B17341, Proposed Tech Specs Surveillance 4.4.5.3.a Re SG Tube Insp Interval

Proposed Tech Specs Surveillance 4.4.5.3.a Re SG Tube Insp Interval

ML20236Y499
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Site:	Millstone
Issue date:	08/06/1998
From:	NORTHEAST NUCLEAR ENERGY CO.
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ML20236Y497	List: B17341, Proposed Tech Specs Surveillance 4.4.5.3.a Re SG Tube Insp Interval ML20236Y495
References
B17341, NUDOCS 9808120262
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Docket No. 50-423 B17341 Attachment 2 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification Steam Generator Tube Inspection interval (TSCR 3-17-98)

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l U.S. Nucl: r R:gul: tory Commission

' 17341\ Attachment 2\Page 2 B

MARKUP OF PROPOSED REVISION Refer to the attached markup of the proposed revision to the Technical Specifications. The attached markup reflects the currently issued version of the Technical Specifications listed below. Pending Technical Specification revisions or Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed markup.

The following Technical Specification changes are included in the attached markup.

. Technical Specification Surveillance 4.4.5.3.a is reworded to identify that the surveillance interval is extended until the next refueling outage or July 1, 1999, which ever date is earlier.

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REACTOR COOLANT SYSTEM STEAM GENERATORS SURVEILLANCE REQUIREMENTS (Continued) .12u m

4.4.5.3 Jnspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a. I Inservice inspections shall be perfqrmed at intervals of not less than 12 nor more than 24 calendar monthsRafter the previous inspection. If

' two consecutive inspections, not including the preservice inspection, l result in all inspection results falling into the C-1 category or if two ,

consecutive inspections demonstrate that previously observed degradation l has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.;

the interval may then be extended to a maximum of once per 40 months; and

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of l the following conditions: 1 l

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1) Primary-to-secondary tubes leak (not including leaks originating h. .i tube-to-tube sheet welds) in excess of the limits of Specification '

3.4.6.2, or 4

2) A seismic occurrence greater than the Operating Basis Earthquake, or

3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or i

4) A main steam line or feedwater line break.

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MILLSTONE - UNIT 3 3/4 4-16 Amendment No. EZ, E9 l

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Docket No. 50-423 B17341 Attachment 3 Mitistone Nuclear Power Station, Unit No. 3 I Proposed Revision to Technical Specification Steam Generator Tube inspection Interval (TSCR 3-17-98)

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August 1998 j

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l U.S. Nucl: r Regul: tory Commission B17341%tt: chm:nt 3\Pcg31 RETYPE OF PROPOSED REVISION Refer to the attached retype of the proposed revision to the Technical Specifications.

l The attached retype reflects the currently issued version of the Technical Specifications. Pending Technical Specification revisions or Technical Specification revisions issued subsequent to this submittal are not reflected in the enclosed retype.

The enclosed retype should be checked for continuity with Technical Specifications prior to issuance.

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REACT 0]LHf1. ANT SYSTEM STEAM 6ENERATORS SURVEILLANCE REQUIREMENTS (Continued) 4.4.5.3 Inspection Frequencies - The above required inservice inspections of steam generator tubes shall be performed at the following frequencies:

a. Inservice inspections shall be performed at intervals of not less than 12 nor more than 24 calendar months

• after the previous inspection. If l two consecutive inspections, not including the preservice inspection, result in all inspection results falling into the C-1 category or if two consecutive inspections demonstrate that previously observed degradation has not continued and no additional degradation has occurred, the inspection interval may be extended to a maximum of once per 40 months;

b. If the results of the inservice inspection of a steam generator conducted in accordance with Table 4.4-2 at 40-month intervals fall in Category C-3, the inspection frequency shall be increased to at least once per 20 months. The increase in inspection frequency shall apply until the subsequent inspections satisfy the criteria of Specification 4.4.5.3a.;

the interval may then be extended to a maximum of once per 40 months; and

c. Additional, unscheduled inservice inspections shall be performed on each steam generator in accordance with the first sample inspection specified in Table 4.4-2 during the shutdown subsequent to any of the following conditions:

1) Primary-to-secondary tubes leak (not including leaks originating from tube-to-tube sheet welds) in excess of the limits of Specification 3.4.6.2, or

2) A seismic occurrence greater than the Operating Basis Earthquake, or

3) A loss-of-coolant accident requiring actuation of the Engineered Safety Features, or

4) A main steam line or feedwater line break.

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• Except the surveillance related to Steam Generator Inspection, due no later than September 24, 1998, may be deferred until the next refueling outage or no later than July 1,1999, whichever is earlier.

l MILLSTONE - UNIT 3 3/4 4-16 Amendment No. E2, ES, Jpp, l 0596 I

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Docket No. 50-423 B17341 Attachment 4 Millstone Nuclear Power Station, Unit No. 3 Proposed Revision to Technical Specification Steam Generator Tube inspection interval (TSCR 3-17-98) i Background and Safety Summary August 1998 l

U.S. Nuciser Regulttory Commission

'B17341\ Attachment 4\Page 1 Background I The latest Millstone Unit No. 3 steam generator tube inspection began on )

September 24,1996, and was complete on October 1,1996. The inspection i results placed the steam generators in category C-2. Technical Specification f Surveillance 4.4.5.3.a establishes an allowable inspection interval of 24 {

calendar months. Without an extension of the interval, Millstone Unit No. 3 must shutdown prior to September 24,1998. This proposed revision requests a one

. time extension to the surveillance interval until the next refueling outage.

Steam generator tubes form a significant portion of the reactor coolant system pressure boundary. Steam generator tube surveillance requirements, including inspection frequencies, were established to ensure that the structural integrity of i the tubes is maintained during plant operation. It can be postulated that extending the inspection interval could allow tube degradation to progress to an extent that would deviate from the guidance of Regulatory Guide 1.121.

However, the discussion below provides the basis for concluding that this will not occur with the interval extended to accommodate the anticipated start date of the f next refueling outage.

SAFETY

SUMMARY

Millstone Unit No. 3 Westinghouse Model F steam generators incorporate second generation design features that provide improved rc:htance to tube degradation mechanisms which have been experienced in prev ous steam generator designs. Notable features include:

e thermally treated inconel 600 tubing which provides improved )

intergranular attack (IGA) and intergranular stress corrosion cracking l (IGSCC) resistance, I e thermal stress relief of short radius u-bends and full depth hydraulic tubesheet expansions which reduce stresses and therefore reduced IGSCC susceptibility, i

e increased and more efficient blowdown capacities, and l

• broached type 405 stainless steel tube support plates which essentially precludes corrosion related tube denting.

Six inservice inspections of the Millstone Unit No. 3 steam generator tubes have been performed to date. Wear adjacent to antivibration bars (AVBs) in the u-bend region of the bundle, and foreign object related wear are the only active damage mechanisms affecting the steam generator tubes. AVB wear is the 1

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U.S. Nucicer R;gulttory Commission

• B17341\ Attachment 4\Page 2 result of tube /AVB impact caused by vibration of the u-bend portion of the tube bundle. Foreign object wear is the result of vibratory interaction between the tube and foreign object. In both cases, the vibration is flow-induced and only occurs while the unit is operating. Extending the calendar duration of Cycle 6 has no effect on the extent or severity of wear in the steam generators, since the j total operating time durirs cycle 6 will not be extended by this proposed I revision. With no change in wear extent or severity, there will be no increase in the likelihood of deviating from the guidance of Regulatory Guide 1.121.

I Additionally, the potential development of primary or secondary side corrosion, as a result of the extended shutdown period, has also been considered. ,

Secondary side intergranular attack (IGA), secondary side intergranular stress l corrosion cracking (IGSCC), primary side IGSCC and pitting are the principal damage mechanisms that were considered; The initiation and advancement of IGA and IGSSC are strongly dependent upon temperature and typically develop after many years at operating temperatures. At shutdown temperatures, no initiation or advancement of these mechanisms is expected to occur.

• Pitting is generally considered to be a high temperature phenomenon, although laboratory data has shown that pitting can initiate at low ismperatures in the presence of faulted water conditions. Secondary chemistry layup guidelines have been established by the industry to minimize the potential for corrosion of steam generator tubes and support structures during non-operational periods. Controlled wet layup chemistry was maintained in accordance with these guidelines throughout the prolonged mid-cycle 6 outage. Although valve repair work did not allow Millstone Unit No. 3 to consistently maintain nitrogen overpressure within the steam generators, the tube bundles were continuously covered with wet layup solution with the exception of a three month period in the Fall of 1996. During the three month period, the water level in steam generator l C was lowered, exposing the u-bend region of the tube bundle. It is

!' known that nitrogen overpressure was maintained during a portion of this t

period, but it cannot be confirmed that it was present during the entire period. In spite of this, tube. pitting is very unlikely to have developed either at the vapor / water interface or in free spans and crevices above the interface under these conditions. In the region adjacent to the vapor / water interface, the presence of controlled layup solution, with its high pH and hydrazine content, is expected to prevent the development of pitting. Similarly, any moisture present above the interface would be residual controlled layup solution, providing protection for these regions.

Once dry, there is no potential for pitting in these areas. In summary,

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U.S. Nucirr R:gul: tory Commission

'B17341\ Attachment 4\Page 3 l pitting is very unlikely to have developed in any region of the tube bundles during the extended shutdown period.

l . Millstone Unit No. 3 was shutdown from April 1996 to June 1998. None of the Millstone Unit No. 3 steam generator tube inspections, including the most recent inspection completed in October 1996 conducted after approximately 6 months in lay-up, have revoaled indications of tube corrosion of any kind. Further, no evidence of tube corrosion was detected during any of the other five inservice inspections performed at Millstone Unit No. 3.

I The proposed revision will not affect any Design Basis Accidents or their consequences, it will not contribute to any new accidents. The only active damage mechanism, affecting the SG tubes, is vibration wear (e.g., adjacent to l an antivibration bar) that occurs during power operation. Since this extension will not increase the actual operating time of Millstone Unit No. 3, the vibration wear will not be increased. Therefore, the extension of the surveillance frequency will not increase the likelihood that a tube or tubes will deviate from i the guidance of Reg Guide 1.121. This change is safe and does not present an I Unreviewed Safety Question.

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s Docket No. 50-423 B17341 Attachment 5 Millstone Nuclear Power Station, Unit No. 3 Steam Generator Tube Inspection Interval (TSCR 3-17-98)

Significant Hazards Consideration and Environmental Considerations l

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U.S. Nucl;ar R:gulatory Commission B17341\ Attachment 5\Page 1 Significant Hazards Consideration NNECO has reviewed the proposed revision in accordance with 10CFR50.92 and has concluded that the revision does not involve a significant hazards consideration (SHC).

The basis for this conclusion is that the three criteria of 10CFR50.92(c) are not satisfied. The proposed revision does noi involve a SHC because the revision would not:

1. Involve a significant increase in the probability or consequence of an accident previously evaluated.

This proposed revision to Technical Specification 4.4.5.3.a for a one time extension to the surveillance interval until the next refueling outage will not increase the potential to impact steam generator tube integrity by allowing a steam generator tube to be degraded and go undetected. The only active damage mechanism, affecting the steam generator tubes is vibration wear adjacent to an antivibration bar that occurs during power operation. Since this surveillance interval extension will not increase the actual plant operating time, the vibration wear will not be increased. If there is no increase in tube degradation, there will be no increase in the probability of occurrence or consequence of a Steam Generator Tube Rupture. The failure of a Steam Generator tube is evaluated within Final Safety Analyses Report Section 15.6.3 and fully bounds this proposed surveillance interval extension.

Thus it is concluded that the proposed revision does not involve a significant increase in the probability or consequence of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any accident previously evaluated.

l This proposed revision to the surveillance interval does not change the operation of any plant system or component during normal or accident conditions. The Final Safety Analyses Report evaluation for a failure of a Steam Generator tube bounds this proposed surveillance interval extension.

Thus, this does not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Involve a significant reduction in a margin of safety.

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The proposed revision to Technical Specification 4.4.5.3.a for a one time extension to the surveillance interval until the next refueling outage will not

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• .s U.S. Nucl:ar Regulatory Commission B17341%ttrchm:nt 5\PIga 2 deviate from the guidance of Reg Guide 1.121. The active damage mechanism resulting in Steam Generator tube degradation currently experienced at Millstone Unit No. 3 has been primarily anti-vibration bar wear and is dependent on power operation. Since this extension will not increase the actual plant operating time, the vibration wear will not be increased.

Thus, it is concluded that the proposed revision does not involve a significant reduction in a margin of safety.

In conclusion, based on the information provided, it is determined that the proposed revision does not involve an SHC.

Environmental Considerations NNECO has reviewed the proposed license amendment against the criteria cf 10CFR51.22 for environmental considerations. The proposed revision does not involve a SHC, does not significantly increase the type and amounts of effluents that may be released offsite, nor significantly increase individual or cumulative occupational radiation exposures. Based on the foregoing, NNECO concludes that the proposed revision meets the criteria delineated in 10CFR51.22(c)(9) for categorical exclusion from the requirements for environmental review.

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