ML20071Q165
| ML20071Q165 | |
| Person / Time | |
|---|---|
| Site: | Sequoyah  |
| Issue date: | 08/01/1994 |
| From: | Shell R TENNESSEE VALLEY AUTHORITY |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| Shared Package | |
| ML20071Q168 | List: |
| References | |
| NUDOCS 9408110108 | |
| Download: ML20071Q165 (10) | |
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te n m..y o,,. 3 rw esce m. an <,u, og, s m ne van August 1, 1994 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of
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Docket No. 50-328 Tennessee Valley Authority
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SEQUOYAH NUCLEAR PLANT (SQN) - ADDITIONAL INFORMATION FOR TECHNICAL SPECIFICATION (TS) 94-09, ALTERNATE PLUGGING CRITERIA FOR STEAM GENERATOR (S/G) TUBING - UNIT 2
Reference:
TVA letter to NRC dated June 28, 1994, " Technical Specification (TS) Change 94-09, Alternate Plugging Criteria for Steam Generator (S/G) Tubing - Unit 2" On July 6, 1994, a telephone conference was held with NRC staff for the purpose of discussing issues relating to TVA's referenced TS j
Change 94-09.
The staf f requested the following information to support l
their review of TS Change 94-09.
1.
A comparison of the SQN interim plugging criteria (as contained in Westinghouse Commercial Atomic Power 13990) with the Farley Nuclear Plant Safety Evaluation Report for Unit 1.
2.
SQN's S/G tube pull plan.
Enclosures 1 and 2 provide the above Items 1 and 2, respectively;
) provides a TVA commitment associated with the implementation j
of SQN's S/G tube pull plan.
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U.S. Nuclear Regulatory Commission Page 2 August 1, 1994 In addition, Enclosure 4 provides a revised commitment page that supersedes the commitment page from Enclosure 4 of TVA's referenced TS Change 94-09 letter. The revised commitment page includes three new commitments (Items 7, 8, and 9) that are based on the additional information contained in this submittal.
Please direct questions concerning this issue to D. V. Goodin at (615) 843-7734.
Sincerely, Y ff,
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R.11. Shell Manager SQN Site Licensing Enclosures cc (Enclosures)*
Mr. D. E. LaBarge, Project Manager i
U.S. Nuclear Regulatory Commission One White Filnt, North 11555 Rockville Pike Rockville, Maryland 20852-2739 NRC Resident Inspector Sequoyah Nuclear Plant 2600 Igou Ferry Road Soddy-Daisy, Tennessee 37379-3624 Regional Administrator U.S. Nuclear Regulatory Commission Region II i
101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323-2711
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ENCLOSURE 1 COMPARISON OF THE SEQUOYAH NUCLEAR PIANT INTERIM PLUGGING CRITERIA AS CONTAINED IN WESTINGHOUSE COMMERCIAL ATOMIC POWER 13990 WITH THE FARLEY NUCLEAR PLANT - UNIT 1 SAFEIT EVALUATION REPORT 4
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Comparison of Sequoyah IPC WCAP and Farley-1 NRC IPC SER Item Farley-1 SER Sequoyah IPC WCAP-13990 Repair limits Repair RPC confirmed ind. >2.0 volts No change and s3.6 volts. Repair bobbin flaw ind. >3.6 volts independent of RPC confirmation.
Operating leakage 150 gpd for voltage based IPC, 150 gpd limit although 140 gpd at Farley-1 for reasons unrelated to IPC.
Bobbin inspection 100% of hot and cold leg TSP 100% of hot leg TSP intersections intersections. EC guidelines typical and all cold leg intersections to of Appendix A.
lowest TSP where ODSCC has been identified.' EC guidelines of App. A.
RPC inspection All bobbin ind. >l.5 volt and <2.0 Same volt. Licensee stated allind. >1.5 volts will be RPC inspected.
RPC sample RPC inspect 100 tubes including Same, except WCAP commits to inspection dents > 5 volts and artifact in>pecting at least 100 TSP indications. Licensee committed to intersections including dents > 5.0 inspect all dents > 5.0 volts.
volts.2 Report unexpected indications Expansion, if required, to be based on such as circumferential or outside the nature and number ofindications.
TSP prior to next cycle.
Report unexpected indications such as circumferential or outside TSP prior to next cycle.
Allowable SLB 5.7 gpm based on SRP methods and 4.3 gpm based on SRP methods and leakage ICRP-2 thyroid dose conversion.
ICRP-30 thyroid dose conversion.
Increased to 22.8 gpm by factor of 4 No reduction in coolant activity.
coolant activity reduction.
BOC bobbin Allindications including RPC NDD Same distribution for and non-detected indications based on SLB analyses draft NUR EG-1477 method with POD = 0.6.
Growth data in Most recent growth rate data.
Same, unless less than about 100 SLB analyses indications, in which case EPRI bounding growth distribution would be applied if more limiting.
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- See enclosure 4 item 7 of this submittal.
See enclosure 4 item 8 of this submittal page 1 of 2
l Item Farley-1 SER Sequoyoh IPC WCAP-1399t Database for burst Most recent correlation including all EPRI database recommended for pressure data, unless a specific error in burst burst correlation. Correlation also correlation test or voltage measurement (called provided for NRC database.
NRC database).
Differences in databases outlined in Section 5.1 of WCAP. WCAP commits to using database accepted by the NRC at the time ofIPC approval.
Database for leak Most recent leak rate data should be EPRI database recommended and rate analysis used for leakage analysis - it is also is more conservative than NRC assumed that this implies NRC database for draft NUREG-1477 leak database including all data without rate methods. Differences in EPRI known test errors.
vs. NRC database outlined in Section 5.1 of WCAP. WCAP commits to using database accepted by NRC.
Probability of Most conservative relative to leakage Log logistic POL recommended for leakage of six forms for POL should be used comparing leakage with allowable correlation in predicting SLB leakage.
limits. Other 5 forms for POL to be provided as sensitivity analyses.
SLB leak rate SLB leakage to be calculated using Same, unless EPRIleak rate versus analysis method draft NUREG-1477 methods.
voltage correlation accepted by NRC prior to IPC approval.
Reporting of SLB SLB leak rate to be reported prior to Projected EOC SLB leak rate must be analysis results restart from refueling outage. SLB shown <4.3 gpm for limiting SG.
burst probability to be reported Additional tubes to be plugged if following completion of the outage.
necessary to meet limit. SLB burst prob. to be calculated and compared to 2.5E-02 value of NUREG-0844. '
IPC assessment Perform an assessment following Not applicable to first time IPC i
report completion of the outage on the implementation such as Sequoyah.
efTectiveness of the IPC methodology and address differences between l
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Tube pull No additional tube pull data required Not addressed in WCAP. (See i
requirements due to prior tube pull data obtained.
enclosure 2 of this submittal.)
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2 See enclosure 4 item 9 of this submittal.
page 2 of 2 1
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a ENCLOSURE 2 SEQUOYAH NUCLEAR PLANT STEAM GENERATOR TUBE PULL PROGRAM i
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Sequoyah Nuclear Plant - Unit 2 F
Steam Generator Tube Pull Program in support of Technical Specification Change 94-09 Upon implementation of the Technical Specification Change 94-09, SON will initiate a tube pull program which is consistent with the EPRI draft " Steam Generator Degradation Specific Management (SGDSM) Tube Pull Program" and consists of the following:
A. Initial steam generator tube pull during Unit 2 Cycle 6 will consist of a minimum of two tubes and a minimum of six tube support plate intersections:
- 1. Tubes will be selected from those locations (excluding peripheral tubes and tubes that are not capable of being pulled) that exhibit the largest accessible indications and those tubes with indications at multiple tube support plate (TSP) locations. Tube selection will focus on large voltage indications (single dominant crack) versus a network of cracks
-l (two or more) about the circumference.
- 2. The goal for each tube pulled is to remove three TSP intersections, which should include TSP intersections with no detectable degradation. However, if an unusually large indication is found at the first or second TSP intersection, the tube would be cut above the large indication to minimize the potential for damage to tha intersection from the pulling operations.
- 3. Intersectiono exhibiting degradation shall be evaluated using pre-and post-pull nondestructive examination (NDE) and metallurgical examination to assess the crack morphology and verify the outside diameter stress corrosion cracking (ODSCC) degradation mechanism.
- 4. Pre pull NDE and laboratory NDE examinations shall be performed on all pulled tube sample intersections to evaluate the degradation morphology.
- 5. Each tube support plate intersection exhibiting degradation will be burst and leak tested.
- 6. Tube support plate intersections that do not exhibit degradation by NDE will be further examined for degradation.
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- 7. Leak rate tests will be performed to obtain data at primary-and secondary-side pressure and temperature conditions corresponding to normal operation and postulated faulted events. Leak rate tests shall l
be conducted using approved standard procedures developed for SGDSM implementation.
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- 8. Burst tests shall be conducted to obtain the maximum pressure i
capacity of the degraded tube. Burst tests shall be conducted using approved standard procedures developed for SGDSM implementation.
- 9. The results from all leak and burst tests shall be made available for inclusion in the industry ODSCC data base maintained by EPRI.
B. Subsequent outage steam generator tube pulls after the Unit 2 Cycle 6 refueling outage will be as follows:
- 1. Pull one accessible tube with three intersections for leak rate and burst testing during alternating refueling outages as necessary to obtain at least one intersection in each of the voltage regions (2 to 4 volts,4 to 8 volts, and 8 to 15 volts).
- 2. The following alternative tube pull schedules may be implemented to obtain leak and burst data: (a) the time for tube pulls may be extended to the next outage if the degradation at any outage is not extensive enough to provide data in any of the indicated voltage regions, or (b) more than one tube may be pulled in an outage when the degradation at any outage extends into more than one of the voltage regions.
- 3. No additional tube pulls are necessary once data have been obtained to verify the degradation mechanism, and one leak rate and one burst test data point have been obtained for each of the three indicated voltage. regions, except for tubes with unanticipated voltage levels substantially higher than those expected for plant specific implementation of the SGDSM criteria for ODSCC.
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ENCLOSURE 3 t
TVA COMMIIMENT i
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TVA Commitment Upon implementation of Technical Specification (TS) Change 94-09 for Unit 2, SQN will initiate a steam generator (S/G) pull program, as described in Item A of Enclosure 2.
The S/G tube pull program will be implemented before restart from the Unit 2 Cycle 6 refueling outage.
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