ML15223A708
| ML15223A708 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Oconee, Crystal River, Rancho Seco, Crane  |
| Issue date: | 01/09/1981 |
| From: | Reid R Office of Nuclear Reactor Regulation |
| To: | DUKE POWER CO., METROPOLITAN EDISON CO., SACRAMENTO MUNICIPAL UTILITY DISTRICT, TOLEDO EDISON CO. |
| References | |
| TAC-08602, TAC-08843, TAC-8602, TAC-8843, NUDOCS 8102030673 | |
| Download: ML15223A708 (21) | |
Text
UNITED STATES o
e.NUCLEAR REGULATORY COMMISSION
< E WASHINGTON. D. C. 20555 January 9, 1981 TO ALL BABCOCK & WILCOX (B&W)
LICENSEES Gentlemen:
Enclosed, herewith, is a request for additional information relating to the B&W 177-FA Owner's Group submittal on asymmetric LOCA loadings.
We have consolidated all our concerns related to all B&W operating reactors in the enclosed request.
We request a response to these requests related to your plant within 60 days of receipt of this letter.
However, to assure that you can be complete in your responses and that all issues can be resolved with one final response, we suggest that the Owner's Group convene another meeting with the staff after you have had the opportunity to review the request and before submittal of the responses. We will be available for such a meeting.
Sincerely, C6ert W. Reid, Chief Operating Reactors Branch #4 Division of Licensing
Enclosure:
Request for Additional Information cc w/enclosure:
Service Lists OCo 6
Enclosure REQUEST FOR ADDITIONAL TNFORMATION CONCERNING B&W 177-FA OWNER'S GROUP SUBMITTAL CONCERNING ASYMMETRIC LOCA LOADINGS DOCKETS NOS. 50-313, 302, 346, 269, 270, 287, 312'& 289
- 1.
QUESTIONS ON CAVITY PRESSURE
- 1. Section 4.3.2 How were friction factors obtained?
- 2. Section 4.3.2 What L/Dh was used for turns not equal to 90 degrees?
- 3. Section 4.3.2 Supply sample calculation of a form loss coefficent.
- 4.
Section 4.3.2 What, if any, equipment is in the path of motion of the venting devices?
- 5.
Section 4.3.2 What uncertainty factors were applied to calculated flow areas? Demonstrate that uncertainty factors were applied in a conservative manner.
- 6. Section 4.3.2 Supply the values of any uncertainty factors that may have been applied to other parameters such as volumes, flow lengths, etc.
- 7.
Section 4.3.2 Supply the initial conditions (pressure and temperature) used in the mass and energy release rate calculations. Demonstrate the generic applicability of these values.
- 8. Section 4.3.2 A drag coefficient of zero does seem unrealistic; however, justification should be provided for the empirical mass multiplier of 2.0.
- 9. Section 4.3.2 Supply drawings showing the reactor cavity, equipment and piping arrangements for the following plants:
- a.
Crystal River - 3
- b.
Three Mile Island I and 2.
1
- 10.
Section 4.3.2 Provide an assessment of the influence of vena contracta effects on the reported results.
- 11.
Section 11.1.3 What is the source document for the ACP and jet inpingement loading data?
- 12.
Section 11.1.3 Assess asyrrnetric cavity pressure loadings for which ACP data were not available, including breaks at the steam generator outlet nozzle.
2
- 2.
QUESTIONS ON THERMAL HYDRAULICS
- 1. In Figures 8.2-1, 4, 7, and 10, identify the interface across which the RV head differential pressures are determined (refer back to Appendix C Figures C-2, C-9).
- 2.
Identify and discuss out-of-RV cavity-pipe break locations considered or evaluated for RV internals hydraulic transients.
Apparently in Section 8.2, RV internals loads. were calculated for the SG cavity break--what was the reason for not identifying this condition in Section 4.4?
Confirm that out-of-RV cavity BOA's in Section 11 are not limiting relative to RV internals hydraulic transients.
- 3.
Were high pressure injection ECCS piping and core flood line check valves considered or analyzed for hydraulic transients?
Discuss load severity relative to RCS break size and location.
- 4.
Discuss the core flood line load severity relative to RCS break location and size.
- 5. In Table 8.2.2, discuss why the RV head differential pressure is higher for a 1.5 A cold-leg break rather than for a 2.0 A cold-leg break.
Would a corresponding relationship (as shown) exist for a 1.5 A SG cavity cold-leg break?
- 6. It is not clear which BOA's are used for each specific analysis for the skirt and nozzle supported plants. In a table, summarize the break location, size, time, and type for each RV internals, ECCS, RCS and cavity analysis (refer to Figures 4.2-3 through 4.2-17, as applicablie).
See questions 7, 8, 9, and 10.
- 7. In Section 4.4, are the break locations analyzed depicted in Figure 4.1-1? Clarify the break locations specified in Section-4.4 (first paragraph, last sentence).
3
- 8.
In Section 8.2, what are the assumed break locations, area, and times.
- 9. In Section 4.4, clarify the last sentence in the second paragraph.
Identify or provide those BOA curves for each break location considered.
Are Figures 4.2-3 through 4.2-17 generic or actual BOA curves?
The stated sentence appears to conflict with use of actual BOA curves mentioned in the first paragraph for the nozzle supported plant (please clarify).
- 10. -In Appendix C, Paragraph 2, how were the 'generic fluid conditionsO established?
- 11.
In Appendix C, for the skirt-supported plant, there are more downcomer nodes than in the Oaudit model* in BA4-10132 P-A.
Discuss this and any other differences and how it will impact RV internals pressures and loads.
- 12. In BAW-10132 P-A, Supplement 1, Figure D-1 (design Model, 16 node downcomer) shows a peak differential pressure 290 psi less than Figure A-1 (audit model, 40 node downcomer).
Based on this indication of lower differential pressures when modeliTng about 50% fewer downconer nodes, justify use of the 20 node CRAFT2 downcomer model for the nozzle-supported plant to evaluate RV internals pressures and loads.
- 13. For the skirt supported plant for a cold-leg RV nozzle break, show a sample calculation for integrating the pressures around the core barrel periphery to get a directional force per unit length. Consider volume nodes 102 through 113 (at one elevation only) in Figures C-2, 3.
- 14.
What effect could a sticking vent valve haveg i.e., late opening/closing (or) failure to open/close, relative to RV internals differential pressures for a hot and cold-leg break?
4 k
- 15.
In estaolishing the generic skirt-supported plant model, what sensitivity studies were conducted among the various skirt-supported plants to define worst case internals differential pressures relative to any design or as-built differences in internals geometry, vent areas, flow loss coefficients, and CRAFT2 model volunme arrangement and flow inertia terms?
- 16.
How were local fluid acceleration (cross-flow drag) forces accounted for in RPV internals loads evaluation (such as lateral forces on the control rod guide tubes and support columns during a hot-leg break)?
- 17.
In Appendix C, qualify and justify the statement in paragraph 2:
8A.
generic set of fluid conditions was establ ished to serve as a conservativie representation of the fluid conditions present in the 177-FA lowered-loop plants."
Were actual or SAR design quoted fluid parameters used?
- 18.
For all skirt-supported plants, provide a table ccmaring those cognizant geometric, thermal, and hydraulic parameters used to justify generic grouping.
What were the main conditions used to establish the similarity between actual plants and the case analyzed for RPV internals and RCS/ECCS hydraulic transients during subcooled blowdown.
Include Davis-Besse 1 parameters for comparison with a nozzle supported plant.
- 19.
In Figures C-9 and C-11 identify downcomer volume nodes adjacent to both hot and cold-leg breaks and in Figures C-2 and C-7 for a hot-leg break.
- 20. Figure C-11 is not interpretable.
Provide a description, with legible volume node numbers, like Figures C-6 and C-7..
- 21. Provide a table of peak differential pressures and loads for each major RV internals component analyzed for both the skirt and nozzle supported plants for each BOA and location evaluated.
Indicate time of occurrence.
5
- 22.
In Appendix C, what are the differences between skirt and nozzle supported plants that require separate CRAFT2 models?
- 23. In Appendix C, paragraph 3, reference is made to simulation of four vent valves.
However, in the-Davis-Besse 1 safety analysis report (SAR) no mention can be found of any vent valves for that plant.
Please clarify.
- 24. Identify 177-FA raised and lowered-loop plants.
What is the relationship between these and the skirt and nozzle-supported plants?
Provide Section 4.1 of BAW-1010 4.
- 25.
In Appendix C, what is the significance in nodal izing volumes that appear to cross the physical boundaries of the lower incore support plate, the flow distributor plate, and the -lower grid in Figure C-9 for the noz.zle-supported plant as compared to Figure C-2 for the skirt supported plant?
To what extent does this influence the direct simulation and proper modeling of the flow inertia terms and the location of minimum area junctions for RV internals differential pressure calculations?
- 26. For both skirt and nozzle-supported plants for both hot and cold-leg RV nozzle breaks, and the 2 A cold-leg SG compartment break (if available), provide (1) comuter listing of the CRAFT2 input for RV internals analysis, and (2) the following pressure transients:
- a.
The differential pressure across the RV upper and lower heads for volume nodes 141-157 in Figure C-2 and nodes 5-10 in Figure C-9.
- b. The differential pressure in the downcomer annulus across volume nodes 97-91, 109-103, and 133-127 for Figure C-2. Provide the same data at the same approximate locations for the nozzle supported plant.
6
C.
Pressures in the RV for Yolzne nodes 153, 141, 97, 91, 127 in, Figure C-2. Provide the same data at the same approximate locations for the nozzle-supported plant.
- d.
Differential pressures across the vent valves across volume nodes 169-86 and 162-80 in Figure C-7.
Provide the same data at the same approximate locations for the nozzle-supported plant.
7
- 3. QUESTIONS ON STPUCTURAL METHODS, MODELS, AND EVALUATIONS 1
Section 4.4 In reference to Table 4.2-2, how are the break opening times determined for the 2A cold-leg break?
- 2. Section 5.1 The comparison indicated in paragraph 4 between Figures D-6 and D-7 is unclear. Explain.
- 3.
Section 5.1.4 Provide the basis for the selection of the composite elements.
- 4. Section 5.1.4 What method was used to calculate the stiffness values?
- 5. Section 5.2 Was pipe whip also considered for breaks outside the RPV cavity?
- 6. Section 5.2.1 Expand the description of the initial conditions used in this analysis. Were the steady-state momentum forces considered in determining these initial conditions?
- 7. Section 5.2.2.7 Discuss in detail the full loop model used to determine the operating condition loads.
B. Section 5.2.2.9 Provide more detail or the iteration procedure (include break area and break time) described in Figure 5.2-1.
- 9.
Section 5.3.6.3 Clarify this discussion.
- 10.
Section 6.0 Was a scram time evaluation performed?
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- 11.
Section 6.2.1.4 Supply the supporting data for the compressive strength value of entedment concrete used for the Rancho Seco Plant.
- 12.
Section 6.2.1.9 Supply a sample calculation demonstrating the procedure outlined for combining seismic and LOCA responses for the erbedments.
- 13. Section 6.5 Was a core barrel shell analysis performed?
- 14.
Section 6.9.3 Essentially four methods of demonstrating acceptability of the Core Flood Line Piping are discussed in this.section:
- 1. response spectrum,
- 2. response soectrum removina failed support,
- 3. time history,
- 4.
time history removing failed supports.
Methods 1 and 3 are acceotable with no further jusitfications; however, methods 2 and 4 require additional quantitative justification.
- 15. Section 6.10.4.5 How was the 7-critical danping incorporated in the dynamic analysis? Specify which modes were critically damped?
- 16. Section 6.10.4.6 Demonstrate with a detailed calculation that the dynamic load factor is 1.15.
- 17.
Section 7.1.1.1 Demonstrate that the strain rates are sufficiently high to justify an increase in yield stress of 10%.
- 18. Section 7.1.1.4 Explain the shear and tension interaction equation development.
- 19. Section 7.2.1 Demonstrate that strain rates are sufficiently high to justify an increase in concrete strength of 25a in compression and 0 in shear.
- 20. Section 7.3.2.1 Justify the use of level D stress limits for tees and branch connections.
- 21.
Section 8.1 Specify the BOA time considered in determining the loadings shown in Table 8.1-1.
- 22.
Section 8.1 The vertical load shown in this and subsequent ficures does not appear to have reached its maximum value. Please discuss in detail.
- 23.
Section 8.1 Identify the break location from Figure 4.1-1 associated with each of the loadings tabulated in Table 8.1-1.
- 24.
Section 10.5 Supply quantitative results and allowable values.
- 25.
Section 10.6 Provide the calculated load value which yields the spacer arid deflection indicated in Table 10.6-1.
- 26.
Section 10.6 Are all owner's group plants supplied with inconel arids? If not, explain?
- 27.
Section.10.6 Were nonlinear inpact elements incorporated into the appropriate computer code to yield the spacer arid deflection? If so were these elements nonlinear-elastic or elastic-plastic? Provide a discussion.
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- 28.
Section 10.7 Are these results applicable for both broken and unbroken loop piping?
- 29.
Section 10.7 Identify which breaks yield which stresses.
- 30. Section 10.7 Are the results in Tables 10.7-1 and 2 consistent with the equations presented in Section 6.77 If not, why not?
- 31.
Section 11.1.2 Explain in detail the calculations performed to reflect the assumed 2A BOA.
- 32. Section 11.1.2 What was the basis for assuning that all the skirt supported plants were hydraulically sitilar to Midland 1 and-2?
The same question applies for nozzle supported plants.
Are specific plant analyses planned in the future?
- 33.
Section 11.1.3 Justify multiplying the 1A break results by 2.0 to obtain ZA results.
Why was a factor of 1.5 chosen instead of some other value?
- 34.
Section 11.1.3 Clarify Figures 11.1-9 thru 11.1-14 with regard to their generation and application.
- 35. Section 11.1.4.1 Justify that removing the gapped. restraints from the model is conservative.
- 36. Section 1.1.4.1 Hydraulic, compartment pressure and jet impingment are mentioned as loading conditions for the RCS structrual analysis.
Were dead weight and thermal loadings also considered in the analysis?
- 37. Section 11.1.4.1 Expand the discussion concerned with factoring the forces at the break plane.
11
.38.
Section 11.1.4.3 Was the broken pipe loop considered for determining steam generator support loadings?
- 39.
Section 11.0 Identif: in detail what constitutes the applied forces and what constitutes the response loads.
Discuss in detail for a specific plant the mathematical model, which includes applied force locations and response locations.
- 40. Section 11.2.1 Supply results data and comparisons.
and 11.2.2 APPENDIX C
- 41.
Section 2 Supply justification that all of the 177-FA skirt supported plants are hydrodynamically equivalent.
Also include a list of differences, if any.
APPENDIX D
- 42. Section 3.3 Is the hydrodynamic mass matrix changed as the annulus changes size during the dynamic analysis?
What experimental/analytical correlations exist for using this technique for this type of analysis?
Have bench-mark cases been completed?
Enclose the results of these analyses.
- 43. Section 5.2.2 Describe in detail how the core-bounce results are used in the system dynamic analysis.
APPENDIX E
- 44.
General Supply the following:
- a.
Boundary conditions used in the analyses, 12
- b. Expanded mathematical model description, and
- c.
Detailed discussion of the applied loads.
APPENDIX I
- 45. General Specify and justify the value of f'c used in the analysis.
- 46. General Discuss in detail all boundary conditions.
- 47. General Define the applied loadings, including application points.
- 48.
General Discuss in detail the procedure used for handling antisymetrical loadings.
- 49.
Section 2 Is using isotropic properties. conservative for loads other than in the hoop-direction?
- 50. Section 3.1 In light of the fuel canal slab strength reduction, discuss bending and shear at discontinuities and boundaries.
APPENDIX J
- 51. Section 1 What effect does the water sass have on impact values?
- 52.
Section 1 Are grid-to-grid-to-baffle and grid-to-baffle inpact stiffnesses verified?
Supply the details.
13
APPENDIX K
- 53.
General Supply the following information:
- a.
Detailed math models
- 1.
Nod alizati on
- 2. Ppplied force locations
- b. Model boundary conditions.
- 54. Section 2.5 Describe the differences between various plants.
- 55. Section 3.1.2 Supply the gap dimensions.
- 56. Section 4.1.1 Provide the basis for selecting the stiffness values for the embedments from the parametric studies.
GENERAL QUESTIONS
- 57.
Seismic and LOCA evaluations do not exist for some of the corponents and piping analyzed. Provide these evaluations.
- 58.
An evaluation of the pump and steam generator supports is not provided for breaks occuring in the reactor cavity.
Supply these evaluations.
- 59.
Some of the major system corponents evaluations were excluded from the evaluations for steam generator comvartment pipe breaks.
Provide these evaluations (include reactor internals).
14
B&W 177-FA OWNERS GROUP PLANT SPECIFIC REQUESTS Arkansas Nuclear One Unit I Requests
- 1. Supply reactor vessel cavity wall drawings and include a detailed discussion of the critical loads, load paths, and functional integrity of the cavity wall. Particular attention should be focused on the location of the overstressed condition.
- 2. Corment briefly on the response of restraints 2, 3, and 4 shown in Figure B-4.
Crystal River Unit 3 Requests
- 1.
Supply reactor vessel support and embedment drawings showing the general configuration.
Provide a discussion of the loading and response of the cavity wall including the linear plate interface.
Demonstrate that stability %ill be maintained and reactor shutdown and coolability will not be impaired.
- 2.
Discuss in detail the modifications proposed for limiting the hot-leg.break area and explain what changes would be anticipated in component response.
Supply drawings showing locations of the proposed modifications.
- 3.
Supply bracket plate and hot-leg restraint drawings with a discussion of the applied loads responses, and consequences of failure.
Discuss in detail the calculation of the allowable loads.
- 4.
Coment briefly on restraints 2 and 3 in Figure 8-5.
- 5. Provide a detailed discussion of the load redistribution predicted to take place in the cavity wall. Supply drawings of the effected area and discuss overall functional integrity of the wall.
Davis-Besse Unit 1 Requests
- 1. Detail the changes made to the skirt-supported models to facilitate the Davis-Sesse 1 analysis. Provide a detailed discussion of the response loads and the mathematical models, especially in the reactor vessel support regions.
- 2.
Supply drawings of the reactor pressure vessel support system and the general plant layout.
uconne Units 1, 2, end 3 Requests
- 1.
Supply reactor vessel support and emfbedment drawings showing the general configuration along with a detailed discussion of the loading and structual response with particular attention to stability and component functions.
- 2.
Supply drawings of the control rod drive mechanism and housing and discuss the applied loads, responses, and modified mechanisms.
Rancho Sec o Requests
- 1.
Supply drawings of the reactor vessel erbeaments detailing the location where the acceptance criteria is exceeded. Discuss the various responses and the effect on overall stability.
- 2.
Comment briefly on restaints 1 and 4 in Figure B-6.
Three Mile Island Unit 1 Requests
- 1.
Discuss in detail the differences between the TMI-1 and CR-3 power plants in the following areas:
- a.
reactor vessel supports,
- b.
reactor vessel support embedments,
- c. control rod drive mechanism housings, and
- d.
proposed modifications for limiting the hot-leg break area.
- 2. Supply hot-leg pipe whip restraint drawings and discuss analytical results with respect to these drawings.
Detail the 1
a consequences of failure of this support.
3.*
Comment briefly on restaint 2 in Figure B-2.
- 4. The following information is requested on the reactor vessel cavity wall:
- a. drawings describing the wall and the surrounding structures;
- b. allowable values based o6 the wall primary functions;
- c. a discussion of the load distribution, load path, and magnitude, along with a comparison of the results to a functional allowable;
- d.
a definitive margin of safety for the wall; and
- e.
definitive ratio of ductility as used in this context.
9 *RR