ML20034H777

Forwards Draft SSAR Section 19H.5, COL License Info, to Support Accelerated Advanced BWR Review Schedule

ML20034H777
Person / Time
Site:	05200001
Issue date:	03/11/1993
From:	Fox J GENERAL ELECTRIC CO.
To:	Poslusny C Office of Nuclear Reactor Regulation
References
NUDOCS 9303220062
Download: ML20034H777 (7)

Text

_ _ _ _ _ _ _ _ - - - - - _ - - - _ _ _ _

e GE Nuclear Errergy se ra i;< w c:rmn

's cm wa san.csv cnsa March 11,1993 Docket No. STN 52-001 Chet Poslusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of the Nuclear Reactor Regulation

Subject:

Submittal Supporting Accelerated ABWR Review Schedule - Section 1911.5 COL License Information

]

Dear Chet:

Enclosed is a draft of SSAR Section 19H.5 COL License Information.

Please provide a copy of this transmittal to Glenn Kelly and Paul Amico.

Sincerely, kb Jack Fox Advanced Reactor Programs cc-Jack Duncan (GE)

Norman Fletcher (DOE)

JF13-56 9303220062 930311 PDR ADOCK 05200001 A

PDR

19H.S COL Liccnco Inforantion The COL applicant shall determine the HCLPF valuco for tha plant-specific /as-designed components corresponding to those generic components defined in Subsection 19H.4.3. The values should be compared to their assumed HCLPF values given in Table 19H.4-6. It should be noted that only the capacities of important contributors (see Section 19.8) need to be determined and compared.

The HCLPF calculations can be made using fragility analysis or the' conservative deterministic failure margin (CDFH) approach recommended in EPRI report NP-6041. The location effects should be taken into account in determining the limiting capacity of l

the same component on different locations.

For structures and components other than the generic components mentioned above, HCLPFs specified in Table 19H.4-6 can be considered achieved when the seismic design adequacy is confirmed if, according to the procedure described in Subsection 2.3.1.2 of Chapter 2, the site-dependent conditions l

are within the site envelope parameters or the site-specific SSE responses are bounded by those considered in the standardized design. Otherwi'se, site-specific HCLPF capacities for these structures and components need to be established.

It is not necessary that in each case the HCLPF exceed the value assumed in the margins analysis of the standardized design. However, depending on the degree of difference and the significance of the component in accident sequences, an evaluation of the site-specific plant level HCLPF capacity may be needed. The level of acceptable seismic margin for the plant should be established in a manner consistent with that used in existing nuclear power plants.

The site should also be investigated for the potential of seismic-induced soil failure (liquefaction, differential settlement, or slope stability) beyond the SSE level in accordance with the approach recommended in EPRI report NP-6041.

In order to increase confidence that the seismic capacities of as-designed structures and components are realiged,in the final l

constructed plant, a seismic walkdown as well as a review of construction drawings and documents shall be performed by the COL applicant. The walkdown procedure should follow the guidelines desertbed in EPRI report NP-6041, including an assessment of potential seismic vulnerabilities, such as marginal anchorage-of equipment and gross deviations from the design documents, and spatial interactions (e.g., operators being disabled due to the failure of the control room suspended ceiling in a seismic event).

W3 " U Ms Movts is swlfe,,,,,

~~

I 23A6100A5 REV A Standird Plant 19H.4 COMPONENT FRAGILITY Failure of the RPV support system would result in excessive RPV dcDection which could induce failure of the connecting pipes. The ultimate capacitrof the 19H.4.1 General AYPer 6 Sysb support system is provided by both the skirt and the

(

stabilizers. In this analysis, the redennce capacity of thchd'!::::. is conservatively'::;;'

? 2 &-

Seismic fragilities of safety.related components

-- ' SW! u ko ic soyyvd din failr were assessed for the following two categories of if:: by tk L iam or anc!torWr.gc. -

&%w The critical failure mode is found to be [ameher yafdw components:

g,o (1) ABWR specific components whose fragility

'-_Sc. Its median ground acceleration capacity evaluation is made according to existing design is _. with a logarithmic standard deviation o? O.33.

The individual factors contributing to the median information.

1 capacity are shown in Table 19H.4-1.

(2) Generic components whose fragilitics are based on the data recommended in Reference 2 or RPV internal Comoonents other data sources as appropriate.

The internal components examined for seismic fragilities include the shroud support, CRD guide 19H.4.2 ABWR Specific Cornponents tubes, CRD housings, and fuel assemblies. Failure of those components could potentially result in inability to insert the control rods to shut down the Detailed seismic fragilby evaluations are reactor.

performed for the following ABWR specific Tables 19H.4-2 through 19H.4-5 show the failure components:

modes and associated median ground acceleration Reactor pressure vessel (RPV) capacities of those components. The contributing factors are also shown in these tables. M= i:.1:

Shroud support Control rod drive (CRD) guide tubes

.i.e.,;c ddg; g:n d :x6.ud:: h 0.15;; OBE

i. x : k O Z k,;da a m J.... L. -. A J

CRD housings Fuel assemblies

': ~ r m f 9 '.

,,iwse v6 h The design seismic loads for these components As noted, th nel assemblics are found to have were calculated directly using a coupled building the lowest sei ic capacity among the RPV internal structures and RPVfmternals model Consequently, components..L '.:: n :f: : Mng oiinc no subsystem dynamic analyses usinginput motions

1. J i'. n :! TE:

.qud::,; :d::-..vuud at support points were required. Therefore, the m i...a :q :::y.. ;. F ;&. L...:hr?: l I

fragility evaluation procedures used for the reactor asandeed h ' -

O building structures as presented previously are also applicabic to these specific components.

19H.4J Generic C$np6hents Reaeter Prmura V-1(RPV)

The failure of the RPV due to an earthquake Detailed fragility evaluations for safety-related results in a sequence similar to a large break components other than those specific components loss-of-coolant accident, with the exception that presented above cannot be made at this stage of there may be no means to provide makeup (i.e.,

certification due to lack of design details.

injection or cooling) to the core. The ABWR RPV The ABWR generic components of interest for is supported by a conical skirt which is anchored to the pedestal with 120 21/2' diameter high-strength this seismic risk analysis are the following-anchor bolts. At an upper elevation, the RPV is laterally restrained by stabiliacrs which are Cable trays connected to the reactor shield wall.

Large flat-bottom storage tanks 19H 4-t Amendment 20 i

T x !

J wl f k

5 MN 2sAstoors 5

i LStandard Phnt uy,

g [cj g f +7 t

epresentative response amplification factor of 5f

-A a '_;eu l-Air-operated valves amping is about 3.8 at relatively high elevatio s.

a $ $ h Heat cdangers he corresponding lower bound median gro ad

{ f ( c; i

Off-site Power (transformers and ceramic a celeration capacityis thus 4.9g. A smaller ca acity j 'g ( g p; insulators) was conservatively used as the seismic fr gility o

Batteries and battery racks for ble trays. The logarithmic standard dei tion is y } g[

Battery chargers / Inverters esti ated to be 0.6.

Electric equipment (chatter failure mode)_

I,f ; j,,;'

t j!

-.J L;i/h. - -.- -

Fo large flat-bottom storage tanks su asdietel y

a.

d J.

Switchgear/ Motor control centers oil sto age tanks, condensate storage t and fire Transformers (not off site transformers) water s rage tanks that are located in e yard and g'j4 Diesel generators and support systems founde on grade, the generic fragil' y is estimated

, d r (%

Turbine-driven pumps based on revious PRAs data comp' d in Reference

• 4 N T a.

Motor-driven pumps 4ml-dwen fants

4. A tot I of 23 data points in his component

'.'>j 1j Small tanks (e.g., standby liquid control tank) category were examined. For ach of the data i *i Motor-operated valves points, a cahacity margin facsor as obtamed as the

%.fety relief, manual, and check valves ratio of the agility level to th SSE peak ground

{ f { 7{

L 0

Hydraulic control units acceleration. The resulting acity margin factors 2

Heating, ventilation, and air conditioning are found to r ge from 1.85 o 10.47, with median e

gII?v ducting value at about

7. Since the ABWR tanks are to be 45 Air handling units /rwm a4r cen4her.s designed followtgg the fle le wall approach, the w

4

)p Piping potential of un er-desi n using the rigid wall approach will not xist, median capacity margin 3.h b..-L' di' AJ m#"

J higher than the edi of the data base is thus 3

e considered achievab, d is estimated to be 7. The Eqcept for the first five con ponents listed were corresponding fragil is 2.lg, which is the product ft agibtv values recommended in Reference adopted the ABWR seismic risk an sis. The of the median capaci margin factor and 0.3g peak generic frag ' ties in Reference 2 wer en based ground accelerati n f the design SSE for the on a review of

'or PRAs and o r fragility data.

ABWR standard p nt.

e associated logarithmic These median cap ' ties are idered achievable standtrd deviatio is ated to be 0.45. The most for the ALWRs with tionaryimprovement in probable failure ode is chorage failure.

the seismic capacitie nponents designed to 0.3g SSE. Detalle escriptio of these component For accum ors, Refe. ec 2 ecommends a 2g categories ar rovided in Anne of Reference 2.

median pea ground acce! ration capacity. The The con onent fragilities and un ainties are recommen d 2g capacity is nservativ: since it is sumiarized in Table 19H.4-6 The st five less than e median value of bout 3.5g of the data components oa the list are discussed below.

base fro which the 2g capaci was estimated. A lesser uservative but more r alistic capacity at e cable tray fragility is taken to be astead 2.5g m peak ground accelera\\ ion, which is still of 2g s recomanended in Reference higher less t an the data bastmeMan s e,is considered capacit ' basedenrecentinform n presented in ach* vible for the ABWR accu ulators. The elated logarithmic standar' deviation is Referen regydag agihty data for three as typical typ of sup or electrical conduits.

ated to be 0.45. The predom' failure mode Since kilure cab sys is typically assumed to allure of supporta.

occur due to sup failure, and the support design is similar toJirst d for supporting conduits, the The seismic fragility for generie air operated conduit sppport fr ility is considered equally valves was estimated based on the achievedtest level applicabfe to cable tra The median fragilities of of main steam isolation valves qualified pr3viously the,tfirce support types ported in Reference 3 for other BWR plants. This achieved test level was jinge approximately from

.5g to 23g in terms of increased by a modest 25% to postulate failure

/ spectral acceleratior Acco. ing to Anner A of (typically stem binding or failure of the att line) in Reference 2, the fundan ental equency of cab! -

terms of spectral acceleration capacity. Considering trays is typically in the 5 to 10 range and amplification through building structures and piping, NN Amendment 5

• 3A4100AS Standard Plant Table 19H.4-6 SEISMIC umu. y '

SUMMARY

Fra n CMAct r

.Ti 1

Capacity (*I Combined NCRF Structure /Comoonent Fallure Mode Am (r)

Uncertainty _ ($)

Reactor Building Wdi Shear 3.2. -h8-0.45

't. f A l Containment Shear J.: 4 0.44

,ig RPV Pedestal Flexural s,0 -%9 0,44 g,go Cegejt,, ^d.;...;.e.y buildingf".

g ig

{ 4 34 f,y o, r, o c....sv.,.-........,.

Reactor pressure vessel Wrrcrt %; _.:...- LA.

$,o -5 9 033

1. g i Shroud support Buckling 1.o -M-036

c. S '?

CRD guide tubes Buckling

1. E -t?-

o,34 0,46-c.18 Plastic vielding 3.5 -3&

0.45 8.2.0.

CRD housing g' Channelbsoleling 1.+ t:9 035 c. s 2. !

Fuel Assemblies

/.60 c.7 + [

Cable trays Support

• ' 3.0 g['

0 l_arge flat-bottom storage tanks Anchorage

/ 2.1

/0.45 0 77 '

' u....:=u

,,yy B

G.45 Air-operated valves Stem l' Gng/ Air line 84 35 0.60 d.7

• Heat Exchanger Anchorage v 2.0 0.45 c.70 Off-site power Ceramicinsulators v 03 0.55 o,as Batteries and battery racks Anchorage /LOF S94&-

0.45 G t.:3 i Battery chargers /Imerters IOF

%.14&-

0.4f6 a.7 7 l Electric equipment (chatter) i function req'd during event Relay chattering W.A. 4:9-V. A GJO u.4 !

function req'd after ewst Relay chattering 2.* G:D-0.50 e.63 [

y

3 3 y,~

=_' (*) 1.KG6-v... w.9 n-.. y -...q 79 0.4f6

c. 4L' Funaional/Struaural l

Switchgear/ Motor control centers Transformers Ft,actional/Struaural 1.T 4:5-0.45 C.

o. GL Diesel generators & support systems Support hf 3:5-0.4M

o. & t-Turbine-driven pumps Anchorage

/2.0 0.45 g,7 a ;

d Motordriven pumps Anchorage / Impeller delle 3He1.6 0.% O M a.s2 i Small tanks Anchorage I.846-0.4f6 CL Motor-operated valves Operstar distortion

'3.0 0.60 a.W an check valves Internal damage G

0.60 e.V i Safety relief.

Pe; 'dL.: c. _ M'"'"

i^F 7WM W M'La

N C.fT '

M'VAC duaing Support 304&

0.60 0 7t.

Air handling units /R..w A.C.

Blade rubbing v'40,,

0.50 e 4'3 :

"5 c4t# Jr4% Po 8"E@ C*

t O.4 f.

~

Y 3,pg C. f.

\\

clrauj *t EM m s+

Lo p s.s

(

Amendment 11

~

M 23AstooAs m

Standard Plant Table 19H.44 SEISMIC FRAGILITY

SUMMARY

(Cont.)

Notes:

Capacities are in terms of median peak ground acceleration.

1.

b. Combined uncertainties are composite logarithmic standard deviations of uncertainty and :andomness componcsts.

The potential for relay chatter was treated in the fo!!owing manner. Only the scram safety function is c

required during a seismic event. This function is fail-safe, so relay chatter would cause a safe state failure (scram) even if relays were employed. For the ABWR, the scram actuating devices are solid state power switches with no failure mode similar to relay chatter. The scram function is supplemented by an alternate scram method (energizing the air header dump valves) to provide diversity. This method uses relay actuation, but no credit was taken for this capability in the scismic analysis. Therefore, there is no potential for relay chatter to prevent safety actions durir.g a seismic event.

Switchgear and motor control centers do include relays whose failure could prevrx. safety actions after the seismic event. It was assumed that the indicated capacity of this equipment (2.5) was more representative than the specific relaf chatter value (2.0) since switchgear and motor control centers are normally qualif with the auxiliary relays in place. Also, the type of auxiliary relays used tend to be the most rugged of relay types and would have a capacity above 2.0. The multiplexer output devicca for ECCS and RHR ope have been assumed to be solid state devices (rather than relays), so the relay chatter failure mode does not I

3PP Y-i.

T L

=

se 19H 4-10 Amendmem 11 c