Text

Forwards Westinghouse Responses to NRC Requests for Addl Info on AP600,including Listing of NRC RAIs
	ML20072N066
Person / Time
Site:	05200003
Issue date:	08/26/1994
From:	Liparulo N; WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	NTD-NRC-94-4276, NUDOCS 9409020308
	Download: ML20072N066 (120)
	v • d • e

r; Westinghouse Energy Systems Ba 355 "

PiSb"'Eh "*""SY'"'a 15230 0355 l Electric Corporation NTD-NRC-94-4276 !

DCP/NRC0195 Docket No.: STN-52-003 !

August 26,1994 Document Control Desk i U.S. Nuclear Regulatory Commission {

Washington, D.C. 20555 ;

A'ITENTION: R.W.BORCHARDT

SUBJECT:

WESTINGilOUSE RESPONSES TO NRC REQUESTS FOR ADDITIONAL l INFORMATION ON HIE AP600 l

Dear Mr. Borchardt:

Enclosed are three copics of the Westinghouse responses to NRC requests for additional information j on the AP600. In addition, revisions of responses previously submitted are provioed.

. i A listing of the NRC requests for additional information responded to in this letter is contained in !

Attachment A. l These responses are also provided as electronic files in Wordperfect 5.1 format with Mr. Kenyon's y copy.

If you have any questions on this material, please contact Mr. Brian A. McIntyre at 412-374-4334.

Nicholas J. Liparuto, Manager !

Nuclear Safety Regulatory And Licensing Activities

/nja i Enclosure i cc: B. A. McIntyre - Westinghouse !

T. 'Kenyon - NRR i

310002 l

- k'k j 9409020308 940826 PDR h ADDCK 05200003 PDR

\

. NTD-NRC-94-4276 ATTACHMENT A AP6(W) RAI RESPONSES ,

SUBMITTED AUGUST 26,1994 220.33 R1 230.37 RI 230.4 i R 1 230.42 R1 230.79 R1 231.29 RI 410.145 >

410.160 410.162 410.196 410.25?

440.61 440.205 440.214 440.218 440.220 440.232 460.5 R1 460.7 R1 460.11 R1 460.15 RI 920.5 952.91 5

J

NRC REQUEST FOR ADDITIONAL INFORMATION E: 9m Response Revision 1

~

E Question 220 33 NUREGCR A 414 reponed that. dunng sescre accident condinons. no leakage wa.s detected from any of the three t uneni electne:d penetration awembhes tEPAs>. under the following condinons t 1) D. G. O'Brien EPA. 361 F.155 psia for 10 das s. (2 p Westmphouse EPA. 41Hr F. 75 psia for 10 days. and (3) Conax EPA. 7tHrF.135 psia for in d.ip. Howeser. the SS AR does not addrew w hat EPAs will be used for the APN10 Provide a conunionent in the SS AR that EPS penetraung containment he at least as strong as the steel containment vessel (Section 3.S.2 of the SS AR).

Response-The eletuical penetration assemblies are desenhed in SSAR Subsection L8.2.1.6 and are depicted in sheets x and 9 of Figme 18.2-4. Their pertonnance under severe accident conditions is desenhed in SS AR Subsection 3X2.4.2.5.

The elettrical penetration assemblies are procured as equipment and the details are dependent on the suppher. The assemblies w dl he quahtied for the cont unment design basis event conditions as desenhed in SS AR Appendix 3D.

The assembhes uill be piocured to be sonilar to one of those tested by Sandia as reponed in NUREG/CR-5334 and will hase ultunate capatities consistent with those demonstrated in the Sandia tests. The ultimate capacity of the EPAs is prunanly detennined by the temperature. The maximum temperature of the containment vessel below the

~

operating deck during a sesere accident is reported in Appendix L of the PRA Report as 315?F. This is signiticantly below the capabihty of the assemblies tested f rom the three suppliers.

SSAR Revision:

Reuse the last paragraph of Subsection 3 1 2.4.2.5 as follows:

Electneal penetrations have a pressure boundary consisting of the sleeve and an end plate containing a series of modules. The prewure capacity of these clernents is large. Tests at pressures and temperatures representative of severe accident conditions are described in NUREG/CR-5334 (Reference 5). w here the Westinghouse penetrations were irradiated, aged then tested to 75 psia at 400"F. Other electrical penetrauon assemblies were tested to higher pressures and temperatures. The penetration assemblies for the AP600 are similar to one of those tested by Sandia as reported in NiiREG/CR-5334 These tests showed that the electrical penetration assemblies would withstand ses ere actiJent conditions.

i I

W Westinghouse 1

l l

. Y

i NRC REQUEST FOR ADDITIONAL INFORMATION WE9 l Response Revision 1 Question 23123 in the Nmeinher 3n.1992 response to Q231.h regarding the lateral carth prewure loads. Westinghouse states that the seismic Catecory I ret.uning strodures and below grade exterior walls are designed for the worst case enselopmp the lateral earth pressure, and that the SS AR will be suitably resised. Westinghouse's response dacs not clearly address the fact that the lateral earth pressures along the walls of the NI are a f unction of the latend extent and charat ter of the bas klill soils. Itased on the above,

a. Speedy. in the SS AR, acceptable ranges of hatklill properties (such as compacted soil density, minunum acceptable decree of cornpacuon, range of si/cs. ete.) f or backtill soils to ensure that the design is adequate.

and

h. .lusuly the use of the Mononobe-Okabe (MO) method for calculating the lateral soil loads on walls of the NI where wall movements relative to the surrounding sod may not develop failure strains in the sod.

Response.

a. The design of the nuclear island is not influenced by backfill properties. Backfill material wdl not be used against the exterior walls of the nuclear island structures. The excavation will have a vertical f ace as descnbed in the following revision to the SSAR. I

b. Please see the response to RAI 220.41 for a discussion of the method for calculating the lateral soil ;

loads '

SSAR Revision:

Add the lollowing Subsection to the SS AR:

1 2.5.1 Excavation and Backfill Excavation in soil for the nuclear island structure.s below grade will use a soil nailing method. Soil nailing is i a method of retaining earth in situ. As the nuclear island excavauon progresses vertically downwani. holes are ;

drilled horizontally into the adjoining undisturbed soil. a metal rod is inserted into the hole, and grout is pumped into each h<de to fill the hole and to anchor the " nail" rod.

As approximately each five feet depth of the nuclear island excavation is completed nominal eight to ten inch diameter holes are drilled horizontally through the venical tace of the excavation into adjacent undisturbed soil.

These " nail" hole.s. spaced horiz.ontidly and vertically on live to six feet centers. are drilled slightly downward at tilleen degrees to the horizontal. A " nail", nonnally a one inch diameter metal bar/ rod. is center located f or the full length of the hole. The nothinal length of soil nails are N)G to 709 of the wall height, depending upon soil conditions. The hole is filled with grout to anchor the rod to the soil. A metal face plate is inst:dled on the exposed end of the nid at the excavated wall vertical surlace. Wehled wire mesh is hung on the wall surface for wall l

W- WestinEhouse

I NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 remforceinent and secured to the soil nail f ace plates for anchorage. A 4J)to Iwi to 5.lHU psi nem-exp.m3ive pea pra'.el shottrete mix is blow n onto the wire mesh to form a nominal four to six inch thick . oil retaining wall.

Installation of the sod ret.uning wall closely follows the progress of the excavation and is f rom the top down, with cash wue mesh reinf orced, shotcreted wall section being supported by the soil " nails and the preceding elevanons of sod nailed wall placements.

Soil naihng as a method of soil retention has been successfully used on excavations up to 55' deep on projects m the US. Soils have been retained for up to W in Europe. The statn .1 Cahfornia CALTRANS uses soil n;uling extensively for excavations and soil retention installations.

The specific soil nailing system is based on actual vul condrions, site conditions and apphed con'truction surt harge loads. The design of the external walls belcw grade does not take any credit for the soil nails. Since the exterior walls below grade are designed for a range of soil conditions. including hard rock and soit rot k. soil nailing has no elfcct on the overall results of the SSI analysis.

The soit nading method produces a venical surf ace down to the bottom of the excavation and is used as the outside forms f or the extenor walls below grade of the nuclear island. Concrete is placed directly against the venical contrcle surf ace of the enavation.

For excavation in rm k, lour to six inches of shotcrete are blow n on to the rock surface. The concrete for the exterior walls is placed against the shotcrete. The shotereie contains a crystalline waterproofing matenal as desenbed in Subsecnon 3.4.1.1.1.

Reuse Subsection 3.4.1.1.1 as show n below:

3.4ci.1.1 Protection frorn External Flooding The probable maximum flood for the APNN) has been established at less than the finished grade as discuwed

[ reuousiv in Section 2.4. The probable maximum flood results f rom site specific events. sut h as nver floothng, upstream d.un f ailure. or other natural causes.

Ilooding does not octor trolil the probable maximum precipitation. Water trom rool drains and/or scuppers.

as well as runoti f rom the plant site and ad.lacent areas. is conveyed to catch basins, underground pipes. or directly to open ditt hes by slopmg the tributarv surf ace area. The site is graded to of fer protection to the seismic Categorv I structures.

The high ground water table interf ace is at two teet below the grade elevation. as discussed previously in Settion 2.4 TheM .e me Catevne y 14nk im*w hu h mte-k.cated 4+b +w-yra+1s4evat i+ 9sueq uotec ted *yaiH44 b w idmpby watert 44

diHe u+embranr* *#*l waiw

tot % Wale ^rtw +*4 mg -Heembrane n4reiH~salletl4m h+ w ir4 mtal *#*1+eruts+1-e41erior

~mhe*4.ek+w-ye a.le.-Watendop+*#e4n~talled 4n-euen+ w astrut4 inh f..mt+ below-gradts Et*f l+ w maHt t* t Htt$4 6a f or 4ht' w alt5 rpl+* d iHg- H604Hbf'aHe% ithd %rtit*l*It 6p* Hf t'4%t%titi +4F tht' ltMI+ +w iHet 4 6H%4 del alit utM

-4Hieras ti,m+itl*4het4mt rete thsouyhout-4heIdesm+e44-4hsplant n_Alyh[y to.-W64h*4rtH+1 the4HasHitulu-hVdo**.talhe pie **me b

- hnist* 4 4-4H4rtllal6+ me httV4Hf miHIllHHH 4 Hit *f leleHti t'tltH IHy-4 het4 ms{Hh 4 Ion i+peittle m%

231.23(RI)-2 W WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION g.. . Eh T I Response Revision 1 '

.-W 4 4,ow4awit.*u+y~++N tsu

. -We,+tha in++ ti+ .n

. --- Im&p4meahdity-

.- &bpd*h4y el w+tle4anding-4mWWuen4++n*ler+i .mWwulitkwe,

.-Inteytity-when++bbied to-planteliat**w The scismic Category I structures below grade are protected against tholing by waterstops and a waterpnioling systern. The waterprooling system is provided by the intnsluction of a cementitious crystalline waterprooting addiuve to the nailed soil retention wall shoicrete or to the shotcrete applied to the nd surf ace as described in Subsection 23.1. For the hori/ontal surf ace under the basemat, the cernentitious crystalline waterpnioling additive is added to the mud mat. The waterpnioting additive is a unique chemical treatment added to the concrete at the time of batching and consists of portland cement, very line silica sand, and vanous active proprietary chemicals. The active chemicals react with the moisture in tresh concrete, and the byproducts of cement hydration cause a catalytic reaction generatmp a non-soluble crywdline formation of dendritic fibers throughout the pores and capillary tracts of the concrete. The concrete is thus sealed against penetration of water or liquid.

i i

1 w wesungnouse 23i.23mn-3

NRC REQUEST FOR ADDITIONAL INFORMATION liii: Nii m :L Response Revision 1 Question 230.37 Section 3.7.2.2 of the SSAR describes the imponance el the mass participation from the high frequency stmetural nuxles of the stick model in the horizontal and. particularly, the vertical duettions due to the rigidity of the containment internal structures. It is not clear how the contributions of the predominant high frequency modes to the structural responses were taken into account in the analyses. Particularly, provide the following intonnation:

a. the cutof f f requencies used in the horizontal and vertical time-history analyses of the fixed base model (the case of structures f ounded on rock site).

b. the cutoit frequencies used in the horizontal and vertie:d SSI analyses using the complex trequency response analysis method, and provide the basis for the cutof f frequencies selected.

c. details of the separate seismic analysis using the coupled containment internal structures and reactor coolant hiop iRCL) lumped-mass model (Page 3.7 5, first paragraph) and the dif ference in the response results between this separate seismic analysis and the original seismic analysis. Was this " separate analysis" done using the fixed base model for the rock site condition!

d. details for considering the high frequency ef fect to thn vertical responses (fortes and moments) of the containment internal structures (Page 3.7-5. first paragmph). Was this consideration applied only to the vertical seismic analysis of the fixed base stmetural nmdel for the rock site condition?

Response: (Revision 1)

a. The cutotf frequency used in the horizontal and venical time-history analyses of the fixed-base model for the hard rott site is 34.0 hertz.

b. The cutoit frequencies 4wed4n for the SSI analyses using the complex frequency response analysis method are 33 hert/ in the horizont2d and vertical directiom ion the sof t mck site, and 15 and 21 hertz in the horizontal and vertical directions, respectively, for the sof t-to-medium stiff soil site 4aalm4torimnial-andwenied-dira4h*N ve-pn livt4y.

Th+,e-uitol14 reipwwiewre -sela4ed-basedwwhe4441owing

--The4Lhes t A-+ut ol f4 reipwney-useJ4+H be4SI-analy se+4or-t he+ol t 4+4-SiteJ+in 4*4wdame-willu he sequisement44-Regulatory-Guide-440.-The-4L4mnsuni-214mnsmi 4f-4)equenue**weddnale4Si analyw4-1 he*4 t 4*anedimn+ti ff* 44 +iteau w w4*4eil4w~ed 4 wa he4naj+ ++ony

mte4+a tural4)*pwo+es of-the-swy4ed*41+tme4me+ystem:

- Thevaltulated-act+{erati*nuime 14storie+for4he*4tatunedium+tilf+oJ+iteare4+ot4ntended to4e "*tand ahuw":-42e*[wweshweleration+ 4)4 Hn4 heha rd4wkr4he* 4 Habmd4lp 4 Ho4Hedium+t il4441-silenare envelope +14or41esign-puqwm 230.37(R1)-1 W WestinEhouse l

l

NRC REQUEST FOR ADDITIONAL INFORMATION y ..

Response Revision 1

-46 6munsnemberamm.,+n.i mbb lisp! wnwnt wre4*ninatwily41* wean.*Lb ire q+w4* ws4ow.+r-d win a

deut +464reituwwhTheretome ton a *44m+4reaquwwymnplaiyhe-4han-4he+utof f4requentwdl h,,en y aninor-elle+ 1w4 hts nasmunauemlw4etw,$nd a**lakhpbanwnis:

These cutoll frequencies for the SSI analyses are selected based on the following reasons:

(1) The 33 hert/ cutoll frequency used in the soit rock case follows the requirement in NRC Regulatory Guide iNL (2) The 15 herti and 21 hertz cutoit frequencies used in the soft-hsmedium stiff soil case are selected based on the major composite natural frequencies of the soil-structures system. The calculated acceleration responses for the soft-hsmedium stiff soil case are not intended to be "st;md alone" in the design of AP600.

Response accelerations from the hard rock, soit rock, and the soft-to-medium stift soil cases are enveloped for design purposes. In the frequency range between the cutoff frequency and 33 hertz, acceleration responses from the soft-to-medium stiff soil case are not controlling.

(3) The maximum member forces and nodal displacements for the soft rock and the soit-to-medium stiff soil cases are c:dculated by converting responses in the frequency domain into responses in the time domain.

Therefore, these responses am dominated by contributions from the low frequency ranges. Furthermom, the complex frequency response analysis method used has considered the entire model mass. Therefore, the " missing mass" adjustments necessitated in mode superposition analyses technique are not applicable.

It is concluded that the SSI responses from the frequency range higher than the cutoll frequencies have negligible effects on the calculated maximum member forces and nodal displacements within each of the soft rock and the soft-to-medium stift soil caws.

c. The analysis referenced in the question is an earlier analysis presented in Revision 0 of the SS AR. The seismic analysis has been revised as presented in Revision I of the SSAR Section 3.7.2.2.

Refer to Revision I of SSAR Section 3.7.2.2. Member forces for the hud rock site are calculated using the fixed base combined nuclear island stick model. The mode superposition time history analysis method is used to calculate the seismic response member forces for the coupled shield and auxiliary buildings and for the steel containment vessel. The response spectrum analysis (RS A) technique is used to calculate the response member forces. of both horizont;d and vertical excitations, for the containment internal structures.

For comparison purposes, seismic member forces f or the containment internal structures are also calculated using the nuxle superposition time history analysis method and compared with the member forces from the RSA methtxl. Biis comparison shows that:

The venical forces determined by the mode superposition time history analysis are approximately 10% to 301 of those calculated by the RSA, and s

+

The horizontal forces determined by the mode superposition time history analysis are approximately 10%

to 301 less than those calculated by the RS A.

230.37(RI)-2 W WestinEhouse

r i

NRC REQUEST FOR ADDITIONAL INFORMATION i-- n Response Revision 1

d. Detd!s fue.idning-the-high-4'requency effeet-to-the-wrtiea! resin - ( fore,,**l-+m*nent4-+4-tiw umtainment4ntanal-4nsures are describ! i : fe+ahme. As descrilwl.-thiwwsleration-e-apg4+ed-to-hoth i dm4wimata! n! edieal-+.eiwn . =di of 'he hed her ~ue:ur:d ode! fm the . mtainment--intemal '

MnMures For the hard rock case, member forces at the coupled auxiliary and shield buildings are calculated using the mode superposition time history technique including all modes up to 34 henz. Because the mass panicipation ,

of the coupled auxiliary and shield buildings is less than 90E the member forces from the time history analysis are verified by comparing with equivalent member forces computed using the response spectrum analysis method with the AP600 design response spectra. The response spectrum analysis is performed using the double sum modal combination and the SRSS co-directional combination method, in conjunction with a 34 hertz frequency cutoff and considemtion of "all missing mass". This comparison shows that the two analytical methods produce ;

I compatible force responses.

In addition, the combined fixed-base nuclear island model is reanalyzed using a 64 hertz cutoff frequency to include a minimum 90% mass participation of the coupled auxiliary and shield buildings. The procedures and models used in this mode superposition time history analysis are the same as those used to establish the design ,

member forces, except the cutoff frequency is 64 henz instead of 34 hertz, and the solution time step used is 0.0025 second instead of 0.005 second. For the coupled auxiliary and shield buildings model, the cumulative participating masses are 99E 99E and 94% for the NS, EW, and Vertical directions, respectively. The member ,

forces from this analysis (Case 2 responses) are compared with design member forces (Case 1 responses) and are shown in Table 230.37-1. i From the comparison in Table 230.37-1, responses are shown to be closely matched except for vertical (axial) !

forces at the two bottom elevations where results including higher mass panicipation are larger. This mismatch is due to the high venical natural frequencies at the lower ponion of the structure near the fixed boundary ,

condition; as a result, a portion of the nodal mass did not participate in the modal time history analysis with 34 !

henz cutoff frequency. . However, these~ venical fon:es are conservative since they are smaller than those calculaterl using the response spectrum analysis method as described above.

In the fixed-base combined nuclear island lumped-mass stick model used for the hard rock site, the steel containment vessel stick and the containment intern:d structures stick are connected to the coupled auxiliary and shield buildings stick through " rigid" beam elements (representing concrete slabs) at elevations 100' and 82.5' respectively. Since these slabs are connected to the exterior walls and all exterior walls at and below grade (elevation 100').are supported by the hard rock site through fixities in boundary conditions, stick-to-stick interaction for the hard rock case is judged to be negligible. Furthermore, the coupled stick model was used in the fixed-base analyses where the various stick models are connected as described above. Any ' stick-to-stick interacticn effect is captured automatically in the responses. Specifically, they are captured in the resuhs from

the mode superpositior. time history analysis with 64 henz cutoff fmquency and from the above response spectrum analysis which considered all " missing mass".

SSAR Revision: NONE I

230.37(RI)-3 i W "

Westinghouse l

i 9

l l

NRC REQUEST FOR ADDITIONAL INFORMATION I l

.n . . ...- )

ire 9#31 Response Revision 1 Table 230.37-1 Comparison of Seismic Forces and Moments Coupled Auxiliary and Shield Buildings Hard Rock Site Condition Elev. Axial Force N-S Shear E-W Shear Torque N-S Moment E-W Moment

01) (x10' kips) (xitf kips) (x10' Lips) (x10' k ft) (x10' k-ft) (x10' k-f t)

Case I Case 2 Case 1 Case 2 Case 1 Case 2 Case i Case 2 Case 1 Case 2 Case 1 Case 2 307 25 25 32 31 1.67 1.69 2.65 1.69 2.49 2.48 5 5 297 73 73 88 85 3.69 3.73 5.47 5.42 5.07 5.06 11 II 284 190 190 223 220 7.09 7.19 9.33 9.25 8.47 8.48 27 27 272 391 391 443 443 11.17 11.35 9.33 9.25 8.47 8.48 52 52 246 631 631 659 656 13.18 12.98 14.45 14.47 14.64 14.76 53 54 241 697 697 722 719 14.29 13.84 15.54 15.45 15.83 15.91 54 54 230 843 843 892 889 16.46 15.49 17.58 17.62 17.85 17.90 68 66 210 1135 1134 1234 1233 19.52 17.53 20.37 20.67 20.14 20.17 93 92 180 1807 1800 1893 1884 22.17 19,58 24.03 24.07 22.66 22.97 785 790 161 2341 2313 2316 2305 23.62 21.49 26.41 26.08 25.42 25.78 953 954 153 2513 2484 2428 2466 25.26 24.43 29.41 28.M 29.63 29.78 695 700 135 2981 2942 2940 2949 27.92 30.41 33.81 33.19 36.41 36.24 958 954 118 3535 3497 3422 3429 29.94 36.24 36.62 37.90 41.13 41.55 1173 1189 100 4200 4359 3937 4154 Note:--~The seismic force ,and moment responses are determined using the male superposition time history technique with 3 components of earthquake input simrlaneously. Case 1 responses are computed using at = 0.005 second and cutoff irequency = 34 hz. Case 2 responses are computed using At = 0.0025 second and cutoll frequency = 64 bz.

230.37(RI)-4 3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

~

=1 Response Revision 1 9 Question 230.41

a. Section 3.7.2.3.3 of the SSAR states that for soil-structure interaction (SSI) analyses, the nuclear island basemat and the periphery walls of the embedded portion of the nuclear island are represented by a three-dimensional finite element model. When the basemat was modeled, has the flexibility of the basemat been considered in the SSI analyses?

b. Evaluate the possibility of the out-of-phase interaction between the shield building, steel containment vessel and containment air battle (Section 3.7.2.3.3 of the SSAR).

Response: (Revision 1)

a. In the soil-structure interaction analysis, the entire nuclear island is represented by a stick model except for the basemat and the embedded ponions of the exterior walls which are modelled with 3D solid elements and shell elements,respectively. This model of the embedded portion models the boundary of the nuclear island, but not the tiexibility of the basemat. However, considering that the basemat thickness is 6 feet and the interior walls are closely spaced, any local flexibility of the basemat in the vertical direction is negligible.

At the 3 slab elevations (grade at 100', 82.5' and basemat at 66.5'), horizontal ngid beam elements are modelled along the exterior wall (shell elements) to simulate the stiffer.ing effect provided by the slab to the wall. At these same elevations, horizontal rigid beams are also used to connect the shell elements with the stick model. The kication of the rigid beam elements at elevations 82.5' and 100' are shown in Figures 230.41-1 and 230.41-2, respectively.

At the basemat (elevation 665), horizont d rigid beams are used:

t 1) at the exterior wall to simulate slab rigidity and to connect the stick model with the exterior wall (as stated above),and (2) to simulate the stiffening effects provided by the internal walls to the basemat.

The kication of the rigid beam elements at elevation 66.5' are shown in Figure 230.41-3.

b. The design conHguration of the steel containment vessel. the containment air haf tle and the shield building is shown in Figures 1.2-12 and 1.2-13.

The steel containment vessel, presented in Section 3.8.2.1.2, is designed as an independent. free-standing structure. The bottom head is embedded in concrete, with concrete up to elevation 100 feet on the outside and elevation 108 feet on the inside. Above elevation 100 feet, seismic gaps are provided between the steel containment vessel and @e shield building.

W Westinghouse u

l l

NRC REQUEST FOR ADDITIONAL INFORMATION siti :nE j

= Response Revision 1 !

1

. k The containment air b;dfle, presented in Section 3.8.4.1.3, is supported from the surface of the steel containment vessel at regular intervals. It will displace together with the containment vessel during a seismic event. A tlexible connection is provided between the air hatile and the shield building roof structure. This nexible connection is designed to accouunodate the ditferential displacement, determined using the absolute sum method, between the containment vessel and the shield building. Therefore, scismic interaction between the shield :

building and the containment vessel through the air baffle is negligible.

The maximum seismic displacements relative to top of hasemat for the shield building and the steel containment vessel are given in tables 3.7.2-8 and 3.7.2-9 respectively. The maximum horizontal seismic displacements relative to the top of basemat, at the top of the containment vessel and the top of the shield building are 0.95 inches and OA2 inches, respectively. The maximum relative displacements between these structures are negligible in comparison with the design gap provided. see Figures 1.2-12 and 1.2-13. There is no out-of-phase interaction between the shield building and the containment vessel / air baffle.

Structure to structure interaction between the steel containment vessel and the shield building through the common foundation during a seismic event is considered, because a coupled model connecting the nuclear island structures to the same foundation is used in the seismic analyses.

SSAR Revision: NONE 230.41(RI)-2 W- WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION

.==

Response Revision 1

~

Rx=

Figure 230.41-1 Foundation of the Seismic Analysis Model for The Nuclear Island, Elevation 82'-6" is _ 7 8

13 94 20 it 21 2t ts N

11 7 8 9 _ 10 6 28 6 "E N # EN 8 27 A

te 3

, n 2 30 t 31 45 45 44 43 4Z 41 W M JB JT 35 35 34 - 33 32 LOCATION OF RIGID BEAM DDENTS 9 EL. 82'-4' NOTES NUMBERED LDES INDICATE RIGID BEAM ELEENTS s

w westingnouse 230.4 nan 3 p

~

NRC REQUEST FOR ADDITIONAL INFORMATION li!i nisi

~

J Response Revision 1 Figure 230.41-2 Foundation of the Seismic Analysis Model for The Nuclear Island, Elevation 100'-0" si - # - as a "

as a

es e 47 _ 00 _ se a N sT 53 _ s4 m_ m ,

I 8 72 si is 3005 0%$$ CDffDI W NI STM e 100'-05 50 gy T4 100 de Te 99 i t 41 n l ur - si so u- us s1 - us us 54 ss - et al . up 7s - Tu j LOCATION OF RIGID BEAM ELEMENTS 9 EL f 00'-O' I

N0m ,

NL79ERED LDES DOICATE RIGID BEAM ELEMENTS !

l i

230.41(R1)-4 3 Westinghouse l

l l

NRC REQUEST FOR ADDlilONAL INFORMATION 3

~

Response Revision 1

Figure 230.41-3 Foundation of the Seismic Analysis Model for The Nuclear Island, Top of Basemat flG __ gy 11 I87 ; g, 8* 1 120 117 1 21 f27 f23 89 19 180 124 I

107 fue 109 11 0 106 203 86 ITS 125 162 163 164 IES . fas . 187 . taa ist 170 ITI f72 105 202 185 177 127 104 201 184 176 128 103 200 1E3 175 129

!47 148 149 150 tsi 152 153 154 155 158 157 158 159 160 161 102 199 iO2 174 130 1 01 ist 181 l73 l33 146 145 444 143 642 141 140 439 i38 437 636 135 134 133 132 LOCATION OF RIGID BEAM ELEMENTS e EL. 66'-6' (TOP OF BASEMAT)

NOTE:

NUMBERED LINES INDICATE RIGID BEAM ELEMENTS n l l

1 230.4 m n-s w westineouse ]

l

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Question 230.42 The following request for additional information pertains to Section 3.7.2.4 of the SSAR:

a. The first paragraph of Section 3.7.2.4 states that the nuclear island SSI responses generated for the analysis and design of seismic subsystems include nodal displacements, nodal accelerations and floor response spectra (FRS). Explain how the structural member forces (forces and moments) used for the structural design were generated for a soil site condition.

b. The last paragraph of Section 3.7.2.4 (Page 3.7-7) states that the selected soil conditions envelop the potential variation of soil properties and, therefore, the guidelines of SRP Section 3.7.2 for the variation of soil properties were not considered. Justify this statement, especially, when structures are founded on sof t soil site for which the variation (uncertainty) in soil properties should be carefully considered.

c. Explain the differences between the two phrases "the time-history SSI analysis using the program SASSI" and "the complex frequency response anal) sis using the program SASSI."

d. When the computer code SH AKE was applied, which soil degradation curve was used7 Response: (Revision 1)

a. The structural member forces used for the structural design for the soil site condition are generated as described in the last paragraph of subsection 3.7.2.1.1.

b. The soft soit profiles considered in the SSI amdysis (see Figure 2A-7 of the SSAR, Appendix 2A) has a linear shear wave velocity prolile varying from 1000 ft/see at the ground surface to 1200 ft/see at 240 ft depth. This profile is considered the minimum best estimate" velocity profile among the soil cases. In order to study the effect of lower bound shear wave velocities associated with the variation in the "best estimate" profile, a reduction of 50 percent in the low-strain shear modulus was analyzed. The resultant shear wave velocity for the lower bound case has a low-strain velocity of 707 ft/see at the ground surface increasing to 850 ftisec at 240 ft depth. This profile was analyzed using SHAKE and the Idriss 1990 soil curves (see response to IMI 230.79). The resultant strain-compatible soil properties were used in the 2D SSI analysis in the NS direction considering the depth to base rock of 120 ft and the water table at grade level. The SSI results of this case (marked as the lower bound soft soil case) are compared with the SSI results of the minimum "best estimate" profile (marked as the soft soil case) and with those of other design profiles (hard rock, soft rock, soft-to-medium soil)in Figures 230.42-1 through 230.42-11. As shown in these figures, the results of the lower bound soft soil case are enveloped by the results of other SSI cases except for small exceedences at very low frequencies (less than i Hertz) with no significant effect on the design responses. Based on these results, the soft soil profile (1000 ft/see at the surface to 1200 ft/sec at 240 depth)is considered to be the minimum best estimate" profile in the site interface conditions.

Tb-NRP-mpsn

f=-wk ' " - -i!-ype" - -

HdrH ' . nf th '" - ig 3 Westinghause 2m2m

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

t NRC REQUEST FOR ADDITIONAL INFORMATION

. p:: :nB p! I' Response Revision 1 '

- c- -

', alie . hat e ! cn perfe m ed r--

E ad range af - :' -.nd - ? " cendi: : ' . tppendi-2,1 ::f P c cctR)! *'r np:.' r.:m p; _ neg . . _ ;, _. .;;p ,;;= _ r .h e :.. e de . gn ' ,

;'r ide b .:gn ; . age f::- ^- nun, p!an! ::c . , pr ' c'

' :f d ,:gn : :!! pr":' - -ing . :^-c'.* r- =

! M 9 fp: " 9 fp:, ' .:

F-- es:25': 'e ' fn- the-p!n + . . & ,: gn ' ed : - - ' . . ' - n f t'. ge - -

- " 'e

rfn. :w;.. . m:; "t.

H cr. ' -- 2.5 equ :re , prop: n' '

> ' i ' - 'h:

g: *

. - - ~ - ', ' . , , e5 ::,

rn ; . t ,

..u,. (t. .ed: ' ni ' >

. ' - 1: 9 prcpe- .-1 g-. *c 5_ : r eq u ' '

!f m fp rn _. _; g. . i r. .

r,. . . t . g ; pp, ,. .p_. i :,,,

c. As discussed in responses to RAI 230.34. Section 3.7.2.4 will be revised and the statements are clarified to read as follows* .

The soil-structure interaction (SSI) analyses of the nuclear island are performed using the program SASSI.

and

SSI analyses are performed using the complex frequency response method with computer program SASSI.

d. As discussed in Section 2A.4, the strain-dependent shear modulus and damping curves used in the free-field SH AKE analysis are presented in Figures 2A-8 for soil materials and in Figure 2A-9 for rock materials. These ,

curves are obtained from references 6. 7 and 8 shown in Subsection 2A.7.

SSAR Revision:

See SSAR revision identified in response to RAI 230.34. *

l 1

?

b 230.42(R1)-2 T Wesunghouse 4

6

n. + ., - a m

NRC REQUEST FOR ADDITIONAL INFORMATION ,

Response Revision 1 Figure 230.42-1 i

2D SASSI Analysis. N-S Direction Lower Bound Soft Soil Profile ACCELERATION RESPONSE SPECTRA 2.0 i . . . i i ii i e i i i i .. . . . . . . ..

- Ri-AX HARD ROCK NI N-S Direction !

- - A2-X SOFT ROCK

-- B2-WX SOFT TO NEDILM NODE 182, FOUNDATION MAT

- - C2-WX SOFT S0ll ;

-- D2-WX LDER BOUND SOFT SJIL 2r DArING 1.5 ,

m I )\

z /

f B \

n 7 I I w i ,

a1g i.

w i j

gi' u '

p 4 A

u

,,'*'.?:.)\l\1\\/\

se .. 4 J

f':.\'1\'.

l .

. , \

',,s; ev, $

y /, s a..,. 1 x

/ \ ,

\ 'N ,

QY-7. M.= :.-- __.

i i e i e iii i i i i i e ii i i

,g e ie i ei 10-1 18e 101 1g2 FREQUENCY-CPS s

23o22mn-3 T westineouse 1

l

I J

l 1

NRC REQUEST FOR ADDITIONAL INFORMATION j l

Response Revision 1 l

i Figure 230A2-2 2D SASSI Analysis. N-S Direction Lower Bound Sof t Soil Profile i

ACCELERATION RESPONSE SPECTRA 4.8 i i i i iiii i i i,i i . . ,,,ii

- Ri-AX HARD ROCX NI N-S Direction

-- A2-X SOFT ROCK

- WX SOFT TD ltDIUM NDDE 3083, Elev. 188'

C2-WX SOFT SOIL

--- 02-WX LOER BOUND SOFT S))L 2r DAl@ING

f. 3. 8 m

z S

s e

W d 2.s u

E l\

I\

d e

s u A/

/ -% !

W {

sd S I'8 *' ' -

.hV pi sh \.

, i A h, ,% ,,, /(., ,

fJ' '% f%,'gt/f N y

'u N.k Q. %w. fi

-- . n .~. =. :-

,g i e i i iiei e i ,ie i ii i i ie i i i i 18-8 188 181 gga FREQUENCY-CPS s

230.42(RI)-4 W Westingh0use

i i

NRC REQUEST FOR ADDITIONAL INFORMATION i

Response Revision 1 Figure 230.42-3 ;

2D SASSI Analysis, N-S Direction 1 Lower Bound Solt Soil Profile ;

i 4

ACCELERATION RESPONSE SPECTRA 4.8 . . . . , , , . . . . i , , ,, , , , , , , ,,

- R1-AX HARD ROCK NI N-S Direction

- A2-X SOFT ROCK

-- B2-WX SOFT TO MEDILM NDDE 3084 Elev. 117.S' i

C2-WX SOFT SDIL -4'

- - D2-WX LOWER BOUND SOFT S3IL 22 DAWING fe 3.8 i in 2 i 0

H I te w '

d 2.8 8 a

< l\

> 1. L y J' J r / l N . \ '-- ~^ '(

' _ Y ny V a 1*8 -

sn j..: .' ';4

,,) %m '*' J f

\

-h $, 0s w_

L-2:' = '-T.-2 ,

i i i e iei i i ie i ie i i i ie i ei

,g e e 18~1 18e ggt 182 FREQUENCY-CPS i

n i

230.42(R1)-5 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION 15 9 in Response Revision 1 t

Figure 230A2-4 ,

2D SASSI Analysis. N-S Direction Lower Bound Soft Soil Profile ACCELERATION RESPONSE SPECTRA 6.0 i i i i . . ii i i i i i iii i . , i i i ii

- R1-AX HARD ROCK NI N-S Direntson '

- A2-X SOFT ROCK

- - B2-WX SOFT TO NEDItM NDDE 3886. Elev. 153'

C2-WX SOFT SDIL

-- D2-WX LDER BOUND SOFT S31L 21 DAWING co 1

s M

z 4.0 0 -

m W

J w

U l U }

'. \ h 2 2.0 l\ ,,)

w I i i a s l M s /.

3 A

l* Dlh:.. % ,' ._' \ \l U'O

'%. .. A

'd

.Q...........,

,g i i i e i t e r i e e iee i e i i e e ii 18-1 10s 101 gg2 FREQUENCY-CPS i

230.42(RI)-6 W -

Westinghouse l

i NRC REQUEST FOR ADDITIONAL INFORMATION IE M iliti

+

Response Revision 1 r y __

g Figure 230.42-5 ,

2D SASSI Analysis, N-S Direction ,

Lower flound Sof t Soil Profile 1

ACCELERATION RESPONSE SPECTRA 6.8 i . . . ,,ii i i i i , , . . , , , , , , , , ;

- R1-AX HARD ROCK NI N-S Direction

- A2-X SOFT ROCK '

-- B2-WX SOFT TO EDIUM NODE 3088, Elev, 188'

C2-WX SOFT SOIL

- D2-WX LDER BOUND SOFT S31L 2r DAW ING 4

m i

e un z 4,8 O

>.e H

st W

1 4

W U '\ , ,

U 4 \ '

[ '

.J '

< ,\

tr 2.8 ' '

$ I !

W # I*N./ } . gf ;

i 11.

U3 v, fv(/S i, .

\<I' C bE.=0.nr.nT, ;;,,,, j

.8 18~1 338 get ig2 {

FREQUENCY-CPS l

1 230.42(RI)-7 3 Westinghouse

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

I l

- NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 230.42-6 '

2D SASSI Analysis. N-S Direction Lower Bound Sof t Soil Profile ACCELERATION RESPONSE SPECTRA 12.0 i i i , iiii i i , iiiie i i ,i,,i.

- Ri-AX HARD ROCK NI N-S Direction

- A2-X SOFT ROCK

- - B2-WX SOFT TO K DIUM NODE 3016, Elev. 387.25'

C2-WX SOFT SOIL

-- D2-WX LDER BOUND SOFT S31L 2r DAWING s

?

a j h

2 B.e ! ,

S / :}

F i

= '

b Il d lII u

lI J h

< i: . I \ \.

m 4'e ')

F u { b'I !

w / \ l Ei p / /f ., \ 'k'

/ f " vl '\ \[ %s__

l

~ 3,%. p. 5,u ' i

.e t o-1 te' let le2 FREQUENCY-CPS l

1 l

230.42(RI)-8 i W ==

Westinghouse J 1

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

i NRC REQUEST FOR ADDITIONAL INFORMATION

~

Response Revision 1 7-Figure 230.42-7 2D SASSI Analysis. N-S Direction Lower Bound Soft Soil Profile ACCELERATION RESPONSE SPECTRA 8.0 , , , , ,,,, , , , , , , , , i,,,,,,.

- Ri-AX HARD ROCK SCY N-S Dermotion

-- A2-X SOFT ROCK ,

-- 82-WX SOFT TO MEDIUM NDDE 3118. Elev. 285.33'

C2-WX SOFT SOIL !

-- D2-WX LDER BOUND SOFT S3IL 2r DAWING f 6.8 ;

ti z

S s

E f d 4.8 g . .

ld m

F 't -

n f-Si.82

, y; g,f.I y :

~ MQ \ y,A

, , i iiii l}$,Ahrl

i W

,,,iii,

.... )jb' N mm:=m.

,,,,,,ii t

1e-2 188 1st is2 FREQUENCY-CPS s

l 230.42(R1)-9 3 Westinghouse ,

. , =. - - . . . . - .

NRC REQUEST FOR ADDITIONAL INFORMATION r

^

g Response Revision 1 Figure 230.42-X 2D SASSI Analysis. N-S Direction Lower flound Soft Soil Profile ACCELERATION RESPONSE SPECTRA B.O . i i . ii. . i i iiiii i i ,iiiii

- Ri-AX HARD ROCK SCV N-S Direction

-- A2-X SOFT ROCK

-- B2-WX SOFT TO NEDIUM NDDE 311S. Elev. 2S5 33'

--- C2-WX SOFT SDIL

-- D2-WX LDER BOUND SOFT SJIL 2r DAWING f 6.0 )

l

@ t

% \

Ya d 4.0 bl ii\\

U l I

< g

.; T I i

< :j 1 1 E \ l I

u w

\ I: p,\

is ,\ , .

I ' . .i '\y ll \ '

Ei * * *

// j ,

\

, ,, g/,b t/\'\,-k(u

. > \,

'* - * ',l

. %y k,%.s"~

, , , ,,,, , , , ,,,i, h-:x.'h...:..T.:

18-1 10e 181 gg2 FREQUENCY-CPS 230.42(RI)-10 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

_.- I Response Revision 1 Figure 230.42-9 2D SASSI Analysis, N-S Direction Lower Bound Sof t Soil Profile l

l l

ACCELERATION RESPONSE SPECTRA 4.0 , , , , , ,,, , , , , , , , , , , , , , , , ,

- R1-AX HARD ROCK CIS N-S Direction l

-- A2-X SOFT ROCK

-- B2-WX SOFT TO NEDILM NODE 3281, Elev, 98.80' .i

--- C2-WX SOFT SOIL ..

-- D2-WX LOWER BOUND SOFT S3IL 2r DAWING l I

l

3. 0 l i.

U z !

l4 1 7 i Lij d 2.0 W

b

.3 I\

$ / \?N r A/ )

U \ \. A

. Y 8 18

.. .,,, , ,-g,

, ;c@

/ f.v' u,Q,s/ x'..g y, f fx,;

u t.,vp NAn. s

.n, ., r

~0ry 0

18-8 10e get ist FREQUENCY-CPS 23o22mu-n T westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 230.42-10 2D SASSI Analysis. N-S Direction Lower Bound Soft Soil Profile ACCELERATION RESPONSE SPECTRA 4.e , , , , , , ., , , . ..>i i i i . . i i .

- R1-AX HARD ROCK CIS N-S Otreotion

-- A2-X SOFT ROCK

- WX SOFT TO E DILM NODE 3293. Elev. 187.17

C2-WX SOFT SDIL

-- D2-WX LDER BOUND SOFT S]IL 2r DAWING i 3.0 (f)

Z O

w ,

b It:

W d 2.B i U

a I s /

\ ~,N S ,

r- A/ T u 1 w \ \h .

g 1. 0 .. ,

, , , ,, ,,,g -V- .j s

',,j -

/\ \

b,'\;'FyQ Q i.

t I t 1 i t I i l i l iif I I

% ==s= f I I I I I I I

',., 1 ,. 1.i 1.-

FREQUENCY-CPS '

1 I

230.42(RI)-12 W WestinEhouse i

.r- ., -

NRC REQUEST FOR ADDiflONAL INFORMATION

. --z Response Revision 1 -

Figure 230.42-11 2D SASSI Analysis. N-S Direction Lower Bound Soft Soil Profile ACCELERATION RESPONSE SPECTRA 6.0 i i i i i iii i i i,,,,, i i , ,,,ii

- R1-AX HARD ROCK CIS N-S Direotton

- A2-X SOFT ROCK l

-- 82-WX SOFT TO MEDILM NODE 3294, Elev. 13S.2S'

C2-WX SOFT SOIL

- D2-WX LOER BOUND SOFT S31L 2r DAMPING

? >

a 1 en z 4.8 8

h

< 1 E

.a U

u

.J

< J g 2.e -

u A W

g ',

l N ~v M

k'^'..L

-'"~ ~j 'N A,v.

.[Is

/, ( g hy

,e , , , , , , , , , ,

18~1 10e ggi g32 ;

I FREQUENCY-CPS 1

[ WE5tinghouse _

l

. I

NRC REQUEST FOR ADDITIONAL INFORMATION j;s: E+;

Response Revision 1 Question 230.79 From the review of Figures 3.7.1-14 and 3.7.1-15 of the SSAR. it appears that the soil shear degradauon curves for the typical soil used in the analysis and design are based on the soil shear degradation model recommended by H.B.

Seed and 1.M. litriss in 1970. A comparison of the shear degradation cunes presented in Figures 3.7.1- 14 and 3.7.1-15 in the SSAR with the current published industry results such as the results published by 1.M. Idriss and Geomatrix in 1990, shows that the Seed-Idriss 1970 curves overestimated the soil strain degradation The staf f anticipates that the use of the Seed-Idriss 1970 curves in the SSI analyses of the NI structures will underestimate the seismic structur:d responses. Provide the basis for using the Seed Idriss 1970 curves in the SSI analyses.

Response: (Revision 1)

The4944d+RAl-wa+4iwwwd4uring*uwdog-anamp NRC qaff-m*1-eemultant nd "': aingl*w.smi hhtelece.mie-analy e 4.n,Aprit-44AW4-aml+ilt4v41iea14arther41* ring *meetmg Aeduled-aHhe+wl of 4 : A >>atten-m.t*w+4.Hid4Al-w#14v-preparel-lollowing-+he-May-nwthw The range of soil degradation curves proposed by Seed & Idriss in 1970 and confirmed in the 1984 study are shown in Figures 230.79-1 and 230.79-2. The average curves corresponding to this range with limiting soil material damping value of 15 percent were used in the seismic SSI analysis of the N1 with the foundation soil sites. The soil degradation curves repor1cd by Idriss (1990) are also shown in these figures. In order to assess the impact of the new soil degradation curves on seismic SSI responses, the case of sof t-to medium soil case was re-analyzed in the EW direction using the Idriss 1990 soil degradation cunes and the 2D SSI model of the nuclear island. The results of 2D analyses are compared with the 3D design responses in Figures 230.79-3 through 230.79-13. As shown in these figures, and depending on the nodal h) cation, some small variations are observed with respect to the frequency and amplitude of the response. These differences are small and are generally covered by the 3D enveloping results.

In light of the results of this parameuic study and considering the fact that the design responses are the envelope of all soil and rock cases, re-analysis of SSI cases using Idriss 1990 curves are not considered warranted, In relation to the above RAI. a concern was raised by the NRC team during the July 1994 audit regarding the adequacy of the range of soil degradation curves for sites with clay contents. In order,to address this concern the site-specific soil degradation curves for the Savannah River Site (marked as Geomatrix and Bechtel) and the Lotung site are compared with the range of Seed & hiriss cunes in Figures 230.79-14 and 230.79-15. The soil materials at these two sites can be characterized as silty sand and clayey sand. As can be observed from these figures, the range in Seed & Idriss' study adequately covers the soil degradation cunes of these sites. On the other hand, the strain-dependent propenies of pure clay materials with high plasticity fall above the range in Seed & Idriss' study.

However, sites with pure high plastic clay wouhl not meet the AP600 interface requirements for minimum shear wave vekeity charisteristics and for hearing capacity.

SSAR Revision: NONE i

W Westinghom

l NRC REQUEST FOR ADDITIONAL INFORMATION 1

~

i N

W Response Revision 1 Figure 230.79-1 2D SASSI Analysis. E-W Direction Soil Degradation Curve Evaluation M .

?

. Idriss,1990

.c 2 e

0.8 e a

& O. /,

%.f

, Seed & Idriss / '

=

(1970,1984) }(,@$6Sjj

v- 0.6 n #.1;

.m >- :py

~ ~ , ,, , . .

o o m m ' % :13"~

2 2 04 4%h a o :::#

V V :fe;^:-

O O -

- r 2 2 j(n', g g g 0.2 ,;

e m :1

.c x: -

mm '

g 0 1 10-4 10-3 10-2 10-1 10-0 Shear Strain, Y - Percent Variation of Shear Modulus with Shear Strain for Sands 230.79(RI)-2 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 230.79-2 2D SASSI Analysis. E-W Direction Soil Degradation Curve Evaluation 28

, .:n , ;

? :<

24 Seed & Idriss .g.(:@fdbe e ',, 'e (1970,1984) J M :.N' 'S$c,' s 20 ' % ' M'* :9 s s. . '

c . g' c

_R o :

o , y ', '

.' x'.~~'3.;^O;p

~

u .;' N o 16

" R$ i:

cn $l: :& . f %{20::.:, ;'if -------

,e 12 , ' ff c:r?

a. !:c W. s'xc..' "

E $ , :0~,,: 35:a';: _'. :. iN - '/

n/'":.5'#" .xN "M r

m #

O 8 , .. ,. .

x8:

o m

Mkh:'$f \ idriss (1990)

- -' - ~ W. y r 4 '

t O

10-4 10 3 10-2 10-1 10-0 f Shear Strain - Percent Damping Ratios W Westinghouse 230.79(R1)-3 .

l l

- ~ .

4 NRC REQUEST FOR ADDITIONAL. INFORMATION Response Revision 1 y

Figure 230.79-3 ;

2D SASSI Analysis, E-W Direction ,

Soil Degradation Curve Evaluation !

ACCELERATION RESPONSE SPECTRA 4.0 , , , , , , , , , , , , , , ,, , , , , , , ,,

- DESIGN NI E-H Direotton

- B2-Y 2D SDFT TD ED. , SEED 1978 CURVES

-- B2-Y 2D SDFT TD MED. , IDRISS 1998 CURVES NDDE 182, FOUNDATIDN NAT t 22 DAWING

" 3. 0 a

z O

4 L ce W :

[;j 2.0 l U i U t.

< ~

f UJ ,

I I A g'g 4\

,f '

ff \) :

b ;

i i i i t

,g t i e t i i e r # i ee e e r s ei 18~1 19e 101 182 ;

FREQUENCY-CPS i h

230.79(RI)-4 3 Westinghouse

?

I

l i

NRC REQUEST FOR ADDITIONAL INFORMATION l 1

= $

Response Revision 1 Figure 230.79-4 2D SASSI Analysis. E-W Direction Soil Degradation Curve Evaluation ACCELERATION RESPONSE SPECTRA 4.e i i e i i i ii i i iiiiii i i i iiiii

- DESIGN NI E44 Direction

-- 82-Y 2D SOFT TO ED. , SEED 1970 CURVES

-- 82-Y 20 SOFT TO ED. , IDRISS 1998 CURVES NODE 3803, Elev. 180' 2r DAWING f 3.8 a

z Fa d 2.8 8

i e T

\

d o- i.e I t.s .-

3 f/s

.!.\g/4

^f\

s~ ~ ~ _

/ , , , , , ,, , , , ,,,,i , , , , , , , ,

18-5 tes ist 182 FREQUENCY-CPS s

W Westinghouse 230.79MS O

NRC REQUEST FOR ADDITIONAL INFORMATION

.m_

Wd Response Revision 1 4

Figure 230.79-5 2D SASSI Analysis. E-W Direction Soil Degradation Curve Evaluation ACCELERATION RESPONSE SPECTRA 4.0 . . . . . i ii . . . . i e i i i i i iiiii

- DESIGN NI E-W Direction

- B2-Y 2D SOFT TD ED. , SEED 1978 CURVES

-- B2-Y 2D SOFT TO ED., IDRISS 1990 CURVES NDDE 3084 Elev. 117.5" 2r DM6IING e

i 3.0 a

5 C

sw dee '

8 <

< , 'y ,

i! aI E i a

w \

a (L 1.8 .i

{d%\ -.

g g ( l l t l l t I f I' f I I f f I I l I I I I

',.. 1 i.e 1.-

FREQUENCY-CPS s

230.7W Westinghouse

r NRC REQUEST FOR ADDITIONAL INFORMATION E+MEi Response Revision 1 Wah Figure 230.79 6 2D SASSI Analysis, E-W Direction Soil Degradation Curve Evaluation ,

e ACCELERATION RESPONSE SPECTRA 6.0 , , , ,,,,, , , , , , , , , , , , , , , , ,

- DESIGN NI E-W Direction -

-- 82-Y 2D SOFT TO ED. , SEED 1578 CURVES i

-- 82-Y 2D SOFT TD MED. , IDRISS 1998 CURVES NDDE 3806, Elev, 153' 22 D W ING ,

in 2 4.e a ,

5 f I a

i x 2. , r5 I >

\

4h _

.g r' . , ,,,i, , , , , , , , , , , , , , , , ,

i 10-1 ige 1g1 102 FREQUENCY-CPS i

l 230.7HRD-7 T Westinghouse 1

l J

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 230.79-7 2D 'sASSI Analysis. E-W Direction Soil Degradation Curve Evaluation ACCELERATION RESPONSE SPECTRA 6.8 . , , , , , ,, , , , , , , , , , , , , , , , ,

- DESIGN NI E-W Direction

- B2-Y 20 SDFT TD ED. , SEED 1978 CURVES

-- B2-Y 2D SDFT TD ED. , IDRISS 1998 CURVES NDDE 3008. Elev. 180' i 2r DAWING f

a m <:

z 4.8 e / .

I m I1 ,

e 8

< b h2.e If \ j N \

% \ '

\

'g l I' s i i i n ii ,i,e in ii i , ,ii,,i 10-1 18e get gg2 FREQUENCY-CPS ;

230.79(RI)-8 W westinghouse i

"7 e - _

NRC REQUEST FOR ADDITIONAL INFORMATION Siu 14 2 Response Revision 1 Figure 230.79-8 2D SASSI Analysis. E-W Direction Soil Degradation Curve Evaluation ACCELERATION RESPONSE SPECTRA 16.e i . . . , i i i i i i . . . i . . , , , , . . .

- DESIGN NI E-W Direction

- B2-Y 2D SOFT TO MED. , SEED 1978 CURVES

-- B2-Y 2D SOFT TO MED. , IDRISS 1998 CURVES NDDE 3e16, Elev. 3e7.2S' 22 DAMPING

, 12.8 lh .--

l z

% )

5 ! ,

d e.e 8 4

< r\

a i \

l

$ J \

U f \

$** 4! \

{9%

gr 17 K

- m -- - --

g a i i f f I I f f 1 I f f f f I i I I i f f f 18-5 tes set 182 FREQUENCY-CPS W Westinghouse 230J9(R&9

. . . . = - - . . . . . . - . - - .- ..

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 230.79-9 2D SASSI Analysis, E W Directica Soil Degradation Curve Evaluation f

ACCELERATION RESPONSE SPECTRA 8.0 i i i i i iii i i iiiiii i i iiiiii

- DESIGN SCY E-H Direction

-- B2-Y 20 SOFT TO ED. , SEED 1579 CURVES

-- B2-Y 2D SOFT TO ED. , IDRISS 1958 CURVES NODE 3118. Elevation 295,33' 2r DAW ING

" 6. 0 a

$ I) 6 \

d 4.0 ti< '

Y tr

/& r*, (

Bi vv

\\ 1,y

/ if %w_._

10-1 108 101 102 FREQUENCY-CPS 230.79(RI)-10 W Westinghouse o . .

-. . . = .. . . - . . - . .

NRC REQUEST FOR ADDITIONAL INFORMATION smusam Response Revision 1 Figure 230.79-10 2D SASSI Analysis. E-W Direction Soil Degradation Curve Evaluation ACCELERATION RESPONSE SPECTRA B.O i i , , , i i, , , , , ,,ii , ,,,,,,i

- DESIGN SCV E-W Direoison

-- B2-Y 2D SOFT TD ED. , SEED 1978 CURVES

-- B2-Y 2D SDFT TD ED. , IDRISS 1998 CURVES NDDE 3115. Elev. 256.33' I

) 2r DAW ING f 6.0 '

=

\

p w

d 4.0 ti 1 f\ T

! \ .,

a. 2.0 , . , .

/ O f f ke % N

, l W___

f f I I I I l f f I i I f f f f f I I f I I f 18'1 10e ggt 182 FREQUENCY-CPS s

230.M R1b11 3 Westingh0use

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 230.79-11 2D SASSI Analysis. E-W Direction Soil Degradation Curve Evaluation ACCELERATION RESPONSE SPECTRA 4.0 . . , , . .,, , , , i,,,, i i i...

- DESIGN CIS E-W Dirmation

-- B2-Y 20 SOFT TO MED. , SEED 1978 CURYES

-- 82-Y 2D SOFT TO MED. , IDRISS 1938 CURVES NODE 3201, Elev. 38.8B' 2r DAWING I 3.0 a

5 1<

5 d 2.e li<

z , ,

1.8

/ Y \

J \lL!\

L% I

' ' ' i i ei , , , , , , , , ' ' ' ' ' ' ' '

,e is-1 tes 3,,

FREQUENCY-CPS l

l 1

1 l

l 230.79(RI)-12 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

-+ - -

"k Response Revision 1 Ren Figure 230.79-12 2D SASSI Analysis, E-W Direction Soil Degradation Curve Evaluation ACCELERATION RESPONSE SPECTRA 4.0 i i i i ii i i > i iiiii i i e iiiii

- DESIGN CIS E-H Dereollon

- B2-Y 2D SDFT TD MED. , SEED 1970 CURVES

-- B2-Y 2D SDFT TD MED. , IDRISS 1998 CURVES NDDE 3203, Elev. 107.17' 22 DAWING i 3.0 a

5 C

w d 2.0 0

a: \

> V\

\

d \\

l[I Nf y

/ we\ v%u 10-8 10s 101 102 FREQUENCY-CPS W Westinghouse 230.M R1F13

NRC REQUEST FOR ADDITIONAL INFORMATION l l

3: - ;; ;

t' Response Revision 1 ;

Figure 230.79-13 2D SASSI Analysis, E-W Direction Soil Degradation Curve Evaluation ACCELERATION RESPONSE SPECTRA 6.0 , , , , , , ,, , , , , , , ,, , , , , , , , ,

- DESIGN CIS E-N Dirmation

-- B2-Y 20 SDFT TD MED. , SEED 1970 CURVES

-- 82-Y 20 SDFT TD ED. , IDRISS 1930 CURVES NDDE 3204, Elev. 135.25' 2r DAW ING l

l z 4.0 0

5 )

e ti k {

E 20 IV. g a.

sn V

\

/ w i

' i i i e i ii i i e i , e ii i e i i iiii 10-1 10s 101 102 i

FREQUENCY-CPS i

230.79(RI)-14 3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION n: z::.

Response Revision 1 1 Figure 230.79-14 2D SASSI Analysis, E-W Direction Soil Degradation Curve Evaluation 10

,I ?;j;;; ' - Geomatrix

- h;; fg (O' to 50',1990) 5 Y, . . k ,[e C E 0.8 ,g,;, x

'@ g 7:%;g \, ,

g ,'o Seed & Idriss

,,.;7,g;'y, (1970,1984) ;l ' x\,

u ,

a w 0'6 -

$ t ' NN['N'; ., i Bechtel U)

> 'b, ' 'l2

\

[(RTF,1993) o o 'L; g,g"; N,

0*4 ~ ;,? <,';

h "3 3 's, 5O 5O Lotung -

,,N[I' \

.9,,

2 2 W'

% % 0.2 e e .

.C .C ,

u) u) 0 <

10-4 10-3 10-2 10-1 10-0 Shear Strain, Y - Percent Variation of Shear Modulus with Shear Strain for Sands 230.M R1F15 W Westingh00Se

. NRC REQUEST FOR ADDITIONAL INFORMATION g.. . :.

w m

2 Response Revision 1 Figure 230.79-15 2D SASSI Analysis E-W Direction Soil Degradation Curve Evaluation 28 i Bechtel (RTF,1993) - .

's ' f.

20 4

/NC "

$ b C ,,97c. x: s E Seed & Idriss s 226 'j'eh!N:2 M

e 16 (1970,1984) _\ S's! "

f,jek

Q*

I

a. ; ~%;t:. s :

w

. c: g:je.c'spc /

y'j;;

' ,., (' , ' ' ' ' Lotung 1

??S :l?

y 8 /'I

NT s-9:s x3 s

\ N Geomatrix

s. ,e (O' to 50',1990) 0 i 10-4 10-3 10-2 10-1 10-0 Shear Strain - Percent Damping Ratios For Sands 230.79(RI)-16 W Westinghouse

l 1

l NRC REQUEST FOR ADDITIONAL INFORMATION yuy +=

s Response Revision 1 Question 231.29 it appears that the Poisson ratio values selected for soils above the water table may not be consistent with values nonnally expected for silly sands of densities high enough to support a shear wave velocity of 1(XX) fps. Evaluate and discuss the ef fect of the assumed Poisson ratio values on the SSI responses.

Response: (Revidon 1) 1 WWa4-+4-4mmmlSe s=: ' ratic ca'ues m-ilmh.fwh4-le .uhmittal-by-Julyv.44%

The case of sof t-to-medium wil profile with the water table and base rock at 120 ft depth was analyzed using the 2D SSI model of the nuclear island in the NS direction. For this soil profile, a constant Poisson's ratio of 0.25 was used for all soit layers. The results of the SSI analysis for this case were compared with the resuhs of the same soil ,

case with the previously used Poisson's ratio of 0.35 in Figures 231.29-1 through 231.29-11. As shown in these figures, the effect of charige in the Poisson's ratio on the SSI responses is relatively insignificant. It should be also noted that the design soil profiles which were identined from a series of 2D SSI analysis cases in the SSAR define the water table at grade level for all soil cases. For these profiles, the Poisson's ratios were adjusted to maintain the P-wave vekrity of water.

SSAR Revision: NONE l l

l i

1 l

l 1

l l

W WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION g .....

~ '

Response Revision 1 Figure 231.29-1 2D SASSI Analysis, N-S Direction Poisson Ratio Evaluation ACCELERATION RESPONSE SPECTRA 4,8 i i e i i e ii e i i iiiii i i e i i i ii X SOFT TO MED,. PO!=. 5 NI N-S Direotton

-- B2-LX SOFT TO ED. POI =. 25 NODE 182, FOUNDATION NAT 21 DAMPING f 3,8 a

5 p

5 W

d 2.8 A 8

d }

\

1.8 k

Y-48 18~1 18e ggt gga FREQUENCY-CPS s

231.29(RI)-2 T westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 231.29-2 2D SASSI Analysis, N-S Direction Poisson Ratio Evaluation 9

ACCELERATION RESPONSE SPECTRA 4.0 i i i i iiii e i i i i i i e i i i ie i i i X SOFT TO ED. POI =. 55 NI N-S Direction

-- 02-LX SOFT TO KD. POI =. 25 r NDDE 3083. Elev. 188' 2E DArING

3.0 a

z O

1 d 2.0 \

8 J

8 g 1.8 w -

W_

18-1 188 188 102 FREQUENCY-CPS i

n i

i l

231.29(RI)-3 9 E I

l

\

t NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 231.29-3 2D SASSI Analysis, N-S Direction Poisson Ratio Evaluation ACCELERATION RESPONSE SPECTRA 4.0 i i i i i i i i i i i i i i i a i i i ' i 8 i *

- B2-X SOFT TO ED. PO!=. E NI N-S Direction LX SOFT TO NED POI =. 25 NDDE 3084, Elev. 127.5' ,

22 DAWING i 3.0 a

Z M

4 !

5 h d 2. 0 i i

I )

E S <

g i.e w I I I I I I f f I I I I I I I I I I I ' I i,-,

I I I i 2.i 1.-

FREQUENCY-CPS s

I 231.2m-4 W Westingflouse

.o - , ~ -

NRC REClJEST FOR ADDITIONAL INFORMATION t;... ..

r n Response Revision 1 Figure 231.29-4 2D SASSI Analysis, N-S Direction Poisson Ratio Evaluation ACCELERATION RESPONSE SPECTRA 6.0 , , . , , ,i, , , , , . . . . . , , , , , ,

- B2-X SOFT TO E D POI =. 35 NI N-S Direction

-- B2-LX SOFT TO MED. POI =. 25 NDDE 3006, Elev. 153' 23 DAWING f

a z 4.0 0

7 a:

w d

< i

[ 2. 8 ,

u E

m f \ j

., i 10-1 18e 101 102 j FREQUENCY-CPS 1 w westingnouse 23i.2mn-s !

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 231.29-5.

2D SASSI Analysis, N-S Direction Poisson Ratio Evaluation ,

ACCELERATION RESPONSE SPECTRA 6.e i i i i i iii i i i iiiii e i iiiiii

- B2-X SOFT TO ED. PO!=. 35 NI N-S Direction

-- B2-LX SOFT TO ED. POI =. 25

[

2r DAWING

?"

a t z 4.0 S

7 5 ,

d O s d

g !

% 2.0 j O -

/

a.

^g .

1s-1 188 ist 182 FREQUENCY-CPS 231.29(R1)-6 T Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 231.29-6 2D SASSI Analysis, N-S Direction Poisson Ratio Evaluation ACCELERATION RESPONSE SPECTRA 12.0 i , , i i i,, , , i i i iii i i i ii .

- B2-X SOFT TO NED. POI =. 35 NI N-S Direation

-- 02-LX SOFT TO NED POI =. 25 NODE 3018, Elav. 307.25' 2r DAWING i

a z 8.0 o

b d

0 N .h.

f 4.0 U

q

\

g

, i

/

e' (

.0 10-1 108 101 102 FREQUENCY-CPS i

23i.2 nan-7 w westinghouse l

l

~

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 231.29-7 2D SASSI Analysis, N-S Direction Poisson Ratio Evaluation ACCELERATION RESPONSE SPECTRA 8.0 i i , i i iii i i i i i iii i i iiiiii

- B2-X SDFT TD NED. PDIs. 95 SCV N-S Direction

- B2-LX SDFT TD IED. PDI=. 25 NDDE 3118. Elev. 295.33' 22 DAWING 1 6.0 5

C n.

W d 4.0 0

J

$ ? I) b 2.0

$ g s

18-1 108 101 102 FREQUENCY-CPS 231.29(RI)-8 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

- .. 4 Response Revision 1 1 Figure 231.29-8 ,

2D SASSI Analysis, N-S Direction Poisson Ratio Evaluation ACCELERATION RESPONSE SPECTRA 8.0 i e i e i iii i i e i i i e i i i i e i i i i

- B2-X SOFT TO E D. POI =. 55 SCV N-S Dirnotion LX SOFT TO ED. POI =. 25 NDDE 3115, Elev. 256.33' 22 DAl@ING t 6.0 a

2 3

4 5

a <.e a t'

N g 2.0 ,'l \

l i i , e e ii e i i i e i i i e i e i i i i i 18'1 19e 101 182 FREQUENCY-CPS s

i W Westinghouse

i I

NRC REQUEST FOR ADDITIONAL INFORMATION

n me Response Revision 1 i Figure 231.29-9 2D SASSI Analysis, N-S Direction Poisson Ratio Evaluation ACCELERATION RESPONSE SPECTRA 4.0 i i , , ii. . i i . . , , , i iii,,.

- B2-X SOFT TO E D. PO!=. 55 CIS N-S Direction

-- B2-LX SOFT TO ED. POI =. 25 NODE 3291. Elev. 98.80' 2r DAWING i 3.0 a

2 3

4 tt: ,

I d 2.0 8

_1 h '

8 g 1.0 v.. _

.8 18'1 108 108 182 FREQUENCY-CPS n

231.29(R1)-10 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION FN Response Revision 1 ,_

Figure 231.29-10 2D SASSI Analysis. N-S Direction Poisson Ratio Evaluation ACCELERATION RESPONSE SPECTRA 4.0 . . . . .,ii i , . . . . . . . . i,,,,

- B2-X SOFT TO NED. POI =. 95 CIS N-S Direction

-- B2-LX SOFT TO LED. POIs. 25 NODE 3293. Elev. 187.17' 22 DAW ING

" 3. 0 a

E C

tu i' d 2.0 S

i Y

E S

g 1.0 v..

b

.8 18'1 19e 101 102 FREQUENCY-CPS s

(

231.29(RI)-11 9

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Figure 231.29-11 2D SASSI Analysis, N-S Direction Poisson Ratio Evaluation ACCELERATION RESPONSE SPECTRA 6.0 i i . . . iii i i iiii

"~

i i i iiiii

- B2-X SOFT TO NED POI =. SS CIS N-S Direction

- B2-LX SOFT TO ED. POI =. 2S 2r. DAWING f

a z 4. 0 0

tr 0

W d \"

$ 2.0 u '

E w

v l

.8 10-1 10s 10t 102 FREQUENCY-CPS I

I 5 I l

l 231.29(RI)-12 W Westinghouse i

I

NRC REQUEST FOR ADDITIONAL INFORMATION k gi Question 410.145 Section 10.3.2.2.1 of the SSAR states that the main steam lines between the steam generator and the containment penetration are designed to meet the leak-before-break (LBB) criteria. He application of LBB in current PWRs is only for the reactor coolant system, w hich has a reactor coolant pressure boundary leakage detection system in accordance with RG 1.45. In order to apply LHB to the nuin steam lines, it has to have a steam line leak detection that is comparable to the reactor coolant pressure boundary leakage detection. Describe the main steam leak detec-tion systems, instrumentation, acceptable leak criteria, and the requirements to be included in the plant technical specifications.

Response

Main steam line leak detection is discussed in SSAR Subsection 10.3.3 which refers to Subsection 3.6.3 for a discussion of the leak before-break application and criteria applicable to the main steam supply system. Subsection l

) 3 states: " As noted in Subsection 5.2.5, the rated capability of the leak detection system for the primary

. wiant inside containment is 0.5 gpm in one hour. This system also detects leakage of 0.5 FPm from the main steam and feedwater lines inside containment.

Main steam line leak detection inside containment is provided by the following monitored parameters that are indicated in the main control room:

Containment sump level monitor Containment air cooler condensate flow monitor ,

Containment humidity Containment atmosphere temperature The leak detection procedure is to set an alarm setpoint at 0.5 gpm in one hour on the sump level monitor. As explained in SSAR Subsection 5.2.5. "The sensitivity of the [containme.it sump] level sensors allows detection of leakage as low as 0.5 gpm within one hour." When the alarm actuates, the operator reviews other monitors (e.g.,

containment temperature and humidity, air cooler condensate flow) to determine the source of the leakage.

Appropriate actions are then taken according to the technical specifications.

Technical Specification B.3.4.7.b, "RCS Operational Leakage " supports leak-before-break for piping 4" and greater; it requires tiiat if leakage greater than 0.5 gpm is detected and cannot be corrected in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the plant must enter LCO 3.0.3 immediately. That technical specification was written specifically for reactor coolant system piping, but the requirements are appropriate for detection of main steam line leakage inside containment. A similar technical specification, w ritten specifically to support leak-before-break for secondary side leakage, will be added, requiring that if secondary side leakage greater than 0.5 gpm is detected and cannot be corrected in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , the plant must enter LCO 3.0.3 immediately.

' ' #5' W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

!I SSAR Revision:

An additional Technical Specification to support LBB for the Main Steam Line piping, will be added in Rev. 2 of the SSAR.

PRA Revision: NONE i

410.145 2 Vj Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question 410.160 Section 9.31.3 of the SSAR describes the use of safety-grade air accumulators or other devices to provide short-term operation of the safety Vated pneumatic salves following loss of air.

a. Explain w hat are the "other device."

b. Provide a list of all of the safety-related pneumatically operated valves (required to change valve position to achieve safe shutdown and accident mitigation) that are furnished with safety-related backup air accumulators and'or other devices,

c. How will the adequacy and reliability of the safety-related backup accumulators and or other devices be ensured? NUREG-1275 Vol. 2 recommends (1) periodic testing of safety-grade backup accumulator check salves for leakage; (2) monitoring and/or alarming accumulator pressme; and (3) serifying the adequacy of safety-related accumulators.

Response-a) SSAR section 9.3.1.3 is reworded as shown below. The words "other desices" have been deleted.

b) There are no safety-related air operated salses that have safety-related air accumulators to support their safety-related functions. There are seseral safety-related valves that have N3 stored in accumulators or inside the s ais e operator. The main steam isotation valves (MSIV) and the main feedw ater isolation s alves (MFIV) store N;. The automatic depressurization system ( ADS) 4th stage valves (if piston operated salves are used) store N; in separate accumulators.

Both the main steam isolation valves and main feedwater isolation valves have pneumatic / hy draulic operators.

The stored energy for closing is supplied by high pressure nitrogen stored in the valve operator. The valve is maintained in its normally open position by high pressure hydraulic fluid that opposes the N2 P'"h*" I"'

emergency closure, redundant safety-related, Class IE solenoid valves are energized by separate safety-related Class IE power sources to dump the high pressure hydraulic fluid to a fluid reservoir.

The AP600 design change report transmitted to the NRC on 2/15/94 indicates that piston-operated gate valves are one of the valve types being considered for the automatic depressurization system fourth stage. These automatic depressurization system salves are normally closed " fail as is" type valves. To perform their safety function they are opened by energizing redundant safety-related, Class lE solenoid valves to align the nitrogen supply stored in separate N; accumulators and to close the normally open vent under the piston while the area abos e the piston is kept sented (see response to RAI 410.162). The safety-related Class IE solenoid valves are energized by redundant safety-related Class IE power sources.

c) The adequacy and reliabWity of the safety-related N; supplies (accumulators / operators) will be provided by:

integrated N; supply (accumulator / operator / piping) leak test during each refueling; 4' 6

W westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION N, supply pressure monitoring and alarm; verification of the adequacy of the N, supply during plant startup testing.

SSAR Revision:

The second paragraph of Subsection 9.3.1.3 will be resised as follows:

The compressed and instrument air system is required for normal operation and startup of the plant.

Pneumatically operated s alves in the plant, which are essential for safe shutdown and accident mitigation, are either designed to actuate to iail-safe position upon loss of air pressure A.-- - fc ' - '->cd, p: -' . "; rp; - * ' A

= f;- ' ' d a ' '-> ' : g a' ~- - - pr; ided -fer, grcd :!+

mm:=!

A ::ge p; ' ' a: .

awnslutwwwide-ped!c re!!::5!c '- -> 'er:w;x - ~'~ e c ch . . fe!!:: ing !;: of a: . or are provided with safety-related air accumulators to provide the air supply for the safety-relate 41 function. Such pneumatically operated valves utilize safety-related solenoid valves to control the air supply. The air accumulators

d -5

'- if required, are included in the system containing the safety-related valves.

410.160-2 W-Westinghouse

~

NRC REQUEST FOR ADDITIONAL INFORMATION Question 410.162 Sec tion 6.3.2.2.7.6 of the proprietary sersion of the SSAR discusses the use of backup salety-related air accumulators for the fourth stage ADS salves. Much of this type of information is typically found in non-proprietary sersions of other SARs. Therefore, revise the SSAR to address and incorporate the following:

a. Include the information provided in the proprietary sersion of Section 6.3.2.2.7.6, regarding the backup safety-related air accumulators for the fourth stage ADS valves, in the text of the non-proprietary version of the SS A R .

h. Include these backup safety-related accumulators in a non-proprietary figure.

c. Revise the SSAR to include information about (1) leak testing the accumulators. (2) seismic qualification of the accumulators, (3) the ability of the accumulators to open the salves against maximum containment pressure.

(4) the capacity of the accumulators. and (5) testing of the accumulators in accordance to RG l.68.3.

d. IE Bulletin 80-01 concerns the operability of the pneumatic supply for ADS valves for licensees of GE HWR facilities. However, the bulletin may be relevant to the AP600 design regarding the use of backup safety-related air accumulators. Do the AP600 backup safety-related air accumulators conform with IE Bulletin No. 80-0l?

Response

a) Westinghouse is discussing the general issue of proprietary classification with the NRC staff. When those discussions are completed the SSAR will be revised as appropriate.

b) The Al%00 design change report transmitted to the NRC on Febuary 151994 indicates that piston operated gate salses are one of the valve types being considered for the automatic depressurization system stage four valves.

In the event that a piston operated gate valve is used, a backup safety-related N2 accumulator configuration would be provided as shown in the attached sketch.

c) The design of the backup accumulators would provide:

1) ne accumulator / line would be leak tested each refueling (pressure decay).

2) The accumulator and the connected lines / salves up to the check isolTion valve would be safety-related, seismic I design.

3) The accumulators have a requirement to open/close the automatic depressurization system stage valve two times with the containment at pressures as high as 45 psig. Note that N2 si supplied at a pressure of about 350 psig.

4) The estimated accumulator capacity is about 200 gallons. Note that this capacity is dependent on the vendor specific valve / operator design.

5) The accumulator wadd be tested in accordance with the Regulatory Guide 1.68.3.

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION di The backup safety-related N, accumulators will conform with IE Bulletin No. 804)l. in particular:

isolation check salve type will be chosen to minimize the potentialleakage; integrated accumulator / operator leak test will be performed during each refueling; the gas accumulator, associated solenoids valses, isolation check valve and piping are Seismic I; Technical Specification will be defined to address the automatic depressurization system stage 4th stage operability.

SSAR Revisic.c NONE i

1 l

1 4

1 1

410.162-2 l W ~

Westinghouse l 1

l 1

NRC REQUEST FOR ADDITIONAL INFORMATION as sa m

BACKUP SAFETY-RELATED ACCUMULATOR CONFIGURATION

5 (6 W W

ee - ( . s. s T T b-VEhi vtN?

f Q NITROCEN SUPPLY H f m

+ A 02 L M to other 4th

"' 9'

COMPART r , v 4TH STAGE VALVE ACCLMULATOR 4TH STAGE VALWS SOLENOID VALVE DMSiON ASSIGNMENT

1 #2 #3 #4 #5 #6 V004A & B A B A B A B V004C & D C D C D C D 4e solenoid volves are shown in their normal positions, whicn are de--energized.

To open tne 4th stage ADS vofve, solenoids #1, 2, 3 & 4 ore energized which chgns the nitrogen supply to under the pistion and keeps tne oreo obove tne pistion vented.

To close the 4th stage ADS volver, solenoids #5 & 6 are energized which oligns the nitrogen supply to over the pistice end keeps the creo under the volve vented.

l 3

i 410.162-3 i W

Westinghouse !

1 i

i e

NRC REQUEST FOR ADDITIONAL INFORMATION i l Question 410.196 loclude the table prosided in the March 18. 1993, response to Q410.27 that lists the safety-related equipment requiring flood protection in the appropriate section of the SSAR. In addition, include the information in the February 9,1993, response to Q435.56 regarding flood protection for I&C equipment in Section 3.4.1 of the SSAR.

Include the caseats regarding information not in the table (i.e.. regarding safety-related equipment above the maximum flood lesel and passive components).

Response

The environmental qualification of safety-related equipment is addressed in SSAR Section 3.11. SSAR Table 3.11-1 lists all safety-related electrical and mechanical equipment and their environmental zones (room numbers). The data in the response to RAI 410.27 was prosided for information and prosides a level of detail beyond that necessary for the SSAR. The check valses and relief valves show n in the response to RAI 410.27 are being added to the list of actise valves in SSAR table 3.11-1.

SSAR Section 3.4.1.1.2 will be resised to address RAI 435.56.

SSAR Revision:

The last paragraph in SS AR Section 3.4.1.1.2 and Table 3.11-1 will be revised as follows:

The AP600 arrangement provides physical separation of redundant safety-related components and sy stems from each other and from nonsafety-related components. As a result, component failures resulting from internal flooding will not present safe shutdown of the plant or present mitigation of the flooding esent. Protection mechanisms are described in Sections 3.6 and 3.11. The protection mechanisms related to minimizing the consequences of internal flooding include the following.

Structural enclosures

Structural barriers

Curbs and elevated thresholds a leak detection systems

Drain systems

Equipment qualification.

0. m W- WestinEhouse

1

. NRC REQUEST FOR ADDITIONAL INFORMATION i

f Table 3.11-1 (Sheet 18 of 28)

Safety-Related Electrical and Mechanical Equipment Operating Envir, Time ;

~

APh00 Zone Function Requim!

Dewription Tag No. (Note 2) (Note I) (Note 5)

ACTIVE DAMPERS: l MCR ISOLATION DAMPERS VBS MD D214 12401 ESF 1 YR MCR ISOLATION DAMPERS VBS MD D215 12401 ESF I YR MCR ISOLATION DAMPERS VBS MD D216 12401 ESF 1 YR '

MCR ISOLATION DAMPERS VBS MD D217 12401 ESF 1 YR MCR ISOLATION DAMPERS VBS MD D220 12401 ESF I YR ,

MCR ISOLATION DAMPERS \ BS MD D221 12401 ESF 1 YR PENETRATION %

PEN ETRATIONS IMECH ANICAL) SEE TABLE 6.2.3-1 PENI TR ATIONS (ELECTRICAL) SEE FIGURE 3 8.2-4 ACTIVE VALVES: ;

CONTAINM ENT ISO INLET CCS PL V200 12306 ESF 5 MIN i LIMIT SWTTCH (CLOSED) CCS PL V200 LC 12306 PAMS 2 %KS i LIMIT SWITCH (OPEN) CCS PL V200-LO 12306 PAMS 2 WKS MOTOR OPERATOR CCS PL V200-M 12306 ESF 5 MIN CONT AINM ENT ISO-OUTLET CCS PL V207 11300 ESF 5 MIN .

LIMIT SWITCH (CLOSED) CCS PL V207-LC 11300 PAMS I YR i LIMfT SWITCH (OPEN) CCS PL V207-LO 11300 PAMS I YR MOTOR OPERATOR CCS PL V207-M 11300 ESF 5 MIN CONTAINMENT ISO OUTLET CCS PL V208 12306 ESF 5 MIN ,

LIMIT SWTTCH (CLOSED: CCS PL V208-LC 12306 PAMS 2 WKS !

LIMIT SWTTCH (OPEN) CCS PL V208-LO 12306 PAMS 2 WKS !

MOTOR OPERATOR CCS PL V208-M 12306 ESF 5 MIN ,

RCS LETDOWN STOP VALVE CVS PL V001 11303 ESF 5 MIN )

LIMTT SWrTCH CVS PL V001-L 11303 PAMS 1 YR MOTOR OPERATOR CVS PL V001-M 11303 ESF 5 MIN RCS LETDOWN STOP VALVE CVS PL V002 11303 ESF 5 MIN i LIMir SWTTCH CVS PL V002-L l1303 PAMS 1 YR l MOTOR OPERATOR CVS PL V002-M 11303 ESF 5 MIN DEMIN FLUSH LINERELIEF CVS PL V042 I1209 ESF: 5 MIN WLS LETDOWN IRC ISOLATION CVS PL V045 11300 ESF 5 MIN SOLENOID V ALVE CVS PL V045-S !!300 ESF 5 MIN LETDOWN FLOW ORC ISO g CVS PL V047 12256 ESF 5 MIN LIMIT SWrTCH (CLOSED) CVS PL V047-LC 12256 PAMS 2 WKS LIMTT SWITCH (OPEN) CVS PL V047-LO 12256 PAMS 2 WKS LETDOWN LINE RELIEF CVS PL V056 ' 11300 ,ESF 5 MIN 410.196-2 W-Westingttouse

NRC REQUEST FOR ADDITIONAL INFORMATION Table 3.11-1 (Sheet 19 of 28)

Safety-Related Electrical and Mechanical Equipment

()perating Envir. Time APbOO Zone Furwtion Requirni Dev ription Tag No. (Note 2) (Note ll (Note 5)

RCS CilARGINO STOP V ALVE CVS PL V081 11303 ESF 5 MIN SOL LNOID V ALVE CVS PL V081 S 11303 ESF 5 MIN M AKLL'P LINE CONT ISOLATION CVS PL V0W 12256 ESF 5 MIN LIMir SWITCll (CLOSED) CVS PL V04 LC 12:56 PAMS 2 WKS LINirr 5% ITC11 IOPENi CVS PL V090-LO 12:56 PAMS 2 WKS MOTOR OPERATOR CVS PL V04M 12256 ESF 5 MIN M AKEUP LINE CON'l .L ATION CVS PL V091 l1300 ESF 5 MIN L IMIT swr!Cil 'CLbd.Di CVS PL V091-LC i1300 PAMS I YR LIMIT SWTICil (OPEN) CVS PL vo91-LO 11300 P A.M S I YR MOTOR OPERATOR CVS PL V091-M 11300 ESF $ MIN llYDROGEN ADDITION CONT ISO CVS PL V092 12256 ESF 5 MIN 1.lMrT SWrTCH (CLOSED) CVS PL V092 LC 12256 PAMS 2 WKS LIMIT SWFTCil (OPEN) CVS PL V092-LO 12256 PAMS 2 WKS 501 ENOID VALVE CVS PL V092-S 12:56 ESF 5 MIN DEMIN WATER SYS ISOLATION CV5 PL V136A 12:55 ESF 5 MIN LIMir SWITCif CVS PL V136A-L 12255 PAM5 2 WKS MOTOR OPERATOR CVS PL Vl36A-M 12:55 ESF 5 MIN DEMIN WATI'R SYS ISOLATION CVS PL Vl36B 12255 ESF 5 MIN LIMIT SWrTCH CVS PL V1368 L 12255 PAMS 2 WKS MOTOR OPER ATOR CV5 PL V136B M 12255 ESF 5 MIN PXS M AKEUP LINE CONT ISO CVS PL Vl71 12454 ESF 5 MIN VALVE 501.ENOID V ALVE CVS PL V171-S 12454 ESF 5 MIN PCCWST 150LATION VALVE PCS PL V001 A 12701 ESF 5 MIN LIMir 5%TTCH (CLOSED) PCS PL V001 A-LC (2701 PAMS WKS LIMir 5 WITCH iOPENi PCS PL V001 A-LO 12701 PAMS 2 WKS SOLENOID V ALVE PCS PL V001 A-S 12701 ESF 5 MIN PCCWST ISOLATION V ALVE PCS PL V0018 12701 ESF 5 MIN LIMIT SWTTCH iCLOSEDI PCS PL V00lB-LC 12701 PAMS 2 WKS LIMir SWTTCH (OPEN) PCS PL V00lB Lo 12701 PAMS 2 WKS SOLENOID V ALVE PCS PL V00lB-S 12701 ESF 5 MIN PCCWST ISOLATION VALVE PCS PL V002A 12701 ESF 5 MIN llMir SWITCH i.CLOSEDi PCS PL V002 A LC 12701 PAMS 2 WKS LIMIT SWTTCH (OPEN) PCS PL V002A-LO 12701 PAM5 2 WKS MOTOR OPI RATOR PCS PL V002A M 12701 ESF 5 MIN PCCWST ISOLATION V ALVE PCS PL V002B 12701 ESF 5 MIN LIMrf SWITCH (CI.OSEDi PCS PL V002B LC 12701 PAMS 2 WKS

PCS PL V0028-LO PAMS 2 WKS LIMIT 5% FTCH (OPEN) 12701 MOTOR OPERATOR PCS PL V002B M 12701 ESF 5 MIN CONT ISOL-AIR S AMPLE LINE PS5 PL V008 11300 ESF 5 MIN CONT ISOL- AIR S AMPLE LINE PSS PL V009 11300 ESF 5 MIN W

Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION i

Table 3.11-1 (Sheet 20 of 28) )

Safety-Related Electrical and Mechanical Equipment Operating ;

Envir, Time AP600 Zone Functius Required Dm ription Tag No. (Note 2l (Note 1) (Note 5)

CONT ISOL-LlQ SAMPLE LINE PSS PL V010A i1300 ESF 5 MIN CONT ISOL-LIQ S ANf PLE LINE PSS PL V010B 11300 ESF 5 MIN CONT ISOL-LIQ SAMPLE LINE PSS PL V0ll 12354 ESF $ MIN CONT ISOL-S AMP 1 E R ETIIRN PSS PL V023 12354 ESF 5 MIN LINE CONT ISOL- AIR S ANf PLE LINE PSS PL V046 12454 ESF 5 MIN CMT A CL INLET ISOLATION PXS PL V002A 11300 ESF $ MIN LIMIT SWITCH (CLOSEDi PXS PL V002A-LC 11300 PAMS I YR t IMIT SWITCll (OPEN PXS PL V002 A LO 11300 PAMS 1 YR SOLENOID V ALVE PXS PL V000A-S 11300 ESF 5 MIN CMT !! CL INLET ISOLATION PXS PL V002B 11300 ESF 5 MIN LIMTT SWTTCH iCLOSED) PXS PL V002B LC i1300 PAMS I YR LIMIT SwTTCil iOPEN) PXS PL V002B-LO I1300 PAMS I YR SOLENOID VALVE PXS PL YOO2B S l1300 ESF 5 MIN CMT A CL INLET ISOLATION PXS PL V003 A 11300 ESF 5 MIN LIMIT S%TTCli (CLOSED) PXS PL ST)03 A-LC i1300 PAMS I YR LINf!T SWITCil (OPEN) PXS PL V003A-LO I1300 PAMS I YR SQL ENOID VALVE PXS PL V003 A S 11300 ESF 5 MIN CMT B Cl INLET ISOLATION PXS PL V003B 11300 ESF 5 MIN LIMIT SWTTCH iCLOSED) PXS PL V003B LC 11300 PAMS I YR LIMIT SWTTCII IOPEN) PXS PL V003B-LO 11300 PAMS 1 YR SOLENO!D VALVE PXS PL V0038 S I1300 ESF 5 MIN CMT A PZR LINE ISOLATION PXS PL V005A 12503 ESF 5 MIN LIMTT SWITCH (CLOSED) PXS PL V005 A-LC 12503 PAMS 2 WKS LIMIT SWTTCli tOPEN) PXS PL V005 A Lo 12503 PAMS 2 WKS MOTOR OPERATOR PXS PL V005 A-M 12503 ESF 5 MIN CMT B PZR LINE ISOLATION PXS PL V005B 12503 ESF 5 MIN LIMIT SWITCH (CLO5LD) PXS PL V005B LC 12503 PAMS 2 WKS LIMIT SWTTCH (OPEN) PXS PL V005B LO 12503 PAMS 2 WKS MOTOR OPERATOR PXS PL V005B M 12503 ESF 5 MIN CMT A DISCH ARGE ISOLATION PXS PL V014 A 11206 ESF 5 MIN LIMIT SWTTCH (CLOSED) PXS PL V014 A-LC 11206 PAMS I YR LIMIT SWTTCH tOPEN) PXS PL V014 A-LO 11:06 PAMS 1 YR SOLENOID VALVE PXS PL V014A-S I1206 ESF 5 MIN CMT B DISCilARGE ISOLATION PXS PL V014B 11207 ESF 5 MIN LIMTT SWITCil(CLOSED) PXS PL V0148 LC 11207 PAMS I YR LIMIT SWTTCH iOPENI PXS PL V0148-LO 11207 PAMS I YR SOLENOID VALVE PXS PL V014B-S 11:07 ESF 5 MIN CMT A DIScil ARGE ISOLATIOY PXS PL V015 A 11:06 ESF 5 MIN LlMTT SWTTCilICLOSEDi PXS PL V015 A LC 11206 PAMS I YR LIMTT SWTTCll (OPEN) PXS PL V015 A-LO 11206 PAMS I YR 410.196-4 W Westinthouse b

NRC REQUEST FOR ADDITIONAL INFORMATION A

Table 3. Il-1 ISheet 21 of 28)

Safety-Related Electrical and Mechanical Equipment Operating Envir. Tirne APfd)0 Zone Function Requiral D w riptinn Tag No. (Note 2) (Note 1I (Note 5)

SOLENOID VALVE PXS PL V015 A-S !!206 ESF 5 MIN CMT B DISCll ARGE ISOLATION PXS PL V015B i1207 ESF 5 MIN LIMir SWTTCH (CLOSED) PXS PL V015B LC i1207 PAMS I YR LIMIT SWITCil iOPEN) PXS PL V015B LO 11:07 PAMS I YR SOLENOID V ALVE PXS PL VOISB S 11207 ESF 5 MIN CMT A DISCFIARGE PXS PL V016A- !!20e ESF 5 MIN CMT B DISCH AROE PXS PL V016B 11207 ESF 5 MIN CMT A DISCHARGE PXS PL V017A 11206 ESF 5 MIN CMT B DIScilARGE PXS PL V017B 11207 ESF 5 MIN ACCUM A DISCH ARGE ISOL PXS PL V027A 11:06 ES F 5 MIN LIMIT 5%TTCil (CLOSED) PXS PL V027A-LC i1200 PAMS I YR LIMIT SWITCH (OPEN) PXS PL V027A-LO 11206 PAMS I YR MOTOR OPER ATOR PXS PL V027A.M i120e ESF $ MIN ACCUM B DISC}l ARGE ISOL PXS PL V027B 11207 ESF 5 MIN LIMIT SWITCil iCLOSED) PXS PL V027B LC 11:07 PAMS 1 YR LIMIT SWITCH (OPEN) PXS PL V027B-LO 11:07 PAMS 1 YR MOTOR OPERATOR PXS PL V0278-M 11207 ESF 5 MIN ACC A DISCHARGE FXS PL V028A 11206 ESF 5 MIN ACC B DISCHARGE PXS PL V023B 11207 ESF 5 MIN ACC A DISCif ARGE PXS PL V029A 11206 ESF 5 MIN ACC B DISCHARGE PXS PL V029B 11207 ESF 5 MIN CMT A STEAM TRAP BYPASS PXS PL V030A 11300 ESF 5 MIN ISOLATION CMT B STEAM TRAP BYPASS PXS PL V030B !!300 ESF 5 MIN ISOL AT'ON CMT A STEAM TRAP BYPASS PXS PL V031 A 11300 ESF 5 MIN ISOL ATION CMT B STEAM TRAP BYPASS PXS PL V03 t B 11300 ESF 5 MIN ISOLATION CMT A STEAM TRAP DISCH ISOL PXS PL V033 A i1300 ESF 5 MIN CMT B STEAM TRAP D15Cil ISOL PXS PL V033B 11300 ESF 5 MIN ORC NITROGEN SUPPLY CONT PXS PL V042 12306 ESF 5 MIN ISOLATION LIMIT SWITCil tCLOSED) PXS PL V042-LC 12306 PAMS 2 WKS LIMfr SWTTCll (OPEN) PXS PL V042-LO 12306 PAMS 2 WKS SOLENOID V ALVE PXS PL V042 5 1230c ESF 5 MIN PRHR HX INI.ET ISOLATION

PXS PL V101 11500 ESF 5 MIN LIMrT 5%TTCH iCLOSED) PXS PL V101 LC 11500 PAMS 1 YR LIMTT SWITCH (OPEN) PXS PL V101-LO 11500 PAMS i YR MOTOR OPER ATOR PXS PL V101-M 11500 ESF 5 MIN W-Westinghouse o

NRC REQUEST FOR ADDITIONAL INFORMATION i&

Table 3.11-1 (Sheet 22 of 28)

Safety-Related Electrical and Mechanical Equipment Operating Envir. Time Al%00 Zone Function Required 1)ew ription Tag No. (Note 2 p (%te ll iNote 5)

PRliR HX DIScil ARGE 150L PX5 PL V108 A i1300 ESF 5 MIN LIMrr SWITCH iCLost Di PXS PL V108A-LC 11300 PAMS I YR I IMIT SWITCH <OPEN) PXS PL V108 A LO 11300 PAMS I YR PRHR HX DISCH ARGE 150L PX5 PL V108B 11300 ESF S MIN LIMir SWITCH ' CLOSED PXS PL V108B-LC l1300 PAMS I YR 1 IMIT SWITCH iOPENs PXS PL. V108B LO l1300 PAMS 1 YR RirfRC SUMP A I5OLATION PXS PL Vil7A 11206 ESF 5 MIN LiMir SWITCH PXS PL Vil7A-L 11200 PAMS I YR MOTOR OPLR ATOR PXS PL Vil7A M 11206 ESF 5 MIN RECIRC SUMP ti ISOLATION PX5 PL Vll?B 11:07 ESF 5 MIN LIMir SWFTCll PXS PL Vil7B-L 11:07 PAMS I YR MOTOR OPERATOR PXS PL Vil7B-M 11207 ESF 5 MIN RECIRC SUMP A ISOLATION PXS PL Vil8A 11:00 ESF 5 MIN LIMIT SWFTCH PXS PL V i l 8 A-L 11206 PAMS 1 YR MOTOR OPERATOR PXS PL Vil8 A M 11206 ESF 5 MIN RECIRC Sl'MP B ISOLATION PXS PL VI18B 11207 ESF 5 MIN LIMTT SWITCH PXS PL VI18B-L 11:07 PAMS 1 YR MOTOR OPER ATOR PXS PL VilRB-M 11207 ES F 5 MIN RECIRC SUMP A PXS PL Vil9A 11206 ESF 24 HRS RECIRC SUMP B PXS PL Vil9B 11207 ESF 24 IIRS RECTRC SUMPA PXS PL V120A !!206 ESF 24 HRS RECIRC SUMP B PXS PL V120B 11207 ESF 24 HR$

IRWST/ SUMP GRAV INJ A ISOL PXS PL V121 A Il20e ESF 5 MIN t (Mir SWTICH < CLOSED) PXS PL V12] A-LC 1120e PAMS I YR LIMir SWTTCH rOPEN) PXS PL V121 A-LO 11:06 PAMS I YR MOTOR OPERATOR PXS PL V121 A-M 1120e ESF 5 MIN IRWST/ SUMP GR AV INJ B 150L PXS PL V121B 11207 ESF 5 MIN LIMir 5%TTCH (CLOSED) PXS PL V121B-LC ll207 PAMS I YR i LIMIT 5%TTCH (OPEN) PXS PL V!21B-LO I1207 PAMS I YR l MOTOR OPERATOR PXS PL VI:lB-M 11207 ESF 5 MIN IRWST IN) A PXS PL V122A 11206 ESF 24 HRS I 1RWST INJ B PXS PL V122B 61207 ESF 24 HRS IRWST INJ A PXS PL V123A 1120e ESF 24IIRS IRWST INJ B PXS PL V123B 11207 ESF 24 HRS l IRWST INJ A PXS PL V124A 11206 ESF 24 HRS IRWST INJ D PXS PL V124B 11207 ESF 24 IIRS 1RWST INJ A PXS PL V125A 11206 ESF 24 HRS 1RWST INJ B i PXS PL V125D 11207 ESF 24 HR$

IRWST GLITTER BYPASS A ISOL PXS PL Vl30A 11300 ESF 5 MIN LIMTT SWTTCH # CLOSED) PXS PL Vl30A-LC 11300 PAMS 1 YR LIMir SWTTCH t.OPEN1 PXS PL V130A LO 11300 PAMS I YR 410.196-6 l W westinghause -

1

NRC REQUEST FOR ADDITIONAL INFORMATION r q F ji]

Table 3.11-1 (Sheet 23 of 28)

Safety-Related Electrical and Mechanical Equipment Operating Envir, Tinie APfdH) Zone function Required D vription TagNo. (Note 2) (Note 1I (Note 5)

Sol ENOID VALVE PXS PL V130A S i1300 ESF 5 MIN IRWST GUTTE R BYPASS B ISOL PXS PL V130B 11300 ESF 5 MIN LIMIT 5% rTCH dCLOSED PXS PL V130B-LC !!300 PAMS I YR LIMIT SWFTCH 40 PENS PXS PL V130B LO 11300 PAMS I YR SOLENOlD VALVE PXS PL Vl30B S 11300 ESF 5 MIN CMT A FILL ISOLATION PXS PL V230A !!300 ESF 5 MIN SOLLNOID VALVE PXS PL V230 A S 11300 ESF 5 MIN CMT B flLL ISOLATION PXS PL V230B l1300 ESF 5 MIN SOLENOID VALVE PXS PL V230B S 11300 ESF 5 MIN ACCUM A FILL / DRAIN ISOL PXS PL V232A 11300 ESF 5 MIN SOLENO!D VALVE PXS PL V232A-S 11300 ESF 5 MIN ACCUM B FilliDRAIN ISOL PXS PL V232B ll300 ESF 5 MIN SOLENOID V ALVE PXS PL V232B-S 11300 ESF 5 MIN PH ADJUST TANK DISCil ISOL PXS PL V301 A 11300 ESF 5 MIN Pil ADJUST TANK DIScil ISOL PXS PL V301B I1300 ESF 5 N11N isT STAGE ADS RCS PL V00l A lit 03 ESF 5 MIN LIMir SWITCH (CLOSED) RCS PL V001 A-LC lico3 PAMS I YR LIMTT SWTTCH <OPEN) RCS PL V001 A-LO lit 03 PAMS I YR MOTOR OPERATOR RCS PL V001 A-M l i t03 ESF 5 MIN 15T ST AGE ADS RCS PL V001B l i tO3 ESF 5 MIN LIMIT SWITCll (CLOSEDi RCS PL V00lB LC l if03 PAMS I YR LIMIT SWTTCH IOPEN) RCS PL V00lB-Lo 11603 PAMS I YR MOTOR OPER ATOR RCS PL V00lB-M 11603 ESF 5 MIN IST STAGE ADS RCS PL V00lc 11603 ESF 5 MIN LIMIT SWITCil (CLOSEDI RCS PL V0ulC-LC 1 It03 PAMS 1 YR LIMIT SWTTCll <OPEN) RCS PL V001C Lo 11t43 PAMS I YR MOTOR OPERATOR RCS PL V001C-M i I t03 ESF 5 MIN IST STAGE ADS RCS PL V001D l it03 ESF $ MIN LIMIT SWTTCil ICLOSED RCS PL V00lD-LC t it03 PAMS 1 YR LIMIT SWITCli (OPEN) RCS PL V001D LO libO3 PAMS I YR MOTOR OPER ATOR RCS PL V00lD M 11603 ESF 5 MIN 2ND STAGE ADS RCS PL V002A lit 03 ESF 24 HR LIMTT SWITCH (CLOSED) RCS PL V002 A-LC l it03 PAMS I YR LIMTT SWITCH (OPEN) RCS PL V002A-LO l it03 PAMS I YR MOTOR OPERATOR RCS PL V002A-M l i t03 ESF 24IIR 2ND STAGE ADS RCS PL V002B 11603 ESF 24 HR LIMfi SWTTCH iCLOSED RCS PL V002B-LC 11603 PAMS I YR LIMIT SWrTCH <OPEN) RCS PL V002B LO l i t03 PAMS 1 YR MOTOR OPFRATOR

RCS PL V002B M 11603 ESF 24 HR 2ND ST AGE ADS RCS PL V002C 11603 ESF 24 HR LIMIT SWTTCH (CLOSEDI RCS PL V002C LC l It03 PAMS 1 YR W Westinghouse om l

l

l l

NRC REQUEST FOR ADDITIONAL INFORMATION l

Table 3.11-1 (Sheet 24 of 28)

Safety Related Electrical and Mechanical Equipment Operating Ernir. Time APf 00 Z<me fum tion Requirwl Dew nption Tag No. (Note 2) (Note iI INote Si 1 lMIT 5% frCil (OPI Ni RC5 PL V002C.t O 11003 PAMS 1 YR MOTOR OPI R ATOR RCS PL VO(CC M 11603 ESF 24 lik 2ND STAGE AD5 RCS PL V002D l it03 [~S F 24 HR LIMr! 5%TTril (CL O51 Dj RCS PL VO(CD LC 11603 PAMS I YR I IMir 5 WITCH iOPIN) RC5 PL V00?D LO l i t03 PAMS I YR MOTOR OPER ATOR RCS PL V002D M l it03 ESF 24 IIR 3RD STAGE ADS RCS Pt V003 A l it03 E5F 24 HR LIMir SWTTCil (CLO5ED: RCS PL V003 A LC i1603 PAMS I YR LIMIT SWTrCil (OPEN) RCS PL V003 A LO I It03 PAMS I YR MOTOR OPI RATOR RCS PL V003 A M l1603 EST 24 HR 3RD STAGi: ADS RCS PL V003B l it03 ESF 24 HR LIMri SWriCil WLOSEDi RCS PL V001B LC 11603 PAMS I YR I.lMIT SWTICll 40 PEN) RCS PL V003B LO 11603 PAMS I YR MOTOR OPERATOR RCS PL V003B M l it03 ESF 24IIR 3RD STAGE ADS RCS PL V003C I1603 ESF 24IIR LIMIT swr 1Cil WLOSED) RCS PL V003C-LC 11603 PAMS I YR LIMIT SWrTC}l (OPEN) RCS PL V003C-Lo 11603 PAMS I YK MOTOR OPE R ATOR RCS PL V003C M 11603 ESF 24 HR 3RD STAGE ADS RCS PL V003D 11603 DF 24 HR l.lMTr 5%TTril irl OSI DJ RCS PL V003D LC l it03 PAMS I YR LIMTF SWTTril (OPEN) RCS PL V003D LO 11603 PAMS 1 YR MOTOR OPERATOR RCS PL V003D M 11603 ESF 24flR 4Til STAGE AD5 RCS PL V004 A 11300 bF 24IIR I!MIT SwirCH (CLOSEDi RCS PL V004 A LC 11300 PAMS I YR LIMIT S4TTCH 40 PEN) RCS l'L V004A LO 11300 PAMS 1 YR 4Til ST AGE ADS RCS PL V004B 11300 ESF 24flR 1 IMIT S%TICH (Closi Di RCS PI V004B LC 11300 PAMS I YR llMrT 5%ITCil (OPl.Ni RCS PL V004B LO 11300 PAMS 1 YR 4Til STAGE ADS RCS PL V004C 11300 ESF 24 ilR LIMir SWTTCH iCLOSEDs RCS PL V004C-LC i1300 PAMS I YR LIMir SWrTCH iOPE N> RCS PL V004C LO 11300 PAMS I YR 4Til STAGE ADS RCS PL V004D 11300 DF 24 IIR LIMrT SWilCil iCLOSE.Di RCS PL V004D LC 11300 PAMS 1 YR 1.lMir 5%TTCH iOPEN) RCS PL V004D LO l1300 PAMS I YR PRDS FQU ALIZATION VALVES RCS PL V006A 11301 ESF $ MIN PRESS EQUAllZATION VALVES krS PL V0068 l1301 ESF 5 MIN PRES'i ! Qt! At lZ ATION V ALVES RCS PL V006C l1302 LSF 5 MIN PRESS EQU AllZATION VALVES RCS PL V006D 11302 ESF 5 MIN AD$ TEST 50LENO!D V ALVE

RCS PL V007A l i t03 LSF 5 MIN ADS TLST SOLENOID V Al,VE RCS PL V0G7n 11603 LSF 5 MIN ADS Tl ST SOLENOID V ALVE RC5 PL VOCC ! ! f 03 LSF 5 MIN 410.196-8 W-Westinghouse

NRC REQUEST FOR ADDITibNAL INFORMATION

= =E E 5 ,

Table 3.11-1 (Sheet 25 of 28) ;

I Safety-Related Electrical and Mechanical Equipment >

.I Operating Envir. Time APh00 Zone Function Required Description Tag No. (Note 2) (Note 1) (Note 5)

ADS TEST SOLENOID V ALVE RCS PL V007D 11603 ESF 5 MIN RCS HEAD VENT ADS VALVE RCS PL Vl52 11300 ESF 5 MIN RCS HEAD VENT ADS VALVE RCS PL V153 11300 ESF 5 MIN RCS INNER SUCTION ISOLATION RNS PL V001 A 11208 ESF 5 MIN LIMIT SWrTCH RNS PL V001 A L 11208 PAMS 1 YR MOTOR OPERATOR RNS PL V001 A-M 11208 ESF 5 MIN RCS INNER SUCTION ISOLATION RNS FL V001B Il208 ESF 5 MIN LIMIT SWITC}l RNS PL V001B-L 11208 PAMS I YR MOTOR OPERATOR RNS PL V00lB-M 11:00 ESF 5 MIN RCS OUTER SUCTION ISOL RNS PL V002A 11208 ESF 5 MIN LIMIT SWITCH RNS PL V002A-L 11:08 PAMS 1 Yk MOTOR OPERATOR RNS PL V002A41 11208 ESF 5 MIN ,

RCS OUTER SUCTION ISOL RNS PL V0028 11208 ESF 5 MIN LIMIT SWITCH RNS PL V002B.L i1208 PAMS 1 YR MOTOR OPERATOR RNS PL V0028-M 11208 ESF 5 MIN RilR CONTROlJISOL VALVE RNS PL V0li 12253 ESF 5 MIN i LIMIT SWirCH RNS PL Voll-L 12253 PAMS 2 WKS MOTOR OPERATOR RNS PL V0ll-M 12253 ESF 5 MIN RNS DISCHARGE CONT ISOL RNS PL V013 - 'i1206 ESF- ' 5 MIN RNS DISCHARGE RCPB ISOL . ' RNS PL V015A 11206 ESF 5 MIN RNS DISCH ARGE RCPB ISOL ' RNS PL V015B . 11207 = ESF' 5 MIN RNS DISCHARGE RCPB 150L- ' RNS PL V017A . 11206.. LESF- 5 MIN l RNS DISCHAROE RCPB ISOL - RNS PL VIOTB - 11207' ESF. 5 MIN RNS HOT LEO SUCTION RELIEF RNS PL V021 - 11206 ESF- 5 MIN RHR PUMP SUCTION HDR ISOL RNS PL V022 12253 ESF 5 MIN LIMIT SWITCH RNS PL V022 L 12253 PAMS 2 WKS MOTOR OPERATOR RNS PL V022-M 12253 ESF 5 MIN IRWST SUCTION LINE ISOL RNS PL V023 11208 ESF 5 MIN LIMIT SWITC11 RNS PL V023-L 15.08 PAMS 1 YR MOTOR OPERATOR RNS PL V023-M 11208 ESF 5 MIN ,

CONTAINMENT ISOLATION SFS PL V034 11206 ESF 5 MIN LIMIT SWITCH SFS PL V034-L 11:06 PAMS 1 YR

MOTOR OPERATOR SFS PL V034-M 11206 ESF 5 MIN CONTAINMENT ISOLATION SFS PL V035 12354 ESF 5 MIN ;

LIMIT SWITCH SFS PL V035-L 12354 PAMS 2 WKS MOTOR OPERATOR SFS PL V035-M 12354 ESF 5 MLN CONTAINMENT ISOLATION

SFS PL V038 12354 ESF 5 MIN LIMIT SWITCH SFS PL V038 L 12354 PAMS 2 WKS i MOTOR OPERATOR SFS PL V03841 12354 ESF 5 MIN )

SFS SUCTION THERMAL RELIEF . ETE PL V048 '11206 . ESP .5 MIN )

4' 88~8 T westinghouse 1

i o

NRC REQUEST FOR ADDITIONAL INFORMATION Table 3.11-1 (Sheet 26 of 28)

Safety-Related Electrical and Mechanical Equipment Operating Envir. Time AIWHi Zone function RequirH) den ription Tag No. (Note 2) (Note 1) (Note 5)

PORV BLOCK V ALVE SGS PL V027A 12406 ESF 5 MIN LIMir 5% fTril (CLOSED SGS PL V027A LC 12406 PAMS 2 WKS LINlrT SWriCil 40 PEN) SGS PL V027A LO 12406 PAMS 2 WKS MOTOR OPERATOR SGS PL V027A-M 12406 ESF 5 MIN PORV Bl OCK V ALVE SGS PL V027B 12404 ESF 5 MIN LIMIT SWITCil (CLO5ED) SGS PL V027B LC 12404 PAMS 2 WKS LINlrr SWTTCH tOPEN 505 PL V02i B-LO 12404 PAMS 2 W KS MOTOR OPERATOR SGS PL V027B M 12404 ESF 5 MIN STEAM LINE COND DR AIN ISOL SGS PL V036A 12406 ESF $ MIN SOLENOID V ALVE SGS PL V036A-5 12406 E5F 5 MIN STEAM LINE CONDENS ATE ISOL- SGS PL V0368 12404 ESF 5 MIN SOLENOID V ALVE SGS PL V036B-S 12404 ESF 5 MIN M AIN 5 TEAM LINE 150LATION SGS PL V040A 12406 ESF 5 MIN LIMIT SWITCH (CLOSED) SGS PL V040A.LC 12406 PAMS 2 WKS LIMrr SWTrCH (OPEN SGS PL V040A-LO 12406 PAMS 2 WKS M AIN STEAM LINE ISOLATION SGS PL V040B 12404 ESF 5 MIN LIMIT SWITril (CLOSED: SGS PL V0408 LC 12404 PAMS 2 WKS LIMIT SWirCil (OPEN) SGS PL V040B-LO 12404 PAMS 2 WKS M AIN FEEDW ATER ISOLATION SGS PL V057A 12406 ESF 5 MIN LIMrr $WirCil (CLO5ED: SGS PL V057A LC 12406 PAMS 2 WKS LIMri SWITCil (OPEN p 5GS PL V057A-LO 12406 PAMS 2 WKS M AIN FEEDWATER ISOL ATION SGS PL V0578 12404 ESF 5 MIN LIMIT SWTTCH (CLOSED: SGS PL V057B LC 12404 PAMS 2 WKS LIMir SWITCil (OPEN) SGS PL V057B LO 12404 PAMS 2 WKS STARTUP F EEDWATLR ISOL SGS PL V067A 12406 ESF $ MIN LIMrT SWITCH SGS PL V067A L 12406 PAMS 2 WKS MOTOR OPERATOR SGS PL V067A-M 12406 ESF 5 MIN STARTUP F EEDWATER 150L SGS PL V0678 12404 ESF $ MIN LIMIT SWTTCH SGS PL vo67B-L 12404 PAMS 2 WKS MOTOR OPERATOR SGS PL V067B M 12404 ESF 5 MIN 50 BLOWDOWN ISOLATION SGS PL V074A 12306 ESF 5 MIN L IMIT SWirCH < Closed) SGS PL V074A LC 12306 PAMS WKS LIMIT SWirCil tOPEN) SGS PL V074A LO 12306 PAMS 2 WKS SOLENOID V ALVE SGS PL V074A S 12306 ESF 5 MIN 50 BLOWDOWN ISOLATION SGS PL V074B 12306 ESF 5 MIN LIMIT SWITril tCLOSED SGS PL V074B LC 12306 PAMS 2 WKS LIMIT SWirCil (OPEN)

SGS PL V074B-LO 12306 PAMS 2 WKS SOLENOID V ALVE SGS PL V074B-S 12306 ESF $ MIN SG SERIES BLOWDOWN 150L SGS PL V075 A 12306 ESF 5 MIN LIMIT SWITCil tCLOSED 5GS PL V075A LC 12306 PAMS 2 WKS 410.196-10 W-Westinghouse

~

f NRC REQUEST FOR ADDmONAL INFORMATION Table 3.11-1 (Sheet 27 of 28)

Safety-Related Electrical and Mechanical Equipment Operating Envir. Time APMM) Zone Function Requimt Dew ription Tag No. (Note 2) (Note 1) (Note 5)

LIMTT SWTTril (OPEN) SGS PL V075 A Lo 12306 PAMS 2 WKS SOLENOID VALVE SGS PL V075A-S 12306 ESF 5 MIN SG SERIES BLOWDOWN ISOL SGS PL V075B 12306 ESF 5 MIN LIMIT SWITCil (CLOSEDi SGS PL V075B-LC 12306 ESF 5 MIN LIMIT SWTTCll (OPEN) SGS PL V075fLLO 12306 PAMS 2 WKS SOLENOID V AI.VE SGS PL V075B-S 12306 ESF 5 MIN STEAM LINE COND DRAIN ISOL SGS PL V086A 12406 ESF 5 MIN SOLENOID V ALVE SGS PL V086A-S 12406 ESF 5 MIN STEAM LINE COND DRAIN ISOL SGS PL V086B 12404 ESF 5 MIN SOLENOID VALVE SGS PL V0868-S 12404 ESF 5 MIN PWR OPERATED RELIEF V ALVE SGS PL V233A 12406 ESF 5 MIN LIMir SWTTCil (CLOSED) SGS PL V233 A-LC 12406 PAMS 2 WKS LIMIT SW TTCil (OPEN) SGS PL V233 A-LO 12406 PAMS 2 WKS PWR OPERATED RELIEF VALVE SGS PL V233B 12404 ESF 5 MIN LIMIT SWITril(CLOSED) SGS PL V233B-LC 12404 PAMS 2 WKS LIMir SWTTCil (OPEN) SGS PL V233B LO 12404 PAMS 2 WKS MSIV BYPASS ISOLATION VALVE SGS PL V240A 12406 ESF 5 MIN LIMir SWTTCll ICLOSED) SGS PL V240A-LC 12406 PAMS 2 WKS LIMfT SWITCil <OPEN) SGS PL V240A LO 12406 FAMS 2 WKS MSiv BYPASS ISOLATION V ALVE SGS PL V2408 12404 ESF $ MIN LIMir SWITCil < CLOSED) SGS PL V240B LC 12404 PAMS 2 WKS LIMIT SWITCil (OPEN SC.S PL V2408 LO 12404 PAMS 2 WKS M AIN TEEDWATER CONT VLC SGS PL V250A 20400 ESF 5 MIN LIMTT SWITCil CLOSED) SGS PL V250A LC 20400 PAMS 2 WKS LIMTT SWITCII (OPEN) SGS PL V250A-LO 20400 PAMS 2 WKS M AIN f EEDWATER CONT VLV SGS PL V250B 20400 ESF $ MIN LIMIT SWTTCil (CLOSED) SGS PL V250B-LC 20400 PAMS 2 WKS LIMIT SWITCil q0 PEN) SGS PL V250B-LO 20400 PAMS 2 WKS ST ARTUP FEEDWTR CONT VLV SGS PL V255 A 12406 ESF 5 MIN STARTUP FEED %TR CONT VLV SGS PL V255B 12404 ESF 5 MIN PRESSURE REG V ALVE A VES PL V002A 12401 ESF 5 MIN PRESSURE REG V ALVE B VES PL V002B 12401 ESF 5 MIN ACTU ATION VALVE A VES PL V005 A 12401 ESF 2 WKS ACTU ATION VALVE B VES PL V0058 12401 ESF 2 WKS OUTLFT ISOLATION VALVE WGS PL V051 12155 ESF $ MIN RCDT CONTAINMENT ISOL IRC WLS PL V004 11300 ESF 5 MIN SOLENOID V ALVE WLS PL V004-5 11300 ESF 5 MIN RCDT CONT AINMENT ISOL ORC WLS PL V006 12256 ESF 5 MIN SUM P CONT AINM ENT ISOL IRG WLS PL V055 11300 ESF 5 MIN '

SOLENOID V Al VE WLS PL V055-S I1300 ESF 5 MIN SUMP CONTAINMENT ISOL ORC WLS PL V057 12256 ESF 5 MIN I

I i

1 1

410.196-11 I W-Westinghouse '

NRC REQUEST FOR ADDITIONAL INFORMATION Table 3. Il-1 (Sheet 28 of 28)

Safety-Related Electrical and Mechanical Equipment Ogwrating Envir. Time APtAl Zone Function Requiral Dev ription Tag No. (Note 2) (Note 1) (Note 5)

SOLE NOID V ALVE WLS PL V057-5 12256 ESF 5 MIN RCDT G AS CONT AINNIENT ISOL WLS PL V067 l1300 ESF 5 MIN SOLLNOlD VALVE WLS PL*/067 5 11300 ESF 5 MIN RCDT GAS CONTAINMENT ISOL WLS PL V068 12256 ESF 5 MIN SOL ENOID V ALVE WLS PL V068-S 12256 ESF 5 MIN SilSCELLANEOl:S:

NON ACTIVE VALVES SEE TABLE 3 2-3 IIEAT EX('ll ANGERS SEE TABLE 3.2-3 T AN KS SEE TABLE 3 2-3 IlYDROGEN RECOMBINER A VLS N1Y E01 A 11300 1 YR llYDROGEN RECOMBINER B V LS MY E0lB 11300 1 YR RLMOTE SilUTDOWN WORKSTATION 12303 NOTE 3 Note 1 RT <Remetor triph ESF iEngmeered Safeguards f eaturen. PAMS IPost Accident Monitonng), ISOL (holation)

Note 2 Zonen identified by room numbers - see Section 1.2, Nontropnetary General Arrangement draw ings Note 3 Not required post-secident Note 4 Only 3 of 16 used for PAMS Note 5 keference Table 3D 4-2 3

410.196-12 W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Question 410.253 ilow is the main steam supply sy stem designed to protect against water entrainment in accordance with Position 2 of Section 10.3 of the SRF?

Response

Water entrainment due to condensation is addressed in the layout of the steamline piping. The layout for the steamline piping inside the auxiliary building provides draining by proper sloping of the lines to enhance condensate collection, and in the use of condensate drains. A 12" condensate pot is provided upstream of the main steam isolation vake to collect condensate in the main steamline.

SSAR Revision:

The fourth paragraph of Subsection 10.3.2.2. I will be revised as follows:

The main steam lines between the steam generator and the containment penetration are designed to meet the leak before break criteria described in Subsection 3.6.3. The portion of the system between the containment penetration and the anchor downstream of the MSIV is part of the break exclusion zone. For each of these cases piping failures need not be postulated.

The layout of the steam piping provides for the collection and drainage of condensate to avoid water entrainment, by the proper sloping of lines and the use of condensate drain pots.

PRA Revision: NONE i

l 1

l l

l I

W

- WB5ilnEh0tlS8 1

NRC REQUEST FOR ADDITibNAL INFORMATION Question 440.61 Provide a description of the design features that minimize shutdown risk. The description should include non-safety-related systems for normal shutdown operations, passive safety-related systems, and the functional capability and availability of the active systems to ensure defense-in-depth, accident mitigation, and core damage prevention capability. The design bases, functions and supporting analyses for each system used to minimize shutdown risk should be discussed.

Response

Sy stems that minimize shutdown risk are discussed in Reference 440.6l-1. The AP600 PR A show s the effectiveness of these systems in minimizing the risk during shutdown conditions. Descriptions of these systems are provided in the appropriate SSAR sections. The SSAR also includes the design basis for each of these systems. The response to R Als 100.11, 440.53 and 440.62 provides additional discussion about features of these systems that minimize shutdown risk.

The response to RAls 440.63 and 440.66 address the analysis of the plant during shutdown conditions.

The response to RAI 140.56 discusses instrumentation available during shutdown conditions.

References:

440.61-1 WCAP-13793, AP600 Systems / Events Matrix, June 1994 SSAR Revision: NONE PRA Revision: NONE 5

440.61-1 W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION c.

Question 440.205 in sequence 24 of the 1.OOP esent tree, PF 203. Figure F-9 of the PRA. the offsite power is not recovered in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months and the PRHRS is initially operating. What is the ef fect of ADS actuation at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter LOOP 7 What is the operator expected to do in this regard? Can RNS be used to remose decay heat? Will a tran> portable generator be available after 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ?

Response

It is very unlikely that loss of both offsite and onsite power will last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Should the loss of of fsite and onsite power continue for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the normal residual heat removal system will not be available due to the unavailability of the ac power. If onsite power becomes available prior to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the startup feedwater pumps would be automatically actuated to remove decay heat. The operator can take actions to use the normal residual heat removal system to remove decay heat after the reactor coolant system is cooled down and depressurized to its cut-in conditions.

If ac power is lost for an extended tirne, it is likely that a generator can be made available within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to sustain decay heat removal with the passise residual heat removal heat exchanger. If the generator cannot be made available, then the automatic depressurization system is automatically actuated to put the plant in a long term core cooling mode before the 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months batteries are exhausted. This automatie depressurization system actuation places the plant in a safe long term shutdown condition using injection from the accumulators, core makeup tanks, IRWST and ultimately from containment recirculation. A generator is not required to support this shutdown cooling mode.

SSAR Revision: NONE PRA Revision: NONE s

W westinghouse 44 .20s-1

l NRC REQUEST FOR ADDITl'ONAL INFORMATION L'E .i,G

...i Question 440.214 Note 16 in the P&lD PXS M6001 and note 22 of PXS M6002 indicate that freeze seal would be used for direct vessel injection lines, accumulator injection lines, and gravity injection lines. Why is ihic necessary? What is the plant condition under which a freeze seal is used? How do you account for this in the PRA?

Response

Freeze seals can be used to temporarily isolate those portions of the lines in the passive core cooling system that connect directly to the primary pressure boundary. They are used to allow maintenance on valves or instnunents in lines that interface directly with the reactor coolant system.

The plant conditions under which the freeze seals could be used are as follows:

Mode 5 (cold shutdown with the reactor coolant sy stem open)

Mode 6 (refueling)

The PRA models accident scenarios in Mode 5. The PRA models maintenance of equipment: the impact is the same regardless of whether enanual vakes or freeze seals are used to provide the isolation. Therefore, use of freeze seals has no impact on the PRA.

SSAR Revision: NONE PRA Revision: NONE s

440.214-1 W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION M

u J

m Question 440.218 in a transient with loss of steam generator heat removal, the PRHRS is used to remose decay heat. In such traraients, w hat are the projected RCS pressure, temperature and lesel? What does the operator need to do to ensure its operation? What instrumentations are available to the operator? During a transient. some RCS insentory may be lost through the pressurizer safety valses. In an overcooling transient, the RCS level could shrink due to thermal contraction. The staff is concerned that there may not be sufficient inventory to support PRHR operation. What is the minimum RCS inventory that can support PRHR operation? What is the elevation of the top of the PRHRS HX? How does it compare with the level in the CMT and the top of the sessel? Can the RCS inventory become lower than the minimum needed in some of the transients? What is the pressurizer lesel at which the PRHRS HX would start draining?

Response

Following a transient with loss of steam generator heat removal capability, the passive residual heat removal heat exchanger is automatically actuated to provide the safety-related decay heat removal function. Due to the passive residual heat removal system operation, the reactor coolant system parameters vary in relation to the initiating event, the number of available heat exchangers, the duration of the passive residual heat removal system operation, and the availability of nonsafety related systems during the transient. SSAR Section 15.2 prosides analyses that demonstrate the performance of the safety-related systems. The projected temperature and corresponding pressure is consistent with the passive residual heat removal system functional requirements described in the response to RAI 440.92. He projected reactor coolant system level is consistent with the design criteria requiring that the passive residual heat removal heat exchanger operation, during postulated non-1.OCA event:, with loss of steam generator decay heat removal capability, shall sufficiently cool the reactor coolant system.

The operator has the capability to monitor the passive residual heat removal heat exchanger operation by means of several instruments. SSAR Table 7.5-1 provides the list of variables available to the operator. Passive residual heat removal operation and the reactor coolant system makeup provided by the core makeup tanks are automatically actuated and require no operator action.

Regarding concerns about the reactor coolant system inventory during the passive residual heat removal heat exchanger operation, it should be noted that two separate kind of events can be postulated:

Reactor coolant system shrinkage due to reactor coolant system cooldown The opening of pressurizer safety valves with loss of primary coolant for a reactor coolant system heatup.

I Section 4.1 of Reference 440.218-1 provides the analysis of the steam line break (a reactor coolant system l couldown event). The results of this analysis indicate that the core makeup tanks provide sufficient reactor makeup to support passive re,sidual heat removal system operation with reactor coolant system temperature as )

low as 212 F without credit for accumulator injection and without automatic depressurization system actuation. l l

l l

440.218-1 '

W-Westinghouse l

1 i

. NRC REQUEST FOR ADDITIONAL INFORMATION I

5l Section 15.2.7 of the SSAR provides the analysis of the loss of main feedwater event (a reactor coolant system heatup event). In this event the pressurizer safety valves open such that some reactor coolant inventory is lost.

The results of this analy sis show that there is sutticient reactor coolant system inventory to maintain pressurizer level w hich is more than enough to support passive residual heat removal system operation.

The passise residual heat removal system will remain water filled until the reactor coolant system level drops to the hot leg. There is no minimum reactor coolant sy stem inventory required to support the passive residual heat removal system operation since the passive residual heat removal heat exchanger will operate with water / steam or just steam inputs. However, for non-LOCA events the reactor coolant system conditions are such that the passise residual heat removal system will operate with water flow from the hot leg.

The elevations requested are as follows:

Top PRHR heat exchanger (top tubes) 130.67 ft Top core makeup tank (CMT) 127.63 ft CMT water elevation to actuate ADS (67% of CMT solume) 120.16 ft Top reactor vessel 113.78 ft Pressurizer bottom 120.60 ft Refer to the responses to RAls 440.103 and 440.126 rerpectively, for information relevant to IRWST insentory during the passive residual heat removal system operation and the effect of the presed::er heaters on the passise residual heat removal system operation.

References:

440.218-1 AP600 Design Change Description Report, February 15, 1994.

SSAR Revision: NONE 1

I 1

l l

1 l

l l

440.218 2 3 Westinghouse 1

l l

NRC REQUEST FOR ADDITIONAL INFORMATION Ouestion 440.220 Hov is the PRHRS HX tube integrity serified? Do you need to perform eddy current tests similar to that for steam generator:,7 How and during w hat plant conditions are these tests done? How do such tests affect the availability of the IRWST!

Response

The design of the passive residual heat removal heat exchanger tube to tube sheet joint is the same as in the steam generator. The inspection of the passive residual heat removal heat exchanger will be performed using the same techniques. These inspections will be performed when the passive residual heat remosal heat exchanger is not required to be available by the technical specifications and the reactor coolant system is depressurized during modes 5 and 6.

Access to the tubes for inspection is through the channel head manway. Since the channel heads are located outside the IRWST, the IRWST is unaffxted by this inspection.

SSAR Revision: NONE s

440.220-1 T Westingt100se l

NRC REQUEST FOR ADDITIONAL INFORMATION Question 440.232 Section A 3 of the PRA defines a small-break LOCA as a break from 3/4 to 4 inches while Section 15.6.5.4B.3.1 of SSAR defines a small-break LOCA as a break size of one square foot or less. Provide the follow ing information:

a. Clarify the inconsistency in the definition of LOCA sizes.

b. What is the size oi the LOCA that a CVS pump is capable of mitigating? How long can this pump perform its mitigating functions? Given such a LOCA, what is the expected plant and operator response to bring the plant to a safe condition?

c. For how long will the CVS be able to inject borated water during an RCS leak or a sery small-break 1.DCA? How much insentory is available? What measures assure that this inventory will be available?

Response;

a. In the AP600 PRA, LOCAs are classified according to the break size, the location of the break, and the ef fect on mitigating systems. He break size was defined on the basis of plant systems required to mitigate the accident. The break sizes that separate the LOCA categories are 10-inch diameter between large/ medium,4-inch diameter between medium /small,3/4-inch diameter between small/very rmll, and 3'8-inch diameter between very small RCS leak. More details of the LOCA classification is provideu N Chapter 7 and Appendix A of the AP600 PRA Report. The RCS leak was addressed in reference 440.2324 and will be incorporated in Revision 2 of the PRA.

The design basis LOCA classification in the SSAR Sections 15.6.5.4A and 4B is based on ANSI 18-2 which disides plant conditions into different categories according to anticipated frequency of occurrence and potential radiological consequences to the public. Based on reference 440.232-2, the RCS design basis breaks were categorized in the SSAR as follows:

large break is defined as a rupture of the reactor coolant sy stem with a total cross-section area equal to or greater than one square foot. This event is considered as a Condition IV cvent (limiting fault) because it is not expected to occur during the lifetime of the plant but is postulated as a design basis accident for design engineered safeguards.

Small break is defined as a rupture of the reactor coolant system with a total cross-section area less than one square foot in which normally operating charging system flow is not sufficient to sustain pressurizer level and pressure. This event is considered as a Condition ill event because it is an infrequent fault that may occur during the life of plant.

b. One chemical and solunge control system makeup pump is capable of supplying sufficient makeup flow to compensate for the leakage from breaks up to 3/8-inch diameter (RCS leak). The chemical and volume control sy stem can maintain the RCS pressure and presssurizer level and permit the operator to perform a W

- WestinEflause O

NRC REQUEST FOR ADDITIONAL INFORMATION controlled shutdow n using the startup feedwater pumps. The normal residual heat removal system is later used to continue the cooldown so the RCS can be completely depressurized and thus reduce the leakage through the break. The PRA evaluation of RCS leak is described in reference 440.232-1.

For breaks sizes from 3/8 to 3/4-inches in diameter hery small LOCA). the chemical and volume control system, in conjunction with the core makeup tanks and the passise residual heat removal system, can provide sufficient RCS makeup for leakage and accommodate RCS cooldown. The normal residual heat removal system is later used to continue the cooldown so that the RCS can be completely depressurized and thus reduce leakage through the break. For small LOCAs, up to two inches, one CVS pump can provide suf ficient makeup to prevent core uncovery. The PRA does not credit CVS operation for mitigating break sizes greater than 3/4-inch diameter because of consideration of limited CVS makeup water supplies. The esent tree model for the very small LOCA event is provided as Figure F-18 in Appendix F, Section F.2.19 of the AP600 PRA Report.

The CVS has sufficient inventory to provide RCS makeup for breaks up to 3/4-inch for 5 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months without operator action. As the operator proceeds to promptly reduce the RCS pressure for shutdown, the amount of inventory being lost through the break will diminish. This reduces the demand for makeup and the CVS pumps will deliver makeup at a reduced rate. The PRA does not model operator actions that would reduce the RCS pressure and allow CVS operation for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This will be evaluated in Revision 2 of the PRA.

The expected plant and operator responses following RCS leaks and very small LOCAs are described in reference 440.232-3.

c. The CVS makeup pumps draw suction from the boric acid tank (BAT). (See SSAR Table 9.3.6-2 for the B AT capacity). The inventory in the BAT will be administratively controlled. In addition to the BAT, the CVS has the capability to take suction from the spent fuel pit. The CVS connection to the spent fuel pit is located approximately two feet below the normal water level and two feet above the spent fuel pit pump suction connection. This arrangement precludes a loss of cooling or inadvertent draining of the spent fuel pit due to CVS operation. The inventory in the spent fuel pit will be administratively controlled.

Conservatively assuming a reduced inventory in the BAT, the BAT and the spent fuel pit maintain sufficient insentory to provide suction to one CVS makeup pump to supply RCS makeup for 5 to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months at 100 gpm.

This does not take credit for operator actions to replenish the inventory in the BAT. In adoition, the AP600 RTNSS evaluation (reference 440.232-4) indicates that the CVS is not RTNSS significant.

References:

440.232-1 ET-NRC-93-3990, " AP600 Reactor Coolant System Leak PR A Evaluation," October 1993.

440-232-2 WCAP-8340, " Westinghouse Emergency Core Cooling System - Plant Sensitivity Studies," July 1974.

440.232 3 WCA P-13793, ," AP600 System / Event Matrix," June 1994.

440-232-4 WCAP-13856, " AP600 Implementation of the Regulatory Treatment of Nonsafety-Related Systems i Process," September 1993.

l l

l 440.232-2 W- WestinEhouse

NRC REQUEST FOR ADDITIONAL INFORMATION A

SSAR Revision: NONii PRA Revision: NON!!

i 6

l l

44 .232-3 W westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION

.. m Response Revision 1 -

IE Ouestion 460.5 Provide the follow mg information " ,urding solid radwastes (Section i1.4):

a. Estimates of solid waste volumes expected to be shipped annually for wet solid wastes and dry solid wastes separately.

b. A discussion of compliance with Position 111.1 ot BTP ETSB 11-3 regarding the storage capacity for accumulated filter sludges.

c. A discussion of compliance with Position 111.2 of BTP ETSB 11-3 regarding storage volume for solidified wastes (both wet and dry solid wastes) available in the plant.

Response: (Revision 1)

a. Estimates of solid waste volumes expected to be generated and shipped annually are separately provided below in Tables 1.A. and 2.A for wet solid wastes and dry solid wastes. These volumes are consistent with SSAR Table 11.4-4. In Table 11.4-4 the volume of wastes to lx shipped accounts for volume reduction of some wastes and addition of the shipping / disposal container. The processing and packaging factors used in SS AR Table 11.4 4 are conservative and may have over-estimated the annual disposal quantities. Below each estimate of annual disposal quantities are the details of the inputs to each category of generation (Tables I.B and 2.B). These waste generation estimates are based on an 18 month refueling cycle For a 24 month refueling cycle, the annual waste generation is less.

Table I.A - Annual Wet Solid Waste Generation and Disposal Quantities Summary Waste Generation Expected Sh5phy Disposal I i

Wet Solid Wastes Volume, ft'AT Volume, ft%T Spent Charcoal and Ion Exchange Resins 250 314.8 Mixed Liquid Wastes 15 17.0 Chemical Wastes }jf0 19.8 Tot:d Wet Solid Wastes 615 351.6 The generation of these wet solid wastes is summarized as follows:

a T Westinghouse M5m

NRC REQUEST FOR ADDITIONAL INFORMATION

"-+ l;";

NI Response Revision 1 1

l l

Table 1.11 - Annual Wet Solid Waste Generation Quantities Details Waste Source Waste Generation Rate Annuali7ed Comments Waste Generation WLS Deep Bed Filter 10 ft' / 6 months 20 ft' Charcoed WLS Deep Bed Filter 40 ft' / year 40 ft' Zeolite 3 x WLS lon Exchanger 30 ft' ea. / year 90 ft' Resin 2 x CVS Ion Mixed Bed 50 ft' /18 months 33.3 ft' one vessel discharged exh Exchanger Resin refueling 1 x CVS Cation Bed 50 ft' / 36 months 16.7 ft' Exchanger Resin 2 x SFS lon Exchanger 75 ft' /18 months 50 ft' one vessel discharged each Resin refueling Subtotal Wet Solid 250 ft' Waste (Spent Resins and Charcoal)

Mixed Liquid Wastes 4 gal./ month for 17 15 ft' mainly contaminated lubricating months + 100 gallons oil during refueling outage Chemical Wastes 200 gal./ month for 17 350 ft' Chemical laboratory and I months + 500 gallons decontamination wastes during refueling outage Total Wet Solid Wastes 615 ft' I

4 !

1 1

460.5(RI)-2 ,

W Westinghouse l l

NRC REQUEST FOR ADDITIONAL INFORMATION

a ==

Response Revision 1 Table 2.A - Annual Dry Solid Waste Generation and Disposal Quantities Suminary Waste Generation Expected Disposal Drv Solid Wastes Volume, ft'/vr Volume, ft3/vr Spent Filter Cartridges 3.6 24.5 4J WGS Charcoal 8.7 9.8 Compactible DAW 4480 953.0 uu4g Non-Compactible DAW 225 363.4 Mixed Solid Wastes 5 7.5 Tot;d Dry Solid Wastes 4722 1358. 2 4 L7&-7 The generation of these dry solid wastes is summarized as follows:

Table 2.11 - Annual Dry Solid Waste Generation Quantities Details Waste Source Waste Generation Rate Annualized Comments Waste Generation WSS Resin Fines Filter 1 x 0.5 ft' cartridge / 0.5 ft' 6" dia. x 30" pleated liberglas, year 25 microns @ 997c efficiency 2 x WLS Filter 4 x 0.3 ft' cartridges /18 0.8 ft' 6" dia. x 18" months CVS Makeup Filter 1 x 0.3 ft' cartridge /18 0.2 ft' 6" dia. x 18" months CVS Reactor Coolant 4 x 0.5 ft' cartridges /18 1.31t' 6" dia. x 30" Filter months SFS Filters 4 x 0.3 ft' cartridges /18 0.8 ft' 6" dia. x 18" months WGS Guard Bed 8 ft' / 36 months 2.7 ft' assumes abnormal charcoal Charcoal replacement WGS Delay Bed 60 ft3 /10 years 6 ft' assumes abnormal charcoal Charcoal replacement High Activity Dry 2 ft'/ month for 17 50 ft $

Compactible Waste months + 40 ft' during refueling outage

[ W85tingh0USB l

1

NRC REQUEST FOR ADDITIONAL INFORMATION mu 4.$

~

Response Revision 1 Low Activity Dry 60 ft'/ month for 17 4000 ft' Compactible Waste months + 5000 f t' during refueling outage Compactible HVAC See Note 1 430 ft' Filters liigh Activity Dry Non- I ft'/ month for 17 25 ft' Compactible Waste months + 20 ft' during refueling outage Low Activity Dry Non- 6 ft'/ month for 17 200 ft' Compactible Waste months + 200 ft' during refueling outage Mixed Solid Wastes 0.2 ft'/ month for 17 5 ft' months + 4 ft' during refueling outage Total Dry Solid Wastes 4722 ft' Note 1: Waste HVAC filters are generated as follows -

Containment purge system (VFS) 13.7 ft' 36.7 ft' granular charcoal / 60 months 4 x 12" x 24" x 24" HEPA / 60 month 2 x 12" x 24" x 24" Pre-filter / 60 rnonths 2 x 12" x 24" x 24" Post-filter / 60 months Rad. Chem. Lab. Supply & Recirculation (VAS) 42.7 f t' 4 x 12" x 24" x 24" HEPA /18 month 4 x 12 x 24" x 24" Pre-filter / 6 month Rad. Chem. Lab. Exhaust (VAS) 21.3 ft' 2 x 12" x 24" x 24" HEPA /18 rnonth 2 x 12" x 24" x 24" Pre-filter / 6 month Radwaste Building (VRS) 240 ft' 60 x 12" x 24" x 24" HEPA / yr Annex Building (VHS) 112 ft $

28 x 12" x 24" x 24" HEPA / yr 429.7 ft' s

l

- 1 I

3 Westinghouse l

I

NRC REQUEST FOR ADDITIONAL INFORMATION IEE Esi

' E Response Revision 1

('nder normal conditions, there are no wastes generated from the second:uy-tilant cycle. If radioactivity is detected in the ste:un genenitor blowdown and reaches a predetermined level (due to primary and second:uy leakage), the blowdown is diverted to the eomismie1*4isher* to the liquid radwaste system for processing as described in SS AR Subsections 10.4.8 and 11.2.2.1.5. Thus. blowdown demineralizer enisMe p+4i4dng resins are not a nonnal source of radwaste requiring shipping and dispos:d. Should plant operation continue with leakage from the primary to the secondary side utilizing the blowdown demineralizers e.elewigw4i4ws, a shipping (disposal) volume of up to 300 it'/ month could be produced by the steam generator blowdown c4eleasateeli4dag system, as indicated in SSAR Subsection 11.4.2.1.

b. The AP600 plam does not include filters that generate sludge-type wastes. Also, tanks and sumps are designed to minimize the fonnation of sludge deposits, and the particulate matter that can cause sludge deposits is transponed to and removed by the cartridge filters in the liquid radwaste system. Therefore, there are no storage provisions for accumulating sludges.

c. Branch Technical position ETSB 11-3 specifies in Position 111.2. that storage areas for solidified wastes should be capable of accommodating at least 30 days of waste generation at nonnal generation rates and that these storage areas should be indoors. The storage durabons of the storage areas in-thshiwaste Buihiing are evaluated for four general types of waste and container categories discussed below.

Spent Ion Exchance Resins and Filter Charcoal in Hich-intecrity Containers (HICs)

Although the shipping or disposal volume is nearly independent of container size (based on equal tilling efficiency. e.g.,907c), the storage duration for the filled HICs is dependent on the number of containers which is indirectly proportional to container size. To be conservative. it is therefore assumed that the spent resins and filter bed charcoal will be dewatered in HICs that will fit into a Type B shipping cask (i.e., the SEG 3-82B. formally HN-200). Nonnally 250 ft'/yr of spent resin and charcoal is expected to be generated (SS AR Table 11.4-4), with an activity of 950 curies (SSAR Table 114-6). This resin can be mixed to pnxluce a uniform specific activity of 3.8 Ci/ft'. A 158 W-ft' HIC filled to 90% would contain about 540 240 Curies, well within the cask's capability. About two four of these HICs are required per year. The three onsite storage casks can each hold one of these HICs and the resulting storage duration is about one and a half years Wnmuhs. De two spent resin container fill stations may also be used for storage until it is necessary to begin lilling another HIC. The two fill stations and two of the onsite storage casks (reserving one onsite storage cask for high-activity tilter drums) provide about one and a half years 42 months of storage. These storage times are in addition to the pre-packaging storage times provided by the spent resin tanks as described in SSAR Subsection i1.4.2.2.1.

Hieh Activity Filter Cartridges in Dmms As indicated in SS AR Table 11.4-4. packaging of the CVS reactor coolant filter cartridges is expected to nonnally generate 36 drums of waste per year. This is based on a generation rate of 4 2-fd-of filter cartridges every 18 months. Rami-on+dann ' n!ume:ric loading af (0t weentral*w one ha!f of a drum 44111edeery44nmalM24tWJLit% 9 % :d hutelnwn-i+ produced e .e y M " :h De high-W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 high-activity filter storage tube module may be used to store all filter cartridges normally generated every 18 months. Thus, after a drum is filled with high-activity filters, encapsulated, and sealed, it may remain in the processing cask fc ab' " :nr Nfre until it is necessary to begin to uge the filter cartridges stored in the storage tube module to clear space for the next batch of spent filter- Therefore, '

a storage duration of about 17 months is available for high-activity filters using only the processing cask.

One of the onsite storage casks could also be used for high activity filter drum storage if necessary.

Other Wastes in Drums t

Based on SSAR Table 11.4-4, about 1I drums are produced each year containing wastes other than high ..

activity filter and mixed wastes. The Radwaste Building (Proprietary Figure 1.2-29) has a packaged waste storage room that may be used to store both drums and boxes. Using two storage locations for palletized i drums stacked three high,24 drums can be stored. His provides about 28 months of storage for the ;

normal expected generation rate. Stacking only two pallets high provides about 18 months of storage.

Without stacking about 9 months of storage is available for the normal generation rate. ;

Mixed wastes are accumulated in drums and are sent to an off site processing facility :pri ! p :f&t :::d

:r:ge bui! ding at an expected rate of about three drums per year.

Wastes in Boxes !

Based on SSAR Table 11.4-4, about 12 444 boxes are generated per year. Ten Faur:== box storage locations are available in the packaged waste storage room. Without stacking and with stacking two and three high, about 1, 2. and 3 years of storage are provided, respectively. ,

Maximum truck loading is expected to be 28 boxes. P.!: equ .:!:- :: i: '

Ten storage locations can accumulate a truck load when stacked three twe high. At the nonnally expected generation rate, it takes 2 years 26 m-'h: to produce a truck load. ;

In summary, indoor storage is provided for all categories of packaged wastes well in excess of 30 days, based on !

normally expected waste generation rates.

SSAR Revision: Table 114-4 is to be revised as follows:

f n

i e

460.5(R1)-6 W

Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION s- ms

+ !!

Response Revision 1 at Table 11.4-4 (Sheet 1 of 3)

Expected Annual Radwaste Volumes influent Vol. Contamer ( l)

Source Form Vol. Reduction (t:33 Type 9 Full Quantny Dispnal Vol. (ftI)

WIS Deep lied hlter Si rnt Activated Char- 20 HIC (2) 90 0.14 25.2 cual WLS Deep Hed lilter Spent Zeolytic Resm 40 HIC 90 0.28 50.4 WIS hm Exchanger #1 Spent llorated Catum to illC 90 0.21 37.8 lied II+Fonn Resm WIS 1.m Exchanger #2 Spent llorated Exed 30 HIC 90 0.21 37.8 lied fl+0H Form Resm WIS lon Exchanger #3 S l ynt Horated %xed 30 tilC 90 0.21 37.8 lied 11+011 Fonn Re:m CVS %xed lied lon Spent llorated %xed 31.1 HIC 90 0.23 41.9 Exchangers Red 12011 Form Resm CVS Cation Bed lon Spent llorated Canon 16.7 -

HIC 90 0.12 21.0 Fxchanger lied il+Fonn Resui SIS lon Exchangers Spent llorated %xed 50 tilC 90 0.15 62.9 Hed II+#)ll Resm WSS Resin Fmes Riter Spent hiter Cartndge 0.5 4:1 Drum 90 0.02 0.1 WIS Filters Spent Filter Cartndge 0.8 4:1 Drum 90 0.03 0.2 CVS Makeup Iilter Spent Filter Cartndge 0.2 41 Drum 90 0.01 0.07 CVS Reactor Coolant Spent Futer Cartndge 1.3 Drum . Note 4 34u 22.523 hiters W SFS Riters Spent Futer Cartndge 0.8 Drum 50 0.21 ! .6 WtiS Guard Red Activated Charcoal 2.7 Drum 90 0.4 3.0 WGS Delay Hed Activated Charcoal 6 Drum 90 0.9 6.R RCA SCAs (2) lingb Actmty Conhactible DAW (2) $0 41 Drum 90 1.9 14.2 460.5(RI)-7 W

_ WestinEhouse

,&+. . -

_s.2- ~6 ._6- - . . . .- .. , . .

NRC REQUEST FOR ADDITIONAL INFORMATION i

u- m;_

Response Revision 1 Table 11.4-4 (Sheet 2 of 3)

Expected Annual Radwaste Volumes influent Vol. Container (1) source Form Vol . Reduction (g3) Type % Full Quantity Disposal Volaft3)

RCA SCA: Other Compactible 4000 6:1 Box 90 8.2 847.7 DAW HVAC Exhaust Filters Pre, HEPA, and 430 644 6:1 Box 90 0.88 m 91.1 ca.4 ,

Charcoal Filters RCA SCAs Higher Activity 25 -

Drum 70 4.9 36.4 Non-Compactible DAW RCA SCAs Lower Activity 200 - B;4 70 3.2 327.0 ;

Non Compactible DAW Luhricants and Mixed Wastea - 5 -

Drum 70 1.0 7.5 l Cleanmg Agents Solids ,

Mixed Wastes - 15 -

Drum 90 2.3 17.0 Liquids Radiochemistry Labo- Chemical Wastes 350 (3) Drum 90 2.6 19 8 retory 1.8 HICs 17.3 44,6 Drums ;

Totals 5337SMS 12.3 Boxes 1710 4 3 t1) Container Parameters Drum internal Volume = 7.35 ft3 Drum Disposal Volume = 7.5 R3 t Hic Internal Volume = 155 ft3 Hic Disposal Volume = 179 ft3 Box Internal Volume = 90 ft3 l Box Disposal Volume = 103 ft3 (2) Acrynoma CVS = Chemical and Volume Control System DAW =

Dry Active, Waste HEPA = High Efbency Particulate Air illC = High integrity Container HVAC = Heating Ventilating and Air Conditioning 460.5(R1)-8 W-Westinghouse e

. . _ - .m. . ,_ .

NRC REQUEST FOR ADDITIONAL INFORMATION

+: A.

+

Response Revision 1

'AW -

Table 11.4-4 (Sheet 3 of 3)

Expected Annual Radwaste Volumes RCA = Radiologically Controlled Area SCA = Surface Contamination Area SFS = Spent Fuel Cooling System WGS = Gaseous Waste Management System WLS = Liquid Waste Management System WSS = Solid Waste Management System (3) Mobile System Concentration = 25:1 and 80% Waste Loading Using Aquaset/Petroset.

(4) One filter cartridge encapsulated in each drum r

i i

I

\

l 460.5(RI) !

T Westinghouse l

.- , ~. . - - .= _. - _. - . _ _J

l l

l 1

l l

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 Question 460.7 Table 2 of Section 11.5 of the SRP, " Process and Ef0uent Monitoring Instrumentation and Sampling Systems "

includes a service water system effluent monitor. The staff notes that AP600 design includes an upstream provision for this monitor in the form of component cooling water system monitor. The staff does not consider an upstream provision as an adequate basis for eliminating a downstream provision for this monitor. Therefore, include a service water system monitor or justify its elimination (Sections !1.5).

Response: (Revision 1)

The Service Water System water inventory is integrated with the Circulating Water System and the cooling tower basin. The Service Water System does not have a dedicated effluent discharge pathway but shares a common cooling tower blowdown system connection with the circulating water system as shown in Figure 10.4.5-1. As shown in Table i 1.5.1, the normally nonradioactive systems which provide potential leak pathways into the service water system include radiation monitors with nominal minimum detectable concentrations of 1.0E-8 uCi/cc. Further dilution in the service water or circulating water systems or in the cooling water blowdown system, as described in Section 11.2.3.3 would reduce the concentration to approximately 3E-1I which is below the minimum detectable concentration for a continuous-type monitor. Therefore, a radiation monitor in the cooling tower blow down (service water ef fluent) will not provide additional information or control of effluent releases.

As discussed in Section 11.5.3, the primary means of quantitatively evaluating the isotopic activities in effluent paths is a program of sampling and laboratory measurements. Table 9.3.4-2, which indicates grab sample locations, identifies a grab sample for the cooling tower blowdown as well as the cooling tower basin and service water basin.

The grab sample allows for measuring of the activities in the effluent path.

SSAR Revision: NONE NRC Disposition Clarify which are the normally non-radioactive leakages that feed into the SWS and which are identified in Table 11.5-1 of the SSAR. Provide this information in the SSAR.

Disposition Response:

The service water system (SWS) configuration is described in SSAR Section 9.2.1. Revision 1.

The only potential radioactive inleakage path to the SWS is from the normally non-radioactive component cooling water system (CCS). The interface occurs at the CCS heat exchangers, which are cooled by the SWS. The CCS ;

has a radiation monitor which is identified in SSAR Table 11.5-1. The design of the SWS includes a radiation 1 monitor with a high alarm to monitor the SWS cooling tower blowdown flow (see also response to RAI 410.110).

The radiation monitor is shokn on SSAR Figure 9.2.1-1. The SSAR will be revised.

l 1

l 460.7(R1)-1 W-Westingt10tise

1 l

l l

NRC REQUEST FOR ADDITIONAL INFORMATION l Response Revision 1 SSAR Revision:

Revise 9.2.1.2.1 to add a new paragraph to follow the third paragraph and read as follows:

A small portion of the service water system flow is normally diverted to waste water through a blowdown flow path upstream of the cooling tower. Blowdown maintains the water chemistry of the service water system.

Revise 9.2.1.2.2 under the heading of Service Water System Valves to add a new paragraph to follow the fourth paragraph as follows:

An air-operated isolation valve is provided in the cooling tower blowdown line. 'This valve allows the operator to regulate the blowdown flow. The valve also provides automatic isolation of blowdown upon loss of offsite power.

The valve fails closed upon loss of control air or electrical power.

Revise 9.2.1.5 to add a new paragraph to follow the sixth paragraph and read as follows:

A radiation monitor with a high alarm is provided to monitor the cooling tower blowdown flow for radioactive leakage into the service water system from the component cooling water heat exchangers.

Revise Table 11.51 to add a radiation monitor listing as indicated below:

Detector Type Service Isotopes Nominal Range SWS-JE-RE001 t Service Water Cs-137 1.0E-7 to 1.0E-2 Blowdown pCi/cc Revise 11.5.2.3.1 to add a new heading and text description. to follow at the end of the subsection, to read as follows:

Service Water Illowdown Radiation hfonitor The service water blowdown radiation monitor (detector SWS-JE-RE001) measures the concentration of radioactive materials in the blowdown flow from the service water system to the waste water system.

The service water blowdown radiation monitor initiates an alarm in the main control room if the concentration of radioactive materials exceeds a predetermined setpoint. Following the alarm, the operator can manually isolate the blowdown flow.

The range and principal isotopes are listed in Table i1.5-1. The detector is a gamma sensitive, sodium iodide, thallium activated, gain stabflized scintillator that views the liquid sample volume.

460.7(R1)-2 W Westingtiouse

NRC REQUEST FOR ADDITl'ONAL INFORMATION Response Revision 1 Question 460.11 Proside the following information regarding the solid radwaste manai;ement s; tem (Section i1.4):

a. Since solidification and encapsulation are not the same. clarify whether either of the abose two options may be used for processing spent resins in addition to a third option, namely dewatering the resins (note that encapsulation is not generally used for processing spent resins). Also, clarify w hether the AP600 solid radw aste management system design desiates from the EPRI Requirements Document for passive reactor designs. The Requirements Document recommends only dewatering for processing the spent resins.

b. Identify the specific design features prosided in the system design to comply with GDCs 60,63 and 64 as they relate to (I) control of release of radioactive materials to the environment from the plant areas w here the solid radwastes are processed, and (2) monitoring radiation levels and leakage.

c. Clarify w hether the description and discussion of acceptability of the portable grouting unit that may be used for processing spent filters is within the COL applicant's scope. If it is within the AP600 design scope, proside specific details of the unit.

d. The staffis concerned that the projected tTables 11.4-4 and 11.4-5 of the SSAR) annual solid radwaste volumes to be disposed (1729 CF for the expected case and 3843 CF for the maximum case) are significantly lower than that actually shipped volume for operating PWRs (EPRI NP-5528, February 1988, Volume 2 - Plants Without Evaporators for the Years 1985 and 1986: 9550 CF). The staff recognizes that the projected solume agrees with the value proposed in the EPRI Requirements Document (1750 CF per year). The EPRI-proposed value depends on following what EPRI regards as sound design and operating techniques outlined in the document (Paragraph B.I.2.2 of Appendix B of Chapter 12) for reducing the shipment of processed sol;d waste volurne.

One of the operating techniques is to asoid solidification and instead use only dewatering for solidifying the w et solid wastes. As stated above. the AP600 design includes solidification as one of the options. The staff is concerned that the storage solume allotted for processed solid wastes may be inadequate if it is to be based on the projected shipment volumes gisen in the SSAR tables. Therefore, provide justification for the projected volumes given in the subject SSAR tables or resise the values as appropriate.

e. Clarify why the AP600 design does not include phase separator tanks, as recommended in the EPRI Requirements Document for passise reactor designs.

f. Section i1.4.1.3 of the SSAR identifies the capability to store processed and packaged solid wastes at the site for at least six months to account for pessible delay or disruption of offsite shipping of the wastes as one of the design objectises of the solid waste management system. However, there is no description of the on-site storage facility in the SSAR. Provide a description of the facility, and clarify whether it conforms with the recommendations identified for such a facility in Section 5.4 of Chapter 12 of the EPRI Requirements Document for passise repetor designs.

W westingtmuse * ""'b'

r l

- NRC REQUEST FOR ADDITIONAL INFORMATION l l

l I Response Revision 1 Response: (Revision 1)

a. References to resin solidification and encapsulation will be deleted from Section i1.4, except as related to space reserved for future or optional solidification facilities. A4.-huli. :m! SS A R 9:5.re' ! ! !.Lt. ! :m al4emativo49-dmA e-*!c , of sp;

1 halew*teing i -

~' dify4c * .ap w4a' th; spe ' - ng:

palitbikwfing age '

E p; c:abhm:p;

c;!rd for a!!;!!!ication, '- . . **e-m4-peovh4al-a -t*+

of-4ho-A puM ;did '. armgc . ' '

De ' ng th; '--:gn '- :-peci fied :h; !':ility 4*puanent -Da

n! the4PMAle.+en-i- rn !! ;anrg.maneer

b. Relative to solid wastes, General Design Criterion 60. Control of Releases of Radioactive Materials to the Environment, requires that means be provided to handle radioactive solid wastes produced during normal reactor operation. including anticipated operational occurrences. Means are provided in the solid waste management system to handle the appheable categories of solid radwaste as indicated in SSAR Section 11.4.

To control the release of radioactive materials to the ensironment, the areas and components in the W ami radw aste building 4, that process or house radioactive solid wastes are located in areas coonw with exhaust sentilation that discharges through HEPA filters, as indicated in SSAR Subsection 9.4.8.2. -les*1dition-load vensd*4immimm.ts : ' MHI?FAr41} tere!c ,-ihed in SE AR c1

an " ! S 2. pr . id*MlW a--Ikum G np: ' -

a--4mw-ae44+4t:. W

' t e:* tem +g-Wnt a-4#w-Aetivity-W*+-ikyer 5 H ag-44a - He umi- Ha - n npaaer a-4mmirv-Aueline-Ta - %d R e.p:- ' - O c ng Dn nh! Haa.1 a-Re pi+* toe-Gleamng-De -O'- -Ha a--Re pirator4 Waning-F: - Da : Mmd a-}4igh-Pre. wee-Mot-Wat*-D= He ah a--4hraeDa en Re :!h Sloped floors and floor drains are provided to collect and thereby To control the release of radioactis e material that could be removed from stored solid waste by water contactr4he ec==! ^ x ^;

'he aduatie bmlding-4mve-+ cJ thrr- H!+ W ho! dup Ame! * ' ^ ' 20 W ga'!n _t hic' egn! !c - the ormr* tion-of-4ht- lr "r d e n- ' !&M gp: 20 mhm4e !n addninn :he 'nnt: : N : du ae imik44ng4hatwetain*4gnific"'i;;uid a! ant:- r t hhi ~ n!!y de ,:gnedecea4 hat +etaire the : - um ikpkkolene : dewitwl-hsh- respr . 'a Q MO E General Design Criterion 63. Monitoring Fuel and Waste Storage, requires that means be provided to detect conditions that may result in excessive radiation levels and to initiate appropriate actions. For the solid waste management system, the wastes with the most potential for high radiation levels are the spent ion exchange resins and filter cartridges. especially those from the chemical and volurr.e control system ion exchangers and filters. The radiation lesels of the spent resin tanks can be *re monitored without entering the rooms. Floor penetrations with shield plugs above the spent resin tanks are provided to allow the radiation levels in the tank 460,11(R 1 )-2 W-Westinghouse

i NRC REQUEST FOR ADDITIONAL INFORMATION3 -

Response Revision 1 pj '

rooms to be monitored by lowering detectors down the outside of the tanks. The shield doors to the spent resin tank rooms are normally locked to prevent inadvertent entry.

As described in SSAR Subsection 11.4.2.3.2, the dose rates of high-activity filter cartridges are measured I during the changeout process when the filter is raised into the high-activity filter transfer cask (but before the bottom cover of the shield cask is secured) using a long-handled radiation probe. The measured dose rate l determines the precautions taken during subsequent handling operations. The high-activity filter cartridges can be transferred into and out of the high-actisity tilter storage tubes using the high-activity filter transfer cask without direct exposure to personnel. The filters in storage can be monitored with minimal exposure at any time through sampling ports (normally closed by shielded plugs) of each storage tube. Ports with shield plugs ,

that may be used for monitoring stored waste containers are also provided on the high. activity filter processing cask and the onsite storage casks.

Dry, solid wastes are normally monitored when received at the radwaste building and are then transferred to the appropriate temporary storage location ar; ('- 2de :: , - high) depending on the measured dose rate as described in SSAR Subsection i1.4.2.3.3. Local shielding can be used within the temporary and packaged waste storage areas to segregate the higher dose rate items and thereby minimize the dose rate in the rest of the storage areas. The radiation levels in the temporary and packaged waste storage areas should be relatively low. The areas can be entered for monitoring at any time. These storage areas have doors that can be locked to control access.

kny!e ' :ds pr Wd r - e - ::er n; rnd r~$ !ng patenn !!y 'ardeus endi Ard - t E,

_g a w _ . : ., r- ._ .y ; . . . . ; . ,7 7 _

up77.,. - ; - xp3; g, -ru ,

, . g: z. . . ;- y aw*smw! ' ;

. ' ' !ac ' * -! ding ; - 5; ed ! cun-' *: dn - t ^c.

Relative to solid radwastes Criterion 64, Monitoring Radioactivity Releases, requires that means be provided for monitoring effluent discharge paths and the plant environs for radioactivity that may be released from normal operations, including anticipated operational oc:urrences, and from postulated accidents. Airborne !

effluents from the auxiliary and radwaste buildings are monitored by an exhaust radiation monitors, as described in SSAR Subsection 11.5.2.3.2. The spent resin waste container fill station . --d ^e high - a ; 'il4*

pr . :ng =k includes provisions for smear survey and decontamination of the external surfaces of the waste containers after filling, as described in SSAR Subsections 11.4.2.3.1 n-d ' ' !.M.2. mpea e!y. Pm nl knew Drums and boxes containing filters and lower-activity dry wastes are also surveyed and decontaminatedt L

^ribed :- SSAR Subuc' - '.2.?.3. A Mobile or portable equipment may be used for clean waste monitoring e " nnd 5:g :nr"--- : urd to verify that wastes segregated and sorted for nonradioactive disposal are nonradioactive. Hand-held survey meters are used to prevent removal of radioactivity from the radwaste

. building by personnel. P" ' -' ' n em pr Wd "'e : per -? : " 're: ^: r: din!:g: "; c' r"ed

c. :- ,o ;3.. % p..:u:ng. The arrangement of the radwaste building allows the corridors and vehicle access i

areas to be very low radioactivity areas, thereby minimizing the need for any decontamination operations.

Liquid wastes generatedgfrom solid radwaste system operations are discharged directly to the liquid radwaste i system (WLS) or are collected by the radioactive waste drain system (WRS) and directed to the WLS for subsequent processing and monitored discharge.

460.11(R1) 3 W- Westln house

l

- 4NRC REQUEST FOR ADDITIONAL INFORMATION

=

Response Revision 1

c. %+-tw4abl+-gmamg-unit Encapsulation of spent filters is by mobile or portable equipment. -i',-*4im mny: alieated-in Fc AR Tat 4+44. i '2 f c'- a !L n 1. * :" 'e p r =d * ; perfenn

. ype viric ' u! :ht deta" -d de-.!;n !!' ' - by !E- " m e;! - W: '

!!! !%Ie4gned ! : prc;; e 2:

, . . ai ..g g r gn , w,.. 535 p, app!: m. u. 9 pit , ., , ..-d the : i*ing P :" '

': pd r" '

c: adding-watee-end ; - nnd pu np ';.pe -d; c *ng. !!! he c'-Jed far ef4kient -

nJ*ing : nd p: np:ng of- ide rage - r er- -- gn ' groot t !!! be d ': , tJ * ^- '

th: ughn'me '. f' g n

ng ep; atix :' *- ' and - N - "'d

'e ' ed m de ---d^""' '-

dd tw!irh:,rged th: agh the p: np nd Fm * ' " ga!!:w. dr ' r~ - a!!!ng und-dun ' ng. W >:!ed : '!Js

.33 p_ .x. .g ..a g:.g .;a - nr*di: J a- - '.

'*'~r"~'"

d. A*4ndi; -d Chapter !2, c -J . !!! " the '""i '3 qu e:ne-' Th catr-the-goalr estabasneo tor low-levet ary and wet waste volumes of I ou tt3 '; r- PWR . - ' ~' ^c p,:fonne th-- hm !4p1 ' "r ; :::r :!'; up; - ting 4tWR p!- ' c

c.

The AP600 is genceally smaller and greatly simplified relative to current plants, including :.:gni - : reductions in the quantity of valves, pumps, and other components requiring maintenance, the generation of solid wastes for the AP600 should be within that produced by the best 10 percent of current plants. -h AP600 uses mobile or portable processing systems for wet and dry solid wastes, allowing the most efficient volume reduction methods and equipment to be used. alw-4+a*

integra'd u!!d - + " nd upp:-' , *. nd fue!!!: . " e ' rd ^ ' * " ; m

pn a r R; ,

-" faeili: -" !! awi4 F' :

rent!y upnating p!- * "-

ehaiqu d -c-ih" :-

Apredi+4M4:ap: - !2 :.c a .i' n:.; ge;;; .,_ , . ry . _ , _ _

, . ; 3.. 34p7_;gm m e c_

The AP600 waste quantities are based on dewatering wet solid wastes and does not include solidification.

u. _ o .a. y yn. , t :u;gc_.+_. .u. u, ;qu . i .

, Solidification of y spent resins would result in no change in the waste volume shipped from the plant. The vinyl ester styrene binder process does not increase the waste volume but fills the space existing between resin beads with binder. Reference to resin solidification will be deleted from the SSAR ,

The overall performance of the nuclear power industry in reducing the volume of solid radioactive waste shipments may be observed by evaluating the historical data provided in NUREG/CR-29071 . The averace solid radioactise waste shipments for all operating PWR power plants from 1978 to 1988 is about 106 x 10~4 m3 /yr per MWe-Hr. ine uniity Kequirements Document (UKU) solid radioactive waste goal of I Du !!bf--i*

equn alent to 1.16 x 1u E 763 ? F "*e "- hr -d - : e W "", p!:

cpc- *ing i -- "- p espany-Jaetor-of Se p --

Thus, the 1978 to 1988 average waste shipments for all PWRs is about nine times greater than the AP6001750 ft3/yr 44RD goal. The average solid radwaste shipments for all PWRs reduces to about 4.4 x 10-5 m3/yr per MWe-Hr from 1985 to 1988, which is about four times the AP6001750 3 l ft /yr ilRD goal. This reduction by over a factor of two shows the results of early efforts Fy !!!!! . : to reduce the volume of solid radioactive waste shipment: d". *^ api;"y n; dH;- ' . t -d den n; dispe ! - *. :!u;r !!" - ---

I NUREG/CR-2907 (BNL-NUREG-51581 Vol. 9) Radioactive Materials Released from Nuclear Power Plants Annual Report 1988, published July 1991.

460.11 (R1)-4 W-Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATIONS -

Response Revision 1 Some plants with aggressive radioactise w aste solume reduction programs did much better than the average.

For cumple. between 1485 and 1988, Diablo Canyon 1 and 2 shipped an average of 1.03 x l(r5 m3/ year per M We- H r. 3 This is about 10 percent less than the AP6001750 ft /yr URD goal. W : Ca! " "'Mf !q!eh has had a program for soTIJ radioactne waste reductmn m place smce mw, snipped anout ; x nw)T59 per-A4 W44 er whk4+4*.-*h > ! .' :ime erw ter-4hr the ! HD-gaa!. Fu- !"F 7 uw! !"RR u he the pryme-was in-64! cpr- ' the shippai- *4ki-cub *w44n '

n! !<pc. g: tater-4hn the !!RD-pah Th. e **mplan*w.tra t e t h:" With aggressive solid radioactive waste programs, current nuclear power plants are being operated within the AP6001750 ft3 /yr IAD goal for solid radioactise waste volume.

It is expected that evaluation of shipped radwaste volumes since 1988 would show continued reductions as **w plants inne. th ir ' minimize their solid wastes *th*.efTer4s.

For the AP600 there is a large reduction in the number of components (pumps, valves, etc) that can become radioactively contaminated. This will result in a large reduction in the generation of solid radioactive wastes due to maintenance operations. The waste generated during maintenance operations is a large fraction of the volume of dry, compressible waste and contaminated equipment. For Diablo Canyon in 1988, this waste accounted for about 89 percent of the total radioactive waste shiprnents. For an AP600 radwaste estimate, this type of waste accounts for about 79 percent of the total expected annual disposal volume (SSAR Table i1.4-4).

hs*kii6*n4o4nbrent4mplif:e ' the iP!M F . feature *4 hat pe- ' ' cnt- at-+e s c'c nf-m: aria!1 - uhr:

than-dig *me! The !cumiry e'"'- two !'< p; x! tapa.

p: thr ugh u . ^' ~'e " *- '

fow-444-pouml-e*pm4w-4r+wwhat-4*gethw-pret - high-them ghpu' .apehi!!! F ^"- cc@ie itene 'ehwm-re "" : !arge-m4kkadu c ge---*"- ' - N ep! 'ed !!h . ^'^ item ,-Thk+-4*4mb

^

i4#w- ' un: !!: -bags e. -d ! ':4!e

e!!d - t und pe- e! pretee+ an gar . ' a e.

The r np:- der +1emk:g : :d dm * *

fac!44 tie; pr ided i ^ ^ r!u * - hulhhag i !!! i erw* .*4he-+#w.e

f ace '- :d ; mpone >

nd uili-alla de entandnation af ite fa- di p a!. :di: '

H e e!ran wmae+willeathm4*editmendte the :ncti:num egrega' cf: .r=Wu ' '

fr: - adioae: . t Thm, -4he,W4m4r 'eg: ate 4w p*44fities n50 - ge ^ ally " n!! : "^h!cet curree' eiear-poweeplant*)

to-mkun - the ai:! : u!!; . ' . - w li:pc ^' n!"ne hv4he++,pon ,#4a4) M. %41M4+mlieaetiu -t '^ragMaeilidee.-pro *4ded-we4he radu *e buikkey*we

, a!n %#1anve-to-th; 10 da

age dura!! :-pribd : "rnr 4 Technien! Pc ' en FTER ! ! 1 He aorag+-duranom.-4e+ four ge va! 'y pe of c!!d n!!: .* t' ed the ::pe.'ed an n! a!u t w4*n. af ES A R T*lde-WY c r %!!r D 'imated PA ged Radu le R n ' - - reevr em are paragon-Spent Ion 1:xchange Kestns 12 months 1 Miter-4'imreoal s

1

- Reducing LLW Generation at Calvert Cliffs, Nuclear News, March 1988.

W-Westinghouse

.m.. _ __ _ - ._ _ - ~m. .m _= - - - . .. . . .

- 6NRC REQUEST FOR ADDITIONAL INFORMATION l

P s Response Revision 1 I

High-Actitil) hlter Lartridges I / months I Uther Wastes m Urums 45 months

3 wastes m tsoxes Jb months

age dwath : ec -c ::h e!; , n! * ^'three r- t '

ge weh 2 P- . dr -!5: w w k-:.:re* age-. '"1 :.pnec n!'m * ' " . ne f r~ . amp! , d r"-d ; u!d heine- -d ! c, .nc "

> age du * - , hi'^- 'e age du- ' en fa - "N - ( for - np!e, ' :m; uld ;' u ., nsag ge era:: r ^ '- . ' :. :ne .

e. Chaple- !2 rr the t hi!!tyRequ . . '-

^^~ . ' : pxik:. p.'ne xpe-^" - en!y fer p!^-' -* e ' chtenh w-we "- " ' ' -

th: ' c;ea 21:+ - - 'udge er sh:rry far:r indie ^'d ' ':R D P- graph c.2.2.2.2 c' Chapir- !2. *- "eJ hshe Ret'--!e fn- -: paragrcph, PWR p! n:: '" .. r 'y c^rtr:dge "N o de no:

m p4 m : "' The AP600 employs only cartridge Elters and does not generate sludges that require settling and decanting.

f. There are three few locations where packaged wastes may be stored until shipped to a disposal facility: (1) the two spent resin container Gil stations, (2) the onsite storage casks,0)e high ^ck '; E!!er pr; e, ng c~

and (3 4) the packaged waste storage room. He per'~- - _ ^r :hc - '

age nr. ^~ h en de .eribe ' ' 'N p: 'n Q-!6n c e ;! Q MO 14dr The following is a physical description of each storage area.

He spent resin container fill stations (SSAR Subsection i1.4.2.5.2)is a er+4we cells with thick concrete walls (See SSAR Figure 1.2-29). " :' c" ^'- ' 'O '^

'er' high, A thick shield cover, with lifting provisions ami-Weldf ugged l pr . r_ _ c n t ad re^ : and : car and de~ ::: ". . , forms the top of the caeh cell. TM-14 atfor:- ^' e'e ^:ica ! !"' 0" . des:gned fn ^!d e- . : ' , ;'^ - nd pn 'ide ? :.pe. .

,d the top nf the ;e"'

The onsite storage casks are described in SSAR Section 11.4.2.2.6 and Table 11.4-12, l Er high - a ::y "N r pr . ng e ' ; de c"rd i- Se iP, S+ e' - ' ' i .2.2. ' t and Tab'e ' ' ' !2 'c'e/

b ,

The packaged waste storage room (SSAR Subsection i1.4.2.5.2)is a shielded, unobstructed area 21 '-v ide ,

'; " '-e !cng and has a clear height for stacking waste boxes or pallets of drums er " 'e "' '^e The mobile systems facility crane :'e!ded 'crk !!f: 'cS '.P. c+re: - ' ' i.2.2.2 rnd '"^c ' ' i !2 !c'a ?), is -

used to handle waste boxes and palletized waste drums into and out of storage (SSAR Subsection 11.4.2.3.3).

Planned positioning of waste containers and portable shielding may be used to minimize the dose rate in the-portions of the area som periodically accessed by personnel. Mobile racks for hanging lead blankets or emi shield panels on casters giay be used 3 ; "^h!: for flexible response to changing conditions in the storage area em.m.

460.11(R1) 6 i W

Westinghouse Y

a- - ,-- c. w ,, ,.w--w. - e, ,% .- ..u-i,. , -- --mp.

7

, .--r-b ---m -r -* -'-

NRC REQUEST FOR ADDITIONAL INFORMATION7 -

Response Revision 1 Although the ALWR Utility Requirements Document (URD) is not a regulatory requirement, -T the storage f acilities are in conformance with URD Chapter 12. Section 5.4-+wepHha: - '

ewing:! ? D N ae*ph L ! .1 ! ! ! ' p w. i;!<d i:!e the high * .4y-fi!!r pr ng . - " - th; '

, ' ap a, ' . -

R*e m 64rae!; the :atent' " :b- r:0 ' id par:s ; r :J - ' pr: ~ hy-*m4-wr. ey ng an b- -

rem +4e4p4w-4hr -s iag-tw:?'c TV , .- f .

?. ! - the '^rd *- sp; ' th+a c.pspmenH4JR1144*as*f4 i 5 3. 2 5l e m m.k! c - ' "" :-wp; ibin', " :he-COL appliaano SSAR Revision:

Revise SSAR Section 11.4.1.3 as follows:

11.4.1.3 Functional Design Basis The solid waste management sy stem is designed to meet the following objectives:

Provide for the transfer and holdup of spent radioactive ion exchange resins and deep bed filtration media from the +*rL++ ion exchangers and filters in the liquid waste processing, chemical and solume control, and spent fuel cooling systems

Provide the means to mix, sample, and transfer spent resins and filtration media to high integrity containers or liners for dewatering or solidification by mobile processing equipment, as required

Provide the means, by mobile processing equipment, to change out, transport, sample, hold up, and package high-actisity filter cartridges from liquid systems in a manner that minimizes radiation exposure of personnel and spread of contamination

Provide the means, by mobile processing equipmes to change out, transport, dry, sample hold up, compact, and package lower-activity filter cartridges fror. ' .luid systems

Provide the means to compact and package spent filters from the plant heating, ventilation, and air-conditioning systems

Provide the means to segregate solid wastes by radioactivity level and to store and process the wastes as appropriate. including segregation and compaction, by mobile processing equipment, to achieve etficient volume reduction Provide the means to characterize and accumulate nonradioactive hazardous and radioactive hazardous (mixed) wastes for subsequent processing s

Provide the means to segregate clean wastes originating in the radiologically controlled area (RCA) and to verify that they are nonradioactive i

1 460,11(R1)-7 !

W Westingtlause '

i l

- SNRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1

Proside the means to sursey, decontaminate. and weigh packaged wastes and to store them for at least six months in the event of delay or disruption of of fsite shipping

Provide the capability. using both pennanent and mobile processing equipment, to limit the packaged volume of primary low lesel waste to 1750 cubic feet per year when augmented by efficient operating techniques to minimize the generation of radioactive and nonradioactive wastes in the radiologically controlled area

Proside the means to package radioactive solid wastes for of fsite shipment and disposal according to applicable regulations, including Department of Transportation regulation, 44 CFR 173 (Reference 1) and NRC regulation,10 CFR 71 (Reference 2).

Provide the means to direct liquid radwas.te from solid radwaste processing to the liquid radwaste system for subsequent processing and monitored discharge Delete SSAR Section i1.4.2.2.15:

44. L 2.2. ! 5 b ah!e G: ng U -

The p -*nH gn ting : ;'nsWAing gr *

' ! pr n;4ng :> ' E!!er dis;x ^ ' - >

s- " -

Nat*4*!!"3 " hen q:: ed b ' - fn - ,-* s Revise SSAR Section i1.4.2.3. las follow s:

11.4.2.3.I Spent Resin Processing Operations Demineralized water is used to transfer spent resins from the wsw ion exchangers and bed E!!=. Although a bypass allows spent media to be transferred directly to a waste disposal container, the normal mode of system operation is to transfer the spent media to the spent resin tanks. Before the transfer operation. it is verified that the spent resin tank selected and aligned as a receiver has the capacity to accept the bed. It is also verified that the resin mixing pump is aligned to discharge excess transfer water through the resin fines filter t, the liquid waste processing sy stem. Either spent resin tank can be designated as a high-activity tank to receive resins exceeding a preset limit

fe e t n;+ -when !T R4W.

During the transfer operation the tank level is monitored and the resin mixing pump is operated, if required, to limit tank water level. The operator stops the transfer when the CCTV camera viewing the sight flow glass indicates an : c n!pe '- - tx that the sluice water is clear and the transfer line is the+fwe flushed of resins.

Af ter the bed transfer, the tank solids level can be checked by operating the resin mixing pump to lower the w ater les el below the solids level. The solids level can be determined by the bottoming out of the ultrasonic surface detector recording. ,

4GO.11(R1)-8 W-W85tiflgh00S8

NRC REQUEST FOR ADDITIONAL INFORMATION9 -

Response Revision 1 Iletween bed transfer operations the water level in the spent resins tank is maintained above the solids lesel.

Demineralized water is supplied for water level adjustment as well as a backup water source for Hushing resin handling lines after resin recirculation and waste disposal container filling operations.

The solids bed can be agitated and mixed at any time by using compressed air or by operating the resin mixing pump in the resin mixing mode. In the resin mixing mode, water is drawn from the spent resin tank via resin retention screens. The water is returned via tank mixing eductors that generate a resin slurry recirculation in the tank equivalent to about four times the resin mixing pump capacity. The solids bed is locally fluidized during this operation.

Before transferring spent media to a waste disposal container, the disposal container is placed into M the twe waste container fill station using the radwaste crane. The waste container fill statione is a shielded enclosure

for the spent resin disposal containers during filling operations. He ;e"~ n!!=: - ; g: - !k;uid

_.g gm;.7 _r.t. r_

94. ._ i _ n . _. .. i . _ . _ .. .,

a ., _.._:.,_ rt, t_ , .

.unteir - 9hg, ?c '-ing, 2nd ' , ' ng are . c d : -e;e"-

Re ? - :tesg +-dpipMram F c-": - ndc '- r nee! !. ""A e fre r t he E!! ' ' 'e !c m

!r '-ing ; ./ H e : !c ^ ; ^ c-- nd cr :d the "", de : r:ng, nd - * ' . . 1 ' ched in the41!! F ' Er "" F nd i met ed 5) " handling dev _ . - t h e r -" - - nd: ' -d ' e . e ": r p: -' 'r5:

,c""'

u!" '

< : hen uwd : rp! u 'd r - .

Th+esn-aning : n't . p r. am4y "c-e-!hed, . anb" 'd 'nen!!y "":d: . nd : dn th:- " t 'edinthe

u. : _ m .a umr__ a _: . , E!!!ng, The resin transfer pump is then started in the recirculation mode. A resin slurry is drawn from the spent resin tank and returned to the same tank. A representative resin sample may be obtained during recirculation or container filling modes 5; cp - ' ng the ' "np!!ng de ce fer : pr : ;n :d ? ! . .

He container fill valve for the appropriate fill station is opened to initiate the filling operation. c!! ';:

opew: n!v if a !h - * 'ch : - the E!!4: cad :ndk"- the the E!! ' ad i "'!rd e - '

. r ": r- Er :-

dewetering-p+::np .aer'ed in 6 e . .: '

, , - " ' " :- :F en:: ire- 'r'r In de '

ng pump

.!F e' ge - 'e- te *'; . "'" !ine. The transfer water flows back to the spent resin tank, thereby presersing the water inventory in the system and retaining any resin fines or dislodged crud within the system.

The resin mixing pump can be stopped at any time during the filling operation. When the solids level nears the top of the container, d " J -d ';= 'e c' nen- - --! Med by a !:'c mic ; na, the fill valve is closed and cycled to top off the container. Excessive water or solids level automatically closes the fill valve.

When the filling operation is complete, the line- Dushing sequence controller is manually initiated to automatically operate the pumps and valves to Hush the resin transfer lines to the waste container or back to the spent resin tank. The resin mixing pump supplies filtered flush water from the spent resin tank. Er d==:cing f*F ' 3Fr'"'ed1*h"" ^ ' !) ' -e " '~ d:23 " ^'" ^^ furt'e- de at: n; a c d " e -d '; : p: np diveharge pr ure d

" "nr an '- - " ndic *:m4re c ;'umpv For dewatered resin shipping and disposal, the waste container is an NRC-approved, high integrity container designed and verified to dewater the solids bed to less than 0.5 percent free water. For solidified resin shipping and disposal, the waste container is a carbon steel liner. He resin is solidified e ::p1"'"rd 5;> c":ng ': :r-* '- "'

NPT app: ed E:-der 4for ;-- np!:, ny! ; ' - 13 rre), using vendor en-"rce' -supplied equipment : : furthe d - -ihed i cuhw4ir ! ! '.2. i '

+: E!!!ng rad b :::c ng cpn: ' n: e np!". , : E!! Had is :e-"r'; dke ved f : : -de

. ": _', _ % . 7

. . :a u . t.

n ng ge rh dc ci ; ' . cu'i :!de- - ' .

u- b; g 460.11(R1)-9 W WestinEhouse 1

l

IWRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 1 a**4,-%4m44hameh-*-tw+4ts):. 'ir!d ; .- Er; ":ine 'id is - *n!!ed :hn gh *F- aFF-epmnag h si: .hicL! a w er-1: a 4 g-*4*na 'r!t ep;- d-eaptw F n!!ed h; ; jib c -

The top and sides of the container are then smear- surveyed through small ports in the shield cover. If contamination is detected, which should be an infrequent occurrence, demineralized water is sprayed on container surf aces through a set of fixed cleaning nozzles.

The cleaning w ater drains to the radioactive waste drain system where the water is directed a nr :r'; np and Quwpe=1 to the liquid radwaste processing system - ' ' * !:t ' ; "n!

^' if;c When-We - *

' dew ! -iErd : '#

1 200 di+*-li"sts fn- 5ipp;ng, or hig.h he m

.'; ,! af pc- -! 4*wlahe-4,h*44h - ~'" ' 3+4n re cpr--*Hn : . r

rn!!cd-fn - - ~' :- Se rmiw* t#4miktiopm4++4*wn-wiule-motstoring opern::- w4 r ed ; . **-!*vi - E .hieh!; c e - - ted tw-u hwshe-#mtwa * .- r*!c- -

ant:n! He p: . er grapple nf er radu 22: r:n. Sen ; etc We the ; ' -

!ine, n! &; . ' -

'ifted n! " re - rd ' de

age t

% -':notmiine dynamat:: '- Se redwa'ae-e+an - :gh or;nnt - - 6: ring-th - !. .r,7 77 ,, ne 7;gu ,_; 23, mi, u ,essmet 4*ee-the- i:: % *

,:.L '* =! > :L : age :.L -

The-ra.!u '

.- .*h,m.p**techbwemoir ; "a! : ' rnn. r ._ _ . , . _ : n,_ , r7 .t_ , . .ag, g so-4 mil *-**.**teeW4sppin g ; ' - Er adu '. crane En ,d!r i. bipping ; :h - ' . a!= 'ndle-4he 4*pph:; mL if thi '- u!d-prove-nec -

fe- trailrr . ' mai: ' c.

Revise SSAR Section I l.4.2.4.1 as follow s:

Il.4.2.4.I Portable and Mobile Radwaste Capabilities Portable or mobile systems can be located in the mobile systems facility ** for processing and packaging chemical wastes. Chemical w astes are normally processed in the radwaste building by a mobile solidification sy stem w hen a batch accumulates in the chemical waste tank, The spent resin processing system provides ineh*les connections to in the cadwee-huikimg mobile systems facility in the auxiliary building train bay to allow spent resins to be delivered to a portable or mobile system for packaging by dewatering or solidification, using radwaste systems provided and operated by contractors.

Connections permit eenteni; : tie-in of resin transfer and dewatering lines that may be required to an optional f acility for w hich space is reserved adjacent to the radwaste building. EF rp:!cna! ' !14ty . u!d be required-for ,

entain-plant.,-wlme reg:: u! !n '- c' dispe n! facilitie*,-4eepse o!!diEcatinn of spen' - Ha .,

w4kli6eason-e*n4m-p*fomw.ein she radu 'e build:ng n high het . :nchi!: ,

Maeilig by .ing a pc-' 'i r mnt : :n : qualineJ44* ding age * - ' : pe nte4h spr * . :- ~ '. . ntai .- n -'hi:prm

eentaine . - Er inside . e ; mtain#e-fd! *

- : an ne

34 s

l 460.11(R1)-10 W-Westinghouse e

1 1

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Question 460.15 Section 9.3.3. I 1.5.3, and i 1.5.4 of the SSAR provide incomplete information on radiological sampling provisions for process and ef fluent streams. For example, the sampling provisions for the waste monitor tank contents, the detergent waste monitor tank contenis, the steam generator blowdown, and the condenser air remosal sy stem have not been identified. Further there is no reference to tritium measurements. Identify how the sample provisions for the liquid and gaseous process and effluent streams for the AP600 design meet the sampling provisions for such streams identified in Tables 1 and 2 of Section !!.5 of the SRP.

Response (Revision 2)

Table 9.3.3-2 of SSAR Subsection 9.3.3 will be updated to include the missing nu local sample points in the gamey ami secondary systems. he ed !*!e i !!! he : " % in M 1" SSAR Tables 9.3.3-1 and 9.3.3-2 include tritium as one of the radioisotopes which is analyzed in both primary and secondary systems sampling.

Sampling is performed to measure those water chemistry and radioactivity characteristics which must be monitored.

The ranges and accuracy of analysis will be appropriate for the water chemistry characteristics being measured.

These monitoring frequencies are selected so there is sufficient time to detect chemistry or radioactivity changes before any adserse affects occur.

SSAR Tables 11.5-1 and i1.5-2 list the radiation detectors in the AP600. Tables 460.15-1 and 460.15-2 identify other SSAR items which corresponds to the SRP Tables I and 2 sampling and monitoring provision items, s

W Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table 460.15 Gaseous Radiological Monitoring and Sampling Number item SSAR Reference 1 Waste Gas Holdup System Section 11.5.2.3.2 Table 9.3.3-2. items 25, 26, 27, 2S Table i1.5-1 2 Condenser Evacuation System Section 11.5.2.3.3 Table 9.3.3-2. item 29 Table 11.5-1 3 Vent & Stack Release Pt. System Section i1.5.2.3.3 Table 11.5-1 4 Containment Purge System Section 11.5.2.3.2 Table 11.5-1 5 Aux. Bldg. Ventilation System Section 11.5.2.3.2 Table 11.5-1 6 Fuel Storage Area Vent System Section 11.5.2.3.2 Table 11.5-1 7 Radwaste Area Vent. System Section 11.5.2.3.2 Table i1.5-2 S Turb. Gland Seal Cond. Vent Included in item 13 System 9 Mech. Vacuum Pump Exhaust included in item 2 (Hogging) System 10 Evaporator Vent Systems N/A i1 Pre-treatment Liquid Radwaste included in item 5 Tank Vent Gas System 12 Flash Tank and Steam Generator Flash tank - N/A Blowdown Vent System Steam generator blowdown - Section 11.5.2.3.1 Table 9.3.3-2, item 8 (See also Table 460.15-2 item 16) 13 Turbine Bldg. VerIt System System description in Section 9.4.9.

14 Pressurizer & Boron Recovery Pressurizer - N/A See note 1.

Vent Systems Boron recovery - N/A 4GO.15(R2)-2 W Westinghouse 9

NRC REQUEST FOR ADDITIONAL INFORMATION I

Response Revision 2 Notes; 1. Pressurizer is vented to sampling system, reactor coolant drain tank or containment atmosphere, s

" .,sm23 3 w wesungnouse

NRC REQUEST FOR ADDITIONAL INFORMATION Responso Revision 2 Table 4tio.15 Liquid Radiological Monitoring and Sampling Number item SSAR Reference 1 Liquid Radwaste (Batch) Effluent Section i 1.5.2.3.3 System Table 9.3.3-2, itenu 12.,16,17,18,19. 20, 21, 22 Table 11.5-1 2 Liquid Radwaste (Continuous) N/A Effluent System 3 Service Water System System description in Section 9.2.1.

4 Component Cooling Water System Section i1.5.2.3.1 Table 9.3.3-2, items 13,14,15 Table 11.5-1 5 Spent Fuel Pool Treat. Syst. Table 9.3.3-2, item 9 6 Equip. & Floor Drain Collection Radioactive floor drain system included in item 1.

and Treatment Systems Clean floor drain system - System description in Section 9.3.5. See item 20.

7 Phase Separator Decant & Holding N/A Basin Systems 8 Chemical & Regeneration Solution Rad./ Chem. lab waste included in item 1.

Waste Systems Regeneration chemical waste - N/A 9 Laboratory & Sample System Included in item 1.

Waste Systems 10 Laundry & Decontamination Waste l2undry - N/A Systems Decontamination waste included in item f.

1I Resin Slurry, Solidification & Spent resin - Table 9.3.3-2, item 23 Baling Drain Systems Solidification & Baling - N/A Table 11.5-1 l

12 Radwaste Liquid Tanks (outside the N/A buildings) g 13 Storm & Underdrain Water Syst. The storm drain system is site specific.

l l

460.15(R2)-4 3 Westinghouse

NRC REQUEST FOR ADDITibNAL INFORMATION Response Revision 2 Number item SSAR Reference 14 Tanks and Sumps inside Reactor Reactor coolant drain tank - Table 9.3.3-2, item 11.

Building Containment sump vents to containment atmosphere.

15 Baron Recovery System Liquid N/A Effluent 16 Steam Generator Blowdown (Batch) Section 11.5.2.3.1 Liquid Effluent System Table 11.5-1 17 Steam Generator Blowdown Section 11.5.2.3.1 (Continuous) Liquid Efnuent Table 9.3.3 2, items 8 Sy stem 18 Secondary Coolant Treat. Waste & See item 20 Turbine Bldg. Drain Systems 19 Ultrasonic Resin Cleanup Waste N/A Systems 20 Non-Contaminated Waste Water & Section i 1.5.2.3.3 PWR Turbine Building Clean Drain Table 11.51 System SSAR Revision: Table 9.3.3-2 is to be revised as shown below:

a 48 .,s<n2ns w wesungnouse

NRC REQUEST FOR ADDITIONAL INFORMATION .

l G

SI .

Response Revision 2 Tab!e 9.3.3 2 (Sheet 1 of 5)

Local Sample Point Not in the Primary Sarnpling System (Normal Plant Operations)

Avail-able Ty pe of Number Sample

Process Measurement Sample Point Name of Points Liquid Sample

1. CVS Boric Acid Tank i Grab pH, chlorine, Quorine, boron, silica, suspended solids radioisotopic liquid, oxygen

2. CVS Boric Acid Batch- 1 Grab Boron, chlorine, Quorine ing Tank

3. CVS Letdown 1 Continuous Radiation monitor (See Section 11.5 Table 11.5-1) 4 Residual Heat Removal 2 Grab Radioisotopic liquid, suspended solids, Heat Exchanger radioisotopic gas, gross speciuc activity, strontium, iron, tritium, hydrogen. I-131, conductivity, pH, oxygen, chlorine, fluorine, boron, aluminum, silica, lithium radioisotopic liquid, lithium radioisotopic particulate, magnesium,' sulfate, calcium, lithium

5. PXS IRWST 1 Grab pH, oxygen, nuorine, boron, conductivity, gross specide activity, sodium, sulfate, r silica ,

6. Main Steam Line i Continuous Radiation monitor (See Section 11,5 Table (Outlet SGl) 11.5-1)

7. Main Steam Line I Continuous Radiation monitor (See Section i1.5 Table (Outlet SG2) 11.5-1) 460.15(R2)-6 W westinghouse .

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

NRC REQUEST FOR ADDITIONAL INFORMATION m

Response Revision 2 Table 9.3.3-2 (Sheet 2 of 5)

Local Sample Point Not in the Primary Sampling System (Normal Plant Operations)

A vail-able Ty pe of Sample Point Name Number Sample * ?rocess Measurement of Points

8. BDS Steam Generator 1 Continuous Radiation monitor (See Section 11,5, Table B!nwdown i1.5-1) 9 . 31 . SFS leops (Upstream 2 Grab Conductisity, pH, chloride, silica, corrosion of SFS Pumps) product metals. gross activity, corrosion product activity, fission product activity, io-dine-131, tritium, turbidity, boron, corrosion product metals, organic impurities

10. 4 PCS W ater Storage 1 Grab Hydrogen peroxide Tank

11. 44-RC Drain Tank i Grab Gross radioactivity and identification and concentration of principal radionuclide and alpha emitters. Dissolved gases. State and federal environmental discharge requirement such as pH, suspended solids, oil and grease, iron, copper, sodium nitrite

12. 44. WLS Degasifier 1 Grab Dissolved gases.

(downstream of degasifier discharge pump)

13. 42. CCS Component Cool- 1 Grab pil, sodium, chloride silica, corrosion ing Surge Tank product metais, corro , a inhibitors

14. 4A CCS Loops 2 Grab ph, sodium, chloride, silica, corrosion (downstream of CCS product metals, gross rawoactisity and pumps) ,

identification and concentration of principal radionuclide and alpha emitters 460 m 2p W Westinghouse I

. . . . . - . - ~ .

- NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 -

Table 9.3.3-2 (Sheet 3 of 5)

Local Sample Point Not in the Primary Sampling System (Normal Plant Operations) t t

Avail- i able Type of Sample Point Name Number Sample

Process Measurement of Points :

~

15. 44, CCS Hot Leg (upstream i Continuous Radiation monitor (See Section 11.5, Table of CCS pumps) 11.5-1)

16. 44. WLS Discharge 2 Continuous Radiation monitor (See Section 11.5. Table 11.5-1)

17. 4. WLS Effluent Holdup 24 Grab Gross radioactivity and identification and Tanks MlD5A,B concentration of principal radionuclide and alpha emmiters

!' "'!S Elihu .uu up 4 % g .. _ .. f , . .

, .._ a ;g,_.g c,_.:;._ . g T-- " 'TD5H . r -ad :wf-pr ::pr' ndir :':" - nd g p t. .. . :.77; i

17.43 WLS Waste Holdup 24 Grab Gross radioactivity and identification and Tanks MT06A,B concentration of principal radionuclide and ,

alpha emmiters 39 "'ES "' " :!hp 4 Geh Gr ::di; -

2n"ide*:c e ::-- nd Tu l "'rr"- H c er- - ' --- 4 pr: :: pad!: ^!!de ---d

!ph: ; '. -

18. A WLS Effluent Monitor 2i Grab Gross radioactivity and identification and Tanks M TU7A,B concentration of principal radionuclide and alpha emitters. State and federal i environmental discharge requirement such as ,

pH, suspended solids, oil and grease, iron, copper, sodium nitrite n

I 460.15(R2)-8 W-Westinghouse

..e., , ,-, - , - _

g

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Tank-41W14 a.n e r n ' '

f pc c:pn! : 8:-

!!de .::W alf4+wnitt c rs. P' nd '.!c n!

eas**nmtaWii: c!:: rg-c ,quirr '

eh g44, .:.p nded .e!!d - :i! nd gr.

in%

. ::ppc , :di: -i:e a

W Westinghouse 460.15(R2W

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table 9.3.3-2 (Sheet 4 of 5)

Local Sample Point Not in the Primary Sampling System (Normal Plant Operations)

A vail-able Ty pe of Sample Point Name Number Sample

Process Measurement of Points

19. la WLS Waste Monitor 21 Grab Gross radioactivity and identification and Tanks M108A.B concentration of principal radionuclide and alpha emitters. State and federal environmental discharge requirement such as pH. suspended solids oil and grease, iron, copper, sodium nitrite 23 W1 c W: *

.S *-

4 "-

4 .h :Ji: c:4@nJ4de shad Tek-MT4*H eenenansen nf pr cipa! radi: c!!de nd c'*

alphe-emnter n dend w.hwnen a c' .rge - _;wir. -

c' pu, 7733_, g . .i:a. .:i .. a .

. .L... o.

20. 24-WLS lon Exchanger i Grab Suspended solids Gr radi: *

('

Pre-filter (downtream) :d irie-+: - :J ; e -- r p: c:pa!

Detspn W ' Tank rad: c': AcA n!;+ m:ner .

21. 2A WLS lon Exchanger i Grab Suspended solids G: Ji: c M ay-*nd A fter-filter ide,4ften' nd r- 'r- -': -f pr: c p.J

-^ c

(downstream)D*+*yent c!!de : > !ph: ; "-

nd W - a t- '-Tek (edwal-wisc --*c! diwharp rye s+eWH . rpc- d -d s:4ie :' :d grec , - appe- ndianwniese

22. 24+- WLS Chemical Waste 1 Grab Gross radioactivity and identification and Tank concentration of principal radionuclide and alpha emitters

23. A WSS Spent Resin Iank i Grab Gross radioactivity, radionuclide (liquid) concentrations G e.*nw&untde 4GO.15(R2)-10

[ W85tlngh0USB

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2

~_ '

T*nk w r: ,

'1M a

460.15(R2)-11 W8MM@00S8

NRC REQUEST FOR ADDITIONAL INFORMATION Response Revision 2 Table 9.3.3-2 (Sheet 5 of 5)

Local Sample Point Not in the Primary Sampling System (Normal Plant Operations)

A vail-able Ty pe of Sample Point Name Number Sample

Process Measurement of Points Gaseous Sample

24. A VES MCR Emergency 2 Grab Air quality, oxygen, carbon monoxide, Air Supply Headers carbon dioxide, contaminants

25. 34- WGS Elfluent I Continuous Radiation monitor (See Section 11.5, Table Discharge to Ensi- 11.5-1) ronment

26. -4 WGS Inlet 1 Continuous Oxygen. hydrogen, moisture A2, WG4leM 4 Gmh Nob!r ga . ':- , ;wtieulide,-tritium

27. AS-. WGS Charcoal Gwml I ContinuousGr Hydrogen Ma -

-h!c-g*Wiae Bed Vault Owilet ah twt **latW4iam

28. 24. WGS Delay Bed 24 Grab Moisture, noble gases, iodine, particulates, Outlets MV02A,B tritium l WGc. Ih4ay44eil.4halet Gmh "" *.

AC 4 rb' g; , x!h ;wtieulaten MVMil triiiam 1 24 Condenser Air Removal I Continuous Radiation monitor (See Section 11.5, Table System i1.5-1) I This column shows methods to obtain a sample for ebenwal analysis.-4441oe,-w49edf) the fu.p . af ::ng4inp*le-ngeeify * ' hwatiee44+ amp! .a!!calom " Grab" means that a grab sample is required for the intended ebenweal analysis. Depending on the sampling condition, this grab sample can be obtained in the laboratory or in the grab sampling unit. " Continuous" means that the required ehemeal analysis is performed via a probe that monitors the sampling stream continuoustb.

1 460.15(R2)-12 3 Westinghouse

NRC REQUEST FOR ADDITIONAL INFORMATION BE Ouestion 920.5 The December 9,1993, response to Q920.1 regarding submittal of the vulnerability analysis indicates that this analysis should be performed by the COL holder. The staff interprets Section 5.2.2.1 of Chapter 9 of the EPRI ALWR Requirements Document for passive plants to mean that the designer should perform an analysis to optimize system design with respect to radiological sabotage protection. Therefore, describe how the design process for the AP600 meets this guidance.

Response

The plant protection system includes both a physical protection system and a security organization. Major features of the AP600 physical protection system which are not site-specific are within the designers scope. The Combined License applicant is responsible for developing the security plan, and for maintaining a security organization.

The designers' scope of the AP600 physical protection system includes: description of the physical protection system, identification of vital equipment, layout of vital area boundaries and protected area enclosures, layout and arrangement of the CAS and SAS, location of controlled access portals, design of security barriers and security hardened walls, and conceptual design of the vehicular barrier system.

Esaluation of vulnerability to radiological sabotage is performed as part of this design process. The vulnerability evaluation will be used to enhance the plant design as required to improve the physical protection system.

A radiological sabotage vulnerability analysis report will be developed by the Combined License applicant and will consider both the physical protection system and the security organization.

SSAR Revision: NONE s

W westinghouse

1 NRC REQUEST FOR ADDITIONAL INFORMATION 1

n= ==: ,

ir 72 !

ur i 1

- I, l Ouestion 952.91 l

l Provide the following information on the CMT test facihty: l

a. Final as-built drawings and piping and iristrumentation diagrams (P&lDs) that include all recent rnodifications to the facility.

b. A current descripuon (or drawing) of the steam diffuser non.le, if still in u se. ,

1 Response: l a.

Up to date, final as-built drawings and piping and instrumentation diagrarr.s for the CMT test facility are l provided in reference 952.91-1, which was transmitted to the NRC via Westinghouse letter NTD-NRC 4244, dated July 29,1994. l b.

A drawing of the CMT steam diffuser is provided in reference 952.91-1, which was transmitted to tre NRC via Westmghcuse letter NTD-NRC-94-4244, dated July 29,1994.

Reference:

952.91-1 WCAP-14132, "AP600 CMT Program - Facility Description Report" SSAR Revision: NONE PRA Revision: NONE l

t 3 Westinghouse 952.91-1 L

NRC-94-4276, Forwards Westinghouse Responses to NRC Requests for Addl Info on AP600,including Listing of NRC RAIs

Contents

Text

SUBJECT:

Dear Mr. Borchardt:

Response

Response

Response

Response

Response

References:

Response

Response

Response

References:

Response

References:

Response

Reference:

Navigation menu

NRC-94-4276, Forwards Westinghouse Responses to NRC Requests for Addl Info on AP600,including Listing of NRC RAIs

Text

SUBJECT:

Dear Mr. Borchardt:

Response

Response

Response

Response

Response

References:

Response

Response

Response

References:

Response

References:

Response

Reference:

Navigation menu

Search