ML20059N539

Forwards Chapter 9 Responses to Request for Addl Info on Ssar for Advanced Bwr,Per DC Scaletti

ML20059N539
Person / Time
Site:	05000605
Issue date:	09/28/1990
From:	Stirn R GENERAL ELECTRIC CO.
To:	Chris Miller NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
MFN-123-90, NUDOCS 9010160153
Download: ML20059N539 (97)

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GE Necker Energy

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Gewar fue como 3 d;

m can. Aen., su u, c4 33n3 l Septembcr 28,1990 MFN No.123 90 -[

Docket No. STN 50-605 i EEN 9052  ;

-Document Control Desk U.S. Nuclear Regulatory Commission Washington,D.C.- 20555 e ,

1 i

Attention: Charles L Miller, Director i

!' Standardisation and Non-Power Reactor Project Directorate i J

Subject:

Subadttal of Responses to Additional Infbrunation as Requested la NRC IMter from Dino C. Saletti, Dated August 15,1990.-

.. . I

Reference:

1. Submittal of Responses (Proprietary Information) to j Additional Information as Requested in NRC Letter from i Dino Scaletti,' dated August.15,'1990, MFN No.124 90,- 1 dated September 28,1990 : _,  !

t

2. Submittal of Responses l Additional Information as(Proprietary Requested m NRC Information)

I.etter from to ~ ,-j Dino Scaletti, dated August 15,1990, MFN No.- 125 90, j dated September 28,~ 1990 P i

Dear Mr. Miller:

4

. Enclosed are thirty four (34) copies of Chapter 9 responses to the subject Request for Additional .

vanced Boilin(g Water Reactor (ABWR).Information RAI)(Enclosure 1) on the Stand '

1 Response to Question 430.197, the ultimate heat sink pot'Jon of Question'430.204 and the (

tabulation for instrument air consumption during normal operation will be provided by October 31,

~

-l' 1990.

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Response to Questions 430.218,430.219,430221,430.222,430.223,430.224,430.225, and 430.226 contain information that is designated as Gmeral Electric Company proprietary information and is N being submitted under separate cover, j

It is intended that GE will amend the SSAR with these responses in a future amendment.

Sincerely, a

.C. Stirn, Acting Manager 1

Regulatory and Analysis SeNices .i

- cc:- F.A. Ross (DOE) I 1

D.C.Scaletti NRC D.R. Wilkins((GE)) - 0 >

J.F. Quirk (GE) f]f Q0 { i

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QUESTION 430.177 ABWR-SSAR Section'9.1.1.1.1,: Nuclear Design, states that j since no credit is taken for neutron leakage,'the value for )

effective multiplication: factors are-really infinite neutron. J

, multiplicationrfactors. . ABWR SSAR Section 9.1.1.3.1, l criticality control,< states ~that'k,gf for both normal and abnormal.stcrage conditions will"be Ioss than or. equal-to .j

.95. However, the:same.section> states'that the new fuel storage area will accommodate. fuel with a kint < 1.35 wi th j no safety; implications. . Resolve this discrepancy. (9.1.1) L

j 1 RESPONSE 430.177- ,' ;e The fuel-storage racks are capable'of. storing l fuel which has an infinite lattice kinf of 1.35, calculated in thatuncon-c 1

j  :,

trolled reactor core-geometry at 20 0C. The storage' rack ge-- i ometry reduces this k ' f to less than;0.95 in the storage. 1 rach. Subsection 9.1.in1.3.1 has been changed accordingly.  !

y QUESTION 430.178-  ;

- ABWR SSAR Section 9.1.1.1.6', Dynamic Analysis, refers =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 O

.1 system. (9.1.1)

.?

RESPONSE 430.178~

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Response'to this question is provided-in" revised Subsection. j 9.1.1.1.6.  :

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QUESTION 430.179  :  :

ABWR SSAR Section 9.1.1'.1.7, Impact Analysis, also refers to. j a nonexistent ABWR SSAR' Section 9'.1'.2.1.7. ' Provide ' impact '  !

analysis for 1REAnt loads up to and including:- a fuel'.assen- .; .

bly and its carrying fixture. (9.1.1)

RESPONSE 430.179 ,

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L Response.to this questionLis'provided inirevised Subsection: '

9.1.1'.1.6. (Note Subsection 9.1.1.1.7 has been. deleted)'.  ;

QUESTION 430.180

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Provide details of assumptions and. input parameters used-in i I

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the criticality analysis for new fuel storage.- Include in-I q

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i formation such as number of racks, their material (e. g., j stainless steel ?), number of fuel assemblies per rack, 3

' neutron-absorbing material and its placement, placement. of- l fuel . assemblies center-to-center . distance between rows and l within rows), an(d effect of spacing on k fg in normal dry a condition or when completely flooded with water., Also,

. clarify whether the spacing ts sufficient to ensure a k gg.

1

-l of. 0.98 or less under optimum moderator conditions (foam, '

small droplets, spray or, fogging) as described in SRP Sec- ,

tion. 9.1.1. . Clarify whether the racks are designed to pre-clude inadvertent placement of a fuel assembly in other than prescribed locations. (9.1.1) .

-RESPONSE 430.180 Response.to this question is'provided in revised subsection.-

]

9.1.1.1.1. ~

QUESTION 430.181 i '

How is the new fuel protected from. internally, generated I

missiles and the effects of moderate or high energy piping or rotating l machinery in;the vicinity of the vault housing 1 the new fuel storage racks. (9.1.1) .;

RESPONSE TO 430.181 l

O' The new fuel vault is' located within the reactor building on the refueling floor (-see Figure 1.2-12 . There are no high  !

energy or : moderate energy pipes or ) rotating machinery  :

located in'the vicinity:of the new fuel-vault. .?

QUESTION 430.182 4 Provide information'on show the design;of;the new fue1~ {

storage facility complies with GDC 61, " Fuel Storage and. :l Handling.and Radioactivity Control." Identify the' l ventilation system provided to handle possible' release off i

radioactivityaresulting.from accidental damage to the-fuel- .

(note- that ABWR SSAR 7.1 does: not describe ~ the radiation  !

L monitoring equipment for'the new, fuel storage areaLas stated- '

in'ABWR SSAR Section. 9.1.1.2).=(9.1.1)

RESPONSE TO 430.182! n The reactor building HVAC system monitors: the building. J exhausts for radioactivity. 'If radioactivity ~is

' encountered, the system is isolated and the SGTS system will 6

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. start operation. This~ prevents any possible release of .

radioactivity from any fuel handling accident. j QUESTION'430.18? f!

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Provide sufficient information and drawings, to determine '!

that the failure of non-seismic systems and structures in  !'

the vicinity of the new fuel storage facility can not'cause an unacceptable increase in ketf. . -(9.1.2) j RESPONSE,TO 430.183 'I i

The new fuel' storage. facility;is located on;the~ refueling )

floor ? of the ' reactor building -(see, Figure 1.2-12) . The  !

reactor building is'a seismic-category I building protecting. l the new fuel from seismic- events and externally generated- l missiles.- There are no'non-seismic systems in the vicinity of the new-fual storage facility. .

QUESTION 434.184 s

Demonstrate ' that the - analysed '.in pact . of a fuel assembly, l

~ including its associated hand 1.ng tool, dropped from a  !

height of 6 feet bounds tho' range of'all possible load drops

-from all possible heights.T For additional guidance on the t required bounding. analysis, see SRP'Section 9.1.2, Item j Q III.2.e.(9.1.2).

RESPONSE TO 430.184 j-  :!

i As discussed near the and of. Subsection 9.1.4.3, light loads such as the blade guide,- fuel support casting, control rod s

l or control rod: guide. tube' weigh considerably less than a- 1 fuel bundle and are administrative 1y controlled-to eliminate 1 the movement of any' light. load,over the fuel. pool above the elevation required for fuel assembly . handling. Thus, the kinetic energy of any light,, load would be less:than a fuel bundle and would have less anage induced.- ,

QUESTION 430.185 ,

Provide sufficient' informdtion and drawings to 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 k gg. . (9.1.2) 1

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l l$ RESPONSE TO 430.185 The, spent fuel; storage pool is located-on the refueling floor of.' the reactor building :(see Figure 1.2-12) . The-reactor. building is a seismic category I building protecting.

,j the.. spent-fuel,from seismic events and externally generated, l missiles.- There are no non-seismic systems in the vicinity '

il of the spent; fuel storage pool. -

QUESTION'430.186' T Provide drawings .and information - pertaining . to' spent fuel iI p transfer' canal capability of the fuel transfer canalt or. I

?other. provisions to prevent a dropped shipping caskofrom causing an unacceptable loss of pool water.(9.1.2)' j?

RESPONSE 430.186  !

. The.: shipping cask'is placed in' a walled of f and drained; The drained volume is, j

r portion of the: spent fuel pool.

-l flooded, and the Seismic category I gates removede . The ' ,;

spent fuel is then transferred.. This process is reversed; to i remove.the cask. The ratio of the-two volumes'is such that '!

f ailure of the . gates 'will.. not lower the water level enough I to be unacceptable. Interlocks on the= main crane prevent j.

t the shipping cask.from being carried over any,other portion- '

~ '!

of the spent fuel storage pool. l QUESTION 430.187 clarify'whether there is.a) an interconnecting fuel' transfer I canal capable of being isolated from the. fuel-' poolo and . 1 adjacent cask _ loading' area, and b)'any high-energy. piping'or j rotating machinery in the vicinity-of.the fuel storage 1 pools. Also, clarify whether_the. racks.areidesigned to i preclude inadvertent placement of a fuel. assembly-;in; other-than prescribed locat:ons.(9.1.2)  ;

RESPONSE TO 430.187-(a) AsL shown in Figure 1.2-12 the spent fuel pool and-adjacent cask loading area are se,parated by seismic category

' I gates.- These gates isolate the-cask 1 1oading' area'from the-spent fuel pool.

(b)- No high energy or moderate energy piping. or; rotating machinery are located in the vicinity _of the spent fuel pool or cask loading area on.the refueling-floor.. .

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l QUESTION 430.188 J

l Describe the function = of the containment pool mentioned sin ' l ABWR SSAR Section 9.1.2.1.5.: (9.1.2) ,j

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RESPONSE 430.188 '

- Section 9.1.2.2(3) has been changed to omit mention of a' ,l containment pool'-

~

' QUESTION 430.189; 4

- What,is the seismic' category'of the gates in the pools?, l (9.1.2) "

RESPONSE TO 430.189- l The gates between the spent > fuel pool and ~ other pools are )

all Seismic Category I.  !

QUESTION 430.190  !

'~

1 Instead of referring to a" specific: GE proprietary report on !

j criticality control for spent fue1L atorage~ (see ABWR SSAR -

Section 9.1.2.3.1), provida datails of assumptions ~and input l

'O parameters fuel storage.

used-in the cpiticality analysis of the spent; Also provide the,Lucertainty_value'~and-asso-q J

ciated ~ probability and , confidence level for the - k gg value  !

determined by the analysis. -Inc1 tde information such as '

number of fuel assemblies stored.in the pool, i center-to-center spacing'between,fust assemblies, material .,

of the racks, neutron absorber used and.its1. placing, and. p k,gf for the abovelcondition;when the--storageEis fully _'

1 loaded and flooded with non-borated water. (9.142):

RESPONSE 430.190 ,I Re.sponse to this. question ,is provided-in revised Subsection' J

9.1.2.3.1.-

• i .

J QUESTION 430.191 . 4

~)

List the specific provisions included-in the design ofithe. i 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) . Iden- l tify the ' corrective actions on - detection of : loss of decay ; l i

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L L, heat removal-capability or excessive radiation levels. Note that for radiation monitoring systems, additionally refer- s encing ABWR SSAR Subsections 11.5.2.1.2.1 and 11.5.2.1.3, if they are applicable, in ABWR SSAR Subsection 9.1.2.4 is suf-fic.ent. (9.1.2)

RESPONSE 430.191-Subsection 9.1.2.4 has been . revised to reference subsection 9 1.3 for the description of the pool linernleakage'detec- .i tion system and water level monitoring system. . Subsection:

9.1.2.4 has also been revised to reference" Subsections 11.5.2.1.2.1 and 11.5.2.1.3 for a description of theiradia- 1 tion monitoring systems.

. i Subsection 9.1.2.4 has been revised to reference ' subsection -

9.1.3 for the . corrective action for loss; of ' decay heat. re-z moval capability and Subsections- 11.5.2.1.2.1 and 11.5.2.1.3 '

for the corrective actions for excessive radiation, levels. I 1

QUESTION 430.192 Provide the results and conclusions'of4 tho' load' drop-analysis which considers dropping of ' one - fuel - assembly and 4 its associated handling tool from a heightLat which it is

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norsally handled above the' spent fuel storage; racks., : ABWR -

i SSAR Subsection 9.1.4.3 ' does not discuss compliance (with. l GDCs 61 and 62; therefore, discuss the.above complianceifor. l the' light load handling system. (9.1.4). ~

RESPONSE TO 430.192 Response to the load drop ana19 sis portion-'of this' question '!

is provided in revised Subsect:.on 9.1'4.3.- . Compliance-with- '

GDC 61 and 62 are discussed in Subsections 3.1.2.6.2.2.1 and'

- 3.1.2.6.2.2.2, respectively.

l QUESTION 430.193 A " slack cable" signal is not considered sufficient indica-tion of a fully seated assembly. Discuss whether positive-

l. vertical position indication wil1~also be.provided. (9.1.4). ]

RESPONSE 430.193 In addition to the slack cable signal,:the elevation:of the grapple-is continuously indicated. Also,,after the: grapple-

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is' disengaged, the position of the upper part of the fuel l bundle can be observed using television.. '

QUESTION 430.194 l

.ABWR'SSAR. subsection 9.1.4.2.2.1, Reactor.: Building Crane, j

indicates that the crane can be used to move new fuel to the '

spent 1 fuel pool-and is also used to handle' the spent fuel ~

cask _over the spent fuel pool and'results of a fallure modes  ;

and. effects analysis demonstrating'the adequacy.of controls' .i and interlocks to prevent compromising _criticalityJor 1 radiological safety. (9.1.4)

RESPONSE TO 430.194 The reactor building crane main hook is L . used to move the- 3 spent fuel. cask, and the auxiliary hook'is'usedtto move new '

fuel from the new fuel vault to the' spent. fuel storage pool.. j Interlocks and procedures prevent tho' main-hook'of;the i reactor building crane, while^ carrying a heavy load, to ,

traverse over the spent fuel pool or thel new fuel storage-.

vault. ,

4 As discussed in Section .15B.1, - FMEAs are provided for ~two- '

[ ABWR systems and one major component: which present _ a?

significant change from past ABWR design =, Specifically, FMEAs are included in' Appendix 15B for:- '

(1) control rod-drive systems (with emphasis ~on the fine motion control rod drive), '

3 (2) essential multiplexing system,- and 4

(3) reactor internal pump. j Regulatory Guide 1.70 requires? FMEAs to 4 bel performed on s i selected subsystems.of' Chapter-6, 7;and19. However,~GE' 'I considers that the plant nuclear safet operationalianalysis >j (NSOA) of appendix 15A and the probab listic~ evaluations'of- -1 Appendix'19D adequately address single failures for those :l systems and components'which are-similar-to past'BWR- :l designs. Since the designs of the ABWR reactor building ~1 crane'iscsimilar to_past= designs,: GE believes that>it-is' unnecessary to perform a FMEA on the reactor building < crane.

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i QUESTION 430.195  !

clarify whether the system design includes interlocks (1) to f

- . ensure correct sequencing of the transfer operation in the  ;

automatic or manual mode, and -- (2) to prevent the! refueling platform and the fuel handling platform moving 'in 'the trans-for area during operations of-the transfer system so'that

- the transferL system will' not = be adversely affacted by 'the .

presence.of either platform. (9.1.4)' ,

t

, RESPONSE 430.195- ,

-Interlocks shall-be provided to ensure correct sequencing of' ,

the transfer operation in the automatic.or manua19 mode..  :

i The terminology <of the various platforms has been clarified ,

in the response.to Question 430.198a. . )

. QUESTION 43041962 '

i ABWR SSAR Tables.3.2-1 (page ,3.2-28)' and 9.1-2 i differ in i seismic classification identification.-for some fuellservic-ing equipment'

. Correct the: discrepancy as' appropriate.. i (9.1.4). '

f RESPONSE 430.196 , '

5 Table 9.1-2 has been changed to state that the. seismic cate-

. gory of the fuel preparation machine-is non-seismic-to agree-with Table 3.2-1. .

i QUEST'ON I 430.198 ABWR SSAR Section 9.1.4 is confusing on;the fallowing de-tails -(9.1.4) 3 Ll

[

(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' any-where under that caption.'It is not clear..what1 constitutes -

the fuel handling platform and whether.itLis distinct from j

.the refueling platform.- l (b)' -ABWR'SSAR Table 9.1-10 refers to:threeL single-failure-proof. cranes the' reactor: building crane, refueling. bridge crane and fuel 1 handling jib crane.: ABWR' s SSAR Subsections 9.'1.4.2.7.1 ' and 9.1.4.3 refer to the auto-  ;

matic refueling machine. (a gantry crane) and the spent-fuel '

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tors mean the same load handling device.

(c) Different subsections in ASWRl 8ection ' 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=spentifuel pool.-

Provide clarification on all the above. Also, provide layout draw-

- ings for all'the storage pools;-including the upper pool _and.

the transfer. canal.

RESPONSE 430.198:

(a), The following terminology will now be used;in the ABWR ,

88AR:

The refuelina elatferrn lis mobile and moves between the -

reactor well pool and the spent. fuel pool. '

The auxiliarv elatform is stationary and located at the reactor vessel, flange.

The under vessel- miatform is located under the reactor vessel and can be rotated.

O- (b)

Table 9.1-10 and subsection 9.1.4. have been changed to use the following terminology for the ' single-failure-proof -

cranes. '

(1) reactor building crane (2) refueling platform crane The jib crane has'been deleted because its function will be performed crane.

by an auxiliary ' hoistLon-the. reactor building y 1

(c) Subsection 9.1.4 has tieen changed: to use' only the term spent fuel pool-when referring to the pool where spent fuel- f j

is stored. . The relationship between'the various pools ist R showls in Figures 1.2-2 and 1.2-11. : EThe~-transfor canal)is-that part of the, spent fuel pool at an elevation.of T. M. 8.

L. 23700 ' between the reactor well pool and the deeper part where the spent fuel is' stored.

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y M. cQUESTION .! 4'J0.199

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• Includel the single-failure-proof characteristic's of 'alli 1 j

. cranes-- used 'in : light - load handling ' -(o note'.that ABWR SSARi subsection . 9.1.4.1 mentions: only hoists; on the L refueling.

Platform). (9.1.4) g RESPONSE TO 430.'199- ,

'Only hoists- are: required on'the.ABWR for light-load handling.

4

' QUESTION 430.200-

._ - : 1 ABWR SSAR. subsection' 7.6.13does not provide; an evaluation of tho' radiation monitoring equipment foritheNrefueling and service: equipment,aststated:in;ABWR'SSAR subsectionl a 9.1.4,,5.4'. Provide the1above,information. TIf itlis:coveredo -i-

-l by some . other radiation 1 monitoring ? systems i(e.. g. , L area - ra--

diation bonitoring . system. and/or; process' and Leffluent moni' toring-system- or both) , ' include -reference lto thosel systems J

1 and the-applicable SSARDSections'in SSAR. subsection

9.1. 4 '. 5. 4 . - ( 9.1. 4 )

y 3 RESPONSE 430.200- ,

Subsection 9.1.4.5.4=has4been changed to identify'the radia-i

,scribed.

tion monitoring equipment and reterencelwhere,it"is de -

QUESTICH 430.201:- '

The. interface criteria of LABWR SSARIS ection! 9.2.15' does not-i include the required interfacelcriteria.for the: design;of' the potable and sanitary water. system. cTo meet /the require-': '

ments of-GDC'60, the design'ofithis systemishouldenot, allow

' for interconnections, between- the potable -'and: sanitary watery system ' and systems . having L the potentiali for containing ? ra- _ ' . .

l+ dioactive:materialsW ProtectionLshould'be provided through  :

the use of air gaps, where necassary.p;Addithese' design cri-

-teria, asJinterfaces,funderjABWR SSARiSectiont9.2.15.

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'RESPONSEc430.201- '

E ;I w -Thalinterfaceiraquireme'ntscfor the potablefand sanitary?

i waterLaystem. of Subsection 19.2 4~has been:added astnew'Sub -- '

-E  ; tsection) 9.2.17.3. One of;these requirements 1willibe the1use' l n '

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active < materials into the potable and sanitary water system.,

QUESTIONJ430[202

.m

Includesthe following interfacesJbesidesLwhat have'beenLal .

ready specified , for ensuring 4 theLultimate heat? sink ((UHS) . sr capability: '1 l passive ' compo(ne)ntsi in. electricalLsystems.' Design;toiaccommodate' single!failu safetyerelated(portionsi from adverse envi(2)" protection! of;2' ronmental1condi tions including those resulting from piping'fallures.t(3)' '

c.

Time-duration ofnUHS-cooling . capability availability.-

(9.2.5, 9.2.15)

~

.v s RESPONSE 430.202 ,

'SubsectilonL9.2.15Lhas been changed toTdiscussJthe; following:; >

r ,-

q (1). Single f ailuresf of < passive - components- in electrical 4

+ W systems will.be discussed.' These; failures;willilead to.'the

' loss oftthe affacted. pump,4 valve orfother component andithe- "

partial-'or; complete loss of' cooling; capability of'that!. divi- '

!sion. However, allisa'faty-related ; heat rejectionLaystems i

areeredundant so thattthe:es'sential cooli(ng function;can be

performedieven with theicomplete loss:of'one division;; ;f y ,

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g (2) Safety-reldted" portions ofI the UHS. shall belprotected ~

from adverse-environmentaliconditionsl including those're-- h j

sulting irom-piping'failuresJbyilocating such componentsiin N t!

a. Seismic Category I building and providing-flood protection-

'inl case of piping failures.L '

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(3)' The ; time . duration' of. UHS 1 cooling ciapabiliky avaikabil .

~ity is-thirty days.-

o -

QUESTION.430.203,  ;

3 g,

. . , c .

1 The ultimate heat sink heat ' . load frequirementstarej identi-fiediby reference toLABWR'SSAR'Tablez9.2-4. .. ThisLeetLot j three tables J(9.2-4a, ' 9.2-4b Land? 9.2-4c) Ride'ntifiesi heat !

'loadst forceach of ' the ? throed reactor- building l cooling l waterl ,

divisions.- These tables de not consider:the case'of alreac- 1 tor.) shutdown at 4 ' hours after ~ai blowdown tosthe' main 4 con s -

Ti

' denser. 5 . Inclusion of the above may;l require Laf highornheat ?

q load . dissipationLcapability for - the LUHS: than ' what hask been t

-1

currentlysestimated (see GE's L responseL toi QuestionTNo.:

4 440.73). L Revise the tables,as appropriate?considering1the.  ;

above' case and: provide the heat load' requirements'-' based?on i +

n 4 o i

1.

2

)

g m

e, Qp'

H a m -: * ,

}{ '

,  :: x  : w . .-.- ._ . -  : -' ~~~ u.--= - ' ;

( '+

j

, , i y .r . Li II

~

the" revised tables'for the1 ultimate heat.sinkf(e. g. ,: tho' '

sum of the heat: loads for allethree divisions,.22of 3).- Are there additional-heat < loads associated with the UNS not car--

+-

ried=by the reactor building cooling' water. system 7 (9.2.5)-

RESPONSE 430.203 4 *

Tables : 9.2-4a, 9.2-4b ? and 9.2-4c ;were changed L in GE letter.

l dated May.31, 1990 to list,the heat loads for,the. case of.a-reactor shutdown at 4' hours'after a blowdown-toMthe. main:

1 condenser.

~

.1

Subsection 9.2.15'has'been changed to' discuss the heat l loado l{

dissipation capability.of'thecUHS during the casesJof=reac- '

tor shutdown atL4thours.after blowdown to the main' condenser! 1 and IOCA.t y '

Therel are Eno. additional heath loads of - the UHS (that are not 1 carried by.the'RCW system. o J

1 /

. QUESTION 430.204: Lg .)

The requirements of,10CFR52? include!theLneedOfor?a' concep- 3 '

tualedesign forJsystems noticonsidered:toibeiwithinLthat

. design scopeLof a= standard nuclear: power plant'. - :- Noisuch

~,

C  ; conceptual ~ design,has,beenfincluded as?part of thelABWR'SSAR1 for either'the; UHS;or tho' interfacing service water, system.,

Provide conceptualLdesigns forethe: UHScand theCinterfacing/

O service water system..(9.2.5)

RESPCNSE 430.204-

,l The conceptual design of the' UHS' will?be, provided Lin a:

future submittal. Subsection 9.2.5:willLnotibeipart of the- '

plant; certification.. .The-design 1ofnthe' reactoriservice- jl water.' system is provided'in: Subsection 9.2.16'andVis;part:of-the plant; certification.

if~

~

QUESTION 430.205~ 4 j

The make-up water- preparation" system is tidentified as out-:

-side the. scope.of'ABWRsstandard plant. :Thisisystem'should: 4 y

E meeti the requirements 'of ' Position; C.2 L of1 Regulatory Guide - ,,

1.29'.-~ Provide an'interfaceLrequirement thatithe failure:of the!make-up. water preparation:systemLwill1not resultfin the-  %

failure-of any~ safety-relatedLstructure,.systemfor compo-nontM(9.'2.8) t i

w N

, 'i, 'N .. .- - - . . - , N-ui ~ - -  :- - - -' 'N

..-.-.~.m.,

M,g m ,-.~..z w_2 - --~y _ m o_.w..c g . ..- u._uw _.ma

y .![' 4 , , q

[g 1

, s .

gm f, ,

+

Q .

f y >

,d si

/

E ,

' RESPONSE 430.205 \ g 1 ,

l m 1 The.make-up water preparationisystemiwill*be'.locatediin'a:- .. 1

' building; that does not; house any: s'afety-relatede structures,'

f'  ; systems or. components. Interfacefrequirements, includingt i '

p flood protection measures. for the make-up ' water' preparation ' ;w TF system are discussed /in' Subsection'9.2.17.c '

?

9 QUESTION 430.206 4 ,

n 4

1. , . i, ,) m

>d clarify how the turbine ' building- coolingt water (TCW); system; o . .;

meets Regulatory Guidei 1.29, Position .C.2 e with respect? to . J q seismic-requirements for'non-safety-related systems.that due '

c toLtheir failure during seismic 7avents mayfadversely" a impact: '

y' structures, systems.or components"important-to safety. ,J 1(9.2.14)- y ,', * '

n .m 1 RESPONSE ~ 430.'206 > 1 s 3

.Partstof the'TCW system are:locat'edfin the Eurbine building. E oh All safety-related-systems.in the turbineHbuilding.are lo-- '

'i cated'in special? areas)to preventlany damage from:

non-safety-related systems duringJoeismic-events.; The parts, , J, of. the L TCW system < outside ; the ; turbine? buildingi are J located:

M away from any safety-related systemsk ' '

S S,

1 3

QUESTION 430.207: -

i

,)

For the TCW system, . provide information -on ,the( following:

items: (9.2.14) ,

o (a) Effact ' of any system component : failure fincluding y rup- N ture of-the atmospheric surge 1 tank onostructures,gsystems5cra '

u,

. components important to safsty.' ,

a 0

(b), Required total coolin'glwater flow and3available cooling / .

water flow; total heat' output'by turbine buildingeauxiliary: -

equipment .and available Leapacityfof: the TCW heatrexchangers.'

e (c) Power cycle . heat sink. to which thejheat from- theJ TCW -!

system is rejected.

'm

RESPONSE 430.207-i S (a) All TCW components.are located in the turbin~e-building.-

Systemsrisportant to safetyiin the~ turbine' building l.are, ,

physically separated'from:TCW system. components._ Failureiof, 4.

any component, including- the~ atmospheric? surge tank,1 'will-not-affect any system importantLto safety.. '

L

.q L

• N, , .+r

^

+

., .. . - - . . . _ _ . _ . . - ~ . . . - . _ . _ . . - . - . -_-__J.--- _ _ . -

. -,_.2. . . - - - . .

.x.- ._,-w_ = ;- -w - - - - - - - - - - -

x-

~ . ~ . ~ .

x 4

i i

j .> -

4 4 (b) ; The TCW.(system hasL been changed'from three:504: pumps: 1 and> heat exchangers to two 100%; pumps and heat exchangers.-

-- The required:TCW flowLis'24,000igpm and there are two pumpt i with a capacity of g9,000 gpa ,each.5 TheLtotalbheat to b(

removed 1s'106;x 10

,l 1

with heat! removal. capacity Btu /hcand =of 130 there x 10agestu/h;each.'

two heat exchangert 1 f , -{

(c)? : Heat. from the TCW systen is: rejected to the power: cycle? t s

j heat sink. '

J QUESTION *430.208-7 ,* l 3

The. system diagrams- lack sufficient 'detal'1': to iascertain-

~

~

whether.orlnot connections.between'the1TCW systenfandc safety-related' water 3 systems exist ~.- Provide assuranceJthat' ,

no suchRconnections to safety-related-systems 1arevprovided- >

or identify: such connections and the : isolation : capabilities provided. Isolation. capabilities shouldJincludeLthe;use'of equipment that;is at-least

• Quality GroupLC and3 Seismic;,Cate-' "

gory L (9.2.14) ,y RESPONSE 430i208' ,

s, _

There are no 3 connecti^ons: 'betweeni ther TCWy sys'tein and-

. safety-related water systems.-  ? '

QUESTION' 430.209 l only ABWR'SSAR. Sections 6 .' 2 ~. 5 J a n d ': 6 . 7 ' d i s c u s si the.

• L Atmospheric! Control' System (ACS) andi Highl Pre'ssure J Nitrogen -

ll System (HPINS) ; - therefore,. correct. SSAR Section 9;3.lc which refers' to the wrong SSAR' sections. for ' discussion of the -

s above systems. Also,.provideTinformationLon the following' items for'the ACS: ,

L (a). Clarification' on appli'cability t of Lsystem{designicrite--

ria 9,J10, and til - (protection againstd single T active component t failure,1 missiles,Tdynamicl effects -.due, to -

piping; failures, tornado-missiles, flooding 1and' seismic-events)1to allinen safety class) system componentst(e.g.1 o nitrogen' storage tanks & vaporisers,Eapplicable' valves '

and piping,' and instrumentation)'.: .(For theseLcriteria, l - see ' SSAR Subsection L 6. 2. 5.'1)'. ' Specify,-if.some_of the"  :

p design : basesJ for ~ thei ACS identifiedin : Subsection. -

L '6.2. 5. 1. are applicableionly forJ th 'e. safety-related l' components of'the system, correct :the~ subsection tas l . appropriate.

1 2 1 q

s J1 5

t

') -

>y , .e .

~

v eiv

• t t=- e ymwe1 w+ve r- t Wf---'- *- W+ = =+=.-#r-, w ve-,- r-- e ---n= -e..-www% e -e+ = - - *w* '=*w-==-'~ **-W'e+ *-'*****m---r - -

..., . . . , ..-. - --..- --......~.-. - .-_ _ . - .- ~ _ .... _.

c p_ .

y >

.m

m , -

w ,9 ,

I i* .!

i f' .

H )

sw b ,ig b) , -JustificationJfor location of the inbo'ardiprimaryl con-tainment5 isolation valves outside'thelcontainment,,  :;

L, which is:a deviation fronlGDC 56,i"Primaryscontainment' i Isolation. "The- af fected Qines are . (1) 2-inch _ N 3-J makeup; lines to the . drywelli and ' wetwell,1(2)- 22-incfi 3 n

purge. suction lines-to:the drywell' and:wetwel17(used;  !

, , for primary containmentninerting'orL.de-inerting;and" connected to : a ' common L 16-inch N2 supply . line)',4 and" (3)

J.; ' '

2-inch and:22-inch purge exhaust lines fromlthe drywell l

and wetwell'.. Wecfind your: response to Question 1Nos. J

.c 430.35candE430.42ldoes notLinclude? justification 1fori ,

4 deviation-from~GDC'56: requirements for the/abovejlines ,

f ' nor;. deviations front GDCi 56 orL55',: ."Reactoricoolant- ,

pressure Boundary Penetrating Containment" requirements i for ' other applicable. lines., Includeijustification' for: '

deviations: from: applicablelGDC forl otherf linestlisted- p

.in ABWR SSAR Table 6.2-7.' <

L. ,d.

-(c) .SSAR' Subsection 6.2.5.217,- which discussesLthe(Flamma H

$1 "'l bility control ; System x:(FCS) ,; does' not provide? suf fi'-

cient details- for us tot conclude. thaththef system com-i ,

plies with thel requirements = of; TMIRAction1 . Item- L II.E.4.1, " Dedicated Hydrogen 1 Penetrations"-of, 1

NUREG-0737. - Therefore, ,' include thel system;: ins Table : j 1

'3.2-1Eand provide detailsEsuch?as;LhowFlongtafter.ICCA '{

[T .and atJwhat concentration 91evaluof? hydrogen the. g recombiner: has to bec activated;u line- sizes cas related' M to ' flowl requirements # 1 and J duration fofg recombiner '

operation.;.Also, identify interface: requirements 1for j

. referencing-ap $

recombiners- (e;plicants=1with1 regard toEthe; externa 1L

g. development?of:proceduraltprovisionsi g L. # to assure availabilityJ off possiblyishareda portable. M

' hydrogen recombiners.betweengsitesn onta, timely; basis-  ?

[ ' and~coordinationiofl surveillance: programs 11n,accordance- ,l with- SRP 6.2.5 acceptance. criterion)II'.12)WW -

. ,1 . ,

..~

4 ABWR:SSAR Tables-6;2-71and 6.2-8:givejallineisize of 4 (d)

. inches and 6 inches traspectivelyj forL theiFCS return linet. Table 6.2-7 o and Figure ) 6.2-400 show elocation- of :

p *' -FCS: primary containmen't inboard isolation'valvesLinside-U the containment' andi outsideL thatcontainment:

respectively; SSAR Section's16.2'. 5.2.7 Eand L 19. A.2.12 3 3' = indicate- portable- a.nd1 permanently!linstalladi r

.recombiners, . respectively.' ResolveFall1the"abovesi N inconsistencies.' .Also, ofjthe4locationiof?all the a y

" primary containment isolationtvalves for!the; system isi i outside the containment, justify:the deviation:from the3 *

! g; D.

f fi l'

s; Y  ; lt!

ew

~ .

.. . .~

m ~

3 .

j 4 s

,-p i j

- GDC 56 requirement: for thefsystem.inboardiisolationu valves.t (1A.2.13,:6.2.5, 9.3;1)-

RESPONSE: 430.209.

1 Subsection 9.3'.1=has been-' corrected perLAmendment . 11.

3 Responses to other' supplemental. questions-are as follows:

(a): The ACS components areIlocated inside_the seismic.Cate-- '

' ~ gory It reactori building, except the nitrogen supply -

(equipment :such as' nitrogen storage' tank and .vaporisers, which are outside the-reactor, building.S.ACS primary containment penetrations-up to'andiincluding;the=second' .

t

' isolation valves are: Seismic: Category Ieconsistent with )

the; primary containment. design l,, Isolation valves are poweredLfrom independent.electricalidivisions;to meet-

. single, failure criteria. . The; reactorJ building i is _

designed ito withstand > and ; protectfeguipment' from ,

tornadoes, missiles,_floodsiand"othergnatural-

- phenomena. _ : ACS _ components ' outside the "reactorf building --

are not designedLtoimeet the=above-mentioned system-

~

.i

, ~ design criteria.

S ub s e ction : '6. 2 . 5; 11 h a s been '

clarified..

• m ,

1. /. '(b) GDC: 55 ' and( GDCs 56 require primary containment 5 penetra- l tions be provided_with:two redundantiisola'tiontvalves I

, .. ' -(one inside and!the other,outside)~. !ACS does.not have.

1 l

primary! containment _ penetrations l that<' communicate to J the reactor, pressure-vessel and is,f.therefore, not part of.the nReactor CoolantMPressu(e! Boundary. ACS:

1 penetrates . the primary containment. -and :: communicates- )

with the:containmentiand drywellsatmospheres.4 Thesel R penetrations"dornot explicitly meet <GDCE55fand GDC 56 since"both isolation 1 valves are:outside(the? primary containment.; ACS' primary containment'penetrationsLdo-

nott extend insidel:the containmentk thus': provision for:

inboard; isolation- valve-DiscnotJ practical'. 4Also, location.of a valve insideithe"containmentLwould L' subject.'it to aLaore severe 1 environment and would!not j be easily accessible for inspection,: surveillance testing and maintenance. ThisJdeviationsfrom GDCK55

-and GDC:567has beentraviewed'forEBWR6 and accepted lbyJ the NRC staff.

Other linesi penetratiing the3 primary containment" listed in--Table 6.2-7 which do1not:explicitlyJmeetfapplicable- H L GDC's:- -(e.g. . instrument' lines) have;also been. reviewed

-and accepted by the'NRC Staff. ' Details"of the review evaluation result' .is reflected iin LGESSAR . II ~ Safety 1

a

~ ~, ,

,(.,i,, ,i, 1 N i. , , , .. -.m , _ + -, -

- ~- '#

p ., w p y.-4 = a.:= ,z;=;;. :a .,::.a...w.==.=:.~ z.; -;.: vu - .=:. - .u . +..- . ,q i > .,

ym , '

s  % ,

-i -

k l 4 Evaluation Report,~NUREG-0979 and Supplement:1", Docket"  !

4 No.:50-447'. . ,

m (c)

~

FCS-has:a dedicated penetration connected to tw'o redun : . 4 dant .(parallel) externally locatederecombiners..'The m . FCS pr;. mary , containment isolation valves" are po_wered .  ;

P , from two' independent' electrical' divisions.. The two

, safety-related recombiners.are. permanently. installed in; '

, .the reactor: building. . ,

of processEand-  !

' recirculation ^ gasses ~'are : controlled and monitored by ac  ;

flow. metering s device. < One? recombinerf is required: to

perform the.. recombination) function. Failure of the '

(

recombiner to achieve flow andvoperating, temperature. ,

will1 result-in manual? actuation:of the. redundant =

recombiner. Per EPG, theirecombiner(isLmanually;actu-J ,;  ;(

ated from'theamain.controliroomLwhen-the hydrogen con - '

centration in'the primary' containment re'achesiabout '

4

, percent by. volume. 'Tne;recombiner isLdesigned tol .

operate continuously for: 60 days Land,Din the covent; of

.intermittently-until core damage, J would - be ? expectedL containment operate;at?leaste lisino<tolonger, isolated. y" Table 3.2.-1 has been orrectedp$r' Amendment 2.

(d) Table 6.2-8 and-Figure'6.2-40!ars' correct.D FCS!suchion- i

>c and return lines penetrating;the. primary , containmentl 1

.are nominal 4-inch andi6-inch pipesyrespectively.L 'Both; 7 a ' inboard and outboard?isolationevalveseofJoachsoffthe s 4 a

~

two-primary containmentLpenetrationsiare locatedc '

j outside. Table:6;2-7thas been corrected.P <

d' v ..

FCS hydrogen'recombinerstare permanently 2 installed  ;

inside ( the L reactor' buiilding.MlSubsectionf19A' 2.12 eis .

correct. Subsinction 6.'2.5.2.7' has1been ? corrected.. .t i

FCS-- primary containme'nt penetrations [dEviatie Efrom GDC' D 56 since. the : location 'of "the :Linboard6and~ outboard .W 4

% isolation-valves areeoutsideItheOcontainmentu Thel 'i 1 justification-for this'designEis?thessame"asLinuitem g (b) above. '

7 a

[, QUESTION 430.210 W

f C1arify which portions of'the.highipressure1 nitrogen. gas' 1 supply system (nitrogen storage bottles, g system piping 'in- "

1 q

1

, V H r

1 l

a l.

l6 'y }

m ,

u4 l o . . . _ , _ _ . . . _ . . .

.m-

'a . - _ a ici --

q

i

q 1

.t

' ~

j 1

i cluding tie linesL betweenisafety-related; divisions and- 1

non-safety-relatedfdivision,1 valves, , instrumentation and-controls) are safety-related..(9.3.1,;6.7)/ }

', RESPONSE'430.210: ,

i The safety-related. portions of the.HPIN system'are:

(1) the nitrogen: storage bottles.and their' headers, >

l 1

. z l

' (2) 'the piping 2.and valves'from F002A,[B, C and D to. ADS accumulators,, , , . . . . .. .

(3) the-piping up to'and including-valves F012A'and,B, ii j

(4) valves F200 and:F208 and1the pipin'g between.them and.

(5) the following;instrumentsLand controls:1PIS001A and1 B .

q and PT002A;and B.

]

. The-non-safety-related-portions:are.t (i (1) the piping,Evalves'and filt'erslfrom7theiAC and'IA: .l interfaces-to,'but!not-including,l valves'F200-and-F012A!- d and B, S '

d (2) the followingqinstruments andt controls EDPS003, . PT004, ' 1 PT005 and=the pressure signal for!PCVF215'

, The signals from PT005 to F012Aland;B[are isolated'from'the, j.

. safety-related signalsifrom!PT002A.and,B1-to~F012A~and B. '

a

.o

1. 2 QUESTION 430.211 ,

N :0 *

-f s" fb

~

L '

o ABWR SSAR~ Figure 6.7-4 show's ' only? one. motor-operated . isola ,

tion valve,on e achi e f 1 theiltiel ilines s between ?. each . 4,, '

safety-related --divisioniand ' the Acommon non-safety.-related - .. , ,

division of the high pressure? nitrogen. gas, supply system' i. '4 (MO-F012A and B) . ThettieJpipingiportion;between'the two a y i, isolation -valves Lis; presumably non-safety-related., Explain how essential! nitrogen'demandtwill be met during"a situation 1 m X[h l3 when there'is a-pipe;ruptureHis(one safety-related. division T m (initiating event),- single' active' component failure insthe..

pF '

~

other' safety-relatedidivisionS(e.fg., isolation valve.'on'the $j' ^

I]y

? applicable tie line. isi open)i and a ; pipe break - ini the ~ '

4

- non-safety-related 3 portion toff.the tie linesi.. (if there is ' i L

such . at portion) . - Alternatelyk provide' two safety-relatedt >

.Q automatic isolationuvalves in. series.on each! tie:line.; ,j h.-

(9.3.1', 6.7) '. *

~.

, l 18 3

L

/

'~

(

L

. e Li N

[.

e ,,. ,' ,

1 "1

w

, , , , . . _ , - , , . . . ,+ ~ - - - - - ~ - - - - -- -~+*-%- -'~ * " ~ ~ ~ ~ ' ' ' ~ ~ ' ' * ' '

_._.m - g_ a;- z u ' sa+.a.a.un. . w

= = c4= =.R w H 7 -' i

,g - ,

% Y y y 2

s u

f. ,

ll

o .

t 4 t:.- ' RESPONSE 430.21$ ,

.. l

A- ,

4

." g ,

The- ADS, accumulators are sized toc permitivalve Loperation' ,

'{ -several times before-new nitrogentmust be1added.. There'are.  !

q check valves at the nitrogen inlet of each ADS ' accumulator. l and check valvasiF008A and B-to' prevent loss lof>nitrogeni- i from the-ADS. accumulators. . M the-postulated: sequence of- .;

. events were to; occur,LPT002A and B would alarm and the posi-s  ;

' tion of valves F012 ; A l and B' woult be .. indicated. . The ; opera- ' '

R tor would'be able to manually cicae the'openavalve,2either. >

F012A or B. ' This would restore' nit rogen to four ADS accumu--

lators if, such resupply; were to: Noone' necessary.9 TheLADS."

t function requires- threa of'signtivalves toj operate: which" would be.available at.'allttimes.J 1

1 , >1

' QUESTION 430.212 , 6; ,

Provide.an FMEA for;the. Nitrogen ~ Gas Supply System 4 (9.3.1,

~

6.7) ' i

~

i '

- .g 7 ,;

RESPONSE 430.212~ ' < t

.m As - discussed' ini Section 115B.1 'and in response Etof Question 430.194, FMEAs: are providedLforstwo ABWRLsystems~and one -

p major. component which present a significant5 change (Lfrom past i Q BWR designs.- . Specifically; FMEAs L are c. included : iniAppendix.

ISB for:

~

4

-(1)^ control rod-drive system '

notion control rod drive),; (withLemphasis,MnIthe' fine '

t n

(2) essential multiplexing; system;-land! ' '

l n - .

(3) reactor internal pump. _

'f l

Regulatory Guide l1.70 requiresi FMEAs to beiperform on se 1 lected' subsystems of; Chapters 6, 7 and 9.: However',' c; GE ; con-siders that the plantinuclear--safety operationalsanalysis 'y (NSOA) of Appendix ~15A and~the probabilisticievaluationsPof.' )

Appendix 19D1 adequately 3 address $singleL failuresUfor those.

systems-and components:whichlare similaratof.past.1BWR~de ' '

[

signs. ,Since theidesigns-of the:ABWR HPIN: system is;similar Li to past' designs, GE;believestthat..itLis unnecessary to-per-form aiFMEA'on;the HPIN.. system.. '! '

i

-)

f

p. .

~~ - - - ~ ~ ~ - - - - - - - - - - - - - - - -

34 y -

4 s 7 7 , q:;

}

qg ,

m

, 1 l-  ;

}

QUESTION 430.213. '

l q

c. , .. Include the, nitrogen gas supply system >in.the ABWR SystemL Tj classificatior: summary Table 3.2-1. (9 3.1,'.6.7) -

J L

, 1 RESPONSE 430.213

• The nitrogen gas supply systems'areT11sted in-Table--3.2-1'Jas- j follows : ~~ ,-

,.]n Table 3.2-1 item no. P9  : High: Phossure Nitrogan' Systemi  ;)

Table 3.2-1 item no. T5 - Atmospheric Control-System. A QUESTION 430.214 1 contrary to what has been' stated in ABWRuSSAR' Subsection; 6.7.1, . there lis only . one non-safety-related . continuous ni- m j 1

trogen supply portion common toJthe-twoiessentia12 supply-di ,

visions (See Figure 6.7-1). Correct: Subsection 6.7.1.- as 'ap-: v1 q

propriate and discuss'the.affact~of' loss of' nitrogen'supplyy lw via the-non-safety-related portion to all the. equipment and. n t

components identified =in SSAR Section'6.7.1/(e. g.,..Pneumat . &

ically operated valves andiinstrumentsicinside the primary; j containment ve'ssel) during normal operation. . Clarify,  ;[

-t whether the pneumatic accumulatoriwhich provides. the backup  ;

operating gas for the main steam 11 solation valve . (See/ SSAR; y subsection 5.4.5.2) is' safety-grade'for each. valve.'LIf not,~ ,

1 i

justify'the design.-(9.3.1,.'6.7) y RESPONSE 430.214 .C l

Subsection-6.7.1 has been revised toistate1that there;is only.one non-safety-related. continuous nitrogen supply por -

-tion common to the two essentia1> supply divisions.. .

Lossr of nitrogen supply via thel non-safety-relaned portion . d of the system during normal operation.would have the follow-ing effects t 3

(1) the MSIV's would close' leading to'a. scram,s -

1 I

(2)' 'the ADS andlreliefnfunction accumulatorscwould be if

' operable for valve' operation foriseveral times'before. l nitrogen. resupply would be needed, (3) testing valves in the:HPCF and RHR: systems.would'not be-available, and t

t 1

p

,3 f

l- . l

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'(4) several s'ampling valves _would.not be.'av'ailable.-

E.

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Loss'of nitrogen. supply would{not(adversely' affect plant c safety.-

The pneumatic accumulator:and0its; inlet' check valve for each R' sain steam isolation-valve are safety, grade. 3 o

QUESTION l430.216 'm 1

. t

, y . .

q Discuss the ' specific features provided :(e.g.L pre- and;-afterf j 1'

filters associated with } compressors, ' particle size, dryer) q for: ensuring that air-or. nitrogen _ supplied by each of:the .  !

applicable systems to components:important to safety! (e.g. ,-

j MSIV's;:SRV's; scram valvestwhichaareclocated outside ther l' containment)- meet < the; quality requirementsu (clean,1 dry and -  :

oil: free)1 of ANSI MC ;11.1-1976Ystandards.t In-this context, ~

the : staff finds GE's justification ' for limiting' particle-  ?

size to 5 microns in the air stream at the instrument .(the l particle size'is mentionedConly(forfthesinstrument air.

1 system) ~ instead- of 7 3 c micronsf as ; required by the above: '!

standardsLunsatisfactory (seelGeneric Letter-88 D

" Instrument Air Supply'. system Affected safety-Related, h Equipment"). JNote that:.the staff will' accept higher than 3=c

' microns only : if. the 4 larger (siza? isl" supported , by . supplier's ' '

-j 6 data for all the2 safety-related= equipment.or componentsathat:

are' supplied' compressed airiorinitrogen:for.their operation.

and there is - assurance ithatithe Ularger Lsize : will , noti cause any. equipment'or component-degradation'with' aging. JAlso,

~9 discuss how all- the above n systems; meets the ' guidelines of j Regulatory Guide < 1. 6 8'.-3,c "Preoperational' . testing of -

y Instrument and E control 1 Airjsystems.'- " Include. the -  :

atmospheric control systemD. since cit - supplies 1 nitrogen ifor; y safety-related components:via* theTnon-essential: portion of; i the nitrogen gasosupply'Esystem'during normal power 1 l l- operation. Include : the : servicei airc system 1 since sitL supplies -1 l air to safety-related compo'nentsjinside: containment- duringL 1 Identify applicablepinterfacee requirements : for9 1~ refueling. o all the nitrogen or: air ' supply systems withs regard to fluid m

.l quality and preoperational testing requirements..-(6.2~.Si

~

l 6.7, 9.3.1) ~ 3

.b RESPONSE 430.216 '

N i

, O Instrument air' system includes two identica1E-oil-less-parallel compressing.. trains L(eacht consisting; of. an ; air: '

suction filter,.a compres'sor unit, after-cooler, andJ d moisture separator) ;y a common receiver!. tank s , andi two:

ei. >1 l 4* <

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.i

, '1 paralleltdryingLtrains (each ocInsisting of: a - profilter, a *i dryer unit,1 ands an- af ter filter) .- , Instrument 1 air is ,

distributed to pneumatic equipment after drying and filtrat ,

tion to no more than 5-microns. Service: air system undergo; l

~

similar filtration andLdrying-processes but hasta'less) < '

a restrictive air quality. ; service air!which servesiasia; ,

-! backup'to instrument air, connects! upstream of the instru : o ment. air's drying trains'and undergo 1another-' drying and; I filtration processeshto meet-instrumentiair'requirementsi before distribution.- The 5; micron particle. size hascnotj #

j ;1 been a problem,infoperating plants anduis:considereds r acceptable.. , , , ,;R' *

.. +1 Atmospheric' control ' system; provides nitrogen-- to : allL pneu i matic equipmentiinside the= primary? containment-during' normal" m operation.. Nitrogenhis,mainnained' dry,. oil-freePandi J filtered before distribution to pneumatic-equipment.- ] f -

All the four; compressed gas Jsystems undergo preoperational1 /

testing in conformance with Regulatory Guide-l'.68.3., Safety; l actuation of;all air-operated valves will be verified *in thel i safe direction on loss ofJpowert and pneumatic medium.. Alli 9 the necessary testing required by Reg.-Guide 1~.68.3 willtbe. 1 l: met. ,

QUESTION 430.217-V , -

1 Provide descriptionMand' figures showing?howLtheffoura [

compressed gas ~ systems (atmospheric control,snitrogen? gas!

supply, instrument air "and 1 service air . systems) c are , .

interconnected. Includejisolation capabilities,Lif applicable, between the essentialrdivisions.ofrnitrogenfgas.  ;

supply system,"and: instrument: air and; service:airisystemsa  !

-RESPONSE 430.217 l

-r s Atmospheric control.. system.normally suppliesinitrogen to'

, HPIN and instrument' air. systems 1:foriall pneumatically:

operated equipment:inside.the. Primary containment..(The HPIN-divisional-safety grade nitrogen supply fromsnitrogencgas

+

= bottles-is.normally isolated from the HPINLnon-safety; grade,' -

.i

=nitrogenssupply by: isolation valves P54-F003A,&LB. WhenLthe<< -l pressure in the.HPIN non-safety grade nitrogenLsupply; drops; to a low' set point, bottled nitrogen supply-valves >P54-F003A! $il

& BLvill-opentand valves P54-F012A'E B1willYclosoito!

. establish a pressure -boundary , between HPIN's safety gradef 1 and non-safety grade nitrogen supplies.- '

j o

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1 1

o- #

e y .

s Y ,

HPIN System interfaces-with instrument air,. system 3through'

~

valve P52-F270. During refuelingioutage:or maintenance when >

1' the primary containment has been de-inerted,- instrument airJ

-system provides air:-to pneumaticallyEoperatedJequipmentc inside- the . containment by manual: closure ' of valve P52-F270 and manual opening-of11nstrument air: supply, valve P52-F257.- ,

4 During normal plant;operationiwhen. atmospheric control '

system. fails-to deliver the required S nitrogen. supply to-the; HPIN system,. instrument air;could;beLused as aishort term i i backup to prevent plant"shutdownL(MSIVfs losing,N2 pressureC 1 Thefamount~of instrumentf

'i will- isolate and cause a " scram) .1 :imaryicontainment air that would be addedtto the pr 91s' not '

considered sufficient:tolde sinert thefcontainment.. , , , .x m

1 I

Instrument ~. air also;provides7 air!torpneumaticallyEoperated; .

equipment outside theiprimary. containment.a Inkths event ( ,

'I that instrument.airisystem pressureEdrops, service air?can .

serve as a backup.. Thef operatory may; manually; open i the-service = air supply valve. 4

~

See attached compressed D gas f systems : interconnection :

schematic. l' j

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QUESTION 430'220- . ,

E  ;

y a compressed air orfnitrogen: supply'sys'tems designed to supply?

~

j fluidLto; equipment or-components' located'inside the' 3 containmentcfor their operation at'no more:thanfdesign basis accident: peak containmentipressure; will ~ notibe- able to .

, '1 perform ;. their intended; function.Lat higher, containment , '

j pressures which.may result under/ degraded-core: conditions.' ,i This; gin turn, say, compromise the-operationfofithe subject' ,

. components.. Address;the:above concern;as'it relates)to thel ,l(

design J of compressed i airL and ' nitrogen ~ gas i systensk '(6.7,

- 9. 3 ~.1) , , =q -

RESPONSE 430.220' a

~ '

I q

,o, 3 operability of safety-related.- pneumaticallyL operated L j' equipment.'insideithe containment?attcontainment pressures. o .g .  ;

higherLthan the) design basis 11sEdiscussedJin. Subsection ,

m4 19E.2.1.2.2.2 paragraph 12(b).m  % (

}

- QUESTION.430.227: ,

W Regarding TMI Action It'em III.D.1.i[(NUREG-0737) concerninga '

]

the integrity'of systemsfoutside containmentflikelypto- H contain' radioactive materiall for? pressurized - water; reactors  !

l and boiling water reactors; provide (information1on $the '

E 1 following . items:! . (lA.2. 34)~

• t QUESTION.430.227a ' ' '

.-.i: i -

l i clarify. :whether theYsystems : thatirequireJperisdicIle'ak) -l L testing: listed L in ABWR: SSAR . Subsection i1A. 2'. 34 ? include - d systems unique to the ABWRedesign.1 JInclude"such systems ifi '

l they;are:n_ot.; currently? includedcin Subsection 1A.2.34. 1 Also, -include; containmentHand . reactor? coolanttsampling- ,

y^

systems lto the-above listt ,

C ,

~

RESPONSEL430.227a 4- i i

There-are1no closed' systems!that couldtcontain highly 2

radioactive fluidstin anismergencycthat1are unique %to thei -

ABWR' design.- t

?

In addition to! the post.accidentisampling system, theTfuelD q pool cooling'andecleanup system:has been addedLto the list' of systems requiring ~ periodic (leak testing..

m 4 3

f J

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? QUESTIONt430.227b

~

. . . ~ .. .. .. - ,

1

, lSincetABWR'SSAR Section"5.2.5-discusses; leak detectio~n j methods' outside - primary containment which include' secondary , , a containment,: . turbine building.and~steamL-tunnel,; rewrite- ,

l M Subsection:1A.2.34~to include all the areas' mentioned above ,

R (current; write-up; refers to: secondary containmentTonly)'.; ~

o,.  ;

'I '

. RESPONSE'430.227b

• u \

Responsejtoathis question is provided'.in revisedISubsection

l y- .1. A . 3 4 .M '

s y

1,

' QUESTION /430.227cc .

'-l- N SSAR: Subsection .1A~.2. 34' states l that' all lines Lwh'ich o pass outside the: secondary containment contain:leakageicontrol j

I systems or! loop seals.and.that these. systems are-discussed. 1

' in'SSAR Section 6 . 5 '. 3 .' . Howeve r ,; ithessil systems ,L 'l particularly, the o loop' < seal"! systems - forf the1 s.acondary ,j containment penetrations, are-not, discussed inithe."SSAR ]

7 --

Section ' 6. 5'. 3. - Discuss,the above systems.1 .

l RESPONSE:430.227c; 7 3, L 1 The references:to " leakage: control systems"Lshould have bee -

'" leakage detection' systems". This : has - been fcorrected:: as? j indicated ' in Subsection ,1A'. 2 ~. 34. 'The . referenc'e ? to iloop ? J seals: for the --secondary containment :penetrationsQ shouldtbe ;1 1 Subsection-6.2.3 rather thanJ 6.5.3~. ' This: has! also been corrected:as indicated in~ Subsection 1.A.2.34.p t- ,

f QUESTION 1 430.227d V: H gy:' .

oc l SSAR1 subsection =1A. 2. 34 ~ indicates vtha'tiun' der?certain-- i

, circumstances '~ an ' af fected1 line associated: with . a y system 1may

'not be: isolated,fromuthe1 secondary containmentEas!part of' corrective' action. : Explain ;under what; circumstances thist ~

will be the case.- '

RESPONSE 430.227d 1

N

]

L. Response to this question,is'providedmin revisedLSubsection J L 1.A.2.34..

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QUESTION. 430.227e- C +

m [

'o '

! d R ,

Explain what the wordst" augmented classHD isystems";mean ini  ?

relation to the - purchase f of! pressure . boundary components- of radioactive waste. systems 3 (See - ABWR SSAR" Subsection?lA.2.34)q

,-n to assure their capability,to?.pr. ovide; integrity. .

.o

• i . - RESPONSE 430.227eJ, M i

li / ^

. '. . a . ,.... -

• b Design- basis L (7)E un. . der Subsection 9 6.2.3'.17 states). " Liquid.

! leakage;from'the secondary 3 containment;to(the:cleanLsone or to:the environment 41s:controlledLby means of.watertloop:

seals;; automatic,shutorf valves;inEseries,jor piping; upgrade y

. to-safety class.". A'pipingLupgrade to safety class augments i a-class D1 system to,a1 class;c orshigher.jsafety. level thus

' assuring a . higher; capability to provide l integrity.' ~ l

, x y QUESTION 430.228 '

o om <

e criteria J or':the f des'ignjbasisf fc ra proteldkihn> from L external -

floodingjshould conform to(Regulatory 1Guidell'.102, " Flood- J t Protection for Nuclear ~ Powerf Plants"s as Ewelli as.-Regulatory 1 Guide.1.59, " Design 1 Basis < Floods"for!NuclearLPower Plants" ' , .

Modify ^the statement 1in ABWR SSARtSectioni:3.4:tojinclude the y commitment tojaeet this-Regulatory. Guide.K(3.4)" l RESPONSE 430.228 ,

.3.Response

4. - to this questionlistp*rovidedLin! 4 revised Subsection. ,

4 o

- ;.i .

QUESTION- 430.229' i H

r . i

~ .

!l Floodt protection .analysi's .Li's ; provided r for theCreactor 1

building- and. control {buildingi only.l eThe ABURfSSAR9 scope V includes structures,.systemprand? components $important:to <

4

. safety in this. area.-LHoweverA portionseofEother! structures,, ' !-[

within , the- scope ; of : the l Plant-specific L applicant = may! house - a systems andscomponentsfimportant;totsafetyn(foriexample,(the .

1 pumps associatedi with -:the ultimate heatLaink) , ' The- SSAR:"

- therefore needs sto sp'e'cifyiasiinterfaceL. criteria % flood >

j protection designEcriteria'for"theseisystems, structuressand a

components similar toathoseDidentified for1internalfand"

+ , 4>

external 1 flooding for the-. systems,. components andJstructuras 1 within the:ABWR SSAR scopen (3.4.1)y ~

~

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RESPONSE 430.229'

. M.

N' ,

Response to this question is provided-in revised 1 subsections; 3.4.1 and:3.4'.1.1.

"j l

QUESTION . 430.230- '

i ABWRt SSAR Subsection 3.4.1.1 '1 references Figure 1.2-28

~y

'(which ; presumably ^ includes a reference to; Figure 1.2-2a) .i Thisy section should : also. referencei Figures L1.2-4 through

-1.2-7'which provide?a more complete' view,offsafety-related components: located below1 the-Edesign- flood ' level'.. ,

's

q.  : Additionally, these figures should be- modifieditoi showithe  !

• location of all watertight; doors used to provide compartmentc separation and the location ofEraised sills for'which credit- ,

is taken.: (3.4.1)L >

RESPONSE 430.230'

. s. , +l\

Response

3 . 4 .1.' 1.1. to this question is proyided1in revised'subsectione V

." Ng QUESTION- 430.231 < ' p" '

m section - 3.4.1.1.2 references Oflooding from a ' feed'w ater line~: i

' break 'in the steam tunnel, . with _ data. fortthe c evaluation?

pro *iided in ' chapter 15.1 1 However, the' -evaluation (is - notl 1 provided in ABWR SSAR ESection ,3.4.1. Provide ~thecfloodi ,

a1 analysis for this high; energy line break. (3. 4 .1)' '

9 9 i RESPONSE,430.231 '

d.

i i

a

Response

3.' 4.1; 1. 2.

to this question-is provided-:in revised:subsectione

~

li QUESTION 430.232 w'" +

4

. . . s i Your response to Question Nos. . 430.73. and ; 430.85.: (submittal; } 'n '^ '"

i dated February' 28,.1990). states? that the:- worst: possibl~e! '

4 flood (circulating water system failure): that canlaffect theb , '

turbine : buildingc would ? result: in'aiflood=leveleslightly\,

higher. than grade cand' thats allD' plant.: safety-related4 M facilities.are protected >a  !

, intrusion ~ (externale flooding) .Explain;howcallLstructures gainst site; surface water!g .'  :;

systems .and components. (ssc) >important tor safetyearek , g protected Dagainst site surf ace .' water < intrusion :- resulting;' 4 '

-from the?above flood 11evel.- -Also, considering'f, access? 'l 4

openings; and penetrations i below ' design- flood L level, betwe'en L 1 the reactor building L and ' turbine < buildingc (see

ABWR7SSARD Table'3.4-2), explain how theJssc important to' safety ,  ;

1

. _ _ _ __'______j

7 ,

t 7 ,

q

, cf  ;

^? \

,iL4

[' ' j= .;(, Y y UI I

-:(2 1

1 t

< . , ' c. -

. p , .. *  ;.,1

-located in the; actor bui ding'are protected fronifloodingn S [%.4k

' inside:the; turbine building.' (3i4.1). '

'i-1

, RESPONSE!430.232 ,

,n All. plant}structuresJarelweather; proofed against-external' 1 flooding 'up to : 8Jcm (3 inches) abovei plant ground ' grade. + '

Also,ithe surface areas'aroundithe' site-structure are sloped}; ',

in such a. matter.to permit water to drain away'from;the site'., g"h

~

e buildings. All the1 penetrations through.the'siteLbuildings) ,

use_ sealsf and?leaktight Econnections. for? protection- againstr external' flooding.: Thet. access.to theJcontrol: building,' -

iJ-;c turbineibuilding,iandLreactorsbuilding*isccontrolled viai -

G enclosed; passagef ways or. water-tight / tunnelst fromithe! \

service . building. ?Any I water ; intrusion : from flooding ;in the 0 FQ 5,'

turbinef buildingtwill be via' the service building, to. otherl j -

structures that. house. safety-related: systems and components.; '

H

. -However,Jwater intrusion!from externalLsources(is: prevented; f

by the:use:of1closedfdoors;and water? tight:structuresGin s- 1 these buildings.; Anylvatoriintrusionninto othe' service:~

• building wil1V flovi into ! tho' floor ' drains. to thei HCW csump; ,

3: ,"p ,

that;is located in theibottom floor of the : service building.J 4 > '1 Subsection;3'.4.l.151Thas!beenrevised"tospecifythatplant[

i structures thatahouse saitety related systems..and; componentst . : a will be7 externallyLwaterproofed up to 8 cm (3. inches) -above? '

M 1 the plant ground grade-ele ation level..

s QUESTION 430.2330 -

4 Discuss -how SSCLimportant ' to safety ; are - protect $d agahLnsh flooding thatimay result.from. failure of1non-safety-related ,

plant equipment::and componentsLlocated-outdoors: ' (eTg.Th condensate : storage tank)l. (3.4.1)4 ' ' O, "

^ .4 -

, . RESPONSE 430.23bi h c'

. . . .  :\ .

ReferEto: theiresponse provided: to Q u e s t i o n 5 4 3 0 .'2 3 2?

l concerningnfloodingofrom;the1 turbine' building.; ci:The 9 t 1

i,

,' i protectionDmeasures? discussed in5the response <to Que'stion t-I

430.232' alsofguard againstLflooding,from outdoor equipmenti -l

.and storage. tanks. Flooding fron-theioutdoor-equipment 31s3 . j not' expected)to penetrate the plant: structures:andiaffect j

'the~ integrity of;the plant'. safety related-equipment./Itiwillj f ~ '

'l spread'over'the(site area and! dissipate into the' ground. W '

yn 11 1

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S QUESTION' 430.234' , ,

i

, ,  : Identify ~ thensafety classification (seismic category,? 4 quality. group) -! fore alliinstrumentation used 'to alerta the .

. operator on. flood' situation.for-performing timely corrective

• actions.'(3.4.1) s q .y s

RESPONSE'430'.234 4 4

' The; sumps Linstirumentations as wellE as flood level monitoring!

are used Lto : alert;. the: operator. on^. flood conditions.stoi s  ; perform timelyacorrective actions. At leastJ.two or more! '

fully 1 instrumented l sumps' . (LCW,s HCW, SDi HSD . . . -'etc. ) . are .

j

. provided in;each1 building-including the drywell and the, -

. control building. Each _. building..has; floorldrains- that;; 1

?

connectLtotitsPHCWt sumpi for transfer f of l thei accumulated ; -

1 liqui,d.JEach drain : sump is. fully instrumented 'as follows:-

3

=

g

- (1)j-Level [switchesfareused'forautomatic;controlof'the, s 4

sump pumps.  ; s n.

- (2) Levellswitches>are used:for annunciation ofssump"

. levels L(HH !andl LL) . . L

. (3)fLeveliswitches are used!in sump pits.except in thef ' #

o ' drywellq% sumps for detection 1of leakage from= sump-tanksA

((4) Flow Jir$1catorsiarei pr'ovidedhin5th's drywell(LCW J and HCW sumps {drainillnesLto monitor; leakage;in the: I

drywell'.1 ,,

(5)Eachjsump[has'two' pumps,Eoneissusedforida'ckupandf y p

for?addedicapacity if needed. -

l

. ThenLcW andXHCW? sumps andJassociaited contro1J-and. p instrumentation ' are ; classified Seismic : Category I while! the" -

- HSDLand;SD sumpsCarefclassified non-Seismic'Categorys I.iThe; ' '

. instrumentation!used to control"and: monitor.the'. operation'of <

- the drain sumps-'are! classified non-essential.1

~

~

H TheEinstrumentationnthatJmonitorslflood' compartments;are normally clas'sified-non-essential

levels in QUESTION 430.2352 m .. ,9

,  ?

.0 -

j Provide- floodin'g 5 analyses L for - applicable plantL areas ~ 'to demonstrate;;thatu safety-related equipment and components t of 4 >

, j g 7 i i

a J

1 ll g.h)

.,_ - .(' , , , , . , [p y , ,, Il e . .~ =' , m = - e- * - - - W * = * ~

" - * * * * * * * * " " - - * * " """"^' " '

iss-w . x x .

a 3,

f

, y* ,

w ,

,. i

+

k.

o

'thesfuel pool ^-cooling and; cleanup system and' safety-rel'ated-i

SSC-in'the fuel' handling area'will-not be adversely affected 4 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, provideL details to? demonstrate. that there ?is no-

. uncontrolled u leak; paths of f radioactive , liquid : from Jthe 'h~

radwaste! building under.: conditions of the worst-case,

' internal-- flood. ;;(3.4 1) - .

. RESPONSE 430.235.-

4

- >+ .. . , :. , '

Response;totthisLquestionEis!provided;in revised Subsection' 3 . 4 .1.' 1. 2 .~ 5.- a n d a inewf Subsections- 3 . 4 3.4.1~.1.314.

~

.1".' 1. 2 . 3 n a n d r

,  ;[

,i h4 P t 4

j

> > m ,

g'~ i t

o

')

g:

I a

I

.l 4

o i

[! , ,, -

7 - - - - - -

---

c. v-~- -

~

f f, d

dBM,. . 2sAsionAc -

L Standard Plant' #

arv e - e s

~

debbw

) 1 program shalllaclude periodic integrated leak L d pass ouIside of the secondary containment contain

' tests at latervals not to exceed each refueling:  ; leakagefonkel systems or loop seals. Dese systems ;

E f

e>

.O

'g cycle. allow tbs SOTS to maintain a negative pressure rela.

~t ive to the environment and thus limit the amount of J ,l ]

Response : . leakage through the secondary costalament. Dese g-s

. systems are discussed in Subsections 4 6 8CPinal **

Leak reduction measures of the ABWR Staa. espected liquid leskoff from equipment outside the ! A,

' dard Plant include a number of barriers to contain.  ;

containment is directed to equipment drals sumps , C ment leakage in the closed systems outside the coa. and processed by the radweste system. Dese makiK taintnest. Dese closed systems include: ple design features of the ABWR Standard Plant. +

Q provide substantial capability to limit any potential '  ;

(1) ResidualHeatRemount,J release to the environment from systems likely to L j contain radioactive material. l

(2) . High Pressure Core Flooder . '

1 Additionally, pressure boundary components of , j (3) 14wPresswe Core Flooder, radioactive waste systems are purchased as'aus.; 1

. c mented Class D systems to assure their capability to c 4 *

. (4) Reactor Core Isolation Cooling implemented provide integrity, j <

{

(5) Suppression Pool Cleanup (suction and return), 1A.235 In Plant Radiation Monitbring s and

[III.DJ.3(3)] s

.y n j (6)' Shutdown Se:vice Water (supply and return).' NRC Posities -

e ~

g *

. Plant procedures will prescribe the method of- (1) Each licensee shall provide'equipmcat and asso. '

, 1eak testing these. systems. The testing will be pet. . cisted tralalag and procedures for accurately dee 1

. formed on a schedule appropriate to 10CFR$0 Ap. termining the airborne lodies concentration in ,m pendix J type B and C penetrations, that is at each areas within the facility where plant personneli ,

refueling outage.; When leakage paths are discov. - may be present during an accident,o 6 cred including during these tests, they will be investi.

gated and accessary maintenance will be performed (2) Each applicant for a fuel. loading license to be( '

to reduce leakage to its lowest practical level. . -Issued prior to January 1,1981 shall provide the : .

equipment, training, and procedures necessary to :

addition, lines which penetrate the ry accurately determine the pressace of airborne t >

]'

centsin contain primary contain isolation i 8 radioiodine in areas withis the plant where plaatj 1 valves which at i ned in a nce with Gen. personnel may be present during an accident. < .

g eral Design Criteria 57 to provide' reliable

,, isolation in the eve e These isolation Response provisions j

l iscussed in Subs n 6.2.4 and tE cent automatic and remote man ng T (1); This is an interface requirement described in -

.2

' Subsection 1A.33. ' \; <

.. hould a small line break develop within a ' (2) Not applicable, inside significant ary containment concu with a , . ". . . . .

~

J active source ter the reactor 1A.2.36 Control Rdom Habitabilityi I

.., water,it would acted b sk' age control systems described la

[III.D.3.4(1))

e son 5.2.5 and the line

- e' may be isolated. 'An i '

y j from such lea radiation uld also is det .

ors which wowd perin t o ioactive material by process NRC Postelen on of la accordance with Task Acdon Plan Item w the y gas treatment system (Subsectico III.D.3.4 and control room habitability, licensees to release to the - ' -- - 'J Alllines which - shall assure that control room operators will be" ']

O ,

C L:

> Amendment s 1A.220L ,

l7 '1 i

L. _ - = = i ' "U

- - z i., ; ..- . i ..=.:: : = .u.- .. ::  :  :- =a . =--~'T--

~

i j

i (7)- Fuel Pool Cooling and Cleanup

[\--)E (8) Post Accident Sampling System *

]

(ii) In addition, all lines which penetrate the-primary containment =

1 are equipped.with inboard- and out boced isolation. valves-that.are l designed in accordance with GDC 55, 56 or'57 to provide' reliable isolation in the eventLof a line break or leakage.:The conteinment isolation provisions are discussed in detail in Subsection 6.2.4, which also identifies all'the system. lines that.

l penetrate-the containment together with their respective

1. sol at ion val ve s. - s c!

Leakage within and outside the' primary containment.are continuously monitored by the leak-detection and isolation syste'n 1 i

(LDS) for breach:in the integrity of the containment. Upon detection of a leakage parameter, LDS will automatically i nitiate the necessary control functions to' isolate the s ource o f. the leak l and alert 5 the operator'for corrective action. The MSL tunnel l area is monitored'for high radiation levels and.for high ambient t emperatures that areLindicative of steam leakage..The turbine

]

building is also monitored for high area ambient temperatures for' MSL leakage. The resulting action causes isolation of the.MSIVs

g and subsequent shutdown of the r,eactor.

(\.,) i The radiation' levels in= the HVAC air exhaust duct s of the reactor building and of the fuel handling; area--are monitored for use in- ,

isolating the secondary containment.1This results1in; closure'of.  :!

the HVAC air ducts in the reactor building, closure.of the:

containment purge and vont lines, and. activation.of the standby 1

]

gas treatment system (SGTS). 1 i

The leak detection methods and associated instrumentation-that j monitor leakage from the reactor, coolant pressure' boundary are' I described in Subsection 5.2.5. ,

q

@For smallline breaks in the secondary containment.that/could

)

cause significant release of rabioactive material-will^ be detected by process radiation monitors in the reactor building HVAC air ducts. As indicated above, a high radiation level will activate SGTS (Subsection 6.5;1) prior to the release'of-radiation to.the environment. Also, any fluid leakage,will drain? ,

into the sumps in the reactor building and will be monitored by sumps instrumentation for level and flow rate. The operator will- .,

be alerted to any abnormal condition for action.. '

I s>

s .

1

, - m ew- r ,"<' ,

n

,. g - ,

1, gg 1

, , c -

' Sinadard Plant ' 2sA410lME :

_ - - rw m TABLE 3.21 -

l CLASSIFICATION

SUMMARY

(Continmed).

k Geosp Quality Safety I. ace. Casal.

Assersace Prinetant commenenta Seisale

Clann* he gd Regnlmmenge Cataned Nales i D Bigh PressureCanFlooder

+

L Reactor pressure vessel injectionline and connected -

1/2' C,5C A/B' '

.B 1. '(g) ' s

- ~

- pipinglacluding supports with--

R

• I

' 2. : Allotherpipingincluding : 2/3. SC,0 supports **

B/C B. Ii ,. (g) :

• - 1 '

3. Main Pump 2 B' SC 'B' I-

4. , Main Pump motor-- ~3 lSC- ~ ,B' I.

5. Valves.outerisolation '1 C SC - A B and within the reactor pressure - - ! .. - (g) vesselinjectionline and i.

a-sonnecsedlines ,

{ 6. Allothervalves 2/3 SC' ;B/C ~B. I< (g)

7. Electricalmodules with safety - 3 C,SC,X / -

related function ' '

-B I y

-n

8. Cable with safety related . 3- 1 g, .

C,3C,X, t B<  !

, :1-D Iask Detection andIseletion 4 System

,' R

. . + .

1. Temperaturesensors . .

U I 3/Nl C.SC, TL - - B/ . 1/- (a) Y..:j

- 2. ;;,x.:_.. :lix 3,?' X t/  :/_ (.) [f g, Pressure transmhters C,5C '

lI[ L(z)

B k '

. 7. -..J^_  !,t' XL 5/  :/_-

(if M

3. /. ? Differentialpressurs -

trarwainers (Dow)

C,SC ' ~

.'B[ {(z)j h-  ;

The ECCS high pressure core flooder spargers are part of the Reactor Pressun Vessel System, see?

= Jtem B1.3.

'b

)

• • Poolsuctionpiping sucsonpipingpom condensatestorage tank, tart Kne topool, pump dscherge pipingandretum Ene topool.

a

)

, y knenomem It '- '

e M 1$ - ,

.= -

,l' s >

8 4' '

1 r p 1 g ,

M '

_ -> !.;_ - - - - - - - - - - --7--

.---,-.,.-----.~~-n ,

w q

'.Qg ,

y

1w; ..- mandard Plant '

arv a - 4 W

\

TABLE 3.21; '

1

!j .

P >

11 i;

>: )~; ., CLASSIFICATION

SUMMARY

(Continued) ' -

j.f ,{ e i y.

o

]c. ,

iJ 7

j p Quality . . 3 <

"&" M

. . . , Group . - ; Quality , .

, 1

. Safety . Loca. Classl.: '

U Assurance ( . - Seismic .

(m Prineinal Comnonenta gb ge gd " Reaulmment' Cataanrvf . % -

i J

E3 lask Detection and Isolation .

System (Continued) . *

,P *

.m

....-.u,__u_..__- _ ,. . .-,iu v

,n

. =-

i

,.,c-

,- y-swieebee.

. 4. 4. i:;;;x cx;;; 3/f' , " ^

!/ ':

?!" ', ,' p: ' ^

. - S. *-- ta ' n v_--~ 2 B' '- .

1

^

.,. . IM.".;; .h' 'k  :..hd... X 'S  !

l' o

-M:

. .,,. ._ ,.. . . . . . - . . . . . . . - . . . ,. ,1., . 'n

- u. ru; ..gg::..  : c,:c .: }_ _

M 4. 42. P96: ::h; msma f%d=4- tA ts

~

 :& l'

N' SC e M f' .15.ls b.'. Vr.lg.,  ; 07C

)i 'L R is/

' 2 /N i : s,sc :

. 0 49 Instrumentlines 3.}  : C,5C ' B :Bi 1 >

g, 1.,X Sampic lines

• 2/NL . C SC ! C/D/.- - . B/- ' 1/~

t

. b 15. Flow transmitters N-

9. ElsAcd Medwh3 t/# ' se, etr, )( - M"-

- k-"

10)r Cables 3/N- ?SC,RZ,X .-. B/~.  ; I/ i

,, j i :l E4 RCICSystem '

i

- i.

't.

1. Pipingincluding supports withi

i. .1/2/3 ' 1 C,SC.

R s

-A/B/C LB 11 - (g) : '

.in outermost isolation valves i '- , '

$  !) 1 j

2. ' Piping including supports . NI O,SC D .-

- 1(g)

Suctionline itom condensate ,

storage tank beyond second . ',

- shut.cff valve and vacuum pump dischargeline from vacuum pump y 1.

l' to containment isolation valves '

L  !

lV

. ,\

' These Jamplelines are totally within containment and radiation monitoringprovides no 1 isalsaanfunedon. l

.g f

'4 ww--a 4 '

' 3.2 16-(- l l

'1 I l hl

.1 3

t,

_ _ . . .. . - . - - - - - - = ~ - - - - - - - - " - ^ ~ ' ' ^ ~ ~ ^ ~

lly!}}'d

.~,.

_%lr pg g' . O:  % '

pqw

@ Jl p , 'u '

teamAnd Plant ,

iL 4

an n f.p , 3 TABLE 3.3-1.

, .. CLASSIFICATION,

SUMMARY

(Coatbued)

,t ,

,,[s N

j I._ %g f , '[ ,

4 ,

Group S Qualty ,

' Losa. Classi. - 'Assmenace Selsade

, _ ..cinal co m anaamas Class das' Bagand: Requirusanage cagena,g ggg, j

' i

, +

.. \

F1 RPV5er*lagEquipment , a L : Steamline plugs N: SC  ;: - < -

i

2. Dryerandseperseer 1

..N SC~ - -- -

strongback and head ,

. , strongback i

F3 RFVInternalServielagEquipment 1_ Controlrodyapple N SC - - - l g F4 Re.NelingEquiposas h

3. h* fueling equipment at.

- pladorm - "; ,

N SC - '--

I-- (bb)L f '1

2. RefueligbsBows N :SC - '

F8 FeelStorage Egelpment L . Fuelstorage racks- =N SC. "

-c <IJ

.newand speat. .(bb)v 2-

2. : Defactive fuelasarage N-N SC' container

.-- (bb) 4

. I o

a G1 RaaeterWaterOsamepSystem  ::

. L Vesselsincludingsupports N .SC: C -i

-. t (81:er/demineraliast) -

' 2.g Regenerativsheit ==haares -N 'SC .C -- '

including supportscarrying a _-

esactor water -

'q!'

3. Osanuprecirculation a N SC C- -- - '

pump,inoters

.j

+ lll ~'

l

_i

a + - . + . , - . --.n,.,-

,a m . -

.e  : ,4 j w - "

- --/<; ..

x 9 1 4 T

. ABWR ' .

wie .

. e==eg pi... -

an a -

SECTION3A -

4 r

CONTENTS 'L

- .msma .

ne na y 3At Fland Protantian 3A1

• g~

3ALI - Plood Protectica Measures for Seimnic J .. - (:

' CategoryIStruawes 3Al"

..~ , . .

-i . . ... . E 4

• 3A1.L1 Plood Protection from Eaternal Sources t

, 14-1 3ALL2 ' Cosopertment Flaading from Postulated Consponset Failures ,

-3A2 3.41111  : Evaluatics of Reactor Buildi$g Flood Evsats . .3A2. ,

3A1111.1 Evaluation of Ploor100(B3P) I 3A2L 141111* '

Evaluation of Ploor 200(B2P)' :3A3- 'y j 3A1L213 Evaluation of Ploor 300(B1P) 3A3 -

~141111.4 Evaluation of Ploor 400(1P) ' 3.44T 3.4.112.1.53 +

Evaluation ofPloor 500 (2P) - -1344

v 141111.6 '

Evaluation of Ploor 600(3P) M5 14111L7 ) EvaluationofPloor700(M4P); ,

asy ,

1411118 " Evaluation of Ploor 800 (4P)'

g3 :-

o 3A ' '"

• PloodingSummaryEvaluatica .

~SAS

.g SA1.112 Evaluation of Contial Budding Flooding Events : j 34 6',- '

P 3AL2 #'

' Permanent Dewatering System 3.46L; i

341 Analveiemi and Test proc.A- - : : g  ; h

c. j 343' laearkens 347 1 cj.

' kk*7 14.u o,o adwaterBlevation

,u7 )'. .y Edd h . .

'3A7 S Eva6L m%see of 9.edwaske SWd A Hoed % Ev* d': 3 Q(s.A. . 1.i. lit.2.3 u w .L k. ,rso w B~ w$pos,qs h j h h6 34 11 7

i 5 g J .

. . . . . . - _ . . . . _ .._ .~

d'-

MN meus mem-A- d Pamme my ,

SECTION3.4 ;3

, FABLES Table nela tar 3.44 Structaw Penetrations, and Aeones Openlap Dedgoed for Nod Proteesion - 3.44 3.42 Reacior Building Acosas Openlop and Poestrations BelowDesign Ploodimol 3.49 4

4 4

i g

e

.t

<a I

3 t

4 a

t t

J 3.4-lil Amenemmet 6

.i

' ~, ** +* ,+ -e._

-m -_ --

ABWR aiaasun as--%d W '

an a 0"

% 3A WATER 2 VEL (FLDOD) DESIGN structures, systems, and comyonents from

> postulated firiedlag. Esism e Category 1 Critula for tha destga basis for protection structures required for safe shutdows remola i

• agalast esternal flooding shall conform to the - accessible during all flood conditions.

• requirements of RO 12. The dulga eritula for protection agalast the effects of compstiment Safety related systems and components are floodlag shall conform to the requirements of flood. protected either because of their location AN$1/ANS.56.11. The desige basis flood levels above the destga flood level or became they are are specified in Table 3.41.

enclosed la relaforced concrete se.lsmic category 3.4.1 Flood Protocelon I structures which have the followlag egiireaea :

' u 3 f r Tais secuos siscosses las flood protectiodf(1) well tbleknesses below flood level of not

, o measures provided for 8eismic category I strued

' ins than two feet tures, systems, and composeats la the ABWR Staa.

~

dard Nuclear Island for both esternal postulated .(2) water 6 ops provided is all construction Booding and postulated flooding from component jolats below flood level; Intlern.

E (3) watutight doors and equipment hatches 3.4.1.1 Ploed Protectica h8ensores for Seismic lastelled below design flood level; and -

r atspry1Streetness ,a f 0 -

• en / The safety related systems and composeats

1. (4 waterproof oostlag u4emal traade* 8'#- U y

• G the ABWR Standard Nuclear ) Waterproofing d taland are identifie%.of focadatless  %

T and la Table 3.2.1. They are either ptotected Seismic Category I structures below grade is T 1 agatast flood damage er are not subject to ?am e accomplisbed priacIpally by the use of water

\by floodina./Tioos protection or safetyTe a e stops at espeasloa and construction jolats. 4e-systems and composeats is provided for all eM!"!: :: :M r ::: ;; n e r n : " :; E g postulatedladesiga described flood levels "ni S.I. Postulated and conditions flooding . 4rt:!!:f " ;r:= r:r;!= :r'r;r !:mrr#

1. ;n?r;:f ng;:n om component faDares la the buildlag compart.

nx. 1 W meats does not adversely affect plaat safety nor A Additloaal specific provisions for flood

• does it represent any basard to the pubtle.

7 . protectica laclude adelaistrative procedures to

'T=W e t.0- 1 3,4. l assure that all watertight doors and batch te structures equipment which and offerbout the safety.related covers are locked la tbs event of a flood f ood '

e identified la Table proteetlos are ' waralsg. If local seepage occurs through the T Descriptions of these stractures are provided to Subsection 3.8.4 a'ad walls, it is controlled by sumps and sump pumps.

3.8.5. Enterior er access openlass a6d In the event of a flood, flood levels take a i peastrations that are below tbs desiga flood relatively long time to develop. This allows level are identified la Table ev44. 6.1-9. ample lead time to perform aseessary emergency 8.4.1.!J Fleed Protesdan grem Esternal ' actions for all accesses which need to be-protected.

Bessess .

The safety related composeats located below o o Selsemic Category I structures that may be the desiga flood level laside a Seismic

' affected by dulga basis floods are dulgaed to y Category I structure are shown is ";n; M i All' .

Q withstand the floods postalsted inh *rdr ?? safety related composeats located ble e- '

a ,,

7 w 4h stractaral provisloes with approachA lacorpdesign ratedflood level are protected la thesalag anlag the tbs hardened protection 9

T plaat design to protect safi ty.related bardened protection approach described above.

4 g y W aV e 2.0 l# "3' d% T' uas sal.

I

@ The types and methods used for protecting the ABWC Safety-related i,

structures, systems and components from external flooding shall  !

hs r }

~

\

confore to the guidelines defined in RG'1.10.t. l

@This section discusses the flood protection measures that are  :

applicable to the standard ABWR plant Seismic Category 1 l structures, systess, and components for both external flooding and postulated flooding from plant component failures.  !

These protection seasures also apply to other structures that '

house syttems and components important to saf ety which fall  :

within the scope of plant specific. ,

O The safety-related systems and components of the ABWR Standard 1 Plant are locatec in the reactor, control, and radwaste buildings l which are iessaile-category 1 s truct ures. These' structures together with thoto identified in Table 3.4-1 are' protected i against external. flood damage., i i

@ In addition t c water stops, wat erproofing of the plant - struct ures  !

that house s af et y relat ed s yt,t ems and component s i s provided up

~

to 6 mm G in) above the plant ground grade- level .t c protect the. '!

external surfacet'from exposure to water.

'i I

l l

1 l

• ,, e

-e-,,..- ,,,>,.sy ,,y.,, e---

MM ma.d -

asasleans - l any a .

3.4.lJJ Compenment Fleeding tres pesenteesd Dars are me interface requirements made apos

• l

-=. - Compeaant Panares the remalader of the plant from possible  !

. . J Sooding la the ASWR standard Plant buildings..  ;

' All piplag, vessels, and heat eschangers' with Other lines, such as stois dralas and normal '

flooding potentialla the reactor building are ' waste lines, laterface with plant yard piplag.  !

seismically qualified with one esception, and However, provisions are made la these lines  !

, oouplete failure of a non seismic task or piping that, should the yard piplag become plugged,'

system is not applicable. The one esception is crushed, or.otherwise inoperable, they will vest (

the radwaste building which costalas no safe onto the ground relieving any flooded whl= 1

shutdows equiposat.

Considering the above alteria and assump.

In accordance wkh Referones 2, leakage oracks ' tions, analyses of piplag falleres and their l L are postulated la any pelat of moderate. energy leoasequesses are performed to dem,onstrate the i '  !

piplag larger than nominal one.lach diameter. ' adequacy of the ABWR design. Dese analyses .

De leakage Bow area is assumed to be a circolar. are provided separately for the~ reactor and  !

L orifice with flow area staal to ens half of the control buildings.

i

' pipe outside diameter multiplied by one. half of l t

the pipe somlaat wall thickness. Resultlag 3.4.tJJ.1 Evaluaales of Ramster Banding t

• . leakage flow rates are approximatodising FleedEvents *

. Equation 3 2 from Referasee 1 with a flow -

H eoefficisai of 0.3, and a normai o,oranes .

Anairsis of poissani nooding wiikin ihe.

m .

e pressure la the pips. . 1

• reactor buildlag is considered on a floor.by.  :

l floor basis.-

De only identined worst case of compartmentl (flooding levolving a high.sacrgy. line T is a , .

. 3.4.1.lJ.lJ Evalentles efyleer100(33p) j feedwater lies break la the steam tussel.. All _

d . ..

. Worst case floodlag on abis floor level would ,

kata kesaccessary from sectiesfor evaluation of this case 15.17 resultat from leakage of the RNR 18'section line -

i

,t k. between the containment wall and the system iso-  ;

No credit is taken for operation of the draia 'lation valva (this applies also to the HPCF,  ;

i samp pumps although they are espected to operate RCIC, and SPCU section lines,'although la during some of the postulated Rooding evoets. . smaller line sises); . Leakage from this soares may cause flooding of the affected RNR heat.

After receivlag a flood detection alarm, the eschenger (HX) room at a rate of 1.04 cubic 1

l i operator has a ten minute grace period to act la meter /minate (275 spa) and may contisme natil F

cases when flooding can be identified and the lies is repaired or equalisation of water  ;

teraiented_ by a remote action from the control level occurs between this room with the suppres-  :

room. la cases levolving visaal laspection to sion pool level Plooding la the room may sause '

identify the specific flooding source la ghe loss of functions for that particular divi.

affected area (escept ECCS areas) followed by a slosal system loop.1 This will not impair the j' remote or local operator action, a minimum of 30 safe shutdown capability of the reactor system.

minutes is provided for the operator. = Flooding of other areas is prevented by water ,

, tight doors. _ gestion lines to other services >

In all lastances of compartment flooding, a : always romala submerged. Other Dooding insi.  ;

slagle failure of an active composeht it dea may result from falleres of other piping considered for systems required to mitigets systems penetrating the RNR' HX rooms for mask sensequeneas of t, pr.Wer floodlas condition. , ' division; these events, however, a detection De smargsasy ears sooling system (BCCS) rooms by sump alarms, are controllable d terminating i are also evaluated on the basis of a loss.of. - flow with closure of valves and shutdows of  ;

soolant sesident (LOCA) and a slagle active pressure sources.

7' failure or a LOCA sombined with a single passive Inilure 10 mientes or more alter the LOCA.

i

) k_.<

i Amenemass 4 I

Sa.3

._-_..a.,_.;.,.__-_..m._ "

. . . - . _ . . _ _ -__._;_ ____.__ _ . _a. . --

,_;.  ;- .  ;~ -; ~ :__ _ , _ ; .:.;;._ , ,_ _. .,_ ., , _ , ,, _i, ,,,_ ; _

i The cynesie effects of postulatcd high energy '.ne breaks in the-  !

7s MLL tunnel area including flood anelysiseare excluded from I i

\ ~- ) o va l ua t i e r., assuminD credit for detection of. leaks-before a line betakt with a gt od accuracy and reliabilit y to permit shutdown e.nd repair. The MSL tunnti area is instrumented with radaation  !

and air t empe rat ure sonit ors that are'used to automatically j isolate the MSL isolation valves upon detection of high abnormel 1 limits.

However, in the event of worst case flooding involving a feedwater line creak, the maximum flow rate from this high energy ,

line break will not exceed 3.6 cubic meters per minute (950.gpm) '

over a 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> period. Refer to Table 15.6-16 for the feedwater line leakage parametert. Wat er discharged from a pcst ulated f eedwater line break will be contained in the seismic category 1  !

structure of .the MEL tunnel area and will not flood any safety related equipment in the reactor building. Th t- flooded area will be allowed-to drain through the floor drains in the t unnel area wh:ch are routed to the HCW sumps in the reactor building for collection end ditcharge.

l i

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o. ABM .. maassas

- into a diffwent divibles below. Minor water ' Flooding may penetrate the fourth a qu' drant at a Dooding dews stairwells saay also esent, mieluum depth,~ and leakage may also ocent through hatches to mondivisloaal areas below.s Flooding la the onwgency electric rooms A,3-and C and the remote sheadown rooms may occur Roergency'dional asserator A,3 and C rooms-

~from leakage or falleros in the % eating and costela coollag water piplag to composeats of ventilating chlued water supply or Emergemey this syst:n. Flooding may occur from failures-NVAC cooling water system. These failures are of s' RCW piplag serving these ecoling aseds at - ,

limited la potential water release by lies a maximum rate of .9 cubic meter /mlaute (240 '

investory and surge task especity and will not spm), which will fill the floor area and weepe esend 3 cable meters (2000 gallons), cessing a into the corridor, with potential cascading down total water depth of 4.$ em (3 2 lachw), the stairwell. The water will spread over the .

side areas on the lower floor while action to Equipment moested en pedestals of 20 em (r) isolate the failed system takes place. ,Intre.

and door allis at area estrances will prevent slos of wate'r lato other divisional' areas is propagallos and provide costelament for this provosted by raised s!!!s on the satry passages.

water quantity estu resposas to the falleres is q meds. Leakage of lebricatlag oil is also possible'  ;

. la the diesel generator rooms, but level indiase ,

Firefightlag activity is all areas of this tion provides a contimulag control on this' ,

level are carried out by manual means at a soures. Even major leakage wul be contained is l

maximum rate of .57 auhde meter /minate (130 gre) the subject rooms due to the small inventory of i and so greater effe:ts than those already; fluid avauable. ;j considered will oecer. Raised door altis prevent i

intrasloa of firefighting water into unaffected Firefighting is tho' diesel generator area j division rated rooms. will be provided by.CO2 releases. L Other.  !

, firefighting will be by hoses but will be of. I Pallures in the CUW and SPCU systems futer l

- (, demiseralizers and associated pipingsmay be of occur limited a/ smaller volumes than those considered, duration.

1 on Floor 200 but will spread over a comparable ' .

y -i area or dreis down to Floor 200 or 100 so that ~ 3.4.1.12.1.8 Evaluaties etFlear800(3F) adequate time is available for detection and , p subsequent system isolation. This Door costals emergemey diesel genwater t1 squipment such as fuellines, control panels,-

El 4 ;

• 3A1.1.1.1A Evalenties of Fleer 400 GF) ~ and air cooling systems. The diesel geserator rooms are protected from fire by CO2, and Flooding from the RHR, HPCF, and RCIC syntans flooding can occur only from cooling water or .

may occur la valve rooms A,3 and C. Manimum ' chilled water lisos at a rate of. 3 cable Dooding is a fallare of the 10' RHR pressere meter /mlaute (34 spa).~ With as available Dow line with leakage flow of 1.34 esbic meter / area of 144 square meters, now over the sill- I misses (354 spa). With a room area of about 34 may occur la by 10 minutes. The water will not square meters Dow over 30 cm (s') sllis oeser la intrade upon other divisloa areas but may begia 3 4 minutes but a mesh larger area is available to descend the stairwell is sorridors and rooms outside these division rooms, so that by the end of the ten miente In the fourth quadrant, leakage may ower ; i response period, a water level of 2 3 cm (one 4 from the feel pool cooling system g* lines. The inch) maximum is seen gesorally outside the maximum leakass rate would be 0.9 embic meter /

affected room. Raised sills os entry doors to minute (240 spa) but as area of several hundred

, other divisions prevent flooding propagation lato square meters is available to spread the water other divisions; separation walls between wblie detection, alarm, and system isolation is' eestrolled necess and clean access areas prevents accomplished.

flood propagatloa to diesel generator areas.

J

"+ . 90 L 3

This floor contains the following equipmen areast

(

/~ a. , The emergency diesel genrrat or A, B, and C equipment k ,' .

m areas including fans, control panels, air storage tanks and  !

associated pipinD. l

.?

The fuel pool cooling and cleanup system consisting of

b. 'l two circulating pumps, two heat exchangers, two filter  !

dominera11rers, instrumentation and associated valves and  ;

piping.

1

c. The' SGTS monit or room, d

d. The stack monitor room. f

e. The MSL tunnel area.

f ar item (a) above, the;three DG equipment areas house-t'he exhaust fans for heating and. cooling'these areas. Flooding can [

only occur from. rupture of the chilled ~ and hot wat er 1 11nes to the; fan cosit.. Howsver,.any flooding is expreted to.be' minimal and 1 will contained within the effected eurbed area. The three l divisional DG equipment areas are separated end mechannically j 1solated.from each other and weter intrusion from a flooded j cosipart m ent is unlikely. '

For item (b) above, the FPC-equipment as well as the fuel- [

handling equipment.are classified non-safetyLrelated. ~The only 6

seismic piping thrit connects with the FPC system belongs to RHR, i

,L which is used to supplement the FPC cooling capability and also. j proviet any eal eup wat er that eay be needed to keepitheif uel 't st orage ' pool tull t o the rin,. The FpC equipment'is located'inL 4 isolated comparteients on the second floor of the reactor buildinD ' l below the fuel pool facilities. During: normal plant operation, area flooding can only occur form rupture:of the 12-inch 111ne j will cause the water contents of botn. surge tanks.to, empty into' a the FPC equipment rooms and cause:ther loss of. fuel pool-coolang. .

1his is not considered detrimental to saf ety since anysdecrease.

.4in tne level of the f uel storage pool -the.t _ may result from. water i

over heating.will be mede up by the'RHR system. (Refer-to:the 'I responses t o Questi ons 410. 34 and 410. 37 for,the discussions on i RHR safety related make-up cabability).  !

The FPC pools are structurally designed f or Seismic Category. l' l and utilize stainless steel liners.to prevent leakage into the  !

pool structure. -Therefore, leakage'from the fuel pool facilities- 1 is unlikely and w111 not impact the -operational capability of' q other equipment. ,

= For the areas identified under items (c). &-(d) above, flooding is not possible because there is no. water ' lines connectionJ to these!

rooms. l 1

For item-(e) above, the steamline tunnel area is isolated from

~

i other areas on this floor through ' sealed doors and firewalls. Any- I flooding in this area will be cont ained and will not propagate i q into other divisional areas. '

1 I

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l L pirenghsing la controlled acesas usas will be done menamtly under esmeroned aamdwa== at a

- Floodtag on this level may also occur from i l room coollag systems or from firefightlag ,

lower Sow sees than the above.  ; offerts. Cooling system failures in air supply,  !

exhaust or filter rooms may allow flooding st j The steam tunnel area is sepwated by ~a the rate of J embic meter /mlaute (80 sys) which i ArewaB and sealed doors so that nooding events will flow out leto adjacent corridor areas If '

l' in that area will not propagate lato mechaalcal madetected for 10 slautes, the appro Imate 3 division rooms os Floor 300, l emble meter (800 gallons) released may stoets a ; }

. depth of a few millimeters over the available  !

3A1.12.14 Beatuasses of plear det OF) floor area; a very limited amount of watw will:

eascado down the stairwells. Divisional areas Flooding eveats at this floor level may - escompasslag the three emergency olestric supph j

' involve' feel oil as well as water. Those , fans and the Rip A exhaust willlaglede raised j divisional rooms associated wkh the emergency sills to preclude water latresion'although water 1 diesel genwater fuel task and cooling system, depth will be slight. Equipment pedestals will.- j have the potential of leakass from the fuel also minimia Soodles impaat on an egelpment.

l eserage tasks. Nas rooms most secommodate leakage of 11.4 subic meter (3000 gallons) for Firefightlag activities is this area would each divialen. Twenty em (a laches) sills os cease water inflow of .57 cable meter /miente entry to these areas successfully sostals all tho' (150 spm) mader sostrolled conditions and, 4 i volume la the tanks. Leakage from these tasks espected water letruslos is no more_than that -

i

will also be monitored through safety grade level . above.'

l Indication and alarm equipment so that protracted l

,j Isakass as well as gross leakage can be SA1.1.11A Evaluaties etFloor 000 (4F) 1 identined. m rooms are protected by CO2 I Erenghting system. Water riooding siay octist Flooding on this floor saa be caused by-- i from the coollag system at about ,15 subic ' rupture of the RCW surge tasks A,3 dt C piplag.~ 9

meter /seinstes (41 spa).' If undetected for ' However, suk tank and its assosiated piping is , <!

, , several hours water may begin cascadios down the located la a separate compartesat which saa be.  !

asarest stairwell but is prevented from entering sealed off la the ovest of assidental Gooding; ether diviales areas by raised sills. The use of raised sills os entry ways will 1 costela the seepage to the flooded area.' Also,' -;

le the 8015 areas, the room sooling equipasst the use of pedestals for egalpment installation may sense Sooding at a rate .15 sobic meter (41 ~ of the RfF supply and subaused fees and for the ~

j 1

spa). Raised sills prevent intrasloa of water : DG C enhamat fans will guard agalaat flooding i lato rooms of anothw division. Flooding may this ogsipment. '

, i also esser dres manual Arefightlag la equipment - _ , y  !

malatenanos aress or from leakage from the' Flooding la the mala reactor hall may esser _l standby Equid sostrol tanks. Maximum task Isak from reactor service opwations, but will be <

l rate will be .1 emble meter /mlaute (25 spa)'so draised lato' service pools. Firenghtlag water r that a response to tank level alarms withig 10 esponded lato this area would oeser at a maximum ' "

mientes .will limit loss to one cubic noter'(or rate of .57 subic meter second (150 sys) but 1, 250 gallons).1.arge floor areas parah spread of will spread over the.large service area '

water at Emised depth. . available.1 bilmor amoests of water may Sad the : 1; j

way to stairwells ~ bst would not impede  ;

m 1'1 T Buelussion of Fleer 100 Odep) operations.1 in the FMCRD panel rooms may oeser SA1.1.1.1J FleedlagSummaryBealensten .,

2 kom Are lag activities at as lapst rate of '

.57 subic meters /alaste (150 spm). Since these Floor.by. floor analysis of potential pipe . '

activkles are massally controlled, any essessive failure genwated Sooding events la the roastor depth of water will be noted sad action takes to buildirt shows the following:

mitigate water intrusion to.other areas.' c' assmamma a 144 j

., .~w , <

' iI

.- . . _ _ . . _ .______.-._______..m_.. _ . _ _ _ _ _ . _ _ . . _ . ~ . _ . _ _ _ _ . .__w . ...- - _ l' .

_--~-.:a~.,.,  ; - ..=, = . a .w. 1- a . ==a .- ,, - .- 9 V , , .

11 M

_ a

' (1)- Where emoaalve flooding' may'escur la a Maniema Sooding may occur from leakass la a . -

L

' divistos rated compartment, propagation to 6. lack chilled water lies at the manimum rate of esber divisions is provosted by watertight .6 cable meter /mlaute (160 gre). Early ;3 doors or seeled hatches. Flooding is one descetion by control room perseasel wiu limit : )

i division is limited to that division and the estent of floodlag which will also be i

' Good water saaset propagate to, ether mitigated by drainage to esterior of the divisions. - building. Total release from tho' chilled water ,

system will be limited to lies lavestory and '

(2) I.eakage of water from large altculatlag. surge tank volume; spiBase of more than 6 subic +

)

water lines, such as reactor buildlag meter (1500 gallons) is salikely.

aoollag water lines may flood rooms'and

! corridors, bat throagh semp alarms and The failure of a coollag water llae'la the -

Isakage defealos systems the control room - mechaalcal room of the building may result is a ~ ,

Is alerted and can costrol flooding by comparable amount of water Sooding that room; _ ' I system isolatloac Divisional areas are . elevation differences and separaties of the ' ,

protected by watertight doors, or where only , . mechaalcal fasesions from the ramalader of the f . .  !

limited water depth ens occur, by raised control buildlag provost pro altis with pedestal mounted equipment withis . , water to ths control area., ' pagatio!

she protested sooma.1 3 a .. . p ':

The evaluation 'of the control buildlag J (3) ' Limited Sooding that may oesar from manual _ ~ floodtog events is summarised as follows: - j 4

Arefightleg or from lines and tasks having . . -

y 1 i limited lavestory is restralasd from ' ._ ,.

Floodlag events that may result from fallare esterlag division areas by raised allis and of the service water systems or firefighting' .

elevation differemens. systems withis the control room do not labibit' i

- plant safety. The proximity of the control area? . ', . !

Therefore, withis the reactor buildlag, to computer and control tasks ensures rapid j 'I internal flooding events as postulated will not response to fire alarms or firehead actuation. O  ?

prevent the safe shutdows of the reactor. Those water lines carrylag water through control .

i j 3.4.L3.2.2 Evalastien er Centroi tellelag areas are shisided to divert water to soa> j Flooding Events .

critical areas.! Service equipment rooms may '

c

. build up limited water levels from cooling water: l or chilled water lies falleros, but elevatloa-. j The control bulldlag is a four story building differences and raised allis provost lettsalos bouslag, gesorally la separate areas, the sontrol ' of water lato control areas; adequate drainage l y

room proper, sentrol and instrument cabinets, and area is available outside the building. Control -

Nai equipment (IfVAC and chillers) asses- room response to those verless levels of <  ;

sary for buildlag occupatles and environmental flooding may estead from system isolation and control for computer sad control equipment. correction to reduction of plant load or'  ;

Water services to the control buildlas comprise ~ shutdown, but sentrol capability is not 'compro.

6 inch fire protection line,6 inch cooling water mised by any of the postulated Aoeding events.  !

lies to the abiller condenser sad a 6 lash ' .

y abilled water bester. Smaller' lines supply 3.d.lJ hrunneset Deweenstaggysesus '- v '

drinklag and saaltary water, and makeup for the j

chilled water system. . Areas with water pipe '

There is so permasset dowatering system '

routed through are supplied with floor drains and - provided for la the Good design. :'

protosalos supplied to roots leakass to the Aoor so that control or competer eqaipment is not 3.4.2 Analytical and Test Procedures - ]

subjessed to water, la thans areas where water '

infusion cannot be tolerated, the. access door. glace the design flood elevation is one foot?

sills are raised, I below the fialshed plant grade, ther is so '

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5

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- _ . . ~ . . _ . _ _ . _ _ _ . . . _ _. . .m..-.__..__

t

'?

3.4.1.1.2.3 Evaluation of Radweste Building. Flooding Events  !

7s .

l

( b The Radweste Building is a reinforced concrete-structure designed-i

(- / -as Seismic Category 1, consist ing of a subst ruct ure.13. 8 set ers below grade and a super- structure 16' meters.above grade. This i building does not contain safety related equipment and is not ,

-contiguous with other plant structures except through a. pipe  !

tunnel. In casi of a flood, the building substructure serves.as a ['

large sump which can collect and hold any . leakage within the building. Also, the medium and large radwaste tankst are housed in1 ,

sealed compartments which are designed to contain any spillage or i leakage from tanks that may rupt ure. The piping that. transfer the liquid waste from the other buildings traverse through a sealed 5 water-tight tunnel to the ~ Radw&ste ' Building at an elevation of

-3,500 mm, which is 3 meters above the Radwaste Building basement j slap.-This tunnel connects to the Turbine and Reactor Buildings. ,

at the same elevation.

)

.The structural clesign of tht's building is. Such that no internal- ,

flooding it expected or will. occur under the worst case condit ions 'f rom t hos e tanks that a.re isolated by the Eeismic. l Categoty 1 compartments..

{

~

riocdit.g from other sources within t he buildir.g - such as : int ernal' radivaste and non redwaste piping, plant = drains, small1 tanks, and pu t.p s is not expected to causehthe-water level to. rise more than- 8

-1 meter abovt the flood depth of 3 meters to.reachLthe tunnel and '

spread radioactive liquid waste, t o other buildings .that house saf e t y relat ed sy st ems.

Thorfore, it can be concluded from the above analysis that there is no uncontrolled leakage path'of radioactive liquid from the  :

Radivis t e Duilding 'under the conditions of worst-case internal ficoding.

l

3. n .1.1. 2. t. Cvelutt ion of Service Building' Flooding Event s . ,

The Cervice Duilding is a non-seismic concrete' structure-centJ:, ting of four floors,- two a'bove grade and two ' below grade. .q lt serves as the main security ehtrance tc the plant and provides l the controlled access t unnels to the Control Building, ' the Turbine Building, and the Reactor Building. This building.does '

nothouseanysafety-related;epuipment.

The connecting access tunnels to the other buildings are below plant grade as indicated in Table 3.4-1. These passage ways are i water tight to prevent seepage into the--tunnels. .Also cthe-l controlled access chambers employ curbsf and: closed doors at both ends of-tunnel that guerd against water. leakage into structures that house saf ety related equipment.

Ths or.ly plant piping that run through this building are those needed for fire prot ect ion, ' wet er s ervi t, ' HVAC heat e rs and ,

chillers, and for draining the sumps. T s building 1has floor I di.Ans anc two sumps (HCW & !CD) f or to lect ing a.nd t ransferring l

[N).

V the l i q uid lead'. e. Under worst-cas,e conditions, flooding f t om line.

i rupt ur e r it unlikely and chn be contained from spreading to the .d structures that house safety relat ed equipment.

w , v

L & O L

Revised (43o.212 TV TABLE 3.4-1  !

i STRUCTURES FENETRATIONS. & ACCESS OPENINCS DESICNED FOR FIDOD FROTECTION V

e 1

^t i 'ST1 w mi Bh N BIAG EERVICE StaC CONT *"_ *N - "^ ~""TE " " Tumagg a m

~

Note 1 Note 2 Note 2 Note 2- Note 2

' 4 DESIGN FIDOD LEVEL 11.700 sum '7.050 mm 7.050 an ' ;7 ,950 num 7.959 mm 4-

}

l- REFERENCE FIANT GRADE 12.000 uun .7.350 mm 7.359 am - 7.350 mm I 7.350 mm 1-BASE SIAS 12TEL -0.200 mm -4.250 mm to -13.150 mm .-6,500 mm 350 mm i

-9,250 aua i i

l

ACTUAL FIANT ORADE 12,000 mm - 7,359 aus - .7,350 sum 7,350 mm 7,350 as i i i i i

- BUIlmING NEIGNT 49,700 am: 16,700 mm '19,700 'am '- l 23.900 mm - 49,350 mm  ;

i l" FERETRATIONS SEISE Refer to None. Refer to- .None None -

1 ' DESIGN FISOD 12TEL ~ Table 6.2 '

- Table 6.2-9 3 f .: [

for RCW Linee  !

j t

ACCESS OPENINGS BEIAW Tunnel from S/S Main Entrance huneet free S/B ~ F1pe Tumme1.. Tummel from S/5- i F

' DESICW FIAOD IRTEL - G -1.450 aus -- ( grade level. G 2.958 mm. NK '

i from R/3 & T/B .G 2.950 mm j

~ (3.500 mm TMSL)_ aree secess free G -3,500 mm -j j- -

S/B G '-7,100 sus , Wete 3 [

t

' .i g

NOTE 1 L Elevettese leeets for .the Reactor Building are .referomeed to Tokyo Besse fee W1 (TML)  !

f g-NerE 1 - The' elevation levels of, these buildings are referenced to the vessel sero reference (4.950 mm below TABL)

~

1 NOTE . 3 - The lines that' run. through the redneste butiding tunnet are -mot esposed to outside groused floodtag. j 4

4 -

=

j

-ABWR m m.aa.

Standard ? am a O

u

[ REACTOR BUILDING ACCESS OPENINGS AND PENE11tA110N Table 3.4 2 , .

7 BELOW DESIGN F140D LEVEL j

fatartsa Namber IJan Namber Elevatian fam) 'l l

Qoan Area Acomes +4800 ' '

i-P 21A RCWSupply - + 1800 .

1 P 21B RCW Supply +1800 l P 21C RCWSupply +1800

- P21A RCW Return ' 3200 0 P 21B ~ RCW Return 3200 1

P 21C RCW Return . .3200 '-1 E.225 HPCF r amie Supply .3300 -

-r .

' E 22C NPCF Condensate Supply - 3200 1 1

,q H0.I3: o l The above openings are between reactor building and' turbine building and . ,

are thus not exposed to ground water or outside pound Sooding. .) '

N '

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Asseenest 4 -

34 0 'l 4 $

< - ~ , ,+-+f -

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j ABWR emandard Pa. e siasiman 1 aw e

.8 produclag the desiga. basis hydroges and backup purge function need not meet this oxygea has ocentred, eritorios, g ( ,(S (2) The hydrogen generation from metal water (10) Composeats of the, system are protected from reaction is defleed la Regulatoey Guide 1.7. postulated missiles and Ma ips whip, as

) #

required to assure proper act on as well as (5) The hydrogen and osygen generation from other dynamic effects such as tornado -  ;

, radiolysis is defined la Regulatory Oulde 4 1.7. missiles and floo,dingMKk.

AG . . isoletleh l8f1 .

$j 1 (11) Thagsystem has the capability to withstand p.a

'(4) The ACS establishes as leert atmosphere throughout the primary containment foBowlsg the ayanmis' effects associated with the gn safe shutdows earthquake without loss of 1 as outage or other occasions when the function. '

containment has been purged with air to as oxygen concentration greater than 3.5 (12) The system is desigend so thatL all.

percent. components subjected to.the primary i containment atmosphere (laboard isolatloa (5) The ACS maletales the primary contalement . 9alves) are capable of withstanding the :

oxygen concentration below the maximum - lemperature and pressure transients , t; permissible limit per Regulatory Guide 1.7 . sesultlag from a LOCA. These components i during normal, abnormal, and accident ' ell! withstand the humidity and radiation . 1

.l conditions la order to assure an inert atmosphere.

enditions la the wetwell or drywell j Jollowlag a LOCA.

]

(6) The ACS also maintalas a slightly positive (13) The ACS is sensafety class except as-i pressure la the primary containment durlag

)

ascessary to assure primary costelement 4 normal, abnormal and accident conditions to lategrity (peastratloss, isolation -

preveat air (oxygea) leakage isto the '

serted volumes from the secondary valves) _ The ACS and FCS are designed and built to the requirements specified in (* '

containment, and provides non essential Section 3.2. I monitoring of the oxygen concentration la ' ,

R the primary contalancat to assure a breathable stature for safe personnel access .

^

or as laert atmosphere, as required. (14) The ACS lacludes the altrogen storage j Essential monitorlag is provided by the tanks, vaporleers,~ valves and piping $

containment atmospheric monitoring system carrying sitrogen to the contalement, (CAMS) as described la Chapter 7. valves and piping from the containment to the SOTS and HVAC (U41) exhaust line, .

(7) The drywell and the suppression chamber will non safety oxygen monitorlag, and all be mixed valformly after the design basis related lastruments and controls. The ACS 1 LOCA due to natural convection and molecular does not laelude any structures housing or -i diffusion. Mixing will be further promoted supporting the aforementioned equipment or j by operation of the costalament sprays, any ducting la the primary aamial====*

1 1

(g) The system is capable of controillag ~(15) The ' system is designed to facilltate. j combustible gas concentrations la the ' periodic laspections and tests. The ACS  :

contalament atmosphere for the design bases LOCA without relying os purging and without can be inspected er tested during normal . l releasing radioactive material to the plant conditions li environacat.

I j

i . .

(9) The system is designed to maintain an inert primary containment after the design bases - '

LOCA assuelag a slagle active failure. The 4s.<

L - ut ws,

, . j

. 1

. 3 a., .. .

l

' 1 BWR

.m y .

-- 1 m ,. i elevatles whle.h would be eovered by post.LOCA she same sins and made from the same sheet snoding for anloading the feet.

to provide uniformity of relief pressws. [ j, 62JJJ pressarecesaret -(6)! The- rupture disks are capable of withstanding full vacuum la the wetwell j I (I) In general, darlag startup, normal, and vapor space without leakage.

u abnormal operatloa, the wetwell and drywell .

'l 1

pressures is malatalsed greater than 0 psig (7) The piping material is carbon steel.' The - 1 to prevent leakage of alt (onygen) lato the design pressure is 10.5 kg/cm83(150 j primary costelament from secondary psi), and the design temperature is 3 costalament but less than the nominal 2 psig 171'C. *

,i stram set point. gufficient margin is .

.a 1 provided such that normal containment 62J.2.7 Itosembiner temperature and pressure fluctuations do not P8N Wg 8g .

g 'j' cause either of the two limits to be reached - (1) Two,gscombiassgeb6de are located is ,, J eonsiderlag variations.lsLlaltlal. accondary costelament. Eac as shW i costatement conditions, lastrumentatlos la Figure 6.2 40, takes suc a from tbs a., 1 errors, operator and equipment responso drywell, passes the process flow through a )

. time, and egulpment performance. heating section, a reactor chamber, and a. l spray cooler. The gas is returned to the .j

wetwell.

!' l (2) Nitrogen 530 kg/m3 (0.75 makeup automatleally psig) positive pressure maletales a - j i

to avoid leakage of alt from the secondary (2) The recomblaers are normallyialtisted on. i lato the primary costatement, high levels as deteralsed by CAMS (if  !

hydrogen Is not present, oxygen' 1

i. (3) The drywell bleed sleikg is capable of concentrations are controlled by sitroges 4

malatainlag the primary containment pressure makeup). ~

l less than 830 kg/m3 (1.25 psis) during the maximum containment atmospheric heating ; 6.2JJ Design Evaluation .-

( ;j which could occur during plant startup. <

j

. The ACs is designed to malatala the 422.2.4 Overpressert Presselles contalement is' en inert condition except for

'altrogen makeup seeded to maintels a positive, (1) The system is designed to passively relieve containment pressure and prevent air.(02 )'

the wetwell vapor space pressure at 5.6 leakage from the secendary lato the primary 2 i kg/cm8 3 The system valves are capable containment. I of being closed from the main control room :I uslag AC power and pecumatic air. ': The primary costalanest atmosphere will be

.-i laerted with alttoges dur'ag normal operation of 1 (2)core The vent thermal system power la the formis of slaed steam .cas so that residu'al l contalament z thebelow will be r,alatalmed plant. Osygen cots 3J volume be passed through the relief piplag to'the percoat measured ra a dry basis.' i stack. ' ,

( - Following as accident, hydrogen concentration -

(3) The faltlal drivlag force for pressure willlacrease due to the addition of hydrogen.

relief is assumed to be the expected from the speelfied design basis metal. water pressure setpoint of the rupture disks. reaction.i Hydrogen concentration will also ~

lacrease due to radiolysis 'Any lacrease la  !

(4) The ruptore disks are constructed of hydrogen concentration is of lesser concera stalaless steel or a material of similar because the containment is inerted.' Due to-corrislos resistance, dilution,-' additional hydrogen moves the  !

operath;g point of the contalement atmosphere (5) A number of rupture disks are procured at farthesiton the envelope of flammability. .'

\

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1

. w a _ . a .-- .L - . - - - - . - . = = --.= =: "-~ LJ'

_ _ - _ _ _ _ _ _ . - _ - _ _ . - - - - _ _ _ _ - . . . - _ - . . . . - _ - . . . - . _ _ . . ~ . _ _ _ _ _ . _ _ _ _ _ _ _

' ABWR naui .

, Standard Plant __

m r-TABLE 6.2 7 (Continued)

~

l CONTAINMENT ISOIATION VALVE INFORMATION -

PLotMAstLITYcolmtotsysTEM 4

vehe No. Tet-PgD1A Tel-FUDIS TepPontA TetPOO3B 1

asAnFw um ase um um i

.l Apphoebb huk ODCM ODCM ODCN ODCM j rum Dw ANo$ DW AMOS DWANOS DW ANOS . ,

k une ame e e, e v i i

ser Y. Y. Ya va i

lashese Canes (e) (a) (a) (a) '

-[

t- 40 -'/0. O. 0

, t TWe C1meh7ess Yu Yes Yes ' Yes "

,\

vehevype om Oe Om Om:

{

Opwww Maw -Mm=- Mew Mew y PrL Aroussion Blee. Eise. Eise. Riee. '1 1

Set. Amumeise Manuel Manual Meeual Manuel . e

-t M Petilish Shot hus hisi - sh.: '

'. ,i MM Miso Ihut hul $byt bg  ;

Peel AM Pheilion

• Opea OPen Open Opse '

PwrFesPteMon As k As k - As k - As k Cast las.84.IN A.K A.K A.K -' A,K L - n.e<mi - ..- <= .. ..

PorSom m (Div) 1 M I< -M' l

^-9 6340.33

'h

• • e e . .

l

. . - . ___ ..,._ __ ._ . . . . _ _ - , _ . . . . . _ . - _ c .. . ~ a . _ , , . . _ ;. 1. : . . ..

J, . . _ .; . . . _ . . . . . _ ,.. a _. .i. n. .. . i

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santu.

. Standard Plant e

_ . . mm e 4

I TABLE 6.2 7 (Contlawed)  :,

i CONTAINMENT ISOLATION VALVE INTORMATION (  !

s I.

FLAMMABILITY CONTROL SWIT.M (Contlemed) j i

e Wehe Me. TW9-P906A T49 M10GB - T49.Pt07A - 749POO75 I i

884R P4 - 6J4 . . 6J4 '6MO 62 4 .:

i AppliseMe Basis ODCM GDCM ODCM ODCM <  ;

Plaid DW ATNOS DW ATWOS DW ATNOS ' DW ATHOS

,p u-Sne r6 y6 /6 /6

a. j e SW~ Yes - Yes Yes Yes  ;

4 lashage Class (a) (e) (e) . (a) laseelse #0 /O O O-D Type C Imak Test Yes Yes Yes ' Yes 1

Valw Type Gate Gene Oete Oete +

Opereew Motor - Motor '~ ~ ..Mosor Maior

\

C- .i l PrLAeeseelen Eke. Elec,  : Biet. - Eke. ...

1 See. Aeemselon Manual Manuel Menvol Manual !3  ;

r boeunal Positima $ hut Shut Shut

-i Shut i

j- Sheedows Pe6hlen Shut- Shut- Shut Shut.

+

Poet Ane Penkien Open OPen Open - OPen ,

PWrIWIPosilien As is Asis Ag is As is

+ *

) f Cees.len, sie.(c) A.K A.K A,K '- AK.

Cleesee flew (see) < 30 < 30 '<30- <30 For aseees (Db)  ! . II ,I 11 s

'i

% ^

(

r G' Amendment 9 6.2 30.M g as 4 g 4 e. .4 g_ "-

eeunes some**eutb eeme dh # '**^**<4

• _

, g id.

, .- , - , ~ . . . . . , . . . _. -

. . . , ~ ~ - . , . ~ . - . _ - . . , . . . .,s.-- . . . . ..4,, _.+,%._. .

, J A.BM .u.a r w san.

67 HIGHPRESSURENrIROGENGAS during as $$E. The bottles are also covered by  !

, SUPPLY SYS'IEM a heavy steel plate, which serves as a barrier J to potential missiles. (

67.1 Functions r

Flow rate and cap'aelty requirements are 1 i The high pressare altrogen gas supply' system divided lato as laitial requirement and a

, l is divided lato two ladependentcostlaucus divisions, with supply. As faitial requirement for

- 1 t4 each divistos containing : ::m'"; - '"-' each ADS SRV provides for actuations of the

- ^' r " y g;'; " a safety related ' valve against drywell pressure. Fifty gallon  ;

3 emergency stored nitrogen supply. The essential accomulators supplied for each mala steam ADS'  ;

)

stored nitrogen aspply is safety Class 3, Seismic $RV actuator fulfill the ateaa valve  ;

Category 1, designed for operation of the mais requirement. The costlaucus supply is divided i steam S/R valve ADS function accumulators, lato safety and mossafety portions. '  ;

He function of the sonsafety related, maksap Compressed nitrogen at a rate adequate to  ;

altrogen gas supply system is: make up the nitrogen leakage of each serviced i' valve is provided by the safety portion. This '

(1)'. relief function accumulators of main steam assumes as alt leakage rate for each valve of 1 . .  :

$/R valves, seth for a period of at least seven days. The '

essential system with associated lines, valves {

(2) pneumatically operated valves and and fittings are classified as Safety Class 3, lastrumsats inside the PCV, Seismic Category I.

}

o (3) leak detection system radiation monitor The sonsafety portlos provides compressed calibration nitrogen at a rate adequate to recharge the ADS

{

i SRV accumulators. The monessential system has j (4) ADS function accumulators to compensate for two pressure control valves to depressurite the t

the leakage from main steam S/R solenoid altrogen gas from the AC system. One is to.

'k valves during normal operation depressurire to 200 psi for the SRV accumulators and the other is to de t

67.2 Systern Description other pneumatic uses. pressurine to 100 psi for  !

Nitrogen gas for the essential system is. The costlevous' supply portion. of the supplied from high pressure nitroge.s gas storage pneumatic system, extending from the AC system ,

bottles. Nitrogen gas for the monessential to the isolation valve prior to the essential "

makeup system is supplied from the mitrogen gas system is not safety related. '

evaporator via the makeup line to the atmospheric .

i control (AC) system. The essential system is D Nossafety piplag and valves of the system are separated into two divisions. There are tiellaes designed to ANSI B31.1, Power Piping Code, and L between the nonessential and each division of the the requireneat of Quellty Group D of- 1 essential system. Each tiellas has a motor Regulatory Guide 1.26. Pressure vessels.and .

j operated shutoff valve. For details, see Figure heat exchangers are designed to ASME Section 1

6.71 and Table 6.71. VIII, Division I.

Each division of the essential system has ten System design pressure is 200 psig with the bottles. Normally, outlet valves from five of - system design temperature at 1500F. .j the ten bottles are kept opea. Esch division-has a pressure control valve to deressurl e the 67J Systein Evaluation nitrossa gas from the bottles.

Vessels, piplag and fittings of the safety The bottles are mechanically rest'ained to: . portion of tbs system'are designed to Seismic preclude generation of high pressurs missiles i

MN

, s.7.I

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~-----ee '- - * - - ' ' - ' ' * * " - ^-'*" ' ' ' * * ' * ' ~"~" "

...-.. a l

MM

-1 mnem -

Stand =sd Finne Raw. a _

i ,

SECITON 9.1 l CONTENTS l l

annien suis . . tan l 9.13 New Fast kernpa 9.11 l

.i' 9.111 Design Bases 9.11 L

! 91Lt.1 Nuclear Design 9.1 1. . <

i 91112 Storage Design ,

.. 9.11 j 911.13 Mechealcaland StromualDesign ;9.11 #

l t

9.111.4i Nrmal. Hydraulic Design 9.11 911.13 Material Considerations < 9.11 d

<' 9.1114 Dynam{q .x-4 9.11 Qwl i

9.112 Facilities Description (New Post Storage)

9.11 I 7

9.1.13 - Safety Evaluation 9.11a e t

\:

9113.1 Criticality Control 9.11a i 9.1.112 Structural Design 9.1.la ~;

91133 Protectico Features of the NewFuel -

Storage Facilities ' 9.1.'1a ' #

9.1J Ramat Femi karagn 9.1 2 .-

9.111 Design Bases 9.12: >

9.111.1 Nuclear Desiga 9.12 -

9.111.2 Storage Design 9.12: j 91213 . Mechaalcaland StruauralDesign p.12 : 'i 1

9.1214 Thermal HydraulicDesign 9.12a '  ;

-i l

-i 9.1.li Amensmem s -

4 I

)

E 7 .

d - l .f

!: . ABWR -- -  !

- Standard Plane m _m SECHON 9.1 e O courrurScco ti

) !4 38C1005 Ellt P.AB  !

9.1317.2 RenaarInternal Pump Servidag - 9.19- l

{

9.1J17.3 Pine Motion Control Rod Drive Servidag 9.110 ,

9.1317A Neutron Monitor Sensor Ser.idag - 9.1 10 -!

i 9.1318 SpecialSwvicing Room / Areas 9.110 l j 9.1319 Heavy Land Handling Egelpment 9.110' {

9.13110- Transportados Paths /Roudag 9.111' ,

{

9.13111 Equipment Operstlag Procedures, ,

Maintenance and Servios - 9.111 i

9.133 Safety Evaluences - 9.112 l 'I

.t 9.114 laspection and Testing - - 9.112 L  ;

9.133 lastrumentatlee Requirements 9.1 12

(-- 'l 9.116 Operadonal Responsibilities .  : 9.1 12:.

9.14 7 asman 9.1 13 i

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9

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-Q ,

' 9.1.vii

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v e-r. . - , - - ,y y - * -r-. , ,v,-w.e- .-y, a,.ww -aw..ww. - - ..+a... e- = . ,,e- - - *

• m. .,. 4 +=, +-.*+-4. - , . -o ,

-...-.7..

l i

O. 9.1.6 Interfaces 1

j 1

9.1.6.1 New Fuel Storage Raeks Criticalit y Analysis , .

]

.v i 9.! 1. 6. E1 Dynamic and Impact Analyses of New Fuel-Storage Racks '!

l 9.1. 6. 3 Spent Fuel Storage Racks Criticality Analysis, 9.' 1. 6. 4 Spent Fuel Storage Racks Load ~ Drop Analysis '

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Standard Plaat-9.1 NEL STORAGE AND HANDIJNG (3) ne blasu between the calculated ruuks and experimental rnalts, as well as the uncertainty

. He asw. fuel storage vault stores a M ears load lavolved la the calculations, are takes lato 1 etnew fuel assemblies. De fuel is stored is the new account as part of the calculational procedure to -)

feel storage racks is the vault which are located as assure that the specific k,glimit is set.

close as practicable to the spent fuel storage pool

work area to facilitate handling durlag fuel' 9.1.1.12 Storage Des [,, '

papersalon. ne new-fuel inspection stand is close to the now. fuel storage vault to minimize fuel- he new fuel storage racks provided la the new fuel .a. 1 transpest distanes. storage vauk provide storage for 40% of oss full core ,

fuelload.

g']

. 2 Spent fut removed from the re. actor vusel must - .

a"p" 3

be mored underwater while awaiting off site transfer. 9.1.1.1J Mechselcal and genneterst Design O  ;

Spent feel storage racks, which are used for this -

t .

purpons, are located at the bottom of the ful storage ne new fuel storage racks contain storage space pool under sufficient water to provide radiological . for fuel assemblies (with channels) or bundles shielding. His pool water is processe.d through the - (without chanasts). ney are designed to withstand l fuel pool cooling and fuel pool and wetwell cleanup 1 all credible static and esismic loadings, ne racks are i (FFWCU) system to provide cooling to the spent - designed to protect the fuel assemblin and bundin 1 fuella storage and for aalatesance of fuel pool from excessive physical damage which may cause the 1 j- water quality, ne spent fuel pool storage capacityis release of radioactive materials la excess of 10CFR20  : 1

' 2105 af the reactor enre. and 10CFR100 requirements, under normal and )

abnormal conditions caused by impacting from either - i ne new fuel and spent fuel atorage rocks are the fuel assemblies, bundles or other equipment.  !

same high density design. ne new fuel rocks can be  :

used for either dry or submerged storage of fuel. ne racks are categorized as Seismic Category 1. j ne design of the spent fuel racks will be described. See Subsection 9.1.2.1.3 for additional discussion of leformation on the new fuel racks will only be design bases and analysis. l presented when the design is different. De detailed i analysis of the rack design is contained in Subsection 9.1.1.1.4 hennal Hydroelle Design 9.11 See Suhaareb 9.111.4 ? i

. 9.1.1 New Fuel Storage 1 f

9.1.1.1J MatedalConsiderstless - t 9J.1.1 Desiga Bases , ,

See Suhaamlan 9.1113.'  ;

9MM Nuclear Design _ a j

,u.1.g e' .w..)br,, ., c.h ,

A fut array of loaded new fuel rocks is designed to

^ T I be subsritical,by atleast $% Ak. .See4mbassaieseMMr. < M .;

i (1) Monte Carlo techniques are employed la ths salsulations performed to assure that ~does

• • " 7"'"C ~

g

. I i

met exoned 0.95 under all normal and -Soo4eheeseisa4MMr- "I <

esaditions. .

80 l 9.1.1.2 F auel= Descdpelos (New Feet .

(2) ne assumption is made that the storage array Storage) is inAsite is all directions. Since no eredit is takes for sentros leakage, the values reported (1) ' De location of the new feel moregs vaak in the ,

as afealve neutros multiplication factors are, reactor bulkhng as shown in Sostica l.2.c  !

is reality, infinite neutron multiplication a demors. s 1

L Amenemset 6  : 9.M -i

')

E.~ C .

O The new fuel _ storage racks.'are purchased. equipment. The= purchase specification for these racks ' will require the . vendor- to provide J

[.' the information, requested in Question 430.180 on criticality analysis and the inadvertent' placement of a fuel assembly in'* -l*

other than prescribed locations. See Subsection 9.1.6.1'for interface requirements.

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1 The new fuel storaDe rackstare' purchased equipment. The purchase

1. ~

specification for the new fuel storage racks will require the l vendor to perf orm confirmatory dynamic and impact analyses. The  !

input excitation for these analyses will. utill:ei the horziontal {

and vertical response spectra provided in new F1Dures 9.1-15 and .

9.1-16. (The SSE response is two . times the '0BE response).

Vertical impact analysis is required because the fuel assembly l1s held in the storage rack by its own weight without any mechnical l holddown devices. Therefore, when.the downward acceleration of' s i the storage rack exceeds ig, contact;between the fuel. assembly j and the storage rack is lost. Horizontal impact analysis is.

l required-because a clearance. exists between.the fuel assembly and l

the storage rack walls.  !

See Subsection 9.1.6.2 for interface requirements.

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(2) The new bel storags rads are top entry reeks ' ' (3) The recha lastede ladiddual solid tabs morses '

L desiped to prededs the possbuky of witinauty compartesats which prodde lateral restraints

• madw awmal and abnormal conditions, ne == the entire bagsk of the hel =-w . r

, _igper tieplate of the het element rests against .

L the module to aredde lateral support. The (4) The weight of the fuel assembly or bundle is; loww tiopiate sim ia the notion of the red, appated mangy by the ree lsen suppen. t -

whiskappens es weight atthe hat. . . ,

(s) De rods are fabrisated too maertak and hr (3) The rack arrangassent is deslaned to provost eonstruaios are la assordanes whh the .

aseidental innersies of fuel assemb!!ss or latestissue of ASTM spes18entions.

, beadles between agacoat reeks. The moregs .

reek is desiped to provide W"i to the , (6) Lead-in guides at the top of the storage spesas

- feelballfor perpling purpassa, provide guidanes of the 8sel during lamanlon.

' (4)' The floor of new feel storegs vaak is aloped te (7) The racks are designed to withstand, while '

a drals loonted at the low potet. This drela . malatalaims the anclear estety design basis, the

removes any asser that may be accidentaDy and ' impoet fores generated by the vertical bee fab

' 1.

• Oi ntrodseed i lato_the vault. The drop of a Ibel assembly from a height of 6 fest.  :!

drain is part of the Seer drain subsystem of the ,

~o Egeid radweste syssess. ' (8) De reek is desiped to withstand a puDep fores ,

y of 1717 kg (4000 lb) and a horisontal fores of '

1

($)- The radiation monitoring equipsest for the = 454 63 (1000 b). Dere are no readDy deflashle new feel storage areas is described in Sesties ~ horizontal foress la essess of 1000 t and, la the .

i 7.1. evest a fuel assembly should Jean, the manimes ;

i lifting fores of the fuel bandling platform- )

9.1.1J Safety Evaluesles papple (assenes theit swiedes fau) is 3000 b. ' 1 l

, 9.1.1J.1 CritientityCoasrel (9) e De osw fuel storage reeks require so periodic j l special testing or inspection for nuclear safety  ;

7be design of the new fuel storage racks provides purposes. i for an effective multipliention famor (k - for both o.

y% aormal and abnormal storage conditions ual to or 9.1.1JJ preteetles Features of the New Fuel

• f% su than 0.91 To ensure desip criteria are met, the Storage Fasilitise l

~ i Jollowing nohnal and abnormal asw fuel storage "

eamih6 were analysed: i De new fuel storses vouk is bonned in the reestor I

bedding. De veuk and rosator building are Seisnie

(1) normal positicales in the new fbel array, and Category I natural phenomena such as tornadoes, tornado missiles, floods and high winds. Fire-(2) onesetric positioning in the new fuel array protection features are deserted la Subsession 9J.1

. and Appendia SA.

Tbs asw fuel agorage area wGI ====adma fast (k. < IJ5 at 3D C in standard eore geometry) with The 'storags reek struaurs is desiped to wisseaad no impliantions. the impaat resuking from a falling weight. Tests using a

a shoulated fbel bundle haus been condemed to verify _

9.1.1JJ StrusawelDesip that the rock ensting can withstand the impea form a bundle dropped from a manimum anowable height (1) The new fuel vaak contains one or more imei- above the artay. Frossderal fuel bandling; I: morage rocks which provides storage for fuel a requirements and equipment desige dictate that so.

==imum of 405 of one fall core fuelload. - more than one bundle at a cimas en be handled over the morage rods and at a manimum height of 6 feet-(2) The new fuel storage rods are desiped to be above the upper reek. Dwefore, the reeks enmast be -

freestas Aias (i.e., no supports above the base), displaced in a mat.nor eenslag arttical spacing as a

, we s , o.i te -

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w -

=-na*.,w w-we4 -t-*a-'4- + +-

- - = . . x .. . u .. a . = .;. = .

m.a= == = = = == =; = == 3 ~~ ~ ~ ~ = ~ ~ -}

L . ABWR is4aswi y m - A rd 7 " n,a-compatible wkk she sevironment of treated water over the entire length of the fuel assembly or

. and ermides a dest, life of so years. bundle. i 1

9.112 Faculties Descripties (Spent Feet

' Starses)

(4) The racks are fabricated fran materials used for construction and are specified in accordi6oce - i

)

with the latest issue of applicable ASTM J

, _ (1)' He spent fasi storage racks provide storage la . speedications at the time of equipment order. 1 she reactor building spent fuel pool for spent imel recahed frees the reactor vessel during the ' (5) Lead in guides at the top of the storage spaces refueling operation. - The spent fuel storage provide guidance of the fbel during insertion. ,!

rocks are tap entry rocks designed to preclude _

the possibliley of criticality under normal and (6) The racks are designed to withstand, while i abeonnal somdialons. He upper tieplate of the ~ maintainlag the nuclear safety design basis, thef j

- fuel elements rests against the rack to provide impact force generated by the vertical free. fall" 1 lateral support. The lower tieplate alts is the drop of a ibel assembly from a height of 6 feet.

- bottom of the sack, which supports the weight

,j; aftheibel.. =(7) The rock is designed t'o withstand a puDup force 1

" of 4000 lb and a horizontal force of 1000 lb.

(2) ne rack arrangement is designed to prevent There are ao readDy denaable horizontal forces J accidentallasertion of fuel assemblies or la excess of 1000 lb and la the event a fuel. i bundles between adjaceat modules. The assembly should jam, the maximum lifting force l storage rack is designed to provide accessibility of the fuel H't platform grapple (assumes ^

to the fuel beu for grappling purposes. limit switches fail)is 3000lb. .

, w .

(3) The locatlos of the spent fuel poolend ehs> (g) ~ The fuel storage racks are designed to handle ' J

. m .3 .x is ahown . irradiated fuel assemblies.1The expected y in Section 1.2 radiation levels arc well below the design levels.- .<

, 9.1.23 Smerty Evelmaties

• The fuel storage facilities will be desigasd to  !

l Seismic Category I requirements to prevent  :

9.1.211 CriticalityCentral earthquake damage to the stored fuel, q; j 0 = . ;;=e m es=s= .e u.: e The fusi siorage poois have adequais waie, 1 Q _

. shielding for the stored spent fuel.~ ' Adequate; I e .

. shielding for transporting the feelis also provided.

9.1.212 Structural Design and Material i Liquid level asanors are lastalled to detect a low pool .;

Compatibility % Ets  : water level,'and adequate makeup water is available  ;

i to assure that the luel wiB act be uncovered should a (1) ne spent feel pool racks provide storage for - leak occur.y ,

l. 270% of the reactor core. . i '

Since the fuel storage racks'are made of 1 (2) ne feel storage racks ere designed to*be acacombustible material and are stored under water, 1 supported above the pool floor by a support there is ao potential fire hazard. The large water i structure. The support structure allows . volume also protects the spent fuel storage racks from .

]

sufficient pool water flow for natural cos. _ potential pipe breaks and associated jet impingement vectica cooling of the stored fust. Sines the loads.

l modales are freestanding (i.e., so supports .

above the base), the support structure also - Post . storage racks are made is accordance with the provides the aquired dynamic stability. latest issue of the applicable ASTM specinestion at the time of equipment order. The storage tubes are:

(3) The racks inciede individual solid tube storage permanently marked wkh identiGcation traosable to ;

compartments, which provide lateral restraints - the agerial certifications. The fuel atorage tube e

Amendmenie s.l.ad

. - . . . . . . . . . . - . . . . .. i

. _. - _ . _ _ _ _ . . _ - _ ~ _ , , - - - ._ _ . _ _ - - _ ._ _

4 f

t

@ The spent fuel Storege' racks are purchased egipment. The I

purchase specificatton for the spent fuel storage racks will-

)

require the vendor.to provide the information. requested in l s Questifon 430.190 on criticality analysis of'.the spent fuel storage including the uncertainity value and associated }

probability.and confidence levelfor the k.cc value.. See l Subsection 9.1.6.3 for intertcee requirements. '

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ABM man sani ha ed Plame . m., a -

essembly is compatible with the savironment of treased water and provides a design life of 60 years, ladsding allowaness for sorrosion. *

~

Regulatory Guide 1.13 is apNcable to spent feel storage facilities. De building costalains the Anet

. storage facilities, locluding the storage racks and pool,is designed to protect the feel from damage caused by, (1) natural events such as earthgaake, high wtads and Sooding, and

,(2) a=l=I damage eaused by dropplag of fuel assemblies bundles, or other objects onto .

saared fuel.

9.1.2.4 Summary of Radielegial Considerosions By adequate destga and careful operational procedures, the safety design bases of the spent fuel storage arrangement are satisfied. Thus, the .

esposure of plant personnel to radiation is malatained well below prblished guidelios values.

Further details of radiological considerations, lecluding those for the spent fuel storage -

attangeneat, are presented la Chapter 12. ' '

t

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I Amendment a p.t.as .)

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i - The pool: liner leakage. detection system and, water level: mon--

, itoring system.are discussed:inisubsection 9.1.3. The cor-rective action forsloss1of'heatiremoval;capabilityvis-in!

subsection 9.1.3.'- The_ radiation monitoring system and the:

corrective = ' action : f or excessive - radiation :.' levels are.

discussed in Subsections- 11.5.2.1.2.1 and 11.5.2.1.3.

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'g ABWR Standard PlanF sam aei

, a- n

. - 9.1,4 Ught Load Handtlag System 1 fue.1 bundle drop. Maalmum deflectica limitations are -

. (Relates to Raftsellag) ;

' imposed on the snaio structures to maintain relative D stiffness of the platform. Welding of the platforms is -

9.1A1 DestgaBases

" la accordance with AWS D141 or ASME Bouer and -.

. Pressure Vessel Code section DC. Gears and bearing .

The fuel.hanaiog syneem is designed to provide a ! E meet AOMA Oear Classification Manual and ANSI :

- safe and effective means for transporting and: 53.3. Materials used in' construction afload bearing banalog fuel tros the thae k reaches the plant natu i ' members'are to ASTM specifications. For perscanel ?

k leaves the plant after post. irradiation cooling. safs  ; safety, OSHA Part 1910179 is applied.- Electrical; x-headling of fool includes design sonsiderations for  ; equipment and controls meet ANSI Cl, National malateining occupationei radiation exposures as low Electric Code,'and NEMA Publication No. ICS1,1 m as practicable during C s "= and handling. '

M01. ; y 1 <

Destga criteria for raajor fool. handling system . . lDe sumulary fust grapple and the s$la islescoplag equipment are povided la Table 9.12 through 9.14/ : fuel grapple have redondant . lifting features and as which list tbs assety class, quality group and seismie - ladicator which soofitas positive grapple category. Where applicable, the appropriate ASME,

' ' ~

engagement.

a:

ANSI, Industrial and Electrical Codes are identified. . . . _ . ,

Additional design criteria are shown below'andi

..The fuel grapple is used for lifting'and transporting ,

expanded further lo Subsectica 9.1A2. ' q; fuel bundles.;1t is designed as a telescoping grapple!

. that can extend to the proper work level and,la ks '

The transfer of new fuel assemblies between the . ; fully retracted state, still maintain adequate water l ancrating area and the new fuel laspection stand; ' abielding over fuele , i and/or the new fuel storage vauh is accomplished; N f, - .. ,

t , log 5. ton auxiliary book on the reactor building- la' addition to redoodsat electricalinterlocks to erans equipped with a suitable grapple. ',, preclude the possibility of raising radioactive material '

out. of the wateri the cables on the auxiliary hoists .

The 1,000 pousd auxiliary hoist on the reactor . ; incorporate an adjustable, removal stop that will Jam .

J building crane is used with'an auxihary fuelgrapple f the boist' cable against some part of the platform to transfer new fuel from the new fuel vault to the _ structure to pavent hoisting whos the free end of the '

fuel storage pool. From this polat on, the fuel will E cable is at a preset distance below water level.i either be handled by the telescoping grapples on the - ,

CII -

^- '

H":; platform ogaw_Ushe..

models.

._ _r:"

y esoa W;  : Provision of a separate cask pit l capable of being' isolated from the fuel storage p' ool, will eliminate the t 0 l potential accident of dropping the cask and rupturing -

l The refueling platform is Seismis Category I froan i the fuel storage pool. Furthermore,' limitation of t a structural standpoint la accordance with 10CPR30, travel of the crane handtlag the cask will preclude :

Appendix A.' Tbs refueling platforza is constructed . < j amplbelJ6esagspodl.

'q in accordance with a quality assurance pogram that .

ensures the design, construction and testing -

! transporting 1 the cash oveq% g01ppgC '

4

9.143 System Deseriptiesi e

'i" requirements are met. Allowable strees due to safe shutdown earthquake.(SSE) loadeg is 120% of yield t

. i , "i . [ 4[ ..

Table 9.15 is a listing of typical tools and servicing ! n or 70% of ultimate, whichever is least.'. A dynamic ? equipmeat supplied with nuclear system. The

' j analysis is performed .on the structures using the ; following paragraphs describe the use of socne of the response spectrum method with load contributions ; major tools and serviclag' equipment and address .

s M7 resulting from each of three' directions actlag ' safety aspects of the design where applicable? , y simultassously belt.g combined by the RMS - , + 1 L

pocedure. . Working loads of the platform structure . Subsection 9.1.5 provides the data that verifies the i are in accordance with the AISC Manual of Steel ABWR Standard Plant heavy load handling systems Construction. All parts of the bolst systems are- l and satisfies the guidelines of NUREO.0612. : '

designed to have a safety lactor of at least ten, based en the ultimate strength of the material,iA. 9.1A2.1 Spent Fuel Cask .

L , , ,

j redundant load path is incorporated in the fuel bolsas -

so that so single component failure could resuk in a =

/- Out of ABWR Standard Plant scope.

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. Standard Plant a- a <

< 9.1AJJ Overhead Bridge Croses 9.1AJJJ New FuelInspecties Standi

( '9.1A.12.1 Reacter Bundlag Crane = Tbc new fuellaspostics stand (Figure 9.1-4) serves

,_ .. J as a support for the new fuel bundles undergoing re 4

'j o De reamor building crane is a seismicaDy analysed " ceiving inspection and provides a working platfors ' i piece of equipment. The erane consists of two erese ' for t"-Id- engaged in perforslag the laparola= j

! ' girders and a trolley which carries two boists. Tbs . .

= .

, runway track, which supports the crane girders, is : . The new fuellaspectice stand consists of a verticalf ,

supported from the reactor bu!! ding walls at eleve- _ guide column, a lih unit to position the work platform ; >

tion 34,600. The troney travels lateraDy on the crase at any desired level, bearing seats and upper clamps ; ,

girders sarrying the main boist and musiliary hoist. , to hold the fuel bundles la position. , j.

. The reactor building crane is used to move all of

• 9JAJJJ f%==aal Belt Wrench -

j the maior components (reactor vesul head, shroud : ._ : _ . .

.. l head and separator, dryer assembly and pool gates) The channel bolt wrench (Figure 9.15)is a manun <

as required by plant operations. The reactor buildc ' aDy operated device approximately 3.% meters (12 ft) i '

' las crame is used for handling new fuel from the roi la overall length.' The wrench is used for removing .

actor building entry hatch to new fuel storage, the ; and installing the channel fastener assembly while the new fuelinspection stand and the spent fuel eenge, feel assembly is held in the fuel preparation machine. : -

' pool.' It also is used for bandling spent fuel cask; . m channel bok wrench has a socket which mates y ,3 The principal design criteria for the reactor building and captures the channel fastener'capscrew.j ,

'?

- crane ar described in Subsectioc 9.1.5. J d 9.1AJJA Chamael Handilag Tool; .'  : -i 9.1A.2J FuelServicing Equipment

~

.m fuel servicing equipment described below has L

The channel. handling tool (Figure 9.16) is used isj '

j eonjunction~with the fuel preparation' machine 1 been designed in accordance with abe criteria listed 1 remove, lastau and transport fuel channels in the .-

la Table 9.12.' Items not listed as Seismic Category aseseos pool.' , - '

1 1, such as hoists, tools and other equipment used for . .

. . ~. ,..

servicing shall either be removed during operation, ' The toolis consposed of a bandling bail 'an N moved to a location where they are not a potential ; lock / release knob, extension shaft, angle guides and hazard to safety related equipment, or seismically re. clamp. arms which engage the fuel channel.1 The j?

strained to prevent them from becoming missiles. '

criamps are actuated (extended or retracted) iby manu. l1 '

auf rotatinglock/ release knob.1 > m 9.1AJJ.1 Fuel Prep Machles The chaand.baOling toolis suspended by its ball: 0 Two fuel preparation machines (Figure 9.13) arel

_$g4 mounted on the wall of th vel seesage pool and are < located on y pool pbotyd "'

frosa a spring balancer ce the'abannel handling boom o I

used for stripping reusab e channels from the spent . . . -

fuel and for reebanneling of the new fuel. The me. *4

' 9.1AJJ.s Fuel

[?

Vacuum Sipper , '

chines are also used with the fuelinspection fimure to provide an underwater inspection capability.

.. ( . . . .. .

~~'

, i The fuel pool vacuum sipper (Figure 9.17)~ pro- [

- vides a means of identifying fuel suspeasd of having? , j Each fuel preparation machine consists of a work ; cladding failuiss. The fbsl pool Vacuum sipper con- : E platform, a irame, and a movable carriage. The1 sists of a ibelisolation aameniaar, Suid onesole, moni6 Nag,Qrame and mova e caniage are located below the = toring console with program controHer an 7 , normal water level in the fuel scenage pool thus pro. f tector and the later connecting tubing and cables. 1 viding a water shield fodhe fuel assemblies being ' : The suspected fuel assembly is placed in the isolation

' handled.~ The fuel preparation machine carriage has : container. A partial Vacuum is established in the gas ,

]

a permanently installed up travel.stop to prevent ' volume above the fuel assembly. The fission product , .I raising fuel above the safe water shield level. . gas leakage is sensed by the beta detector and moni '

j

.toring console. > l f

• l Asneedsment 7 ' , p 3.se l

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a . . , W.J.t . - - . - - - - - " 2 2- - ' - +- "- --" " - '

-ABWR siui wi

- Standard Pin =# - a- a 9.1AJJ4 General purpose Grapple - A radiation hardened portable underwater closed ,

circuit television into thecamera is provided. The camer' I

The general purpose grapple (Figure 9.18) is n' - may be lowered reactor vessel and/or fue handling tool used peasrally with the fuel. Tbc grap - eserees pool to assist in the inspectice and/or ma,nte.

pie can be attached to the jib erane to headle fuel mance of these areas.

during channelias. .

1 g .

~

A general purpose, plastic viewing aid is provided io  ;

gw f 9_.1AJJ.1 Jib Casas Sont on the water surface to provide better visibility. . ,

. The sides of the viewing aid are brightly colored to j.

The jib crane (Figure 9.19) consists of a aBow the operator to observe it la the event of filling motor drivea swing boom monorall and a motor. . with water and sinking. A portable, submersible-type, driven trouey with as c.lectric hoist. The jib eranc is underwater vacuum cleaner is provided to assist in re- '!

mounted along the edge of the reactor building fuel moving cred and miscellaneous particulate matter - )

storage pool to be used during refueling operations, from the pool Doors or reactor vessel. Tbs pump and < l Use of the jib crane leaves the refueling platform or the filter unit are completely submersible for ex.

fuel.bandling platform free to perform general fuel tended periods. Tbc Alter ' package'is capable of shuffling operations and still permit uninterrupted being remotely changed, and the Akers will St into a fuel preparation la the work area. Upon boisting, standard shipping container for offsite burial. Fuel tbe fitat of two ladepeadeatly adjustable limit . pool tool accessories are also provided to meet servic. j switches automaticaDyinterrupts hoist power at the . lag requirements. A fuel sampler is provided. his is . 1 maximum safe uptravellimit. When the jib crane is to be used to detect defective fuel assemblies during  ;

used in the handling of hazardous radioactive materi- open vessel periods while the fuel is in the core.' The ,

als that must be kept below a speciGe water level, a fuel sampler head isolates ladividual fuel assemblies ; l 5:ed mechanical stop is installed on the boist cable / by sealing the top of the fuel channel and pumping to prevent further W- when that levelis reai:hed.] . water from the bottom of the fuel assembly, through ' '

ReN ellag the fuel channel, to a sampling station, and returning ' ,

9.1AJJJ *d ~ . natform  ! the water to the primary coolant system. After a ' '

l

• soaking" period, a water sample is obtained and is o'

( Refer to Subsection 9.1A.2.7 for a description of radiochemically analysed to determine possible fuel l bundle leakage.

~ thef(t Nettb, platform.

9.1AJJ.9 dilag Boom 9JAJJ RanctorVesselServicing Egalpment A channel handling boom (Figure 9.1 10) with a . The essentiality and safety classifications, the quality spring. loaded balance realis used to assist the oper. group, and the seismic category for this equipment stor la supporting a portion of the weight of the are listed in Table 9.13.'.Following is a description of channel as k is removed from the fuel assembly. The' - the equipment designs in reference to that tabic.

boom is set between the fuel preparation machinesJ :

With the channel handlias tool attached to the reel,' 9.1 AJJ.1 Reacter Vessel Sesvice Teels the channel may be conveniestly moved betwesa the l

fuelpreparation machines. These tools are used 'wben the reactor is shot down l and the reactor vessel head is being removed or rein- '

9JAJA servicing Aids , stalled. Tools in this group are:

General area underwater lights are provided with Stud HandlagTool a sukable reflector for illumination. Seitable light l

support brackets are furnished to support the lights Stud Wrench in the reactor vessel to allow the light to be posi.

.I tioned over the area being serviced independent of Not Ranner.

the platform. Local area underwater lights are small I l diameter lights for additionalillumination. Drop Sag Thread Proteaar

! Eghts are asad form-inada where aseded.

i

~

nread Proteaar Mandrel l

\

, s Amenemme 4 9.14b l

^

e tw - --e .+ , - , , - . . ~ . , - - . - - . - - . - . - , - - . . .-,..,--.nn -ve---. ~ . - .. -- -

7, 7 g j

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. 9MM . Standard Plant J namend a- a 2

, , :l

. _ =. . .

h service. ne pedestals have mods which engage three l

a BushissWrench , _ ,

evenly spaced stud holes la the head flange. The i q

' SealSurface Protamor flange surface rests on replaceable wear pod made of ' '

al== 6= ,  ;

's .

? - Stud Elongelos Measuring Rod -

W ~

When resting on the pedestals,'the head Gange is ' :l Dial ladicator Elongation Measuring Devise appeosimately 3 h above the Door to aBow access to : j

l. the seal surface for inspection and 0 ring replacee ,i Head Gelds Cup ,

meat. .

$ RIPImpeRer / Shah AssemblyTool s The pedesial structure is a carbon steel weldsent .

~!

oosted with as approved palet.- k has a base with bok , j Impeter Storage Rack. holes for snousting k to the concrete Soor.

g j

. . .. .. . n- .. .. ,

The tools are designed for a 60 year life la the spesi ? f A seismic analysis was made to determine abs seine 1

- Bed environment. I.!Alag tools are designed for a mic foress imposed onto the pedestals, Soor anchors,  ;

safety factor of to or better with respect to the uitle ? using the Door response spearum method. The struc t .

- mate strength of the material used.- When carbon - ture is designed to withstand these calculated forces -

steel is used, it is sither. bard chtone plated, . and meet the requiressats of AISC. A parkerlaad, or coated with an approved palat perL . . . . . '

.{

Regulatory Guide 1.H. 9JAJJJ Nand Sted Rack . ,

9-a 9JAJJJ Steamilae Plus The head'sted rock is.used for transporting andI (

storage of eight reactor pressure vessel studs.. It is" ' ,

o The sieantine plugs are used during reactor  ; suspended from the reactor buuding crane book when O refueling or servicing; they are inserted la the steam lifting studs from the retpor well to the operatingi ,

outlet soules from inside of the reactor vessel to Aoor. ,

l

( prevent a Dow of water from the reactor wellinto the -

mila steamline during' servicing of safety relief'

. The r'ack is made of aluminum to resist corrosion.

_ 7 ,

ar valves, main laalmu valves, or other components'of and is designed for a safety faaor of 5 with respect to '4 the mais steamlines, while the reador water level is tlw ultimate strength of the material.c :s at the refueling level. The steam lies plus design:

+

provides two seals of different types. Each one is in ' The structors'is designed la accordance with the A

, dependently capable of holding full head pressure. ' Aluminum Constructica Manual

• by the Aluminum ' ~~ '

The equipment is constructed of corrosion resistant A=ad=6

. materials. All calculated safety factors are 5 or 4 r

d better. De plus body is designed in accordance with .9JAJJJ DryerandSeparaterStreegback ' .j the Aluminum Construaion Maamar by the Alumi - .' ( . .. ..

s' aum Maad*6 tf The Dryer and Separator Strongback is a lifting ,

o device used for transporting ths.'essam dryer or the 9JAJJJ Shrend Head Stud Wrench 1 shroud bead with the steam separators between the N reactor vessel and the storags pools.' The strongback F This is a hand. held tool for tightenias and loosse- la a cruciform shaped I. beam structureiwbich has a las the shroud head stads.' It is designed for a - hook bos with two book pins in thelcester for sagage >

v 60. year Efe and is made of aluminum for easy hast l ment with'the~rencor building crane'aister book._ The M ding and to resist corrosion. Calculations have been ' strongback has a socket with a pneumatically oper.

performed to saaGrum the design. sted pin on the and of each arm for sagsgies k to the '

four lik eyes on the steam dryer or shroud head.? ";

9JAJJA HandMelangFedestal .

j f _. . . _ . . .

The strongback has been designed such that one Three pedestals are provided for mounting on the book pie and one mala been of the cruciforse wiB be" refueling Soor for supporting the reactor vessel head capable of carrying the totalload and so that no single and strongback/ carousel daring periodfofy J.

M4dec .

mr -:

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. L. u__ . _ ~ . = . = = := = . . = L. = . = L- ,

_ _ _ _ _ _ _ _ _ _ _ _ _ 7 s u , ' '

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. u gy 5

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g R' Rs..a..a pi..e ,

a I: aomponest failure wl!! cause the load to drop or (4) Storage wkham RPV Hand During reactor op.E s  !

sming - " "i out of an essentially level atti. . eration, the carousel is atored on the refueling , '

sede. De safety factor of aB Ming members is to or Soor. .

u bemer in reference to the skimme breaking strength h of she assarials.

~

The stroogback,'with ks lifting composeats,is ~ ti 2

.. .. designed to meet the Crass Manufacturers As.  ?{

The erscrure is designed la assordance with W sociation ofAmerica, ep iamh No.10. The i Mammal af Shell Constructice* by AISC. The son. . design provides a 15% impact allownses and a :  ;

j plated assembly is proof tested at 125% of rated  : safety factor of to la reference to the ultimate -

. loed, and au structural welds are magnetic particle  ! strength of the material used. Aher completion d hopeasd sharload test. of welding and before painting, the lifting as- j 2

, sembly is proofload tested and au load affected - 4 L

9,1A2.5.7 Mead 0. 7^fcareensi - welds and lift plas'ars magnetic. particle is.

spamed.L +

3 The RPV bead strongbad/carouselis as inte.

, g grated piece of equipment consisting of a-  ; De steel structure is designed la accordance ,j eraciform shaped strongbad, a circular monorailf  ; with the Manuelq(S#eef Consoussion by AISC. j and a circolar scrage tray, i Alumlaum structures are designed la accor. . O

. . dance with the Ahnminum Consouesion Manuel 'l The stressbeck is a bos. beau structure which has - bythe Almaisum Aa-4 ash ,

d a book bos with two book pies in the esoter for en. .

gegement with the reactor service crane sister book. The strongback is tested la accordance with '

Each are has a lifs rod for engagement to the four1

, American National Standard for overhead hoists .

Eh lugs on the RPV head.- The monorail is mounted f ANSI B30.16, Paragraph 161.2.2.2 and such that s en cuensions of the stromsbeck arms and four addi. : one book pia and one mala beam of the struc.

tional arms equally spaced between the strongback - ture is capable of carrying the total load, and so arms. De monorail circle matches the stod circle of- 4 that no single component failure will cause the L :y 1 the reactor vessel and it serves to suspend st'ud load to drop or swing uncontrollably out of an: '

tensioners and out. handling devices. The storage - essentiaDy level attitudei De ASME Boiler and ' j s

tray is suspended form the ends of the same eight ; t Pressure Vessel Code, Sectica IX, Welder Qual. ' '

. arms and surrounds the RPV flangef A manifold is . ification is applied to aD welder struerures. A mounted underneath the book boa for' distributing , 'l bydrastic and passmatic pressures to equipment RegulataEGaide 1.54 m 4 travellag 'on_ ~ths. monorail. Thel head. . >>

a . . .

4 strongbed/ carousel serves the fogowing functicas: - General compliancs or akersate' assessment for Rag.5 9 '

alatory Guide 1.$4,which provides dcaign eriteria for ~ -

(1) IJhiep of VassalHead: The arongback,w protective seatings, may be found la S=hwsk 6.1.2.

enspended from the Reactor Building,mies6 '

L erans mais hook, wiu transport RPV bead plus a.7 _

.9J.4.tila Vessel Serviclag Equipasset

)

the enrousel with au hs attachments between , ,

a >

the seamor vessel and storage on the pedestals. The instrumsat 'stroOeck att' ached to the C-auxiliary hoist is used for servicing the local er?

(2) Tenslanlag of Vassal Hand Clannrn: De tar. .

range monitor - (LPRM), soures range moaltor H i essel, when supported on the RPV head on,the > (SRM), and latermediate range moskor (IRM) dry . , n

, geesel, wiE carry tensiosas, ks own weight, ths ' tubes should they reqaire replacement.s Thel q strongback, storage of sets, washers, thread : strongback laitiauy supports the dry tube ~isto the 4 M preensters, and ma==i=*=dt ools and equipment. vessel. The incore dry tube is- then decoupled from ; 'i (5) starage whh RPV Hand: The carousal,when the strongback and is guided lato place'while beingi w supported by' the isistrument handlias tool.! Final) stored with the RPV bead holding pedestals, incore insertion is accomplished from below the~ reac. <

p sorries the sameload for(2)above. tor, vessel. The lastrument handtlag acolis

, , A ')

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k\

. ..._..,,.L , . . , . . . . . . _ . , _ . . u . , _ .n . . ~ .a . . :~ . ., 1_u ; ; _ . . _ . .. .-...._.V

i t ABM Rimsulard Plame

4siewi  ;

a. n -

1 Jg 4 4; p., 5.

s.ac,ed to.e exe am ,,- _., ,. 4 .s is used for removing and lastalling LPRM fined '

service platform. The auxiliary platform is Seismic (

, lacore dry tubes as wB as handling the SRM and ! Category L i IRM drytubes.

g,1_ ._a saarH tin. _ - --- -

,,, pg , ,,,,,

9.L42.y AsheNagEgnapment ter flange level working a ce for la vessellaspec.

tion and reactor internals servicias, and permits ser.

Fuel movement and reactor serviclog operations vicing access for the Adl vessel diameter. Typical op- t are performed from platforms which span the erstions to be performed are laservice inspections gueling, servicing and storage cavities. De reactor No hoisting equipment is provided with this platform, fcpM m iding is suppbod with at ama-aak refuel' -

as this function can be performed from the refueling

&for fuelmovement and servicing e.eemase -

platform or auxiliary platform. De platform operates j form for servicing operations from tA vessel llange on tracks at the reactor vessel flange level and is low.

level. ered into positics by the reactor building crane using ,

@ er Ph(t$om the dryer / separator strongback. De platform weighs ,

9J.42.7.1 Aasensele Rahell approximately 4,000 hs..and features 5 ft wide work ~

p areas and motorised travel.' The platform power is ' ,

ne assemesis re fa gastry crane, supplied by a cable from the refueling floor elevation.  ;

which is used to tra fuel and reactor compo. 1 seats to and from pool storage and the reactor- 9.1.4.2.7J Fuel AssemblySampler

[4 ~4ssel The platform spans the fuel r= d .

i vesselpoolf as bedded tracks la de refueling floor. The fuel assembly sampler (Figure 9.1 11) provides i A telescopias mast and grapple suspended from a a means of obtaining a water sample for radiochemi-troDey system is used to lift and orient fuel bundles cal analysis from fuel bundles while lastalled la the for placement in the core or storage rack. Control of core. The fuel assembly sampler consists o_f a sam-the platform is from na operator station on the - pling station, two sampling chambers and intercos-refueling floor- ascting tubing. The sampling chambers are kmered -

over four adjacent assemblies and samples are ob- -

/ ,

A position indicating system and travellimit com - . tained of the water in the fuel chanacts. -

poter is provided to locate the grapple over the

~

vessel core and prevent =liklaas with pool obstacles. - 9.1.4.2.8 Storage Equipment Two auxiliary hoists of 480 kg and 500 kg capacity i_ (approsisately 1,000 hs.), one main and one auxil. SpeciaDy designed equipment storage racks are pro .

l. lary monorail trolley mounted, are provided for ' vided.' Additional storage equipment is listed on' L incore servicing. The grapple la its fully retracted ; Table 9.15 For fuel storage racks description and -

position provides sufficient water shisiding over the . -fuel arrangement, see Senaaiaas 9.1.1 and 9.L2.

active feel during transit. The fuel grapple boist has >

a redundant load path so that so single composest Defectidfuel assemblies are placed la special fuel failure will result in a fuel bundle drop. Interlocks - storage containers, which are stored la the equipseat

on the platform

(1) prevent hoisting a fuel assembly storage rack, both of which are~ designed for the de<

l' over the vessel with a control rod removed; (2) pre 1 factive fuel.' These may be used to isolate leaking or , l vent satision with fuel pool walls or other structures; defective fuel while la the fuel pool and during ship - 1 (3) limit travel of the fool grapple; and (4) laterlock ping. Channels can also be removed from the fuel .

gn grapple hook sagagement with hoist load and boist bundle while is a defective fuel storage container.

.l r

ap poser. .

l Q The feel pool sipper may be used for out of core -- )

f e

9.L42.7J Animary anMmesi Platterm/ - wet sipping at any tims. ney are used to desea a dem o

fective fuel bundle while drcalatlag water through the gM As assuary putrorm is povided to allow versatil- fuel bundle in a closed system. De ball on the son-

,- g ley of operations. This platform operates 'over the tainer head is designed not to de into the fuel grapple.

L v Raastor Building pool and povides as additional f work men for reactor servicing. ~ A 500 lb sepacity choist is provided for reactor servicing taska. Part of; Amemensesa 9.1 es t

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. Standard Plant a- n 1 9.1 AJ.9 Under Reactor Vessel Servicing . 9.1AJ.10J ' Arrival'et Fuel on Slee c

' Equipment o;" m ..m . . .

. 1

~ . The new fuelis delivered to the plant on flatbsd The primary functions of the under reactor vessel  : truck or raucar. The new fuel is delivered to the re- 1 servicing equipment are to: (1) remove and instau ? ceiving stations la the reactor building through the ' ,

control rod drives; (2) lasta5 and remove the neutron rail and truck entry door. There, the incoming new..  :

l detectors; and (3) remove and lastall RIP Motors. - fuelis unloaded, inspected,'and prepared for use.- ~

Table 9.14 lists the equipment and tools required for .

,.' ' i servicing. Of the equipment listed, the~ equipment 9.1AJ.10J Refueling Pruedure handling platform and the CRD handling equipmenti s are powered pneumaticaDy. L A general plant refueling and servicing sequence di-.

agram is shown in Figure 9.1 12. Fuel, handling pro - a

' The CRD handling equipment is designed for the < cedures are shown in Figures 9.113 and 9.114 and ' -(

G removal and installation of the control rod drives: described below. Typical reactor building layouts are i , 4 from their housings. This equipment is used in coni . shown in Section 1.2 and component drawings of thel [

junction with the equipment-handling platform. It is ; principal fuel handling equipment are shown in Fig. .  !

designed in accordance with OSHA 1910.179, and ures 9.13 through 9.1 11.'

American lastitute of steel Construe ion, AISC. . L .

.The  ; der

_:?.desfel form provides a work i When the reactor is sufficiently cooled, the dryweU ( <

head and vessel head are removed by the reactor -

ins surface for equipmen personnel performing i building crane and placed in their respective storage; 1 work in the under vessgl area. It is a polar platform o ; areas; The reactor building crane a_ad cruciform. , 3 capable of rotating 360 This equipment is designed shaped strongback will be used to handle the RPV ' .

-in accordance with the applicable requirements of head and attachments. The strongback is designed so :

OSHA (Vol 37, No. 202,'Part 1910N), AISC, that no single component failure will cause the load tol ANSI C.1, National Electric Code, 3 drop or swing uncontrollably out of an essentiaUy hor-irontal attitude.- -

+

The spring reelis used to pull the incore guide'

.. )

l tube seal or incore detector into the incore guide- The strongback' attaches to the crane sister book by j tube during incore serymns. means of an integral hook box and two book pins.~;

Each pin is capable of carrying the rated loads. Each -

s -

The water seal cap is designed to prevent leakage , main beam of the cruciform is capable of carrying the i of primary coolant from incore detector housings < rated load.: " ~

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during detector replacementi lt is designed to indus- i . . . __

- trial codes and man ~ufactured itom corrosion resis L LOn both ends of each les are adjustable lifting rods,"

tant material. y su'spended vertically to attach the lifting legs to the t i RPV head.LThess"are for adjustment for even '

. The incore flange seal test plug is used to deter. . ' four point load distribution and allow for some flexi. +

mine the pressure lategrity of the incore flange - bility in diametrical location of the lifting lugs on thef O. ring seal. It is constructed of corrosion resistant 'H head. &

material.

1

. ._ y . j t The maximum potential drop height is at the point c i 9.1AJJO Fuel Handilag Tasks ;where the head gets lifted vertically from the vessel : " J Land before moving it horizontaHy to the head storage The fuel handling and transfer system provides a ,: ' pedestals.H safe and effective means for transporting and han. .

,s . .

.dling fuel from the time it reaches the plant untilit!

The shroud head load and the steam dryer load will: 4 l leaves the plant after post irradiation cooling. The - both be lifted with the dryer / separator strongback.1 following subsections describe the' integrated fuel ' ' 1 transfer system which ensures that the design b-L L Tl e t This 'strongbhck i[a cruciform shape:with; i of the fuel handling system and the requirements of . . box shaped at the four ends. Each hocket box has two-Regulatory Guide 1.13 are satisfied.

compartments to accommodate the two different' lug -

spacings on the dryer and on the shroud head.-

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Pneumatically operated lifting pins will penetrate the - receiving station within the reactor building. The-sockets to engage the liAing lugs.

~

crates are unloaded from the transport vehicle and ex- -1 n . amined for damage during shipment.~  !

' ~

Each of the above strongbacks are load tested at 3 1259E: rated load. At this test, measurements are The crate dimension are approsisately 32 X 32 X ' ~l taken before test load, seder test load and after ret - 216 inches. Each erste ontains two fuel bundles sup- 'j leasing load, to verify that deflections are within ac- ported by an inner metal catainer. Shipping weight . j ceptable limits. A magnetic particle test of structural of each unit is approximately 30W Ibs.' The receiving welds is performed after the load test to assure struc. station shallinclude a separate area where the crate .

turalintegriry,_ cover and the inner metal container can be removed ' l from the crate.' Both inner and outer shipping con-yh seal esists around the vessel opening to seal the. tainers are reusable. Handling during uncrating is ac- (

Il from th ;;r ;::' le ehe-meentimet eddi. complished by use of the reactor building cranes. The wig '"%esel water is umped into the pppee. pool. Once inner container is tihed to a position which is almost

" 3 he ;;- g:' is filled, tbe dryd and separator are vertical, while the fuel bundles are unstrapped and re.

gremoved and transferred to their storage arensi moved from the container with the reactor building . ~

t

- QM11% thin the :;;: ;r! using the dryer / separator - crane.' They are then transported to storage in the ~

strongback. The tools are used in these and subse- new fuel storage racks located in the new fuel storage .

quent reactor serdcing operations'are listed in Table vault or to the new fuel inspection stand locatt.d on -

9.12. Once access to the core is possible, the the refueling floor. ,

refueling platform can relocate and move assemblies  ;

to and from the pool storage racks ' Simultaneously - The actualinspection of tbc new fuelis normally -!

RIP, CRD hydraulic system, and the neutron moni- . deferred until all the reusable containers are emptied ;

toring system may be serviced from beneath the : and the area around the'new fuel vault cleared. 'At :

vessel. -

that time, the individual fuel bundles are removed .

kg4 -..2; from the vault, inserted in the new fuel inspection '

During refueling. the ausemeeie refuelingnaelu'ne ' stand, dimensionally and visually inspected and re- , ,

transfers the spent fuel from the core to thn spent turned to the storage vault to await assembly with?

fuel pool. The spent fuel assembly is placed in the ~ channels.' The new fuelinspection stand accommo .

fuel preparation machine, where its channel is teT dates two fuel assemblies at one time, f moved and fitted to the new fuel bundle previously .

1 l- laced in the machine. During channeling, the spent 9.1.4.2.10.2.1.2_.Channeling New Fuel el bundle is placed in the storage racks by the asse-l: f84 "Mnasic refuelingmeehies. The h refuelingNew fuel is unloaded from the new fuel vault -

l maahese then (Jaces another new fuel bundle in the transported to the fuel racksin t fuel pool. Us ,J fuel preparation machine for channeling. ' _ channeling new fuel is'done ocurrently with g

dechanneling spent fuel Two fuel preparation -

When refueling and servicing are completed, the , - chines 'are located in th nel pool, one use c steam separator assembly is replaced in the. vessel,j dechanneling spent fuel the other to channel new - '

the steamline plugs removed and the steam dryer re6 ' - fuel. The procedure is as follows:1 Using the enso. ~ 4 turned to the vessel. At this point, the gates are in- ma6ie refuelin smashine; a' spent fuel bundle iNr"a~aI .

g stalled, isolating the reactor well from the upper - ported to the fue prep machine; The channel is un-gools. The reactor well is then drained to the main bolted from the bundle using the channel bolt wrench.

condenser. With the reactor well empty, the vessel The channel handling tool is fastened to the top of the . ,

and drywsD beads are replaced. : c channel and the fuel prep machine carriage is lowered

. removing the fuel from the channel. The channel is -

9.1.4J.10J.1 New Fuel Preparation then positioned over a new fuel bundle located in fuel' 4

prep machine No. 2 and the process reversed. The c 9.1.4.2.10.2.1.1 Receipt and Inspection of New channeled new fuel is stored in the pool storage racks -

Feel ready for insertion into the reactor.'

The incoming new fuel will be delivered to a Amdess 4 M.4g

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' 9.1AJ.102.1J Equipawat Preparaties - r Each stud is tensioned and its out loosened la a -- <

i series of two to three passes.' Finally, when the' sets L '

t Another impedient in a succeuful refueling outage : ' are loose, they'are backed off using a nut runner until - d is equipment and new fuel readiness.' Equipment ( only a few threads engagei The out is hand rotatedh_

a.

j'

. long lying dormant must be brought to life. AU tools,' ' free from the stud and the nuts and washers are grapples, slings, strongbacks, stud tensioners, etc., . . placed in the racks provided for them on the carousel; _

l will be given a thorough inspection and operational: < When all the nuts'and washers are removed,' the; ;l check, and any defective (or well worn) parts will be , vessel stud protectors and vessel head guide caps are -

replaced. ' Air boses on grapples will be checked.- inmallad. '

Crane cables wiu be routinely inspected. All acces-- >

'l .

~

, 1 sary maintenance will be performed to preclude Next, the head, strongback and carousel are 'trans. 1 outage emession due to equipment failure. : ported by.the Reactor Building crane to l' ea he'ad L holding pedestals on the refueling floor. The head '

9.1A.2.10J.2 Reacter Shutdown holding pedestals keep tbc vessel head elevated to fa.

-cilitate inspection and 0 ring replacement.. m The reasser is shut down according to a prescribed . ,. 11 T . . p procedure. During cooldown, the reactor pressure vesselis vested and filled to above flange level to The studs in line with tbc fuel trans

moved from the venel and placed in tYe rack prov promote coohag. . vided for them. The loaded rack is transported to thel " '

refueling floor for storage. Removal of these studs c1

9.1A.2.10.2.2.1.Drywell Head Removal provides a path for fuel mowment.

• Immediately after cooldown the work to remove; ' 9.1A.2.10.2JJ Dryer' Removal the drywou head can begin. The dryweu head will be ; .. ' f '_ e attached by a quick disconnect mechanism. To: The dryer. separator strongback is lowered by the - ,)

remove tbc head, the quick disconnect pins are with( reactor building cranc _and attached to the' dryer lif6 ,

drawn and stored separately for reinsertion when the head is replaced.' The drywell head is lifted bp the -

ing lugs.; The dryer is lifted from the reactor vessel e y and transported underwater to its s ctage location'in  ;

eweshead vilding crane to its storage space on the?

[ (.),g6 refueling o' or. The drywell seal surface protector is the Dt-/8 %

"7- adjacent to the reactorwell. 0 j

.instaued before any other activity proceeds in the re.

9.1A.2.10.2JJ Separator Removel'

+

3 actor well area.

Lin preparation for the separator removal, the

1. s

(?>%1) 9.! A.2.10.2.2.2 Iteactor Well Servicing steamline plugs are installed in the four mais steanic nozzles.: The separator is,then' uabolted from the

When the drywell head has been removed, several ' l shroud using' shroud head bolt 'wrenchesGWhen the .

]

pipe l' ass 'are exposed. These lines penetrate the re- unbolting is accomplished, the dryer separator- g actor wou through openings. The piping must be re- strongbeck is lowered into ths vessel and attached to i moved and the openings scaled. There are also vari. l the separator lifting lugs. The separator is lifted from :

'ous vent openings which must be made watertight, the reactos vessel and transported underwater tolthe' storage locolon in the a n ; 'jdjacent to the ;

Water levelin the veuel is now lowered to flange reactor well.

O level in preparation for head removal.

1D/g}?it

+ m .

,z i M 9.!A.2J0JJA Vuel AssemblySamplingl c I

. 9.1A.2.10JJ Reactor Vessel Opealag '

9.1A.2.10.2J.1 Vessel Head Removal

.During reactor operation, the core offgas' radiation ; J ~

level is monitored. If a rise in'offgas activity has been , a L

noted, the reactor core may be' sampled during shut o ,

The combination head strongback and ' carousel 4 i down to locate any leaking fuel assemblies. The fuel; J stud tensioner is transported by the Reactor Building - ' sample isolates up to a 4. bundle array in the core. ' j crane and positioned on the reactor vessel head. This stops water circulation'through the bundles and L .

4 allows fission products to concentrate if a b'undle~is =  ?

~

defective. ~After 10 minutes,'a water sample is ' taken ;

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' Ef fissi:a prodret a:alpais. Ifi defective bundle is boist and auxiliary boist prevent boisting of a fuel as.J '

f. nd, k is transferred to the]uel asesege pool and c . sembly over the core with a control rod withdrawn;-  ;

i

/. . ; stored is a special defective Ibel storage container to - i interlocks also prevent withdrawal of a blade with a . F FI . minimize background activityin they pool.: fuel assembly over the core attached to either the fuel 4

..'E , speht' 4estk ~~ , . grapple _or auxiliary.boists. Isterlocks block travel c ,

g ,

9.1;4J.102.4 Refbellos and Reactor Servicing  : over the reactor in the startup mode. * ' '

1' '

nW 900l L The gate isola t rguely from the reactor The refueling platforia contains a system that 'isdi. _

1 well is removed, thereby interconnecting the uppes cates position of the fuel grapple over the core.' Thei

, pool areas. The ' refueling of the reactor can now readow, in' the local control room, matches the core 1 4 d

c _ begia.1 - artangement cellidentification ' numbers.- The posig u L

.' ,9,1.4.2.10J.4.1 Refuellag

, . 7' .

tion indicator is accurate withis 1/5 inch, relative to -

y;

' O actual posillon, and minimizes jogging required to : '

. . correctly place the grapple over the core, . y ,

' During a normal outage, approximately 25% of the s JQi_,.. _

D. ._ _

,l' fuelis removed from the reactor vessel,25% of the - To m'ove fuel, the fuel grapple is aligned over thei '

fuel is abuffled in the ' core (generally from peripheral; ' fuel assembly, lowered and attached to the fuel? mi s; to center locations) and 25% sew fuel is installed. ! bundle bail.,The fuel bundle is raised out of theggt / H JT actual fuel bandling is done with the essemetic' : moved through the refueling slot to tb apper pool,7 " peh ,

f verueling,meabina. It is used as the principal.means - positioned over the storage rack and ' red into thei lj

" f transRrting fuel assemblies between the reactor

~

rack. Fuel is shuffled and new fuel is moved from the m .;

gW 11 and the. upper

' transport device.

I;it also serves as a boist and l upper pool to the reactor vesselin the same manner.

provkles an operator with work . l t i j

l

[ surface for ahnost all the other servicing operations. 9.1.4J.16J.$ Vessel Closure? o-The' platform travels on track extending along each : N >

.e of the reactor well an(pool'and supports' tbs; ,..Tbc following steps; when performed, will retura ,

'q refueling grapple auxiliary b5tsts. Tbc grapple is sus. the reactor to operating condition.LThe procedures E i pended from a trolley that can traverse the vedth of i tbc platform.

are the reverse of those described in the preceding . , 7 7[M l

/'

3- l sections._' Many steps are performed in parallel and:

, not as listed.

'fM

( 'C ",

' The refueling platform has two 480 kg (1/2 ton)'- t

~

auxiliar  !*

0*8 '4e.y boists. C._ "": , 0.4. '. n..Z u-- -95 pint oist h a6r.

. (1)~ Core verifiution . the core position of each fuel *A'"

. assembly must be verified to assure the desired '

mally can be used with appropriate Or pples to core configuration has been attained.' Underwa. -

bandle control rods, guide tubes, fuel support pieces,; ' ter TV with a _ video tape is ~ utilized.LCable'op-; ,l sources and other internals of the core. The auxil. .  : tionalJ%

lary hoist can also serve as means of handling other ' . m ' "

equipment within the pool. A second auxiliary boist - . (2) ~ FMCRD testsi the. control rod drive' timing,;

is mounted on the platform trolley.~ ' .Lt friction and scram tests'are performed as reE The. platform control system permitis j _ quired."

1 M

"l=

4 l[4 variable. speed, simultaneous operation of all three . (3)L Replace separator.'

j platform motions. Maximum speeds are: ' . ..

,_ g .,1 .

(1) bridge 66 fpm (20 m/ min)-

l - (4)g Bolt separator and remove four steamline plugs. ), M e (5) - Replace steam dryer, m

• f  ; t

.(2) trolley 'l '

~

33 fpm (10 m/ min)- . .

g

'(6) "Installpustgates.' '

(3) grapple boisti 39 fpm (12 m/mia)

~

(7) . Drain reactor well.' _

l A single operator can' control all the motions of the .

J 4 platform required to handle the fuel assembliesL (8) N Remove drywell seal surface coveringt open during refueling. Interlocks on both the grapple &

e drywell vents. 'N s

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. 'eables on the y ' oke are attached to the head sad se u (9) Replacevesselsuds.= cured and the closwe outs are dissagaged. The cadL is ass raised and transferred lato the ensk pit."

00) lastall reactor vesselhead.

~

Tlie~ cask is moved to a position over the esater of

03) lastat vessel head piping and insulation. , the cask pk and slowlylowered into the cask ph_.matil

' k r' ests on the cask pk Soor.

02) Hydro test vessellf required.

The cask lifting yoke is lowered 'undt e hengaged '

03) lastaB drywon head;1eak check. from the cask trunnions and the closure be 'd lifted <

1 off the cask. The closure head and yoke are saved .

04) InstaB shield pings. -

- loto the cask washdown pit for storage. The ea' l -

gates between the cash pit and i

{7+

elsessage -

05) Snowpaalgates,f 2are removed and spent fueltranaf om the storage <

racks to the eask la started.1 .

06) Startap tests - the reactor is returned to full '

g -

power operation. Power is increased graduaDy? 'th fuel pool la a series of steps unit the reactor is operating Spent feelis transferred maderwater from,as.

us is y toggI- the cask us at rated power. At specific steps during the ap- ; gr proach to power, the incore flux monitors are mounted on the 9-k r '

s Pgg e Wbes the cask is filled with spent fusi, .

onlibrated.- '

' tween the ensk pit and thejuel asesage'poo  ;,

placed. De closure head is Nplaced 'on the cask and : ' '

y 9JA.2.88.3 Departurv erFuel From S6te the lift yoke engaged with the cask trunnions. The loaded cask is raised, transferred to the cask The empty cask arrives at the plant ce abe special washdown pit, and slowly lowered to the pk Door.' H Satbed railcar or truck. %e personnel shipping bar.

rier and transfer impact structure are removed fross - f.The jask is ebecked by health physics personnel and - -

the large casks and stored outside the rail entry door. decontamination is performed la the cask washdown '

Heakh physics perseasel check the cask emerior to. pit with high pressure wstor sprays, ehemicals and f

determine if decontamination is necessary. Decos. : hand scrubbing as requit sd to clean the eask to the < f":.

tamination, if required, and washdown to remove seed dirt,is performed before removal of the cask - - level required for transpc rt. Cooling from the transport vehicle. The R/B equipment entry airlock door is opened and the cask with'its -

? available _ia the dessesas sinasies>eemi la threvent

? cooling is required duris g decontam activities.

"g l wp. sosseag' ~

The' remaining closure : uts'are replaced and tighty @g' '

transport device moved into the building. The rail Teasd.: Smsar tests are pe ' formed to. verify sisasing to n b

car ertreekis blockedis pr=&a 1

~ a s .oEske transportation regoements.  ; M y 1

The airlock door is closed and the ensk is lospected for shipping damage.

_ Ds sisamed cask is lowered frees the refueling floor to the reactor building entry lock onto ensk skids with (

qj

+ j the roastor building creas and mounted on the trans. ,

Tbs eask ecoling system of the transport vehicle is = 1 port vobiele. The cask ecoling system'of the transport j diseconected. The cask yoke is removed from ks vehicle is ceasemed to the ensk and the ensk internal; i jD

~

morage posklos oc the Ratbed and attaebed to the J pressure and temperature are mookored. _When they?

ensk treacions. Tbs yoke engagesset, car brakes , are at equHibrium ecoditions,' the ensk is ready for-and wheel blocks and clearances for eask tBt and lik ? 1 shipment. De personnel barrier and impact strue.

era chectadi The cask is tihed to the vertical posi 1 ture are replaced. De reactor bauding airlock facihty

> tion with combined main hoist lih and trousy move. '  :)

doors are opened and the eask and transport device meet. With the cask lo a vertical position, abe cask is ~ are moved out of the roamor bsBding. '

Rhod approminately 5 A odf the transport devies skid - &

moendag trussions to clear the upper coolant duct.-

h/m ensk is moved up to the refalingspiese 9Jnoor and thasof Feel - Raadung

~

A 3 Sadhty Itvalsation 1

1. [, "he cask ' 4 wt and slowlylow. - ,

ered to abe Door of the,usalt. Closure head lihing -

, ) en pit "

Aamnedsammu, $

~.

9.M  !

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t

ABM Starvtant Plant s u sia w i i a- a ho Safety aspects (evaluation) of the fool servicing aer auxiliary boist en the trouey aseeeeeber bohren ,.

equipment are discussed in Subsecsion 9.1.4.2.3 and bewirmaeorelle nose three boists are precluded ,

safety aspects of the refueling equipment are dis- from operating simukaneously because control power

. cussed throu6 b out theb 9.1.4.2.7. In addialco, is available to only oos of them at a time. The two a followlog summary safety evaluation of the auxiliary bolsts have load cells wkh interlocks which j fuel-bandliog system is povided below. prevent the boists form moving anything as heavy as a {

fuelbendis.  !

The fast pep machine teamves and installs chan-nels with au parts remaining underwater. Echani- The two auxiliary boists have electrical interlocks cal stops prevent the carriage from liftlog the feel which pevent the littlog of their loads higher than a j bundle or assembly to beight where water shielding specined limk. Adjustable mechanicaljam stops on i is not asificient. Irradiated channels, as weg as smau the cables back up these interlocks.

parts such as bolts and springs, are stored underwa- ---

l ter. De spaces in the channel storage rock have I l center posts which prevent the loading of fuel buo-dies into this rack, rwhich Deis spent fuel linked in travel so kbandhng crans is a low cannot carry a abipping cask over stored fuel Also, ks height is limited such that the cask cacaot be lifted up on the operating ,

nere are no nuclear safety probleins associated floor. Dus the cask cannot roll into the fuel pool if 1 I

with the bandling of new fuel bundles, slagly or la gaccidentauy dropped.

pairs. Equipment and procedures prevent an accu. -

mutation of more than two bundjeti n any tarash Further, a series of watertight gates are provided I pca'Cioth such that the usk never exceeds the 30.ft drop dealgaweg- l The aseemetic refueling mashine is designed to criteria. The cask is moved to tb 'n'; _ ._Dd J prevent it form toppling int /the pools during a $$E. gated off and thelendes pit filled water , el f Redundant safety laterlocks, as well as limit switches, then is the fuel asessee Pool connected to the SpWN ,

are provided to prevent accidentally running the and the fuel transfer begun. When the cask is loadge -

grapple loto the poolwans. Tbc grapple undized for tb fuel "sem08 Pool is gated closed and the cask N. ,- ,

fuel noement is on the end of a telescoping ost, a procedure reversed. A cask decontamination At full retraction of the mast, the grapple is suffi. pk areais provided. "

l

['

5 ciently below water surface, so there is no chance of raising a fuel assembly to the point where k is inade. Light loads such as the blade guide, fuel support quately sbicided by water. The grapple is boisted by casting, control rod or control rod guide tube wei8h redundant cables inside the mast, and is lowered by considerably less than a fuel bundle and are adminis-pavity. A digital readout is displayed to the opera- tratively controued to eliminasa the movement of an (,

tor, showmg him the exact coordinates of tbc papple light load over abe fuel pool above the elevat' re-I over the core. quired for fuelassImbly MHS Thus, the kinetic l energy of any light load would be less than a fuel i ne mast is suspended and gimbated from the trol- boodle and would have less damage induced. Sec.

I ley, near its top, so that the mast can be swung about the axis of platform travel, in order to remove tbe ondly, to satisfy diin6 components over NUREGfuel pool 0554, the equipment are desigBWo ha grapple from the water for servicing and for storage. seet the single fauure - criteria. #

l The grapple has two independent books, each op- in sammary, tbs A -'8-ha system complies with

=

@g s' erated by an air cylinder. Engagement is indicated to General Design Crkeria 2,3,4,5,61, and 63, and ap-the operatoe. Interlocks prevent grapple disengage-

.g,,

plicable portions of10CFR50.

ment until a ' slack cable' signal from the lifting cables inchcates that the fuel assembly is seated. De The safety evaluation of the new and spent feel stor-slack cable indication is also used to determine if s age is praemased in Subsections 9.1.1.3 and 9.1.2.3.

M fuel boodle is lodged in a position other than its

$ normal, seated positionin the core- 9.1.4.4 lospection and Testing Regelrunents In 9.1.4.41 Inspestlos to the main boist on the trouey, thereg e t Aad= s 9.3.sa l.

, y w .;;: . ,

I r

.i I' i

@ :In' addition to the slack cable signal, th's.Lelevationj of f tho '  !

grapple is-continuously indicated.- 'Also, fatter.the grapple: -'

is disengaged, the position'.of the-upper.part of,the fuel-bundle can be observed using television.' ~

+ <

4 The spent fuel . storage racks are - purchased equipment. ; ! The 9 purchase; specification for these racks willarequire-the?

vendor'to provide ~the-information; requested.'in Question; '

3 430.192-pertaining to load-drop; analysis.: .See Subsystem 9 9.1.4.3:forl interface requirements. J, m .,,

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ReSmeBag and servidag equipment is subject to the ! electrical and/or mechaalcal functions are opera.

1i.  : aria controls d quality assarance, incorporadas the ; tional . m >

'sequiramama of federal regulation 10CFR$0, Appea-Q . -

as B &:- : 1 f:'l ' m rr-M e ri:- " - Passive unksisuch as the fuel storage rocks, are vi.

1se,dhe,feeIl storage rack (kfueling platform,and

'suaDyinspealed prior to ase.; '

m. m-,__.

.__m.a

• have an additional set of engiacering specified

• qual. '

' Fuel bandilag and vessel servicing equipment by requkements' that ideatify h': ; ::'r ' features prsoperatlosal tests are described la Subseetioni abich require specine OA verification of compliance - H.2.12. ; -

+

• t ,

4

{> aodraming requirements. , , _ ..

9.1M Instrumentaties Requireasots '

For composeats einmalGod as American Society M Pldtfarm ,

Mechaalcal Engineers (ASME) Section RI, the abop 9.145.1 Amsmaanse gneashine t

. eparation must secure and maintain as ASMR Tr esap, which requires the submittal of an acceptable _ _

.l

! The assomatic refueling,plaMors e 3

Ashm quality plan and a corresponding procedural

~

tion indicator system that'i@nforms the operator which

=====1.

J eore fuel cou the fuel grapple is accessing.

Interlocks ?

g 1  ; and contro; room monitor are provided to pavent the :

AddhionaDy, the shop operation must submit to # ' fuel grapple from operstlog la a fuel cell where the U toquest ASME audits and composest inspections k l control rod is not la tbs proper oriestation for ; ' '

j by resident state code laspectors. Prior to shipment,; , refueling. 4 i sesry component langwdan kers is reviewed by OA -- _. , _ .. . , 1 sapervisory personnel and combined into a summary . ' Additionally, there is a series of mechanically acti ' d poduct qualuy checklist (POL) By issuance.of the; voted switches and relays that providef monitor indi. 1 POL, verification is made that_ all quality requires sations on the operator's console for grapple limits,7 mama have been confirmed and are on record in the boist and cable load conditions, and confirmation that podmot's historicalfile. s ' the erapple's book is either engaged or released.

f bi .

.~ .

.. -i 9.lA42 Testing . A~ series of load cells is installed to provide auto { -!

matic shutdown whenever threshold limits are sai '!

QuaBRestion testing is performed on refuehag had . esaded for sither the fuel grapple or abe auxiliaryz servicing equipment prior to maki unit production." 1boist units.S E ~

d~'

i 4 '

Test specificadoes are defined by the responsible - . < .

3 design engineer and may include sequence of operar - 9.1A5.2 FeelSupport Grapple.' '

eions, load capacky and life cycles tests. These test -

it -

f medvities are performed by as indepeadent seat engi- , Akhough tbs fuel support grapple is not' essential to 1

seerlag group and, la many cases, a full desi 1p 8 , safety, it has as instrumentadon system enesisting of 7 1 soview of the podna is conducted before and after - mechanical swkebes and indicator li bts. 'Inis hystem 6 q

- she goalification testing cycle. Aay design changes  ; provides the operator with a positive lodication that J j assaing function, that are made after the'comrle; the grapple is paperly aligned and oriented and that! ,,

tion of qmalincados testing, ars' requalined by test or; , the grappling meebasism is sitbar emended or re -

entnatanna.  !

~' ,L traced. A ,R

, ';  ; 1 Philemal tests are performed in the shop pior to - l 9 JAB.3 Other she abipment of productico volts and generally lac ..

m

- clade slearical tems, leak tests, and sequeses of op .

' ' Refer to Table 9.15 for addialocal refueling and - i eratises easts.  ; servicing equiposos not requiring instrumentation. A j 1 g*j T N;d When the molt is rossived at the site, it is laspeasd ' 91A8.4 Radiation Montecetag -

to ensus ao damsgo has occurred during transit or 2 _- _ ,

morage. Prior to use and at periodicintervals,each- f The*pdiation monitoring equipment for the h' piece of agalpment is agala .g

- sectico7A1 tested to ensure theWlfA,,

refueling an I

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t c The- radiation, monitoring; equipment? for. theifuely handling ;

area is.discussediin' Subsection- 11.522.1'.2.1.. The fueltarea~ >

j ventilation: exhaust radiationi monitoring'?.is1 discussed in e

. subsection 11.5.2;1.3.- "

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(($) QA Program to' Monitoriand Assure

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Implementation and Compliance of Heavy Load - a

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4  %- l Handhng Operations and Controls. -

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' ti- Genere! Electric Standard Application for <

1 N, Acacsor fuel, (NEDE 24011 P A, latest approved - Hj

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'9.1.6,-Interfaces. , 29 1

s

'f 3 -9.1.6.1 New Fuel Storage Racks' Criticality: Analysis  ; ,  ;

q The. applicantiref erencing the ABWR design > shall' provide th'eLNRC4 .

.confirmatcry criticality analysis as7 required-by Subsection- .

9.1."1.I'.1. <

m, c

L9.1.6'.~2. Dynamic.and Ispect;An,alyses'of;NewiFuel Storage.' Racks'

.J a

y..

.The applicant; ref erencing the ABWR' designishal1 ~ providelthe .NRC'

~

confirmatory; dynamic andtimpact almlysesiof thefnewl fuel'. storage l :1 racks. -See Subsection 9.1.1'.1.6. 4

- 4  :)

E i 9.1.'6.3l Spent 5 Fuel'_ Storage l Racks Criticalith Analysis!

W .

The applicant referencing'the ABWR detiinish'all-' provide The.NRC I Y U L

confirmatory criticality analysis as required'by Subsection '

1 .

J

~

. - . 9.1. 2. 3.1. ,

7 t - 4 -

3 9.1.6.~4 SpentiFuel Storage Racks Load Di*op Ana!ysis: ,

The-'applicantTreferencing.the.ABWR design shal1+ provide:the NRC'.'

u , , .. e . .['.  :?

.confirmat ory1.oad drop' analysis asi required byeSubsectioni, "" s t

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- V l Table 9.12 r:'1 J

\- = FUEL SERVICING EQUIPMENT 1 h F f Essentist Sassey

' Compeneet Classis. < Q assis. , Quality Selsene ',

!' . No. Identificatis - enties L s enties / Group! ,.Catsgery *

- (s) .'. Gs)i - ' - (c).'. ^ (d) -

, a,

=1' FuelPrepMachine NE 'OL ,' E . f NA-;. ,

e <

2? New Pust zmpaction stand .

se

'NE- ;O lEi

/ NA! -'

y. '

S 3 ChannelBokWrenchi NE .O E. , NA ~ M' .t 4 ' Channel-HandlingTool ( NE . ~ , 'O- 'LE- , .NA"

' t s y I-5' _ FuElFoolVacuum Sipper NE' '

O- ,

LE :NA- <

v t v ' .

. . 4i. ;E

.- 6 . c General-Purpose Orspple .. . NE - y0 , ?E >

3NA ;.

(7 JibCrane s

0' NE: 'O E

' I 3)k[. -

s

'8: Aulematic Refuelingl HOL LE:' I! '

g, $

Mahine Plathemi *a ' NEo '

t 9 - Channel-Handling M$ chine NE *

.O1 E: fNAL ,

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- NE '= . Nos Essential ' M (a)

L PE = Passive Essential: ,

• 4

-(b) 0.4 Other ,'

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o, J' (c) B ' =' : ASME Code Section HI Class 2 -  :!

!- D"= ANSI B31.1 : .j:,

E' = ElectricalCodes Apply

"-' j~

'i I t= _ IndustrialCode Applies u-g , .

' s ,o (d) NA = : No Seismic Requirements I- = :- Class I .

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Table 9.1 10 i -

<. 4 .

=;

h;  ; , SINGLE FAILURE. PROOF CRANES j

f g -

1. Reactor building crans ,

M-= HOP %

2. : Refuelingbaidge erone (',

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SECHON 9.2

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CONTENTS (continued)J .

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- 21 tis  : P.ast -

l 911612 Power Ocaeration Design Bases 9.2 12.1 7 1 g- . <

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7; 91162 System Deswiptics ' 9.2 12.1 7 , .;

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' 9sut1 , OcaeralDesaiptics

~9.2-12.11 .

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< 9116.2.2-' . Componest Description - 9112.1: .J l 911613- , System Operation 9.2 12.1 - ,

l 1

q 9116,3 9.2 12.2 -;3 -j

. Safety Evaluatid. a - .

9116.4J Tests andlaspections? s- <

9.2 12.2 - s

,3 9116.5 Instrumsstation Application ' 9.2 12.2 e

,r 9117 jatufass '

9.2 13 *

.w 9117.1 < Ultimate Heat Sink Capability:- ^ ^

9.2 13 J,

\ ' .d 9117.2 Makeupwater ystem capability - 9.2 13 :

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s { e 9.2 WATER SYSTEMSL 9.2.1 Statlos Service Water System

                                                                               =

9.2.9 Makeup Water System (Condensate)! 9J.9.1 Design Bases l' j

                                                                                                                              ~

The functions normally performed by the '(1) The makeup water. condensate system (MUWC); L station service water system are performed by the shall provide condensate quality water for; 1 systems discussed is Subsection 9.2.11. . both normal and emergency operations _ when . ' '~' E ..

                                                                        '                                         required.:                                                                                                                 ,

9.2.2 ClosedCoollagWater, System ' p' f (2) The MUWC system "shall provide's required -jI

                                       . Tbs functions normally performed by the closed -                         water' quality as follows: >

cooling water system are performed by the systems . . .. . .

                                                                                                                                                          ..f.'               . -

• 1 eiscussed is Subseetions 9.2.11, 9.2.12, p.2.13, Coaductivity(p S/cm)110.5 at 25'C , ~{

4 and 9.2.14.' Chlorides, as C1(ppm) La g.02 i b . ,1 n

                                                                                                               <pH                     , : 5.9 to 8.3 at 25 C1 . c ." . .. .                                                                 !

9.2J DemineralizedWaterMakenp Conductivity and pH' limits shall be appliedH 7 , System x ' after, correction for dia' solved CO2 . -(The

                                                                                                              < above limits'shall be' met at least 90% of                                                                              j

the time.);

J

                                     - ne functions normally performed by the domine                                                                                        >                   >

eralized water makeup system are performed by the , , systems discussed in Subsections 9.2.8,9.2.9 and ? - (3) %e MUWC system shall supply water for the K > 9.2.10. uses shown in Table 9.21/c ' 9.2.4 Potable and SanitaryWater . (4)lThe MUWC system is ot safety related.

                                                                                                                ~

d Systems . .

                                                                                               . ' 5 The condensate stora                                                                                                              @

teh(ea) ' capacity of 2,110 m3.ge tank sh'all have a?

                                                         ~

Out of ABWR $taqdarc , Plant Scope. S ee S SThis capacity was-- M 9 7.l'T.3 fey odora e 44 Mh f. i determined by the capacity required by'the 'f 9.2.5 Ultimate Heat Sink--- 7 uses shown in Table 9.2 2. ~ c ilc a

                                                                                                                             .i, .                     .M.                           ;;p .J                                            A Out of ABWR Standard Plant scope. See . (6) All tanks, piping'and other equipment shalli 3
                                ] Subsection 9.2.17.1 for interface' requirements.                             -be made o,f corrosion.resistaat materials!                                                      '

s , . . . , 9.2.6 Condensate Storage Facilities -  : (7) The 'HPCF an_d' RCIC ins. + ,trumentation,which j , .  ; and Distribution System' laitiates the automatic switchover of HPCF o d' and RCIC suction from the CST header to the. The functions of the storing and distribution  : suppression pool; shall be"designe'd;to:  ; of condensate are described in Subsection 9.2.9. . _ safety grade requirement's (including - 0

Installation with 'accessary; seismic: R 9.2.7 Plant Chilled Water Systems support). ,' J'. , -

                                                                                                                                                                                                                                   @q L                                         The functions of the plant chilled wateri 9.2.9J System Description]                                                                                 /

1: 4 system are performed by the systems described in .

                                                                                                                                          .1        . . .                                 .           ..        . . _

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                                 ' Subsections 9.2.12 and 9.2.13.

De MUWC P&ID is'shown in Figure 9.2 4 This 3t , system includes the following m ; -

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9.2.8 MakeupWaterSystem(Preparation) j 1 J (1) 'A condensate storage ta'ak (CST) is provid. 1 Out of.ABWR Standard Plant scope.1 See . ed. It is' of concrete ' construction .with 'a < 1 Subssetion 9.2.17.2 for laterface requirements. . stainless steellining. The volume is shown - g L , ;in Table 9.2 3. c(2) The following pumps take' suction from the

                                                                                                             ' CST:

4 Amendment 11 .9.21. 'l g j l

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          ..                     Standard Plant               .                                                                                                                       aw a                      '

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                              / evaporator. If the temperature of the?shillhd - 9J.14J System Descripticaj water drops below a specified level, theT                                                                             l'..
                                                                                                                      ~ ' '

eestroller automatically adjusts the positloa~ of L 92.142J General Description < the compressor inlet guide vases.t Flow switches ? .

                                                                                                                                                         'V
    ~
                               . prohibit the chiller from operstlag unless there                        : The TCW system is illustrated on Figure; is water flow through both~ evaporator and' 9.2 6.: The system is a slagle loop system and1 sendsaser. .                                                      consists of one surge tank, Lone chemical <

a d di tio n ~ t a n k,f --- ' ---- ' "-- '--' ' 9.2.14TurbineBelldingCoolingWaterSystem -- '------ (connected la paratiel)Tand' PU e

                                              <                                                    associated coolers, piplag, valves, controls,i                                                                      C' 92J4.1 Design Roses )

• and lastrumentation. Heat is removed from the1 @

TCW system and transferred to the non. safety - 40 9JJ4.1.1 gadsty Desiga tases i - related turbine service water system (Subsection ' 9.2.16). J The turbine halldlag cooling water (TCW) .' ,

             'j                  system serves ao' safety function and has ao - . ; A TCW system sample is parlodically taken . .

safety design basis, < ' for analysis to assure that the. water quality , y f:,,

                                          ,       .      .                                       meets the chemical specifications.;                                              o 0

92.14.12 pesar Generstles Design Bases ' ' i

                                                                                           . 92.142J Composest Descripties -
                              . (1) The TCW spitem provides corrosion lahibited,1 deminerallaed 'coollag water to all turbinet                  . Codes and staddards applicable to.the TCW L island auxiliary equipment listed in Table: : system are listed la Table 3.21.1 The system is ?                                                                                       1 9.2 11.                                                  designed in accordance with. quality group D;                   '

specifications.

g . . (2) During power operation, the TCW system . operates to provide a continuous supply'ofi ' The chemical addillos tank is located la the i J

                                      ' cooling water, at a maximum temperature of" turbine building in close proximity to the :TCW.                        ~

I 1058F to the turbine island auxiliary system surge tank.1 ' 1 equipment, with a service. water talet  : 400 ? -- .- . temperature not saceedlag 958F. . De TCW pumps are de% capacity sach and are ? constant speed electric motor driven, borizontal: , o (3) The TCW system is' designed to permit the- centrifugal pumps.L The three pumps are; ' q maintenance.of any single active component; connected la parallel with common suction and(

                                   ' ' witbout latstruption of the cooling),'dischar e lines.'                                               <

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                                                                             ,.                      ! The TCW heat exchangers are 40% capacity 3                                                                        ,       .

(4) Makeup to the TCW system is designed to L each and ' arc designed to have the TCW water 4d '" permit costlevous system operation'with Leirculated on the ~shell side and the power cycle i design' failure leakage and to permit 1 ; heat' sink. water ' circulated on the tube side. 1 expeditions post maintenance system refill.- J The'aurface area'is based on normal heat load. '

                                                                                                                                               ~                       ..

(5) The TCW system is designed to have an; The TCW surge tank is an' atmospheric carbon 6 f l~ atmospheric surge _ task located at tho' steel tank located at the highest point in thes X highest potat la the system.  ; TCW system. The surge tank is provided with a

                                                                                               'Isvel control valve that controls' makeup water g

(6) The TCW system is designed to have a higher ' addition.- ' ~

                                                                                                                                                                                                                  $3 d pressure than the power cycle' heat slak.
                                                ~

t . K 'I water to easure leakage li from the TCW : 92.14JJ System Operaties system to the power cycle beat sink in the . 4

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?- - event a tubs leak occurs la the TCW system - During normal power operation, gg .one of the'- H beat exchanger, these 50% capacity TCW system pumps circulate y U8 L Ansseneet11 ' 9.2 10

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There are-no connections between-the-TCW' system'and ar.y safety-related. systems.. 9 "two pumps.with..a' capacity of-29,000 gymieach :tgo heati , u-exchangers' with - heat removal capacity ofl 130 x 10 ' Btu /h

                    =each*:
                                                                                                                         -#                      - j-Those partsL of the TCW system LinL the; turbine ' buil_ ding Eare t
                    =1ocated tems.

in areas-that do not,contain'any-' safety-related sys - are All'= safety-relatedisystemsrin.the turbine building ,

                                                                                                                                                     ..l-located linVspecial areas:to: prevent!any damage;from:                                                                       j non-safety-related 'systensa during- seismic ' events s  Those -                                       .

parts: of the TCW system outside tho' turbine ' building c are'lo-cated away from any' safety-related systems.y i s f

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l 92J7 lineerhees " l 92.17J tasm.a. llent Slak Capahinty

                                                                                                                                             .e -                                                                               , ,
                               . The shimate beat slak shall be capable of-
                              . dissipating reactor decay beat and. essential-                               ,,                              >&J ecoling system beat loads after a normal reactor                                  '

0;, shutdown or a shutdown following an accident,j 4

                             . including LOCA.( The 'emosat-of heat to be c                                                                                                                                                                    X
                              . dissipated ander normal and accident conditions                              .
                                                                                                                                                                      ?;                                    '                                                             ',.

E is listed la Table 9.2 4.-- '

                                                                                                                                                                                                                                                    ?                            -'

1 The altimate heat slak and'aoy pumps, valves,, f- 4 , m - structures, or other composeats that remove heat  ; C. . 4 , i j 4 from safety systems shall be designed to Seismic l ,

                                                                                                                                                                                                                                                                                       -4 Category I and ASME Code Section Ill, Class 3,i Quality Assurance B, Quality Group C, IEEE.279,'

and IEEE 308 requirements. The~ safety related y" portions'aball be protected fross flooding : , spraylog, steam laplagement, pipe whip, jet" '

                                                                                                                                                                                                                                                                                     "a
                             - forces, missiles, fire and the effect of failure .                                                                                                                       .
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4 of any son. Seismic I equipment.1The safety ' related portions of these systems shall be' ( f' designed to meet the above mentioned design bases i , during a loss of offsite power.; Ti le safety: M related portions of these systems aball be'- ' ' i @ designed to perform their required cooling ,

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                             - functica assumlag a slagle active failure la any                                                  .        E'                                                                                                                                  '

mechanical or electrical system. JThe divisions $$ , of these systems shall be mechaalcally and. .

                                                                                                                                          'pa p                         4      .         -                                                                      " *.

electrically ' separated.1 " f 90/, l 9J.17J MakeupWater System capability

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g  ; p' q%." The raw water treatment and preparation of < ,

                                                                                                                                                                                                                                             '                    '               'd the domineralized water is sent to the makeup-s water system (purified) described in Subsection!

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Amendment 11 9.2-13. ll

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                                   . = - . . = . . - - . - - - . ~ - . - . -        - . - . . ,

b i O @ Revise first sentence to readt' The ultimate heat sink (subsection 9.2.5) shall.be-capable of dissipating rector decay heat and essential cooling systen heat loads after a normal reactor shutdown at four hours after a blowdown to the main condenser or a shutdown-following an accident, including locA.

            @ ingle failures of1 passive components in elect 8

will lead to the loss of-the affected pump, valve or other. components and the partial or complete loss of cooling capa-bility of that division.. However, all safety-related heat rejection systems are redundant so that.the essential cool-ing function.can be performed even'with the complete loss of one division. Safety-related portions of the UHS shall--be-located in seis-alc category I ' structures and protected against adverse en-vironmental conditions by being capable of. withstanding, without loss of safety function, the following events: (1) .the most severe natural phenomena,-appropriate with site conditions, but with no two or=nore phenomena. occurring. O simultaneously, (2) site-related events that historically have' occurred or might occur during the plant = lifetime,. (3) reasonably probable combinations of less severe natural phenomena and/or site-related events,"and (4) a single failure of man-made structural features. Each division shall be physically separated and protected ~ ' against flooding in. case of piping:fatlures asidiscussed in section 3.4. Cooling capability shall be provided for thirty days. There dition toare thenoRcW other heat loads associated with the'. uhs in ad-system. O The makeup water preparation system shall be located in a s building which does not contain.anyl safety-related struc =

              .tures, systems or components. .If the system.is not' avail-                        w able, dominera11:ed water can be obtained.from mobile equip-nent. _ The system shall be designed so that any failure,in the system, including any that cause flooding, shall not' result in the failure of anytsafety-related structure,>

system or component. 1 x

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o 0 9.2.17.3; Potable .and ~ Sanitary Water System The potable and sanitary water system shall be, designed with no interconnections with systems having:_the potential for containing radioactive materials. - Protection shall be. Provided through the use of air gaps, where necessary. (See Subsection 9.2.4.).

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MM 31pndard Plant 3M6)ttAH < n,.. . .I i

                                                                           ' SECTION 9.3 j

CONTENTS (Continued) -

   'O                            kilen                                         3 118                                                    . East' 9 3.10.2 -           System Ducription                                                                   93 13                                              l i

93.103 SafetyEvaluation 9 S 13.1 j 9 3.10.4 Tests and laspections - 95 0.1 93.10.$ instrvmentation Application - '93 13.1 j 9J.11 Kine Inleetlan Rwatem . 9S13.1 = f 9.3.11.1 ' Design Bases 9 S 13.1 - f 9.3.11.2 Safety Evaluation 9 } 13.1 I

       .                                                                                                                                                                                     i 93.113              Test and Inspections                                                              - 9.3 13.1 -                                          )
                                                                                                                                                                                      <i 9.3.11.4-           Instrumentailon                                                                      9513.1                                              ,

93,l1 Interfaces  ! q,3,h,I Radionc4ive Dmiw hohe Systawi (- {

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l 1 1 i l. 93 iiib i Amenement 12 1 i

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