ML20045A905

Application for Amend to License DPR-49,consisting of Request for TS Change (RTS-197A),correcting Errors & Improving Consistency of Rewrite of TS Section 3.6, Primary Sys Boundary.

ML20045A905
Person / Time
Site:	Duane Arnold
Issue date:	06/04/1993
From:	Franz J IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation
Shared Package
ML112380515	List: ML112380514 ML20045A905
References
NG-93-2104, NUDOCS 9306150273
Download: ML20045A905 (26)

Text

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l Iowa Electric Light and Power Company June 4, 1993 .

2ons r.ra m .2n. NG-93-2104 1j Dr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation  !

U. S. Nuclear Regulatory Commission [

Attn: Document Control Desk .!

Mail Station P1-137 Washington, DC 20555 g

Subject:

Duane Arnold Energy Center Docket No: 50-331  ;

Operating License No: DPR [

Request for Technical Specification  :

Change (RTS-197A): Revision to TS l

-Section 3.6, " Primary System Boundary" 7

Reference:

1) Letter NG-92-5326, Franz (IELP) to Murley (NRC),

dated December 31, 1992  :

File: A-117  !

Dear Dr. Murley:

In Reference 1, Iowa Electric Light and Power Company (IELP) requested .

revision to the Technical Specifications (TS) for the Duane Arnold Energy  !

Center (DAEC). Subsequent to that letter, some erroneous references and  ;

inconsistencies were identified. This letter corrects those errors and improves the consistency of the rewrite of TS Section 3.6.

.i This letter and attachments constitute a resubmittal of the original request  :

for TS revision (RTS-197, Reference 1), and' supersedes that' submittal in its  ;

entirety. ,

A copy of this submittal, which includes our analysis of significant hazards consideration, is being . forwarded to our appointed state official pursuant to  ;

the requirements of 10 CFR 50.91. l Should you have any questions regarding this matter, please contact this office.

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140050  ;

P 9306250273 93060 Y

.fDR ADOCK 05000 PDR 31 fg /U \: l-General Office

• P.O. Bos 351

• Cedar llapids, Iowa 5240G

• 329/398-4421

• r

Dr. Thomas E. Murley

~ June 4, 1993 NG-93-2104 Page 2 This letter is true and accurate to the best of my knowledge and belief..

IOWA ELECTRIC LIGHT AND POWER COMPANY

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'. , \

By /j67 . Ah .I JohnF.F'ranz/ j

/ ice President, Nuclear i State of Iowa I (County) of Linn .f

- 1 Signed and sworn to before me on this N day of s } k nb ,

1993, by -d ohn F. F rano ,.

9 g  :

l (l%Y% M l/hfNj i

Notary Public in and for the Stite of Ioda

\

%$ MARY mme- mcHnEOWEALl  :

J cru f IQ45 :i Commission Expires ' .

JFF/TWP/pjv~ j Attachments: 1) Evaluation of Change Pursuant to 10 CFR 50.92

2) Proposed Change RTS-197A to the Duane Arnold Energy Center Technical Specifications  !

3) Safety Assessment .

4) Environmental Considerations  ;

cc: T. Page [

L. Liu (w/o attachments) -! '

L. Root (w/o attachments)

R. Pulsifer (NRC-NRR) i J. Martin (Region III) i NRC Resident Office S. Brown (State of Iowa)  ;

DCRC s

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RTS-197A Attachment I to l NG-93-2104 i Page 1 of 24 EVALUATION OF CHANGE PURSUANT TO 10 CFR 50.92

Background:

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In 1991, an independent evaluation of the Technical Specifications (TS) for the  !

Duane Arnold Energy Center (DAEC) was conducted as part of .the DAEC TS  !

Improvement Program. A portion of the Program included comparison of the DAEC  ;

TS with TS from similar plants, Standard TS (NUREG-1202, July 1986), and the .

draft Improved Technical Specifications (NUREG-1433). Based on this comparison, the current DAEC TS Section 3.6, " Primary System Boundary," has been rewritten i and is the subject of this submittal . Specifically, the proposed' changes-  !

contained in this submittal will revise the Limiting Conditions For Operation  ;

(LCO) and Surveillance Requirements (SR) for Thermal and Pressurization Limitations, Coolant Chemistry, Coolant Leakage, Safety and Relief Valves, Jet Pumps, Jet Pump Flow Mismatch, Structural Integrity, and Shock Suppressors . ,

(Snubbers). In addition, definitions for IDENTIFIED tEAKAGE, UNIDENTIFIED .

LEAKAGE, TOTAL LEAKAGE, and DOSE EQUIVALENT I-131 are being incorporated into TS  ;

Section 1.0, " Definitions." '

i Iowa Electric Light and Power Company, Docket No. 50-331  !

Duane Arnold Energy Center, Linn County, Iowa

(

Date of Amendment Request: June 4, 1993 i

Description of Amendment Request: l The proposed amendment revises DAEC TS Sections 1.0 and 3.6 to provide additional  !

definitions and improve the clarity and consistency of LCOs and SRs for Primary j System Boundary. The majority of the changes being proposed are consistent with  ;

comparable Specifications in the Standard TS (NUREG-1202). The other changes are ' --

editorial or administrative in nature. i Definition 41, IDENTIFIED LEAKAGE  ;

The existing DAEC TS do not define IDENTIFIED LEAKAGE. This definition is being )

added to improve clarity and consistency.with LCO 3.6.C, " Coolant Leakage." This '

definition is consistent with the. guidance provided by Standard TS.

i Definition 42, TOTAL LEAKAGE The existing DAEC TS do not define TOTAL LEAKAGE. This definition is being added ,

to improve clarity and consistency with LC0 3.6.C, " Coolant Leakage." j i

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RTS-197A Attachment I to NG-93-2104 Page 2 of 24 Definition 43, UNIDENTIFIED LEAKAGE lhe existing DAEC 15 do not define UNIDENTIFIED LEAKAGE. This definition is being added to improve clarity and consistency with LCO 3.6.C, " Coolant Leakage."

This definition is consistent with the guidance provided by Standard TS.

Definition 44, DOSE EQlllVALENT I-131 The existing DAEC TS do not define DOSE EQUIVALENT I-131. This definition is being added to improve the clarity and consistency with LC0 3.6.B, " Coolant Chemi s t ry." This definition is consistent with the guidance provided by the Standard TS.

TS Section 3/4.6, Primary System Boundary The existing Applicability and Objective sections are editorially revised to use initial caps for Reactor Coolant System. This is consistent throughout the Section 3.6 rewrite.

TS Section 3/4.6.A, Thermal and Pressurization Limitations This section is being revised to improve clarity and provide consistency within -

the DAEC TS as well as to adopt the specific language of the Standard TS. The existing TS do not specify clearly and concisely the actions to be taken when a specific LCO or SR is exceeded. The proposed revisions to this section eliminate that problem. A summary of the proposed changes follows:

Existing TS Section 3.6. A.1 is being editorially revised to make the wording consistent with existing SR 4.6.A.1.

Existing TS Section 3.6.A.2 is being revised to delete the last part of the LCO which provides a commitment to update Figure 3.6-1 six months prior to 16 effective full power years. This commitment is not required in the LCO since it does not verify system OPERABILITY requirements or provide any additional infonnation assisting in the operation of the plant or mitigating any accidents. This commitment has been relocated to the-Bases Section.

Existing TS Section 3.6.A.3 is being editorially revised to identify the location where temperature readings are to be taken before reactor vessel head bolting studs are placed under tension. In addition, editorial changes are being made to be consistent with the existing Bases.

Proposed TS Section 3.6. A.4 is being added. ' The existing TS do not specify the actions if the temperature / pressure limits are exceeded. The

5 RTS-197A Attachment I to NG-93-2104 Page 3 of 24 addition of this LC0 provides time limits for bringing temperature /

pressure back within specification requirements and performing an engineering evaluation. It requires shutdown only if the plant cannot comply with the specific actions.

Existing TS Section 3.6.A.5 has been revised to provide additional guidance, consistency, and to incorporate specific information from the Standard TS. The LC0 will specifically identify the MODES of operation i which apply to the recirculation pump. The' current TS did not provide  ;

this information. Existing TS Sections 3.6.A.4 and 3.6.A.5 were 1 editorially revised to provide clarity and consistency within the DAEC TS i using the guidance provided by the Standard TS and were combined into a  ;

single section.  !

Existing SR 4.6. A.1 has been editorially revised to provide consistency and clarity within this section of the DAEC TS. In addition, the word

" logged" has been replaced with the word " recorded." This is discussed in more detail in Attachment 2.

Existing SR 4.6. A.2 has been revised to delete the last two paragraphs which discuss when the last specimens were withdrawn and when the next ones are scheduled'to be withdrawn. This type of information should not be contained in the SR or LCO. This information has been incorporated into the Bases Section. A SR has been added which requires specimens to be removed in accordance with 10CFR50, Appendix H. This SR is in accordance with Generic Letter 91-01, " REMOVAL OF THE SCHEDULE FOR THE WITHDRAWAL OF REACTOR VESSEL MATERIAL SPECIMENS FROM TECHNICAL SPECIFICATIONS."

Existing SR 4.6.A.3 has not been changed.

Existing SRs 4.6. A.4 and 4.6. A.S have been reorganized. The existing information has been maintained and itemized under the proposed SR 4.6.A.4.

TS Section 3/4.6.8, Coolant Chemistry This section is being revised to provide clarification and consistency within the DAEC TS as well as adopt specific language of the Standard TS. The entire existing Section 3/4.6.B is being either revised or new LCOs and SRs added. A summary of changes are as follows:

Existing TS LC0 3.6.B.1 has been revised and divided into three different -

LCOs. The existing LC0 contained information which was difficult to read and understand. This proposed revision does not change the actual intent

l l

RTS-197A Attachment 1 to  !

NG-93-2104  !

Page 4 of 24 of the existing LCO but made it clearer.

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Existing TS LCO 3.6.B.2.a is being revised. The existing TS contains information that will be easier to understand in tabular format. In addition, the LCO references a steaming rate of 100,000 pounds per hour.

This LC0 was revised to place appropriate information in a new table which references a temperature associated with a MODE of operation. This is  ;

more meaningful to plant personnel than rates in " pounds per hour." 'The j other information contained in this LC0 has been relocated in other TS  !

within this section. '

Existing TS LC0 3.6.B.2.b has been - revised. Much of the information l contained is incorporated in new Tables 3.6.B.2-1 and 4.6.B.1-1. As -l stated above, more meaningful plant MODE conditions have been used instead l of rates in " pounds per hour." l i

Existing TS LC0 3.6.8.2.c has been incorporated into proposed LCOs within i this section. i Existing TS LCO 3.6.B.2.d has been incorporated into several proposed LCOs within this section. l Existing TS LC0 3.6.B.3.a has been revised. This information is either-provided in the proposed LCOs or in the new Table 3.6.B.2-1.

Existing TS LC0 3.6.B.3.b has been revised and incorporated into proposed l LCO 3.6.B.2.a.3. This LC0 provides guidance for continuous conductivity .l monitoring and incorporates the . Standard TS shutdown statement.

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Existing TS LCO 3.6.B.4 has been incorporated into the individual proposed LCOs.

Proposed TS LCO 3.6.B.2.d has been added, requiring an engineering -

evaluation be performed to verify structural integrity if the limits

, specified in Table 3.6.B.2-1 are exceeded.

Existing SRs 4.6.B.1.a through 4.6.B.1.h have been revised. and  ;

incorporated into proposed Table 4.6.B.1-1 and the proposed SRs. .

Editorial changes have also been made with the SRs patterned after those l of the Standard TS.

4 Existing SR 4.6.B.2, 4.6.B.2.a, and 4.6.B.2.b have been incorporated into proposed Table 4.6.B.1-1 and the proposed SRs. Editorial changes have been made to the SRs'to be consistent with Standard TS. ,

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l RTS-197A Attachment 1 to' j NG-93-2104'  !

Page 5 of 24  :

Proposed TS SRs 4.6.B.2.a, 4.6.B.2.b, 4.6.B.2.c, and 4.6.B.2.d have been '

added to provide additional guidance for obtaining samples as specified in ,

Table 3.6.8.2-1. -!

Existing SRs 4.6.8.3.a and 4.6.8.3.b have been revised and retained as '

proposed SR 4.6.B.2.e and 4.6.B.2.f. The specific monitoring locations i have been relocated to the Bases Section.  !

TS Section 3/4.6.C, Coolant Leakage '!

This section is being revised to clarify existing LCOs, SRs, add specific

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l shutdown requirements and provide consistency with the DAEC TS and Standard TS. .;

A summary of the proposed changes are as follows:

Existing TS LCOs 3.6.C.1, 3.6.C.1.a, 3.6.C.1.b, and 3.6.C.1.c have. been li editorially revised to provide clarity. The editorial changes consist of capitalizing defined terms and replacing existing words to be consistent-with Standard TS. i l

Existing TS LC0 3.6.C.3 has been renumbered to LC0 3.6.C.2. Proposed TS- 1 LCO 3.6.C.2 has been editorially revised to provide clarity and consistency with the DAEC TS and the guidance provi_ded in the Standard TS. -;

In addition, a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> action statement has been added. This- will. allow

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leakage to be brought back within its limits before a shutdown action is j initiated. The proposed shutdown ' action has also been. revised to- t incorporate the guidance provided in the Standard TS.  ;

I Existing LC0 3.6.C.2 has been renumbered to 3.6.C.3. Existing LCO 3.6.C.2 does not provide either a reference or specific requirements that define Sump System OPERABILITY. Therefore, proposed TS LCO 3.6.C.3 is being  ;

revised to reference the applicable section of the DAEC TS Table 3.2-E 'j which defines Sump System OPERABILITY. l Proposed TS LCO 3.6.C.4 is being added to state specific actions.to take 1 in the event that the Sump System is inoperable. In addition,-a shutdown requirement is being added to require being in at least HOT SHUTDOWN ,

within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within' the- following 24

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1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> if the Sump System cannot be restored to 0PERABLE status within 24  ;

hours. This addition is consistent with the guidance provided in the' l Standard TS.  !

Proposed TS LCO 3.6.C.5 is being added to state specific actions and shutdown requirements to take in the event neither the Sump System nor the Air Sampling System is OPERABLE. This revision is consistent with the.  ;

guidance provided in the Standard TS.  ;

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RTS-197A Attachment I to j NG-93-2104 i Page 6 of 24 Existing SR 4.6.C.1 is being revised to use initial capital letters for . t the Reactor Coolant System and Sump System. Th.is is an editorial change which is consistent with the rest of the DAEC TS.

Proposed TS SR 4.6.C.2 is being added to verify OPERABILITY of the Sump  !

System in accordance with Table 4.2-E. The existing TS SRs do not  !

currently contain this requirement. l Existing SR 4.6.C.2 is being renumbered to 4.6.C.3. The existing SR does j not define requirements to verify Air Sampling System OPERABILITY in the ,

event that the Sump System becomes inoperable. The revision to existing i SR 4.6.C.2 (now proposed SR 4.6.C.3) consists of verifying the Air Sampling System is OPERABLE in accordance with-Table 4.2-E.  !

The Bases Section 3.6.C & 4.6.C have been revised to reflect the proposed l changes. j

.t TS Section 3/4.6.0, Safety and Relief Valves This section has been revised to clarify existing LCOs, SRs, add specific shutdown requirements, and to provide consistency with the DAEC TS and the 'I guidance provided by the Standard TS. A summary of the proposed changes are as I follows:

4 ,

Existing TS LCO 3.6.D.1 is being revised to use proper MODE titles. For  !

consistency, all defined terms are to be identified in the DAEC TS in all caps. In addition, a note was added to state that SRVs which perform an i ADS function must also satisfy the OPERABILITY requirement as specified in '

Specification 3.5.F.

Existing TS LC0 3.6.D.2.a is being revised to clarify the LC0 requirements  !

in the event that the safety function of one relief valve becomes  ;

inoperable. l

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Existing TS LC0 3.6.D.2.b is being revised to clarify the LCO requirements in the event that the safety function of two relief valves become  !

inoperable. j Existing TS LC0 3.6.D.3 is being revised to clarify and state the shutdown ,

requirements when TS LCO 3.6.D.1 or 3.6.D.2 is not complied with.

Existing TS SR 4.6.D.1 is being revised to be more consistent with the j guidance provided in the Standard TS. The revision also clarifies the '

specific requirements for pressure testing, removal, and ' replacement of safety and relief valves.

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RTS-197A Attachment 1 to

?= NG-93-2104 Page 7 of 24 ';

i Existing TS SR 4.6.D.2 is being editorially- revised' by capitalizing *

"0PERATING CYCLE" since it is a defined term in the DAEC 15.

Existing ~ TS SR 4.6,0.3 is being editorially revised by capitalizing l "0PERATING CYCLE" since it is a defined term in the DAEC TS. In addition, t the footnote is being deleted as it is a superfluous statement. j Existing TS SR 4.6.D.4 is being revised editorially by replacing the word i

" required" with the word "specified". This change is consistent with the DAEC TS and the guidance provided by Standard TS.  ;

i The Bases Section 3.6.D & 4.6.0 have been revised to reflect the proposed  ;

changes. -j

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TS Section 3/4.6.E, Jet Pumps This section has been revised to clarify existing LCOs, SRs, . add specific i shutdown requirements, and to provide consistency within the DAEC TS as well as  !

with the guidance provided by the Standard TS. A summary of the proposed changes  ;

are as follows:

Existing TS' LC0 3.6.E.1 is being revised to refer to defined MODES of-operation. LC0.3.6.E.1 also contains a statement that, . if a specific. I surveillance cannot be' met, an additional surveillance is to be performed.  !

t within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This proposed Amendment relocates this information in.

its entirety to proposed SR 4.6.E.1.c. ~!

Existing TS LC0 3.6.E.1.a and 3.6.E.1.b have been revised and renumbered  !

to proposed LC0 3.6.E.1.a, 3.6.E.1.a.1, and 3.6.E.1.a.2.

These proposed changes are being made to provide clarity within the LCO. .j t

Proposed TS LC0 3.6.E.1.a.2 has been editorially revised. In addition, a i shutdown requirement has been proposed to be consistent with the guidance j provided by the Standard TS and to eliminate unnecessarily . cycling the l plant to the COLD SHUTDOWN condition as currently required in the DAEC TS.

Editorial changes are made in existing TS SR 4.6.E.1. - The word.  !

OPERABILITY is a defined term in the DAEC TS and is to appear in capital  :

letters . Instead of abbreviating recirculation, the proposed change - ,

spells it out.

Existing TS SR 4.6.E.1.a and 4.6.E.1.b have minor editorial changes made f as noted~in Attachment 2 providing consistency throughout the TS. 'i 1

Proposed - TS SR 4.6.E.1.c has been moved from existing LCO 3.6.E.1 as  !

discussed above. i e-s,- - ,y --y

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

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RIS-197A Attachment I to  ;

NG-93-2104 j Page 8 of 24 l l

Existing TS SR 4.6.E.2 has not been changed. j Existing TS SR 4.6.E.3 has been editorially changed for clarity. (

i Existing TS SR 4.6.E.4 has not been changed.

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TS Section 3/4.6.F, Jet Pump Flow Mismatch i This section has been revised to clarify existing LCOs, SRs, add specific j shutdown requirements -and to provide consistency within the DAEC TS .and the  ;

guidance provided by the Standard TS. A summary of the proposed changes are- as follows: l Existing TS LCO 3.6.F.1 has been divided into two itemized sections. 'l proposed as TS LCOs 3.6.F.1 and 3.6.F.2. Minor editorial changes were  !

made to each LCO in order to allow it to stand alone. These minor changes  ;

do not change the intent or requirements of the existing LCO. j Proposed TS LCOs 3.6.F.3 and 3.6.F.3.a have been added as clarification -

and for consistency with the guidance provided by the Standard TS. The addition of this LC0 allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for the recirculation pump speeds to be restored within the above limits. The current TS 'does not allow any time to restore the system to within the limits before taking further  ;

action. 'f 1

Existing TS LCO 3.6.F.2 was revised and renumbered as proposed LC0 l 3.6.F.3.b. <

i Existing SR 4.6.F.1 has been editorially revised to provide additional clarification and consistency by replacing the words " checked and logged" i with " verified."  !

Existing SR 4.6.F.2 has been editorially revised by changing the word  !

" Specification" to " Surveillance Requirement." The number referenced is  !

a Surveillance Requirement number and is identified accordingly. l TS Section 3/4.6.G, Structural Integrity  !

This section has been revised to clarify existing LCOs, SRs, and specif.ic shutdown requirements, add LCOs, and provide consistency with the DAEC TS and the j guidance provided by Standard TS. A summary of the proposed ~ changes are as l follows: j Existing TS .LC0 3.6.G.1 has been revised to provide clarity 'by i specifically identifying when structural integrity is required and to also -l 1

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RTS-197A Attachment 1 to l NG-93-2104 j Page 9 of 24 )

l correct the reference to ASME Section XI Code Cl ass . 1, 2, and 3 1 components.  !

Proposed TS LC0 3.6.G.2 has been added, providing specific actions for j Class 1 and Class 2 components when they do not conform to the ASME i Section XI requirements. This proposed change is a clarification in that  !

the existing TS does not provide specific actions if the Class 1 or Class  !

2 component does not meet TS LC0 3.6.G.I. The proposed wording is j consistent with the guidance provided in the Standard TS. i Proposed TS LCO 3.6.G.3 has been added, providing specific actions for l Class 3 components when they do not conform to the ASME Section XI l requirements. This proposed change is a clarification in that the existing TS does not provide specific actions if the Class 3 component does not meet 15 LCO 3.6.G.I. The proposed wording is consistent with the i guidance provided in the Standard TS. l t

Proposed TS LC0 3.6.G.4 has been added. The existing TS do' not include  !

actions to be taken in the event that a Class 1, 2, or 3 component (s) cannot meet the structural integrity requirements when above 212 F. l l

Existing TS SR 4.6.G.1 has been revised to include testing requirements of

Class 1, 2, and 3 pumps and valves. These requirements were incorporated

, into this SR from existing SR 4.6.G.2, which will be deleted.  ;

1 Existing TS SR 4.6.G.I.a is being deleted. The information contained in  !

the existing SR does not provide any guidance or verification of equipment j j OPERABILITY. This information is already included in the Bases Section. {

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, Exnting TS SR 4.6.G.2 is being deleted. The requirements for pump and  ;

valve testing are being relocated to TS LCO 3.6.G.1 above. I i

• Existing TS SR 4.6.G.2.a is being deleted. The information contained in  !

the existing SR does not provide any guidance or verification of equipment j OPERABILITY. This information is already included in the Bases Section. j Existing TS SR 4.6.G.3 has been revised and renumbered to proposed SR i 4.6.G.2. The word " augmented" replaced the word " inservice" which is a i more grammatically correct and accurate description of DAEC's program. l i

TS Section 3/4.6.H, Shock Suppressors (Snubbers) f This section has been revised to clarify existing - LCOs, SRs, and provide  ;

consistency within DAEC TS and the guidance provided by the Standard TS. A i summary of the proposed changes follows:  ;

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i RTS-197A Attachment 1 to

.NG-93-2104 Page 10 of 24 Existing TS LC0 3.6.H.1 is being revised to state the specific MODES and  :

conditions in which the LC0 is applicable, capitalize OPERABLE, and .j provide other editorial changes. l l

Existing TS LC0 3.6.H.2 is being editorially revised 'to correct the '!

Specification. number referenced and to abbreviate Limiting Conditions for, (

Operation (LCO) as it normally appears.

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l Add Table 4.6.H-1 to the existing TS. This Table is being added to '!

provide requirements for snubber visual inspection intervals for the j number of unacceptable snubbers. This revision is being made as a result '!

of NRC Generic Letter 90-09, " Alternate Requirements for Snubber Visual l Inspection Intervals and Corrective Actions." l i

Existing SR 4.6.H is being editorially revised by inserting the word {

" augmented" to clarify that the.DAEC is an augmented inspection program.

In addition, references to Surveillance Requirements 4.6.H.5 and 4.6.H.6 are being added.  ;

a Existing SR 4.6.H.1 for visual inspections.is being revised. The revision l is being made to ensure that the DAEC TS comply with NRC- Generic Letter.  !

90-09. 3 Existing SR 4.6.H.2 is being revised to conform to the guidance provided j by Standard TS. The language of the Standard TS is clearer and provides-  :

expanded and specific requirements _ for determining the next inspection l Interval for unacceptable snubbers. In addition,' the proposed SR requires j a review and evaluation be performed and documented to justify continued operation with an unacceptable snubber.

Proposed SR 4'.6.H.3, " Transient Event Inspection" i.s being added. The .;

existing SRs do not have this section. . The addition of this section i provides specific guidance in the event :that a' potentially ' damaging transient occurs. If one does occur, the new SR requires that a review of. j operational data or a visual inspection of' the system (s)- be performed. -j The addition of this SR is consistent with the guidance provided by the i Standard TS.

Existing SR 4.6.H.3 is being changed to proposed SR 4.6.H.4. In addition, "0PERATING CYCLE" is being changed to .all caps. OPERATING CYCLE is a defined term and is to appear in all caps. The word " Specification" is  ;

being changed to " Surveillance Requirements." - The. referenced numbers in.  !

this SR are being renumbered due to the addition of new SRs. The footnote is also being deleted. It contains superfluous information that is.not j needed to perform any SR or LCO action. [

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RTS-197A Attachment.1 to [

NG-93-2104  ;

Page 11 of 24 ,

Existing SR 4.6.H.4 is being changed to proposed SR 4.6.H.5. In addition there were some editorial changes made. This makes the proposed SR 4.6.H.5 consistent with proposed SR 4.6.H.6.

• Existing SR 4.6.H.5 is being changed to proposed SR 4.6.H.6. This is an editorial change. j Add SR 4.6.H.7, " Functional Testing of Repaired and Replaced Snubbers."  !

The addition of this SR ensures that repaired or replacement snubbers meet the functional test criteria before installation in the unit. ,

i Add SR 4.6.H.8, " Snubber Service Life Replacement Program." This SR l ensures that the service life of the snubbers is monitored, ensuring that the service life is not exceeded between surveillance inspections.

Existing SR 4.6.H.6 is being deleted. The intent of the requirements for ,

this surveillance are incorporated into proposed SR 4.6.H.7 and ' SR l 4.6.H.8.

The Bases Section 3.6.H & 4.6.H have been revised to reflect the proposed-changes. l Basis for proposed no significant hazards consideration determination: f The Commission has provided standards (10CFR50.92 (c)) for determining whether a_ significant hazards consideration exists. A proposed amendment to the '

facility's operating license involves no significant. hazards consideration.if-operation of the facility in accordance with the~ proposed amendment would not (1) involve a significant increase in the probability or consequences of an accident .q previously evaluated; (2) create the possibility of a new or different kind o_f accident from any accident previously evaluated; or (3) involve a significant'-

reduction in a margin of safety.

THERMAL AND PRESSURIZATION LIMITATIONS All components of- the Reactor Coolant System (RCS) are designed to -

withstand effects of cyclic loads due to system Temperature / Pressure (T/P) ,

changes. These loads are introduced by startup (heatup) and shutdown; t (cooldown) operations, . power transients, and reactor trips. The LC0 -

limits the T/P changes during heatup and cooldown, within the' design ,

assumptions and the stress limits for cyclic operation. .

Figure 5.3-1 of the DAEC UFSAR, shows three operating limit curves,  :

s

-n , , -- , ,

RTS-197A Attachment 1 to NG-93-2104 Page 12 of 24 including ir :* tion shif t of the core beltline region curves to their position at - life (32 full power years). The three curves represent three speci' .onditions: a) system hydrostatic and leakage tests, b) non-nuclear

• atup or cooldown and low level physics tests, and c) core critical operation. The t.urves were established by requirements of Section III, Appendix G, of the ASME Code and by 10CFR50, Appendix G.

Each T/P limit curve defines an acceptable region for normal operation.

The usual use of the curves is operational guidance during heatup or.

cooldown maneuvering, when T/P indications are monitored and compared to -

the applicable curve to determine that operation is within the allowable region.

The LCO establishes operating limits that provide a margin to brittle failure of the reactor vessel and piping of the Reactor Coolant Pressure Boundary (RCPB). The vessel is the component most subject to brittle failure. Therefore, the LCO limits apply mainly to the vessel.

The T/P limit curves are composite curves established by superimposing limits derived from stress analyses of those portions of the -reactor vessel and head that are the most restrictive. Across the span of the T/P limit curves, different locations are more restrictive, and thus the curves are composites of the most restrictive regions.

The heatup curve represents a different set of restrictions than the cooidown curve because the directions of the thermal gradients through the vessel wall are reversed. The thermal gradient reversal alters the location of the tensile stress between the outer and inner wall.  !

I A possible consequence of ' violating the LCO limits is that the RCS is {

operated under conditions that could have resulted in brittle failure of  !

the RCPB, possibly leading to a non-isolable leak or loss-of-coolant  ;

, accident. In the event t,ase limits are exceeded, an evaluation must be i performed to determine the effect on the structural integrity of the RCPB l components. ASME Code,Section XI, Appendix E provides a recommended j methodology for evaluating an operating event that causes an excursion outside the limits.  ;

Comparison of the pressure versus temperature limits in UFSAR Figure 5.3-1 .

with intended normal and upset operating conditions, shows that the limits  !

will not be exceeded during any foreseeable operating condition. Reactor operating procedures have been established such that actual transients will not be more severe than those for which the vessel was designed. Of i

~

the design transients, an upset condition produces the most adverse temperature and pressure condition with a minimum fluid temperature of i

i l

'I RTS-197A Attachment 1 to l NG-93-2104 l Page 13 of 24 1 250 F and a maximum pressure peak of 1180 psig. Scram automatically ,

occurs with initiation of this event, prior to the reduction in the fluid 3 temperature, so the applicable operating limits are given by UFSAR Figure  !

5.3-1 curves A and A'. For a temperature of 250 F, the maximum allowable l pressure at end of life exceeds 1400 psig for the intended margin against  ;

nonductile failure. The maximum transient pressure of 1180 psig is j therefore within the specified allowable limits. i The average rate of reactor coolant temperature change during normal l heatup and cooldown is limited by operating procedures to 100 F in any one j hour period. During emergency and faulted conditions, the cooling rates  ;

may exceed this value os o result of rapid blewdown due to postulated  ;

valve malfunction or rupture accidents. The operator can compare the i actual heatup and cooldown thermal and pressure cycle history for any }

given period of actual plant operating time with the reactor vessel cyclic l design bases. This comparison will give, at any desired time, the status  !

of actual vessel cyclic history and design cyclic requirements. l i

The revision discussed .above is editorial in nature. The existing information does not provide the control room operator with any prudent i action or guidance in the operation of the plant, mitigation of any j accident, nor does it affect any procedural steps in the Emergency j Operating Procedures. The proposed revision will not result in any loss  :

of regulatory control since DAEC still meets the requirements specified in  !

10CFR50, Appendix H. ,

i The proposed LCOs and SRs provide additional guidance, clarification, and l consistency within the DAEC TS as well as utilizing the guidance provided by the Standard TS. The existing DAEC TS do not provide specific actions ,

in the event the temperature / pressure limits are exceeded. The proposed {

LCO would allow 30 minutes to restore temperature / pressure limits. Once j restored, an engineering evaluation is to be performed to determine any l

. effects of the out-of-limit condition on the structural integrity of the  !

RCS. If no effects are identified, operation is continued. If any of the l above actions cannot be complied with, a reactor shutdown is initiated. l Most violations of the temperature / pressure limits will not be severe, and  !

the activity can be accomplished in a controlled manner. Besides  !

restoring operation'within limits, an evaluation is required to determine  ;

if RCS operation can continue. The evaluation must verify the RCPB l integrity remains acceptable and must be completed before continuing i operation. Several methods may be used, including comparison with pre-  !

analyzed transients in the stress analyses, new analyses, or inspection of l the components. [

Detailed stress analyses have been made on. the reactor vessel for both L

P i

f

+

RTS-197A Attachment 1 to NG-93-2104 Page 14 of 24 steady-state and transient conditions with respect to material fatigue.

The results of these transients are compared to allowable stress limits.

Requiring the coolant temperature in an idle recirculation loop to be within 50 F of the operating loop temperature before a recirculation pump is started ensures that the changes in coolant temperature at the reactor vessel nozzles and bottom head region are acceptable.

Heating and cooling transients throughout plant life at uniform rates of 100 F/hr were considered in the temperature range of 100"F to 546 F and were shown to be within the requirements for stress intensity and fatigue limits of Section III of the ASME Code (1971 Edition including Summer 1972 Addenda).

The coolant in the bottom of the vessel is at a lower temperature than that in the upper regions of the vessel when there is no recirculation flow. This colder water is forced up when recirculation pumps are started. This will not result in stresses that exceed ASME Code, Section 111 limits when the temperature differential is not greater than 145 F.

The minimum temperature of the fluid retained by a component can be used as a conservative estimate of metal temperature in evaluating the margin from the temperature at which the NDT properties were measured.

Additional margin can usually~be shown by calculating the temperature of the metal for the condition and area of concern.

The addition of the SR implementing Generic Letter 91-01," REMOVAL 0F THE SCHEDULE FOR THE WITHDRAWAL OF REACTOR VESSEL MATERIAL SPECIMENS FROM TECHNICAL SPECIFICATIONS" is within the requirements, approval, and guidance provided by the NRC. The addition of this SR does not involve a significant increase in the probability or consequences of an accident previously evaluated. This statement is based on the fact that the regulatory requirement of 10CFR50, Appendix H will remain in effect in the TS. Therefore, removal of any references to the specimen withdrawal will not result in any loss of regulatory control since any changes to this schedule are controlled by the requirements of 10CFR50 Appendix H.

Based on the addition of the previously mentioned SRs, several of the existing SRs are either being revised or deleted. These proposed changes are considered to be editorial in nature. These proposed changes are being made based on applying human factors concepts to minimize the potential for confusion, provide additional guidance not specifically

~

provided in the previous TS, and provide consistency within the DAEC TS and the guidance as provided by the Standard TS. The transient associated with recirculation pump startup is a nonlimiting event based on:

l RTS-197A Attachment 1 to i NG-93-2104' l Page 15 of 24 i

1. Generdl Electric Standard Application for Reactor Fuel-United  !

States Supplement, NED0-24011-P-A-US, and  :

.i

2. Duane Arnold Energy Center Single-Loop Operation, NEDD-24272. j The proposed changes greatly enhance the safety significance of the I existing TS. These changes do not impact the safety analysis 'or any j calculations or parameters utilized in the licensing bases for DAEC. In  :

addition, the proposed changes do not relax any NRC regulations as j contained in 10CFR50. Therefore, the proposed changes do not involve a- l significant increase in the prc,oability or consequences of an accident i previously evaluated.  !

COOLANT CHEMISTRY ,

The LCOs and SRs have been revised by placing the information in proposed j Tables or by formatting in accordance with the guidance provided by the Standard TS. These revisions improve clarity, are consistent with the.

current industry practices, and provide additional guidance not. i specifically stated in the existing DAEC TS. The proposed changes do not change any safety . analysis, parameters used in developing the safety .  !

analysis or any intended function for any_ safety related equipment.

• I t

COOLANT LEAKAGE i Reliable means are provided to detect leakage from the- nuclear system barrier inside the drywell. . Limits' are established for nuclear system leakage rates so that appr'opriate action can be taken before the integrity ,

of the nuclear system process barrier is unduly compromised. l 1

The DAEC design includes a- nucl. ear system leak detection, . isolation, .

processing, and makeup system. This system provides. for leakage control capability. The capability of this system is discussed in Section 3.6.C  !

of this submittal.

?

The functions of the leak detection system are accomplished under normal [

operation or postaccident conditions so that normal (10CFR20) or accident (10CFR100) offsite dose limits do'not exceed established values'and in.a-manner in which the core and the containment cooling continuity are not impaired.

The leakage considered here is limited to that water. or steam released ,

from the nuclear system process barrier inside the primary containment

.- (d rywell) . Leakage inside the drywell 'is-treated separately from; leakage  ;

elsewhere in the plant because the drywell contains a high concentration I

~

'wp- a p .i,, A M $

I RTS-197A Attachment I to ,

NG-93-2104 Page 16 of 24 of nuclear system piping and is totally inaccessible during reactor l operation.

If a leak occurs, the drywell will contain the released matter that will i be present in the liquid, gaseous, and vapor phases. This will result in -

the collection of water in the sumps, a possible increase in drywell temperature, pressure, and relative humidity, an increase in the air-conditioning heat load, and an increase in the radioactivity of the drywell atmosphere. The closed limited volume of the drywell enhancas the.

detection sensitivity.

• There are 6 different methods .used to detect leakage in the primary f containment. These are outlined in the Section 3.6.C discussion of this

submittal. The different drywell parameters provide diverse methods for l determining if an increased leak rate exists within the drywell. The  !

allowable leakage rates have been based on the predicted and  ;

experimentally determined behavior of cracks in pipes, the ability to make i up coolant system leakage, the normally expected background leakage due to equipment design, and the detection capability of the various drywell monitors. I i

i Based on the behavior of cracks, a 5 gpm leak rate limit has been assigned  !

to UNIDENTIFIED LEAKAGE and 20 gpm to IDENTIFIED LEAKAGE totaling a 25 gpm l TOTAL LEAKAGE. Experience has shown that the normal leak rate is 4 gpm  :

into the equipment drain sump and 0 to 0.5 gpm into the floor drain' sump. j 1

i The sump working capacities and sump discharge capacities are large enough  ;

to accept the design leak rates. The sump working capacity is the amount  !

of water between the low level pump trip and the high-high-level alarm _  !

point. The equipment drain sump (approximate working capacity, 450 gallons) and the floor drain sump (approximate working capacity, 225 gallons) are drained by two 50 gpm pumps each. This pump capacity permits ,

one pump in each sump to remove the design total leakage because of the -

possibility that most of-the leakage could flow into one sump.  :

The criterion for establishing the total leakage rate limit is based on-the makeup capability of the CRD and RCIC Systems and is independent of l the Feedwater System, normal ac power, and the emergency core cooling  !

3 systems. The CRD System supplies 42' gpm into the. reactor vessel. - The ]

RCIC System can supply 425 gpm through the feedwater sparger to the j reactor vessel. The total leakage rate limit is set at less than 0.1 of this value, or 25 gpm.

The proposed changes revise a shutdown requirement when the Reactor Coolant Leakage exceeds the LCO limits, by allowing a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period to 1

-. . _ , _ - . . - . _- ._ _ ___ _ ___ _ ___ _ _ - _ _ _b

i i

RTS-197A Attachment 1 to I NG-93-2104 ,

Page 17 of 24 l t

bring the leakage back within limits. If the leakage cannot be reduced  !

within the required time, a shutdown requirement to HOT SHUTDOWN and  ;

eventually COLD SHUTDOWN is initiated. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period is a justified  !

and accepted time frame by the NRC. The 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> is adequate time to allow  !

the leakage to be reduced and brought into. compliance with the limits  !

specified in the TS. The probability of an accident exceeding the safety j analysis during this 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> period is minimal and therefore does not i increase the probability or consequences of an accident. The existing TS l require the reactor to be placed in COLD SHUTDOWN when the leakage limits (

are exceeded. The proposed shutdown requirement is consistent with the  ;

other shutdown requirements within the DAEC TS and the guidance provided l by the Standard TS.

l The addition of two LCOs provides specific guidance in the event the Sump l System and/or Air Monitoring System are inoperable. The proposed changes i are consistent with current plant practices, however, the wording of the j existing LCOs is somewhat confusing. Applying human f actors concepts, the  !

existing LCOs have been revised to provide clarity and avoid potential i confusion. The SRs have been revised to incorporate the changes made to  !

the associated LCOs. These proposed changes do not affect the assumptions ,

utilized to support the plant safety analysis or change any mitigation- l factors which have been credited in the DAEC licensing bases. l i

Other proposed changes, basically editorial, are also being made to  !

4 provide consistency and clarity within the DAEC TS and to utilize the {

4 guidance provided by the Standard TS. Consequently, these changes do not  !

involve a significant increase in the probability or consequences of an  ;

accident previously evaluated.

. SAFETY AND RELIEF VALVES  !

The nuclear system pressure relief system includes two safety and six l safety / relief valves located on the main steam lines within the drywell l

, between the reactor vessel and the first isolation valve. The safety  !

! valves provide protection against the overpressure of the nuclear system i 3

and discharge directly to the interior space of the drywell. The  !

safety / relief valves, which discharge to the suppression pool, provide the  !

following three main functions-

.i

1. Overpressure relief operation. The valves are opened to limit the pressure rise and prevent spring safety valve opening. )

i

2. Overpressure safety operation. The valves augment the spring safety valves by opening. in order to prevent nuclear system overpressurization.

J

• . - -, ,,,,m, =,- w - y $

i l

i i

RTS-197A Attachment 1 to- l NG-93-2104 l Page 18 of 24 }

t

3. Depressurization operation. The required valves are opened  !

automatically or manually by indirectly operated devices as part of l the protection system for small line breaks.

The main steam lines, in which the safety / relief and safety valves are ,

installed, are designed, installed and tested in accordance with the l applicable codes as discussed in the DAEC UFSAR Section 3.2. [

t The operational objective of the nuclear system pressure relief system is  !

to prevent the opening of the spring-loaded safety valves during normal i plant isolations and load rejections. The safety design bases are as l follows: -;

~

1. The nuclear system pressure relief system prevents the  ;

overpressurization of the nuclear system to prevent the failure of  !

the nuclear system process barrier because of pressure.

2. The nuclear system pressure relief system provides automatic nuclear l system depressurization for small breaks in the nuclear system  !

occurring with maloperation of the HPCI System so that the LPCI and i the Core Spray Systems operate to protect the fuel barrier. l i

3. The safety / relief valve discharge piping is designed to accommodate j forces resulting from relief action and is supported for reactions i due to flow at maximum relief valve discharge capacity so that system integrity is maintained.

4. The nuclear system pressure relief system is designed for testing f prior to nuclear system operation and for periodic verification of  !

the operability of the nuclear system pressure relief system.  !

During power generation, the design bases are as follows: l

1. The nuclear system safety / relief valves prevent the opening of the  !

spring-loaded safety valves during normal plant isolations and load 1 rejections.  !

i

2. The nuclear system safety / relief. valves discharge to the suppression  ;

pool below the water level to condense the exhaust steam.

{

3. The safety / relief valves will properly reclose following a plant  !

isolation or load rejection so that normal operations can be resumed i as soon as possible. i i

The ASME Code requires overpressurization protection for vessels designed j to meet Code Section III. The code _ permits a' peak allowable pressure of i

.,--g- m o e-wew- ~w ---v

RTS-197A Attachment 1 to NG-93-2104  :

Page 19 of 24 l 110% of vessel design pressure (1375 psig for a 1250 psig vessel). The i tode specifications for safety valves additionally require that the lowest }

safety valve setpoint be at or below vessel design pressure (1250 psig)  ;

and the highest valve setpoint be at or below 105% of vessel design  !

pressure (1313 psig). The safety / relief valves are set to open by self- l actuation (overpressure safety function) in the range from 1110 to'1140 l psig, and the safety valves are set to operate at 1240 psig. These '

settings satisfy the ASME Code specifications for the setpoints of the j safety valves.  ;

For DAEC, the transient produced by the closure of all main steam line I isolation valves represents the most severe abnonnal operational transient f resulting in a nuclear system pressure rise when direct scrams are  ;

ignored. The plant is assumed to be operating at the turbine generator i conditions at a maximum vessel dome pressure of 1025 psig. The analysis hypothetically assumed the failure of the direct isolation valve position scram. The reactor is shut down by the backup, indirect, high-neutron- l flux scram. For the analysis, the self-actuated setpoints (safety function) of the safety / relief valves are assumed to be in the range from ,

1121 to 1151 psig, and the safety valves are assumed to operate at 1252 l psig (setpoint +1%). The safety / relief and safety valves open to limit  !

the nuclear system pressure rise to 1275 psig. The analysis indicates that the design valve capacities are capable of maintaining adequate _  ;

margin (100 psi). below the peak ASME Code allowable pressure in the l nuclear system (1375 psig). The safety valve capacity in conjunction with j safety / relief valve capacity limits the peak nuclear system pressure at l the bottom of the vessel. The resulting criterion to the ASME Code limit ,

ensures adequate protection against excessive overpressurization for the nuclear system process barrier even for this hypothetical reactor isolation event. l The analysis that forms the basis for the evaluation of the pressure. I relief function of the nuclear system pressure relief system appears in  ;

i the DAEC UFSAR Chapter 15. In summary, the opening of a relief valve or safety valve allows steam to be discharged into the primary containment.

The sudden increase in the rate of steam flow reaching the reactor vessel  :

! causes the reactor vessel coolant inventory to decrease. The result is a l mild depressurization transient. l l

The small amounts of radioactivity discharged with the steam are contained j inside the primary containment; the situation is not significantly i different, from a radiological viewpoint, than that encountered in cooling [

the plant using the relief valves to remove decay heat. This transient is i

! a nonlimiting event (DAEC UFSAR reference 2 of Section 15.0); accordingly, only the foregoing narrative description of the event is provided. l l

- - - - - - --- w - w m -- - - ei

i I

1 RTS-197A Attachment 1 to l NG-93-2104 Page 20 of 24 As seen by the above discussion, none of the proposed changes deviate from the current safety analysis. The existing LCOs have been revised to l include clear, concise, and specific shutdown requirements. In addition,  !

a footnote has been added to provide a reference that some of the relief  !

valves also perform ADS functions. These proposed changes provide i consistency within the DAEC TS and utilize the guidance provided in the l Standard TS. In addition, human factors concepts have been applied to i these LCOs in order to minimize potential confusion.  ;

The existing SR, requiring that 1 safety and 3 relief valves be checked l per OPERATING CYCLE, has been revised incorporating the wording provided j by the Standard TS. The proposed SR provides specific and detailed I requirements ensuring that the safety and relief valves are tested in ,

accordance with manufacturer's recommendations. The existing SR does not l contain this level of detail. The proposed changes are an enhancement. l These changes, along with the editorial changes, provide additional .,

guidance not specifically provided in the existing TS. The proposed  !

changes do not change any parameters or calculations used in the current i safety analysis.  ;

These changes provide additional assurance that the safety and relief j valves will perform their intended function as described above.  ;

I JET PUMPS DAEC has two external recirculation loops each discharging high pressure I flow into an external manifold from which individual recirculation inlet {

lines are routed to the jet pump risers within the reactor vessel. The j remaining portion of the coolant mixture in the annulus becomes the driven t flow for the jet pumps. This flow enters the jet pumps at the suction  !

inlet and is accelerated by the driving flow. The driving and driven i flows are mixed in the jet pump throat section resulting in partial  ;

pressure recovery. The balance of recovery is obtained in the jet pump j diffusing section as referenced in DAEC UFSAR Section 5.4. The adequacy of the total flow to the core is discussed in DAEC UFSAR Section 4.4. Jet i pump operating experience has shown that the design is sound and that the -

j jet pump operation is stable and predictable.  ;

From a safety analysis standpoint, there is no specific jet pump analysis. l Any jet pump event is enveloped into the safety analysis of _the l recirculation system. The only other event involving the jet pumps is j the cracking of the hold down beams. This. issue was brought before the  !

industry in the NRC issued Bulletin 80-07. This bulletin required, in I addition to performing examinations of these hold down beams, that a surveillance program to monitor jet pump performance be initiated and l i

1

RTS-197A Attachment 1 to NG-93-2104 .

Page 21 of 24 continued until plant TS could be changed. Iowa Electric initiated and l performed the required monitoring until NRC approved and issued TS  ;

Amendment 158, dated April 28, 1989. This approved Amendment was developed using the guidance provided by General Electric issued SIL No. .

33, " Jet Pump Beam Cracks." This SIL discusses the jet pump beam failure problem and provides recommendations for modifications to the TS in order to improve detection of any impending failure of these beams.

The proposed changes to this section do not change any of the surveillance requirements approved by the NRC with regard to this concern.  ;

i The existing TS requires that if an engineering evaluation determines that  :

a jet pump is inoperable, the reactor must be brought to COLD SHUTDOWN.

The proposed TS would require the reactor to be brought to HOT SHUTDOWN l instead of COLD SHUTDOWN. The intent of the TS action / shutdown statements is that if a system is inoperable and a MODE change is required, that the  ;

reactor be placed in a MODE to which the LC0 does not apply. The TS ,

require jet pumps to be OPERABLE in the RUN and STARTUP MODES. Therefore,  ;

HOT SHUTDOWN would be the first MODE in which the Jet Pump is not required to be OPERABLE. Making this change eliminates the requirement to cycle the plant to colder temperature conditions than what is actually needed, t thus not adding any unnecessary thermal stresses to the system. This l proposed change does not alter the safety analysis, calculations, or  ;

parameters used to support the licensing bases of DAEC.

In addition, several editorial changes are being proposed to provide  :

additional clarification and consistency with current plant practices. ,

These changes do not change the intent of the existing TS and therefore, .j do not involve a significant increase in the probability or consequences [

of an accident previously evaluated.

JET PUMP FLOW MISMATCH 5 Reference the above section, " Jet Pumps" for additional discussion on .

safety analysis.

i The existing TS requires that if recirculation pump speed is outside its  !

limits, the pump must be tripped and single loop operation (SLO) entered.  !

The proposed TS allows 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to bring the recirculation pump back to ,

within its limits prior to initiating SLO. The 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> is based on the  ;

low probability of an accident occurring during this time period, on a  !

reasonable time to complete the action, and on the frequent core '

monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected. In addition, DAEC has been analyzed for SLO in NEDD- t 24272 dated July 1980, with SLO approved by TS Amendment 119 dated May. i t

,.,.s _., -

, - . _ _ y

u ..-.-.a

(

a RTS-197A Attachment 1 to NG-93-2104 _ v Page 22 of 24 l l

1985. The proposed change is more conservative and is encompassed within t the SLO envelope. This change does not result in deviating or departing

• from any existing calculations, safety analysis, or accident mitigation  ;

actions, and is consistent with the guidance provided by the Standard TS.  !

Editorial changes have been proposed to provide clarity, consistency, and additional guidance not included in the existing TS. I i

a  !

STRUCTURAL INTEGRITY  :

Three LCOs were added to this TS stating the thermal and pressurization  !

limits and actions to be taken if those limits for any ASME -Section XI Class 1, 2, or 3 component (s) are exceeded. In the event one of the i subject component (s) cannot be restored to within these limits, an action

• statement allows isolation of the component (s) with specific restrictions. i stated in each separate LCO. The existing TS do not contain these LCOs. i This change will enhance the existing TS and ensure that the Class 1, 2, ,

and 3 component (s) perform in accordance with the ASME Boiler and Pressure ~ i Vessel Code and applicable addenda as required by 10CFR50.55a(g), except  ;

where specific written relief has been granted by the NRC. _The proposed changes do not alter programs previously approved by _ the NRC,  ;

calculations, nor safety analysis utilized by DAEC. The proposed action  ;

statements are consistent with the guidance provided in the Standard TS.

The two SRs being deleted are not actually.SRs. They contain information j which is located in the Bases. .The deletion of these SRs will not  ;

eliminate any testing or verification of system OPERABILITY requirements or-actions. These proposed changes do not degrade the' intent of the. ,

structural integrity TS, any safety related ~ components, or alter any -

intended safety functions of any equipment. i The editorial changes are intended to enhance the TS by providing clarity i and casistency. The guidance provided by the Standard TS was utilized in i making these proposed changes. These changes comply with the guidance and l examples provided in 51FR7751 dated March 6,1986, that are considered not l likely to involve a significant hazards consideration. i e

SHOCK SUPPRESSORS (SNUBBERS) i As determined by the Staff, the alternate schedule for visual inspections ,

proposed in Table 4.6.H-1' provides the same level . of confidence as the j existing schedule. The actions required by the existing TS as a result of -

finding snubbers inoperable remain the same.

Several LCOs and SRs have been revised. The intent of these revisions is l Y ret - e. - -- - - --

r i

i RTS-197A Attachment I to l NG-93-2104  ;

Page 23 of 24 {

to provide clarity and consistency throughout the TS section. These changes are editorial in nature as discussed in 51FR7751 dated ~ March 6, j 1986. These changes utilize the guidance provided in the Standard TS for i the shutdown requirements. These changes do not alter DAEC TS for safety ,

functions, plant equipment, or calculations in the safety analysis. j This proposed Amendment also adds SRs for Transient Event Inspections, .;

functional Testing of Repaired and Replaced Snubbers, and Snubber Service  !

Life Replacement Program. The existing TS do not contain these subject  :

SRs. The proposed SRs will provide assurance that the snubbers perform l their intended function. The addition of these SRs will also ensure that the snubbers will assist the associated supported system in performing its intended function as required in the DAEC FSAR.

i

1. The proposed changes do not involve a significant increase in the i probability or consequences of an accident previously evaluated. l l

The proposed changes discussed in this section are provided to enhance the l overall quality and safety significance of the existing DAEC TS. The proposed TS do not change any accident analysis, plant safety analysis, l calculations, degrade existing plant programs, modify any functions of i

~

safety related systems, or accident mitigation functions DAEC has l previously been credited with. Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an j accident previously analyzed. j

• I The proposed changes to the Bases Section 3.6 and 4.6 reflect the above  !

changes and include various editorial corrections. These changes have no  !

effect on the consequences of a previously evaluated accident. l

2. The proposed changes do not create the possibility of a new or different kind of accident from any accident previously evaluated.  ;

l The proposed changes do not alter any plant parameters, revise any safety {

limit setpoint, or provide any new release. pathways. In addition, the i proposed changes do not modify the operation or function of any safety j related equipment, nor do they introduce any new modes of operation,- l failure modes, or physical changes to the plant. The proposed changes do  !

not change any plant parameters or transient responses assumed in the Design Bases of the plant and therefore, do not create the possibility of a new or different kind of accident from any accident previously_  !

evaluated.

The proposed changes to the Bases Section 3.6 and 4.6 reflect the above  ;

changes and include various editorial corrections. Therefore the proposed  ;

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RTS-197A Attachment 1 to '

NG-93-2104 Page 24 of 24 changes and corrections do not create the possibility of a ' new or  !

different kind of accident from any accident previously evaluated.  ;

3. The proposed changes do not involve a significant reduction in a margin of i safety.

The proposed changes do not require any modifications to existing plant systems or equipment, Emergency Operating Procedures, safety limit i settings, or parameters utilized in the licensing bases for the! safety l analysis. These proposed changes are being made to enhance TS Section 3.6 1 by clarifying and making LCOs and SRs consistent throughout-the section. ,

In addition, several LCOs and SRs have been added, providing additional information that did not exist in the current TS. As discussed above, the i proposed changes do not change any safety analysis or any accident mitigation actions for which DAEC has previously taken credit. Therefore, j the proposed changes do not involve a significant reduction in a margin of

• safety.  !

The proposed changes to the Bases Section. 3.6 and 4.6 reflect the above ,

changes and include various editorial corrections. These changes.do not involve a significant reduction in a margin of safety.

In conclusion, the Commission has provided guidance concerning the application of standards in 10CFR50.92 by providing certain examples i (March 6,1986, 51FR7751) of amendments that are considered not- likely to-involve a significant hazards consideration. Although the majority of the  ;

proposed changes are directly comparable to the examples, several other proposed changes are not; however, as stated above, the changes do not involve a significant hazards consideration.  :

Local Public Document Room Location: Cedar Rapids Public Library, 500' First ,

Street SE, Cedar Rapids, Iowa 52401 Attorney for Licensee: Jack Newman,-Kathleen H. Shea, Newman and Holtzinger,- i 1615 L Street NW, Washington, D.C. 20036 l l

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