ML20101S491

Proposed Tech Spec Changes,Updating Inservice Insp Program for Pressure Retaining Components & Supports to Comply w/10CFR50.55

ML20101S491
Person / Time
Site:	FitzPatrick
Issue date:	01/28/1985
From:	POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK
To:
Shared Package
ML20101S474	List: ML20101S473 ML20101S480 ML20101S491
References
NUDOCS 8502050437
Download: ML20101S491 (9)

Text

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ATTACHMENT I e

Proposed Technical Specification Changes Related to the Inservice Inspection Program New York Power Authority James A. Fitzpatrick Nuclear Power Plant Docket No. 50-333 January 28, 1985 8502050437 850128 PDR ADOCK 05000333

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3.6f(cont'd)' .

4.6-(cont'd)

F. Structural Intearity F. Structural Intearity

-The structural integrity of the 1. . Nondestructive inspections shall Reactor Coolant System shall be be performedLon the ASME Boiler maintained:at<the level' required and Pressure Vessel Code Class 1,

.by the original acceptance stan-- 2 and 3 components and supports dards'throughout the life of the in accordance with the requirements Plant. of the weld and support inservice in-spection program. This inservice inspection program is. based on an NRC approved edition of, and addenda to Section XI of the ASME Boiler and Pressure Vessel Code which is in effect 12 months.or less prior to the beginning of the inspection interval.

2. An augmented inservice inspection program is required for those high stressed circumferential piping joints in the main steam and feedwater lines larger-than 4 inches in diameter, where no restraint against pipe whip is provided. The augmented inservice inspection program shall consist of 100 percent inspection of'these welds per inspection interval.

G. Jet Pumps G. Jet Pumps Whenever the reactor is in the Whenever there is recirculation flow l startup/ hot standby or run modes. . with the reactor in the startup/ hot all jet pumps shall be operable. standby or run modes, jet pump oper-If it is determined that a jet pump ability shall be checked daily by is operable,-the reactor shall be verifying that the following conditions placed in a cold condition within do not occur simultaneously:

24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Amendment No.

144

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3.6 and 4.6 BA888-(cont'd).

not' required to be operable (reactor. Several locations on the main steam coolant temperature 212*F.and the. lines and feedwater. lines are not reactor vessel vented or the reactor' restrained to prevent pipe whip in vessel head removed). ' Permitting. the event of pipe failure at these physics testing and operator locations.,'The physical layout training-under these conditions within-the drywell precludes re-would not. place the plant in an straints at these points. Unre-Unsafe condition. strained high stress areas have been identified in these lines where F. Structural Inteority breaks.could result in pipe whip such that the pipe could impact the A pre-service inspection of the primary containment wall. Augmented ASME Code Class 1 components was inservice inspection of these weld performed after site erection to locations shall be performed during assure the system was free of gross each inspection period.

defects. An initial inspection program as detailed in Appendix F In addition, visual inspection in ac-of the FSAR was developed and cordance with the approved ASME code will based on an approved edition of the be made during periodic pressure and ASME Code. hydrostatic tests of critical systems.

The inspection program specified en-compasses the major areas of the The program has been expanded to in- vessel and piping system within clude the requirements of later, the drywell. The inspection period is' approved ASME Code editions and based on the observed rate of defect addenda as far as practicable. The growth from fatigue studies sponsored importance of these inspections is by the AEC.

recognized, and efforts to develop practical new alternative methods These studies show that thousands of assuring plant inservice integrity of stress cycles, at stresses be-will continue. This inspection pro- yond any expected to occur in a gram should assure the detection of Reactor Coolant System, were required

. problem areas well before they re- to propagate a crack. The test present a significant impact on safety.

Amendment No.

153

JAFNPP PAGES 157 THROUGH AND INCLUDING 162 HAVE BEEN INTENTIONALLY LEFT BLANK f

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Atendment No.

157-162

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y ATTACHMENT II Proposed Technical Specification Changes Related to the Inservice Inspection Program New York Power Authority James A.-FitzPatrick Nuclear Power Plant Docket No. 50-333 f

January 28, 1985-1 6

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I. Description of the Proposed Chances

~The proposed changes to the FitzPatrick Technical Specifications relate to the Inservice Inspection (ISI) program for pressure retaining components and their supports.

Specifically, the following changes are being proposed:

On page 144, Section 4.6.F.1 is completely rewritten to read, "1. Nondestructive inspections shall be performed on the ASME Boiler and Pressure Vessel Code Class 1,2'and 3 components and supports in accordance with the requirements of the weld and support inservice inspection program. This inservice '

inspection program is based on an NRC approved edition'of and addenda to Section XI of the ASME ,

Boiler and Pressure Vessel Code which is in '

effect'12 months or less prior to the beginning of the inspection interval."

On page 144, modify Section 4.6.F.2 to read, "An' augmented inservice inspection program is required for those high stressed circumferential piping joints in the main steam and feedwater lines larger than 4 inches in diameter, where no restraint against pipe whip is provided. The augmented inservice inspection program shall consist of 100-percent inspection of these welds per inspection interval."

On page 153 modify Section F to' read, "Aspre-serviceJinspection of the ASME Code Class I. components was performed after site erection to assure-the-system was free:of gross defects. ' An initial inspection program as detailed in

'Apppendix F ofethe FSAR was developed and based on an_ approved-edition of'Section XI of the ASKE Code.

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The program has been. expanded to include the requirements of later, approved ASME Code editions and addenda as far as practicable. The importance of these inspections is. recognized, and efforts to develop practical new alternative methods of assuring plant inservice integrity will continue. This inspection program should assure the detection of problem areas well before-they represent a significant impact on safety.

Several locations on the main steam lines and feedwater lines are not restrained to prevent pipe whip in the event of pipe failure at these locations. The physical

-layout within the drywell precludes restraints at these points. ' Unrestrained high stress areas have been-identified in these lines where breaks could result in pipe whip such that the pipe could impact the primary containment wall. Augmented inservice inspection of these weld locations shall be performed during each inspection period.

In addition, visual inspection in accordance'with the approved ASME Code will'be made during pressure and hydrostatic. tests of critical. systems.

The inspection program specified encompasses the major areas'of the-vessel and piping systems within-the'drywell.

The inspection period is based on observed rate of defect ~ growth from fatigue studies sponsored by the AEC.

These studies show that thousands of stress cycles, at stresses.beyond any expected to occur in a Reactor Coolant System, were required to propagate a crack. The test"...

Delete ~ Table 4/6-l'on pages 157 through 162.

IIi> Purpose of the' Proposed Chances

'The proposed changes are.necessary to allow the revision of the.FitzPatrick weld and support ISI program to comply with the requirements of the later editions of the ASME Boiler'and Pressure-Vessel' Code,Section XI.

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The revision-of theLISI Program and this amendment are required by 10 CFR 50.55 a (g)

(4) (ii) and.(5)(i).- This revision of the

. inspection program is to allow the incorporation of~ improved examination techniques and sampling plans that may have been develop 6d during the previous 120 month interval.

-III. Impact of the Proposed Chances The ISI program provides continued assurance of structural integrity of the pressure-retaining components of critical systems that contain radioactive and/or high pressure fluids, or that provide for safe shutdown of the plant in either normal or accident conditions. The proposed changes will allow the program to reflect industry experience and increased understanding of service induced failure mechanisms, and detection. This experience and-knowledge is reflected in the require-ments of newer editions of Section XI of the ASME Code approved for use by the.NRC. The revised inspection pro-

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gram represents an expanded and more clearly defined sampling plan of ox-amination of pressure retaining comp-

onents'and their supports. For these reasons, operation of the FitzPatrick Plant in accordance1with the proposed amendments would not:

(1) involve a significant increase in-the L probability or consequences of an acci-(f ~ .. dent previously evaluated;1or create the possibility of a new or (2) ' different kind of accident from any l -

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< accidentipreviously evaluated; or L (3) involve a significant reduction in a margin L ,

.of safety..

t-4IV. Implementation'of the' Chances 9 Implementation of the changes,Jas proposed, s< will'not; impact the fire protection. pro--

p .g ram at FitzPatrick, nor-will the changes impact the environment.

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.n V. Conclusion The incorporation of these changes

a) will not increase the probability or the consequences of an accident or malfunction of equipment important to safety as evaluated previously in the Safety Analysis Report (SAR):

b) will not increase the possibility of an accident or malfunction of a type other that that evaluated previously in the SAR:

c) will not reduce the margin of safety as defined in the basis for any Technical Specification;

.d) does not constitute'an unreviewed safety

question; and e) involves no significant hazards considerations, as defined in 10 CPR 50.92.

VI.

References:

1. James A..FitzPa' trick Nuclear Power Plant Safety Evaluation Report (SER).

2.- ' James 1A. FitzPatrick Nuclear Power' Plant Final. Safety Analysis Report (FSAR), Rev.

1, July 1983.

3. ASME Boiler and Pressure Vessel Code Section XI.

J L 4. 10CFR50.55. '

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