ML20141C444
| ML20141C444 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 04/02/1986 |
| From: | Westafer G FLORIDA POWER CORP. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20141C450 | List: |
| References | |
| 3F0486-02, 3F486-2, NUDOCS 8604070250 | |
| Download: ML20141C444 (17) | |
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e; Power C ORPCR AT TON April 2, 1986 3F0486-02 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555
Subject:
Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Snubber Optimization Approval Request
Dear Sir:
to this letter references thirteen documents, transmittals, and meetings since the Fall of 1984 related to the Florida Power Corporation (FPC) plans to optimize the reactor coolant (RC) snubber arrangement for Crystal River Unit 3 (CR-3). These plans are based on the leak-before-break (LBB) concept as applied te the primary system of pressurized water reactors which is soon to be formalized as a revision to General Design Criterion 4 (GDC-4).
At the February 27, 1986 meeting (Reference 13), ten questions raised by the NRC staff reviewers were discussed among representatives of NRC, FPC, Babcock and Wilcox ( B&W), Brookhaven National Laboratory, and Gilbert Commonwealth. These ten NRC questions and the elements of response to each question as discussed in the February 27, 1986 meeting are shown in.
All but questions 1, 2, 3, and 5 were considered resolved at the meeting. Subsequent review by the NRC of a Franklin Institute report resolved questions 1 and 2.
Question 3 was resolved after an NRC review which indicated that the approach used by B&W appeared to be the same as used by other nuclear steam supply system manufacturers.
An additional study / analysis was performed by B&W to provide supplementary information on Question 5.
Results of this study and analysis are included as Attachment 3.
On March 19, 1986, the NRC requested additional information on questions 9 and 10 which is included as Attachment 4.
We believe the information provided the NRC since the Fall of 1984 and further supplemented by this transmittal should constitute a sufficient technical basis for NRC's completion and issuance of the Safety Evaluation Report.
As of April 1,
- 1986, installation of the CR-3 optimized snubber / link bar configuration has been completed for RC pumps A and B and d
is underway for RC pumps C and D.
8604070250 060402 00 3 ADOCK 05 g2 g i gDR GEN ERAL OFFICE 3201 Thirty-ft,..o. oumn soutn e P.O. Box 14o42, St. Petersburg, Florida 33733 e 813-866-5151
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April 2, 1986 3F0486-02 Page 2 With this transmittal, FPC has completed its response to all outstanding questions from the NRC related to the pump support configuration planned for use at CR-3.
We request NRC endorsement of these plans by April 30, 1986 to provide an orderly startup for CR-3.
Sincerely, e
G. R. Westafer Manager, Nuclear Operations Licensing and Fuel Management EHD/feb Attachments 4
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i FPC 3F0486-02 ATTACHMENT 1 LIST OF REFERENCES
i ATTACHMENT 1 List of References 1.
B&W Report, BAW 1847, dated October 1984, subject the B&W Owners Group Leak-Before-Break Evaluation of Margins Against Full Break for RCS Primary Piping of B&W Designed NSS.
2.
FPC letter to NRC, Westafer to Denton, dated February 1,
1985 (3F0285-02),
subject Request for Exemption from a
Portion of 10 CFR 50, Appendix A, General Design Criterion 4 (GDC-4).
3.
Meeting on August 5, 1985 among NRC, Babcock and Wilcox, and FPC representatives to present FPC plans and define NRC staff needs for information to support the FPC request of Reference 2 above.
4.
FPC letter to NRC, Westafer to Denton, dated August 30, 1985 (3F0885-24), subject Re-evaluation of CR-3 Reactor Cooling System Loads Utilizing Leak-Before-Break Concept to Remove Reactor Coolant Systen Main Loop Pipe Break Protective Devices; transmitted B&W Report prepared for FPC, subject Evaluation of Reactor Coolant System Loads and Component Support Margins Resulting f rom 'Optimi zed Reactor Coolant Pump Support Configuration.
5.
FPC letter to NRC, Simpson to Denton dated September 27, 1985 (3F0985-26), subject Transmittal of Report Related to Request for Exemption from a Portion of 10 CFR 50, Appendix A,
(GDC-4);
transmitted B&W Report, Document ID 51-1159048-00, prepared for FPC, subject Safety Balance Assessment for Elimination of Reactor Coolant System Main Loop Pipe Break Protective Devices.
6.
B&W Report, BAW 1847, Rev.
1, dated October 7,
1985, same subject as in Reference 1; revised report issued to address comments and questions raised by NRC staff reviewers.
7.
FPC letter to NRC, Westafer to Denton, dated October 29, 1985 (3F1085-13), subject Transmittal of Report Related to Request for Exemption from a Portion of 10 CFR 50, Appendix A,
General Design Criterion 4; transmitted report prepared by FPC, subject Assessment of CR-3 RC Leak Detection System, File:
SP 83-133, dated October 25, 1985.
8.
Meeting on October 31, 1985 among
9.
NRC summary issued by H.
Silver dated November 13, 1985 of the reference (8) meeting.
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FPC letter to NRC, Westafer to 'Denton, dated January 13, 1986 (3F0186-12), subject Additional Information Regarding Request for Partial Exemption from General Design Criterion 4 11.
FPC letter to NRC, Westafer to Denton, dated January 16, 1986 (3F0186-18),
subject Technical-Specification Change Request No. 142; proposes to remove the tabular list of snubbers. from the Technical Specifications in accordance with NRC guidance provided in Generic Letter 84-13.
12.
FPC letter to NRC, Simpson to Denton, dated January 21, 1986, subject Snubber Optimization Approval Request.
13.
Meeting on February 27, 1986 among NRC, FPC, B&W, and Gilbert Commonwealth, Lynchburg, Va.,
to discuss ten questions raised by NRC.
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'l a-ATTACHMENT 2-h Meeting Among NRC, FPC, B&W and Gilbert Commonwealth Lynchburg, Virginia ELEMENTS OF RESPONSES TO NRC QUESTIONS OF 2/21/86 V
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February 27, 1986 i
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NRC OUESTION 1 The reactor coolant system analysis is based on a one-half system structural model with appropr,iate boundary conditions.
B&W has stated that this approach has been verified by comparing satis-factorily the results from a half-system model to a model containing both loops.
However, a diagram of the full system indicates that the symmet~ry is only approximate.
The concertis is whether the half-system model of the reactor coolant cyctem and its boundary conditions provides correct results when compared to f
the full system analysis.
ELEMENTS OF RESPONSE TO OUES R C 0
B&W Repo':t 32-1103808 considered the CR-3 RCS pipiny arrangement.
O Conclusion - A hal.f loop model can be used to determine the dynamic responce of the full system.
O CR-3 RC purps have supports with:
Differing Orientations Spring Rates Length.s O
Differences have been med611ed.
b Both the North and South half loops have been analyzed.
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NRC OUESTION 2 The analysis of the RCS is based on a model which combines the reactor building interior concrete and the reactor coolant system.
The concern is whether the properties of the reactor building have been properly determined and are properly reflected in the combined model.
ELEMENTS OF RESPONSE TO OUESTIOMJ O
Roactor building interior concrete model properties used by B&W have been reviewed recently by Gilbert Commonwealth.
O A 10.6% veight difference was identified.
O Weight diffetence results in approximAtely a 7% change in seismic stress at the highest stressed location.
f O
Results in an approximate 1/2% increase in the total design stress at the highect stressed location.
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NRC OUESTION 3 The analysis of the RCS subjected to dead weight, and other distributed loading, is performed by imposing concentrated forces at the mass joints.
The equivalent moments (known as fixed-end moments) appear to be neglected in this approach.
No justifica-tion is provided for this approach.
ELEMENTS OF RESPONSE TO OUESTIONS 3 0
RCS structural model includes local deflection of the RV and OTSG nozzles.
O End moments are accounted for by modelling the nozzle flexibility.
O Lumped masses at discrete locations along the piping length produce conservative moments.
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NRC OUESTION_1 The damping value for the RCS components was taken as 2% of critical, per RG 1.61.
However, for the pumps, a total of 32 large bore snubbers were replaced by a total of four smaller snubbers and four struts, thus reducing significantly the physical sources of damping in the structure.
Justification is needed for not choosing a lower damping value for the optimized configuration than that specified in RG 1.61.
ELEMENTS OF RESPONSE TO OUESTION 4 O
Damping of the RCS piping system has been reduced by support removal.
O RC pump and motor assembly is a complex arrangement with:
Rotating Elements Various Attached Piping Systems Gaped Interfaces Bolted Connections Complicated Structural Members O
2 percent damping is considered appropriate for the pump, motor stand and motor assembly per RG 1.61.
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NRC QUESTION 5 An evaluation is needed on the structural stability of the optimized RC pump support as-built configurations when subjected to compressive loading.
Uncer certain types of such loading, the proposed configurations appear to be structurally unstable indicating that the supports may not be effective under actual loading, thus causing the allowable stresses to be exceeded.
ELEMENTS OF RESPONSE TO QUESTION 5 o
The RCS structural analysis has pump support elements modelled as pinned bars.
o No moments or perpendicular loadings are transferred by these elements.
o Support r6 embers are positioned so they complement existing support already provided by the piping, o
The RCS structural model is a 3 dimensional model with a 3 dimensional loading.
o Masses are input at centerline locations representative of the actual structure.
O Earthquake inputs are calculated for 3 directions.
o The piping primarily supports the pumps; constant support hangars provide additional support for dead weight.
o Allowable stresses are not exceeded.
NRC OUESTION 6 Clarification is needed on the method of determining the com-posite damping values and the seismic response spectra for structural models with different damping through the structure.
ELEMENTS OF RESPONSE TO OUESTION 6 O
Composite damping values are calculated based on a mass and mode shape weighted technique.
O Method referred to as strain energy weighted.
O Input seismic response spectra is from the FPC Environmental and Seismic Qualification Guide Specifications and Data SP-5095.
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NRC OUESTION 7 Clarification and justification are needed for the generation of different flexibility matrices for the seismic analyses of different earthquake components.
ELEMENTS OF RESPONSE TO OUESTION 1 0
Two flexibility matrices are generated for seismic because of half loop model.
O Boundary conditions are different in the half loop models.
O For X and Y direction earthquakes:
RV and wall centerlines are fixed for Z translation and rotation of about the X and Y axes.
O For Z direction earthquake:
RV and wall centerlines are fixed for X and Y translation and Z rotation.
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NRC OUESTION 8 Clarification is needed on the boundary conditions used for the thermal analysis of the half-system model.
ELEMENTS OF RESPONSE TO OUESTION 8 O
Boundary conditions maintain veetical RV centerline O
RV and wall centerlines are:
Fixed for Z direction translation Fixed for X direction rotation O
RV rotates and transmits loadings to base.
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NRC OUESTION 9 Clarification is needed that steady-state hydraulic loads include flow induced vibration due to pump operation (NUREG/CR-1319,
" Cold Leg Integrity Evaluation"), and that these loads are appropriately combined with dead weight and thermal loads.
ELEMENTS OF RESPONSE 9 O
Stress is 1.7% of operating primary stress allowable.
O Stress amplitude is well below the endurance limit.
4 O
Stresses are negligible in the ASME Code fatigue analysis.
Usage factor would be increased by less than 1%
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NRC OUESTION 10 Clarification is needed that local stresses in the RC pump casing resulting from the attachments of the supports to the casing have been included in the stress evaluation of the casing.
ELEMENTS OF RESPONSE TO OUESTION 10 0
RC pump supports are not integral attachments to the pump casing.
O The pump supports are attached to a restraint ring attached to the motor stand.
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0; ATTENDEES: 2/27/86 Name Organization E. H. Davidson FPC, L'icensing.
Larry Tittle FPC, Engineering Santo Ferrarello Gilbert / Commonwealth Engrng.
Paul Schmitzer Gilbert / Commonwealth Engrng.
Paul Bezler BNL/HRC-Harley Silver NRC Jim Canning B&W Bob Allen B&W Randy Schaefer B&W Mark Hartzman U.S. NRC l.
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