ML19199A123

Forwards Safety Evaluation for Pool Dynamic Structural Load Aspects of Mark I Containment Long-Term Program. Util Actions to Restore Original Design Safety Margin Acceptable Submittal of Tech Spec Change within 90 Days.

ML19199A123
Person / Time
Site:	Quad Cities
Issue date:	02/15/1986
From:	Zwolinski J Office of Nuclear Reactor Regulation
To:	Farrar D Commonwealth Edison Co
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ML19199a122	List: ML19199A123
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Download: ML19199A123 (35)

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' .

I UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, 0. C. 20555 February 15, 1986 Docket No. 50-254/265 Mr. Dennis L. Farrar Director of Nuclear Licensing Corrrnonwealth Edison Company Post Office Box 767 Chicago, Illinois 60690

SUBJECT:

MAPK CONTMINMENT LONG TERM PROGRAM Re: Quad Cities Nuclear Power Station, Units 1 and 2 We have completed our post-implementation audit review of the Quad Cities Station, Units 1 anc 2 Plant Unique Analysis Report (PUAR) for the Mark I containment long term program. Enclosed are the staff's Safety Evaluations for the pool dynamic and structural load aspects of this program. Technical assistance was provided by Brookhaven National Laboratory for the pool dynamic load audit review and by Franklin Research Center for the structural audit review. Copies of their technical evaluations art also enclosed.

The staff has determined that all but a few of the modifications made by Corrrnonwealth Edison Company are in accordance with the generic acceptance criteria contained in Appendix A of NUREG-0661, Mark I Containment Long Term Program and its supplement. Where deviations from the acceptance criteria specified in NUREG-0661 have been taken, they have been found acceptable.

Therefore, -..i1e staff has concluded that the Corrrnonwealth Edison Company PUAR analysis verified that the containment modifications made have restored the original design safety margin to the Mark I containment at the Quad Cities Station, Units 1 and 2. This action completes our review of this issue.

The vacuum breakers on Mark I containments, which were the subject of Generic Letter 83-02, are not considered within the scope of the Mark I containment long term program. This issue is being reviewed independently.

8603030411 8 54 PDR ADOC O PDR p

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• I Mr. Dennis L. Farrar February 15, 1986 Within 90 days of receipt of this letter it is requested that you submit any Technical Specification changes required as a result of the Mark I contai n ment modification s you have made; (i.e., torus temperature monitoring system changes, torus to drywell differential pressure control changes, torus water level chan ges).

1_ JJ Sincli J n . Zwolinski, Director B oject Directorate #1 Divisio n of BWR Licensing

Enclosures:

1. Safety Evaluation - Pool Dynamics load

2. Safety Evaluation - Structural Review

3. BNL Technical Evaluation Report

4. FRC Technical Evaluation Peport cc w/er.closures:

See next p3ge

I t Mr. Dennis L. Farrar Quad Cities Nuclear Power Station Colffllonwealth Edison Company Units 1 and 2 cc:

Mr. B. C. 0 1 Brien President Iowa-Illinois Gas and Electric Company 206 East Second Avenue Davenport, Iowa 52801 Robert G. Fitzgibbons, Jr.

Isham, Lincoln & Beale Three First National Plaza Suite 5200 Chicago, Illinois 60602 Mr. Nick Kalivianakis Plant Superintendent Quad Cities Nuclear Power Station 22710 - 206th Avenue - North Cordova, Illinois 61242 Resident Inspector U. S. Nuclear Regulatory ComJission 22712 206th Avenue NJrth Cordova, Illinois 61?42 Chairman Rock Island County Board of Supervisors Rock Island County Court House Rock Island, Illinois 61201 Mr. Gary N. Wright Nuclear Facility Safety Illinois Department of Nuclear Safety 1035 Outer Park Drive, 5th Floor Springfield, Illinois 62704 Regional Administrator, Region III U. S. Nuclear Regulatory Conmission 799 Roosevelt Road Glen Ellyn, Illinois 60137

fE:B 15 1986 Mr. Dennis L. Farrar Within 90 days of receipt of this letter it is requested that you submit any Technical Specification changes required as a result of the Mark I containment modifications you have made; (i.e., torus temperature monitoring system changes, torus to drywell differential pressure control changes, torus water level changes).

Sincerely, John A. Zwolinski, Director BWR Project Directorate #1 Division of BWR Licensing

Enclosures:

1. Safety Evaluation - Pool Dynamics Load

2. Safety Evaluation - Structural Review

3. BNL Technical Evaluation Report

4. FRC Technical Evaluation Report cc w/enclosures:

See next page DISTRIBUTION .*

Doc1cet,'. .fl1 e NRC POR .

Local PDR BWD#l Reading RBernero, w/o TERs DELO EJordan, w/o TERs BGrimes, w/o TERs JPartlow, w/o TERs RBevan CJamerson, w/o TERs ACRS (10)

FEltawila, w/o TERs HShaw, w/o TERs DVassallo, w/o TERs GHolahan, w/o TERs QC File DBL: PD#l CJamerson

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t DBL: PD#11>1lJ RBevan:t-1 lf/86 DBL:PD#l JZwolinski r;r*

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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO MARK I CONTAINMENT LONG-TERM PROGRAM POOL DYNAMIC LOADS REVIEW QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NOS. 50-254/265

1.0 INTRODUCTION

In July 1980, the staff issued a report, NUREG-0661, "Safety Evaluation Report, Mark I Containment Long-Term Program," to address the NRC acceptance criteria for the Mark I containment Long-Term Program, which are intended to establish desigr basis loads that are appropriate for the anticipated life of each Mark I boiling water reactor (BWR) facility, and to restore the originally intended design safety margins for each Mark I containment system.

Since the issuance of NUREG-0661, the Mark I owners submitted additional reports in which they provided additional justification for the adequacy of: (1) the data base for specifying torus wall pressure during condensa tion oscillations; (2) the consideration given to asymmetric torus loading during condensation oscillations; and (3) the effect of fluid compressibility in the vent system on pool-swel1 loads. As a result of the staff's and its consultant's (Brookhaven National Laboratory (BNL)) evaluation of these reports, Supplement 1 to NUREG-0661, dated August 1982, has been issued.

2.0 EVALUATION Co1m1onwealth Edison Company submitted a Plant Unique Analysis Report (PUAR) on the pool dynamic loads for the Mark I containments for Quad Cities Station.

Units I and 2. This report provides a description of the specific application of the generic Mark I pool dynamic loads and methods for Quad Cities and the plant unique loads used in assessing the capability of the containment and components to acconmodate the pool dynamic loading phenomena. The BNL was contracted to review the PUAR for compliance with the staff's acceptance criteria and to e valuate the acceptability of any proposed alternative load specification.

A sunmary of the BNL review and status for each of the pool dynamic loads is presented in the attached report titled "Technical Evaluation of the Quad Cities Plant Unique Analysis Report. 11 As indicated in the report, Conmon wealth Edison has adopted all but a few of the generic criteria. For those few exceptions alternative criteria were proposed. The BNL evaluation of these criteria is included in the attached report. Based on its review, the staff endorses the BNL evaluation and conclusion.

I

3.0 CONCLUSION

S The staff has completed an assessment of Quad Cities Station, Units 1 and 2 against generic acceptance criteria contained in NUREG-0661 and its supplement, and has also reviewed those few areas where alternative criteria have been proposed. In addition, the staff has completed its review of those areas where additional information was relegated to the plant unique review. In each of these areas the staff has concluded that the pool dynamic loads utilized by the licensee are conservative and, therefore, acceptable.

Principal Contributor: F, Eltawila Date: February 15, 1986 Attached: Technical Evaluation Report, dated September 1984, prepared by BNL

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 SAFETY.EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO MARK I CONTAINMENT LONG-TERM PROGRAM STRUCTURAL REVIEW QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 DOCKET NO. 50-254/265

1.0 INTRODUCTION

The capability of the boiling water reactor (BWR) Mark I containment structures and piping systems to withstand the effect of hydrodynamic loads resulting fro a loss of coolant accident (LOCA) and/or a safety relief valve (SRV) discharge was not considered in the original design of the structures.

The resolution of this issue was divided into a short-term program and a long-term program.

Based on the results of the short-term program, which verified that each Mark I containment would maintain its integrity and functional capability when subjected to the loads induced by a design-basis LOCA, the NRC staff granted an exemption relating to the structural safety requirements of 10 CFR 50.55(a). The study was reported in NUREG-0408, "Mark I Containment Short Term Program".

The objective of the long-term program was to maintain a margin of safety when the Mark I containment structures and piping system are subjected to additional hydrodynamic loads. The detailed guidance of the long-term pro gram are contained in the NRC Safety Evaluation Report, NUREG-0661, "Mark I Containment Long-Term Program" and its supplement which describe the generic hydrodynamic load definition and structural acceptance criteria consistent with the requirements of the applicable codes and standards.

To fulfill the objective of the long-term program, Cormionwealth Edison Company (the licensee) has completed all modifications on the Quad Cities Units land 2 containments and torus attached piping. The adequacy of these modifications was documented in a report prepared by Nutech Engineers, Inc., titled, Plant Unique Analysis Report, Quad Cities Nuclear Generating Station Units 1 and 2, Revision 0.

The Franklin Research Center (FRC) was contracted to review the structural adequacy issue for compliance with the staff's acceptance criteria.

2.0 EVALUATIOf\

The Mark I long-term program of the Quad Cities Station, Units land 2 was described in the plant-unique analysis report prepared by Nutech. This report describes modifications performed on containment structures and torus 8603030426 860215 PDR ADOCK 05000254 P PDR

attached p;pjng at the Quad Cities Station. Units 1 and 2. Areas covered by the report include the torus shell. external support system. vent header system. internal structures. torus attached pipings, SRV lines and vent pipe penetrations. The materials, design and fabrication requirements of the modifications were in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Division 1.Section III with Addenda through Surrmer 1977 and Code Case N-197. "Service Limits for Containment Vessels".

Modifications were performed in accordance with the requirements of Section XI of the same code. To determine the appropriate code allowable service limits for the specified loading combinations. the report followed guidelines of NUREG-0661 and the Ger,eral Electric Company (GE) report. NED0-24583-1, "Mark I Containment Program Structural Acceptance Criteria Plant Unique Analysis Application Guide." The portion of the report applicable to loadings and loading combinations was audited by BNL, and results of that audit are discussed in a separate Safety Evaluation. Using the properly determined loadings and loading combinations. the pool dynamic loads used by the licensee are conservative and, therefore, acceptable. Results of the analyses were sunvnarized to show that modifications are adequate under various loading combinations.

The adequacy of the modified containment structures and torus attached piping was audited by the FRC. FRC developd audit procedures for all Mark I long term program users, which is described in detail in the FRC TER-C5506-308, "Audit Procedures for Mark I Containmerit Long-Tenn Program - Structura 1 Analysis." Th review performed by FRC followed this document closely.

Results and conclusions of this effort were reported in FRC TER-C5506-325.

11 Audit for Mark I Containment Long-Term Program-Structural Analysis for Operating Reactors- Conmonwealth Edison Company, Quad Cities Nuclear Generating Station Units 1 and 2." The audit verified analyses by examining mathematical models and loading combinations used, and surrmarized the results to see whether the modifications met the required criteria. A check list was compiled to ensure the completeness of the auditing. The staff has reviewed the FRC report and concurs with its conclusions that the modifications meet the Mark I Containment long-Tenn Program objective. An augmented fatigue evaluation method for ASME Code Class 2/3 piping was developed by MPR for GE in MPR Report-751, titled, "Augmented Class 2/3 Fatigue Evaluation Method and Results for Typical Torus attached on SRV Piping System". dated November 1982. This report was reviewed by the staff *and the conclusion that all torus piping systems have a fatigue usage of less than 0.5 during the plant life is acceptable for the Quad Cities Units 1 and 2.

3.0 CONCLUSION

S The modifications performed at the Quad Cities Nuclear Generating Station Units 1 and 2 followed the guidelines of NUREG-0661 and its supplement and met the respective requirements of Sections III and XI of the ASME Boiler and Pressure Vessel Code and are, therefore, acceptable. The licensee's analyses have been verified by the FRC audit and approved by the staff under the LOCA and SRV discharge loads.

Principal Contributor: H. Shaw Date: February 15. 1986 Attached: Technical Evaluation Report prepared by Franklin Research Center, dated June 21

• 1985

TECHNICAL EVALUATION OF THE QUAD CITIES NUCLEAR GENERATING STATION UNITS 1 AND 2 PLANT-UNIQUE ANALYSIS REPORT John R. Lehner George Bienkowski Reactor Safety Licensing Assistance Division Department of Nuclear Energy Brookhaven National Laboratory Upton, New York 11973 September 1984 FIN A-3713 BNL-04243 8603030438 860215 PDR ADOCK 0,000254 P PDR

ABSTRACT Th;s Techn;cal Evaluation Report (TER) presents the results of the post-implementation audit of the Plant Unique Analysis Report (PUAR) for the Quad Cities Nuclear Generating Station Units 1 and 2. The contents of the PUAR were compared against the hydrodynamic load Acceptance Criteria (AC) contained in NUREG-0661. The TER summar izes the aud;t findings (Table 1), and discusses the nature and status of any exceptions to the AC, identified during the audit (Table 2).

ACKNLEDGEMENTS The cognizant NRC Technical Monitor for this program was Dr. Farouk Eltawila of the Containment Systems Branch (OSI) and the NRC Project Manager was Mr. Jack N. Donohew of the Technical Assistance Program Management Group of the Division of Licensing. Mr. Byron Siegel of erating Reactors Branch No. 2 (DL) was Lead Project Manager.

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List of Acronyms AC Acceptance Criteria BNL Brookhaven National Laboratory BWR Boiling Water Reactor co Condensation Oscillation OBA Design Basis Accident DL Division of Licensing OSI Division of Systems Integration FSI Fluid Structure Interaction FSTF Full Scale Test Facility LOR Load Definition Report LOCA Loss-of-Coolant Accident LTP Long Term Program NRC Nuclear Regulatory Corrmission PUA Plant-Unique Analysis PUAR Plant-Unique Analysis Report QSTF Quarter Scale Test Facility RFI Request For Infomation SER Safety Evaluation Report SPTMS Suppression Pool Temperature Monitoring System S/RV Safety/Relief Valve S/RVDL Safety/Relief Valve Discharge Line STP Short Tenn Program TAP Torus Attached Piping TER Technical Evaluation Report

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Table of Contents Page No.

ABSTRACT t ACKNo.lLEDGEMENTS ii LIST OF ACRONYMS hi

1. INTRODUCTION 1

2. POST-IMPLEMENTATION AUDIT

SUMMARY

3

3. EXCEPTIONS TO GENERIC ACCEPTANCE CRITERIA 11

4. CONCLUSIONS 18

5. REFERENCES 19

.. 1 y ..

1. INTRODUCTION The suppression pool hydrodynamic loads associated with a postulated loss of-coolant accident (LOCA) were first identified during large-scale testing of an advanced desig pressure-suppression containment (Mark III). These additional loads, ich had not explicitly been included in the original Mark I containment design, result from the dynamic effects of drywell a1r and steam being rapidly forced into the suppression pool (torus). Because these hydrody namic loads had not been considered in the original design of the Mark I con tainment, a detailed reevaluation of the Mark I containment system was required.

A historical development of the bases for the original Mark I design as well as a summary of the two-part overall program (i.e., Short Term and Long Term Programs) used to resolve these issues can be found in Section 1 of Refer ence 1. Reference 2 describes the staff's evaluation of the Short Term Program (STP) used to verify that licensed Mark I facilities could continue to operate safely "1ile the Long Term Program (LTP) was being conducted.

The objectives of the LTP were to establish design-basis (conservative) loads that are appropriate for the anticipated life of each Mark I BWR facility (40 years), nd to restore the originally intended design-safety margins for each Mark I containment system. The principal thrust of the LTP has been the development of generic methods for the definition of suppression pool hydrody namic loadings and the associated structural assessment techniques for the Mark I configuration. The generic aspects of the Mark I Owners Group LTP were com pleted with the submittal of the wMark I Containment Program Load Definition Re port" (Ref. 3) and the NMark I Containment Pr*'Jram Structural Acceptance Guide" (Ref. 4), as well as supporting reports on the LTP eperimental and analytical tasks. The Mark I containment LTP Safety Evaluation Report (NUREG-0661) presented the NRC staff's review of the gener;c suppression pool hydrodynamic load de*ition and structural assessment techniques proposed in the reports cited above. It was concluded that the load definit;on procedures utilized by the Hark I Owners Group, as modified by NRC requirements, provide conservative estimates of these loading conditions and that the structural acceptance crite ria are consistent with the requirements of the applicable codes and standards.

The generic analysis techniques are intended to be used to perform a plant-unique analysis (PUA) for eacn Mark I facility to verify compliance with the acceptance criteria (AC) of Appendix A to NUREG-0661. The objective of tnls study is to perform a post-implementation audit of the Quad Cities plant-unue analysis (Reference 5) against the hydrodynamic load criteria 1n NUREG-0661.

2. POST-IMPLEMENTATION AUDIT

SUMMARY

lhe purpose of the post-implementation audit was to evaluate the hydrodynamic loading methodologies \llllich were used as the basis for modifying the pressure suppression system of the Quad Cities Nuclear Generating Station Units 1 and 2. lhe Quad Cities PUAR methodologies (Reference 5) 111ere compared with those of the LDR (Reference 3) as approved in the AC of NUREG-0661 (Reference 1). lhe audit procedure consisted of a moderately detailed review of the plant unique analysis report (PUAR) to verify both its completeness and its compliance with the acceptance criteria. A list of requests for further information was submitted (Reference 6), and answers were obtained at a meeting with the licensee (Reference 7).

Table 1 su1TVTiarizes the audit results. It lists the various load categories specified in the AC, and indicates plant-unique information through the references, in the right-hand column, to the notes -,ich follow in the text.

CRITERIA u,Z w w :::c

_J wO ID .... u en

....0 Of- <t <l'.

u ..... <l'. Zo I

C>W MET NOT 0U Z.J a: CJ::

W Wen MET a.

a._

t,- CL z

J u a. -'<l LOADS z <l'. <t <l CONTAINMENT PRESSURE 6 TEMPERATURE 2.1 \,/"" '

VENT SYSTEM THRUST LOADS 2.2 V POOL SWELL TORUS NET VERTICAL LOADS 2.3 v TORUS SHELL PRESSURE HI STORIES 2.4 /

I

.,:. VENT SYSTEM IMPACT AND DRAG 2.6 v I I

IMPACT AND DRAG ON OTH ER STRUCTURES 2.7 \,/

FROTH IMPI NGEMENT 2.8 V"" .t, P O OL FALLBACK 2.9 v LOCA JET LOCA B UBBLE DRAG 2.14.1 2.14.2 v 3 v"' 3 VENT HEADER DEFLECTOR LOADS 2.10 v TABLE 1. LOAD CHECKLIST FOR POST-IMPLEMENTATION AUDIT

iii

{0 .,

CRITERIA w w :r::

wO

_J m u en o.- ._ <t  ::z <t

• u NOT 0 o:O J-Cl llJ MET Z-l wO:: 0 Wen MET a.. ._ a.. z 0:

> u a.. ..J a.

<l <l .

Z<! <t LOADS CONDENSATION OSCILLATION TORUS SHELL LOADS 2.11. i v"' "

LOADS ON SUBMERGED STRUCTURES 2.14.5 V""

VEN T SYSTEM LOADS 2.11.3 v :3, 't, t

D OWNCOMER DY NAM IC LOADS 2.11.2 v U'1 CHUGGING TORUS SHELL LO/\OS 2.12.1 v" 4 LOADS ON SUBMERGED STRUCTURES 2.14.6 v :SJ "',

VENT SYSTEM LOADS 2.12.3 /

LATERAL LOADS ON DOWNCOMERS 2.12.2 /

TABLE 1. (CONTINUED)

(0 <L-CRITERIA w

.J W

C lO 0 (D t-u (f) 0 1--

u 1-- <{ Zo

<( <{ w NOT I

0 1--

(9 W MET Z.J a: 0:: 0 l..d (f)

MET LIJ a.. z o: a..

a. t-a..

=.J u ...J <t Z<t LOADS <(

T*QUENCHER LOADS DISCHARGE LINE CLEARING 2 .13. 2 /

TORUS SHELL PRESSURES 2.13.3 /

JET LOADS ON SUBMERGED STRUCTURES 2.14.3 v

0\

.,/

AIR BUBBLE DRAG 2.14.4 1, 7 THRUST LOADS ON T/Q ARMS 2.13.5 v S/RVDL ENVIRONMENTAL TEMPERATURES 2.13.6 TABLE 1. (CONTINUED)

CRITERIA wO Z w

...J W :C t-w m <tU (/)

0 I-I u f- <( z<(

o:O w

I-e>w NOT 0 MET Z..J UJ 0:: 0 Wen 0: MET a. ._ a..

_,a.

> u a.

2 <l: ct ct <t DESCRIPTION I SU PRESSION P O OL TEMPERATURE LIMIT 2.13.8 v 2

SUPRESSION POOL TEMPERAT URE MONITORING SYSTEM 2.13.9 v 8 DIFFERENTIAL PRESSURE CONTROL "II SYSTEM FOR THOSE PLANTS USING A 2.16 DRYWELL-TO-WETWELL PRESSURE DIFFERENCE AS A POOL SWELL v' 9

• MITIGATOR SRV LOAD ASSESSMENT BY 4 IN* PLANT TEST 2.13.9

/ ,o TABLE 1, (CONTINUED)

Notes to Table 1 Number 1 The Acceptance Criteria do not provide a separate procedure for calculating pool swell impact on spherical structures such as the main vent-to-vent header junction in Quad Cities. In the PUAR, the spherical junction was modeled as a series of cylinders with axes along the main vent centerline. Acceleration drag, buoyancy and velocity drag were calculated using AC methodology for cylinders.

This procedure was found acceptable.

2 For some structures, Region I froth loads were calculated using the high-speed QSTF movies. This alternative is outlined in Appendix A of the AC.

3 Instead of the equivalent cylinder procedure specified in the AC to calculate acceleration drag volumes on sharp cornered submerged structures, the PUAR selected alternate modeling of the structures and used published acceleration volumes. Tite discussion in Section 3.1 explains 'tiny this procedure was found acceptable.

4 To calculate CO and post-chug loads on the torus shell as well as on submerged structures, the 50 individual load harmonics were combined using a random phasing technique instead of the absolute surmiation specified in the AC. lhe discussion of Section 3.2 describes why this alternate method was found acceptable

Number 5 To account for FSI effects during CO and chugging submerged structure loads, the AC suggested adding torus boundary accelerotions directly to local fluid accelerations. Instead, the applicant used a method which calculated FSI acceleration fields anywhere in the torus based on knowing the boundary accelerations. This method, ..tlich has been accepted during previous PUAR reviews, is discussed in Section 3.3.

6 The analytical model to calculate SRV torus shell loads approved in the AC was modified slightly before being applied to Quad Cities. The purpose of the modifications was to more closely bound the pressure traces observed in the lltinticello tests on lilklich the model is based.

These changes have been found acceptable. SRV tests conducted in the Drsden plant further confirm that the analytically obtained loadings for Quad Cities are conservative.

7 For SRV air bubble drag loads, the applicant reduced the AC bubble pressure bounding factor of 2.5 to 1.75. This still bounded peak positive bubble pressure and maximum bubble pressure differential from the Monticello test data. Dynamic load factors were derived from Dresden's in-plant SRV test data. These modifications have been found acceptable and are discussed 1n more detail in Section 3.4.

8 The new SPTMS is acceptable. As stated in Section 1-5.2 of the PUAR, the applicant has conmitted to perform a separate analysis demon strating that delayed operator action *based on SPTMS readings will not cause the supp r ession pool temperature to exceed the limit specified in NUREG-0783.

\ I Number 9 In order to reduce torus shell pressures caused by OBA pool swell. a minimum positive pressure d;fference of 1.0 psi is maintained between the Quad Cities drywell including the vent system. and the torus air space. According to Technical Specifications. the plant is required to come to shutdo if the main AP system fails.

10 SRV tests performed in the Dresden plant * .tiich is very similar to Quad Cities, were used to confirm that the analytically derived SRV shell loads for Quad Citi s are conservative and to deduce dynamic load factors for submerged structures.

3. EXCEPTIONS TO GENERIC ACCEPTANCE CRITERIA Quad Cities Units 1 and 2 are two of several plants analyzed by NUTECH Engineers, Inc. based on an essentially conmon hydrodynamic loading methodology (Fermi, Duane Arnold, Monticello and Dresden are other plants in this group).

The methodology differs from the generic acceptance criteria of NUREG-0661 in four major areas "'1ich are listed in Table 2.

In -ttat follows, each of these areas is discussed in detail, and the bases for the resolutions of the differences indicated.

Table 2: Issues Identified During Audit as Exceptions to the Generic Acceptance Criteria Issue No. Description Status Resolved

1. Use of acceleration drag volumes Wlich X differ from those approved in the AC to determine drag on sharp cornered struc tures.

2. Phasinj of load harmonics used to analyze X structures affected by CO and post-chug 1 oads.

3. FSI methodology used for CO and chugging X submerged structure loads.

4. Use of calibration factors developed from X Dresden in-plant tests for use in defining SRV sut.. arged structure drag loads.

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3.1 Accelerat;on Drag Volumes for Sharp Cornered Structures The Acceptance Criteria 2.14.2 section 2b in NUREG-0661 states that drag forces on structures with sharp corners (e.g. rectangles and "I" beams) must be computed by considering forces on an equivalent cylinder of diameter Deq*21/2 Lmax "'1ere I.max is the maximum transverse dimension. The intent of th;s criterion is to provide a conservative bound (based on very lim ited data) that includes non-potential flow effects such as vortex shedding on both the acceleration drag due to hydrodynamic mass and the "standard" drag pro portional to velocity squared. Since the dominant load for the Ring Beam (the primary non-cylindrical structure) is ar-leration drag, the issue concerns only the hydrodynamic mass or acceleration vclume and not the drag coefficient in the Quad Cities plant-specific case.

The PUAR states that "published" acceleration drag volumes listed in Table 1-4.1-1 are used for sharp edged structures rather than the equivalent :yl inder specified in the acceptance criteria. The detailed response to a Request for Information (Item 1) explains that modeling of the actual structures is neces sary, and in particular, forces on the web of the r;ng beam are obtained by modelling the beam by a circumscribed rectangle. In order to evaluate the implications of this modelling, sample calculations were performed on the ring beam for the post-chug loading condition.

A direct application of the Quad Cities PUAR methodology leads to an accel eration volume of 7.19 ft3 for in-plane forces on the maximally loaded ring beam segment. A model that 1110re accurately rep.resents the interference effect but uses acceleration volumes from Tablel-4.1-1 gives a volume of 3.93 ft3, thus providing a substantial margin for possible non-potential flow effects.

For the out-of-plane forces, the PUAR model given an acceleration volume of 66.4 ft3, wiile a similar modelling of the structure but using more realistic interference corrections yields a transverse acceleration volume of 61.5 ft3.

While this leaves very little margin for non-potential flow corrections, in the parameter range of CO and post-chug acceleration spectrum lilflere the major energy is concentrated, the flow is expected to be very nearly potential. In addition.

the use of single mode dynamic load factors, as explained 1n response to RFI item 2, provides additional substantial conservatism up to a factor of 2. The conservative application of the AC equivalent cylinder model and interference correction to the hydrodynamic volume alone, while retaining the real volume for the "effective buoyancy 11 effect, gives an effective acceleration volurr.e of 106.8 ft3, wtiich yields out-of-plane loads 611 higher than those predicted by the -

PUAR. Because of the parameter range in the Quad Cities plants and the conser vative application of these loads, the potential non-conservatism on the accele ration volumes is adequately balanced by the conservatisms in the interference corrections and the load application.

On the basis of these comparisons we conclude that while the direct use of "published" acceleration volumes for sharp edge structures may not 1n general lead to conservative loads, the PUAR methodology for the application of these loads to the relevant structures, has sufficient conservatism to bound any hy drodynamically produced stresses that could arise in these structures.

3.2 CO and Post-Chug Harmonic Phasing The OBA condensation oscillation and the post-chug load definitions on the torus shell and on submerged structures, accepted in the NUREG-0661, were based on data from a series of blowdowns in the FSTF facility (NEDE-24539), subject to additional confirmatory tests reported in the General Electric Letter Report Ml-LR-81-01 of April 1981.

The condensation oscillation load definition as described in NED0-21888 is based on taking the absolute sum of 1 Hertz components of a spectrum from Oto 50 Hz. Three alternative spectra are to be calculated with the one producing maximum response used for load definition. The procedure was found acceptable in the supplement to the SER (NUREG-0661), because the demonstrated high degree of conservatism associated with the direct su1T1nation of the Fourier components of the spectrum was sufficient to compensate for any uncertainties concomitant with the data available. The post-chug load definition is based on bounding FSTF chugging data but otherwise follows similar procedures to those used in the J, CO load definition.

The PUAR uses a factor of .65 to multiply the CO and post-chug loads com puted on the basis of the absolute sum of the harmonic components. The justifi cation is based on comparisons of measured and predicted stresses in the FSTF facility using statistical studies of different phasing models (References 8, 9, 10, 11). The factor .65 is chosen to give 841 non-exceedance probability with a confidence level of 90\. The PUAR does use an additional spectrum, Alternate 4, for the CO loading, based on test Ml2 from the supplementary tests reported in the letter report Ml-LR-81-01. The information in Table 1-4.1-4 of the PUAR provides additional justification to show that the computed loads (using the .65 factor and Alternates 1 through 3) bound the measured stresses at critical points in the FSTF facility by 11\ for axial shell stress to 691 for column force. The use of Alternate 4 in the Quad Cities plants provides an additional conservatism of about 20% to the shell response. The use of random phasing 1n the time domain for TAP (Volumes 6 and 7 of the PUAR) coupled with a factor of 1.3 for alternates 1, 2, and 3 and a factor 1.15 for alternate 4 1s consistent with the results of Reference 10 and conservatively bounds all FSTF data.

The procedures are a conservative application of the phasing design rules evaluated in Reference 12 and are therefore found acceptable.

3.3 FSI Methodology for CO and Chugging Drag Loads A detailed discussion of the method used to account for FSI effects on con densation oscillation and chugging submerged structure loads 1s provifed in Ref erence 13. The methodology described in this note is used to compute accelera tion fields across a submerged structure anywhere in the torus resulting from rSI, based on knowing the torus boundary acceleration. The method is presented as an alternative to the NRC Acceptance Criteria suggestion of adding the bound ary accelerations directly to the local fluid acceleration to account for FSI effects since the latter is deemed too conservative.

The review of the method outlined in Reference 13 has shown it to be rea sonable and acceptable. The equations derived for fluid accelerations and pres sure fields are plausible approximations for the conditions prevailing in the suppression pool. Assumed boundary conditions including the dr1 .,g one at the torus wall are suitable. Overall trends as well as the acceleration fields de picted in the selected results appear reasonable. Therefore, the alternate pro cedure used to account for FSI effects on submerged structures is considered ac ceptable in this application.

3.4 Calibration of SRV Drag Loads Based on In-Plant Tests The staff requested clarification of the detailed procedures used to derive the calibration factors from in-plant tests for SRV submerged-structure loads.

On the basis of this response, as 11 as those* provided in other PUAR reviews of NUTECH plants, the staff considers the procedures as an acceptable modifica tion of the AC.

lne SRV bubble pressure data from llt>nt1cello tests 1s shown to be bounded using a bounding factor of 1.75 instead of the 2.5 specified 1n the AC. In the Quad c;ties plants, dynamic load factors are derived on the basis of Dresden in-plant tests. (Dresden 1s very similar to Quad C1t1es 1n all essential param eters.) A bounding DLF value of 2.5 is then used for all submerged structures.

lne staff considers these procedures to be a reasonable app11cat1on of the in-plant test results, and considers any potential uncertainties associated with the limited data base to be bounded by other conservatisms associated with the design load calculation procedures.

4. CONCLUSIONS A post-implementation pool dynamic load audit of the Quad Cities PUAR has been completed to verify compliance with the generic acceptance criteria of NUREG-0661. Four major differences between the PUAR and the AC were identified along with some other minor issues needing additional clarification. Based on additional information supplied by the applicant, as detailed 1n the previous section, a11 of these issues were resolved. The review of the Quad Cities PUAR has been completed with no issues or concerns outstanding.

• 7

5. REFERENCES References cited in this report are available as follows:

Those items marked with one asterisk (*) are available in the NRC Public Document Room for nspection; they may be copied for a fee.

Material marked with two asterisks (**) is not publicly available because it contains proprietary information; however, a nonproprietary version is avail able 1n the NRC Public Document Room for inspection and may be copied for a fee.

Those reference items marked with three asterisks (***) are available for purchase from the NRC/GPO Sales Program, u. s. Nuclear Regulatory Coflllliss1on, Washington, o. c. 20555, and/or the National Technical Information Service, Springfield, Virginia 22161.

All other material referenced is in the open literture and is available through public technical libraries.

(1) "Safety Evaluation Report, Mark I Long Term Program, Resolution of Generic Technical Activity A-7 NUREG-0661, July 19JO.***

11 (2) "Mark I Containment Short-Term Program Safety Evaluation Report",

NUREG-0408, December 1977.***

(3) General Electric Company, "Mark I Containment Program Load Definition Re port", General Electric Topical Report NED0-21888, Revision 2, November 1981.*

(4) Mark I Owners Group, MMark I Containment Program Structural Acceptance Cri teria Plant-Unique Analysis Applications Guide, Task Number 3.1.3", General Electric Topical Report NED0-24583, Revision 1, July 1979.*

(5) "Quad Cities Nuclear Power Station Units 1 and 2 Plant-Unique Analysis Re port", Vols. 1-7, Prepared for Co1T1t1on\l!ealth Edison Company by NUTECH Engi neers, Inc., May 1983. **

(6) Attachment to Letter from J. R. Lehner, BNL to F. Eltawila, NRC,

Subject:

Dresden Units 2 & 3 and Quad Cities Units 1 & 2 Request For Information, June 27, 1984.

(7) R. Rybak to H. Denton letter dated August 24, 1984, "Response to Questions concerning Mark I Conta;nment Plant Unique Analysis, NRC Docket Nos. 50-237/249 and 50-254/265". .'

(8) General Electr;c Company, MMark I Containment Program, Evaluation of Har monic Phasing for Mark I Torus Shell Condensation Oscillation Loads M ,

NEDE-24840, prepared for GE by Structural Mechanics Associates, October 1980.

(9) NEvaluation of FSTF Tests Ml2 and MllB Condensation Loads and Responsesu ,

SMA12101.04-R001D, prepared by Structural Mechanics Associates for General Electric Company, 1982.

• I (10) R. P. Kennedy, "Response Factors Appropriate for Use with CO Harmonic Re sponse Combination Design Rules, SMA12101.04-R002D, prepared by Structural 11 Mechanics AssociatJs for General Electric Company, March 1982.

(11) R. P. Kennedy, "A Statistical Basis for Load Factors Appropriate for Use with CO Harmonic Response Combination Design Rules,u SMA 12101.04-R003D, prepared by Structural Mechanics Associates for General Electric Company, March 1982.

(12) G. Bienkowski, wRev1ew of the Validity of Random Phasing Rules as Applied to CO Torus Loads", Internal BNL Memo, August 1983.

(13) A. J. Bilanin, "Mark I Methodology for FSI Induced Submerged Structure Fluid Acceleration Drag Loads", Continuum Dynamics Tech. Note No. 82-15, June 1982.*

1 ENCLOSURE 2

• SALP -

prepared by the Contah*nt Systeas Branch Ev1lU1tton QUf\O CITIES Crtteri1 Category Narrative Description

1. NINgeaent lnvolWJll!nt 2 IMnage..ent took positive steps to assume tily resolution of the I

issue.

t

., a --a: ,_ a:::--*--**-- I I ------ .

2 Sound Technical Understanding of the issue. Worked closely with the staff *nd its consultant toward resolution of the issue.

espons 1veness Met with the staff and Its cons utant shortly after receiving the RAJ.

2 n

. forceaent History N/A wnts N/A N/A r11n1ng