ML20065A752

Forwards Assessment of Unresolved Safety Issue Task A-46 & Final Draft Responses to Selected Questions of 820825,1115 & 1231 Commission Info Requests

ML20065A752
Person / Time
Site:	05000447
Issue date:	02/17/1983
From:	Sherwood G GENERAL ELECTRIC CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
References
REF-GTECI-A-46, REF-GTECI-SC, TASK-A-46, TASK-OR JNF-009-83, JNF-9-83, MFN-035-83, MFN-35-83, NUDOCS 8302220152
Download: ML20065A752 (48)

Text

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GENhR AL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY,175 CURTNER AVE., SAN JOSE, CALIFORNIA 95125 MFN 035-83 MC 682 JNF 009-83 (408) 925-5040 i

February 17, 1983 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Attention: Mr. D.G. Eisenhut Division of Licensing Gentlemen:

SUBJECT:

IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II)

DOCKET NO. STN 50-447 ASSESSMENT OF UNRESOLVED SAFETY ISSUE (USI) TASK A-6 AND REVISED DRAFT RESPONSES Attached please find an assessment of USI Task A-6 and final draft responses to selected questions of the Commission's August 25, 1982, November 15, 1982 and December 31, 1982 information requests. Only modifications (new or revised) to the response of the referenced letters are provided.

Sincerely, Glenn.G. Sherwood, Manager Nuclear Safety & Licensing Operation Attachments cc: F.J. Miraglia (w/o attachments) C.0. Thomas (w/o attachments)

D.C. Scaletti L.S. Gifford (w/o attachments) 8302220152 830217 PDR ADOCK 05000447 A PDR

LISTING OF ATTACHMENTS PROVIDED Attachment hmber Subject 1 Assessment of USI Task A-46 2 Draft Responses-to Procedures and Test Review Branch Questions 3 Draft Responses to Effluent Treatment Systems Branch Questions 4 Draft Responses to Power System Branch Questions 5 Draft Responses to Reactor Systems Branch Questions 6 Draft Responses to Equipment Qualification Branch Questions

.' ATTACHMENT NO. 1 ASSESSMENT OF UNRESOLVED SAFETY ISSUE TASK A-46

4 GESSAR II 22A7007 238 NUCLEAR ISLAND REV.,4},

APPENDIX 1B CONTENTS (Continued)

Section Title Page 1B.2.11 Hydrogen Control Measures and Effects of IB-27 Hydrogen Burns on Safety Equipment (Task A-48)

! 1B.2.11.1 Issue Description 1B.2.11.2 1B-27 NRC Activities 1B.2.11.3 1B-28 Industry Activities and Resolution Status 1B-29 1B.3 CONCLUSION 1B-33 i

TABLES Table Title 1B-1 Unresolved Safety Issues

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'i 1B.l.2 Objective The unresolved safety issues were initially identified in NUREG-0510 (Identification of Unresolved Safety Issues Relating to Nuclear Power Plants", January 1979). These  ;

issues are updated quarterly in NUREG 0606 (" Unresolved Safety Issues Summary"). The quarterly update provides current programmatic and schedule information and includes information relative to the implementation status of each issue for which technical resolution is complete.

The overall objective of this appendix is to comply with the Atomic Safety and Licensing Appeal Board decision (ALAB-444) that the Safety Evaluation Report (SER) for each plant should contain an assessment of each significant unresolved generic safety issue. The assessment should include a summary description of relevant investigative programs and the measures devised for dealing with the issues on the subject plant.

1B.1.3 238 Nuclear Island Applicability The unresolved safety issues outlined in NUREG 0606 include all issues for which technical resolution is not considered complete by the NRC. Several apply only to pressurized water reactors , :: opplies vul, t Operating-auclear-power r '

y- plants, and one applies only to boiling water reactors with a Mark I containment. The remaining unresolved safety _.

issues which are applicable to the 238 Nuclear Island are given in Table 1B-1. The number of the generic task in the NRC program addressing each issue is given along with the section in which each issue is discussed.

i GEII-C 1B-2

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, GESSAR II 22A700 238 NUCLEAR ISLAND REV. ,

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TABLE 1B-1 UNRESOLVED SAFETY ISSUES Applicable Unresolved Safety Issue NRC Task Number Section Waterhammer A-1 1B.2.2 Reactor Vessel Materials A-ll 1B.2.3 Toughness Systems Interaction in Kaclear A-17 1B.2.4 Power Plants Safety Relief Valve Pool A-39 1B.2.5 Dynamic Loads Seismic Design Criteria A-40 1B.2.6 Containment Emergency A-43 1B.2.7 Sump Reliability Station Blackout A-44 1B.2.8 Shutdown Decay Heat A-45 1B.2.9 Removal Requirements 7 Safety Implications of A-47 1B.2.10 ~~

Control Systems i Hydrogen Control Measures and A-48 1B.2.ll Effects of Hydrogen Burns on Safety Equipment Se u w e. Q a d [% on ck A-46 l B. 2.12 Gg-s g # ksnGpus '^j

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fA L 1B.2.11.3 Industry Activities and Resolution Status ,

(Continued)

With respect to Task A-48, it is concluded that the 238 Nuclear Island can be operated, with no additional hydrogen control systems, without undue risk to the health and safety of the public.

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o pnce;8 \ B. 2. I2 SEtsmc QUAuF1cArrou of Eaus P/1ENT I/d OPERArtuG )LANrs [] ASit b-VC) 13.2.IZ.l Issur DESCRIPT/o/V The design criteria and methods for the seismic qualification of mechanical and electrical equipment in nuclear power plants have undergone significant change during the course of the commercial nuclear power program. Consequently, the margins of safety provided in existing equipment to resist seismically induced loads and perform the intended safety functions may vary considerably. The seismic qualification of the equipment in operating plants must, therefore, be reassessed to ensure the ability to bring the plant to a safe shutdown condition when

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subject to a seismic event.

,32.12.2 NRC WM

18) ri-is naen or %sc A % rue ^)RC i s AT7Er$PDAX, to establish an explicit set of guidelines that could be used to judge the adequacy of the seismic qualification of mechanical and electrical equipment at all operating plants in lieu of attempting to backfit current design criteria for new plants. This guidance will concern equipment required to safely shutdown the plant, as well as equipment whose functions is not required for safe shutdown, but whose failure could result in adverse conditions which might impair shutdown functions.

hLSC), T7tE N W WiLL. E S T N lh I, d rl GOiDEL i AJES FoR US E IM REQOn Lt FV/A)G E Q uiPticiar WHosE SEtsts tc. QUAL )FtCA TICAJ l-l A s 8twtG F&Mb To BE IN A DEQ UATE

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& Pai,acent._ oeaee7we or ins < ASL is rc,ca,socc THE ADEQvAC Y OF THE SEISttic. QUA L I Ft C A i'ior)

OF EQuipr1ENT It\1 o PERnri^)G PLANrs WH ICH HAVE Nor BEEb3 bESIC h)E D To c.oRRE^TT' 5'EISt1tc DES /C= A) 2LS NLtt2AR l5LAN/), Hou)LvEk

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STAND ARDS . lHE IS . designed using current seismic design critierta, and methods for seismic equipment qualification are to be latest codes and standards. Requirements for seismic equipment qualification include IEEE 344-1975 and Regulatory Guides 1.92 and 1.100. Standard Review Plans 3.2.2, 3.9.2, qualification 3.9.3, and 3.10 have also been considered in the efforts.

%ce rne 23e w aa tstnnoun,uwv DGStGNED V SudG THE NOST Y P T V) D A TE C RITERIA , THERE. IS A)O NEED Yo DE t tR nioE THE ROEQt)ACV of 77/E SE/SNtz G OA u f/LA T;'O^)

I M L.I E o of 6 ACKF I t I t A)(, CORREA)T DES JG A) CRI TERii ItJ A 0 D i TIO^), SINCE C U R REA)I' CRITERi A (5 /9LREADY vsED FOR THE $EIsNtC Cx uAttfic 19rion, THERE / S A/D EQ ui PttE^>T' k) H O 5& QudL t Fil A TIOA) SHOOL t> BE //w M EQ uA1 YtTH RESPECT To IASK bIb li IS CONCLODED Il1AT THF- 2.38 h)OCiEAR. lSLAAID C A A> BE CPE/' A TE O VJ ITHooT' UA)D UE RtSK To THElEMD4 /wt SMQ Cf 00YlUt

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ATTACHMENT NO. 2 DRAFT RESPONSES TO PROCEDURES AND TEST REVIEW BRANCH QUESTIONS

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5 640.04 Modify Section 14.2.7.3 of your FSAR to indicate the level of (14.2.7) conformance of your intitial test program with the following regulatory - -

guides: (1) Regulatory Guide 1.68.1; (2) Re (3) Regulatory Guide 1.95, Position C.5; (4)gulatory Regulatory Guide Guide 1.68.2; 1.108, Position C.2.a; (5) Regulatory Guide 1.128, Position C.4; (6) Regulatory -

Guide 1.140, Position C.5.

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K Regulatory Guide 1.68.1, Revision 0, Dated December 1975

Title:

Preoperational and Initial Startup Testing of Feedwater and Condensate Systems for Boiling Water Reactor Power Plants This guide describes in detail the type and nature of BWR feedwater and condensate sy' stem tests that are acceptable to the staff.

Evaluation The preoperational and startup testing of the Feedwater and Condensate Systems are not within the GE scope of services) there- ,

fore, this guide is not applicable ,to-GESSAR 4E.

However, cama *4 hapter 14 states that a comprehensive testing program will be developed to ensure that all nuclear safety-related equipment and systems will perform in accordance with their design criteria. As individual systems are completed, they will be tested, reviewed, and approved according to predetermined and written <

procedures. In general, all procedures will be developed in accordance with NRC publications such as Regulatory Guide 1.68.1.

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W Regulatory Guide 1.68.2, Revision 1, Date,d July 1978

Title:

Initial Startup Test Program to Demonstrate Remote Shutdown Capability for Water-Cooled Nuclear Power Plants This guide describes an initial startup test program acceptable to the NRC staff for demonstrating the capability to shut down the hot operating and cold reactor from outside the control room by verifying that i

(1) the nuclear power plant can be safely shut down from out-side the control room; (2) the nuclear power plant can be maintained in the hot shutdown condition from outisde the control room; and (3) the nuclear power plant has the potential for being

, safely cooled from hot and cold shutdown conditions from outside the control room.

Evaluation The Remote Shutdown System is designed with the capability to accomplish the objectives of the test program outlined in the regulatory guide.

The system described in GESSAq ection 7.4 provides remote control for reactor systems needed to carry out the shutdown function from outside the main control room and bring the reactor to cold condi-tion in an orderly fashion.

The system provides a backup variation to the normal system used in the main control room permitting the shutdown of the reactor from outside the control room when feedwater is unavailable and normal heat sinks are lost (turbine and condenser).

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GESSAR II 22A7007 4

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, Regulatory Guide 1.68.2, Revision 1, Dated July 1978 (Continued)

Activation of the relief valves and the Reactor Core Isolation Cooling (RCIC) System will bring the reactor to a hot shutdown condition after scram and isolation. During this phase of shutdown, thesuppressionpo'olwillbecooledasreEkiredbyoperatingthe Residual Heat Removal (RHR) System in the suppression pool cooling mode. Reactor pressure will be controlled and core decay and sensible heat rejected to the suppression pool by releasing steam through the relief valves. Reactor water inventory will be main-tained by the RCIC system. This procedure will cool the reactor and reduce its pressure at a controlled rate until reactor pressure becomes low enough to discontinue RCIC operation.

- % design basis assumptions for the Remote Shuthwn System do not laelude control room damage so massive that all ECCS capabilities are 1**t e in f act , the design assumes all necessary cooldown and

8ikal heat removal equipment remains operational. Paragraph C4.C is interpreted to apply only with the reactor scraused - not t with failure to scran. .

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0

/.% 96 Jh Q4HF Regulatory Guide 1.95, Revision 1, Dated January 1977

Title:

Protection of Nuclear Power Plant Control Room Operators Against an Accidental Chlorine Release i

This, guide describes assumptions acceptable to the Ragn atory staff to be used in assessing the habitability of the control room during and after a ?ostulated external release of chlorine. It also des-cribes requirements for control room isolation and emergency procedures.

Evaluation l

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ww 231 a In the GESSER6 design, chlorine is identified by human detection in cccordance with gParagrap'h C.7 of Regulatory Guide 1.78. For cpecific cases in which a plant would be sufficiently close to a railroad or highway, analysis will be carried out to show whether or not the chlorine limits stated in Regulatory Guide 1.78 will be exceeded due to a postulated accident. If the limit could be exceeded, chlorine detection devices will be placed in the control room and intake ducts. The plan would also include an automatic isolation system. .

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GESSAR II 'aa-238 NUCLEAR ISLAND Rev. 0 f,D,lCb T$$3m6 Regulatory Guide 1.108, Revision 1 Dated August 1977

Title:

Periodic Testing of Diesel GeSarator Units Used as Onsite Electric Power Systems at Nuclear Power Plants This regulatory guide describes diesel generator design features and testing provisions which are acceptable for verification of availability and reliability of standby power sources.

Evaluation The design of the Standby Diesel Generator Systems are in conform-aree with the subject regulatory guide with the following exception to paragraph C.1.b(5): The diesel generator surveillance system is not required to have a first-out annunciation feature because annunciation of individual protective trips give the operator adequate information for correct action.

With regard to the proposed Division 1 and 2 standby power sources, I the staff required GE to conform to the position outlined in Standard Review Plan Appe.7 dix 7 BTP EICSB2, Diesel Generator

• Reliability Qualification Testing. GE has stated its ccmmitment to a qualification program in conformance with this commitment.

v GESSART~ states that readiness of the diesels is of prime importance and will be demonstrated by periodic testing. The testing program will be designed to test the ability to start the ESF system loads cs well as to run under the load long enough to bring all compo-nents of the system into equilibrium conditions. Full functional tests of the automatic control circuitry will be conducted on a pe'riodic basis to demonstrate correct operation.

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. O l.t.ia t fL,3e&& Regulatory Guide 1.128, Revisipn 1, Dated October 1978

Title:

Installation Design and Installation of Large Lead Storage Batteries for Nuclear Power Plants This regulatory guide describes an installation design for class IE batteries that is in general agreement with IEEE-484-1975.

Evaluation The 125-volt de systems are divided into four Class IE divisions.

Each system has de battery, battery charges, and load center dis-tribution panels. These are designed as Class lE equipment in accordance with applicable classes of IEEE Standard 308-1974. The plant design and layout from these de systems will provide physical separation of the equipment, cabling, and instrumentation essential to plant safety. Each system is located in its own ventilated room and all the components are housed in a safety class structure. The battery rooms are independently ventilated to keep the gases pro-duced due to charging of the batteries below an exposure concentration.

Therefore the Class lE battery' installations are in accordance with the Regulatory Guide with the following interpretation of

, Position C.1:

l The area in the immediate vicinity of the battery vent is l excluded from the area defined by the phrase. . . . . "at any location within the battery area."

l Fire detection instrumentation is part of the fire protection system rather than the battery system.

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.3EieT96 Regulatory Guide 1.140, Revisionf, Dated M.=wh -i.da

Title:

Design, Testing, and Maintenance Criteria for Normal Ventilation Exhaust System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants This guide presents methods acceptable to the NRC staff for imple-menting the Commission's regulations in 10CFR50 and in Appendices A and 1 to 10CFR50 with regard to the design, testing, and maintenance criteria for air filtration and adsorption units installed in the normal ventilation exhaust systems of light-water-cooled nuclear power plants. This guide applies only to atmosphere cleanup sys-tems designed to collect airborne radioactive materials during normal plant operation including anticipated operational occur-rences and addresses the atmosphere cleanup systems including the various components and ductwork in the normal operating environ-ment. This guide does not apply to post-accident engineered-safety-feature atmosphere cleanup systems that are designed to mitigate the consequences of postulated accidents. Regulatory Guide 1.52, j Design, Testing, and Maintenance Criteria for Post-Accident Engineered-Safety-Feature Atmosphere Cleanup System Air Filtration and Adsorption Units of Light-Water-Cooled Nuclear Power Plants,

! provides guidance for these systems.

Evaluation The design described in GESSAR ection 9.4.5 makes provision for air filtration and adsorption units in the Containment Purge, Exhaust, and Pressure Control System only. However the need for l

filtration and adsorption units is expected to be determined on a site unique basis. If filtration and adsorption units are installed in the containment Purge, Exhaust, and Pressure Control System, they will comply with this guide. The determination of whether filtration and adsorption systems should be included on normal effluent sys-f tems is not interpreted to be within the scope of this guide.

Therefore, this guide is considered not applicable unless the j determination is made that filtration and adsorption are required.

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, ATTACHMENT NO. 3 DRAFT RESPONSES TO EFFLUENT TREATMENT SYSTEMS BRANCH QUESTIONS 1

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460.11 Provide additional information on the following items for the ESF (6.5.1) filters of the standby gas treatment system (SGTS) and the control --

building: _

a. State whether instrumentation for measuring flow rates through the ESF filter systems will be provided in accordance with Regulatory ' _ . .

( Guide 1.52, Revision 2 (March 1978).

b. Indicate the type of recording device whicn will be provided for recording pertinent pressure drops and flow rates in the control rooms. - -

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460.14 Provide additional information on the following itaas applicable to '

(11.3) the gaseous waste management systems: -

a. Sface your systes description, tables and figures in Chapter 9 of your FSAR do not clearly indicate whether there are provisions for both HEPA and charcoal adsorbers' for the reactor bt11 ding pressure control mode and purge exhaust, provide the appropriate information ___

relating to filter units for the reactor building. -

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exhaust filter units, other than the primary containment purge and exhaust system. All ventilation exhaust have process radiation monitors in ~

-.. the exhaust stream that will detect the release of radioactivity. - -

In event that high level of radioactivity is detected, the ventilation

' exhaust wi 1 automatically be shut off and the Standby Gas Treatment -~

___ System will utomatically actuated to ventilate that area. -

. The above applies to the secondary containment buildings; that is. -

Shield Building Annulus. ECCS/RWCU Pump Rooms of the Ac::iliary

"~ Building and the Fuel Building and the primary containment. -

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, 9.4.3.4 Inspection and Testing Requirements (Continued) filters, f ans and redundant components to assure system avail-ability. The tests include determination of differential pressures and filter efficiencies, control setpoints and signals, alarm func-tioning, modulation valve performance, airflow races, damper func-tioning, airflow switch operation, isolation butterfly valve functioning and thermal performance of heaters and coolers. Test connections are provided for sampling and monitoring the above-noted categories of performance.

The balance of the system is proven operable by its use during operation. Standby equipment can be tested to ensure proper opera-tion on demand. Equipment layout provides easy access for inspec-tion and testing.

9.4.3.5 Instrumentation Application Instrumentation and controls for the Auxiliary Building pressure control systems [ Figure 9.4-3 (K-16 3) ] are designed for automdtic operation. The system fans are started from manual pushbutton stations in the main control room. Airflow failure, sensed by an airflow switch, actuates an alarm, which starts the standby fan and repositions the associated dampers.

from Mt Auxiltav Buddi E C CS Aven Nssov oCedd sb radioactivity. A high Exhaustairgiscontinuouslymontoredfgo level of activity or an ECCS operating signal automatically starts the SGTS, stops the supply and exhaust f ans, closes their asso-ciated dampers, closes the air supply isolation valves and directs the exhaust air to the SGTS.

The ECCS recirculating fan coil cooling units for RHR pump rooms A, B and C, RCIC, HPCS and LPCS pump rooms are interlocked to start when the pump they protect is started. Also, manual override from pushbutton stations in the main control room is provided.

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IT 440.14 hh (Continued)

. 11.5.1.1.2 Systems Required for Plant Operation The radiation monitoring systems (RMS) provided to meet these objectives are:

(1) for gaseous effluent streams -

(a) plant vent discharge, (b) offgas exhaust vent, radwaste building ventilation RMS, and (c) ,

4 (d) turbine building ventilation RMS; (2) for liquid effluent streams -

(a) radwaste effluent RMS and

~

(b) service water effluent to cooling pond RMS; (3) for gaseous process streams -

(a) offgas pretreatment RMS, (b) offgas post-treatment RMS, and (c) carbon bed vault RMS; and (4) for liquid process streams -

(a)

RHR service water system RMS (loops A and B) and (b) closed cooling water RMS.

_ $ A pp vt Cow I NS p ohJbbb 11.5-3

' us. - m 1 238 NUCLEAR ISLAND Riv . )( l '

460l4 CEve 11.s.2.1.4 Auxiliary Building Exhaust Radiation Monitoring 9

This system monit e k the radiation A4Waher levelECCS 4 SmlMwa exterior Ar to the Juxiliar e a MIsuve Us>% %y3eM.

4

?

Building ventilation ed aust d etg Wdt system consists of two redundant. instrument subsystems,, channel A and channel B, which are physically and electrically independent of each other. Each channel consists of a local detector, a converter and a main control room radiation monitor. Power for channel A is supplied from 120-vac RPS Bus E. Power for channel B is supplied from 120-vac RPS Bus F.

Each radiation monitor provides two trip circuits: one for upscale (high) radiation or an inoperative circuit and one for downscale.

The upscale / inoperative trip of channel A initiates opening of the exhaust to the SGTS valve, closing of the exhaust to the plant vent valve, closing of the ECCS corridor exhaust valve, and the closing of the RWCU corridor supply valve for Division 1. The same trip also initiates startup of t'he SGTS, Division 1. The trip of channel B monitor initiates the actuation of the corre- -

sponding valves for Division 2 and startup of the SGTS for Division 2.

High radiation and downscale control room annunciators are actu-ated by the signals from the monitors. Each control room radiation monitor visually displays the radiation level.

INSERT s 7

(FROM ll.s.2.1.s standby Gas Treatment nadiation sonitoring System W9fT Pass) This system monitors the radiation level at the SGTS exhaust duct.

l The detectors are physically located downstream of the exhaust and heat removal fans and dampers on the exhaust ducts for Division 1

and Division 2.

l 11.5-11

INSERT ON PAGE 11.5-11 t

The exhaust from the Auxiliary Building electrical areas, corridors, steam tunnel and Elevator Tower HVAC System (Figure 9.4-4a) is thenugh two louvered roof vents and is not monitored. Only the steam tunnel has a potential for gaseous radioactive releases requiring monitoring. The steam tunnel is isolated from the rest of the auxiliary Turbine Building steam tunnel via the Seismic Interface Restraint Structure. Monitoring a gaseous releases from this area will be accomplished by Turbine Building vent monitoring. The Turbine Building vent monitoring is the responsibility of the Applicant. The control rod drive maintenance area source has been determined to be not significant. The remaining areas exhausted by this system contain no radioactive sources and are isolated from the potentially radioactively contaminated areas of the Auxiliary Building.

l

i

_M 60.14 . . . ..

. _. er State whether the source terms you have used to evaluate off-site doses due to a postulated failure of the off-gas system are consistent with Branch Technical Position ETSP 11-5 (July 1981).

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t% % 11J 101 22A7007 238 NUCLEAR ISLAND Rev. 0

, Table 15.7-4 GASEOUS RADWASTE SYSTEM FAILURE SYSTEM RUPTURE (DESIGN BASIS ANALYSIS)

FISSION PRODUCT RELEASE TO ENVIRONMENT Isotope Ci Isotope Ci Isotope Ci Cr23 1.39E-2 H3 1.01E-3 Rul03 1.07E-7 24 3.02E-2 Cl4 9.81E-5 105 1.08E-6 25 2.96E-2 Na24 4.18E-6 1-6 4.65E-9 Il31 9.98E-3 P32 4.01E-8 Agl10m 2.25E-7 132 1.28E-1 Cr51 9.70E-7 Tel29m 2. 76 E- 8 133 7.43E-2 Mn54 4.69E-8 129 3.98E-6 134 2.87E-1 56 1.06E-4 131m 1. 86 E- 7 135 1.19E-1 ,

F 59 1.47E-7 131 2.76E-7

.CoS8 8.58E-6 132 1.08E-6 Kr83m 9.00E+1 '60 3.74E-7 Co187 4 . 8 8E- 8 85m 1.54E+2 Ni65 6 . 8 4 E- 7 188 1. 7 2E- 8 85 8.24E-2 2n65 2.71E-9 Co189 6.29 87 4.19E-12 RsPb88 1.42E+2 140 8.44E-8 88 4.88E-12 , 89 6.80E+1 141 2. 8 4 E- 8 89 1.52E Q 'ir89 2.96E-2 142 4.31E-8 90 90 4.34E-5 Lal40 1.08E-4 3.355}13__

Xel31m 8.56E-1 91 8.12E-2 142 1.09E-8 133m 8.81 92 1.30E-3 Cel41 2.00E ~

133 2.14E+2 Y90 3.51E-7 143 9.87E-7 135m 3.87E+2 91m 1.61E-2 144 7 .6 6 E- 8 135 6.85E+2 91 1.75E-5 Nd147 1.18 E- 8.

137 1.87E+3 92 5.31E-6 W187 1. 9 0 E- 8 138 1.26E+3 93 3.48E-6 Np289 1. 81E- 3 139 3.54E+3 Zr95 1.83E-7 140 3.17E+3 97 1.73E-6 Nb95 1.60E-4 Mo99 1.69E-6 Te99m 8.97E-4 101 1.14E-3 15.7-37

I

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'460 18 i ,' ' Provide additional information on the following itses applicable to

y, ,, Item 111.0.1.1 of NUREG-0737:

c. Gescribe the leak reduction measures which will be incorporated into ._

_nur design.

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@E@SAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1A.77 INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT LIKELY TO CONTAIN RADIOACTIVE MATERIAL FOR PRESSURIZED-WATER REACTORS AND BOILING-WATER REACTORS (NUREG-0737 Item III.D.1.1) (Cont'd)

Response

L 4 a k d edtoo b -a.a.s u v4o o f O 4:. )

iwelwoke Who 238 Nuclear Islandp rrrider a number of barriers to 1 containment leakage in the closed systems outside the  !

containment. These closed systems include:

l

a. Residual Heat Removal,

b. High Pressure Core Spray,

c. Low Pressure Core Spray,

d. Reactor Core Isolation Cooling,

e. Suppression Pool Cleanup (suction and return), and

f. Shutdown Service Water (supply and return).

provsJcJ h4k ApphcJ Plant specific procedures will prescribe the method of leak testing these systems. The testing will be performed on a schedule appropriate to 10CFR50 Appendix J type B and C penetrations, e v c. that t e. is at each refuelh' g outage.W q- he n eakagepathsgdiscovere Md durin ch these tesdsgw11/1 be investigated -eguts when necessarybaintenance will be performed to reduce leakage to its lowest practical level.

INSERT / ~

Additionally, pressure boundary components of radioactive waste systems are purchased as augmented Class D systems to assure their capability to provide integrity.

1A.77-2 l

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% a.A6EMem Hines whi& p/netrate the primary contaiment contain primary con-taiment isolation valves whim are designed in accordance with General Design Criteria 55, 56 or 57 to provide reliable isolation ' netion event- 8 of line breaks. Rese isolation provisions are discussed i

( 6.2.4 and contain both automatic and remote manual-closing valves.

thould a small line break develop witnin a space inside the seconaary contaiment concurrent with a significant radioactive source term in tne reactor wateg, it tion would 5.2.5be detected and the line by the may Leakage oe isolated. Control SystemsAny release of described in radioactive material from such leaks would also be detected by Irocess

~~

radiation monitors ich tion would permit operation of the Standby Gas 6.5.1) prior to release to tne enviroment.

-- Treatment System (

All lines which pass outside of tne secondary contaiment contain Leakage

~~ Control Systems or loop seals. Rese systems allw the SGTS to maintain a negative pressure relative to the enviroment and tnus limit tne amount of leakage t ugh the secondary contaiment. Rese systems are tions 6.5.3. Finally, expected liquid leakoff fraa discussed in

~_ _.

equipnent outside the contaiment is directed Rese to multipleequipnent design drain features su:apsof and processed by the Ra&aste System.

the 238 naclear Island provide substantial capability to limit any potential release to the enviroment from systems likely to contain radioactive material.P __ _e__

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t ATTACHMENT NO. 4 DRAFT RESPONSES TO POWER SYSTEMS BRANCH QUESTIONS 1

l

)

'd ,

430.11 Describe in Section 8.3.1.4 of your FSAR, the cable spreading area and (8. 3.1')

~ the separation of cables in this area with respect to the requirements contained in Section 5.1.3 of IEEE Std. 384-1974 as modified by Regulatory Guide 1.75. State whether: (1) this area contains high-energy equipment

' such as switchgear, transformers and rotating equipnent or piping (both high and moderate-energy) which could be a potential source of

'- missiles or pipe whip; (2) flammable materials are stored in this area; (3) power cables are routed through this area; and (4) redundant cable spreading areas are utilized. Provide the cable tray plan for this area and the electrical equipment room areas.

Resnute See a+$acbek3l2eefS,

. .F.sAR. .jsetA 1-3. f. y. 2.s.2. & b. e dd XQ m/e O % st 9 4 =% a s4v., J., ,t b n 2%ii2 L e f .

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1

RESPONSE

The GESSAR II design does not have cable spreading areas as defined by IEEE 384, Section 5.1.3. According to that definition, a cable spreading area is a space or spaces adjacent to the main coni.rol room where instrumen-tation and control cables converge prior to entering the control, termina-tion, or instrument panels. In application a cable spreading room allows the design of the trays and cabling external to the control room to proceed independently of the layout of the control room cable trays and panels. Mismatches develop in the grouping of the plant cables with respect to their required grouping for the control room. This mismatch is corrected by making the required routing transitions through inter-connected trays in a cable spreading area. This is not the case for the GESSAR II Design.

For the GESSAR II design the field cables, starting at the end devices, enter the conduit and tray system and are routed back to main cable trays which are routed through the auxiliary building to the control building.

This conduit and tray system may be visualized by thinking of each end device as being at the end of a tree branch. The cable then follows a branch back to a limb or a series of limbs and then finally to the trunk or main cable tray. These main tray runs were located in the best areas available for cable trays. Avoiding such things as piping systems, high energy equipment, flammable material sources and maintenance hazards was considered as tray routes were selected.

By the time a main cable tray penetrates the three hour fire rated wall of the control building, there are no cables to gather up. They are all in the appropriate tray. Each cable remains in the same tray until it peels off in a branch (tray or conduit) to the proper PGCC termination cabinet. All trays are solid metal with solid metal covers.

A listing of the power cables passing through the cable tray areas of the

, control building is attached. Each cable has been marked with a service number based on the following descriptions:

Service Number Description 1 120VAC Load on the AC instrument bus 2 125VDC Load on the DC instrument bus 3 Current transformer output leads 4 120VAC Miscellaneous power (panel space heaters, description panel fear, etc.,)

5 480VAC Instrument bus transformer feed 6 480VAC Regulating transformer feed 7 480VAC Feeder to control building MCC 8 480VAC Load in control building 9 480VAC Load on control building MCC All of the catile's either serve loads or access power supplies located in

! the control bu.ilding. Cable service numbers 1 through 4 are for 120VAC or 125VDC maximum. Cables with service numbers 5 through 9 are 480VAC.

I I

l l

JEM:pc/LO2025-1 j 2/2/83

f

~f The cable tray areas are specifically designed to safely contain all types of cables which are routed through them. The areas are ventilated and have smoke venting systems. The rooms are separated from the control room by two hour equivalent fire rated walls and separated from the remainder of the plant by three hour rated walls. The rooms have fire detection systems and sprinkler systems. They are secured and not in any traffic pattern. The power trays, of which there are a total of three, are the top trays in the stacks. All trays are solid metal with solid metal covers. Any internally generated electrical fire would be confined to the originating tray. The divisional separation is such that if a fire should develop in one of the rooms, the plant operator would know immediately which safety systems he could plan on using to maintain cooling to the plant and which systems could be shutdown to commence fire fighting. He would also know that the remote shutd wn panel would still be available for use.

In short, the control building cable tray areas are the safest and best place to route cables within the control building. Since the loss of either a power or control cable degrades a system, attempting to route a cable elsewhere just because it is a power cable only exposes the system to additional possibilities of damage and subsequent curtailment of operation.

See a+%c hed GESSA R Tl +ex+ revision fo r en cid n + 1onal d isc us sen.

JEM:pc/LO2025-2 2/2/83

l.

f Po we r Cab les Passin cr ihmuh 7%e confro/ Bui/ din C%le / m Area Divibibn / A a.

. .. W.h.Ih .??. .. .. ... ?* CW C'lBF-Al -82lC-4000-DIU3 T - .458 NI-D23R-4803-ilV3 9 .'229 AI -P45 -4000-DIU3 F i.303 '

Al -P45 -4018-DIU3 9 1.003 AI -P45 -40iP-D1U3 '" F I.003,:

Al -P45 -4020-DIU3 8.003 $lfeec/' 2,eonduciate/p/

Al -P45 -4021-DIU3 F 1.003 Al -P45 -4022-Df 883 s I.003_J Al -P61 -4003-DIU3 I .158 Al -Ril -4001-DIU3 f .33S Al -Ril -4007-DIU3 S~ .423 Al -R24 -4021-DIU3 7 f.003 1 Al -R24 -4022-DIU3 7 1.003 i r Al -R24 -4093-DIU3 7 f.003 L l fee o Al -R24 -4099-DIU3 7 1.003 Al -R24 -4I00-DIU3 7 1.003 i '

Al -R24 -4101-DIU3 7 1.003 J Al -R24 -4137-DIU3 + .338 Al -R28 -4002-DIU3 E. .467 Al -R28 -4009-DIU3 tL. .467 O i vicion 2 A r-ea Cable hb. Se.py,'e c Diameier "~' *I=5'5 500'~NDU3 ' 'M8 Al -D23/1-4D03- 12~~ .

Al -833R-4076-NDU3 T .l58 Al -Ri s -4005-D2b. k ~ 723 Al -Ril -4806-D2V3 f A.l -333R-4077-NDU3 ? .458

.423 Al -C5f C-4025-NDU3 .158 Al -Ril -4077-22U3 5- .?30 Al -Bl?Z-4001-NDU3 1

Al -R24 -4023-D2U3 7 1 003 ' 7 .158 -

AI -R24 -4029-D2U3 7 Al -D2iR-4005-NDU3 / .158 1.003 i Al -D2iR-4006-NDU3 I .158 Al -P24 -4 t l2-D2U3 7 1.003 j r i Al -D2iR-4009-NDU3 /

Al -R24 -4 f l3-D2U3 7 1.003 !l' lteed Al -D2f R-40ll-NDU3 I

.l58 Al -R24 -4114-D2V3 7 f.003 i .158 Al -D2iR-4012-NDU3 I Al -R24 -4115-D2U3 7 1.003 ,

Al -D2i A-4013-NDU3

.l58 Al -R24 -4 539 D2U3 4 .338 l .358 Al -R28 -4003-D;U3 2.

Al -D21R-4014-NDU3 i .l58

.g67 Al -D2iR-4015-NDU3 I .158 Al -D2fR-4016-NDU3 / .458 Al -D2fR-4017-NDU3 I . l58 i Al -D2iA-4018-NDU3 / .158 i Al -D2iR-4019-NDU3 / .158 AI -X93 -4049-NDU3 F

.194 8 Ag , ~$1 0 2-ND 3

• Al -X93 -4050-NDU3 F .194 -

'I Al -X93 -4056-NDU3 Y .194 ' Al -D21R-4024-NDU3 / 'l58 Al -X93 -4057-NDU3 7 .194 s Al -D2fR-4026-NDU3 i 'l58 Al -X99C-4002-NDU3 1 .358 i - Al -D2f A-4023-NDU3 i ~158 Al -X99C-4007-NDU3 / .558 i Al -D2fR-4030 NDU3 / '158 Al -X99C-4009-NDU3 / .158 8 Al -D2iR-4032-NDU3 / 'l58 Al -X99C-40f0-NDU3 1 . 358 e RI -D2iR-4039-NDU3 / 'ISS Al -X99C-40ll-NDU3 / .158 i Al -D2fA-4036-NDU3 I '153 Al -X99C-4012-NDU3 I .358

• Al -D2fR-4038-NDU3 / 158 GI -X99C-4014-NDU3 / .l58 ' Al -D2iA-4090-NDU3 / 'ISS Al -X99C-4015-NDU3 / .158 i Al -D2iA-4042-NDU3 1 'ISS Al -X99C-4017-NDU3 / .458 Al -D2fR-4094-NDU3 / 158 Al -D21R-4045-NDU3 / 'l58

~

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Di vision & Area (ContW) \

Al -D21A-4046-NDU3 / .358 Al -PS3 -40ll-NDU3 I .158 Al -D21A-4047-NDU3 / .158 gg .P53 -4012-NDU3 i .358 AI -D2iA-4048-NDU3 / .158 ' Al -P53 -4013-NDU3 L 158 Al -F42R-4019-NDU3 / .196 4000-NDU3 5 . 1.158 Al -P42A-4027-NDU3 / .558 -gAl g .g-Ril g g .4003-NDU3 T .542 Al -F42R-4031-NDU3 / .358 Al -Rif -4004-NDU3 E .615 '

Al -G33R-4039-NDU3 I .458 Al -Rif -4008-NDU3 f .423 Al -G33R-4061-NDU3 / .158 Al -Ri f -40ll-NDU3 E' .423 Al -G33R-4062-NDU3 I gg .pgg .4012-NDU3 C .467

.858 '

Al -G3 W U26-NDU3 / .423 Al -Ri f -4013-NDU3 Y .423 Al -Ril -4015-NDU3 E .467 al -G41A-4031-NDU3 Al -G41W39-NDU3

/

/

_ .158 '

.558

• Al -Rif -4051-NDU3 8" .615 Al -G4 f A-4040-NDU3 I .158 Al -Ril -4082-NDU3 3- .467 Al -G46A-4008-NDU3 1 .158 - GI -RII -4033-NDU3 5~ .467 Al -P39 -4028-NDU3 I .194 ' Al -RII -4084-NDU3 E- .338 Al -P39 -4029-NDU3 i .194 8 Al -Ril -4096-NDU3 ( .338 Al -P39 -4039-NDU3 I .423 ' AI -R24 -4000-NDU3 7 1.003 Al -P39 -4041-hDU3 l .423 Al -R24 -4044-NDU3 7 I.003 Al -P44 -4026-NDU3 / .158

• Al -R24 -4045-NDU3 7 1.003 Al -P44 -4028-NDU3 I .358 i Al -R24 -4I34-NDU3 4 .338 Al -P44 -4029-NDU3 1 .542 i Al -R28 -4004-NDU3 7-. .615 al -P44 -4030-NDU3 / .403 Al -R28 -4007-NDU3 ? .615 i Al -P44 -4032-NDU3 / .158 : Al -R42 -4060-NDU3 2 .615 Al -Pv4 -4033-NDU3 / Al -R42 -4061-NDU3 ?_ .615

.542 : Al -T4IB-4052-NDU3 I .158 Al -P44 -4036-NDU3 1 .158 : Al -T4fB-4054-NDU3 al -P44 -4037-NDU3 / .858 (

I .158 Al -T4tB-4056-NDU3 I l.003 AI -P44 -4038-NDU3 9 .358 i Al -T4fB-4058-NDU3 I Al -P44 -4040-NDU3 I .158

• AI -T4tB-4C50-NDU3 I

.l58 Al -P48 -4013-NDV3 l . l58 f .358 AI -PS3 -4008-NDU3 I Al -T4fD-4020-NDU3 i .358

.158 l Al -T4tD-4022-NDU3 I .158 al -P53 -4009-NDU3 I .358 '

Al -PS3 -4010-NDU3 I .358 Al -X73R-4014-NDU3 ) .358

GESSAR II 22A7007 238 NUCLEAR ISM Rev 0 O

, for 4'34%9 )

'r,

8. 3.1. 4. 2. 3. 2 other safety-malated systems (Continued)

I (8) Detailed design basis, description, and safety evalua- l tion aspects for a power generation control complex 1 (PGCC) system shall be as comprehensively documented and presented in GE Topical Report, Power Generhtion l

Control Complex, NEDO-10466A and its mnendments.

k PGCC consists of control room panels, racks, floor

~

sections, and termination cabinets. The floor sections are divided into ducts and the termination cabinets have metallic barriers to separate redundant Class 1E wiring.

The floor section ducts are designed so that each duct g ,

acts as a raceway and has adequate fire barriers and will contain wiring of only one division. The ducts havesolidmetalwallsandfloofand#removablesolid metal cove [g

! Cable access to the two PGCC areas is provided through two cable tray l

rooms located on either side cf the control room. Each cable tray room is marked on drawings as containing two divisions. In actuality, Divisions 4 and 3 are imbedded in the. walls of the Divisions 1 and 2 cable trcy rooms, respectively. The imbedded depth is sufficient to provide a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rated separation.

The cable tray rooms do not contain any high energy equipment, rotating equipment, or piping which could be a potential source of missile or pipe whip. No flammable materials are stored in,!these rooms. : Low voltage power cables (V3) are routed through the cable tray rooms to provide power for lighting transformers, regulating transformers and instrument buses within the control building. The areas are utilized for cable tray and conduit routing only, no other major equipment is housed within the cable tray rooms.

See Figures 8.3.30, 8.3.31 and 8.3.32 for physical layouts of the area.

l

O 1

ATTACHMENT NO. 5 DRAFT RESPONSES TO REACTOR SYSTEMS BRANCH QUESTIONS s

+

I 440.05 In Section 5.4.7.1.5 of your FSAR you provide a discussion of the reactor (5.4.7) heat removal (RHR) system alternate shutdown cooling mode in which water is discharged through the automatic depressurization system (ADS) valves. Provide, or make reference to, test data confirming that the ADS valves used in your design can pass sufficient water in this mode for the most limiting conditions. Include a discussion of the appli-cability of the particular tests which you reference.

hesponse_

The description of the alternate shutdown cooling flow path presented in Section 5.4.7.1.5 has been superceeded by the EPG.

The EPG require the RHR/LPC loops with heat exchangers to be placed in suppression pool cooling.

The other low pressure injection pumps LPCS & LPCI "c" are used to force water through the S/RV.

Utilization of water directly from the suppression pool which has not been passed through the heat exchangers is also desireable from the consideration of avoiding RPV N T or head tension limit.

Test on the S/RV to verify their abiltty to pass suitable quantities of water for alternate shutdown cooling have been performed, The results of these tests are presented in General Electric Report Number NEDE 24988-P ti.tled

" Analysis of Generic BWR Safety / Relief Valve Operability Test Results", Tests were perfomed in response to NUREG-0737.

440.16 The ECCS contains manual as well as mot:r-opsrated valves. There is (6.3.2) a possibility that annually operated valves might be left in the

'3, wrong position and remain undetected prior to the occurrence of an accident. Examples of such valves include those pairs of normally closed valves which are in the test / drain lines between the HPCS, LPCS and LPCI isolation valves. Provide a list of all manually-operated valves in the safety-related reactor systems, including their location and type. Discuss the methods which will be used

' to minimize such an occurrence. It is our position that you provide indication in the control room for all critical ECCS valves (manually or actor-operated).

A- E M d11 manually-operated valves in the safety-related reactor systema r i- ridin, O.;i; 1::;tirr r' ti "6 have indica on in

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vsc e. 1 SYS VALVE MPL NO LINE LOCATION VALVE TYPE RER E12-F010 20" RER 19-EAA 20" Gate - Band operated E12-F029A 18" RER 7-BAB 18" Gate - Hand operated E12-F029B 18" RER 13-BAB 18" Gate - Hand operated E12-F029C 18" RER 21-BAB 18" Gate - Hand operated E12-F039A 12" RER 10-EAA l 12" Gate - Hand operated  ;

E12-F039B 12" RER 16-EAA 12" Gate - Hand operated  !

E12-F039C' 12" RER 22-EAA 12" Gate - Hand operated '

LPCS E21-F007 12" LPCS 3-EAA 12" Gate - Hand operated E21-FF121 4" LPCS 6-BAB 4" Globe - Hand operated .

HPCS E22-F036 12" EPCS 4-EAA 12" Gate - Hand operated E22-FF124 4" EPCS 20-EAB 3" Globe - Hand operated RCIC E51-FF210

• 6" RCIC 2-EAB 6" Gate - Hand operated E51-FF211 8" RCIC 1-AAB 8" Gate - Band operated E51-FF222 3" RCIC 10-EAB 1" Globe - Hand operated HPCS SW P40-FF001 8" CSSW 1-AKC 8" Buttarfly - Hand pperated (ESW Div 3r P40-FF002 8" CSSW 2-AKC 8" Butterfly - Band operated ESW P41-FF001A 10" ESW 3-ADC 10" Buttarfly - Hand operated (Div 1 & 2) P41-FF001B 10" ESW 43-ADC 10" Butterfly -Hand operated P41-FF002A 10" ESW 4-ADC 10" Butterfly - Hand operated P41-FF002B 10" ESW 44-ADC 10" Butterfly - Hand operated P41-FF006A 10" ESW 4-ADC 10" Butterfly - Hand operated l P41-FF06B 10" ESW 44-ADC 10" Butterfly - Hand operated

' Each of the above valves is monitored by the Performance Monitoring Systems (PMS) for individual alarming of "not fully open". These valves are then grouped by system and division with a status light in the control room for system level indication of " manual valve misaligned". In addit;.un to the status light, a connection is made to the " system out of service" alarm such that an alarm results whenever the status light is on. '

1 1

2 The positions of normally closed safety-related reactor system manually-operated valves which are not part of an ECCS loop (such as on the test /

drain lines between isolation valves, etc.) are monitored implicity by the Reactor Coolant Pressure Boundary and ECCS System Leakage Detection System described in subsection 5.2.5. If a pair of these normally closed valves were inadvertently left open, the leakage detection system would activate an alarm in the control room and allow the operators to take appropriate corrective actions. The capability of the leak detection system is summarized in the following table:

Leakage Rate Within Location Detection Activates Alarm Drywell Drywell Floor Drain Sump 5 gpm*

Drywell Equipment Drain Sump 25 gpm External to Containment Floor Drain Sump 5 gpm Drywell (Within Containment Equipment Drain Sump 25 gpm Containment)

Outside of Area Temperature 25 gpm Containment (Aux. Building, Floor Drain (HPCS, RHR, RCIC, 5 gpm Main Steam and LPCS Pump Rooms)

Tunnel, and Turbine Building)

• The sensitivity within the drywell is 1 gpm within I hour. The 5 gpm is for activation of control room alarm.

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ATTACHMENT NO. 6 DRAFT RES.'ONSES TO EQUIPMENT QUALIFICATION BRANCH QUESTIONS I

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't 7'T 4.O 7 (3.9)

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4. The GESSAR II FSAR does not contain substantial discussion about the development of hydrodynamic loads for purposes of equipment qualifi-cation or how the loads are handled in the qualification. The limited discussion of this subject in Section 3.9 indicates that the hydro- -

dynamic loads will be represented by response spectra. But no dis-cussion is presented as to how the response spectra are developed or how the hydrodynamic loads are combined with seismic loads. If these loads are combined by performing an SRSS summation, the l results may be less than conservative. Discuss more thoroughly.

• l the treatment of hydrodynamic loads, seismic loads, and their I combination. -

l b . Persistently throughout Section 3.9 the statement is made that if l equipment can be shown to have natural frequencies greater than 33 Hz, it can be considered rigid. This, of course, may not be true if the equipment is subjected to hydrodynamic loads which have I a frequency ~ content greater than 33 Hz. Where hydrodynamic loading l is mentioned, a frequency of 60 Hz as the cutoff frequency should be provided since hydrodynamic loads often contain higher frequencies.

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rav. 0 y 3.9.2.2.2.17 Other Nuclear Island ASME III Equipment of interest at a rate no greater than one octave per minute. If no resonances are located, then the equipment is considered as rigid and single frequency tests at every 1/3 octave frequency interval are acceptable.

Also,f.ife all 34 nateal frequencies 0 m c. To d 4 4 6o H3 of Vev the equipmente kyh ed e\**doj i

are greater than 33 Hgg the equipment may be considered Tigid and analyzed statically as such. In this static analysis, the dynamic forces on each component were obtained by concentrating the mass at the center of gravity and mul iplying the mass by the appropriate floor acceleration.ht 33 Hz. The dynamic stresses were then added to the operating stresses and a determination made of the adequacy of the strength of the equipment. The search for the natural fre-quency is done analytically if the equipment shape can be defined mathematically and/or by prototype testing.

If the equipment is a rigid body while its support is flexible, the overall system can be modeled as a single-degree-of-freedom system consisting of a mass and a spring. The natural frequency of the system was computed; then the acceleration was determined from the floor response spectrum curve using the appropriate damping value. A static analysis was then performed using this acceleration value. In lieu of calculating the natural frequency, the peak acceleration from the spectrum curve was used. The critical damping values for welded steel structures from Table 3.7-1 were employed.

In case the equipment cannot be considered as a rigid body, it can be modeled as a multi-degree-of-freedom system. It is divided into a sufficient number of mass points to ensure adequate representa-tion. The mathematical model can be analyzed . a modal analysis technique or direct integration of the equatic- 3r motion. Spe-cified structural damping is used in the analys.o unless justifica-tion for other values can be provided. A stress analysis was per-formed using the appropriate inertial forces or equivalent static loads obtained from the dynamic analysis of each mode.

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3.9-54

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2'71. 0 4 (S.I0) i 8.'The GE position on fatigue effects due to hydrodynamic loading should be discussed. The argument for using only one OBE-intensity earthquake instead of five as stipulated in the IEEE 344-1975 Standard for seismic fatigue evaluation may be acceptable for certain plants. The use of only one 08E, however, should nBt be used on a generic basis. It should rather be . justified for each plant where only one OBE is used.

k. The qualification ' program should address the degree of aging or env.ironmental degradation that pieces of equipment could potentially incur prior to the occurrence of dynamic loading.

The program should assure that the equipment has undergone its maximum expected amount of aging before the dynamic loads are appl.ied in the qualification of the equipment. Surveillance and maintenance programs needed to assure that the equipment does not age to a degrees worse than qualified to should be described. .

.C. Sequential testing"needs to be discussed more thoroughly.

The discussion should make clear that seismic and hydro-dynamic tests follow other environmental testing on the ,

equipment. The sequence of exploratory seismic and hydrodynamic loads and how this sequence, properly quali-fies the equipment for all loads incurred during the life of the equipment should be included in the discussion..

R.e.s p o w S t. A i

The GE method for handling fatigue effects due to hydrodynamic loading is presented in Section 4.4.2.5 of NEDE-24326-1-P.

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Cl.It is stated that-closely spaced modes are combined by either the Double Sum Method or an algebraic sum of such modes. The Double Sum Method is acceptable according to Regulatory Guide 1.92, but an algebraic sum.could be inappropriate. If the

. modes are added algebraically, cancel.lation would occur among modes having opposing signs. This cancellation could result in a non-conservative calculated total response. Justify the use of the algebraic sum method.

b. Two deviations from SRP 3.7.3 criteria are given. Justification for allowing these deviations should be provided.

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