ML20040C761

Forwards Advance Info to Be Included in Next FSAR Amend. Includes Text Changes on Preoperational & Startup Tests

ML20040C761
Person / Time
Site:	Byron, Braidwood, 05000000
Issue date:	01/06/1982
From:	Tramm T COMMONWEALTH EDISON CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
NUDOCS 8201290211
Download: ML20040C761 (74)

Text

{{#Wiki_filter:Commonwealth Edison [ On'a First Nation;l Pl'2a. Chicago, Illinois ) Address Reply to: Post Office Box 76) 'N ,' Chicago. Illinois 60690 January 6, 1982 co r p Mr. Harold R. Denton, Director RECEIVED 't g Of fice of Nuclear Reactor Regulation g ggjgg g [ U.S. Nuclea r Regulatory Commission g,..ggy g. y n.. jgt Wa shi ng' t on, DC 20555 V tum nce a /' @A' r.ac

Subject:

Byron Station Units 1 and 2 ( Braidwood Station Units 1 and 2 t-Advance FSAR Information NRC Docket Nos. 50-454/455/456/457

Dear Mr. Denton:

This is to provide advance copies of information which will be included in the Byron /Braidwood FSAR in the next amendment. Attachment A to this letter lists the information enclosed. One (1) signed original and fif ty-nine (59) copies of this letter are provided. Fi f teen (15) copies of the enclosures are included for your review and approval. Please address further questions to this o f fice. Very truly yours, Y0 f4" T. R. Tramm Nuclea r Licensing Administrator Pressurized Water Reactors Attachment N // $5E'E8Ed'o?88i:L PDR

ATTACHMENT A LIST OF ENCLOSED INFORMATION I. FSAR QUESTION RESPONSES New: 022.30 Revised: 010.50 022.50 022.79 022.71 040.108 022.73 040.176 040.89 371.5 241.6 423.23 241.7 423.33 423.24 423.34 423.30 423.42 423.37 (partial) II. PSAR TEXT CHANGES ppg. 7.2-33, 34, 7.3-63, 65 S.G. Water level pre-op tests: pg. 14.2-25 DC Power pg. 14.2-26 Vital Bus Independence Verification pg. 14.2-34 Fuel Pool and Cleanup System pg. 14.2-62 Primary Safety and Relief Valves Startup test: pg.'14.2-88 Power Ascension Section 11.4 changes (VRS) Revised pages for Table 6.2-58 III. Miscellaneous Items ICSB Agenda Item 52 i ETSB Open Item 1 CPB Open IN m 5 (Revised Response) i l

i B/B-FSAR 4 4 ) CPB ITEM #5 (Revised Response) On-Site Fuel Inspection Program On-site inspection of fuel assemblies, control rods, etc. will be described in the detailed written plant procedures. Surveillance of fuel and reactor performance is routinely conducted on Westinghousa reactors. Power distribution is monitored using the excore fixed and incore movable detectors. Coolant activity and chemistry are followed which permits early detection of any fuel clad defects. A routine fuel inspection program shall be established to provide information on irradiated and discharged fuel at each refueling. The program will primarily involve visual examinations of selected assemblies. Typically, visual examinations will be made on lead burnup and special test assemblies, as well as 5 to 10 percent of the fuel dis-charged. Visual observations should include, but not be limited to, crud buildup, rod bowing, grid strap conditions, missing components, etc. Additional fuel inspections should be performed depending on the results of operatiohal monitoring, including coolant activity, and the visual fuel inspections. Os This information will be added to subsection 4.2.4.5. I i h G

= B/B i' t. d ICSB Agenda Item 52 Discuss the plans and schedule for complying with Regulatory Guide 1.97, Revision 2. Describe the conformance of the present design. Summary of Response The applicant reiterated that design changes are presently being developed to meet the requirements of Regulatory Guide 1.97, Revision 2. When the design is complete, it will be sub.aitted to the NRC staf f for review. Post-accident monitoring instrumentation will be installed to satisfy Regulatory Guide 1.97, Revision 2, or justification will be provided for any alternatives. s O i l l i l l O

9 .e B/B e ETSB Open Item #1 (6.5.1.2) A memo and calculation follow, s l l l t I l

SARGENT &LUNDY 3/Ob ~ INTER OFFICE MEMORANDUM ~ o k from W. B.__Easc 1 x3876 Date Janua ry__5.,__19.8 2 __ d Projee No._01149244 68 3/ 8.4.D 0 Dept /Div._ECChanical/RVAC Spee. No. File No. Page No. Client... commonwea 1 th _Ed i no n c m _ Sen, nyron/_Braidwona Unit 1 K. _2_ Subject Pffluent Tr ea_tm en t_S ys t ems __Branc h_Q u e s tion s_Rais e d in the_SER, To: _E_. R. Crass - (1/ 3 ) - 3J K. J. Green - (1/1) - 22 cc: W. C. Cleff - (1/1) 22 S. N. Planjery - (1/1) - 31 In response to the Evaluation and Findings, Section 6.5.1.2 of the SER, we have completed our analysis of the relative humidity upstream of the charcoal adsorbers contained in the filter plenums serving the non-accessibic areas of the Auxiliary Building. The results can be su c.arized as follows: (m For design extreme summer, design extreme winter, and intermediate weather conditions the maximum relative humidity of the exhaust air upstream of the non-accessible area charcoal adsorber has been calculated as being less than 25% (22%, 420%, 20%, respectively). A copy _of the calculation is attached for your reference. HBP: jab At tachment - t S }

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• s1

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• ~

= A:tutes en **ain Eteam Isolation 1*J = Act ates en R1r. Tceiwater Isolation TI = Achatso m Cer.tairca nt Egrey Actuation l T2 .= Act.ates cn Contalry.ent Vent. Isolation I 20/21 A = Autt atic (Atr, P ore 111e or Flectrical) Operstion J M = %rual Opration I RM = R-t:,te V.c.ual rieret!on 22 IA = Ir.et r.ent Air I I 23 C=e Tirae 6.2 29 I.* [ 11 MIN. " Valves w!!! be placed as close to the contalement as practical. IE - Alt'cugh the data lietei le or.ly given for 17 nit i valves, the data applies to ? nit 2 valves se well. Tmo ntiel systems are thone eyatms which hay be uoed l'o110 wine a containwrtt, m, isolation etc al. Essentini syntes may be teolated m containamt l. taa'att m stenale an noted in Columi 19, but timf r inolation valves are 9upplied with 1R powar to remit ivrotr3 rnnumi rrcirning if rtvluim!.

E/B-ESAR 1 7.2.2.3.5 Steam Generator Water Level /~('\\j The basic function of the reactor protection circuits associated with low-low steam generator water level is to preserve the steam generator heat sink for removal of long-term residual heat. Should a complete loss of feedwater occur, the reactor would be tripped on low-low steam generator water level. In addition, redundant auxiliary feedwater pumps are provided to supply feedwater in order to maintain residual heat removal af ter trip. This reactor trip acts before the steam generators are dry. This reduces the required capacity, increases the time interval before auxiliary f eedwater pumps are required, and minimizes the thermal transient on the Reactor Coolant System and steam generators. Therefore, a low-low steam generator water level reactor trip circuit is provided for each steam generator to ensure that sufficient initial thermal capacity is available in the steam generator at the start of the transient. Two-out-of-four low-low steam generator water level trip logic ensures a reactor trip if needed even with an independent failure in another channel used for control and when degraded by an additional second postulated random failure. A spurious low signal from the feedwater flow channel being used for control would cause an increase in feedwater flow. The mismatch between steam flow and feedwater flow produced by the spurious signal would actuate alarms to alert the operator of the situation in time for manual correction. If the condition (se) continues, a two-cut-of-four high-high steam generator water l level signal in any loop, independent of the indicated feedwater flow, will cause feedwater isolaticn and trip the turbine. The turbine trip will result in a subsequent reactor trip if power is above the P-7 setpoint. The high-high steam generator water level trip is an equipment protective trip preventing excessive moisture carryover which could damage the turbine blading. In addition, the three element feedwater controller incorporates reset action on the level error signal, such that with expected controller settings a rapid increase or decrease in the flow signal would cause only a small change in level before the l controller would compensate for the level error. A slow change in the feedwater signal would have no effect at all. A spurious low or high steam flow signal would have the same effect as high or low feedwater signal, discussed above. A spurious high eteam generator water level signal from the protection channel used for control will tend to close the feedwater valve. However, before a reactor trip would occur, l two-out-of-four channels for a steam generator would have to l indicate a high water level. A spurious low steam generator water level signal will tend to open the feedwater valve.

Again, before a reactor trip wculd occur, two-cut-of-four channels in a i

loop would have to indicate a low-1cw water level. Any slow (~N drift in the water 1cvel signal will permit the operator to (_) respond to the 1cvel alarms and take corrective action. 7.2-33 l l

B/B-FSAR 4 t Automatic protection is provided in case the spurious high level (T reduces feedwater flow sufficiently to cause low-low level in the ( ) steam generator. Automatic protection is also provided in case the spurious low level signal increases feedwater flow sufficiently to cause high level in the steam generator. A turbine trip and feedwater isolation would occur on two-out-of-four high-high steam generator water level in any loop. ( 7.2.2.4 Additional Postulated Accidents Loss of plant instrument air or loss of component cooling water is discussed in Subsection 7.3. 2. Load rejection and turbine trip are discussed in further detail in Section 7.7. The control interlocks, called rod stops, that are provided to prevent abnoraal power conditions which could result from excessive contrcl rod withdrawal are discussed in Subsection 7.7.1.4.1 and listed in Table 7.7-1. Excessively high power operation (which is prevented by blocking of automatic rod withdrawal), if allowed to continue, might lead to a sa fety limit (as given in the technical specifications) being reached. Before such a limit is reached, protection will be available from the Reactor Trip System. At the power levels of the rod block setpoints, safety limits have not been reached; and therefore these rod withdrawal stops do not come under the scope of safety-related systems, and are considered as control systems. () 7.2.3 Tests and Inspections The Reactor Trip System meets the testing requirements of IEEE Standard 338-1971, as discussed in subsection 7.1. 2.19. The testability of the system is discussed in Subsection 7. 2. 2. 2.3. The initial test intervals are specified in Chapter 16.L. Written test procedures and documentation, conforming to the requirements of IEEE Standard 338-1971, will be available for audit by responsible personnel. Periodic testing complies with requirements for periodic testing of protection system actuation functions as discussed in subsections 7.1.2.13 and 7. 2. 2. 2. 3. t ( l O 7.2-34 l l n

~ B/B-FSAR TABLE 7.3-2 (Cont'd)- NUMBER NUMBER OF OF CHANNELS NUMBER FUNCTIONAL UNIT CHANNELS TO TRIP 3. FEEDWATER LINE ISOLATION a. Safety Injection See Item No. 1 of Table 7.3-1 b. Steam Generator 4/ loop 2/ loop l High-liigh level l i 2/4 on any Steam Generator c. Low Tavg (on P4, see 1/ loop any two Table 7.3-3) loops O S l I (i

-1, i

I L, O I ii 7.3-63 e

---

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l B/B-FSAR d TABLE 7.3-3 (Cont'd) i FUNCTION DESIGNATION INPUT PERFORMED Defeats the manual bypass 2/4 T setpoin$above av of steam dump block-reenergize steamline rate block P-14 2/4 Steam gener-Closes all feedwater con-ator water level trol valves and isolation above setpoint on valves any steam generator Trips all main feedwater pumps which closes the pump discharge valves ? I Actuates turbine trip 'l i O 7.3-65

B/B-FSAR 11.4.3 Volume Reduction System Description 0) ( The volume reduction system (VRS) is composed of a fluidized bed dryer, a dry waste processor, and associated equipment. Figure 11.4-7 indicates the operation of thc VRS. The fluidized bed dryer system processes the concentrates or bottoms from the radwaste evaporators. Table ll.4-2.provides the quantity of concentrations to be processed. The evaporator bottoms containing typically 10 to 25 wt% dissolved salts are stored in the waste liquor storage tanks. When the concentrates are to be processed, they are pumped by a positive displacement, variable-speed pump from the waste liquor storage tanks to the fluid bed dryer nozzle where the slurry is air-atomized and injected into the fluidized bed dryer. The fluidized bed is electrically heated by means of vertically-mounted tubular heaters attached to the outside surface of the chamber. Fluidizing air is supplied by the fluidizing air blower operating in a closed-loop circuit. The air is electrically heated and passed upward through the fluid bed vessel, fluidizing the bed particles. The resultant dry anhydrous product from the fluidized bed dryer is a free-flowing spherical shaped salt (300 to 400 micro-inch diameter) containing the trace quantities of' radioactive material present in the influent liquid waste feed strewn. The granulated, solid waste product is dischargad from the fluidized bed into the product storage hopper. ( ') The overhead air stream from the fluid bed vessel, consisting of hot gases and fine salt particles, passes through a gas / solids separator, wherein the bulk of the fines and ash (from'the dry waste processor) are removed from the gas stream and discharged to the product storage hopper. DAW consisting typically of contaminated wood, paper, cardboard, cloth, rubber, and plastics such as polyethylene ig processed in the dry waste processor. Approximately 3435 ft of this material will be processed annually. Contaminated waste oil is also processed in the dry waste processor. DAW enters the VRS by passing through a metal detector and a chopper-shredder. It is stored in one of two hoppers until a batch is accumulated. The DAW is then pneumatically transferred to the dry waste processor where it is reduced to ash. The dry waste processor is also a fluidized bed system. The fluidizing air carries the ash to the gas / solids separator where it is combined with the fines from the fluidized bed dryer. ) v 11.4-8i

B/B-FSAR The gases leaving the gas / solids separator are directed jg to a venturi scrubber / condenser unit. In this unit, the t I gas containing the fines and ash is first contacted with \\w water in a high-energy venturi scrubber, resulting in removal of virtually all of the remaining particulate. The gas and any remaining particulate then pass through the packed scrubber section where water vapor evaporated in the fluid bed dryer is condensed. The gaseous effluent from the scrubber / condenser goes through an exhaust gas heater, an absolute filter, and then to the blower for recompression and recycling. A small quantity of the recirculating air stream is continuously taken from the system. It is processed by an absolute filter, a charcoal adsorber, and a second absolute filter. It is then released to the auxiliary building vent system for discharge to the atmosphere. The salts resulting from the drying of the evaporator concen-trates and the ash resulting from the DAW and contaminated oil are stored in a product storage hopper. These wastes are metered to a solidification system for immobilization in polymer binder prior to shipment to a burial site. This solidification system, which is located in the radwaste building, is completely independent from the wet waste solid-ification system in the auxiliary building. Table 11.4-3 indicates the interfaces between the VRS and the other plant ( systems. For more detailed information about the VRS, see Topical Report Number AECC-2-P (Topical Report " Radioactive Waste Volume Reduction System," October 15, 1979). J ll.4-8j

.:.\\ 47.h i A p O r U v i l TADLC 11.4-3 PI. ANT INTE14 FACT:5 1{Iy!! SOI.ID RADt!ASTE SYSTI:M ESTIMATED I.INE TEMPERATURE ESTIMATED caoSS stADIOIsoTOPE INTER:'ACC NU:-13ER LINn* ~ ,AN3 DESC3IPTION N3!MSI:R SI"AU PRESSURC (*P) PLOW RATE DATCII SIZE CONCrt.~.*3AT ION * * *

• A 1.

Spent renin OWXX9A 1\\ in. 50 psig '90 120 gpm up to 400 gal 370 pCi/CC 1 2. Evaporatcr concentrates OwX143nA 1% in.

• 35 psig 190 15 gpm 22.5-36. 5 gal"'-

0. 5' isci/CC 3.

Drum processing unit drain Otc '11 AA 3 in. ATMOS 110 256 gpm 25 gal 0.5* pCi/CC 4. Decanting tant overflow .OWX153DA 2 in. ATMOS 90 120 gp6 intermittent 50 Ci/CC 5. Drum processing unit vent OVF29AA 1h in; -2 in. H C 100 20 scfm continuous 0.05 pCi/CC * " 2 6. Decanting tank vont OVF28AA 2 in. -2 in. I! C 90 20 scfm continuous 0.05 a.C i/CC"

• 2 7.

Flush and decontamination OWX14GDA 1 in. '52 psig 110 50 gpm 25 gal 0.16 pCi/CC 9 ?, ? s D 51 lw . Typical 1;r.c numrects, for Unit B line numbers the last letter is D. 1 Directly to.the drum from a rceirculation loop via a motoring pump. Mostly Xc - 131, only when the drumming system is in operation and i l resin is in the decant tank and averaged over 16 hours.

• • • • Based on Table 11.1-11.-

j + eee. e i k

1 i. f') \\ G. .B/B-FSAR 5 Table 11.4-3 (Cont'd) t Interface Number Line Line Pressure Temperature Flow Rate Batch Cross Radioisotope and Description Number Size Size Concentration j

8. Volume Reduction System 1

a. Evaporator concentrates OVR80A 1

35psig 190 30gpa 3500 gal 0.5 uCi/Cc I

b. Waste Oil OVR69A 2

25psig 150 50gpm 140 gal' 1 ow >I

c. Purep Seals f. Dilution i

water OPW60A 2 50psig 120 75gpm .[e Primary water tent

d. Decon water supply OPMK7A 2

135psig 100 100gpm 600 gal primary water

e. Cooling water supply OWSJ4A 4

140psig 110 156 pm continuous 0.0 9

f. Cooling water return OWSJ6A 3

140psig 124 156gpn continuous 0.0

g,g.

g. Service air supply 2SA67A 1%

115psig 120 82sefm continuous 0.0 i

p.

h. Filte. red exhaust OVR084B 4 Atmos 175 477sefmm continuous 5.6 x 10-4 ci/Cc p

i. Drainc OWF69A 3

50psig 150 50gpm intermittent 2.7 x 10-2pci/Cc l, J. Decon water return OVR123A 2 50psig 180 50gpm 150 gal so.05pci/Cc

(,

k. Drumming unit litlet 0987A 8

7.2psig 880 500/hr 550lb. (min. ) 8pci/Cc t ?,4 t N n ?. -a 'J .1 h .,d ..j [...1

a ID/B-FSAR t. () TABLE 11.4-1 (Cont 'd) PROCESSING EQUIPMENT QUANTITY DESIGN CAPACITY MATERIALS Caustic tank 1 1000 gal 304SS Decon tank 1 450 gal 304SS Contaminated oil tank 1 175 gal CS Bed storage and transfer hopper 2 2900 lb 304SS ~ Trash hopper 2 1500 lb CS/Fe Waste liquor storage tank 2 3500 gal 316L-SS Fluid bed dryer 1 0.41 gpm 347SS/Inconel 625 Drs" caste processor 1 83 lb/hr 3475S O Trash conveyor 1 Rubber /CS Trash elevator 1 20 lb/ min CS Waste feed pump 1 58 gph 316L-S,S Waste recire. pump 2 10 gph 316L-SS Decon. pump 1 50 gpm 304SS Dryer feed pump 1 24.8 gph 316L-SS Condensate pump 1 22 gpm 316SS Contaminpted oil pump 1 4.38 gph CS Scrubber preconcentrator recire. pump 1 28 gpm 316L-SS Caustic additive pump 2 20 gpm 304SS Scrubber preconcentrator 1 16.8 gpm 316L-SS/Inconel 625

• ()

Secondary scrubber 1 1142 scfm 316L-SS ll.4-9a 'E= ac.. m -h-o e e ce ** * -*

• B/B-FSAR TABLE 14.2-13 h

DC POWER U~ (Preoparational Test) l l L. Plant Condition or Prerequisite Prior to core load. h biectiva To verify proper operation of the batteries, battery chargers, switchgear and t alama of the 125 vde oysten. Test Sunnry - /s preoperationsi test will be run on the 125 vde syste::t includin;; the batteries, chargera, and distribution centers. The battery capacity t<1ll be verified vich the battery charg r electrically disconnected. Individimi cell. voltage readings will be taken at pariodic intervale during the espacity teat to ensure that individual cell lirits are not exceeded. DC loads uill be operated at a voltage equal to the rnininta neceptnble battery tereinal voltnae. A perfort:rince test will be conducted on the battery cherxars to . verify its voltage regulation. The test will verify the proper eettings of the low volt. ige slem, high de output voltage trip of ac input brewer, s: bui unt'.ervcitage alarm, ground dm.tector alam, and breth: trip O

• fail alam, alntns. The battery ch;rger will be cepabic of charging its a nocista!

battery within twenty-four houra while supplying pow r to expected ster-dy state, nomal plant operating loads. Acceptance Critoels ?.sch battery can carry its design load for a specified tice. Inter 1ceks, relay 3, r.eters, and other de ccaponent.a function :ss designed. Inverters 9111 be teated in the instenment pWer tent. l O G 14.2-25 p 4 s.Q s @CP

• • <O

& ben 22_ i q__m _ m _ __ n. g Ng}on om e nm m, e. . & -Czo _e e

B/B-FSAR TABLE 14.2-14 VITAL BUS 7t:DEPEt:D3NCE VERIPICATION O' (Preoperational Teot) Pinnt_ Condition or Prerequisite Prior to plant operation. Test ebjective_ To verify the existence of independence among redundant onsite.pouve sources and their load groups. Tent Suecary The plant vital bused vill be independently powered fecn each of the three - possible power nources. Each losd group (Sat'oguards Train) will be tested with the other load grcup co:pletely disconnected. Each test will include in,jection of sinulated accident signale, startup of the diesel generater and load group under test, sequencing of loais, and r.he functionni perfor:ance of the loads. Each test will be of sufficient duration to achieve etable or.arsting conditions. During esch test the d-c and the onsite a-c buses and related loade.not under teot will be monitored to verify absence of voltsge nt theae busca and Icade. ( Acceptsoce Criterla EdCh redundJat ondite power Source and itB load group cHu function Without dependence upon any other redundant load group. W 9 I a 14.2.-26 '3-

n/B FSAR AH2NDM3NT 23 qQ* 'DHCCMBER 1981 TABLE 14.2-22 FijEL POOL C00LIMG AUD CL1WTJP SYSTF.ti (Preoperational Test) Plant Cond_1 tion or Prerequisite Prior to core lond. Test Obfoetive. To ensure the proper operation of all equipment, controla, interlocka and nierms associated with the Puel Pool Cooling and Cleoning System. To ensure the operation of the Skimter Loop. To perfore a Spent Fuel tool lera innpsc-tion using the leak detection sight ;;1anaes and measure any leskage detected. To deitonstrate flow through the Spent Fuel Fool Demineralizers and Heat Ex-changer 1 cops. Test Sunnsry Tests will be perforced to verify flow through the Spent Fuel Fool Demineral- .izera, Heat Exchanger loops plus the loop feca the Refueling Weer Storssa Tank z . through the Spent Fuel Pool Deminatalicers and Filters and bs::k to the Ref:el-ing Water Storage Tanko., The.line fecm the Refueling Vater Storage tank to t"f the Fuel Cask Fill will aleo be tested. The Spent Fuel Fool will be filind using the Refueling Water Stors3e Touk and the Refueling Uater Purification Pumps. Alarm setpointo uill he checked and valvoo, instruments and controls tested. The leakage detection system util be used to m2asure any leakage. de tecteil. The anti-alphon device will be tested using n lou point drain on the Rent Exchangers. Acceptance criterin Demonstrate the ability of the tystem to fill, circulate and drain the Spin. i Fuel Podl. Dee.anstrete the Skirt. iter loop operates to reenvc debris f ree the surf sco of the Spent Fuel Fool at a c.inicara of 100 spr. thenush clean filters. l The high and low level clares operate. Ths high tettparature alarm operat.u. I Verify the anti-alphon featura on tha return line. Verify the capability te l cireninta sater tetuacn the Refueling WAtor Storage Tanks and the Spant Fuel Pool Decinerali::ers, i 14.2-34 u 7' -u .m,-,...-,...,,.,.. m..

B/B-FSAR O TABLE 14.2-50 PRIMARY SAFETY AND RELIEF VALVES (Preoperational Testi Plant Condition or Prerequisite Prior to core load. Test Objective To verify the setpoints and measure the seat leakage of the pressurizer safety valves. To verify proper actuation and operation of the Power Operated Relief Valves. Test Summary Setpoint verification and seat leakage will be determined in-plant for the pressurizer safety valves. Three specific features of the Power Operated Relief Valve p/ Logic will be tested. They are the Auto-Low Temperature g, Control, Valve Response time and Enable Functions f rom the pressurizer pressure transmitters. Acceptance Criteria The valve setpoints are in accordance with Technical Specifications, Chapter 16.0. O 14.2-62

B/B-FSAR TABLE 14.2-76 P01(ER ASCEtOSIDi{ (Startup Test) V) Plant _ Condition or Prerequiteites 5 Core load through 100% pouer operation. Test Objec_ttva .'To conduct lou power phystrs tasting,and testing at various power level plateaus while increning pow e to 100%. Te s t s eu' a ry Following initial critics 1 tty low power physics testing will be accomplished. At approxiestely 30, 50, 75, and 100% reactor power the unit will. be held for other systote testing. Startup Tenta Conducted et Specific Fouar Platos_ue Tdhle fjo._. Tit 1_e_of Startup Test Powe r I. eve'.s 14.2-62 Initial Core Loud O% 14.2-63 Control Rod Driven 0% 14.2-64 Rod Position Indicators 0% 14.2-65 Reactor Trip Circutt 0% 14.2-66 Rod Drop fleasurem2nta 0% , O 14.2-67 Incore P1m: Monitor System 0% ' v '/ 14.2-6B liuclea-Ipstriu.entation ' 0%,30%,50%,75?.,10'H 14.2-69 Reactor ecolsnt System Pressure 0% 14.2-70 Reactor Coolent System Flow 0% 14.2-71 Pressuriter Effectiveneas 0% 14.2-72 trater Chemistry 0% 14.2-73 Radiation Surveys 50%,1 W.~ 14,2-74 Effluant Padtstion Monitor 0%,50%,1CM 14.2-75 Initini Critienitty 0% 14.2-77 Moderator Temperature Reactivity Coefficient Hessurem nt Os 14.2-78 Control Eod Eeoetivity 1.' orth Mr.ssurorrent 0% 14.2'-70 fioron Pesetivity 1' orth He.twrmnt 0% i 14.2-80 Flux Distribution Haasurements 5% 14.2-81 Pseudo Rod gjection 0%,3 0!.' ( 14.2-82 Po.:sr Renettv1ty Coeffictent 30%,5 0%,7 57.10.'*; 14.2-83 Coro Perfurcance Evaluation 30%,50%,75.,1.:2 t 14.2-84 Pius Any natry Evelustion 30%,503 14.2-85 Turbine Trip 100% 14.2-86 finutdown from outside the Contcol Room 10'. 14.2-87 Loss of Offatte Powar 10% 14.2-S8 10% Losd Saing 30%,75%,'.0 Z 14.2-89 50% Lrad Ser;uence 100% 14.2-90 RTD Croas Cs11bration 0% p _Accepto.. Crit erI4 t The plunt la tahe'n_ to approximately 10D% pt.wcr end all startup test cc pleted, reviewd and approved. ~ l'4.2-88

B/D-FSAR QUESTION 010.50' "four response to 0010.33 concerning the effects of flooding resulting from a failure of the circulating water system transport barrier is incomplete. You have not provided an adequate response to items (4) & (5) of 0010.33. Our concern is for the consequences of a major circulating water system leak in the turbine building caused by failures of such non-seismic Category I components as the main water headers or expansion joints to the condenser coupled with failures of their corresponding butterfly isolation valves. The potential exists to flood the turbine building basement to the water level elevation of the cooling tower basin (Byron) or the cooling pond (Braidwood) by simple gravity draining from these large reservoirs. " Describe the designs and locations with the aid of drawings, if necessary, of the watertight barriers provided to prevent floodwater leakage from the turbine building to the auxiliary building or any other safety-related enclosure. Include a discussion of the consideration given to passageways, pipe chases and/or cableways joining the flooded space to space containing safety-related system components. As an example, discuss the means of preventing floodwater from entering the main steam tunnel and eventual 3y reaching (~} the auxiliary building at its termination with the main sj steam tunnel near the safety valve room. Include in the discussion water exiting the turbine building al er above m grade level and entering other safety-related enclosures through watertight barriers removed for maintenance."

RESPONSE

In the event of a circulating water line break which cannot be isolated, the turbine building could theoretically be flooded to grade level at Byron and to five feet below grade At Braidwood. Damage to turbine building equipment will not prevent safe shutdown of the plant because no essential equipment is located in the turbine building. The auxiliary building is completely watertight below grade at the turbine builoi.-e auxiliary building interface except for the main steam tunnel. Watertight closures prevent flooding of.the main steam tunnel from affecting the auxiliary feedwater tunnel, the containment, or any other auxiliary building areas. AV 0010.50-1

B/B-FSAR O The only safety-related items which will be affected by turbine building flooding are the main steam isolation valves (MSIV's). With a loss of power, the MSIV's will fail as is. The steam lines are automatically isolated on high containment pressure or low steamline pressure signals. If the event which damages the circulating water piping also causes a significant break in a main steam line, the resulting decrease in pressure will cause MSIV closure prior to MSIV inoperability. In the event that the turbine system remains intact following a circulating water piping failure, the turbine will be tripped and the turbine stop valves will isolate the steam system. Failure of the MSIV's will not have an adverse affect. The only lines which would remain open are the gland steani lines and the steam jet air ejectors. These are relatively small lines and would not drain the steam generator. One fully severed circulating water pipe would provide [\\ sufficient flow into the turbine building to flood the MSIV's N/ in about 10 minutes. This is not realistic because a guillo-tine break of this pipe would require an eight foot lateral motion of the pipe. A large break in the circulating water system would quickly be evident to operating personnel and action would be taken to secure the main steam system. However, as discussed above, the only concern in this case is the possibility of gross failure of the main steam piping in conjunction with the circulating water pipe f ailure and this event results in MSIV closure prior to flooding. If the main steam system is intact, the MSIV's may fail open without impact on plant safety. O .0010.50-2

B/B-FSAR h OUESTION 022.30 J " Penetrations P2, 3, 7, 9, 14, 15, 5, 6, 8, 10, 22, 25, and 48 have been listed in FSAR Table 6.2-58 as having met the requirements of GDC 57. However, these penetrations service systems inside containment that include components which are not Seismic Category I and/or not Safety Class 2. There-fore, in accordance with the provisions of SRP Section 6.2.4.ll.9.c and d, the systems inside containment associated with these penetrations are not acceptable as closed systems and cannot be considered as one of the isolation barriers. Provide information demonstrating that the design provisions for these penetrations meet the requirement for two acceptable isolation barriers in series."

RESPONSE

Penetrations P-2 and P-3 have been sealed closed because the station heating is no longer utilized in the containment. Penetrations P-7, P-9, P-14, and P-15 are the inlet and outlet penetrations for the essential service water system. These pene-trations meet the isolation requirements of General Design Crite-rion 57 for closed systems. The essential service water penetra- /~N, tion piping and isolation valves are seismic Category I Safety (,) Class 2. The essential service water cooling coils and associated piping inside containment are Seismic Category I Safety Class 3. Safety Class 3 design provides adequate assurance that the system will remain closed to reactor coolant and the containment atmos-phere. For this system, there is no difference in allowable stresses or design and fabrication procedures between Safety Class 2 and Safety Class 3. The only substantial difference g" t e and restricts the nondestructive testing to redtograpMy~Ehile _) ,."T is the Safety Class 2 requires more detailed QA documentation ' R,C Safety Class 3 will allosFPea;trent cr-::; net 4c testing,ys alter-natives._3 The safety classification ob the Reactor Containment s { Fan Coolers is discussed in the response to Question 022.73. f,q.,, g Penetrations P-5, P-6, P-8, and P-10 serve the chilled water N is 'fD'Act,q system. This is a Category II Grade D system and, as such, not considered a closed system. Penetrations P-6 and P-10 currently *(e-meet GDC 56 requirements with an automatic isolation valve outside containment and a check valve inside containment. Table 6.2-58 has been updated to include these valves. Penetrations P-5 and P-8 will meet GDC 56 requirements with the addition of a second automatic isolation valve inside contalment. This addi-tional valve will be either a fail closed hydraulically actuated valve or a motor operated valve powered from a different emergency power division than the existing isolation valve. (~s ( ) Penetrations P-22, P-25, and P-48 serve the component cooling j system. This system has been upgraded. The piping, valves, i 022.30-1 l

B/B-FSAR i O gers are nokJSeismic Category I Safety Class B. and both the tube and shell side of the excess letdown heat exchan-FSAR Sections 3.2 and 9.3 will be updated to reflect this revision to the safety class of this equipment. As a closed system, these penetrations 4 j meet the requirements of GDC 57 and are listed as such in Table i 6.2-58. l l i l l O 1 i l 4 i 1 O Q22.30-2 .,... -., ~.., - - -..,

m.. v B/B-FSAR ) QUESTION 022.50 "NUREG-0737, Item II.E.4.2 pertains to containment isolation dependability. Describe specifically how each paragraph of this NUREG-0737 item is satisfied."

RESPONSE

Position 1) Containment isolation system designs shall comply with the recommendations of Standard Review Plan Section 6.2.4 (i.e., that there be diversity in the parameters sensed for the initiation of contain-ment isolation). The following parametets are monitored for the initiation of containment isolation: Automatic Safety Injection Containment Pressure Steamline Pressure Pressurizer Pressure. Position 2) All plant personnel shall give careful consideration to the definition of essential and nonessential [~h systems, identify each system determined to be non- \\m l essential, describe the basis for selection of each essential system, modify their containment isolation designs accordingly, and report the results of the reevaluation to the NRC. All systems penetrating the containment were designed to the requirements of General Design Criteria 54, 55, 56, and 57. Essential systems have been defined in Table 6.2-58. Position 3) All nonessential systems shall be automatically isolated by the containment isolation signal. All systems not required for hot shutdown are automatically isolated by the containment isola-i tion signal. Position 4) The design of control systems for automatic contain-5 ment isolation valves shall be such that resetting l l the isolation signal will not result in the automatic reopening of containment isolation valves. Reopening of containment isolation valves shall require delib-erate operation action. l (, The individual control circuits are designed i (, to prevent automatic loss of containment isolation due to the resetting of the isolation signals. I l 022.50-1

B/B-FSAR P The containment setpoint pressure that initiates O osition 5) containment isolation for nonessential penetrations must be reduced to the minimum compatible with normal operating conditions. The containment isolation setpoint pressure is 5 psig. This value is used in all analyses of the capability of the containment to withstand and contain the results of postulated line breaks. Operating plant experience indicates that use of this setpoint pressure will not result in unnecessary isolation signals. Analytical results show that the containment pressure and offsite releases will stay well below limits and that safety systems will work properly with this setpoint. Position 6) Containment purge valves that do not satisfy the operability criteria set forth in Branch Technical Position 6-4 or the Staff Interim Position of October 23, 1979 must be sealed closed as defined in SRP 6.2.4, Item II.3.f during operational conditions 1, 2, 3, and 4. Furthermore, these valves must be verified to be closed at least every 31 days. (A copy of the Staff Interim Position is enclosed as.) The containment purge valvec are closed whenever the reactor is not in the cold shutdown or refuel-ing mode. These valves will be put under adminis-trative control per ANSI N271-1976. See response to Question 22.54. Position 7) Containment purge and vent isolation valves must close on a high radiation signal. A high radiation signal, separate from the con-tainment isolation signal, will close the contain-ment purge and vent isolation valves. See response to Question 22.55. l O. (_) 022.50-2 i

B/B-FSAR O. QUESTION 022.71 "The following valves are listed in FSAR Table 6.2-58 as failing to the 'as is' position upon loss of power. Justify why the valves fail 'as is' rather than fail closed (reference SRP Section 6.2.4.11.56. Penetration Valve System 28 1CV8100 RCP S.eal Water Return Line 28 1CV8112 RCP S,eal Water Return Line 71 1CV8105 Charging Line 5,8 IWOO20A,B Chilled Water Lines 6,10 lWo006A,B Chilled Water Lines 21 1CC9414 Component Cooling Water Line 21 1CC9416 Component Cooling Water Line 24 2CC685 Component Cooling Water Line 24 1CC9438 Component Cooling Water Line 25 1CC9413B Component Cooling Water Line"

RESPONSE

Revised Table 6.2-98 indicates that the above valves fail in the "as is" position. The "as is" position for these valves is the safe position. The response to 0022.30 describes the revision to be made to penetrations 5 and 8 for meeting GDC 56 requirements. Based on the changes to these penetrations, failure in the "as is" position for Valves lWp020A, B is acceptable. j i I i [h (./ Q22.71-1 l 1

B/B-FSAR QUESTION 022.73 "FSAR Table 3.2-1 indicates that quality measures equivalent in intent to those in Quality Group C will be applied to the reactor containment fan coolers. It is our position that the reactor containment fan coolers must be designed, fabricated, erected and tested to Quality Group B standards, as recommended by Regulatory Guide 1.26 (SRP Section 6.2.2.11.66. Provide information on how you will comply with this position."

RESPONSE

The reactor containment fan cooler (RCFC's4 coils are AS.ME S.e.c t i o n I I I, Class 3 components. The difference between Quality Group B and C (ASME Section III, Class 2 and Class 31 is in the type of nondestructive testing required. ASME Section III, Class 2 requires radiographic testing, ASME Section III, Class 3 nondestructive testing requiring magnetic particle, dye penetrant or radiographic testing. Allowable stresses, design and fabrication require-ments for ASME S.ection III, Class 2 and Class 3 are the same. .The fan coolers have been designed to meet seismic and other safety-related Quality Group C requirements. In addition, the coils are functional during normal operating conditions and are redundant, such that only two of four coolers are required for post-accident heat removal. O O22.73-1

B/B-FSAR QUESTION 022.79 " Describe, in detail, with text and figures, the perma-nently installed leakage surveillance system used for continuous pressurization between the closure flanges of electrical penetrations. Provide the system pressure and state whether or not the system will be employed at all times during normal plant operations. State which electrical penetrations, if any, are not serviced by this system; provide the testing provisions for these pene-trations. State whether the use of this system is to be in lieu of conventional Type B testing provisions. Describe the way in which leakage measured with this system is to be added to the total of Type B and Type C leakage. Describe the high leakage alarm and its redundancy and single failure characteristics." REpPONSE The permanently installed leakage surveillance system is in service continuously. Nitrogen is used to provide pressurization between the closure flanges of all electrical penetrations. The system normally operates at 50 psig. This system is used in lieu of conventional Type B testing. (N Periodic testing will be performed to insure that the system i is leak tight. This testing is done in accordance with

IEEE-317 and will establish that the leak rate of this system is less than 10-6 cm3 sec.

This leakage rate is / insignificant in comparison with the total allowable contain-ment leak rate. The system is equipped with instrumentation to detect and alarm on high or low pressure and high flow. The alarms are located in the main control room. The nitrogen system is non-safety-related and therefore, redundancy and single failure criteria do not apply. Operation of the system is not required to maintain the containmcNt integrity. l O LJ Q022.79-1

B/B-FSAR ~ (~T QUESTION 040.89 \\s-) "The FS.AR text and Table 3.2-1 states that the components and piping systems for the diesel generator auxiliaries (fuel oil system, cooling water, lubrication, air starting, and intake and combustion system 4 that are mounted on the auxiliary skids are designed seismic Category I and are ASME Section III Class 3 quality. The engine mounted components and piping are designed and manufactured 'o DEMA standards, and are seismic Category I. This 's not in accordance with the Regulatory Guide 1.26 position that the entire diesel generator auxiliary systems be designed to ASME Section III Class 3 or Quality Group C. Provide the industry standards that were used in the design, manufacture, and inspection of the engine mounted piping and components. Also show on the appropriate P&ID's where the Quality Group classification changes from Quality Group C."

RESPONSE

All auxiliary equipment up to the engine is designed to ASME S.ection III with the exception of the intake and exhaust pipes. The engine and engine-mounted components are designed f~ in accordance to DEMA Standards. The generator is manufactured (s) in.accordance to the Electrical Manufacturers Standards. All interconnecting piping between the auxiliary skids and the engine is designed in accordance to ASME. Engine mounted piping is designed to the Cooper-Bessemer standards which are indicated as DEMA Standards. The piping furnished is ASTM A106, Grade B and is acceptable under ANSI B31.1. The piping has been evaluated for seismic loads. Wall thicknesses are dependent on both the seismic and service l conditionc. The piping meets or exceeds ANSI B31.1 design l requirements. The intake and exhaust pipes are fabricated from ASTM A 155 KC 65 CL and ASTM A 155 CM 65 materials respectively. This piping is not fully in conformance with ASME Section III Class 3 requirements. These requirements are used to i insure that piping will perform the function of transporting i liquids and steam under pressure. In this case, the pipes l in question are not required to withstand pressurizativn beyond 5 psi. Because of this, the nondestructive testing required of pressure pipe was not performed. This pipe is required to remain intact and functional during seismic, LOCA, and normal operating conditions. The design requirements i and fabrication procedures specified for these pipes are l adequate to ensure that the diesel generators will be operable '( ) as required for safe plant operation. Q40.89-1

B/B-FSAR QUESTION 040.108 " Assume an unlikely event has occurred requiring operation of a diesel generator for a prolonged period that would require replenishment of fuel oil without interrupting operation of the diesel generator. What provision has been made in the design of the fuel oil storage fill system to minimize the creation of turbulence of the sediment in the bottom of the storage tank. Stirring o/ this sediment during addition of new fuel has the potential of causing the overall quality of the fuel to become unacceptable and could potentially lead to the degradation or failure of the diesel generator."

RESPONSE

A filter has been provided on the fill lines to the diesel oil storage tanks. The filters are rated 5 microne 98% removal. In addition, filters have been provided on the discharge of each diesel oil storage tank transfer pump. The rating of those filters is also 5 micron, 98% removal. I Diesel oil will be trucked on site immediately whenever a prolonged run is anticipated. The diesel oil tanks are s ) top filling tanks which will be maintained full by transfer N' from the tank trucks to minimize turbulence at the bottoms of the tanks. The twin diesel oil tanks supplying the Unit 1 emergency diesels, during periods of prolonged operation, will be replenished by refilling one tank with the other tank in service and allowing the refilled tank to settle for 12 hours. O Q40.108-1

B/B-FSAR

RESPONSE

There are two redundant and independent 4-kV emergency buses and each has two levels of undervoltage protection: ll loss of power, and 24 degraded grid voltage. The relays will be connected to the existing potential transformers on the bus. The first level of undervoltage protection is provided by induction disk type undervoltage relays. The second level of undervoltage protection is provided by instantaneous undervoltage relays with delayed drop-out. The voltage and time set points will be determined from an analysis of the voltage requirements of the safety-rela".ed loads and actual field measurements of bus voltages under various motor-starting conditions. The approximate pick-up voltage for the first level of protection is 70% of rated voltage. The preliminary setting for the second level of undervoltage protection is 92% of rated voltage. There is a 10 second time delay to avoid alarms on transisnts, and if the degraded voltage is not corrected within 5 minutes, the bus will automatically disconnect from the offsite power source and connect to its onsite diesel generator. During a sustained degraded grid voltage condition, the (~ subsequent occurrence of a SI (accidenti signal will (imme-(s)] diately4 trip the offsite power supply to the 4 kV ESF buses. Testing will be conducted during refueling outages so spurious trips during testing will not affect plant operation. The circuit will be designed to prevent automatic load shed-ding of the emergency power buses once the onsite sources are supplying power to all sequenced loads on the buses. The load shed interlock feature will use the "b" contact of the respective diesel generator breaker. This interlock will defeat the load shedding feature while the loads are being fed from the onsite power source. The load shed feature will be reinstated when the diesel generator breaker is open and the loads are fed from the offsite source. The Applicant has completed a preliminary voltage study for the Byron Station auxiliary power system. The tentative setpoint for the second level of undervoltage relay was calculated as 3804 volts. The voltage analysis performed to calculate the setpoint is based on the following conditions: 1 - Grid voltage at minimum anticipated voltage () 2 - All 4160/480 volt transformers are set on tap 3 (4160/480V4 040.176-5 l

B/B-FSA,R O

3. - All 4160/480 volt transformers are loaded to 80% of their rating 4 - SA,T (TR 6142-1 & 142-24 is carrying shut down and LOCA loads 5 - Bus loads are calculated using motor brake horse power.

The above setpoint is based on relay accuracy of +2% and a minimum voltage of 90% on the terminal of Class lE motors. O a O 040.176-6

BYRON-FSAR QUESTION 241.6 " Pipeline Settlement & Seismic Resoonses 1. Provide the final ground surface profile along the pipeline on Plate 3. 2. Although the weight of the pipeline, as stated, is less than the weight of the excavated soil; settle-ment along the pipeline should be anticipated in areas where site grade has been raised by filling and the compressible soil underneath the pipeline has variable thickness and compression characteristics. Provide settlement estimates of the pipeline located in the areas identified as Areas of Concern No. 11 and 12. Actual testing data should be used in the analysis. 3. The pipeline as shown on Plate 1, is approximately 3 miles long and extends from the River Screen House on the Rock River to the Essential Service Cooling in the plant site area. The soil supporting Tower the pipeline has variable properties and has thick-nesses varying from about three feet to about 100 7-feet over bedrock. The seismic amplification charac-( j) teristics are affected by the thickness and properties of the soil deposit. Provide analytical results showing the seismic amplifications along the pipeline and discuss their impact on the pipeline design. 4. Poorly graded, loose, non-plastic soils were encountered at Areas of Concern No. 11 and 12. Provide an eval-uation for the liquefaction potential and seismically induced settlements of these soils. Since surface water could percolate around the edges of the cohesive cover, the degree of saturation for the soil beneath ,the pipeline should be considered in the analysis."

RESPONSE

1. The final ground surface along the pipeline is shown in Figure 2.5.G-3 and is labeled on this figure as either the ground surface, 1977, or existing ground surface. 2. Estimated settlements induced by the fill and backfill in Areas of Concern Nos. 11 and 12 range from 0.25 to 0.75 inch in Area of Concern No. 11 based on the variations in soil thickness and are on the order of 0.5 inch in Area of Concern No. 12. \\/ 3. The shear modulus of the soils underlying the pipeline were estimated utilizing the data in Figures 2.5-83 Q241.6-1

BYRON-FSAR O) and 2.5-89 and mean effective over-burden soil pressures ( (Table 0241.6-1). In design of the buried piping, the variability of the supporting soil strata has been accounted for by conser-vatively choosing the design particle velocity and the apparent shear wave velocity. 4. Recent studies (Chaney 1978, and Martin et al, 1978) slow that the resistance to liquefaction increase substantially following reduction of the degree of saturation to levels below 99%. Chaney states that a degree of saturation in excess of 99% must be achieved before liquefaction occurs in less than 1000 cycles. Martin et al, shows that the stress ratio required to cause liquefaction in 10 cycles for loose sands (relative density 45%) increases by approximately 100% to 200% when the degree . of saturation decreases from 100% to 99% and 98%, respectively. The ground water table was not encountered within the soils which underlie the pipeline in Areas of Concern Nos. 11 and 12. The moisture content determined by testing samples obtained during the investigation along the pipeline range from 2.5% to 17.4% with a mean value of 10.6%. Assuming a minimum void ratio of 0.60 for (v) the loose sands, this moisture content corresponds to a mean degree of saturation of 47%. Air filled pore space, therefore, makes up approximately 20% of the soil matrix, i.e., for a 10-foot thick deposit to become saturated, a water column of 24 inches must infiltrate and remain in the soil. Since liquefaction will not occur, the settlement caused by seismically induced loads were calculated based on Silver and Seed, 1971, and Pyke et al, 1975. The estimated maximum settlements, according to these procedures, in Areas of Concern Nos. 11 and 12 are 1.5 and 0.5 inches, resp 6ctively. Since the soil profiles in the sections in question appear to be relatively homogeneous with respect to permeability characteristics, i.e., obvious impervious layers were rarely encountered in the borings, and since the bedrock contains numerous joints and fractures, most of the water that infiltrates the soils along the pipeline route should be quickly lost by percolation through the near-surface soils and joints and fractures in the bedrock. During summer months, transpiration and evaporation will account for additional loss of soil muisture. Therefore, it appears that the gechydro-(~'/ logical conditions in the area are not conducive to N \\_- the development of perched water conditions or saturation of the subsurface soils. The soils underlying the pipeline, therefore, are not susceptible to liquefaction. 0241.6-2

i I BYRON-FSAR C ESFERENCES w 1. R. C. Chanel, " Saturation Effects on the Cyclic Strength of Sands," Earthquake Engineering and Soil Dynamics, Volume 1, American Society of Civil Engineers, New York, pp. 342-358, 1978. 2. G. R. Martin et al, " Effects of System Compliance on Liquefaction Tests," Journal of the Geotechnical Engineering Division, Volume 104, No. GT4, American Society of Civil Engineers, pp. 463-479, 1978. 3. R. Pyke et al, " Settlement of Sands Under Multidirectional Shaking." Journal of the Geotechnical Engineering Division, Volume 101, No. GT4, American Society of Civil Engineers, pp. 379-398, 1975. 4. M. L. Silver, and H. B. Seed, " Volume Changes in Sands During Cyclic Loading," Journal of the Soil Mechanics and Foundation Division, Volume 97, No. SM9, American Society of Civil Engineers, pp. 1171-1182, 1971. 0241.6-3

.(.) f-s ( ) TABLE 2 .6-1 -~' SHEAR MODULUS FOR SOILS ALONG THE ESWS PIPELINE EFFECTIVE MEAN DYNAMIC DEPTH O'VERBURDEN PRESSURE, EFFECTIVE PRESSURE SHEAR MODULUS FEET PSF PSF Gmax, KSF l Upland Section 5 650 430 1,330 i 10 1,300 870 1,880 15 1,850 1,230 2,240 20 2,400 1,600 2,560 25 2,950 1,970 2,840 30 3,500 2,330 3,090 35 4,050 2,700 3,320 40 4, 60 C' 3,070 3,540 m N Flood Plain Section ? u-5 670 450 1,910 y i 10 1,040 690 2,360 15 1,400 930 2,740 ^ 20 1,760 1,180 3,090 25 2,130 1,420 3,390 30 2,490 1,660 3,670 35 2,850 1,900 3,920 40 3,220 2,150 4,170 45 3,580 2,400 4,410 50 3,940 2,630 4,610 60 4,670 3,110 5,020 70 5,390 3,600 5,400 80 6,120 4,080 5,750 90 6,850 4,570 6,080 100 7,570 5,050 6,400 110 8,300 5,530 6,700 115 8,660 5,770 6,840

e BYRON-FSAR QUESTION 241.7 " Concrete Cracks During our site visit of May 19, 1981, cracks were observed in the mat connec 3.ng the two essential cooling towers. Investigate and determine the cause of the cracks."

RESPONSE

Cracking was observed on the two longitudinal short walls between the cooling towers and on the end transverse walls. Two types of cracks were observed on the longitudinal short walls: shrinkage cracks affecting the surface of the concrete; and cracks at the construction joints between the short walls and the end transverse walls. The shrinkage cracks are narrow, less than 0.001 inch, and random in nature. The construction joints cracks are caused by the seasonal thermal movement of the end transverse walls. The end short walls are not part of the lateral load resisting system. At the end transverse walls, vertical cracks were observed. These are shrinkage and thermal cracks. Horizontal cracks in these walls, at about 2 feet from the top surface of the mat foundation, are thermal cracks caused by the contraction s of the hot water basin. Both the shrinkage and the thermal cracks are on the order of 0.001 inch wide or less. On the outside surface of the end transverse walls and on the longitudinal short walls, some diagonal cracks and weo : shaped cracked concrete were observed. These are localized 3 cracks caused by the alternating thermal movements of the end transverse walls. These localized cracks do not affect the structural resisting capability of the cooling tower. Additionally, the towers are founded on rock, and no settlement pattern of cracks can be seen. Shrinkage cracks and horizontal thermal cracks are typical of structures the size of these cooling towers. The observed crack widths are less than the limiting crack width of 0.013 inch allowed in Section 10.6 of the ACI 318-77 " Commentary on Building Code Requirements for Reinforced Concrete." The cracks observed do not reduce the structural resistance capability of the towers. Therefore, no structural repair is needed. Spalled concrete at the construction joint will I be. repaired and a flexible sealer will be placed at the joint between the longitudinal short walls and the end transverse walls. t Q241.7-1 l

w. BRAIDWOOD-FSAR O QUESTION 371.5 "Are there any outside surface tanks containing radioactive i liquids that could spill and runoff directly to surface water? If there are such tanks, provide analyses to show dilution factors and travel times at the nearest downgradient portable water supply (either surface or oroundwater, whichever is critical in terms of concentra- '- io n i. Provide discussions for both pathways."

RESPONSE

There are only two outside surface tanks which may contain radioactive liquids. These are two identical 450,000-gallon capacity refueling water storage tanks which are located immediately west of the fuel handling building on a 6-foot thick reinforced concrete mat. The refueling water storage tank is a reinforced concrete cylindrical structure consisting of 2-foot thick walls lines on the inside with a % inch stainless steel liner. The refueling water storage tanks are considered as Category I structures (and leaktight4 and are discussed in S.ubsection 3.8.4. The primary water storage tank and tank foundation is designed for the seismic (3 \\,) load condition for OBE and SSE. S O Q371.5-1

B/B-FSAR f QUESTION 423.23 g " Modify Subsection 14.2.4 to address the following items: 1. Inclusion of the entire initial test program (both preoperational and startup tests). 2. Incorporate the response to Item 423.6 into this sub-section. Ensure that all data from unsuccessful tests will be 3. recorded to permit post-test analysis. 4. State how test procedure modification (both major and minor) is accomplished. Note that the technical specifi-cations will require that minor temporary changes to procedures covering test activities of safety-related equipment must be approved by two members of the plant management staff, at least one of which holds a Senior Reactor Operator's License or, the affected unit. Since most, if not all, startup tests affect safety-related systems, this requirement applies to startup test pro-cedures. (It does not apply to preoperational tests conducted before fuel loading.) Therefore, indicate that minor changes to startup test procedures will [',l be made in accordance with technical specification s/ requirements for safety-related systems."

RESPONSE

Subsection 14.2.4 will be revisaed to incorporate the following: 1. The entire test program will be described to include both preoperational and startup tests. 2. The response to item 423.6 will be included in the text. 3. Unsuccessful results will be recorded to permit posttest review and resolution. 4. Modification of preoperational test procedures is accom-plished by the System Test Engineer for minor changes. Minor changes are defined as those which do not change the intent of the test procedure. Subsequent review of minor changes is conducted by the On-Site Review Board during post-test review. Major test procedure modifications require the approval of the On-Site Review Board and Project Engineering prior /~' to implementation. Major and minor modificatione to ctirtup C; test procedures will be performed in accordance with T >*:- least nical Specifications which require two signatures at one of which holds an SRO license on the affected unit. 0423.23-1

l B/B-FSAR O QUESTIOtt 423.24 "The response to Item 423.7 is not adequate. Revise Subsection 14.2.5 to include the applicable information l from the referenced documents." RESPO'1SE Modification and rework on systems that is required to resolve test deficiencies is controlled by the On-site Review Board during post test review and by Project Engineering who has responsibility for final test acceptance and approval. Project Engineering may specify additional test requirements to resolve test deficiencies prior to final test approval. Se.e revised Subsection 14.2.5. k O O Q423.24-1

B/B-FS.AR (} QUESTION 423.30 "The response to Item 423.10 is not acceptable. Expand existing test abstracts, or provide additional abstracts, to demonstrate the operability of the noted systems and components in accordance with Regulatory Guide 1.68 (Revision 26, Appendix A. The following additions /correc-tions are to be addressed in addition to those listed in Item 423.10.

1. n. (74 Fire protection systems 4.t.

Performance of natural circulation tests of the reactor coolant system to determine that design heat removal capability exists. NUREG-0694, "TMI Related Regiirements for New Operating Licenses", Item I.G.1, requires applicants to perform "a special low power testing program approved by NRC to be conducted at power levels no greater than 5 percent for the purposes of providing meaningful technical information beyond that obtained in the normal startup test program and to provide supplemental training". To comply with this requirement, new PWR applicants have committed to a series of natural circulation tests. To date, such tests have been performed at the Sequoyah 1, North Anna 2, and Salem 2 facilities. Based on the success of the programs at these plants, the staff has concluded that augmented natural circulation training should be performed for all future PWR operating licenses. Include description of natural circulation tests that fulfill the following objectives: Testing The tests should demonstrate the following plant characteristics: length of time required to stabilize natural circulation, core flow distri-bution, ability to establish and maintain natural circulation with or without onsite and offsite power, the ability to uniformly borate and cool down to hot shutdown conditions using natural circulation, and subcooling monitor performance. Training Each licensed reactor operator (R0 or SRO who N performs RO or SR0 duties, respectivelyn should (j participate in the initiation, maintenance, g

23.30-1

B/B-FSA.R () and recovery from natural circulation mode. Operators should be able to recognize when natural circulation has stabilized, and should be able to control saturation margin, RCS. pressure, and heat removal rate without exceeding specified operating limits. If these tests have been performed at a comparable prototype plant, they need be repeated only to the extent necessary to accomplish the above training objectives. 5.w. Containment penetration cooling system. On those penetrations where coolers are not used, provide a startup test description that will demonstrate that concrete temperatures surrounding hot penetrations do not exceed design limits.

RESPONSE

Item 4.t. Natural circulation tests of the Reactor Coolant System will be conducted to determine that design heat removal capability exists. These tests will be conducted within guidance proposed by the s reactor manufacturer (E.P. Rake (WI letter to R.R. Denton (USNRC4, dated July 8, 1981, NS,EPR-24654. Item 5.1. Later. 4 h l O Q423.30-2

a B/B-FSAR fN 22. Table 14.2-61, Reactor Containment Crane and Hoists. 1-(',) Modify the test abstract to describe load testing to be performed to meet the requirements of Regulatory Guide 1.68, Appendix A, Part 1.o.(14.

RESPONSE

Item 1: " Reactor Protection" S,ensor-to-tripping of the reactor trip breakecs will be measured. Assurance that the total reactor protective system response time is conservative with respect to the accident analysis assumptions will be provided external to the detailed preoperational test procedure. Item 5: " Component Cooling System" Table 14.2-16 will be revised to remove the sentece "S, urge tank relief valves will be bench tested at the plant site." Acceptance Criteria for bench testing of the surge tank relief valves will therefore not be added, r The component cooling surge tank relief valves (3) are outside the preoperational test program and are covered under ASME Section XI, IWV-3500, Category "C" valves. Item 6: " Containment Spray System" Table 14.2-18 will be revised to state that the paths for the air-flow test of the containment spray nozzles will overlap the water-flow test paths of the pumps at the connecting spool pieces. Item 7: "AF Pumps" Table 14.2-19 has been revised to identify the " prime movers" in question and commits to 5 con-secutive, successful, cold starts per pump. Item 8: " Primary S.ampling System" The Test Summary section of Table 14.2-20 will be revised to include verification of flow paths, holdup times and sampling procedures. Item 10: " Fuel Pool Cooling and Cleanup System" l 't a. The Test Summary section of Table 14.2-22 \\/ is revised to specify the other flow paths. See revised Table 14.2-22. 0423.33-7

a f B/B-FSAR b. Test objectives, a test summary, and accpetance gx g,) criteria have been provided for the anti-siphon feature in the return line and for the operation of the filters and demineralizers in purifying the Refueling Water Storage Tanks. Test objectives and a test summary have been provided for the leakage detection system. See revised Table 14.2-22. Item 11: " Fuel Handling System Testing" Table 14.2-23 has been revised to incorporate the following changes: a. The system description has been moved from the Test Summary section to the Test Objective section. b. A description of load testing for the Fuel Handling System hoists and cranes to meet the requirements of Regulatory Guide 1.68, Appendix A, Part 1.m. (44 revision 2 was added to the Test Summary section. Item 12: " Diesel Generators" () Table 14.2-25 has been revised to state that data is taken to conform to Regulatory Guide 1.108, Rev. 1, Reg. Position C2. a. (14, (34, (44, (64, (96 and 2.b. See revised Table 14.2-25. Item 13: " Diesel Fuel Oil Transfer Pump" Table 14.2-26 has been revised to state that each fuel oil transfer pump will deliver fuel oil to each diesel generator in excess of the maximum demand, as indicated in Subsection 9.5.4. See revised Table 14.2-26. Item 14: " Tables 14.2-28, -29, -30, ECCS - Safety Injection Pumps, Centrifugal Charging Pumps, RHR Pumps" The response to item 423.12 sub-item 21 will be revised to read "see Table 14.2-33" instead of "see revised Tables 14.2-26, -29 and -30". The Safety Injection Pumps and RHR pump abstracts will be revised to read refueling water storage tanks instead of reactor water storage tanks. Tables 14.2-23, 29 and 30 will be revised by deleting O "and flooded" from the first sentence of the Test 0423.33 8

B/B-FSAR f-S.u.mma r y. Also, the second sentence of the Test (S) Spmmary paragraph 2 (on Tables 14.2-28 and -294 will be deleted. This will eliminate the existing inconsistencies. Item 15: " Table 14.2-31, ECCS - Accumulators" The Test Spmmary section of Table 14.2-31 will be revised to state that proper operation of the nitrogen fill, vent valves, accumulator drains and accumulator makeup will be verified. Item 16: " Diesel Generator Ventilation System" The Diesel Generator Room Ventilation system does not contain filtration or absorption units. Item 17: "H Recombiners" 2 Table 14.2-40 has been revised to state the test will be performed to demonstrate the capability of the system to operate properly, at conditions near as possible to those given as standard per the manufacturer. The hydrogen analyzer monitor will be demonstrated (O ) to function properly. S.e.e revised Table 14.2-40. Item 18: " Containment Ventilation System" Table 14.2-41 has been revised to replace " specialized" with " operated". In addition, the following acceptance criteria has been added to Table 14.2-41: The containment recirculation fan motor current will be demonstrated to be within its design value at accident conditions by measuring air density, temperature, humidity, fan speed, air flow and motor current and making engineering extrapolation to accident conditions. See revised Table 14.2-41. Item 19: "MSIV's" Table 14.2-42 has been revised to indicate full travel closure times of each valve and acceptance criteria expanded to include all items in test summary. Item 20: " Primary Safety and Relief" a. Table 14.2-50 has been revised to state that three specific features of the Power Operated (G"'T Relief Valve Logic will be tested. They 0423.33-9

_. = _. ~. D/B-FSAR I' are the Auto-Low Temperature Control, Valve response time and enable functions from the pressurizer pressure transmitters. See revised Table 14.2-50. b. In-plant preoperational testing of Primary System safety valves will include setpoint verification and seat leakage measurement. S,ee revised Table 14.2-50. Item 21: "St.eam Generator S.a.fety & Relief Valves Table 14.2-51, Acceptance Criteria, has been revised to include all components identified in the Test Summary. Item 22: " Polar Crane" The construction tests conducted for the Polar Crane meet the requirements of R/G 1.68 Appendix A, Part 1.o.(I4 and will not be repeated during the Preoperational Testing. O G G O Q423.33-10

B/B-FSAR ( ) QUESTION 423.34 j "The response to Item 423.13 is not acceptable. Section 9.3 states that some valves in the compressed air system, namely certain containment isolation valves, power-operated main steam relief valves and auxiliary feedwater flow control valves, fail in the safe position on loss of air. The i operability of safety-related equipment and processes would i be compromised if these valves. failed in the unsafe position. Demonstrate proper operation of these valves in accordance with the testing requirements of Regulatory Guide 1.80."

RESPONSE

s The containment isolation valves'for the compressed air systems, the power operated main steam relief valves and auxiliary feed-water control valves will be tested to insure they fail to their designed safe position on loss of air. t g O w W N 6 A w 1 O Q423.34-1 i .-v

B/B-FSAR D C QUESTION 423.37 5- "Our review of licensee event reports has disclosed that many events have occurred because of dirt, condensed moisture, or'other foreign objects inside instruments o and electrical components (e.g., relays, switches, breakerst. e Describe any tests or inspections that will be performed or any administrative controls that will be implemented .iuring the initial test program to prevent similar com-ponent failures."

RESPONSE

Administrative controls will be implemented during the initial test program to maintain the level of cleanliness in accordance with Regulatory Guide 1.39, Revision 2. i l l l l t I r Q423.37-1

B/B-FSAR ('s -} ' ~' b. The Test Summary section has been changed such that all rods falling outside of the two-sigma limit in drop times will be retested a minimum of three times each. I tem 4 : " Reactor Coolant System Flow" a. Table 14.2-70 will be revised to state that data will be taken to ensure pump performance, rota-tional speed and indicated flow are consistent with performance curves, b. Every test procedure includes a prerequisite that all instrumentation used in the test to measure acceptance criteria must be within current calibration intervals. Therefore, no modification to Table 14.2-70 will be made, c. Vibration monitoring of the reactor coolant pumps will be done using 2 IRD pickups mounted to the motor supports (90 degrees apart in the horizontal plane). In addition, baseline vibration data on the pumps will be obtained using a portable IRD vibration measurement unit. These will be taken at bearing points on the pump motor (in 3 directions, where possible) during the preopera-C, tional test. Item 5: " Initial Criticality" Table 14.2-75 has been revised to incorporate the following changes and additions: a. Source range nuclear instrumentation shall indicate at least 1/2 count per second prior to startup and the source range signal-to-noise ratio will be greater than 2. i b. The NSSS vendor will provide predictions of boron concentration and control rod positions for initial criticality. Rod withdrawal or boron dilution will be termirated if actual values are seen to be deviating-from predicted values until the source of the deviation is evaluated. c. The reactivity addition sequence prescribed per NSSS vendor recommendations to ensure that criti-cality will not be passed through on a reactor period shorter than 30 seconds. Item 6: " Power Ascension" The requested information has been incorporated in the test summary of Table 14.2-76 (Power Ascension). 0423.42-5

s + B/B-FSAR i Item 7: " Pseudo Rod Ejection" g ')' g Table 14.2-81 has been revised to incorporate the following additions: a. The most reactive RCCA will be withdrawn for this test. b. The worth of the most reactive RCCA will be veri-fled to be conservative with respect to the acci-dent analysis. Item 8: " Power Coefficient of Reactivity" Table 14.2-82 has been revised to incorporate the ~ following addition: Reactor thermal power will be determined using calori-metric data'. The asssociated reactivity changes will be determined using the reactivity, computer, T,yg recorded, and Delta T recorder. Item 9: " Turbine Trip" The turbine trip startup test will document the plant response to:the transient. Comparison of results gg( ) to predicted values will be made external to the procedure. Item 10: " Shutdown from Outside Control Room" Table 14.2-86 will be modified to show data will be obtained from outside the control room to verify a plant hot standby condition following shutdown and that the plant can be maintained stable for at least 30 minutes. The Byron /Braidwood safe shutdown is designed for hot standby. Item 11: " Station Blackout" Table 14.2-87 has been revised to state that a blackout l of at least 30 minutes is an acceptance criteria. I See revised Table 14.2-87. Item 12: "104 Load Swing" Table 14.2-88 will be revised to include the response to 0423.19 sub-item 15. Additionally acceptance criteria will be included to address overshoot, under-shoot, and oscillations. l 0423.42-6

1 l B/B-FSAR l i Item 13: "50% Load Reduction" I Table 14.2-89 will be revised to incorporate the response to Q423.19, sub-item 16. i i I f l l i f I O i 4 i 4 t f f 9 1 l 1 t 1 E t t l 4 i e 1 i j. Q423.42-7 i l i e 4 6 . - - - - ~ - - - - -}}