ML20140F788

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Forwards Draft Generic Ltr Requesting Applicants of Plants W/Mark II Type Containments to Describe Design Provisions to Accommodate Loads in Suppression Pool Due to Operation of Primary Sys Pressure Relief Valves
ML20140F788
Person / Time
Site: Nine Mile Point, Susquehanna, Columbia, Limerick, LaSalle, 05000000, Zimmer, 05000037, Shoreham, Bailly
Issue date: 04/10/1975
From: Tedesco R
Office of Nuclear Reactor Regulation
To: Deyoung R, Moore V
Office of Nuclear Reactor Regulation
Shared Package
ML20140F372 List: ... further results
References
FOIA-85-665 NUDOCS 8604010163
Download: ML20140F788 (32)
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UNITED STATES NUCLEAR REGULATORY COMMISSION W ASHIN GTON. D. C. 20555 g

R. C. DeYoung, Assistant Director for Light Water Reactors, Group 1, RL V. A. Moore, Assistant Director for Light Wata.r Reactors, Group 2, RL STANDARD LETTER TO APPLICANTS CONCERNING PRIMARY SYSTEM PRESSURE RELIEF VALVE LOADS FOR PLANTS WITH MARK II TYPE CONTAINMENTS Enclosed is a draft of a generic letter requesting applicants of plants with Mark II type containments to describe their design provisions to accommodate loads in the suppression pool resulting from operation of the primary system pressure relief valves. These loads are due to two distinct phenomena. First, pressure waves are generated within the suppression pool when, on first opening, relief valves discharge high pressure air followed by steam into the pool water. These are referred to as the steam vent clearing loads. Second, steam quenching vibrations can accompany extended relief valve discharge into the pool if the pool water is at an elevated temperature. ,

We have maintained periodic contact with GE on a generic basis regarding their progress in resolving these problems. Specific reviews of plants with Mark I type containments are being coordinated by Operating Reactors (OR) using a standard letter transmittal to each licensee. GE has stated in a letter to R. L. Iedesco from I. F. Stuart, dated March 10, 1975, that it has no contractual involvement in either the design or supply of Mark II containments in the U. S. We consider it necessary that a standard letter be transmitted to all Mark II applicants requesting information on loads due to relief valve actuation which may not have been fully considered at the time that CP's were issued. It is recommended that each applicant be informed of this outstanding issue and that their documentation be completed in a timely manner. The inference in the GE letter would indicate that this category of plants is proceeding on the basis of relief valve discharge line designs and load analyses done by their respective architect /

engineers. This precludes a strictly generic review in this area and we therefore recommend that a standard letter, of the type enclosed, should be sent to the applicants for plants with Mark II type containments as listed below:

AFFECTED MARK II PLANTS Shoreham Bailly, Unit 1 Zim=er La Salle Hanford 2 Susquehanna, Units 1 & 2 e

Limerick, Units 1 & 2 Mine Mile Point 2

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7 9604010163 860114 PDR FOIA hb l

FIRESTOB5-665 PD9 l ,

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R. C. DeYoung V. A. Moore APR 1 0 1975 The review of relief valve loads for each plant covered by the standard letter should be initiated by a TAR from DRL to DTR and should include CSB, SEB, and MEB. The CSB contact for this effort will be Dr. L. Slegers.

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Robert L. Tedesco, Assistant Director for Containment Safety Division of Technical Review

Enclosure:

As stated cc: E. Case F. Schroeder A. Giambusso S. Hanauer R. Boyd t

R. Maccary G. Lainas J. Glynn J. Kudrick J. Shapaker L. Slegers R. Cudlin C. Grimes C. Anderson K. Goller RL B/C's 4

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SAMPLE LETTER TO APPLICANT WITH MARK II CONTAIhW NT Gentlemen:

An ongoing review area for BWR plants with pressure suppression-type containments has been the capability of the suppression pool retaining structures to tolerate loads due to operation of the primary system pressure relief valves. Experience at several operating BWR plants has indicated that loads due'to relief valve actuation may not have been fully con-sidered in the structural design of the suppression chamber. In addition, the General Electric Company is now preparing to start a series of small-scale relief valve tests which will be used to verify analytical predictions

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of tnese loads as applicable to all classes of plants.

Pool dynamic loads due to relief valve operation are due to two distinct phenomena. First, pressure waves are generated within the suppression pool when, on first opening, relief valves discharg'e high pressure air followed by steam into the pool water. These are referred to as steam vent clearing loads. Second, steam quenching vibrations can accompany extended relief valve discharge into the pool if the water is at an elevated temperature.

Enclosed are specific requests for information pertaining to these effects which we will require to evaluate your design with regard to these phenomena.

We believe that these developments warrant further consideration of your particular design of the Mark II containment at this time. Therefore, we are requesting that you provide us with the status of your design and planned course of action to ensure that your design will reflect the latest available information. This issue, which was not fully considered at the time that the CP was issued should be resolved prior to issuance of an OL.

The attached list of questions is provided for the preparation of your response.

You should provide us within 30 days after receipt of this letter a schedule for submittal of the requested information. Please contact us if you desire additional discussion or clarification of the material requested.

Enclosure:

Request for Additional Information

REQUEST FOR ADDITIONAL INFORMATION RELIEF VALVE LOADS

1. Specify the number of safety relief valves, their design flow rate, I

and discharge line size. Provide a listing of operating conditions l

under which these valves would be operated either manually or auto-matically. Describe, with the aid of drawings , the routing of the discharge line to, and orientation in, the suppression pool, and the design of the discharge line exit.

2. Provide the load specification for the suppression chamber structure to accommodate actuation of one or more safety relief valves.
3. Provide the design load capability for the suppression chamber e

structure.

4. Provide justification for the load specification given in (2) above by the use of appropriate experimental data and analysis. If the General Electric (GE) Company is responsible" for specifying these loads, a statement to that effect is sufficient.
5. Identify, with the aid of drawings, any components or structures in the suppression pool region, other than the bounding walls of the suppression chamber, and the location of such components relative to the relief valve discharge line exits. Discuss the structural capability of these components to accommodate loads due to relief valve actuation.
6. Estimate the maxieum number of single and multiple relief valve openings j over the life of your plant.

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Request for Additional Information 1 i

7. Identify the maximum temperature limits of the suppression pool with the reactor at power. This temperature limit should include provisions for the testing requirements of relief valves.
8. Specify the operator actions that are planned when specified temperature limits are exceeded.
9. Present the temperature transient of the suppression pool starting i

from the specified limits in (1) for the following transients:

(a) main steam line isolation; (b) semi-automatic blowdown; and, (c) stuck open relief valve.

For purposes of this analysis, the minimum water level should be assumed in the suppression pool.

10. The temperature instrumentation that will be installed in the pool and the sampling or averaging technique that will be applied to arrive -

at a definitive pool temperature.

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SUSQUSHANUA SSS CONTAIUMENT MS/RV LOAD DEFINITICE PROGRAM Meeting Agenda Date:

Place: US:!RC Offices Bethesda, Maryland Attendees: USURC, Kraftverk Union, Bechtel, Stanford Research Institute International, PP&L Agenda:

I. Cpeninr.; Remarks E.M. Mead /PP&L A. History of SESS Contain=ent

3. Modificatiens C. Mari: II Progra D. SRI Effert -

E. K fJ Effort F. DAR II. SRI Preser.tation A. Overview B. ME/RV Load Mitigetion Assescment Dr. G. R. Abrahamson/ SRI Dr. G. R.

  • brahntzen/ SRI C. Bubble Physics E:: peri =ents Dr. T. J. Connoll; /33.I Dr. A. T. Leonard / SRI Dr. G. R. Abrahausc:./SSI III. Eechtel Assessment J. A. Weyandt/Bechtel A. D sigr. Margin Maximization E. C.uencher C. Suoquehennu Pregra:

IV. MS/RV Lead Deff rition Dr. Becker/K~..'U A. Szisting Detc Buce Dr. Koch /KWU-E. Susqwehr,nna 1:ni ,te 'suencher

(%rr.21 PerCerr:ance)

C. Design Mothelolou D. Second Pop E. Verilicction Fro; ram Dr. Simon 2/L.'U l (Kari Lain 'Iest) l

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Method used to Calculate Pool

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' ' t A. Overview J. A. Weyandt/Bechtel I i

~ .B . . Compute.-m,Qele Dr. B. Ilarpor/Bechtel  ;

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Dr. H. Safwat/Bechtel C. Application VI. SSSS quencher Design J.A.Weyandt/Bechtel f A. Codes B. Hold Down (Slip Joint) f i

VII. Wrap Up E.M. Mead /PP&L

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VIII. Cor.ents USNRC  ;

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SUSQUEIMNI A SES CONTAIIHGNT MS/RV LOAD DEFI!!ITION PROGRAM PP&L 03JI:CTIES

, 1. Overall explaration of entire FP&L Containment Program including Mark II interface and scheduloc.

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a. Review or past hictory to arrive at present ' course.
b. Future activities and schedule.

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2. PP&L will discuss load definitions rather than structural response.

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3 Explain PP&L approach to total design margins.

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1 4 Explain FPLL unique quenchor design.

5 ~;xplain existing analytien1 an:1 test basis to support unique quencher cpproach.

j 6. Explain PP&L unique verification program at Karlstein.

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i 7 Explain St .r.feri Resecrch International bubble physics experimants.

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1. Cc=ents frcm NRC on Overall program and schedule.

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2. Cc=:r.ts freir ;aC on the Karlstein verification program for
Susquehanr.e SES uniquesuencher.

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1. Give proprietsr/ dccu.c.
ntatien to the IRC that is supportive to

+he Susquehsr.r.3 SES unique quencher design.

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SRV AND LOCA LOADS

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STRUCTURAL ANALYSIS FLOW CHART ~

FORCING FUNCTION

. PRES 3URE T.H. I SRV OR LOCA

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l NOOAL FORCING FUNCTION i

. IH.FOR CONTAINMENTANADSIS -

, ,(ANSYS) FINITE ELEMENT PROGRAM

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1 CONTAINMENT DISE RESR _

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(CECAP) (POSTPROCESSOR) ** I 4 +

N.BAR Sit 6WA CONTAIMENT ACCEL

ON CONC. SECTIONS RESP T.H.

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i r DESIGN ASSESSMENT l AND l MARGIN FACTORS -

(CE 920) ,

(CE 92I ) (CE 789) ,. (CE 789)

(CE918 ) MSPEC MSPEC 1r i r 3 r CONTAINMENT RESR RESP. SPECTRA FOR REACTOR BLDG EVALUkTION S CMA FOR MOUE GE.,NSSS l PIPING & EQUIP., E TC. EVALUATION I l .o A COMPUTER PROGRAM TO CONVERT PRESSURES TO FORCE $

o* A COMPUTER PROGRAM TO CALCULATE ACCELERATIONS FROM l DISPLACEMENTS .

,- 1 Name: Dr. Manfred Becker Present Position: Senior Supervising Engineer Kraftwerk Union  ;

Berliner Strasse 295-299 l D-6050 offenback (Main) FRG l Education: Technical University Aachen _

Post-Graduate work and Ph.D.

Institut fur Luft-und Raumfahrt

. Technische Hochschule Aachen (Institute fo'r Aeronautics and Space Research, Technical University Aachen)

Experience: 1972 - Present - Kraftwerk Union ,

1965-1972 Assistant at Institute for Aeronautics and Space Research, Technical University 1958-1 % 5 Student at Technical University Aachen 4

Name: Jack A. Weyandt Present Position: Assistant Project Engineer Susquehanna Proj ect Bechtel San Francisco Power Division Education: BSME - University of California - Berkelery General: Registered Professional Engineer - State of California Experience: 1965 - Present - Bechtel Corporation 1951 - 1965 - Kaiser Engineers O

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Name: B. E. Harper Present Position: Head of Mechanical Analysis Groups Mechanical / Nuclear /PlantDesignStaffs

- Bechtel San Francisco Power Division Education: Ph.D. University of California M.S. Mass. In?titu*ve of Technology B.S. University of Mississippi General: Registered Professional Engineer - California Publications with American Nuclear Society and American Society of Mechanical Engineers Experience: Bechtel Power Corporation 1970 - Present.

Atomics International 1958-1970 MIT Research Laboratories 1954-1958 U.S. Navy - Electronics Officer 1951-195h e

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l Name: Girish Shah Education: Received M.S. and Engineer's degrees from Stanford University in 1965 and 1966.

General: Has Brofessional Licenses in states of California and Pennsylvania

-Experience: Joined Bechtel in June 1965. Has worked on Palisades, Turkey Point, Point Beach, Arkansas and Quanicessee Nuclear Plants. Deeply involved in the design of of Prestressed PWR Containment.

Present Position: Civil Group Supervisor on SSES project p

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Name: Wm. L. Penn Education: BS, Mechanical Engineering, U.S. Naval Academy, JD, University of San Francisco Experience: 1965 - 68 Engineering Department officer of a nuclear powered submarine 1968 - 70 Chief Engineer of a nuclear powered submarine 1970 - 72 Special Assistant to Commander-in-Chief Atlantic for undersea warf are, nuclear powered (submarines),

special weapons , and intelligence 1972 ,- 77 In charge of Bechtel Standard Safety Analysis Repori (BESSAR)

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Name: George Abrahamson Present Position: SRI International Director, Poulter Laboratory l Education: Stanford University Ph.D. Engineering Mechanics, 1958 General: Member, ANS, ASME Experience: 1953 to present - SRI International 1958 - 1968 - Head, Engineering Mechanical Group 1%9 - present - Director, Poulter Laboratory Poulter Laboratory is a group of 70 persons specializing in theoretical and experimental research in explosions, shock waves, and structural responses.

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e Name* A. Leonar.'

Present Position: Research Scientist in computational fluid dynamics, NASA Ames Research Center, Moffett Field, California Consulting Associate Professor, Mechanical Engineering, Stanford University Consultant, SRI International and Nuclear Energy Division, General Electric Company Education: California Institute of Technology, B.S. Mechanical Engineering Stanford University, M.S. Engineering Science, Ph.D. Nuclear Engineering General: Registered Professional Engineer (Nuclear), State of California Member, American Physical Society, Society for Industrial and Applied Mathematics Experience: 1963-66 RAND Corporation, Santa Monica, Cal., Technical Staff 1966-73 Assistant and Associate Professor in Nuclear and Mechanical Engineering, Stanford University ,

1973-Present NASA Ames Additional Consulting: RAND Corporation, Calspan Corporation, Buffalo, N.Y.

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Name: T. J. Connolly Present Position: Professor and Associate Chairman Mechanical Engineering Department Stanford University Stanford, CA 9h305 Education: B.Ch.E. Syracuse University M.S.Ch.E. Carnegie Tech.

PhD, Ch.E. California Institute Tech.

General: Registered Professional Engineer, State of California Member, American Nuclear Society Experience: Engineering Faculty, LULA 1950-59 Engineering Faculty, Stanford, 1959 - Present One year leaves at:

Atomics International Joint Establishment for Nuclear Energy Research (Norway)

Kernforschungszentrum Karlsruhe l

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e Name: E. M. Mead Present Position: Project Engineering Manager - Susquehanna Steam Electric Station, Pennsylvania Power & Light Co. ,

2 North 9th Street, Allentown, Pennsylvania Education: Bucknell University B.S. in Electrical Engineering General: Registered Professional Engineer, Commonwealth of Pennsylvania Member, American Nuclear Society Member, American Society of Mechanical Engineers Member, Institute of Electrical and Electronic Engineers Experience: 1963 to Present - PP&L Co. - Nuclear related work 1952 to 1%3 - PI&L Co.

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l Name: H. W.' Holland '

Present Position: Nuclear Mechanical Project Engineer - Susquehanna Steam Electric Station, Pennsylvania Power & Light Co. , 2 North 9th Street, Allentown, Pennsylvania Education: B.S. ME Drexel University l M.S. ME Stanford University l General: Registered Professional Engineer, State of Pennsylvania Member, American Nuclear Society Member, American Society of Mechanical Engineers Member, American Welding Society Experience: 1972 to present - PP&L Co. - Susquehanna Steam Electric Station Design 1971 - Mobil oil corporation - New Fefinery construction Co-ordination l

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J' Name: Dietrich Gobel Present Position: Kraftwerk Union Offenbach (West Germany)

Berliner Str. 297-299 Education: Technische Hochschule Darmstadt Allgemeiner Maschinenbau Experience: Employed by Kraftwerk Union Since April 1972 h

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4 History A. SRV stuck open during startup in April lW2

1. Suppression pool water beated to 170 F in 30 minutes.
2. Large pulsating forces developed.

3 30 sq. in. leak in bottom 12ner plate. .

I B. NRC issued Bulletin 74-14 to all BWR Operators November 14, Igrf4, alerting them to the "Wurgassen effect."

C. CE identified several dynamic loading conditions in January,1975, one of which was SRV discharge thermo-hydrodynamic phenomeca, concerning Mark II design criteria.

D. PP&L altered Containment construction sequence to ascertain the effect of the new phenomena.

E. ' Disk force formed with PE/Bechtel/PP&L in March, lW5 F. Part of the brute force approach, the following civil-structural modifications were incorporated into the Susquehanna in May, Igrf5

1. Added 60 vertical reinforcing bars in suppression chamber (each containment).
2. Added embedments and anchor bolts in suppression chamber walls and Diaphragm Slab.

3 Diaphragm Slab reinforcement changed from 45 to 90 to increase uplift loadings.

G. In June, Igrf5, the Mark II Owners Group were fomed to define suppression pool dynamic loads and find ways of mitigating them.

H. A generic Dynamic Forcing Function Infcrmation Report (Rev.1) was jointly issued by GE/Sargent & Lundy/ Mark II Owners to the NRC on March 15, 1976.

I. More supportive data issued in DFFR Rev. 2 on September 1,1976.

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  • TE PROBLEM
1. It is our opinion, that the greatest load amplitudes for containment are from the SRV's.
2. Mark II program has an SRV mitigating device called a quencher.

3 Bere is a testing program task (s) for this mitigating device.

4. We felt these program tasks have 'not been timely for our construction schedule.

5 We were concerned that some testing is not directly within Mark II control.

6. Mark II lead plants presently need all the resources available to ccurplete their analysis, testing and licensing. This is for completion of the short term program. Socn the resources will be turned toward the long term program. -

7 On April 21,19TT these items were informally discussed with the NRC.

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OUR APPROACH  ;

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1. All quencher designs are based on KWU data. .

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2. PP8d. has purchased from KUU a quencher design for SSES.  !

3 PP&L has contracted with KWU for additional verification testing of their device to provide additional design margin verification.

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.-____L____.--.___- .__- - _ _____.  : - . - - _ _ _ . _ - - _ - .

- - - - ..o a (Continued) l NWU 'fESTS AND REPORTS l

Antivity Doctamentation Stdtus

1. Fometion an'd Oscillatica of a sperical pag bubble AEG - Report No. 2241 Complete 1

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2. Analytiqal model for clarification
of pressum gulaation in the .

westve *1 aftqr , vent clearing AEG - Report No. 2208 Complete l

3 Testa on mix d cod.ensetton with

! motel quenchers K'WU - Report No. 2593 ,Complate

k. CondensationYand vent clearing ,

testa at GKMjwith guenchere KWU - Report No. 2594

Complate
5. Concept and (esign of the pressum relief system with quenchors KWU - Report No. 2703 Complete
6. KKB vent clearing with quencher KWU - Report No. 2796 Complete '

, 7 Tests on con (ensation with j quenchers whdn subestgence of l quencher ams is shallow ,

Kv0 - Report No. 2840 4

I Complete 8.

j KKB - Concept' and task of passure ~

j relief systes, KWU - Report No. 2071 Complate l 9. Experimenta1'opproach to vont

  • j clearing in a' model tank KWU - Report No. 3129 .

i Complete j 10. XXB.- Specification of blowdown j

tests during non-nuclear hot ~

functional test - Rev. I dated l Oct. 4, 19716

  • KWU/V822 Report, Complete

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(Continued)

KW 'IESTS AND REPQR1B .

Astivity

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PDocumentation Sitatus l 11. Anticipate data for blowdown i

tests with ressure relief system -

j during the on-nuclear hot functions 1

^ te st a t ny ear power station Brune- -

buttel aKW - Lport No. 3141 C9mplete

12. Hesulte of $he non-nuclesr hot functions 1! tests with the pressure relief systda i in the nucigar power j station Brunshuttal kW . Deport No. 3267 Complete

! 13. Analysis of.'the loads measund on j the pressure relier system during the non-nuclear hot functional test at KKB, 'KW - Deport No. 3346 Complete l 14. KKB - Listi[ng of test parameters and important test data of the non-l uuclear hot

  • functional tests with

! the pressure relief pystem IKW - Working Report i a521/kO/77 c9mplete

!; 15. MKa - Spect'fication or additional i tests fo testing of the pressure

! relief valves during the nuclear i start-up Reir. 'l JKW/V 822 in, l 'Not Numbered Complete .

i 1 16. KKB .- Results from nuclear start-up ,

i testing of pmssure relief system 'KW - Working poport j ;R Ik2-136/76 cmplete 1

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(Continued) j KW 'IESTS AND REPORTS Activity Documentation Status 17 Nuclear Power Station Phillipsburg L -

< Unit 1 Not Functions 1 1 Inst -

l Specificapion af pmssure relief valve.

1 tests as,pl1 as emergency cooling and"

] wetwell cpoling systema KWU/V822/RF'13,.

Not Numbered ' Complete

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18. Results of the non-nuclear hot ,
fimetionag tests with the pressure j relief sypten in the nuclear power

, station Phillipsburg , KWU'- Working Report-I R 182-38/77 6 Complate

19 KKBI - Listing of test parameters envi important test data of the non-
nuclear hot functional tests with the i pressurer3elief system KWU - Working Report l R521/1s1/77 ,Completo I

Air oscinat' ions during vent clearing i 20.

with single pad double pipes ' AEG - Report No. 2327 Complete l

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. . SUSSIE!! ANNA STEAM EI2CTRIC STA9"05 Conte'nts of DAR Submittal to NRC 1.0- General Information 1.1 Purpose of Report January,1W8 1.2 History of Problem _ January,1978 13 Plant Description January, 1W8 1.4 Quencher Discharge Device January,1W8 2.0 Summary 30 Transient Description 31 safety / Relief valve (s/av) Discharge January,1W8 32 Loss-of-Coolant Accident (LOCA) January,1W8 4.0 Lead Definition 4.1 S/RVLoads 4.1.1 Physical Process Description '

January, lW8 4.1.2 Parametric Effects Januaryy 1978 4.1 3 sss quencher Design Parameters January,1W8 4.1.4 Maximum Backpressure January,1W8 k.l.5 Vent clearing Pressure January, lW8 4.1.6 Loads on Pool Boundaries and -

t Submerged Structures January, lW8 k.l.7 Costaiment Boundary Loads January, 1978 l s 4.1.8 h rged Structures January, lW8 4.1 9 Quencher Load Specification January, 19T8 4.1.10 Subsequent Actuation (Second Pop) January, 1978 4.1.11 Dermal Performance January, 1978 4.2 LOCILeads

'4.2.1 LdCA Loads Associated with Pool Svell January,1W8 4.2.2 Oondensation Oscillation and Chugging Loads January,1W8 4.2 3 Long Tem IOCA Loads Jursary,1978 4.2.4 IOCA Loading Histories for -

SSES Contaiment Caponenta January,1978 O

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contents'of DAR (Continued) Submittal to NRC i

- 4 3 An=iln= Pressurization January, 1978 4.4 KWU Testing Program - January,1978 4 5 Mark II Owners supporting Program January,1W8

~ 50 Design Assessment

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51 Leeds -

. January,1W8 52 Design Load ccabinat' ions '

5 2.1 containment 5 2.2 34 1978 Structural Steel 34 1978 523 Liner Plate 34 17T8 5 2.4 Downcomer Pipes 525 Quencher Support 34 1W8 5 2.6 Piping Systems 34 1978 527 3419T8 NSSS (RPV, Recir'c. Line & Pumps) 3q 19T8 5 2.8 squipment e

34 1978 53 camponents Evaluated 531 containment structure

. 532 Liner Plate ' 3419T8 533 Downecaer Pipes 34 1978 534 Quencher support 34 1978 535 Piping Systems 3419T8 536 sSas 3417T8 537 Equipment 34 1978 34 17T8 -

5h Beactor malaing Evaluation 34 1978 6.0 Design Methodology and Criteria 6.1 concrete structures January, 1978 6.2 Steel Structures _

January,17T8 63 Liner ' Plate -

January, 1978 6.4 Downecaer Pipes January',1978 l 6 5 Quencher Support '

1 January, 1978  !

6.6 Piping Systems -

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, Centents of DAR (Continued) Sahnittal to NRC 67 NSSS s 34 1978 I .8 6 squipment 3q1W8 6 9 Reactor an11 ding January /341978*

70 Analysis Methodology i

T.1 Load Definition January,1W8 T.2 containment .

January,1978 T3 Downcomer Pipes Jama ry,1W8 .

T.4 quencher support January, 1978 T.5 Piping 34IW8 T.6 sssS 34IW8 T.T Equipment '

34 1978 T.8 anactor Building January /34IW8 8.0 Margins Evaluation , 3qIW8 90 ccurputer Program Verification 7 7 /34 1978 10.0 References January /3R IW8 11.0 NRC Questions and Answers January /34IWIS i

  • January /34 IW8 signifies partial complation for the first .

submittal (In the esse af the reaetor bnilaing, the structure analysis will be c6 before the systems analysis).

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REASONS.FOR GOING TO QUENCHER v %,.-s .'? .

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g 1 L FOR MSRV DISCHARGE

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- S.l_OWER t LOADS ON CONTAINMENT, STRUCTURES, PIPING & EQUIP g -

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9 INCREASED DESIGN MARGINS AND REDUCED BACKRi ilN.G' .

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SINCREASED POOL OPERATING TEMPERATURE '

n, l . FEWER OPERATION pESTRICTIONS '

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,  % .1 S LINEH PLATE STIFFENING NOT REQUIRED i .

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j G MARK 111 PLANTS USING QUENCHERS _ -

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8 ADEQUATE BACKGRbUND FOR QUENCHER DESIGN , s s

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9 LOWER "SECOND POP" LOADS .

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, G CONTROL SCHEDULE IS BETTER ,

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