ML20212C072

Structural Evaluation of Vacuum Breakers (Mark I Containment Program),Oyster Creek Plant, Supplementary Technical Evaluation Rept

ML20212C072
Person / Time
Site:	Oyster Creek
Issue date:	08/04/1986
From:	Carfagno S, Con V, Triolo S CALSPAN CORP.
To:	NRC
Shared Package
ML20212C077	List: ML20212C072 ML20212M326 ML20212M356
References
CON-NRC-03-81-130, CON-NRC-3-81-130 TAC-07944, TAC-7944, TER-C5506-319, TER-C5506-319-S01, TER-C5506-319-S1, NUDOCS 8608080138
Download: ML20212C072 (20)

Text

- - _ _

- - ATTACHMENT TO SAFETY EVALUATION s:FFLErr;TARY TECHNICAL EVALUATION REPORT ,

FRC PROJECT C5506 NRC DOCKET NO 50-219 NRC TAC NO 0794.; FRC ASSIGNMENT 12 NRC CONTRACT NO. NRC-03-81-130 FRCTASK 319 STRUCT"EAL EVALUATIO!; CF THE VAC'JJM EFIAJIES (MAFE. I CO :TA!!;!'.E!;T FRO 3Fa.;

GE"IFA' F"ELIC UTILITIES CYSTEF CREEK FLA';T

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Prepared for FRC Group Leader: v . 1; . Cer.

fj Nuclear Regulatory Commission NRC Lead Engineer. E. Sha.

Washington. D.C. 20555 l

August 4, 1986 l

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l This report was prepared as an account of work sponsored by an agency of the United States Government. Neitner the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal hability or l 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 thirc party would not inf ringe privately owned rights.

Prepared by: Reviewed by: Approved by:

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- FORDCRD This. Technical Evaluation Report-was prepared by Franklin Research' Center

- under a contract 1with the.U.S.' Nuclear Regulatory Cor.i.ission (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 estahlished by

the'NRC.

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. 1. INTRODUO!!'ON l

~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 hydrodynaric 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 Contain:ent 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 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 cs-built acceptability of the vacuum breakers.

Franklin Research Center (FRO) b.s been retained by the NRC to evaluate the acceptability of the structural analysis techniques and design criteria used in the plant-unique analysis (PUA) reports of 16 plants. As a part of this review, the structural analysis of the vacuum breakers has been reviewed cnd documented in this report.

The first part of this report (Sections 1 through 4) consists of generic information that is applicable to all affected plants. The second part of the report (Sections 5 and 6) provides a plant-specific review, which pertains to the Oyster Creek plant.

1,1 GENERIC BACKGROUID 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 1963, the NRC issued Generic Letter 83-08 (2], requesting commitments from affected utilities to provide

TEF-Cli16-31E analytical results. The licensees responded to the NRC request by develeping appropriate force functions sirulating the anticipated hydrodynamic -loads, and

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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 fun: tion is to review the stress analysis submitted by a 4 licensee.

1.2 VACUUM BTJ.Ak'IR FUNCTION During steam condensation tests on B'4R Mark I containments, the wetws11-to-drywell vacuum breakers cycled repeatedly during the transient phase of steam blowdown. This load was not included in the original load continations

- used in the design of the vacuum breakers. Consequently, the repeated irpa:t of the pallet on the valve seat and body created stresses that may impair its 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 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 flow from the wetwell to the drywell when the wetwell is pressurized and the drywell is depressurized slowly.

A vacuum breaker can be internally or externally mounted. Figures 1 and 2 illustrate locations of vacuum breakerr.

Schematics of typical GFE and AM vccuum breakers are illustrated in Figures 3 and 4.

A typical pressure differential vacuum breaker during a LOCA is provided in Figure 5.

Table 1 lists the various vacuum breaker types and the plants affected by them.

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-Table 1. : Vacuum Breaker Types and Affected Plants-Vacuum Breaker Plant

,f x7 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 Termi Unit 2 GPE 24 in (Internal) Hope Creek' AM 18 in (Internal) Monticello Quad Cities Units 1-and 2 AM 18 in (External) Dresden Units 2 and 3 Millstone Unit 1 Oyster Creek Vermont Yankee-AM 18 in (External) FitzPatrick Nine Mile Point Unit 1 x

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2. EVALUATION CRITEK*A To evaluate the design of th'e vacuum breakers, the affected licensees 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 N0 for Class 2' Components, 1977 Edition, including'the summer 1977 addenda (4), have~been used to evaluate-the

- structural integrity of the vacuum breakers.

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3. DESIGN LCADS The loads acting on the Mark'I structures and on the-vacuar b~reaker 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 CDI used a dynamic model of a Mark I pressure Continuum Dynamics, Inc, (CDI).

suppression system,,which.was capable of predicting pressure transients at With this dynamic model and theffull-specified locations-in the vent system.

scale test facility data, load definition resulting in pressure differential This across. the vacuum breaker disc was quantified as a function of time.

issue has been reviewed and addressed by the NRC [6).

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4. S!TsESS EVA1.UATICN I'

To determine structural integrity of the vacuum breaker, the licensees have employed standard analytical techniques, including the finite eierent method, to calculate stresses of critical components of the vacuum breaker under various des:gn loadings. Loads resulting from the hydrodynarie l

phenomenon were compared with those values specified in the AS.E Codes (4:.

For illustration purposes, a schematic drawing of the moving parts of a'..

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, respective 1,..

l The model in Figure 7 was used to investigate the dynamic response follow:n; impact.

A typical model for stress analysis of the vacuum breaker disc is sho n in Figure 8. Loading inputs to this model are the displacement time histcr:es l

that wtre obtained froc the impact model analysis.

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5. PLANT-SPOCIFIC REVIEW: OYSTER CREEK PLANT

5.1 BACKGROUND

INFORMATION o Vacuum breaker type: 18-in Atwood-Morrill (external) o Two wetwell-to-drywell vacuum breakers are located on each of seven external wetwell-to-drywell vacuum relief piping assemblies.

o In addition to the watwell-to-drywell vacuum relief lines, there is

-one wetwell-to-reactor building vacuum relief piping assembly. This assembly contains two vacuum breakers that pern.it air to flow from the reactor building air space into the wetwell.

5.2 STRESS ANALYSIS RESULTS An analysis of the torus-to-drywell vacuum breaker valves was performed to determine the structural integrity of the valves under the chugging transient load. The analysis in Reference 8 indicated that certain vacuum breaker parts could become overstressed. Consequently, the Licensee will replace the following parts with parts having different material properties to achieve acceptable stress levels:

o counterweight arms o disk arms o disk arm keys o valve shaft I o counterweight hubs o counterweight arm hub keys.

I According to Reference 8, GPU Nuclear requests deferment of modifications until the Cycle 12 refueling outage, at which time 7 of the 14 valves will be modified. The remaining 7 valves will be completed during the Cycle 13 outage.

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6. CONOLUSIONS A review has been conducted to dete mine the structural integrity of the vacuu breakers at the Oyster Creek 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 vacuu-breaker, and the following conclusion is drawn from the review:

I o Some vacuu breaker parts may become overstressed during the chugging transient. The Licensee will replace the counterweight a ms, disk arms, disk are keys, valve shaft, counterweight hubs, and counter-I weight arm hub keys with parts of different materials so stress levels will be acceptable.

l o The Licensee intends to defer modifications for 7 out of the 14 valves until the Cycle 12 refueling outage and the remaining 7 until the Cycle 13 outage.

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.7. REFERENCES a

1. NURIG-0661

" Safety Evaluation Report, Mark I Containment Long-Term Program Resolution of Generic Technical Activity A-7," Office of Nuclear Reactor Regulation, USNRC July 1980

2. D. G. Eisenhut "USNRC Generic Letter 83-80, Modification of Vacuum Breakers on Mark I Containment" February 2, 1983

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 S. NEDO-21888 Revision 2

" Mark I Containment Program Load Definition Report," General Electric Co., San Jose, CA November 1981

6. D. B. Vassallo, NRC Letter with Attachment to H. C. Pfefferlen, BWR Licensing Prograes, GE

" Evaluation of Model for Predicting Drywell to Wetwell Vacuum Breaker Valve Dynamics" December 24, 1984

7. General Electric Company Letter MFN-159-82 to NRC October 29, 1982

8. P. B. Fiedler (GPU)

Letter with attachment to J. A. Zwolinski (NRC)

Subject:

Oyster Creek, Mark I Containment Torus-to-Drywell Vacuum Breakers GPU Nuclear May 23, 1986