ML20059A393

Forwards Request for Addl Info Re GE Application for Certification of Advanced BWR Design.Responses to Encl 1 Requested by 900831 & Responses to Encls 2 & 3 by 900928

ML20059A393
Person / Time
Site:	05000605
Issue date:	08/15/1990
From:	Scaletti D Office of Nuclear Reactor Regulation
To:	Marriott P GENERAL ELECTRIC CO.
References
NUDOCS 9008230027
Download: ML20059A393 (26)

Text

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August 15, 1990 Docket No. STN 50-605 Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy General Electric Company 175 Curtner Avenue San Joss, California 95125

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SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE ABifR DESIGN During the course of the review of your application for certification of your Advanced Boiling Water Reactor Design, we have identified a need for additional information. Our request for additional information, contained in the enclosures, addresses the areas of SRP Sections 3 and 9, (Enclosures 1 and 2) reviewed by the Plant Systems Branch, and SRP Section 13 (Enclosure 3) reviewed by the Safeguards Branch. We request that you provide your responses to Enclosure 1 by August 31, 1990, and your responses to Enclosures 2 and 3 by September 28, 1990. If you have any concerns regarding this request please call me w (301) 492-1104. ,

i Sincerely, Oriainal sianed by D. Scaletti Dino C. Scaletti, Project Manager Standardization Project Directorate Divisj on of Reactor Projects - III, IV, V and Special Projects Office of Nuclear Reactor Regulation >

Enclosures:

As stated i cc w/ enclosures: l

- See next page '

. DISTRIBUTION:

Docket? Rile, JKudrick CMiller PShea  !

NRC PDR RArchitzel CMcCracken l PDS Reading TChandrasekaran ACRS(10)

DScaletti BMendelsohn PMckee

• See previous concurrence f PDS *PDS E NtB PShe CMiller fo TIT /90 08/15 90 08/15/90

/0 b 9008230027 900815 t i PDR ADOCK 05000605 i \-

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August 15 1990 l

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Dockkt No. STN 50-605 i

Patrick W. Marriott, Manage.r i Licensing & Consulting Services  !

GE Nuclear Energy j General Electric Company -

j 175 Curtner Avenue j San Jose, California 95125  !

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION FOR CERTIFICATION OF THE ABWR DESIGN During the course of the review of your application for certification of your Advanced Boiling Water Reactor Design, we ,

have identified a need for additional information. Our request  !

for additional information, contained in the enclosures, i addresses the areas of SRP Sections 3 and 9, (Enclosures 1 and 2) reviewed by the Plant Systems Branch, and SRP Section 13

. (Enclosure 3) reviewed by the Safeguards Branch. We request  ;

l thr.c you provide your responses ta Enclosure 1 by August 31, '

1t90, and your responses to Enc.1osures 2 and 3 by September 28, 2990. If you have any concerns regarding this request please

<all me on (301) 492-1104.

t.

l Sincerely, l

Oriainal sianed by D. Scalmill Dino C. Scaletti, Project Manager Standardization Project Directorate Division of Reactor Projects - III, IV, V and Special Projects office of Nuclear Reactor Regulation

Enclosures:

l As stated cc w/ enclosures:

See next page I

. DISTRIBUTION:

Docket File JKudrick CMiller PShea NRC PDR RArchitzel CMcCracken PDS Reading Tchandrasekaran ACRS(10)

DScaletti BMendelsohn PMckee P,Dh PDS $ PDS 1 D5 cat)etti PShe CMiller 08/15/90 08/15/90 08//[/90

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UNITED STATES

. . [ '$ f' '$6,i NUCLE AR REGULATORY COMMISSION

$'% 8 W ASHINGTON. D. C. 20555 tugust 15. 1990 e...*

Docket No. STN 50-605 Patrick W. Marriott, Manager Licensing & Consulting Services GE Nuclear Energy General Electric Company 175 Curtner Avenue San Jose, California 95125 Dear Mr. Marriott

SUBJECT:

REQUEST FOR ADDITIONAL INFORMATION REGARDING THE GENERAL ELECTRIC COMPANY APPLICATION FOh CERTIFICATION OF THE ABWR DESIGN During the course of the 20 view of your application for certification of your Advanced Boiling Water Reactor Design, we have identified a need for additional information. Our request for additional information, contained in the enclosures, addresses the areas of SRP Sections 3 and 9, (Enclosures 1 and 2) reviewed by the Plant Systems Branch, and SRP Section 13 (Enclosure 3) reviewed by the Safeguards Branch. We request that

.you provide your responses to Enclosure 1 by August 31, 1990, and your responses to Enclosures 2 and 3 oy September 28, 1990. If you have any concerns regarding this request, please call me on (301) 492-1104.

Sincerely, c /$r.c ( /N Dino C. Scaletti, Project Manager Standardization Project Directorate Division of Reactor Projects - III, IV, V and Special. Projects Office of Nuclear Reactor Regulation

Enclosures:

As stated cc w/ enclosures:

See next page

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...a General Electric Company Docket No. STN 50-605 cc: Mr. Patrick W. Marriott, Manager -

Licensing & Consulting Services GE Nuclear Energy General Electric Company 175 Curtner Avenue San Jose, California 95125 Mr. Robert Mitchell General Electric Company

• 175 Curtner Avenue San Jose, California 95114 Mr. L. Gifford, Program Manager Regulatory Programs GE Nuclear Energy-12300 Twinbrook Parkway Suite 315 Rockville, Maryland 20852  ;

Director, Criteria & Standards Division Office of Radiation Programs U. S. Environmental Protection Agency 401 M Street, S.W. '

Washington, D.C. 20460 Mr. Daniel F. Giessing i Division of Nuclear Regulation and Safety Office of Converter Reactor Deployment, NE-12 Office of Nuclear Energy Washington, D.C. 20545 t

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ENCLOSURE 1 REQUEST FOR ADDITIONAL INFORMATION PLANT SYSTEMS BRANCH 430.177 ABWR SSAR Section 9.1.1.1.1, Nuclear Design, states (9.1.1) that since no credit is taken for neutron leakage, the value for effective multiplication factors are really infinite neutron multiplication factors. ABWR SSAR Section 9.1.1.3.1, Criticality Control, states that k for both normal and abnormal storage conditions will $$g less than or equal to .95. However, the same section states that the new fuel storage area will accommodate fuel with a k 1 ResolvethisbSkc<repa.35withnosafetyimplications.

ncy.

430.178 ABWR SSAR Section 9.1.1.1.6, Dynamic Analysis, refers (9.1.1) to ABWR SSAR Section 9.1.2.1.6, which does not exist.

Provide the results of a dynamic analysis of the new fuel storage system.

430.179 ABWR SSAR Section 9.1.1.1.7, Impact Analysis, also (9.1.1) refers to a nonexistent ABWR SSAR Section 9.1.2.1.7.

Provide impact analysis for imoset loads up to and including a fuel assembly and its carrying fixture.

430.100 Provide details of assumptions and input parameters (9.1.1) used in the criticality analysis for new fuel storage.

Include information such as number of racks, their material (e.g., stainless steel ?), number of fuel assemblies per rack, neutron-absorbing material and its placement, placement of fuel assemblies (center-to-center distance between rows and within rows), and effect of spacing on k in normal dry condition or when completely fIbbdod with water. Also, clarify whether the spacing is sufficient to ensure a k of underoptimummoderatorconditionsif$am,0.98orless small droplets, spray or fogging) as described in SRP Section 9'1.1.

Clarify whether the racks are designed to preclude inadvertant placement of a fuel assembly in other than prescribed locat4ons.

430.181 How is the new fuel protected from internally generated (9.1.1) missiles and the effects of moderate or high energy piping failures? State whether or not there is moderate or high energy piping or rotating machinery in the vicinity of the vault housing the new fuel storage racks.

430.182 Provide information on how the design of the new fuel (9.1.1) storage facility complies with GDC 61, " Fuel Storage and Handling and Radioactivity Control." Identify the ventilation system provided to handle possible. release of radioactivity resulting from accidental damage to the fuel (note that ABWR SSAR Section 7.1 does not describe the radiation monitoring equipment for the-new fuel storage area as stated in ABWR SSAR Section 9.1.1.2).

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i' 430 183 Provide sufficient information and drawings to (9.1.1) determine that the failure of non-sessmic systems and structures in the vicinity of the new fuel storage facility can not cause an unacceptable increase in k,gg.

430.184 Demonstrate that the analyzed impact of a fuel (9.1.2) assembly, including its associated handling tool, dropped from a height of 6 feet bounds the range of all possible load drops from all possible heights. For additional guidance on the required bounding analysis, see SRP Section 9.1.2, Item III.2.e.

430.185 Provide sufficient information and drawings to (9.1.2) determine that the failure of non-seismic systems and structures in the vicinity of the spent fuel storage facility can not cause an unacceptable increase in N

eff' 430.186 Provide drawings and information pertaining to spent (9.1.2) fuel storage sufficient to verify the isolation capability of the fuel transfer canal or other provisions to prevent a dropped shipping cask from causing an unacceptable loss of pool water.

430.187 Clarify whether there is a) an interconnecting fuel (9.1.2) transfer canal capable of being isolated from the fuel pool and the adjacent cask loading area, and b) any high-energy piping or rotating machinery in the vicinity of the fuel storage pools. Also, clarify whether the racks are designed to preclude inadvertent placement of a fuel assembly in other than prescribed locations.

4J0.188 Describe the function of the containment pool mentioned (9.1.2) in ABWR SSAR Section 9.1.2.1.5.

430.189 What is the seismic category of the gates in the pools?

(9.1.2) 430.390 Instead of referring to a specific GE proprietary (9.1. 2 ) report on criticality control for spent fuel storage (see ABWR SSAR Section 9.1.2.3.1), provide details of assumptions and input parameters used in the criticality analysis of the spent fuel storage. Also, provide the uncertainty value and associated probability and confidence level for the X value determined by the analysis. Includeinfor$$kionsuch as number of fuel assemblies stored in the pool, center-to-center spacing between fuel assemblies, material of the racks, neutron absorber used and its placing, and k for the above condition when the storageisfulffgloaded and flooded with non-borated water.

ENCLOSURE 1

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l 430.191 List the specific provisions included in the design of (9.1.2) the spent fuel pool to comply with GDC 63, " Monitoring Fuel and Waste Storage" (e.g., pool liner leakage  :

detection, water level monitoring and radiation '

monitoring systems). Identify the corrective actions on detection of loss of decay heat removal capability or excessive radiation levels. Note that for radiation i monitoring systems, additionally referencing ABWR SSAR Subsections 11.5.2.1.2.1 and 11.E.2.1.3, if they are applicable, in ABWR SSAR Subsection 9.1.2.4 is i sufficient. '

430.192 Provide the results and conclusions of the load drop ^

(9.1.4) analysis which considers dropping of one fue> assembly ,

and its associated handling tool-from the height at which it is normally handled above the spent fuel pool -

storage racks. ABWR SSAR Section 9.1.4.3 does not

, discuss compliance with GDCs 61 and 62; therefore, i

discuss the above compliance for the light load handling system.

430.193 A " slack cable" signal (ABWR SSAR Section 9.1.4.3) is ,

(9.1.4) not considered sufficient indication of a fully seated assembly. Discuss whether positive vertical position indication will also be provided.

430.194 ABWR SSAR Subsection 9.1.4.2.2.1, Reactor Building (9.1.4) Crane, indicates that the crane can be used to move new '

fuel to the spent fuel pool and is also used to handle '

the spent fuel cask. Discuss the provisions for preventing movement of the spent fuel cask over the spent fuel pool and results of a failure modes and ,

effects analysis demonstrating the adequacy of controls and interlocks to prevent compromising criticality or radiological safety.

430.195 Clarify whether the system design includes interlocks (9.1.4) to ensure correct sequencing of the transfer operation '

in the automatic or manual mode, and to prevent the refueling platform and the fuel handling platform moving in the transfer area during operations of the transfer system so that the transfer system will not be adversely affected by the presence of either platform.

430.196 ABWR SSAR Tables 3.2-1 (page 3.2-28) and 9.1-2 differ (9.1.4) in seismic classification identification for some fuel servicing equipment. Correct the dicarepancy as appropriate.

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430.197 Provide an enlarged legible version of ABWR SSAR Figure I (9.1.4) 9.1-12, " Plant Refueling and Service Segaenca". J 430.198 ABWR SSAR Section 9.1.4 is confusing on the following (9.1.4) details: .

(a) ABWR SSAR Subsection 9.1.4.2.3.7 and.9.1.4.2.3.8 refer to a fuel handling platform; but it is not  ;

described anywhere under that caption. It is not clear ,

what constitutes the fuel handling platform and whether it is distinct from the refueling platform.

(b).ABWR SSAR Table 9.1-10 refers to three single-failure-proof cranes: the reactor building crane, refueling bridge crane and fuel handling jib crane. ABWR SSAR Subsections 9.1.4.2.7.1 and 9.1.4.3 ,

refer to the automatic refueling machine (a gantry crane) and the spent fuel handling crane. It is not clear which of the above descriptors mean the same load handling device.  :

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(c) Different subsections in ABWR Section 9.1.4 refer to the fuel storage pool, reactor building fuel storage pool, fuel pool and spent fuel pool. It is not clear whether all the above descriptors mean the spent fuel pool. Provide clarification on all the above. Also, provide layout drawings for all the storage pools, ,

including the upper pool and the transfer canal, t

430.199 Include the single-failure-proof characteristics of all (9.1.4) the cranes used in light load handling (note tnat ABWR SSAR Subsection 9.1.4.1 mentions only the hoists on the refueling platform). ,

430.200 ABWR SSAR Subsection 7.6.1 does not provide an (9.1.4) evaluation of the radiation monitoring equipment for the refueling and service equipment as stated in ABWR L SSAR Subsection 9.1.4.5.4. Provide the above  !

information. If it is covered by some other radiation L monitoring systems (e.g., area radiation monitoring -

system and/or process and effluent monitoring system or ,

both), include reference to those systems and the p applicable SSAR Sections in SSAR Section 9.1.4.5.4..

43C.201 The interface criteria of ABWR SSAR Section 9.2.15 does (9.2.4) not include the required interface criteria for the design of the potable (9.2.15) and sanitary water system. To meet the requirements of GDC 60, the design-l of this system should not allow for interconnections

! between the potable und sanitary water system and l systems having the potential for containing radioactive L materials. Protection should be provided through the l use of air gaps, where necessary. Add these design

• l criterja, as interfaces, under ABWR SSAR Section 9.2.15.

ENCLOSURE 1

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430.202 Include the following interfaces besides what have been (9.2.5) already specified for ensuring the ultimate heat sink (9.2.15) (UHS) capability: (1) Design to accommode.te single failures of passive components in electrical systems.

(2) Protection of safety-related portions from adverse environmental conditions including those resulting from piping failures. (3) Time duration of UHS cooling capability availability.

430.203 The ultimate heat sink heat load requirements are ,

(9.2.5) identified by reference to ABWR SSAR Table 9.2-4. This set of three tables (9.2-4a, 9.2-4b, and 9.2-4c) identifies heat loads for each of the three reactor '

building cooling water divisions. These tables do not consider the case of a reactor shutdown at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> after a blowdown to the. main condenser. Inclusion of the above may require a higher heat load dissipation capability for the UHS than what has been currently estimated (see GE's response to Question No. 440.73).

Revise the tables as appropriate considering the above case and provide.the heat load requirements based on the revised tables for the ultimate heat sink (e.g.; ,

the sum of the heat loads for all three divisions, 2 of 3). Are there additional heat loads associated with the UHS not carried by the reactor building cooling  ;

water system? i 430.204 The requirements of 10CFR52 include the need for a

, (9.2.5) conceptual design for systems not considered to be within the design scope of a standard nuclear power plant. No such canceptual design has been included as part of the ABWh SSAR for either the UHS or the interfacing servi:e water system. Provide conceptual designs for the UHS and the interfacing service water system.

430.205 The make-up water preparation system is identified as (9.2.8) outside the scope of ABWR standard plant. This system should meet the requirements of Position C.2 of Regulatory Guide 1.29. Provide an interface requirement that the failure of the make up water preparation system will not result in the failure of any safety-related structure, system or component.

430.206 Clarify how the turbine building cooling water (TCW)

(9.2.14) system meets Regulatory Guide 1.29, Position C.2 with respect to seismic requirer nts for non-safety-related- "

systems that due to their ftilure during seismic events may adversely impact structures, systems or components importent to safety.

ENCLOSURE 1

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.' 430.207 For the TCW system, pro' tide information on the  !

(9.2.14) following items (a) Effect of any system component I failure including rupture of the atmospheric surge tank l

• on structures, systems or components important to '

safety. (b) Required Total cooling water flow and available cooling water flow; total heat output by turbine building auxiliary equipment and available ,

capacity of the TCW heat exchangers. (c) P6wer cycle l heat sink to which the heat from the TCW system is rejected.

430.208 The system diagrams lack sufficient detail to ascertain (9.2.14) whether or not connections between the TCW system and '

safety-related water systems exist. Provide assurance that no such connections to safety-related systems are provided or identify such connections and the isolation capabilities provided. Isolation capabilities should include the use of equipment that is at least Quality Group C and Seismic Category 1. .

430.209 Only ABWR SSAR Sections 6.2-5 and 6.7 discuss the (9.3.1) atmospheric control svstem (ACS) and high pressure nitrogen systems therefore,(6.2.5) correct SSAR Section 9.3.1 which refers to the wrong SSAR sections (1A.2.13) for discuosion of the above systems. Also, provide information on the following items for the ACSt (a) Clarificction on applicability of system design criteria 9, 10 and 11 (protection against single active ,

component failure, missiles, dynamic effects due to piping failures, tornado-missiles, flooding and seismic events) to all . won-safety class system components .

(e.g., nitrogen storage tanks, vaporizers, applicable valves and piping, and instrumentation). (For these '

criteria, see SSAR Subsoction 6.2.5.1.) Specifica~'y, if some of the design bases for the ACS identified An Subsection 6.2.5.1 are applicable only for the safety-related components ot' the system, correct the subsection as appropriate.

P (b) Justification for location of the inboard primary.

containment isolation valves outside the containment, which is .a deviation from GDC 56, " Primary Containment Isolation." .The affected lines are (1) 2-inch nitrogen make-up lines to the drywell and wetwell, (2) 22-inch purge suction lines to the drywell and wetwell (used-also for primary containment inerting or deinerting and connected to a common 16-inch nitrogen supply line),

and (3) 2-inch and 22 inch purgo exhaust lines from the drywell and wetwell. We find your response to Question Nos. 430.35 and 430.42 does not include justification ENCLOSURE 1

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lines nor deviations f rom GDC 56 or 55, " Reactor i Coolant Pressure Boundary Penetrating Containment" requirements for other applicable lines. Include justification for deviations from applicable GDC for othar lines listed in ABWR SSAR Table 6.2-7.

(c) SSAR Subsection 6.2.5.2.7, which discusses the flammabiliflT control system (FCS), does not provide sufficient . .ils for us to conclude that the system complies wiin Lae requirements of TMI Action Item II.E.4.1, " Dedicated Hydrogen Penetrations" of NUREG-0737. Therefore, include the system in SSAR Table 3.2-1 and provide details such as; how long after

• LOCA and at what concentration level of hydrogen the recombiner has to be activated; line sizes as related .

to flow requirements; and duration of recombiner operation. 7.lso, identify interface requ.rements for  !

referencing applicants with regard to the external recombiners (e.g., development of procedtral provisions '

to assure availability of possibly shared portable hydrogen recombiners between sites on a t!mely basis  ;

and coordination of surveillance programs in accordance j with SRP 6.2.5 acceptance criterion II.12).

(d) ABWR SSAR Tables 6.2-7 and 6.2-8 give a line size  ;

of 4 inches and 6 inches respectively for the FCS return line; Table 6.2-7 and Figure 6.2-40 show location of FCS primary containment inboard isolation ,

valves inside the containment and outside the containment respectively; SSAR Sections 6.2.5.2.7 and ,

1$.A.2.12 indicate portable and permanently installed t recombiners, respectively. Resolve all the above.

inconsistencies. Also, if the location of all the i primary containment isolation valves for the system is l

outside.the container.,t, justify the deviation from the GDC 56 requirement for the system inboard isolation  ;

valves.

430.210 Clarify which portions of the high pressure nitrogen l (9.3.1) gas supply- system (nitrogen storage bottles, system ,

l (6.7) piping including. tie lines between safety-related divisions and non-safety-related division, valves, instrumentation and contro.l.s) are. safety-related.

l 430.211 oBWR SSAR Figure 6.7-4 shows only one motor-operated l (9.3.1) isolation valve o each af the tie lines between each (6.7) safety-related division and the common non-safety-related divinion of the high pressure nitrogen gas supply system (MO-F012A and B). The tie piping portion between the two isolation valves is presumably non-safety-related. Explain how essential ENCLOSURE i

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. nitrogen demand will be met during a situation when  !

there is a pipe rupture in one safety-related division (initiating event), single active component failure in  :

the other safety-related division (e.g., isolation valve on the applicable tie line is open) and a pipe  ;

break in the non-safety-related portion of the tie i lines (if there is such a portion). Alternately,  ;

provide two safety-related automatic isolation valves .

in series on each tie line, i 430.212 Provide an FMEA for the Nitrogen Gas Supply System.

(9.3.1)

(6.7) 430.213 Include the nitrogen gas supply system in the ABWR (9.3.1) System classification summary Table 3.2-1. '

(6.7) 430.214 Contrary to what has been stated in ABWR SSAR .

(9.3.1)

Subsection 6.7.1, there is only one non-safety-related (6.7) continuous nitrogen supply portion common to the two essential supply divisions (See Figure 6.7-1). Correct Subsection 6.7.1 as appropriate and discuss the effect of loss of nitrogen supply via the non-safety-related  ;

portion to all the equipment and components identified  ;

j in SSAR Section 6.7.1 (e.g., pneumatically operated '

! valves and instruments inside the primary containment l vessel) during normal operation. Clarify whether the -

pneumatic accumulator which provides the backup

• operating gas for the main steam isolation valve (See s SSAR Subsection 5.4.5.2) is safety-grade for each valve. If not, justify the design.

l 430.215 Provide enlarged and legible size piping and (9.3.1) instrumentation diagrams for the instrument air and ,

(9.3.6) service air systems (SSAR Figures 9.3-6 and 9.3-7), '

(9.3.7) which clearly indice.te alt the components served, .

safety and non-safety-related portions, and isolation l

provisions betwecn the safety and non-safety-related portions; a table showing instrunent air consumption during normpi plant operation. Explain the statements .

in SSAR Siosections 9.3.6.1.1 and 9.3.7.1.1 which indicate that the containment penetrations (secondary l containment penetrations) for the instrument and service air systems are equipped with 7"f ficient isolation valves to satisfy the single-!'ailure '

l criterion (the SSAR figures do not indicate this).

Under the " Location" column for Item P.4 (Instrument / Service Air Systems), Sub-item 5 of ASWR SSAR Table 3.2-1 (Page 3.2-33), include turbine building, radwaste building and service building since ENCLOSURE 1 i

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,' some of the" components of these systems are located in l these buildings. Also, identifv the design feature of )

safety-related air-operated valves outside the '

containment to handle the loss of air supply by the nonsafety-related instrument air system during normal i plant operation. J 430.216 Discuss the specific features provided (e.g., pre and (9.3.1) after filters associated with compressors, particle (6.2.5) size, dryer) for ensuring that air or nitrogen supplied  ;

(6.7) by each of the applicable systems to components important to safety (e.g., MSIVat SRVs; scram valves which are located outside the containment) meet the '

quality requirements (clean, dry and oil free) of ANSI l MC 11.1-1976 standards. In this context, the staff  :

finds GE's justification for limiting maximum particle size to 5 microns in the air stream at the instrument (the particle size is mentioned only for the instrument air system) instead of 3 microns as required by the 4 above stahdards unsatisfactory (See Generic Letter 88-14 " Instrument Air Supply System Affected Safety-Related Equipment"). Note that the staff will accept higher than 3 microns only if the larger size is upported by the supplier's data for all the safety-related equipment or components that are supplied compressed air or nitrogen for their operation and there is assurance that the larger size will not -

cause any equipment or component degradation with aging. Also, discuss how all the above systems meet the guidelines of Regulatory Guide 1.68.3, ,

"Preoperational Testing of Instrument and Control Air Systems." Include the atmospheric control system since it supplies nitrogen for safety-related components via the nonessential portion of the nitrogen gas supply system during normal power operation. Include the ,

service air system since it supplies air to safety-related components inside the containment during refueling. Identify applicable interface requirements for all the nitrogen or air supply systems with regard to fluid quality and preoperational testing requirements. ,

430.217 Provide description and figures showing how the four (9.3.1) compressed gas systems (atmospheric control, nitrogen gas supply, instrument air and service air systems) are interconnected. Include isolation capabilities, if applicable, between the essential divisions of the nitrogen gas supply system and instrument air and service air systems.

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.' 430.218 Clarify whether the instrument air system supplies (9.3.1) backup air to the nitrogen consumers located inside the *

(9.3.6) primary containment during normal plant operation when '

the nitrogen gas supply pressure drops below the specified set point. If so, justify supply of backup air instead of backup nitrogen inside the containment during normal operation when the containment has to be maintained inert.

430.219 Clarify whether both air compressors of the service air (9.3.1) system operate simultaneously whenever the demand for (9.3.7) service air exceeds 50 percent of the peak air consumption.

430.220 Compressed air or nitrogen supply systems designed to (9.3.1) supply fluid to equipment or components located inside .

(6.7) the containment for their operation at no more than design basis accident peak containment pressure will not be able to perform their intended function at '

higher containment pressures which may result under  !

degraded core conditions. This, in turn, may compromise the operation of the subject components.

Address the above concern as it relates to the design of compressed air and nitrogen gas supply systems. ,

430.221 Provide system P& ids for the radioactive drain transfer (9.3.3) system, which clearly show the safety-related portions-(9.3.8) of the system and the primary containment isolation valves. Provide a description of the loop seal design  !

for the secondary containment penetrations for the system which includes (but is not limited to) survivability under various modes of reactor conditions (e.g., transients, accidents) and safety claqsification (seismic category and Quality Group). Also,, provide design and expected flow capacities and sOmr  :

c4pacities.

430.222 Pr3 vide information regarding the effects t olockage (9.3.3) in any portion of the drain system, including potential (9.3.8) overflow paths.

430.223 Are the level switches for each sump of the radioactive ,

(9.3.3) drain transfer system (e.g., ECCS pump rooms, fuel (9.3.8) handling area, steam tunnel) redundant and safety-related. Do the level switches annunciate an alarm and provide level indication in the control room in case of rising water level? If they are not designed as stated above, justify the design. Also, include the sump level switches in ABWR SSAR Table 3.2-1 under " Radioactive Drain Transfer System."

Further, identify which flow transmitters located in the secondary containment under, " Leak Detection and Isolation System" in SSAR Table 3.2-1 are non-safety-related.

ENCLOSURE 1 e

.J 430.224 ABWR SSAR Subsection 9.3.8.2.1 indicates that the (9.3.3) capacity of the nonsafety-related radioactive drain (9.3.8) transfer system, in conjunction with the placement of i safety-related equipment on raised pads or grating,  ;

precludes the adverse consequences of flooding on I safety-related equipment and components. However, SSAR Subsection 3.4.1.1.2 states that the ABWR design does not take any credit for operation of the drain sump pumps to provide flood protection. Resolve the above inconsistency, realizing that the drain transfer system has to be safety-related if its operation is to be credited for flood protection of safety-related equipment and components. ]

J 430.225 Identify the system design features and their safety (9.3.3) classification (i.e., seismic category, quality group) ,

(9.3.8) provided to prevent backflooding of safety-related '

equipment rooms (e.g., ECCS equipment rooms).

i 430.226 Provide an interface requirement for the drainage [

j (9.3.3) systems for non-radioactive liquid wastes prohibiting (9.3.8) any connections to the radioactiva drain transfer system.

430.227 Regarding TMI Action Item III.D.1.1 (NUREG-0737)

(1A.2.34) concerning the integrity of systems outsida containment likely to contain radioactive material for pressurized water reactors and boiling water reactors, provide information on the following items: ,

(a) clarify whether the systems that require periodic leak testing listed in ABWR SSAR Subsection 1A.2.34 include systems-unique to the ABWR design. Include such systems if they are not currently included in Subsection 1A.2.34. Also, include containment and reactor coolant sampling systems to the above list.

l (b) since ABWR SSAR Section 5.2.5 discusses leak detection methods outside primary containment which include secondary containment, turbine building and steam tunnel, rewrite Subsection 1A.2.34 to include all the areas mentioned above (current write-up refers to secondary containment only).

(c) SSAR Subsection 1A.2.34 states that all lines which

!=

pass outside the. secondary containment contain leakage-control systems or loop seals and that these systems are discussed in SSAR Section 6.5.3. However, these  :

systems, particularly, the loop seal systems dor the secondary containment penetrations, are not discussed in the SSAR Section 6.5.3. Discuss the above systems.

ENCLOSURE 1

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.' (d) SSAR Subsection 1A.2.34 indicates that t:nder certain circumstances an affected line assoc.iated with a system may not be isolated from the secc:dary containment as part of corrective action. Explain under what circumstances this will be the case.

(e) Explain what the words " augmented class.D s(stems" mean in relation to the purchase of pressure L andary

• components of radioactive waste systems (See ABWR SSAR Subsection 1A.2.34) to assure their capability to provide integrity. -

430.228 Criteria for the design basis for protection from (3.4) external flooding should conform to Regulatory Guide i 1.102, " Flood Protection for Nuclear Power Plants" as well as Regulatory Guide 1.59, " Design Basis Floods for-Nuclear Power Plants". Modify the statement in ABWR SSAR Section 3.4 to include the commitment to meet this i Regulatory Guide.

430.229 Flood protection analysis is provided for the reactor (3.4.1) building and control building only. The ABWR SSAR scope includes structures, systems and components important to safety in this area. However, portions of other structures, within the scope of the plant-specific applicant may house systems and components important to safety (for example, the pumps associated with the ultimate heat sink). The SSAR therefore needs to specify as interface criteria flood  ;

protection design criteria for these systems, -

structures and components similar to those identified for internal and external flooding for the systems, components and structures within the ABWR SSAR scope. ,

430.230 ABWR SSAR Subsection 3.4.1.1.1 references Figure 1.2-2 (3.4.1)

(which presumably includes a reference to Figure 1.2-2a). This section should also reference Figures 1.2-4 through 1,2-7 which provide a more complete view of safety-related components located below the design flood level. Additionally, these figures should be modified to show the location of all watertight doors used to provide compartment separation and the location of raised sills for which credit is taken.

430.231 Section 3.4.1.'1.2 references flooding from a feedwater (3.4.1) line break in the steam tunnel, with data for the <

evaluation provided in Chapter 15.1. However, the evaluation is not provided in ABWR SSAR Section 3.4.1.

Provide the flood analysis for this hich energy line break.

ENCLOSURE 1

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. 430.232 Your response to Question Nos. 430.73 and 430.85 (3.4.1) (submittal dated February 28, 1990) states that the I worst possible flood (circulating water system failure) .

that~can affect the turbine building would result in a

• flood level slightly higher than grade and that all ,

plant safety-related facilities are protected against site surface water intrusion (external flooping). .

Explain how all structures, systems and components ,

(SSC) important to safety are protected against site ,

surface water intrusion resulting from the above flood level. Also, considering access openings and ,

penetrations below design flood level between the reactor building and turbine building (See ABWR SSAR Table 3.4-2), explain how the SSC important to safety located in the reactor building are protected from flooding inside the turbine building. ,

430.233 Discuss how SSC important to safety are protected (3.1.1) against flooding that may result from failure of non-safety-related plant equipment and components located outvfors (e.g., condensate storage tank).

430.234 Identify the safety classification (seismic category, (3.4.1) quality group) for all instrumentation used to alert the operator on flood situation for performing timely ,

corrective actions.

430.235 Provide flooding analyses for applicable plant areas to (3.4.1) demonstrate that safety-related equipment and r components of the fuel pool cooling and cleanup system ,

and safety-related SSC in the fuel handling area will t not be adversely affected by any postulated flooding; include flooding analysis for the radwaste and service buildings in so far as they relate to other structures

, which house SSC important to safety. Also, provide details to demonstrate that there is no uncontrolled leak path of radioactive liquid from the radwaste building under conditions of the worst-case internal flood.

w ENCLOSURE 1

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l ENCLOSURE 2 REQUEST FOR ADDITIONAL INFORMATION .

PLANT SYSTEMS BRANCH l I

430.236 Since the service building is a nonsafety-related (9.4) structure, justify its inclusion in the list of locations of some electrical modules and cables performing a safety-related function and some.

safety-related valves and dampers of NVAC systems.

Also, justify non-safety quality group classification for "other safety-related valves and dampers" for NVAC systems (see ABWR SSAR Table 3.2-1, Page 3.2-29).

430.237 Explain the words "high efficient section" occurring in 1 (9.4) SSAR Subsection 9.4.1.1.3, second paragraph. If the l above words mean HEPA filter, include it in SSAR Figure 9.4-1, and provide a table listing compliance status >

including justification for non-compliance with each of ,

the applicable guidelines identified in Positions C.1 '

and C.2 of Regulatory Guide 1.140 for control building normal ventilation exhausts.

430.238 Clarify whether (1) the two redundant safety-related (9.4.1) trains of the control room equipment HVAC system are totally independent and whether each has 100 percent -

capacity and (2) the three subsystems of the essential  :

electrical HVAC system (SSAR Subsection 9.4.1.2.3) are

• totally independent so that' failure of any one subsystem will not compromise the availability of the remaining two subsystems. Also, explain what Essential Chiller Room C (SSAR Subsection 9.4.1.2.3.2) means L since the HECW system presumably has only two safety-related chiller trains.

430.239 Provide complete system P& ids including safety (9.4.1) classification changes (i.e., seismic category and quality group) for the control building HVAC system (i.e., SSAR Sections 9.4.1.1 and 9.4.1.2). The P& ids r should show among other things (1) monitors located in the system intakes that are capable of detecting radiation and smoke, (2) capability for isolation of nonessential portions by two automatically actuated dampers in series and (3) provisions for isolation of the control room upon smoke detection at the air intakes. Also, provide complete flow diagrams for all modes of control building HVAC system operation (i.e.,

normal, accident, smoke / toxic gas removal) showing l among other things flow rates and component description tables for the building HVAC system (SSAR Figure 9.4-1 l is illegible in parts and is also incomplete).

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9 430.240 SiAR Subsection 9.4.1.1.3 states that the emergency '

(9.4.1) recirculation system includes an electric heating coil  !

whereas SSAR Figure 9.4-1 shows only a hot water system '

connection to a heating coil. The above figure additionally shows three HECW divisions whereas SSAR Subsection 9.2.13 mentions only two HECW divisions.  !

Resolve the above inconsistencies. Also, clarify -

whether the normal recirculation unit and the hot water '

system are safety-related, since their availability during the emergency mode of operation is vital to maintaining proper environmental conditions in the ,

control room and at the safety-grade filter train (Note that there is no description of the hot water system in the SSAR. This should be provided). '

430.241 Clarify whether the system air ~ intakes are provided (9.4.1) with tornado missile barriers.

430.242 For the turbine building ventilation system, provide *

(9.4.4) (1) complete system P& ids including safety classification changes and isolation and monitoring i devices, (2) complete system flow diagrams showing flow rates among other things, and (3) component description tables. Also, identify the corrective operator action following annunciation of alarms upon detection of high radiation in the building ventilation exhaust.

430.243 For the reactor building ventilation system, provide the followingt l a. Complete system P& ids including safety classification changes, isolation and monitoring devices for secondary l

containment (e.g., radiation monitors ~in the secondary

, containment ventilation exhaust, spent fuel pool and l essential equipment room area exhausts), essential l

electrical equipment, essential diesel generator, I drywell purge and reactor internal pump control panel 3 room HVAC subsystems. '

b. Some of the SSAR figures (e.g. , Figures 9.4-3, 9.4-4) have illegible portions; there is no figure in the SSAR r for the mainsteam/feedwater tunnel HVAC subsystem; SSAL Figure 9.4-3 for secondary containment HVAC subsystem does not show servicing of rooms housing redundant L

equipment for some essential systems; and the figures do not specify flow rates. Provide enlarged and legible size complete flow diagrams showing flow rates among other things for each subsystem (for guidance on  :

contents for requested response, see GESSAR-II HVAC system flow diagrams provided in the GESSAR-II SAR).

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.' c. Component description tables for each subsystem.

d. FMEA for each subsystem.

e. Description of isolation devices including safety classification, redundancy and source of power supply to the devices for all nonsafety-related HVAC subsystems that interface with safety-related structures, systems and components (SSC) (e.g., secondary containment HVAC '

subsystem, drywell purge supply / exhaust subsystem). *

f. Specific design characteristics for meeting GDC 4 l requirements for safety-related HVAC subsystems. l

g. Table listing compliance status with each of the  !

applicable guidelines of Regulatory Guide 1.140, '

Positions C.1 and C.2 including justification for .

non-compliance for the normal ventilation exhausts f: n '

the secondary containment and drywell purge 9ubsystr n (ABWR SSAR Subsection 9.4.5.1.2 refers to filters ! 'as secondary containment normal exhaust system, but dow- s~

discuss what kind thece are).

h. Discussion of smoke removal operation for applicable HVAC subsystems including how the affected area will be

isolated from other unaffected plant areas. Also, include the impact of applicable HVAC subsystems on safe l >

or alternate shutdown capability for a fire event in a "

plant area serviced by one of the applicable subsystems.

430.244 ABWR Subsection 9.4.5.4.2 states that each divisional (9. 4. 5) HVAC system consiste of two supply fans, two exhaust fans, and two recirculation units. However, SSAR ,

Figure 9.4-4 shows only one recirculation. unit per i division. Also, the-figure shows three HECW divisions  ?

supplying chilled water to the respective division room coolers; but SSAR Section 9.2.13 describen only two >

divisions for the HECW system. Resolve-the above discrepancies realizing that the safety-related support systems for the three diesel generators have to be completely independent of each other.

l 430.245 Confirm'that each supply and exhaust fan (of the l

.(9.4.5) essential electric equipment room HVAC System) mentioned e above'is a 100% capacity fan. '

l' 430.246 Discuss how the essential electric equipment HVAC

.(9.4.5) subsystem meets GDC 17 " Electric Power Systems" as it reltites to the protection of essential electrical l

components of the subsystem from failure due to the accumulation of dust and particulate materials (see SRP Section 9.4.5, Acceptance Criterion II.4 for required contents of response to this item).

ENCLOSURE 2 e - . - - . _ _ .L- _.

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. 430.247 Subsectioli 9.4.5.4.,5 does not discuss temperature .

(9.4.5) control. Provide a discussion of the method and instrumentation provisions for temperature control.

430.248 Provide a discussion of the means used for maintaining (9.4.5) the rooms cooled by the essential electrical equipment HVAC system at positive pressure. .

430.249 provide assurance that the air intake elevation for the (9 4.5) essential diesel generator HVAC system is greater than 20 feet above grade or discuss the methods for protecting electrical panels from dust and particulate materials.

430.250 ABWR SSAR Subsection 9.4.5.5.2 states that the two supply (9.4.5) fans for each of the three diesel generators take air from the outside and distribute it to the diesel ,

generators. Clarify whether there is a common header.for '

all the diesel generators for intake air. '

If there is, justify such a design.

430.251 Provide 3rawings for the drywell purge supply / exhaust (9.4.5) system and a discussion of the interfaces to the secondary containment HVAC system and to the standby gas. '

treatment system, 430.252 Discuss the sensor location and actuation setpoint for i (9.4.5) the exhaust radiation monitor for the drywell purge supply / exhaust system as they relate to preventing +

unanticipated radioactive releases.

430.253 Since there is a separate wetwell purge supply / exhaust (9.4.b) system for the ABWR, include a description of that system in the SSAR. Note that all the information requested above for the drywell purge system should be included in the description of the wetwell purge system.

430.254 ABWR SSAR Subsection 9.4.5.6.1.2 states that the drywell (9.4.5) purge system only operates during plant shutdown.

Correct the above statement since it will operate also during inerting, deinerting or pressure control of the primary containment. Also, discuss how both the drywell and wetwell purge supply / exhaust subsystems together meet Branch Technical Position CSB 6-4 " Containment Purging During Normal Plant Operations."

430.255 ABWR SSAR Subsection 9.4.5.1,2 states that two fr.n coil (9.4.5) units provide cooling to the steam tunnel. ExplGin how the air is cooled.

430.256 ABWR SSAR Subsection 9.4.5.8.2 states that each division (9.4.5) of the reactor internal pump (RIP) control panel room '

HVAC subsystem contains two recj'culation units. This  ;

doen not agree with Figu'.e 9.4-5. Resolve this discrepancy.

ENCLOSURE 2

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j 430.257 -ABWR SSAR Subsection 9.4.5.8.3 addresses the <

(9.4.5) non-essential equipment HVAC system instead of the RIP control panel HVAC system. Provide a safety analysis which addresses the proper system, including a discussion i of the effects of loss of ventilation on the RIP control ,

panel.  ;

430.258 For the radwaste control room and balance of the radwaste >

(9.4.5) building HVAC systems, provide (1) complete P& ids showing safety classification changes, isolation and monitoring devices, (2) complete flow diagrams showing among other things flow rates, and (3) component description tables.

Also, clarify whether any affected space is isolated by safety-related devices.

43C.259 ABWR SSAR Subsection 9.4.6.2.2 states that one "Ldwaste (9.4.6) building HVAC supply and exhaust fan are norLally operating and the other of each type (i.e., for the >

radwaste control room and the balance et the radwaste building) is on stondby. SSAR Subsection 9.4.6.3  ;

': mentions prevision for automatic start of the standby unit. However, SSAR Subsection 9.4.6.5.2 indicates that only an alarm is actuated by low flow in the exhaust fan discharge duct, and that ventilation must be restarted manually. Clarify whether the standby fan is started on failure of the operating fan. If not, provide justification.

{

430.260 Provide a failure modes and effects analysis for the l (9.4.6) radwaste building HVAC system which shows that the normal direction of air flow from areas of low potential contamination to areas of higher contamination will not be reversed for the failure of any active component.

1 430.261 For both of the radwaste building HVAC system zone (9.4.6) exhausts, provide tables listing compliance status

• including justification for non-compliance with each of the applicable guidelines identified in Positions C.1 and C.2 of Regulatory Guide 1.140.

430.262 For the service building ventilation system, provide (9.4.8) complete system P& ids including safety classification changes, isolation and monitoring devices, (2) component description tables, and (3) compliance with applicable guidelines of Regulatory Guide 1.140 for the system exhaust. Also, provide legible and enlarged portions of the SSAR Figure 9.4.7 which are currently illegible; include flow rates in the figure. 430.263 Provide enlarged and legible version of the drywell cooling system (9.4.9) P&ID (SSAR Figure 9.4-8).

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ENCLOSURE 2

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- 430.264 Identify the HVAC system that will service the remoto (9.4) shutdown panel area that will be uses for providing alternative shutdown capability following certain. fire events.

430.205 Identify interface requirements as they relate to HVAc i (9.4) systems for plant areas which do not fall within ths ABWR design scope but which may impact the SSC that are within the ABWR scope. Also, provide interface requirements for the technical support center (TSC) HVAC system.

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2NCLOSURE 3 kEQUESTlFOR ADDITIONAL INFORMATION SAFEGUARDS BRANCH i

910.17 General Electric Co. stated that in-plant security.

communications requirements- of 10 CFR 73.55(f) are outside.the setp> of the ABWR Standard Design-and would F be the responsibility of,the certification Osers. NRC Information Notice 83-83, "Use of Portable Radio Transmitters Inside Nuclear Power Plants," discussed concerns about the potential for radio frequency interference (RFI) from portable radio transmitters to  ;

cause reactor. system malfunctions and spurious actuations. A capability for continuous communication between security personnel on pa: col within vital ~ areas of the-plant and the security alarm stations-is required by 73. 55(f) (1) . Common practice is to use hand held radios to meet this requirement; LAs noted in Information Notice 83-83, administrative prohibitions ,

on the use of portable radios in certain areas ofLthe j

-plant may not. adequately resolve the concern e .

particularly for new designs that r .)ue extensive use of solid state devices in instrumentation and cons.ol-circuits.- The ABWs Licensing Review Basis (August, 19E7) stated that the ABWR SSAR will not provide b

details but will' identify design requirements for 73.55(f). Please address design requirements to assure l that means can be provided for continuous communication ,

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! between security personnel stationed within, or on l

patrol within, vital areas of the plant'and the security alarm stations, without interference with plant instrumentation and control. l 910.18 Generic Lecter 37-08 states that ' an uninte:cuptible 4 <

power supply is preferred for alarm annunciator

equipment and non-portable communications equipment. '

Industry standard' ANSI /AMS-3.3-1988 states that. i intrusion detection aids ie,g., door alarms, fence alarms, and the alarm assessment (closed circuit-television) system) should Olso be supplied with' l, uninterruptible power. Regulatory Guide 5.65 notes u L that an uninterruptible power supply for electrical

?. locking. devices on vital area doors is an acceptable method for providing the prompt access to vital l

b equipmont required by 10 CFR .73.55(d) (7) (ii) . Section 8.3 of the A-..d SSAR discusses onsite power systems, including non-class 1E vitalt AC power 'for important h non-safety related loads, but makes no mention of security system power requirements. The draft l'

~

E EPRI-ALWR Requirements Document quoted in ABWR SSAR Li Appendix 19B says that the security power subsystem  :

shall be a non-interruptible power source.. Therefore, we request you.to discuss whht provisions for these l

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security systems have-been provided in the standard- '

design, and provide interface criteria that will allow-

. the security requirements for these systems to be  ;

accomplished without adversely interfering with safety '

systems.  ;

910.19 Explain why the environmental conditions parameters of ABWR SSAR Appendix 3I should not apply to the design and qualification of security access control ' systems. ,

Consider desirability of operable card reader-controlled door locks in the event of a pipe break, 'I such as occurred at Surry (NRC Augmented Inspection Team Report 50-281/86-42).

910.20 . The list.of vital areas and vital equipment in . ,

(13.6.3) Subsection 13.6.3.3 appears to include all of the=

reactor coolant pressure boundary, including appropriate motor cor. trol centers and power supplies; systems required for mitigation of transients; and support systems (e.g., cooling water, instrumentation, j control power) necessary for these systems to' operate;

, as well as other safety related systens. Are there any i exceptions to this statement?

910.21 . Subsection 13.6.3.4 specifies door alarms only i (13.6.3) for doors at card reader locations. All doors and hatches connecting vital-areas to non-vital areas should be alarmed (e.g., balanced magnetic sw. itches wich tamper-safe cabling), not just doors at card-  !

reader locations, with the alarm hardware being on the i vital side of the door.

910.22 10 CFR. 73.55 (d) (7) also requires provisions to (13.6.3) accommodate the poteN.ial need for rapid ingress'or L egress. Emergency Axits shouldl include provisions for

! exiting without use of keys or card readers. Please p include appropriate language in Subsection l'3.6.3.4.

g 910.23 Certain rooms are identified in Cubsection'13.6.3.6, (13.6.3) Bullet-Resisting Walls and Doors, Security Grills and-L Screens, as: i i ... a particularly high security zone. Specific.

l precautionary measures have been incorporated into h

the= building design to minimize forcible' access to-this area."

l- This seems to confuse two requirements of 10 CFR 73.55.

K Bullet-resisting barriers are required by 10 CFR

73. 55 (c) (6) for the control room. Accert '7 to 10 CFR 7 3. 55 (c) (1) , access to all the vital areas identified in~ Subsection 13.6.3.3 requires pasoage through two -

physical barriers of sufficient strength to meet the performance requirements of 10 CFR 73.55(a). As noted ENCLOSURE 3 h

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' ? 'in. Regulatory Guide 5.65, a vital area barrier is to be j constructed of-materials that provide delay to forcible i access from non-vital areas. 1 910.24- The revision to Subsection 13.6.3.6 needs further (13.6.3) clairification. Will the design of air exhausts, NVAC gratings, and other man-sized (i.e., 96 square inches)

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openings, in all physical barriers that separate vital areas from non-vital areas, satisfy the criterion in ,

NUREG-0908 and Regulatory Guide 5.65 that'the integrity of a vital area barrier containing them not be decreased? '

s 910.25 The meanings of some statements in Section 13.6.3 are (13.6.3) unclear and may be unnecessary '

(a) Subsection 13.6.3.1, Introduction, includes "the capability for detection of inoperability of vital -;

equipment" as c concern of'the physical security der,ign. requirements, is this what it was Jeant to say? This is not typically a physical security  ;

function. What portions of Chapter 7 discusses this?

(b) What is the intent-of the~last sentence of Subsection 13.6.3.3:

"Hence, access control.is considered separately."?-

(c) The interface requirements of Subsection 13.6.3.7, Compatibility with the Remainder of the Plant, would be covered in the site security plans -

required by 10 CFR 50.34(c) and (d).- Section 13.6.2 already states that the security plans are out of the scope of the ABWR Standard Plant design certification which means they would be required-to be provided by applicants referencing the J certified design. Of the eleven items = listed, ,

, only (3 appears to be a unique ABWR. interface.

requirement. If 13.6.3.7 is intended to clarify what additional' security requirements those applicants would need to satisfy, the list is ,

incomplete, as it omits lighting and other  :

requirements.

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ENCLOSURE 3 I

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