANGES - LESS RESTRICTIVE (continued)

Ley.2 -

11PCI Logic System Functional Test (ITS SR 3.3.5.1.8)

(cont.)

- Safety and Relief Valve Setpoint Verification and Inspection Tests (3 tests)(ITS SR 3.4.3.1)

The majority of problems associated with failures of the Diesel Generator and Emergency Service Water automatic actuation are related to personnel or procedural errors. The single exception was a failure in a diesel generator output breaker. The failures associated with the llPCI logic system functional test include the failure of the turbine control valve to open due to the failure of a newly installed relay, the failure of a pump suction motor-operated valve to cycle (the valve is routinely cycled by the IST Program and would have been detected at another time), and the failure of the turbine stop valve to close due to a sticking limit switch. The failures associated with the IIPCI System cycle operability test were mainly associated with the inability to reach rated flow within the speciSed time of 30 seconds. In each case, the system responded within the analyzed 45 seconds. These and the other failures associated with this test would have been identined during the performance of similar quarterly testing. The failures associated with the SRV setpoint verincation and inspection tests include numerous it.:tances of as-found valves lifting more than 1% below the specified setpoint and a single failure of an SRV being above the 1% setpoint tolerance (see ITS change in setpoir. tole:ance from -l% to -3%).

For each of these tests, the nature of the failures, corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension.

The equipment performance supports interval extensions from 18 to 24 months, with a maximum proposed interval of 30 months in each case.

DAEC 8 Revision E

l l

DISCUSSION OF CilANGES JTS 3.6.4.2: SECONDARY CONTAINMENT ISOLATION VALVES / DAMPERS ADMINISTRATIVE CIIANGES Ai All refonnatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. The reformatting, renumbering, and rewording process involves no technical changes to the CTS.

Editorial rewording (either adding or deleting) is made consistent with the NUREG.

During NUREG development certain wording preferences or English language conventions were adopted which resulted in no technical changes (either actual or interpretational) to the CTS. Additional information has also been added to more fully describe each subsection. This wording is consistent with the NUREG. Since the design is already approved by the NRC, adding more detail does not result in a technical change.

A2 Two new Notes were added to ITS 3.6.4.2 Actions (Notes 2 nd 3).

Note 2 provides explicit instructions for proper application of the Actions for TS compliance, in conjunction with ITS Specification 1.3 " Completion Times,"

this Note provides direction consistent with the intent of the existing Actions for inoperable isolation valves / dampers.

Note 3 facilitates the use and understanding ofthe intent to consider any system

, afTected by inoperable isolation valves / dampers which is to have its Actions also apply ifit is determined to be inoperable. This clarification is consistent with the intent and interpretation of the CTS, and is therefore considered an administrative presentation preference.

A3 ITS 3.6.4.2 Required Action D.1 has been modified by a Note stating,"LCO 3.0.3 is not applicable." If moving irradiated fuel assemblies while in Mode 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in Mode 1,2, or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement ofirradiated fuel assemblies would not be a sufficient reason to require a reactor shutdown. In addition, by adding an exception to LCO 3.0.3 for the suspension ofirradiated fuel movement in Mode 1, 2, or 3, the plant would still be required to shutdown after 4 or 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , as applicable, if secondary containment was not restored to Operable status per ITS Required Action C.1 and C.2 and suspend irradiated fuel movement per Required Action D.I. Currently, this is the intent of CTS 3.7.k.3. Therefore, this change is considered administrative.

DAEC l Revision E 1

DISCUSSION OF CllANGES ITS 3.6.4.2: SECONDARY CONTAINMENT ISOLATION VALVES / DAMPERS TECIINICAL CilANGES - MORE RE",TRICTIVE Mi ITS 3.6,4.2 Required Action A.2 adds a requirement to verify the penetrations which were isolated by ITS 3.6A.2 Required Action A.1, are isolated every 31 days.

The 31 days is reasonable because the valves are operated under administrative controls and the probability of their misalignment is low. This Action is modified by a no'e that applies to valves and blind flanges located in high mdiation areas, and allows them to be verified by use of administrative means. Allowing verification by administrative means is considered acceptable, since access to these areas is typically restricted. Therefore, the probability of misalignment of these valves, aner the mitial verification has been performed under Administrative Controls, is low.

The addition of new requirements constitutes a more restrictive change.

M2 ITS 3.6.4.2 aods a new Applicability, a new portion of Condition D, and an appropriate Required Action P.3, for Operations with a Potential for Draining the Reactor Vessel OPDRVs. Secondary containment isolation valves / dampers are now required to be Operable during OPDRVs to provide mitigation if an inadvertent vessel draindown event occurs. The new Applicability and the addition of the Conditions and Required Action is an additional restriction to phmt operation and constitutes a more restrictive change.

M3 ITS SR 3.6.4.2.1 was added to verify the isolation time of each power operate automatic SCIV/D is within limits. This test helps to ensure the SCIV/D function as assumed in the safety analysis. The addition of this SR constitutes a more restrictive change. (3.6.4.2-9) l M4 ITS 3.6.4.2 Action D adds a requirement during Core Alterations that Core Alterations be "Immediately" suspended if secondary containment isolation vaives/ dampers are inoperable. Although the CTS contains the requirement (CTS 3.7.K.3) to suspend movement ofirradiated fuel, it does not contain the requirement to suspend Core Alterations (which includes not only fuel movement, but control rod movement also). Immediately suspending Core Alterations minimizes the probability of a fission product release if a reactivity event occurs while the secondary containment isolation valves / dampers are inoperable. Imposing a time limit to suspend these activities is a more restrictive change.

DAEC 2 Revision E l

i

DISCUSSION OF CilANGES FFS 3.6.4.2: SECONDARY CONTAINMENT ISOLATION VALVES / DAMPERS TECIIN! CAL CilANGES - RELOCATIONS Ri CTS 3.7.k.1 requires secondary containment isolation valves / dampers be maintained Operable if the fuel cask is being moved in the reactor building. CTS 3.7.k.3 requires fuel cask movement to be suspended if a secondary containment isolation valve / damper is inoperable and the associated penetration is open. These requirements will be relocated to plant procedures governing heavy loads. The procedures governing heavy loads provide assurance that an appropriate level of safety is provided. Any changes to these procedures will be evaluated in accordance with the DAEC 10 CFR 50.59 program. This change is consistent with the NUREG.

R2 The details in CTS Defmition 16 that constitute Secondary Containment Integrity with respect to SCIV/Ds have been relocated to the Bases. These details are not necessary to ensure the SCIV/Ds are ma'ntained Operable. The requirements ofITS 3.6.4.2 Required Actions and Surveillance Requirements are adequate to ensure the SCIV/Ds will be in the proper position during accident conditions. Changes to the Bases will be controlled in accordanec with the proposed Bases Control Program described in Chapter 5 ofITS.

TFCIINICAL CIIANGES - 1 ESS RESTRICTIVE ly ITS 3.6.4.2 adds Action B when one or more penetration flow paths with two

, SCIV/Ds are inoperable. With two SCIV/Ds inoperable,4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months are allowed to isolate the penetration flow path. CTS 3.7.K.3 requires the phmt to begin shutting down with two SCIV/Ds in a line inoperable. This change will allow a period of time to restore the secondary containment to Operable status in order to preclude an unnecessary plant shutdown or suspension of movement ofirradiated fuel assemblies, Core Alterations, and OPDRVs. The 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Completion Time for two SCIV/Ds inoperable in one or more penetration flow paths is commensurate with the importance of maintaining secondary containment during applicable Modes or Conditions. This time period also ensures that the probability of an accident (requiring secondary containment Operabil%y) occurring during periods where secondary containment is inoperable is minimal. Allowing this extended time to potentially avoid a plant shutdown or immediate suspension of movement of irradiated fuel assemblies, Core Alterations and OPDRVs, is reasonable and does not represent a significant decrease in safety.

i DAEC 3 Revision E

DISCUSSION OF CIIANGES ITS 3.6.4.2: SECONDARY CONTAINMENT ISOLATION VALVES / DAMPERS TECilNICAL CH ANGES - 1 ESS RESTRICTIVE (continued)

L2 CTS 4.7.k.1 requirements for automatic initiation testing of the Secondary Containment Isolation Valves / Dampers (SCIV/Ds) stipulates a " simulated" test be performed. The phrase " actual or," in reference to the automatic isolation signal, has been added to ITS SR 3.6.4.2.2. This allows satisfactory automatic SCIV/D isolations for other than Surveillance purposes to be used to fulfill the SRs.

Operability is adequately demonstrated in either case since the SCIV/D itself cannot discriminate between actual" or " test" signals.

l .cy.2 Generic Letter 91-04, Chances in Technical Snecification Surveillance Intervals to Accommodate a 24-month Fuel Cycle, describes NRC requirements for preparing such license amendment requests. The Generic Letter indicates that the NRC staff has geneacally reviewed the extension of surveillance intervals from 18 to 24-months and found that "the effect on safety is small because safety systems use redundant electrical and mechanical components and because licensees perform other surveillances during plant operation that continn that these systems and components can perform their safety functions. Nevertheless, Licensees should evaluate the efTect on salety of an increase in 18-month surveillance intervals to accommodate a 24-month fuel cycle. This evaluation should support a conclusion that the effect on safety is small."

The Generic Letter specifies the following specific items for review:

Steam Generators Not applicable to DAEC 4

Instrument Drift Addressed independent of this review by

. the DAEC Setpoint Control Program Annendix J Exemption TS Amendment No. 219 addressed DAEC adoption of Option B to Appendix J. Ne additional review is required in this evaluation.

In addition, the Generic Letter indicates Licensee's should review the effect on safety of the extension of other surveillances to ensure that it is supported by historical maintenance and surveillance data.

Data was collected for a ten-year period from January 1986 to January 1996 of all deficiencies which occurred for the surveillances for which a frequency extension DAEC 4 Revision E

p. 4. J 4 -w- Ja 6 #e., , e gg JJ, 4 _,a4g, L z,_ac...a_..JW 4.Aa@+,4.w h 4,4myJ ,A&c .a, pi.-nADs3%a -C _a-. ; - 1 w-Jla aE..,.& _

DISCUSSION OF CHANGES ITS 3.6.4.2: SECONDARY CONTAINMENTISOLATION VALVES / DAMPERS TECilNICAL CIIANGES - LESS RESTRICTIVE (continund)

Ley.2 is being sught. The ten-year period was selected to ensure a broad overview of (cont.) long term perfonnance and because a similar comprehensive review was performed in 1986 for preceding years to support changes from 12-month to 18-month intervals.

As a supplemental check, the database for 10CFR50.65 (Maintenance Rule) compliance was reviewed to confirm that equipment performance overall was compatible with a decreased surveillance frequency. The DAEC program for Maintenance Rule includes targets for safety system train availability and reliability compatible with assumptions in the DAEC Probabilistic Safety Analysis (data for the Maintenance Rule is limited to the period since 1991).

Data for the following surveillance tests were reviewed:

Description CTS Section ITS SR SilLC Squib Valve Firing 4.4.A.2.b 3.1.7.7 Sul.C Flow Verification 4.4.A.2.c - 3.1.7.8 SDV Vent and Drain Cycling 43.B.3 3.1.8.3

"' actor Mode Switch Channel 4.1.A. I 3.3.I.13 isnctional RPS Response Time 4.1.A.2 33.1.18/3.3.1.19 MSL Radiation Monitor Logic System 4.2.D.2.c 33.6.l.9 Functional ATWS RPT Logic System Functional 4.2.G.2 3.3.4.2.4 RPT Breaker Response Time 4.2.G 3 33.4.13/33.4.1.5 SV Setpoint Verification 4.6.D.1 3.43.1 SRV Setpoint Verification 4.6.D. I 3.43.1 SRV Manual Opening 4.6.DJ 3.43.2 IIPCI Low Pressure Flow 4.5.D. I .e 3.5.1.6 CS Logic System Functional 4.2.B.2.a 33.5.1.9 RiiR Logic System Functional 4.2.B.2.b 33.5.1.9 Containment Spray Interlock Logic 4.2.B.2.c 33.6.1.9 System Functional llPCI Logic System Functional 4 2.B.2.d 3.3.5.1.9 IIPCl!RCIC Suction Transfer 4.5.D. I .f 3,5.1.7/3.53.5 (relocated)

ADS Logic System Functional 4.2.B.2.c 33.5.1.9 ADS Simulated Automatic Actuation 4.5 F.I .a 3.5.1.8 i

ADS Valve Manual Opening 4.6.D3 3.5.1.9 RCIC Low Pressure 4.5.E.1.e 3.53.4 Drywell to Torus Leak Test 4.7.E.4 3.6.l.1.2 DAEC 5 Revision E i

l-

1 i

DISCUSSION OF CIIANGES ITS 3.6A.2: SECONDARY CONTAINMENT ISOLATION VALVES / DAMPERS TECIINICAL CilANGES -!.ESS RESTRICTIVE (continued)

, b2 CY 1

i (cont.)

Description CTS Section ITS SR j PCIV Simulated Automa!4 Actuation 4.7.B. l .a 3.6,l.3.6 (Groups 1 - 6,8,9)

ICIV Logic System Functional Test 4.2.A.2.a - g 3.3.6.1.9 (Groups 16)

EFCV isolation 4.7.B. I .c 3.6.1.3.7 LLS Valve Manual Opening 4.6.D.3 3.6.1.5.1 LLS Logic System Functional 4.2.B.2.g 3.3.6.3.6/3.6.1.5.2 Secondary Containment integrity 4.7.J. l .a 3.6.4.1.3 SCIV/D Simulated Automatic 4 /.K.I 3.6.4.2.2 Actuations 5 BOT Simulated Automatic Actuctior. 4.7.L.I.d 3.6.4.3.3

. River Water Supply Simulated 4.5.J. l .a 3.7.2.4 Automatic Actuation ESW Automatic Start w/ DG 4.8.E.1.a 3.7.3.2 SFU Simulated Automatic Actuation 4.10.A.3 3.7 4.3 Control Buildmg Positive Pressure 4.10.A.3 3.74.4 LOOP /LOCA Test 4.8.A.2.b 3.8.1.13 Battery Service Discharge 4.8.B.I.c 3.8.4.7 in each of these tests, no train failures were identified by performance of the reference cyclic test during the ten-year period reviewed. In each case, the system performance was within targets established under the Maintenance Rule. This combination of no test failures and acceptable syster performance is viewed as a strong indicator that interval extension is acceptable without more detailed review.

For six Surveillance Tests, more than one failure was identified during performance of the test during the ten year interval. These tests were singled out as requiring further review prior to extending the interval.

Diesel Generator and Emergency Service Water Automatic Actuation (ITS SR 3.7.3.2)

' llPCI System Cycle Operability Test (ITS SR 3.5.1.6) 1IPCI Logic System Functional Test (ITS SR 3.3.5.1.8)

DAEC 6 Itevision E

DISCUSSION OF CilANGES ITS 3.6.4.2: SECONDARY CONTAINMENT ISOLATION VALVES / DAMPERS 4

~ TECIINICAL ClIANGES - LESS RESTRICTIVE (continued)

LCY.2 '

Safety and Relief Valve Setpoint Verification and Inspection - ^,

- (cont.) Tests (3 tests)(ITS SR 3.4.3.1) -

The majority of problems associated with failures of the Diesel Generator and -

Emergency Service Water automatic actuation are related to personnel or procedural errors. The single exception was a failure in a diesel generator output breaker. The failures associated with the 11PCI logic system functional test include the failure of the turbine control valve to open due to the failure of a newly installed relay, the failure of a pump suction motor-operated valve to cycle (the valve is routinely cycled by the IST Program and w6uld have been detected at another time), and the failure of the turbine stop valve to close due to a sticking limit switch. The failures associated with the llPCI System cycle operabihty test were mainly associated with the inability to reach rated flow within Tae specilied time of 30 seconds. In each case, the system responded within the analyzed 45 seconds. These and the other failures associated with this test would have been identified during the performance of similar quarterly testing. The failures associated with the SRV setpoint verification and inspection tests include numerous instances of as-found valves lifting more than 1% below the specified setpoint and a single failure of an SRV being above the 1% setpoint tolerance (see ITS change in setpoint tolerance from -l% to -3%).

For each of these tests, the nature of the failures, corrective actions that were

taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension. ,

4 The equipment performance supports interval extensions from 18 to 24 months, with a maximum proposed interval of 30 months in each case.

n J

DAEC 7 Revision E q, -----

--ne .,-e , --p , - - . - .,-,-,-,7 m , - - . . c , , , ,

(;

DISCUSSION OF CHANGES

, ITS 3.6.4.3: SBGT SYSTEM ADMINISTRATIVE CIIANGES -

Ar All reformatting and renumbering is in accordance with the NUREG. As a result, the ITS should be more readable and more understandable by its users. The reformatting, renumbering, and rewording process involves no technical changes to

. the CTS. .

Editorial rewording (either adding or deleting) is made consistent with the NUREG.

During NUREG development certain wording preferences or English language

, conventions were adopted which resulted in no technical changes (either actual or interpretationd) to the CTS. Additional information has also been added to more fully describe cach subsection. This wording is consistent with the NUREG. Since 3

the design is already approved by the NRC, adding more detail does not result in a technical change.

- A2 ITS SR 3.6.4.3.2 is being added to clarify that the tests specified in the Ventilation Filter Testing Program (VFTP) must be completed and acceptable for the Standby Gas Treatment System to be Operable. ITS SR 3.6.4.3.2 requires performing required SBGT filter testing in accordance with the VFTP which is described in ITS 5.5.7, Ventilation Filter Testing Program (VFTP). This change in the location of the technical requirements for SBGT filter testing to ITS 5.5.7 is in accordance with the fbrmat of the NUREG. Any technical changes to the requirements for SBGT

filter testing will be addressed with the content ofITS 5.5.7. Therefore, moving all details for performing required SBGT filter testing to ITS 5.5.7 is an administrative change.

4 A3 CTS 3.7.L.1 contains exceptions (Reference to CTS 3.7.L.3 and CTS 3.9.D) to when both trains of the SBGT System have to be operable. This list of exceptions is not necessary in the ITS since individual TSs contain stand alone Operability and Action provisions for inoperable systems. Since the CTS provisions are built into the ITS, this change is considered administrative.

Ai This proposed change deletes the explicit requirement to verify the Operability of -

the remaining tran of SBGT when one train of SBGT is inoperable. This verification is an implicit part of using Technical Specifications and determining the appropriate Conditions to enter and Actions to take in the event ofinoperability of-Technict.1 Specification equipment.

4 DAEC i Revision E

, - - .-.--n.--,_,. - --

DISCUSSION OF CIIANGES ITS 3.6.4.3: SBGT SYSTEh1 ADhilNISTRATIVE CHANGES (continued)

A4 Therefore, the explicit requirement in the CTS to ve-ify the Operability of the (cont.) remaining SGTS when one train of SGTS is inoperable is considered to be tmnecessary for ensurcing compliance wkh the applicable Technical Specification Actions in the ITS. This is consistent with the NUREG and is considered to be administrative in nature. {3.6.4.3-3}

A3 ITS LCO 3.6.4.3, Required Action C, and LCO 3.6.4.3, Required Action E.1, have been modified by a Note stating that LCO 3.0.3 is not applicable. If moving irradiated fuel assemblies while in hiode 4 or 5, LCO 3.0.3 would not specify any action. If moving irradiated fuel assemblies while in Mode 1,2 or 3, the fuel movement is independent of reactor operations. Therefore, in either case, inability to suspend movement ofirradiated fuel assemblies would not be sufficient reason to require a reactor shutdown. In addition, by adding an exception to LCO 3.0.3 for the suspension ofirradiated fuel movement in Modes 1,2 or 3, the plant would be required to be shutdown after the allowed restoration time has expired, per ITS Required Actions B,1, B.2 and D.1, and suspend irradiated fuel movement per Reauired Action C.2.1 and E.1. Currently, this is the intent of CTS 3.7.L.3.

Therefore. this change is considered administrative.

A3 CTS 3.7.L.3 states that with one train of SBGT inoperable, fuel handling can continue provided the remaining SBGT train is Operable. Therefore, when both SBGT trains are inoperable, the "provided" statement is not met and fuel handling must be suspended in accordance with CTS 3.7.L.3. ITS LCO 3.6.4.3, Action E,is being added to specifically state to suspend Core Alterations and movement of i Tadiated fuel assemblies in the secondary containment, if both SBGT subsystems are inoperable during these conditions. This addition is a presentation preference and is considered administrative.

TECllNICAL Cil ANGES - MORE RESTRICTIVE Mi CTS 3.7.L.1, Standby Gas Treatment System, is applicable at all times when Secondary Containment Integrity is required. ITS LCO 3.6.4.3 has , m applicability statement that is identical to the Applicability statement for ITS LCO : 6.4.1, Secondary Containment and LCO 3.6.4.2, Secondary Containment isolation Valves / Dampers. ITS 3.6.4.3 Applicability, a new portion of Conditions C and E and Required Actions C.1 (as it relates to OPDRVs only), C.2.3 and E.3 add new provisions for OPDRVs. The SBGT System is now required to be Operable during DAEC 2 Revision E

1 l

1 DISCUSSION OF CilANGES ITS 3.6.43: SilGT SYSTEM

. TECilNICAL CilANGES MORE RESTRICTIVli (continued) ,

Mi OPDRVs to provide mitigation if an inadvenent vessel draindown event occurs.

(cont.) 'lhe new Applicability and the addition of the Conditions and Required Actions are ;

additional restrictions to plant operation in the TS and constitute a more restrictive l change.

M2 CTS 3.7.L.3 states that with one train of SilGT inoperable, operation can continue provided the remaining SilGT train is Operable, 'lherefore, when both SilGT trains are inoperable, the plant must be shutdown in accordance with CTS 3.7.1.3. ITS 3.6.4.3 Action D is added to require entry into ITS LCO 3.0.3 immediately. LCO 3.0.3 requires the plant to be in Mode 2 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months . Therefore, the plant is required to decrease power at a faster rate than is currently re9i red. Therefore, the addition of this requirement is considered more restrictive.

M ITS 3.6.4.3, Action L .idds a requirement during Core Alterations that Core i Alterations be "Immediately" susreended if SilGT system is not restored within the allowed Completion Time. Although the CTS contains the requirement (CTS 3.7.L.3)ic sui;>end movement ofirradiated fuel, it does not contain the requirement to suspend Core Alterations (which includes not only fuel movement, but control rod movement also). Immediately suspending Core Alterations minimizes the probability of a fission product release, imposing a time limit to suspend this activity is a more restrictive change.

'l ECilNICAL CllANGES - IWI.0L ATIONS N u i S 3.7.L.1 requires the SilGT system t, be Operable if the fuel cask is being moved in the reactor building (by reference to secondary containment). CTS 3.7.L.3 requires that if one train of the SilGT System is inoperable and not restored within the Completion Time, suspend reactor buliding fuel cask movement. The reference to fuel casks will be relocated to plant procedures governing heavy loads. ,

The orocedures govemn.g heavy loads provide assurance that an appropriate level of safety is provided. Any changes to these requirements will be evaluated in accordance with the DAEC 10 CFR 50.59 program. This change is consistent with i the NUREG.

R3 CTS 4.7 L.l.f requires the SilGT System drains to be inspected quarterly for adequate water level in loop seals. This iurveillance serves to ensure prevention of excess bypass leakage, llecause other means exist to detect if a SBOT System drain loop seal has blown out (such as area radiation monitors) and because the DAEC DAEC 3 Revision E

DISCUSSION OF CilANGES i

ITS 3.6.4.3: SBGT SYSTEM TECilN.jCAI, CllANGES - REI OCATIONS (continued)

R2 design of the SBOT System is not significantly different from that of other BWRs, (cont.) this Surveillance is not necersary in Technical Specihcations and can be relocated.

Changes to the relocated requirements in plant precedures will be evaluated in accordance with the DAEC 10 CFR 50.59 program. This change is in accordance with the NUREG. ;

TECilNICAl, Cil ANGE - I.ESS RESTRICTIVE 1,i CTS 3.7.1 3 with one train of standby gas treatment inoperable nnd not restored within 7 days, requires the unit to be placed in Cold Shutdown enJ he suspension ofirradiated fuel movement. ITS 3.6.4.3, Required Action C.1, will also allow placing the Operable SBGT subsystem in operation as an ahernative to suspending mo /cment ofirradiated fuel assemblies. This attemative is less restrictive than the existing requirement. Ilowever, the proposed attemative ensures that the remaining subsystem is Operable, that no failures that could prevent automatic actuation have ,

occurred, and that any other failure would be readily detected. This change is consister t with the NUREG.

12 ITS SR 3.6.4.3.2 has been modified by a Note that delays entry for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months into the Conditions and Required Actions. During the nerfbnnance of this SR, the Note is necessary because due to a cross-tie duct between the two SBOT subsystems, the flowpath through the SHOT subsystem not being tested must be isolated, making it inoperable, to estchlish conditions necessary to ensure the tested SilGT subsystem meets the filter train differential pressur; 'equirements of the VFTP. The ability to

draw a vacuum on Secondary Containment is maintained by the subsystem under test. One hour minimizes the amount of time the SBGT subsystem is inoperable while providing enough time to perform the required testing. The allowance provided by the Note avoids potential entry into LCO 3.0.3 (Condition D) during required routine Surveillances and d e ing demonstration of Operability for a previously inoperable subsystem under LCO 3.0.5. The addition of the Note to this SR is considered acceptable due to the low probability of an event requiring system actuation during this one hour time fnune balanced against the need to perfonn Surveillances to demonstrate Operability and avoidance of an unnecessary plant shutdown while performing testing for Operability un_ - 1.CO 3.0.5, While the addition of a time limit on the allowance could be considered to be a more restrictive change to the CTS, the addition of this Note is considered to be a less t,.strictive change relative to the NUREG. This is being done for overall conservatism in characterizing this change.

DAEC 4 Revision E

>- --v.n-.- ,-y...-r------%- ,- - _-~ - ,r .,----%-- .r -- - - w ee +w c- w#~+ - ~ w---r=i --n --=+

DISCUSSION OF CllANGES ITS 3.6.4.3: SBGT SYSTEM Tl?CllNICAI. Cil ANGE - I.ESS RESTRICTIVE (Continued)

L2 Ilased upon discussions with the Stafron September 9,1997 regarding our response (cont.) to the StalTs Request for Additional Information (RAl) of February 24,1997 (Ref.

NO 971597, September 5,1997), the proposed change to add Note 2 to ITS SR 3.6.4.3.2 has been withdrawn. {3.6.4.3 5) 1.cy.2 Generic Letter 91-04 Charmes in Technical Snecification Surveillance ingah to Acconmiodate a 24 month Fuel Cvele, describes NRC requirements for preparing such license amendment requests. The Generic Letter indicates that the NRC staff has generically reviewed the extension of surveillance intervals from 18 to 24-months and fbund that "the efTect on safety is small because safety systems use redundant electrical and mechanical components and because licemcees perfbrm other surveillances during plant operation that confirm that these systems and components can perform their safety functions. Nevertheless, Licensees should evaluate the effect on safety of an increase in 18-month surveillance intervals to accommodate a 24 month fuel cycle. This evaluation should support a conclusion that the etTect on safety is small."

The Generic 1.etter specifies the fbliowing specific items for review:

Steam Generators Not applicable to DAEC

Instrument Drift Addressed independent of this review by the DAEC Setpoint Control Program

Apnendix J Exemption TS Amendment No. 219 addressed DAEC adoption of Option 11 to Appendix J. No additionalieview is required in this evaluation.

In addition, the Generic Letter indicates Licensee's should review the effect on safety of the extension of other surveillances to ensure that it is supported by historical maintenance and surveillance data.

DAEC 5 Revision E

DISCUSSION OF CllANGES ITS 3.6.4.3: SilGT SYSTEhi TFCilNICAl, Cll ANGES - I ESS RESTRICTIVli (continued)

Ley.2 Data was collected Ihr a ten year period from January 1986 to January 1996 of all (cont.) deficiencies which occurred for the surveillances for which a frequency er. tension is being sought. The ten year period was selected to ensure a browl overview of long term perfbrmance and because a similar comprehensive review was perfbnned in 1986 lbr preceding years to support changes from 12 month to 18-month intervals.

As a supplemental check, the database for 10CFR50.65 (hiaintenance Rule) compliance was reviewed to continn that equipment performance overall was compatible with a decreased surveillance frequency. The DAEC program for hiaintenance Rule includes targets for safety system train avai 't ilty and reliability compatible with assumptions in the DAEC Probabilis Safety Analysis (data fbr the hiaintenance Rule is limited to the period smee 1991).

Data lbr the ibilowing surveillance tests were reviewed:

1)escription C 1 S Section I'lS SR SilLC Squib Valve l' iring 4.4.A .2.b 3.1.7.7 SilLC i low Verincation 4.4. A.2.c 3.1.7.8 SDV Vent and Drain Cycimg 4.3.11.3 3.1.8.3 Reactor Mode Switch Channel 4. L A. ] 3.3.1.13 l'unctional RPS Response Time 4.1. A .2 3.3.1.1 K/3.3.1.19 MSL Radia!Wn Monitor Logic System 4.2.D,2.c 3.3.6 1.9 l'unctional A 1 WS RPI I ogic System i unctional 4.2.G.2 3 3.4.2.4 RP I' Inreaker Response ' lime 4.2.0.3 3.3.4.1.3/3.3.4.f.5 SV Setpoint Veritication 4.6.D. I 3.4.3.1 SRV Setpoint Verification 4.6.D.1 3.4.3.1 SRV Manual Opemng 4.6.D.3 3.4.3.2 ilPCI Low Pressure Flow 4.5.D. I .e 3.5.1.0 CS 1 ogic System i unctional 4.2.D.2.a 3.3.5.1.9 Rilk L ,ic System i unctional 4.2.II.2.b 3.3.5.1.9 Containment Spray Interlock Logic 4.2.ll.2.c 3.3.6.1.9 System l'unctional llPCI Logic System Functional 4.2.ll.2.d 3.3.5,1.9 ilPCl:RCIC Suction Transfer 4.5.D. I .! 3.5.1.7/3.5.3.5 (relocatedl ADS 1.ogic System functional 4.2.ll.2.e 3.3.5.1.9 ADS Simulated Automatic Actuation 4.5.F.1.a 3.5.1.8 ADS Valve Manual Opening 4.6 D.3 3.5.1.9 DAEC 6 Revision E

DISCUSSION OF CilANGES ITS 3.6.4.3: SilGT SYSTEh!

TECilNICAl. CilANGES - 1 ESS ItF.STitlCTIVE (continued) 1 cy.2 (cont.)

Description C1 S Section iiSSR RCIC 1.ow Pressure 4.5E I .c 3.5.1.4 Drywell to lorus 1.cak lest 4.7 E 4 3.6.1.1.2 ,

PCIV Simulated Automatic Actuation 4.7.lL I .a 3.6.l.3.6 l (Groups I . 6,8. 9)

PCIV 1.ogic System i unctional lest 4.2.A.2.a g 3.3.6.l.9 (Groups 16) til CV isolation 4.7.ll. l .c 3.6.l.3.7 1.1.S Valve Manual Openmg 4.6.D.3 3.6.l.5.1 1.1.S 1.ogic System l'unctional 4.2.ll.2.g 3.3.6.3.6/3.6.1.5.2 Secondary Containment Integrity 4.7.J. l .a 3.6.4.1.3 SCIVID Simulated Automatic Actuations 4.7.K.I 3.6.4.2.2 SilGT Simulated Automatic Actuation 4.7.1,. l .d 3.6.4.3.3 River Water Supply Simulated 4.5.J. l .a 3.7.2.4 Automatic Actuation 1SW Automatic Start w/ DG 4.8.l!.l.a 3.7.3.2 SI'U Simulated Automatic Actuation 4.10.A.3 3.7.4.3 Control fluilding Positive Pressure 4.10. A.3 3.7.4.4 1,00P/1.OCA lest 4.8.A.2.b 3.8.l.13

!!attery Service Discharge 4.8.ll. l .c 3.8.4.7 l In each of these tests, no train failures were identified by performance of the reference cyclic test during the ten year period reviewed. in each case, the system perfmw.ance was within targets established under the hiaintenance Itule. This combination of no test failures and acceptable system performance is viewed as a strong indicator that interval extension is acceptable without more detailed review.

For six Surveillance Tests, more than one failure was identified during performance of the test during the ten year interval. These tests were singled out as requiring further review prior to extending the interval.

Diesel Generator and Emergency Service Water Automatic Actuation (ITS Sit 3.7.3.2)

+

llPCI System Cycle Operability Test (ITS Sit 3.5.1.6)

IIPCI Logic System Functional Test (ITS Sit 3.3.5.1.8)

DAEC 7 Revision E

I DISCUSSION OF CllANGES ITS 3.6.4.3: SBOT SYSTEM TEClINICAL CllANGES - 1 ESS RESTRICTIVE (continued) i i

ley.2

Safety and "clief Valve Setpoint Verification and Inspection (cont.) Tests (3 tests)(ITS SR 3.4.3.1)

The majority of problems associated with failures of the Diesel Generator and Emergency Service Water automatic actuation are related to personnel or i

prowdural errors. The single exception was a failure in a diesel generator output breaker. The failures associated with the llPCI logic system functional test '

include the failure of the turbine control valve to open due to the failure of a newly incalled relay, the fcilure of a pump suction motor-operated v .lve to cycle (the valve is routinely cycled by the IST Program and would have been detected at another time), and the failure of the turbine stop valve to close due to a sticking limit switch. The failures associated with the llPCI System cycle operability test were mainly associated with the inability to reach rated flow within the specified time of 30 seconds. In each :] 4 t.. system responded within the analyzed 45 seconds. These and the other l4m uociated with this test would have been identified during the performance of s..uilar quarterly testing. The failures associated with the SRV setpoint verification and inspection tests include numerous instances of as found valves lifting more than 1% below the specified ,

setpoint and a single failure of an SRV being above the 1% setpoint tolerance (see l 1 TS change in setpoint tolerance from -l% to -3%).

For each of these tests, the nature of the failures, corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension.

The equipment performance supports interval extensions from 18 to 24 months, with a maximum proposed interval of 30 months in each case.

Lw.: Generic Letter 91-04, Opnces in Technical Soecification Surveillance Intervals to Accommodate a 24 month Fuel Cvele, describes NRC requirements for preparing such license amendment requests. The Generic Letter indicates that the NRC staff has generically reviewed the extension of surveillance intervals from 18 to 24-months and found that "the efTect on safety is small because safety systems use redundant electrical and mechanical components and because licenseer. perfonn other surveillances during plant operation that confirm that these systems and components can perform their safety functions. Nevertheless, Licensees should evaluate the effect on safety of an increase in 18-month DAEC 8 Revision E l

,. . - = - - - - - --_- - . .

DISCUSSION OF CllANGES ITS 3.6.4.3: SilGT SYSTEhi TECilNICA1.CilANGES 1.ESS RESTitiCTIVE (continued) 14c.2 surveillance intervals to accommodate a 24 month fuel cycle. This evaluation (cont.) should support a conclusion that the elTect on safety is small."

The Generic Letter specifies the following specific items for review:

Steam Generators Not applicable to DAEC

Ingrument Drift Addressed independent of this review by the DAEC Setpoint Control Program '

Appendix J Exemption TS Amendment No. 219 addressed DAEC adoption of Option 11 to Appendix J. No

, additional review is required in this evaluation, in addition, the Generic Letter indicates 1.icensee's should review the effect on safety of the extension of other surveillances to ensure that it is supported by historical maintenance and surveillance data.

Data was collected (br a ten year period from January 1986 to January 1996 of all deliciencies which occurred for the surveillances fbr which a frequency extension is being sought. The ten year period was selected to ensure a broad overview of long term perfbrmance and because a similar comprehensive review was perfbrmed in 1986 lbr preceding years to support changes from 12 month to 1H-month intervals.

As a supplemental check, the database for 10CFit50.65 (hlaintenance llule) compliance was reviewed to confirm that equipment performance overall was compatible vith a decreased surveillance frequency. The DAEC program ihr hiaintenance llule includes targets Ihr safety system train availability and reliability compatible with assumptions in the DAEC Probabilistic Safety Analysis (data fbr the hiaintenance Itule is limited to the period since 1991).

DAEC 9 Itevision E

DISCUSSION OF CilANGES ITS 3.6.4.3: SilGT SYSTEM TECilNICAl, CllANGES I.ESS Rl!STRICTIVE (continued) 1.m.2 Data for the fbilowing surveillance tests were reviewed:

(cont.)

IX scription C'l S Section IiS SR CS Simulated Auto 4.5. A. I .a 3.5.l.7 Actuation 1.PCI System Simulated 4.5.A.3.a 3.5.1.7 Automatic Actuation llPCI Simulated Automatic 4.5.D. l .a 3.5.1.7 Actuation RCIC Simulated 4.5.li. l .a 3.5.3.5 Automatic Actuation SRV Pressure Switch 4.2.11.2 g 3.6.l.5.2 System functional iorus/ Reactor fluildmg 4.7.D.3 3.6.1.6.3 Vacuum lireaker SilGT liypass Darnper 4.7.t.. l .c 3.6.4.3.4 in each of these tests, no train failures were identified by performance of the reference cyclic test during the ten year period reviewed. In each case, the system perfbrmance was within targets established under the Malatenance Rule. This combination of no test failures and acceptable system performance is viewed as a strong indicator that interval extension is acceptable without more detailed review.

The equipment perfbrmance supports interval extensions from annual to 24 months, with a maximum proposed interval of 30 months.

DAEC 10 Revision E

i DISCUSSION OF CilANGES ;

ITS 3,6: CONTAINMENT SYSTEMS IIASES ;

The liases of the CTS for this section (pages 3.7 22 through 3.7-43) have been completely replaced by revised liases that reflect the fonnat and applicable content ofITS Section 3.6, consistent with the NUREG. The revised liases are as shown in the ITS Ilases. ,

1 i,

i 4

i DAEC 1 Revision E

DISCUSSION OF CilANGES TO NUREG 1433 I SECTION 3.6- CONTAINMENT SYSTEMS -

Pl. ANT SPECIFIC CIIANGES (continued)

Pu ne second frequency to NUREG SR 3.6.1.8.1 requires the Suppression Chamber-to Drywell Vacuum Breakers to be verified closed after they may have been opened. This frequency is not needed. Surveillances must be continually met (ITS SR 3.0.1), thus if the vacuum breakers are open and the Surveillance Frequency is not due yet, the SR would still be considered not met, and appropriate Actions taken. There are many other instances where valves are required to be closed, and verified closed on a periodic basis. If these other valves are cycled (e.g., ECCS

valves), plant administrative controls ensure they are left in the correct position; a "special frequency" of the Surveillance is not required. in addition, these vacuum breakers have position indication in the control room, and are continuously monitored by control room operators. If conditions exist for the vacuum breakers to be potentially opened (e.g., venting the drywell), control room operators would be alert to the possibility and ensure the vacuum breakers were closed at the completion of the evolution.

The second and third frequencies to NUREG SR 3.6.1.8.2 require a functional test Pn of the vacuum brcokers (i.e., cycle the vacuum breakers) within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months aller the vacuum breakers have cycled, or after an operation that may have caused them to cycle.

Since the vacuum breakers are designed tu operate and assumed to function aller a LOCA blowdown, their operation as designed aller some other minor steam release from the SRVs should not raise questions regarding the immediate Operability of the vacuum breakers. Furthem1 ore, the steam quenching from the discharge of an SRV has been enhanced by the addition of"T" - quenchers since this frequency was first imposed. Steam discharged to the Torus, resulting in increased wetwell pressure and vacuum breaker opening, might pose a long tenn equipment degradation issue, rather than any immediate Operability concern. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months frequency would be meaningless to detect long term degradation, while the nomial 31 day frequency would more than suffice for this concern.

In addition, a review of vacuum breaker failures was perfomied and noted that no failures were due to the valves not opening. Thus it is not appropriate for DAEC, which does not have these current frequencies, to verify the vacuum breakers will open af ter they havejust opened.

DAEC 4 Revision E l

. , , . . ~ , . . . , .-. - - . . . , . _ . . . . - - - -,

--r. .._._,_c

,-, , . _ .e., .. - - - . -

DISCUSSION OF CilANGES TO NUREG 1433 SECTION 3.6- CONTAINMENT SYSTEMS EANT SPECIFIC CilANGES (continued)

Pu His Specification was deleted because the function does not exist at DAEC.

Pn ITS 3.6.3.2 Applicability was revised for clarity and to reflect the D AEC specific licensing basis. Additionally, the change achieves consistency with the 13ases of the NUREG for the Applicability of the Primary Containment Oxygen Concentration Specification.

Pu, NUREG SR 3.6.2.3.1 requires that every 31 days, each RIIR suppression pool cooling subsystem manual, power operated, and automatic valve in the flow path that is not locked, scaled, or otherwise secured in position, be verified in the correct position or can be aligned to the correct position. The cunent DAEC TS does not contain this SR and the valve lineup check is not in current licensing basis. DAEC has adequate controls on the manual valves such that monthly checks are not required. The controls are as follows:

1. Administrative controls are adequate to ensure manual valves are maintained in the proper position. These controls include:

a. Independent valve lineup verifications following outages when a system has been taken out ofits nomial lineup.

b. Independent licensed operator preparation and. verification of tagouts and independent placement and verification of placement of these tagouts.

2. The Locked Valve Program at DAEC requires that all Safety System manual valves that could prevent the fulfillment of the safety function of the system shall be locked in their proper position as indicated on the Locked Valve Listing when the system is required to be Operable. Therefore, additional, periodic requirements to check manual valve positions are unnecessary.

3. Verifying rome system valve positions will require entry into radiation areas and will result in increased dose. Not perfomling these position checks will help maintain dose ALARA.

DAEC 5 Revision E l

I l

DISCUSSION OF CilANGES TO NUREG 1433 l SECTION 3.6--CONTAINhiENT SYSTEhiS l Pl. ANT SPECIFIC Cil ANGES (continued) l Pn, 4. hilspositioned open vents and drains will be detectable by water on floors,

)

(cont.) increased sump leakage or by decreases in tank inventories much sooner '

than the 31 day surveillance. Additionally, periodic walkdowns (required by CTS 6.8.5.1 and CTS 6.8.5.2 and ITS 5.5.2) would detect mispositioned i open vents and drains. !

5. Verifying positions of remotely operated hiOV's ensures that major diversion flowpaths (i.e., minimum flow, test, etc.) are properly aligned.

Pn NUREG 3.6.2.4, RilR Suppression Pool Spray, was deleted since the DAEC specific analysis does not nssume operation of diis system. See DAEC split report for furtherjustification for deletion.

Pa NUREG 3.6.2.5. Drywell to-Suppression Chamber Differential Pressure, was deleted since the DAEC specific analysis does not assume a differential pressure is maintained between the drywell and the suppression chamber.

Pn, NUREG 3.6.3.2, Drywell Cooling System Fans was deleted since the DAEC accident analysis does not assume Drywell Cooling System fans are available to assure adequate mixing.

Pm NUREG SR 3.6.3.4.2 requires that every 31 days each CAD subsystem manual, power operated, and automatic valve in the flow path that is not locked, scaled, or otherwise secured in position, be verified in the correct position or can be aligned to j the correct position. The current DAEC TS does not contain this SR and the valve lineup check is not in current licensing basis. DAEC has adequate controls on the manual valves such that monthly checks are not required. These controls are as follows:

1. Administrative controls are adequate to ensure manual valves are maintained in the proper position. These controls include:

a. Independent valve lineup verifications following outages when a system has been taken out ofits nonnal lineup.

DAEC 6 Revision E l

,~ ,______~~_ _..

DISCUSSION OF CllANGES TO NUREG 1433 SECTION 3.6--CONTAINMENT SYSTEMS jl6FT SPECIFIC CilANGES (continued) l Po2 b. Independent licensed operator preparation and verification of tagouts (cont.) and independent placement and verification of placement of these tagouts.

2. The Locked Valve Program at DAEC requires that all Safety System manual valves that could prevent the fulfillment of the safety function of the system shall be locked in their proper position as indicated on the Locked Valve Listing when the system is required to be Operable. Therefbre, additional, '

periodic requirements to check manual valve positions are unnecessary.

3. Verifying some system vulve positions will require entry into radiatic.) areas and will result in increased dose. Not performing these position checks will hcip maintain dose ALARA.

Pn Clarification provided to reflect the DAEC specific design and tenminology.

Editarial changes made including renumbering, as necessary.

P2 NUREG SR 3.6A.2.1 requires every 31 days that each secondary containment isolation manual valve and blind flange that is required to be closed during accident conditions, be verified closed. The current DAEC TS does not contain this SR and the valve lineup check is not in current licensing basis. DAEC has adequate controls on the manual valves and flanges such that monthly checks at not required. These controls are as follows:

1

1. Administrative controls are adequate to ensure manual valves and flanges ire maintained in the proper position. These controls include:

a. Independent valve lineup verifications following outages when a system has been taken out ofits nomial lineup.

b. Independent licensed operator preparation and verification of tagouts and independent placement and verification of placement of these tagouts.

DAEC 7 Revision E l

t DISCUSSION OF CilANGES TO NUREG 1433 SECTION 3.6--CONTAINMENT SYSTEMS Pl. ANT SPECIFIC CllANGES (continued)

P22 (continued)

c. Flanges are only positioned using maintenance work control documents.

2. The Locked Valve Program at DAEC requires that all Safety System manual valves that could prevent the fulfillment of the safety function of the system shall be locked in their proper position as indicated on the Locked Valve Listing when the system is required to be Operable. Therefore, additional, periodic requirements to check manual valve positions are unnecessary.

3. Verifying some system valve positions will require entry into radiation areas and will result in increased dose. Not perfomiing these position checks will help maintain dose ALARA.

P23 ITS SR 3.6.4.1.1 has been revised to verify all secondary containment equipment hatches are closed. The temiinology "and sealed", as currently stated in the NUREG has been deleted. As stated in the NUREG Base.;," scaled"in this context has no connotation ofleak tightness. The intent of this requirement is to verify that there is some form of seal (gasket) between the door and the flange. DAEC is deleting this requirement to verify the seals for the following reasons:

1, The subject seals are inspected in accordance with plant procedures any time the equipment hatches are opened or when any maintenance is perfomied.

2. Gross seal leakage or degradation will be identified as a result of the secondary containment leak rate test.

3. The subject seals cannot be entirely inspected and verified unless the equipment hatches are opened in order to view the entire seal. To open the equipment hatch to verify the seal will require the reactor to be brought to Mode 4 since opening the hatch will breach secondary containment integrity.

DAEC 8 Revision E l

-- , - . - = _ _ - - . -_ -- - .- __ .. . - - - _ . - - .

DISCUSSION OF CllANGES TO NUREG 1433 SECTION 3.6--CONTAINMENT SYSTEhiS Pl ANT SPECIFIC CIIANGES (continued)

P 24 NUREG 3.6.1.3 Note I to Actions contains bracketed provisions that prohibit purge valve flow paths from being unisolated intermittently under administrative i controls. This restriction is being deleted in the DAEC ITS since it is not in the CTS, the purge valves will close during a LOCA event and these valves are not currently opened routinely during plant operation.

t Pu Per our Response to NRC Question 3.6.1.3-7 and the DAEC's prior commitment under Amendment #219, the second frequency of STS SR 3.6.1.3.7 (ITS SR ,

, 3.6.1.3.4) will be retained. (3.6.1.3-7)

P2,, NUREG SR 3.6.1.6.1 for manually opening the LLS valves, contains a frequenc> of I8 months on a Staggered Test Basis for each valve solenold. Since DAEC only has one solenoid per valve, testing on a Staggered Test Basis is not applicable.

Pn NUREG 3.6.2.1 for suppression pool average temperature contains limits based on IRhis being > [25/40] divisions of full scale on Range 7. TilERMAL POWER in the range of 1% RTP is not readily quantified with much accuracy. While Range 7 on IRhis approximates 1% RTP, this power level can also be approximated from SRhis and even by determining the point of adding heat. These acceptable options are desired to be maintained in plant procedures, with the CTS 1% RTP requirement ma:.-tained in ITS. Therefore, the LCO and Actions have been modified to reflect ,

the 1% RTP requirement.

P:s At DAEC, gaseous nitrogen is used in the CAD System. 1 P,2 NUREG SR 3.6.1.1.2 contains a second frequency of 9 months aller two consecutive tests fail (and continues until two consecutive tests pass). This testing is not in the DAEC current licensing basis and is not being adopted. This new test frequency would require mid-cycle testing which would require a plant shutdown due to the current testing method. The current test method involves blocking open DAEC 9 Revision E l

. - . . . . . . . - . . - , . -, .~. - ,

_ _ _ _ _ _ _ ~ . _ _ _ _ _ __ __ _ _ . .._ ._ _

i DISCUSSION OF CilANGES TO NUREG 1433 SECTION 3.6.-CONTAINMENT SYSTEMS PLANT SPECIFIC CilANGES (continued)

P3 the suppression chamber-to-reactor building vacuum breakers to establish initial (cont.) conditions, pressurizing and maintaining the drywell at 1.2 psid above suppression chamber pressure, and then measuring the rate at which suppression chamber pressure increases.

~

Pn The acceptance criteria of restricting flow to s 1 gph for EFCVs is not applicable at DAEC. Per UFSAR 6.2.6.3.3, the acceptance criteria is observing a marked c'ecrease in flow rate.

Pu Per our Response to the Staffs RAI on this Note (Ref. NG-97-1597) and our meeting with the Staff on September 9,1997, this change has been withdrawn.

(3.6.1.3 8)

Pn DAEC uses the tenninology " vacuum breaker assembly" in the CTS and this is being n tained in the ITS. At DAEC there are two vacuum breaker assemblies located in parallel in two branch lines off a single penetration going into the suppression chamber. Each vacuum breaker assembly contains two vacuum breaker valves, an air operated butterfly valve and a check valve. ITS 3.6.1.6 has been modified to reflect the DAEC terminology without changing the intent of the TS.

l Pn CTS 3.7.11 allows the Containment Atmosphere Dilution System to be inoperable for 7 days with no attemate hydrogen control available. At DAEC, veming of the Primary Containment through the Standby Gas Treatment (SBGT) System for hydrogen control, even though not required ir. the original licensing basis, can be used as an attemate means of hydrogen control. Since the SBGT System is a TS required function with its own Operability requirements, which are more restrictive than CAD (i.e., enter LCO 3.0.3 upon loss of both SBGT subsystems), it is not necessary to include NUREG Action B.l. Because of the diversity within the hydrogen control function et the DAEC (i.e., inerted containment), the NUREG has been modified to recognize that TS Action is not required until the entire CAD System is inoperable (i.e., either because both nitrogen injection subsystems are inoperable or the common nitrogen bank is inoperable). The CTS allowance of 7 days has then been implemented in the ITS to require establishing Operability of one nitrogen injection subsystem and the nitrogen storage bank and establishment of a " required" flowpath to both the Drywell and Suppression Pool.

2 DAEC 10 Revision E l

DISCUSSION OF CilANGES TO NUREG 1433 -

SECTION 3.6--CONTAINMENT SYSTEMS i PLANT SPECIFIC CilANGES (continued)

Py in addition, the NUREG has been changed to reflect the CTS Mode of (cont.) Applicability of the reactor in " reactor power operation"(i.e.,21% RTP) rad the primary containment inerted per CTS 3.7.1.1. (ITS 3.6.3.2). [ Note: Bect.use the requirements of CTS 3.7.1.1 bound those of 3.7.11.1, they will be used.] This change is necessary as the design of the DAEC CAD System can not, by '.tself, inert the primary containment (Ref. UFSAR 6.2.5.3). Thus, the CAD system is not Operable until the containment is already inerted. Consequently, the LCO Applicability has been changed to be consistent with that for the mntainment to be inerted (ITS 3.6.3.2); otherwise, the plant could never startup frm.. Modes 3 or 4, per LCO 3.0.4. (3.6.3.1 1 } and {3.6.3.1-5)

Pu NUREO 3.6.2.3, RilR Suppression Po al Cooling, has been changed to reflect DAEC current licensing basis assumphons for Operability. UFSAR Section 6.2.1.3.3.2 and Table 6.216 documents the long tenn primary containment response ihr a LOCA. The LOCA analysis shows that one RilR pump with one heat exchanger (with 2 RilRSW pumps) discharging via the LPCI injection valve is sufficient to limit peak suppression pool temperature to 200 'F. In response to NUREG-0783, which imposed local and bulk suppression pool temperature limits, NEDC 22082 (and Supplement 1) and NEDE-30051 (for power uprate and reduced RilRSW flow) were generated. These analyses document suppression pool temperature response to certain non-limiting LOCAs (small breaks) and transients (stuck open SRV). In order to meet the requirements of NUREG-0783 for suppression pool temperature limits, one k>op of suppression pool cooling is required, consisting of two R1IR pumps, one heat exchanger and 2 RilRSW pumps.

The DAEC analysis suppmts the changes to the NUREG by requiring a minimum of two RilR pumps to be Operable in Modes 1,2 and 3. Requiring subsystem Operability ensures that the necessary heat exchangers and pumps are available. In addition, a new Action has been added (Action D) to provide a restoration time for when both suppression pool cooling subsystems are inoperable. The proposed 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Completion Time is consistent with the time provided in the NUREG when both RI1RSW subsystems are inoperable. The time is considered appropriate since an immediate shutdown has the potential for causing plant trip or transient. The 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months provides some time to restore one of the subsystems prior to requiring a shutdown (thus precluding the potential to cause a plant trip or transient) yet is short enough that it does not significently increase the probability of an accident occurring during this additional time DAEC 11 Revision E l 1

l l

l DISCUSSION OF CllANGES TO NUREG 1433 SECTION 3.6--CONTAINMENT SYSTEMS PI, ANT SPECIFIC CllANGES (continued)

P 33 ITS SR 3.6.4.3.2 has been modified by a Note delaying entry into the Conditions and Required Actions for one hour. During the performance of this SR, the Note is necessary because due to a cross-tie duct between the two SBOT subsystems, the flowpath through the SBGT subsystem not being tested must be isolated, making it inoperable, to establish conditions necessary to ensure the tested SBOT subsystem meets the filter train differential pressure requirements of the VFTP. The ability to draw a vacuum on Secondary Containment is maintained by the subsystem under test. One hour minimizes the amount of time the SBOT subsystem is inoperable while providing enough time to perform the required testing. The allowance pmvided by the Note avoids potential entry into LCO 3.0.3 (Condition D) during required routine Surveillances and during demonstration of Operability under LCO 3.0.5. The addition of the Note to this SR is considered acceptable due to the low probability of an event requiring system actuation during this one hour time frame balanced against the need to perform Surveillances to demonstrate Operability and avoidance of an unnecessary plant shutdown while performing testing for Operability under LCO 3.0.5. {3.6.4.3 or 5}

P. 3 NUREO SR 3.6.2.3.2 has been revised to change ">" to "2". DAEC's RilR pumps are cach required to be able to deliver a minimum flowrate of 4800 gpm. The addition of the " equal to" allowance is less restrictive but the additional margin is so small that it is insignificant and, therefore, is acceptable.

Pn NUREG 3.6.1.3 Action I has been changed by modifying the Required Action to initiate action to suspend OPDRVs. This condition is only applicable for the RllR Shutdown Cooling valves. Therefbre, the only "OPDRVs" that need to be suspended are those associated with the RilR Shutdown Cooling System.

P3 The one hour requirement to maintain 2 0.25 inch vacuum has been deleted.

DAEC's phmt procedures require a 10 minute hold to allow conditions to stabilize before taking readings. The 10 minute stabilization has been adequate in the past to ensure adequate equilibrium conditions are obtained and maintaining conditions ihr one hour would cause undo burden on refuel outage scheduling without any real safety benefit. Additionally, DAEC does not have instrumentation which can record Secondary Containment vacuum continuously. DAEC uses the average of four manometers with one manometer mounted on each face of the Reactor Building to negate the efTects of wind direction. These manometers must each be read locally and then manually averaged to determine the Secondaiy Containment vacuum value.

DAEC 12 Revision E l

DISCUSSION OF CllANGES TO NUREG 1433 .

SECTION 3.6- CONTAINhiENT SYSTEhfS P1, ANT SPECIFIC ClIANGES (continued)

Pw Revised ITS 3.6.1.3 second Applicability to more clearly state what valves are !

applicable (Shutdown Cooling System Isolation Valves). The second Applicability statement as originally written was confusing as to when it applied and to what it applied. The revised Applicability statement defines the hiodes and System to which this statement applies. For DAEC, the only instmmentation required to be Operable in hiodes 4 and 5 is Reactor Low 1.evel which will close the Shutdown Cooling System Isolation Valves.

Po 4 NUREG SR 3.6.1.3.8 requires hiSIV isolation time to be "2" and "s" the DAEC specific times. Current DAEC testing requires isolation tina "between 3 and 5 seconds." DAEC will change "2" to ">" and "s" to "<."

Pu NUREG SR 3.6.1.3.11 requires testing of the explosive squib every I 8 months on a STAGGERED TEST llASIS. CTS does not have any requirements for testing TIP Squibs but DAEC has been testing them per the ash 1E program. This change will be consistent with current DAEC practices and is still a more restrictive change.

Pu The goveming code for the IST Program (Oh!.10/Ohi 1, Paragraph 1.3.4.3) requires operability testing for primary containment vacuum relief valves to be perfonned every 6 months, unless historical data indica'es a requirement for more frequent testing. The functional testing specified by ITf. SR 3.6.1.6.2 meets the intent of the " operability tests" discussed in the cale kcause this test ensures the vacuum breaker assembly valves open properly to perfonn its design function. This agrees with the NUREG llases.

Pc Per our Response to NRC Question 3.6.2.2 1, this change has been withdrawn.

(3.6.2.2-1 }

Pu NUREG SR 3.6.4.1.1 and SR 3.6.4.1.4 are being deleted. hiaintaining a negative pressure in the secondary containment during nonnal plant operations is not assumed by any of DAEC's accident analyses, nor is attainment of a negative 1/4-inch required within a given time period to mitigate the consequences of any accident.

Pc Correction of typographical error. [CRF 9108]

DAEC 13 Revision E l

DISCUSSION OF CHANGES TO NUREG 1433 SECTION 3.6-CONTAINhiENT SYSTEhtS PLANT SPECIFIC CilANGES (continued)

Py, llased upon discussions with the Staff on September 9,1997 regarding our response to the Stafrs Request for Additional Information (RAI) of February 24, 1997 (Ref. NG 97-1597, September 5,1997), the proposed change to add Note 2 to ITS SR 3.6.4.3.2 has been withdrawn. {3.6.4 3-5)

Po This Note has teen deleted since it is not needed; purge valves are not required to be Operable in himies other than 1,2, and 3. The Applicability of this LCO is only in hiodes 1,2, and 3: Modes 4 and 5 are only applicable for Shutdown Cooling System isolation valves when the associated instrumentation is required to be Operable per LCO 3.3.6.1.

Pu NUREG SR 3.6.1.3.1 has been deleted since DAEC is not required to maintain the purge valves scaled closed. CTS 3.7.11.4 requires "... purge valves may not be opened so as to create a flow path from the primary containment while PRlhiARY CONTAINMENT INTEGRITY is required except for inerting, de-inerting, vent / purge valve testing, or pressure control." This CTS requirement is being retained in the Note to NUREG SR 3.6.1.3.2 (ITS SR 3.6.1.3.1).

P.w NUREG 3.6.1.3 Actions G and 11 are in brackets and have been deleted since they are not applicable to DAEC In the Mark I containment design, the Refuel Floor is part of the secondarv containment, not the Primary Containment where the purge / vent valves are kicated.

Po3 NUREG SR 3.6.1.3.14 requires that the combined leakage rate through hydrostatically tested lines that penetrate the primary containment be verified to be within limits given in the Primary Coniainment Leakage Rate Testing Program. At the DAEC, the leakage from tests conducted with water is added to the air leakage totals to demonstrate that total leakage is within acceptable limits. The Technical Evaluation Report (TER) for the DAEC's Containment Leakage Rate Testing, dated hiarch 17,1982, states that this is acceptable. The TER states perfomiing tests with water and adding the resuhs to the air leakage totals to detemiine compliance with leakage limits is consemitive with regards to the requirements of Appendix J. ITS SR 3.6.1.1.1 requires the performance ofleakage rate testing in accordance with the Primary Containment Leakage Rate Testing Program and thereby adequately addresses testing of hydrostatically tested lines, making this SR redundant and unnecessary, Therefore, NUREG SR 3.6.1.3.14 has been deleted from the ITS.

DAEC 14 Revision E l

. ~_ _ _ - _ _ _ . ._ .-_ -_. _ _ _ __ _ ___ ___ _.

DISCUSSION OF CilANGES TO NUREG 1433 SECTION 3.6--CONTAINMENT SYSTEMS Pl. ANT SPFCIFIC CHANGES (continued)

Psi SR 3.6.1.2.2 is being revised to require testing of the airlock door interlocks at an interval of 24 mc Ss. This is consistent with Appendix J Option B, which allows for an extension oi e overall airlock leakage test frequency to a maximum interval of 30 months. The Note to the SR is deleted since it will no longer be required due to the frequency change. This change is consistent with Generic Traveler TSTF-17.

P 32 NUREG SRs 3.6.1.3.6 ITS SR 3.6.1.3.3 and 3.6.4.2.2 (ITS SR 3.6.4.2.1) are being revised for clarity. The purpose ofITS SRs 3.6.1.3.3 rad 3.6.4.2.1 are to ensure the '

valves and dampers will isolate in a time period less tisan or equal to that assumed in the safety analysis. There may be valves and dampers creMted as containment isolation valves which are power operated that do not receive a containment isolation signal. These power operated valves and dampers do not have an isolation l time as assumed in the accident analysis since they requ:re operator action.

Therefore, restricting the SR to power operated automatic isolation valve and damper time testing reduces the potential for misinterpreting the requirements of this SR while maintaining the assumptions of the accident analysis. This revision is consistent with the Generic Traveler TSTF-46, Rev.1.

{3.6.1.3 11 and 3.6.4.2-9).

P 33 ITS SR 3.6.4.1.2 has been modified from NUREG 1433 to retain the current licensing basis for the DAEC. Dermition 16.a of the CTS requires that "At least one door in each access opening is closed." Because the DAEC design has secondary containment penetrations with three access doors, the NUREG wording was modified to allow multiple doors to be open that do not compromise secondanj containment integrity. Closure of either an inner or outer access opening ensures that any other access opening will not prevent the Secondary containment from performing its intended safety function. (3.6.4.1-2}

Pu in 60 FR 49495 dated September 26,1995, the NRC published an amendment to Appendix J to 10 CFR Part 50. This amendment became effective on October 26, 1995 and revised Appendix J to allow Licensees the choice of complying with either the new perfonnance based requirements (Option B) or previously existing perspective requirements (Option A). Regulatory Guide 1.163," Performance -

Based Containment Leak - Test Program," was issued to provide guidance on the implementation of Option B. The CTS and NUREG aie being modified consistent with DAEC Request for Technical Specification Change RTS-269 (NG-95-2985) dated December 22,1995 and NRC letter from Christopher 1. Grimes (Chief Technical Specifications Branch, USNRC) to David J. Modeen (Director, DAEC 15 Revision E l

DISCUSSION OF CilANGES TO NUREG 1433 SECTION 3.6-CONTAINhtENT SYSTEMS Operations and Management, NEI) dated November 2,1995, implementing Option 11.

l l

l 1

4 l

4 x

DAEC 16 Revision E l

i DISCUSSION OF CilANGES TO NUREG 1433 BASES SECTION 3.6--CONTAINMENT SYSTEMS Pl. ANT SPECIFIC CllANGES Pi ne plant specific nomenclature, number, reference, system description, or analysis description was used to reflect DAEC (additions, deletions, and/or changes e.re included).

P2 llases revised ihr enhanced clarity, to correct typographical errors, or to be consistent with similar phrases in other parts of the Bases.

P3 Note 4 to the Actions and Conditions A and il ofITS 3.6.1.3 have been modified from the current NUREG cxception for purge valve leakage not within limits, to include exceptions also to MSIV leakage not within limits. These exceptions are acceptable since other Actions in ITS 3.6.1.3 address these exceptions. The Bases have been changed to reflect the changes to the Specifications.

P4 At DAEC the reduced pressure test is not used and the reference to this Type A Testing is deleted.

P3 DAEC was not licensed to the GDC's. Ilowever, the DAEC has been evaluated to show that the intent of each GDC is substantially met. The appropriate UFSAR Section that documents these evaluations is referenced in place of the GDC itself.

P6 A discussion has been added to the Applicable Safety Analyses for the Bases of ITS 3.6.1.4 (Drywell Air Temperature). This discussion merely reflects the Current 1.icensing Basis, as discussed in UFSAR 6.2, and as such, is acceptable.

P7 A discussion has been added to the Background section of the liases fbr NUREG 3.6.4.3 explaining that the 0.25 inches water gauge negative pressure in Secondary Containment is an average of four manometer type pressure gauges, and that ').25 inches water gauge negative pressure includes margin to a negative pressure that ensures zero exfiltration. This is ime since maintaining 0.25 inches negative pressure under calm wind conditions ensures a negative pressure (with respect to the lowest pressure building face with wind present) under worst case conditions.

Ps The Bases have been revised to reflect changes to the associated technical specification.

'l!

DAEC 1 Revision E l

DISCUSSION OF CilANGES TO NUREG 1433 BASES SECTION 3.6--CONTAINMENT SYSTEMS PIANT SPECIFIC CilANGES (continued)

P, Amendment No. 201 to DAEC CTS allows the functions of"inerting or pressure adjustment" as part of the drywell vacuum breakers intended function and is acceptable so inclusion of"inerting or de inerting containment"into the llasis for ITS LCO 3.6.1.7 is acceptable.

Po i Changes to References made to reDect specific DAEC requirements.

Pn The sentence in the Background section of the Bases for 3.6.4.1 (Secondary Containment) regarding a possible control volume pressure rise due to pump and motor heat has been deleted. This sentence was confusing and did not add value to the discussion.

Pn Exemples of ha' hes subject to the Surveillance Requirements has been added to the Bases discussion for SRs 3.6.4.1.1 and 3.6.4.1.2 Po The Bases for the 31 day Frequency for SR 3.6.4.3.1 has been changed to reHect that preventing moisture build-up in the charcoal is the primary reason for the surveillance. This agrees with the preceding statement in the Bases. Also, the reference to excessive vibration has been deleted, since vibration data is not collected during this surveillance.

Pn ITS SR 3.6.1.6.3 frequency discussion has been changed to reflect the ability to perform this surveillance at any time and to agree with the assumption of a 12 month calibration interval in the determination of the magnitude of equipment drill in the setpoint analysis.

Pn This sentence is being deleted. Current DAEC operating practice involves controlling the passive Secondary Containment boundary as a single general category. No component testing for the individual passive devices was relied on in the 1.icensing Bases for the DAEC Administrative Control Procedure (ACP) 1410.6, Temporary Modification Control, contains the appropriate precautions / procedures to ensure secondary containment penetrations are adequately controlled. Changes to this procedure will be evaluated in accordance with the DAEC 10 CFR 50.59 program. {3.6.4.2-7) l Pa A statement was added to Bases LCO 3,6.4.2 stating that a utility penetration may be open and Secondary Containment maintained Operable as long as it can meet the negative pressure Surveillance Requirements.

DAEC 2 Revision E l i

i

DISCUSSION OF CilANGES TO NUREG 1433 BASES SECTION 3.6- CONTAINhiENT SYSTEh1S P1, ANT SPECIFIC CilANGES (continued)

Pn Per our Response to the Staff's RAI on this Note (Ref. NG 97-1597) and our ;

meeting with the Staff on September 9,1997, this change has been withdrawn.

(3.6.2.3-5 } { 3.6.3.1-4 }

i Pa in 60 FR 49495 dated September 26,1995, the NRC published an amendment to ;

Appendix J to 10 CFR Part 50. This amendment became efTective on October 26, 1995 and revised Appendix J to allow Licensees the choice of complying with either the new performance based requirements (Option B) or previously existing perspective requirements (Option A). Regulatory Guide 1.163,"Perfonnance -

Ilased Containment Leak - Test Program," was issued to provide guidance on the impJementation of Option B. The CTS is being modified in the NUREG consistent w.in DAEC Request for Technical Specification Change RTS 269 (NG-95-2985) dated December 22,1995 and NRC letter from Christopher 1. Grimes (Chief Technical Specifications Branch, USNRC) to David J. hiodeen (Director.

Operations and hianagement, NEI) dated November 2,1995, implementing Option B.

Pn, Changed all .eferences to "the NRC Policy Statement" to its associated Section in 10 CFR 50.36.

Pa 2 To verify the setting of the suppression chamber to-drywell vacuum breakers, personnel must physically enter the suppression chamber which is not possible with the containment inerted. Therefore, it is impossible to perform this surveillance with the reactor at power and this statement in the Bases is inappropriate and has been deleted.

P 2i The specific requirement for the subsystems to be powered from two safety related independent power supplies has oeen deleted since the design of the system already reflects this. This statement is not used in other LCO Bases where the system is designed with independent power supplies (e.g., Bases 3.6.4.3,

" Standby Gas Treatment System").

DAEC 3 Revision E l

DISCUSSION OF CilANGES TO NUREG 1433 SECTION 3.6--CONTAINMENT SYSTEMS PLANT SPECIFIC CllANGES Pi NUREG SR 3.6.1.2.1 has been revised to reflect DAEC Request for Technical Specification Change RTS 269 (NG 95-2985, dated Dec. 22,1995).

P2 The Drywell Pressure LCO (NUREG 3.6.1.4) has been deleted. The NUREG LCO is based on the initial assumption of.75 psig in the safety analysis, and is required in Modes 1,2, and 3. A recent GE evaluation shows that an initial drywell pressure of 2.0 psig is acceptable for ensuring containment pressure design limits are not exceeded. This LCO is not needed since the RPS high drywell pressure scram will trip the unit prior to exceeding 2.0 psig, effectively placing the unit in Mode 3.

While the RPS trip is not required in Mode 3, the DAEC Emergency Operating Procedures (EOPs) will govem actions if the drywell pressure exceeds 2.0 psig (cfrectively bounding the 2.0 psig limit). The EOPs will require entry into the RPV Control and Primary Containment Control actions. These actions require steps to be taken to reduce primary containment pressure to less than 2.0 psig and to cooldown the reactor at nomial coo!down rates to Mode 4 if pressure carmot be reduced to less than 2.0 psig.

P3 Conditions A and B contain in brackets, an exception for purge valve leakage not within limits in addition to purge valve leakage, MSIV leakage is also added as an exception to Conditions A and B. Changes to Condition D will replace secondary containment bypass leakage rate not within limits, with a requirement for one or more penetration flow paths with one or more MSIVs not within leakage limits.

These changes are rnade to account for MSIV leakage in addition to purge valve leakage, which is a.:cced for in Condition E. Sermtdary Containment bypass leakage is not in the DAEC current licensing basis. At DAEC leakage from hydrosta ically tested lines is included in the Type C test total as required by CTS 4.7.A.I.c.4. Note 4 to the Actions has also been modified to indicate that the PCIV leakage described is only fbr MSIVs or purge valves, as these are the only valves for which leakage require nents are contained in ITS 3.6.1.3. Additional leakage testing requirements are contained in ITS 3.6.1.1 and 5.5.12 (Primary Containment Leakage Rate Testing Prognun). (3.6.1.3-2)

DAEC 1 Revision E l

i i

DISCUSSION OF CllANGES TO NUREG 1433 -

SECTION 3.6 -CONTAINh1ENT SYSTEhtS Pl. ANT SPECIFIC CilANGES (continued)

P4 The time to restore MSIV leakage to within limits has been changed to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , consistent with the time to restore an inoperable MSIV Oar reasons other than leakage)in Action A. Action A allows 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months to isolate the afTected main steam line when an MSIV is inoperable due to a reason not involving leakage. This could include a MSIV that will not automatically isolate (which means it is essentially fully open). Action D was modified to include MSIV leakages (Generic Change UWR-15, C4), and it appears to not have fully been changed to allow the 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months in Action A, which is the Action that would have been entered for a leakage problem prior to the generic change.

P3 At DAEC the containment purge valves are located outside containment and there are no penetration isolation devices inside containment. Therefore, the allowance in NUREG 3.6.1.3 Required Action E.2 for these devlces is not needed and is deleted.

I'6 The changes to NUREG SR 3.6.1.1.2 are made to reflect testirg methods currently in place at DAEC. The current testing method verifies the average rate of suppression chamber pressure change over a 10 minute period is < 0.009 psi / minute.

DAEC UFSAR Section 6.2.6.3.5.1 assumes a dilTerential pressure of greater than 1 psid.

P7 NUREG SR 3.6.1.3.3 and SR 3.6.1.3.4 require each primary containment manual isolation valve and blind flange that is located either inside or outside primary containment, and that is required to be closed during accident conditions, is verified to be closed every 31 days. DAEC will not adopt these SRs for the following reasons:

l. The current DAEC TSs do not contain these SRs and the valve lineup checks are not in current licensing basis.

2. Administrative controls are adequate to ensure manual valves and flanges are maintained in the proper position. These controls include:

a. Independent valve lineup verifications following outages when a system has been taken out ofits nonnal lineup,

b. Independent licensed operator preparation and verification of tagouts and independent placement and verification of placement of these tagouts.

DAEC 2 Revision E l

DISCUSSION OF CllANGES TO NUREG 1433 t SECTION 3.6--CONTAINMENT SYSTEMS

l>LANT SPECIFIC ClIANGES (continued)

P7 c. Flanges are only positioned using controlled maintenance work (cont.) documents, ,

3. The Locked Valve Program at DAEC requires that all Safety System manual valves that could prevent the ftdfillment of the safety function of the system shall be locked in their proper position as indicated on the Locked Valve Listing when the system is required to be Operable. Therefore, additional, periodic requirements to check manual valve positions are unnecessary.

4. Out of position, open, manual valves and flanges on primary containment penetrations may cause increased drywell nitrogen usage and should be detected by current trending ofliquid nitrogen inventory.

Ps NUREG SR 3.6.i.3.12 has not been used in the DAEC ITS submittal since the current license does not include this requirement. This type ofleakage is part of the overall containment leakage and no special limits apply.

P, The current licensing basis for leak rate testing of the MSIVs is contained in CTS 4.7.A.l.b. These CTS allowances are retained in ITS SR 3.6.1.3.9. The maximum leakage from any one MSIV shall not exceed 100 scth at a test pressure of 24 psig and the combined maximum pathway leakage rate for all four main steam lines shall

, not exceed 200 scfh at a test pressure of 24 psig. If a MSIV exceeds 100 scth, it will be restored to 5: 11.5 scih.

Poi At DAEC, the 18 inch purge valves are pennanently blocked to restrict opening to 30'. Therefore, NUREG SR 3.6.1.3.15 is not applicable and can be deleted.

Pn At DAEC there are only two LLS valves and Condition B ofITS 3.6.1.5 has been revised to reflect this design.

DAEC 3 Revision E l

NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS i

ADMINISTRATIVE CllANGES

[ The proposed changes involve the refonnatting, renumbering, and rewording of the TS and Baset.

These changes, since they do not involve technical changes to the Current TS (CTS), are administrative. All of the administrative changes contained in the Discussion of Changes for this chapter :'r addressed by this evaluation.

. The DAEC has evaluated these proposed CTS changes and has determined that they involve no significant hazards consideration. This determination has been made in accordance with the criteria set forth in 10 CFR 50.92, based on the following considerations:

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

These proposed changes are administrative, including refomiatting, renumbering, and rewording of the CTS and Bases. These changes do not involve technical changes to the CTS. These changes to the CTS are being made in order to be consistent with the choice of style and language in NUREG 1433. During deveh,pment of the NUREG, certain wording

preferences or language conventions were adopted. These proposed chmiges are administrative in nature and do not impact initiators of analyzed events. They also do not impact the assumed mitigation of accidents or transient events. Therefbre, these changes do not involve a significant increase in the pmbability or consequences of an accident previously evaluated.

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

These proposed changes do not involve a physical alteration of the plant (no new or ditTerent type of equipment will be installed) or change the methods goveming plant operation. The proposed changes will not impose any new or ditTerent requirements or eliminate any existing requirements. Therefore, these cnanges do not create the possibility of a new or ditTerent kind of accident from any accident previously evaluated.

DAEC 1 Revision E

~ -- . _ _ _ __ __

I l

. . l NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS ADMINISTRATIVE CliANGES (continued)

3. Does this change involve a significant reduction in a margin of safety?

The proposed changes do not involve a physical alteration of the plant (no new or difTerent ,

type of equipment will be installed) or change the methods gov; ming plant operation. The propowd changes will not impose any new or different requirements or climinate am existing requirements. As a result, the proposed changes will have no impact os g afety analysis assumptions. Therefore, these changes do not involve a significant redu e . .a a margin of safety.

DAEC 2 Revision E

.. - - - . . . - - - - . - - - - ~. .-

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS TECifNICAL CilANGES - MORE RESTRICTIVE The proposed changes incorporate more restrictive changes into the CTS by either making current requirements more stringent or adding new requirements which currently do not exist. All of the more restrictive changes contained in the Discussion of Changes for this chapter are addressed by this evaluation.

The DAEC has evaluated the proposed CTS changes and has determined that they involve no significant hazards consideration. This determination has been made in accordance with the criteria set forth in 10 Cl:R 50.92, based on the following considerations:

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

These proposed changes provide mere restrictive requirements than previously existed in the CTS. The more restrictive requirements will not result in operation that will increase the probability ofinitiating an analyzed event. The new requirements either do not change, or in some instances may decrease, the probability or consequences of an analyzed event.

These changes will not invalidate assumptions relative to mitigation of an accident or transient event. These changes have been reviewed to ensure that no previous accident evaluations have been adversely affected. Therefore, these changes will not involve a significant increase in the probability or consequences of an accident previously evaluated.

DAEC 3 Revision E

NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECilNICAL CilANGES - MORE RESTRICTIVE (continued)

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

The more restrictive requirements imposed by these changes will not alter the plant configuration (no new or difTerent type of equipment will be installed), Any resulting changes in the methods goveming plant operation will be consistent with assumptions made in the safety analyses. Therefore, these changes will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The more restrictive requirements imposed by these changes either increase or do not afTect the margin of safety. These changes do not impact any safety analysis assumptions.

Therefore, these changes will not involve a significant reduction in a margin of safety.

J 1

DAEC 4 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TEC1INICAl, CllANGES - RELOCATIONS The proposed changes relocate requirements from the CTS to licensee controlled documents.

These changes are labeled " Technical Changes - Relocations." All of the relocation changes contained in the Discussion of Changes for this chapter are addressed by this evaluation.

The DAEC has evaluated the proposed CTS changes and has determined that they involve no significant hazards consideration. This determination has been made in accordance with the criteria set forth in 10 CFR 50.92, based on the following considerations:

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

The proposed changes relocate requirements from the CTS to licensee controlled documents. The proposed changes are consistent with NUREG-1433, which was approved by the NRC Staff. The NRC staff concluded that the change controls for proposed relocated details and requirements provide an acceptable level of regulatory authority. Any future changes to the licensee controlled documents containing relocated requirements will be evaluated in accorduce with the DAEC 10 CFR 50.59 program. Additionally, changes to the TS Bases are subject to the requirements of the TS Bases Control Program in the Administrative Controls Section of the ITS. Since any changes to licensee controlled documents will be evaluated in accordance with the DAEC 10 CFR 50.59 program, no increase in the probability or consequences of an accident previously evaluated will be allowed without pric. NRC approval. Therefore, these changes will not involve a significant increase ia the probability or consequences of an accident previously evaluated.

DAEC 5 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS TECIINICAI, CilANGES - REl.OCATIONS (continued)

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

The proposed changes relocate requirements from the CTS to licensee controlled documents. These changes will not alter the plant configuration (no new or different type of equipment will be installed) or change the methods governing plant operation. These changes will not impose different requirements or eliminate existing requirements.

Adequate control of these requirements will still be maintained. These changes will not alter assumptions made in the safety analysis or licensing basis. Therefore, these changes will not create the possibility of a new or different kind of accident fam any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed changes relocate requirements from the CTS to licensee controlled documents. These changes will not reduce a margin of safety since they have no impact on any safety analysis assumptions. These changes will not impose different requirements or climinate existing requirements. Since any future chances to the licensee controlled documents will be evaluated in accordance with the DaEC 10 CFR 50.59 program, no reduction in a margin of safety will be allowed without prior NRC approval. Therefore, these changes will not involve a significant reduction in a margin of safety.

4 DAEC 6 Revision E

NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECilNICAl, CilANGES - IISS RESTRICTIVE (1,cy.2 Labeled Comment / Discussion for ITS 3.6.1.1)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

Based on the evaluations conducted to satisfy Generic Letter 91-04 requirements,1) no time dependent failures were found,2) other more frequent testing would have found many of the failures that were discovered during cycle tests, or 3) there was no loss of function (i.e. no common cause failures were found). Thus, the availability of systems required to mitigate the consequences of an accident are not adversely impacted. The proposed change does not increase the probability of any accident, as the overall system reliability is not reduced, nor are the performance of these surveillance tests considered to be accident initiators. Therefore, the probability or consequences of previously evaluated accidents are not significantly increased.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or change the methods governing normal plant operation. Foi each of these tests, the nature of any failures, corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

DAEC 7 Revision E

't

NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6--CO. Al'NMENT SYSTEMS IEC11NICAL CIl ANGES - 1.ESS RESTRICTIVE (1,CY 2 Labeled Comment / Discussion for ITS 3.6.1.1) (continued)

3. Does this change involve a significant reduction in a margin of safety?

Although the proposed enange will result in an increase in the interval between surveillance tests, the impact on system availability is considered small based upon other more frequent testing, the availability of redundant systems or equipment, and the fact that there is no evidence of any existing equipment failures that would impact the availability of the system. Therefore, the proposed change will not significantly impact the availability or reliability of the plants systems or their ability to respond to plant transients and accidents. Therefore, the proposed change will not involve a significant reduction in a margin of safety.

DAEC 8 Revision E l.

l .- .

NO SiONIFICANT liAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS

)):CHNICAL CHANGES - LESSJJESTRICTIVE (continued)

(L ii sheled Comment / Discussion for ITS 3.6.1.2)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evahiation is provided for the three categories of the significant hazards consideration standards:

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

This change does not result in any hardware changes. The primary containment air lock interlock is not assumed to be an initiator of any analyzed event. The role of the interlock is to ensure the primary containment boundary is maintained, thereby limiting consequences.

Failure of the interlock during testing could result in a loss of primary containment Operability. Since the proposed change reduces the frequency of challenge to the interlock, the probability of a loss of primary containment Operability during the Modes when primary containment is required (LCO 3.6.1.1) is reduced. The Operability of the interlock has no effect on the consequences of an accident previously evaluated because no credit is taken for it in the mitigation of an accident. Therefore, this change does not involve a significant increase in the probability or consequences of a previously evaluated accident.

DAEC 9 Pevision E l

. - ~ . _ - -. . . . - . . , . .

!. I' l,

l

-j E -

NO SIGNIFICANT HAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECHNICAL CH ANGES - LESS RESTRICTIVE ,

(L iLabeled Comment / Discussion for ITS 3.6.1.2) ( continued)

~

2.- Does the change create the possibility of a new or different kind of accident from any ,

- accident previously_ evaluated?

The proposed change does not necessitate a physical alteration of the plant (no new or ,

. different type of equipment will be installed) or changes in parameters governing normal plant operation. The proposed change will still ensure the interlocks remain Operable when 7

required. Tims, this change does not create the possibility of a new or different kind of . j

~

accident from any accident previously evaluated.

, ( 3. : Does this change involve a significant reduction in a margin of safety?

This change reduces the challenges to primary containment Operability during Modes when :

primary containment is required to Operable. Funher, proving the Opernbility of the air

lock interlock at more frequent intervals serves no useful purpose since no enhancement to

, safety is gained by unnecessarily testing the interlock. From the standpoint of primary containment Operability and a reduction of unnecessary testing, the proposed change

. represents an enhancement to safety. As such, no significant reduction in a margin of safety is involved with this change.

i.

I E

A J-

)

i 1

f.

i i:

i DAEC .. 10 Revision E

- - __ . _ _ . . . ___- ._._____,u____.._.. _ n ._ _ _ _ - . _ . . _ . _ , , _ ._ _ _ _ . . . . .

NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECliNICAL CllANGES -i.ESS RESTRICTIVE (continued)

(L iLabeled Comment / Discussion for ITS 3.6.1.3)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

This change relaxes the Completionri te from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 8 hoors to isolate the affected penetration (s) if one MSIV in one or more penetrations is inoperable. The proposed change does not increase the probablity of an accident. The time allowed to isolate the penetration by use of a de-activated automatic valve, blind flange, etc. is not assumed to be an initiator of any analyzed event. Therefore, this change does not involve a significant increase in the probability of an accident previously evaluated. The MSIVs willisolate, along with the other PCIVs to control leakage from the primary containment dunng accidents. Allowing 4 additional hours to isolate the MSIVs will not significantly increase the consequences of an accident. The chances of an event occurring are the same in the second 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period as they are in the first 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period. Also, the consequences will be the same for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months as for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The additional 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , however, will allow additional time to repair the inoperable MSIV and possibly avoid a shutdown. Shutting down the plant is a transient which puts thermal stress on components and could increase the chances of challenging safety systems. This change will not alter assumptions relative to the mitigation of an accident or transient event. Therefore, this change will not involve a significant increase in the consequences of an accident previously evaluated.

DAEC 11 Revision E

NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS IECIINICAL CIIANGES - LESS PESTRICTIVE (L iLabeled Comment / Discussion for!TS 3.6.1.3) (continued)

2. Does the change create the possibi!ity of a new or different kind of accident from any accident previously evaluated?

This change relaxes the Completion Tinie from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for isolating the afTected penetration (s) if one MSIV in one or more penetrations is inoperable. The additional 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months that the MSIVs are not isolated will not create the possibility of an accident. The chance of an event occurring which would require the MSIVs to be isolated and a failure occurring which would prevent the Operable MSIV to close is remote. Also, this change will not physically alter the plant (no new or different type of equipment will be installed).

The changes in methods goveming normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or difTerent kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

This change relaxes the Completion Time from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for isolating the afTected penetration (s) if one MSIV in one or more penetrations is inoperable. The margin of safety is not significantly reduced because the chances of an event occurring are the same in the second 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period as they are in the first 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period. The additional 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allows more time to repair the inoperable MSIV and avoid the potential for a phmt shutdown.

Isolating the MSIV penetrations will require a reduction in power and has the potential for tripping the plant. A reduction in pcwcr or a plant trip is considered a transient due to the thermal etTects it has on plant equipment. During the additional time allowed, a limiting event would still be assumed to be within the bounds of the safety analysis, assuming no single active failure. Therefore, this change does not involve a significant reduction in a margin of safety.

DAEC 12 Revision E i

h l

l NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS IECIINICAI, CilANGES - 1,ESS RESTRICTIVE (continued)

(L 2Labeled Comment / Discussion for ITS 3.6.1.3)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categoriu of the significant hazards consideration standards:

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

This change would decrease the surveillance frequency of the purge isolation valve leakage test so that it is required to be performed every 184 days instead of every 3 months.

The proposed change does no' affect the purge valve design or timction. A failure of a purge valve is not identified as an initiator of any event. Therefore, this proposed change does not involve an increase in the probability of an accident previously evaluated. Since the change impacts only the frequency of verification and does not result in any change in the response of the equipment to an accident, the change does not increase the consequences of any previously analyzed accident.

2. Does the change create the possibility of a new or difTerent kind of accident from any accident previoucly evaluated?

This change does not result in any changes to the equipment design or capabilities, or to the operation of the plant. Since the change impacts only the frequency of verification and does not result in any change in the response of the equipment to an accident, the change does not create the possibility of a new or difTerent kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

This change impacts only the frequency of purge valve leak rate testing. DAEC experience has shown (from current successful 3 month testing) that a 184 day leak rate test is acceptable. herefore, the change does not involve a significant reduction in a margin of safety.

DAEC 13 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECilNICAL CilANGES - LESS RESTFICTIVE (continued)

(LyLabeled Comment / Discussion for ITS 3.6.1.3)

-_ DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has.been performed in accordance with the criteria set fonh in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration stancards:

- 1. - -- Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

- This change would allow additional time to isolate a primary containment penetration if both isolation devices are inoperable. Primary containment isolation is not considered as an initiator of any pre,iously analyzed accident. Therefore, this change does not significantly increase the probability of such accidents. The proposed change allows additional temporary operation with less than the required isolation capability. However, the consequences of an event that may occur during the extended outage time would not be any different than during the currently allowed outage time for other loss of containment integrity situations. Therefore, this change does not significantly increase the consequences of any previously analyzed accident.

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

This change does not result in any changes to the equipment design or capabilities or to the operation of the plant. Further, since the change impacts only the required action completion time for the system and does not result in any change in the response of the equipment to an accident, the change does not create the possibility of a new or different kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in margin of safety?

This chance impacts only the required action completion time for inoperable valves that provide containment isolation. The methodology and limits of the accident analysis are not-affected, nor is the containment response. Therefore, this change does not involve a significant reduction in a margin of safety.

DAEC 14 Revision E

NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TEClINICAL ClIANGES - 1.ESS RESTRICTIVE (continued)

(L. Labeled Comment / Discussion for ITS 3.6.1.3) -

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This detennination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

'l. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

This change would allow additional time to isolate a primary containment penetration if one or more penetration flow paths have one or more containment purge valves not within leakage limits. Primary containment isolation is not considered as an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probability of such accidents. The proposed change allows additional temporary operation with less than the required isolation capability. However, the consequences of an event that may occur during the extended time would not be any ditTerent than during the currently allowed time for other loss of containment integrity situations. Therefore, this change does not significantly increase the consequences of any previously analyzed accident.

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

This change does not result in any changes to the equipment design or capabilities or to the operation of the plant. Further, since the change impacts only the required action completion time for the system and does not result in any change in the response of the equipment to an accident, the change does not cre 'te the possibility of a new or ditTerent kind of accident from any previously analyzed awident.

3. Does this change involve a significant reduction in a margin of safety?

This change impacts only the required action completion time for ir. operable valves that provide containment isolation. The methodology and limits of the accident analysis are not significantly affected, nor is the containment response. Therefore, the change does not involve a significant reduction in a margin of safety.

DAEC 15 Revision E

1 NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECIINICAl, CilANGES - LESS RESTRICTIVE (continued)

(L 3Labeled Comment / Discussion for ITS 3.6.1.3)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

1. Does the change involve a significant increase in the probability or consequences of an accident presiously evaluated?

This change extends the time to isolate single PCIV penetrations containing excess flow check valves (EFCVs) from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The proposed change does not increase the probability of an accident. The time allowed to isolate the penetration is not assumed to be an initiator of any analyzed event. The EFCVs isolate containment, along with the other PCIVs to control leakage from the primary system during accidents. Allowing 8 additional hours to isolate these penetrations will not significantly increase the consequences of an accident. The EFCVs on instrument and small pipe diameter penetrations limit the amount ofleakage that can occur. The chances of an event occurring are the same in the additional 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period as they are in the first 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period. This change will not alter assumptions relative to the mitigation of an accident or transient event. Also, the consequences of an event occurring during the proposed 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period are the same as those during the current 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period. Therefbre, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

5 s

DAEC 16 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS .

SECTION 3.6-CONTAINMFNT SYSTEMS TEC1INICAI, C11ANGES - LESS RESTRICTIVE (L 3Labeled Comment / Discussion for ITS 3.6.1.3) (continued)

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

This change will not create the possibility of an accident. This change extends the time to isolate penetrations containing excess flow check valves (EFCVs) in penetrations containing only one PCIV from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The additional 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months that the penetrations are not isolated will not create the possibility of an accident. The chances of an event occuning which would require containment isolation in the additional 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is remote. Also, this change will not physically alter the plant (no new or ditTerent type of .

equipment will be installed). The change in methods governing normal plant operation is consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction :n a margin of safety?

This change extends the time to isolate penetrat ons containing excess flow check valves (EFCVs) in penetrations containing only one PCIV from 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The margin of safety is not significantly reduced beccase the chances of an event occurring are the same in the additional 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months period as they are in the first 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months period. During the additional time allowed, a limiting event would still be assumed to be within the bounds of the safety analysis, assuming no single active failure. Therefore, this change does not involve a significant reduction in a margin of safety.

DAEC 17 Revision E

NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS TECIINICAL CIIANGES - LESS RESTRICTIVE (continued)

(L 6Labeled Comment / Discussion for ITS 3.6.1.3)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

The proposed change does not involve any hardware change. The reasons that the large primary containment purge and exhaust isolation valves may be opened are not assumed in the initiation of any analyzed event. Expanding the reasons these valves may be opened does not affect any assumptions of the accident analyses and still ensures the time period these vaives may be opened in Modes 1,2, and 3 is limited. In addition, these purge and exhaust valves are capable ofclosing in the erwironment following a design basis accident.

Thus, the consequences of an accident are not affected by this change. This change will not alter assumptions relative to an accident or transient event. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evale ited.

2. Does the change create the possibility of a new or difTerent kind of accilent from any accident previously evaluated?

This proposed change will not involve any physical change to plant systems, structures, or components (SSCs), or the manner in which these SSCs are maintained, modified, tested, or inspected. The change in methods governing normal plant operation is consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or ditTerent kind of accident from any accident previously evaluated.

DAEC 18 Revision E l

-. -.. - . . ~ . - - . - - - .- . . . - - . - - . - -.. -.--_ - . -. . . - . . - . . - - . -

i !

NO SIGNIFICANT llAZARDS CONSIDERATIONS ,

' SECTION 3.6-CONTAINMENT SYSTEMS b TECIINICAl, C11 ANGES - LESS RESTRICTIVE .

. _(L 6 Labeled Comment / Discussion for ITS 3.6.1.3) (cantinued) -

3. Does this change involve a significant reduction in a margin of safety?

The proposed change expands the reasons the primary containment purge and exhaust '

isolation valves may be opened in Modes 1,2, and 3. This change does not involve a reduction in the margin of safety since these valves are capable of closing in the environment following a design basis accident. This change does not affect the current

. safety analysis assumptions. As such, no question of safety exists. Therefore, this change does not involve a significant reduction in a margin of safety.

t e

i

.4 4

s 4

DAEC 19 Revision E-9-w d M *4 T' er [ye- e/ -may r - ggt---s- w -by9 ap mag- -

e-ms, e ----

. NO SIGNIFICANT HAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS TECHNICAL CHANGES - LESS RESTRICTIVE (continued)

(L, Labeled Comment / Discussion for ITS 3.6.1.3) -

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. He following evaluation is provided for the three categories of the significant hazards consideration standards:

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

The phrase " actual or," in reference to the automatic isolation signal, has been added to the system functional test surveillance test description. This does not impose a requirement to create an " actual" signal, and does not eliminate any restriction on producing an " actual" signal. While creating an " actual" signal could increase the probability of an event, existing procedures and 10 CFR 50.59 control of revisions to them, dictate the acceptability of generating this signal.' The proposed change does not affect the procedures goveming plant operations and the acceptability of creating these signals; it simply would allow such a signal to be utilized in evaluating the acceptance criteria for the system ftmetional test requirements. Therefore, the change does not involve a significant increase in the probability of an accident previously evaluated. ' Since the method ofinitiation will not affect the acceptance criteria of the system functional test, the change docs not involve n significant increase in the consequences of an accident previously evaluated.

DAEC. 20 Revision E

- ~ , - - . _ . .

C NO SIGNIFICANT liAZARDS CONSIDERATIONS l SECTION 3.6--CONTAINMENT SYSTEMS TECIINICAL CIIANGES - LESS RESTRICTIVE (L 7Labeled Comment / Discussion for ITS 3.6.1.3) (continued)

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

The proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant. Therefore, it does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

, Use of an actual signal instead of the existing requirement, which limits use to a test signal, .

will not alTect the performance or acceptance criteria of the Surveillance. Operability is adequately demonstrated in either case since the system itselfcannot discriminate between

" actual" or " test" signals. Therefore, the change does not involve a significant reduction in a margin of safety.

4 I

l DAEC 21 Revision E 4

]

l

1 NO SIGNIFICANT l{AZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS -:

TEC11NICAL CilANGES - 1.ESS RESTRICTIVE (continued) - -l (L, labeled Comment / Discussion for ITS 3.6.1.3)

Per our Response to the Staffs RAI on this Note (Ref. NG-97-1597) and our meeting with the StafT ,

. on September 9,1997, this change has been withdrawn. {3.6.1.3 8} ,

4 F

i DAEC 22 Revision E

l NO SIGNIFICAN1 IIA 7ARDS CONSIDERATIONS -

SECTION 3.6-CONTAINMENT SYSTEMS TECliNICAL CliANGES - LESS RESTRICTIVE (continued)

(Ley.2 Labeled Comment / Discussion for ITS 3.6.1.3)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This detennination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hantrds consideration standards:

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

Ilased on the evaluations conducted to satisfy Generic Letter 91-04 requirements,1) no time dependent failures were found,2) other more frequent testing would have found many of the failures that were discovered during cycle tests, or 3) there was no loss of function (i.e. no common cause failures were found). Thus, the availability of systems required to mitigate the consequences of an accident are not adversely impacted. The proposed change does not increase the probability of any accident, as the overall system reliability is not reduced, nor are the perfomtance of these surveillance tests considered to be accident initiators. Therefore, the probabi! 2y or consequences of previously evaluated accidents are not significantly increased.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or change the methods governing normal plant operation. For each of these tests, the nature of any failures, corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

DAEC 23 Revision E

l

' NO SIGNIFICANT 11AZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECilhlCA15 CilANGES - LESS RESTRICTIVE ~

- (Lcy.2 Labeled Comment / Discussion for ITS 3.6.1.3) (continued)

3. Does tids change involve a significant reduction in a margin of safety?

Although the proposed change will result in an increase in the interval between

'i surveillance tests, the impact on system availability is considered small based upon other

~

more frequent testing, the avail ability of redundant systems or equipment, and the fact that there is no evidence of any existing equipment failures that would impact the availability of the system. Therefore, the proposed change will not significantly impact the availability or reliability of the plants systems or their ability to respond to plant transients and accidents. Therefore, the proposed change will not involve a significant ,

reduction in a margin of safety.

i M

e l

\

s I

)

DAEC- y Revision E, 4

t c-r-p , gr--, , , .- . -< . . . ,

l

~ NO SIGNIFICANT HAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS JT:CHNICAL CilANGES - LESS RESTRICTIVE (continued)

(Lcy.2 Labeled Comment / Discussion for ITS 3.6.1.5)

DAEC has evaluated this proposed CTS change and has determined that it involve', no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

Based on the evaluations conducted to satisfy Generic Letter 91-04 requirements,1) no time dependent failures were found,2) other more frequent testing would have found many of the failures that were discovered during cycle tests, or 3) there was no loss of function (i.e. no common cause failures were found). Thus, the availability of systems required to mitigate the consequences of an accident are not adversely impacted. The proposed change does not increase the probability of any accident, as the overall system reliability is not reduced, nor are the performance of these surveillance tests considered to be accident initiators. Therefore, the probability or consequences of previously evaluated accidents are not significantly increased.

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

The proposed change does not involve a physical alteration of the plant (no new or ditTerent type of equipment will be installed) or change the methods governing normal plant operation. For each of these tests, the nature of any failures, corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

DAEC 25 Revision E t

NO SIGNIFICANT HAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS TECIINICAL CHANGES - LESS RESTRICTIVE (Lcy.2 Labeled Comment / Discussion for ITS 3.6.1.5) (continued)

3. Does this change involve a significant reduction in a margin of safety?

Although the proposed change will result in an increase in the interval between surveillance tests, the impact on system availability is considered small based upon other more frequent testing, the availability of redundant systems or equipment, and the fact that there is no evidence of any existing equipment failures that would impact the-availability of the system. Therefore, the proposed change will not significantly impact the availability or reliability of the plants systems or their ability to respond to plant transients and accidents. Therefore, the proposed change will not involve a significant reduction in a margin of safety.

l l

l DAEC 26 Revision E

i NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECilNICAL CIIANGES - LESS RESTRICTIVE (continued)

(L iLabeled Comment / Discussion for ITS 3.6.1.6)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three caiegories of the significant hazards consideration standards:

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

The proposed change involves the number of reactor building-to-suppression chamber vacuum breaker valves that may be inoperable and the associated Completion Times before a reactor shutdown is required. The proposed change will make a distinction between loss of function (containment integrity and venting capability) which still requires initiating action within one hour and loss of redundancy for a function which must be recovered within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The existing specification fails to make this distinction between loss of function and loss of redundancy. The probability of an accident is not increased because these vacuum breaker valves are not considered the initiators of any accidents previously evaluated. The consequences of an accident will not be increased because the proposed change will provide assurance that both the containment integrity and venting capability functions are available or restored within one hour. The proposed change could allow continued operation for up to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months without redundant capability for these functions; however, the 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months Completion Time takes into account the redundant capability afforded by the remaining vacuum breaker, the fact that the Operable vacuum breaker valve is closed, and the low probability of an event that would require the vacuum breaker valves to be Operable during this period. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

DAEC 27 Revision E

NO SIGNIFICANT HAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECIINICAl, CHANGES - i .ESS RESTRICTIVE (L iLabeled Comment / Discussion for ITS 3.6.1.6) (continued)

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

This proposed change will not involve any physical changes to plant systems, structures, or components (SSCs), or the manner in which these SSCs are operated, maintained, modified, tested, or inspected. Therefore, this change will not create the possibility of a new or difTerent kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a .uargin of safety?

The proposed change involves the number of reactor building-to-suppression chamber vacuum breaker valves that may be inoperable and the associated Completion Time before a reactor shutdown is required. The proposed change will make a distinction between loss of function (containment integrity and venting capability) which still requires initiating action within one hour and loss of redundancy for a function which must be recovered within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months . The existing specification fails to make this distinction between loss of function ar.d loss of redundancy. The proposed change will provide assurance that both the containment integrity and venting capability functions are available or restored within one

, hour. The change does not affect the current analysis assumptions. Therefore, this change does not involve a significant reduction in a margin of safety.

DAEC 28 Revision E

__- ~. . _ _ _ - . _ _ . . . . _ - _ _ _ _ _ _ _. - ._. . _ _ _ . . _ _ _ _ __

NO SIGNIFICANT IIAZAllDS CONSIDEllATIONS SECTION 3.6 -CONTAINhiENT SYSTEhtS IECllNICAl, CllANGES I.ESS I(ESTRICTIVE (continued) l (L 21.abeled Conunent/ Discussion Ibr ITS 3.6.1.6)

DAEC has evaluated this propsed CTS change and has determined that it involves no significant hamrds consideratic'n. 'ihis determination has been perfbrmed in accordance with the criteria set fbrth in 10 CFR 50.92. The follow;ng evaluation is provided for the three categories of the significant hanud> considnation standards:

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

1 Thf.s change would decrease the surveillance frequency of the vacuum breaker valve position verification so that it is required to be performed every 14 days instead of every 7 days. 'the p:oposed chage does not afTect the vacmmi breaker valve design or fimetion .

A failure o f .4 vacuum breaker valve is noi identified as an initiator of any event. Therefore, this proposed change does not involve an increase in the probability of an accident previously evaluated. Since the change impacts only the frequency of verification and does not result in any change in the response of the equipment to an accident, the change does not increase the consequences of any previously analyzed accident.

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

This change does not result in any changes to the equipment design or capabilities, or to the operation le plrnt. Since the change impacts only the fiequency of verification and does not result . y change in the response of the equipment to an accident, the change does not create t. ossibility of a new or dilTerent kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

This change impacts only the frequency of verification of the vacuum breaker valve position. DAEC experience has shown that a change to 14 days to verify that a vacuum breaker valve is closed, is not a significant change in operating practice and that the proposed test fiequency is acceptable. Therefore, the change does not involve a significant reduction in a margin of safety.

DAEC 29 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6 -CONTAINMENT SYSTEMS ,

Tl?CIINICAl, ClIANGliS -I.ESS RiiSTRICTIVE (continued)

(13 Labeled Commen'/ Discussion for ITS 3.6.1.6)

DAEC has evaluated this proposed CTS change and has detennined that it invokes no significant hazards consideration. This detennination has been performed in accordance with the criteria set Ibrth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hantrds consideration standards:

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

This change does not result in any hardware or operating procedure changes. The reactor building-to-suppression chamber vacuum breaker position indication is nc ;sumed in the initiation of any analyzed event. The requirements for % vacuum breaker posit ion indication does not need to be explicitly stated in the Technical Specifications. To peribmi the verifications and tests required for the Surveillance Requirements c." Specification 3.6.1.6, the capability to determine vacuum breaker position must be available. If the capability to detemiine vacuum breaker position is not available, these verifications and tests cannot be satisfied and the appropriate actions must be taken for inoperble vacuum breakers in accordance with the Actions of Specification 3.6.1.6. As a result, accident consequences are unaffected by this change. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The possibility of a new or different kind of accident froia any accident previously evaluated is not created because the proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant.

DAEC 30 Revision E

NO SIGNil lCANT llAZAltDS CONSIDERATIONS l SECTION 3.6-CONTAINMENT SYSTEMS TECilNICAl.CIIANGES 1.ESS RESTRICTIVE (lo 1.abeled Comment / Discussion for ITS 3.6.1,6) (continued)

3. Does this change involve a significant reduction in a margin of safety?

The proposed deletion of th: vacuum breaker position indication requirements from Technical Specifications does not impact any margin of safety. The requirements for the vacuum breaker position indication does not need to be explicitly stated in the Technical Specifications. To perform the verifications and tests required for the Surveillance ,

Requirements of Specification 3.6.1.6, the capability to detennine vacuum breaker position must be available. If the capability to determine vacuum breaker position is not available, these verifications and tests cannot be satisfied and the appropriate actions must be taken for inoperable vacuum breakers in accordance with the Actions of Specification 3.6.1.6.

Therefore, this change does not involve a significant reduction in a margin of safety.

~DAEC 31 Revision E

.i NO SIGNIFICANT IIAl.ARDS CONSIDERATIONS !

SEC110N 3.6--CONTAINMENT SYSTEMS 1

[

IEClINICAL CllANGES - 1.ESS RESTRICTIVE (continued)

(L 4Labeled Comment / Discussion for ITS 3.6.1.6)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant ;

hazards consideration. This determination has been performed in accoidance with the criteria set :

lbrth in 10 CFR 50.92. The fcDowing evcluation is provided fbr the three categories of the significant hazards considera! on standards:

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? ,

DAEC currently performs this surveillance every 3 months in accordance with the Inservice Testing Program. This change could potentially decrease the surveillance frequency of the vacuum breaker valve functional test verification so that it is allowed to be perfonned up to every 6 months (allowable by IST Program) instead of every 3 months (as currently perfbrmed).1he proposed change does not afTect the vacuum breaker valve design or ibnetion. A failure of a vacuum breaker valve is not identified as an initiator of any event.

Therelbre, this proposed change does not involve an increase in the probabilite cf an 4 accident previously evaluated. Since the change impacts only the frequency of w ilication and does not n:sult in any change in the response of the equipment to an accident, the change does not increase the consequences of any previously analyzed accident.

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

This change does not result in any changes to the equipment design or capabilities, or to the ,

operation of the phmt. Since the change impacts only the frequency of verification and does not result in any change in the response of the equipment to an accident, the change does not create the possibility of a new or different kind of accident from any previously analyzed accident.

3. Does tnis change involve a significant reduction in a margin of safety.

This change impacts only the frequency of the functional test verification of the vacuum breaker valve. DAEC e.<perience has shown (from current successful testing on a quarterly basis) that a 6 month functional test verification is acceptable. Therefbre, the change does not involve a significant reduction in the margin of safety.

DAEC 32 Revision E

. _ _ _ _ _ _ _ _ _ . - . - _ . _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ . _. m l

l NO SIGNIFICANT llAZARDS CONSIDERATIONS ,

SECTION 3.6-CONTAINMENT SYSTEMS i

TECIINICAl, CilANGES 1 ESS RESTRICTIVE (continued)

(l.ic.2 Labeled Comment / Discussion for ITS 3.6.1.6)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. 'lhis detennination has been perfonned in accordance with the criteria set ;

forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the signifkant hazards consideration standards:

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

llased on the evaluations conducted to satisfy Generic Letter 91-04 requirements,1) no '

time dependent failures were found,2) other more frequent testing would have found many of the failures that were discovered during cycle tests, or 3) there was no loss of function (i.e. no common cause failures were found). Thus, the availability of systems required to mitigate the consequences of an accident are not adversely impacted. The proposed change does not increase the probability of any accident, as the overall system reliability is not reduced, nor are the perfbrmance of these surveillance tests considered to be accident initiators. Therefore, the probability or consequences of previously evaluated accidents are not significantly increased.

2. Does the change create the possibility of a new or difTerent kind of accident from any

, accident previously evaluated?

The proposed change does not involve a physical aheration of the plant (no new or different type of equipment will be installed) or change the methods governing nonnal plant operation. For each of these tests, the nature of any failures, corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptabfe conditions for interval extension. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

DAEC 33 Revision E 1

- , - - - - - - . , - . - , - . . c - - - - . ,n, , .-,-,,n . -,, -- . . , , _ , . - - - ----,,,n . . , , - -

i NO SIGNil lCANT llAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECilNICAI, CilANGES - 1.ESS RESTRICTIVE (L ic.2 Labeled Comment / Discussion for ITS 3.6.1.6) (continued) i

3. Does this change involve a significant reduction in a margin of safety?

Although the proposed change will result in an increase in the interval between

surveillance tests, the impact on system availability is considered small based upon ottier more frequent testing, the availability of redundant systems or equipment, and the fact that there is no evidence of any existing equipment failures that would impact the availability of the system. Therefore, the proposed change will not significantly impact the availability or reliability of the plants systecs or their ability to respond to plant transients and accidents. Therefore, the proposed change will not involve a significant t reduction in a margin of safety.

DAEC 34 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEh1S TEC1INICAL Cll ANGES - 1.ESS RESTRICTIVE (continued)

(1%v 1 abeled Comment / Discussion for ITS 3.6.1.6)

DAEC has evaluated this proposed CTS change and has detennined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

The proposed change will not result in any hardware changes. The opening setpoint of the Reactor Iluilding-to Suppression Chamber Vacuum lireakers is not assumed to be an initiator of any analyzed event. Existing operating margin between plant conditions and actual phmt setpoints is not significantly reduced due to this change. As a result, the proposed change will not result in unnecessary plant transients. The role of the Reactor fluilding-to Suppression Chamber Vacuum Ilreakers is in mitigating and thereby limiting the consequences of accidents. The Allowable Values (and corresponding Trip Setpoints) have been developed to ensure that the affected components remain capaHe of mitigating design basis events as described in the safety analyses and that the results and consequences described in the safety analyses remain bounding. Additionally, the proposed change does not alter the plant's ability to detect and mitigate events. Therefore, this change does not involve a significant increase in the probability or consequences of an accident previously evaluated.

2. Does the change create the possibility of a new or different kind of accident from an; accident previously evaluated?

The proposed change does not create the possibility of a new or ditTerent kind of accident from any accident previously evaluated. This is based on the fact that the method and manner of plant operation is unchanged. The use of the proposed Allowable Values (and corresponding Trip Setpoints) does not impact safe operation of the DAEC in that the safety analysis limits will be satisfied. The proposed Allowable Values (and corresponding Trip Setpoints) involve no system additions or physical modifications to systems in the plant.

These Allowable Values (and corresponding Trip Setpoints) were developed using a methodology to ensure the affected instrumentation remains capable of mitigating accidents and transients. plant equipment will not be operated in a manner different from previous operation, except that setpoints will be changed. Since operational methods remain DAEC 35 Revision E I

- - . .___-- . - - - - _ - - - _ - . - -- .- -. -- - -~ ... -.

I NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6--CONTAINhtENT SYSTEMS .

TECllNICAl, Cll ANGES - I.ESS RESTRICTIVE (14v I abeled Comment / Discussion for ITS 3.6.1.6) (continued:

2. (continued) unchanged and the operating parameters have been evaluated to maintain the plant within existing design basis criteria, no difreient type of failure or accident is created.

3. Does this change involve a significant reduction in a margin of safety.

The proposed Allowable Values (to be included in the ITS) ano the CTS Trip Setpoints (to be included in plant procedures) have been established by the DAEC Instrument Setpoint Methodology which is based on the General Electric (GE) Instrument Setpoint

-Methodology. The NRC has reviewed and approved the use of the GE Instrument Setpoint Methodology. The setpoint evaluation used the uncertainties associated with the DAEC instrumentation and actual DAEC physical data and operating practices to ensure the validity of the resulting Allowable Values and Trip Setpoints. The methodologies used are based on combining the uncertainties of the associated channels and take into account calibration accuracies. The use of these methodologies for establishing Allowable Values and Trip Setpoints ensures design or safety analysis limits are not exceeded in the event of transients or accidents. As such, this proposed change does not involve a significant reduction in a margin of safety.

DAEC 36 Revision E

--r--+--m--y.-T. my,m- -y c -- y-y. .pp ,7,~7,. -y ,- , y,.- - _. ,, , py,.._ , .,,.m .

ew,_,..m.7 9. ..g,mw y, _

~ _ - _ - - - - - .. -_ - . . .-

NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS i TECilNICAL CllANGES - 1 ESS RESTRICTIVE (continued)

(L iLabeled Comment / Discussion for ITS 3.6.1.7)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. His determination has been performed in accordance with the criteria set i forth in 10 CFR 50.92. The following evaluation is provided for the three categorie= of the significant hazards consideration standards:

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

This change would decrease the surveillance frequency of the vacuum breaker position verification so that it is required to be performed every 14 days instead of every 7 days. The proposed change does not affect the vacuum breaker design or function . A failure of a vacuum breaker is not identified as an initiator of any event. Therefore, this proposed ,

change does not involve an increase in the probability of an accident previously evaluated.

Since the chance impacts only the frequency of verification and does not result in any change in the response of the equipment to an accident, the change does not increase the consequences of any previously analyzed accident.

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

This change does not result in any changes to the equipment design or capabilities, or to the operation of the plant. Since the change impacts only the frequency of verification and does not result in any change in the response of the equipment to an accident, the change does not create the possibility of a new or ditrerent kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

This change impacts only the frequency of verification of the vacuum breaker position.

DAEC experience has shown that a change to 14 days to verify that a vacuum breaker is closed, is not a significant change in operating practice and that the proposed test frequency is acceptable. Therefore, the change does not involve a significant reduction in a margin of safety.

DAEC 37 Revision E 3

,_ - - , .-__.r,. . . _ _ . _ -- . , , , . _ . . , . - ,,r,... . . . _ _ , --

l l

NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECIINICAL Cil ANGES - I.ESS RESTRICTIVE (continued)

(L 2Labeled Comment / Discussion for ITS 3.6.1.7)

I DAEC has evaluated this proposed CTS change and has detennined that it involves no significant hazards consideration. This determination has been perfbnned in accordance with the criteria set fonh in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

This change does not result in any hardware or operating procedure changes. The l suppression chamber-to-drywell vacuum breakers are not assumed in the initiation of any analyzed event. The requirements for the vacuum breaker visual inspection do not need to be explicitly stated in the Technical Specifications. The performance of the verifications i and tests required for the Surveillance Requirements of Specification 3.6.1.7 and SR ;

3.6.1.1.2 ensures the Operability of the vacuum breakers. As a result, accident consequences are enafTected by tnis change. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The possibility of a new or ditTerent kind of accident from any accident previously evaluated is not created because the proposed change does not introduce a new mode of phmt operation and does not involve physical modification to the plant.

3. Does this change involve a significant reduction in a margin of safety?

The proposed deletion of the vacuum breaker visual inspection n quirements from Technic.d Specifications does not impact any margin of safety. The requirements for the vacuum breaker visual inspection do not need to be explicitly stated in the Technical Specifications. The performance of the verifications and tests required for the Surveillance Requirements of Specification 3.6.1.7 and SR 3.6.1.1.2 ensures the Operability of the vacuum breakers. As a result, the Operability of the vacuum breakers will be maintained to satisfy the associated SRs of Specification 3.6.1.7 without the need for explicit visual inspection requirements in the Technical Specifications. Therefbre, this change does not involve a significant reduction in a margin of safety.

DAEC 38 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS Tl!CilNICAL CilANGES -liSS RESTRICTIVE (continued)

(L 3Labeled Comment / Discussion for ITS 3.6.1.7)

DAEC has evaluated this proposed CTS change and has determined that it invoh es no significant hazards consideration. This determination has been performed in accordance with the criteria set fbsth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

1. Does the change involve a significant increase in the probability or consequences of an I accident previously evaluated?

This change does not result in any hardware or operating procedure changes. The suppresnion chamber to drywell vacuum breaker position indication instrumentation is not assumed in the initiation of any analyzed event. The requirements for the vacuum breaker position indication instrumentation do not need to be explicitly stated in the Technical Specifications. To perform the verifications and tests required for the Surveillance Requirements ofITS 3.6.1.7, the capability to determine vacuum breaker position must be available, if the capability to determine vacuum breaker position is not available, these verifications and tests cannot be satisfied and the appropriate actions must be taken for inoperable vacuum breakers in accordance with the Actions ofITS 3.6.1.7. As a result, accident consequences are unaffected by this change, Therefore, this change will not involve a significant increase in the probability or consequences of m .cident previously evaluated.

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

lhe possibility of a new or different kind of accident from any accident previously evaluated is not created because the proposed change does not introduce a new mode of plant operation and does not involve physical modification to the plant.

DAEC 39 Revision E

NO SIGNIFICANT llA7ARDS CONSIDERARONS SECTION 3.6-CONTAINMENT SYSTEMS TEC1INICAL Cll ANGES - 1.ESS RESTRICTIVE (L 3Labeled Comment / Discussion for ITS 3.6.1.7) (continued)

3. Does this change involve a significant reduction in a margin of safety?

The proposed deletion of the vacuum breaker position indicaton instrumentation requirements from Technical Specifications does not impact any margin of safety, The requirements for the vacuum breaker position indication instrumentation do not need to be explicitly stated in the Technical Specifications. To perform the verifications and tests required for the Surveillance Requirements ofITS 3.6.1.7, the capability to detemiine vacuum breaker position must be available. If the capability to detennine vacuum breaker position is not available, these verifications and tests cannot be satisfied and the appropriate actions must be taken for inoperable vacuum breakers in accordance with the Actions of ITS 3.6.1.7. As a result, the capability to determine vacuum breaker position will be maintained to satisfy the associated SRs ofITS 3.6.1.7 without the need for explicit instmmentation requirements in the Technical Specifications. Therefore, this change does not involve a significant reduction in a margin of safety.

DAEC 40 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS TECilNICAI, CilANGES -i.ESS RESTRICTIVE (continued)

(L iLabeled CommerNDiscussion for ITS 3.6.2.1)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

The proposed change involves elimination of a requirement to perform an external visual inspection of the suppression chamber whenever there is indication of relief valve operation with the local suppression pool temperature reaching 200 F or greater. The probability of an accident is not increased because performance of a visual inspection following a relief valve operation is not considered as an initiator of any accidents previously evaluated. The consequences of an accident will not be i:. creased because the basis ibr deleting this surveillance is that testing has demonstrated that there are no undue loads on the suppression pool or its components at elevated temperatures and pressures when SRVs discharge through " quenchers" (spargers). This testing is discussed in NEDO-30832.

" Elimination of Limit on BWR Suppression Pool Temperature for SRV Discharge with Quenchers," dated December 1984. Each of the DAEC relief valve discharge lines tenninates in a T-quencher (sparger). Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

DAEC 41 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECHNICAL, Cll ANGES - LESS RESTRICTIVE (L iLabeled Comment / Discussion for ITS 3.6.2.1) (continued)

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

This proposed change will not involve any physical changes to plant systems, structures, or components (SSCs), or the manner in ,vhich these SSCs are operated, maintained, modified, tested, or inspected. Therefore, this change will not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed change involves climination of a requirement to perform an external visual inspection of the suppression chamber whenever there is indication of relief valve operation with the local suppression pool temperature reaching 200"F or greater. This change will

< not reduce the margin of safety because testing has demonstrated that there are no undue loads on the <uppression pool or its components at elevated tamperatures and pressures when SRVs discharge through " quenchers" (spargers). This testing is discussed in NEDO.

30832, " Elimination of Limit on BWR Suppression pool Temperature for SRV Discharge with Quenchers," dated December 1984. Each DAEC relief valve discharge line tenninates in a T-quencher (sparger). As a result, the change does not alrect the current analysis assumptions and adequate assurance of suppression chamber integrity will be maintained.

Therefore, this change does not involve a significant reduction in a margin of safety.

DAEC 42 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIGNS SECTION 3.6--CONTAINMENT SYSTEMS TFCIINICAL CilANGES - 1 ESS RESTRICTIVE (continued)

(ly Labeled Comment / Discussion for ITS 3.6.2.1)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This detemiination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

i

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

a This change would allow the suppression pool average temperature limit of greater than or paual to 110' F for scramming the reactor to be changed to creater than 110 F. The suppression pool average temperature limit is not considered an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probability of such accidents. The proposed change would allow continued operation at exactly 110 F.

Ilowever, the consequences of an event that may occur at slightly greater than i 10 F would not be any different than an event that occurs at equal to 110 F since the UFSAR assumes a 120 F initial suppression pool average water temperature prior to a LOCA blowdown of the RPV, which ensures average water temperature will not exceed 170 F. Therefbre, this change does not significantly increase the consequences of any previously analyzed accident.

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

This change does not result in any changes to the equipment design or capabilities, or to the Operation of the plant. Therefore, this change does not create the possibility of a new or ditTerent kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

This change would allow the suppression pool avemge temperature limit of greater than or equal to 110' F for scramming the reactor to be raised an insignificantly small amount to greater than 110' F. The UFSAR assumes a 120 F initial suppressiori pool average water temperature prior to a LOCA blowdown of the RPV, which ensures average water temperature will not exceed 170 F. Therefore, the change does not involve a significant reduction in a margin ofsafety.

DAEC 43 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS ,

TFCIINICAL CllANGES -I.ESS RFSTRICTIVE (continued) :

(14 Labeled Comment / Discussion for ITS 3.6.2.1)

DAEC has evaluated this proposed CTS change and has detennined that it involves no significant hazards consideration. This determination has been perfbnned in accordance with the criteria set fonh in 10 CFR 50.92. The following evaluation is provided fbr the three categories of the significant hazards consideration standards:

1. Does the change involve a significant increasc in the probability or consequences of an accident previously evaluated?

This change would allow the suppression pool averagc. temperature limit of greater than or caual to 120' F for depressurizing the reactor to be changed to creater than 120" F. The suppression pool average temperature limit is not considered an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probability of such accidents. The proposed change would set..... the reactor to not be manually depressurized (a severe transient on the reactor vessel) at exactly 120* F. Ilowever, the consequences of an event that may occur at slightly greater than 120' F would not be any difTerent than an event that occurs at equal to 110 F since the UFSAR assumes the initial suppression pool average water temperature prior to a LOCA blowdown of the RFV is equal to 120 F, which ensures everage water temperature will not exceed 170" F.

Therefore, this change does not significantly increase the consequences of any previously analyzed accident.

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

This change does not result in any changes to the equipment design or capabilities, or to the operation of the phmt. Therefore, this change does not create the possibility of a new or different kind of accident from any previously analyzed accident.

DAEC 44 Revision E

NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECilNICAI,CilANGES-i.ESS RESTRICTIVE (continued) l

. (lo Labeled Comment / Discussion for ITS 3.6.2.1)

3. Does this change involve a significant reduction in a margin of safety?

'this change would allow the suppression pool average temperature limit of greater than or equal to 120' F for depressurizing the reactor to be raised an insignificantly small amount to ,

greater than 120' F. The UFSAR assumes the initial suppression pool average water -

temperature prior to a LOCA blowdown of the RPV is equal to 120'F, which ensures average water temperature will not exceed 170' F. Because the initial conditions of the ,

UFSAR analysis are preserved, the change does not involve a significant reduction in a margin of safety.

DAEC 45 Revision E

,.,r,--.,r. ,n .--n . - - - ~ -n- --, -- , .-, , - ,. ., - -

. _ - - _ - _. . . = . .- - . .- -- -. __ -_.=. _.-. - _--- -- - --

NO SIGNIFICANT llA7ARDS CONSIDERATIONS .

SECTION 3.6-CONTAINMENT SYSTEMS TECl!NICAL CilANGES - LESS RESTRICTIVE (continued)

(ly Labeleo Comment / Discussion for ITS 3.6.2.2) i DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been perfonned in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

'this change will allow 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ( CTS allows I hour) to restore suppression pool level when

it is found outside limits. The suppress;on pool level is not assumed to be an initiator of any previously analyzed accident. The role of the suppression pool is in the mitigation of accident consequences. The proposed change would allow temporary operation when suppression pool level is not within limits, llowever, since the only change is in the amount of time the level is outside the limits, the consequences of an event that may occur during this time would not be any different than with the cunent recuirements. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

The proposed che.nge will not involve any physical changes to plant sy stems, s*.ructures, or components (SSCs), or the manner in which these SSCs are operated, maintained, modified, tested, or inspected.1herefore, this change will not crease the possibility of a new or different kind of accident from any accident previously evaluated. .

i DAEC 46 Revision E

.m-y**-'.- e.m-e3 yq s'> wsm .+ c- ,.m-._...%-m - , ,,eq.- , , _y w.%,-,._.w. '-a- w -rvP me'gr-*-e---e e-wm-e t-n&e-- ' - < *-m-+eN'rwvp-9 w- e

NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TEClINICAl> Cll ANGES - 1.ESS RESTRlCTIVE (L iLabeled Comment / Discussion for ITS 3.6.2.2) (continued)

3. Does this change involve a significant reduction in a margin of safety?

The change provides a Completion Time of two hours when suppression pool level is not within required limits. 'Ihe proposed time is acceptable based on the small probability of an event requiring the unavailable capabilities. The proposed time will provide a reasonable time to attempt restoration of the suppression pool water level without placing the plant in a shutdown condition. The exposure of the plant to the small probability of an event requiring the suopression pool level to be within required limits during the 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months time is ,

insignificant ano offset by the benefit of avoiding an unnecessary plant shutdown.

Therefore, this change does not involve a significant reduction in a margin of safety.

DAEC 47 Revision E

P NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS l

TEClINICAl, CllANGES -I.ESS RESTRICTIVE (continued)

(ly 1.abeled Comment / Discussion for ITS 3.6.3.1)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. 'lhe following evaluation is provided for the three categories of the significant hazards consideration standards:

1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated? ,

This change would relax the surveillance frequency of the CAD nitrogen bank volume verification to every 31 days from once per week. The proposed change does not affect the nitrogen bank design or function. A failure of the nitrogen bank is not identified as an initiator of any vent. Therefore, this proposed change does not involve an increase in the probability of n1 accident previously evaluated. Fmthermore, since the change impacts only the frequency of verification and does not result in any change in the response of the equipment to an accident, the change does not increase the consequences of any previously analyzed accident.

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

This change does not result in any changes to the equipment design or capabilities, or to the operation of the plant. Since the change impacts only the frequency of verification and does not result in any change in the response of the equipment to an accident, the change does not create the possibility of a new or ditTerent kind of accident from any prse iously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

This change impacts only the frequency of verification of the nitrogen bank volume.

The nitmgen banks in the CAD System are dedicated for use in that system. Control room indication is available for bank pressure and outside air temperature which are used to detemiine available volume. Therefore, the change does not involve a significant reduction in a margin of safety.

DAEC 48 Revision E

, ,,,r-- --,---,,-7., y-r,-,,-y- - ,, ,.s,---- , , y ,- e-- ---.---4--- y v.- r--, . ,. - =,-r--*< ,-. - - - .

NO SIGNIFICANT llAZARDS CONSIDERATIONS j SECTION 3.6--CONTAINMENT SYSTEMS ,

TECilNICAL CilANGES -I.ESS RESTRICTIVE (continued)

(L i1 abeled Comment / Discussion for ITS 3.6.3.2)

DAEC has evaluated this proposed CTS change and has detennined that it involves no significant hazards consideration. This detennination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

The proposed change allows 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months aller exceeding 15% RTP to inert the drywell instead of the current requirement of aller placing the Mode Switch in Run (approximately 5-10%

RTP). On a shutdown, the proposed change will also allow de-inerting the drywell 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months prior to reducing themial power to <l5% RTP instead of the CTS requirement of ,

prior to taking the Mode Switch out of Run. If Actions crc not met to restore Oxygen Concentration to within limit, the ITS only requires thern al power to be reduced to < 15%

RTP within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months instead of the CTS requirement to be in Startup/Ilot Standby. The oxygen concentration is not assumed to be an initiator of any previously analyzed accident.

Thercibre, the probabihty of an accident previously evaluated is not significantly increased.

The consequences of an accident are not significantly increased since the hydrogen generation rate is independent of power level, and the CAD System is still Operable and can remove combustible gas mixtures if needed.

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

The proposed change i stroduces no new mode of plant operation and it does not involve any physical modification to the plant. Normal operation of the plant does not involve any manipulation of the oxygen concentration limit. The changes in methods governing normal plant operation are consistent with the current safety analysis assumptions. Therefore, this change will not create the possibility of a new or ditTerent kind of accident from any accident previously evaluated.

DAEC 49 Revision E i

i NO SIGNIFICANT llAZARDS CONSIDERATIONS ,.

SECTION 3.6--CONTAINMENT SYSTEMS TECilNICAl, Cil ANGES - I.ESS RESTRICTIVE (L, Labeled Comment / Discussion for ITS 3.6.3.2) (continued)

3. Does this change involve a significant reduction in a margin of safety?

ne margin of safety is not significantly reduced since 1) hydrogen generation rate is independent of power level,2) the CAD System and the capability to vent through the SilOT System are available to reduce combustible gas concentrations, if needed, and 3) there is a low probability of an accident that generates hydrogen.

DAEC 50 Revision E I

NO SIGNIFICANT liAZARDS CONSIDEllATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECilNICAL CliANGES - LESS RESTRICTIVE (continued)

(L iLabeled Comment / Discussion for ITS 3.6.4.1)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This detennination has been perfonned in accordance with the criteria set forth in 10 CFR 50.02. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

'lliis change would allow up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to restore secondary containment in the event that it became inoperable. Secondary containment Operability is not considered as an initiator of a iy previously analyzed accident. This time period (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) also ensures that the probability of an accident (requiring secondasy containment Operability) occurring during periods where secondary containment is inoperable is minimal. Allowing this extended time (from I hour to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ) to potentially avoid a plant shutdown is reasonable and does not represent a significant decrease in safety. Therefore, the probability of a previously analyzed event that may occur during this extended time would not be significantly increased. Secondary containment isolation capability is still maintained by utilizing the secondary containment isolation valves, dampers and associated actuation instrumentation.

The consequences of an event that may occur during this time would not be any different than during the currently allowed time fbr other loss of secondary containment integrity situations. Therefore, this change does not significantly increase the consequences of any previously analyzed accident.

DAEC 51 Revisior' E ,

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TEClINICAl, Cil ANGES - 1.ESS RESTRICTIVE (L iLabeled CommerdDiscussion for ITS 3.6.4.1) (continued)

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

This change does not result in riy change to the equipment design or capabilities or to the operation of the plant. The change impacts only the Required Action Completi on Time for restoring secondary containment. The probability of an accident (requiring secondary containment Operability) occuning during periods where secondary containment is inoperable is minimal. Tiw change does not result in any change in the response of the equipment to an accident. The change does not create the possibility of a new or different kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

This change impacts the Required Action Completion Time for restoring secondary containment. The methodology and limits of the accident analysis are not affected, and the secondary containment response is unalTected. Since the proposed compensatory boundary essentially meets the original criteria and provides leakage characteristics similar to currently approved comly:nsatory boundaries, the change does not involve a significant reduction in a margin of safety.

DAEC 32 Revision E

- = ~ , . -___m___ _ _ _ . _ _

t NO SIGNil'ICANT llAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECilNICAL CliANGES - LESS RESTRICTjyl (continued)

(L 2Labeled Comment / Discussion for ITS 3.6Al)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hamrds consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

1. Does the change involve a significant increase in the probability or consequences of an uccident previously evaluated?

This change would remove a specific restriction to perform a surveillmee of the secondary containment prior to refueling. This test of the secondary containment is not considered as an initiator of any previously analyzed accident. Therefore, this change does not significantly increase the probaoility of any accident. The appropriate plant conditions for performance of this surveillance will continue to be controlled to assure the potential consequences of any accident are not significantly increased. This control method has been previously duermined to be acceptable as indicated in Generic Letter 91-04. Therefbre, this change does not significantly increase the consequences of any previously analyzed accident.

2. Does the change creatc the possibility of a new or difTerent kind of accident from any accident previously evaluated?

This change removes a specific restriction on the plant conditions for perfbrming a surveillance, but does not change the method of perfommnce. The appropriate plant conditions for perfonnance of this surveillance will continue to be controlled to assure the possibility for a new or different kind of accident is not created. This control method has been previously determined to be acceptable as discussed in Generic Letter 91-04.

Therefore, this change does not create the possibility of a new or difTerent kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

The margin of safety considered in determining the appropriate plant conditions for perfonning this surveillance will continue to be controlled to assure that there is no significant reduction in the margin of safety. This control method has been previously determined to be acceptable as discussed in Generic Letter 91-04. Therefore, the change does not involve a significant reduction in the margin of safety.

DAEC 53 Revision E

NO SIGNIFICAN f IIAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS TECIINICAl, CllANGES - LESS RESTRICTIVE (continued)

(Lcy.2 Labeled Comment / Discussion for ITS 3.6.4.1)

DAEC has evaluated this proposed CTS change and has detennined that it involves no significant hanrds consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

llaced on the evaluations conducted to satisfy Generic Letter 91-04 requirements,1) no time dependent failures were found,2) other more frequent testing would have found many of the failures that were discovered during cycle tests, or 3) there veas no loss of function (i.e. no common cause failures were found). Thus, the availability of systems required to mitigate the consequences of an accident are not adversely impacted. The proposed change does not increase the probability of any accident, as the os erall system reliability is not reduced, nor are the performance of these surveillance tests considered to be accident initiators. Therefore, the probability or consequences of previously evaluated accidents are not significantly increased.

2. Does the c) mge create the possibility of a new or different kind of accident from any accident pt .viously cvaluated?

The proposed change does not involve a physical alteration of the plant (no new or difTerent type of equipment will be installed) or change the methods governing normal plant operation. For each of these tests, the nature of any failures, corrective actions that were taken, system redundancy, or detectability of the failuies by other mid-cycle testing resulted in acceptable conditions for interval extension. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

DAEC 54 Revision E

~. _. ___ _ _ .. _ _.___

i NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS :

T1? CLINICAL C11ANGES - LESS RESTRICTIVE (Lcy.2 Labeled Comment / Discussion for ITS 3.6.4.1) (continued)

3. Does this change involve a significant reduction in a margin of safety?

Although the proposed change will result in an increase in the interval betwee 1 surveillance tests, the impact on system availability is considered small based upon other more frequent testing, the availability of redundant systems or equipment, and the fact that there is no evidence of any existing equipment failures that would impact the availability of the system. Therefore, the proposed change will not sigri.ficantly impact the availability or reliability of the plants systems or thei ability to respond to plant transients and accidents. Therefore, the proposed change will not involve a significant reduction in a margin of safety.

t DAEC 55 Revision E :

- . - - . =. _. .-- __ - - _ _ . . ..- .- .- . . _ . .

NO SIGNIFICANT liAZARDS CONSIDERATIONS ,

SECTION 3.6--CONTAINMENT SYSTEMS TEClINICAl, CIIANGES -I.ESS RESTRICTIVE (continued)

(La Labeled Comment / Discussion t'or ITS 3.6.4.2)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following eva!uation is provided for the three ca:egories of the significant hazards consideration standards:

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

This change would allow 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to isolate a secondary containment penetration if one or more open penetration flow paths contain two inoperable SCIV/Ds. Secondary containment isolation is not considered as an initiator of any previously analyzed accident.

Therefore, this change does not significantly increase the probabiPy of such accidents. In this condition, the CTS would basically require a plant shutdown be initiated immediately.

The consequences of an event that may occur during the extended time would not be any different than during the time currently allowed for other loss of secondary containment integrity situations. Therefore, this change does not significantly increase the consequences of any previously analyzed accident.

2. Does the change create the possibility of a new or difTerent kind of accident from any accident previously evaluate 67 This change does not result in any changes to the equipment design or capabilities or to the operation of the plant. Since the change impacts only the Required Action Completion Time for the system and does not result in any change in the response of the equipment to an accident, the change does not create the possibility of a new or ditTerent kind of accident from any previously analyzed accident.

3. Does this change involve a significant reduction in a margin of safety?

This change impacts the Required Action Completion Time for inoperable valves that provide secondary containment isolation. The methodology and limits of the accident

- analysis are not affected, and the secondary containment response is unafTected. Therefore, the change does not involve a significant reduction in a margin of safety.

DAEC 56 Revision E

r NO SIGNIFICANT 11AZARDS CONSIDERATIONS SECTION 3.6-CONTAINMLWT SYSTEMS IEClINICAl, C11ANGES - 1 ESS RESTRICTIVE (continued)

(L 2Labeled Comment / Discussion for ITS 3.6.4.2)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accorduce with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

1. Does the change involve a signi6 cant iricrease in the probabili'y or consequences of an accident previously evaluated?

The phrase " actual or," in reference to the automatic isolation signal, has been added to the system functional test surveillance test description. This does not impose a requirement to create an " actual" signal, and does not climinate any restriction on producing an " actual" signal. While creating an " actual" signal could increase the probability of an event, existing procedures and 10 CFR 50.59 control of revisions to them, dictate the acceptability of generati ng this signal. The proposed change does not affeet the procedures governing plant operations and the acceptability of creating these signals; it simply would allow such a signal to be utilized in evaluating the acceptance criteria for the system functional test requirements. Therefore, the change does not involve a significant increase in the probability of an accident previously evahiated. Since the method ofinitiation will not affect the acceptance criteria of the system functional test, the change does not involve a significant increase in the consequences of an accident previously evaluated.

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

The proposed change decs not introduce a new mode of plant operation and does not involve physical modification to the plant. Therefore, it does not create the possibility of a i new or different kind of accident from any accident pxviously evaluated.

I j,

3. Dees this change invoh e a significant reduction in a margin of safety?

Use of an actual signal instead of the existing reqairement, which limits use to a test signal, will not affect the performance or acceptance criteria of the Surveillance. Operability is adequately demonstrated in either case since the system itself cannot discriminue between

" actual" or " test" signals. Thereface, the change does not involve a significant reduction in a margin of safety.

DAEC 57 Revision E ,

l l

l

- - . ._-----__\

NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECIINICAL CilANGES - LESS RESTRICTIVE (continued)

(Ley.2 Labe:ed Comment / Discussion for ITS 3.6.4.2)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

Based on the evaluations conducted to satisfy Generic Letter 91-04 requirements,1) no time dependent failures were found,2) other more frequent testing would have found many c,f the failures that were discovered during cycle tests, or 3) there was no loss of function (i.e. no common cause failures were found). Thus, the availability of systems required to mitigate the consequences of an accident are not adversely impacted. The proposed change does not increase the probability of any accident, as the overall system reliability is not reduced, nor are the performance of these surveillance tests considered to be accident initiators. Therefore, the probability or consequences of previously evaluated accidents are not significantly increased.

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

The proposed change does not invclve a physical alteration of the plant (no new or ditTerent type of equipment will be installed) or change the methods governing nonnal plant operation. For each of these tests, the nature of any failures, corrective actions that were taken, system redunduncy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

DAEC 58 Revision E l

l l

NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECllNICAl, CilANGES - I.ESS RESTRICTIVE (1 cv.2 Labeled Comment / Discussion for ITS 3.6.4.2) (continued)

3. Does this change involve a significant reduction in a margin of safety?

Although the proposed change will result in an increase in the inte - n surveillance tests, the impact on system availability is considere< . inon other more frequent testing, the availability of redundant systems or egr , rnt xu ) fact that there is no evidence of any existing equipment failures that w o pacti availability of the system. Therefore, the proposed change will not si <2m mpact the availability or reliability of the plants systems or their ability to re> .ei w . at transients and accidents. Therefore, the proposed change will not involve o - icant reduction in a margin of safety.

t l

l I

DAEC 59 Revision E l

J e

NO SIGNIFICANT liAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECHNICAL, CIIANGES - 1.ElS RESTRICTIVE (continued)

(L Labeled Comment / Discussion for ITS 3.6.4.3)

DAEC 1,as evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

1. Does the change involve a significant increase in the probability or consequences of an accident ineviously evaluated?

The proposed change will allow placing the Operable SBGT subsystem in operation as an alternative to suspending movement ofirradiated fuel whenever SBGT subsystem Operability requirements cannot be met. The proposed change does not increase the probability of an accident because the inoperability of one SBGT subsystem and continuous operation of the redundant SBGT subsystem is not considered to be an initiator of any analyzed accident. The proposed change does not increase the consequences of an accident because, in lieu of suspending the potential for releasing radioactive material to the secondary containment, placing the Operable SBGT subsystem in operation mitigates the consequences of an accident by ensuring that the remaining subsystem is Operable, that no failures that could prevent automatic actuation have occurred, and that any other failure would be readily detected. Proper operation of only one SBGT subsystem is sufficient to mitigate the consequences of any analyzed accident. Therefore, this change will not involve a significant increase in the probability or consequences of an accident previously evaluated.

I DAEC 60 Revision E

NO SIGNIFICANT 11AZAROS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS '

TECliNICAL CIIANGES - I ESS RESTRICTIVil (L iLabeled Comment / Discussion for ITS 3.6.4.3)(contin ied)

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

The proposed change will not involve any physical changes to plant systems, structures, or components (SSCs), or the manner in which these SSCs are operated, maintained, modified, tested, or inspected. Therefore this change will not create .he possibility of a new or different kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed change will allow placing the Operable SBGT subsystem in operation as an alternative to suspending movement ofirradiated fuel whenever SBGT subsystem Operability requirements cannot be met. The proposed change does not result in a significant reduction in a margin of safety because it allows operations which have the potential for releasing radioactive material to the secondary containment to continue only if the system designed to mitigate the consequences of this release is functioning. Proper operation of only one SBGT subsystein is sufricient to mitigate the consequences of any analyzed accident. Therefore, this change does not change any of the assumptions in the accident analysis and does not involve a significant reduction in a margin of safety.

l DAEC 61 Revision E l

.1 1

NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TECIINICAL CHANGES - LESS RESTRICTIVE (continued)

(L 2Labeled Comment / Discussion for FFS 3.6.4.3)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

The proposed change will allow a Standby Gas Treatment (SBGT) subsystem to be declared inoperable and to delay entry into the associated Condition and Actions for up to I hour to allow testing of SBGT to meet the requirements of the Ventilation Filter Testing Program (VFTP). For the I hour, the function is normally maintained by the SBGT subsystem under test. SBGT is not assumed to be the initiator of any previously analyzed l cvent. Therefore, the proposed change does not involve a significant increase in the probability of any previously analyzed accident, Since the ftmetion is normally maintained by the SBGT subsystem being tested, assuming that the subsystem being demonstrated to be Operable, will infact, be proven Operable, the proposed change does not involve a significant increase in the consequences of any accident previously evaluated.

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

The proposed change will not involve any physical changes to plant systems, structures, or components (SSCs) or the manner in which these SSCs are operated, maintained, modified, tested or inspected. Therefore, this change will not create the possibility of a new or ditTerent kind of accident from any accident previously evaluated.

3. Does this change involve a significant reduction in a margin of safety?

The proposed change recognizes that the SBGT subsystem is unavailable during this required surveillance testing and ensures that the time in this condition is restricted. This is acceptable since the time duration for the testing is short relative to the probability of an event requiring system operation and since the function is normally maintained by the l SBGT subsystem being tested, assuming that the subsystem being demonstrated to be Operable, will in fact, be proven Operable. Consequently, this change does not involve a significant reduction in the margin of safety. (3.6.4.3-5)

DAEC 62 Revision E I

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6-CONTAINMENT SYSTEMS TFCIINICAL CilANGES - 1 ESS RESTRICTIVE (continued)

- (Lev.2 Labeled Comment / Discussion for ITS 3.6.4.3)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant hazards consideration. This determination has been performed in accordance with the criteria set forth in 10 CFR 50.92. The following evaluation is provided for the tluee categories of the significant hazards consideration standards:

-1. Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

13ased on the evaluations conducted to satisfy Generic Letter 91-04 requirements,1) no time dependent failures were found,2) other more frequent testing would have found many of the failures that were discovered during cycle tests, or 3) there was no loss of function (i.e. no common cause failures were found). Thus, the availability of systems required to mitigate the consequences of an accident are not adversely impacted. The proposed change does not increase the probability of any accident, as the overall system reliability is not reduced, nor are the performance of these surveillance tests considered to be accident initiators. Therefore, the probability or consequences of previously evaluated accidents are not significantly increased.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or change the methodr. ,0veming normal plant operation. For each of these tests, the nature of any failures, corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension. Therefore, this change does not create the possibility of a new or different kind of accident from any accident previously evaluated.

4 4

DAEC 63 Revision E

NO SIGNIFICANT IIAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS TECIINICAL CliANGES - 1.ESS RESTRICTIVE (LcY.2 I,abeled Comment / Discussion for ITS 3.6.4.3) (continued)

3. Does this change involve a significant reduction in a margin of safety?

Although the proposed change will result in an increase in the interval between surveillance tests, the impact on system availability is considered small based upon other more frequent testing, the availability of redundant systems or equipment, and the fact that there is no evidence of any existing equipment failures that would impact the availability of the system. Therefore, the proposed change will not significantly impact the availability or reliability of the plants systems or their ability to respond to plant transients and accidents. Therefore, the proposed change will not involve a significant reduction in a margin of safety.

DAEC 64 Revision E s

NO SIGNIFICANT HAZARDS CONSIDERATIONS -

SECTION 3.6 -CONTAINMENT SYSTEMS TECHNICAL CIIANGES - LESS RESTRICTIVE (continued)

(L ica Labeled Comment / Discussion for ITS 3.6.4.3)

DAEC has evaluated this proposed CTS change and has determined that it involves no significant

. hazards consideration. This determination has been performed in accordance with the criteria set fonh in 10 CFR 50.92. The following evaluation is provided for the three categories of the significant hazards consideration standards:

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

Based on the evaluations conducted to satisfy Generic Letter 91-04 requirements,1) no time depender.t failures were found,2) other more frequent testing would have found many of the failures that were discovered during cycle tests, or 3) there was no loss of function (i.e. no common cause failures were found). Thus, the availability of systems required to mitigate the consequences of an accident are not adversely impacted. The proposed change does not increase the probability of any accident, as the overall system reliability is not reduced, nor are the performance of these surveillance tests considered to be accident initiators. Therefore, the probability or consequences of previously evaluated accidents are not significantly increased.

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

The proposed change does not involve a physical alteration of the plant (no new or different type of equipment will be installed) or change the methods governing normal plant operation. For each of these tests, the nature of any failures, corrective actions that were taken, system redundancy, or detectability of the failures by other mid-cycle testing resulted in acceptable conditions for interval extension. Therefore, this change does not create the possibility of a new or difTerent kind of accident from any accident previously evaluated.

DAEC 65 Revision E

NO SIGNIFICANT llAZARDS CONSIDERATIONS SECTION 3.6--CONTAINMENT SYSTEMS -

ITECilNICAL Cil ANGES . LESS RESTRICTIVE (L ic.2 Labeled Comment / Discussion for ITS 3.6.4.3)-(continued)

3. Does this change involve a significant reduction in a margin of safety?

Although the proposed change will result in an increase in the interval between I

> surveillance tests, the impact on system availability is considered small based upon other more frequent testing, the availability of redundant systems or equipment, and the fact that there is no evidence of any existing equipment failures that would impact the -

availability of the system. Therefore, the proposed change will not significantly impact .

the availability or reliability of the plants systems or their ability to respond to plant transients and accidents. -Therefore, the proposed change will not involve a significant i reduction in a margin of safety.

4 i

i DAEC 66 ' Revision E o

J l

-,e -.-i-mv -e ,, , --ea -- aw, c ;,- -- m -,+ , enm,-,----eA ,,---r- +v -ew,--r-,, -,--- w - - -

ENVIRONhiENTAL ASSESSh1ENT ClIAPTER 3.6--CONTAINh1ENT SYSTEhtS These proposed TS changes have been evaluated against the criteria for and identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21. It has been detcrmined that the proposed changes meet the criteria for categorical exclusion as provided for under 10 CFR 51.22(c)(9). The following is a discussion of how the proposed TS changes meet the criteria for categorical exclusion.

10 CFR 51.22 (c)(9): Although the proposed changes involve changes to requirements with respect to inspection or surveillance requirements; (i) the proposed changes involve no Significant llazards Consideration (refer to the No Significant llazards Consideration section of this Technical Specification Change Request),

(ii) there is no significant change in the types or significant increase in the amounts of any ellluents that may be released offsite since the proposed changes do not affect the generation of any radioactive ef11uents nor do they affect any of the pennitted release paths, and (iii) there is no significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed changes meet the eligibility criteria for categorical exclusion set forth in 10 CFR 51.22(c)(9). Based on the aforementioned and pursuant to 10 CFR 51.22(b), no environmental assessment or environmental impact statement need be prepared in connection with issuance of an amendment to the Technical Specifications incorporating the proposed changes of this request.

DAEC 67 Revision E i

I

Enclosure 2 to ,

NG-97-2097 I

Relocated Items Matrix for the 4

DAEC Improved Technical Specifications s

9

Relocated items Matrix 18-Nov-97 US DOC CTS Description Location 1.0 R1 1.5 Operability Definition Bases 2.0 R1 1.1.D Definition of" Top of Active Fuel" UFSAR 2.0 R2 6.7.3 Notification of VP on SL violation QAPD 3.0 R1 1.26 25% extension convenience Bases 3.1.2 R1 3.3.E3 Perform reactivity difference analysis Bases 3.1.3 R1 3.3.A.2.e Details on how to disarm CRDs Bases 3.1.4 R1 4.3.D.1 Details on scram time testing Bases 3.1.5 R1 4.3 A.2.a CR accumulator level switch Deleted by DOC L3 surveillance 3.1.7 R1 4.4.A.2.a Surveiltance on operation /setpoint of IST Program SLC reliefs 3.1.7 R2 4.4.A.2.b Details on how to verify SLC flow Bases 3.1.7 R3 4.4 A 1 Details on SLC pump loop testing Bases 3.10.3 R1 3.9.A.3.b Details conceming shutdown margin Bases testing 3.10.3 R1 3.9 A 3.a.2 Single rod / drive mechanism Bases withdrawal 3.10.3 R2 3.9.A.3 How to disarm control rods Bases 3.10.4 R1 3.9.A.3 a.2 Single rod / drive mechanism Bases withdrawal 1

l

DOC CT3 Description Location US 3.10.4 R1 3.9.A 3 b Details conceming shutdown margin Bases testing R2 3 9.A.3 How to disarm control rods Bases 3.10.4 3.10.5 R1 3.9 A 3.b Details conceming shutdown margin Bases testing R2 3.9 A 3 How to disarm control rods Bases 3.10.5 R1 4.12.C.1.b Verify MCPR following significant Deleted by DOC L2 3.2.2 pourchange 3.2.3 R1 3.12.B LH: TRM 3.3.1.1 R1 Table 4.1-1 note f Calibrate LPRMs using the TIP Bases system 3.3.1.1 R1 Table 4.1-1 note b 1/2 decade overlap for SRMs/lRMs Bases 3.3.1.1 R10 Table 4.1-1 notes g. k, I Testing requirements for RPS trip g: Bases functions k, i: Deleted by DOCS A9, A14 3.3.1.1 R11 Table 3.1-1 note i Additional trip functions actuate EOC- Bases RPT system 3.3 1.1 R12 2.1 Relocate " Limiting Safety System UFSAR Settings" 3 3.1.1 R12 Table 3.1-1 Relocate " Trip Level Settings" UFSAR 3.3.1.1 R12 2.2 Relocate" Limiting Safety System UFSAR Settings" 3.3.1.1 R2 Figure 2.1-1 Core power vs. recirculation flow UFSAR 3.3.1.1 R2 2.1.A.1 APRM flow biased high scram UFSAR equation 2

ES- DOC CTS Description Location 3.3.1.1 R3- Table 3.1-2 RPS response times UFSAR 3.3.1.1 R3 3.1.A RPS response Smes UFSAR 3.3.1.1 R4 31.A notes *," When not to place RPS channels in Bases trip 3.3.1.1 R5 Table 3.1-1 note b IRMs automatically bypassed when Basas mode sw. in RUN 3.3.1.1 R6 Table 3.1-1 note c APRM inoperability requirements Bases 3.3.1.1 R7 Table 3.1-1 note e MStV closure trip bypassed when Bases mode sw not in RUN 3.3.1.1 R8 4.1.A.2 Staggered test basis of RPS functions Bases 3.3.1.1 R9 Table 3.1-1 Turbine First Stage Pressure Bases Permissive 3.3.1.1 R9 Table 4.1-1 Turbine First Stage Pressure Bases Permissive 3.3.1.2 R1 3.9.B.1 SRMs be inserted during Core Alts UFSAR 3.3.1.2 R2 3.9.B.3 Get 3 cps by 2 of 4 assemblies Bases loaded neat 4 SRMs 3.3.2.1 R1 Table 3.2-C APRM, IRM, SRM, SDV & Recirc TRM Flow Rod Blocks 3.3.2.1 R1 Table 4.2-C APRM, IRM SRM, SDV & Recirc TRM Flow Rod Blocks 3 3.2.1 R2 Table 3.2-C note a RBM bypassed when periphera! rod Bases selected 3.3.2.1 R3 Table 4.2-C Channel check RBM upscale and Deleted by DOC L8 j downscale functions 3

[TS DOC CTS Description Location 3.3.2.1 R4 Table 4.2-C note c Include RMC multipk:xing system Bases input in RBM CFT 3.3 2.1 R5 Table 3.2-C Relocate " Trip L.evel Settings" UFSAR 3.3.2.1 R6 4.3.C.1.c Details on RWM testing Bases 3.3.2.1 R6 4.3.C.1.d Details on RWM testir'g Bases 3.3 2.1 R6 4.3.C.1.b Details on RWM testing UFSAR 3.3.3.1 R1 4.2.H Extra PAM instruments TRM 3.3.3.1 R1 3.2 H Extra PAM instruments TRM 3.3.3.1 R1 4.2.F Surveillance instruments TRM 3.3.3.1 R1 3.2 F Surveillance instruments TRM' 3.3.3.1 R2 Table 3.2-H Descriptive details for various PAM Bsses instruments 3.3.3.1 R2 Table 4.2-H Descriptive details for various PAM Bases instrumen'.s 3.3.3.1 R3 Table 4.2-H notes b e Testing method for Rad Monitors and Bases H2/02 Monitors 3.3.3.1 R3 Table 3.2-H note b Normal for Containment H2/02 Bases Monitoris Standby 3.3.3.2 R1 3.10.B Verify Remote Shutdown Panels UFSAR locked once/ week 3.3.3.2 R1 4.10.B Verify Remote Shutdown Panels UFSAR locked once/ week 3.3.4.1 R1 Table 4.2-G EOC-RPT Response Times UFSAR 3.3.4.1 R2 Table 3.2-G Relocate 7tip Level Settings" UFSAR 4

{TS DOC CTS Desedntio_n Location 3.3.4.1 R3 Tabke 3.2-G note b Trip system description for ATWS- Bases RPT and EOC-RPT 3.3.4.2 R1 4.2 G ARI functions TRM 3.3.4.2 R1 3.2.G ARI functions TRM 3.3.4.2 R2 Table 3.2-G Relocate " Trip Level Settings" UFSAR 3.3.4 2 R3 Table 3.2-G note b Trip system description for ATWS- Bases RPT and EOC-RPT 3.3.5.1 R1 4.2. B.2.f Perform LSFT on Safeguards TRM Systems Area Coolin; 3.3.5.1 R2 Table 3.2-B Relocate " Trip Level Settings" UFSAR 3 3.5.1 R3 Table 3.2-B notes c, d Desenptive material for HPCI pump Bases trips 3.3.5.2 R1 Table 3.2-B notes c, d Desenption for RCIC pump tnps Bases 3.3.5.2 R2 Table 3.2-B Relocate " Trip Level Settings" UFSAR 3.3.6.1 R1 3.2.A.1.b note

Describes when to place channels in Bases trip 3.3.6.1 R10 Table 3.2-A RCIC and HPCI isolation function TRM 3.3.6.1 R11 Tabte 3.2-A HPCl/RCIC steam low pressure trip UFSAR reset setpoint 3.3.6.1 R2 Table 3.2-B Relocate " Trip Level Settings" UFSAR 3.3.6.1 R2 Table 3.2-A Relocate " Trip Level Settings" UFSAR 3.3 6.1 R3 Table 3.2-A notes b, c, e Descriptive material Bases 3.3.6.1 R4 Table 3 2-A note o 2 MSL tunnel temperature sensors Bases per MSL required 5

HS DOC CTS Description Location 3 3.6.1 R5 Table 4.2-A notes ##,a,c Descriptive information on various a: Basas SRs c De_ :M by DOC A9

1. UFSAR 3.3.6.1 R6 Table 3.2-A notes d, f, j Descriptive information for sarious f: Bases functions d, j: UFSAR 3.3 6.1 R7 Table 4.2-A note ### RPV Level-Lo and DW Pressure-Hi UFSAR common to RPS/ECCS 3.3.6.1 R8 Tab!e 3.2-D note c MSL Rad Monitors trip Mech. Vac. Bases Pump 3.3.6.1 R9 Table 4.2-D note a Method for CFT for MSL Rad Deleted by DOC A9 Monitors 3.3 6.2 R1 3.2.A.1.b note

Describes when to place channels in Bases trip 3.3.6.2 R2 Table 3.2-A Relocate " Trip Level Settings" UFSAR 3.3.6.2 R3 Table 3.2-A note c Respective signals start the SBGT Bases system 3.3.6.2 R4 Table 4.2-A note ### RPV Level-Lo and D'N Pressure-Hi UFSAR common to RPS/ECCS 3.3.6.3 R1 2.2.1 Relocate" Limiting Safety System UFSAR Settings" 3.3.8.1 R1 Table 3.2-8 Relocate " Trip Level Settings" UFSAR 3.4.1 R1 3.3.F.3 Methods to exit the Exclusion Region Bases 3.4.1 R2 3.3.F.5.b Reqs for opening lower speed recire UFSAR pump disch viv 3.4.1 R3 3.3 F.4.c Methods to ensure idle loop is Bases isolated 6

ES CTS Description Location D4r'.

3.4.2 R1 4.6.E.1 Performance of Jet Pump SRs after Deleted by DOC A3 abnormal changes 3.4.2 R2 4.3 F.4.a Update baseline data ASAP ' Bases 3.4.2 R2 4.6.E.4 Update baseline data ASAP Bases 3.4.3 R1 4.6.D.2 Disassemble and inspect 1 SRV per ISI Program l cycle 3.4.3 R2 4.6.D.3 How to verify SRV is manually Bases opened 3.4.5 R1 Table 3.2-E notes a, b Descriptions of Sump System & Air B'ses Sampling System 3.4.6 R1 Table 4.6.B.1-1 Requirements for sampling and gross iodine: Deleted by anary;is DOCL5 filter Bases 3.4.9 R1 3.6.A.2 Meet restrictions of operating curves, Deleted by DOC L3 vent RPV 3.4.9 R2 4.6 A 1 Details on when RCS temperature Bases SRs can be stopped 3.4.9 R3 4.6 A 1 Areas for RPV temperature Bases monitoring 3.4.9 R3 4.6.A.2 Areas for RPV temperature Bases monitoring 3.4.9 R4 3.6.A.4.b Perform engineering evalif P-T limits Bases exceeded 3.4.9 R5 4.6 A.4 Record surveillance results OAPD 3.4 9 R5 4.6.A.3 Record surveillance results QAPD 3.4.9 R5 4.6.A.2 Record surveillance results QAPD 7

f US DOC CIS Description Location 3.5.1 R1 4.5.H.1 Test LPCI and CS line pressure UFSAR switches 3 5.1 R2 3.5.1 ES Compartments Cooling & TRM Ventilation 3.5.1 R3 4.6.D.3 Manualoperation of each relief valve Bases 3.5.1 R4 4.5 A 3.c Pump operability and MOV IST Program operability tests 3.5.1 R4 4.5.D.1.b Pump operability and MOV IST Program operability tests 3.5.1 R4 4.5.A.3.b Pump operability and MOV IST Program operability tests 3.5.1 R4 4.5.A.1.c Pump operability and MOV IST Program operability tests 3.5.1 R4 4.5.A.1.b Pump operability and MOV IST Program operability tests 3.5.1 R4 4.5.D.1.c Pump operability and MOV IST Program operability tests 3 5.1 R5 4.5.A.3 e Verify RHR valve panelinstruments Bases operate normal 3.5.1 R6 4.5. D.1.f Verify HPCI suction can be Bases transferred 3.5.1 R7 4.5 F.1.b Leak test ADS N2 accumulator Bases, IST Program check valves 3.5.2 R1 4.5.H.1 Test LPCI & CS line pressure UFSAR switches 3.5.2 R2 4.5.1 SRs for all ECCS room coolers TRM 6

ES ' DOC CTS Description Location 3.5.2 R3 3.5.G.3 Operability reqs for RHR and CS Bases pumps in Mode 4/5 3 5.3 R1 4.5.E.1.a include ROC auto-start on low water Bases level signai 3 5.3 R2 4.5.E.1.c Pump operability and MOV IST Program operability tests 3.5.3 R2 4.5.E 1.b RCIC Pump operability and MOV IST Program operability tests 3.5.3 R3 4.5.E.1.f Verify RCIC suction can be Bases transferred 3.5.3 R4 3.5.! ES Compartments Cooling & TRM Ventilation 3.6.1.1 R1 1.15 Blind flange and manway details Bases 3.6.1.3 R1 1.15 PCIV details Bases 3.6.1.3 R2 4.7.B.1.b.1 Close & reopen normally open power IST Program operated PCIVs 3 6.1.3 R3 4.7.B.1.b.2 Requirement for pc.ver to be <75% IST Program for MSIV testing 3.6.1.3 R4 3.7.B.4.a List of containment vent / purge valves Bases and groups 3.6.1.3 R4 4.7.A.1.c List of containment vent / purge valves Bases and groups 3.6 1.5 R1 4.6.D.3 How to verify LLS valve has manually Bases opened 3.6.1.6 R1 3.7.D.1 Details on operable vacuum breakers Bases 3.6.3.1 R1 3.7.H.1 Details on operabie CAD system Bases 9

US DOC CTS Descrio lon Location 3 6.3.1 R2 '4.7.H.1 Test CAD system annually UFSAR 3.6.3.1 R3 3.7.H.2 Determine CAD system contains UFSAR minimum req N2 3.6.4.1 R1 3.7.J.1.d Maintain secondary containment if UFSAR cask being moved 3.6.4.1 R2 4.7.J.1.a Requirements for wind concltions i Bases ar.d filter train 3.6.4.2 R1 3.7.K.1 SCIVs operable to move cask UFSAR 3.6.4.2 R2 1.16 SCIV details Bases 3 6 4.3 R1 3.7.L1 SBGT operable to move cask UFSAR 3.6.4.3 R2 4.7.L1.f Inspect SBGT system drains UFSAR 3.7.1 R1 4.5.C.1.b SRs for RHRSW pumps and valves IST Program 3.7.1 R1 4.5.C.1.a SRs for RHRSW pumps and valves IST Program 3.7.2 R1 4.5.J.1.d SRs for RWS pumps and valves IST Program 3.7.2 R1 4.5.J.1.b SRs for RWS pumps and valves IST Program 3.7 2 R1 4.5.J.1.c SRs for RWS pumps and valves IST Program 3.7.3 R1 4.8 E.1.b SRs for ESW pumps and valves IST Program 3.7.3 R1 4.8.E.1.c SRs for ESW pumps and valves IST Program 3.7.4 R1 4.10.A.3 Details on demonstrating the SFU Bases system operable 3.7.4 R1 4.10 A2.d Details on SFU system UFSAR demonstration of operability 3.7.4 R2 4.10.A.3 Requirements for wind conditions Bases 10

.- .=.

ES . DOC CTS Description Location 3.7.8 R1 ' 3.9.C.1.a.1 Place loads above pool /RPV in safe Bases configur

ion 3.8.1 R1 4.8.A.2.a.2 Record survei!!ance results QAPD 3.8.1 R1 4.8 A.2.b Record surveillance results QAPD 3.8.1 R2 4.8 A.2.c inspect DGs once per operating cycle UFSAR 3.a.1 R3 4.8.A.2.a.1.b Manualloading of DGs for Bases surveillance testing 3.8.1 R3 4.8.A.2.a.1.a Manual starting of DGs for Bases surveillance testing 3.8.1 R3 4.8.A.2.a.2 Manual starting of DGs for Bases surveillance testing 3.8.3 R1 4.8.A.2.e Record DG fuel monthly and after QAPD DG use 3,8.3 R2 4.8.A.2.a.1.c Coeck proper air compressor UFSAR operation 3.8.4 R1 4.8.B.2.a Monitor battery room H2 TRM concentration 3.8.4 R1 3.8.B.2.a Provide portable battery rm TRM ventitation equipment 3.8.4 R2 3.8.B.1 24 volt batteries and chargers TRM operability 3.8.4 R2 3.8.B.2.d 24 voit batteries and chargers TRM operability 3.8.4 R3 4.8.B.1.c Det. specific gravity & voltage during Bases, OAPD disch test 3.8.4 R4 3.8.B 2.b Perform cross-train checks SFDP 11 l

i

a US DOC CTS Description Location 3.8.6 R1 4.8.B.1.b Measure and rccord Battery cell QAPD parameters 3.8.7 R1 3.8.C.2.b Details on operable distnbution Bases system 3.8.7 R1 3.8.C.2.a Details on operable distritGc:- Bases system 3.8.7 R1 3.8.C.1 Details on operable distribution Bases system 3.8.7 R2 4 8.C.1 inspect each essential AC breaker UFSAR 3.8.7 R3 3.8.B.2.b Perform cross-train checks SFDP 3.9.1 R1 4.9.A.1.a Hoist load setpoints UFSAR 3.9.1 R1 4.9.A.1.b Hoist load setpoints UFSAR 4.0 R1 5.6 Seismic design details UFSAR 4.0 R1 5.3 RPV description UFSAR 4.0 R1 5.4 Containment description UFSAR 4.0 R2 1.37 Definition for Site Boundary UFSAR 5.0 R1 6.4.2 Fire Protection Program UFSAR 5.0 R1 6.1.2 Fire Protection Program Fire Plan 5.0 R10 6.8.5 Preventative and corrective UFSAR maintenance program 5.0 R11 6.9.2 notes ". *** Radiation measurement distances Added to ITS per NRC question 5.0-8 5.0 R12 6.9.4 b Radiological Environmental ODAM Monitoring Program 12

ES DOC CTS Description Location 5.0 R13 6.9.5 Source leakage tests TRM.QAPD 5.0 R13 7.9.5 Source leakage tests TRM,QAPD 5.0 R14 6.10 Record retention QAPD 5.0 R15 6.11.1.a Startup Report QAPD 5.0 R16 6.6.1.b Reportable Event Action QAPD 5.0 R17 6.15 Process Control Program QAPD 5.0 R18 4.2.1 Monitoring gas downstream of off- TRM gas recombiners 5.0 Ri8 3.2.1 Monitoring gas downstream of of'- TRM gas recombiners 5.0 R19 3.14.B Uquid holdup tank instrumentation ODAM 5.0 R19 4.14. B Liquid holdup tank instrumentation ODAM 5.0 R19 4.14. A Liquid holdup tanks ODAM 5.0 R19 3.14.A Liquid holdup tanks ODAM 5.0 R2 6.1.3 QA Program QAPD 50 R20 4.8 A 2.d Testing new and stored DG fuel oit Bases 5.0 R20 4.8.A.2.f Testing new and stored DG fuel oil Bases, QAPD 50 R20 4.8.A 2.g Testing new and stored DG fuel oi! Bases 5.0 R3 6.11.2.a.2 LHGR TRM 5.0 R4 Table 6.2-1 Minimum shift crew and license UFSAR requirements 13

US DOC CTS Description Location 5.0 R5 6.3.4 Training requirements for plant UFSAR management 5.0 R6 6.4.1 Staff training requirements UFSAR 5.0 R7 6.5 Review and audit functions QAPD 5.0 R8 6.8.1.12 lodine Monitoring Program UFSAR 5.0 R8 6 9.1 Radiation protection procedures UFSAR 5.0 R9 6.8.2 Review and approval process QAPD 5.0 R9 6.8.3 Temporary change process QAPD CTS 3.11 R1 3.11 River Level TRM CTS 3.2.D R1 3.2.D Radiation Monitoring instrumentation ODAM CTS 3.5 B R1 3.5.8 Containment Spray TRM CTS 3.6.B.2 R1 3.6.B.2 Chemistry TRM CTS 3.6.G R1 3.6.G StructuralIntegrity TRM CTS 3.6.H R1 3.6.H Snubbers TRM,OAPD CTS 3.7.M R1 3.7.M MechanicalVacuum Pump TRM,ODAM CTS 4.11 R1 4.11 River Level TRM CTS 4.2.D R1 4.2.D Radiation Monitoring Instrumentation ODAM CTS 4.5.B R1 4.5.B Containment Spray TRM CTS 4.6.B.2 R1 4.6.B.2 Chemistry TRM CTS 4.6.G R1 4.6.G Structuralintegrity TI<M 14

ES DOC CTS Description Location CTS 4 6.H R1 4.6.H Snubbers TRM,QAPD CTS 4.7.M R1 4.7.M Mechanical Vacuum Pump TRM,ODAM TRM T 3.31 R1 Table 3.2-G ARI trip Level settings UFSAR TRM T 3 3.2 R1 Table 3.2-C Rod Block tnp Level settings UFSAR TRM T 3.3.3 R1 Table 3.2-H Various details for PAM instruments TRM Bases TRM T 3.3 6 R1 Table 3.2-F Total channels for Surveillance inst UFSAR TRM T 3.3.7 R1 3.2.1.1, Note " Location of hydrogen sampling points TRM Bases TRM T 3.3.7 R1 4.2.1.1.b Perform inst cal for hydrogen TRM Bases monitors TRM T 3.7.2 R1 4.6.H.4 Required engineering evaluation for TRM Bases snubbers TRM T 3.7.2 R2 4.6.H.8 Documentation requirements for QAPD snubbers TRM T 3.7.4 R1 3.7.M.2 Max allowed rad release rate for ODAM mech vac pump TRM T 3.8.1 R1 4.8.B.1.c Record specific gravity / voltage each TRM Bases 24VDC cell TRM T 3.9 R1 6 9.5.B Reporting requirements for sealed QAPD sources 15 w _ _ . - . _ - -. =

- _ ._ - _

NG-97-2097, Forwards Response to RAIs Re DAEC Request to Convert TSs to Improved Tss,Per NUREG-1433

Contents

Text

References:

Enclosures:

4.1 BACKGROUND

Navigation menu

NG-97-2097, Forwards Response to RAIs Re DAEC Request to Convert TSs to Improved Tss,Per NUREG-1433

Text

References:

Enclosures:

4.1 BACKGROUND

Navigation menu

Search