2CAN079803, Forwards Changes to Unit 2 TS Bases.Attachment to Ltr Contains Brief Description of Each One of Associated TS Bases Changes

Forwards Changes to Unit 2 TS Bases.Attachment to Ltr Contains Brief Description of Each One of Associated TS Bases Changes

ML20236Q274
Person / Time
Site:	Arkansas Nuclear
Issue date:	07/13/1998
From:	James D ENTERGY OPERATIONS, INC.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20236Q277	List: 2CAN079803, Forwards Changes to Unit 2 TS Bases.Attachment to Ltr Contains Brief Description of Each One of Associated TS Bases Changes ML20236Q278
References
2CAN079803, 2CAN79803, NUDOCS 9807200129
Download: ML20236Q274 (5)

Text

.

. o.

O Entergy Operations,Inc.

1448 S.R. 333

( __

RussoMe, AR 72801

. Tel501858 5000 j July 13,1998 l

L 2CAN079803 - 1 l

l' U. S. Nuclear Regulatory Conunission . <

I l De_imaat Control Desk l Mail Station OPI-17 l Washington, DC 20555

Subject:

Arkansas Nuclear One- Unit 2 Docket No. 50-368 License No. NPF-6 Technical SpeciScation Bases Changes i

~ Gentlemen:

Attached are changes to the Arkansas Nuclear One - Unit 2 (ANO-2) Technical Specification Bases. The Attachment to this letter contains a brief description of each one of the associated Technical Specification Bases changes.

As these changes modify the Bases of the Technical Specifications, a "No Significant Hazards Determination" is not required. A review of the changes in accordance with 10 CFR 50.59 has concluded that an unreviewed safety question is not involved as a result of these changes.

We request that these changes be issued with the next ANO-2 Technical Specification amendment. Should you have any questions regarding these changes, please contact me or a member ofmy staff.  ;

1 Very truly yours, j W

ale E. J s f Actingpir or, Nuclear Safety off, i DEJ/rd k Attachment 9807200129 990713 PDR ADOCK 05000368 P PDR

U. S. NRC July 13,1998 2CAN079803 Page 2 l

! cc: ' Mr. Ellis W. Merschoff l Re8i onal Administrator U. S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-8064 NRC Senior Resident Inspector Arkansas Nuclear One  !

P.O. Box 310 i London, AR 72847 )

Mr. William D. Reckley NRR Project Manager Region IV/ANO-1 & 2 U. S. Nuclear Regulatory Commission NRR Mail Stop 13-H-3 One White Flint North 11555 Rockville Pike l Rockville, MD 20852

)

l t

i I

l' Attachment to 2CAN079803 Page1 of3

, The following Arkansas Nuclear One - Unit 2 (ANO-2) Technical Specification (TS) Bases

' char %es have been reviewed in accordance with 10 CFR 50.59. Each of the associated j evaluations concluded that an unreviewed safety question is not involved as a result of the associated changes. A brief description of each of the associated changes is listed below.

TS Bases 2.1.2 (name B 2-2) t The proposed change to the ANO-2 TS Bases 2.1.2 (page B 2-2), Reactor System Coolant l Pressure, adds a footnote to the RCS design codes that allow use oflater ASME Section III l Codes as long as the code section is reconciled. This change has been evaluated under 10 CFR 50.59 with the resulting conclusion that an unreviewed safety question is not l involved.

f l The proposed changes are necessary to properly reflect the actual design qualifications that f

were necessary for many ANO systems. Since construction, many systems have been repaired under the ASME Section XI program The ANO-2 repair / replacement program has been updated to the 1992 code edition. Per the 1992 edition of the ASME Section XI IWA-4170, titled " CODE APPLICABILITY," in subsection (d) it states the following:

l (d) " Alternatively, an item to be used for replecament may meet all or portions of the

requirements oflater editions and addenda of the Construction Code or Section III when the Construction Code was not Section III, provided that the following requirements are met.

l (1) The requirements aff'ecting the design, fabrication, and examination of the item to be l

used for replecamant are reconciled with the Owner's specification through the Stress j Analysis Report, Design Report, or other suitable method that demonstrates the item j is satisfactory for the speci6ed design and operating conditions.

t 4 l (2) Mechanical interfaces,6ts, and tolerances that provide satisfactory performance are compatible with system and component requirements.

(3) Materials are compatible with installation and system requirements."

ANO has a Design Engineering Standard that satis 6es the above ASME Section XI IWA-4170 requirements.

l l TS Bases 2.1.1 (name B 2-4)

A proposed change has been made to TS Bases 2.2.1 (page B 2-4), located under the section  ;

titled Steam Generator Pressure-Low. The proposed change removes the specific value listed l for the full load steam generator (S/G) pressure. This change has been evaluated under j 10 CFR 50.59 with the resulting conclusion that an unreviewed safety question is not involved.

Attachment 12 ,

2CAN079803 Page 2 cf 3 The Bases for this speci6 cation indicates that the full load operating po:nt for S/G pressure is approximately 900 psia. The actual value of S/G pressure varies from cycle to cycle depending on several variables including the Reactor Coolant System (RCS) hot leg temperature and S/G tube plugging In an attempt to inhibit the S/G secondary side stress corrosion cracking, one of the measures taken at ANO-2 was a reduction in the RCS. hot leg temperature. This reduction in the RCS hot leg tory-dure resulted in a reduced full load S/G secondary side pressure.

ANO-:. TS 3.4.5 requires periodic iaW ons i of the S/G tubing and the repair or removal of tubes from service when they meet the repair criteria. A reduction in the S/G heat transfer surface area occurs as each tube is removed from service. The reduced heat transfer surface area has produced a reduction in the fullload S/G secondary side pressure. The Bases for this specification will continue to require an adequate operating margin between the full load operating S/G pressure and the S/G Pressure-Low setpoint. This requirement will remain regardien of the actual value of the full load operating S/G pressure.

TS Bases 2.2.1 (pane B 2-7)

The TS Bases for TS 2.2.1 (page B 2-7) was modi 6ed under the DNBR-Low section to allow j the Integrated Radial Peaking Factor-High auxiliary trip setpoint to be raised from "s 4.28" to {

"s 7.00". This change has been evaluated under 10 CFR 50.59 with the resulting conclusion that an unreviewed safety question is not involved.

At power levels below approximately 20%, the core protection calculators (CPCs) use a fixed axial shape. The core average axial shape is multiplied by the planar radial peaking factors for each plane to obtain the hot pin power distribution. The average of the hot pin power distribution is multiplied by the tilt to obtain the integrated one-pin radial peak (PI). The use of a conservative 6xed axial shape in conjunction with typical radial peaking factors can result in integrated peaks exceeding the previous wide band limit.

Analyses performed for Palo Verde Nuclear Generator Station (PVNGS), demonstrated the conservatism of the fixed axial shape relative to an adverse 20% power shape from the CPC uncertainty analysis. The analyses also confirmed the validity of the CPC thermal hydraulic

. tuning process for an extended wide band trip limit of 1.28 to 7.00. Since ANO-2 uses the same CPC algorithm as PVNGS, the conclusion is applicable to ANO-2. This trip limit can be raised since it is not credited for any safety analysis events.

f i l

l l

I

An=cha==d to 2CAN079803 Page 3 of 3 TS Bases 3/4.2.1 and 3/4.2.4 (names B 3/4 2-1 & B 3/4 2-3)

The change to TS Bases 3/4.2.1 and 3/4.2.4 (pages B 3/4 2-1 & B 3/4 2-3) describes the COLSS uncertainty factors in general terms rather than listing speci6c values. Tim COLSS program contains appropriate uncertainties for the planar radial peaking factor naasurement, engineering design factors, state parameter measurement, soAware algorithm modeling, computer processing, rod bow, and core power measurement. The specific values for the l associated uncertainties are oRen dependent on analytical methodologies and therefore require change from cycle to cycle. This change has been evaluated under 10 CFR 50.59 with the resulting conclusion that an unreviewed safety question is not involved.

l l This description is consistent with that contained in Revision 1 of NUREG-1432, " Standard Technical Speci6 cations for Combustion Rap +dg Plants" and similar to how other CE l plants have addressed these values.

TS Bases 3/4.6.3 (name B 3/4 6-4)

This change to the ANO-2 TS Bases 3/4.6.3, (page B 3/4 6-4) Containment Isolation Valves, revises a reference to the procedure containing the containment isolation valve table. This change has been evaluated under 10 CFR 50.59 with the resulting conclusion that an unreviewed safety question is not involved.

L l Generic Letter 91-08 provided guidance for relocating the containment isolation valve TS tables to plant procedures. ANO submitted a change request by letter dated July 22,1993, l (2CAN079302) to remove the containment isolation valve table from the TS. Amendment

154 to ANO-2 TS implemented the proposed chaage. Amendment 154 revised the bases to

! reference the containment isolation valve table located in Procedure 2203.005. Subsequently, i ANO determined the scope'and significance of the containment isolation table warrants ,

locating it in operations department administrative procedure. Since there is not a

( 7equirement to maintain the specific procedure number in the TS Bases and because the

(

. procedure number could change again, the procedure number was removed. This change revises the bases to indicate the containment isolation valves have been relocated to the plant l procedures.

l This change also corrects a grammatical error in the second paragraph on page B 3/4 6-4.

The first word of the second paragraph was inadvertently changed from "The" to "In" during the development of Amendment 154. This change co:Tects this error by implementing the previous wording.

l l

l I