ML20107J955

Application for Amend to License NPF-8,requesting Unit 2 Cycle 11 TS Change for SG Tubes Remaining in Svc

ML20107J955
Person / Time
Site:	Farley
Issue date:	04/23/1996
From:	Dennis Morey SOUTHERN NUCLEAR OPERATING CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20107J957	List: ML20107J955 ML20107J962
References
NUDOCS 9604250330
Download: ML20107J955 (10)

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W '* i Soutnern Nuclxr Operating Comp:ny  !

' J'- Pos! Offico Box 1295

. .. Birmingham Alabima 35201

' Telephone (205) 868 5131 o... uo,.y Southern Nudear Operating Company

~ Vice President Farley Project April 23, 1996 the southem electnc system

. Docket No.:' 50-364

. U. S. Nuclear Regulatory Commission i - ATTN.: Document Control Desk Washington, D. C. 20555 2-Joseph M. Farley Nuclear Plant - Unit 2 L* Tubesheet Region Cycle 11 Specific Plunning Criterion Ladies and Gentlemen:

l On April 22, 1996, Southern Nuclear (SNC) submitted a Unit 2 Technical Specification j change concerning L* Tubesheet Region Plugging Criterion. On April 22,1996, SNC l requested and NRC Staff approved Enforcement Discretion for Farley Unit 2. The basis for l this Enforcement Discretion is described in SNC letter to the 'NRC dated April 23,1996 which described the Unit 2 cycle 11 specific L* criteria. l Enclosed is SNC's request for a Unit 2 Cycle 11 only Technical Specification change for Steam Generator Tubes remaining in service that meet the L* criteria described in the SNC

} April 23,1996 Enforcement Discretion letter.  ;

E It is requested that the NRC approve the Unit 2 Cycle 11 specific L* Technical Specification change. Attachment I contains a Significant Hazards Consideration analysis. SNC has determined that the proposed license amendment will not significantly affect the quality of the human environment. Attachment 2 provides page change instructions and revised technical specification pages and bases. Attachment 3 provides corresponding marked-up existing l technical specification pages and bases pages.  !

, A copy of the proposed changes is being sent to Dr. D. E. Williamson, the Alabama State l

Designee, in accordance with 10 CFR 50.91(b)(1).

4 If there are any questions, please advise.

l 9604200330 960423 Respectfully submitted, PDR ADOCK 05000364 P

, PDR gg j Dave Morey Sworn to amisubscribed before me thgidayof A k ,1996.

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U. S. Nuclear Regulatory Commission . Page 2 Attachments:

1. Significant Hazards Evaluation

2. Technical Specification Pages

3. Marked Up Proposed Technical Specification Pages cc: Mr. S. D. Ebneter, Region II Administrator Mr. B. L. Siegel, NRR Senior Project Manager Mr. T. M. Ross, FNP Sr. Resident Inspector Dr. Donald E. Williamson, State Department of Public Health -

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Attachment 1 4

Sinnificant Hazards Evaluation p

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CYCLE 11 La TUBESHEET REGION PLUGGING CRITERION FARLEY NUCLEAR PLANT UNIT 2

. SIGNIFICANT HAZARDS CONSIDERATION ANALYSIS INTRODUCTION As required by 10 CFR 50.91(a)(1) this analysis is provided to demonstrate that a proposed license amendment for the J. M. Farley Nuclear Plant Unit 2 steam generators involves no significant hazard consideration. In accordance with the three factor test of 10 CFR 50.92(c),

implementation of the proposed license amendment (Technical Specification Change) was analyzed using the following standards and found not to: 1) involve a significant increase in the probability or consequences of an accident previously evaluated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or,3) involve a ,

significant reduction in a margin of safety.

Operation with 6 or more tubes that fail to meet the F* technical specification but meet the criteria for L' as described below is acceptable. The basis for this is provided in SNC submittal " Joseph M. Farley Nuclear Plant Unit 2 L* Tubesheet Region Plugging Criterion" dated April 22,1996. All structural integrity and leakage requirements of draft Reg. Guide 1.121 continue to be met for the tubes in question. The " Joseph M. Farley Nuclear Plant Unit 2 L* Tubesheet Region Plugging Criterion" technical specification submittal dated April 22, 1996 supports our safety basis for continued operation. It should be noted that the values of L* criterion in the " Joseph M. Farley Nuclear Plant Unit 2 L* Tubesheet Region Plugging Criterion" dated April 22,1996, bound the values of the Unit 2 cycle 11 specific L* criterion associated with this submittal. This submittal is intended to support a cycle specific one time ,

technical specification change using the measured L* criterion presented below. i 1

Unit 2 Cycle 11 Specific L* Criterion I Parameter Measured L* Base L* critericin criterion for Cycle 11 Minimum distance of SRE 2.07 inches 1.84 inches Maximum number of distinct degradation areas in a band 5 15 Maximum inclination angle within a single band 15 degrees 30 degrees l Maximum crack length .39 inches .5 inches Minimum distance of SRE from the bottom of the transition 1.45 inches .56 inches roll to the top of the indication l

Provided below is a excerpt from the " Joseph M. Farley Nuclear Plant Unit 2 L* Tubesheet l Region Plugging Criterion" technical specification submittal dated April 22,1996 to support our safety basis for this technical specification change. This execrpt combined with the margins noted above from the base L* c:iterica form the basis for this significant hazards evaluation.

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. BEGIN EXCERPT

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. DESCRIPTION OF THE AMENDMENT REQUEST

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The L* criterion assesses indications of degradation within the roll expanded region of steam generator tubes based upon location of the indication, the axial length of the indication and the inclination angle of the indications. The tubes to which the criteria are applied must meet the following restrictions:

i e Each tube must be inspected using the RPC probe or other advanced NDE inspection method for a minimum distance of 3.1 inches below the bottom of the roll transition or the .  ;

top of the tubesheet, whichever is lower. This is termed the inspection region.

• - A minimum length of sound roll expansion of 0.50 inch plus allowance for eddy current

! measurement uncertainty extending down from the bottom of the roll expansion or the top

of the tubesheet, whichever is lower, must be established. This is termed L* length.

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• For tubes where a single degradation band is evidenced within the inspection region, an aggregate distance of sound roll expansion of 1,74 inch minimum plus allowance for eddy current measurement uncertainty, must be established. This aggregate distance includes i

the L* length. Single band degradation is limited to 0.50 inch axial length, with an inclination angle of not more than 30 degrees from the vertical, and limited to 15 distinct l degradation areas within the band. SNC has administratively limited the number of distinct degradation areas in the band to 15, although the Westinghouse support documentation justifies up to 30 such areas.

. For tubes where multiple degradation bands (maximum of two) are detected within the inspection region, an aggregate distance of sound roll expansion of 1.87 inch plus allowance for eddy current measurement uncertainty, must be established. Multiple degradation bands are limited to 0.5 inch axial length each, with an inclination angle of no 4

more than 30 degrees from the vertical, and limited to 15 distinct degradation areas within each band. SNC has administratively limited the number of distinct degradation areas in the band to 15, although the Westinghouse support documentationjustifies up to 30 such areas.

• The L* criterion is limited to 600 tube ends per steam generator.

If any of these items is not validated, the tube must be repaired.

The total length of aggregate sound roll expansion needed to prevent pullout during all plant conditions for the single and multiple band degradation cases is termed the pullout load react'.on length (PLRL). This term is equivalent to the L* distance.

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. # r j EVALUATION i i- ,

. The proposed license amendment (Technical Specification change) addresses the action  ;

required when degradation has been detected in the top portion (within the F* distance) of the  :

mechanically expanded portion of steam generator tubes within the tubesheet region. The i presence of the tubesheet enhances the integrity of degraded tubes in that region by precluding

tube deformation beyond the expanded outside diameter. The roll expansion of the tube into the tubesheet provides a barrier to significant leakage for through wall cracking of the tube in the expanded region. L* criterion approved for implementation at similar plants include provisions which account for the support of the tubesheet in the portion of the tubes below the i top of the tubesheet. The previously approved (at Farley Unit 2 and similar plants) F* criterion was not established to address degradation in the top portion of the tube expansion.

F*/L* Pluuning Criteria Comoarison 5 As stated previously, the L* criterion is essentially an extension of the existing F*

methodology. WCAP-11306 Rev. 2 provides the basis for the F* criterion currently licensed at Farley Unit 2. Both criteria (F* and L*) provide for stmetural integrity of the tube during all plant conditions by verification of SRE lengths. The SRE lengths provide for resistance 4

load capabilities such that the most limiting postulated pressure developed end cap loading defined in Regulatory Guide 1.121 is balanced. In the F* criterion, a SRE length of 1.54 inch plus allowance for eddy current uncertainty measurements must be established below either the bottom of the roll transition, or the top of tubesheet, whichever is lower. The F* distance of ,

1.54 inch is established based on faulted condition loadings and represents the limiting RG  !

1.121 loading. The F* criterion precludes primary to secondary leakage during all plant  !

conditions. The L* criterion addresses tubes which contain axial or nearly axially oriented eddy current indications within the F* length. The L* criterion includes a minimum length of SPE immediately below the bottom of the roll transition or top of tubesheet, whichever is lower, which precludes significant leakage during all plant conditions. This distance plus allowance for eddy current measurement uncertainty is termed L* length. The SRE length above a crack is summed with the SRE length below the crack tc ensure structural integrity

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during all plant conditions. This distance of total aggregate SRE is termed L* distance. In both the F* and L' evaluation bases, circumferential separation of the tube is postulated below either the F* or L* distance. Verification of the SRE F* or L* distances provides for structural integrity even for the postulated condition of a tube separation below the F* or L

• distance. Tube rupture can occur if a F* or L* tube is postulated to separate and lift out of the tubesheet due to the end cap load. Verification of the F* or L* distance prevents tube hft from occurring and therefore prevents the possibility of a tube rupture even for a tube postulated to become separated below the F* or L* distance.

L' Evaluation Discussion l

. The proposed change designates a portion of the tube for which tube degradation of a defmed

type does not necessitate remedial action except as dictated for compliance with tube leakage i

limits as set forth in the Farley Nuclear Plant Unit 2 Technical Specifications. As noted above, 1 the area subject to this change is in the expanded portion of the tube within the tubesheet of

the steam generators. The length of mechanical expansion required to resist significant leakage

- for all normal operating and postulated accident conditions is designated L* length. Tubes whose integrity is ascertained using L* length will not contribute significantly to off' site dose.

In addition to determining that the expansion of the tube within the L* length is not degraded, ,

4 a determination of the condition of tube degradation in the portion of the tube below the L*

l length must be made to verify that the degraded tube has sufficient structural integrity, i The proposed amendment would modify Technical Specification 3/4.4.6 " Steam Generator Surveillance Requirements" which provides tube inspection requirements and acceptance

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criteria to determine the level of degradation for which the tube may remain in service. The proposed amendment would add definitions required for the L* criterion and prescribe the portion of the tube subject to the criteria. The proposed Technical Specification changes

accompany this analysis.

1 ANALYSIS 1

Conformance of the proposed amendments to the standards for a determination of no significant hazard as defined in 10 CFR 50.92 (three factor test) is shown in the following:

1. Operation of the Farley Nuclear Plant Unit 2 steam generators in accordance with the proposed license amendment does not involve a significant increase in the probability or

. consequences of an accident previously evaluated.

The supporting technical evaluations of the subject criteria demonstrate that the presence of the tubesheet enhances the tube integrity in the region of the hardroll by precluding tube deformation beyond its initial expanded outside diameter. The resistance to both tube rupture and tube collapse is strengthened by the presence of the tubesheet in that region. The result of i 1 the hardroll of the tube into the tubesheet is an interference fit between the tube and the i

tubesheet. Tube rupture can not occur because the contact between the tube and tubesheet

, does not permit sufficient movement of tube material. In a similar manner, the tubesheet does not permit sufficient movement of tube material to permit buckling collapse of the tube during postulated LOCA loadings.

The type of degradation for which the L* criterion has been developed (cracking with an axial or near axial orientation) has been found not to significantly reduce the axial strength of a tube.

An evaluation including analysis and testing has been done to determine the strength reduction for axial loads with simulated axial and near axial cracks. This evaluation provides the basis 4

for the r.cceptance criteria for tube degradation subject to the L* criterion.

The SRE L* length is sufficient to preclude significant leakage from tube degradation located below the L* length. The existing Technical Specification leak rate requirements and accident

. analysis assumptions remain unchanged in the unlikely event that significant leakage from this region does occur. Any leakage from the tube within the tubesheet at any elevation in the tubesheet is fully bounded by the existing steam generator tube rupture analysis included in the

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. Farley Nuclect Plant Final Safety Andysis Report. A conservative leakage allowance for each L* tube is provided to determine the impact of L* criterion upon offsite doses in the event of a

, postulated double ended guillotine break of the main steam line outside of containment, but upstream of the main steam line isolation valves. Since Farley Unit 2 has implemented the Interim Plugging Criteria (IPC) for ODSCC at the tube support plates, projected steam line break (SLB) leakage at the end of the next successive operating cycle must be evaluated. Per l Generic Letter 95-05, plants implementing the IPC can utilize SLB leakage limits higher than i i the originally assumed 1.0 gpm primary to secondary leakage value provided an analysis of offsite doses consistent with the Standard Review Plan methodology is performed This

! analysis performed for the Farley Unit 2 plant indicates that primary to secondary leakage of i 11.2 gpm in the faulted loop (0.1 gpm in the intact loops) will result in offsite doses at the site boundary ofless than 10% of the 10 CFR 100 guidelines. The total projected SLB leakage

' from all leakage sources must remain below this value. Per attachment 4 addressing the L*

methodology, the number of tube ends to which L* criterion can be applied is limited to 600

per steam generator. Using a bounding SLB leakage allowance per L* tube, the SLB leakage component from 600 L* tube ends will be less than 0.33 gpm in the faulted loop. The ,

2 proposed L* criterion does not adversely impact any other previously evaluated design basis l accident. . As the current Unit 2 IPC SLB leakage has been calculated to be less than 2 gpm in the faulted loop, a SLB leakage margin of over 9 gpm is provided for this cycle As noted above, tube rupture and pullout is not expected for tubes using the L* criterion. In i addition to the L* length, a minimum length of SRE below the identified degradation must be established. The aggregate L* distance of SRE provides the structural integrity to prevent tube pullout. Conservatively, it is assumed that the degraded band length does not provide any support in resisting tube pullout.

Therefore SNC concludes that Operation of the Farley Nuclear Plant Unit 2 steam generators

, in accordance with the proposed license amendment does not involve a significant increase in 4 the probability or consequences of an accident previously evaluated.

2. The proposed license amendment does not create the possibility of a new or different l l- kind of accident from any accident previously evaluated.

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' Implementation of the proposed L* criterion does not introduce any significant changes to the plant design basis. Use of the criterion does not provide a mechanism to result in an accident initiated outside of the region of the tubesheet expansion. The structural integrity of L* tubes will be maintained during all plant conditions. Any hypothetical accident as a result of any tube

degradation in the expanded portion of the tube would be bounded by the existing tube rupture accident analysis. Ifit is postulated that a circumferential separation of an L* tube were to occur below the PLRL, tube stmetural and leakage integrity will be maintained during all plant conditions. Verification of the L* distance of non-degraded tube roll expansion prevents the  !

postulated separated tube from lifting out of the tubesheet during all plant conditions.

Verification of the L* criterion prevents tube displacement of any magnitude, and therefore,

, postulated axial cracks existing a minimum of 0.5 inch from either the bottom of the roll transition or top of tubesheet, whichever is lower, from migrating out of the tubesheet.

j Therefore, SNC concludes that the proposed license amendment does not create the possibility of a new or different kind of accident from any accident previously evaluated.

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3. The proposed license am:ndment does not involve a significant reduction in a margin ofsafety.

. 1 The use of the L* criterion has been concluded to maintain the integrity of the tube bundle commensurate with the requirements of draft Regulatory Guide 1.121 under normal and postulated accident conditions. The safety factors used in the verification of the strength of the degraded tube are consistent with the safety factors in the ASME Boiler and Pressure Vessel Code used in steam generator design. The L* length has been verified by testing to be greater than the length of roll expansion required to preclude significant leakage during normal and postulated accident conditions. The leak testing acceptance criteria are based on the primary to secondary leakage limit in the Technical Specifications and the leakage assumptions used in the FSAR accident analyses. The L* distance provides for structural integrity during all plant conditions.

Implementation of the L* criterion will decrease the number of tubes which must be taken out of service with tube plugs or repaired with sleeves. Both plugs and sleeves reduce the RCS flow margin, thus implementation of the L* criterion will maintain the margin of flow that would otherwise be reduced in the event ofincreased plugging or sleeving.

Therefore, SNC, concludes based on the above, it is concluded that the proposed change does not result in a significant reduction in a loss of margin with respect to plant safety as defined in the Final Safety Analysis Report or the bases of the FNP technical specifications.

CONCLUSION Based on the preceding analysis, it is concluded that operation of Ae Farley Nuclear Plant Unit 2 steam generators in accordance with the proposed amendment dces not involve a significant hazards consideration as defined in 10 CFR 50.92.

4 END OF EXCERPT Based on the analysis provided above it is concluded that operation of the Farley Nuclear Plant in accordance with this proposed amendment for the remainder of Cycle 11 does not involve a

significant hazards consideration as defined in 10 CFR 50.92.

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a Attachment 2 Revised Technical Specification Pages l

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