ML17326A787
| ML17326A787 | |
| Person / Time | |
|---|---|
| Site: | Point Beach, Surry, Turkey Point, Robinson, Cook, La Crosse, Zion  |
| Issue date: | 10/29/1980 |
| From: | Lainas G Office of Nuclear Reactor Regulation |
| To: | Anderson T WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| References | |
| REF-GTECI-A-02, REF-GTECI-RV, TASK-A-02, TASK-A-2, TASK-OR NUDOCS 8011210443 | |
| Download: ML17326A787 (36) | |
Text
Docket Nos~0~3IP'50-31 6, 50-261, 50-295, 50-304, 50-250, 50-251, 50-409, 50-280. 50-281, 50<<266 and 50-301 OCTOBER 2 9 1980 Hr. Thomas H. Anderson Hanager Nuclear Safety Westinghouse Electric Corporation Box 355 Pittsburgh, Pennsyl v'ani a 15230
Dear Hr. Anderson:
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We have completed our review of your submittals "Asymmetric LOCA Loading" (TAP A-2) WCAP 9628 and 9748 dated November 1979 and June 1980, respectively.
Enclosed is our request for additional information.
Please propose a schedule for addressing each of the items we have identified.
As discussed during a telephone conversation on October 3, 1980, you will propose a meeting, within a few weeks of receiving our request, if you require further clarification.
James J.
Shea (301/492-7231) will coordinate the NRC staff effort.
Sincerely, pngmal s<gnel bg
Enclosure:
Request for Additional Information cc w/enclosure:
See next page Gus C. Lainas, Assistant Director for Safety Assessment Division of Licensing
- SEE ATTACHED YELLOW FOR PREVIOUS CONCURRENCES l
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~NRC FORM 318 (9-76) NRCM 0240 4V.S. GOVERNMFNT PRINTING OFFICE: l979.289.369
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Docket Hos. 50-315, 50-316, 50-261, 50-295, 50-304, 50-250, 50-251, 50-409, 50-280, 50-281, 50-266 and 50-301 t1r.
Thomas H. Anderson Chairman llestinghouse Owner's Group Manager Nuclear Safety llestinghouse Electri c Corporati on Box 355 Pittsburgh, Pennsyl vani a 15230 Qear Hr. Anderson:
Me have completed our review of your submittals "Asymmetric LOCA Loading" (TAP A-2) HCAP, 9628 and 9748 dated November 1979 and June 1980, respectively.
Enclosed is our request for additional information.
Please propose a schedule for addressing each of the items we have identified.
As discussed during a telephone conversation on October 3, 1980, you will propose a meeting, within a few weeks of receiving our request, if you require further clarification.
James J.
Shea (301/492-7231) will coordinate the NRC staff effort.
Sincerely.
Enclosure:
Request for Additional Informat ion Gus C. Lainas, Assistant.Director for Safety Assessment Division of Licensing cc w/enclosure:
See next page OFFICE P SURNAME OATEP
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DISTRIBUTION:
Dockets 50-315, -316; -261, NRC PDR 50-315, -316, -261, LPDR 50-315, -316, -261, TERA 50-315, -316, -261, NSIC 50-315, -316, -261, NRR Reading ORB ¹5 Reading ORB ¹1 Reading ORB ¹3 Reading DEisenhut RPurple TNovak RTedesco GLainas JRoe OELD OIaE (3)
DCrutchfield JShea SNowicki (Ginna)
HSmi th SVarga KParrish SNiner (D.C.Cook 1/2)
DNei ghbors (H.B.Robinson 2,
DWigginton Zion 1/2) t<Grotenhuis (Turkey Point 3
RAClark PKreutzer CTrammell (Point Beach 1/2)
RBosnak ACRS (16)
JHeltemes, AEOD GCwalina KKniel EChelliah, 440 Gray Files (12)
Xtra Cys (6)
-304,
-304,
-304,
-304,
-304,
-295,
-295,
-295,
-295,
-295, Surry 1/2)
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-250,
-250,
-250,
-250,
-250,
-251, -409, -280, -281,
-251, -409, -280, -281,
-251, -409, -280, -281,
-251, -409, -280, -281,
-251, -409, -280, -281,
-266, -301
-266,
-301
-266,
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-266,
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-266,
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 October 29, 1980 Dock'et Nos. 50-315, 50-316, 50-261, 50-295, 50-304, 50-250, 50-251, 50-409, 50-280, 50-281, 50-266 and 50-301 Mr. Thomas M. Anderson Manager Nuclear Safety Westinghouse Electric Corporation Box 355 Pittsburgh, Pennsyl vani a 15230
Dear Mr. Anderson:
'e have cooyleted our review of your submittals "Asymmetric LOCA Loading" (TAP A-2) WCAP 9628 and 9748 dated November 1979 and June 1980, respectively.
Enclosed is our request for additional information.
Please propose a schedule for addressing each of the items we have identified.
As discussed during a telephone conversation on October 3, 1980, you will propose a meeting, within a few weeks of receiving our request, if you require further clarification.
James J.
Shea (301/492-7231) will coordinate the NRC staff effort.
Sincerely,
Enclosure:
Request for Additional Informati on us
. Lain, Assistan for Safety Assessmen Division of Licensing irect r cc w/enclosure:
See next page
cc w/enclosure:
Harry H. Voigt, Esquire
- LeBoeuf, Lamb, Leiby and MacRae 1333 New Hampshire Avenue, N. W.-
Suite 1100 Washington, D.
C.
20036 Mr. Michael Slade 12 Trailwood Circle Rochester, New York 14618 Rochester Committee for Sci ent ifi c Informat i on Robert E. Lee, Ph.D.
P. 0.
Box 5236 River Campus Station Rochester, New York 14627 J effrey Cohen New York State Energy Office Swan Street Building Core 1, Second Floor Empire State Plaza
- Albany, New York 12223 Director,,Techni ca 1 Devel opment Programs State of New York Energy Office Agency Building 2 Empire State Plaza
- Albany, New York 12223 Rochester Public Library 115 South Avenue Rochester, New York 14604 Supervisor of the Town of Ontario 107 Ridge Road West
- Ontario, New York 14519 Resident Inspector R. E. Ginna Plant c/o U. S.
NRC 1503 Lake Road
- Ontario, New York 14519 Mr. T. Campbell, PC-2 Westinghouse Electric Corporation Box 355 Pittsburgh, Pennsylvania 15230 Director, Technical Assessment Di v'isi on Office of Radi at i on Programs (AW-459)
U. S. Environmental Protection Agency Crystal Mall k2 Arlington, Virginia 20460 U. S. Environmental Protection Agency Region II Office ATTN:
E IS COORDINATOR 26 Federal Plaza New York, New York 10007 Herbert Grossman, Esq.,
Chairman Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D.
C.
20555 Dr. Richard F. Cole Atomic Safety and Licensing Board U. S. Nuclear Regulatory Commission Washington, D. C.
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Emmeth A. Luebke Atomic Safety and Licensing Board U. S. Nuclear Regulatory Coranission Washington, D. C.
20555 Mr. Thomas B. Cochran Natural Resources Defense Council, Inc.
1725 I Street, N.
W.
Suite 600 Washington, D. C.
20006 Mr. Leon D. White, Jr.
Vice President Electric and Steam Production Rochester Gas 8 Electric Corporation 89 East Avenue Rochester, New York 14649
cc w/enclosure:
Mr. Bruce Churchill, Esquire Shaw, Pittman, Potts and Trowbridge 1800 M Street, N.W.
Washington, D. C.
20036 Mr. Mi.ll)am Guldemond USNRC Resident Inspectors Office 6612 Nuclear Road Two'Rivers, Wisconsin 54241 Joseph Mann Library 151,6 Sixteenth Street Two Rivers, Wisconsin 54241 Mr. Glenn A. Reed, Manager Nuclear Operations Misconsin Electric Power Company Point Beach Nuclear Plant 6610 Nuclear Road Two Rivers, Misconsin 54241 Mr. Gordon Blaha.
Town Chairman Town of Two Creeks Route 3
Two Rivers, Misconsin 54241 Ms-Kathleen M. Falk General Counsel Wisconsin's Environmental Decade 302 E. Washington Avenue Madison, Wisconsin 53703 ll Director, Technical Assessment Division Office of Radiation Programs (AW-459)
U. S. Environmental Protection Ag'ency Crystal Mall f2 Arlington, Virginia 20460 U. S. Environmental Protection Agency Federal Ac'tivities Branch Region V Office AT1'N:
EI S COORDINATOR 230 S. Dearborn Street Chicago, Illinois 60604 Chairman Public Service Comission of Wisconsin Hill Farms State Office Building Madison, Misconsin 53702 Mr. Sol Burstein Executive Vice President Wisconsin Electric Power Company 231 West Michigan Street Milwaukee, Wisconsin 53201 G. F. Trowbridge, Esquire Shaw, Pittman, Potty, and Trowbridge 1800 M Street, N.W.
Washington, D. C.
20036 Hartsvi 11 e Memorial Library Home and Fifth Avenues Hartsvi 1 1 e, South Carol ina 29550 Michael C. Farrar, Chairman Atomic Safety and Licensing Appeal Board Panel U. S. Nuclear Regulatory Comission Washington, D.
C.,20555 Richard S.
Salzman Atomic Safety and Licensing Appeal Board Panel U. S. Nuclear Regulatory Coomission Washington, D. C.
20555 Dr.
W. Reed Johnson Atomic Safety and Licensing Appeal Board Panel U. S. Nuclear Regulatory Comnission Washington, D. C.
20555 Mr. J. A. Jones Senior Executive Vice pres.
Carolina Power and Light Co.
336 Fayetteville Street
- Raleigh, Nor th Carolina 27602
cc w/enclosure:
Mr. Robert Lowenstein, Esquire Lowenstein,
- Newman, Reis and Axelrad 1025 Connecticut Avenue, N.W.
Suite 1214 Washington, D.
C.
20036 Environmental and Urban Affairs Library Florida International University Miami, Florida 33199 Mr. Norman A. Coll, Esquire
- Steel, Hector and Davis 1400 Southeast First National Bank Building Miami, Florida 33131 Mr. Henry Yaeger, Plant Manager Turkey Point Plant Florida Power and Light Company P. 0.
Box 013100 Miami, Florida 33101 Mr. Jack Shreve Office of the Public Counsel Room 4, Holland Building Tallahassee, Florida 32304 Admi nistrator Department of Environmental Regulation Power Plant Siting Section State of Florida 2600 Blair Stone Road Tallahassee, Florida 32301 Resident Inspector Turkey Point Nuclear Generating Station U. S. Nuclear Regulatory Coranission Post Office Box 971277'uail Heights Station Miami, Florida 33197 Dr. Robert E. Uhrig, Vice Pres.
Advanced Systems and Technology Florida Power and Light Company Post Office Box 529100 Miami, Florida 33152 cc w/enclosure:
Robert J. Vol'len, Esquire 109 North Dearborn Street Chicago, Illinois 60602 Dr. Cecil Lue-Hing Director of Research and Development Metropolitan Sanitary District
'f Greater Chicago 100 East Erie Street Chicago, Illinois 6061 1
Zion-Benton Public Library District 2600 Emmaus Avenue Zion, Illinois 60099 Hr. Phillip P. Steptoe I sham, Lincoln and Bea 1 e Counselors at Law One First National Plaza 42nd Floor
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Box -12194 Research Triangle Park; N.
C.
27709 1
Dr. Forrest J.
Remick 305 East Hamilton. Av nue State College, Pennsylvania 16801 Rick Yonter 617 Piper Lane Lake Villa, Illinois 60046 U. S. Nuclear Regulatory Commiss'ion Re.-,ident Inspectors Office Po.",t Office Box 288 Deerf ield, I 1 linois 60015 Hr; J.
S, Abel Director of Nuclear Licensing Commonwealth Edison Company Post Office Box 767.
Chicago, Illinois 60690 I
h
cc w/enclosure:
Mr. Robert W. Jurgensen Chief Nuclear Engineer American Electric Power Service Corporation 2 Broadway Hew York-, New York 10004 Gerald Charnoff, Esquire Shaw, Pittman, Potts and Trowbridge 1800 M Street, N.W.
Washington, D. C.
20036
'l Citizens for a Better Environment 59 East Van Buren Street Chicago, Illinois 60605 Maude Preston Palenske Mei~orial Library 500 Market Street St. Joseph, Michigan 49085 Mr. D. Shaller, Plant Manager Donald C.
Cook Nuclear Plant P. 0.
Box 458 Bridgman, Michigan 491'06 U. S. Nuclear Regulatory. Commission Resident Inspectors Office 770 Red Arrow Highway Stevensvi lie, Michigan,"49127 William J. Scanlon, Esquire 2034 Pauline Boulevard Ann Arbor, Michigan 48103 Mr. John Dolan, Vice President Indiana and Michigan Electric Co.
Post Office Box 18 Bowling Green Station, New York, Hew York 10004
cc w/enclosure:
Mr. Michael W. Maupin Hunton and Williams Post Office Box 1535 Richmond, Virginia 23213 Mr. J. L. Wilson, Manager P. 0.
Box 315 Surry, Virginia 23883 Swem Library College of William and Nary Will iamsburg, Virginia 23185 Donald J.
Burke, Resident Inspector Surry Power Station U. S. Nuclear Regulatory Comission Post Office Box 959 Wi lliamsburg, Virginia 23185 Nr. J.
H. Ferguson Exec.
Vice President
- Power Virginia EleCtr'ic and Power Co.
Post Office Box 26666 Richmond, Virginia 23261
qUESTIONS ON WESTINGHOUSE OWNERS GROUP ASYMMETR I C LOCA LOADS EVALUATION
I.
QUESTIONS ON CAVITY PRESSURE 1.
On page 8-5 of the report it is stated that obtaining the maximum flow rate from the hypothetical break is conservative with respect to the cavity pressure.
It is noted that decreasing the temperature has the effect of decreasing the specific enthalpy of the effluent.
'Justify in terms of mass and energy re'lease rates that selecting tne lowest temperature vessel/core inlet temperature for calculating the mass and energy release rates is conservative with respect to cavity pressurization.
2.
On page 8-10 of the report it is stated that sensitivity calculations were made to encompass plant to plant variations, and resulted in a
2X variation in peak cavity pressure.
Provide specific numeric variation in the four parameters listed on page 8-10.
3.
Justify friction factor input to TND listed in Tables 5
and 6 of Appendix 8 as zero for all;but three valves.
4.
How were the form loss coefficients in Tables 5 and 6 of the Appendix 8
'I calculated?
Provide an example.
5 ~
Pl ovide junction numbers on Fi gures 7 and 8 of Appenoix 8 so that junctions can be related to the input listing.
Also, show assumed direction of flow.
6.
How can it be determined from the input listings to TMD (Tables 5
and 6
of Appendix 8) that revers'e flow form loss coefficients are accounted for?
Alternately, show that reverse flow form loss coefficients have been accounted for.
7.
Were any sensitivity calculations made for flow inertias?
8; What specific uncertainty factors were applied to flow areas, volumes and form loss of coefficients?
9.
Supply drawings for the following plants showing equipment locations within the subcompartments and sufficient wall details to verify the generic applicability of the models used for the steam generator and pump s ubcompart ment:
1.
POINT BEACH 2 2.
SURREY 1
3.
ZION 2
'10.
In reference to page 8 of the. Phase C report, please answer the following questions on the inspection port plugs:
I a.
What value of friction between the plug and port wall was used in the calculations'?
(Equation on page 2-2) b.
What, is the mass of each Boro Silicone cube?
c.
What are the missile effects of the Boro Silicone cubes when blown out?
d.
Which plants require sand plug modifications?
11.
Page 2-1 last paragraph:
Provide the mass energy release rates used in the reactor cavity pressurization analysis.
12.
Page 2-3 first paragraph:,
Pi'ovide a sample calculation of the form loss coefficients.
13.
Page ii:
Designation for Tu"key Point 3 and 4 throughout remainder of report appears to be FPL not FLA as on page ii.
Are FLA and FLP des i gn at ions i nterch ang cable?
14.
For Figures 4, 9, 14, 19, 24, 19, 34 and 39 of. Appendix C, please provide assumed direction of flow and junction numbers.
l5.
Provide, flow path data for each reactor cavity similar to Table 5 and 6
of Appendix B.
I I.
QUESTIONS ON THERMAL HYDRAULICS l.
In modeling the downcomer annulus for MULTIFLEX, what dimensional distance was input for rarefa~tion wave traverse across the lower plenum?
Specify the values used. for generic two, three and four-loop p'lant groupings and any differences between the Phase B
RCL and Phase C
'P V i nter na 1 s mode 1 s.
3.
Are Y-directional hydraulic loads (90 out-of-plane to the x-inlet axis) considered in the. beam representation (noding) of the core barrel?
I Discuss importance on a plant-by-plant basis and between the original beam and the advanced beam MULTIFLEX models.
In NCAP-0748 on page 2-8, the last paragraph, does "Figure 1" refer to Figure 2.2-1?
4.
In WCAP-9628, Section 3.2-2, identify those five plants selected for the RCL hydraulic analyses.
Are WCAP-9628 (Section 3.0)
RCL MULTIFLEX models different than those used for WCAP-9748 (Section 2.2)
RPV internals hydraulic analyses?
Clarify and discuss the Phase 8 and Phase C plant, specific and generic MULTIFLEX models and identify the parameters in Table 3.1-2 of WCAP-9628 that were used in each model and calculation (see question 8).
5.
Nas the generic plant grouping consistent for both Phase B ana C analysis I
of hydraulic transients'?
How does Table 3.1-4 of WCAP-9628 relate to the five MULTIFLEX models of Sect ion 3.2-2 of NCAP-9628 and the eight hydraulic calculations of Se~tion 2.2 of NCAP-9748?
- 6. 'escribe the procedures employed in applying generically modeled plant MULTIFLEX results to plant, specific hydraulic analyses of RPV internals and the RCL.
7.
Are RPV internals pressure losses or flow attenuation effects considered during blowdown when using MULTIFLEX? Specifically, how are the flow resistance coefficients determined and utilized for the downcomer/thermal 4
shield split, the diffuser plate, core support plates, thimble guides, fuel bundles, etc.?
8.
What are the differences (NULTIFLEX inputs) involved in the eight separate RPV internals hydraulic calculations in WCAP-9748, Section 2.2'?
How many different NULTIFLEX models are employed and what pre they'?"
9 In WCAP-9748, Section 4.0, was check valve hard closure considered in the accumulator ECCS and RHR lines in the component evaluation?
Di scuss the implication of such a check valve failure relative to break size and location.
10.
In WCAP-9628, Table 3.1-2, how is the penetration coefficient "Beta inlet/DC joint" calculated or determinea'?
Give example.
What is the significance of a 'high value'r a 'low value', comparatively?
Is this coefficient sensitive to NLTIFLEX rigid or flexible wall model assumption.
ll.
Provide justification and purpose for use of the NULTIFLEX advanced beam model for the Phase C
RPV internal hydraulic transients.
12.
In WCAP-9748, Section 2.2, justify not analyzing' RPV hot-leg nozzle break for RPV internals loads.
Are lateral forces on the guide tubes or support columns in the upper plenum expected to be significant'?
Could these forces contribute to the up-lift forces on the upper core plate or be a consideration. for rod in ertion?
13.
In a table, summarize the pip~ break type, area and location for each analysis.
Specify whether th<: pipe break was used'for cavity/subcompartment, RPV internals, RCL and ECCS piping loads analysis, etc.
14.
In WCAP-9628, Section 3.1, provide the basis that establishes the lowest fluid temperature results in the maximum hydraulic loads for both I
Phase B RCL and Phase C
RPV, internals.
Discuss'and identify the physical mechanism(s) that cause such a condition.
15.
In WCAP-9628, Table 3.1-2, describe the meaning of the 'Upflow or Downflow'olumn.
Discuss any significance relative to Item 3 on page 1-7 of WCAP-9628 and the first paragraph on page 1-3 of WCAP-9748.
16.
Confirm that out-of-RPV cavity pipe breaks will not impose RPV internals loaas that are worse than the RPV cold leg nozzle 'limited break area'ondition considered in WCAP-9748, Section 2.2.1.
17.
Describe any significant variance in RPV internals geometry. which exist on a plant-by-plant basis and how these effects were used to establish plant specific and the generic plant NULTIFLEX models for the Phase B and the Phase C studies.
18.
Provide a table of peak differential pressures and loads for each major internals component analyzed for each generic plant group and each plant specific analysis and indicate time of occurrence for. the Phase C studies.
19.
Provide a copy of the input listing and noding model schematic for NULTIFLEX for the cold leg in-cavity pipe break which was used to evaluate RPV internals loads for each generic plant group in WCAP-9748 (Section 2.2).
20.
For WCAP-9748, Section 2.2, provide the following time-history plots {for the cold-leg RPV nozzle in-cavity break) for each of the generic plant gl oups
~
0 (a)
Differential pressure across the core barrel at two points, 0
and 180, at the same elevation.
(0 is defined as the broken cold leg inlet)
{b)
Differential pressure across the upper core plate (c)
Differential pressure across the lower core plate
{d)
Pressure in the downcomer annulus at the orientation for Item (a).
II I.
QUESTIONS ON STRUCTURAL METHODS,
- MODES, AND EVALUATIONS Gen era 1 Phase B
Section 1.4 1.0 It was stated that pipe, whip could possibly occur on sever al plants but design protection was iot.included in the report.
Please define pipe whip.
(Is pipe whip the formation of a plastic hinge or the possibility of damage to other components and/or supports'?
or what?)
Why are the design protections not included?
Section 2.2 2.0 Expana on steady-state an> transient loadings.
How.are they calculated and how are they applied (i.e., location and direction)?
3.0 Provide details of the more realistic evaluations of.the subcompartment press urizat i ori s, exc luded from the load combinations.
Section 2.3 4.0 Do mass points shown in Fi gure 2.3. 1. 1 rerro.sent the total numoer of mass points used in the ar alysi s?
Justify that the number of mass points used were sufficier t wit4 regard to compo",ent and forcing function frequencies.
5.0 Where is Figure 2.3.1-4?
6.0 Please provide justificat~ons for the use of 4X critical damping in the system analysis.
How is this damping incorporated in the analysis?
7.0 What code was used to determine the support loads? 0 8.0 Were static and dynamic results combined?
Section 2.4 9.0 Justify cri teri a of 90K 'cri tical buck ling load for compression in component supports and FY/vI3 for average shear stresses.
- 10. 0 To s uffici ent ly rev i ew thi s report, we require Refere nce 13.
\\
Phase C
Section 1.0
- 11. 0 Provide page 1-4.
Section 2.1
- 12. 0 Supply justification for using break areas smaller than those for a double offset quillotine break.
How are these areas determined?
- 13. 0 Justify and desicpate the various=break times used in the analyses.
Section 2.1.2
- 14. 0 Referring to Table 2.1-1, how was the peak horizontal force calculated and at what location on the RPV was this load applied.
Al.so, is the peak horizontal load the only loao acting on the PPV?
Section 3.0 15.0 Provide a discussion of the effect of a Hot Leg Nozzle break on the reactor internals respon e.
Section 3.1
16.0 Justify the selection of the 3 generic models developed for the reactor vessel and internals analyses.
Detail the possible differences within each group.
Please be quantitative.
Section 3.1.1 17.0 Provide greater detail concerning the stiffness restraining effects of the RCL (i.e., linear and/or nonlinear 'effects and loop components considered)
'18. 0 Provide complete definition of finite element representations, detailing nonlinearities, general plant layouts, and locations of all nodal masses and siffnesses considered for each generic model (see Figures 3.l-la thru c).
- 19. 0 Provide a better definition of the applied loads, including application points on the mode)s, directions with respect to plant
- layout, and specific types of loadings (dead weight, strain energy release, jet impingement, etc.).
Section 3.2
- 20. 0 Supply allowaole values for the component loads ana stresses presented in Tables 3.2-1, 2, 3, 4, and 5.
Section 3.3
- 21. 0 Supply allowable values for the component loads
..nd deflections presented in Tabl e 3.3-1.
- 22. 0 Provide a>>
i".,.essment of fuel assembly component buckling for the ap pl ied 1 oadi ng cons i d ered.
Section 3,4
- 23. 0 Provide a better definition of the applied
- loads, including application points on th~ models, directions with respect to plant
- layout, and specific typ~ s of loadings (deadweight, strain energy release, jet impingement. etc.).
Section 3.4.2.1
- 24. 0 Provide a discussion of t.he experiment from which the modified frequency i s obtained.
Section 4.0
- 25. 0 Supply the allowable load and stress values referred to in Tables 4.2-1 and 2.
Section 4.3.2
- 26. 0 Supply additiogal response spectra plots similar to Figure 4.3-1 including a discussion'of the time history displacement.
Appendix A
27.0 Supply results of the primary shield wall analysis with comparisons to al lowabl e values.
Appendix A.3
- 28. 0 Present a comparison of the Westinghouse calculated peak pressures to reactor cavity design pressures.
10
IIk
Phase B and Phase C
- 29. 0 Are the designations FPL and FLA for the same owners 30.0 Provide an evaluation of the combined effects of seismic and LgCA loadings.
No internals analysis exist for out of cavity pipe breaks.
Explain.
32.0 Provide an ECCS piping. support evaluation for the appropriate loading conditions.
Plant Specific QUESTIONS
0.
C.
Cook 1 and 2
The D..
C.
Cook Units 1 and 2 are being treated as structural:ly identical sister plants in this evaluation.
The fuel loads presented are slightly different for the two units.
An explanation of this difference is therefore being requested.
Additional uestions 1.0 Supply RCP support drawings and include a discussion of the critical loads and load paths associated with the support.
2.0 Explain in detail the reason for the differences in fuel assembly responses between Units 1 and 2.'
H.
E. Robinson, Questions 1.0 Supply RCP support drawings and include a discussion of the critical loads and load paths associated with the support.
Detail modifications which could be made to eliminate the overstressed condition in the support.
2.0 Supply the following in ormation from the K. B. Robinson fuel system analysis:
(a) core plate motions (b) gri d loads (c) fuel tube and thimble stresses (d) critical gric buccal'ng boa.'e) stress allowables (f) reactor core and fuel system geometry.
3.0 Provide reactor pressure vessel support drawings to include the primary shield wall restraints a ong with a discussion of the calculated loads and the critical load paths.
Zion Units 1 and 2
(}uest i on s 1.0 Supply RCP support drawings and include a discussion of the critical loads and load paths associated with the support.
Detail all modifications needed to eliminate the overstressed conditions.
}
II 2.0 Provioe the following information for the Zion loop 4 hot leg safety injection line:
(a) isometrics (b) support drawings (c) hanger locations (d) design and material parameter s
(e) anchor locations and motions (f)
- a discussion of tne results of the asymmetric loads evaluation, critical stresses and modifications needed to eliminate the over stressed conditions.
I 3.0 Supply drawings and a description of the various load paths associated with the CRDil components.
(a) lifting leg (b) pl at form Include details of tne evaluation performea and the generation of allowable loads and stresses.
Turne Point Units 3 and 4
questions 1.0 Supply steam generator su)port drawings and include a discussion of the critical loads and load paths associated with the support.
Detail all proposed modifications ne.ded to eliminate the overstressed conoitions.
2.0 Provide detailed informat on concerning the maximum imbedment loads, the pnysical structure and thi generation of the allowable load values.
R.
E. Ginna questions 1.0 Supply details concerning the geometry, the loading and the generation of the allowable stresses in the core barrel girth weld area.
Discuss the implications of an overstressed conditions in this area..
II 2.0 Provide the following information for the R.
E. Ginna 2-inch safety injection lines:
(a) isometrics (b) support drawing, (c) hanger location (d) design and mate-ial properties to>
anchor location, and motions (f) a discussion of the results of the asymmetric loads evaluation, critical stress' and modifications needed to eliminate the overstressed cor ditions.
3.0 Supply reactor coolart pump support drawings and include a discussion of the critical loads and load paths associated with the support.
4.0 Detail the refinemen',s needed in the out of,cavity analysis to show steam generator support integrity.
1 4
Surr Units 1 and 2
uestions 1.0 Supply RCP support drawings and include a discussion of the critical loads and load paths associated with the support.
Outline all proposed modifications needed to eliminate predicted overstressed conditions.
2.0 Provide PPV support'rawings and include a discussion of the critical loads and load paths.
Discuss the generation of the allowable load and stress valves.