ML20040H019
| ML20040H019 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/29/1982 |
| From: | Adensam E Office of Nuclear Reactor Regulation |
| To: | Parker W DUKE POWER CO. |
| References | |
| NUDOCS 8202170010 | |
| Download: ML20040H019 (40) | |
Text
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DISTRIBUTION Docket File: 50-413/414
& 29 E DEisenhut LB#4 R/F g
A KJabbour t
MDuncan E
Docket Kos.: 50-413
[g RECE!VED and 50-414 RVollmer FEB 2 1982>
r'C JKramer C
n mn nrumum RMattson 74 8m*m a Hr. Willian 0. Parker, Jr.
RHartfield, MPA
\\"$
Vice President - Steam Production OELD g
Duke Power company 01&E (3)
D P.O. Box 33189 bcc: LPDR TERA Charlotte, North Carolina 28242 NRCPDR ACRS (16)
Dear tir. Parker:
Subject:
HdC Mechanical Engineering Branch (HEb) Age'Ja Iteus for Discussion with Duke on Catawba Station Attached (Enclosure 1).is a list of items which the McB would like to discuss with you in a series of meetings. The first meeting is scheduled for the week of March 29, 1962, in Charlotte. We anticipate that other questions and concerns may arise as a result of the acetinus. Thus, the attached list should not be Consioered a final or complete list of items to be resolved prior to issuing a Safety Evaluation Report.
If you require any clarification of this' letter, please contact the project Gandger, Kahtdn Jabbour, at (301) 492-7621.
The reporting and/or record keeping requirements contained in this letter af fect fewer than' ten respondents; therefore, oho cleararce is not required under P.L. 9u-bil.
Sincerely, l
l l
l Elinor G. Adensau, Chief Licensing Brancu #4 Division of Licensing l
Enclosure:
As stated l
cc: See next page 8202170010 820129 PDR ADOCK 05000413 A
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. CATAWBA Mr. William 0. Parker Vice President - Steam Production Duke Power Company P.O. Box 33189 Charlotte, North Carolina 28242 cc: William L. Porter, Esq.
North Carolina Electric Membership Duke Power Company Corp.
P.O. Box 33189 3333 North Boulevard Charlotte, North Carolina 28242 P.O. Box 27306 Raleigh, North Carolina 27611 J. Michael McGarry, III Esq.
Debevoise & Liberman Saluda River Electric Cooperative, 1200 Seventeenth Street, N.W.
Inc.
Washington, D. C.
20036 207 Sherwood Drive Laurens, South Carolina 29360 North Carolina MPA-1 P.O. Box 95162 James W. Burch, Director Raleigh, North Carolina 27625 liuclear Advisory Counsel 2600 Bull Street Mr. F. J. Twogood Columbia, South Carolina 29201 Power Systems Division Westinghouse Electric Corp.
Mr. Peter K. VanDoorn P.O. Box 355 Route 2. Box 179N Pittsburgh, Pennsylvania 15230 York, South Carolina 29745 Mr. J. C. Plunkett, J r.
Janes P. O'Reilly, Regional Administrator NUS Corporation U.S. Nuclear Regulatory Commission, 2536 Coun'tryside Boulevard Region II Clearwater, Florida 33515 101 Marietta Street, Suite 3100 Atlanta, Georgia 30303 Mr. Jesse L. Riley, President Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 28208 Richard P. Wilson, Esq.
Assistant Attorney General l
S.C. Attorney General's Office P.O. Box 11549 i
Columbia, South Carolina 29211 Walton J. McLeod, J r., Esq.
General Counsel South Carolina State Board of Health J. Marion Sins Building 2600 Bull Street Columbia, South Carolina 29201 l
l
CATAWBA Walton J. McLeod, Jr., Esq.
Mr. William O. Parker, Jr.
~ Mr. William O. Parker, Jr.
G:n:ral Counsel Vice President -
Vice President -
SC Stste Board of Health Steam Production Steam Production J. M:rion Sims Building Duke Power Company Duke Power Company 2600 Bull Street P.O. Box 33189 P.O.-Box 33189 Columbia, SC 29201 Charlotte, NC 28242 Charlotte, NC 28242 William L. Porter, Esq.
M,r. Jessee L. Riley, President North Carolina Electric Duke Power Company Carolina Environmental Study Membership Corp.
P.O. Box 33189-Group Chtriotte, NC 28242 854 Henley Place Charlotte, NC 28208 J. Michael McGarry, III, Esq.
James W. Burch, Director Debsvoise & Liberman Nuclear Adivsory Counsel 1200 Seventeenth Street, NW.
2600 Bull Street Washington, DC 29036 Columbia, SC 29201 N2rth Carolina MDA-1 Mr. Peter K. VanDoorn P.O. Box 95162 Route 2, Box 179N Raleigh, NC 27625 York, SC 29745 Mr. R. S. Howard Powar Systems Divisicn Wastinghouse Electric Corp.
P.O. Box 355 Pittsburgh, PA 15230 Mr. J. C. Plunkett, Jr.
NUS Corporation 2536 Countryside Boulevard Clearwater, FL 33515 Richard P. Wilson, Esq.
Assistant Attorney General S.C. Attorney General's Office P.O. Box 11549 Columbia, SC 29211 North Carolina Electric Membership Corp.
3333 North Boulevard P.O. Box 27306 Reloigh, NC 27611 l Saluda River Electric l
Cooperative, Inc.
207 Sherwood Drive Laurens, SC 29360 i
ENCLOSURE 1 AGENDA ITEMS FOR MEETING WITH DUKE POWER COMPANY ON CATAWBA STATION MECHANICAL ENGINEERING BRANCH 3.2 CLASSIFICATION OF STRUCTURES, COMPONENTS, AND SYSTEMS 3.2.1 Seismic Classification Table 3.2.1, page 1 Explain the reasons behind the classification of the " Waste Evaporator Feed" and the " Vent. Unit Cond. Drain".
Table 3.2.1, page 2 Provide safety related category for RCP Motor Oil Fill Tank.
3.2.2 Mechanical System Quality Group Classification l
3.2.2.1, page 3.2-3 Where has an orifice of size greater than 3/8" ID been used to protect a component in the reactor coolant pressure boundary?
3.2.2.1, page 3.2-4 The applicant states that Safety class 2 applies to components of the reactor coolant pressure boundary not covered in Safety Class 1.
What components of the reactor coolant pressure boundary are not Safety Class 1?
Table 3.2.1-2, page 1 Auxiliary Steam System references Note 22.
Where is Note 22?
It is also referenced on pages 3, 7, 8, 11, 13, and 14.
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Table 3.2.1-2, page 4 Explain the reasons behind the ASME Code and Safety Class classifications of Fuel and Control Rod Assemblies and Burnable Poison Rod Assemblies and Control Rod Drive Mechanisms.
Table 3.2.1-2, page 4 Explain the ASME Code and Safety Class classificati.on of the reactor vessel internals.
Table 3.2.1-2, page 6 The table shows the Nuclear Sampling System as NNS, Code III-3 but not seismically designed.
I.s this consistent?
Table 3.2.1-2, page 9 In the' Diesel Generator Engine Starting Air System the starting air tanks are seismically designed yet the air compressors are n o 't.
How many starts can be provided by the air tanks?
Table 3.2.1-2, page 15 Footnote 21 references Engineering Justification Report SES-JR-10. Provide a copy of this report.
Table 3.2.2-3 Footnote 1 states that Duke Power Company established ao
" effective code date" for the station in accordonce with 10 2-
1 i
CFR 50.55a, reviews and may-elect to comply with portions or all the latest versions of the above cedes.
This i s acceptable as long as different versions of the ASME Code.
are not used for evaluation of the same system or component.
Provide a commitment to this effect.
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3.6.2 Determination of Break Locations and Dynamic Effects Associated with the Postulated Rupture of Piping 3.6.2.1.1, page 3.6-8' In the primary loop, what size breaks are postulated for the design of pipe whip restraints?
What size breaks are postulated in the primary loop for determination of compartment pressurization and asymmetric loads?
If breaks for either case are less than full size, provide justification.
3.6.2.1.1.1, page 3.6-9 The FSAR states that breaks are postulated at locations where the cumulative usage factor exceeds 0.2 for normal and upset operating conditions.
(a)
Breaks are to be postulated where the cumulative usage factor exceeds 0.1.
Show that your analysis complies with the SRP.
~
(b)
The criteria is evaluated under loadings resulting from normal and upset operating conditions including the OBE.
Change this to comply with the SRP.
3.6.2.1.1.2, page 3.6-9 Identify the analysis, experimental data or physical restraint characteristics which have been used to limit pipe displacement.
For and breaks with areas other than the full cross-sectional area of the-pipe.
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3.6.2.1.1, page 3.6-10 Item (b) mentions encased or. jacketed piping.
Where has it been used?
Show that the required inservice inspection can be performed.
3.6.2.1.1, page 3.6-11 Item j)'provides criteria for describing the effects an impingement jet has 6n various sizes of target pipe.
It is the staff's position that this criteria has not yet been demonstrated to be valid when applied to jet impingement.
Therefore, either:
(1)
Remove this cr.iteria from the FSAR and provide a commitment to properly evaluate the cases where it was used; or (2)
Demonstrate through experimental and analytical data that such criteria are, in fact, conservative and correct.
(Note: The comparison to the whipping pipe recently proposed on the McGuire docket is not accepta.ble.)
3.6.2.1.2, page 3.6-11 Item j) indicates that certain exceptions to the criteria may be taken.
In addition to indicating where these exceptions are taken, provide the analytical or experimental basis (es) for the exceptions.
.5
3.6.2.1.2.1, pages 3 6-13 to 16 This pipe break criteria is not in compliance with the latest version of SRP 3.6.2 (NUREG-0800).
3.6.2.1.2.1, page 3.6-16 Item c)2)i) indicates the criteria for selecting intermediate pipe break locations in Duke Class E, G, and H piping systems.
It is the staff's position that, since Table 3.2.2-3 indicates that these piping systems are not designed for seismic loadings, pipe breaks should be postulated so as to' clearly demonstrate that failure of the system will not result in any loss of capability of assential systems and components to withstand the further effects of any single active component failure and still perform all functions required to shutdown the reactor and mitigate the consequences of the postulated piping failure.
Therefore, provide a commitment to meet this position.
- 3. 6. 2.1. 2. 3, p a g e 3. 6-17 Item a)3 states circumferential breaks are assumed to result in pipe separation of one diameter displacement of ruptured piping sections unless physically limited by piping restraints.
Show where piping restraints are used to limit pipe displacements.
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3.6. 2.1. 2. 3, p a g e 3. 6-18 Identify in Item a)4 where limited pipe displacement, line restrictions, flow limiters, positive pump - controlled flow, and the absence of energy reservoirs are used to reduce the jet discharge.
- 3. 6. 2.1. 2. 3, p a g e 3. 6-18 In Item a)5 are all-possible targets of-the whipping pipe examined?
- 3. 6. 2 '.1. 2. 3, p a g e 3. 6-19 In Item c)4 elaborate on the statement, "Throughwall cracks are not postulated inside containment."
3.6.2.2.1, page 3.6-21 The computer codes SATAN-IV and THRUST should be included in the list of computer codes in Section 3.9.1.2.3.
3.6.2.2.2.1, page 3.6-23 Is the discharge coefficient equal to something other than 1.0 for any. conditions?
3.6.2.2.2.3, page 3.6-25 In the first paragraph, the applicant states that a dynamic load factor of 2.0 shall be used in the absence of an
(
analysis justifying a lower value.
What kind of
(
0 analysis i s performed?
How low a value is used?
Justify any values less than 2.0.
i 3.6.2.3.1, page 3.6-25 In Item 1) of General Criteria for Pipe Whip Evaluation, what kind of analysis justifies a value lower than 2.0?
In Item 2) nonlinear pipe and restraint material properties may be used as applicable.
Where have nonlinear properties been used and what values are used?
In Item 3) all targets of a whipping pipe must be looked at.
Provide assurance that this has been done.
3.6.2.3.3, page 3.6-26 The last sentence references Section 3.9.3.1.5 for a discussion of where restraints have been designed to function as supports.
Provide locations of where this has been done and analysis techniques used.
3.6.2.4, page 3.6-26 The second paragraph discusses the process pipe making up the pressure boundary.
What size is the process pipe?
3.6.2.4.2, page 3.6-28 I
Provide details of field welds between the guard pipe attachment and reactor building anchor section as discussed in Item b.
8
3.6.2.4.4, page 3.6-29 This section refers to Section 6.6 for a discussion of access for inservice inspection.
Item 3)- of Section 6.6.8 (page 6.6-2) indicates that access ports will be provided where possible.
In addition to indicating which welds are not examinable,because of a lack of access, provide an engineering basis for why the ports are not possible.
Are there any " break exclusion" regions used in any seismic Category I piping systems?
Ta b l e 3. 6.1 -1 Provide details of the Safety Injection System and Main Steam System relative to the pipe break protection method (c).
Table 3.6.1-3, page 2
-Explain comparison of SAR Section 3.6.1.1.2/with NRC criteria.
What analyses have been performed to show insufficient energy to develop sudden failure?
In Item a) are the consequences of breaks in the excluded lines l
considered?
What criteri'a are used to postulate breaks l
and cracks, and in which systems?
Table 3. 6.1 -3, pages 2 and 3 This item notes Duke's exception to postulating terminal end breaks at the shut off valve which separates the pressurized and non-pressurized portions of a piping run.
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It is the staff's position that this is the only logical location for such a terminal end.
Postulating the break elsewhere in the pressurized side limits the length of whipping pipe available.
Postulating the break anywhere in the nonpressurized side would not result in a release of energy.
Therefore, provide a commitment to meet this position or alternatively, treat the entire piping system as if it were a high energy piping system and postu, late breaks and provide protective measures accordingly (i.e.,
assume the shut _off valve to be normally open.)
Table 3.6.1-3, page 3, SAR Section 3.6.2.4 Provide an example of the analysis con. ducted to assure that Duke penetration designs are acceptable.
T a b t e '3. 6.1 -3, pages 3 and 4, SAR Section 3.6.2.1.2.1 The criteria is a deviation from the SRP.
A minimum of two break location's is required.
Table 3.6.2-1 In Item 11 are there no intermediate breaks postulated in the pressurizer surge line?
l Table 3.6.2-2 Are all penetrations Duke Class B?
If not, are they evaluated for emergency conditions?
I.
l Table 3.6.2-3 Define variables -- for example, F F t, F
etc.
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Table 3.6.2-4 This table provides stress allowables for pipe rupture restraints.
It is the staff's position that the strain limit for such restraints is one half of the ultimate strain.
Therefore, provide assurance that your stress based criteria is always as conservative as the above strain criteria.
Table 3.6.2-6, page 1 Provide stresses fop' rerouted pipe at break-locations 1NI-122-048, -049, -050 and 1NI121-050, -051, -052, -053.
-Table 3.6.2-6, pages 1 and 2 What are the allowable stresses?
Table 3.6.2-8, page 4 Provide info.rmation on re. routed pipe for break locations noted above.
Figures 3.6.2-6,
-7,
-8 Provide details of welds in penetration areas.
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i 3.7.3 Seismic Subsystem Analysis 3.7.1.3, page 3.7-3 Justify critical damping values for' bolted steet j
structures.
3.7.1.3, page 3.7-3 The fourth paragraph discusses the CRDMs and their seismic supports.
What portions of the CRDMs are seismically designed?
3.7.2.1.1.3, page 3.7-9 What criteria is used for determining significant modes?
3.7.2.5, page 3.7-14 The applic' ant stares "When the ground response spectra are used the acceleration values corresponding to 20 Hz are used as a minimum,value for the design of piping and l
components.
The acceleration values at 20 Hz are greater than the values corresponding to a rigid system and therefore are conservative."
Provide an example of this design method for piping and components.
l 3.7.2.7 page 3.7-15 l
The method of combining modal responses is acceptable only if the absolute value of "the product of the responses of I
the modes in each group of closely spaceo modes and a coupling l
. I2-
.he factor c" is added to the square root.of the sun.
squares of all modes.
The equation should be:
N N.
N S
j J
2+2 R
R C
R
=
R g
g T
1
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j=1 K=M.
i=K+1 J
- 3. 7. 2. 7, p a g e 3. 7-15 What is the duration of the operating basis earthqGake?
5 3.7.3.1.2, page 3.7-20 i
Several referenced figures, Figures 3.7.3-1, 3.7.2-1, are missing.
Provide a timetable for their inclusion in the FSAR.
3.7.3.2, page 3.7-28 How many eartnquake cycles are postulated for BOP piping?
The number of cycles for NSSS components is not given.in Table 3.9.1-1 as stated in the first paragraph. Provide this information.
3.7.3.5, page 3.7-29 For which seismic piping has the equivalent' static load method of analysis been used?
Provide an exanple of this analysis.
/3
3.7.3.8, page 3.7-29 The applicant states that,the specific analytical procedures used in qualifying the pipe depend on its size, temperature, structural frequency, and other factors as discussed in this section.
What criteria is used in qualifying the pipe?
3.7.3.8.1, page 3.7-30 A period of 0.033 seconds corresponds to a frequency of 30.3 cps.
In Regulatory Guide 1.60 the cutoff is 33 cps.
Justify the use of the lower frequency as 'the cutoff.
3.7.3.8.3, page 3.7-32 and Table 3.7.3-1 Provide a detailed example of the " Alternate Analysis of Flexible Piping" including the evaluation of Seismic Anchor Motion.
Explain the use of.the 15/8 factor and the 2g and i
39 limits on page 3.7-33.
i 3.7.3.9, page 3.7-33 The method for analyzing multiply-supported components is unacceptable.
The appropriate method was either a response spectrum that envelopes the response spectra at all support elevations or multiple response spectra inputs.
Provide a l
commitment to this method of analysis.
l l
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e 3.7.3.9, page 3.7-34 The stresses caused by differential seismic motion of piping are secondary stresses for piping, but are primary stresses for pipe supports. They are not secondary stresses as noted.
Provide a commitment that your analysis reflects this.
3.7.3.13, page 3.7-35 Provide an example of how the seismic boundary is protected in cases where the seismic and non-seismic piping connect.
EDITORIAL COMMENTS 3.7.3.4, page 3.7-29 The applicant states that rigid equipment / support systems have natural frequencies greater than 33 Hz as noted in Section 3.7.3.1.
Is this a correct reference?
I I
5
3.9 MECHANICAL SYSTEMS AND COMPONENTS 3.9.1.1, page 3.9-1 Emergency conditions are omitted here.
Does this mean that the Small Steam Break and Small LOCA in Table 3.9.1-1 (page
- 2) are evaluated using faulted limits?
3.9.1.1, page 3.9-7 How many cycles of OBE earthquake are used in the analysis?
3.9.1.1, page 3.9-8 The first caragraph lists conservative assumptions.
Assumption a) is that "the reactor is initially in a hot, zero power condition."
Is this actually conservative?
3.9.1.2, page 3.9-9 Are al.l computer codes used in the analyses included in this section?
Provide the referenced documents used for verifi-ca" tion of all codes?
3.9.1.2.1, page 3.9-9 This section' states that " analyses must be in accordance with the functioral and structural analysis requirements set forth in the applicable Duke specifications."
Provide an example of the functional and structural analysis requirements set forth in applicable Duke specifications.
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3.9.1.4.3, page 3.9-14
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The applicant. states that "The cozppnent upper and lower lateral supports are inactive'dur'ing plant heatup, cooldown s
and normal plant operating conditions.
However, these r,
restraints become activewhen' thel. plant is at power and i
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, under the rapid motions of the' reactor coolant loop
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c)o ponents that occur from the d9n'anic loadings and are ss
.s represented by stiffne,ss matrices 'and/or individual tension or compression spring members in'the dynamic model.
The analyses are" performed;at the full' power condition."
s What comped nt does this paragraph refer to?
Is the full g
t power; condition a normal operat,ing condition?
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3.9.1.4.3, page 3.9-15 9s The applicant states that "The modal amplitudes are then converted to displacements in t 0 global coordinate system the horresponding m, ass point.
From these and applied to data the forces, moments,.deflectiohs, rotations, support reactions and piping stresses a r.e calculated,for all significant modes."
What criteria is used to determine the significant modes?
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3.9'.1.4.3, page 3.9-15 Justify the use of 4% critical damping for the loss of coolant accident?
3.9.1.4.3, page 3.9-16 The last paragraph refers to Figure 3.9.1-2.
Provide this figure.
3.9.1.4.4, page 3.9-19 The third paragraph says that seimsic analyses are performed individually for the reactor coolant pump, the pressurizer and the steam generato.r.
How is the connected piping handled for these components?
3.9.1.4.4, page 3.9-19 The ta'st sentence of the third paragraph says that quali-tication of the rea,ctor pressure vessel is by a " static stress analysis based on loads that have been derived from dynamic analysis."
Provide details of this analysis and justify its use.
3.9.1.4.4, page 3.9-19 In the third paragraph, the applicant states that the
" Seismic analyses for the steam generator and pressurizer are performed using 2 percent damping for the OBE and 4 percent damping for the SSE."
Justify the use of 4%
critical damping for the SSE.
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3.9.1.4.5, page 3.9-20 The applicant states " Pipe displacement restraints installed in the primary shield wall limit the break opening area of the vessel nozzle pipe breaks.
An upper bound break area of 85 square inches was determined, taking into account the primary shield wall pipe restraints and vessal and pipe relative motions from similahplant analyses.
Detailed studies have shown that pipe breaks at the hot or cold leg reactor vessel nozzles, even with a limited break area, would give the highest rea'ctor vessel support loads and the highest vessel displacements, primarily due to the influence of reactor cavity pressurization."
Provide supporting evidence for an upper bound break area of 85 square inches.
Were the detailed studies which showed that pipe breaks at reactor vessel nozzles iive highest vessel support loads and displacement performed for the Catawba plant?
If performed for a similar plant, show that they are indeed similar to Catawba.
3.9.1.4.6, page 3.9-23 The first paragraph of item 1 identifies stress limit criteria which are not in compliance with Appendix F of the ASME Code.
Provide a commitment to the stress limit criteria of Appendix F of the ASME Code.
_. /1
3.9.1.4.6, page 3.9-23 The applicant states "If plastic componer.t analysis is used with elastic system analysis or with plastic system analysis, the deformations and displacements of the individual system members wilL be shown to be no larger than those which can be property calculated by the analytical methods used for the system analysis."
Indicate when this has been used.
3.9.1.4.7, page 3.9-24 The first paragraph. indicates that elastically determined stresses wilL be compared against inelastic limits.
This approach is not one of the methods listed in Appendix F of the ASME Code.
Provide an example of.the analyses and provide assurance that this method is at least as consecvative as those in Appendix F.
EDITORIAL COMMENT The applicant states " Dynamic Seismic analysis for the SSE is performed on this piping' utilizing the response spectrum l
method in accordance with USNRC Regulatory Guide 1.92."
Is this a correct reference?
3.9.1.4.7, page 3.9-24 Provide your interpretation of jurisdictional boundaries as they pertair to NF supports.
Justify your position.
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3.9.2.1, pages 3.9-24 and 25 More detail is needed for the NSSS and BOP preoperational vibration testing program.
The applicant has not given a clear description of the BOP acceptance criteria for steady-state piping vibrations.
The NSSS program has not been adequately described.
What are the acceptance criteria for steady-state and transient vibration testing? Will snubbers be checked?
To what transients will the piping be subjected?
Which lines will be instrumented? What types of instrumentation will be used?
If not instrumented, how will the visual observations be performed?
What is the criteria for testing high flow velocity steam lines?
3.9.2.1.1, pages.3.9-24 and 25 The second paragraph of this section refers to Table 3.9.2-1 which lists piping systems subject to steady-state vibration testing.
Standard Review Plan 3.9.2 states that "The Systems to be monitored should i,nclude a) ASME Code Class 1,
2, and 3 systems, b) other high-energy piping systems inside Seismic Category I Structures, c) high-energy portions of systems whose failure could reduce'~ tee functioning of any Seismic Category I plant feature to an unacceptable safety level, and d) Seismic Category I portions of moderate-energy piping systems located outside containment.".
Confirm that all piping systems in the above categories are in Table 3.9.2-1.
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3.9.2.3, page 3.9-32 The first full. paragraph mentions three plants which provide prototypical data.
Which of the three plants -Indian Point is the valid prototype for 2, Trojan or Sequoyah 1 Catawba?
3.9.2.3, page 3.9-33 Item 3 discusses vibration of the upper internals.
The second to the last sentence says that " Applying a 5%
increase in level to the high factors of safety deduced from Sequoyah 1 data results in adequate margins for Catawba i
upper internals.".
Wh'at are the margins for the upper internals?
3.9.2.4, page 3.9-33 Provide Figure 3.9.2-1.
3.9.2.5, page 3.9-35 The first full paragraph states that "For faulted conditions, stresses are above yield in a few locations.
l For these cases only, some inelastic stress limits are e
applied.". Define the inelastic stress limits and indicate where these limits have been applied.
3.9.2.5 Previous analysis for other nuclear plants have shown that l
l certain reactor system components and their supports may be i
l l
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22.
l l
2 subjected to previously under-estimated asymmetric loads under the conditions that result from'the postulation of ruptures of the reactor coolant piping at various locations.
The applicant has described the design of the reactor
. internals for blowdown loads only.
The applicant should also provide information on asymmetric loads.
It is, therefore, necessary to reassess the capability of these reactor system components to assure that the calculated i
dynami'c asymmetric leads resulting from these postulated pipe ruptures will be within the bounds necessary,to provide high assurance that the reactor can be brought safely to a cold shutdown condition.
The reactor system components that require reassessment shall include:
a.
b.
Core supports and other reactor internals.
c.
Control rod drives, d.
ECCS piping that is attached to the primary coolant piping.
e.
P.rimary coolant piping.
f.
Reactor vessel supports.
l The following information should be included in the FSAR about the effects of postulated asymmetric LOCA loads on the above mentioned reactor system components and the various cavity structures.
-23
4 1.
Provide arrangement drawings of the reactor vessel support systems in sufficient detail to show the geometry of all principal elements and materials of cort +ruction.
2.
If a plant-specific analysis will not be submitted for your plant, provide supporting information to demonstrate that the generic plant analysis under consideration adequately bounds the postulated accidents at your facility.
Include a comparison of the geometric, structural, mechanical, and thermal-hydraulic similarities between your facility and the case analyzed.
Discuss the effects of any differences.
3.
Consider all postulated breaks in the reactor coolant piping system, including the followint locations:
a.'
Steam line nozzles to piping terminal ends, b.
Feedwater nozzle to piping terminal ends.
c.
Recirculation inlet and outlet-nozzles to recircu-Lation piping terminal ends.
4.
Provide an assessment of the effects of asymmetric l
pressure differentials
- on the systems and components 1
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- Blowdown jet forces at the location of the rupture (reaction forces), transient differential pressures in the annular region between the component and the wall, and transient differential pressures across the core barrel within the j
l reactor vessel.
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Listed above in combination with all external loadings including. safe shutdown earthquake loads and other faulted condition loads for the postulated breaks desc:'. bed above.
This assessment may utilize the following mechanistic effects as applicable:
a.
Limited displacement' break areas.
b.
Fluid-structure interaction.
c.
Actual time-dependent forcing function.
d.
Reactor support stiffness.
e.
Break opening times.
5.
If the results of the assessment on item 3 above indicate loads leading to inelastic action of these systems or displacement exceeding previous design limits, provide an evaluation of the inelastic behavior f
(including strain hardening) of the material used in the system design and the effect of the load trans-mitted to the backup structures to which these systems are attached.
6.
For all analyses performed, include the dethod of analysis, the structural and hydraulic computer codes employed, drawings of the models employed and comparisons of the calculated to allowable stresses I
and strains or deflections with a basis for the allowable values.
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7.
Demonstrate that safety-related components will retain their structural integrity when subjected to the combined loads resulting from the loss-of-coolant accident and the safe shutdown earthquake.
8.
Demonstrate the functional capability of any essential piping when subjected to the combined loads resulting from the loss-of-coolant accident and the safe shutdown earthquake.
3.9.3, page 3.9-40 and 41 The third paragraph on page 3.9-40 and the second full paragraph on 3.9-41 refers to ~Section 4.5.2 for design Loading conditions for core support structures.
Thi.s section discusses Reactor Internals Materials.
Provide the appropriate reference.
3.9.3.1.1, page 3.9-41 The first paragraph refers to the Emergency Condition in Table 3.9.3-1 and to the stress limits for each loading combination in Tables 3.9.3-2,
-3,
-4,
-5 and
-6.
The Emergency Condition is missing from Table 3.9.3-1 and no strass limits are given for the Emergency Conditions.
Provide the missing information.
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3.9.3.1, pages 3.9-40 to 43 This section does not address the criteria used to assure the functional capability of essential systems when they are subjected to loads in excess of those for which Service Limit B limits are specified.
By essential systems are meant those ASME Class 1, 2 and 3 and any other piping systems which are necessary to shut down the plant following, or to mitigate the consequences of an accident.
Provide such criteria.
3.9.3.1, pages 3.9-40 to 43 This section does not cover bolts.
Provide service limits for bolts.
3.9.3.1.5, page 3.9-43 The first paragraph refers to Tables 3.9.3-1.1 and 3.9.3-12 for load combinations for supports, restraints, anchors, and snubbers.
However, these tables are indicated as only being l
applicable for Duke Class A,
B, C, and F items.
Provide the load combinations for Wes.tinghouse items.
3.9.3.1 Sections 3.9.3.1.5 (page 3.9-43) and 3.9.3.4 (page 3.9-51),
while discussing supports and restraints, do not provide sufficient information with respect to snubbers. Provide the
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basis for selecting the location, the required load capacity, and the structu.al and mechanical performance parameters of all safety-related snubbers (mechanical and hydraulic) and the method of achieving a high level of operability assurance including:
1.
A description of the analytical and design methodology utilized to develop the required snubber locations and charscteristics.
2.
A discussion of design specification requirements to assure that required structural and mechanical performance characteristics and oroduct quality are achieved.
3.
Procedures, controls to assure correct installation of snubbers and checking the hot and cold settings during plant startup tests.
4 Provisions for accessibility for inspection, testing and repair or replacement of snubbers.
3.9.3.1.5, page 3.9-43 The second paragraph says that " loads for each loading
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combination are combined algebraically except that components which contain positive and negative values are combined to assemble the worst case load combination." Provide an example of what is done here.
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3.9.4.3.1, page 3.9-58 The first full paragraph says that "th'e dynamic behavior of the reactivity control components has been studied using experimental test data and experience from operating reactors."
Provide details of the tests and data.
3.9.S.1, page 3.9-62 The first full paragraph refers to Figure 3.9.S-1.
Provide this figure.
3.9.5.1, page 3.9-63 The third paragraph. discusses the energy absorbers.
Provide details of the analysis.
What do they look like?
How much deflection is there?
3.9.6, page 3.9-68 The applicant must provide a commitment that the inservice testing of ASME Class 1, 2, and 3 components will be in r
l accordance with the revised rules of 10 CFR, Part 50, Section j
50.55a, paragraph (g).
3.9.6, page 3.9-68 Any requests for relief from ASME Secti~on XI should be submitted as soon as possible.
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o 3.9.6 There are several safety systems connected to the reactor coolant pressure boundary that have design pressure below the rated reactor coolant system (RCS) pressure.
There'are also some systems which are rated at full' reactor pressure on the discharge side of pumps but have pump suction below RCS pressure.
In order to protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high pressure RCS and the low pressure systems.
The leak tight integrity of these valves must be ensured by periodic Leak testing to prevent exceeding the design pressure of the l,o w pressure systems thus causing an inter-system LOCA.
Pressure isolation valves are required to be category A or AC per IWV-2000 and to meet the approriate requirements of IWV-3420 of Sectio'n XI pf the ASME Code except as discussed below.
Limiting Conditions for Operation (LCO) are required to be added to the technical specifications which will require corrective action; i.e.,
shutdown or system isolation when the final approved leakage limits are not met.
- Also, surveillance requirements, which will state the acceptable leak rate testing frequency, shall be provided in the technical soecifications.
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Periodic leak testing of each pressure isolation valve is required to be performed at least once per each refueling outage, after valve maintenance prior to return to service, and for systems rated at less than 50% of RCS design pressure each time the valve has moved t r ora its fully closed position unless justification is given.
The testing interval should average to be approximately one year.
Leak testing should also be performed after all disturbances to the valves are complete, prior to reaching power operation following a refueling outage, maintenance, etc.
The staff's present position.on leak rate limiting conditions for operation must be equal to or less than 1 gallon per minute for each valve (GPM) to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function and give an indication of valve degradation over a finite period of time.
Significant increases over this limiting valve would be an indication of valve degradation from one test to another.
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Leak rates higher than 1 GPM will be conside' red if the leak rate changes are below 1 GPM above the previous test leak rate or system design precludes measuring 1 GPM with sufficient accuracy.
These items will be reviewed on a case by case basis.
_ ji/ -
The Class 1 to Class 2 coundary will be considered the isolation point which must be protected by redundant isolation valves.
In cases where pressure isolation is provided by two val'ves, both will be independently leak tested.
When three or more valves provide isolation, only two of the valves need to be leak tested.
Provide a list of all pressure isolation valves included in your testing program along with four sets of Piping and Instrument Diagrams which describe your reactor coolant system pressure isolation valves.
Also discuss in detail how your leak testing program will conform to the above staff position.
Table 3.9.1 -1, page 1 Explain Note la as.it refers to the Inadvertent Auxiliary Spray transient.
Table 3.9.1-3 The footnote indicates that a test may be performed in lieu of an analysis to determine ASME Code compliance. Provide the criteria for such tests.
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Tables 3.9.1-3, 3.9.3-7, and 3.9.3-8 What are the attowable stresses?
Table 3.9.3-6 Note 4 to Table 3.9.3-6 indicates that the design 1
requirements are not applicable to parts contained within the valve or w'hich are not part of the pressure b5undary.
1 It i s the staff's position that the valve disc is a part of the pressure boundary.
Therefore, indicate the design criteria for valve discs when subjected to "Pmax"-
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d 3.2.2, page 3.2-1 Are the safety and relief valve piping on the main stream line classified as safety-related?
Are there any other piping > 21/2" connecthdtothemainsteam Line up to the outermost containment isolation valve?
If so, what is its safety classificatior?
Explain the design of these portions of structures and systems that form an interface between-Seismic category I and non-Seismic Category I features.
What GA requirements are applied to those systems, structures, and components?
(Q210.1)
Provide a discussion of your compliance w/R.G. 1.29.
3.6.1.1.1, page 3.6-1 The FSAR stated that " reactor coolant piping i s restrained such that the lateral displacements of the broken ends of the pipe are less than the pipe watt thickness."
Provide the assumptions and the analytical results to verify this statement.
l The FSAR states that system response due to breaks in the RCS are " accommodated directly by the supporting structures of the reactor vessels the steam generator, and the reactor coolant pumps including two additional pipe supports."
Provide the assumptions and analytical results to justify the statement.
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3.6.2.1.1, page 3.6-7 Describe the analysis performed to verify the integrity and operability of the isolation valves for a pipe break beyond the restraint.
i 3.6.2.1.1.1, page 3.6-9 Show locations of pipe whip restraints on the reactor coolant piping and for which breaks in the RCS they are designed.
3.6.2.~1.1.2, page 3.6-10 A Longitudinal break (Break 7) at 50* elbow on the Intrados assumes a break area less than the cross-sectional area of the pipe.
Provide the analytical and experimental bases for'the Limited break.
(Reference 1 does not contain the assumptions).
3.6.E 2,2.3, page 3.6-24 In the equation used to calculate jet impingement loads, explain the use of "cosf ".
Explain how the total cross-sectional area of the jet at the target structure (Aj) is calculated.
3.6.2.3.3, page 3.6-26 Describe what buckling criteria and limits are used in the design of pipe whip restraints.
3.6.2.5, page 3.6-29 Item c) - The locations of pipe whip restraints are not shown
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in the Figures 3.6.2-9 thru 3.6.2-36.
Provide locations of all pipe whip restrai.nts, jet barriers, and enclosures.
3.7.2.'14, page 3.7-36 Providd justification that for the Catawba SSE, the fuel assumbly displacements are large enough to result in no damping values less than 10%.
3.9.2.3, page 3.9-32 The applicant states that "since catawba has a slightly higher flow rate than Trojan, vibration levels due to the core barrel excitation are expected to be somewhat greater,than those of Trojan."
Provide assurance that the vibration levels in the will net cause. fallare of the. rea cAct-reactor 4 neutron shield bottings which might result in the dropping of the neutron shielding pads.
Unit 1 incorporates a Model D3 steam generator.
Unit 2 incorporates a Model D5 steam generator.
Provide assurance that the Model D3 l
S/G tubes are adequately designed to prevent failure and adverse secondary side Leakage.
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- s 3.7.3.6 page 3.7-29
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This section refers to 3.7.2.6 for a discussion of the method Westinghouse uses to account for three components of earthquake riot. ion.
Foe the N555 piping, hob 'are the three components of earthquake motion handled in 'the seismic.
i analysis?
3.9.3.1 A table of stress criteria and design loading combinations similar to Tables "3.9.3-1 and Table 3.9.3-2 is required for core support structures and component supports.
l 3.9.3.3.1, page 3.9-50 The ASME code is referenced several times. Any reference to the ASME code should specify the part of the code being referenced.
3.9.5.2,page 3.9-65 The applicant has not ' included asymetric loads in the list of design loding-conditions for the reactor internals.
Assurances must be provided the the reactor l
internals have been analyzed for asymetric loads.
Table 3.9.2-2 Table 3.9.2-2 lists the maximum deflections for reactor internal support structures. The allowable and the no-loss-of function deflections are the same for the upper barrel (radial) component.
Provide assurances that-this provides -
l for an adequate margin of safety.
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