ML20077P945
| ML20077P945 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/16/1991 |
| From: | Shelton D CENTERIOR ENERGY |
| To: | |
| Shared Package | |
| ML20077P944 | List: |
| References | |
| 1927, NUDOCS 9108200141 | |
| Download: ML20077P945 (9) | |
Text
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Docket Number 50-346 I
'. . License Number NPF-3 '
Serial Number 1927
. Enclosure Page 1 APPLICATION FOR AMENDMENT l TO !
FACILITY OPERATING LICENSE NUMBER NPF-3 DAVIS-BESSE NUCLEAR POVER STATION UN1T NUMBER 1 l
Attached are the requested changes to the Davis-Besse Nuclear Pover Station, Unit Number 1 Facility Operating License Number NPF-3. Also included is the ;
Safety Assessment and Significant flazards Consideration. '
The proposed changes (submitted under cover letter Serial Number 1927 CollC e rfil l Technical Specifications Section 3/4.4.5, Steam Generators Technical Specifications Bases 3/4.4.5, Steam Generators
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-1 By:
D.
I~tL//l 3 /'v '
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C.' Shel ton , Vice Presik rit -
l 1
Nuclear - Davis-Besse l 1
Sworn and subscribed before me on this 16th day of August, 1991. I Yhf 0h ' r AA Notary PublR, State of Ohio IIVELDl.. DIG $S NOIAMU30J OL\iC0F0HO f% D T M aSph4My4136j 910.3200141 910816 PDR ADOCK 05000346 P PDR
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Docket Number 50-346.
License Number NPF-3 Serial Number 1927-Enclosure Page 2 The following information is provided to support issuance of the requested change to the Davis-Besse Nuclear Pover. Station, Unit Number 1 Operating License Number NPF-3, Appendix A, Technical Specifications 4.4.5 and Bases 3/4.4.5.
A. Time Required to Implement: This change is to be implemented within 45 -
days after the NRC issuance of the License Arendment.
B. Reason for Change (License Amendment Request Number 91-0012): Revise the current TS and Bases to allow the use of Babcock and Vilcox-(B&V) sleeving process for Steam Generator tube repair.
C. Safety Assessment and Significant llazards Consideration: dee Attachment 1.
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Docket Numbac 50-346 :
I License Number NPF-3 Serial Number 1927 Attachment 1 >
Page 1 i
SAFETY ASSESSMENT AND SIGNIFICANT HAZARDS CONSIDERATION FOR ,
LICENSE AMENDMENT REQUEST NUMBER 91-0012 i
TITLE:
License Amendment Request (LAR) to modify Technical Specification (TS) 3/4.4.5, Steam Generators, to Allow the Use of the Babcock and Vilcox (B&V) ;
Sleeving method for Steam Generator Tube Repair.
DESCRIPTION: ;
The purpose of these proposed changes is to modify Davis-Besse Nuclear Power Station (DBNPS) Operating License NPF-3, Appendix A, Technical Specification ,
(TS) 3/4.4.5, Steam Generators. The proposed changes vould revise Surveillance Requircment (SR) 4.4.5 and TS Bases 3/4.4.5, Steam Generators, ;
to allow use of the B&V Sleeving method for steam generator (SG) tube ,
repair. Since 1994, B&V has installed more than 1400 Alloy 600 Inconel !
sleeves into Once-Through Steam Generators (OTSG).
Surveillance Requirement 4.4.5.4 states that a steam generator tube '
containing a defect is a defective tube. A defect is currently defined as an imperfection of such severity that it exceeds the plugging limit, which ,
is equal to 40 percent of the nominal SG tube vall thickness. Presently, ,
all tubes exceeding the plugging limit are taken out of service by plugging.
Repairs by means other than plugging are not currently addressed in the TS. ,
Toledo Edison proposes to utilize tube sleeving as an alternative to ~
plugging. The advantage of tube cieeving versus tube plugging is that the '
tube vill remain in service with-the structural integrity of the tube maintained and only a small reduction in flow and heat transfer capabilities. The repaired tube functions in the same manner as the :
original tube. The tute sleeving method referenced by this T.S. change is described in proprietary B&V Nuclear Services Company Topical Report .
BAV-2120P, OTSG 80" Hechanical Sleeve Qualification (Alloy 690). This )
topical report was submitted to the Nuclear Regulatory Commission (NRC) for !
approval by B&W 1etter dated March 26, 1991. This topical report addresses +
sleeve design, qualification, installation methods, non-destructive ,,
examinations and ALARA' considerations. Use of the B&V sleeving method is ,
contingent upon NRC approval of' Topical Report BAV-2120P.
The following TS changes are proposed to support the use of the B&V tube sleeving process:
- Change SR 4.4.5.2a.1 to note that the 100 percent inspection of ,
tubes or tube sleeves that previously had detectable vall pane- ,
i trations (>20%) does not include those that have had the detected vall penetrations (>20%) sleeve repaired; I
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Docket Number 50-346
'. . License Number NPF-3
. Serial Number 1927
. Attachment 1 Page 2 Add new SR 4.4.5.4a.1 to define Tubing or Tube to include a tube sleeve;
- Change SR 4.4.5.4a.3, Degraded Tube Criteria, to exclude a tube that has the affected area repaired by sleeving from being considered a degraded tube;
- Change SR 4.4.5.4a.5, Defect Criteria, to redefine Defect criteria to include tube sleeving as a repair method;
- Change the title of SR 4.4.5.4a.6 from " Plugging Limit" to " Repair Limit" and add tube sleeving as an alternative to tube plugging. Also add reference to B&V Topical Report BAV-2120P;
- Change SR 4.4.5.4b to allow sleeving as an alternative to plugging tubes that exceed the repair limit; Change SR 4.4.5.5b.3 to add a requirement to report repaired tubes as well as plugged tubes; Change TS Table 4.4-2 to include repair by sleeving as an alternative-to plugging:
Change the TS Bases 3/4.4.5, steam Generators, to add tube repair by sleeving and address degradation detection at the roll expanded
-areas and sleeve end.
SYSTEMS, COMPONENTS AND ACTIVITIES AFFECTED:
The system affected by these proposed changes is the OTSG. The components affected by these proposed changes are the OTSG tubes (there are over 15,000 tubes in each OTSG). The activity affected by these proposed changes is the repair of OTSG tubes.
SAFETY FUNCTIONS OF THE AFFECTED SYSTEM, COMPONENTS AND ACTIVITIES:
The OTSGs convert the thermal energy of the reactor coolant into steam for use in the turbine generator and act as a heat sink for the reactor. The OTSG tubes, acting as a major barrier against fission product release to' the environment, provide a flow path for the reactor coolant while transferring its heat to the secondary side fluid. Defects in the OTSG tubes are L presently repaired by plugging the tube in such a manner as to separate the l defective tube area from the reactor coolant, thereby preventing reactor coolant (which may contain fission products) from entering the secondary side.
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. Docket Number 50-346 License Number NPF-3 Serial Number 1927
-Attachment 1 Page 3 EFFECTS ON SAFETY:
The general tube sleeving procedure involves inserting a tube of smaller j diameter (the sleeve) inside the tube to be repaired. Sleeves span a !
defective or degraded region of a tube and are mechanically joined to the parent tube by a roll expansion at the tube's end areas, thereby maintaining j the OTSG tubing primary-to-secondary pressure boundary under normal-and I accident conditions. Thus, sleeving leaves the repaired tube functional.
This is in contrast to plugging, which removes the heat transfer surface of the plugged tube from service and reduces the reactor coolant system (RCS) flow available for reactor core cooling. The installation of a sleeve affects the heat transfer capability of the tube containing the sleeve to a j
lesser degree than plugging. Therefore, a large number of sleeves can be l installed without significantly affecting either reactor coolant system flow !
rate or plant operating efficiency (as compared to plugging), and the l service life of a OTSG that is experiencing degradation can be extended.
The principal accident associated with this proposed change is the steam l generator tube rupture accident. The environmental effects associated with I a severe steam generator tube rupture are discussed in the Updated Safety J
Analysis Report in Section 15.4.2, Steam Generator Tube Rupture. .For this ;
occurrence, fission products contained in the RCS would be released to the l secondary system. Some of the radioactive noble gases and iodine would be .
released to-the atmosphere through the ccndenser air removal system and l steam line safety valves. Use of the B&V tube sleaving process vill al'iov the DBNPS to repair degraded OTSG tubes such that the function and integrity of the tube is maintained. Topical Report BAV-2120p describes in detail the analytical methods used for design and qualification of the B&V tube sleeve.
The topical. report also contains the results of the sleeve design verification which included analysis and confirmatory testing to demonstrate the acceptability of the OTSG sleeving technique. The design and operating conditions (including transient conditions and cycles) specified for'the sleeve in the topical report bound the DBNPS OTSG design conditions.
The tube sleeve is spacifically designed to repair OTSG tubes which are exhibiting failures in the upper tubesheet, upper tube span, or at the 15th.
tube support plate. The tubes have a nominal outer diameter of 0.633 inches and a vall thickness of 0.034 inches. The sleeve has a 0.045 inch minimum vall thickness. The resulting nominal outer diameter (CD) is 0.525 inch.
Material for the sleeve is thermally treated Alloy 690 Inconel, which has superior corrosion resistance, compared to Alloy 600 Inconel, as detailed in the topical report.
The structural analysis of the sleeve demonstrates that its design meets the ASME Boiler and Pressure Vessel Code (ASME Code)Section III criteria for the design conditions of pressure, temperature, and flow listed in Topical Report BAV-2120P and establishes the minimum reactor coolant pressure boundary vall thickness requirements. Vibration testing and analysis were performed to demonstrate the adequacy of the sleeved tube for the 40 year life of the OTSG. Fatigue loadings used during the qualification testing of the sleeve joints were established per ASME Code requirements to veri'y the
Docket Number 50-346 License Number NPF-3
, Serial Number 1927 Attachment 1 Page 4 integrity of'the sleeve over the design life at the plant. Fatigue testing consisted of axial load cycling, vibration cycling, pressure cycling and thermal cycling. The sleeve is designed to accommodate all fatigue that the tubes may experience due to normal plant conditions and all anticipated
-transients specified for OTSGs, The tubesheet and free span joints are mechanical seals produced by roll expanding the sleeve into the tube. The structural integrity of the joints was proven by subjecting sleeve / tube specimens to a series of tests representing service conditions. These samples vere fatigue tested, tensile tested, thermal cycled, and leak tested to qualify the joints by experimental stress analysis per ASME Code Section III, Appendix 11.
Corrosion testing has demonstrated the corrosion resistance of the sleeve and the sleeve / tube joints.
The TS do not specifically limit the number of OTSG tubes which can be plugged while retaining acceptable primary flow rates. As discussed in Topical Report BAV-2120P, the thermal output effect of installing a total of 10,000 sleeves into an OTSG has been analyzed. The analysis also considered 5000 sleeves in each of the two OTSG's. In both cases, the calculated loss of OTSG steam'superheat was minimal. The amount of superheat produced remained greater than 50'F, vell above the minimum design value of 35' superheat. The reduction in the primary side flow rate is a maximum of 2.2 percent with all 10,000 sleeves installed into one OTSG. Utilizing sleeving as a tube repair method to this magnitude at the DBNPS is unlikely due to the excellent primary and secondary chemistry control programs at DBNPS.
This is evidenced by the small number of tubes which have previously required repair by plugging. The specific effect on primary side flow rate by the exact number of steam generator tubes sleeved would be. reviewed under 10CFR50.59.
An analysis has been performed by B&W in accordance with the guidelines of Regulatory Guide 1.121, Bases for Plugging Degraded PVR Steam Generator Tubes, to establish the sleeve defect plugging criterion. The plugging limit for the sleeve is calculated to be a 70 percent-through-vall defect, which is the same plugging limit as the unsleeved OTSG tube. However, to allow for possible defect growth between inspections, a plugging limit of 40 percent of the original sleeve vall is established.
l l The DBNPS currently has a program for inservice inspection of OTSG tubes l based on a modification of Regulatory Guide 1.83, Revision 1, Inservice l
Inspection of Pressurized Vater Reactor Steam Generator. Tubes. Routinely scheduled inservice OTSG tube examinations are utilized to detect defects which develop.
Eddy current testing (ECT) is used to detect the presence of defects in OTSG tubes. Eddy current testing detects the presence of defect-caused variations in the effective electrical conductivity and/or magnetic permeability of the tubes. Prior to installing a sleeve, the parent tube vill be inspected by ECT to establish the condition of the tube. Following ,
the sleeves' insertion, the sleeve vill be inspected again using ECT to establish a reference baseline.
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Docket Number 50-346 i License Number NPF-3 Serial Number 1927
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-Attachment 1 Page 5 l
Eddy Current Testing can be used to detect defects as small as 20 percent through-vall penetration in all areas of the sleeve and tube except the roll expanded areas and the sleeve ends. This is due to the change in tube / sleeve diameters in these locations. .The reduced detection capability at the sleeve ends is considered acceptable primarily because: 1) this region of the tube sleeve is inspected prior to sleeve installation when the 20 percent through-vall sensitivity is available, 2) a baseline ECT signal is determined immediately after sleeve installation such that any subsequent change in the signal can be evaluated should it occur, and'3) the 40 percent through-vall degradation is detectable, and since this is the plugging. criterion, no action (other than routine reporting of the
" degraded" tube) is required in the affected range of the 20-40 percent through-vall. The DBNPS will evaluate, and as appropriate, implement better testing methods which are developed and validated for commercial use so as to enable detection of degradation as small as 20 percent through vall without exception. Until such time as 20 percent penetration can be-detected in the roll expanded areas and sleeve ends, inspection results vill be compared to those obtained during the baseline sleeved tube inspection. I l
-Defects which have been spanned by a sleeve need not be considered for determination of inspection result categories per SR 4.4.5.2, Steam Generator Tube Sample Selection and Inspection. For the case in which the degraded _ tube has_been spanned by a sleeve, further tube vall penetrations ,
in the parent tube (from the bottom of the uppermost -rolled joint to the top of the lower-most rolled joint) are considered inconsequential since that portion of the tube no longer constitutes the reactor coolant pressure boundary. Any degradation in the parent tube in the area spanned by the sleeve does not affect the integrity of the pressure boundary and therefore, does not require the same degree of scrutiny as a vall penetration greater than 20 percent in a portion of the tube that does constitute the pressure boundary. The-mandatory inspection requirement still applies to a sleeved tube which has been subjected to a random full length examination and has been found to have a vall penetration greater than 20 percent in either the portion of the tube which is not spanned by the sleeve or in the sleeve itself.
Accordingly, based on the above discussion, it is concluded that there is no adverse effect on safety by revising the Technical Specifications to allow use of the B&V sleeving method as described in Topical Report BAV-2120P for OTSG tube repair.
SIGNIFICANT HAZARDS CONSIDERATION The Nuclear Regulatory Commission has provided standards in 10 CFR 50.92(c) for determining whether a significant hazard exists due to a proposed amendment to an Operating License for a facility. A proposed amendment involves no significant hazards consideration if operation of the facility in accordance with the proposed changes vouldt (1) Not involve a significant increase in the probability or consequences of an accident previously evaluated; (2) Not create the possibility of a new or different
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. . Docket Number 50-346 i License Number NPF-3 Serial Number 1927 l Attachment 1 1 Page 6
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kind-of accident from any accident previously evaluated; or (3) Not !nvolve i a significant reduction in a margin-of safety. Toledo Edison has reviewed l the proposed change and determined that a significant-hazards consideration i does not exist because operation of the Davis-Besse Nuclear Power Station, _j Unit Number 1, in accordance with these changes vould: l la. Not involve a significant increase in the probability of an accident :
previously evaluated because, as discussed in Topical Report- :
BAV-2120P, the OTSG tube sleeve has been analyzed and tested to i operating and design conditions which bound the original tube; the j structural integrity _of the OTSG tubes is maintained by the sleeving .,
process; the sleeve is less susceptible to the identified corrosion
_l failure mechanisms of the original tube because of the use of an ';
improved material, Alloy 690 Inconel; and the continued integrity of l the sleeve vill be verified by the TS OTSG inspection requirements. !
Therefore, there is no significant increase in the probability of i the steam generator tube rupture accident which is associated with ;
the proposed changes.
lb. Not involve a significant increase in the consequences of an i accident previously evaluated because, at discussed in Topical- i Report BAV-2120P, the OTSG tube sleeve has been analyzed and tested i
to operating and design conditions which bound the original tube. ;
The steam generator tube rupture accident and its consequences are ?
associated with these proposed changes. The structural integrity of :
the OTSG tubes for containing fission products is maintained by the i sleeving method. In addition, the sleeve is less susceptible to the i identified corrosion failure mechanisms of'the original tube because !
of the use of an improved. material, Alloy 690 Inconel. The continued integrity of the sleeve vill be verified by the TS OTSG [
inspection requirements and plugging of a defective sleeve vill'be l available. Accordingly, there is no significant increase in radiological consequences of an accident previously evaluated.
2a. Not create the possibility of a new kind of accident from any- h accident previously evaluated because the OTSG tube sleeve functions is essentially the same manner as the original tube. The purpose of +
sleeve repair is to repair a degraded-0TSG tube in order to maintain the function and integrity of the tube. The sleeve repair method ;
has been analyzed and tested and meets OTSG design conditions, i Continued tube sleeve integrity is verified by TS inspections. ;
Repairing.the tube to a serviceable condition utilizing this sleeving method does not create the possibility of a new kind of ,
accident. }
2b. Not create the possibility of a different kind of accident from any ,
accident previously evaluated because the OTSG tube sleeve functions l 1s essentially the same manner as the original tube. The purpose of l
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the sleeve repair is to repair a' degraded OTSG tube in order to j maintain the function and integrity of the tube. The epaired i sleeve has been analyzed and tested and meets OTSG design conditions. Continued tube sleeve integrity is verified by TS :
t- inspections. Repairing the tube to a serviceable condition utilizing this sleeving method does not create the possibility of a different kind of accident.
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. Docket Number 50-346 License Number NPF-3 Serial Number 1927 Attachment 1 Page 7
- 3. Not involve a significant reduction in a margin of safety because r the-reactor pressure boundary of the tube is maintained by the installation of the sleeve. The sleeve plugging limit has been calculated to be a 70 percent through-vall defect which is the same plugging limit as the unsleeved 0TSG tube. Defects are detectable at 40 percent in the tube sleeve. Therefore, there is no significant reduction in a margin of safety. The thermal output effect of installing tube sleeves has been analyzed (as described in Topical Report BAV-2120P) with a slight reduction in RCS flov and full-power steam superheat shown to result, llovever , this reduction is significantly less than that of a tube that has been plugged.
Therefore, there is no significant reduction in a margin of safety.
CONCLUSION:
On the basis of the above, Toledo Edison has determined that the License Amendment Request does not involve a significant hazards consideration. As this License Amendment Request concerns a proposed change to the Technical Specifications that must be reviewed by the Nuclear Regulatory Commission, this Litense Amendment Request does not constitute an unreviewed safety question.
ATTACHMENT:
Attached are the proposed marked-up changes to the Operating License, f