ML17329A420

Application for Amend to License DPR-58,revising TSs to Allow Alternate Plugging Criteria to Establish Operability of SG Tubes.Westinghouse Nonproprietary Rept WCAP-13188 & Proprietary Rept WCAP-13187 Encl.Proprietary Rept Withheld

ML17329A420
Person / Time
Site:	Cook
Issue date:	03/20/1992
From:	Fitzpatrick E INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	Murley T NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML17329A421	List: ML17329A420 ML17329A422 ML17329A423 ML17334B415
References
AEP:NRC:1166, NUDOCS 9203260124
Download: ML17329A420 (15)

Text

ACCEK,ERATED DISTMBUTION DEMONSTPWTION SYSTEM "f. ~ ..p, REGULATOO INFORMATION DISTRIBUTION STEM (RIDE) r kiq i"'

ACCESSION NBR:9203260124 DOC.DATE: 92/03/20 NOTARIZED: YES DOCKET FACIL:50-315 Donald, C. Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 AUTH. NAME AUTHOR AFFILIATION FITZPATRICK,E. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele RECIP.NAME RECIPIENT AFFILIATION MURLEY,T.E. Document Control Bra@eh (Document Control Desk)

SUBJECT:

Application for amend to License DPR-58,revising TSs to allow alternate plugging criteria to establish operability of SG tubes. Westinghouse nonproprietary rept WCAP-13188 6 proprietary rept WCAP-13187 encl. Proprietary rept withheld.

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Columbus, OH 43216 AEP: NRC: 1166 Donald C. Cook Nuclear Plant Unit 1 Docket No. 50-315 License No. DPR-58 TECHNICAL SPECIFICATIONS CHANGE TO ALLOW ALTERNATE PLUGGING CRITERIA U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. '20555 Attn: T. E. Murley March 20, 3.992

Dear Dr. Murley:

This letter and its attachments constitute an application for amendment to the Technical Specifications (T/Ss) for Donald C. Cook Nuclear Plant Unit 1 in accordance with 10 CFR 50.90. The purpose of this license amendment request is to obtain authorization to use an alternate plugging criteria to establish operability of Cook Nuclear Plant Unit 1 steam generator tubes.

The T/Ss now require steam generator tubes to be plugged or repaired when the degradation exceeds 40% tube wall penetration as determined by non-destructive examination.'e believe that flaw indications within the bounds of the tube support plate with a bobbin voltage less than or equal to 1.5 volts may remain in service. Flaw indications greater than 1.5 volts but less than or equal to 4.0 volts may remain in service if a rotating pancake coil (RPC) probe inspection verifies that axial outer diameter stress corrosion cracking (ODSCC) is the degradation mechanism. Flaw indications with a voltage greater than 4.0 volts will be plugged or repaired. We also believe that a tube can remain in service if the signal amplitude is less than or equal to 4.0 volts, regardless of the depth of penetration, if the projected end-of-cycle distribution of crack indications is verified to result in primary-to-secondary leakage less than 120 gpm in one steam generator during a postulated steam line break. Westinghouse WCAP-13187, "D.

C. Cook Unit 1 Steam Generator Tube Plugging Criteria for Indications at Tube Support Plates," provides the development of the 4.0 volt alternate plugging criteria.

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~ AEP: NRC; 1166 A detailed description of the proposed changes and our analyses concerning significant hazards considerations are included in to this letter. Attachment 2 contains the proposed revised T/Ss pages. Attachment 3 contains the marked-up copies of the exi.sting T/Ss. The Westinghouse reports, WCAP-13187, "D. C, Cook Unit 1 Steam Generator Tube Plugging Criteria for Indications at Tube Support Plates" (Proprietary), and WCAP-13188, "D. C. Cook Unit 1 Steam Generator Tube Plugging Criteria for Indications at Tube Support Plates" (Non-Proprietary), are in Attachment 4.

Enclosed in Attachment 5 are a Westinghouse authorization letter, CAW-92-282, accompanying affidavit, Proprietary Information Notice, and Copyright Notice. Because WCAP-13187 contains information proprietary to Westinghouse Electric Corporation, it is supported by an affidavit signed by Westinghouse, the. owner of the information. The affidavit sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of Section 2790 of the Commission's regulations. Accordingly, it is respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR Section 2790 of the Commission's regulations.

Correspondence with respect to the copyright or proprietary aspects of the items listed above or the supporting Westinghouse affidavit should reference CAW-92-282 and should be addressed to N. Liparulo, Manager of Nuclear Safety and Regulatory Activities, Westinghouse Electric Corporation, P. 0. Box 355, Pittsburgh, Pennsylvania 15230-0355.

We believe that the proposed changes will not result in (1) a significant change in the types of effluents or a significant increase in the amount of any effluents that may be released offsite, or (2) a significant increase in individual or cumulative occupational radiation exposure.

These proposed changes have been reviewed by the Plant Nuclear Safety Review Committee and will be reviewed by the Nuclear Safety and Design Review Committee at their next regularly scheduled meeting.

In compliance with the requirements of 10CFR50.91(b)(1), copies of this letter and its attachments have been transmitted to the Michigan Public Service Commission and the NFEM Section Chief.

Please contact us ifNuclear you have any questions concerning this license amendment request. Regulatory Commission staff review and

Dr. T. E. Murley AEP: NRC: 1166 approval of these changes are requested prior to May 22, 1992.

This is the minimum amount of time necessary to prepare for implementation of the alternate plugging criteria during the Unit 1 refueling outage beginning approximately June 23, 1992.

This document has been prepared following Corporate procedures that incorporate a reasonable set of controls to ensure its accuracy and completeness prior to signature by the undersigned.

Sincerely, Vice President eh Attachments cc: D. H. Williams, Jr.

A. A. Blind - Bridgman J. R. Padgett G. Charnoff A. B. Davis - Region III NRC Resident Inspector - Bridgman NFEM Section Chief

ATTACHMENT 1 to AEP:NRC:1166 NO SIGNIFICANT HAZARDS CONSIDERATION EVALUATION IN SUPPORT OP THE ALTERNATE PLUGGING CRITERIA to AEP:NRC:1166 Page 1 INTRODUCTION A license amendment is proposed to preclude unnecessarily plugging steam generator tubes due to the occurrence of outer diameter initiated stress corrosion cracking (ODSCC) at the tube support plates in the Cook Nuclear Plant Unit 1 steam generators. Using the existing Technical Specifications (T/Ss) steam generator tube plugging criteria of 40% tube wall penetration as determined by non-destructive examination (NDE), many of the tubes with crack indications would needlessly have to be removed from service. The alternate plugging criteria for tube support plate elevation ODSCC occurring in the Cook Nuclear Plant Unit 1 steam generators may result in tubes with both partial and through-wall cracks returning to service. In the limiting case, it is demonstrated that the presence of through-wall cracks alone is not reason enough to remove a tube from service.

DESCRIPTION OF THE AMENDMENT REQUEST As required by 10 CFR 50.91 (a)(1), an analysis is, provided to demonstrate that the proposed license amendment to implement an alternate steam generator tube plugging criteria for the tube support plate elevations at Cook Nuclear Plant Unit 1 involves no significant hazards considerations. The alternate plugging criteria involves a correlation between eddy current bobbin coil signal amplitude (voltage) and tube burst. and leakage capability.

The plugging criteria is based on testing of laboratory-induced ODSCC specimens, extensive examination of pulled tubes from operating steam generators (industry vide), and field experience from leakage due to indications at the tube support plates (world vide).

Specifically, crack indications with bobbin coil voltages less than or equal to 4.0 volts, regardless of indicated depth, do not require remedial action if postulated steam line break leakage can be shown to be acceptable. Crack indications with bobbin coil signal amplitudes exceeding 4.0 volts must be either plugged or repaired. Bobbin coil signal amplitudes greater than 1.5 volts, left in service at the discretion of the owner, must be inspected if using a motorized rotating pancake coil (RPC).

The proposed amendment would modify T/Ss 3.4.5 "Steam Generators,"

3.4.6, "Reactor Coolant System Leakage," and the associated bases.

These proposed changes provide tube inspection requirements and acceptance criteria to determine the level of degradation for which a tube experiencing ODSCC at the tube support plate elevations may remain in service in the Cook Nuclear Plant Unit 1 steam generators.

to AEP:NRC:1166 Page 2 EVALUATION Steam Generator Tube Inte rit Discussio In the .development of the alternate plugging criteria, Regulatory Guides (RG) 1.121, "Bases for Plugging Degraded PWR Steam Generator Tubes," and 1.83 "Inservice Inspection of PWR Steam Generator Tubes," are used as the bases for determining that steam generator tube integrity considerations are maintained within acceptable limits. Regulatory Guide 1.121 describes a method acceptable to the NRC staff for meeting General Design Criteria 2, 4, 14, 15, 31, and 32 by reducing the probability and consequences of steam generator tube rupture through determining the limiting safe conditions of tube wall degradation. Tubes with unacceptable cracking, as established by inservice inspection, should be repaired or removed from service by plugging. This RG uses safety factors on loads for tube burst that are consistent with the requirements of Section III of the ASME Code. For the tube support plate elevation degradation occurring in the Cook Nuclear Plant Unit 1 steam generators, tube burst criteria are inherently satisfied during normal operating conditions by the presence of the tube support plate. The presence of the tube support plate enhances the integrity of the degraded tubes in that region by precluding tube deformation beyond the diameter of the drilled hole. It is not certain whether the tube support plate would function to provide a similar constraining effect during accident condition loadings. Therefore, no credit is taken in the development of the plugging criteria for the presence of the tube support plate during accident condition loadings. Conservatively, based on the existing data base, burst testing shows that the safety requirements for tube burst margins during both normal and accident condition loadings can be satisfied with bobbin coil signal amplitudes less than 6.8 volts, regardless of the depth of tube wall penetration of the cracking. Regulatory Guide 1.83 describes a method acceptable to the NRC staff for implementing GDC 14, 15, 31, and 32 through periodic inservice inspection for the detection of significant tube wall degradation.

Upon implementation of the plugging criteria, tube leakage considerations must also be addressed. It must be determined that the cracks will not leak excessively during all plant conditions.

For the alternate tube plugging criteria developed for the Cook Nuclear Plant Unit 1 steam generator tubes, little or no leakage is expected during normal operating conditions even with the presence of through-wall cracks. Industry wide, the crack morphology of SCC at tube support plate intersections is best described as a short, tight, axially oriented microcrack separated by ligaments of non-degraded material. The same morphology is judged to be present to AEP:NRC:1166 Page 3 in the Cook Nuclear Plant Unit 1 steam generators. Tube pull examination results from 1983 indicated axial SCC in its early stages. The depths of the evidenced degradation (determined by destructive exam) showed the maximum depth of penetration to be approximately 10% through-wall. The RPC testing performed during the 1989 outage has confirmed that axially oriented SCC cracks exist at the tube support plate intersections. Based on the RPC testing results and relatively small amplitude bobbin voltages coupled with low bobbin voltage growth rates, it is concluded that axially oriented ODSCC best defines the degradation morphology occurring at the Cook Nuclear Plant Unit 1 tube support plate intersections. Additional tubes will'e pulled during the next refueling outage (summer 1992) for destructive examination to confirm the tube degradation phenomena occurring at the tube support plates. No leakage during normal operating conditions has been observed in the field at similar plants for crack indications with signal amplitudes less than 7.7 volts. Additionally, no primary-to-secondary leakage at the tube support plate has been detected in U.S. plants. Relative.to the expected leakage during accident condition loadings, the limiting event with respect to primary-to-secondary leakage is a postulated steam line break event. Laboratory data for pulled tubes from other plants and model boiler specimens show limited leakage for indications under 10.0 volts during a postulated SLB condition (See Section 9,0 of WCAP-13187).

Additional Considerations The proposed amendment would preclude approximately 10 manrem occupational radiation exposure that would otherwise be incurred by plant workers involved in tube plugging or repair operations. The proposed amendment would minimize the loss of margin in reactor coolant flow through the steam generator in LOCA analyses. The proposed amendment would avoid loss of margin in reactor coolant system flow and therefore assist in demonstrating that minimum flow rates are maintained in excess of those required for operation at full power. Reduction in the amount of tube plugging required can reduce the length of plant outages and reduce the time that the steam generator is open to the containment environment during an outage, thereby minimizing airborne contamination and exposure.

In addition, the required 100% bobbin coil inspection to be performed at each outage as outlined in WCAP-13187 will help to identify new areas of concern which may arise by. providing a level of inservice inspection that is far in excess of the current T/Ss requirements.

to AEP:NRC:1166 Page 4 NO SIGNIFICANT 1ULEARDS ANALYSIS Ve have evaluated the proposed T/Ss change and have determined that it does not represent a significant hazards consideration based on the criteria established in 10 CFR 50.92(c). Operation of the Cook Nuclear plant in accordance with the proposed amendment will not:

Involve a si nificant increase i the robabilit or conse uences of an accide t rev ous eva uated Testing of model boiler specimens for free standing tubes at room temperature conditions show burst pressures in excess of 5000 psi for indications of ODSCC with voltage measurements as high as 19 volts. Burst testing performed on pulled tubes from other plants with up to 10 volt indications show burst pressures in excess of 5900 psi at room temperature.

Correcting for the effects of temperature on material properties and minimum strength levels (as the burst testing was done , at room temperature), tube burst capability significantly exceeds the RG 1.121 criteria, requiring the maintenance of a margin of three times normal operating pressure differential on tube burst if through-wall cracks are present. Based on the existing data base, this criteria is satisfied with bobbin coil indications with signal amplitudes less than 6.8 volts, regardless of the indicated depth measurement. This structural limit is based on a -95%

lower tolerance limit (LTL) confidence level of the data.

The 4.0 volt plugging criteria compares favorably with the structural limit considering expected growth rates of ODSCC at Cook Nuclear Plant Unit l. Alternate crack morphologies correspond to 6.8 volts so that a unique crack length is

'an not defined by a burst-pressure-to-voltage correlation.

However, relative to expected leakage during normal operating conditions, no field leakage has been reported at other plants from tubes with indications with a voltage level of under 7.7 volts (as compared to the 4.0 volt proposed alternate tube plugging criteria proposed in this submittal).

Also, a qualitative assessment is made between the beginning-of-cycle (BOC) 4.0 volt tube plugging criteria and the current 40% allowable tube wall penetration plugging criteria at Cook Nuclear Plant Unit 1. An ODSCC-degraded tube support plate intersection with a 4.0 volt bobbin coil response is expected to burst at approximately 7400 psi, using the mean curve of Figure 9-2 of WCAP-13187. While the -95't LTL curve is used in the application of the plugging criteria, the mean curve must be used for this specific comparison in order to adequately compare the two data sets used. Per WCAP-13187, to AEP:NRC:1166 Page 5 a comparison of the material properties at 650 P and room temperature condition properties showed that the elevated temperature properties are approximately 0.86 of the room temperature properties. Therefore, the temperature-adjusted burst pressure for a 4.0 volt bobbin coil indication is expected to be approximately 6400 psi.

Figure ll of NUREG-0718 plots the burst pressures of thinned 0.875 x 0,050 inch steam generator tubes. At 40% actual uniform wall thinning, extending 0.75 inch in axial length, the burst pressure is 6800 psi. The NUREQ test data is obtained at a temperature of 600 P, compared to the Westinghouse data noted above, which is adjusted for 650 F.

The NUREG results at 40% actual thinning are comparable to the 4.0 volt BOC criteria expected burst pressure (6800 psi versus 6300 ps+ . The burst pressure for non-thinned tubes with partial depth cracks up to 0.75 inch in length is slightly lower than for uniform thinning up to depths of about 60%. Also, NUREG-0718 information can be used to estimate the burst pressure for a tube which has been slotted, simulating an axial crack. The expected burst pressure for a 40% deep, 0.75 inch long EDM slot using NUREQ-0718 is approximately 6000 psi. Therefore, it is judged that the margin of safety corresponding to the current 40'%y NDE depth based plugging criteria is not significantly reduced upon implementation of the bobbin coil voltage criteria at

.Cook Huclear Plant Unit 1.

Relative to the expected leakage during accident condition loadings, the accidents that are affected by primary-to-secondary leakage and steam release to the environment are:

feedwater system malfunction, loss of external electrical load and/or turbine trip, loss of all AC power to station auxiliaries, major secondary system pipe failure, steam generator tube rupture, reactor coolant pump locked rotor, and rupture of a control rod drive mechanism housing. Of these, the major secondary system pipe failure is the most limiting for Cook Huclear Plant Unit 1 in considering the potential for off-site doses. Upon implementation of the alternate plugging criteria, it will be verified on a cycle-by-cycle basis that the distribution of cracking indications at the tube support plate intersections is such that primary-to-secondary leakage would result in site boundary doses within a small fraction of the 10 CFR 100 guideline, i.e., 30 rem thyroid, during a postulated steam line break event.

Data indicates that a threshold voltage of 2.8 volts would result in through-wall cracks with the potential to leak at steam 'ine break (SLB) conditions. Application of the to AEP:NRC:1166 Page 6 proposed plugging criteria requires that the current distribution of number of indications versus voltage be obtained during each refueling outage. The indicated bobbin coil voltage is then combined with the rate of change in voltage measurement to establish an end-of-cycle (EOC) voltage distribution and, thus, leak rate during SLB pressure it differential. If is found that the projected SLB leakage for degraded intersections planned to be left in service exceeds 120 gpm, then additional tubes will be plugged to reduce projected SLB leakage below 120 gpm. Monte Carlo analyses results based on the Cook Nuclear Plant Unit 1 growth rate and assumed eddy current uncertainties indicate that over 4000 indications, all with a (BOC) bobbin coil voltage of 2.0 volts, 'ould contribute less than 1 gpm leakage at SLB conditions. Based on the inspection results from the last outage (1990), indications left in service are expected to have a total predicted SLB leak rate of 0.1 gpm at EOC conditions.

2) Create the ossibilit o a new or different kind of accident from an revious anal @ed Implementation of the proposed amendment does not introduce any significant changes to the plant design basis. Use of the criteria does not provide a mechanism that could result in an accident outside of the region of the tube support plate elevations. Neither a single nor multiple tube rupture event would be expected in a steam generator in which the plugging criteria has been applied (during all plant conditions). The bobbin coil signal amplitude plugging criteria is established such that neither operational leakage nor excessive leakage during a postulated steam line break condition are anticipated.

Indiana and Michigan Power Company will implement a maximum leakage rate limit of 150 gpd (0.1 gpm) per steam generator to help preclude the potential for excessive leakage during all plant conditions upon application of the alternate plugging criteria. The current technical specification limit on primary-to-secondary leakage at operating conditions is a maximum of 1.0 gpm (1440 gpd) for all steam generators or a maximum of 500 gpd for any one steam generator. The RG 1.121 criteria for establishing operational leakage rate limits that require plant shutdown are based upon leak-before-break considerations to detect a free span crack before potential tube rupture. The 150 gpd limit provides for leakage detection and plant shutdown in the event of the occurrence of an unexpected single crack resulting in leakage that is to AEP:NRC:1166 Page 7

'egulatory Guide 1.121 acceptance criteria (Item 3 of Section 3.2 of WCAP-13187) for establishing operating leakage limits are based on leak-before-break considerations such that plant shutdown is initiated if the leakage associated with the longest permissible crack is exceeded. The longest permissible crack is the length that provides a factor of safety of three against bursting at normal operating pressure differential, A voltage amplitude of 6.8 volts for typical ODSCC corresponds to meeting this tube burst requirement at the -95% LTL uncertainty limit on the burst correlation.

Alternate crack morphologies can correspond to 6.8 volts so that a unique crack length is not defined by the burst pressure versus voltage correlation. Consequently, typical burst pressure versus through-wall crack length correlations are used below to define the "longest permissible crack" for evaluating operating leakage limits.

The single through-wall crack lengths that result in tube burst at three times normal operating pressure differential and SLB conditions are about 0.44 inch and 0.84 inch, respectively. Nominal leakage for these crack lengths would range from 0.1 gpm to 4 gpm, respectively, while lower 95't confidence level leak rates would range from about 0.01 gpm to 0.5 gpm, respectively.

An operating leak rate limit of 150 gpd wi11 be implemented in application of the alternate tube plugging criteria. This leakage limit provides for detection of 0.4 inch long cracks at nominal leak rates and 0.6 inch long cracks at the -95t LTL confidence level leak rates. Thus, the 150 gpd limit provides for plant shutdown prior to reaching critical crack lengths for SLB conditions at leak rates less than a -95t LTL confidence level and for three times normal operating pressure differential at less than nominal leak rates.

3) Involve a si nificant reduction in a mar in of safet The use of the alternate plugging criteria at Cook Nuclear Plant Unit 1 is demonstrated to maintain steam generator tube integrity commensurate with the requirements of RG 1.121.

Regulatory Guide 1.121 describes a method acceptable to the NRC staff for meeting GDCs 14, 15, 31, and 32 by reducing the probability of the consequences of steam generator tube rupture. This is accomplished by determining the limiting conditions of degradation of steam generator tubing, as established by inservice inspection. Tubes with unacceptable cracking should be removed from service. The most limiting effect would be a possible increase in leakage during a steam to AEP;NRC:1166 Page 8 line break event. Once the alternate plugging criteria is applied, excessive leakage during a steam line break event is precluded by verifying each fuel cycle that the expected end-of-cycle distribution of crack indications at the tube support plate elevations would result in minimal and acceptable primary-to-secondary leakage during all plant conditions. This helps to demonstrate radiological conditions are less than a small fraction of the 10 CFR 100 guideline.

In addressing the combined effects of a loss-of-coolant accident (LOCA) and a safe shutdown earthquake (SSE) on the steam generator component (as required by GDC 2), it has been determined that tube collapse may occur in the steam generators at some plants. This is the case as the tube support plates may become deformed as a result of lateral loads at the wedge supports at the periphery of the plate due to the combined effects of the LOCA rarefaction wave and SSE loadings. The resulting pressure differential on the deformed tubes may cause some of the tubes to collapse.

There are two issues associated with steam generator tube collapse. First, the collapse of steam generator tubing reduces the RCS flow area through the tubes. The reduction in flow area increases the resistance to flow of steam from the core during a LOCA which, in turn, may potentially increase peak clad temperature. Second, there is a potential that partial through-wall cracks in tubes could progress to through-wall cracks during tube deformation or collapse.

Consequently, since the leak-before-break methodology is applicable to the Cook Nuclear Plant Unit 1 reactor coolant system primary loops, the probability of breaks in the primary loop piping is sufficiently low that they need not be considered in the structural design basis of the plant.

Excluding breaks in the RCS primary loops, the LOCA loads from the large branch line breaks were analyzed for Cook Nuclear Plant Unit 1 and were found to be of insufficient magnitude to result in steam generator tube collapse or significant deformation.

Regardless of whether or not leak-before-break is applied to the primary loop piping at Cook Nuclear Plant Unit 1, any flow area reduction is expected to be minimal (much less than 1t) and PCT margin is available to account for this potential effect. Analyses results show that no tubes near wedge locations are expected to collapse or deform to the degree to AEP:NRC:1166 Page 9 that secondary-to-primary in-leakage would be increased over current expected levels. For all other steam generator tubes, the possibility of secondary-to-primary leakage in the event of a combined LOCA and SSE event is not significant.

In actuality, the amount of secondary-to-primary leakage in the event of a combined LOCA and SSE is expected to be less than that currently allowed, i.e., 500 gpd per steam generator. Furthermore, secondary-to-primary in-leakage would be less than primary-to-secondary leakage for the same pressure differential since the cracks would tend to close under a secondary-to-primary pressure differential. Also, the presence of the tube support plate is expected to reduce the amount of in-leakage.

Addressing RG 1,83 considerations, implementation of the alternate plugging criteria is supplemented by 100%

inspection requirements at the tube support plate elevations having ODSCC indications, reduced operating leak rate limits, eddy current inspection guidelines to provide consistency in voltage normalization, and rotating pancake coil inspection requirements for the larger indications left in service to characterize the principal degradation mechanism as ODSCC.

As noted previously, implementation of the alternate plugging criteria will decrease the number of tubes which must be repaired or taken out of service by plugging. The installation of steam generator tube plugs reduces the RCS flow margin and, thus, implementation of the alternate plugging criteria will maintain the margin of flow that would otherwise be reduced in the event of increased tube plugging.

Based on the above, it is concluded that the proposed change does not result in a significant reduction in margin with respect to plant safety as defined in the Final Safety Analysis Report or any bases of the plant Technical Specifications.

CONCLUSION Based on the preceding analysis, it is concluded that using the tube support plate elevation bobbin coil signal amplitude alternate steam generator tube plugging criteria for removtng tubes from service at Cook Nuclear Plant Unit 1 is acceptable and the proposed license amendment does not involve a Significant Hazards Consideration as defined in 10 CFR 50.92.