ML20040C764

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Forwards Advance Info to Be Included in Next FSAR Amend. Includes Revised Fire Protection Items
ML20040C764
Person / Time
Site: Byron, Braidwood, 05000000
Issue date: 01/07/1982
From: Tramm T
COMMONWEALTH EDISON CO.
To: Harold Denton
Office of Nuclear Reactor Regulation
References
NUDOCS 8201290215
Download: ML20040C764 (31)
v • d • e


Text

{{#Wiki_filter:--.. 'N Coedmonwealth Edison - ) che First No.ional Plaza. Chicago. Illinois C" Add ess Riply to: Post Office Box 767 CtVcago, litinois 60690 January 7, 1982 g q s Mr. Harold R. Denton, Director g !?OCEWED Of fice of Nuclear Reactor Regulation

]M 237982 U.S. Nuclear Regulatory Commission T.

yj cp // Washington, DC 20555 (

Subject:

Byron Station Units 1 and 2 p s Braidwood Station Units 1 ah /2' Advance FSAR In fo rmation ~ l' NRC Docket Nos. 50-454/455/456/457

Dear Mr. Denton:

l This is to provide advance copies o f 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 fifty-nine (59) copies-of this letter are provided. Fifteen (15) copies of the enclosures are included for your review and approval. Please address further questions to this office. Very truly yours, N. 4"* T.R. Tramm Nuclea r Licensing Administrator. Pressurized Water Reactors Attachment 4 Qool s /I O!hh0kS4 A PDR i

ATTACHMENT A LIST OF ENCLOSED INFORMATION I. FSAR QUESTION RESPONSES New: CSB,T6.2-16 Revised: 005.3 022.15 040.83 022.25 II. FSAR TEXT CHANGES Pg. 8.3-9 (per Q40.83) Pg. 11.4.3 (VRS) Pg. 11.5-10 Pg. ll.5-10a III.- MISCELLANEOUS ITEMS Revised Fire Protection Items 1, 2, 4 CEB Items 1 through 10 MEB Item 14 ASB Item 8 1 copy of P&ID M-59 for CSB

D/B t i 11RC-P.M. SEARS /CMEB REQUEST FIRE PROTECTION OPEN ITEMS REQttBST 1. 1 Provide a description of the extent of fire proofing and fire barriers in the Control Building Complex. State whether or not all rooms have two means of exit. RESPCNSE Except as noted below, all rooms have two means of exit: The HVAC equipment rooms on El. 439'-0" in Cable Rooms A and G; the cable riser areas on El. 439'-0" and 451'-0"; the Records Room, Computer. Rooms and Storage Room on El. 451'-Od; the Security Control Center on El. 451'-0", Upper Cable Spreading Areas B and G on El. 463'-9". Fireproofing of exposed steel in the Control Building complex covers all columns and floor framing steel on El. 439'-0", 451'-0", and 463'-0". The steel decking is also fire proofed in all these areas except for the 463'-5" slab between column rows 11 and B and L and M, and 23 and 25 and L and M. Below these two areas are the Security Control Center and kitchen. All fire doors and 3 hour fire rated and all walls and floors O are three hour fire rated in both directions excluding the interior walls and ceilings of the Rooms underneath Upper Cable Spreading Areas A and F. The stairwells are provided with three hour fire walls and doors. The control room filter units are located in area A and F in the Upper Cable Spreading Area. O

w-i B/D NRC-P.M. SEARS /CMEB REQUEST w FIRE PROTECTION OPEN ITEMS REQUEST 2 2 Provide the locations of fire detectors in the control room. RESPON SE Main Control Panels 1(2) PM0lJ through 1(2) PM06J are provided with their own supply and exhaust an ducts. There are eight other ducts exhausting into the control room on each unit. All 28 supply and return ducts on each. unit are provided with a smoke detector. Smoke detectors are also installed above the egg crate ceiling. The remaining panels in the control room are not ventilated and are not provided with smoke detectors. O i l U t

7 - B/B NRC-P.M. SEARS /CMEB REQUEST FIRE PROTECTION OPEN ITEMS REQUEST 4 - State that all fire stop penetration seals will have a fire rating equal to the fire barrier they are penetrating.

RESPONSE

All fire stop penetration seals will have a fire rating at least equal to the fire barrier they are installed in. Fire stop assemblies have been or will be qualified in tests for the fire rating required using ASTME119-or IEEE383-1974 standards and insurance underwriter requirements. F L O O ~ l

'~ D/B CEB ITEMS 4 1 l le Demonstrate compliance with all requirements of NUREG-l 0737, II.B.3, for sampling, chemical and radionuclide analysis capability, under accident conditions. R?SPCNSE The' liquid sampling equipment located in the sample room on elevation 401 feet of the auxiliary building is designed for post accident analysis of primary coolant. An in-line chemical analysis panel (CAP) and a grab-sampling panel (both manuf actured by Sentry Equipment Corporation of Oconomowoc, Wisconsin) have been purchased for each unit., The grab-sampling panel is designed to create fi::ed' volumes (cooled and depressurized) of diluted and undiluted primary coolant samples. This panel is also capable of giving a controlled volume of diluted strip gas. These samples can be analyzed in the station's gas chromatograph or sent to an offsite lab when necessary. /"') ~ The CAP is capable of performing inline analysis of pH, (_) dissolved oxygen, chloride, hydrogen, and specific conductivity of primary coclant during normal and post accident conditions. The grab-sampling panel is the manual backup for the CAP. A diluted primary coolant grab-sample is required for boron and radionuclide analysis. This sample or an additional sample can be analyzed in the laboratory for hydrogen, oxygen, and chlorides. Special casks have been purchased for these samples, so that they can be safely handled, The iso-topic analysis shall include iodines, cesium, and non-volatiles. A minimum of one diluted grab-sample can be taken per day for the duration of the post accident period. Samples are taken from the reactor coolant system by direct extraction from the reactor coolant system piping through dedicated sample taps and sample lines connected to the High Radiation Sampling System. Samples taken from the contair. ment atmosphere are extracted directly from ~ the containment atmosphere through containment air sampling panel system piping and containment penetrations. No isolated auxiliary systems are required to be in l operation in order for the reactor coolant and containment atmosphere samo. les to be taken.

'^ i B/B O) (_ RESPONSE #1 (Continued) A diluted primary coolant sample can be drawn via a direct tap, delivered to the laboratories, and analyzed-within 3 hours of the time that the sampling decision is made. The radiation exposure to any individual will be less.than the exposure limits specified in GDC 19. The shielding design is based on sources specified in Regula-tory Guides 1.4 and 1.7. The primary coolant volume for this post accident shielding design is the same as the normal primary coolant volume. Item #2 below discusses the shielding and doses in more detail. Laboratory equipment has been purchased that will allow identification and quantification of isotopes which include iodides, cesians, noble gases and nonvolitals. The accuracy, range and sensitivity of these instruments and the CAP are adequate for describing radiological and chemical status of the primary coolant. Procedures are being developed for testing and calibrating the measuring equipment. The procedures will also define the dilution requirements (if required). for a containment air, a diluted primary coolant (10-3 dilution), and a strip gas sample. Table 1.1 gives the expected concentrations of a few radioactive isotopes using Regulatory Guide 1.4 and 1.7 source assumptions. y 1 A hood in the high level laboratory will be designated ~ as!a high activity hood. The owner has authorized (in October, 1981) modification to this hood. Shielding shall be designed and installed prior to fuel load. Dilution capabilities and procedures will be similar to those used at Commonwealth Edison's. operating stations. The following design features are included in the High Radiation Radiation Sampling System: a. Purging and flushing is designed into both the liquid sampling and containment atmosphere sampling portions of the High Radiation Sampling' System. Purging and flushing of panel piping a" semple lines can be done prior to and subsetie2 i to drawing of the sample. b. stainless steel piping for all s:7,2% ,nes and piping'of the system, c. fifteen tube diameter radius bends in field sample 1

lines, O)

(_

? D/B l () '\\ > RESPONSE #1 (Continued) d. heat tracing of containment atmosphere sample j l lines to reduce iodine plateout, purging and flushing of lines to reduce plateout f l e. and exposures, f. multiple sample taps for reactor coolant sampling to insure representative sampling and diverse l sample points, g. optimum panel sample line sizes to limit crud buildup, h. and a waste drain tank and pumps for collection of liquid sample effluents before return to contairament during accident conditions or to the chemical drain tank or volume control tank under normal operating conditions. Ventilation air from the High Radiation Sampling System's liquid sample panels and chemical analysis panels located in the Primary Sample Room is exhausted through the Engineered Safety Features Filter Systems which is (% disco-sed in FSAR Subsection 6.5.1 and 6.5.1.1.2. The (_/ filter system includes provisions for prefilters, HEPA filters and charcoal adsorbers. Inline hydrogen analysis of the containment atmosphere is descriLcd in Item E.30 of Appendix E. The information given above indicates Byron /Braidwood Stations' compliance or plans for compliance to NUREG-0737, II.B.3, Post Accident Sampling. All planning, procedores, equipment installation, and shielding will be completed 6 months prior to receiving an "above 5% power license." Testing and training will be completed prior to receiving this license. O 1 \\_/ o

B/B CEB ITEMS Table 1.1 Maximum Expected Concentrations O in Lab,oritory Samples (uCi/cc)* Diluted ** IStrip@ Containment @ Primary Gas Air Isotope Coolant Sample Sample 477 201 Kr*85 29.5 12.4 Kr85 Rb88 126 Sr89 2.14 Sr90 .20 Y90 3.70 Zr'95 5.19 Mo99 5.28 Il31 158 79.0 33.3 TE132 4.40 Il32 166 83.0 35.0 Il33 231 116 48.8 5950 2500 Xel33 Csl34 .11 1350 572 Xel35 Csl37 .30 t 1490. 10450. 4410.

  • '1 hour after shutdown, post accident with 100% cladding failure and R.G.

1.4 and 1.7 source

    • Stripped primary coolant that is diluted by a factor of 1000 0 100% core nobles and 2.5% of core halogens O

S

B/B l CEB ITEMS (Continued) 2. Provide sufficient shielding to meet the requirements of GCD-19, assuming Reg. Guide 1.4 source terms.

RESPONSE

i The design basis post accident sources for the post accident sampling system assumes 100% fuel cladding damage which results in the most conservative liquid concentrations and containment air concentrations. Regulatory Guide 1.4 sources are placed in the normal operating primary coolant volume (i.e.,100% nobles, 50% hologens, and 1% of all other isotopes). Containment air contains 100% nobles and 25% hologens. Credit i for containment spray was taken. The panel manufacturer designed the panels with shield-ing for the front of the panel only. The grab-sampler and the CAP have 7 inches of lead shot and 1 inch of steel. The containment air sampler has 3 inches of steel shielding. The radiation shielding surrounding these panels are designed to have a contact dose of 100 mrem /hr for occupational areas and'1 rem /hr for other areas. i The shielding design for the liquid sampling panels are shown in figures 1, 2, and 3. The shielding for the air sampling system is similar. The transportation casks satisfy NRC shipping requirements (i.e., 10 mrem /hr at l' meter). Figures 4 and 5 give the designs of the liquid and air sampling transportation casks. This information will be included in A'ppendix E (Section E.21) when the sampling' system's shielding is completed. The shielding design exceeds the requirements of GDC-19. Table 2.1, which is an estimate of the doses that are experienced by a trained operator performing all of the tasks, give the estimated whole body and extremity doses for taking a diluted primary coolant sample. The trip gas sample gives doses that are 30% greater. The doses received while taking a containment air sample are expected to be half the diluted liquid sample doses. If a single operator took 4 samples (2-primary coolant, 1-strip gas, and 1 contsiament airl, he would receive a whole body dose of 300 mrem. Laboratory procedures and local-l ized shielding will be utilized to maintain doses (to lab workers) well below the allowable levels in GDC-19. O 1

A [J TABL 1 I t

SUMMARY

OF ESTIMATED INTEGRATED DOSES t 87 ml DILUTED REACTOR COOLANT SAMPLE PROCEDURE * (Assumes one individual performs all tasks) Time Dose Rate Whole Body Extremities Required Range Dose Dose (min) (mrem /hr) (mrem) (mrem) Step _ Assemble & instruct trained personnel 7.5 ~0 0.0 0.0 Purge *** incoming lines to the panel 7.9 0-142 18.9 25 Purge *** the panel to sampling 3.7 142-264 16.1 25 Fill the 100 mi bottle (87 ml) 3.2 264-321 17.1 25 Isolate the bottle within the cask 0.5 321 2.7 6 Pull cart & close lid 0.1 <30 0.1 0.5 Flush the activity from the panel 17.1 321-0 23.2 30 Deliver samples to lab <20 10-1 1.0 1.0 s TOTALS 60.0** 79.1 102.5 times a degassed design basis primary coolant sample. The contact ~ Dilution is 10 for 1 ml of this liquid should be less than 5 rem /hr and 200 mrem /hr et 10 centimeters. The estimated maximum time is 90 minutes. The extra 30 minutes is to allow for contingencies, i.e., loss of offsite power. With primary coolant i

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D/B a CEB ITEMS (Continued) ('] U 3. Commit to meet the sampling and analysis requirements of Reg. Buide 1.97, Rev. 2.

RESPONSE

A. Post Accident Sampling Capability - As stated in the B/B - FSAR Appendix A, the requirements of this guide (Reg. Guide 1.4) have been adhered to in all pertinent sections; thus, we already met the requirements for post accident sampling capability sited in Reg. Guide 1.97, Rev. 2. B. Post Accident Analysis Capability (on sitel 1. A post accident radionuclide analysis portable system (PARAPS1 will be operational on-site for gamma-ray spectrum analysis in a post-accident situation. 2. The B/B Post Accident High Radiation Sample System will analyze for pH, hydrogen, oxygen, and boron concentration (.as an added section the O High Radiation Sampling Systeml. 4. Verify that all electrically powered components associated with post accident sampling are capable of being supplied with power and operated, within thirty minutes of an accident in which there is core degradation, assuming loss of off site power.

RESPONSE

The electrically powered components of the High Radiation. Sampling System are supplied from 4160V non-ESF buses which are capable of being manually connected to 4160V ESP buses using a safety related tie breaker. Further discussion of the auxiliary a-c power system can be found in FSAR Section 8.1 l 5. Verify that valves which are not accessible for repair l af ter an accident are environmentally qualified for the conditions on which they must operate.

RESPONSE

D) The valves in the High Radiation Sampling System are (,, designed for the environment in which they must operate. i l

D/B CEB ITEMS (Continued) 6. Provide a procedure for relating radionuclide gaseous and ionic species to estimate core damage. l

RESPONSE

A study is presently being made to estimate core damage ucing isotopic identification and quantification. The table in item #1 above gives the expected maximum concen-tration in the laboratory samples of 100% cladding failure. These levels are at least 1 order or magnitude greater than the design basis normal operating concentrations (1% failed claddingl. 7. State the design or operational provisions to prevent high pressure carrier gas from entering the reactor coolant system from on line gas analysis equipment, it it is.used.

RESPONSE

The reactor coolant module of the High Radiation Sample O System's liquid sample panel takes a reactor coolant sample which then is separated into a depressurized reactor coolant sample and a pressurized stripped gas (from the reactor coolant sample). An 80 psig argon supply is provided to the reactor coolant module for the separation function. The argon pressure is further reduced in the module to provide a 10 psig stripped gas sample from the reactor coolant sample. The depressurized reactor coolant sample and the pressurized stripped gas sample are then routed to the chemical analysis panel where analyzers for oxygen, chlorides, ph and conductivity, and chromatographs both ion and gas are located. Detailed sampling operation procedures are used for taking the reactor coolant sample, separating the pressurized strip gas and depressurized reactor coolant, and routing them to the chemical analysis panel. The procedures and low supply pressure of the argon gas will prevent any gases from entering the reactor coolant system. (q _s

n D/B' () 8. Provide a method for verifying that reactor cociant dissolved oxygen is at 0.1 ppm if reactor coolant chlorides are determined to be 0.15 ppm.

RESPONSE

A Rexnord dissolved oxygen meter will be the method for determining that oxygen is less than 0.1 ppm if reactor coolant chlorides are determined to be greater -han 0.15 ppm. 9. Provide information on (al testing frequency a.".d type of testing to ensure long term operability of the post accident sampling system and Ob) operator trair.ing requirements for post accident sampling.

RESPONSE

A. Testing frequency of the post accident system will be at a minimum yearly. All analysis functions will be located. B. Formal initial training will be given to a"l personnel responsible for the post accident system. Retraining O will be given, as necessary, to maintain c",mpetance. I 10. Provide additional information on the condensa e cleanup system.

RESPONSE

See revised Subsection 10.4.6. O

B/B h MEB ITEM 14 Inter-system LOCA/ Periodic Leakage Testing The'following valves have been identified as RCS system isolation valves requiring periodic leakage testing. The list is tentative. The present Byron piping design allows for leakage testing of individual " backup" check valves only. ' tion leakage testing for " primary" isolation check valves is done simultaneously (not individually) for a gross leakage with a gobal limit of 5 gpm. It is our understanding MOV's with position indication in the control room can be deleted from the program / list. This includes all MOV's on the list. O O e t O o

B/B Byron Inter-system LOCA Isolation Valves Check Valves providing Backup Check RCS pressure isolation Valves High Head Inj. Cold Leg 1 SIB 900A 1S18815 ISI8900B 1S18815 ISI8900C 1SI8815 1FI8900D ISI8815 RHR Pump Discharge Hot Leg ISI8949A ISI8841A ISI8949C 1SI8841B SI Pump Discharge Hot Leg ISI8949A ISI8905A ISI8949B ISI8905B 1SI8949C 1SI8905C 1SI8949D ISI8905D Accumulator Cold Leg .1SI8948A ISI8956A O c) ( ISI8948B ISI8956B ISI8948C 1SI8956C 1SI8948D ISI8956D SI Pump Discharge Cold Leg ISI8948A ISI8819A ISI8948B ISI8819B ISI8948C 1SI8819C 1SI8948D ISI8819D RHR Pump Discharge Cold Leg ISI8948A ISI8818A ISI8948B 1S18818B l 1SI8948C 1S18818C l 1SI8948D ISI8818D 1 l l O l

B/B ASB OPEN ITEM #8 O Subject Auxiliary Feedwater System Reliability Analysis - Third Recalculation. Summary of Results ~ LOOP 9.2X10 Unavailability Per Demand Assumptions 1. LOOP to both units 2. Credit taken for manual breaker closure to Bus 141 from Bus 241 3. All other assumptions per NUREG-0611 Changes From Second Recalculation () 1. LOOP assumed to both units rather than to only the unit under study ~ Di,esel generator gnreliability assumed to be 1X10 2. rather than 3X10-3. The components of the fault tree that lead to a top event titled " Failure to Supply Bus 141 from Bus 241" are: a. Failure to close breakers from 241 to 141 (two subcomponents of operator action and breaker f ailure to closel -- 1.2X10-2 l b. Failure of 2A diesel generator to start - l 1.0X10-2 l c. Failure of diesel generator breaker to auto close to Bus 241 - 1.0X10-3 l The analysis currently in the FSAR will be revised to reflect these changes. i 1

9 3/E-FSAR a including diesel-generators and distribution equipment, satisfies single-failure criteria. p) (_ The loss of either offsite power, onsite power, or power generated by the unit will in no way preclude the availability of the other power sources. Class lE equipment is environmentally qualified to meet the intent of IEEE 323-1974 to the maximum extent possible (see Section 3.11). All Class lE equipment and systems important to plant safety are designed to permit periodic testing during power operation in accordance with General Design Criterion 18. Preoperational and initial startup test programs for the standby power system and its supporting systems are described in Chapter 14.0. Each safety-related electrical control and power system including all safety-related equipment, is designed to facilitate periodic functional testing during plant operation. The testing of these systems and components are in accordance with IEEE Standards 279-1971 and 338-1975. Tests have been designed to demonstrate performance reliability of each diesel-generator with respect to the expected parameters of operation and environment during a design-basis accident and/or a loss of offsite power. Prototype qualification made on one diesel-generator consisting of 300 start and sequential loading tests with no more than three failures is performed to demonstrate type reliability. Loading 7g for prototype testing is the kW equivalent of the ECCS loads. (' ') Testing of each of the remaining three diesel-generator sets purchased by Commonwealth Edison included at least two starting and loading tests. Plant preoperational testing is described in Chapter 14.0 and availability testing during normal plant operation of the diesel-generators is in accordance with those provisions set forth in IEEE 387-1972. Each diesel-generator is R periodically tested as described in the Technical Specifications. 8.3.1.3 Physical Identification of Safety-Related Equipment 8.3.1.3.1 General Color coded nameplates or labels are used to distinguish between Class lE and non-Class lE components and between components of different division, as shown in Table 8.3-4. Cables are routed through independent cable tray systems according to the separation criteria stated in Subsection 8.3.1.4.3. 8.3.1.3.2 Raceway Identification The cable tray system is distinctively identified throughout the plant in accordance with IEEE Standards 279-1971 and 384-1974. Each tray is marked with a colored alpha-numeric segregation code consisting of characters which are 2 inches high (or maximum bv 8.3-9

D/C-FSAR AMENDMENT 36 JANUARY 1982 11.4.3 Volume Reduction System Description O) \\s-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 the VRS. The fluidized bed dryer system processes the concentrates or bottoms from the radwaste evaporators. Table 11.4-2 provides the quantity of concentrations to be processed. The evaporator bottoms containing typically 10 to 25 wt% dissolved salts are storeu 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 l 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. Fluid-izing 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 l particles. The resultant dry anhydrous product from the fluid-ized 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 stream. The granulated, solid waste product is discharged from i the fluidized bed into the product storage hopper. I O \\,s/ The' overhead air stream from the fluid bed vessel, consisting of hot gases and fine salt particles, passes through a gas / solids l separator, wherein the bulk of the fines and ash (from the dry waste processor) are removed from the gas stream and discharged i to the product storage hopper. I DAW consisting typically of contaminated wood, paper, cardboard, cloth, rubber, and plastics such as polyethylene is grocessed in the dry waste processor. Approximately 29,000 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 l chopper-shredder. It is stored in one of two hoppers until a j batch is accumulated. The DAW is then pneumatically trans-ferred to the dry waste processor where it is reduced to ash. l The dry waste processor is also a fluidized bed system. The fluidizing air carries the ash to the gas / solids separ-'.or where it is combined with the fines from the fluidized ced dryer. l The gases leaving the gas / solids separator are directed to a venturi scrubber / condenser unit. In this unit, the ga's con-taining the fines and ash is first contacted with water in a high-energy venturi scrubber, resulting in removal of virtually ) 11.4-81

m.e D/B-FSAR AMENDMENT 36 JANUARY 1982 O A high radiation level signal initiates automatic closure of the valve located in the component cooling surge tank vent line to prevent gaseous radiation release. 11.5.2.3.2 Component Cooling Water Monitors Radiation detectors 1RE-PR009, 2RE-PR009, and ORE-PR009 continuously monitor the component cooling system for leakage of reactor coolant from the reactor coolant system and/or i the residual heat removal system. Detector 1RE-PR009 is interlocked with the component cooling surge tank 1CC0lT vent valve ICCRCV017, and detector 2RE-PR009 is interlocked with the component cooling surge tank 2CC0lT vent valve 2CCRCV017. Detector ORE-PR009 is interlocked with both vent valves, ICCRCV017 and 2CCRCV017. 11.5.2.3.3 Steam Generator Blowdown Detectors 1RE-PR008 and 2RE-PR008 monitor steam generator blowdown for Unit 1 and 2 respectively. Steam generator blowdown sample flow is normally routed through the steam generator blowdown sample panel, OPS 0lJ, and on O to the radiation monitor. Automatically on high radiation, detector 1RE-PR008 interlocks to close steam generator blowdown sample valves 1PS179A through D to terminate sample flow to the sample panel and radiation monitor. A similar interlock i exists between detector 2RE-PR008 and valves 2PS179A through D. I Termination of sample flow on high radiation protects personnel in the high level laboratory where the sample panel is located. Subsequent sampling of steam generator blowdown can be accomplished by manually redirecting sample flow to the primary sample room. Sequential isolation of steam generator blowdown can be used to determine which steam generator may be leaking. 11.5.2.3.4 Blowdown Filters The flow from the blowdown mixed-bed demineralizers is normally sent to the condensate storage tank. Detectors ORE-PR0l6 through 19 are interlocked with the blowdown after filter discharge valves OWX119A through D and blowdown monitor tank inlet valves OWX58A through D. Automatically on high radiation, the flow from the blowdown mixed-bed demineralizers is redirected to the blowdown monitor tanks. 11.5.2.3.5 , Gross Failed Fuel' Monitor l Radiation detectors 1RE-PR006 (and 2RE-PR006 for Unit 2) continuously monitor the CVCS letdown line downstream of 11.5-10

e B/B-FSAR AMENDMENT 36 JANUARY 1982 () the letdown heat exchangers and upstrem of the mixed bed demineralizers. The CVCS is described and the piping and instrument diagram shown in Subsection 9.3.4 and Figure 9.3-4, respectively. High radiation alarms are provided' in the main control room to alert the operator of an abnormal increase in gross gamma activity in the letdown stream. Grab sample features are included on the monitor skid for laboratory analysis of primary coolant letdown activity. In addition, process sampling of the letdown line is available at the high radiation sampling system described in Subsection 9.3.2.1 and Appendix E.21. 11.5.2.3.6 Miscellaneous Process Liquid Monitors A high radiation signal from other liquid detectors listed in Table 11.5-2 will be annunciated in the control room. 11.5.2.4 Sampling The following subsections present a detailed description of the radiological sampling procedures, frequencies, and objectives for all plant process and effluent sampling. This sample program provides the means to show compliance O' with the technical specifications as presented in Chapter 16.0 for the process radiation monitoring and radwaste systems. 11.5.2.4.1 Process Sampling Subsection 12.3.4 describes the inplant airborne sampling system. The sample frequency, type of analyses, analytical sensitivity, and purpose of the sample are summarized in Table 11.5-3 for each liquid process sample location, and in Table 11.5-4 for each gas process sample location. The analytical procedures used in sample analysis are presented in Subsection 11.5.2.4.4. These O 11.5-10a

D/D CSB QUESTION T6.2-16 Where did this mass and energy release data come from. What analytical model or computer code was used in generating tie data. RES P.)NS E A review of this table indicates that although the table is titled as being applicable for a steamline break, the data appears to correspond to release rates for a liquid line break. The applicant is preparing a revised table to replace existing Table 6.2-16. When the revised table is available, appropriate references to the source of the data will be included. O l ([]) ~ - =

WE.T w BYRON-FSAR 3 QUESTION 005.3 5 9 "As noted in Section 5.2.1.1 of the FS.A,R, the control ( valves for the Byron Station Units 1 and 2 are not in conformance with 10 CFR Part 50, Section 50.55a, Codes and S.tandards. These components are constructed to Section III, Class 1, of the ASME Boiler and Pressure Vessel Code, 1971 Edition, through the Su.mmer 1972 Addenda, whereas, the regulation requires the components to be constructed to the same code and edition through the Winter 1972 Addenda to the code." "In order to assess the acceptability of these control valves, identify those portions in the 1972 Winter Addenda to the code with which the control valves are not in compliance."

RESPONSE

As stated in Subsection 5.2.1, the Byron control valves are designed and fabricated in accordance with ASME Code S.ection III, 1971 Edition through the Summer 1972 Addenda. Westinghouse believes that an acceptable level of quality () and safety in the design and fabrication of these valves has been achieved by conformance with the S.u.mmer 1972 Addenda of the Code. i However, Westinghouse is performing a review of the differences between the S.ummer 1972 Addenda of the Code to which the Byron control valves were' designed and the Winter 1972 Addenda required by 10 CFR 50.55a. The significance of any differences identified will be assessed and explained. The results of this review will be available by February 15, 1982. O 005.3-1

.______e_________ B/B-FSAR ~ '~l OUESTION 022.15 "In accordance with the NRC request to all applicants. contained in Task Action Plan A-2 of NUREG-0609, " Asymmetric i Blowdown Loads on PWR Primary Systems," provide the following information for the reactor coolant system break that results in the peak differential pressure loads within the reactor cavity: a. Peak and transient loading on the reactor pressure vessel that includes forces and moments separated into their X,Y,Z components. b. Provide projected areas used to calculate these loads and the location of these areas shown on detailed plan and elevation drawings. " NOTE: This is a reiteration of the request made in previous question number QO22.3.j."

RESPONSE

a. This item is answered in the response to Question 110.62 () and the revision to S.ection 3.9 of the FSAR. b. The drawings requested were provided under separate cover in response to Question 22.16. O Q22.15-1

y ~ B/B-FSAR O QUESTION 022.25 "FSAR'Section 6.5.2.2 states ' Containment spray injectio'n and caustic education...will continue until...the low-low level alarm of the RWST is annunciated. Containment spray injection and caustic addition may then be terminated, and the operating personnel may transfer the containment spray pumps from the injection to the recirculation made by first closing the motor-operated valves in the suction line from the RWST, the water and caustic lines to'the eductor, and then opening the motor-operated valves in the suction lines from the containment sumps.' State clearly whether transferring the containment spray pumps from the injection to the recirculation mode involves stopping and restarting'the containment spray pumps as implied in the above statement because the valves in the suction line from the RWST are closed before the valves in the suction lines from the containment sumps are opened."

RESPONSE

The containment spray pumps do not have to be stopped when O transferring from the injection mode to the recirculation mode 1 of operation. The response to Question 450.2 from the Accident Analysis Branch provides a more detailed discussion on this subject. i l O Q22.25-1 1

D/C-FSAR ..g, 7 NRC Question No. 040.83 "Your response to Q040.61 is not acceptable. We require that the system include automatic emergency override of the test mode which would require disconnecting the D/G from the bus while it is*on test at full load. Demonstrate proper operation during D/G load shedding including a test of loss of the largest single load and of complete loss of load per RG 1.108 position C.2.a(4)."

Response

We agree that the automatic emergency override of the test mode does require disconnecting the diesel generator from the bus while on test at full load. Therefore, the diesel-generator breaker control logic includes a trip of the breaker upon receipt of a safety injection signal coincident with the diesel generator operating in the test mode. This will leave the diesel generator R idling in the emergency control mode (governor automatically set at 600 RPM and voltage regulator automatically set at 4160 Volts), ready to provide power to the bus in the case of a loss of offsite power. In addition, FSAR section 8.3.1.2 has been revised. i I 1 e O -}}

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