ML20081C065
| ML20081C065 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 10/05/1983 |
| From: | Wiesemann R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19268E290 | List: |
| References | |
| CAW-83-81, NUDOCS 8310310083 | |
| Download: ML20081C065 (6) | |
Text
-__.....
,o NuclearTechnology Division Water Reactor Westinghouse Electric Corporation Divisions 8c,333 PittsburghPennsylvania13230 i
October 5, 1983 CAW-83-81 1
Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C.
20555 APPLICATI0M FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE I
REFERENCE:
Georgia Power Company letter to NRC dated October 1983
Dear Mr. Denton:
The proprietary material for which withholding is being requested in the reference letter by the Georgia Power Company (GPC) is further identified in an affidavit signed by the owner of the proprietary information, The affidavit, which accompanies this Westinghouse Electric Corporation.
letter, sets forth the basis on which the information may be withheld from
]
public disclosure by the Commission and addresses with specificity the con-siderations listed in paragraph (b)(4) of 10CFR Section 2.790 of the Com-mission's regulations.
The proprietary material for which withholding is being requested is of the same technical type as that proprietary material previously submitted with application for withholding CAW-83-80.
Accordingly, this letter authorizes the utilization of the accompanying affidavit by GPC.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-83-81, and should be addressed to the undersigned.
Very truly yours, I
Robert esemann, Manager Regulatory & Legislative Affairs
/bek E. C. Shomaker, Esq.
Office of the Executive Legal Director, NRC cc:
S 8310310083 831025 q
DR ADOCK 05000424
}1 PDR
CAW-83-80 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
ss COUNTY OF ALLEGHENY:
Before me, the undersigned authority, personally appeared John D. McAdoo, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Corporation (" Westinghouse") and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
h g
^ \\w %%W, -
D. McAdoo, Assiitant Manager Nuclear Safety Department l
l l
l Sworn to and subscribed before me this N T4 day of J.d..tt-i 1983.
3 j
I l
';.,,l[,;...f;T lotN,,P'f]3%U:Uc w, u.um
- ,l'
.;.19?C I
.' CAW-83-80 (1)
I am Assistant Manager, Nuclear Safety Department, in the Nuclear Techno-logy Division, of Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withhcid from public disclosure in connection with nuclear power plant licensing or rule-making proceedings, and am authorized to apply for its withholding on behalf of the Westinghouse Water Reactor Divisions.
(2)
I am making this Affidavit in conformance with the provisions of 10CFR Section 2.790.of the Commission's regulations and in conjunction with the Westinghouse application for withholding accompanying this Affidavit.
(3)
I have personal knowledge of the criteria and procedures utilized by Westinghouse Nuclear Energy Systems in designating information as a trade secret, privileged or as confidential commercial or financial information.
(4)
Fursuant to the provisions of paragraph (b)(4) of Section 2.790 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be with-held from public disclosure should be withheld.
(i)
The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
(ii)
The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westing-house has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, i
utilizes a system to determine when and whether to hole certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
5055Q: 1D/093083
.=. --
. CAW-83-80 Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
(a)
The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage j
over.other companies.
J (b)
It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the I
application of which data secures a competitive economic advan-tage, e.g., by optimization or improved marketability.
(c)
Its use by a competitor would reduce his expenditure of resour-ces or improve his competitive position in the design, manufac-ture, shipment, installation, assurance of quality, or licensing a similar product.
(d)
It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)
It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)
It contains patentable ideas, for which patent protection may be desirable.
5055Q:10/093083
. CAW-83-80 (g)
It is not the property of Westinghouse, but must be treated as proprietary by Westinghouse according to agreements with the owner.
There are sound policy reasons behind the Westinghouse system which include the following:
(a) The use of information by Westinghouse gives Westinghouse a competitive advantage over its competitors.
It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.
(b)
It is information which is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c)
Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
(d)
Each component of proprietary information pertinent to a parti-cular competitive advantage is potentially as valuable as the total competitive advantage.
If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competi-tive advantage.
(e)
Unrestricted disclosure would jeopardize the position of promi-nence of Westinghouse in the world market, and thereby give a market advantage to the competition in those countries.
5055Q:lD/093083
g CAW-83-80 (f)
The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and main-taining a competitive advantage.
(iii)
The information is being transmitted to the Commission in confidence and, under the provisions of 10CFR Section 2.790, it is to be received in confidence by the Commission.
(iv)
The information sought to be protected is not available in public sources to the best of our knowledge and belief.
(v)
The proprietary information sought to be withheld in this submittal is that which is appropriately marked in " Technical Bases for Eliminating large Primary Loop Pipe Ruptures as the Structura: Design Bases for the South Texas Project," dated September 1983, prepared by S. A. Swamy and J. J. McIr.erney.
The subject information could only be duplicated by competitors if they were to invest time and effort equivalent to that invested by Westinghouse provided they have the requisite talent and experience.
Public disclosure of this information is likely to cause substantial harm to the competitive position of Westinghouse because it would simplify design and evaluation tasks without requiring a commensurate l
investment of time and effort.
Further the deponent sayeth not.
l l
5055Q: 1D/093083
)
o MT-SME-3082 ENCLOSURE B TECHNICAL BASES FOR ELIMINATING LARGE PRIMARY LOOP PIPE RUPTURES AS THE STRUCTURAL DESIGN BASIS FOR' ALVIN W. V0GlLE UNITS 1 AND 2 Prepared by:
S. A. Swamy J. J. McInerney September,1983 4
h.
M APPROVED:
APPROVED:
J. M. Chirigos, Manage?
E. R." Johnson, Manager Structural Materials Structural and Seismic Engineering Development APPROYED:
M W. 5. Brown, Manager Mechanical and Systems i
Licensina i
l i
8310310086 831025 PDR ADOCK 05000424 A
0 0
OUTLINE
~
Section
-1. 0 INTRODUCTION 2.0 PIPE GEOMETRY AND LOADING i
3.0 FRACTURE MECHANICS EVALUATION 4.0 LEAK RATE PREDICTIONS 5.0 FATIGUE CRACK GROWTH ANALYSIS
6.0 CONCLUSION
S
7.0 REFERENCES
APPENDIX A 0886s:10 2
4
1.0 INTRODUCTION
)
1.1 Purpose The current structural design basis for the reactor coolant system (RCS) primary loop requires that pipe breaks be postulated as defined in the approved Westinghouse Topical Report WCAP 8082, Reference 1.
In addition, protective measures for the dynamic effects associated with RCS primary loop pipe breaks have been incorporated in the Alvin W. Vogtle plant design.
However. Westinghouse has demonstrated on a generic basis that RCS primary loop pipe breaks are highly unlikely and should not be included in the structural design basis of Westinghouse plants (see Reference 4). The purpose of this report is to demonstrate that the generic evaluations performed by Westinghouse are applicable to the Vogtle plant.
In order to demonstrate this applicability, Westinghouse has performed a comparison of the loads and geometry for the Vogtle plant with envelope parameters used in the generic
- analyses (Section 2.0); fracture mechanics evaluation (Section 3.0);
determination of leak rates from a through-wall crack (Section 4.0), fatigue crack growth evaluation (Section 5.0); and conclusjons (Section 6.0).
1.2 Scope This report applies to the Vogtle plant reactor coolant system primary loop piping.
It is intended to demonstrate that specific parameters for the Vogtle plant are enveloped by the generic analysis performed by Westinghouse in WCAP-9570 (Reference 5) and accepted by the NRC as noted in a letter from Harold Denton dated May 2,1983 (Reference 6).
I e
1.3 Objectives i
The conclusions of this report (Reference 5) support the elimination of RCS primary loop pipe breaks for the Vogtle plant.
In order to validate this conclusion the following objectives must be achieved.
OB86s:10 3
m
~
a.
Demonstrate that Vogtle plant parameters are enveloped by generic Westinghouse studies.
b.
Demonstrate that margin exists between the critical crack size and a postulated crack which yields a detectable leak rate.
c.- Demonstrate that there is sufficient margin between the leakage through a postulated crack and the leak detection capability of the Vogtle plant.
d.
Demonstrate that fatigue crack growth is negligible.
1.4 Background Information Westinghouse has performed considerable testing and analysis to demonstrate that RCS primary loop pipe breaks can be eliminated from the structural design basis of all Westinghouse plants. The concept of eliminating pipe breaks in j
the RCS primary loop was first presented to the NRC in 1978 in WCAP 9283 (Reference 7). This Topical Report employed a deterministic fracture mechanics evaluation and a probabilistic analysis to support the elimination of RCS primary loop pipe breaks.
l This approach was then used as a means of addressing Generic Issue A-2 and j
Asynnetric LOCA Loads. Westinghouse perfonned additional testing and analysis l
to justify the elimination of RCS primary loop pipe breaks. As a result of this effort, WeAP 9570 was submitted to the NRC. The NRC evaluated the technical merits of this concept and prepared a draft SER in late 1981 endorsing this concept. Additionally, both Harold Denton and the ACRS have endorsed the technical acceptability of the Westinghouse evaluations".
Specifically, in a May 2,1983 letter (Reference 6) Harold Denton states that
... it is technically satisfied with Westinghouse Topical Report 9570 Rev.
- 2....." Additionally, the ACRS stated in a June 14, 1983 letter (Reference
- 8) that "... there is no known mechanism in' PWR primary piping material for developing a large break without going through an extended period during which the crack would leak copiously."
e i
0886s:10 4
[
I
The NRC funded research through Lawrence Livermore Natforal Laboratory (LLNL) to address this same issue using a probabilistic approach. As part of the LLNL research effort, Westinghouse performed extensive evaluations of specific plant loads, material properties, transients, and system geometribs to demonstrate that the analysis and testing previously performed by Westinghouse and the research performed by LLNL applied to all Westinghouse plants including Vogtle (References 2 and 3). The results from the LLNL study were released at a March 28, 1983 ACRS Subcommittee meeting. These studies which are applicable to all Westinghouse plants east of the Rocky Mountains, determinet. the mean probability of a direct LOCA (RCS primary loop pipe break) to be 10-10 per reactor year and the mean probability of an indirect LOCA to be 10-7 per reactor year. Thus, the results previously obtained by Westinghouse (Reference 7) were confirmed hy an independent NRC research study.
The above studies establish the technical acceptability for eliminating pipe breaks from the. Westinghouse RCS primary loop. The LLNL study has been shown applicable to Yogtle plant by inclusion of plant specific data. This report will demonstrate the applicability of the Westinghouse generic evaluations to the Alvin W. Yogtle plant.
l l
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0886s:10 5
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e 2.0 PIPE GEOMETRY AND LOADING i
A segment of the primary coolant hot leg pipe is shown in Figure 1.
This segment is postulated to contain a circumferential through-wall flaw. The inside diameter and wall thickness of the pipe are 29.0 and 2.45 inches, respectively. The pipe is subjected to a normal operating pressure of[
]+
+a c.e psi. The design calculations indicate that the junction of the[
jls most highly stressed. At this location
+a,c.e tne axial load, F, and the total bending moment M, are[
] Tips (including
+a,c.e the axial force due to pressure) and[
jln-kips, respectively. The method
+a,c.e of obtaining these loads can be briefly summarized as follows:
The axial force F and transverse bending moments, M and 11, are chosen y
7 for each static load (pressure, deadweight and thermal) based on elastic-static analyses for each of these load cases. These pipe load components are combined algebraically to define the equivalent pipe static Based on elastic SSE response spectra analyses, loads F, Mys, and M s
23 amplified pipe seismic loads, F ' Myd' Mzd are obtained. The maximum d
pipe loads are obtained by combining the static and dynamic load components as follows:
F=
F
+
F 3
d 2
M=
M
+M y
7 where M
+
M My-ys yd M
M My=
zs zd The corresponding geometry and loads used in the reference report (Reference
- 5) are as follows:
inside diameter and wall thickness are 29.0 and 2.5 inches; axial load and bending moment are[ ~
]Tnchkips.
+a,c.e The outer fiber stress for Vogtle is[~
~]tsi, while for the reference report
+a,c,e it is[
]ksi. This demonstrates conservatism in the reference report which
+a,c,e makes-it more severe than the Vogtle project.
0886s:10 6
3.0 FRACTURE MECHANICS EVALUATION
~
3.1 Global Failure Mechanism i
Determination of the conditions which lead to failure in stainless steel must be done with plastic fracture methodology because of the large amount of deformation accompanying fracture. A conservative method for predicting the failure of ductile material is the[
]+7his methodology has been shown to be applicable to ductike p
+a.c.e through a large number of experiments, and will be used here to predict the critical flaw size in the ;-imary coolant piping. The failure criterion has 1+
ta c.e been obtained by requiring [
(Figure 2) when loads are applied. The detailed development is provided in Appendix A, for through-wall circumferential flaw in a pipe with internal pressure, axial force and imposed bending moments. The[
'-~]forsuch
+a,c.e a pipe is given by:
+a,c.e f
7 0886s:10
~
+a,c.e t
1
'N i;}
'N s
I F
g
}.
The analytical model described above accurately accounts for the piping interr.al pressure as well as imposed axial force as they affect [
]+In order to validate the model, analytical predictions were compared
+a,c.e-with the experimental results [9] as shown in Figure 3.
Good agreement was i.-
s found.
3.2 Local Failure Mechanism The local mechanism of failure is primarily dominated by the crack tip behavior in, terms of crack-tip blunting, initiation, extension and fit: ally crack instability. ' Depending on the material properties and geometry of the pipe, flaw size, shape and loading, the local failure mechanisms may or may not govern the ul timate failure.
\\
The stability will be assumed if the crack does not initiate at all.
It has from a been accepted that the initiation toughness, measured in terms of JIN 4'
/ J-integrahfresistance curve is a material parameter defining the crack O; f initia tion.
If, for a given load, the calculated J-integral value is shown to i
be less than J of the material, then the crack will not initiate.
If the IN
'initiationicriterion is not met, one can calculate the tearing modulus as p;
N defined by the following relation:
,(
' T T
,y E
app da 2
- f y
l s
+
\\
0886sv10 8
i s
p..
where T,pp - applied tearing modulus E = modulus of elasticity of =[
](flow stress)
+a,c.e a = crack length
]+
+a,c.e E
In summary, the local crack stability will be established by the two step criteria:
J<J IN if J > J T,pp < Tmat 3y 3.3 Results of Crack Stability Evaluation Figure 4 shows a plot of thc[
- ]as a function of through-
+a,c.e wall circumferential flaw length in the[
lof the main coolant piping.
+a,c.e This[
$3was calculated for Vogtle ' data of a pressurized pipe at[
+a,c.e
~
]with ASME Code
+a,c.e minimum [
, Nroperties. The maximum applied bending moment of) l
+a,c.e in-kips can be plotted on this figure, and used to determine the critical flaw length, which is shown to be approximately [
]+ inches. This is considerably
+a,c.e
+a,c.e larger than the [
l+ inch reference flaw used in Reference 5.
+a,c.e I
0886s:10 9
~
i
] Therefore, it can be concluded that a postulated [
'] Inch
+a,c.e through-wall flaw in the Vogtle loop piping will remain stable from both a local and global stability standpoint.
l l
l I
l 0886s:10 10 l
l
4.0 LEAK RATE PREDICTIONS Leak rate calculations were performed in Reference 5 using an initial through-wall crack [
]+The
+a,c.e computed leak rate was[
] based on the normal operating pressure of[.'
]*
+a,c.e psi. [
~ '*This computed leak rate [
]+ - +a,c.e
]
significantly exceeds the smallest detectable leak rate for the plant. The Vogtle plant has a RCS pressure boundary leak detection system which is consistent with the requirements of Regulatory Guide 1.45 and can detect leakage of 1 gpm in one hour. There is a factor of[
]between the calculated
+a,c.e leak rate and the Vogtle plant leak detection systems.
i 0886s:10 11 y
_..~,-.w, T
._w.,_,,-e
5.0 FATIGUE CRACK GROWTH ANALYSIS To determine the sensitivity of the primary coolant sys' tem to the presence of small cracks, a fatigue crack growth analysis was carried out for the[
]$egion of a typical system. This region was selected
+a,c.e because it is typically one of the highest stressed cross sections, and crack growth calculated here will be conservative for application to the entire primary coolant system.
A finite element stress analysis was carried out for the[
^
]8f a plant typical in geometry and operational characteristics to any
+2,c.e Westinghouse PWR System. [
l+All normal, upset +a,c.e and test conditions were considered, and circumferentially oriented surface flaws were postulated in the region, assuming the flaw was located in three different locations, as shown in Figure 5.
Specifically, these were:
-4 Cross Section A:
Cross Section B:
+a c.e Cross Section C:
Fatigue crack growth rate laws were used[-.
~ ~
~ ' ~ ~ ~ ~ ~ ~
' ~ ~
]?The law for stainless steel was
+a,c.e derived from Reference 10, with a very conservative correction for R ratio.
The ratio of minimum to maximum stress during a transient is:
0886s:10 12
4.48 h = (5.4 x 10-12) g 1nches/ cycle eff where k,ff = Kmax (1-R)0.5 R=K IK min max
+a,c.e The calculated fatigue crack growth for semi-elliptic surface flaws of circunferential orientation and various depths is summarized in Table 1, and shows that the crack growth is very sna11, regardless[
+a,c.e 3+
0886s:10 13
~
6.0 CONCLUSION
S This report has established the applicability of the g,eneric Westinghouse evaluations which justify the elimination of RCS primary loop pipe breaks for the Yogtle plant as follows:
a.
The loads, material properties, transients and geometry relative to the Vogtle RCS primary loop are enveloped by the parameters of WCAP 9570.
b.
The critical crack length at the worst location in the RCS primary loop is[
] This is significantly greater than the[
]*
+a, c,.:
inches stable crack used as a basis for calculating leak rates in WCAP 9570.
c.
The leakage through a[ ] crack in the RCS primary loop is[
]$ased
+a,c,c on WCAP 9570. The Vogtle plant has a RCS pressure boundary leak detection system which is consistent with the requirements of Regulatory Guide 1.45 and can detect leakage of 1 gpm in one hour.
Thus, there is a factor of[
]between the calculated leak rate and
+a, c,(
the Yogtle plant leak detection systems.
l d.
Fatigue crack growth was determined for postulated flaws and was found to be extremely small over plant life and, therefore, is considered insignificant.
Based on the above, it is concluded that RCS primary loop pipe breaks should not be considered in the structural design basis of the Alvin W. Vogtle plant.
0886s:10 14 l
7.0 REFERENCES
1.
WCAP 8082 P-A, " Pipe Breaks for the LOCA Analysis,of the Westinghouse Primary Coolant Loop," Class 2 January 1975.
2.
Letter from Westinghouse (E. P. Rahe) to NRC (W. V. Johnston) dated April 25, 1983.
3.
Letter fror.; Westinghouse (E. P. Rahe) to NRC (W. V. Johnston) dated July 11, 1983.
4.
Letter from hestinghouse (E. P. Rahe) to NRC (R. H. Vollmer) dated May l
11, 1983.
l 5.
WCAP 9570, Rev. 2, " Mechanistic Fracture Evaluation of Reactor Coolant l
Pipe Containing a Postulated Circumferential Through-Wall Crack," Class 3, June 1981.
6.
Letter from NRC (H. R. Denton) to AIF (M. Edelman) dated May 2,1983.
7.
WCAP 9283, "The Integrity of Primary Piping Systems of Westinghouse Nuclear Power Plants During Postulated Seismic Events," Class 2, March, 1978.
l 8.
Letter from ACRS (J. J. Ray) to NRC (W. J. Dircks) dated June 14, 1983.
9.
Kanninen, M.
F., et al, " Mechanical Fracture Predictions for Sensitized Stainless Steel Piping with Circumferential Cracks" EPRI NP-192, Septerrber 1976.
- 10. Bamford, W.
H., " Fatigue Crack Growth of Stainless Steel Piping in a Pressurized Water Reactor Environment" Trans. ASME Journal of Pressure Yessel Technology Vol.101, Feb.1979.
0886s:10 15
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+a,c.c 4
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APPENDIX A
.+
+a.c,:
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0886s:10 17 4
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TABLE 1
~
~
+
FATIGUE CRACK GROWTH AT (40 YEAPS)
+^*C
FINAL FLAW (IN)
P INITIAL FLAW (IN) 1-
+a,ca u
i 0.292 0.31097 0.30107 0.30698 0.300 0.31949 0.30953 0.31626 0.375~
0.39940 0.38948 0.40763 0.425 0.45271 0.4435 0.47421 w
e l
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,-,,,s.,
M k., i c
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t 0
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9 C P1 2
h 5
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M
=
e ip P
tna loo C
ro t
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r g
I i ss.
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F i
c, -
a
+
(
se t;,
1 4 t-;'
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!!l,iJl<j
<1!.
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4 +a,c,:
///////
j 2a Neutral Axis f
e
~ + a,c.e r
Figure 2 L Stress Distribution
.1 S
20
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Y a,C,e M
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FIGURE 3 COMPARISON ON ;
PREDICTIONS WITH EXPERIMENTAL RESULTS 21
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. +a,C,e PREDICTED [
]
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+a,c.e l
n
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)-
I
+
Flaw GE04TRY I
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+
i 1
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CRACK LENGTH, INCHES FIGURE 4 CRITICAL FLAW SIZE PREDICTION
}
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O G
e O
U.
N
+n O
S e
6O EC (J
naa 44-ten en C
E 44 4w
=
b m
5 23
s a,co E
Ju>M rn UE oc:9 E
z?4 ui m
(.
>Z EOeC M
u<
Eu STRESS INTENSITY FACTOR RANGE (c.1X (KSI-[IR.)
g Fig. 6 REFERENCE FATIGUE CRACK GROWTH CURVES FOR
{'
}+a,c.e 24
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2
ENCLOSURE C Design-Impact of Elimination of Postulated Double-Ended Pipe Breaks in-the RCS Primary-Loop STRUCTURES,' SYSTEMS, COMPONENTS ETC. TIL\\T ARE IMPORTANT TO ELIMINATION OF DOUBLE ENDED PIPE BREAKS IN THE RCS PRIMARY LOOP IMPACT RCS Pipe Whip Restraints-Deleted from design Jet Barriers Deleted from design Reactor Cavity / Primary Shield Wall No Change Steam Generator Subcompartment/
.No Change Secondary Shield Wall RCS Component Supports No Change
- Emergency Core Cooling Systems No Change Containment Systems No Change Reactor Coolant Pressure Boundary No Change Leakage Detection Systems (Regulatory Guide 1.45 commitment)
Environmental Qualification Program No Change Quality Assurance Program No Change FSAR Future revision to reflect the change
- Structures are sized based on the combination of SSE loads and preliminary loads associatri with pipe break effects.
Final calculations that do not include pipe break loads will show increased seismic margins.
1 Enclosure D Postulated RCS Primary Loop Pipe Breaks and Associated Pipe Whip Restraints Per Unit Postulated Break Associated Whip Locations Per Loop Restraint Per Loop
- 1.
Reactor vessel inlet nozzle 1.
Primary shield wall cold leg restraint (wagon wheel) 2.
Reactor vessel outlet nozzle 2.
Primary shield wall hot leg restraint (wagon wheel) 3.
Steam generator inlet nozzle 3.
Hot leg restraint 4.
50 elbow on the intrados 4.
(Same as 3)
(longitudinal slot) 5.
Steam generator outlet nozzle 5.
Crossover leg vertical run restraint Crossover leg steam generator-side elbow restraint 6.
Reactor coolant pump 6.
Crossover leg pump-inlet nozzle (pump suction) side elbow restraint 7.
Crossover leg closure weld 7.
Crossover leg steam generator-side elbow restraint Crossover leg pump-side elbow restraint 8.
Reactor coolant pump outlet 8.
None
- 6. Restraints per loop