Text

Forwards Final Draft Responses to Radiation Assessment Branch,Effluent Treatment Sys Branch,Auxiliary Sys Branch & Power Sys Branch 820825 Info Request
	ML20070H328
Person / Time
Site:	05000447
Issue date:	12/17/1982
From:	Sherwood G; GENERAL ELECTRIC CO.
To:	Eisenhut D; Office of Nuclear Reactor Regulation
References
	JNF-54-82, MFN-194-82, NUDOCS 8212230328
	Download: ML20070H328 (179)
	v • d • e

{{#Wiki_filter:y GENER AL h ELECTRIC NUCLEAR POWER SYSTEMS DIVISION GENERAL ELECTRIC COMPANY.175 CURTNER AVE.. SAN JOSE. CALIFORNIA 95125 MFN 194-82 MC 682, (408) 925-5040 JNF 54-82 December 17, 1982 U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation Washington, DC 20555 Attention : Mr. D.G. Eisenhut, Director Division of Licensing Gentlemen:

SUBJECT:

IN THE MATTER OF 238 NUCLEAR ISLAND GENERAL ELECTRIC STANDARD SAFETY ANALYSIS REPORT (GESSAR II) DOCKET N0. STN 50-447 Reference : Letter from G.G. Sherwood (GE) to D.G. Eisenhut (NRC) pertaining to draft responses to the Commission's August 25, 1982 request for additional information on GESSAR II, dated November 10, 1982. Attached please find final draft responses to the Commission's August 25, 1982 information request. Only modifications (new or revised) to the responses of the referenced letter are provided. If a response is not included in one of the attachments, then the response provided on November 10, 1982 should be considered as the final draft response. Attachment No. I summarizes the status of these responses. Except for the Structural Engineering Branch response, this transmittal completes the Commission's August 25, 1982 request for additional information on GESSAR II. As indicated on Attachment No.1, the Structural Engineering Branch responses will be provided early January 1983 in conjunction with the audit. An amendment 1 is scheduled for January 1983 to formalize the responses. Sincerely, 03 D*T Glenn G. Sherwood, Manager b , lq Nuclear Safety & Licensing Operation GGS:td Attachments cc: F.J. Miraglia (w/o attachments) C.0. Thomas (w/o' attachments) D.C. Scaletti L.S. Gifford (w/o attachment:.) i 1 8212230328 821217 PDR ADOCK 05000

/ /' / ATTACHMENT NO.1 Status of Final Draft Responses to Commission's 8/25/82 Request Branch Final Draft Response Hydrologic and Geotechnical Engineering 11/10/82 Letter i Chemical Engineering 11/10/82 Letter Radiation Assessment Attachment No. 2* Effluent Treatment Systems Attachment No. 3 Auxiliary Systems Attachment No. 4 Power Systems Attachment No. 5 l Procedures and Test Review Attachment No. 6 l Structural Engineering Submittal in early Jan.1982 in connection with SEB 11/30/82 - 12/2/82 audit action items

Attachments to this letter l

1 i l \\

ATTACHMENT NO. 2 DRAFT RESPONSES TO RADIATION ASSESSMENT BRANCH QUESTIONS 1

[ ( 471.04 Revise Section 12.1.1.3.1 of your FSAR to show compliance with Regulatory (12.1.1) Guide 8.8, Revision 3 as you state in Section 1.8. ...a.. t E + A sews <.; 2 SEc f-n c,w ! (2. l!. l. 3.1 ,,o.11.k.s_.w vesed.as i.a 4e wi d a % %i ut%4O mv k~.vy. j I 4 l

238 NUCLEAR ISLAND

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t cxT C N ew FsR. 4'71.04t) 12.1.1.1 Design and Construction Polic'ies (Continued) o I. (Description on onsite inspections to determine.that the design cnd operation keeps radiation exposures ALARA is, where required, the responsibility of the Applicant.) ...l.1.2 Operation Policies l (Description of operational policies to maintain occupational doses ALARA is the responsibility of the Applicant.) 12.1.1.3 Compliance with IDCFR20 and Regulatory Guides 8.8, 8.10 and 1.8 Compliance of the Nuclear Island design with Title 10 of the Code ?f Federal Regulations, Part 20 (10CFR20), is ensured by the ccmpliance of the design and operation of the facility within the guidelines of Regulatory Guides 8.8, G.10 and 1.8. ) 12.1.1.3.1 Compliance with Regulatory Guide 8.8 The design of the Nuclear Island fully meets the intent of Revision,Y f Regulatory Guide 8.8, and reflects the commitment of General Electric and its subcontractors. Examples of compliance with all items in Section C.3 of the regulatory guide are deline-ated in Subsection 12.3.1 of this SAR. Design features of the Nuclear Island allow the Applicant to comply easily with the recommendations of Subsection C.4 of the guide. For instance, p ; visions are made in systems such as the Reactor Water Cleanup Syctem (RXCS) to allow flushing of the piping in shielded cubicles before entry, and to use remote reach rods. Breathing air headers are provided in areas where past experience indicates airborne radioactivity has been a problem. Design provisions allow for r; note creration of fuel handling and radwaste cask filling. 12.1-2 2 ft. I 1

p '~ att.nc : ; Ou'r posttion in' Section C.1.dd) of Regulatory Guide 8.8 states that ~ (12.1.2) n licensees should propose designs which incorporate features to maintain iG' occupational doses to "as low as reasonably achievable" (ALARA) duri."g ^: deconnissioning. We state in Section 12.1.2 of the Standard Review Review Plan (SRP) that our determination of the acceptability of the

proposed design will be based on our evaluation of your proposed measures for assuring that occupational doses during decommissioning will be

'~ ALARA. Accordingly, describe in Section 12.1.2.1 of your FSAR, your proposed design considerations for miniatzing radiation doses during decommissioning including, for example, a description of your proposed provisions for major equipment removal from the drywell, process equi;wncnt removal through hatches or removable sections of shield walls and kn:ck-out walls. Twsc>5 Y.56\\\\ " md. [i dr ~~ ~ CO N emd a4 s b.ra do vn P A i i: Ai t-tis 7 wt. c w t,e op.orat.d w.A. m ainht a 1 witti AtaM e_c pes u r s wtil atto s evu b c.sstst in ac.Li.v;ag 4La M e.x p esuv u dev MeaA de-g y. c. s,s, E s = -,. 4 F b.s w kr L.,i1)..s.ist u matet : ;pl s c e m m.i s si n i ng es posuves du.<i a3 d e c..-.w.s,. ; d I*" d' " fd.ow J "c l 3 ioc.tud the ) Fo(lodig' 60 Pro.,isi., 6 A ra: n. 3, NL:o3, m d d e u b - t aab3 3 o ; f. d. d ripi n g. (k) basts" af eg aq a.*t to miaWhe. b k fluP oF radi ch Me. vial.awd % Facilitat Ask.g.P cmd %gs. i O sh.etting soLich p<..,:J trotedim bis 3 ~uie-

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t o 471.06 Provide an estimate of the airborne sources of radioactivity in the (12.2.2) reactor containment during normal plant operation, including the . assumptions you use. Describe in Section 12.2.2.2 of your FSAR, the l8 saximum expected airborne sources in accessible areas of the reactor containment following relief valve venting. Estimate the dose to personnel l at the travelling in-core probe (TIP) drives while the operating personnel l are leaving the containment following relief valve venting, including l the assumptions you use. Raceawm o-M A A. f.M eg +. smW Scb n..d.t.t - A c. cms < routive assess ~.nT J e rbo<e a ct.f ,w F.ie ~ y c m W o m.st J rtng wo r m i pl a# op emien w f u-Fr u m e.d by a.ssaming

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PA6 l l 471.07 In Section'12.2.2.3 of the FSAR, you state that other potential (12.2.2) airborne radioactivity could occur during vessel head venting and fuel movement. Explain why the entrapped radioactive gases, collected under the vessel head, could not be vented or exhausted via the gas treatment systes prior to vessel head removal. 8 es D o nse I b 4. rafp4d b d uca, irlMsi. v eu.1 h eal c 64 v.mt.4 f el o-I MI $ gnE45 e 6 k esd m e.o.l. I F is mu-.4J L y n af s (<. 6 ;e tL2 tLts pa ehlee so<n07 b< Po\\ta.aed du<i63 e.h sMng or o% < o pva% ~Lt 4 in v olve. remo<.Iof % vecrel had. i -,----,----,-nr.~.

I t i 471.08 In Section C.1.e of Regulatory Guide 6.8, we recomend the use of low (12.3.11 cobalt and low nickel bearing materials for primary coolant piping, tubing, vessel internal surfaces and other components in contact with the primary coolant. Indicate in Section 12.3.1 of your FSAR, the cobalt and nickel content of such materials. Describe in this section. the steps you have taken to eliminate cobalt and nickel from such surfaces. State whether the following design features were considered: (1) selection of alternative materials, other than Ste111te, for hard facings of wear materials; (2) limiting the cobalt content in stainless steel to a specified maximum such as 0.05 percent for reactor internals; and (3) ~ limiting the cobalt content in stainless steel in contact with the primary coolant to a maximum cobalt content of 0.2 percent for uses other than reactor internals. If these measures were considered, indicate what actions you took in this regard. ~ .e Ty p.I o n.k t awa eot..tt.c.,& sis of L p<io a p l maierials wkih ar in cenhT J.44 %e cocle.d oa.g=. h, % M e_ g.

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. 471.08 ca d;nw 0 cewiewt in siatnl.ss st..(s aseJ wis;de A v ssel h.s mei 3..n li vnit.) to c.2 7.<c.it ; he weser, ( previous . review of many vn. tea.(s c ertificdiens indicat.J an average. c abelt c o nten-t of o nly c.15 perce,d i% ausienitk stn' %l..se sieeIs, All-Cr-F .Ileys. sac.k as Inc w 16oo and raco.1 x iso whi d hav e_ bgb vucle I cowi # are. as d in some re actor o ss.1in h nal c. 9 n.nt.s. "T6 s. unteri.ls us J in argli cati.ws For u liic.L % r .. _ sf.d.I ave rey viremenis to 6 , disp..J c suc.4 as po s. ,s; n, sr. cift c ther.nal estawsiew c.4.r.ch:stics ale n3 w% adepv4fe c o vre s1.n vests %ce) ad Fo< ~kt& ao su;fal,lc alter =fa lew -ntd el m.t. eri.I is a va;lable. Co L.It e. t.at in Iwcon.( uro use.d in A bl .,s.di t. ;. li mitect to 0.o5 e.< cent. Ste.Ilite is n, d b L.<1 Rc t o g,

coge ments

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g T' 1 j i .._ j.___ _ m. 471.09 Provide a table of primary system components (e.g., the reactor pressure , ~".' vessel internals, clad, fuel, the recirculation loop piping and the feedwater ' piping downstream of the CCS) which are in contact with the reactor coolant showing the corrosion of the material (e.g., producing areas in units of square feet, a description ~"_-- * *' stainless steel, zirconium, Ste111te, Inconel or carbor. .. ' steel) the proposed cobalt content Ifrijs ex "-''-?, cobalt and the corrosion rate (in og/drr.mo) pressed as a weight percent of for each material. Additionally. _. ' :.,, ? 'd.D..".'.71' provide a table of the various siaterials used in the primary system 'e' indicate their contribution to the cobalt in the primary system, expressed ' l:,..'* e'...r. ,as a percentage; the total contributions should equal 100 percent. h G.s po ers e _Th_.wost a ynbulse. s%.d of ce L.lf inph to b BW ( y ]sys k Atck 4.a.v.ailelle w n.s c a v tsd u.t u ad e 97 a.uusl ig af. %._. Ele drk Po w << heseer d h fitute. t . b alis.of.. h 6 a.s.h d w ere. pk(;sl..d. s.s, E (MT_ alp-2263 y (f.efer mee..ll ._.. ),. hts eq ert ce ins Mw. ._ y 4.h's c W5 b ested sh N t..$nr S s Ce C t. dt. /f. RS l E uhrt<b upend +a % cool.2, e Aal+ na edel [. h 254. MARherit N 'e CC 4 5 aset C m4.. S. adH d covyesiew yui. cs. oud c_ ome pond.( ng c eEalD i . 19 & ede.s. 'to % sys%. 8b ..e.. v

4 ( p A. i 2 . = - _;...____.a....... i. ____ 5__....._ 471.10' Provide the results of cost / benefit analyses evaluating the effects of i[. reducing the cobalt content of cobalt contributing materials and components (e.g., the reactor vessel internals at core vicinity, the reactor pressure.. vessel cladding, the primary recirculation loop and the feedwater piping). This cost / benefit evaluation should be done for the cobalt content reduced to 0.25, 0.10 and 0.05 wei g. reent. In addition, correct the radiation survey data for the _..pA housing in Table 12.2-19 of r-your FSAR, which indicates 3000 ar/hr before cleaning and 4000 ar/hr l_... after cleaning. ~. - -, - - bNb . [ ~.. __.5.~.I. ~....'Th. hse,d-htat Y kh.wlN . conemia3 b_.. ~ p .t . Ms claatsw 4.colo a.lt tra.q e vbd..cjg esi b n. i 3.. .. IMMegude. Io. sw..ppevi gua.dtiaisve. pre.dicSon of. flie. ........... k e n eOis #.ce_du.ciw3. %e..cobolt c o niegts d cobaIt... . be.a/i$ mabals..Eesh e.nalybl models assume 3 ~~ ~ Ai %el6td J vadO5m leve.ls o n A p p ', n 3 i s ~ 9 .. da e e ctly cov< etat I w$ h co6ff_ to c.oncedrab o f t k r e a d o y-wab. Aece,& eprah3pl<ntekta., . h6 wever, reveal nu.merou.s c edralcion, to h aca.q-a l .h on i n C. b d's q 'lM 5 d wcel W b eye. k e. TY e.w 't M. Yo.dl a i OW wets hes been yword. w L;le % cc,buH - to c oncen - tm%n has. b. n d e c< assw3 o r dakle. u d.

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l if F.vo<.ble, le..d to i sp..iation.S col,<at co t t l an, h s.+s, The. afiev-cleaning dose rate. in Table _ 12..Z -19 will be. covvedeA Fram 4 oco * % c L zooo*^/h as skewa A a.thc.k.d marksp. om

GESSAR II 22A7007 238 NUCLEAR ISLAND R2v. 0 / Table 12.2-19 [ RADIOACTIVE SOURCES IN CONTROL ROD DRIVE SYSTEM-Control Rod Drive Radiation Survey Data Gamma Dose Measured at Contact MR/hr Before Cleaning After Cleaning Component Maximum ' Average Maximum Average Spud 10,000 600 110 Filter 23,000 3,500 20,00 300 1 eco collet Housing 3,000 1,800 1 d,, 0 00 70.0 outer cylinder 1,200 60 80 40 Strqiner 8,000 1,800 1,000 500 Flange 1,000 200 400 150 Control Blade Principal Isotopes Curies (135 GWd/Te 7-Days Cooled) ( Isotope Ci/ Blade ( Cr51 1.4E5 Mn54 9.lE3 Fe55 1.6ES CoS8m 7.7E3 CoSB 8.BE3 Co60 1.lE5 Ni63 5.0E3 Total 4.4E5 0 6 12.2-68 3/3 l

QUESTION 47I. Il In Section 12.3.2.3 of your FSAR, you state that the SFCU circulation pumps are located in an open corridor at the minus 32 foot elevation and that during operation, dose rates in the pump area.are less than 1 mr/hr. However, you further state that during an isolation transient, dose rates in this area temporarily increase to 700 mr/hr and that due to the nature of the event, egress from the area can be accomplished yell before dose rates reach this level. Explain how an individual in this area will know that the dose rate is increasing so that egress can be accomplished in sufficient time. RESf0N3E 47/. // na jkb

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CESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 418 ( 12.3.2.3 Plant Shielding Description (Continued), 'g operation, dose rates in the pump area are less than 1 mR/hr. During an isolation transient, however, dose rates in the area temporarily increase to 700 mR/hr. Due to the nature of the event, egress from the area can be accomplished well before dose rates reach this level. Access to equipment in this area is not required during this occurrence.A" div!As/ i" $12 nill be' 5*Y $' = re ~ /o< /~~*"" U arc =- n *'id!*** ~ *~;r;'y d.sa r.Js ?.v in rc si sous senser, a. mte,i dide tiya.1/raryatG...fauo% a /* r d . n y'm.ve4 '~~ The two redundant SGTS filter units are located in separately shielded cubicles in the Fuel Building. Shielding of the fan from the filter unit provides a lower radiation area for fan maintenance. Operation of the SGTS does not require entry into the SGTS filter area. ( (5) Control Room - The dose rate in the control room is much less than 1 mR/hr during normal reactor operating con-ditions. The outer walls of the Control Building are designed to attenuate radiation from radioactive materials contained within the Shield Building and from possible ~ airborne radiation surrounding the Control Building following a LOCA. The walls provide sufficient shielding to limit the direct-shine exposure of control room per-sonnel following a LOCA to a fraction of the 5-Rem limit ( as is required by 10CFR50, Appendix A, Criterion 19. Shielding for the outdoor air cleanup filters is also provided to allow temporary access to the mechanical equipment area of the Control Building following a LOCA, should it be required. ( (6) Radwaste Building - Shielding for the Radwaste Building is designed to limit radiation levels in the open cor-ridors, control room and HVAC and Electrical Equipment 12.3-38

i 471.12 In Section 12.3.2.3 of your FSAR, you state that the dose rate in the (12.3.2) control room is much less than 1 mr/hr during normal reactor operating conditions. However, you show radiation levels in the control room and in the control building to be 1 to 5 mr/hr in the control building radiation zone map drawings (Figures 12.3-16, 12.3-17, 12-3-18 and 12.3-19). Correct this dircrepancy and revise the zone map drawings, as required. h.seowse tk. u L w.< hoeA1u ve.mm i i z.s -\\ t., n.s-s1 lR& ~ \\%; al \\t.% -\\% s% A A. LL Lan $w \\ m\\d'e.sm hs N tut-to cA' LLMu .4 4L-u-am \\as,

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( 471.13 In Section 12.2.2.1 of your FSAR, you state that your basis for release (12.2.2) is, among others, 24 drywells purges per year 365 hours between each purge. Explain why this basis for estimating the average I-131 release i was chosen recognizing that you state in Section 9.4.5.2.2 of your FSAR that the drywell purge system functions only during plant shutdown. b G.5 O C M5 9-e p ukwe h [ u Ve.~ w du*e-( M (A. m f C M Ak 9.- Y @L P c k e.a. n to Yw was. a s - u v., s - % m.s,-a.a r e.l e s. via. + 6 pa A In vic.o of A '/ -F-ac.t t-ku+ rei Meca. -ths oltained ar _ smatl c amp r.J b As Fre-ohr v e leo.t e_ pa % ao Fu.A%< re% men't of % assump-hs. was. 4&mpte.d. i 9 t

i l l 471.14 In Section 12.3.2.3 of your FSAR, you state that access to the fuel transfer (12.3.2) tube is through a hatch shielded by a stepped composite concrete and lead shield plug. It is our position that all accessible portions of the plant near the spent fuel transfer tube and/or canal must be shielded during fuel transfer. Refer to our position in Section C.2.a of Regulatory Guide 8.8 which states that extraordinary design features are warranted for very high radiation ams. Using removable shielding for this purpose is acceptable. In this r: gard, the removable shielding shall be such that the resultant contact radiation levels shall be no greater than 100 rads per hour. All accessible portions of the spent fuel transfer tube shall be clearly marked with a sign stating that potentially lethal radf ation fields are possible during fuel transfer. If removable shielding is used for the fuel transfer tubes, it must also be expiteitly marked as described above. It is our position that if permanent shielding is not used, local radiation monitors capable of providing audible and visible alams must be installed to alert personnel when the temporary fudl transfer tube shielding is removed during fuel transfer operations. Accordingly, provide the following additional information: State whether an interlock is provided to prevent spent fuel passage a. when the shield plugs at the 11 foot and 26 foot elevations are open. M S rs owsf Se RW 3ds e. do a (2.12.3 (Tn se d Ib 6) b. State whether unique caution signs (i.e., (1) high radiation area; and (2) potentially lethal radiation fields are possible during fuel transfer) will be provided. @.c s p ow 14. M %d I d%h ILS. 2 d (Tw s4 d TMs Aa.wdC.)

c. Indicate the thickness of the spent fuel transfer tube shielding on Figure 19.3.12.3-6 of your FSAR at the 26.5 foot elevation. h 5 p ow R \\ . C d% C Mh G S O dg mehg g 6 oF4Lsy M -4 dg A 5 ohm AA 5rpw Os ~ %4 J i=g e t s, s. t 2.3 - 6. Provide a description of your proposed shielding and access controls for access to the fuel transfer tube valve room in the annulus area. d. 19.3.12.3-6) (annulus access at elevation 11'-0", Figure 9.Q.3 p aw J-4L_ S -e-o. as sa) s,\\, M 5,_ \\q.3,g_,3 aJ ny

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GESSAR II 22A7007 238 NUCLEAR ISIAND rov. O l ( 12.3.2.3 Plant Shielding Description (Continued) (' domineralizer cubicle, which is infrequently required, is via a stepped shield plug at the top of the cubicle. The bulk of the piping and valves for the filter deminer-alizers is located in an adjacent shielded valve gallery. Backflushing and resin application of the filter domin-eralizers are controlled from an open corroder, where dose rates are less than 1 mR/hr. The RWCS backwash ? receiving tank is also separately shielded from the other components of the RWCS, including the tank dis-charge pump, which allows maintenance of the pump without direct exposure to the spent resins contained in the backwash tank. A shielded labyrinth entry to the back-wash tank cubicle reduces the dose rates at the entry to less than 1 mR/hr. I The fuel storage pool is shielded to maintain accessible areas around the pool at less than 1 mR/hr. Transfer of fuel from the Reactor Building to the Fuel Building is 1 through an inclined fusi-transfer tube. A reinforced concrete structure surrounding the tube reduces radiation

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' levels to less than 5 mR/hr during fuel transfer. Access to the fuel transfer tube is through a hatch shielded by ( __ a stepped, composite concrete and lead shield plug. f T e main steamline, RWCS piping and RHR piping are routed through the shielded steam tunnel. The steam tunnel gugge,y t-np4 walls are designed to reduce the dose rates to less than ($ 1 mR/hr in areas along the sides of the steam tunnel and less than 5 mR/hr both above and below the steam tunnel. Shielding of the Traversing Incore Probe (TIP) modules is provided by wing walls and a stepped, shielded floor at the (+)l1 f t 0 in elevation above. During normal 12.3-34 1

CESSAR II 22A7007 238 NUCLEAR ISLAND R;v. 0 ( 12.3.2.3 Plant Shielding Description (Continued) ( operation, dose rates in the pump area are less than 1 mR/hr.. During an isolation transient, however, dose rates in the area temporarily increase to 700 mR/hr. Due to the nature of the event, egress from the area can be accomplished well before dose rates reach this level. Access to equipment in this area is not required during this occurrence. The two redundant SGTS filter units are located in separately shielded cubicles in the Fuel Building. Shielding of the fan from the filter unit provides a lower radiation area fer fan maintenance. Operation of the SGTS does not require entry into the SGTS filter area. s.,c m ' C ( (5) Control Room - The dose rate in the control room is much less than 1 mR/hr during normal reactor operating con-ditions. The outer walls of the Control Building are designed to attenuate radiation from radioactive materials contained within the Shi' eld Building and from possible airborne radiation surrounding the Control Building following a LOCA. The walls provide sufficient shielding to limit the direct-shine exposure of control room per-sonnel following a LOCA to a fraction of the 5-Rem limit (, as is required by 10CFR50, Appendix A, Criterion 19. Shielding for the outdoor air cleanup filters is also provided to allow temporary access to the mechanical equipment area of the Control Building following a LOCA, should it be required. (6) Radwaste Building - Shielding for the Radwaste Building is designed to limit radiation levels in the open cor-ridors, control room and I!VAC and Electrical Equipment 12.3-38

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l l I 471.15 Describe the shielding for protection of personnel on the platform at elevation 47'-2" in the upper drywell area from radiation exposure which could occur during passage of the spent fuel over the reactor vessel flange to the fuel pool gate. R= rp owsg S4 <- sas4.) Alo sa_ cn ow 12,3.2.a rit.14 In Table IAA-2 of your FSAR, you indicate a source term of zero percent noble gases, 50 percent halogens, and 1 percent all remaining. This mix corresponds to a source representative of depressurized reactor water. State whether a pressurized water source was used for the shielding design of the post-accident sampling station and for estimating personnel exposures for this activity. In l - this regard, we state in NUREG-0737 that a source mix representative of pressur-t ized water is 100 percent noble gases, 50 percent halogens and 1 percent all remaining. It is our position that this pressurized water source should be used as the basis for establishing the shielding design of the post-accident sampling station and for estimating personnel exposures during the taking, transporting and analyzing of reactor water samples. i ayd peysevmeI PJf"'

Response

The following release terms were used as a basis for shielding $alculations. 1 Core Inventory Reactor Coolant Noble Gases 100% Halogens 50% All Others 1% Containment Atmosphere Noble Gases 100% Halogens 25% l

l I 471.17 In paragraph (4) of Item II.B.2 of NUREG-0737, we state that you should submit post-accident dose rate maps for potentially occupied areas and indicate the projected doses to individuals who must be in vital areas for certain necessary occupancy times. Accordingly, provide post' accident radiation zone maps and the estimated doses received by indiv-assigned to perform the following functions: I a. Operate three manual valves in the auxiliary and fuel building (IAA.2.C). R.EL 5 pohf4. A ck. d Ww-SecmonJ l '2. 3. 5 \\23.6 NW s 4 b. Obtain reactor coolant and contairment gas samples in less than 1 hour, i 84 5 p ehi( c estimated exposures to personnel as a result of sampler operation e:_ + --- ; _ _.. y m e M --i _n+.L..., one hour after the accident are as follows: Source 3 ft from sampler mR Liquid Sample 17.5 Mt* Gas Sample 60 Mit*A m .msure due to background radiation is unique and will be supplied by the , cant. .usuming a sample collecting time of 10 minutes. J

Perform radiocheefcal/ chemical analyses of samples in less than 2 hours. c. In addition, specify the location of. the post-acciJent sampling and sample analysis areas. besponse$ The / c ca 'i e ws of thv ecsi-acc.iclevd sa-plm3 and tb %p e A. o_p g i da l e u avsTL u p l w, i l o e. a v + a. aw spcified b y l t-w ill pce;Je < st mah J d uet recuived y p<t, s c ww + l e de rm i ng a w u l s.es # th e b y s ~ p es. ,,..,,,..n,- ,---,-,,_..,--yw--

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GESSAR II 22A7007 238 NUCLEAR ISLAND R3v. O IS SI g I (' 12.3.f References 1. N. M. Schaeffer, " Reactor Shielding for Nuclear Engineers", TID-25951, U. S. Atomic Energy Commission (1973). 2. J. H. Hubbell, " Photon Cross Sections, Attenuation Coefficients, and Energy Absorption Coefficients from 10 kev to 100 GeV", NSRDS-NBS29, U. S. Department of Commerce, August 1969. 3. " Radiological Health Handbook", U.S. Department of Health, Education, and Welfare, Revised edition, January 1970. 4. " Reactor Handbook"" Volume III, Part B, E. P. Blizzard, U. S. Atomic Energy Commission (1962). 5. Lederer, Hollander and Perlman, " Table of Isotopes", Sixth Edition, (1968). 6. M. A. Capo, " Polynomial Approximation of Gamma Ray Buildup Factors for a Point Isotropic Source", APEX-510, November 1958. k 7. " Reactor Physics Constants", Second Edition, ANL-5800, y,' U.S. Atomic Energy Commission, July 1963. 8. ENDF/B-III and ENDF/B-IV Cross Section Libraries, Brookhaven National Laboratory. 9. PDS-31 Cross Section Library, Oak Ridge National Laboratory. 10. DLC-7, ENDF/B Photo Interaction Library. e 12.3-49/12.3-50 l

enA l i 471.18 Provide your response to Item II.F.1.3 of NUREG-0737. (In containment G3 high range radiation monitors). (GE will provide in September 1982) 4 A esponse. This ' n f e v d ri e n ~ ti b<- tre v;d'I b r %* h ap p emt t }* 9 D G 1 /

r (. ..'471.19[..In Section IAA.2 of your FSAR, you state that it is not necessary for - operating personnel to have acess to any place other than the control { foca and three manual valves in the auxiliary and fuel buildings to operate the equipment of interest during the 100 day period. You also state in Section 1AA.3.3 of your FSAR that necessary shutdow nd post- - accident operations are performed from the control room, ex)antift for the. several manual valves cited above. Revise this section of your FSAR to indicate the required personnel access to the post accident sampling station and the sample analysis area, as stated in NUREG-0737. b 45. Po nSe .1 .5. c.h 14 A.2 J 1 A A.3,3 k.akm. vees. J .. fo m du J, & __g, od - a c ci d mt. sa.*r i a.s 5b+) cM l x .ud %e sa.Q.an(ys;sar<.s ,+ d\\ caf d. m as a th e. bed avk-vps. M =. ( e t 9

GESSAR II 22A7007 238 WUCLEAR ISLAND REV. 7 ( 1AA.2 8120tARY OF SEIELDING DESIGN REVIEW (Continued) rooms and pumps and valves per Table 1AA-1. All vital equipment will be environmentally qualified. It is also shown that this exposure envelope is not time dependent after about 100 days. c) It is not necessary for operating personnel to have an the contqol Room e-acc[sstoanyp1isce er n% fs sna ysu ae n the, *Auxilia,ry and Fuel B tie s accider sa ngs ree manual va ve to operate the equipment of interest during the 100 day period. The manual valves are for essential service water supply (one in each division) to the hydrogen mixing blowers of the Combustible Gas Control system and a Drywell Bleed-off Vent System valve. These valves are considered eccessible on a controlled exposure basis. Direct shine from the containment is less than one R/hr within four hours post-accident. d) The control room is designed to be accessible post-accident. e) Access to radwaste is not required, but the Radwaste Building is accessible since primary containment sump discharges are isolated and secondary containment sump pump power is ched at the onset of the accident.

Thus, fission products are not transported to radwaste.

The hydrogen control system is operated from the Control Room; the 238 Nuclear Island does not have a containment isolation reset control area or a manual ECCS alignment area. These functions are provided in the Control Room. {.I n b m d ion v lude.d do ih pod-accidsd sa%7 wy shit w %J Ii N <=%,l,.,.t.o. m w ill 1e p.,o AJ ky A ff w La f y ts ekM s, -% <. d&l d cit s wssia.1 whis f.II,ws. es n+ g

GESSAR II 22A7007 238 NUCLEAR ISLAND REV. 4 1AA.3.3 Post Accident Operation (Continued) ) r For purposes of this review the plant is assumed to remain in the safe shutdown condition. 1 The basis for this position is that the foundation of plant safety is the provision of sufficient redundancy of systems and logic to assure that the plant is shutdown and that adequate core cooling is maintained. Necessary shutdown and post-accident operations are nerformed from the Co s%, W p st==ide.fsmp3 sts46%, A saw Room, except for the/paveral manual valves note [!= - earl er. t ? ) t t 9 O O O 4 lAA-10 30 105C9 1 c-

8 471.20 in Section C ef Regulatory Guide 8.19, we state that you should provide assessments of the annual occupational doses in man-rems, principally during the design stage. We further stata that as a result,of the dose assessment process, we expect that various design changes and i fr.novations to reduce radiation doses will be incorporated in your design. We designate certain detign features in Section C.2.e of ( Regulatory Guide 8.8 which shx1d be considered in the crud control effort. Accordingly, state whether the following design features were considered in your proposed design and indicate what actions you took: Nigh temperature filters (i.e., magnetic filters) for crud removal a. from the primary coolant during reactor operation. Rc39oh3e l 'ig _ie radure. i rs. an. n oi owplem ealed en 9 5+ h A lWJunhs Id 1 k 4 I$M. f. S n a m s - w r a + ;,, m y u s u..r.4 m _%,m3 F, ed-ak gualdy rA pla ds, w t p +r J-F,ord Labr draits, h. u p se.,+Iy no d -.n sw h d L - nt w.K mir4d 4. cowbl d redini,e wildup. l k 1 l l r ..______._-__m-_, _ _. _ - _.. _ _ _...

f 6. Stainless steel piping and heat exchanger tubing downstream of the condensate cleanup system. .h3p6Ws-C, Ahtd<ss x4 eel n e_sgeeDed Few ha.<3.ex elsang e ua (bc v.s.._wk e., s..l D h _ n e e es.:e 7 't o. M M % 1. e. [a rkes_Meef _ (w6 <A ku O.v.r e shw . ve --....} b w. _c eth slYa sbY S $w y: pi a$ uk ts us f wt& M St-h<t\\tnce. d.. ..s y s$+m, c. Reduction of corrosion by sinimizing the internal surfaces of the primary system. .--22-s00 & ..L.. yen I,tk p ; m a v7 s y s.te m.c m f o v< d e4 alt d<d.J Ly a p,r m..dr__o fb c. A. s L d; g . rs N O em. g M N g.g y M e b Nh - .- 3 *MI M 4 ce. A F 4A.f..h C d wsk

h. S u ra t &).

c oot . __._o..___ l l

i I d. Reduction of personnel exposure during in-service inspection by reducing the amount of weld footage; e.g., using forged sections as opposed to forged-welded plant sections of pressure systein components. ksQ6%.JA i ml..,pe. L.,L.. r. , u s e d v i... n.ci; c 1 in. .<d.< io limiute soE..f N. lengtad.aal w eIJs ~ ki L v. p ir, c u.<.a u supction.. A %.<.Ly veJoce ewp.,u<es # n<uoa p<F.,-i.3 su.<,ic. n <t<... o

e. Reduction of the fron and cobalt content in the reactor coolant water by increasing the efficiency of the reactor water purification systems and by increasing the cleanup flow rate. .[ Ra-5 9 ew s-4. A pru.%t <<.d.< w.b c l ..p,7,t.m is ad.y o.f

t. m.d ensiin3 wd.< 9o.lity re-puurem.nfs. Vk, k I;~sta.hans.T y as.&

un cedainfie, reg a vdi n3 %.,,e L..t - # cruJ be.tJug,thu syd., e s co n s;J,r.J 1. b nsar o ph ou., a n c. p.d to bal acin3 -J..,7.FC. J.N ..ss 4p. d 1.s, b l.d t h,- I .Ff t e t.-cy .mJ syd.m co st. I t kas ,4.t h..n J. -.nstrai.J ht, der. as.J c I< a w op s y sie-u ra <ify ~;ll be e % c.t e in reducia3 c.< u d - r e laisd radi cdiow 1+v e Is, l l l l

l f. Provisions for injecting oxygen into the feedwater line. h 5 p 4 W S-4 Provisiews Fo< iyee.%wof.<y3ew into Fe dw4 % h v wot b..a impl.m.nhd m h destn. It h.s nef leen Jemenskt d Ad 3 implem.ntdion 4 hs f duv._ w;tl l., efkt-t re d ucin3 crud. low'ilclup. Two v e t ni Ive in +2ds <f etys.n in). diew a1. e p va %3 ple nts p r u tJ s d. n o eu ; <lence 8a be n. 41.1.fT.d ra di dien lev ls. T b incenieve,W ony, on Fev osy3.n taj.ct..n is I w.y ;n a pket s'.% Fe rm.<J-pu-c.A beder draios as a. < e s u lt of A F. d ha+ A F..hb et y S o n < en%1 hi k, 3 I

ATTACHMENT NO. 3 r l DRAFT RESPONSES T0 l EFFLUENT TREATMENT SYSTEMS BRANCH QUESTIONS I l

( 4&c.c)7 150.39 ?i nvtie s isole in Sectian t.3 of jour Fin :: esetn1. e :ssi p (1.3) features at tae 1 t ;ui $, ;35eous sid_ toli t at teiste sf ite s.1:n i m l i.i.2)

osition of Requistory Gut !e 1.143, Reviilan 1 Mct:oer 11.*11. Jus:iff (11.3) each :ost tian for 4hich in exception is t.iten.

If f -f1r-stian i 5 =c m :e 2 (11.4) in 3t'ter sectlans of the FSAR for the lajivtaust lia s, cross-rerereses to trese c.ections is acceptable. 'de cocit ser c:Spl1ance.et th Sectf on C.5 of Revistory Guide 1.143 to be esseitt:1. Verify.aetrer fou :stisfy our acceptance criteria for concentratto1s of radiotctice const:ti.ents in accordarce eith item It of : action 15.7.3 of tre Standard Review Pisa (SRP). Our position is that limiting doses to 0.5 recs, as stated in Section 11.3.2.20 of your FSAR, is nct an acceptaole altercative. 'R e spoM 3El 460.09 Reoju,la,bg Gtdde 1.145 Retnisoni(ect i<t7e) furnishes Seis mic cke.r: Sn aggie (awe r ca.cceptakle To The STAR. G E ha.s i m ple ment eM'# d.Il T Ac periDons N RC of This qaric(e wit,k the ey ception of 0.2.13. ( Lt.3 Thoes portions of the saaecus radusste tnatment systens that are intended to store ce delay the release of sassous radioactne waste, including portions of structures j housing these systems, shculd be demaned to the seismic desism triteria saves in regulatory position 5 of this guide. Foi the systems that normally operate at peessures above i 1.5 atmospheres (absolute), theos criteria should apply to isolation vahoe, equipment, interconnecting pipias, and components located between the upstream and downstream valves uand to isolate these cc.mponents froen the rest of the system (e.g., waste ans storage tanks in the FWR) and to the building houang this equipment. For systems that operate near ambient pressure and retain anses on charcoal adsorbers, theos cnteris should apply to the tank succort element (e.g., charcoal delay tanks in a BWR) and the buildaag_ housing the tanks.

I k %;gr). y d.o a.y y ed ; 5.Jns aa. e c y, J l l te-s4,w,v,.o Aw w u>cct w e4./; A) no ukea.19dequenciee betwee4 2 and3r#eds. r 31 b) a stvess in tAe suppas.tr, hased as n demoofa/ Static efu/se/en?' foo ce egens/tb the oBE /ess E4eh /.33 EtAres the a//owable.rdws leve/ af dist manesu/efrtee/ Consfrecho'e" 75' edtion R?c aid c) the tanks are / ace.k~f v4 6de Jssenraf af the AvX/t9 /roo.w. theer. 'a,L w Y - - k a v f -l c A b a e a u ( A s a a a 'i a' N I. %.rg A nebe -A p -e.4 # M,rrops4 l u> U Ao % s & & lc ~ 9 1. w a 4 4.7 ys %.

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u a enu, ~ ca "tu ey ~a ss, w) -f to 4p-s. m. I. aaJehm Eu n o m e=-ze se~1.r W W (e.hAL m % pau ob) M tLo -TsD.~s { tLo as m 6 u.M rew ca.et.24w s p =$ y A ( n 0 l o % s t h r y m L A &u '1 ns e_ s 4 zS, oo op.kft2<. Cso loopts M eq wm. T r&, u 9 '!!d a/sne. t x ETSB Us.S~ M 74 f

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0 l DO tnitL Ek O_ny e_m i cufi% <r{ S. Th Ir. 7.3 of th SRP G ogpis& m-L 0:1.4 a w .1. m &a, m a.s. kw h

~ -- W C F CRAUN O CO or..s: Jg3 MOTF.5 ~ ~ l cusTante pacts e racc i .som s302*f ,,AE4.RAtus ^( oatt av ettu @E3TloM 460 10 Add sections for effluent radiation r.onitors and engineered safety feature (ESF) filters in Table 3.21 of your FSAR. Also add to this table, under appropriate sections, the recociners in the of f-gas system 'and the process radiction c;onitors themselves. f, 165f0/155 460./0 i.r ised afr/in Gr.y X p nen A d<fi.e EWlaeset n/ta% owauif&iky A.%r spuritse, of 7.He 3. 2-/. Esf f?/fer.r are. Irr G ~p XXXI, Sta. fly Ga.s Treehre f Jpfess,ef 7'l/e 3.z-/. TAe ree. liner.r sa a Me. eff~ys.ryu1ks are saafadeaf

M M. pressare re.sse./s in Smyr i

x* off <> sy fw,.F Taki 3.2-/. 7Ae.,onssis r<h L m.-ifer.r y dem,ss/re,r.re reefade/wi0 7Ye sAefris./ needule.s in 6 reap ky / mess 44. frau /% iAr Jysfem ef Taife JC2-/. e f ,) e </ O f C "i - ( v. ~.,, g e 's..e e

. ( ? (, '( Table 3.2-1 l EQUIPMMNT CLASSIFIC5 TION (Continued) l Quality i l Group Quality Safet Classi~d Principa't component, Class ( Incationc a f fication Requirement Category Comments I &"au y*e *<'ou*" Ny'$s'ffNY,eiferiej) 1. Electrical modules e. win sank 2 Ac X N/A B I j 3 _seeeml4ne-end=reme-u

t-M

.yentilation:.monitqse fanc/,*.es O.,.(, le s m ou 6ok.e) 2. Cable miswiW:dteamlifriAntpi 2 A,C ,X N/A B I y es mentainment=vepEITW% 3*7E/ isontrars /Lne rien g cO gn i ,68 ' N XI RHR System l 1. Heat exchangers - primary side 2 A B B I MM tn M h 2. Heat exchangers - secondary 3 A C B I side a 3. Piping within outermost 1,2 C A/B B I (g) isolation valves 1 4. Piping beyond outermost 2 A B B I (g). isolation valves t 5. Pumps 2 A B B I 6. Pump motors 2 A N/A B I l i u 7. valves - isolation, LPCI line 1 D,A A/B ~ B I (g) <4 .o B I (g) 8. Valves - isolation, other 2 D,A B 1 i l l l

c.. c:

3 __.,;_

460.11 Provide additional information on the following items for the ESF (6.5.1) filters of the standby gas treatment system (SGTS) and the control building: State whether instrumentation for measuring flow rates through a. the ESF filter systems will be provided in accordance with Regulatory l Guide 1.52, Revision 2 (March 1978). 9.a_S p> s w 3.e._ ?;4e isa.r,%< Ah.o. Ar ~ wy filrtss ~A4r.*.yA Me. E.SF ....... fiD ' ')Y***

S"' N' #M f* 'A
l A7 5 *M"A"Y

_ _ _..n.s parf *F :Ke 6aseea scff/~enf A%.;faris, sy.s71ru,. (GEMS). .... N* '"'S'"*W'Yi** f*' m#1y f4's' rstes tar.ujA Me. d'sF Y . _. _ ft/ter sp ame Ev W c 7%/ &i/droy. eL. t d,. A ir. c n,, 9 ._ @B Mc) System 1.s p>.. vide / a,,ee.e4,. e wi?% sfey/sto>y. side _ /.52fevlsie. 2 (Mar l. J17s) eneyt Av d e rea,,,.J g p f s a ~ discused is, .We. &# as., pth.2). _..._m b. Indicate the type of recording device which will be provided for recording pertinent pressure drops and flow rates in the control rooms. R L s y o.w s e, -__ ,.....~ tkie pr.ridad for v.e.edioy. 14s f/a, ^ a. .:)7Ae type..f rdiny. re,Ji.AA Ms.ESF ft/tsy,rysfes.s Ar Me m t - -. _ J ndas 7%9 Ges>S. de appAa f :.s p.vt of Me 47 p A6 ree.ediy devise M,pr vide / Ar ns.ediy de per7%e f rusrs deviss, is mf oreedd i,. fAa conf,J r.ehaf.y $t 3G75. A yss.rNy afrys afandl .it will. t. dis.ly.rArt if ei%r.f He poe,.aty aia,<< Me y f s A /r.,.

e., pa,fi...t,,,s.ve Jr.p., ~.Ae.ri.s x i y.~: )Ar. rua.rJiy la vte. N pr.,ide/& v.A.y da park, ?pressne d<.y.s ea esc.rp e-Ar-,diy f

.. J f t ~ w e s i K < e.,,-/, s r n d e /<vte* 1.r o..f needel stuee. N ahadly unif wi# a f.,.a'i If4.s fa<tif -M< ye,. fig -~;f>.s f 4 e ra t< raa Ja.s if.1 f. war %.,i/i.y. dif,,a. l t l

~. _ . = _ _ _. -.. - _. Since the explanations given in Table 6.5-1 of your FSAR indicating c. how you satisfy positions C.2.j and C.4.b of the regulatory guide cited in Item (a) above are unclear, explain how replacements of either all or part of the filter train will be accomplished when this is required. Also explair how the filter train components will be maintained by servicr.trsonnel located outside the housing. Indicate whether the ESF atm. phere cleanup system will be totally enclosed. fb s p ows e-J _ Z,. N n,e /L>. A m-o

e s.

r ~ s(nr a

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~ e-2-j Both the standby gas treatment unita'and the emergency air makeup clean-ing units are not removable as intagunit. Eigh activity accumulating alements can decay safely in place prior to removal as safe radvaste. ~~-"" Removal of the charcoal will be done pneumatically *into standard solid radwaste costainers with minimum exposure to operating personnel. 4 ., 'F, -r___.. r -- l l b 4h coher% d ~ Me) n h.' 4 L AL WA wK u ~ ~ ~ s /A

J d 4._

m = 'OP nE G"i" 4~.ch N ,m. ws e" IC/'s.S to a e s r pq l s.e r" g,'.o,u, l l \\ .,s., l i ff *s

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s pao 2 t om efu ., J p.,4 C F BRAUN O CO PAeas 51 w,=ue 4t.d:c.A' do,<y-l 4,% A2,,e.ch rme 4 f' /r ,L. of 'e.m GM w.S e,e '. s. y% m = ,'A n a. JLC-s.& / A,.. = u- ~ se.,e LC-hs n._ aj h s. &774.>a,r! = ,~ N, Ah H = Ap. 2._a s.4s ./L s' - .a_3.a d Y i ~ A s nx 2 v 2 m J JA.4 A y'3).- L 1 A.c s-- = NA'd al.a r &d Ma*>*so Asw.4 )b1 A. J sd vdr ,41 >. A = K-- > a. m - ,Y a J% 4-rr 2,4$a r. / e,'/n a e<m. SC 4.f-. J' Z,.de 4 fA. <, na r 0 ='= C'L - e ,4k a JC. 1J-Afd. _ n/ elJ. E- 's x,s k4-,._ A />% w h J-e 4/ '>n2 es: J s' ~ A 4.a r '= J A. de.L 'L. A u.s - a n. ..a. ~, K s.n ,.,s.! 4, a .a J J l l ~ J M .A,4 m -r:6-e X4.- d e, '= ~ -~ a. +% ?*/.4..,.. %.de s - u;~ .mc- -a7'::t, I 4 M ~ /_4 'i &l = ( 4%, + ? 2--- .,. A / ., -- 4.,e. ,w =' ~ f -,,. A Ac. .; AL L 4 ~ f ( = as n AWm A m./i rL/anam n. // Jol 'L & / --L'Vd. A.,1-A. Aw S2,*,m.

v. m IF i

l ,[ ] ~l l l f[ l l l ~ d. State whether duct and housing leak tests will be performed in accordance with the provisions of Section 6 of ANSI N 510-1975 and in accordance with position C.2.1 of the regulatory guide cited in item (a) above. b-i i b m!ed aA E ' I I i i 1 ,:: J' 7,5,,,, _ fsk _= ? -n %And~, r s,. k.sypb l h.s. Yd Y l s. gu , y,,,, H d I',>,', I .pd.d./G.14A34 hie.!d M M> M ~bL\\ n i w OA\\o ,s.>6 ~-

c. e w

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m- = 1 N h,,$ .dlg$ ,, =. _ > p m 'fa88 =as,se j N NO c e,*L/$=.s u) s'J/ '2 s w -=> 5.A-e m.) kN1 h 8 D-W% G '~ ~~~ _ n t e. a.> ~ 4 v% ji A Ade n c.0Le 0.* aLV n 6 41 M_l 0 A i _u f.e xh_ it s.w r_ s Cd M. V-4t 9/e M 4 Mfa/tet, man /O ~~ Ce e a ..rc.,y t) i lre re-, 4, /,1r 4 f/e u,o edt-s.I 1b e l.._ o With regard to the position C.3.b of this regulatory guide, state e. whether the manual overtemperature cutoff switches for the air heaters will be accessible following a postulated loss-of-coolant accident (LOCA). Note that the temperature set point should not exceed 225 'F per ANSI N 510-1975. 1 - - - -.. _ - -.. _. :. ... k.S > nW 5- *-.... _ .,. wift r<y to r%< puift.~ c.s.k.f 4 nac saykfay ea.We Asz,, l Anaist z (Av.4 /r7q,),, au/ AN.5 / At SOT -/T74, a.dowsfra.wricyermkre .J fr a rteJa aus M.rJJ n. Ma air A r'er.s i /ia .f M n.a / I .nrfe, pre t.are e.dwff aws/ des. 7xe.,a utw.fic..vertc.pner.e ra aderf a 1// a.f le <asess.t/< f nty ps A/Ja/ x.oca. Thi.r is se;Mes M.y >,./.s.u.Jiu /y..d af, n.f ry.a,< e.,, ./ Al/s.atu e.

a. y asti.e..f. ydkl. 7As b y er t r< an.tpeter'.sh AV a f seceedzas *F ra-y l

nnst N ro7-/s76. TAe applie.at a,ill de rep,rd f. c.,y/y. l i \\

,_ 460.12 Provide information on sou'rce terms for tha following items: (11.1) a. Provide the appropriate data for the items listed in Chapter 4 of NUREG-0016, Revision 1 (January 1979). For those items for which information has already been provided elsewhere, cross-references to the applicable sections are acceptable. ~ . h A S p ch.S-4 4 A.\\ pom = T 13 0 p w Y., t) F\\c%<\\ vvsaw 4.3 g o olwdttsn c. 50 Ccf p ........ _ _.__. 5 0. C G / p... .b.. -_ t) .h y t A 'L. = s s.s ao" iss./uc 4 r a.1. sb.w. pM(,_. = a. d. _.. R P V Iw

s. n, n e w s.

._ __.. _.2).7 ~ 4,(. _ ${...s.j._No\\d4p'Ti m frow h.c~m 5 ~_ A h aW e ~s... _. sse<~ =. o.o.0.tt cr 3 _.._ _.A). S-e.-e_..Gs 3Aa._E. S e ekt ow JL1

o r-

.. c k S 3. e n p 1 u w. o-F. o W .c t r_. s r i_...p.a.<Pe % w c.a_.._._....__..) _.._._/. U _...- Wo. 9 z.-A.e s

hv.. _.4o _.o. o i. Ob

. L_. _. - Ita_A.u..Xe...&_k.G o... t 2. ci /,s-e e is . s, t x i o.~ 5 C L /3 o. c... .2 J..a4 Waw. hvbm..C.oA c 3 et/y. h % w o f._ c L.= 7 A, 5. b =.

3) Ma s; ofe 6 ccod 1

4 \\.=-3 *p, o4 g =- 6 5

F v c.

. ab ow cos~1c,s.v = 'a 0 3 7 cc. / K h V.4 a--= 9 3 cc / 0 vn bur o6 u MW. cv3 o 3ew ds " - l L.s n. .5--e+ -. 3 s(c).M.o ws.n.,l 46c-b~b N a b.t c 2 _... e p &.t-bk a.c

s. l m__u...

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g I 1 l f b. Release data for tritium from operating BWR's does not support your conclusions regarding release via: (1) the gaseous pathway as compared to the liquid pathway; or (2) the total release. In ~ ~ fact, for a number of operating BWR's, tritium releases are sign-iiticantly higher than your estimate. Accordingly, verify your estimates for tritum release via the gaseous and liquid pathways '~~ ~ ~ ~ using actual release data. Verify and correct the N-16 concentration given in Table 11.1-4 c. of your FSAR. Additionally, verify and correct, as appropriate, the reactor water concentrations for Na-24, P-32, Cr-51, Mn-54 and Zr.-65 since these are significantly lower than the corresponding concentrations given in NUREG-0016, Revision 1. d. Add Fe-55 to Table 11.1-5 of your FSAR. ... _... -....-.....SA.s 9 0^ V-s .u.- 6 Aq r.ghtwA-.A*

n tf_1 Th4T.__BAK.

EEEv_ RE20:t31a...yAGV_. . _S o.._.44. 5-A_TLV M'L..W_slWAurC_hC_$_h4mycLg. _.. ' kultalui.THE. A 4K.h@Wp.fCU' _ffw_+G I Tn.1T ...LM..u o76R.... n Ati.L f +3f 0._ _.C..h:J O.' $ $.l_Nlh. ' 'ML T.M.. $U... c A $. (, di &.' !4.Yp \\ ..___..L....._.. . _ G..,_MA.. WlLt.._.GJf445.. Tim.5.__.41. LA..To.s.now .n/- ((,. A-f.44. Tost. .... _.. _W.M.(C.L. _.C.0 Gued.TLST\\ oM._.or _ (a 0 MQ '.A._C ,-_1Co A _ Ls o wse. 6-1 M M & /[ toa cu._wuu._s.w cu w _i.LL.5__rp..._..__. L. -.._ _._ _ 5.H h t..*T31 K. T o u p u d1 8 ?.1 _ l.L G M f_ d . i... _ _......G9 4KNrg ATnog N i.. _. _M M

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t 4 _.___.P..__~$~1_._. 2.0..x.l 0.. C 4 ".g. .,0 x \\Q'S CC-TI .... lA # lh.O ~ ~ L _ _... _..._ _ .*. (de.w ). Fe -65' '9.Dx 104 h-GS 2.o no -^ Tp s s.s.{d.An ovt t

460./3 ct ' 550.13 Pravite attitt.;nsi information an the fall:wl.,) Ite.,s i:2i tetst e :s : e (11.2) 11;a14 ste.nsnaje, tat systas: a. Provide the Ifquid waste In;uts in gs11cns per dty (G70), avers;:J on a fearly basis, of waste generation for low conductivity.tnd ht;h coad.ctivity wastes to be used for evalusting Ilquid effluent relesses and relatad off-site doses. In addition to the waste strea.ns you have identifled as destJn basis inputs in Table 11.2-4, you shau11 also f aclude the resin rinse and cleanup phase separatar decant insuts. State the primary coolant activity frictions for each of the Individus1 streus for these tao waste subsystems. f6ffdM$e,' ff 0./3 av %'u J Q &csy-Nano l& &?;4 ee.a f a cA= 8 $ a TaMa H. z-9. e e e .1 i I -t ) 7 3 2 I

h b(, lY O b. Ye'ar laputs for creitcal 1 siorstory nsste,11;orst:cf.s u < t :. s ; l liandry dralan are Ice in c:m; sri 5:n vita tne corres:and'. ; si.es ( =.- i /f l gt <!a in NtJELC0tS, Resist an 1, on s :er resc ar 'a sif s. and cJerect. 25 s;propriate, t.9esa inpts. l

Response

Lausd+y c{ rain ihyd is in ap.eemesf udA nexta-ooir. vslues r eparted or ssng.re > ta.Levato+y wash waten and chenaca/ cuarte input ai//de Aines ocu. changed to correspo nd to O M

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n e !..; i..a. :.2 s :.. - 3 I 11.2-21

r 4 60,i3 C Since you htye considered c.1f t?e deep bed re;entrent systtai (se 1 C. condensate cles.wp snj ys*J have also stated that the co.a. dentate cleanup systen is within the sppiteant's scope, indicate wr. ether ussje of the deep bad regenerant systea for condensate clescup is an interface requirec. tat. Adaitic. tally, Indicate ghetner ul traualc resin cleaning is also an laterface requirement. $ESPcNSE: 966./3 e W fo NOU6 c/or a cbOnmAd W r.z.93 (Pa y u 4t). 7h & @a /h eh e a m; A of c6 & crme i.u) u D +ca nupyma cuaei.r.4 t e I 1/

^ 460.I3 cl d. Since the filtered detergent wastes.may be directly disc 5sried 1.1:a the circulating water discharge canal, state the fraction af detergent wastes that you expect to be discharged in a year to the circulating water discharge canal. [93f 0f7S t': #fo0,/3k a, w a s k u.a.g,is ?y ' m a A'-y Jairs & ir w k s sp /.su lige4 & & ,w g. Se 46 25 Y d>'e l'A m -A st. &*1'?b'

La w&& 4 =% eJ,,...2 f);,..,1g,..n -sa

&=?& m. d!^- M YM W f T. s. i q>eby wJ@w & ao a.!=.==My 4)& L wowb m a Zl.a & ) sepakof4D p =+ a ea-wa e a p/. M N3 74 6r* ...a. X&. _e??4 % nesw- : =n n==LJb i Ar ='-/ A a rany = %4A,. 8/ M ..L I I . p. i i' ) .p.. a e .T

4(oo, I3 e

e. Indicata = hat yoe,esn by a ' waste collector subsystea* u.nt:s f:.

refer in SectIon 11.2.2.2 of your FSAR; we do r.ot find f: 3Isc,isa t anywhere. Respnse :

Wa ste ce ilecM' wa,s cha.%ed. To Low Gedwentig m tevised. gaa.ystk I l. 2.1. 2..

( GE.SSA.tfREVistod W L-11.2.2.2 High conductivity Subsystem This subsystem collects and processes dirty radwaste, i.e., water of relatively high conductivity and solids content. Floor drains, and condensate demineralizer regeneration solutions, are typical of wastes found in this subsystem. These wastes are collectc-d, chemically adjusted to a basic pH as required, and concentrated in { a forced-circulation concentrator with a submerged, steam-heated , l element to reduce the volume of water containing contaminants and L to decontaminate the distillate. The distillate is demineralized to remove soluble contaminants carried over from the evaporator Q t cond.u.ct wJef l and then combined with the as.ew este ecd M Ur subsystem just prior to ( the second waste demineralizer, unless high conductivity dictates recycling. Conductivity instrumentation on the effluent of the distillate demineralizer determines the routing of this stream to either the waste demineralizer or the distillate tank for recycle. t.o u> If desired, the distillate need not be combined with the e l b= Iubsystem. This alternate route is through the excess f water tanks (Subsection 11.2.2.3) to condensate storage or dis-charge from the plant. b l

-.v...._... ..y_~

.. - _ _ L - --

^~~ ~ - ~ ~ - ~ ' - - - - ~- 6 0.13 f f. Since the excess water ts.-k collects excess water fr:132:a : + 1:. 4 and high conductivity subsyste.ns, explain how yoJ can selectivelf prevent discharge of excess water frcm the los conductivity sos /ste, during the tlno when excess water from the high conductivity sasjste.a is discharged to the envf ron ent. If you cannot pre.e.it dischse;e of low conductivity wastes to the enviror.nent at all times. t.*e4 include the appropelate fraction of waste dischsege frca this subsystea to tne e.,vf ror.nent. Respehse : 1 The eseess'.wate* Tanks are sAnwn th riy'ee it..t - a x wird dedicated spaf Jon#CdS. Discht,tgv pigheg 'gs APP &hjed Je ) cAat. excess water from the law cendnatierty sw/sysfeN canne/Je oliselarged to the enrinment(&tayowph lt.1. 2. s ). ) t 4 e 7

hb Y. l$ 8 460.13 3 S'ac* your PH d'53r"$ for t"* "25t' 50575t"5 ir* 'o'

2>*1

'at radosste system, indicate.hether the e:pi;nent that you nsaa lis:ed l on page 11.2-30 of your FSAA is for both untts or ane:ner ie is on a 9er unit basis. i Tes pon se : Iwf.. md i.w ye e.vteA n yage, u.2.-so is 4

a. single u. wit radwaste sq rtem.

ne eyrjeent /irt is exentially the same foo. a. siny e e dna.1 aa7 radansie 1 l sy sim. espacity or. rome epairment would inarense aliQTIy y a. Aant. t. u.s'sT %:sf t19 i ~ ) ~ l. 0 g l ~. 8 r e

== ~ 460.t 3 h h. Descrue tr.e prsvisions fse prese. tl. ; t.7:oite211e1 re:n;e; :/ 1 1 radiositive.uterials due to spills;e in bullsf.,;s or frei sat.::ar tan'as if tne latter is within your s:c;e. If tr.ese pensisio,s ofit be described in your response to Question 450.01, a cross-referecce to the relevant partian of Section 11.Z f s s::4ptole. $(: 060./3 k. d SeSa's o'0 & /2. s. t. i .AMk 94/4. pm & 4 e o 9 ) t .a ^

a P

N

r 1 </td.i3 r. (, 4 l 1. ProvJe,'the corce.itratt :as. sf radt :c 4cildes in t.9e et:2 ss esita 5 :; s; ta34.. ferify and correct, as sp;:r:griate,tre 29:,e af endf a rt:I v i :/, 'n curles, for 1-131 and the total curtes in ene c3ccenested was e tank given in Table 12.2-13 of your FIAT. teydhe: 96o t3 i The GoNcedra.l' irs oA t.adibrrac//i/us to, t4e exces.r wafs sto'yye t.1eb is reporfed a i% led ta2$u niis isL/e.so>// Le adoled to' C4e reutreol rersion a f psna. Type 7 stj. bic correcY/bbs _ fes-b Z-/3/ concenk?%o in L/e donceeba.teof wa$h 7% L asifa/ss he made a xle i.ewred psa r v 1 2 e h i i / r a. v.

Table 12.2-13 (Continued) EXCESS W/ ITER TRNKS A 8/( 4 Source Volume = 100000 gal. _y Tota 1 Cutien = 0. f6 Soluble Fission Insoluble Pission Activation lla l oqon s Products Products Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies lif f-313 3.BE-04 Sit-89 I.5 E-0 2. 'l 11 - 9 5 /.? E-0 V NA-24 3.6 E-0 V flR - 8 4 f.8E-O f SR-90 f.3 E-0 3 ZR-97 6.oE-06 P-32 8.sE-O f" !:14 - 8 5 O. SR-91 3.o E-0 3 NB-95 6.O E-0 ( CR-51 E.6E-03 U 1-111 9.0 E-0 2 SR-92 2.f E-0 3 RU-103 9.o E-O f MN-54 2.fE-04 I-1 32 2.0 E-0 2, Y-90 /.3 E-0 3 RU-106 f.VE-0 5' MN-56 /.6 E-0 3 I-133 d/. I E-0 2, Y-91M f.5E-0 3 RH-103M i.0 E-0 5 CO-58 Z.T E-0 Z NS I-134 7.ol:-0 V MO-99 2.f E-0 V RH-106 f.d/E-O f CO-60 3.L E-0 3 p l,'; I-135 /. 6 0-0 2, TC-99M f.f E-0 3 LA-140 J.6 E- 0 2. FE-59 4.b E-0 4

)-

TC-101 3.g E-0 6 CE-141 /.fE-06" NI-65

6. oE- /O TOTAL 1.7E-O f TE-129M

/.fE-0 3 CE-143 /.fE-05" ZN-65 /.ZE-OS' 7,; ';,' TE-132 /.9 E-0 2 CE-144 /.TE-Of ZN-69M V.? E-0 6 y CS-134 9.f E- 0 V PR-143 /.8/E-0 V AG-110M 3.SE-0V i$ U CS-136 3.8 E-0 V ND-147 N.fE-05" W-187

9. fE-O f CS-137

/.Y E-0 7 CS-138 1.Z E-0 V TOTAL 3.7 E-0 2. TOTAL 3.8 E-0 2. BA-137M /.3 E-0 3 BA-139 /.2 E-0 3 BA-140 3.f E-0 2 HA-141 f.f E-O f BA-142 /.OE-0 6 NP-239 2.2E-0/ TOTAL J.1E-0l. $ [' . d, i

Table 12.2-13 (Continued) CONCENTRATED WASTE TANK A700 Source Volume = 3000 gal. normal, 25,000 gals. full Total curies = 29 Soluble Fission Insoluble Fission Activation llalecens Products Products Products Isotope Curies Isotope Curies Isotope Curies Isotope Curies BR-83 2.5E-04 SR89 9.4E-02 ZR-95 1.2E-03 NA-24 4.0E-04 BR-84 1.9E-06 SR-90 8.9E-03 ZR-97 7.5E-06 P-32 4.4E-04 BR-85 4.7E-10 SR-91 -' 8E-03 NB-95 3.7E-03 CR-51 1.4E-02 w I-131 2.6 1 SR-92 . 6E-04 RU-103 5.2E-04 MN-54 1.5E-03 I-132 6.2E-0 Y-90 8.9E-03 RU-106 8.7E-05 MN-56 9.5E-05 s I-133 1.5 - 0 Y-91M 3.3E-03 RIl-103M 5.2E-04 CO-58 1.7E-01 Ec I-134 1.0E-0 MO-99 7.3E-02 Ril-10 6 8.7E-05 CO-60 .2.0E-02 OE y I-135 7.3E-02 TC-99M 1.7E-03 LA-140 1.8E-01 PE-59 2.6E-03 yy TC-101 2.7E-07 CE-141 1.0E-03 NI-65 5.6E-07 wx TOTAL 2.8 1 TE-129M 7.9E-03 CE-143 3.lE-05 ZN-65 8.4E-05 sr, w^ TE-132 5.9E-02 CE-144 1.2E-03 ZN-69M 5.0E-06 CS-134 6.lE-03 PR-143 6.9E-04 AG-110M 2.3E-03 k CS-136 1.9E-03 ND-141 2.2E-04 W-187 1.6E-03 O CS-137 9.4E-03 CS-138 3.9E-06 TOTAL 1.9E-01 TOTAL 2.lE-01 BA-137M 9.4E-03 BA-139 5.4E-05 BA-140 1.6E-01 BA-141 6.8E-07 BA-142 1.3E-07 TOTAL 1.0 - 0 Q wt 3? 5 O-

4& c.B e d J. Indicate m'et?er yhr estiantei rel ttle s s J 'c :;;: li ; :::t s..e 025: ;;. ;e to 11;uid ef fl.ents are based on dest).15 ssis res *.:- : terns provided in Tables 11.1-2 sn3 tt.t-3 sf fcur FSA4. If. at, use reactor coolant scarce tarms const stent =t ta t'e 3mses in v. G-Cni. Response. (4o.133) ';1Jlu ti.i-2. a n, t -3 u pd tb w % M a o._ L_. _i_.i s n J_. _~ r_ ~:., y. cow.dJ>% e'tu m % & s& =n mwni a us !-4 mig s p%. a!'=f,uu) a ~ %>7 3 u s n h v. e g _s q% tio.Ly ~ 4 g 1 at e -l- >ua > A f l D A.L - a. h 4 ti,2,3 h ll / u

-.-..-....-..-----..-.-...-.-.--.-e.---G.=::-_M-i 460.14 Provide additional information on the following items applicable to "~ (11.3) the gaseous waste management systems: a. Since your system description, tables and figures in Chapter 9 of your FSAR do not clearly indicate whether there are provisions for )- both HEPA and charcoal adsorbers for the reactor building pressure control mode and purge exhaust, provide the appropriate information relating to filter units for the reactor building. 3 p O W.5-R. i The filter unit, marked future, on Figure 9.4.7 is to filter - - ~--~- tha containment exhaust if operational msasurement of radioactive emission indicate that filtration is needed to ~ - " - - - - - - meet Appendix I limits. This exception to the GESSAR PDA requirement for the filter ~" - } unit was negotiated betwesn TVA and the NRC for the Hartsville and Phipps Band STRIDE units. (GE to provide exact reference). The Nuclear Island. design provides space and provisions for the addition of the filter units. ..... =. =.. -= Total airborne effluent releases of nbie gases, including Ar-41, I t,. ~~ - " - - ' - tritium and C-14 and some of the particulates given in Table 11.3-8 of your FSAR, are not consistent with NUREG-0016, Revision 1, and are lower than corresponding releases for radionuclides cited in this document. We assume that you have not taken any credit for particulate ...Q....-- removal by HEPA filters in the building exhaust systems since you state in Section 1.8 of your FSAR that the need for HEPA':, and charcoal absorbers will have to be decided on a site. specific basis. Accordingly, verify that your estimated releases are conservative. You should note tnat using an off-gas release rate of 25,000 Cf/sec for noble gases after a 30 minute delay is not consistent with the basis provided in - - - ~ -- NUREG-0016, Revision 1. A release rate of about 53,000 C1/sec is ~ aporopriate according to this document. You should also note that the caption for Table 12.2-22 is misleading since the annual airborne releases from the various sources for evaluating the environmental g impact should be used for total plant release and corresponding 7 off-site gaseous effluent doses. Either correct the caption for Table 12.2-22 or revise the contents of the table so as to reflect expected releases rather than design basis releases. Revisions to Table 11.3-8 should be Coordinated with corresponding revisions tc gaseous effluent dose estimates given on page 11.3-25.

u r (i'. 5S Pc MS4 b M' NoSLG GA 5, TD.mun y AnAcou A t M wASE R rrES iWTW2 - D CW*0 E AA/S is.g, iuv. t n,p a- ?, o Aw@r s An* SAst6 o u 9 M. [Su. THE pon. 30 st.m ss. TH$'NvMfbM. 15 ~2 L' A t0 c.3 < 1o 6 uhe c n mm ws ca. i, 2 w. c. s - ~ - ' A oeTon k #esco.wnow of TASLE 17 '1 - T2-is f~oun>o ~'^ - ^2 ^ r- / IrJ PAR A c,eAPd li. 3. A. G. E riTLs i s e n Adc.O ~ To " pr s v.,. rJ B Me s A J N grc Rp.C A it 7.W, of N ehr. e,Wt 5 t IO1.6 5 - F.:A G 9 s.a N M r.v ~f mg ?,KT e J At.s A7 4J 3' Add flow rate measuring devices for the monitors and samolers for c. all the airborne effluent release pathways. 9 a.-s e m s-e ...y.. The 4'oslow;n... }...e...f..6.....u...e..w~t....$t...w.. wcu. js.....a...<...e... ca.n..e..m.....t...%...cAes.e.p.e..d... w Jek. s clue.t. : 6ow<-..........o...te m.....e.. a...s... ..i.n...}..d...eo...i..c. e..s........i..n..... t...h.. e..... s a.m...p.....]..,.s... - )...o....f.. gas....y.r..e...t...<..e...o tm ewt (sce.... F...:.3.... 7... 0, - t o n. )..=

6...

f ( se F 3

7. 9....... )o.. b)..

- Q) o ff as ,a,.t...t....v..e...a..tmewt- .....-..e.. ~ -..g Ei : 7 G-lob)... ,,,...S.....o.fC. as....v...e..*....t.....p..te.... t...e..c... ( ( -3 3 .e.m... t..n...a...t..i.o...n.... dh.:..k. a...v..ge... _ m.....o...n.. ib....s.... .@- conta.i..n..m.. est

u..

~[~..(5)... 5hi.eM. wn..tdus HN.%..-..lE 5..- W h @S - ')... S 3 Tht...io.l. Wi.g...pikWne s d4.... net..have, f. low.. men.s.yo'. W . ekO tes.. ckdi.cak.A d6tshc.t%...d.....c... g....f(qqc.s..s.... haA'a,ti.o..rs... W on d._o..h h... mo4imuvn toAcA.[10WS CLssetia. tea w(S tiu. che.Ct 3 b u v, g, g, %tk ves g.t..e..tNe, e..t..w....t..s'. \\ ai.i. on.. dtgj. ... Oddut..i..o..n..c..J.......h... o.w...... w.as..u....a... - !.. M t9iS.C'a....h %C5e d.3.%15...W..%....Wdhe k.....y?.WCaN... Sto.PC.: ...6..)... f x..e..L...... B.. u.d... d...s...... . hN A C. ..t. ...m..e.....T.., s...{og....E. d...w...r. (.....Y..n....o...d. L...... ..1....) Con.M.. ( w... ...Q. ).....Q....,u.. w\\. s.a...r.Oud.d...i...n.[.. hN A C........ .S f0.)..MN\\.dS... MC >.. ..I.'i).... t 9 . T...h...e... Sta.v.db..q..(.>as .T...r...e.atm...e..d.....,.....%st..em..anA. %....... RoAcaste. B..o.d\\d.g ..h. v...A...c... s...e..\\s.t.. e. m.. ar..e.. not..eA Q.. u ..t...w.....G.......o....n... T. 6 - \\ t a )... as.n. a.o. i. . h.....e. to.. mowin....cie9.5e5 %.).k...N5W\\ek.... ..htt....ca5%er..

s-d. Since the off-gas system is located in the turbine building which l~~~ is not within the scope of your design, state whether the design of j ~ - - ' - - ~ ~ ~ the off-gas system lies within your scope. If not, state whether the off-gas system you have described is an interface requirement .) for the balance of plant. e State whether the source terms you have used to evaluate off-site doses due to a postulated failure of the off-gas system are consistent with Branch Technical Position ETSP 11-5 (July 1981). f. State whether the seismic criteria for the proposed off-gas system will conform to Section C.5 of Regulatory Guide 1.143. In responding to this question, a cross-reference to another section of your FSAR is acceptable. ~ 1 . @JL,.3 p DW J R. - ~ G.- y6s ~ * .._._.f.* . A_. % W EJ. M 7"* R? FGAG. _ s.vs,7 re.._ $ c c.aic A _ _'To. M.yp.3,45M.uij si. __._.LZL h..MN._ % 5. W99 EV M..St NCad.fhit.vrt4_ dE Ahl htThk.c.o W6W d T, j _.. _ _.____ _.. 5_$ d %..W5.W.cEf GM S vS#e...% _ Bf._)_mO.%'T"..tM Aw A c.T v6 c.*m.@MA' 7 : _T*4..t L ss M_o.: s::r a,,np htHT...Ta.Npr 7%. Bc.+v e..e_.y(cHea4L tos.Tip s {*_$f.G.. M S4.L4 "TO.._..h W. Ch.0 J

44 0.15 ? 160.15 - Prsvide tuttf onal Infomation on t.ne foll:41ng ite :s isol tent e :s :,,e (11.4)! solid radwaste systes: j a. Provide the isotopic breakdown of the total curie content of %et solid wastes that are espected to be shipped annually to a licensad burial site, accounting for the mintmuss decay available during starne I. prior to shipment. The total should include contributtons from: (1) evaporator bottoms associated with hl h conductivity and detergent 3 wastes; (2) spent resins associated with reactor water cleanup, radwaste, regenerant condensate deep bed, fuel pool and suppression pool cleanup deafneralizers; and (3) filter sludges. Provide an estimate of the nur:bar of containers which will be shipped annually. &SPOA/SE,' Ydod,/f A Wuht'b'$ Y Nf & /0 u 1. 2. 3. i. c M,ua b + L a L N. 7716LE //, 't'~3 &)A!W& s yesn uktu s M 41 F H, 2 -M e ) 9 /2 ~

460.ll b-b. Expertence alth operating 3WA's indicates that a det; '2e4 contensa e pollshing systes can 3entrate a st ntficantly nt;her ut. e af solidt!f M i " wet

solid wastes (i.e. stout 4L.CCO cu3tc feet for a 14C0 !*wt pl ant) than that presented in Table 11.4-2 of your FSAR. Accordingly, verify that your inputs to Table 11.4-1 of your F5AR are correcc.

1 R ES?ck)sC: Altoo.15 6 Ocw fk. 8 t d A 's ott U S Q "* A ~ h- /7so. to,13eo M J/7 wi/h tb j !' N ar . 84 f. //. 'l-2n... - L d y pE / '._~..T.~~ i i~ ~ OM ~ . au wreay JWJA. ..n arue,h:2 duj. bd, A Ap&p %. 7leith p g e- .+..m. e. ... + m . ee m 4 8 j L. .,/. ... _. i. ) D ). /3 m

~' .-........-._..-...,...,__:.r.y._. kh0/5 a

c. Add the suppression pool cles.9up wastes la Section 11.4.L of yaw psA.t A hk ka bh A S ' 5 d 'w h kkoso :

1.4.1 Design Bases 11.4.1.1 Power Generation Design 3ases 3 l The solid waste management ays .. provides the capa i for I solidifying and packaging wa tes from the reactor water cleanu suprmnen Peel cresw spyt'eti I system, the fuel pool coolir g and cleanup system,4 the Etquid rad waste system, resins, and pa ticulate wastes from the conden e cleanup system. Wastes from the q tems will cons' spent resin, evaporator bottoms, diatomaceou.4 earth, and other filtering 3 media. The solid waste management system also provides a means of c'om-3 pacting and packaging miscellaneous dry radioactive materials, such as paper, rags, contamir.ated clothing, gloves, and shoe coverings and for packaging contaminated metallic materials and incompres-sible solid objectives such as small tools and equipment parts. ) 'The solid waste management system is designed so that failure or ) mintenance of any frequently used component shall not impair sys-tem or plant operation. Storage is provided ahead of process units to allow hold-up in case of delay for maintenance. Drum capping and sample retrieval are performed locally. The operating philosophy of the solid radwaste control system is manual start and automatic stop with all functions interlocked to provide 7. a fail-safe mode of operation. ti.4-1 /#

-- :^.=-- =- -~ ~ m (60.15d. c'eseelea your pravt st2ns for c: 9 f ij.1ti 3 rice i t'ac. at:41 ?2si u n 11 ET5311-3, Revision 2 (July lHL). four !ts:rt; tion sent a tac 14:a: (1) the curds and dratas3e provisions for centstalag r23!:settye sollts; (2) a refarence ts the peccess centrol pecge:1 as an 1.1:erface re:;uirennt; (3) heat tracing for tylporstar concenteste pipioj 4 and tanks tnat are Ilkely to solidify at 3-bient tes; erat 1res; (l) flush!.y coanections. hte ver appropriate; (5) tne direct venting of at;ut;nent.htch uses canpressed 13ses for tne transcort of resins or filters slu!ges; (S) the appropelate was:e storage capacities for tanks acct.nulating spent resins feca the reactor water cleanup system and other sources and (11ters slu'ges in accordance with our position in the besnch technical posittoa cited above; and (7) the value of the available este storsge area for both the high and low-level wastes. t i Response: w e. e s l.. (i) sec.tik. sa.s.i.i et sect. ihm 9 3.3 discus p ea vls/**5 4.*c..nta w wg vai\\ adive a pills. 4, d.4 wiup a p4,uu,- + A~e., M / L (a) a.g s+ Ut+ y7 9 1,<<<- A %g =. 'Gb -.1.. Q LJp e :m~ G cc .h* hl, p W re 'wd c % w --- e -. evils t4 n 'el M 4 e = f =. (3) Weat. t eac'in9 of c.onechtrate piping aeid tasiks is shown on Pi ess li.2-aj and st.z -z h. t (O PeeJerions -Fo* pe re awd equipmemi 4L.tking as.e close.nbed iw 0km@ev n a.wd skeuw en s&e spec:i4ie P 4 to +oe t.kd agsteen. (y) an sJr m. 3 5-n4 At suu e=p- ~J a4 (n. .m. t4 aa. & G dr M ve.J.spJ~o s.3%rg 4, am <at~. s= mv,it.2-4. m. (6) storage capadities few spent vesh an.( c.\\es.w p phs.se separators are *Peeted. in Figines ll.1. - I b. (7) ststa.ge ca.pai. sty 4e, solic[ waste is 115t.ussed ) pavag raph 11.9. f.2. in A

460./5 e 3. Ada sa laterface ra.pte. tent to csitral t. a rilaise.it st sar e dusts generated during the c: pi: tion pr: cess tar

2ry* soli 3.as es, i

T Rsspa.sg: so,ir e / /-f

41. = - -

- _. _ i :..a%- _.. ~. y- .. ny -. 1-/ L _ : / /.t_^ . L : '. - ~~ t_ f,n.---, ,., ~ . /,. _ / y,_ 9 ///Ayma n the s;nu,,u Awit/ . b R> y4M( ~

l

] Ji &. f a u. v. a. sa. 0 e p i ) l 0 /4

r ( 46k16 Provide additional information on the following items appMcable to the (11.5) proc'ess and effluent and radiological monitor'ing and sampling systems: Provide in tabular columns, the sampling frequency, the minimum anal.ysis a. frequency and the sensitivity in C1/cc for the following airborne effluents and process streams: 1. Grab sampling for the principal gamma emitters and tritium for the plant vent, turbine building vent and radwaste building ventilation system effluents. 2. Grab sampling for the principal noble gas gamma emitters for the off-gas system, the drywell purge system and the fuel building ventilation system effluents. 3. Grab sampling for iodine in process streams for the off-gas treatment system; the drywell purge system; the auxiliary, fuel, radwaste and turbine buildings vent systems; the evaporator vent systems; and the pre-treatment liquid radwaste tank vent gas systems. 4. Continuous sampling of the effluents for fodines, particulates l and gross alpha emitters for the plant vent, turbine building ' vent and radwaste building vents. = ~ Your sampling and analysis frequencies and sensitivities for Items (1) through (4) above should be consistent with the appropriate frequencies 1 and sensitivities in NUREG-0473. Revision 2 (February 1980). State whether the turbine building monitoring and sampling provisions are within the applicant's scope. . _.. ele.spo % n ~~ The, W c.hdud., 5 fed 6...._tn.. t_.i..o..n...s_....f..o.r radio..t gi..c..a.l..... e..fAuents.......[o r .g.a...b..... MM. ....$thufw. .,. min.hgyg.. o.ys.gsis.. f.egge.neq...om4..he_.. ~ '._,....S.h5.8N.h....%h041d-he Sf0.V.@.ed e....h. ..the. afpWcfwt_ ckfinj....hj5.. 3. ....s. ubd*d-of the-waste. sayny......g aa. a=Lps..ggam...th..

....Cb.n.(bemann.....w im N1 1.a.t.

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b. For liquid effluents and process streams: ~ 1. Add your proposed grab sampling provisions for.the service water and the detergent drain tank effluents to Table 11.5-6 of your FSAR. 'l 2. Add your grab sampling provisions in the process liquid streams for the compo, tent cooling water system and the laboratory and sample system waste systems in Table 11.5-4 of your FSAR. Clearly indicate whether the fuel pool filter-demineralizer includes both spent fuel and refueling pools. 3. It is our position that your grab sampling and the associated analysis should identify the isotopic composition and determine the concentrations of the principal radionuclides and determine the concentration of the alpha emitters in addition to determining the gross radioactivity for all liquid effluents and process streams. 4. Explain what you mean by the waste sample tanks and the floor drain sample tank to which you refer in Table 11.5-6 of your FSAR. We find these references to be unclear since the discharge to the environment from the liquid radwaste system can only be from either the excess water tank or the detergent drain ~~ tank according to your system description. -~ 5. Add the radionuclide Fe-55 to the isotepic analyses of effluent ~ - '~ and process streams. ~ ~ .R.a.3 paw: pit _..... -.=...,.m. The...yd. 5**.f.Wj...pMi%...ior A-4ti"ged %.bK.a.fe. ,_.[ @..L*d '" Nit'. IL'S 48 - Ik"...p..50'M.\\Ig.....?fo#Cn6_.:lef.SerW *- ... o,te c.. w m h e_ w h 6._ %..a.gp.Wtod.s Scope.,, W s...'.yD.k...M/".h.}... yid).).i.CM.... f k.8... CC9Cf f A..CCc ... M f / hst'39. 3. W "6.. % ISPcfM9'.'.g.. Mk>en.f%...).$t.m. W'%te. Mp.se.ms...w.i 9..b. 3..n., ...t..T.p.'Aa n.ts seef.e... ..T...w...e_.....-[u.e..l.......\\ o.\\....-f.. H.. e..r.. - dLes n e m.in......e.. v ' b lha., 5!.u. CS..M,W}...h.....in.e,iwb .b. c..+. h......w...... s..{.e. ath....e...\\ _ cm.t (c. h..eM.]...yo...c.. D t m......b....s. a...mp.in....g...a.. d W aso.s..... t4 wA,ses Sc cmkokedwL.c.. on. centeo.G a G __........... _e. w &....G.1..h. e ctppu t e_. a.c y.,= m.._.n.-...s.-. ...s...c..o... g-m d *h.e...._..w...a..s....t..e...., s...a. m....?.t.e..,.bks a..m. ct the., f...i..c.e....r....6....<..c. 6....... . _. 3., . %. e. e n g l .t.a. m.K. -t d e.._..4.3n B,te, i i. 5 - L,......u be deteted..

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c. State whether the design criteria for the radiological effluent monitors will conform with the manufacturer's standard per ANSI N13.10 (1974) and the staff's position on quality assurance in Sections C.4 and C.6 of Regulatory Guide 1.143, Revision 1. If not, provide justification for any deviations. W S (> 5 %.L R. .1t is b%s.n at Ms the., wwewe a e48u.at meses wia cefum vW ^2I H l3 50 U9").. 9^c8-M 6 Cote *f 5Ngh... b {r.o. o..m...s..... rat'on Twe. m+.a ef GE wy=L =nes wm bewnea-g#e. geht. z

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o l l N 460.17 Since the radiological consequences resulting from the release of t contaminated liquid to the environs due to a postulated failure of the liouid tank are dependent upon site specific geological and hydrological i parameters, provide justification for not leaving the evaluation of the off-site radiological consequences within the applicant's scope. l Our understanding of your proposed nuclear island is that your scope of. work should be only to supply the source terms, In this regard, i your assumption that iodine is the critical isotope which will determine whether radionuclide concentrations at the nearest surface water supply in an unrestricted area will be within the limits of 10 CFR Part 20, is not valid. (In general, the long-lived isotope Cs-137 is the critical i sotope. ) L. 1 A4 ".a c WAT C. Gl %wvt 1 64 Ne u ed JV T 4 f-A N ALs/ Sis FdL Postub-4No TM W-

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ATTACHMENT NO. 4 DRAFT RESPONSES TO AUXILIARY SYSTEMS BRANCH QUESTIONS

( 310.07 In addition to the possible missile sources you hav.e identified, (3.5.1) verify in Section 3.5.1.2 of your FSAR, that your analyses insidt containment have included the reactor vessel head bolts and the automatic depressurization syst.em (ADS) accumulators. Rn.sp6w3c kg acca 4hs oi do A S. b A 4 tv.p.7 e s. ij W HwM.S .rA N ow J' 3. 5 /.1, ). ,1 ?"d R . 3 5 I 7. 7-...As ckst s c vs W u m d 4 r i 4 4 % ( s)

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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. -Aust' 3.5,1. 2,4 Evaluation of Potential Gravitational Missiles Inside Containment Gravitational missiles inside the containment have been considered as follows: S2itmic Category I systems, components, and structures are not potential grcvitational missile sources. Nun-seismic items and systems inside containment are classified as follows: a. Cable Tray j All cable trays for both Class IE and non-class IE circuits are seismically supported whether or not a hazard potential is evident. b. Conduit and Non-Safety Pipe nan f uP/*l W /F 'I % cL%s.I E condusT /5.5E/Jm/t9// S /S 'AGN71/IGO As A th'IEN'TI s9c /tA2Mgn t0 $RFE7f- %1<% GyVIfin64. ALL REAcfDJ2 /WO M AJ-W ETf cia 55 P/9trilo I b S F! S/ h t t('s/ly ANN fh?$ (AJtfd b& 00GC2fft&1 OF A0WhTC GoiLOtO(o, c. Equipment for Maintenance 8 J All other equipment, such as hoists, that is required during maintenance will either be removed during operation, moved to a location where it is not ) a potential hazard to safety related equipment, or seismically restrained to )- prevent it from becoming a missile. l l p w-

238 NUCLEAR ISLA!!D R v. O f,3.5.1.1.2.2 Missile Analyses (Continued) I divisional equipment makes the design acceptable. All safe shutdown functions in the Reactor Island design have redundant backups and these redundant items are separated either by considerable distance or a missile- , proof barrier. Based on this, the probability of a valve bonnet missile striking both Division 1 and 2 vital targets for safe shutdown is extremely low making the resultant probability much less than 10-7 times per year. ~ (2) Valve stems - All the isolation valves installed in the reactor coolant systems have stems with a back seat which eliminates the possibility of ejecting valve stems even if the stem threads fail. Since a double failure of highly reliable components would be required to (l produce a valve stem missile, the overall probability of occurrence is less than 10-7 times per year. Hence valve stems can be dismissed as a source of missiles, moosmte ensegy vessels less % z75 pai9 4 to.0 7 mru M~ c. red; ole missile seurse. (3) Pressure Vessels gThe pneumatic system air bottles are designed for 2500 psig to AStiE Code Section III require-ments. The bottles are not considered a credible source of missiles for the following reasons: (a) The bottles are fabricated from heavy-wall rolled steel. (b) The operating orientation is vertical with the ends facing concrete slabs. The bottles are topped with steel covers thick enough to preclude penetration by a missile. -3s _ (c) The fill connection is protected by a permanent steel collar. 3.5-9

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 3.5.1.2.1 Rotating Equipment (Continued) l' By an analysis similar to that in Subsection 3.5.1.1.1,.it is concluded that no other items of rotating equipment inside the containment have the capability of potential missiles. All other are incapable of achieving an overs e condi ,m.v fonemmoie, fHc AD$ ACMMOfAMb O DrstGuto N n. 2.co Ps 6 upter Prnsort-3.5.1.2.2 Pressurized Components i (4 215 Pts) to Assyte* SECT 4ME TW~ monts wo n,4. herefere aar cessotad A Credibl. moslic suwce. 4 Identification of pot al missiles anc their conseq6ences outside containment ar ecified in Subsection 3.5.1.1.2. The same con-clusions be drawn for pressurized components inside of contain-ment.g one additional item is control rod drives (CRD) under the reactor vessel. The CRD mechanisms are not credible missiles. The CRD housing supports (Section 4.6) are designed to prevent any significant nuclear transient in the event a drive housing breaks or separates from the bottom of the reactor vessel. Since these hc,usius supports are in close proximity to the drive housing and the supports have been designed specifically for the separation event, there is no reason to consider the CRD mechanisms as credible missiles. 3.5.1.2.3 Missile Barriers and Loadings credit is taken in some cases of rotating and pressurized com-ponents generating missiles for missile-consequence mitigation by ( structural walls and slabs. Penetration of the following walls M =1=bs by potential missiles is not considered credible: (I) drywell wall, (2) weir wall, (3) upper pool walls and floor, (4) reactor pedestal, and (5) other interior walls and slabs. 3.5-15

410.16 In your letter dated February 12, 1982, you state that the review base for Section 4.6 of your FSAR is the Clinton plant. Revise your FSAR to include the additional information provided on the Clinton docket in the course of the Clinton review, including that additional information which was submitted to close.the opcn items in this portion of the Clinton SER.

Response

The appropriate GESSAR II sections will be revised to include the additional information provided on the Clinton docket in the l course of the Clinton review pertaining to the Scram discharge l system. The following sunnarizes this information: 1) Modifications will be implemented (and GESSAR II revised accordingly) to the Scram discharge system that will comply with the criteria enumerated in the Generic Safety Evaluation Repo-t - BWR Scram Discharge System. 2) A discussion will be included in GESSAR II describing the effects of a drive / cooling water pressure control valve failure (closed or open) 3) Control Rod Drive (CRD) system specifications presently ccmply to the requirements of NUREG-0619 for the deletion of the CRD return line. D:monstration by test, that CRD flow to the reactor vessel is equal to or greater than the boil-off rate as discussed in NUREG-0619, is no longer a requirement. 4) GESSAR II Figure 4.6-5 will be updated to provide a complete P&ID of the control rod hydraulic system as modified. M o d 8x c Ah t u^.I d6 k GG.5.5 A(L TL M sk o-o -4 4 b y g.a..r. m e

W S E E n_p3 ZR@hXl/ 138 NUCLEAR ISLAND Rsv. O M ~caAus d Tog +10.16 4.6.1.1.2.4.2.2 Accumulator Charging Pressure (Continueu; D0 ring normal operation, the flow control va1ve maintains a con-stant system flow rate. This flow is used for drive flow and drive cooling. 4.6.1.1.2.4.2.3 Drive Water Pressure Drive water pressure required in the drive header is maintained by the drive pressure control valve, which is manually adjusted from the control room. A flow rate of approximately 16 gpm (the sum of the flow rate required to insert 4 control rods) normally passes from the drive water pressure stage through eight solenoid-operated stabilizing valves (arranged in parallel) into the cooling water header. The flow through two stabilizing valves equals the drive insert flow for one drive; that of one stabilizing valve equals the drive withdrawal flow for one driva. When operating a drive (s), the required flow is diverted to the drives by closing the appropriate stabilizing valves, at the same time opening the drive directional control and exhaust solenoid valves. Thus, flow through the drive pressure control valve is always constant. . low indicators in the drive water header and in the line down-stream from the stabilizing valves allow the flow rate through the stabilizing valves to be adjusted when necessary. Differential pressure between the reactor vessel and the drive pressure stage m is indicated in the control room. i &> ear [ (8) J-4.6.1.1.2.4.2.4 Cooling Water Header The cooling water hea der is located downstream from the drive / cooling pressure valve. The drive / cooling pressure control valve is manually adjusted from the control room to produce the required drive / cooling water pressure balance. 4.6-15

1 IMSGC c,A.( ved.ve (PcVb If the ages set were to fall to a full-open position, the cooling water pressure would increase and the drive water 4~ pressure would decrease. The resulting cooling water pr.es-sure increase could cause control rods to drive inward. The existence of rod drifts would be alarmed to the control room operator for appropriate action. The resulting drop in drive water pressure would make normal control and notch movements impossible but would not affect the ability of the scram function. Conversely, if the 9994 PCV were to fail to a full-closed position, the cooling water pressure would decrease while the drive water pressure would increase. The reduction in cooling water pressure (and flow) would eventually lead to high CRD temperatures being alarmed in the control room. t The CRD system's scram function would not be affected by the increase in drive water pressure. i 1 bek b )c % m e,,, d 23:~ ~ %;s, i t y& + s rw b w W. ic J eel M- ~ v ~JA~s% 0 A pcv w~id ELL tl 4.c 4 46. f cw.cAe A. ,g e 4

GESSAR II 22A7007 238 NUCLEAR ISLAND Rsv. 0 4.6.1.1.2.4.2.4 Cooling Water Header (Continued) The flow through the flow control valve is virtually constant. Therefore, once adjusted, the drive / cooling pressure control valve will maintain the correct drive pressure and cooling water pres-cure, independent of reactor vessel' pressure. Changes in setting of the pressure control valves are required only to adjust for chahges in the cooling requirements cf the drives, as the drive ceal characteristics change with time. A flow indicator in the control room monitors cooling water flow. A differential pressure indicator in the control room indicates the difference between reactor vessel pressure and drive cooling water pressure. Although the drives can function without cooling water, seal life is chortened by long-term exposure to reactor temperatures. The tem-perature of each drive is indicated and recorded, and excessive temperatures are annunciated in the control room. 4.6.1.1.2.4.2.5 Scram Discharge Volume The scram discharge volume consists of header piping which con-nects to each HCU and drains into an instrument volume. The header piping is sized to receive and contain all the water discharged by the drives during a scram, independent of the instrument volume. During normal plant operation, the scram discharge volume is empty and vented to atmosphere through its open vent and drain valve. When a scram occurs, upon a signal from the safety circuit these vent and drain valves are closed to conserve reactor water. Lights in the control room indicate the position of these valves. During a scram, the scram discharge volume partly fills with water' discharged from above the drive pistons. After scram is completed, the CRD seal leakage from the reactor continues to flow into the scram discharge volume until the discharge volume pressure equals the reactor vessel pressure. A check valve in each HCU prevents reverse flow from the scram discharge header volume to the 4.6-16

uraann 11 azuwryu 238 NUCLEAR ISLAND .s-R:Lv. 0 l ~ l 4.6.1.1.2.4.2.5 Scram Discharge volume (Continued) i s N N drive. When the initial scram signal ir_ cleared from the reactor s s protection system (RPS), the scram discharge volume si'gpal iy ovdr-ridden with a keylock override switch, and'-the scram dishharge volume is drained and returned t'o atmospheric pressure. 'A s I Remote manual switches in the pil6't valve solenoid circuits allow the discharge volume vent andsdrain valves to be tested withdut e disturbing the RPS. Closing the $ cram discharge volump valves. allows the outlet scram valve?aeats to be leak-tested b'y timing.the ~ ~ er / accumulation of leakage inside the scram discharga volume. (s N m ~. ~ (Seven liquid-level switches activated by six transmitters con-nectedtotheinstrumentvolume,monitopthevolumefo'h\\nbnormal water level. They are set at three different levels. At the lowest level, a switch actuates to indicate-that the volume is not com-pletely empty during post-scram draining or to"i'nd'icite that the volume starts to fill through leakage eccUmulation at other tines ,-y during reactor operation. At the second level ;two switches pro-4 duce a rod withdrawal block to prevent further withdrawal of any L em controlrodwhenleakageaccumulat)sto,.halfthecapacityofthe ~' instrument volume. The remaining four switches,<are inter 6cnnected + with the trip channels of the Reactor' Trip dystem and will initiate ~ ~ ' a reactor scram should water accumulation fill the instrument Js '

Y N

volume. p -s a 4.6.2.1.2.4.3 Hydraulic Control Units q 'N ) Each hydraulic control unit (HCU).furnishqs pressurized water, on j sign.1, to a drive unit. The drive then po0itions its control s ro'd as required. cperation of the electrical sy' stem that supplies f scram and normal control rod positioning signals to the HCU is described in Subsection 7.7.1.2 (Rod Control and'Information System). ~ w 4.6-17 +' ,-~n---e we- -. ,,--,m

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5; 2 n, -.}. o -- - - s,. msgy s s e s., 's ' s ~ ~ v (t s .;y g 4, t -ey ~ R J edundant Scram Discharge Volume (SDV) vent and drain valves are ~ tf. provided.:: p;rt :f t.'.: ""! r;dificatic.; ea- 'n t u % ::

b t: .

The redundant SDV valve configuration assures that no single active failure can result in an uncontrolled loss of reactor ( coolant. An additional solenoid operated pilot valve a ntrols tne redundant vent and drain valve. The vent and drain system is therefore sufficiently redundant to avoid a failure to isolate the ~ SDV due to solenoid failure. The vent and drain valve's opening .and closing sequences are controlled to minimize excessive hydro-dynamic forces. ~ ~ All SDV piping is raquired to be continuously sloped from its hit point to its low point. A vent line is pro ided as part of the 5: ram discharge system to l assure proper drainage in preparation for scram reset.

2. _C ::

-rkis prMr is t ;;n.1 a dedicated vent line with a nonsubmerged discharge to the atmosphere. Furthermore, additional vent capability is provided by the vent line vacuum breakers. The vacuum breaxers are required to have a differential pressure no greater than 5 inches of water. s e \\ "~ Tne SDV vent and drain lines are required to be dedicated lines that discharge into the Radwaste System. Vacuum breakers on the SDV vent line and shut-off valves on the SDV vent and drain lines ' preclude water from siphoning back into the SDIV from the Radwaste System.

The SDV a'fid associated vent'and drain piping is classified as impor uit to safety and required to meet the ASME Section III L,

Class II and Seismic Category I requirements. l' u T \\ ,\\ n. \\ w % f N, N g, I \\ b s ,i s \\,

CESSAR II 22A70'07 ~ 338 N0 CLEAR ISLAND Rsv.'O .f /. l ,t 4.6.1.1.2.5.3 Scram (Continued) + V The CRD accumulators are necessary,to ecrum the ' control rods Wit in the required time. Each drive, however, has an internal ball-check valve which allows reactor pressure to be admitted under the drive piston.. If the reactor is above 600 psi, this valve ensures rod insertion in the event the n'ecumulator is.not charged or the inlet scram valve fails to open. The insertion time, however, will be slower than the scram time with a properly functioning scram d system. The CRDS, with accumulators, provides the following scram perform-ances at full power operation, in terms of average elapsed time after the opening of the RPS trip actuator (scram signal) for the drives to attain the scram strokes listed: From Full-Ou'c (Notch Position 48) To: Notch Position 44 28 12 Stroke (in.) 12 60 108 Time (sec) 0.28 0.91 1.620 4.6.1.1.2.6 Instrumentation The instrumentation for both the control rods and control rod drives is defined by that given for the red control and information system. The objective 6f the rod control and information system is to provide the operator with the means to make changes in nuc-lear reactivity so that reactor power level and power distribution can be controlled. The system allows the operator to manipulate %gVn(l control rods. I 'The design bases and further discussion are covered in Chapter 7, " Instrumentation and Control System." 4.6-23

r (NSW 1 b ( Diverse, and redun' dant level sensing instrumentation on the Scram Discharr.e Instrument Volume (SDIV) is provided for the automatic scram tunction. SDIV water level is measured by utilization of both float sensing and pressure sensing devices. Instrument taps ( have been relocated from the vent and drain piping to the SDIV to protect the level sensing instrumentation from the flow dynamics in the scram discharge system. Each SDIV has a redundant instrument loop. A one-out-of-two twice logic is employed for the automatic scram function. Thi. instrumentation arrangement assures the automatic scram funct.on on high SDIV water level in the event of a !I The SDIV scram level instrumentation arrangement and trip logic allows instrument adjustment or surveillance without bypassing the scram function or directly causing a scram. Eacn level instrumen: can be individually isolated without by:assir.g the scram function. 3 =: r t -: ' - -r t- ; = b + '.gic h -... W + Technical S - 7. _ 4 Specifications will ensure that the scram function is not bypassec during repair, replacement, adjustment or surveillance of any system component. Supervisory instrumentation and alarms such as accumulator trouble, scram valve air supply low pressure, and scram discharge vclume not drained alarms, are adequate and permit surveillance of the scram system's readiness. j c k

410.19 dn your letter of February 12,1985! you state that the new and spent (9.1.1) Tirei st6 cage fact 11ttes which you propose for your nuclear islend are (9.1.2) the same as_those for the Perry Nuclea'r Power Plant. However, your FSAR describesthich density _ ne@ and spent fuel storage facilities which were not evayuated auring the Perry review. Correct this apparent, discrepancy.[ t 4t0. tot Resvouse _ sTouws. v e s THe sessaksew % spear vuet. IM he u OENstTf. SotwE. mootFiCG TiGO W LLL- % M\\(o QewtT of M Tim lontis-up witt ocev(t es A d 4l0,Z'5 ReSCbn5f. ,f % W'E he LgTTer2. OF M Y2,.19 8L WMF27 "" Respect w Fust_ SToreM, \\J 3 NW CF N Tus Gss500.tr M W M W FoQsu\\u G P A 6*M-

R;v. O 238 NUCLEAR ISLAND a f, 9. AUXILIARY SYSTEMS _ FUEL STORAGE AND HANDLING 9.1 THE THE NW FUEL. N.JC W FUEL STod.'%5 RAac.5 ARE g CAJ S4ms. IMH CErc(Ty De6tG*0, THE. NEW POET. 72PKX.5-FOR. Erruen t*,y OR Sut5 morse. 5,70 GEE OP FUEEt. BE 05EO j M FoW46 t.ou, Desc.RtBE. The Desi6e op 'These l AAC IDdMTtow, NE Dd rvWTton {-Ca DS luce The-f WEW 5 A R&.xs wtu oww Be OlesemTeO Uwed.R. d NPSotrt.m% he. CE.TAtt eo Wi$th ci: The.PN<. MW 8S <CWTAtWO Un Sec.Tww 9.(.2. @ Scery jt.ef f $ T O /2c k y d, 9,I.\\ M69 REL. $T0QP65 l,l,) @SW bb , - Nuclear D?$t W C4 l, l, I., l '~ A full array of loaded new fuel racks is designed to be subcritical, by at least 5% ak. Neutron-absorbing material, as an intecral part of the design, is employed to assure that the calculated k,gg, including biases and uncertainties, will not exceed 0.95 under all normal and abnormal conditions, or 0.98 under optimum moderation. (a) Monte Carlo techniques are employed in the calcu-d es not lations perfcrmed to assure that heff exceed 0.95 under all normal and abnormal conditions. (b) The assumption is made that the storage array is infinite in all directions. Since no credit is i taken for neutron leakuge, the values reported as effective neutron multiplication factors are, in reality, infinite neutron multiplication factors.

(c) The biases between the calculated results and i experimental results, as well as the uncertainty involved in the calculations, are taken into account I as part of the calculational procedure to assure that the specific k,ff limit is met. %l. l, (,7 bIOPA 6 M M6M k) The new fuel storage racks provided in the new fuel stor. age vault provide storage for 30% of one full core e fuel load. t k The new fuel modules are designed and arranged so that fuel assemblies and bundles can be handled efficiently during refueling operations. 3.l.\\.\\.9 MEGt%tA 60 STRUCTVW DEwrJ (Figure 9.1-1) .3 g m The new fuel storage racks contain storage space for 30% f of one full core of fuel assemblies (with channels) or bundles (without channels). They are designed to with-stand all credible static and seismic loadings. \\ rn~S% i _' i m e The racks are designed to protect the fuel assemblies and ea I l g{ {j' ',! bundles from excessive physical damage which may cause the release of radioactive materials in excess of 10CFR20 pG and 10CFR100 requirements, under normal and abnormal ,' p i$ conditions caused by impacting from either fuel assem-Nh! blies, bundles or other equipment. (4% , % The racks are constructed in accordance with the Quality 'j Assurance Requirements of 10CFR50, Appendix B. l l The racks are categorized as Safety Class 2 and Seismic Category I, t ..a_.aam ~ (W

9,1.1, t. & Thh L -HYN c. D#5t60 $k Sa;t,aw 9,1.1,!,4 Fort SWE h5Co.53luh,, 9o I al, I, f NwiCQ-I(Au CmssderaTouns 6e< Sectse C,1,7., \\.s' ForL 015cussivo 1

9. \\. I \\.0 Dywome c pears us 1fe sec7 vn 9.I,2.,!,4 F0/L DiscosseA

/ 9 h!].7 Impact hy5:5 Gee SecTtow 9. t.2., t l Fo/t. Dt$cusu g

4

9. l.1,7 Facilities Description (MEu) F064. ST0 AAtrE.)

(1) The location of the new fuel storage facility within the complex is shown in Section 1.2. (2) The new fuel storage racks are top entry racks designed to maintain the new fuel while precluding the possibility of criticality under normal and abnormal conditions. The upper tieplate of the fuel element rests against the module to provide lateral support. The lower tieplate sits in the bottom of the rack, which supports the weight of the fuel. (3) The rack arrangement is designed to prevent accidental insertion of fuel assemblies or bundles between adjacent racks. The storage rack is designed to provide accessi-1 bility to the fuel bail for grappling purposes. Nominal fuel spacing from center to center is 6.56 inches by 6.56 inches. (4) The floor of the new fuel storage vault is sloped to a drain located at the low point. This drain removes any water that may be accidentally and unknowingly introduced into the vault. The drain is part of the floor drain subsystem of the. liquid radwaste system. (5) The radiation monitoring equipment for the new fuel storage area is described in Subsection 7.1.1.6.2.

9.1.1.3 S3foty EvoluOtlen h 9.1.1.3.1 Criticality Control The design of the new fuel storage racks, which includes neutron-j absorbing materials, provides for an effective multiplication factor (k,ff) for both normal and abnormal storage conditions equal to or less than 0.95. To ensure design criteria are met, the following normal and abnormal new fuel storage conditions were analyzed: I \\ i l

73unaisL%Ln3xm 9.1.1.3.1 Criticality Control (Continued) (1) normal positioning in the new fuel array, and ) (2) eccentric positioning in the new fuel array (Figure 9.1-2). new fuel storage area will accommodate fuel (k 1 1.35 at inf ine 20*C in standard core geometry) from a multi-unit BWR facility with no safety implications. 9.1.1.3.2 Structural Design (1) The new fuel vault contains one 13x17 fuel storage rack, which provides storage for a maximum of 221 fuel assemblies or bundles. (2) The new fuel storage rack are designed to be supported above the vault floor by a support structure. Since the racks are freestanding (i.e., no supports above the base), the support structure also provides the required stability. (3) The racks include individual solid tube storage compart-ments which provide lateral restraints over the entire length of the fuel a:ssembly. (4) The weight of the fuel assembly or bundle is supported axially by the rack lower support. (5) The racks are fabricated from stainless steel. Materials used for construction are specified in accordance with the latest issue of applicable ASTM specifications at the time of equipment order. In ^ ' r

m MLMMU 238 NUCLEAR ISLAND R;v. 0 9.1.1.3.2 Structural Design (Continued) (6) The nominal center-to-center spacing for the fuel assemblies or bundles between rows is 6.56 inches. The maximum spacing between racks is 2.0 inches. (7) Lead-in guides at the top of the storage spaces provide guidance of the fuel during' insertion. (8) The rocks are designed to withstand, while maintaining the nuclear safety design basis, the impact force generated by the vertical free-fall drop of a fuel assembly from a height of 6 ft. (9) The rack is designed to withstand a pullup force of 4000 lb and a horizontal force of 1000 lb. There are no readily definable horizontal forces in excess of 1000 lb and, in the event a fuel assembly should jam, the maxi-mum lifting force of the fuel-handling platform grapple (assumes limit switches fail) is 3000 lb. (10) The new fuel storage racks require no periodic special testing or inspection for nuclear safety purposes. 9.1.1.3.3 Protection Features of the New Fuel Storage Facilities The new fuel storage vault is housed in the Fuel Building (Sub-section 3.8.4). The vault and Fuel Building are Seismic Category I structures. The Fuel Building provides protection from severe natural phenomena such as tornadoes, tornado missiles, floods and high ~1nds. Fire protection features are described in Subsection 9.s.1 and Appendix 9A. The stcrage rack structure is designed to withstand the impact resulting from a falling weight. Tests using a simulated fuel bundle of G.c correct weight and size have been conducted to O_ g __ -

CESSAR II 22A7007 238 NUCLEAR ISLAND R2v. 6 9.1.1.3.3 Protection Features of the New Fuel Storage Facilities ] (Continued) verify that the rack casting can withstand,the impact from a bundle dropped from a maximum allowable height above the array. Procedural fuel-handling requirements and equipment design dic-tate that no more than one bundle at a time can be handled over the storage racks and at a maximum height of 6 ft above the upper ~ rack. Therefore, the racks cannot be displaced in a manner causing critical spacing as a result of impact from a falling object. The five-ton general-purpose building crane can traverse the full length of the fuel building. A corridor is provided along the shield building side (not over) of the pools and vault; roof hatches are provided in the vicinity of the FPPCU equipment. This permits removal of major equipment by way of the hatch, thus elim- .inating the need to move these components along or over the pools and vault. The shipping cask cannot be lifted or moved above the new fuel vault because of inadequate clearance. Should it become necessary to move major loads along or over the pools, administrative controls will require that the load be moved over the empty portion of the spent fuel pool and to avoid the area of the new fuel storage va 11t. New fuel is carried to the new fuel vault and placed in the storage rack using the fuel-handling platform. During positioning of new fuel into the new fuel racks, the grapple is always above the upper fuel rack casting, and the grapple interfaces only with the fuel bundle bail and could not engage the fuel rack. Thus, the transfer devices used for new fuel handling to the new fuel vault cannot impose uplift loads on the rack castings. l M

.. :.e.. : ~ SE D "- 0,a m y) 9,[,2 6msI( F0EL cp,2., l Dt55troM L2ASE55 c),1,2. t, \\ MocLen/_. Nw (1) A full array in the loaded spent fuel rack is designed to be subcritical, by.at least 5% Ak. Neutron-absorbing material, as an integral part of the design, is employed to assure that the calculated k,ff, including Mases and uncertainties, will not exceed 0.95 under all normal and abnormal conditions. (a) Monte Carlo techniques are employed in the calcu-lations performed to assure that k,ff does not exceed 0.95 under all normal and abnormal g conditions. s (b) The assumption is made that the storage array is infinite in all directions. Since no credit is l i l l t l

~ taken for neutron leakage,,the values reported as effective neutron multiplication factors are, in reality, infinite neutron multiplication factors, n (c) The biases between the calculated results and d e experimental results, as well as the' uncertainty involved in the calculations, are taken into account as part of the calculational procedure to assure that the specific k,gg limit is met. % l,7,I.1 Storage ~ /%) The fuel storage racks provided in the spent fuel storage pool provide storage for 326% of one full core fuel load. g i k The fuel storage racks provided in the containment pool provide storage for 68% of one full core fuel load. g The spent fuel racks are designed and arranged so that fuel assemblies and bundles can be handled efficiently during refueling operations. Y, /,2.. /. 3 %ECW( $pp'. Structural i re 9.1-1) The spent fuel storage racks in the Fuel Building and Containment contain storage space for 394% of one full core of fuel assemblies (with channels) or bundles (without channels). They are designed to withstand all credible static and seismic loadings. g l ~ [ The racks are designed to protect the fuel assemblies e and bundles from excessive physical damage which may cause the release of radioactive materials in excess of 10CFR20 and 10CFR100 requirements, under normal and abnormal conditions caused by impacting from either fuel assemblies, bundles or other equipment.

h The racks are constructed in accordance with the Quality Assurance Requirements of 10CFR50,' Appendix B. h The racks are categorized as Safety Class 2 and Seismic Category I. M The pool level is maintained by structura,1 concrete walls with a stainless steel liner. The bottoms of the pool gates are sufficiently high to maintain the water level -g over the spent fuel storage racks for adequate shielding and cooling. All pool fill and drain lines enter the, 4 pool above the safe shielding water level. Redundant o anti-siphon vacuum breakers are located at the high point of the pool circulation lines to preclude a pipe break from siphoning the water from the pool and jeopardizing the safe water level. 1

y L_ ..The rcsko includa individuni colid tube otercga ccmpart-1- I _______..,......mento, which prcvida 1ct':rol rastrointo ov0r tha entiro. L ._____1.ngth of the fu.1 as..mbly or bundl.. Eu I ..._ __..___..The weight of the fuel assembly o'r b'undle is supported g._..'..___ axially by the rack fuel support. 'The racks are. fabricated from stainless steel. Materials ~ ~ ~ ~ ~ ~ ~' ~~ ~ used for construction are specified in accordance with the latest issue of applicable ASTM. specifications at the time of equipment order. l g ,The nominal center-t6-icenter ' spacing for the fuel assemblies or bundles between rows is 6.56 inches. The ,f , T " + 3 pacing between racks is 2.0 inches. - ~ ~ _. 4: @W& -.2 _.___..3 ) i , Lead-in guides at the top of the storage spaces provide ~ lguidanceofthefuelduringinsertion. l i r-e i i = fi The racks art. designed to withstand, while maintaining i l the nuclear safety design basis, the impact force gen-s erated by the vertical free-fall drop of a fuel assembly s from a height of 6 ft. 1 The rack is designed to withstand a pullup force of 4000 lb and a horizontal force of 1000 lb. There are no ,readily definable horizontal forces in excess of 1000 lb and in the event a fuel assembly should jam, the maximum . lifting force of the fuel-handling platform grapple (assumes limit switches fail) is 3000 lb. The fuel storage eks are designed to handle irradiated ~ ~ ' fuel assemblies. The expected radiation levels are well I~ below the design, levels.4 g__.---_. r = I r-g-_... 9 a.. @66. .e .66 t I l k

l I I" I I W l s 1 e. I e r 1r e E. l b l ..w In accordance w!ta Regulatory Guide 1.29, the high density fuel sto' rage l ,j racks are designated Seismic Category 1. Structural integri y of the rack "I. ... has been demonstrated fcr the load combinations below using linear elastic . design methods. The applied loads to the rack are (1) dead loads, which are weight of rack and fuel assemblies, and hydrostatic loadst (1) live loads - effect of lif ting an empty rack during installation; =.. _.,.... - (3) eneraal leads - the uniform thornal expansics due to pool temperature ~ cl,anges1 e (4) seismic forces of OBE and $$t; { (5) accidental drop of fuel assembly from maximum possible height (6 ft above rack); and (6) postulated stuck fuel assembly causing an upward fi..ce of 3000 lb. ~

7...

t s t i j I e .s { g ? l s 3 . _ +.. =- 1...._ I . [... -* y**~ 9 .a e l g s = -.. '... 1 .,.._..[. .p . f._......_. - ? I 1 l t ._L -t_._____.. .) l 3 I - ~, I n 8 i l l f _e_..__'.. I .,..-.m.....r .. ~. - - - -. - - a..a. Se .g 6. N. .= 'n -w

'F ,.*:mRZ22.g \\ l.L Sns M Mh)Y. ne lead esebiastions seasidered in the reek desian are (- (1) live loads: E, (2) deed leads plus est: 5, () I ~ (3) dead loads plus SSE: and (4) dead lands plus fuel drop Thermal lands were not included in the above combinations because they were 4 megligible due to the destga of tl.a rack (i.e., the rack is attached only st its base and is free to expand / contract under pool temperature chanavs). ) l De loads experienced under a stua.k fuel assembly condition ere less than those esiculated for the seismic conditions and, therefore, luve not been inc1r. fed as a load combination. i The storage rac*es are attached to the support structure by bolties, sufficient to counteract the tendency to everturn from hertsontal loads and to lif t ( free vertical leads. He analysis of the rack assumed an adequate supporting structure, and loads were generated accordingly. W ci. i 2.1.5 Stress analyses were performed by classical methods based upon shear and soments developed by the dynamic method discussed in Subsectaer. Using the given loads load conditions and analytical mothe"ds, seresses were cal- .culated at critical sections of the rack and compared to acceptance criteria referenced in ASME Section !!! subsection IIT. Compre,esive stability was i calculated per AI51 code or light sage structures. i ( De loads in the three orthogonal directions were considered to be acting simultaneously and were combined using the SRSS method sugested in Regulatory Guide 1.92. De load's due to the OBE event are approximately 90% of those ~ ,1 4 q j --w .y ,p . ~....... -.

_i.rc i. h h ~ g&& .o g.g>?+ Mr2YcJy) . - m e.t. d.11 able..tres.,evels,or. g ere,,,e,,ig. therefore making the est sweat the limiting lead sendition except for stability.

uhpre SSE acceptance ariteria of $72 of critical bucklins strength ta

' ting. Wnder fuel drop leading tendition. the acceptance criterion is that, althouth g deformation may occur. E,gg must remain e0.95. The rock is designed such that.. should the drop of a fuel assembly damage the tubes and dislodge a plate of poison material, the K,gg is stiti e0.95 as required. The effect of the gap between the fuel and the storage tube i.a* been taken into account on a local effect basis. Dynamic response analysis shows that che fuel cortacts the tube over a large portion of its length, thus pr venting an overloaded conditten of both fuel and tube. i f

,[

i Vertical impact load of the fuel onto its rest has been considere:. conservatively as being slowly applied without any benefit for strain rate effects. E l l l l i i

( k,h,( J$.1 setal-IITDgAUI.1C 9331018 . :M g::e-The bish density. fuel storage rack is desisned to provide suffletent astural l esavection eeelant flew to remove 64,000 Stu/hr/ bundle of decay heat. l The fuel bundle rests la the storage opsco with the lower tieplate 'estendtag l through,the hele la the support plate. The water needed to cool each fuel bundle flees b= estural streulatten up through the lever tieplate and the i handle. Ceraer siete are provided la the suppcrt plat.e ta permit aseling of ( slee anauter gap between a channeled fuel bundle and storage tube wall. / 6 The outer and central. tube revs are supported on the base by the fittiass. ~ which have large coel' ant helas is all four sides.' These holes, as well as 'tbe rossining rows of tubes, are ope's to the water planum formed between the 'a A I l i l. I .__L... ~ .hase plate anJ the fuel supports. This plenwt has four large openings in the g l j l base pl. ate, permittint w. iter ftev'frue the support structure below. ~ J...... --.--. *.- The ' support structure must be designed to provtu an ade'guate flow' rite to . prevent viter" reachin,t excessive temperatures (212*T). The flew rate is dependent oc the decay heat 10.:d. the at lesses throudh the structure a w: the losses threuth the rac'k and busJ1s. '"-~ ~ [ $f,3 g i Tigur and Table re supplied to the es1Af ty to allow proper stai ~ ~~ ~ g ~~t"~~l of the flow area required la the support structure. ~ e n ' - i n =- - u.. 4 - {- - - : : _, r _. _ -- -- n -- - - g a ~" to the storage pool, the bundle decay heat la removed by recirculation flev to an outside heat exchanger se that a favorable peel temperature ase be - ( M ,8 1 L I

~ .h *G'"~' (*; l? ...n [ ..i .e (3 is i na q a st = g u g .2 g is{

8..

..t I I ts j ( ,.I. u ( ii j i i i ,o e as u u u g-Plow vsLocity tanovan susets wust C 4.r-3 Figure cubic Equation Coefficien versus Flow Yelocity Through Sundle and Water Temperature Increase Across Sundle C i-. ( 8 .*M E.8 9+ m --. g , que ..p-.. -e ,--e .,_.,,-,_--------s-- ,.,--,,----,,nw---

5 -- - -- -- M-- _ 9!P (- &y_. l8 f asistained. Although the destga poet emit-temperature (to heat eschanger) is . far below seiling, the e'eelaat toegerature within the reek eem1d reach the 1 IIulk beillas tempers'ture if the naturally induced bundle flow is met ensegh (due t,ea b'igh flow resistance) to carry away the decay heat Benerated [tythespentfuel. It is necessary to evaluate the rate of the naturally 'etreuteted flev to determine the maximum rack exit temperar. ore. The parameters which will affect the water flow through tne high-density fuel storage rock and ceasequently the water temperature emiting the top of the storage space are (1) hole s'pe through the fitting; (2) ' flow area through the base plateg i (3) flow resistance through the bundles l (4) height of the module alove the pool linerg I (5) support structure restriction to horisontal flow under module; a vi (6) leading pattern of fuel ta pool (e.g., fresh fuel loaded La center of stray would result la higher cooling water exit temperatures. The analysis was performed with the bunda s flow channel la place, since th14 is the most restrictive (all bundle coeling flow mus: ent'er through the levet tieplateer1{1ce). Also, beat is generated in the water space between the channel and the tube by Samma capture la the water and met 1, thus creating a seed for seditional flow e,wnings into this space. Beat generation rates for l the BWR h ad. Arradiated 44 cud /Ne and cooled seven days were calculated using the ORICtX computer code. These rates ares ,_e. Sundle 63.000 stu/hr 5-gr Channel 752 Stu/hr 5 0 Space 2.510 Stu/hr 2 stainless Doral Tube 256 Stu/br (. I i .b W s e e o - e * * ** eme.e q> eop. gem e

o-O' ~ E ( 2a ao case does the coettag water emit temperature at the top of the rack . approach'bellias. With exit water at 11S*F and the peel return water tempera-sure at 100*F. the c1' adding temperature will be 122'T and the seral tube center-line temperature will be 10$'F. N.. ors torivenein Te.,,erstur. roer..se aeross the sundle '- 4,12IA,\\ These factors are listed la previous sections. The magnitude of the effects of each of these facters la discussed below. The fellowing rel.:tionships, sach relating te one of'the facters are used to selve the cooling water temperature increase as it flows upward through the bundle. The driving force to generate flow through the bundle is given by: IMt + 1 + D,f 'b ~ 1 h,MQ h, 1 20 u 6)8e p (144 e This force is egual to i*.e various pressure drops the water encourters in getting up through the bundle, or Y !N, wkre IN is the sua of these b g g pressure drops givra below: 2 .,. 2.,q,2) fy f ndle ead o.s, 1 2' Roles in Castings: R, = + 2kb 4 e I l N .mrss. i . gemeesp*W 9W8'M N '*** =, e. .e=.-- e me em a-.-

O ( K Igl +l pI'b~ f f 42 'br = Base Plates . L'i i .5 4.e ( i E f a + b)fLa EI l V* Area Weder Modules N = a abs (A ) b 1 Reduction of area under module due to support structure 2 a 1 b)J V IA 2 n (A j b 7 g m g g l,1 li E

I~

N,, = 23 c Definition of the above terms la given in Subsecti 4.4.2 The factor a, . the ratio of flow rates to the previous quadrants and the flow rate to the k quadrant in question. A quadrant in this case is one-fourth of the bundles in a module which is 169/4 = 42 bundles for a 13x13 module. A quadrant of a module is used, since the support structures essentially divids the module fato four equal areas. Assume the sedule quadrant in question is four quad-rants from the edge of the pool array. The cooling water to this quadrant' must flow horizontally und.ar the four other module quadrants and supply cooling water to these modules. If the heat load in tech quadrant is equal, then the flow to the outer quadraet is five times the flow to the quadrant la que, scion m 5. As we move eleser to the guade.ut in.;uestion, a, becomee 5, 4, ( and a3 3, 2 sad finally, 1. I I

Thus, is the sun of the five pressurs loss factors given above. These i

J relationships may be summed up as a cubic equation having the following form: ,..=. ( j ! ( l 9 - ~ * * *

9

'***--.y ,e, e e e ..g-. sr

m - NF* f o t . - D* i - 1 ?.' y = 0, since there are no V, terms; l I Nt + f + e i a = h, 1-I **d b a g. 2C o' y (1444 e Fr,r a given geonstry of fuel, talet water temperature and heat from the bundle = and S will be constants and 9 is the only coefficient that changes. Thus, under the conditions above, defining 4 will set the value of T. The cubic was h g solved for a se les of arbitrary values for 4 and the results plotted in Figure ret's in; to this plot and knowing the value fcr 8 permits rapid determination of Y, and the temperature increase across the badle. .t erfu*f-i is prennted for the case where the module is supported 4 in. above the floor and the support structures occupy 25* of the orar. under the module. Using the reistionships above and the other factorr. as defined, the amount each factor cotdributes to e is as follows: Factor 1 sundle Mead less 0.257 Base plate. 0,0137

9.0127 Notes la Castings Area Under Module 0.0188

,), Reduction of Area under Module 0.0513 Total 0.354 A { e

~ = = 1.12" x 10*I $ = -4.06 x 10'3 .g ~ and the equation to be solved is } 0.354Vf+1.129x10'3'b - 4.64 x 10*3 = 0 . l"I Referring to Figureb for a soefficient of 0.354, v = 0.231 f t/see and b the temperature increase of the cooltag water is 12.3'r. It can be noted that the miniana temperature increase will be determined when 0 = 0.257, which ta the pressure drop through the bundle alone, and this increase will be 11.1*F. The effects of changtag the design parameters can be quickly deternised veing the above relationships. The results obtained will be conservative due to the high bundle heat loads assumed nd assumptions made as to sedule loca-( tion is the pool. Design of the storage tube is the module, the support castings, the support plate, and the base plate la fined. However, details of the module support structure will probably vary between facilities. For convenience, the cosfficients are tabulated la Tablehfor various module heights above the floor and for reductione of this area {'due to the presence of 4-1 a. l I e-9 s !l l ~

7 y' ~ ~..... ( ) Table @ 8 CCPIC IQL'ATION COEFFICIENTS FOR VARIOUS NEICHTS OF !cDCLE A30VE THE FLDOR Als VARIOCS REDLTT10rs OF ARIA SE1 VEIN 3 No0CLE AS FLOOR L i Cobic Equation Coefficient rea e action Abov loor Area L'nder (in.) .L9.1 2E_ E E _ Module 0.0185 3.73 x 10~3 16 0.0135 0.0154 0.0362 9.86 x 10~3 10 0.0265 0.0301 0.0513 1.88 x 10-2 8 0.100 4.37 x 10-2 l 6 0.0736 0.0836 i 1 jA [x 100. r where area reduction = 3 If it is assumed that the module is 10 in. above the floor, with a 20% area ( reduction in the area between module and floor and with all other parameters as given above, then the temperature increase is ds emined as follows: 9 i Area Under Module 0.00986 Area Reduction Under Module 0.0301 Base Faste 6.0137 i Boles in Castin8s 0.0127 Sundle 0.257 Total 0 0.323 'f.14 From Figure Yb = l'.238 ft/see and*tenperature increase is 11.8'F. When PS.e dest;;r. results in values that fall between the data given in Table an interpolation can be performed to get the correct value. 9-/ t. ^ e e 94 mgSDg. m "N W- --

=i--,

-o o e =w %, m. .y

we ge o r $41 o YS tem of Terme

A '

Projected free area n'erest to horisontal flew beneath module, exclusive rE sf stru:aure, retteed for one quadrant of module (sn.2). -. s { ~ Area normal to horisontal flev beneath the module, facteding structure,- A ratteed for one quadt.at of module (in.2), Flow area through base plata/42 (in.2, 3 Ag Holearcathroughfitting=f(D,) 0.4 in.2, A2 Flow area through bundles = 15.353 in.2, 4 Arbitrary area used in bandle frictian correlation = 10 in.. Width of t.maule perpendieu4:e to direction of flow Ut). a k natie flew at.upport point a s. fiev at modute go.drant refereme24; a, for this calculation a = 4 and a, = 1, 2, 2, 4. b 5.isht of module above pool floor (ft) C Orifice coefficient = 0.41 C, spec.fic heat of water = 1.0 scullb *T. f Fractmen factor = 0.0085. t Cravitatior.a1 constant *32.2 ft/sec, 3 ~ Naad loss through bundle $t I 0). E 2 b s, se.d loss through h 1.s in fittings (ft n 8) 2 3-i g. l [ N-g. y OhN- .gg,gqgg

- - - - "B8 h+

{F ee ao. e e, e

O

@"4+W4@4e 6 - e-

eeg, og 4 w

,c-- 4- - - --,m, ,,,---ns.

}. 7 l i i. _ _ _ _. l t. i s. 1 l l i 1 i .:u_ i l i T' i t l I .( e s h.. r r..y. k_(, .'j,, s 'l l L 1 e 4 e ) ,. s-l 1 i j w-- Need less through ac a la base plate (ft 5 0). p 2

W I'*d I.oss through area under module (f t N 0).

h 2 m. W,7 Need less through area esader module restrictions (ft N 0). 2 h, Effective depth of sold water ever entrance point into bundle = 13.5 ft in thie an g le. K Need loss coefficient due to bends = 0.45. b K Nu/ less due ta area contraction = 0.11. c 3 1 Intercept la a versus a correlation = 63.45 lb/f t. 3 M Slope of a versus t correlation = -0.0145 lb/fc

T.

w... e Density of water = 62.00 lb/ft3 (at 100*T). e q Neat evolution rate from bundle = 68 Stu/sec. t Inlet water temperature (100'F). Y I'I'"I'T.of water through bundle (ft/sec). b f g t t e t y_.. t 3 l .1.. e 1 e. ..._._.s...... M-t i i .... _... _ _.. 4 i l i l i 8 ,i I-. 8 s 5 1 I g -] H, i i n I i r i i i. l i 3 i .__.1 _... _ - _ _. + _ _. - - -.... _.. -.. _ t i i ,.. -.. m.

a...

- ~ ~ ~ ~ - * - - -

l l j l j 4'!P' 0 ?, I i i i i 4 I i g l I i 24 i i i. s I l l 1 i k i f ., k.h n'd W- __ .....f k.. e t l 1 + 1 I 1 i (_..... I ..__y .p_.M. y_.,g.. _ g,.; .w 64* 'Ag ,_ Y 4 as h.c....

g..

All structural material used la the fabrication of the MDFS$ is *4pqm 22-scainless stee This mater $d was chosen due to its corrosion resistance and its abilit to be formed and welded with consistent quality. Boral plates. ~~~ ' " ~ ~ ~ ~ used as a neutron absorber, are sa integral nonstructural part of the basic ~ ~~ ~' fuel storage tube. These plater are sandwiched between the inner and outer ' ~ ~ ~ ~

{

wall of the stor. de tube and are est subject to dislocation or removal. deliberate or inadverseat. The inner and outer walls of the storage tube are [ welded together at each end, thereby isolating the Boral plates from direce -- - contact with spent fuel pool water.[t the norial pool water ohrating ~ ~ temperatures of 60 to 150*F, there is no significant deterioration or corro- -- h sion of stainlass steel or Boral. ..m_.m._ .m. .._-_.e. . =. = - - l I j i i i ) l i. l T. l i I t t 4 l I l W .==- l I. I i l i ._ m ~ - e i l I ._.l.- -. _. i e .r I I ,p t. 1 g i i .g .- p . __. g -l T i _.. 7 ~ l l ..___..q ___ + 4-- .j__ _ t_.. _...i s i l I n l j 1, I l l - -- I I 8 l I f ? g j_ _.--..q.._....._....--.i.-g-I } a i I ~ ~ - ~ ~ ' " ~ ~ * - ~ - n., r. w

y g l l i I l I 3 7 l I l 3 l 8 I I l i e t i 1 %huf g- @A j Aet w g t. l g j j i ~ .b Presence of the neutron absorber material in the fabricated fuel storage module will be verified by visual, examination and dine.nsional inspcetion. The - I tube design allows visual verification that material exists itt the space pre- - 8 ..l vided for placemenc of the neutron absorber material. The thickness of the Boral plates is nonstandard, providing a statistical significance between the

thickness of the Borsi and commercially produced aluminum or steel sheets.

therefore enabling confirmation of Boral presence by dimensional inspection. In addition, use of non-Boral plates in fabrication of fuel storage tubes will - cause tube deformation. Deformed tubes will not be accepted by inspection or by fabrication fixtures used for assembly of a fuel storage module. Dimen- .__._, _., sional reinspection at neutron absorber plate locations can be performed at L,,_, _ _,,_ _ the pool site. These data would be compared with the dimensional results obtained during fabrication of the individual tubes and of the module assembly. - q A visual reinspection of each Boral plate location can also be performed. __.; Acceptance of the above inspections will ensure that Boral plates are contained - ___.,,, in the fuel storage module sufficient to maintain the neutron multipihation __... factor at or less than 0.95 with a 95% confidence level. -s ,,M 'the storage tube and the integral neutron absorber .... ___._-____ material are permanently marked with identification traccable to __._ _ the material certifications. The fuel storage tube assembly con- .taining the neutron absorber material is compatible with the I b environment; of treated water and provides a design life of 40 r - T l years, including allowances for corrosion. i i .I i I p_ - . 1!. t .-. [- a '.l .I l L=. -i L s.mosion data and industrial experience confim that aluminum and Boral have ' acceptable corrosion-resistant properties for the proposed applicacion.2' ~ .y-- - J~ ~ ~ , J_ l i 1 n_'_ l L / =_ - g pft,c,Qi" -~~

~ l l l I l ! 68!P" i i i i-2B. i. e i. 1 ( t t L i 8 6 1 l \\ I n ~, t

n. 7 T. w~c. MA~ l_.w~~s

,.. ' r. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ - - - - ~ ~ i I 8 1 t t ,t l _. _e__- _.... -.-.. q ) l i .g i _. 9,D.lkI-** a=n= ~ M .. %,h.....Q es% 971 5, Input Ekt.itatisa The iiigh density fuel. storage rack was analysed using the CE standard 3ER/6 ~ ' ~ ~ seismic response spectra. The spectra were datived by detemining the dynaalc - ~-~ j i I i s. -- a. '....... _ = 7.. e t 3 i, i i -( 1._.. + i i t I I i .A.._.... t t

4. _.'

l i t .r-- ....s... ...._....a t 4 t a .i e t' w. _. = s. l .-4 l l 7 i l l l s 6 !_-.--.--e-l 3 l i e i .I i i H' .i. 1 t 1 "T. .s.. .. _ _.. - = -,4

3. -.--..-p........'~~-

.u _.. - = f, l l i l p... I. e w. 6.-. ... -. = * - I p l g 9. ..-..t il e ,,,.1 l l l 1 I I \\ l -[. O h l l_g i 5 i i i i

q .l ~ ' response of the GE etandard plant ever a wide range of soil coditiens with a 2.33 ASE excitatten. This implie; that the rack is designed fur alw worst site condition, asd considerably highar seismic margin is achievable for more .. optimal sites; esmesesMt. ' M.i.." Mode 11r.g kne (1) Norizontat Yarious pool arrangunents were analysed to determine the worst-case hydro-dynamic mass effect. The case chosen is that of ef.ght racks in a re-angular pool with e 14-in. spacing to the wall and 2-in. spacing between the modu *es. 7%is arrangement was modeled as two lumped mass cantilever columns as shov= in FigurehhN~rf Kodes 1 through 12 represent the group of eight racks and Nodes 15 through 25 represent the pool wall to account for int raction effects. Mode 12 represents the joint of the tube and fitting in the rsek. s (2) Y_ertical .I 5.cause of the ench...J.c natural frequency ('e33 Es) in the vartical direction. the vertical reaponse force was determined statically and no special modeltag is necessary. 8 a WNWW hi-7~ Analysis The tots! asas matrix of each rack for the analysis is equal to its structural mass satrix plus the hydrodynaale asce matrix. Conservative structural damping values of Regulator / Cuide 1.61 are used without any added danping due to fluid

==i-- effects. The VATER-01 cesputer program (CE C.rapany Froprietary) was used to l determine the hydrodynami: mass of one rectanguisy body inside another rectangu-lar body. . :.~.- The'noverning dpsaic equation for a water-filled rectangular container subject to ground excisation.iar (Ns + Mh)lx) + (c)lAl + (a) {ul = -Ofs + Mh)!y) l l ) ...... -.. -. -. ~

- - ee e

,,,wm, ,,,e ,w-- ,-,w----------,,--,n--- ,r---

o ft""Y Oi t (. 9 i. ( l g E, w

G q p q pl 983 %

.Y g 1-i g g O O I 4 >t. 4 la 18833 m O l .e - SK73 m t i1 q >4 I e O g t i1. 4 i. 1.?A8 A M O, O I A I i O O,,, I e. i e O I t i21 4 p.

==.7.17 A I ( O O, 2 'O O I d >. t ise sta.m i e O I < >3. t itt -sa w A @ i=, x,,..V.. 3., .f << <n

y Q

s. _ _. m. e.. u. 9,t-4 1 .c ,g Figure MSEA Computer Model

f G

-gr a e l i I ' ' ' ~, .=ye. e===w e.. - ---- e. .s

=

~ there '..s .. h.,*/. ( 30s = a dingunal matria represenstag temped structural asas Ist = a mondiagonal hydrodynamic' asse matria which causes eeupled .t.- setten ameeg the racks and between the rocks and the pool well; j P i ..T.- ( ~i ~ a and k = dasping ' nd stiffnees astrices, r6spectirely, of' the system; a

a. I. a = the acceleration, velocity and' displacement vectors of the system l

relative to the support acties; i y = t' e rock base support acceleration excitation. a The response spectnan analysis method la the DTSEA computer program is used to calculate the response forces due to the OBE er SSE horizontal accelera-( tica. The horizontal response spectra used is shows la Figur 4-6 for the g frequen y range of 1 to 30 Es. g j.,3 u esta=aas {, e*E s

eae s

3 l s 1 1 l 28 s ( ~ 1 I m. 8 g i t t t _ t I t t t t e g a w is as se ao as s a a e s. sasousescr eese 1 '.} pigure -4 [. ~ Worisoots1' Spectrum (CE Company Proprietary) .g g - - ~ ~ e

-r g l's)e ~5[ l l , s =. 4 '( Tare:f" Nsults er

Wstas the taput and methods described above, the maxima response forces at the tube-to-fitting connection for each rock are as shown below l

t: Vertical Besping M Noment Shear Force l ' twent (I) (in.-16) (1b) (1b) 7 5 5 esE 2 1.83 x 10 1.62 x 10 1.90 x 10 I 5 5 SSE '4 1.96 a 10 1.79 x 10 2.30 a 10 Tne first natural frequency of the rack with base fixed la 12.8 Rs. 5.. ~ Y: 9,I.I,I,7 ~ "'Ac'^"'**" , Q / g*f g *-4rt:1 Vertteal tacact Analysis I

l, Ycrtical impact analyais is required because the fuel assembly is held ta (i

the storage rack by its own might wf %ut any mechanical holddown devices. I. Tasrefore, when the deviarard acceleration of the storage rack exceeds 1.03 (. (g = gravitational acceleration), contact between the fuel assembly and the storage rack is lost. Impact escurs when fuel-assembly / storage-rack contact la re-established. Although it is very unlikely for most utility plaats to have storage rack vertical seismic accelerations exceeding 1.0. such 3 3=qe acceleratima values can occur fee certata wafavorable combinations of sitas and building designal thus, the need for fertical impact analysis. ^ '.1.0 Input Excitation The taput encitation consists of several representative acceleration time histories at the refueltas pool flocr. These time histories are generated from using actual site and plant structure acdels. The time histories have been normalised to 0.15' OgE and 0.30g SSE for those plaats uhose seismic require-S ~ ments are less than 0.15g DBE and O'.30s SSE. "For those plants whose seismic I seguirements are more than 0.15g Ost and 0.30s s$r. the time histories were not reduced. The time histories are then time scaled and amplitude scaled , -... ~... w ,...-,.m. .-,.---,_,-----g

ouch that a breed spectrum is obtained. The above procedure has been adopted ( in etter to enceapass all known plaats where the potential for utilising the ,high density storage rack *autess. he enveloped spectra of the time histories are shown in Figure g,ph i l 1,;.i.F Analy' sis Method ( - e'a-- fepset analysis requires a somlinest systes andel, the time history direct integration method is used to determine the systes response. The. integration scheme uses a central difference procedure. The integration time step sise is selected snell enough to ensure seaveyance to the correct solution. e t ~ OBEAesat 3 I E:: E 3I s g l, I g I J f f I f I' I I 3 3 e S s to SS to 4 ss FnEQUt8sCT teal a l i ' f Figure 7 Tertical spectrum (CE Company Proprietary). I (

9. l-0

~ -_-,y

,e y W n ('- M. ,.t-3. ( . M Hede11ag Per the versteal impact analysis of the feel bundle, three types of elements ~ are used: Q) Imped mass gap element to represeas the fuel assembly, ) (2) the variable water mass element se represent the fuel rack support plate and fuel interaction effects; and (3) the linear spring-damper al m at to represent the top flange of the I-beam in subfloor suoport dystem, which consists of a series of fabricated I-beams arrarged la a rectangular array. The webs and bottom flanges of tte I-beams are calculated to be extremely stiff fo the vertical direction and are represented by the same rigid base as the fuel storage pool floor. l The hydrodynamic effects result in a lowered natural fregaency of the fuel (. asstably and are accounted for in the analysis. The pet entage of critical desping ta taken to 73 for gst condittoa and at for ogf conditions. =1 . C.1.4 Design Adeguacy Evaluation Asseing that all of the fuel bundles stored la a fully filled high density rack will vibrate in-phase with behavior similar to the response of a single fuel bundle, the dynamic design loads of the high denalty fuel rack with 169 fuel bundles were calculated and used for destgr. structural odeguacy {' evaluaties. . 4,3v8 Norizontal Impact Analysis yy - Borizontal impact analysis is required be.:ause a clearance t.tists between tht fuel assembly and the high density fuel storste rack tube walls. 'It is e l l

~ PL*V apected that the fuel sasamblies have equal probability of belas la see, tact with any one of the four sides. Dus, the tapact forces are espected se be equally likely on all four surfaces of the fuel rack and fuel ansembly. De randos nature of the sorisontafiapacts thus precludes significant gross " leading of the fuel storage rack. hus, only localised deformations of the fuel storage rack cells need to be cessidered. { .dc3:2st laput Excitation For horisontal impact analysis, the fuel assembly is modeled as a series of parallel beans representing the channel and the fuel rods stToorted by spacers. De local flexibility of the fuel storage rack cells is determined through a standard structural analysis using the SAF-4 computer code. s ha hydrodynaal: effect.s between the fuel assembly and the storage rack are accounted for by a fluid mass etenent. A lumped-mass gap element is used between the fuel assemble and the fuel storage rock. D e lumped-mass gap element is also used between the fuel rods and the channel to si:sulate the clearance between the fuel rods and the channel. D e upper and lower tieplates o, are assumed to be attached to ths channel due to the tight fit between the tieplater. and the channel. He overall bending stiffness of the fuel channel am sapresented by linear spring damper elements. gecause of the relatively I high gross bending stiffness of the fuel rack to the horisental direction. tne t :k 1: considered fixed as far as gross bending is concert.ad. Adequacy Evaluation . :rontal impact response determined using the model and analysis method

r...

( a..... ? z' ova is used to determine the dynaste leads on the fuel ra.:k cella. De l'ocal loading on the individual storage rack cells is used for desigu er ..-t -at adeq.nacy ev,1uation. w m# a me

tl t =~ m r.m ga TusAcA '"v'Y Tr.

'm rvv W v Facilities Description $54'wir F0FA W )

.'l 9.1. 2Mg (1) The spent fuel storage racks provide storage in the containment and spent fuel pools for spent fuel received from the reactor vessel during the refueling operation. 4 The spent fuel storage racks are top entry racks designed 3i to maintain the spent fuel while precluding the possibil-ity of criticality under normal and abnormal conditions. The upper tieplate of the fuel element rests against the The lower tieplate sits rack to provide lateral support. in the bottom of the rack, which supports the weight of the fuel. The rack arrangement is designed to prevent accidental (2) insertion of fuel assemblies or bundles between adjacent The storage rack is designed to provide modules. accessibility to the fuel bail for grappling purposes. Nominal fuel spacin; from center to center is 6.56 inches by 6.56 inches. The location of the spent fuel pool and the containment (3) pool within the complex is shown in Section 1.2. l

s 410.25 In Section 9.1.3.2 of your FSAR, you describe the chemistry of the (9.1.3) water with regard to its compatibility with the-alwiinum storace racks. GS Revise this section of your FSAR to be consf stent w"tn your new hf gh - ~ density stainless steel racks described in Section 9.1. f your FSAR. 4b26 RES # - TW GESSAL TcEXT HA5 N Co m 7eo To Delete The Re.peece To Awmiuvm pace.sLsce Arramo). k)E HAUC EELEcce.o Nctr 70 C+iAN65 00a WATEG-CHEWtbTAy RGtpvia.eme as. l l

bbbbAM 11 22R700W 238 NUCLEAR ISLAND R5v. O e 9.1.3.2 System Description (Continued) drained from the inclined transfer tube during downward fuel trans-fer, as well as the volume of water above the skimmer weirs, which drains from the pools following a temporary loss of circulation. Clarity and purity of the pool water are maintained by a combinr.- tion of filtering and ion exchange. The filter-demineralizers maintain total dissolved heavy element content (Cu,'Ni, Fe, Hg, etc.) at 0.1 ppm or less with a pH range of 6.0 to 7.5 for com-patibility with - - fuel storage racks and other equipment. l4to.25 Each filter unit in the filter-demineralizer subsystem has adequate capacity to maintain the desired purity level of the pools under normal operating conditions. The flow rate is designed to be approximately that required for two complete water changes per day for the fuel transfer and storage pools. The maximum system flow rate is twice that needed to maintain the specified water quality. Water may be returned to condensate storage after being filtered and demineralized. The FPCCU System is designed to remove suspended or dissolved impur-ities from the following sources: (1) dust or other airborne particles; (2) surface dirt dislodged from equipment immersed in the pool; g3) crud and fission products emanating from the reactor during refueling; l (4) debris from inspection or disposal operations; and (5) residual cleaning chemicals or flush water. l 9.1-20 l ../

b ae-m== - -- 410.29 Provide the same information for the fuel handling system as is . _ sv.6.4) requested in Question 410.17 for the leak detection system since your FSAR is not consistent with the Perry FSAR. b a Respomse ( N$}. ch Me %,_a,, hQ _Q % ew as a Pwm -ed kdismO ,o w SS3 s car 33 W A 0-3 yA rA+.s r ek 4 -R,- ,1 7 m3 4kees p % 4 - L d a s k % ) d. ~ 4, ,,e t A h -t a. l 7 A evo m2), .6

.m-ATTACHMENT NO. 5 DRAFT RESPONSES TO POWER SYSTEMS BRANCH QUESTIONS i l l 1 1

l .I 430.27 Provide the specified operating voltage range of the Class LE l (8.3.2) de loads. Provide the maximum equalizing charge voltages for the Class 1E batteries and the de system minimum discharge voltage at the end of the two hour design discharge. Provide the ra'.ing of the Division 3 battery charger and indicata the nisaber of ce?ls in each Class IE battery. State whether the Division 3 battery charger will be affected by the voltage sag which occurs when the HPCS pump is started on the diesel-generator. s get dfMR2 7 des /len/ g. 3. 3. /. / f Sy. 8 3- /2 The number of cells in each patte bank (either Class 1E or non-Class 1$w) isMnoi11sffer/eee o'/ V/Clow /' L and4 e $s 60 c 't v r' ~** l the non-se The operating voltage range for Division 3 (NPCS) Class 1E de loads is 112.5V to 137.5V with 125V de nominal voltage. The maximum equalizing charge voltage for Division 3 (HPCS) 125Vdc battery is 137.4 volts. Voltage at the end of two-hour design discharge will be provided by the applicant. Division 3 battery charger is rated for 240/480V AC input with 132 volts (nominal), 100 amps de output. Division 3 de battery has g 60 cells. i The charger is also capable of automatically regulating output voltage within 11/2% of its rated value at any load between 0 and 100%, with the ac power feeding the charger deviating from the rated voltage by 110%. Thus thi Division 3 battery charger will not be affected by the voltage sag which occurs when the HPCS pump is started on the DG. The 125V DC battery will be able to maintain the bus voltage. All 6t loads connected on the division 3 dc bus are rated for operation inthevoltagerangeof112.5Vto1$.5V. S W e o m e e e O

h l ROcctor Fivo indGpandCnt 125 VDC sy0tems cro providcd to cupp y propriato. Island normal and emergency DC powar for czch unit as ap The fifth Four of the five 125 VDC systems are Class lE power. system supplies non-Class lE power. 7., i n' emergency The DC power systems provide adequate power for stat o ll modes of auxiliaries and for control and switching during a operation. The operating voltage range of Class lE de loads is 110V to 140v. +'s 1 The maximum equalizing charge voltage for Class 1E batteries is 140Vdc. The de system minishm discharge voltage at the end of the two hour discharge period is 1.83vdc per ce The 125 VDC systems provide a reliable control and switching power source for the Class lE systems. All batteries are sized so that required loads will not exceed 80% of nameplate rating, or warranted capacity at end-of-installed-( life with 100% design demand. Each 125 VDC battery is provided with two chargers, each of which is capable of recharging its battery from a discharged state to a fully charged state while handling the normal, steady-state DC load. Battery sizes are specified as: (1) Battery E, Division 1 - 1950 A-hr at 8-hr rate; 2080A for 1 min l l l (2) Battery F, Division 2 - 1500 A-hr at 8-hr rr.se: 1620A for 1 min (3) Battery G, Division 3 (EPCS) 400 A-hr at 8-hr rate: 500A for 1 min l (4) Battery H, Division 4 - 425 A-hr at 8-hr rate: 550A for 1 min (5) Raetarv .T nn"A4"deiaa*1 '*:^ ' ' - - ' ' " * - - - - - ' -

af 3 >g A 52l'

: E i s

41 is e r 2.. A - :ss a 5 - si-zn I EGs S 5 .y 1

i[

y=J li!} -j' a-- j y

i!!

y=='

lij

--lll 4s 4i p

jig +< j:!g = {! illi!ii i l!!il:ii l!!$i y! ai 9,1 lin a 'Ili i lig .I Y-s i .3'! 3l ll ija'il 'ili i liu a i liin 1 nn i i .lli!il j$r l,ti,, il $= 111,i p j--j , lli i i j fik ill! f-E' ill! i= i'!! lllj \\ 'si 141 is' '~ t;l !3-li's -1,91 1 til eci i 12,isees i i ol i 1 t_S_! i- ~ ija i A 'j;. li'I - llli = i l 1i j i i i g . =! =_! .li 1

qi

!.n I qj hf E[$ 'h,'jhj". IGA f~ IS[Sj',l s! 4 ..ll{lil =n-- = #' !&l5~~".'!jjjj j gijj 1 d l-llja' a l i,, .,1;7ili L.,,!..ylili-i m idi il.i=!" ili11' iiill 'll iii n = r lirl! W !!!!I. i. ii. =. e lii fb e:! I ll I il n ii;' bdl 'dld.! E a -l i:: g=

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= i.

ji i

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430.41 Diesel-generators with a high degree of reliability are an essential (8.3) part of the safety systems for nuclear power plants. Accordingly, provide a discussion of the level of training which will be required for the applicant's personnel to ensure that diesel-generator reliability levels inherent in your nuclear island will be maintained. ( As applicable, state your recomendations for the types of personnel to be trained; i.e., operators, maintenance crew, quality assurance personnel and supervisors. In your discussion, identify the amount and kind of training you recommend for each of the above categories and the type of ongoing training program you recomunend to assure optimum availability of the diesel-generators, Discuss the level of education and einfaum experience requirements you recomumend be met for the various categories of operations and maintenance personnel associated with the emergency diesel-generators. RESPowse as7a m +s.4L,s u e ~y % Aeph,ciI,6~. %J\\Ve g rovtw b A+ +Le ebw c a ct -+Le el t a..sel - $4w ev ameo - cL A.ms % dov ov v wsil su 8p h +o 4L4. A g

i 430.54 Demonstrate that the control room and the remote shutdown pandl (9.5.3) illumination levels under emergency conditions are in conformance with the applicable sections of NUREG-0700. 1 roLwb r i S. po d a -k - Ac n, m a. xAl W N 7 I g e- -y ,w. ,-y- ~~.,_.e,- -,--,,e-

I 430.61 You show on Figures 9.5-10 and 9.5-11 of your FSAR, the day tank (9.5.4) vents terminating somewhat outside the diesel-generator room. However, it is not clear from Figures 9.5-10 and 9.5-11 nor from Figures 1.2-18 through 1.2-22 of your FSAR, exactly where the Divisions 1, 2, and 3 day tank vents terminate. Accordingly, provide additional infor4ation on these vents. Show vent the terminations on appropriate i views in Figures 1.2-18 through 1.22 of your FSAR and provide detafis of the terminations which show that they are protected from tornados, floods and the effects of severe, weather conditions. i kJ pch SAE. l Division 1 and 2 day tank vents terminate 6-inches beyond building wall, These terminations are protected by the roof over i into the dock area. Division 3 vent the dock area and by the wall around the doc % area. j i terminates on a 45 degree down sicpe at the outside surface of the Bird screens cover all three Diesel Generator Building wall. l All three terminations will be locat9d on draving K-036 terminations. ,/ (Figure 1.2-19) and the attached Section H-H added to drawing X-037 (Figure 1.2-20). l m l ~ e I i l l

su"as C P' CRAUN C CD R W Christiancen General Electric GESSAR { Project 6382 - OUND 1 QUESTIONS September 29, 1982 San Jose f: o u e.w r ic a 4. s o.6 r \\ D i ' h*I~ 1 RocP Et..Ulko*9 _,,_,__ t <p( E.22-Cool-IS Pl.A*A P A5:'A'rSTu st, 1 4s'si. sow M. '/ 4 g.os c. &:.:..g 23.. d4 '- t o' v' i \\s hlDU MMO TaguI4&Ih5 As 'M s. t g/ pout >H weiei c i,rqN ) W A L L. A N n C o w u_ s r 0 m gg I E s. Ws TH Ri@ $6.Rt:(.N 2 E'J. M E Lr1 W.Te1. W e'hiA. htAS 0AkvANittD ".'aTEEs. i 3 J Jy--- ~I. _'SECTION H-H i o e e g. e e. j E

P Identify all high and moderate-energy lines and systems which will Discuss the measures which 430.62 be installed in the diesel-generator room. (9.5.4) will be taken in the design of the diesel-generators to protect (9.5.5) the safety-related systems, piping and ctmponents from a postulated (9.5.6) Our concern is failure of either a high or moderate-energy line. (? t ' (9.5.8) the availability of the diesel-generators when needed. M5p sh54._ Theee are ne h r'a h -ensecr u finss in s'k dia.s sl~ a n.nsee.+ae ran ma. c a s. s <.chian s. t..

r dieu /

The. modaea}s. ~ enneau lins.s are. in n sk ask tube oil sfae +,de a4. fsis.Io ll. s ye.nF s ss a.n sie./ ss.edae. wa.he and 4h z outomen+ desak sas4sms. AII'; h. %h. mads. ede c en Wnes & on/a ss.saals+2 win m.is.

nLs ceL+;vs d i v i s r a n,s.

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im, (_ ct The rrfiniinven o, v a n.( s -la, of fuel do be, sheed 'E-1o es dies./ .u ne <a he :im /i Lt. Aaseel l on 7 daqs ,su'pp ig. 6.f'.rn ax >'m e<h spie.ific. 4'u s.1 *o i I .c o n s a l p }.toit see, 9, s, q,1, I (q e. The diesel ILiel oil aualb sfanscem whl:,, / { mod be m t+ I: th c. A s trn s ceelfic.= /:in \\ \\ D 91s for Dlc.s e/ fue/ oi/s. t 2. c/i:=c./ -ru < /, \\ wi+h no + lc.s.s +/,nn 35 t e f o n t. numhc \\ Tht. o bas <, meef +h t, etou l,a m e n 4 o 5 2c.n Cuidr /./61, l \\ I N C, The 60P fuel of f -frans{te pump minimun - m-c.opac.iiy ch *.Il be., no1 Ie..s1 thoo t'ne. l sne cl[iN fus.l oii co n s v m p /s'.>n af +h c. d it.,sel e n slnt ahle), d.:> t 1 na + exct t el '2 O,St Gr o nd.s of fu n./ o<e ne f haru n>.~te w si.. hos e.,sa.. The minimum di.s csces A u s-Shall bl twf /c 51 ff,cn /L psi c p, or,i / v 2 r- ~ ~ ' ')n d 2T Pits he Ow 3. Cbm pli.-ne e i: by ihc /.)1-331ican+,

GUESTION 430.78 (9.5.5) Provide a detailed discussion of how the diesel generator cooling water systems functions in the steney mode to maintain Jacket we.ter tempera-tures above ambient temperatures to enhance the diesel engine start capability. Your discussion should address how the jacket water is heated, how heated water is circulated through the diesel engines and the design jacket water temperature at the anticipated ambient temperatures of the diesel-generator teams. Identify any excess capacity in the jacketwaterheatingsystem. The operation of the Division 3 diesel generator cooling water system during standby requires additional discussion since *here is an apparent lack of heated jacket water under forced circulation in this mode.

RESPONSE

Division 3 diesel-generator cooling water system is _Cd to maintain the engine in a warm staney condition in accordance with the quick start reliability requirements. The specific details of the system functions to achieve this will be provided by the applicant depending on the type of the keepwarm system furnished for a particular engine. Dur s n..,e, -th t. s+ondbl rnac/c o f th e. diese f - yn < ro lor k j a c.b.T Wa1ie g.,ceec L eL. 0 rn a i ' + s o'n o bon o m br'e n l- + t m pe n hoe s y mans soa h e~ h a Jec he D1/ I oF a n ele.c4rlc _ s e.k t-a~l 2. on el on ele e.frie iLn,e e nin h< ai en fan DW 3. Th e he.a4 act w o.fer is

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g' C F SRAUN O CO , e' ~ ,p_.christinnaan w General Electric GESSAR Proj ct 6382-P P. ROUND 1 QUESTIONS September 29, 1982 (- _ San Jose QUESTION / RESPONSE 430.93 (9.5.5) QUESTION 430.93 7 vide enlarged and more detailed plan and elevation views of the l Division 3 Diesel Generator Air Start System Air Compressors. Show the intake, the exhaust, the cooling system and the fuel supply for the diesel engine-driven compressor. Incorporate these enlarged views into the appropriate drawings in Section 1.2 of your FSAR. RESPONSE 430.93 The diesel engine driven air compressor is an air-cooled type and requires no cooling water. The fuel supply is provided by a tank locally mounted on the air comprest:or base. The air intake is through a filter mounted on the compressor head. The diesel engine exhaust is piped to the Diesel Generator Building stack. The attached detail will bc added to drawing K-035 (Figure 1.2-18). 0 9 C o S I j e e L

O

r7s g F CRAUN O CD so 33 R W Christiansen General Electric GESSAR Project 6382-: ROUND 1 QUESTIONS September 29, 196; ( Q ocsTros 4-33.9.3 D \\ Fuct TAwk ~ AtR .,- r iwy4x e. MIM -)- FTM , ENGiN E-c m ust -:i.si. ',--, . E, To stack. e m. ! ATR To I I; .fR ECE tVERS . x l. ,3 l . _$. b . !n t ( e %../ l l l l _. -- -4 _ - _- %.=_l-_-t 0_ L ~ DETAIL i STARTN G AIR CCMPREGSOR E 2'2 - S 001-9 . Aie-cocLED DtESEL ENGINE - ORWEN k., O 1

-() 430.101 In Section 9.5.7.4 of your FSAR, you refer to alarms for low oil (9.5.7) pressure, high oil temperature and low oil level. However, none of these alarms are shown on Figure 9.5-16. Further, you show these alarms on Figure 9.5-17 in addition to a low oil tenperature'alarn, a lube oil high temperature and a high pressure alarm associated v.itn a relief valve and an extra lube oil low pressure alarm. None of i these alarms are described in the text of your FSAR. Revise Figures 9.5-16 and 9.5-17 to agree with the text antt/or revise the text to agree with Figures 9.5-16 and 9.5-17. R.c.sp E R ~" Ih WO r De.e. i[r'c ci fee H,c. Div Ionel 2 dksel-ornera k luhe oil

m. / c <r, 5.:t Mc onnunu*c &

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QUESTION 430.105 (9.5.7) One of-the recommendations in MUREG/CR-0660 is for prelubrication of the diessi engines prior to starting, thereby minimizing wear due to a lack of adequate lubrication at the time of starting. The keepware circuit shown on Figure 9.5-16 provides continuous prelubrication to the Divi-sions 1 and 2 diesel angines, except for the turbochargers and the uppe-part of the diesel engine. Show that this lack of prelubrication does not impair diesel engine operation or reliability. If the Divisions 1 and 2 diesel engines will be manufactured by DeLaval, t% vise your lubrication system P&I diagrams to show vendor modifications to provide drip lubrication to the turbocharqsr thrust bearings. State whether vendor modifications to the governor lube oil circuits have beer, or will be, incorporated. If the Division 3 diesel generator is manu-factured by E2, show that the recommendations of MI-9644 have been incbrporated. (Refer to Item (c) of Question 430.110.) i RES*0NSE The implementatioa of MI-9644 recommendation to be answered by the app 11 cant. Division 3 diesel generator has a continuously operating soakback pump which provides lubrication to the turbo-charger parts in the standby condition. Di vi.rio n /d c dimsp./ emines \\ %.fh e s/em N co di nuou.sh resrur 12.ec{ f re fubt i ca hsn Sy'Nh f i ec/ fiec/ un Aves &>e Vencdr con d'erneono? AY 13 S i& is not reyvirec/ c*Oh'a& O m SW"'D 9 90 -t ha ines possth/6 /ve dteMs ink thb cy /in er&'cou/c/ j i so me e ti 9lnMLL t-h e f4e en n e on ci subsquen t- \\ resulf J n detmage , Sh;r& Con fI r inaho n cf co ./i6nce fo }he j CnY /3 fbe //ccry;( r U / i~C" l l l l i l t i i l MP:csc/Il0146-39 11/4/32 i I

430.111 Revise Figure 9.5-10 of your FSAR, to show the conplete combustion air (9.5.7) intake and exhaust systems. Alternative'y, provide a new P&1 diagran sho, sing these s.ystems, including..all three civisions. Snow s11 instrumentation and controls associated.<itn tne systens. n The. combath*on o,k inta ke onc/ evh.w.s ? = v.s /cm: shown ih Fifure. 9 5-10 comp /c /c.. 4// th: lc..in,cn f n lion nn d c on dro /.s c:.s ce a c h d wrYn 4hc sys /em e ha/ cre no/ .:iro on art. Ly _ooplic.cnf, A d,VTa ez o A'c. / pri s.s v ee. c?=agt.

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( 43C.112 Describe the instrumentation, controls, senscrs and alams provided (9.5.8) in the design of the diesel engine combustion air intake and exhaust system which alert the operator when parameters exceed ranges recon-mended by the engine manufacturer and describe any operator action required during alars conditions to prevent harmful effects to tne ( diesel engine. Discuss systems interlocks provided. __R_e3 pow K The. Div I ad 2. aIinNIe f(Hre.s acc. pro v;c'...] wi+h a diffe.c e nfio./ nees.sucu a a ve< a cro s t inciver!Vl w.noyn m&'whvsi (www at er, n e c ch fiNee Ac clihano/ ins 1%m e n la Mn conko /.s s ene ers, anot a /cem.c movions'h &e de sion of ff:c diesc./ onade, combuslion t Otn /0/0 &c on el L Vha v.sl .5Wsltm QtL h A p/ico'nCas vs.qvies d. u d,es ';,,z a.

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t 430.115 In Section 9,6.8.3 of your FSAR, you briefly discuss the effects of (9.5.8) decreases ir, barometric pressure on diesel engine performance. Expand this discutsion to be more specific as to the effect of decreasing barometri'.: pressure. State the maximum tornado-induced pressure change, in unitr, of psi per second, the diesel engines can withstand without signif'.cantly affecting performance. State the minimum barometric i pressures (in. of Hg regulating from a hurricane) at which the diesel engines can operate for: (1) up to one hour; and (2) for extended periods without degrading output or causing engine problems. In your response, discuss the three diesel-generators. R4 3 pohS E h 00 l'L on f ws ll lbL l2 " G, !Y htO V C/L. G. I'n O tte 5pcclllC. Nl3c.t)ssion On N1L e ffe cf-Of de cre.c:in'q beroen e frie, pec.ss ue t, on die sr./ ensint &lf'forn3 CMC L

1} was spt cifit.d Hmt M t, d5.s e /- e osint.

Shou /cl br O bl<- fo S fo ncl C. mc es a cula'Ja t' YOffindQ. /bhunol Pet.S$ vrt, Chanat. Of 2 FS .be e sceci>d for' 'DW / an o! 1 dase/~ ninerafse. m r ~ -..w..r. m.s1c. u..,+., m, v1 ,,,s,,,' 4 c--3 3 es t,. see rs uc. sc esion 3 3. 2.. e e A' d & oALS1 Abe ~ Io irlani-t& ah6Lifa o1 1oon 1kc5 a l b a u l d. . % - 1,, 6b de 1Le>51 1 k m o,ifi u a M & % e ' r>, A w o h a m(4<, H ol% 'kel a wakU. 6 ~ l 6 l

430.117 Snow by analysis that a potential fire in the Division 2 and Division 3 (9.5.8) diesel-generator building occurring with a coincident single f ailure of the fire protection system, wi41 not degrade the quality of the diesel co :bustion air, thereby permitting the remaining diesel-generatcr to provide its full rated power. A3 ~3c'.s S t_ i i Po+1nNa/ f,,c in +h e 6:vi.s on 2 an d.s di e seI - 9enaeadce bus /dha w i// no + c/eacoc?e i %c quaJ1% oC +h e dit..s J c.om 6u.s H.>n o,h % eb.,. .a m ; + h *n s h s. re.ma adso e diuJ-en eea+w~ fo eravido 14.s 4h//e +<.J wuer 5 fo llaarn's ts.aws.s i Ea J, Divis*an 2 i lno'spenabn f hm and .3 Da Mr Ido:.,e s c e s. i,sa la.,e.J fem eai o Mu~ an d sai// be. MJ/1 e -Ec a ch> u~ in c asr aF o er.aren a. a f fr,c e si 1 ' & s.e ha e'Idiny. Yht la e.a. N.a n ,S tex cab in/ats for a+ a s e m,l,I< .s o u t. 1 +h & i C.AJA oh $s? L sA O1 Ss bsA / /Sn % lh L am a.<n,h s h ok eifat.< wi// 4+ Lt. aPhukl, See ris v, c 1., ir lo ca 4' n of osv i r : ooh t ri +a k.e. an ol Fs's 1 2. 1 9 fue Oo's 3 l l

ATTACHMENT NO. 6 = 9-DRAFT RESPONSES TO PROCEDURES AND TEST REVIEW BRANCH QUESTIONS

ATTACHMENT NO. 6 c, s DRAFT RESPONSES TO PROCEDURES MD TEST REVIEW BRANCH QUESTIONS l

64O.07 1 1. Emer64ncy response information system (ERIS) tests. u . h._-...R.A-sow 5-t. a-2 -Yk h> hce .. FSA12.. u t.1 do.o c yi k L_ NA G9.- 3 kb3 .) I 2 l' l. O 4 i i______ m. Reactor water sampling system tests. Verify that the test v.ill l be adequate to verify flow paths, holdup times and procedures. ~ ( 8 .R.n.s p aw sQ. _. ga.c 4%M _ 4e sk.r _( __ TM o S d.s d o n 1 4. '2.. L'2 r 3 1. Preoperational testing to determine expansion, vibration, and dynamics n. effects for: (1) ASMI Code C1 css 1, 2, and 3 systems; (2) other high-energy piping systems inside seismic Category I structures; (3) high-energy portions of systems whose failure could reduce the functioning cf any seismic Category I plant feature to an u.ucceptable level; and (4) seismic Category I portions of moderate-energy piping systems located outside containment. c ._._.R.a s r> m s-r 6

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ML20070H328

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