ML20040C828
| ML20040C828 | |
| Person / Time | |
|---|---|
| Site: | Byron, Braidwood, 05000000 |
| Issue date: | 01/09/1982 |
| From: | Tramm T COMMONWEALTH EDISON CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| TASK-2.F.1, TASK-TM NUDOCS 8201290282 | |
| Download: ML20040C828 (95) | |
Text
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C:mm:nw0dth Edison
'O . j O . ore First NItional Piare. Chicago, litenois C Addrsss Flepty to Post Off'ce B3x 767 Chicago, lilinois 60690 t- @
RECEIVED 9 January 9, 1982 E-2 JAN 281982> . -
'Y 'DY Mr. Harold R. Denton, Director Of fice of Nuclear Reactor Regulation Q' . /
U.S. Nuclea r Regulatory Commission Washington, DC 20555
Subject:
Byron Station Units 1 and 2 Braidwood Station Units 1 and 2 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 fifty-nine (59) copies of this Oletter are provided. Fif teen (15) copies of the enclosures are included for your review and approval.
Please address further questions to this office.
Very truly yours, f / . 2[Aes n -
T. R. Tramm Nuclear Licensing Administrator Pressurized Water Reactors Attachment 1
iI 4
O ATTACHMENT A List of Enclosed Information I. FSAR Question Responses New: 010.40 Revised: 022.6 022.55 022.8 022.58 022.12 022.72 022.13 022.79 022.25 040.120 022.29 110.61 022.30 110.63 022.39 221.3 022.50 281.4 022.54 421.22 II. FSAR Text Changes Table 3.2-1 (Classification of Structures)
Table 3.8-1 (Cont. Penetrations)
Table 6.2-9 (Per 022.12)
(~' Revised Pages for T6.2-58 (3)
V) . Subsection 7.6.9 Figure 7.6-6 Table 11.4-2 (VRS)
New Subsection 11.5.2.2.15 (VRS)
Table 11.5-1 (VRS)
Page 15.4-26, Table 15.4-1 (Boron dilution)
Appendix A: Regulatory Guides 1.40, 1.89,' l.100, 1.121 Appendix E: II.F.1 III. Miscellaneous Items l MED Iter 6 Fire Protection Item 8 ICSB Item 7.3.2.7 CSB Item: Process Radiation Isolation l
/ B/B C]
1 MEB Open Item 46 - Functional Capability of ASME Code Class 2 and
(~^ 3 Stainless Steel Elbows i RESPONSE 8 This subject corresponds to agenda items Bll/N17 from the meeting i held with the Mechanical Engineering Branch during the week of May 11, 1981.
I At that meeting, item Bil was completely resolved. Bll deals a with toe functional capability of essential piping designed by 1 Sargent & Lundy. Sargent & Lundy conservatively designs essen-I tial piping using functional capability criteria outlined in i GE Topical Report #NEDO-21985, September, 1978. This report I has been evaluated and approved for use by the Mechanical Engineering Branch of the NRC. Thus, for balance of plant i piping designed by Sargent & Lundy, this item is closed. This
- is documentation in the meeting notes for the May MEB meeting.
. Item N17 is concerned with the functional capability of ASME
. Code Class 2 and 3 stainless steel elbows for Westinghouse designed piping. This item was not resolved at the May
- MEB meeting. However, Westinghouse recently agreed to a
- resolution of this item with the Mechanical Engineering Branch.
- 3 The applicant agrees to resolve this item for Byron /Braidwood
. in the same manner in which it was resolved for Comanche Peak
- (reference the October 15, 1981 letter from H. C. Schmidt of
. Texas Utilities Services, Inc. to Dr. Harold R. Denton of the
. NRC), as described below:
3 In response to the Safety Evaluation Report (SER) open item
- concerning the functional capability of ASME Code Class 2 and 1 3 stainless steel elbows, Commonwealth Edison proposes the
- following stress limits be used to screen these elbows for
- acceptable functional capability:
B 1 2t
+ 2 U1 g
1 1.8 Sy 9 where B1 = (-0.1 + 0.4h) and 01 By 1 0'.5 and B1 =
0.5 for B2 = 1.0
- i . 1.3/(h2/3) for g > 90' 0 L but not h B2* /0.895/(h.912) for'= = 90 less than 1 1.0 for a o =0 I
! Linear interpolation for.O < a g < 9 0 ',
4 where h=t and = o is the angle of the bend.
r 1
O, Other terms are as defined in NC/ND-3600 of Section III of li ASME Ccde. .
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D/B
(' There are no Class 2 and 3 stainless steel elbows or bends
( with D g/t > 50.
Commonwealth Edison proposes to demonstrate that these stress ILnits have been satisfied by reevaluating' the stresses using the above criteria for a random sample of Class 2 and 3 stain-less steel elbows in the piping systems listed below:
- a. Safety injection
- b. Chemical and volume control
- e. Containment spray The sample will encompass the full temperature and size range of stainless steel piping. Twdnty-five percent of elbows in each system will be sampled. For those systems where 25% of the elbows is less than five, a minimum of five elbows will be evaluated.
O 8
O
B/B O
. ... e. e.
Response to fire protection open item #8 Byron and Braidwood Stations operate with six rotating shifts.
To meet the requirement of four drills / year / brigade there will be a minimura of 24 drills / year. Using an average of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> per drill. from the time the alarm sounds until the time an operator can return to the job, there would be a minimum of 120 mr n-hours / year that station operation would be affe cted.
The minimun of 120 hours0.00139 days <br />0.0333 hours <br />1.984127e-4 weeks <br />4.566e-5 months <br /> comes from the fact that any other members of a fire brigade who would happen to be onsite would also respond to the alarm. This is a large number of hours to neglect operating duties in lieu of fire fighting drills. Two drills per year per brigade with one mandatory will only interrupt 60 hours6.944444e-4 days <br />0.0167 hours <br />9.920635e-5 weeks <br />2.283e-5 months <br /> of operation that is more appropriate.
Another point of consideration is the possible development of an apathetic outlook toward the drills due to the frequency.
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B/B f( ICSB OPEN ITEM 7.3.2.7 Auxiliary Feedwater System Flow Indication The power source for the Train "B" Auxiliary Feedwater Flow instruments will be changed from ESF Bus 12 to a Battery Dacked supply. This change will allow the operator to maintain control room indication of Train "B" flow and determine if the diesel driven pump should be-turned off to eliminate the possibility of an over-cooling transient. Based upon the above, the applicant does not believe that a Battery Backed supply is re-quired for the Train "B" flow control valves.
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B/B CSB Item: Process Radiation Isolation O
The process radiation system has been equipped with two isolation valves in series outside of containment to provide isolation for Penetration P-52. This was done on these 1 inch lines to insure that the valves are maintainable and operable during plant operation.
The valves are located very close to the containment. The line is not postulated to break or crack due to the very low stress level of this part of the system. This arrangement is acceptable per SRP 6. 2. 4,Section II . 3.d and SRP 3. 6. 2.
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s_s R/B-FSAR TABLE 3.2-1 (Cont'd) fr PRINCIPAL STRUCTURES, SAFETY QUALITY SYSTEMS, AND COMPONENTS CATEGORY GROUP ELECTRICAL
$. AP - Auxiliary Power 480 V a-c and abcve'
- a. All a-c auxiliary power equipment necessary for Category I items to perform their safety ,
function including ESF switchgear, MCCS, transformers, buses and cables - which in-clude the following: 1 N/A IE
- b. Other a-c auxiliary power equipment including unit auxiliary transformers, system auxiliary transformers and their low voltage connections to the ESF switchgear II N/A Non-IE
- 6. AR - Area Radiation Monitoring II N/A Non-IE
- a. Safety-Related Equipment I N/A IE
- b. All other equipment II N/A Non-IE
- 7. AS -Aux 11
- ory steam (Including Heating Boiler) II D Non-ZE S. BR - Boron Thermal Regeneration
- a. Moderating Heat Exchanger I C N/A
- b. Letdown Chiller Heat Exchanger Tube Side I C N/A Ns ,s/ .Shel) Side II D N/A
- c. Letdown meheet Heat Exchanger Tube Side I B !a/A Shell Side I C N/A
- d. Chiller Pumps II D N/A
- e. Chiller Surge Tank II D N/A
- f. Chiller II D Noa-IE
- g. Thermal Regeneration Domineralizers I C N/A
- h. Chiller Pump Motors II N/A Non-IE i
- 9. CB - Condensate Booster II D Non-IE
- 10. CC - Component Cooling'
- a. Component Cooling Heat Exchangers I C N/A
- b. Component Cooling Pumps I C N/A
- c. Component Cooling Surge Tanks I C N/A
- d. Component Cooling Pump Motors I N/A IE
- 11. CD - Condensate (Excluding Condensate Booster)
(Including Makeup and Overflow) II D Non-1E
- 12. CF - Chemical Feed and Handling (Hydrazine. Asconia , Morpheline, Sulphite and other miscellaneous chemicals) II D Non-IE
- 13. CO - Carbon Dioxide (Includes Fire Protection and Non-IE Generator Purge) II D
- 14. CO - Code call. Public Address, Telephones, Gate TV,
- Gate operators, Evacuation Alarm, Station Security, etc. II N/A Non-IE.
- 15. CS - Containment Spray *
- a. Containment Spray Pumps I B N/A
. b. Containment Spray Pump Motor I N/A IE l c. Spray Additive Tanks I B N/A l d. Spray Eductors I B N/A N
3.2-5
E/B-FSAR 4 e
U TABLE 3.2-1 (Cont'd) s' PRINCIPAL STRUCTURES, SAFETY QUALITY g w\ SYSTEMS, AND COMPONENTS CATEGORY GROUP ELECTRICAL
\
\ws/
) 22. DO - Diesel Fuel 011* (Supply and Transfer) r Diesel Fuel Oil Storage Tanks I C N/A
- a. I C N/A
- b. Diesel Fuel Oil Transfer Pumps
- c. Diesel Fuel Oil Transfer Pump Motors, Inside I N/A IE Diesel Fuel Oil Transfer Pump Motors, Outside II D ,
d.
II D Non-IE
- 23. DV - Feedwater Heater Miscellaneous Drains and Vents EC - Chemical Cleaning, Equipment and Pipe II D Non-IE }
- 24. r
- 25. EF - Engineerca safety Features Logic Testing and Actuation
- I N/A IE I
II D Non-IE
- 26. EH - Turbine EHC
- 27. EM - Environs Monitoring (Including Strong Motion - Non-IE Seismic Instrumentation) II N/A II D Non-IE ,
- 28. ES - Extraction Steam II N/A Non-IE
- 29. EW - Welder Outlets
- 30. FC - Fuel Pool cooling and Clean-Up (See Table 9.1-2)
Spent Fuel Pit Heat Exchanger I C N/A
- a. C N/A I
- b. Spent Fuel Pit Pump II N/A Non-IE
- c. Spent Fuel Pit Pump Motor D N/A Skimmer Pump II
- d. N/A Non-IE Skimmer Pump Motors II
- e. D N/A Spent Fuel Pit Filter II
- f. II D N/A j
- g. Spent Fuel Pit Demineraliser C N/A -
- h. Refueling Water Purification Pump (one pump only) I Non-IE '
- 1. Refueling Water Purification Pump Motor II N/A
- 31. FH - Fuel Handling and Transfer Nuclear I N/A N/A
- a. New Fuel Storage Racks M48 4 M
J
['N i b. Spent Fool Storage Racks I N/A 4 tion-IE k_,/ c.
d.
Fuel Handling Building Crane Manipulator Crane I
I N/A N/A Non-1E Non-IE
- e. Spent Fuel Bridge Crane II N/A Non-IE
- f. New Fuel Elevator
- g. Fuel Transfer System Fugl Transfer Tube & Flange I N/A N/A Conveyor System N/A (Fuel Building Side) I N/A II N/A N/A j Remainder of System
I C N/A Seismic Qualified Areas II D N/A Centrifugal Pumps FW - Main Feedwater l
33.
Outside Containment up to Isolation Valve II D Non-IE e.
- b. Inside Containment up to and including C IE
- I
' Isolation valve I 34. GC - Generator Stator Coolinq Non-IE (Including Excitation Cubicle Cooling) II D CD - Grounding and Cathodle Protection II N/A Non-IE 35.
II D Non-IE 3 6 .* GS - Turbine Gland Seal Steam
- 37. GW - Radioactive Waste Gas C Non-IE (Excluding Off Gas) including the following: I l
I C
- a. Waste Gas Compressor I C
- b. Gas Decay Tanks D II
- c. Gas Analyzer 30.-HC - Holsts. Cranes, Elevators, and Manlifts
- % (All except Fuel Handling and Transfer System)
I N/A Non-IE
- a. Containment Building Crane f ( ,/ b. All other equipment II N/A Non-IE D Non-IE
- 39. HD - Feedwater Drains - Turbine Cycle II 3.2-7 I
as R/B FSAR f<-s TABLE 3.2-1 (Cont'd) i N'w PRINCIPAL STRUCTURES. SAFETY QUALITY SYSTEMS, AND COMPONENTS CATEn0RY GROUP ELECTRICAL
- 60. PM - Main Control Room Panels
- a. For Safety Related Equipment I N/A IE
- b. Others II N/A Non-IE
- 61. PR - Process Radiation Monitoring
- a. Safety Related Eouipment I C It
- b. All other equipment II O Non-IE
- 62. PS - Process Sameling Primary & Secondary Systam Including Chiller Ecutpment (Samp. Cond.
6 Monitoring Assemblies)
- a. Primary Sampling Remote Air Operated Valves, I C fun-!E
- b. Primary Sampling Containment Isolation Valves, I e !E All other equipment II D IMn-II:
c.
- 63. PW - Primary water II D Non-IE
- 64. RC - Reactor Coolant System (Not including Pressurizer System)
- a. Reactor Vessel I A N/A
- b. Steam Generator Tube Side I A N/A Shell Side I B N/A
- c. Reactor Coolant Loop Stop Valves I A M/A
- d. Pressure Boundary Piping and Fittings I A N/A
- e. Reactor Coolant Pump RCP Casing I A N/A Main Flange I A N/A Thermal Barrier I A N/A Thermal Barrier Heat Exchanger I A N/A
- 01 Seal Housing I A N/A 02 Seal Housing I B N/A Pressure Retaining Bolting I A N/A
- f. RCP Motor I N/A Non-IE
- 65. RD - Control Ro t Drive I N/A Non-IE
- a. Full Length *RDM Housing I A N/A
, b. Part Length CRDM Housing I A N/A
- c. CRDM Head Adapter Plugs I A N/A
- d. Thermal Sleeves I B N/A
- e. Control Rod Drive Mechanism I N/A N/A
- 66. RE - Reactor Building & Containment Equipment Drains to Radwaste (Including Reactor Coolant Drains and Pumps)
- a. Reactor Coolant Drain Tank II D N/A
- b. Reactor Coolant Drain Pumps II D N/A
- c. Reactor Coolant Drain Pump Motors II H/A Non-IE
- 67. RF - Reactor Building & Containment Floor Drains to Dadwasto (Including Sump Pumps) II D Non-!E
- 69. RH - Residual Heat Removal Pumps *
- a. Residual Heat Removal Pump I B N/A
- b. Residual Heat Removal Pump Motors I N/A IE
- c. Residual Heat Exchangers Tube Side I B N/A Shell Side I C N/A
- 69. RP - Reactor Protection
- I N/A IE
- 70. RY - Reactor Coolant Pressuriter System
- a. Pressuriser I A N/A
- b. Pressurizer Relief Tank II D N/A
- c. Pressurizer Heaters II N/A Non-IE Pressurizer Safety Valves I A N/A
/'"Sg d.
- e. Pressurizer Power-Operated Relief Valves I A N/A
{'w 3.2-9
m
.- g 4
I l B/B-FSAR TABLE 3. 2-1 (Cont'd) 7-~ .
l ?
\ f PRINCIPAL STRUCTURES, EAFEty OUALITY SYSTEMS, AND COP.PONENTS CATEGORY GROUP Etrc',pICAL
- d. Cooling coils I C nA
- e. Supply Filters ! N/A * 'A
- f. Controls & Instrumentation I N/A It
- g. E1cetric Heaters II N/A !: n-It
- h. Humidifiers !! N/A t,-I t
). Charcoal Falter Houginq ! N/A
- r,- t t
- k. Ductwork anal D.nepers ! N/A It
, m. Utility Exhaust Fans II N/A 'on-It
- 89. VD - Diesel-Cenerator Room Ventilatinn' I N/A IE
- a. Diesel-Generator Room Fan Motor I l 90. VE - Misc. Electric Equipment Room Ventilation
- I N/A IE
- 91. VF - Containment Building, Auxiliary Building Filtered Vents II D Non-IE
- 92. VH - Pumphouse Ventilation II N/A Non-IE
- 93. VI - Radwaste 6 Remote Shutdown Control Room HVAC II N/A Non-IE
- 94. VN - Containment Building, 6 Auxiliary Building Non-Filtered Vents II D Non-IE
- 95. VJ - Machine Shop Ventilation II N/A Non-IE
- 96. VK - Switchyard Relay House HVAC II N/A Non-IE j 97. VL - Laboratory HVAC II N/A Non-IE
- 98. VP - Primary Containment ventilation *
- a. Reactor Containment ran Coolers I N/A N/A
- b. Reactor Containment Fan Coolers Motors I N/A IE
/N c. CRDM Exhaust Fans I N/A N/A
/ I' d. CRDM Exhaust Fans Motors II N/A Non-IE k ' ~ ' , e. Reactor Cavity Vent Fans II N/A N/A
- f. Reactor Cavity Vent fans M'o tors if N/X on-I'6
- g. CRDM Booster Fans II N/A N/A
- h. CRDM Booster Fans Motors II N/A Non-IE
- 1. Manipulator Crane Fans II N/A N/A
- j. Manipulator Crane Fans Motors II N/A Non-IE
- k. Containment Charc. Filter Unit II N/A N/A
- 1. Containment Charc. Filter Fan II N/A N/A
- m. Cont. Charc. Filter Onit ran Motor II N/A Ncn-IE
- n. RCFC Ess. Service Water Coils I C N/A
- o. RCFC Chill Water Coils II D N/A
- p. Ductwork Dampers and Supports I N/A N/A l
- 99. VQ = Primary Containment Purge
- a. Post-Loca Purge Filter Unit II N/A N/A
- b. Purge Supply & EXH Fan II N/A N/A
- c. Purge Supply 6 EXH Fan Motor II N/A Non-IE
- d. Purge Supply & EXH Filters II N/A N/A
- e. Purge Controls & Instrumentation II N/A Non-IE
- f. Post-Loca Purge Contr. 6 Instr. II N/A Non-IE 100. VS - Service Building, HVAC II N/A Non-IE 101. V7 - Turbine Building, HVAc II N/A Non-IC 102. VV - Miscellaneous ventilation II N/A Non-IE 103. VW - Radwaste Facility Ventilation II N/A Non-IE 104. VX - Switchgear Neat Removal *
- a. For Class IE Switchgear Rooms' I N/A 11
- b. Other II N/A Non-IE 105. WE - Aux. Bldg. Equip. Drain Radwaste Reprocessing
,_ 6 Disposal II D Non-IE
/ 5 t,
1 106. WP - Auxiliary Building Floor Drain Radwaste
'm/ Re recessing & Disposal II D Non-IE
'qu pment Drain Tank II D Non-IE 3.2-11
ms .
\
7 B/B-FSAR TABLE 3.2-1 (Cont'd)
PRINCIPAL STRUCTURES. SAFETY QUALITY SYSTEMS, AND COMPONENTS CATEGORY GROUP ELECTRICAL I II N/A Non-IE i 107. WC - Cland water j 108. wL - Raw water Including Cooling Lake Makeup .II D Non-IE and Blowdown 109. wM - Makeup Demineralizer ,
D Non-IE (Including Effluent and Flushinal II -
110. Wo - Chilled water
- I C IE
- a. Control Room System II D Non-IE
- b. Remainder of System i II D Non-IE 111. WS - Non-Essential Service water II D Non-IE 112. ww - well water t 113. wx - Solid Radwaste Reprocessing & Disposal (Wet and Dry) (Including Drumming and Resin b Removal)
- a. Blowdown Demineraliters II O N/A
- b. Blowdown Profilters . II D -N/A
- c. Blowdown Monitor Tanks IT D N/A
- d. Blowdown Monitor Tank Pumps II D N/A Non-!E
- e. Blowdown Monitor Tank Pump Motors II N/A
- f. Spent Resin Storage Tank I C N/A
- g. Laundry Drain Tanks __ II D N/A 1 h. Release Tank II D N/A
- 1. Radwaste Evaporators II D Non-IE
- j. waste Evaporator Monitor Tanks II D N/A
- k. Spent Resin Flushing Pump II D N/A
- 1. Spent Resin Flushing Pump Motor II N/A Non-IE
- m. Liquid Radwaste Filters II D N/A-l
114. WY - Laundry Equipment & Floor Drains Radweste Non-IE Reprocessing 6 Disposal II D II D Non-IE 115. w2 - Chemical Radwaste Reprocessing & Disposal II D N/A
- a. Chemical Drain Tank I N/A IE 116. Electrical Penetration 117. Interconnecting Cable between Safety Category I I N/A IE Electrical Equipment 4
118. All other electrical equipment shown on II N/A Non-IE Figures 8.3-1 and 8.3-2 119. Reactor Vessel or Core-Related I N/A N/A
- a. Reactor Vessel Shoes and Shims
- I N/A N/A
- b. -Control Rod Drive Mechanism Seismic Support Tie Rod Assemblies I N/A N/A
- c. Control Rod Guide Tubes I N/A N/A
- d. Reactor Vessel Internals I N/A N/A
- e. Full Length Control Rod Clusters II N/A N/A
- f. Burnable Poisons II N/A N/A
' g. Primary and Secondary Sources I N/A N/A
- h. Irradiation Sample Holder N/A II N/A
- 1. Control Rod Drive Mechanism Dummy can Assemblies i i 1
3.2-12
TABLE 3.8-1 CONTAINMENT PENETRATIONS VERTICAL HORIZONTAL PENETRATION WALL ELEVATION SKEW SKEW NUMBER SIZE THICKNESS (ft-in.) AZIMUTH ANGLE ANCLE CESCRIPTION P-1 24.0 0.688 407 0 97*-30 ft 0* O' Containment spray P-2 16.0 0.844 407 0 75'-00 ft 0* 0* Spare P-3 16.0 0.844 407 *) 72'-30 ft 0* 0* Spare P-4 28.0 0.365 407 0 67*-30 ft 0* 0* Test connection P-5 22.0 0.875 407 0 65'-00 ft O' 0* Chilled water P-6 22.0 0.875 407 0 60*-00 ft O' 0* Chilled water !
P-7 30.0 1.000 403 0 105*-00 ft 0* 0* Essential service water P-8 22.0 0.875 403 0 102*-30 ft 0* 0* Chilled water
< P-9 30.0 1.000 403 0 100*-00 ft O' 0* Essential service water P-10 22.0 0.875 403 0 97*-30 ft 0* O* chilled water P-11 10.0 0.365 403 0 95*-00 ft 0* O' Reactor b1dg. & cont. l Equipment drains to radwaste p-12 16.0 0.844 403 0 72*-00 ft 0* 0* H2 rionitoring system l p-13 16.0 0.844 403 0 67*-30 ft 0* 0* OFF gas system 1.000 l P-14 30.0 403 0 62*-30 ft 0* 0* , Essential service water e P-15 30.0 1.000 403 0 57*-30 ft O' 0+ ' Essential service water y 7 P-16 24.0 0.688 399 0 75*-00 ft O' 0* Containment spray e
" P-18 16.0 0.844 399 0 65'-00 ft 0* 0+ Spare @
g 1 p-19 16.0 0.844 399 0 60*-00 ft O' 0* Spare >
" P-21 16.0 0.844 395 0 127*-30 ft O' 0* Component cooling "
P-22 10.0 0.365 395 0 122*-30 ft 0* O' Component cooling P-23 16.0 0.844 395 0 117*-30 ft O' 0* OFF gas system P-24 16.0 0.500 395 0 112*-30 ft 0* 0*
l Component cooling P-25 16.0 0.500 395 0 107*-30 ft O' 0' Component cooling P-26 12.0 0.406 395 0 102*-30 ft o' 0* Safety injection P-27 10.0 0.365 395 0 97*-30 ft 0* 0* Reactor coolant pressurizer l P-28 8.0 0.322 395 0 72*-30 ft O' 0* Chemical and volume control l P-29 16.0 0.844 395 0 67*-30 ft 0* 0* Spare F-30 10.0 0.365 395 0 62*-30 ft 0* 0* Makeup demineralizer P-31 16.0 0.844 395 0 57*-30 ft O' 0' H2 monitoring system P-32 10.0 0.500 191 'O 125'-00 ft O' 0' Fuel pool cooling sad cleanup P-33 14.0 0.375 391 0 120*-00 ft 0* O' chemin i aad volume control P-34 12.0 0.406 391 0 115*-00 ft 0* 0* rire protec tion P-36 22.0 0.875 391 0 105*-00 ft O' 0' Sparc P-37 14.0 0.375 391 0 100*-00 ft O' 0* Chemical und volume control P-39 80 0.322 391 0 75'-00 ft 0* O' Instrument air supply P-41 12.0 0.375 391 0 65'-00 ft 0* O' Chemical and volume control P-42 22.0 0.875 391 0 60*-00 ft O' 0* Spare P-43 16.0 0.844 391 0 57*-30 ft O' 0* Spare P-44 10.0 0.365 387 0 127*-30 ft O' 0* Reactor coolant P-45 16.0 0.844 387 0 122*-30 ft O' 0* Spare l \
/
g TABLE 3.8-1 (Cont'd)
VERTICAL HORIZONTAL PENETRATION WALL ELEVATION SICEW SKEW NUMBER SIZE THICKNESS (ft-in.) AZIMUTH ANGLE ANGLE t'ESCRIPTION P-47 10.0 0.365 387 0 112*-30 ft O. 0* Waste disposal l P-48 10.0 0.365 387 0 107*-30 ft 0* 0* Component cooling P-49 22.0 0.875 387 0 102*-30 ft O. 0* Spare l P-50 24.0 0.688 387 0 97*-30 ft 0* 0* Safety injection P-51 24.0 0.688 387 0 72*-30 ft 0* 0* Safety injection P-52 16.0 0.844 387 0 67*-30 ft O. 0* Process radiation monitoring 0* l P-53 14.0 0.375 387 0 62*-30 ft O.
0*
Chemical and volume control P-54 22.0 0.875 387 0 57*-30 ft 0* Spare P-55 10.0 0.365 383 0 125*-00 ft 0 0* Safety injection P-56 10.0 0.365 383 0 120*-00 ft 0 0* Service air P-57 8.0 0.322 383 0 115*-00 ft 0* 0* Fuel pool cooling and cleanup P-59 16.0 0.375 383 0 105'-00 ft 0 0* Safety injection P-60 14.0 0.375 383 0 100*-00 ft O. 0* Safety injection P-61 16.0 0.375 383 0 75*-00 ft O. O' Spsre P-63 16.0 0.844 383 0 65*-00 ft 0* 0* Spare P-64 16.0 0.844 383 0 60*-00 ft O. 0* Spare P-65 10.0 0.365 379 0 127*-30 ft O' 0* Reactor b1dg. and cont. drn. as 0*
to rad. equip. drains lD P-66 24.0 0.688 379 0 122*-30 ft O. Safety injection 4 s P-68 24.0 0.688 379 0 112*-30 ft O. 0* Residual heat removal m O P-69 P-70 16.0 12.0 0.844 0.375 379 379 0
0 107*-30 102*-30 ft ft 0
0*
0*
0*
OFF gas system Process sampling l$
P-71 14.0 0.375 379 0 97*-30 ft 0* 0* Chemical and volume control P-72 16.0 0.844 379 0 72*-30 ft 0* O' Spare P-73 16.0 0.844 379 0 67*-30 ft C' 0* Safety injection P-74 16.0 0.844 379 0 62*-30 ft 0* Spare P-75 24.0 0.688 379 0 57*-30 ft 0 0 .* O' Residual heat removal P-76 34.0 1.000 390 0 Note 1 -9*-30 ft 0* Feedwater P-77 54.0 1.375 386 6 -4*-45 ft O' Main steam Note 1 P-78 54.0 1.375 386 6 -4*-45 ft O' Main steam Note 1 P-79 34.0 1.000 390 0 '+8*-15 ft 0* Feedwater Note 1 0*
P-80 12.0 0.375 388 0. Note 1 +9*-30 ft Steam generator blowdown P-81 12.0 0.375 386 6 Note 1 9*-30 ft 0* Steam generator blowdown P-82 12.0 0.375 385 0 9*-30 ft 0* Steam generator blowdown Note 1 P-83 12.0 0.375 383 6 Note 1 9*-30 ft 0* Steam generator blowdown P-84 34.0 1.000 390 0 Note 2 -9*-30 ft 0* Feedwater P-85 58.0 1.500 386 6 Note 2 -4*-45 ft 0* Main steam P-86 58.0 1.500 386 6 Note 2 r4*-45 ft 0* Main steam P-87 34.0 1.000 390 0 Note 2 +8*-15 ft 0* Feedwater P-88 12.0 0.375 388 0 0* Stean generator blowdown Note 2 +9*-30 ft P-89 12.0 0.375 386 6 Note 2 +9*-30 ft 0* Steam generator blowoown P-90 12.0 0.375 385 0 Note 2 +9*-30 ft 0* Steam generator blowdown
/
v
TABLE 3.8-1 (Cont'd)
VERTICAL HORIZONTAL PENETRATION WALL ELEVATION SgEw SKEW NT.;MB ER SIZE THICFNESS (ft-in.) AZIMUTH ANCLE ANGLE DESCRIPTION P-91 12.0 0.375 383 6 Note 2 +9*-3C ft 0* Steam generator blowdown .
P-92 28.0 0.375 * -
0* O' Safety injection, cont, spray P-93 28.0 0.375 * -
0* 0* Safety injection, cont, spray P-94 16.0 0.500 474 0 108*-00 ft 0* 0* Mini-flow purge exhaust P-95 60.0 1.000 462 0 123'-00 ft 0* O' Containment purge exhaust P-96 14.0 0.375 462 4 132*-45 ft 0* 0* Mini-flow purge supply P-97 60.0 1.000 462 4 O' 139*-00 ft 0* Containment purge supply P-98 24.0 0.688 392 6 Note 3 -4'-10 ft o' Fuel transfer tube P-99 16.0 0.844 390 9 Note 1 - 3*- 30 ft 0* Feedwater P-100 16.0 0.844 390 9 Note 1 +3*-30 ft 0* Feedwater P-101 16.0 0.844 390 9 Note 2 -3*-30 ft O' Feedvater P-102 16.0 0.844 390 9 Note 2 +3*=30 ft 0* Feedwater E-1 24.0 0.500 435 0 120*-15 ft 0* 0* Reactor coolant pump p E-2 E-3 18.0 0.438 439 439 3
3 123*-45 ft 0 0* CRD Fan y E-4 12.0 0.406 127'-15 ft 0 0* Misc. ESF instrumentation e
." E-5 12.0 12.0 0.406 0.406 439 439 3
3 130*-45 134'-15 ft ft 0*
0*
O' 0*
Reactor cont. fan cooler Misc. control 15 e I >28 e
E-6 12.0 0.406 439 3 137'-45 ft O' 0* Misc. control E-7 12.0 0.406 439 3 141*-15 ft 0* 0* Process instr.
E-8 12.0 0.406 439 3 120*-15 ft O' 0* Misc. power E-9 '
12.0 0.406 435 0 123*-45 ft 0* 0* Pressurizer heater E-10 12.0 0.406 435 0 127*-15 ft 0* O' Pressurizer heater E-11 12.0 0.406 435 0 130*-45 ft 0* O' Misc. power E-12 12.0 0.406 435 0 134*-15 ft O' 0' Misc. control E-13 12.0 0.406 435 0 137*-45 ft 0* 0* Misc. instrumentation E-14 12.0 0.406 435 0 141*-15 ft O' 0* Neu.ron monitoring E-15 12.0 0.406 421 9 0*
E-16 12.0 120*-15 ft 0* Reactor cont. fan cooler !
0.406 421 9 123'-45 ft 0* O' Pressurizer heater E-17 12.0 0.406 421 9 127*-15 ft 0* 0* Pressurizer heater E-18 12.0 0.406 421 9 130*-45 ft 0* 0* Misc. control E-19 12.0 0.406 421 9 134*-15 ft 0* 0* Misc. power l E-20 24.0 0.500 421 9 137*-45 ft 0* 0* Reactor coolant pump E-21 18.0 0.438 421 9 141*-15 ft O' 0* Crd fan, Itg. and weld recep., {
cav. fan E-22 12.0 0.406 417 6 120*-15 ft 0* C* Neutron monitoring E-23 12.0 0.406 417 6 123'-45 ft O' 0* Incore-flux mapping det. instru-E-24 mentation 12.0 0.406 417 6 127*-15 ft 0* 0*
E-25 12.0 0.406 417 6 Process instrumentation E-26 12.0 130*-45 ft 0* 0* Misc. instrumentation E-27 0.406 417 6 134*-15 ft O' 0' Spare 12.0 C.406 417 6 137*-45 ft 0* 0*
E-28 18.0 Misc. ESP instrumentation l 0.438 417 6 141*-15 ft 0* O' Crane feed E-29 12.0 0.406 439 3 0* 0*
E-30 12.0 0.406 439 3 174*-45 ft 0* 0*
Misc. instrumentation E-31 178*-15 ft Spare E-32 12.0 0.406 439 3 181*-45 ft o' 0* Control rod drive unit 12.0 0.406 439 3 185*-15 ft O' 0' Control rod drive unit a
a e e e
.~
, m ,, , ;
/ $
[ 4 \
U
/
! \ /
k' v )' i v w/
TABLE 3.8-1 (Cont'd)
HORI2ONTAL VERTICAI. SKEW PENETRATION WALL ELEVATION SKE4 ANGLE DESCRIPTION NUMBER SIZE THICKNESS (ft-in.) AZIMUTH ANGLE 0* Reactor coolant pump E-33 24.0 0.500 439 3 188*-45 ft O.
12.0 439 3 192*-15 ft O. O Spare E-34 0.406 O'
E-35 12.0 0.406 439 3 195'-45 ft 0 Process instrumentation 12.0 0.406 435 0 174*-45 ft D. O' Spare E-36 O' E-37 12.0 0.406 435 0 178*-15 ft 0 0*
Control rod drive unit E-38 12.0 0.406 435 0 181*-45 ft O. Control rod drive unit 435 0 185'-15 ft 0' Control rod drive unit E-39 12.0 0.406 12.0 435 0 188*-45 ft O O '.
0* Reactor cont. fan cooler E-40 0.406 O' l E-41 12.0 0.406 435 0 192*-15 ft 0* Misc. power 12.0 0.406 435 0 195'-45 ft 0 O' Neutron monitoring E-42 0*
E-43 12.0 0.406 421 9 174'-45 ft o. O' Misc. control E-44 12.0 0.406 421 9 178*-15 ft 0* Misc. control O. 0* Misc. power E-45 12.0 0.406 421 9 181'-45 ft 0* Misc. instr.
E-46 12.0 0.406 421 9 185*-15 ft 0 421 9 188*-45 ft O. 0* Reactor coolant pump E-47 24.0 0.500 i,,E-48 12.0 0.406 421 9 192'-15 ft 0 O, Misc. power R
,E-49 e E-50 12.0 12.0 0.406 0.406 421 417 9
6 195'-45 174*-45 ft ft O.
0*
O 0,
Reactor cont. fan cooler Neutron monitoring l?9 0; E-51 12.0 0.406 417 6 178*-15 ft 0* O' Process instrumentation $
0.406 417 6 181*-45 ft 0* O Spare w E-52 12.0 0.406 417 6 195*-15 ft 0* 0* Spare E-53 12.0 417 6 188'-45 ft 0* 0* Spare E-54 12.0 0.406 0*
E-55 12.0 0.406 417 6 192*-15 ft 0 Spare 417 6 O. 0* Spare E-56 12.0 0.406 195*-45 ft 1-1 6.0 0.432 427 6 120'-00 ft O' 4 4,'-45 ft Containment pressure monitoring I-2 6.0 0.433 452 6 139*-00 ft 0* 45 ft 4,45 ft.
Containment pressure monitoring I-3 6.0 0.432 427 6 137'-45 ft O' Containment pressure monitoring I-4 6.0 0.432 452 6 118'-00 ft O' 4, -4 5 f t 4,-45 ft containment pressure monitoring I-5 6.0 0.432 427 6 141'-15 ft 0* Dead wt. test stand Note 1: Unit 1 - See Plan 8, 5-936 and M-197-1.
Unit 2 - See Plan 12, 5-936 and M-197-5.
Note 2: Unit 1 - See Plan 7, 5-936 and M-197-1.
Unit 2 - See Plan 11, 5-936 and M-197-5.
Note 3: Unit 1 - See Plan 9, 5-936 and M-197-1.
Unit 2 - See Plan 10, 5-936 and M-197-5.
I
_ . _ _ _ _. I
(p) v'
'P\
t i B/B-PSAR t are provided through use of two PORV's to mitigate any poten-
- tial pressure transients. The protection system is required only during low temperature water solid operation when it
- is manually armed and automatically actuated.
- 5.2.2.11.1 System Operation i The pressurizer power-operated relief valves are both supplied
. with actuation logic to ensure that an automatic and indepen- l
- dent RCS pressure control backup feature is provided for
, -tho operator during low temperature operations.
. This system provides the capability for additional RCS
. inventory letdown, thereby maintaining RCS pressure within
- allowable limits. Refer to Subsections 5.4.7, 5.4.10, 5.4.13,
, 7.6.9, and 9.3.4 for additional information of RCS pressure r and inventory control during other modes of operation.
The basic function of the system logic is to continuously monitor RCS temperature and pressure conditions whenever plant operation is at low temperatures (%350' F). An auctioneered system temperature will be continuously converted to an allowable pressure and then compared to the actual RCS pressure. The system logic will first annunciate a main board alarm whenever the measured pressure approaches within a pre-g- si determined amount, thereby indicating a pressure transient is occurring. On a further increase in measured pressure an s ,/ actuation signal to the power operated relief valves when required will prevent pressure temperature conditions from exceeding allowable limits.
5.2.2.11.2 Evaluation of Low Temeerature Overpressure Transients pressure Transient Analyscs
! Section III, Appendix G of the ASME Code, establishes guide-lines and limits for RCS pressure primarily for low temperature conditions (<350' F). The relief system discussed in Subsection 5.2.2.11.1 satisfies these conditions as discussed in the following paragraphs.
- Transient analyses were performed to determine the maximum
- pressure for the postulated worst case mass input and heat input events.
2 The worst case mass input transient analysis was performed
- assuming the inadvertent actuation of two charging pumps, a which, in combination with letdown isolation, pressurizes a the RCS. The results show that at an RCS temperature of 70' F a maximum pressure of 525 psig will be reached.
l 5.2-6a i
O) i x_ -
l l
4 1
j 1 .
.!, i, . .
B/B-FSAR t
1 4
! TABLE 6.2-9
) !' O.942 FT2 SPLIT RUPTURE - 30% POWER - STEA.58LINE STOP
- ' VALVE FAILURE / ACCIDENT SECUENCE l !
- i TIME OF OCCURRENCE 1
EVENT (sec) i Main feedwater isolation 25.7 i 38.5
- Steamline isolation Start fan coolers 58.74 5 t Containment peak temperature 59.0, f Start sprays 137.4
)
i 5 Steam generator dryout 595.0
. Containment peak pressure 387.0 I
i 4
. r 9
t I
i I
i 6.2-74 i
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$4 43 Air 1/3 n-205-3 IVQO14 Outside Tee NIN. Globe N Closed Closed Closed N/A N/A N/A R R N/A 4 i
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$6 69 Aar & My 3 UES M-4 7-2 OfCC62 OUTSIDE TES egne But. F 1. E CLCSrD Closed CL y t D AS 15 y g ,, gg g
$6 2) Air & My 3 VES M-41-2 0 0 ~.0 6 3 'OUTSIDE TES 54.0 Bat. Fi .MO CLosrD C;.OSTD CLo3ED AS 15 T A gg 1 56 23 Air 6 My 3 VES 47-2 OfG064 OUTSIDE TEE mtwe Sut. F1 m CLOSED CLohED CLOFED AS 15 9 g M,g gg g
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.T A 3pt IE 8 56 52 Air Mi 7s*1g+ IPR 001A Ou'tside Yes 1.4 Globe AO/s. ,' Open ' Closed Closed closed PW*e 11 Globe ho/a Open Closed Closed closed T A RN IE 8 RAS:?" TT 56 52 Air #1 M-78-1g IPR 0019 Outside Yes 3.5' 7 A RM 5 Air n-78-10 1PR03] Outside Yes 2.3 G1dbe AO/S lOpen Closed Closed closed IE i
(IVTURE) 56 52 ti Check N/A 'N/A N/A N/A N/A N/A N/A N/A N/A 5 56 52 Air' tL M-78-10 1 PR0 3d Inside Yes N/A' Globe AO/S Open Closed Closed Closed T RM M 1E FYL V EN 56 12 H + Air 3/8 M-68-7+ IPS22 A Outside Yes T RM M IE 3/8 M-68-7 1PS221A Outside Ya4 Globe AO/S Open Closed Closed Closed T M 1E fD 'IiO3 56 12 Mf+ Air M-68-7 1PS23d A Outside Y*9 Globe AO/S Open Closed Closed Closed RM (FL"IFI) 56 12 .g2+ Air 3/ 6 IPS23l A Inside Ye9 N/A Check N/A- N/A N/A N/A N/A N/A N/A N/A N/A 56 12 H2+ Air 3/8 M-68-7 T RM M
'56 31 H2+ Air 3/8 M-68-7 1PS221 B Outside 'e' Globo AO/S. Open closed Closed Closed T RM M IE 56 31' 3/8 M-68-7 1PS22' B Outside Ye9 Clobe AO/S Open Closed Closed Closed M IE H2+ Air 1PS231'8 Outside Yes Globe AO/S 'Open Closed Closed Closed T RM IE 56 3L H2+ Air 3/8 M-68-7 N/A N/A N/A N/A N/A N/A 56 3L H2+ Air 3/8 M-68-7 1PS23' B Inside '*9M/A Check N/A N/A N/A f
! +Dravian ?epicting these systems are being developep. These sysiers will comly with CDC-56.
I e
e
- _ ,_m ._
g
() B/B-ESAR During plant shutdown, the accumulator valves are in a closed position.
When the RCS pressure is above the SI Unblock Pressure, an alarm will sound in the main control rocm for any accumulator isolation valve not fully open as indicated by the valve stem limit switch.
7 7.6.7 Switchover 'from Iniection to Recirculation 3 The details of achieving cold-leg recirculation following safety injection are given in Subsection 6.3.2.8 and on Table 6.3-7.
Figure 7.6-4 shows the logic which will be used to open the sump valves automatically.
7.6.8 Reactor Coolant System Loop Isolation Valve Interlocks, 7.6.8.1 Description The purpose of these interlocks is to ensure that an accidental startup of an unborated and/or cold, isolated reactor coolant loop results only in a relatively slow reactivity insertion rate.
The interlocks are required to perform a protective function.
Therefore, there are:
- a. Two independent limit switches to indicate that a valve is fully open.
d
- b. Two independent limit switches to indicate that a l valve is fully closed.
l
- c. Two differential pressure indicating switches monitoring a flow orifice in each line which bypasses a cold leg loop isolation valve indicate: 1) the valves in the line are open, 2) the line is not blocked, and 3) the pump is running.
7.6.9 Interlocks For RCS Pressure Control During Low Temperature Operation -
The basic function of the RCS pressure control during low temp-erature operation is discussed in Subsection 5.2.2.11. As noted in Subsection 5.2.2.11 this pressure control includes manually armed semi-automatic actuation logic for two Pressurizer l Power Operateo Relief Valves (PORV's). The function of this actuation logic is to continuously monitor RCS temperature and pressure conditions, with the actuation logic only unblocked by a manual arming switch when plant operation is at a tempera- [
ture below the Reference Nil Ductility Temperature (RNDT).
The monitored system temperature signals are processed to l
()
()
generate the reference pressure limit program which is compared to the measured pressure.
7.6-3
("N B/B-FSAR V .
9 7-~s This comparison will semi-automatically open the PORV when B
(') the device is manually armed as necessary to prevent pressure conditions from exceeding allowable limits. See Figure
? 7.6-6 for the block diagram showing the interlocks for RCS B pressure control during low temperature operation.
) As shown in'this' figure, the station variables required for
? this interlock are channelized as follows:
}
- a. Protection Set I
? 1. wide range RCS temperature from hot legs and 5 2. wide range RCS system pressure (PT 405) .
- b. Protection Set II
- 1. wide range RCS temperature from cold legs.
- c. Protection Set IV
- The wide range temperature signals, as inputs to the Pro-tection Sets I and II continuously monitor RCS temperature
$ conditions whenever plant operation is at a temperature below
. the RNDT. In Protection Set I, the existing RCS hot leg wide f range temperature channels will supply through an isolation
) \- device continuous analog input to an auctioneering device, a which is located in the Process Rack of Control Rack Group
- 1. Th" lowest reading will be selected and input to a functi%n generator which calculates the reference pressure
, limit program considering the plant's allowable pressure and temperature limits. Also available from Protection Set I is the wide range RCS system pressure signal which is sent through an isolation device to Control Rack Group 1.
r The reference pressure from the function generator is compared 5 to the actual RCS system pressure monitored by the wide range f pressure channel. The error signal derived from the difference
- between the reference pressure and the actual measured pres-
- - sure, will first annunciate a main board alarm when-h ever the actual measured pressure approaches, within a pre-l determined amount, the reference pressure. On a further in-crease in measured pressure, the error signal will generate
- an annunciated actuation signal. The actuation signal avail-f able from Control Rack Group 1 will control PORV "A" whenever a manually armed permissive signal from Control Group 4 is present. The manually armed permissive to the PORV's.
p actuation device is a signal which is turned on when the MCB two position maintained arming switch is placed in the L
ARM position. When it is in the BLOCK position the actua-tion signal at temperature greater that the range of concern, This will prevent unnecessary system actuation when at n\m,/ normal RCS operating conditions as a result of a failure in the process sensors.
7.6-4
1
- B/B-FSAR O The arming switch is placed in the ARM position when the low auctioneered RCS temperature signal reaches a low set-point value which is indicated by an annunciated actuation signal. The monitored generating station variables that generate the actuation signal for the "B" PORV are processed in a e
- O i
l .
- l i
4 1
O i
7.6-4a
/'"N B/B-FSAR b
$- similar manner. In the case of PORV "B", the reference
- temperature is generated in Control Rack Group 4 from the 3 lowest auctioneered wide range cold leg temperature, the
- auctioneering device deriving its inputs from the RCS g wide range temperature in Protection Set II, and the actual
- measured pressure signal is available from Protection Set
. IV. Therefore, the generating station variable.= used for PORV "B" are derived from Protection Sets that are independent
- of the Set from which generating station variables used for PORV "A" are derived. The error signal derivation itself 3 used for the actuation signals is available from the Control Group.
- Upon receipt of the actuation signal and with the arming
- switch in the ARM position, the actuation device will auto-matically cause the PORV to open. Upon sufficient RCS inventory letdown, the operating RCS pressure will decrease,
- clearing the actuation signal. Removal of this signal causes the PORV to close.
- ~
' ~
7.6.9.1 Analysis of Interlock l Many criteria presented in IEEE 279-1979 and IEEE 338-1971 standards do not apply to the interlocks for RCS pressure
- control during low temperature operation, because the inter-locks do not perform'a protective function but rather provide
- automatic pressure control at low temperatures as a backup to the operator. However, although IEEE-279 criteria do not
, apply, some advantages of the dependability and benefits of an IEEE-Std-279 design have occurred by including the pressure
, and temperature signal elements as noted above in the Pro-
- tection Sets and by organizing the control of the two PORV's
- (either of which can accomplish the RCS pressure control function) into dual channels wherever practical. Either of the two PORV's can accomplish the RCS pressure control function.
- The design of the low temperature interlocks for RCS pressure b control is such that pertinent features includes.
- a. No credible single failure at the output of the l
. protection set racks, after the output leaves
- the racks to interface with the interlocks,
. will prevent the associated protection system a channel from performing its protective function
- because such outputs that leave the racks go through an isolation device as shown in Figure 7.6-6 and because there are no chared components between channels.
- b. Testing capability for elements of the interlocks l , within (not external to) the Protection System j N.
l l
7.6-5
j B/B-FSAR d
- j is consistent with the testing principles and l , methods discussed in Subsection 7.2.2.2.3. It a should be noted that there is an annunciator
- which provides an alarm when there is low li autioneered RCS
.I l h
it i Il i
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BYRON /BR AIDWOOD ST ATIONS FIN AL S AFETY AN ALY SIS REPORT S
1 FIGURE 7.6-6 DI AGRA'l SH0'.llfiG GEt1ERATlfiG PL/J:T VAR" ULE PROCESSItiG FOR LO'.1 TEMPERATURE IflTERLOCKS FOR RCS PRE 55UP.E
o o .
o o TABLE 11.4-2 EXPECTED AND DESIGN BASIS ANNUAL VOLUMES OF i
I (UNITS 1 AND 2) SOLID WASTE MANAGEMENT SYSTEM OUTPUT
- Solid Waste Processed by the Solid Radwaste Management System and Quant
- ties of Processed Waste Requiring Onsite Storage or Offsite Disposal With and Without The Volume Reduction System WITHOUT VOLUME REDUCTION SYSTEM WITH VOLUME REDUCTION SYSTEM EXPECTED DESIGN EXPECTED DESIGN TYPE OF WASTE VOLUMES DRUMS VOLUME DRUMS VOLUME DRUMS VOLUME DRUMS w
3 3 3 Deep Bed Resin 1,520 ft 413 1,520 ft 413 1,520 ft 413 1,520 ft 413 g O i Disposable Filter 3 3 3
Z Elements 75 ft 190 75 ft 190 75 ft 190 75 ft 190 Evaporator 3 3 3 326 concentrates 16,850 ft 4580 18,690 ft 5140 16,850 ft 294 18,690 ft 3 7 Dry Active Waste 36,'220 ft 580** 36,220 ft 1160** 36,220 ft 7 36,220 ft 73 boxes 73 boxes 73 boxes 73__b_ oxes 3 3 54,665 ft 5763 56,5G5 ft 6903 54,665 ft 904 56,505 ft 936 Total 73 boxes 73 boxes 73 boxes 73 boxes t
- The values given are approximate.
- Not Solidified.
i f .
C/B-FSAR O)
\
Tne detector wetted parts are required to maintain gressure
- boundary integrity during abnormal conditions. The detector is y not required to meet performance requirements for this perica.
. For post-LOCA monitoring, the samples will be collected manually
' and analyzed in the laboratory.
11.5.2.2.11 Miscellaneous Tank Veht Effluent l
Detectors ORE-PR025A, B and C (air particulate, gas and I2 L. channels respectively) monitor miscellaneous tank vent effluent.
I Iligh radiation is annunciated in the , main control room.
f 11.5.2.2.12 Radwaste Area Vent Exhaust r;
4 Detectors ORE-PR026A, B and C (air particulate, gas, and I2
< channels respectively) monitor the radwaste area vent exhaust.
The radwaste area vent exhaust is ducted to the auxiliary building vent stack.
High radiation is annunciated in the main control room.
11.5.2.2.13 SJAE/ Gland Steam Exhaust Detectors 1RE-PR027A, B and C (air particulate, gas and I2 channels, respectively) and 2RE-PR027A, B and C monitor.the off-gas system
) exhaust. Automatically on high radiation in the off-gas exhaust a
( stream, l'ypass valves and the off-gas vent filter system 0 ($G0lS 0 are energized, s
j 11.5.2.2.14 Gas Decay Tank Effluent l
$ Detectors ORE-PR002A and B (gas channel) monitor the radiation level
) of the gas decay tank discharge to the auxiliary building vent stack.
l Automatically, on high radiation in the gas decay tank discharge, valve OGWRCV014 closes.
11.5.2.2.15 VR System Areas and Cubicles Ventilation Exhaust f
Detectors ORE-PRO 40A, B and C (air particulate, gas and I2
) channels respectively) monitor the ventilation exhaust from 9 the Volume Reduction equipment areas and cubicles, a
The radiation monitor is interlocked with the volume reduction ventilation exhaust fans OVW10C and OVW14C, associated bypass, filter inlet and outlet dampers. Automatically on high radia-tion the bypass dampers close and the fans start to route the
(
exhaust through the filter unit.
I o Refer to Subsection 9.4.3.3 for a description of the radwaste
( building ventilation system.
1 3 x f
f 11.5-9 1
1
D/B-FSAR g 11.5.2.2.1,6 Miscellaneous Process Monitors l g Miscellaneous other monitors shown in Table 11.5-1' monitor the process as indicated.
- High radiation is annunciated in the main control room.
- 11.5.2.3 Liquid Effluent Monitors
- 11.5.2.3.1 Process Liquid Monitor Liquid Radwaste Effluent g Detector ORE-OR001 monitors liquid radwaste effluent and is s interlocked with release tank discharge valve OWX353.
- On high radiation in the liquid radwaste effluent, the release tank discharge valve is closed automatically.
- 11.5.2.3.2 Component Cooling Water Monitors
, Radiation detectors 1RE-PR009, 2RE-PR009, and ORE-PR009 con-
, tinuously monitor the component cooling system for leakage of
. reactor coolant from the reactor coolant system and/or 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 2CC01T O^ vent valve 2CCRCV017. Detector ORE-OR009 is interlocked with
. both the vent valves 1CCRCV017 and 2CCRCV017.
G O
ll.5-9a l
TABLE 11.5-1 (Cont'd)
O PANEL SEISMIC RADIATIOM TYPE OF TYPE OF TYPE OF SENSITIVITY RANCE DETECTOR IOCATION CAT. OF DETr.CTOR *tO. SERVICE CMANNEL DETECTOR MEAS. (Uc/ce) (cpm) PANEL NO. DWG. NO. DETECTORS SETPOINT REMARKS ORE-PR031A,2,C Control Room " Jut- Air Part. *II B Scint. Cross B 10 5 10h-107 OPR31J M-832-12 I ~<2s background Eedundant with side Air Intake A Cas .B Scint. Cross a 30 6-10-2
~
II 5 1C'-10 1 7 ORE-PRO 32A,8 lodine Na! y(I-131) 10 10 -10 Interlock ref.
11.5.2.2.8 ORE-PR032 A,3,C Control Room Out- Air Part. B Scint. Cross B 10~I 10-5 10 -10 77 OPR32J M-832-12 I 12s background Interlock ref.
side Air Intake A Cas B Scint. Cross B 10-6 10-2 10 -10 7 11.5.2.2.8 Iodine Na1 y(1-131) 10"II--10-5 10 -10
-11 5 ORE-PR0 33 A,B,C Control Room Out- Air Part. B Scint. Cross B 10 102 10 -10 7I OPR33J M-832-19 I ~<2: backgrGund Redundant with side Air Intake B Cas B Scint. Cross B 10 10~ 10 -10 ORE-PkO14A,B Iodine Nat y(I-131) 10"Il-10~ 10 -10 Interlock ref.
11.5.2.2.8 ORE-PR034 A,B,C Contr ol Room Out- Air Part. B Scint. Cross B 10~1 -10 OPR34J M-832-19 I 12s background Interlock ref.
side Air Intake B Cas a Scint. Cross B 10"6-10~f1010-10-10 7 11.5.2.2.8 g Iodine Nat y(g.131) 10 -II-10~I 10 -10 I 5
h ORE-PR035A,B,C Control Rm Turb. Air Part. B Scint. Gross B 10NI-10'2 10Ig-10 OPR35J M-832-12 I 12s background Redundant with o Bldg. Air Intake A Cas B Scint. Cross B 10 g10-10~$
10g -10 7 10 -10 -
ORE-Ph036A,B >
Iudine Na! 1(1-131) 10~ j ORE-PP036 A,6,C Control Rm Turb. ~$ 10Ig-10 9 Air Part. B Scint. Cross B 10 -10 OPR36J M-832-12 I -<2 background b Bldg. Air Intake A Cas B Scint. Cross B 10 -10 Iodine Na! y(I-131) 10~h10-2 10" -10 ~5 10 -10 7 I
ORE-PR0 37A,B,C Control ps. Turb. Air Part. B Scint. Cross B 10 -10~ 10-10hOPR37J g M-832-14 I S2s background Redundant w{th Bldg. Air Intake B Gas B Scint. Cross B ~
ORE-PR038A,8 Iodine NaI y( I-131) 10~II-10-2 10~ -10~5 101 -10 7' CRE-PP038 A,B.C Control Rm. Turb. Air Part. B Scint. Cross B 10~ -If2 I
10 g-10 OPR38J M-832-14 I $2s background Bldg. Air Intake B Ces B Scint. Cross B 10*II-10 101 -10 #7 Iodine Na1 v(I-131) 10- -10~ 10 -10 IRE-PR021 Pipe Tunnel Air Part. B Scint. Cross B 10 *II-10*$ 10 -10 1PR21J M-831-10 I $ 2s background 1 7 Gas B Scint. Cross B 10-6-10-2 10 -10
~
Iodine Na! y(1-131) 10~II-10-5 10I -10 2KE-PR021 Pipe Tunnel Air Part. B Scint. Cross B 10~II-10~I 10I -10 2PR21J M-831-3 I 1 2s background A,b,C I
Cas B Scint. Cross B 10~0-10-2 10 -10 Iodine Na1 y(1-131) 10' I-10-5 101 -10 7
, ORE-PR040 VR System Areas Ala Pa r t .
l A,B,C 6 Cub. Ventila-B Scint. Cro n R 30 -10~ 100-10 OPR40J II 12 background Interlock ref.
. Gas B Scint. Cross B 10-6-10-2 10*-107
-10~ 10"-10 11.5.2.2.15
r"'s B/B-FSAR U
Dilution During Cold Shutdown .
)
' In the event of an inadvertent dilution during cold shudown, the source range nuclear instrumentation will detect a doubling of the neutron flux by comparison of the current source range flux to that of approximately 10 minutes earlier. Upon detec-tion of the flux doubling, an alarm is sounded for the operator and valve movement in the CVCS, to terminate the dilution and start boration from the RWST is automatically initiated. These automatic actions are carried out to minimize the approach to criticality and regain the loast shutdown margin g Dilution During Hot Shutdown and Hot Standby In the event of an inadvertent dilution during hot. shutdown or hot standby, the source range nucl~ ear instrumentation will detect a doubling of the neutron flux, automatically initiate valve movement to begin boration and terminate the dilution, and sound an alarm for the operator. No operator action is required to terminate the transient and minimize the approach to criticality.
Dilution During Startup For dilution during startup the minimum time required for the shutdown margin to be lost and the reactor to become critical is 32.0 minutes.
4 Dilution During Full Power Operation ,
- a. With the reactor in automatic control, the power and temperature increase from boron dilution results in insertion of the rod cluster control assemblies and a decrease in the shutdown margin. The rod insertion limit alarms (low and low-low settings) provide the operator with adequate time (of the order of 110 minutes) to determine the cause of dilution, isolate the primary grade water source, and initiate reboration before the tctal shutdown margin is lost due to dilu-tion.
- b. With the reactor in manual control and if no operator
- action is taken, the power and temperature rise will cause the reactor to reach the overtemperature AT trip setpoint. The boron dilution accident in this
- case is essentially identical to a rod cluster control
- assembly withdrawal accident. The maximum reactivity 9 insertion rate for boron dilution is approximately 0.78 pcm/sec and is seen to be within the range of insertion b rates analyzed. Prior to the overtemperature AT trip, an overtemperature AT alarm and turbine runback would be actuated. There is adequate time available 15.4-26
l
() B/B-FSAR
'(of the order of 108' minutes) after a reactor trip for O- the operator to determine - the cause of. dilution, iso-late the primary grade water' sources and initiate l
reboration bef ore the reactor can return to criticality.
15.4.6.3 Radiological Consequences There are only minimal radiological consequences associated with a chemical and volume control system malfunction that results in a decrease in boron concentration in the reactor 4
i
)
O -
5 15.4-26a l
- -.~--_,:,..-_---,,----_-.-.-_____
B/B-FSAR l
O l TABLE 15.4-1 (Cont'd)
Time Sequence of, Events for Incident which Cause Reactivity and Power Distribution Anomalies TIME ACCIDENT EVENT (sec.)
l Startup of an Dilution begins 0 inactive reactor .
coolant loop at an Operator terminates dilution 5040 incorrect temperature minimum margin to criticality occurs I CVCS Malfunction
.3 that results in'a decrease in the,,.
boron concentration d in'the reactor s coolant
, , 1. Dilution during* Dilution begins O refueling Operator isolates source of 5172 dilution; minimum margin to criticality occurs
,s ;
- 2. -Dilution during Dilution begins O cold shutdown Operator isolates source of Later l
dilution; minimum margin to criticality occurs O 15.4-47 s
- -._____m__.______. - . _ . _ . _ _ _ _
Q B/B-FSAR V
() TABLE 15.4-1 (Cont'd) l TIME SEQUENCE OF EVENTS FOR INCIDENT WHICH CAUSE l REACTIVITY AND POWER DISTRIBUTION ANOMALIES TIME ACCIDENT EVENT (sec.)
- 3. Dilution during Dilution begins 0 hot standby Operator isolates source of Later l dilution; minimum margin to criticality occurs
- 4. Dilution during Dilution begins O startup Operator isolates source of 1920
() dilution; minimum margin to criticality occurs
- 5. Dilution during full power oper-l ation l
f
- a. Automatic Dilution begins 0 l
reactor control Shutdown margin lost 6600
- b. Manual Dilution begins 0 reactor control Reactor trip setpoint reached 120 for-overtemperature AT l
Rods begin to fall into core 122 0 -
Shutdown is lost (if dilution 6600 continues af ter trip)
.,\
B/B-FSAR O REGULATORY GUIDE 1.40 Current Issue: Rev. O, March 16, 1973 l QUALIFICATION TESTS OF CONTINUOUS-DUTY l MOTORS INSTALLCD INSIDE THE CONTAINMENT OF WATER-COOLED NUCLEAR POWER PLANTS
[
NSSS Scope It is the Westinghouse position that motors inside containment comply with the qualification control requirements of Criterion III to Appendix B to 10 CFR 50. These requirements are satisfied by qualification as described in WCAP-8587 and its supplement which contains appropriate EQDP's (Equipment Qualification Data Packages) for Westinghouse supplied continuous duty motors within
- . the containment. We therefore feel we are in compliance with the objectives of Regulatory Guide 1.40.
(s)
Non-NSSS Scope The Applicant complies with the requirements of Regulatory Guide 1.40 with the clarification to the Regulatory Position identified and justified below:
Regulatory Position Cl To the extent practicable, auxiliary equipment that will be part of the installed motor assembly should also be qualified in accordance with IEF.E Standard 334-1971.
Applicant's Position
, Comply with regulatory position, in that to the extent prac-ticable, auxiliary equipment essential to the safety function of the installed motor assembly will be qualified in accordance with IEEE 334-1971.
l Justification of Applicant's Position
- Nonesuential auxiliaries have no safety function and should be excluded from the requirements.
O g
A1.40-1
m ,
B/B FSAR REGULATORY GUIDE 1.89 Current Issue: Rev. O, November 1974 QUALIFICATION OF CLASS lE EQUIPMENT FOR NUCLEAR POWER PLANTS NSSS Scope For Westinghouse _NSSS Class lE Equipment, Westinghouse will meet the requirements of IEEE 323-1974, "IEEE Standard for Qualifying Class lE Equipment for Nuclear Power Generating Stations" includ-ing the Nuclear Power Engineering Committee (NPEC) Position Statement of July 24, 1975, and Regulatory Guide 1.89, by an appropriate combination of any or all of the following: type testing, operating experience, qualification by analysis and on-going qualification. This committment will be satisfied by implementation of the final approved version of WCAP-8587.
Non-NSSS' Scope The extent of the Applicant's committment to comply with the :
requirements of Regulatory Guide 1.89 is presented in Subsections ,
3.11.2 and 8.1.16.
9 0 .
A1.89-1
B/B-FSAR REGULATORY GUIDE 1.100 Current Issue: Rev. O, March 1976 SEISMIC QUALIFICATION OF ELECTRIC EQUIPMENT FOR NUCLEAR POWER PLANTS NSSS Scope The Applicant is in compliance with the objectives of Regulatory Guide 1.100. The Westinghouse program for seismic qualification of safety-related electrical equipment is delineated in WCAP-8587, Revision 1. This program is currently under review by the NRC.
For further details, refer to Section 3.10.
Non-NSSS Scope
) The Applicant complies with the objectives of Regulatory Guide 1.100. The Applicant's approach to seismic qualification of Class lE equipment is discussed in Section 3.10.
6 O
A1.100-1
/
B/B-FSAR Position C.2.b In cases where sufficient inspection data exists to establish a degradation allowance, the rate used will be an average time-rate determined from the mean of the test data. The combined effect of these requirements would be to establish a maximum permissible primary-to-secondary leak rate which may be below the threshold of detection with current methods of measurement.
Westinghouse has determined the maximum acceptable length of a through-wall-crack based on secondary pipe break accident loadings which are typically twice the magnitude of .armal operating pressure loads. Westinghouse will use a leak rate associated with the crack size determined on the basis of acci-dent loadings.
Position C.3.e(6)
Westinghouse will supply computer code names and references rather than the actual codes.
Position C.3. f (1)
The 40% T.S. limit is a reference limit for Westinghouse steam generators. Regulatory Guide 1.121 analyses have not been completed for model D4 and D5 steam generators used in Byron /
O Braidwood. These analyses will be completed prior to first refueling and at that time the T.S. limits will be re-evaluated and this information can then be included in the Regulatory Guide 1.121 position, if necessary.
Position C.e.f(4)
~
Where requirements for minimum wall are markedly different for different areas of the tube bundle, e.g., U-bend area versus straight length in Westinghouse designs, two plugging limits may be established to address the varying requirements in a manner which will not require unnecessary plugging of tubes.
I i
A1.121-2
D/B-FSAR O E.30 ADDITIONAL ACCIDENT-MONITORING INSTRUMENTATION (II.F.1)
() POSITION:
- 1. Noble Gas Effluent Monitor (II.F.1-1)
- a. Auxiliary Building Vent Stack Two General Atomic Company wide-range monitors will be installed on the auxiliary building vent stacks (final release points), one monitor per stack. The monitor _7 has a range for 4gdioactive gas concentration of 1 x 10 PCi/cc to 1 x 10 pCi/cc. The monitor is designed to meet lE requirements and is qualified to IEEE 323-1974.
The wide-range gas monitor meets the requirements of ;
Table II.F.1-1 of NUREG-0737. The monitor includes the following: ic nozzles, one for normal condi-two at tions operating isokineg/mjn 2ft and one for high range condi-tions operating at 0.06 ft / min; sampling rack (reference discussion of II.F.1-2); sample conditioner, operating only at high range conditions to filter out large concen-trations of radioiodine and particulates; and the wide-range gas detectors assembly, consisting of three radio-active gas detectors, a low-range detector (Model Number RD-521, a mid-range detector RD-72-01) and a high-range detector (Model Number RD-72-02). Each monitor system has a microprocessor which utilizes digital processing techniques to analyze data and control monitor functions.
O, Control room readouts include a che.rt recorder and an RM-23 remote display module for all monitor parameters.
- 1. The calibration techniques and procedures including the energy dependence of the detectors will be provided to meet the requirements of NUREG-0737.
- 2. The monitors will receive power from ESF buses.
- 3. The plant release calculations are as follows:
l Plant release rates are calculated from the iso-topic analysis in accordance with the following
( expression:
Curies /Sec =
4 X(curies /cc of sample) x (Flow-cfm) x (2.33x10 cc/ft3) 60 se:/ min where:
(curies /cc of sample) is determined from the isotopic analysis as being the time-corrected sum of the concentrations of the measured radio-nuclides, and the flow is the stack effluent
('di flow rate.
E.30-1 l
N-B/B-FSAR i (o,) Subsequent plant release rates may be calculated
, from the effluent monitor readings in accordance with the following expression:
Curies /Sec =
monitor readings (CPM) x xCurie/Sec x F_
cpm Fi i where:
, Curie /Sec is the ratio of the release rate calcu-cpm lated from the isotopic analysis above to the monitor reading when the isotopic analysis sample was taken; and the ratio of F/Fi is the ratio of the effluent flow to the effluent flow when the isotopic analysis was taken.
The isotopic analysis of the grab sample will establish the correlation between effluent monitor reading and plant release rate.
- b. Main Steamline N' Four General Atomic Company RD-12 detectors will be pro-vided for each of the four main steamlines upstream of the safety yand relief valves +3 The range of the monitor is 1 x 10 PCi/cc to 1 x 10 uCi/cc. The monitor is
- designed to meet lE requirements and is qualified to IEEE 323-1974. The monitors will be mounted external to the main steamline piping and corrections made for the loss of low energy gammas.
The detectors are connected to local mounted micropro-cessors that collect and store data. Main control room mounted remote readout modules and strip chart recorders are connected directly to the microprocessors to provide information to the operator during and following an acci-
- dent. Readings can be obtained by the operator at 15-minute intervals if required.
( 2. Sampling and Analysis of Plant Effluents (II .F .1-2 )
) The Genere.1 Atomic, Company wide ra7ge gas monitor includes h a sampling rack for collection of the auxiliary building i vent stack particulate and radiciodine samples. Filte- holders
, and valves are provided to allow grab sample collection for
- isotopic analyses in the stations' counting rooms. The sampl-8 ing rack is shielded to minimize personnel exposure. The (N
\s,)
sampling media will be analyzed by a gamma ray spectrometer which utilizes a Ge(Li) detector. Filter cartridges will
, be reverse blown with air to purge interferring noble gases.
E.30-la
i i
O B/B-FSAR The sampling system design will be such that radiation expo-sures are within the requirements of GDC 19 as stated in NUREG-0737. The sample media used for both iodine and partic-ulates sample collection will meet the requirements for effec-tive adsorption and retention as stated in NUREG-0737.
1 i
i l
4 l
O - E.30-lb '
i i
l s B/B-FSAR l
? i
- QUESTION 010.40 Provide a response to questien QO10.17 and include the folicuing in your response. Provide the results of analyses of the effects :n safety-related systems of failures in any high or moderate energy piping system in a::or-dance with the J. F. O' Leary letter of July 12, 1973, as defined in 3rta:n Techniczi Fasition ASB 3-1, Appendix C. Provide a table which identifies the method of protection provided all safety-related systems listed in FSA?.
Table 3.6.1 from failuras of any high or moderate ener;y systa s listed .
in FSAR Table 3.5-2. Include figures depicting the locacions Of failures nlative to the systems of FSAR Table 3.5-1 giving dimensions, locaticns and ,
protective method for each postulated break or crack, in a high ' cr mcderate energy system. Include the assumptions u ed in your analysis such as flowrates through postulated cracks, pump racm areas, su p capacities, and ficor drainage system capacities.
RESPONSE
The attached report has been generated to address high/ moderate energy line break outside containment.
G I
I Q10.40-1
~
. B/B
/3
. HIGH/ MODERATE ENERGY LINE BREAK OUTSIDE CONTAINMENT O I INTRODUCTION To insure safe and reliable operation of the Byron and Braidwood Nuclear Power Stations, the possibility of high or moderate energy line breaks have been considered in the design. This report documents a confirmatory study of the potential high and moderate energy line breaks which demonstrates that all design features necessary to mitigate the results of line breaks have been in- .
corporated.
II SCOPE This report considers potential high and moderate energy line breaks outside containment which could af fect safe'ty related systems. Non-safety related areas, such as the turbine building, were not investigated because damage to or failure of equipment in these areas will not affect the function of safety related systems.
The possible effects considered in this report are structural loads due to pressurization, increases in pressure, and temperature j
which could affect environmental qualification of equipment, 1
and damage due to pipe whip. Flooding is a potential effect but is not addressed in this report. A separate report will be issued to demonstrate that high and moderate energy line breaks will not cause flooding which would adversely affect the plant safety.
O 4
010.40-2
B/B -
S .
Because of variations in requirements, techniques, and failure
)
(
j effects, high and moderate energy lines are addresse'd separately.
l Similarly, the pipe whip, subcompartment pressurization, and environmental analysis all have somewhat different approaches.
The following sections are divided to reflect these distinctions.
III HIGH ENERGY LINE ANALYSIS Standard Review Plans 3.6.1 and 3.6.2 were followed in defining and identifying high energy lines. High energy lines are those for which either:
- 1. The service temperature is greater than 200'F; or
- 2. The design pressure is greater than 275 psig.
Only a limited number of systems in the auxiliary building meet either of these criteria. The following systems have been
(}
identified as containing high energy lines in the auxiliary building: -
Chemical and volume Control (CV)
Auxiliary Steam (AS)
Steam Generator Blowdown (SD)
Radioactive Waste Processing (WX)
Boric Acid , (AB)
Main Steam (MS) !
(AF)
. Recidual Heat Removal (RH)
Safety Injection (SI)
Systems which are normally not used or at reduced temperature and pressure are not necessarily required to be considered as high energy lines. A guideline has been established (Branch Technical Position MEB 3-1) that if the system is at high energy O
G l
010.40-3
s.
B/B l
V) . .
) conditions less than 2% of the time, it may be considered a
. moderate energy line and its normal conditions applied to the i line break analysis. On this basis, the last three systems (AF, RH, SI) are not considered as high energy. systems. The Byron /Braidwood AF system is not considered a high energy system because it is.used only under loss of offsite power conditions and not normal startup as at some other plants.
Subcompartment pressurization is investigated for all lines with temperatures above 200*F. Lower temperature lines do not have the potential for flashing to steam and thus will not increase the pressure of a subcompartment in the event of a, break. Pressur-ization is of concern only in small subcompartments with relatively
(~)'
\_- large high energy lines or subcompartments with limited pressure relief venting.
High energy lines below 200*F have only minor effects on the environmental conditions. The absence of steam and the ability to drain warm liquid from the break area limits the temperature rise from these breaks. The auxiliary building HVAC has sufficient capacity to accommodate these iower temperature breaks.
i Breaks of higher energy lines may influence the expected maximum temperature in some areas of the auxiliary building even if high pressures do'not result. The auxiliary building contains several large areas with high energy lines that are not subject to pressurization but are investigated for environmental effects.
I l wp 010.40-4
B/B
?'
s Certain postulated break locations in high energy piping systems
\- are used to investigate the potential foi damage due to pipe whip. The guidelines in Standard Review Plan 6.2.2 are used to determine the number and locations of the pipe breaks. Pipe restraints are added as re uired to prevent damage to structures and safety related equipment.
IV MODERATE ENERGY LINE BREAKS ,
Moderate energy lines do not cause subcompartment pressurization.
The low temperature of a moderate energy line insures that no steam will be produced and the pressure within the subcompartment will remain atmospheric.
Moderate energy line breaks will not cause increases in the
() environmental temperatures and pressures. The reduced break area applicable to these breaks and the absence of steam allow the auxiliary building HVAC to maintain temperatures within those specified in the environmental qualification program.
Moderate energy line breaks do not result in pipe whip.
4 O 1 010.40-5
B/B 0.-
1 V SUBCOMPARTMENT PRESSURIZATION .
3 For the purpose of protecting subcompartments from overpressuri-f zation, the CV, AS, SD, WX, MS and FW systems were traced through h the auxiliary building and all subecmpartments containing high energy lines were identified. Potential breaks were identified f
j at pipe joints and fittings. The most severe break in the sub-
) compartment was analyzed. If the system consisted of only a h straight run (no joint or fitting) no' break was postulated.
) The main steam (MS) and feedwater (FW) systems are routed entirely 3 in an enclosed tunnel in the auxiliary building. The limiting
- break in this tunnel is a main steam line rupture. FSAR
? Section C3.6 fully describes an analysis of a break.in this tunnel.
& The remainder of the auxiliary building was surveyed level by level to identify all subcompartments which could be pressurized r by high energy line breaks. Figures 1 through 5 identify all j areas containing high energy lines. The zone numbers do not k correspond to environmental qualification zones (Section 3.11).
j Figure 1 represents elevation 346'-0". Zone 1, the recycle I waste evaporator room, has been analyzed and the results are
! reported in Section A3.6 of the FSAR. Zones 2 and 3, letdown 3 reheat heat exchanger rooms, have been analyzed and the results t are reported in Section A3.6. The assessment'in A3.6 addressed Zone 3, the more limiting zone.
010.40-6
B/B f .
1 Assessment of Zones 4 and I () Figure 2representselevation364'-0".
5, positive displacement charging pump areas, has not been completed.
The limiting line break in these areas will be a 3" CV line which operates at maximum conditions of 285 psig and 290*F. Zones 11
)
and 12, blowdown condenser rooms, have been analyzed and the results are reported in Section A3.6.
Zones 15 and 16, letdown 3 Figure 3 represents elevation 383'-0".
4 heat exchanger rooms, have been analyzed and the results are reported in Section A3.6. Zone 18, the auxiliary stemn line 8
i piping tunnel, has been analyzed and the results are reported in Section A3.6. .
Zones 20, 21, and 22, the Figure 4 represents elevation 401'-0". .
O1 surface condenser rooms, have been analyzed and the results have been reported in Section A3.6. Zone 23, boric acid tank room, has been analyzed and the results have been reported in Section A3.6.
\(
Figure 5 represents elevation 426'-0". Assessment of Zone 25, the volume control tank area, has not yet been completed. The limiting line break in this area is an 8 inch CV line which has Zones 26,
- a maximum operating condition of 75 psig and 250*F.
1 a 27, and 28, radwaste evaporator rooms have been analyzed and the J results are reported in Section A3.6.
V 010.40-7
B/B (m~,
) VI ENVIRONMENTAL QUALIFICATION .
3 A program to document the environmental qualification of electrical
) equipment is underway for Byron /Braidwood. The scheduled completion 5 date for this program is June, 1982. This program will establish 3 that '.he equipment required to safely shut down the plant will be 9 operable under potentially adverse environmental conditions.
8 One of the potential causes of severe environmental conditions is 6 a break or crack in a high or moderate energy line. This could
! cause an increase in pressure, temperature, or humidity or a flooding condition in the area of the break. Flooding is addressed in a
- separate report and will not be discussed here.
- The basic design of the Byron /Braidwood stations includes features
- to mitigate the impact of line breaks on the ability to safely
- t, shut the plant down. Some of the features are
' 0 1. Essential safety systems are redundant or backed up by other safety systems.
. 2. The effectiveness of the redundancy is, protected by separation of redundant systems to the greatest extent possible.
2 3. Walls and compartments have been included to both protect
. equipment and to isolate breaks.
y 4. Large high energy lines such as main steam, feedwater, a and auxiliary steam are partially or completely enclosed l
() in protective tunnels in the auxiliary building.
010.40-8
. B/B 5.
) Efforts have been made to minimize the number of high energy lines in areas containing safety related equipment and to
~
) minimize the Eze and length of high energy lines. For example, j Byron /Braidwood uses motor and diesel driven auxiliary feed-
}
water pumps rather than turbine driven pumps, thereby elimi-J' nating the associated high energy steam lines.
i
) The zones identified in Section V for high energy line breaks
) analysis are included in the environmental zones. The subcompart-3 ment transient conditions calculated in the pressurization analysis 3 are used for qualification of equipment. If the equipment is not l
l qualified for the predicted conditions, failure may be acceptable in the event of the postulated break. If this approach is used, b(~%
y the equipment will be checked to verify that failure is acceptable i
i 7
in the event of a postulated break in other lines in the subcompart-
) ment.
The areas identified in Figures 1 through 15 but not discussed in l
Section V, are not considered as subcompartments due to the relative size of the high energy lines, the volume of the area, and the f available flow areas. Table 1 lists these areas and the most limiting line associated with the areas. The analysis has not'yet f
l
) been completed to determine the maximuu temperatures which would p result from high energy line breaks in these areas. Because of the h large areas involved, the size and energy of the pipes involved, and the limited amount of safety related equipment in the areas, the l
(
l 010.40-9
B/B
/~'T .
\wsl. .
}
(_,/ impact on the equipment qualification is not expected to be major.
Moderate energy lines are not expected to impact the equipment qualification parameters. For lines with operating temperatures significantly above the normal area temperature, the crack flow rate and potential for heat transfer will be checked to insure that su'ficient f HVAC capacity exists to prevent failure of required safety related equipment.
I VII PIPE WHIP -
The methodology employed in the analysis of pipe whip is explained in detail in FSAR Section 3.6.2. Standard Review Plan 3.6.2 is followed. As discussed in the previous section, plant design features eliminate most pipe whip concerns. '
l 1 Of the systems listed in Section III, the main steam and feedwater o
are of most concern due to the large size and high pressure. The postulated break locations and the resulting restraiht locations fer the main steam and feedwater lines in the auxiliary building (main steam tunnel) are shown in FSAR figures 3.6-43 and 3.6-44.
The remaining systems for which high energy line breaks must be postulated (CV, AS, SD, WX, AB systems), the lines in many cases are not highly stressed or do not have the potential of impacting safety systems. The CV system, which contains high pressure lines, has been investigated and nine pipe restraints have been added.
Q10.40-10 -
3 ( Currently, all lines four inch diameter and larger have been
- analyzed. Portions of the smaller piping in high energy systems a remain to be analyzed and are currently under assessment. Only
- a small portion of this piping is expected to require restraints and the required restraints will be small and easily installed.
I
- VIII CONCLUSION 1 The subject of high and moderate energy line breaks as addressed a in Standard Review Plan 3.6.1 and 3.6.2 is a part of the Byron /
2 Braidwood design process. The basic arrangement of the plant and s the safety systems considers t'e potential for pipe breaks. As described above, the safety concerns have been addressed, incorporated
,; into the design, and in many cases, documented in the FSAR.
l:
The fact that work in this area is ongoing is not the result i
!. of omis,sion or disregard for safety standards, but the result
- o of the status of construction and the work required to respond to
,: revised and additional regulatory positions. The continuing work
- is, to a large extent, confirmatory. The schedule for this work
,: is consistent with the scheduled fuel load dates.
l b
G l
Q10.A O-ll
i O . .
i TABLE 1 .
i HELB (Tyoical)
- Zone System Size Temperature ('F) Pressure (psig) !
6 SD 6" 556 . 1092 ,
i 7 SD 6" 556 1092 8 SD 6" 556 1092
~
9 SD 6" 556 1092 10 AS 3" 300 50 CV 3" 290 285
~
l E 14 CV 3" 290 285 17 SD 3" 556 1092 i
19 CV 8" *250 75 24 AS 1" 300 50 i
l l
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1
9 D/B-FSAR AMENDMENT 36 JANUARY 1982 supplier has qualified the valves'for mechanical Is and seismic loading by analysis, and has proven
\- ') the operability of the valves through normal and emergency environmental conditions by actual test.
B.l.d: The containment isolation provision for the purge system lines are designed to Section III, Class 2, and Category IE electrical requirements. Inboard and outboard isolation valves (redundant valves 4 are supplied by Division 11 and 12 power respectively.
Operators are of an air / spring design, fall the 1 valve to the closed position upon 1 :ss of air or power, and are testable from the Control Room. The containment isolation provisions of the purge system therefore, meet all standards appropriate to Engineered Safety Features.
~
B.l.e: The purge system isolation valves close automatically on receipt of an ESF actuation signal. No external energy source is required to close the containment
, isolation valves. They are of a spring return design and will fail to the closed position upon loss of air pressure or electric power.
B.l.f: The specified maximum closure time for the containment g purge isolation valves is 5 seconds.
u
\- B.l.g: The containment mini-flow purge exhaust intake is
,.8 inches in diameter, located 73 feet above the
, operating floor and approximately 2 feet 6 inches from the face of the containment wall. Due to this distance, it is unlikely that following an accident, any debris would blow as high as the mini-flow exhaust intake.
To insure that debris or damaged ductwork does not
! Impair the isolation of the miniflow system following a lOCA, a debris screen will be added to the miniflow supply duct such that the system between the isolation valve and the screen will be capable of withstanding seismic.and LOCA conditions. -
1
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! 022.6-2 l
. p/B-FSAR
. QUESTION 022.8 O " Provide information on the required instrumentatior: to monitor containment atmosphere temperature and sump water temperature in the post-accident environment. Include the instrument range, accuracy, and response t!mo."
RESPONSE
Modifi.ation of the station design to implement Regulatory Guide . 97, Revision 2 is not complete. Postaccident monitoring instrumentation will be installed to satisfy Regulatory Guide 1.97, Revision 2; or justification will be provided for any alternatives. When the design is complete, i t will be submitted to the NRC staff for review. (This is the same commitment made to the Inctrumentation and Control System branch of the NRC to resolve ICSB Agenda Item 52.)
se t
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k Q22.8-1 T
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s
_B/B-FSAR QUESTION 022.12 "Prbvide an analysis demonstrating that the assumed times for full operation of the RCFC system and containment spray system in the containment functional analvses are conserva-tive, i.e., 40.0 and 45.0 seconds, resoectively, for the LOCA cases (FSAR Tables 6.2-6, 7, and 8) and 40.0 and 88.0 seconds, respectively, for the MSLB (FSAR Table 6.2-9).
RESPONSE
In the event of a LOCA or main steam line break in conjunction with loss of offsite power, the diesel generators will be at full power within ten seconds. Table 8.3-1 in the FSAR shows that the RCPC's are immediately loaded onto the emergency buses. The table also shows that containment spray.oumps are available at 25 seconds after loss of offsite pcwer or if not zequired' at 25 seconds, the pumos will be available at 50 b seconds. Delay times in Table 8.3-1 may be exceeded by 2 seconds. This will account for switching and signal transmission times.
The RCFC's will operate at full power 5 seconds after loading onto emergency power. The containment spray pumps operate at-full power 2.05 seconds after startup. The containment isolation valve requires 10 seconds to open. The valve is immediately loaded onto the ESF buses. The containment spray system is kept full at least to the 407' elevation (isolation valve) by the RWST. It will require less than 29 seconds to fill the spray I system and achieve full flow.
O O22.12-1
. B/B-FSAR
~h In the LOCA or MSLB case, the RCFC's will require the following (G
startup times after loss of offsite oower:
. s Diesel Startup 10 Seconds s RCFC Speedup 5 Seconds s Allowance for Signals 2 Seconds TOTAL 17 Seconds
, .This is well within the 40 sec'onds allowed in the analyses.
i For the LOCA case, the availability of power will be the limiting factor in the containment spray startup. The spray
- actuation signal (containment pressure) will be received before the spray pumps are loaded onto the emergency power.
. The containment spray valve will be open when the cump.is started. The containment spray will be available after the following times:
Pump Loading 25 Seconds
. Pump Startup 2 Seconds
- Spray System Filling 29 Seconds .
Allowance for ignals 2 Seconds TOTAL 58 Seconds This time is longer than the 5 seconds used in the analysis.
However, it should be noted that gravity flow will begin fill-ing the spray system when the valve is open because the RNST
~
level is 50 feet above the isolation valve. The pump will be partially effective during startup. The spray will initiate from the lower rings prior to full system flow. For these reasons and the margin shown by the analyses, the current assessment is considered satisfactory.
The results of the MSLB analysis shown in FSAR Figures 6.2-13 and 6.2-14 is an updated analysis (submitted in FSAR Amendment 31, l
' ('N June 1981). By error, Table 6.2-9 was not ucdated at this time.
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Table 6.2-9 has been updated to reflect the correct accident Q22.12-2 l
,, B/B-FSAR O
i sequence and timing.
() As discussed above the RCFC startup time of 58.7 seconds is very conservative. The containment spray will be available
- s -
- = after the initiation (high pressure) signal is received plus
- J the following times
Valve Opening Time 10 Seconds il Pump Startuo Included above
- Spray System Fill 29 Seconds
- Allowance for Signals 2 Seconds TOTAL 41 Seconds
- As shown in Figure 6.2-13, the containment pressure will
- reach 20-22 psig and initiate the spray at between 50 dnd 80
- seconds after the break. The total time for spray initiation
- is less than 121 seconds. This is conservatively represented N g
. by the 137.0 second time as shown in the revised Table 6.2p9.
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B/B-FSAR .
l C)
Orestion 022.13 U " Justify the containment spray system heat removal assump-tion (duration and flow) used in the containment function-al analyses in light of Subsection 6.5.2., which indicates that containment spray flow may be terminated when a low-low level is reached in the RWST and/or that manual switch-over from the injection to the recirculation may involve stopping and starting of the containment spray flow (see Question 022. 25) . "
RESPONSE
The containment spray system is designed to operate following a LOCA to reduce the elemental iodine concentration of the con-tainment atmosphere and to raise the pH of the' containment sump, by adding NaOH, to ensure that the iodine removed from the con-tainment atmosphere will be retained in the sump solution.
The objectives are completed in approximately 30 minutes, at which time the spray injection phase is terminated. The sys-tem is then isolated from the RWST and plant valves are aligned for recirculation operation. (It should be noted that after 30 minutes most of the heat removal from containment is provided by the reactor containment fan coolers, which are
(% Sprays are not required
(_) safety grade for long forheat term Byron /Braidwood).
removal. Nevertheless, the containment sprays will be separated for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a LOCA before they will be terminated.
The RHR and SI systems are designed to operate following a LOCA to cool the reactor core. These systems are switched from injection to recirculation at approximately 30 to 40 minutes and remain in operation for the remainder of the ac-cident. Additional fuel clad failure is not postulated while l these systems are operating.
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D '
D/B-FSAR AMENDMENT 36
[N Y JANLARY 1982
, QUESTION 022.25 g "FSAR Section 6.5.2.2 states ' Containment spray injection g and caustic education...will continue until...the low-o low level alarm of the RWST is annunciated. Containment 3
spray injection and caustic addition may then be terminated, e and the operating personnel may transfer the containment a spray pumps from the injection to the recirculation b mode by first closing the motor-operated valves in the j L'2ction line from the RWST, the water and caustic lines i to the eductor, and then opening the motor-operated y valves in the suction lines from the containment sumps.'
o State clearly whether transferring the containment spray a Pumps from the injection to the recirculation mode involves a . stopping and restarting the containment spray pumps o as implied in the above statement because the valves e in the suction line from the RWST are closed before e the valves in the suction lines from the containment sumps are opened." -
RESPONSE
The containment spray pumps will be run for at least 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> following a LOCA. During this time, switchover of pump I~h suction from the injection to the recirculation mode of kJ u ope' ration will be manually initiated and completed.
. The containment spray pumps do not have to be stopped when transferring from the injection mode to the recirculation mode of operation. The response to Question 450.2 from the Accident Analysis Branch provides a more detailed discussion on this subj ec t.
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, Q22.25-1
l O B/B-FSAR AMENDMENT 36 JANUARY 1982
( _- QUESTION 022.29
" Describe the containment isolation provisions for the k test connection penetration (P-4) and the spare penetrations listed in FSAR Table 3.8-1. Specifically include the design criteria connecting piping."
RESPONSE
The leak testing test connec'. ion penetration (P-4) and the spare penetrations are closed with welded cover plates. The design criteria for these cover plates are equivalent to the containment liner.
Drawing M-197-2 Revision G now contains the design information for 1PC-4. This penetration will be used for the integrated leak rate test. The penetration will be sealed off between the pipe and sleeve with a steel plate welded to both members.
There is a blind flange outside containment which will seal l the pipe when it is not being used for leak rate testing. '
, Drawing M-105-3 shows the piping arrangement (lPC-4 E P-4).
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- 022.29-1
i B/B-FSAR AMENDMENT 36 !
()
JANUARY 1982 s and both the tube and shell side of the excess letdown heat exchan-O gers are now Seismic Category I Safety Class B.
3.2 and 9.3 will be updated to reflect this revision to the safety FSAR Sections
, class of this equipment. As a closed system, these penetrations
- meet the requirements of GDC 57 and are listed as such in Table 6.2-58. .
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B/B-FSAR O
QUESTION 022.39
- I
" Provide in FSAR Table 6.2-58 the missing distances to the outside containment isolation valves (Column II). Additionally, provide i
evidence that all containment isolation valves located outside containment have been placed as close to the containment as practical, as required by GDC 55, 56, and 57, since some of the distances listed in FSAR Table 6.2-58 appear to be excessive."
1 I ,
RESPONSE
Revised FSAR Table 6.2-58 now lists the distances from the containment The valves were to the outer isolation valve on a particular line.
placed as close as practical to the containment with respect to the physical. arrangements of the plant, barriers, and obstacles. Isolation i r valves are being added in the off gas system to reduce the distances to the outside containment isolation valves. When the design is
- complete, the changes will be added to Table 6.2-58.
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l B/B-FSAR AMENDMENT 36 !
. JANUARY 1982
(-
( Position 5)
The containment setpoint pressure that initiates !
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) Contain5?ent 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 Attachment 1.)
The containment purge valves 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. These valves will be verified to be closed at least once every 31 days by checking position indication in the control room. 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.
O Q22.50-2
B/B-FSAR O
, QUESTION 022.54 9
" Verify that the normal containment purge system isolation
- valves (lVQ001A,B, and IVQ002A,B) and post-LOCA purge s system isolation valve (lVQ003) will be. sealed closed (as
- defined in SRP Section 6.2.4 ll.3.f)during the operational y modes of power operation, startup, hot standby, and hot
. shutdown."
~
RESPONSE
D The containment purge valves will be locked closed by the adminis-
- trative procedure of interrupting power to the valve at the 3
circuit breaker (i.e., the circuit breaker will be racked out) s and tagging the breaker "out of service." Inadvertant operation
- of the purge valves requires violation of procedures prohibit-
- ing both the operation of tagged-out equipment and the contain-
- ment purge system. Tagging out at the breaker is considered
- equivalent to a mechanical lock because in both instances positive action is used to prevent the valve from receiving power and an
- administrative procedure is required to return the breaker to service.
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(' .s B/B-FSAR QUESTION 022.55
" Provide information demonstrating, how CRP Section 6.2.4 II.7 will be met. This criterion concerns how system lines t which provide an open path from the containment to the environs should be equipped with radiation monitors that
{ are capable of isolating these lines upon a high radiation-signal."
RESPONSE
Area radiation detectors 1RE-AR0ll and 1RE-AR012 are interlocked with containment purge isolation valves lvo001A and B, and IVQ002A and B, and containment mini-purge isolation valves IVQ003, IVQ004A and B, and IVQ005A, B, and C. Upon detection l of high radiation levels, and containment ventilation isolation
- signal will be initiated and the above menti,ned valves closed.
It should be noted that the containment ventilation isolation signal is separate from either the Phase A or Phase B containment isolation signal as shown on page 24 of Table 6.2-58.
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- G Q22.55-1
O- ' AMENDMENT 36 B/B-FSAR JANUARY 1982 vO level is relatively insensitive to the recombiner initiation time. As demonstrated by the procedural outline in item (b),
operation of the recombiners requires only valve and switch operation. No difficulty is anticipated in initiating hydro-gen recombination within the four hours.
~ .
The hydrogen recombiner is one of the components on a list of common compon.ents and systems which has been developed to insure that all required equipment will be operational when Unit 1 is started. The hydrogen recombiner is, to a large extent, a self-contained package. The piping system required for the crosstie between units is unique to the system and will be in place. Normal and emergency power supplies to the recombiner and system valves will be provided as required by the original system design. -
The hydrogen recombiners are physically separated in the i plant by approximately fifty feet and several walls. The recombiners are powered from separate emergency power sources. The crossover valves are powered and arranged such that failure of one emergency power source will not
, prevent the remaining recombiner from being used on either unit.
'd. The maximum operating conditions for the recombiners have been reevaluated. The recombiners have been specified and are being qualified to operate with containment atmosphere temperatures of up to 225'F and pressures up to 21 psig.
The recombiner will. withstand system pressures of up to 60 psig while not operating. After a LOCA, the containment temperatures will be reduced to the operating levels in l
less than 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />.
l The piping section of the recombiner is designed to ASME Section III, Class 2 standards with exception taken to the l high temperature allowables based on the limited time of i operation at high temperature (code case 1481) .
! When the hydrogen recombiner section of the FSAR is updated, the recombiner information will be corrected to reflect the increased environmental qualification and to show that the recombiner will have a minimum capacity for hydrogen recom-l bination and gas flow of 70 scfm.
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022.58-3
D/B-FSAR
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QUESTION 022.72
() "Concerning the containment isolation design of the hydrogen recombiner lines to and from containment:
, a) Verify that the following containment isolation valves have positive position indication in the control room and are remote manually operable from the control room in accordance with SRP Section 6.2.4 11.5.c and ANSI N271-1976 Paragraph 4.2.2 and 4.2.3:
00G059 00G063 00G061 00G064 00G062 00G065 b) Describe the isolation provisions for the hydrogen recombiner discharge lines (00G45B 3 and 00G43B 3).
Although the normally open valves (00G060 and 00G066) in these lines are supplied with power from emergency buses, they must receive an automatic containment isola-tion signal, be remote manually operable from the control room, and have positive position indication in the control room to be acceptable as containment isolation barriers."
RESPONSE
The following valves make up part of the containment isolation barriers for the hydrogen recombiner and have positive position indicators locally mounted in the auxiliary building: 00G059, 00G061, 00G062, 00G063, 00G064, and 00G065. These valves are also remote manually operable. The revised P&ID for the hydrogen recombiner indicates compliance for the valves (00G060 and 00G066) to serve as isolation barriers. The hydrogen recombiner is not operated during modes 1 through 4, and the containment isolation valves are sealed closed (i.e., the breakers are racked out) . The panel is accessible following a LOCA. There-fore, there is no need to have positive indication in the main control ' room. Specific operator action following an accident is required to utilize a hydrogen recombiner.
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Q22.72-1 4
B/B-FSAR
/w QUESTION 022.79_
in detail, with text and figures , the permanently
" Describe, ~
installed leakage surveillance system used for continuous pressurization between the closure flanges of electricalProvide penetrations.or not the system will be enployed at all times during if normal State which electrical penetrations, plant operations.any, are not serviced by.this system; provide the testingS provisions for these cenetrations.this system is to be in lieu of con provisions. Describe the way in which leakage measured with this system is to be added to the total of Type B and Describe the high leakage alarm and its ._
Type C leakage. redundancy and single failure characteristics."
RESPONSE
The permanently installed leakage surveillance system is in service continuously. Nitrogen is used to provide cressurization The between the closure flanges of all electrical penetrations.
This system is used in
()systemnormallyoperatesat50psig. l lieu of conventional Type B testing. Surveillance of this I system and the calibration interval of its instrumentation will be performed at the same interval as periodic Tyoe B and I C tests.
The measured leakage will be added to the total Type B leakage, taking into account instrumentation accuracy.
The leakage rate of the electrical oenetrations is expected to l
l:
1 be insignificant in comparison with the total allowable contain-
! ment leak rate. The system is equipped with instrumentation to The detect and alarm on high or low pressure and high flow.
h The nitrogen system
- alarms are located in the main control room.
t is non-safety related and therefore, redundancy and single failure criteria do not apply. Operation of the system is not required to
() maintain the containment integrity.
l 022.79-l
(
N B/B-FSAR AMENDMENT 36 y! JANUARY 1982
] QUESTION 040.120 J
" Expand your description of the diesel engine starting system. The F5AR test should provide a detail system description of what is shown on Figure 9.5.3. The FSAR text should also describe: 1) components and their function, 2) instrumentation, controls, sensors and alarms, and 3) a diesel engine starting sequence.
In describing the diesel engine starting sequence include the number of air start valves used and whether one or both air start systems are used."
RESPONSE
General The engine is started using compressed air (250 psi) ,
furnished by two separate motor driven air compressors mounted on the starting air skid. Each compressor pumps air through a check valve past a relief valve through a refrigerant type dryer into an air tank. This is a dual system with either half capable of starting the engine.
Compressed air from the starting air tanks is applied to gs the starting air control valves, on the engine, which are
() controlled by starting air solenoid valves. When the starting air control valves open, starting air is supplied to both banks of air start valves and air distributors. One start valve is located in each cylinder head and all are controlled by the air distributors.
Normally both starting air control valves open simultaneously and air is taken from both tanks. If a malfunction occurs and one of the control valves fail to open, the crossover piping will admit air to the other bank of cylinders. As the air tanks lose pressure, the compressors start to replenish the air supply. If both compressors are inoperative, the l
air tanks have sufficient capacity to provide four starts with the existing pressure in the tanks.
Each individual air tank has sufficient capacity to provide for four starts.
The piping between the air start valve and the air control valve is continuously purged with combustion air from the turbocharger discharge. This compressed air, approximately 300*F in temperature, prevents the possibility of an explosion due to a leaky air start valve and prevents any moisture condensation buildup in the air start piping. In the unlikely event that some condensation occurs when the
( ) turbocharger purging air cools to ambient temperature, the l line filter in the air start will collect the moisture.
Q40.120-1
- B/B-FSAR AMENDMENT 36 JANUARY 1982 3 In addition, the filter contains a sight glass to show any water collected. The arrangement of_the air start piping prevents any possible condensation from accumulating on 1
the air start valves. Cooper Bessemer states that they have not experienced any problems to date with corrosion of the starting air system as designed.
Starting Air Controls Air aoplied to the starting air control valve is blocked but f'ows out an alternate port through a check valve to the manual control air valves. If all four shut-off valves are open, control air flows through the shuttle valve to 3 _the engine controls and shutdowns and through filters to the interlock valves. If the turning gear is disengaged,
, these interlock valves are open and air flows to the solenoid valves. If these solenoid valves are activated, air flows through them and a shuttle valve, thus admitting air to L the distributors and air start valves in each cylinder head.
i When the starting signal is turned off, air vents from the distributors and cylinder heads through the orificed check valves and out the vents in starting air control valves.
Air Compressors i
A panel on the air tank contains a pressure gauge, pressure switches to start .and stop the compressor automatically, 4
test valves and connections and shutoff valves. The compressors can also be started by manually setting the switch on the local
, control panel to the " HAND" position.
Eachcompressorisdrgvenbyu 15 hp motor and delivers 32.2 cfm to the 96 ft air tank. Due to the interconnecting piping, both tanks should be depleted at the same rate.
Therefore, both compressors will start to replenish the i
air supply at approximately the same time. The pressure switches on each air tank panel start the compressor at
. 240 psi falling and stop it at 250 psi. Prqssure relief L
valves are set at 265 psi.
040.120-2
i 1
B/B-FSAR t
- QUESTION 110.61
' " Expand Appendix Al to include a commitment to comply
- with Regulatory Guide 1.121."
a j RESPONSE
'g The -equested change was made in Amendment 36. !
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, Q110.61-1 i
.-----...,,,,-.m.,--, --,__n.,,- n.-,.n_,,,,. , - - - .- . , . , , _ _ , - - -- ,_,n_,,,,.,,---. - + , - , _ - - , - - - - . , , _ _ _ .
() B/B-FSAR C'i "(a) During initial system heatup and cooldown, at specified temperature intervals for any system which attains operating temperature, verify the snubber expected thermal movement.
" (b) For those systems which do not attain operating temperature, verify via observation and/or calculation that the snubber will accommodate the projected thermal movement.
"(c) Verify the snubber swing clearance at specified heatup and cooldown intervals. Any discrepancies or inconsistencies shall be evaluated for cause and corrected prior to proceeding to the next specified interval.
"The above described operability program for snubbers should be included and documented by the pre-service inspection and pre-operational test programs.
"The pre-service inspection must be a prerequisite for the pre-operational testing of snubber thermal motion.
This test program should be specified in Chapter 14 of the FSAR."
_/ RESPONSE The Pre-Operational Snubber Program will be defined by March 1, 1982.
However, a summary description of the program as presently conceived is given below.
The preservice inspection program will include the visual examination of all hydraulic and mechanical snubbers in-stalled on safety-related systems. The inspection will verify that the snubbers are installed correctly, and are L undamaged.
A list of all snubbers (both hydraulic and mechanical) on safety-related systems will be developed. Documentation will be provided to record the inspections conducted on .
each snubber. The documented inspection will be conducted no longer than 6 months prior to the preservice testing requirements.
For hydraulic snubbers the fluid will be verified to be k' at the recommended level and not leaking.
O Q110.63-2
O\
U B/B-FSAR O During hot functional testing, snubber thermal movements for systems whose operating temperature exceeds 250' F will be verified. The thermal monitoring program will be included in the test program. The thermal monitoring program consists of visual verification of snubber movements, as indicated on the snubber, from room temperature to maximum operating temperature. If maximum operating temperature is not attained during testing, the amount of movement expected will be calculated by multiplying the movement indicated on the )
snubber by the ratio of the temperature rise to the test ,
temperature (AT /6 g . If snubber movement differs from the expected mo9emen)t by more than 1/8 inch, an assessment I
will be made to verify that the snubber will satisfy its design function for the design load.
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O Q110.63-3 l
O V
B/B-FSAR QUESTION 221.3 9 "In our Safety Evaluation Report on WCAP-9500 ' Reference 9 Core Report 17X17 Optimized Fuel Assembly' the staff 9 required that those plants using the Westinghouse Improved 3
Thermal Design Procedure (ITDP) supply additional informa-
- tion on the plant specific application of the ITDP.
)
Since the applicant is using the ITDP to perform their 3
thermal-hydraulic analyses, they must comply with the q following:
5 (1) provide the sensitivity factors (Sg ) and their D range of applicability; if the S, values used in the Byron /Braidwood analyses 9 (2) are different than those used in WCAP-9500, then 9 the applicant must re-evaluate the use of an uncer-9 tainty allowance for application of equation 3-2
' of WCAP-8567, ' Improved Thermal Design Procedure' and the linearity assumption must be validated;
- (3) provide and justify the variances and distributions -
for input parameters; (4) justify that the nominal conditions used in the O analyses bound all permitted modes of plant operation;
- (5) provide a discussion of what code uncertainties, including their values, are included in the DNBR f analyses;
- (6) provide a block diagram depicting sensor, processing equipment, computer and readout devices for each parameter channel used in the uncertainty analysis.
- Within each element of the block diagram identify
- the accuracy, drift, range, span, operating limits,
- and setpoints. Identify the overall accuracy of each channel transmitter to final output and specify the minimum acceptable accuracy for use with the new procedure. Also identify the overall accuracy
' of the final output value and maximum accuracy I requirements for each input channel for this final output device; and I (7) If there are any changes to the THINC-IV correlation, or parameter values outside of previously demonstrated
- acceptable ranges, the staff requires a re-evaluation
- of the sensitivity factors and of the use of equation
- s 3-2 of WCAP-8567."
a Q221.3-1
B/B-FSAR O'" RESPONSE The response to this question will be provided by April 1, 1982 and will be generated in the following manner:
) 1. Operating parameter uncertainties, i.e., instrumentation
- uncertainties, will be determined using the approach
- outlined in the Westinghouse report, " Westinghouse Reactor
- Protection System / Engineered Safety Features Actuation
- System Setpoint Methodology" (Proprietary), submitted on the Donald C. Cook Nuclear Power Plant Unit 2, Docket No. 50-315 via Indiana & Michigan Power Company letter
- (J. Tillinghast) June 22, 1978, to NRC (E. Case). This a methodology was reviewed by the staff and approved
, via NRC letter (S. Varga) February 12, 1981, to Indiana 9 & Michigan Electric Company (J. Dolan).
L 2. Nuclear and thermal parameter uncertainties, and fuel s fabrication parameter uncertainties will be determined as noted in WCAP-8567, " Improved Thermal Design Procedure" (Proprietary) and approved by the NRC Staff.
. 3. The uncertainties for the operating parameters, nuclear and thermal parameters, and fuel fabrication parameters
/~D will be combined statistically as noted in WCAP-8567,
\- / " Improved Thermal Design Procedure" (Proprietary) and WCAP-9500, " Reference Core Report 17 X 17 Optimized Fuel Assembly" Section 4.4 as approved by the NRC Staff.
- 4. Information requested concerning process blowk diagrams
. will be provided in a manner similar to that provided
' in Westinghouse letter WS-TMA-1837 (T. Anderson) June 23, 1976, to NRC (S. Varga) concerning WCAP-8567.
O O221.3-2
B/B-FShR OhESTION281.4
- 'Your FSAR did not indicate that the refueling water storage tank, the boric acid mi,:ing tank, the chemien1 additive tank, and the su=p tank will be sampled. Confirm that these tanks will be sampled according to standard Review Plan 9.3.2.*
RESPONSE
The refueling water storage tank is sampled in accordance The boric acid mixing tank with Technical Specifications.
and reactor cavity cump have provisions to obtain sa=ples on demand. Due to its mode of operation, The it is not required spray additive tank to' sample the chemical additive tank.
is sampled in accordance with Technical Specifications.
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0261.4-1
O BYRON-FSAR QUESTION 421.22 "Section 17.1.2.2 of the standard format (Regulatory Guide 1.70) requires the identification of safety-related struc-tures, systems, and components controlled by the QA program.
You are requested to supplement and clarify Table 3.2-1 of the Byron Station FSAR in accordance with the following:
- a. The following items do not appear in FSAR Table 3.2-1.
Add the appropriate items to the table and provide a commitment that the remaining items are subject to the pertinent requirements of the FSAR operational QA program or justify not doing so.
- 1. Biological shielding within the. auxiliary building and fuel handling building
- 2. Missile barriers within the auxiliary building, fuel handling building, and other buildings and structures, as appropriate.
- 3. Non-safety systems that penetrate containment and
' /~h are an extension of the containment boundary
\-s up to and including the containment isolation l valves.
I 4. Fuel assemblies.
- 5. Control rod assemblies.
- 6. Core support structure.
- 7. Reactor vessel internals other than items 4, 5, and 6 above.
8.
Pressurizer spray nozzles.
- 9. Steam generator steam flow restrictors.
- 10. Reactor coolant system piping.
- 11. Reactor coolant pump seals and ancillary system components required for the reactor coolant pump to perform its safety function.
- 12. Reactor coolant pump seal bypass orifice.
O Q421.22-1
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BYRON-FSAR O
- 13. Pressurizer safety valves.
- 14. Pressurizer relief valves.
- 15. Pressurizer block valves.
- 16. Pressurizer control valves.
- 17. Other valves and piping within the reactor coolant pressure boundary.
- 18. Condensate storage tank, AFW system.
- 19. Instrument air accumulators and lines for valves that perform a safety function.
- 20. Spent fuel pool concrete structure.
- 21. Spent fuel pool liner.
- 22. CVCS letdown orifices.
() 23. Fuel transfer tube.
- 24. Control rod drive mechanisms.
- 25. Conveyor system and control.
- 26. Cask handling crane.
- 27. Refueling machine.
- 28. Spent fuel handling tool.
- 29. Interconnecting piping and valves of all systems in Table 3.2-1 that are classified as seismic Category I. Appendix B is applicable to those portions of the system required to perform a safety function.
- 30. Steel liner of the reactor building.
- 31. Supports for the containment Heat Removal System.
- 32. Piping and valves of the Containment Heat Removal System.
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- 33. Combustible Gas Control System a) H2 monitoring system b) H2 analyzer c) Supports
- 34. Containment Emergency Cooling System
- a) Cooling coils b) Ductwork dampers and supports
- 36. Raceway installations (.i.e., conduit and cable trays and their supports) containing a.c. and d.c. Class IE cables and other raceway installation whose failure could damage other safety-related items. (.See II-5, 19, and 44.)
- 37. Battery racks.
- 38. Diesel generator governor, voltage regulator, I and excitation system.
- 39. Underground cable system, cable splices, connectors, and terminal blocks. (See II-44) .
- 40. Control room HVAC system (See II-88) h) Humidifiers i) Condenser j) Charcoal filter housing
- k) Ductwork and dampers
- 1) Valves with safety isolation function m) Utility exhaust fan n) Electrical modules with safety function o) Cable with safety function
- 41. Containment Spray' (See II-15) e) Containment spray piping f) Valves g) Spray nozzles
- 42. Radiation monitoring (fixed and portable).
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- 43. Radioactivity monitoring (fixed and portable).
- 44. Radioactivity sampling (air, surfaces, liquids).
- 45. Radioactive contamination measurement and analysis.
- 46. Personnel monitoring internal (e.g. , whole body counter) and external (e.g., TLD system).
- 47. Instrument storage, calibration, and maintenance.
- 48. Decontamination (f acilities , personnel, and equipment).
- 49. Respiratory protection, including testing.
- 50. Contamination control.
- 51. Radiation shielding.
- 52. Foundation for refueling water storage tank.
- b. Enclosure 2 of NUREG-0737, ' Clarification of TMI N, Action Plan Requirements '
(November 1980) identified numerous items that are safety-related and appro-priate for OL application and therefore should be on Table 3.2-1.
These items are listed below. Add the appropriate items to Table 3.2-1 and provide a commitment that the remaining items are subject to the pertinent requirements of the FSAR operational QA program or justify not doing so.
NUREG-0737 Enclosure 2 Clarification Item
- 1. Plant safety-parameter display console. I.D.2
- 2. Reactor coolant system vents. II.B.1
- 3. Plant shielding. II.B.2
- 4. Post accident sampling capabilities. II.B.3 O
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- 5. Valve position indication. II.D.3
- 6. Auxiliary feedwater system II.E.1.'l
- 7. Auxiliary feedwater system initiation and flow. II.E.1.2
- 8. Emergency power for pressurizer heaters. II.E.3.1
- 9. Dedicated hydrogen penetrations. II.E.4.1
- 10. Containment isolation dependability. II.E.4.2
- 11. Accident monitoring instru- ~
mentation. II.F.1
- 12. Instrumentation for detection of inadequate core cooling. II.F.2
') 13. Power supplies for pressurizer v relief valves , block valves , and level indicators. II.G.1
- 14. Automatic PORV isolation. II.K.3(1)
- 15. Automatic trip of reactor coolant pumps. II.K.3051
- 16. PID controller. II.K.3(9)
- 17. Anticipatory reactor trip on turbine trip. II.K.3(12)
- 18. Power on pump seals. II.K.3(25)
- 19. Emergency plans (and related equipment) . III.A.l.1/III.A.2
- 20. Equipment and other items
' associated with the emergency support facilities. III.A.l.2
- 21. Inplant I2 radiation monitoring. III.D.3.3
- 22. Control room habitability. III.D.3.4 Q421.22-5
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RESPONSE (
A. TABLE 3.2-1 ITEMS '
As discussed in Section 3.2, Table 3.2-1 lists princi-pal structures, systems, and components and identifies the safety category and quality group of these items.
Although it is not feasible to include each component in this table, an attempt has been made to illustrate the approach used in assigning the safety category and quality group. As noted in Section 3.2.1.1, the safety category of each piping run, valve, etc. and the divisions between Safety Category I and II can be deter-mined from the P&ID's. (The response to Question 212.7 g lists all P&ID's in_the FSA?. In addition, one full size set of P&ID's was tran c mitted to the NRC in March, 1979, along with the responta to Question 212.7.) . Each of the items identified in Eart of this question has s been investigated. The results are listed below by item number.-
- 1. The safety category for the Auxiliary Building, Fuel Handling Building, and other structures is listed in Table 3.2-1 under Section I-1 and I-2.
This safety category alr lies to internal gstructures with'in the i _ led buildings.
(3 2. See Item 1 response.
\_/
'3. All containment penetrations and isolation valves are Safety Category I regardless of the safety classification s of the system. Because of the large number of penetrations involved, these are not A identified individually i'n Table 3.2-1. The containment penetrations and isolation valves are listed in Table 6.2-58 and may be found on the system P&ID's.
- 4. The fuel assemblies are designated by Westinghouse as ANS Safety Class 2 components. They are designed and fabricated under the full provisions of a i
10CFR50 Appendix B quality assurance program. They will also be subject to the pertinent requirements of the operational quality assurance program.
- 5. The control rod assemblies are designated by Westinghouse as ANS Safety Class 2 components. They are designed and f abricated under the full provisions of a 10CFR50 Appendix B quality assurance program. They will also be subject to the pertinent requirements of the operational quality assurance program.
) 6. See revised Table 3.2-1. This is included in item l
II.119.d.
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- 7. See revised Table 3.2-1. This is included in item !
II.119.d )
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- 8. This is included in item II.70.a. l l
- 9. This is included in item II.70.a of Table 3.2-1.
- 10. See revised Table 3.2-1, Item II.64.d.
- 11. See revised Table 3.2-1, Item II.64.e.
- 12. This is included in Item II.64.d.
- 13. See revised Table 3.2-1, Item II.70.d.
- 14. See revised Table 3.2-1, Item II.70.e.
- 15. These valves are Category I. As noted in Section 3.2.l.1, the safety category of piping and valves may be found on the system P&ID's. As previously noted, Table 3.2-1 illustrates the Safety Category and Quality Group break-down for principal structures, systems and components only.
- 16. These valves are Cctegory I. As noted in Section 3.2.1.1,
- the safety category of piping and vc1ves may be found on the system P&ID's. As previously noted, Table 3.2-1 illustrated the Safety Category and Quality Group break-down for principal structures, systems and components only.
- 17. See revised Table 3.2-1, Item II.64.d.
- 18. As noted in Table 3.2-1, Item II-11, the Condensate System is Safety Category II. However, pertinent requirements of the operational QA program will be applied to this system. As shown in Item II.3, the Auxiliary Feed Water System is Category I.
- 19. No air operated valves perform a safety function on the Byron /Braidwood stations.
- 20. See Item 1.
- 21. See Item 1.
- 22. See Item 1.
- 23. See revised Table 3.2-1, Item II.31.g.
- 24. See revised Table 3.2-1, Item II.65.
O 25. See revised Table 3.2-1, Item II.31.g.
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- 26. The cask is handled by the Fuel Handling Building 3 crane which is listed in Table 3.2-1 (Item II-31e) .
- 27. See revised Table 3.2-1.
- 28. This is included in Item II.31.g.
- 29. As noted in Section 3.2.1.1, the safety category of piping and valves may be found on the system P&ID's.
As previously noted, Table 3.2-1 illustrates the Safety Category and Quality Group breakdown for principal structures, systems and components only.
- 30. See Item 1.
- 31. The safety category for supports is the same as the pipe or equipment being supported. Items I-2a and I-2b in Table 3.2-1 are being revised to clarify this.
- 32. See Item 29.
- 33. a. H2 monitoring system will be added to Table 3.2.1.
Tnis is a Category I system. ,
- b. H2 analyzer will be added to Table 3.2-1. This is a Categcry I system.
() 34.
c.
a.
See Item 31.
As shown in Table 3.2-1, Items II-98n and II-98o, the RCFC Essential Service Water Coils are i Category I.
- b. As discussed in response to Item 31, the dampers and supports necessary for operation of Category I systems are designed as Category I.
- 35. a. The hydrogen recombiners listed in Table 3.2-1, Item II-53a.
- b. Not applicable.
- c. See Item 29.
- 36. See Item 31.
- 37. Thds equipment is already covered by Item II.19 of Table 3.2-1.
- 38. This equipment is covered under Item II.20 of Table 3.2-1.
- 39. Item II-44a includes all equipment necessary for Category I Instruments and Control to perform their safety func-O~ tion, l
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DYRON-FSAR Os g 40. Although not all components in the control Room HVAC system were designed as Category I components, the pertinent provisions of the operational phase QA program will be applied to this entire system.
Control room HVAC equipment is treated as follows:
- h. Humidifier will be added to Table 3.2-1.
- i. The Condenser is included in Table 3.2-1, Item II-11.a.
- j. Charcoal Filter housing will be added to Table 3.2-1.
- k. Ductwork and dampers are not included in this table. The safety category may be found from the P&ID's or equipment lists as noted in Section 3.2.1.1.
- 1. The control room HVAC does not have safety isolation valves.
- m. The utility exhaust fan will be added to Table 3.2-1.
- n. Electrical modules wit'h safety functions are
() o.
included in Table 3.2-1, Item II-88f.
Cables with safety functions are included in Table 3.2-1, Item II-88f.
- 41. See Item 29.
- 42. The fixed equipment is covered under Item II.6 of Table 3.2-1. Although not all components of this system were designed as Category I components, per-tinent equipment of the operational phase QA program will be applied to this system. Portable equipment is purchased under site QA and thus is not included in Table 3.?.-l. -
l l 43. The fixed equipment is covered under Item II.61 of
- Table 3.2-1. Although not all components of this system were designed as Category I components, pertinent requirements of the operational phase QA program will be applied to this system. Portable equipment is purchased under site QA and thus is not included in Table 3.2-1.
- 44. This equipment is covered under Item II.61 and 62 of Table 3.2-1. Although not all components of this l
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S system ware designed as Category I components, pertinent requirements of the opeational phase QA l program will be applied to this system.
0421.22-9
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'h 45. This item does not constitute principle structures, systems, or components of the plant, and thus is not included in Table 3.2-1.
- 46. See Item 45.
- 47. See Item 45.
- 48. See Item 45.
- 49. See Item 45.
- 50. See Item 1.
- 51. See Item 31.
B. NUREG-0737 ITEMS NUREG-0737 Items are addressed in Appendix E. As discussed previously, Table 3.2-1 gives general information on the safety classification of equipment corrections required as a result of NUREG-0737 items have been made to Table 3.2-1 but not all NUREG-0737 items correspond to Table 3.2-1 items.
The items listed have been investigated and are discussed here.
) 1. The plant safety-parameter display console is being s,/ installed at Byron /Braidwood stations. (See Section E17) This is included in Table 3.2-1, Item II.60.a.
- 2. The Reactor Coolant System Vents are discussed in Section E.19. This is included in Table 3.2-1, Item II.64.a.
- 3. NUREG-0737, Item II.B.2, Plant Shielding, is a design review of the plant shielding and postaccident.
radiation levels and therefore, does not directly involve plant structures or equipment. As discussed in response to Part A.1 of this question, safety category of shiciding is equivalent to that of the building in which it is located. -
- 4. The post accident sampling capabilities are discussed in Section E.21. This system is included in Table 3.2-1, Item II.62, and is a Category I system.
- 5. The indication of relief and safety valve position is explained and the safety category requirements discussed in detail in Section E.24. The instrumentation is the same safety category as the valves it is provided for, i.e., Category I.
() 6. The auxiliary feedwater system is discussed in Section E.25 and included in Table 3.2-1, Item II.3.
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- 7. This issue will be responded to later in Section E.26.
T Table 3.2-1, Item II.3 shows that the auxiliary feedwater s l system is Category I.
- 8. Emergency Power Supply for Pressurizer Heaters is discussed in Section E.27. Pressurizer Heaters are included in Table 3.2-1, Item II .70.c. A Category I emergency power supply is available for these heaters.
It is covered in Item II.S.a of Table 3.2-1.
- 9. Dedicated Hydrogen Penetrations are provided as -
discussed in Section E.28. The hydrogen recombiner systems included in Table 3.2-1, Item II.53.a.
- 10. Containment Isolation Dependability is discussed in Section E.29. The penetration isolation is a function of may systems and as such does not appear on Table 3.2-1.
- 11. Accident Monitoring Instrumentation is discussed in Section E. 30. Any modifications made as a result of this review will be under the pertinent provision of the QA program, including the operational phase thereof.
This equipment will be Category I.
- 12. Instrumentation for detection of inadequate core cooling (NUREG-0737 Item II.F.2) will be addressed in Section E.31. Any modifications made as a result of this review will be under the pertinent provision of the QA program, including the operational phase .
- thereof . This equipment will be Category I.
- 13. Power supplies for pressurizer relief valves, block valves and level indicators are addressed in Section E.32. These items are included in Table 3.2-1, Item II.44.a.
) 14. Automatic PORV isolation as addressed in Section E.49.
Any modifications made as a result of this review will be under the pertinent provisions of the QA program, including the operational phase thereof. This equipment will be Category I.
- 15. Automatic trip of reactor coolant pumps is addressed in Section E.52. Any modifications made as a result of this review will be under the pertinent provisions of the QA program, including the operational phese thereof.
This equipment will be Category I.
- 16. The PID controller is discussed in Section E.54. Any modifications made as a result of this review will be
("3 under the pertinent provision of the QA program, including
(,,/ the operational phase thereof. This equipment will be Category I.
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- 17. The anticipatory reactor trip on turbine trip is discussed in Section E.57 and included in Table
() 18.
3.2-1, Item II.69.
Power on pump seals is addressed in Section E.66 and included in Table 3.2-1, Items II.5.a and II.20.
- 19. This subject will be addressed in Appendix E. Procedures and equipment required to resolve this item will be under the pertinent provisions of the operational phase QA program.
- 20. This subject will be addressed in Appendix E. Procedures and equipment required to resolve this item will be under the pertinent provisions of the operational phase QA program.
- 21. The inplant iodine radiation monitoring system is discussed in Section E.78. Any modifications made as a result of this review will be under the pertinent provisions of the QA program, including the operational phase thereof. This equipment will be Category I.
- 22. HVAC System Equipment necessary to maintain Control Room habitability is identified in the FSAR Table 3.2-1, Section II-88, a through f. The system is designed to maintain habitability with the equipment provided.
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