ML20212A824
| ML20212A824 | |
| Person / Time | |
|---|---|
| Site: | Cooper  |
| Issue date: | 07/24/1986 |
| From: | Triole S CALSPAN CORP. |
| To: | NRC |
| Shared Package | |
| ML20210K247 | List: |
| References | |
| CON-NRC-03-81-130, CON-NRC-3-81-130 TAC-07934, TAC-7934, TER-C5506-418, TER-C5506-418-S01, TER-C5506-418-S1, NUDOCS 8607290218 | |
| Download: ML20212A824 (22) | |
Text
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- l SUPPLEMENTARY TECHNICAL EVALUATION REPORT NRC DOCKET NO. 50-298 FRC PROJECT C5506 NRC TAC NO. 07934 FRC ASSIGNMENT 12
! N RC CONTRACT NO. N RC-03-81 130 FRC TASK 418 STRUCTURAL EVALUATION OF THE VACUUM BREAY2RS (MARK I CONTAINMENT PROGRAM) t NEBRASKA PUBLIC POWER DISTRICT COOPER NUCLEAR STATION TER-C5506-418
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Prepared for ^^ Nuclear Regulatory Commission FRC Group Leader: V. N. Con Washington, D.C. 20555 NRC Lead Engineer: H. Shaw g
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July 24, 1986
~ I This report was prepared as an account of work sponsored by an agency of the United States t , Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, appa-ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.
Prepared by: Reviewed by: Approved by: k Y Yn s f W Akf A
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i Date:Pbb Date: 7- M -9 fe Date: %d 'N l FRANKLIN RESEARCH CENTER
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- _ ,4 TER-C550G-418 CONTENTS Section Title Page 1 INTRODUCTION . . . . . . . . . . . . . 1 1.1 Generic Background. . . . . . . . . . . 1 1.2 Vacuum Breaker Function . . . . . . . . . 2 2 EVALUATION CRITERIA. . . . . . . . . . . . 9 3 DESIGN LOADS . . . . . . . . . . . . . 10 4 STRESS EVALUATION . . . . . . . . . . . . 11
- 5 PLANT-SPECIFIC REVIEW: COOPER NUCLEAR STATION . . . . . 15 Background Information.
5.1 . . . . . . . . . 15 5.2 Stress Analysis Results . . . . . . . . . 15 6 CONCLUSIONS. . . . . . . . . . . . . . 18 7 REFERENCES . . . . . . . . . . . . . 19 l&
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i r y TER-C5506-418 FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center i under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical assistance in support of NRC operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC.
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! 1. INTRODUCTION In a latter state of the generic resolution of the suppression pool dynamic load definition of the Mark I Containment Long-Term Program, a potential failure mode of the vacuum breakers was identified during the chugging and condensation phases of hydrodynamic loadings. To resolve this issue, two vacuum breaker owner groups were formed, one for those with General Precision Engineering (GPE) vacuum breakers, the other for those with Atwood-Morrill (AM) vacuum breakers.
The issue was not part of the original scope of the Mark I Containment Long-Term Program as described in NUREG-0661 (1]. However, vacuum breakers have the function of maintaining containment integrity and, therefore, are subject to Nuclear Regulatory Commission (NRC) review. In a generic letter
't dated February 2, 1983 (2], the NRC requested all affected plants either to submit the results of the plant-unique calculations which formed the bases for modifications to the vacuum breakers or to provide the justification for the as-built acceptabili'y of the vacuum breakers.
F Franklin Research Center (FRC) has been retained by the NRC to evaluate b
+ the acceptability of the structural analysis techniques and design criteria y used in the plant-unique analysis (FUA) reports of 16 plants. As a part of E
] this review, the structural analysis of the vacuum breakers has been reviewed and documented in this report.
UT b The first part of this report (Sections 1 through 4) consistsofgenericf information that is applicable to all affected plants. The second part of the I l report (Sections 5 and 6) provides a plant-specific review, which pertains to the Cooper plant. i [ 1.1 GENERIC BACKGROUND f
; In 1980, the Mark I owners and the NRC became aware of the vacuum breaker damage during full-scale test facility testing and of the potential for damage
( during actual LOCAs. Two vacuum breaker owner groups, General Precision Engineering (GPE) and Atwood-Morrill (AM), were formed to develop action plan for resolving this issue. In February 1983, the NRC issued Generic Letter 83-08 (2), requesting commitments from affected utilities to provide i l
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f TER-C5506-418 l l 4 analytical results. The licensees responded to the NRC request by developing appropriate force functions simulating the anticipated hydrodynamic loads, and then performing stress analyses that used these loads. With respect to loading, the NRC has reviewed and issued a staff position as indicated in Section 3. FRC's function is to review the stress analysis submitted by a i licensee. 1.2 VACUUM BREAKER FUNCTION During steam condensation tests on BWR Mark I containments, the wetwell-to-drywell vacuum breakers cycled repeatedly during the transient phase of steam blowdown. This load was not included in the original load combinations used in the design of the vacuum breakers. Consequently, the repeated impact i, of the pallet on the valve seat and body created stresses that may impair its I capability to remain functional. A vacuum breaker is a check valve installed between the wetwell and the drywell. Its primary function is to prevent the formation of a negative pressure on the drywell containment during rapid condensation of steam in the I drywell and in the final stages of a LOCA. The vacuum breaker maintains a wetwell pressure less than or equal to the drywell pressure by permitting air
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ci flow from the wetwell to the drywell when the wetwell is pressurized and the drywell is depressurized slowly. E A vacuum breaker can be internally or externally mounted. Figures 1 and i 4 2 illustrate locations of vacuum breakers. ; l Schematics of typical GPE and AM vacuum breakers are illustrated in i Figures 3 and 4. 'l A typical pressure differential vacuum breaker during a LOCA is provided I in Figure 5. f Table 1 lists the various vacuum breaker types and the plants affected by then. l I
") TER-C5506-418 IN TRAN AL.
Vacuum SAEAJLEA WETWE L AIRSPACE ,, r-MAIN VEN A Tg { , .
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TER-C5506-418 f**J Table 1. Vacuum Breaker Types and Affected Plants Vacuum Breaker Plant 4 GPE 18 In (Internal) Brown Ferry Units 1, 2, and 3 Pilgrim Unit 1 Brunswick Units 1 and 2 Cooper Hatch Units 1 and 2
- Peach Bottom Units 2 and 3 Duane Arnold Fermi Unit 2 GPE 24 in (Internal) Hope Creek AM 18 in (Internal) Monticello Quad Cities Units 1 and 2 b
AM 18 in (External) Dresden Units 2 and 3
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'!i 2. EVALUATION CRITERIA To evaluate the design of the vacuum breakers, the affected licensees 1' follow the general requirements of NUREG-0661 (1) and those of " Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis
;; Application Guide" [3). Specifically, the requirements of the ASME Boiler and Pressure Vessel Code, Section III, Subsection NC for Class 2 Components, 1977 Edition, including the summer 1977 addenda (4), have been used to evaluate the structural integrity of the vacuum breakers, 1
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- 3. DESIGN LOADS The loads acting on the Mark I structures and on the vacuum breaker are based upon the Mark I Program Load Definition Report (5) and the NRC Acceptance Criteria (1]. The loads acting on the vacuum breaker include gravity, seismic, and hydrodynamic loads. The hydrodynamic forcing functions were developed by Continuum Dynamics, Inc. (CDI). CDI used a dynamic model of a Mark I pressure suppression system, which was capable of predicting pressure transients at specified locations in the vent system. With this dynamic model and the full-scale test facility data, load definition resulting in pressure differential across the vacuum breaker disc was quantified as a function of time. This issue has been reviewed and addressed by the NRC [6].
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- 4. STRESS EVALUATION l
.l To determine structural integrity of the vacuum breaker, the licensees +, l1 have employed standard analytical techniques, including the finite element jj method, to calculate stresses of critical components of the vacuum breaker l under various design loadings. Loads resulting from the hydrodynamic phenomenon were compared with those values specified in the ASME Codes (4]. l .
! For illustration purposes, a schematic drawing of the moving parts of all jq components other than the actual disc of the Atwood-Morrill valve and of the corresponding finite element model are shown in Figures 6 and 7, respectively.
1 1 The model in Figure 7 was used to investigate the dynamic response following
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! A typical model for stress analysis of the vacuum breaker disc is shown
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.s that were obtained from the impact model analysis.
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TER-C5506-418 i INSIDE DISC SURFACE i I'
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- 5. PLANT-SPECIFIC REVIEW: COOPER NUCLEAR STATION
5.1 BACKGROUND
INFORMATION o Vacuum breaker type: 18-in GPE (internal) l o There are 12 vacuum breakers within the wetwell. ll o Vacuum breakers are located on the main vent / vent header I intersection: two on each of six intersections.
! 5.2 STRESS ANALYSIS RESULTS
,' Stresses in critical vacuum breaker components were analyzed using an ANSYS finite element model. The pallet, hinge shaft, and hinge arm were {" analyzed for hydrodynamic loading due to the chugging transient, including i pallet impact loads based on pallet impact velocities determined in Reference
;'- 7. Stresses in the hinge arm studs and valve seat bolts were based on the
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response of the pallet and hinge arm to the induced loadings. Table 2 r l .. - lL , provides a summary of the critical streses in the vacuum breaker components (8). Allowable stresses were based on References 9 and 10. It is evident j from Table 2 that vacuum breaker stresses were within allowable limits. Therefore, no modifications were necessary. [ I! P *y
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I , i e s'.) TER-C5506-418 l Table 2. Critical Stresses in Vacuum Breaker Components Stresses (psi) Hinge Hinge Hinge Arm Fallet Seat i Pallet Arm Shaft Studs Bolts y Material SA-516 SA-516 SA-479 SA-516 SA-320 Gr 70 Gr 70 Gr 70 B8 -{ MX19 Allowable S h* 17,500 17,500 25,000 17,500 15,000 Service Level A 22,260 15,549 26,959 14,220 15,403 Allowable (1.5 x S h )* 26,250 26,250 37,500 26,250 22,500
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Stress Ratio 0.85 0.59 0.72 0.54 0.68 Service Level B __ 22,260 15,632 26,155 14,371 15,403 N Allowable (1.65 x S h )* 28,875 28,875 41,250 28,875 24,750 Stress Ratio 0.77 0.54 0.66 0.50 0.62 , r 7 a Service Level C 22,260 15,714 27,351 14,522 15,403 Allowable (1.8 x S h)* 31,500 31,500 45,000 31,500 27,000 g
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Stress Ratio 0.71 0.50 0.61 0.46 0.57 l
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- Allowable stresses are based on References 9 and 10.
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- 6. CONCLUSIONS A review has been conducted to determine the structural integrity of the vacuum breakers at the Cooper plant. The design loads associated with the hydrodynamic phenomena have been reviewed and addressed by the NRC in Reference 6. This review covered only the structural analysis of the vacuum breaker, and the following conclusion is drawn from the review:
o The analytical methods used to evaluate stresses of critical components have been reviewed and judged to be adequate: the stress
. results are within the allowables as shown ia Table 2 and, therefore, the existing design is structurally adequate.
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TER-C5506-418
- 7. REFERENCES
- 1. NUREG-0661
" Safety Evaluation Report, Mark I Containment Long-Term Program Resolution of Generic Technical Activity A-7," Office of Nuclear Reactor Regulation, USNRC t July 1980 1
- 2. D. G. Eisenhut j "USNRC Generic Letter 83-80, Modification of Vacuum Breakers on Mark I l Containment" February 2, 1983 i
- 3. NEDO-24583-1
" Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide," General Electric Co., San Jose, CA October 1979
- 4. American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section III, Division 1, " Nuclear Power Plant Components," New York, 1977 Edition and Addenda up to Summer 1977
.. 5. NEDO-21888 Revision 2
" Mark I Containment Program Load Definition Report," General Electric Co., San Jose, CA November 1981 7
. 6. D. B. Vassallo, NRC q
Letter with Attachment to H. C. Pfefferlen, BWR Licensing Programs, GE
" Evaluation of Model for Predicting Drywell to Wetwell Vacuum Breaker
[ ' Valve Dynamics"
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December 24, 1984
- 7. " Improved Dynamic Vacuum Breaker Valve Response for the Cooper Nuclear j
'" Station, Revision 0," C.D.I. Technical Note 82-31 /
Continuum Dynamics, Inc., Princeton, New Jersey September 1982
- 8. J.M. Pilant y Letter with Attachment to D. B. Vassallo (NRC) l
Subject:
Mark I Containment Program Plant Unique Analysis Report Nebraska Public Power District g April 29, 1983
- 9. ASME Code Case N-62-2 (1621-2), " Internal and External Valve Items, Section III, Division I, Class I, 2, and 3 Line Valves," Appe ved May 15, f 1980
- 10. ASME Code, Section III, Subsection NC-3500, up to and including the i
Summer 1977 Addenda i I I }}