ML19268D430
| ML19268D430 | |
| Person / Time | |
|---|---|
| Site: | Wolf Creek, Callaway, 05000000 |
| Issue date: | 11/17/1982 |
| From: | Wiesemann R WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML19268D428 | List: |
| References | |
| CAW-81-79, CAW-82-67, NUDOCS 8212070347 | |
| Download: ML19268D430 (10) | |
Text
O Westinghouse Electric Corporation Power Systems eaxass PittsburthPemsylvana 15230 November 17, 1982 CAW-82-67 Mr. Harold R. Denton, Director Docket Number 50-482 Nuclear Reactor Regulation 50-483 U. S. Nuclear Regulatory Commission Phillips Buildino 7920 Norfold Avenue Bethesda, Maryland 20014
SUBJECT:
Steam Generator Tube Plugging Margin Analysis for the Standardized Nuclear Power Plant System (SNUPPS) WCAP-10043, February,1982 (Proprietary)
REF.:
SNUPPS Application for Withholding, Patrick to Denton, November,1982
Dear Mr. Denton:
The proprietary material for which withholding is being requested by Standardized Nuclear Power Plant System (SNUPPS) is proprietary to Westinghouse and withholding is reouested pursuant to the provisions of paragraph (b) (1) of Section 2.790 of the Commission's regulations.
Withholding from public disclosure is requested with respect to the subject information which is further identified in the affidavit accompanying this application.
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-81-79 and was accompanied by an affidavit signed by the owner of the proprietary information, Westinghouse Electric Corporation.
A copy of affidavit CAW-87-79 submitted to j1stify the previous material is attached and is eaually applicable to this material.
Accordingly, this letter authorizes the use of the proprietary information and affidavit CAW-81-79 by SNUPPS in Wolfe Creek, Unit 1 and Calloway, Unit 1.
8212070347 821203 PDR ADOCK 05000482 E
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NS-RAW-82-430 CAW-82-67 Mr. H. R. Denton November 17, 1982 Correspondence with respect to this anplication for withholding or the accompanying affidavit should reference CAW-82-67 and be addressed to the undersigned.
Very truly yours, Robert A. Wiesemann, Manager Regulatory & Legislative Affairs
/pj Enclosure cc:
E. C. Shomaker, Esq.
Office of the Executive Legal Director, NRC k
4 CAW-81-79 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
ss COUNTY OF ALLEGHENY:
Before me, the undersigned authority, personally appeared Robert A. Wiesemann, 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 (I' 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:
!10^ 4 Robert A. Wiesemann, Manager Regulatory and Legislative Affairs Sworn to and subscribed before me this day of
- ? t,,..
,1 1981.
}, os y i.
Notary Pub'lic,
1
' CAW-81-79 (1)
I am Manager, Regulatory and Legislative Affairs, in the Nuclear Tecnnology Division, of Westinghouse Electric Corporation and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public dis-closure, 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 con-junction with the Westinghouse application for withholding ac-companying 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)
Pursuant 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 in-formation sought to be withheld 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.
1 CAW-81-79 (ii)
The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public.
Westinghouse. has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and wnether to hold 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.
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 com-petitive 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 consti-tutes a competitive economic advantage over other companies, g
(b)
It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
, CAW-81-79 (c)
Its use by a competitor wculd reduce his expenditure of resources or improve iiis competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.
Ili (d)
It reveals cost or price irjormation, production cap-acities, budget levels, or 'hommercial s trategies of Westinghouse, its customers or suppliers.
(e)
It reveals aspects of past, present, or future West-inghouse or customer funded development plans and pro-grams of potential commercial value to Westinghouse.
(f)
It contains patentable ideas, for which patent pro-tection may be desirable.
(g)
It is not the property of Westingnouse, 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 such information by Westinghouse gives Westinghouse a competitive advantage over its com-petitors.
it is, therefore, withheld from disclosure to protect the Westinghou-' competi tive position.
. _ _ _ CAW-81 -79 (b)
It is information wnich 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 particular competitive advantage is potentially as valuable as the total competitive advantage.
If competitors acquire components of proprietary infor-mation, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
(e)
Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition in those countries.
(f)
The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
i CAW-81-79 (iii)
The information is being transmitted to the Commission in
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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 I
public sources or available information has not been pre-viously employed in the same original manner or method 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 " Steam Generator Tube Plugging Margin Analysis" for the Virgil C.
Summer Nuclear Power Plant Unit No.1, WCAP-9912, Revi-sion 2 (Proprietary) being transmitted by South Carolina Electric and Gas Company letter Application for Withholding Proprietary Information from Public Disclosure, Nichols to Denton, November 1981.
The proprietary information as sub-mitted for South Carol'ina Electric and Gas Company, Vi'.gil C.
Summer Nuclear Station use is expected to be applicable in other licensee and applicant submittals in response to cer-tain NRC requirements for justification of the steam generator tube plugging margin.
This information is part of that which will enable Westing-house to:
(a)
Provide documentation of the analyses, method and test-ing for determining plugging margin.
i 4 CAW-81-79 (b)
Establish the minimum wall thickness in compliance with Regulatory Guide 1.121.
(c)
Establish the stress limits versus thinning of the remaining tube wall.
(d)
Establish the maximum allowable leakage in support of the leak-before-break criteria.
(e)
Assist the customer to obtain NRC approval.
Further this information has substantial commercial value as follows:
(a) Westinghouse plans to sell similar information to its customers for purposes of meeting NRC requirements for licensing documentation.
(b)
Westinghouse can sell support and defense of the tech-nology to its customers in the licensing process.
Public disclosure of this information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to pro-vide similar analytical documenta td in and licensing defense services for commercial power reac uars without commensurate expenses.
Also, public disclosure of the information would enable others to use the information to meet NRC require-ments for licensing documentation without purchasing the right to use the information.
~
,< CAW-81 -79 The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended for system design software development.
Further the deponent sayeth not.
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STEAM GENERATOR TUBE PLUGGING MARGIN ANALYSIS FOR THE WESTINGHOUSE STANDARDIZED NUCLEAR POWER PLANT SYSTEM (SNUPPS)
A. P. Villasor, Jr., Ph.D.
November, 1982 APPROVED-
- d. L. Houtman, Manager Applied Structural Mechanics Work Performed Under Shop Order No. YNGP-24401
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e WESTINGHOUSE ELECTRIC CORPORATION Nuclear Energy Systems P.O. Box 355 Pittsburgh, Pennsylvania 15230
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This report describes the analysis to determine the plugging margin for the Westinghouse Standardized Huclear Pcwer Plant System (SNUPPS) steam generator (!!odel F) tubing.
Based on the results, a minimum tube thickness requirement of the nominal wall is ad,e
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established in accordance with the guidelines of USNRC Regulatory Guide 1.121.
Assuming {
]allowancefor continued tube wall degradation, a plugging margin of 53%
of the nominal wall is recommanded.
With discrete wedgu uusd to support the TEPs, the
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F SNUPPS = Standardized Nuclear Power Plant System ASIG
= American Society of Mechanical Engineers E
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= Antivlb atlen bars EC
= Eddy-Current FDB
= Flow distribution baffle FIV
= Flow int'uced vibrations FLB
= (muin) Feedline t reak (accident)
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- Loss-of-Coolant Acc.ident (primary)
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I Nak clad temperature PCT
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United States Nuclear Regulatory Commission USNRC
=
Westinghouse Electric Corporation W
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-vi.
TABLE OF CONTENTS Section Title Page 1
INTRODUCTION 1
1.1 Regulatory Requirements for Tube Plugging 1
1.2 Scope of the SNUPPS Plugging Margin Analysis 2
2 INTEGRITY REQUIREMENTS AND CRITERIA 7
2.1 Functional and Safety Requirements 7
2.2 Tube Bundle Integrity Requirements 8
2.3 Locally-Degraded Tube Integrity Requirements 9
2.4 Tube Stress Classification 10 2.5 Criteria and Stress Limits 15 3
LOADS AND ASSOCI'T'.D ANI. LYSES 19 3.1 Normal Operating Loads 19 3.1.1 Upset Load 20 3.2 Accident Condition Loads 20 3.2.1 LOCA Loads 21 3.2.1.1 LOCA Rarefaction Wave Anslysis 22 3.2.1.2 Rarefaction Wave Induced Tube Loads 27
- 3. 2.1. 3 Rutfaction Vave Induced TSP Loads 32 3.?.1.4 LOCA Shaking Loads 32 3.2.2 FLB/SLB Loa 6 40 3.2.3 SSE Loads 43 3.2.3.1 Seismic Model 43 3.2.3.2 Seismic Analysis Output 47 4
RESULTS OF ANALYSES AND EVALUATION 51 4.1 Functional Integrity Evaluation 52 4.1.1 Level D Service Condition Stresses 52 4.1.2 Primary Flow Area Reduction 55
-vii-
TABLE OF CONTENTS (CONTINUED)
Section Title Pace 4.2 Minimum Wall Requirements for Degraded Tubes 59 4.2.1 Normal Plant Conditions 60 4.2.2 FLB/SLB + SSE 60 4.2.3 LOCA + SSE 61 5
BURST STRNGTH REQUIREMENTS 65 5.1 Leak-Before-Break Verification 69 5.2 Margin to Burst Under Normal APj 75
- 5. 2.1 Effect of Bending on Burst Strength of Tube 76 5.2.2 Tube with a Thtw-Wall Degradation 77 5.2.3 Thinned Tube 78 6
PLUGGING MARGIN RECOMMENDATION 79 7
APPENDIX 81 7.1 Deviation of Lower Bound Tolerance 81 Limits for Strength Properties
-vi i i -
LIST OF ILLUSTRATIONS (CONTINUED)
Figure Title Pace 4-1 Typical Wedge Group Arrangement for Tube Support 53 Plate 4-2 Schematic of a Tube-Tube Support Plate Crush Test 54 4-3 Correlation Between Tube Ovality and Collapse 62 Pressure 5-1 Plot of a Typical Leakrate *.est (SGTLR #30, t=0.379 67 Inch) 5-2 Correlation Betveen Axia'. rack Length versu Leakrete, 68 for Model F Tubing Under.',ormal Operating AP 5-3 Relationship Between Normalized Burst Pressure and 71 Axial Crack Length of SG Tibing 5-4 Minimum Expected Burst Strength of Model F Inconal 600 72 Thermally-Treated Tubing 5-5 Variation in Margin to Burst as a Function of R,/t for 74 Thermally-Treated 0.688" OD x 0.040" t Tubing
-X=
LIST OF ILLUSTRATIONS Figure Tit 1e Pace 1-1 Cutaway View of a Model F Steam Generator 3
'2 Schematic of a Model F Steam Generator Tube Bundle 4
i-Geometry 3-1 Tuoe Model for LOCA Rarefaction Wave Analysis 23 3-2 Differential Pmssure Time-Histories at Various Nodes E6 Following a LOCA 3-3 LOCA Rarefaction Wave Tube Horizontal Displacement (UX) 30 vs Time for flode 6 to Node 9. Inclusive 3-4 LOCA Rcrefaction Wave Tube Bending Homent (MZ) vs Time 31 in Element 2 to Element 5, Inclusive 3-5 Reactor Coolant Loop Model for LOCA Analysis 33 3-6 Steam Generator Displacements Due to a Steam Generator 34 Outlet Nozzle Break 3-7 Model of the Tube Bundle for LOCA Shaking Analysis 36 with Nede Numbering 3-8 Model of the *ube Bundle for LOCA Shaking Analysis with 37 Element Numbering 3-9 SNUPPS SSE Response Spectra 42 3-10 Seismic Model of the SNUPPS Steam Generator with 44 Node Numbering 3-11 Seismic Model of the SNUPPS Steam Generator with Element 45 Numbering 3-12 Seismic Model of the U-bend Showing Elemnt Numbering 46
-ix-
LIST OF TABLES Title Paoe 2-1 Tube Stress Classification 12 2-2 SNUPPS Tube Strength Properties for R.G.1.121 14 Analyses (0.688"0D x 0.040"t) 3-1 LOCA Rarefaction Tube Bending Stresses 28 3-2 LOCA Rarefaction Tube Rotations at Top TSP 29 3-3 LOCA Shaking Tube Stresses 38 3-4 LOCA Shaking Tube Rotations at Top TSP 39 3-5 SSE Tube Bending Stresses 48 3-6 Maximum Tube Support Loads Due to SSE 49 4-1 Summary of Maximum Tube Support Plate Wedge Loads 56 5-1 Summary of Leakrates of Axially-Cracked Model F Tubing 66 o ' b' c Under Normal Operating AP [
]
4 5-2 Burst Pressure Test Data on Axially-Slotted Model F 70 Tubing at Room Temperature 7-1 Lower Tolerance Limits of Strength Properties for 83 the SNUPPS Tube
-xi-
SECTIOM 1 INTRODUCTIO11 1.1 Reculatory Requirements for the Plueging The heat transfer area of steam generators in a PWR nuclear steam supply system (NSSS) comprises over 50%
of the total primary system pressure boundary.
The steam generator tubing therefore represents a major barrier against the release of radioactivity to the environment.
For this reason, conservative design criteria have been established for structural integrity of the tubing under the postulated design-basis accident condition loadings in accordance wita Section III of the AS!!E Boiler and Pressure Vessel Code (hereinafter designated as the Code).
Over a period of time under the influence of the operating loads and environment in the steam generator, some tubes may become degraded in local areas.
To determine the condition of the tubing, inservice inspection using eddy-current (EC) techniques is performed in accordance with the guidelines of USt1RC Regulatory Guide 1.83.
Partially-degraded tubes with wall thicknesses greater than the minimum acceptable tube wall thickness are satisfactory for continued service.
Also, the minimum required tube wall thickness is 1
adjusted to take ca e of possible discrepancies in the EC probe and to annular an operational allowance for continued tube degradation until the next scheduled inspection.
The USURC Regulatory Guide 1.121 describes an acceptable method for estcblishing the limits of tube degradation beyond which tubes will be repaired or r2 moved from service.
The amount of degradation as recorded by the EC testing is customarily expressed as a percentage of the design nominal tube wall thicknsss, and the ecceptable degradation is referred to as the tube plugging margin.
1.2 Scoon of the SMUPPE Plugging Margin Analysis This report describes the results of analysis performed for the Westinghouse Standardized Nuclear Power Plant System (SNUPPS) steam generator tubing in order to establish the tube plugging margin.
Each SHUPPS unit has a 4-loop HSSS wnich includes the Model F steam generator.
A cutaway view of a Modcl F steam generator is shown in Figure 1-1.
Figure 1-2 shows a schematic drawing of the tube bundle which consists of 3626 U-tubes made of Inconel-600 (SB-163) alloy.
Some of the earlier SNUPPS units have both the mill-annealed and thermally-treated tubing.
Lateral support for the tube is provided by the seven (7) tube support plates (TSP) approximately 40 inches apart in the straight region of the bundle.
In the U-bend area, the out-of-plane motion of tube bends is limited by coupling the U-bends with three sets of 2
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FIGURE l-1: CUTAWAY VIEW OF A MODEL F STEAM GENERATOR -
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The nominal tube is 0.688" OD x 0.040" t.
The minimum tube wall requirements were calculated in accordance with the criteria of USNRC Regulatory Guide 1.121, entitled " Bases for Plugging Degraded PWR Steam Generator Tubes".
The basic requirements consist of:
1)
In the case of tube thinning, stresses in the remaining tube wall are to meet applicable stress limits during normal and postulated accident condition loadings, and 2)
In the case of tube cracking, with or without any thinning, the maximum allowable leakage during normal operation is to be limited consistent with lenk-before-break criteria.
Additional requirements consist of verifying the margin to burst under normal operation and margin against collapse during a LOCA.
The question of fatigue failure under cyclic bending stresses is covered in the validation of leak-before-break.
In connection with the tube bundle integrity evaluation, it should be noted that both the safety and functional requirements are to be satisfied.
The safety requirement which is the basis of the Regulatory Guide 1.121 criteria governs the limiting safe condition of localized tube degradation, as established by inservice inspection, beyond which tubes should be repaired or removed from service.
In contrast, the functional requirement applies to the overall degradation of the tube bundle in terms of its heat removing capability and the impact on the peak clad temperature due to the primary coolant flow restriction through the tube bundle following a LOCA, which is evaluated in conjunction with SSE.
Although both the safety and functione.1 requirements were found satisfied, the subject matter of this report deals mainly with the safety requirements associated with the plugging margin criteria in Regulatory Guide 1.121.
Specific criteria and the corresponding allowable limits and/or margins associated with the safety and functional requirements are discussed in Section 2.
Details of tube loadings during the various plant conditions are discussed in Sections 3 and 4 with the related analytical results and evaluations.
Section 5 contains the discussion of leak-before-break verification and burst strength requirements.
Finally, the recommended tube plugging margin is set forth in Section 6.
SEcTion 2 INTEGRITY REOUIREMENTS AND CRITEPIA The steam generator tubing represents an integral part of the primary system.
In the event of a primary loss-of-coolant (LOCA), the tubing provides the necessary heat sink, initially for the core cooldown and later for maintaining the plant in the safe shutdown condition.
Thus, it is important to establish the structural integrity of the steam generator tubing so that the tube bundle can sustain the loads during normal operation and the various postulated accident conditions without a loss of function of safety.
2.1 Functional and Safety Pecuirements Tube walls may be affected by a number of different factors such as environment-induced corrosion (including intergranular attack and stress-corrosion cracking), erosion due to the fluid friction, and fretting wear from mechanical and flow-induced vibrations.
The wall loss due to general erosion or corrosion has been conservatively established and is assumed to be more or less uniform for the entire tube bundle during the plant operating period.
However, a potential for additional wall degradation may exist locally in some tubes, near the top of the tubesheet and in the region of tube-tube support plates (TSP) intersections, because of a higher potential for chemical concentrations and/or relative motion in these regions.
Based on steam generator operational history, the whole bundle may be subjected to only a small, but probably a more or less uniform, tube wall loss over the total operating period of the unit.
On the other hand, some tubes of the bundle may degrade locally to the extent that either the removal of these tubes from service or local repair to restore integrity is sufficient for continued safe operation of the unit.
Because of these two distinct modes of tube degradation, it is possible to separate the functional and safety requirements into those af fecting the integrity of (1) the overall tube bundle, and (2) a locally-thinned or degraded tube.
Tube associated with these modes of degradation are referred to as the " median" and the
" locally-degraded" tube.
The median tubing corresponds to the minimum expected strength properties of the overall tube bundle and represents a tube with the end-of-design life minimum wall, which may be the drawing minimum less the design basis erosion / corrosion allowance.
The end-of-design life conditions assume a general corrosion on the
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2.2 Tube Bundle Integrity Requirements These requirements are based on the assumption that removal of a small number of tubes from service does not impair the structural and functional capability of the overall tube bundle. In the event of extensive tube plugging, plant derating and/or reanalyses associated with functional requirement verification may be necessary.
However, reLnalyses for the verification of structural integrity of the tube bundle as a whole will not be required since almost all of the deactivated tubes would physically remain in the tube bundle, thus maintaining the structural characteristics of the ?,ube bundle practically intact.
Specifically, the following two criteria are to be satisfied, assuming the median tube properties:
1)
For Level D Service Conditions, the primary stresses do not exceed the stress limits specified in Appendix F of Section III of the Code.
2)
The loss of tube bundle flow area due to the combination of the cross-sectional distortion and/or collapse of a 1imited number of tubes a' C loads does not due to the postulated increase the primary flow resistance of the
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system 2.3 Locally-Degraded Tube integrity nequirements As previously indicated, the potential for tube wall degradation other than due to nominal erosion-corrosion may exist at certain locations in the tube bundle.
Even though such localized degradation is known to be confined over a small 9
portion of the tubing (and hence of no adverse consequence to the functional capability of the bundle), it is to be assessed from the viewpoint of a potential tube rupture, if the associated depth of penetration is relatively large.
Therefore, to show that there are no safety consequences as a result of random tube bursts, a conservative bound on acceptable degradation for continued operation must be established along with the in-service inspection and leakage monitoring requirements for the detection of degraded tubes.
Guidelines in Regulatory Guide 1.83 for EC inspection and Regulatory Guide 1.121 for tube plugging margin calculations provide the bases for determining the limiting safe condition of a locally-degraded tube.
For tube degradation in excess of the established plugging margin, it is required that the tube be repaired or removed from service (by plugging or otherwise) in order to provide continued safe operation.
The intent of Regulatory Guide 1.121, as applicable to this analysis, is summarized below:
In the case of tube thinning due to the mechanical and chemical wastage, and generalized intergranular attack, stresses in the remaining tube wall are shown to be capable of meetint the applicable requirements with adequate allowance for the EC measurement uncertainties and assumed continued erosion-corrosion until the next scheduled outage.
The strength requirements are specified in terms of allowable primary stress limits and margins against burst during normal operation and collapse following a LOCA.
Por tube cracking due to fatigue and/or stress corrosion, a specification on maximum allow.2ble leak rate during normal operation must be established such that the associated crack will not lead to a tube rupture during a postulated worst case accident condition pressure loading.
If the leak rate exceeds the specification, the plant must be shutdown and corrective actions taken to restore integrity of the unit.
2.4 Tube Stress Classification For plants in seismic regions, the most limiting loads for establishing the tube integrity are imposed 4,c during the Level D service conditions; There are two general considerations which must be accounted for in determining the classification of stresses; namely, the location in the structure and the nature of the loading.
- a, c,
~
]The
_ tube stress classification for various locations in the tube bundle under the different types of loadings TABLE 2-1: TUBE STRESS CLASSIFICATION
- a. b, (1).MedianTube (2) Thinned Tube is summarized in Table 2-1.
The notation P refers m
to general primary membrane stress, P refers to b
primary bending stress and 0 refers to secondary stress.
At the top TSP, a distinction is made between bending stresses in median tubes and locally-thinned tubes.
In the U-bend region the anti-vibration bars couple the tubes for motion out of the plane of the U-bend so that out-of-plane bending is resisted by the entire bundle.
c, c A distinction is made between self-excited, flow-induced vibration (FIV) stresses and flow-induced vibration from other causes.
A self-excited vibration mechanism could be established if flow velocities exceed criteria values for fluidelastic vibration.
When the vibration amplitude increases, however, the amount of damping in the vibrating tube also increases.
The vibration amplitude of cyclic bending stresses are limited by ac the amount of damping in the system.
TABLE 2-2:
SNUPPS TUBE STRENGTH PROPERTIES FOR R.G.1.121 ANALYSES (0.688" OD x 0.040" t) 3, uma T
I i
.m..
G, C 2.5 critaria and stronn f.imitn A,C A summary of these calculations is given in a*c Table 2-2.
l W
- 4,b,*
l'oreel and Upret Plant Conditions The primary-to-secondary pressure differential P
should not produce a primary membrane stress g
in excess of the yielo stress of the tube material at operating temperature; that is, a b.
P 1S m
y Portulated Accident conditione Loadings associated with a primary (LOCA) or a secondary side (SLD/FLD) bicwdoun, concurrent with the SSE, should be accommodated uith the margin determined by the stress limits specified for Level D Service Conditions in
- o,c Appendix F of the Code {
Por Locolly-Tbinned Tubirg o., b, c P
1 smaller of (2.4 S 0.7 S )
=
g, P
+P 1
b A b,?
Por the fledian Tubing P
m_
P
+P b"_
~
m Since the tube has regions of plastic deformations, the chake factor K is introduced in determining the allow.ible stress.
This constant is a function of the cross-sectional o, b,e
~
dimensions of the tube.
As far as the consideration of the secondary and peak stresses in the evaluation of a locally-thinned tube is concerned, it is noted that the effects of these stresses will be manifested into racheting, fatigue and/or corrosion-fatigue types of mechanisms associated with tube cracking if that should A,6 occur.
L.
SECTIOM 3 LOADE AMD ASSOCIATED AMALYSEE In establishing the safe limiting condition of a tube in terms of its remaining wall thickness, the effects of loadings during both the normal operation and the a,c postulated accident conditions must be evaluated.
3.1 Morral Operati ng Leads The limiting stresses during normal and upset operating conditions are the primary membrane stresses due to the primary-to-secondary pressure differential AP across the tube wall.
During normal g
operation at 100% full power, the pressures are as follows:
Prirary Eidet Reactor coolant pressure, Pg = 2250 psia Secondary Sidet Steam pressure, P
= 1000 psia o
at 100% power is thus The pressure differential APg 1250 psi.
3.1.1 Upnet Load However, the maximum operating conditicn APg occurs during a loss-of-load transient when:
Primary side pressure, Pg = 2650 psia Secondary side pressure, P
= 975 psia g
Hence, A Pg=Pg - P,= 1675 psi.
3.2 Accident condition Loads For the faulted plant condition evaluation, the postulated Level D Service Condition events are:
Loss-of-Coolant Accident (LOCA), main Steam Line Dreak (SLB), main Feed Line Break (FLB) and Safe Shutdown Earthquake (SSE).
The tube integrity evaluation is performed for the blowdown loads in
~
conjunction with the SSE loads;f a,e u
The tube loadings were maximized by v
s assuming these events to initiate when the plant is 3
operating at 100% full power condition.
3.2.1 LocA Loadn LOCA loads are developed as a result ot' transient flow and pressure fluctuations following a postulated D,
a,c
~
main Coolant pipe break.
(
I b
I 1
I.
N
.\\ ' '
,i
\\
\\
)
s
\\
-s s
q 2,*
.r r-y I
s s
s i
i 3
i
(
\\
l "l
I.'
o' t
\\
. l I
/
l
/
.)
I h
e
-I
. l t
J
\\
i j
3.2.1.1 1GCA h-dacid pn Have Analysen i
The princip>l tube loading during a LOOA is ca*laed by gs the raref achon wave in -the primary fluid.
E e
4 i
1 5
)
\\
'b a, b, s I
l FIGURE 3-1: TUBE MODEL FOR LOCAL RAREFACTION WAVE ANALYSIS
. a,e Figure 3-1 shows the node and element
~
numbering for a typical single tube model which was analyzed using the NECAN program.
o,e L
a, c,
_]
A,3 a, b, a i
l The pressure time histories to be input in the structural analyses were obtained from transient thermal-hydraulic (T/H) analyses using the MULTIPLEX - a, b.a Code.
a A, b, e.
FIGURE 3-2: DI:FERENTIAL PRESSURE TIME-HISTORIES AT VARIOUS PODES FOLLOWING A LOCA
- aic 4
In addition to the pressure bending loads, the rarefaction wave analysis includes the presture membrane stresses du:e to the primary-to-secondary AP and the effect of fluid friction and centrifugal g
forces.
3.2.1.2 Rarefaction wave induced Tube toads ihe maximum tube bending stresses and rotations at t
the top TSP are summarized in Tables 3-1 an6 3-2, respectively, for the various cases analyzed.
Figures 3-3 and 3-4 show the time-history variations of the in-plane horizontal displacements and bending moments, respectively, at selected nodes of the largest bend radius tube.
Comparison of these results lead to the following two major inferences.
2, L.
TABLE 3-1:
LOCA RAREFACTION TUBE BENDING STRESSES
, a, b,,,
b Due to pinned bordary assumption, no bending stmsses result at this'1ocation.
TABLE 3-2: LOCA RAREFACTION TUBE ROTATIONS AT TOP TSP a b.o e
%T %
a, b, c, I
l I
FIGURE 3-3:
LOCA RAREFACTION WAVE TUBE HORIZONTAL DISPLACEMENT (UX) VS TIME FOR NODE 6 TO N0DE 9. INCLUSIVE.
a, b, c, i
1 4
a l
FIGURE 3-4:
LOCA RAREFACTION WAVE TUBE BENDING M0 MENT (MZ) VS TIME IN ELEMENT 2 TO ELEMENT 5, INCLUSIVE GA J
3.2.1.3 Rarefaction Wave Induced TSP Loads The tube motion due to the LOCA rarefaction wave induced loading is restrained at the TSP locations, a,c resulting in reaction forces in the plates.
A, b,c 3.2.1.4 LOCA shaking Loadn Concurren' with the rarefaction wave loading during a LOCA, the tube bundle is subjected to additional bending loads due to the shaking of the stecm generator caused by the break hydraulics and reactor o,b,e l
FIGURI 3-5: REACTOR COOLANT LOOP MODEl FOR LOCA ANALYSIS
- 3 3-
a,',s c
I i
i 1
FIGURE 3-6: STEAM GENERATOR DISPLACEMENTS DUE TO A STEAM GENERATOR OUTLET N0ZZLE BREAK a, c coolant loop motion.
F l
9 5
I i
I i
e i
i l
1 i
t i
_]
g eW A,bic FIGURE 3-7: MODEL OF THE TUBE BUNDLE FOR LOCA SHAKING ANALYSIS WITH N0DE NUMBERING.
o, b,:
7 I
l FIGURE 3-8: MODEL OF THE TUBE BUNDLE FOR LOCA SHAKING ANALYSIS WITH ELEMENT NUMBERING
- 3 7-
TABLE 3-3:
LOCA SHAKING TUBE STPISSES a, b, c
- Due to pinned boundary asswtion, no bending stresses result at this Ircation.
. ~.. - - - -. - - -. -,. ~. _.....
r'
.1-.
?-
- %h s
-v,
.F-TABLE 3-4: LOCA SHAKING TUBE ROTATIONS AT TOP TSP R
% b d, i
d,o I
2 1
k p
W'
- ?
)
- i N&
a
. if e
"5
.u 8
p,
- =:.,,,,.a.u~g.=c
,---....n.n.
,e,n;~.,.
ya.
...,. ~ - -,
6 w
tir
[
The WECAN model with the node and element numbering used for the LOCA shaking analysis of the tube bundle is shown in Figure 3-7 and in Figure 3-8.
The maximum bending stresses in the tube U-bends (both the nominal and median geometries) and the maximum tube rotations at the top TSP are summarized
~
in Tables 3-3 and 3-4, respectively.
3.2.2 FLB/SLB Loads Dr; g the postulated FLB/SLB accidents, the rredominant primary tube stresses result from the P
loading.
The peak differential pressures for 1
these events were obtained from the results of
~ A'h""-
transientblowdownanalyses.f_
r These secondary side blowdown transients are based on an instantaneous full double-ended rupture of the d,?
main feedline/steamline.
M a, b, c 2
In addition to the primary pressure stresses, axial bending stresses in the tubes are developed as a result of flow-induced vibrations and tube-baffle o,
interaction.
~
I J
19,M710
.t,'e, '-
f I
l I
i i
i
?
I I
4
}
a,bic FIGURE 3-9: GNUPPS SSE RESPONSE SPECTRA
_o, 3.2.3 SSE Loads Seismic (SSE) loads are developed in the steam generator as a result of the motion of the ground
~
- 'b'O duringanearthquake.[
Because of the
~
~ differences in the SNUPPS peripheral support designs for the tube support plates (TSP), two separate analyses were performed: desigitated Plant 1 Site and Plant 2 Site.
The response spectra used in these analyses are shown in Figure 3-9.
3.2.3.1 Sei nnie flodel The analyses were performed using the WECAN computer _
code.
m 19,M711 a, b, c.
i t
t l
FIGURE 3-10: SEISMIC MODEL OF THE SNUPPS STEAM GENERATOR WITH N0 NUMBERING 19,647 12 a, b,.:.
~
e d
i i
l i
i i
I
~
FIGURE 3-11: SEISMIC MODEL OF THE SNUPPS STEAM GENERATOR WITH ELEMENT NUMBERING 4, 6, c FIGURE 3-12: SEISMIC MODEL OF THE U-BEND SHOWING ELEMENT NUMBERING a,c The node and element
~
~ numbering details of the model are shown in Figures 3.10 and 3.11, respectively.
- a,b,a i
i i
i j
Details of the element numbering of the
~
~ mathematical model of the U-bend region are shown in Figure 3.12.
The e. ode numbering is the same as was shown in Figure 3-7.
3.2.3.2 is:snic Analysis output In addition to the displacements, velocity and acceleration of each node point, the seismic solution provides the stresses in each element as well as a, o' e support wedge reaction loads on the TSP's.
F TABLE 3-5: SSE TUBE BENDING STRESSES o. b, l
=*
TABLE 3-6: MAXIMUM TOBE SUPPORT LOADS DUE TO SSE
- d.h.c
~
i O --
._ a,c The analysis output pertinent to the subject evaluation consists of the tube bundle stresses and the in-plane TSP loads.
The maximum (axial) stresses in both the nominal and median tube, and the TSP loads are summarized in Tables 3-5 and 3-6.
- 4,c respectively.
% a SPCTIO11 4 Rest 1LTE OF At1ALYSES AMD EVALUATIO11 Loads and stresses generated from the analyses described in the previous section were used to verify the following requirements:
(1)
Functional requirements associated with the overall tube bundle integrity during and following the Level D Service condition loadings, that is:
a,c (2;
Safety requirements on a locally-degraded tube; viz.,
- a., t.,
4.1 Punctional Integrity Evaluation a, c The evaluation consisted of verifying that the tube primary stresses and the reduction in the primary flow area of the tube bundle under the limiting faulted loads were within the specified acceptance limits.
4.1.1 Level D Service Condition Stresses
_. c,c
~
~
This
~ loading condition is most limiting for the case of locally-degraded (thinned) tubing and is considered later in the determination of the minimum required thickness.
a,c a,e Resultsofthe[
analyses discussed in the previous section were used to compute the maximum stress intensity in the tube U-bends.
a, c
~
19,947 17 h
DETAIL 8 7
r O-
-O-DETAIL A gSUPPORT P.,
-Q c==s c===h [
c=== c===
WR APPE R %,
4$-
4>-
H>-
SHELL 1
I t
b h
M I
I 7
r jfjf 7
..N...
i
......,----..._x_
y, h$
DETAll A DETAIL B FIGURE 4-1: TYPICAL WEDGE GROUP ARRANGEMENT FOR TUBE SUPPORT PLATE _
1s,e47.t s
- 4. b, c A - PLATE SOLID RIM B - PLATE BROACHED PERFORATED REGION C - PLATE FIXTURE (OUT-OF PLANE RESTRAINT)
D - DIAL INDICATOR CAGE E - 12" WEDGING.
FIGURE 4-2: SCHEMATIC OF A TUBE-TUBE SUPPORT PLATE CRUSH TEST
- a,bi; 4.1.2 P r i ra r'_' Flew Area P eclu c t i c n The in-plane T5P loads due to LOCA and SSE are transmitted to the chell through the supports of the a' b ' o tube support plates.
a, b, :.
- Originally, there were 4 plant orders for S110P PS.
Only the earlier tuo, Callaway I!o. 1 and Uolf Creek, are being built.
The other tuo were cancelled. -
TABLE 4-1:
SUMMARY
OF MAXIMUM TUBE SUPPORT PLATE WEDGE LOADS
,0,b,c D
i i
a,:.
a, b, c Table 4-1 sumnarizes the individuc1 wedge loads along a'**
with the contact loads.
- a,C
.J A,C
- G,Ie. C 1
4, h, C l
s i
I i
i W
- a,c Thus, the functional requirements
" cre met by the S11U5PS !!odel F stecn generators.
4.2 f t i n i no, tin 11 Pecuirenents for Decrnantion Tubos
- G h, ?
~
e i.
- G.'o,:
4.2.1 l'b rn al Plant conditions a, b. :
4.2.2 FT.n / SLB + S S E a,b,c a, b,3 4.2.3 rocA+gse 2,b.:
The collapse pressure is significantly affected by tube ovality.
A number cf correlations using limit analysis theory have been developed to predict collapse strength of ovalized tubes.
A correlation uas found to be quite accurate for the thermally-treated (or stress-relieved) tubing, believed to be due to its less anisotropic yield properties compared to thct of as-manufactured tubing.
The validity and conservatism of this ts.so te b
a.i, a CORRELATION BETWEEN TUBE OVALITY AND C FIGURE 4-3:
analytical correlation was verified against the results of room temperature collapse pressure tests on mill-annealed 0.75 in. OD :: 0.043 in t,
and 0.875 in. OD :: 0.050 in. t oval tubes.
Figure 4-3 shows the comparison of analytically predicted (normalized) collapse pressures with those obtained frcm the tests.
a, 'o, a M
SECTIOff 5 BURST STRE!1GTH REOU IP Ef1E11TS In addition to the limite on allowable stresses and margin to collapse due to external pressure discussed previously, the following requirements on the burst (pressure) strength capability of the degraded tubing is also to be shown as satisfied:
A,b,c M
M TABLE 5-1:
SUMMARY
OF LEAKRATES OF AXIALLY-CRACKED MODEL F TUBING UNDERNORMALOPERATINGAP}
}
c, b, e.
4 mu e
9 6
w.mi.=
a,b,c
~
FIGURE 5-1:
PLOTOFATYPICALLEAKRATETEST(SGTLR#30, SSA4M1 a,beC FIGURE 5-2: CORRELATION BETWEEN AXI AL CRACK LENGTH VERSUS LEAKRATE. a.,b c FOR MODEL F TUBING UNDER NORMAL OPERATING AP 4_
- A b,?
5.1 Lenk-Defere-Break "erification The rationale behind this requirement is to limit the maximum allowable (primary-to-secondary) leak rate during normal operation such that the associated crack length (through which the leakage occurs) is les: than the critical crack length corresponding to the maximum postulated accident condition pressure loading.
Thus, on the basis of leakage monitoring during normal operation, it is assumed that an unstable crack growth leading to tube burst would not occur in the unlikely event of the limiting accident.
For the SMUPPS units, the maximum technical allowable leakrate is 0.35 gpm per st:eam generator.
Results of four leakrate (0) tests in Table 5-1 were used to determine the maximum allowable crack length (L) through the nominal wall during normal operation corresponding to this specified limit, conservatively assuming that the entire leakage is associated with a _,
single crack.
Beyond this crack length, the leakage vould exceed the TABLE 5-2: BURST PRESSURE TEST DATA ON AXIALLY-SLOTTED H0 DEL F TUBING AT ROOM TEMPERATURE 0,b,c M
wwm a, b, c 1
FIGURE 5-3: RELATIONSHIP BETWEEN NORMALIZED BURST PRESSURE AND CRACK LENGTH OF SG TUBING sew se a, b, C.
FIGURE 5-4: MINIMUM EXPECTED BURST STRENGTH OF MODEL F INCONEL 600 THERMALLY-TREATED TUBING technical specification limit, requiring a plant shutdown for a corrective action.
a, b, e
~
~
The results are plotted in Figure 5-3.
Since all previous tests were on mill-annee. led material, the results in Table 5-2 of testing on thermally-treated tubing was included in Figure 5-3 to verify that the lower bound (shown by the solid line) established by the broad deta base is applicable to the evaluation of thermally-treated S!10PPS tubing.
a, b, o Aonlicability to Thinned Tubing The applicability of leak-before-break is also to be verified for the case of a tube with crackinc
_ A,c
_ superimposed on thinning.
W
.p.s is.ww a, b,:
~
1 FIGURE 5-5: VARIATION IN MARGIN TO BURST AS A FUNCTION OF (/t FOR THERMALLY-T"IATED 0.688"0D x 0.040"t TUBING d, d, 0, b, 0
~
t 5.2 Margin to Eurst Un6er Ferral P g According to the Regulatory Guide 1.121 guidelines, a factor of safety (PS) cf 3 is required against bursting under the normal operating pressure a, b, c.
differential.
M d, b.C A,C
~
5.2.1 Fffect of Bendinc on Burst Strenetb of Tube 4,d AA i
\\
5.2.2 Tube t'ith Tb ru-ita11 De9fodoti C,b,0 4
k e
I l
l f
?
6 e
6 6
5.2.3 Thinned Tube For the case of a predominantly thinning mode of tube degradation; i.e.,
no thru-wall cracking and hence no leakage, the minimum tube vall thickness is establishedF
- a, b,:.
L 4
s' Thus, the previously established minimum tube wall l,
~
- 4,e.,
meets t$e applicable burst strength s,,,
~
\\
requirement.
\\
s A
% ss h
I '
73-s s,
\\
's
SECTIOf! 6 PLUCGIl1C ?!APGIll P.ECOfif'r?IDATICM Based on analyses in the previous sections, a minimum wall abe
- is,n,ecessary to satisfy the stress
~
~
1imit and strength requirements of US:RC negulatory Guide d, 6, 1.121.
The allowable degradation incorporates additional allowances for any additional degradation under continued operation until next scheduled inspection and the measurement uncertainties using the EC probes.
An estimate of the degradation allowance can be made based on the history of similarly designed and operated units and
- o, h c.
e the projected inspection interval.
Thus, the recommended tube plugging margin for S!!UPPS is 53 percent of nominal wall; i. e. ; 0.021 inch,
which exceeds the plugging margin of 40% (0.016 in.) allowed by the ASME Cooe Seution XI, Paragraph IUB 3521.1 in lieu of anclyses, s
SECTION 7 APPENDIX 7.1 Deviation of Lower Bound Tolerance Limits for Strenoth Propertie s Expected strength properties to be used for the SNUPPS tubinp evaluation were obtained from statistica1 analyses of tensile
~ ~ "' b' #
test data of actual production tubing.
e M
- c b, c, Table 7-1 sumarizes the calculations of statistical analyses of test data of the mill-annealed and thermall.y-treated Inconel-600 tubing for SNUPPS.
TABLE 7-1: LOWER TOLERANCE LIMITS OF STRENGTH PROPERTIES FOR THE SNt!PPS TUBE o.b,C m
m