ML20082M896
| ML20082M896 | |
| Person / Time | |
|---|---|
| Site: | Satsop |
| Issue date: | 11/22/1983 |
| From: | Sorensen G WASHINGTON PUBLIC POWER SUPPLY SYSTEM |
| To: | Knighton G Office of Nuclear Reactor Regulation |
| References | |
| G3-83-893, GL-82-14, GO3-83-893, NUDOCS 8312060322 | |
| Download: ML20082M896 (47) | |
Text
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Washington Public Power Supply System Box 1223 Elma, Washington 98541-1223 (206)482 4428 Docket No. 50-508 November 22, 1983 G03-83-893 Director of Nuclear Reactor Regulation ATTN: Mr. G. W. Knighton, Chief Licensing Branch No. 3 U. S. Nuclear Regulatory Commission Washington, D. C. 20555
Subject:
NUCLEAR PROJECT 3 RESPONSES TO NRC QUESTIONS In accordance with the guidance of Generic Letter 82-14, the Supply System hereby submits 40 copies of responses to the NRC's requests for Additional Information as shown.
In preparing this submittal it was necessary to revise several large Tables from the FSAR. Since as a practical matter it is quite dif-ficult to include a copy of each for each copy of this letter, the NRC Licensing Project Manager will receive three copies of each for Distribution.
This situation has been discussed with the NRC Licensing Project Manager for WNP-3.
8312060322 831122 PDR ADOCK 05000 (
) -
Mr. G. W. Knighton Page 2 NUCLEAR PROJECT 3 RESPONSES TO NRC QUESTIONS If you require additional information or clarification, the Supply System Point of Contact for this matter is Mr. D. W. Coleman, Licensing Project Manager (206/482-4428 ext. 5436).
Sincerely,
.' W fy G. C. Sorensen, Manager Regulatory Programs GCS/kh Attachments: 1. NRC Question No. 430.21
- 2. NRC Question No. 430.36
- 3. NRC Question No. 430.37
- 4. NRC Question No. 430.61
- 5. NRC Question No. 430.64
- 6. NRC Question No. 450.6
- 7. NRC Question No. 450.12
- 8. NRC Question No. 460.1
- 9. NRC Question No. 471.6
- 10. NRC Question No. 480.13
- 11. NRC Question No. 480.17
- 12. NRC Question No. 480.18
- 13. NRC Question No. 480.21 ,
cc: P Christofakis - Ebasco NYO N. S. Reynolds - 0 & L J. A. Adams - NESCO D. Smithpeter - BPA A. Vietti - NRC A. A. Tuzes - CE Ebasco - Elma WNP-3 Files
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Question No.
430.21 Discuss the precautionary measures that will be taken to assure (SRP 9.5.4) the quality and reliability of the fuel oil supply for emergency diesel generator operation. Include the type of fuel oil, impur-ity and quality limitations as well as diesel index number or its equivalent, cloud point, entrained moisture, sulfur, particulates and other deleterious insoluble substances; procedure for testing newly delivered fuel, periodic sampling and testing of onsite fuel oil (including interval between tests), interval of time between periodic removal of condensate from fuel tanks and periodic system inspection. In your discussion include reference to industry (or other) standard which will be followed to assure a reliable fuel oil supply to the emergency generators.
Response
The specific standards that will be followed to assure a reliable supply of fuel oil to the emergency diesel generators are:
(1) ASTM-D240-65, Standard Method of Sampling Petroleum and Petroleum Products.
(2) ASTM-0975-74, Standard Classification of Diesel Fuel Oils.
(3) Manufactures Diesel Fuel Recomendations, Maintenance Instruction 1750 Rev. D, March 1973, EMD - General Motors Corporation.
The fuel is No. 2-D and meets the requirements of the following specifications:
Requirement Specification 2 Analysis Method
- 1. Cetane Number 50 min, 57 max ASTM-D-613 ,
- 2. 90% Boiling Pt 600 max ASTM-D-86
ASTM-D-86 Final Boilding Pt 625'F max ASTM-D-86 t
Distillation Recovery 99.0% ASTM-D-86
- 3. Total Sulphur .2% max not to ASTM-D-129 or exceed legal limit ASTM-D-1552
- 4. Copper Strip Corrosion No. 1 Strip or ASTM-0-130 3 hr # 212*F better Modified
- 5. Carbon Residue .15% max ASTM-0-189 or 10% bottoms ASTM-D-524 l
- - - - . ~ . _ _ _ . . ..,,____ . _ . , , __ _ . . _ _ _ _ _
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. l Question No.
430.21 Response (Cont'd)
Requirement . Specification 2 Analysis Method
- 6. Ash, wt% trace max ASTM-D-482
- 7. Water and Sediment trace max ASTM-D-96
- 8. Cloud & Pour Point ASTM-D-975 ASTM-D-2500
- 9. Flash Point 150*F min ASTM-D-93
- 11. Filtration Cleanliness 1.3 mg/l max on ASTM-D-2270 0.80 Micron Filter
- 12. ASTM Color 1.0 max ASTM-D-1500
- 13. Viscosity, 100*F 2.4 min ASTM-D-445 Kinenatic, est. ----
- 14. Gravity, API 37 max ASTM-D-287
- 15. Thennal stability 16 hr 15 mg max ASTM-D-2274
- 16. Aniline Point, F 115 min ASTM-D-611
- 17. Neutralization Number 0.2 max ASTM-D-974 Representative samples shall be collected from new delivered fuel. -
Representative samples of diesel fuel from the diesel fuel storage tanks will also be collected quarterly (every 92 days).
Viscosity, API gravity, water and sediment, and flash point will ;
be determined for each sample. If the diesel fuel does not meet '
the manuf acturer's specifications it will be replaced or, in the i case of water and sediment, treated to remove imputities.
See the response to Q430.17 concerning the removal of water frora the diesel fuel storage tanks.
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. t Question No.
430.36 Provide the design dew point temperature (OF) for the compressed (SRP 9.5.6) air as it leaves the refrigerant air dryers. Show that the design dew point will be at least 100F lower than the lowest possible ambient temperature ,in the diesel generator room. Discuss tne procedures that will be implemented to ensure that compressed air dew point design is maintained and that moisture does not collect in the air receivers.
Response
The compressed air design dew point temperature is 35'F (1.7'C).
This temperature is maintained in the heat exchanger automatically by a thermostat that controls the operation of the refrigeration system. The moisture is collected in the condensate separator and is periodically drained (see response to Question 430.37). As described in Subsection 9.4.5.2.1, tne lowest ambient temperature in the DG room is 60*F during all conditions of operation and varying outside air temperature. This is 25'F higher than the automatically maintained design dew point temperature of 35'F.
Table 9.5.6-1 will be amended to reflect the design dew point of the diesel generator compressed air system.
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133N-1 WN3 FSAR A TABLE 9.5.6-1 DIESEL GENERATOR STARTING SYSTEM EQUIPMENT DESIGN DATA
- 1. Air Compressors Operator motor,15 HP, 3 ph, 60 Hz, 460V 1 Design Pressure, psig 700 Design Temperature, F 139.
ntity/ Engine 2' Seismic Category I
- 2. Air Receivers 3 54.7 Volume ft Design Pressure, psig 700 i I1 Design Temperature, F 139 Quantity / Engine 2 Seismic Category I ASME Section III
- 3. Air Dryees
/" Type Refrigerant 700 l1
, (~ Design Pressure, psig Design Temperature, F 139 Quantity / Engine 2 Seismic Category I l
- 4. Shutdown Air Accumulator .
Volume, Cu. inches 230 I
Design Pressure, psig 700 Operating Pressure, psig 670 l t.
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Design Temperature, F 140 Quantity / Engine 1 ASME Section III l
Destew o.w 35 Pouoty f~
G N30.3G Ge:M I70 9.5-42 Amendment No. 1, (10/82) l
. i Question No.-
430.37 Describe the provisions in the design of your compressed air (SRP 9.5.6) system which prevent accumulation of dirt and oil in the receiver and/or other parts of the system.
Re sponse
- The Ingersoll-Rand refrigeration type air dryers are installed .
between the air compressors and receivers. They are utilized to remove moisture, dirt, oil vapor and other contaminants from com-pressed air. The moisture, dirt and oil vapor are collected in the air dryers (condensate separator) and periodically drained and cleaned in accordance with the procedure provided in Vendor Instruction Manual.
Subsection 9.5.6.2 will be amended to reflect the response to this question.
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,. 1322W-2 WNP-3 FSAR and foreign materials. Both air starting systems are used for normal starts.
Therefore, the combined capacity of the dual train system is 10 cold start cycles without recharging the air receivers. The average expected cranking cycle duration is three to three and one half seconds. A limit switch is l provided to IIait the durat' ion of cranking cycle to a maximum of seven seconds. If this duration exceeds seven seconds the air supply for cranking i will be shutoff. The air receivers are provided with a pressure switch to atart and stop the compressor as required. Each air train is provided with a low pressure alare. The compress' ors are not required during the starting
- operation or during diesel engine operation. 'No complete automatic g ref rigerant type air dryers on the discharge side of the air compressors ' '
_control.d provide moisture-f A check ulve reelocated air by reducing upstreamthe of air thedewpoint for ensures air receiver starting and that a
{
lu broken line in the non-safety portion of the piping will not result in a W sudden loss of air. The main air start control mlve and air start solenoid 2 , wives are used for startup of the system.
H A shutdown air accumulator with a solenoid operated shutdown 1alve is provided, which holds enough air trapped by a check valve on the upstream of the accumulator to close the fuel racks to shut the diesel engine down even if both starting air systems lose pressure. The design data for major components of this system is shown in Table 9.5.6-1.
9.5.6.3 Saf ety Evaluation The portion of the diesel generator starting system between the diesel generator and the first check im1ves upstream of the air receivers is classified as Nuclear Safety Class 3 and designed to ASME Section III, Class 3 seismic Category I requirements to ensure system operation during an SSE.
The non-saf ety portions of the starting air system are designed in accordance with the applicable codes and standards listed in Table 9.5.4-2. The system y is shown on Figure 9.5.6-1.
The air starting systems for one diesel are physically and electrically separated f rom those for the other diesel to assure that no single failure can cause malfunction of both divisions of standby ac power. l1 The single failure criterion is satisfied and significantly enhanced by having a dual train air starting syeten for each diesel generator.
i The dual train starting system ensures that a tailure of any components of any , ,'
train cannot cause loss of system ability to start the diesel, to supply i emergency power, so as to saf ely sitigate the consequences of an accident and saf ely shutdown the reactor. A failure in the non-saf ety portion of the system piping will not have any impact in the saf e operation of the system. A ,
f ail ure mode and ef fects analysis of the DGSS is presented in Table 9.5.6-2.
9.5.6.4 Inspection and Testing Requirements The system will be operated and tested initially with regard to flow path, flow capacity and mechanical operability in accordance with Section 14.0. To
.Th D
G430.37 Sc N S7/
9.5-40 Amendment No. 1, (10/82)
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- Question 430.37 INSERT 1 Moisture, dirt, oil vapor are removed from the compressed air. They are collected in the condensate separators, which are periodically drained and cleaned in accordance with.the Vendor Instruction Manual.
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. t Question No.
430.61 Discuss your inservice testing and inspection program for the (SRP 10.2) motor operated steam extraction valves such as provided for the turbine governor, control, interceptor, and reheat stop valves.
Response
Inservice inspection is not covered by SRP 10.2 for turbine extraction motor operated valves. The acceptance criteria (10.2.II.3) is directed at extraction steam non-return valves limiting steam flow so turbine speed stablizes. The motor oper-ated valves do not' perform this function.
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- Question No.
430.64 In Section 10.4.1.4 you have discussed tests and initial field (SRP inspection but not the frequency and extent of inservice inspec-10.4.1) tionof the main condenser. Provide this information in th.e FSAR.
Response
Inservice inspection is performed in two ways:
a) Continuous monitoring of condensate to detect leaks and;
' b) Periodic (refueling outage intervals) inspections to assess conditions of tubes in known, industry wide problem areas, and to assess condenser air in leakage.
The Supply Systen will continue to follow industry efforts to im-prove condenser leakage detection and will select additional in-spection methods as the worth of these inspections is proven.
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- Question No.
450.6 The operating procedures for responding to a steam generator tube (15.6.3) rupture are currently an open issue on CESSAR. Resolution of the bases for analysis of this accident must either be accomplished for CESSAR, or on a plant specific basis for WNP-3. In order to resolve the issue for WNP-3, the following question must be an-swered. The current operating procedures call for the operators to steam the affected steam generator to prevent overfilling. the present steam generator tube rupture accident evaluation in the FSAR assumed tht no releases occur from the affected steam genera-tor after 30 minutes. Describe why the CESSAR-FSAR evaluation is bounding for a steam generator tube rupture event in light of the operator action guidance.
Response This issue has been addressed on the CESSAR-F Docket via letter LD-83-066, from A. E. Scherer to D. G. Eisenhut, dated July 22, 1983, " Confirmatory Item 18 Steam Generator Tube Rupture Event".
Additionally, guidance for operator actions during emergencies is contained in CEN-152, " Emergency Procedure Guidelines" which have been approved by the NRC via a Safety Evaluation Report forwarded by letter from D. G. Eisenhut to R. W. Wells, dated July 29, 1983. The WNP-3 emergency procedure will be developed from these guidelines, which include specifics concerning Steam Generator Tube Rupture.
Once confirmatory Item 18 is accepted by the staff the Supply Sys-tem will reflect the most limiting SGTR event in the WNP-3 FASR, as appropriate.
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Question No.
Because WNP-3 intends to reference CESSAR for certain accidents,
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450.12 demonstrate how the interface requirements for CESSAR are met.
Response ,
Chapter 15, as presently written, does not' identify any inter-face conditions as specified in CESSAR System 80. It should, however, be noted that the FSAR sections which describe those systems that are used to mitigate the accidents described in Chapter 15 do reference and denote which interface requirements are met. Those sections also note those interface requirements which have been modified, as required. Furthermore, FSAR Sec '
tion 1.9 lists all CESSAR interface requirements, cross-refer-ence to interface requirements, FSAR compliance status with the CESSAR requirements, and the FSAR sections wnich address the NSSS requirements.
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o Question No.
460.1 Supply information relating to the effluent radiation monitors (11.5.2.4.2) for steam generator blowdown flash tank vent and steam seal
. gland steam condenser ventilation which the FSAR indicates as later or proyide a schedule for submittal of this information.
Response
As committed to in Letter #G03-82-1085, dated October 22, 1982.
FSAR Section 11.5 will be updated as shown to include the re-quested information.
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WNP-3
= *1633W-4 FSAR
'~N- pre-established setpoints an annunciation is made througn the system
) CRTs and event typer. If the activity exceeds ths high radistion alarm setpoint, or if the monitor fails, as determined oy the local microprocessor, a contact closure is made at the local microprocessor which is used to automatically terminate the waste gas discharge.
The receipt of these alarms will alert the operators to analyze additional gas samples to determine the reason for the alarm. The records of the total quantity of radioactive material released is c. sed in writing the reports required by Regulatory Guide 1.21. The alarm setpoints are selected in consideration of the requirement to prevent activity concentrations at the plant boundary or beyond from exceeding 10CFR20 limits, and to support the release limits set in the plant technical specification. The setpoints may be adjusted continuously '
over the entire range of the monitor. The range of this monitor was selected to span the expected range of radioactive gas concentrations expected in the waste gas, c) Steam Generator Blowdown Flash Tank Vent Radiation Monitor The steam generator blowdown flash tank vent radiation monitor provides plant operations personnel with an indication and record of contamination of the Secondary Steam System and tne potential for release via the steam generator blowdown flash tank vent. This contamination could occur due to leakage of primary reactor coolant into the secondary coolant through a steam generator, y d/edt.r -
I 7'~'j This monitor is an ambient type monitor located next to the_ steam of /
generator blowdown flash tank vent line 6BD12-200tz: El. ':22 ft. Lv the Turbine Building. I mon or i colli ted w th a 1 d sh ld t ,,jr
' red a th effe o back ound The ashie ed por ion o the de ctor sa u bstru ted v ew of he ve t line The mbie Snha _efon 1 .5.2_ 2 l nitor in de cri d i _
. 1 The measured activity level is automatically transmitted to the system )
! computer where it is recorded and available for display. If the l activity exceeds setpoints an annunciation is made through the system l CRTs and event typer. Receipt of these a'larms will alert the operator (
to the possibility of contamination of the Secondary Steam System and indicate the need for additional sampling and further action.
The alarm setpoints are selected above plant background to give the greatest sensitivity for possible contamination without causing frequent false alarms'. These setpoints may be adjusted continuously over the entire range of the monitor.
l f) Auxiliary Condensate Flash Tank Radiation Monitor The auxiliary condensate flash tank radiation monitor providu_ plant operations personnel with an indication and record of contamination of l the Auxiliary Steam System and the potential for release via various vents in the Auxiliary Steam and Condensate System. This contamination 6
ould occur due to inleakage f rom the various radioactive systems that i2
$( e serviced by the Auxiliary Steam System.
bk4b0s i Scd $55At 11.5-23 Amendment No. 2, (12/82)
h l L,%.14 As0 cn}
VNP-3 yta +,. uu, 5.,g 16334-5 FSAR ALW blq
% .14 9 .
This monito is an ambient type monitor located next tothejxiliary gondensate asn/Innkti: :a: d ;;;e; . 6 1..., L.le.u. ua w Lue 59912-?nn e- the 335 ft 1e"21. '%e erniter ir cr114-~-d it' 2 Irrd 2
- hi;1d
- : ::d::: ::: cff::: ;: :-:: ..;;: _ d . Th: :::ti:ll:f ;:::irr :f th: d::::::: h:: :: :::i::::::22 ci:1 :f : r f12:5 :: '; : The ambient monitor is descrioed in Subsection 11.5.2.3.
The measured activity level is automatically transmitted to the system computer where it la recorded and available for display. If the activity exceeds setpoints an annunciation is made tnrough the system CRTs and avent typer. Receipt of these alarms will alert the operators to tne possibility of contamination of the Auxiliary Steam and a Condensate System and indicate the need for additional sampling and further action.
The alarm setpoints are selected above plant background to give tne greatest sensitivity for possible contamination without causing frequent false alarms. These setpoints may be adjusted continuously over the entire range of the monitor.
- 6) Steam Seal Gland Steam Condenser Exhaust Radiation Monitor Tne steam seal gland steam condenser radiation monitor provides plant operations personnel with an indication and record of' contamination of the Secondary Steam System and the potential for release via the steam ,
segl gland steam condenser vent. This contamination could occur due to (. .
leakage of primary reactor coolant into the secondary coolant through a steam generator. 34 m to.
rce The monitor is an ambient type monitor located next to the vent line (bAE10-022) d.I SL. 155 ft. of the Turbine Building. _
m itor 's colli ted ich a ead ield 'o r uce t ef et o 2 ack round The u hie ed po ion the etec or na an obst ete vi of e vent ine. Ihe bien monit r is escr ed i Sub cti 1 .5.2. .
The measured activity level is automatically transmitted to the system computer where it is recorded and available for display. If the activity exceeds setpoints an annunciation is made enrougn the system CRTs and event'typer. Receipt of these alarms will alert the operators to tne possibility of contamination of the Secondary Steam System and indicate the need for additional sampling and further action.
The alarm setpoints are selected above plant background to give the greatest sensitivity for possible contamination witnout causing frequent false alarms. Tnese setpoints may be adjusted continuously over tne entire range of the monitor.
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ll.5-23a Amendment No. 2, (12/82)
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wra.3 FSAR TAnlE 11.5-1 (Cont 'd) .
Typical Design t'a me Background Alarm Set- Aast oma t ic Sampler Activity Sensitivity Range (Imtrument Tag Number) points Actions
_Q [ml/hr Co-60) J pe Measured tB Background PCi/cc . j Ci/cc initiated location Duty Administration Building i 2.5 Two Cross # 1 10-9 pCi/cc Disch2rge Nadiation l u l 0-I'3 to 3al0*IU Alarm inside Gintinuous Stage Particu- Sr-90 t mR/hr lx10~5 3a10*9 tbnitor (NE-RH-0008 & Only llR-51 Airborne late Co-60II CE-RH-cuo9) & Gas 1s10-6 ,c; fee g,gg-7 to 5:10-7 Kr-85 Co-60(g) 1 ma/hr talo-2 3:30-6 Condenjar Hechanical 2.5 One 1x10'0 pCi/cc Vicuum rump Discharge 1
Cross # lul0~I to 1m10 Alarm Line Continuous Stage Cas Xe-133 1 mR/hr 1 10-2 1:10*3 Nadtstson Monitor Only 6AEl63024 Aarborne Co-60I3
( k ti- At.-1400)
Wsste Cza Discharge 2.5 One 1 10-2 pCi/cc kidistian Monitor 1
Cross # tal0-3 to 2x10~I Alarm. 6CHl- Batrh Stage . Cas Kr-85 m*/hr 1 10'3 2x100 (NE-c 'w8) Airborne Termina- IlFR Co-60U) tion of Discharge Steam Generstor i Ambient Blowdown Flash Tann Vent CrossI N lul0*I to Above ambient Alarm Line Continuous Nxdiction Monttor -
- /0@/44(./) 1:10' mR/hr background Goly 68012-200 2
Ausiliary Con.lensate Flash 1 2.5 Aiab ie nt Cross ? J a * :^ - tal0'I to (.irt 4 A $9-03 7 lank Hidiatinn Honitor Above ambient Alarm h nx? Continuous
- */gc ,e 1 10% mR/hr background Only Acad,-kn*-
Co luis- [
M ::'
nated Sterm Citnd Seal Steam 1 2.5 Ambient Crossi g Isl0'I to Exhaunt Radiation Above ambient Alarm 1.ine Continuong Honitor Colue 7pg eg/pgy 1 100 mR/hr background only 6AElu-022
/
ate Hot M.e c n o ne Saiup Discharge i 1.5 Samplar None N.A.
Nample r I N F -Htt-UO l o)
N.A. H.A. None In*.ide a:ont'e nuous t:st-5 8 Waste Man.ageiwnt System i 1.5 Liquid CrossY lul0-6 ,ggfe, 3,gg-7 to Isi s cha r ge 5:10~S Alarm s.ine hatch Cs-137 I eM/hr 5 10-2 5x104 Termina- 68.S)-080 H.e.1 s.eI a on Hon a t or Co-60II Ikt-w1-621B) tion el WNP-)
D'scharge bl.S )-101 W N I'- 5 E .. e am o n l'I .e n t tilla.nt 1 ..e 1.i gee l d Cross y inlu~D p Ci /r e- Sal 0~ to 5 : 1 88 Alana l'ou t i nes.m H.e.l e.it s on Houstor 1.i nee tut.-i w-MM sus Cs-l)7 ( I mR/hr 5 10*2 Sz illO Unly brid / l -
Co-holl8 ogn ,;3 ho w utu+ rs:uy &Q,
' 2-'"
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Question No.
471.6 Regulatory Guide 1.70 and the SRP (NUREG-0800) state that the (12.5.2) description of the health physics instrumentation should include the instruments sensitivity. You provided the type of radiation the instrume,nt detects and not the instrument sensitivity in Table 12.5-1. Provide the requested information.
Response
As committed to in Letter #G03-82-1085, dated October 22, 1982.
FSAR Section 12.5 will be updated as shown to provide the re-quested informatiohn.
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1401W-1 WNP-3 FSAR 6,4[~7 f . [a s c. n c : s' TABLE 12.5-1 WNP-3/5 ?LANT HEALTH PHYSICS INSTRUMENTATION Number / Plant _ T,ype, Sensitivity Range
&si "---
1 Multichannel Gamma Spectrometer N/A with Ge(Li) or HPGe Detector l1
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1 Multichannel Gamma Spectrometer -dilemme - N/A with NaI(T1) Detector l1
^
2 Low Background Alpha-Beta .2; h , " rte N/A Proportional Counters
, vie /e /
1 Liquid Scintills. tion Counter J eee.- N/A
,w w /o /
4 End Window G-M Counting Systema h N/A Im P/bo-14 Ion Chamber Dose Rate Meters - ; . , 10-3-5 R/hr 4 Extendable Probe High Range Dose 10-3-104 R/hr Rate Meters 2 Neutron Dose Equivalent Rate Meters ta/L Neweeen- 10'3-10 res/hr l1 Sood 2.;-, perkoo ca,* 0-50,000 cpe 14 G-M Survey Meters with Thin Wall, End Window or Pancake Probe 10Cdpon/t00 sow' Portable Alpha Counters air 0-500,000 cpa 2
' $00l}onllooeN
- -, 0-50,000 cpa 10 Friskar Type Personnel Contamination Monitors -
9, g,'$.'e r N/A l1 l 2 Portal Radiation Monitors {,-
iKiti Ci/cc Se-9 o 6 Portable Constant Air Monitors ,., g- 4 ,, g ,,
0-10 10-4 C1/cc 1 1
0-500 mR 500 Direct Reading Personnel Dosimeters Various ranges 100 Direct Reading Personnel Dosimeters- to 100 R High Range e /es:
e s ees;/,%;/y wns maa / ar earsed
- 1. 74 a.
v.4 as v'u s ir- masc aqrz,
/ +4e 4ia T< e 6 se * /
"Deef/ 11a diofagir /PES$/v n/
- c (Ec = /*'a r. : As s~' tt 'e .
12.5-23 Amendment No. 1, (10/82)
Question No.
480.13 In Appendix 6.2A, it is stated that the WATEMPT code is an exten-(6.2.3) sion of the CONTEMPT code. Identify the modifications that have been made to the CONTEMPT code to assure a conservative calucla-tion of the shield building annulus pressure response.
Response The modifications to the code which assure a conservative calcula-tion of the Shield Building annulus pressure response are de-scribed in a proprietary description of the WATEMPT code. This proprietary code was also used to analyze the performance of the Shield Building Ventilation System for St Lucie Unit 2 (Florida Power & Light Co.) and its approval by the staff was benchmarked via St Lucie Unit 2's Safety Evaluation Report (NUREG-0843).
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Question No. j 480.17 Identify all valves in FSAR Table 6.2.4-1 that are greater than 10 l (6.2.4) feet from the containment wall. Provide the rationale for their locations and justify that they have been olaced as close to the containment as practical, as required by L C 55, 56, and 57.
Re sponse All the containment isolation valves in FSAR Table 6.2.4-1 that are greater than 10 feet from the containment are listed on Table 480.17-1. The valves listed are all located in the penetration area adjacent to the containment building. These containment iso-lation valves have been located as close to the containnent as possible while allowing adequate space requirements for operabil-ity and maintenance FSAR Table 6.2.4-1 will be updated as shown.
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QUESTION 480.17 TABLE 480.17-1 CONTAINMENT ISOLATION VALVES Length front Penetration Valve No. Penetration (ft.) Service Number 2MS-VD001 31 SGI MS Line 1 2MS-VE091 31 SG1 MS Line 1 2MS-R010 13 SGI MS Line 1 2MS-R011 20 . SG1 MS Line 1 2MS-P016 38 SGI MS Line 1 2MS-VE055 14 SGI MS Line 1 22-VE001 39 SGI MS Line 1 2MD-VE087 42 SGI MS Line 1 2MS-VD002 31 SG1 MS Line 2 2MS-R007 13 SG1 MS LIne 2 2MS-R008 20 SGI MS Line 2 2MS-P012 38 SGl MS Line 2 2MS-VE138 14 SGI MS Line 2 2MD-VE002 39 SGI MS Line 2 2@-VE088 42 SGl MS Line 2 2MS-VD003 31 SG2 MS Line 3 2MS-VE092 31 SG2 MS Line 3 2MS-R019 13 SG2 MS LIne 3 2MS-R020 20 SG2 MS Line 3 2MS-P023 38 SG2 MS Line 3 '
2MS~-V E082 14 SG2 MS Line 3 2MO-VE003 39 SG2 MS Line 3 210-VE089 42 SG2 MS Line 3
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QUESTION 480.17 TABLE 480.17-1 (Cont'd)
CONTAINMENT ISOLATION VALVES Length from Penetration Valve No. Penetration (ft.) Service Number 2FW-VD038 16 SG2 DWNC E FW & Aux FW 11 2FW-VD039 13 SG2 DWNCMR FW & Aux FW 11 2AF-VE135 23 SG2 DWNC E FW & Aux FW 11 27 SG2 DWNCMR FW & Aux FW 11 2AF-VD036 2AF-VE132 23 SG2 DWNCMR FW & Aux FW 11 2AF-VD039 27 SG2 DWNCMR FW & Aux FW 11 2FW-VD023 16 SGl DWNC E FW & Aux FW 12 2FW-VD024 13 SGI DWNCMR FW & Aux FW 12 2AF-VE133 23 SGl DWNC E FW & Aux FW 12 2AF-VD040 27 SGl DWNCMR FW & Aux FW 12 2AF-VE134 23 SGI DWNC E FW & Aux FW 12 2AF-VD035 27 SGI DWNCMR FW & Aux FW 12 2SI-VQ019 14 HPSI Loop 2A 13 2SI-VQ020 17 HPSI Loop 2A 13 2SI-VQOl6 17 HPSI Loop 28 14 2SI-VQ017 14 HPSI Loop 2B 14 2SI-VQ013 19 HPSI Loop 1A IS 2SI-VQ014 16 HPSI Loop 1A 15 2SI-VQ008 19 HPSI Loop 18 16 2SI-VQ009 16 HPSI Loop 1B 16 l
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! 2SI-VS054 15 SHTDN COOL Suction 28 2SI-VUO55 12 SHTDN COOL Suction 28 i
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Question No.
480.18 Provide a tabulation of those containment isolation valves for (6.2.4) which provision has been made to allow them to be individually leak tested, in the correct direction.
Response .
Containment Isolation Valves which have provisions for leak test- )
ing are shown on Table 480.18-1. i FSAR Figures 6.2-36b - k, m, o and p and Table 6.2.4-1 will be revised as shown. '
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TABLE 480.18-1 PENETRATION VALVE SERVICE 21 2CS-VS091 Containment Spray 21 2CS-VS037 Containment Spray 22 2CS-VSO92 Containment Spray 22 2CS-VS038 Containment Spray 23 2CS-8002 Containment Spray 24 ,
2CS-8001 Containment Spray 27 2SI-VS060 Safety Injection 27 2SI-VUO61 Safety Injection 27 i ISI-VP097 Safety Injection 28 2SI-VS054 Safety Injection 28 2SI-VUO55 Safety Injection 28 ISI-VP091 Safety Injection 29 2SI-VQ169 Safety Injection 29 2SI-VQ115 Safety Injection 30 2CC-8522 Component Cooling 30 2CC-V; 012 Component Cooling 31 2CC-BS21 Component Cooling 31 2CC-VH023 Component Cooling 32 2CC-B526 Component Cooling 32 2CC-8524 Component Ccoling 33 2CC-BS25 Component Cooling 33 2CC-8523 Component Cooling 34 2HA-VS002 Hydrogen Analyzer 34 2HA-VS008 Hydrogen Analyzer 34 2HA-V5009 Hydrogen Analyzer 34 2HA-VS010 Hydrogen Analyzer 34 2HA-VS012 Hydrogen Analyzer 35 2HA-VS021 Hydrogen Analyzer 35 2HA-VS027 Hydrogen Analyzer 35 2HA-VS028 Hydrogen Analyzer 35 2HA-V5029 Hydrogen Analyzer -
35 2HA-VS031 Hydrogen Analyzer 36 2SL-VP102 Sampling 36 2SL-VP103 Sampling 38 2IA-VE020 Instrument Air 38 2IA-VE021 Instrument Air 39 2SA-VH007 Station Air 39 2SA-VH008 Station Air 40 . 2CH-VP011 CVCS 40 2CH-VP077 CVCS 41 2CH-VQ040 CVCS ;
41 2CH-VQ064 CVCS 42 2CH-VP029 CVCS 42 2CH-VP030 CVCS 42 2DI-VS043 Dionized Water 42 2DI-VSO44 Dionized Water 44 2CH-VWO37 CVCS 44 2CH-VWO40 CVCS
TABLE 480.18-1 (Cont'd)
PENETRATION VALVE SERVICE .
45 2CH-VSO41 CVCS 45 2CH-VS501 CVCS 46 2SI-VQ173 Safety Injection 46 ISI-VP183 Safety Injection 47 2SI-VQ172 Safety Injection 47 ISI-VP181 Safety Injection 50 2PC-VW010 Fuel Pool Cleanup 50 2PC-VW011 Fuel Pool Cleanup
. 51 2PC-VW008 Fuel Pool Cleanup 51 2PC-VW009 Fuel Pool Cleanup 54 2NG-VE057 Nitrogen Supply 54 2 NG-VE058 Nitrogen Supply 56 2SL-VP104 Sampling 56 2SL-VP105 Sampling 56 2SL-VP106 Sampling 56 2SL-VP107 Sampling 60 2PV-B009 Plant Ventilation 60 2PV-B010 Plant Ventilation 65 2PV-B014 Plant Ventilation 65 2PV-V015 Plant Ventilation 68 2PV-8016 Plant Ventilation 68 2PV-8017 Plant Ventilation 68 2PV-8018 Plant Ventilation 70 2PV-B109 Plant Ventilation 70 2PV-B110 Plant Ventilation 73 2PV-Bill Plant Ventilation 73 2PV-8112 Plant Ventilation 73 2PV-Bil3 Plant Ventilation 75 2PV-Bil4 Plant Ventilation 75 2PV-8115 Plant Ventilation 80 2PV-8019 Plant Ventilation ~
80 2PV-V021 Plant Ventilation 81 2PV-B164 Plant Ventilation 81 2PV-B064 Plant Ventilation 82 2EC-B009 Chilled Water 82 2EC-B010 Chilled Water 83 2EC-B011 Chilled Water 83 2EC-B012 Chilled Water
. 90 2CH-VSO42 CVCS 90 2CH-VSO44 CVCS 90 2CH-VS047 CVCS 90 2CH-VS049 CVCS 93 2CH-VQ005 CVCS 93 2CH-VQ006 CVCS 95 2FP-VH003 Fire Protection 95 2FP-VH006 Fire Protection 96 250-VS126 Sump Dishcarge 96 2SD-VS127 Sump Discharge
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\p Figures 6.2-36 through 6.2-36r and Figures 3.8.2-1 through 5 show the various piping and electrical penetrations, respectively. Mechanical and electrical l2 penetrations that will not be subject to Type B tests are those that rely on welds for sesling an'd are designed such that the penetrations are subject to the Integrated Leak Rate Test conditions. Leakage from those penetrations will be included in the overall leak rate measured during the Type A tesc.
Test of the personnel access lock after each entry under some conditions is described below.
6.2.6.3 containment Isolation valve Leak Rate Test Table 6.2.4-1 also provides a listing of all containment isolation valves and identifies those that require and those that are exempt from leak rate tast Type C. Type C leak testing is performed as for Type B leak testing described in Subsection 6.2.6.2a or b.
Type C Leak test should be completed prior to Type A tests. In the event that a Type C test is not completed prior to the Type A test, any Type C penetration path test leakage not accounted for in the Type A test shall be added to the measured leakage determined by the Type A test.
Figures 6.2-36 through 6.2-36r show theT:rrr ;::::t t: b; ;;; rid:d f:: :::ing n
^
- f :::::i-- nt i: letic: ::17 : ::bj :t te Ty;: C ::: ting 21;;; with th di :: tic: ef pre rri stier te be d fer th: ::d Type C tests are performedbylocalpressurization(:Eachvalvetobetestedshallbeclosedby
[m b} normal operation without any preliminary exercising or adjustments and pressurized to not less than the peak accident pressure of 39.4 psig. tn the c.or r ec.*
The combined leakage of penetrations and valves subject to Type B and C cL(recMcm testing shall be less than 0.60 of La.
6.2.6.4 Scheduling and Reporting of Periodic Tests The Type A, B and C leakage tests will be completed prior to any reactor operating period. After the initial preoperational leak testing, periodic tests shall be performed in accordance with the following schedule:
a) A set of three Type A tests shall be performed at approximately equal intervals during each 10 year service period. The third test of each set shall be conducted when the plant is shut down for the 10 year plant in-service inspection.
b) Type B tests shall be performed during each reactor shutdown for refueling, or other convenient intervals, but in no case at intervals greater than two years. Air locks shall be tested at six monnth intervals. Air locks which are opened during such intervals shall be tested after each opening.
c) Type C tests shall be performed during each reactor shutdown for refueling, but in no case at an interval greater than two years.
v.
4; w anem vake arrangemeas aca inaicake +bcse
{coecdr va.Wes & Web M &nhms we ben prodded Oc mdM b a\ lea.k -\ ee'q.
gebef 6.2-174 Amendment No. 2, (12/82)
r PENETRATION NO./ INSIDE OUTSIDE SERVICE CONTAINMENT CONTAINMENT NOTES
- 0 :
$' STEEL CONCRETE CONTAIN. 4.; SHIELD M VESSEL .';. , BUILDING
- m ,.
13,14,15,16 ' . '~
HPSI TO i O M COLD LEGS - - -
- ' a fo.,
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,, M 17,18,19,20
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CONT SPRAY gI r -
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I, r 1;
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STEEL LINER
_ _ 1 :-
RECIRC -[- .a ~ SHIELD BUILDING SUMP . . . ..' .....I. .* : .
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AMENDMENT NO.1 (10/82)
FIGURE WASHINGTON PUBLIC i[fhd
- " POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6.2 36b Nuclear Proj.ects 3 & 5 FINAL SAFETY ANALYSIS REPORT
/
S c Wi77
PENETRATION NO./ INSIDE OUTSIDE SERVICE CONTAINMENT CONTAINMENT NOTES l 4:9 STEEL M CONCRETE SHIELD CONTAIN. WC VESSEL BUILDING 4 9a . /?
25 '00 "!O 05'"-
, ~
FIRE O O PROTECTION k[] ' a 9 r TYPICAL TYPICAL FOR EACH .'2 FOR EACH OF FOUR -
'h OF FOUR LINES 9 - LINE 26 ..
SG SG SAMPLE V
2.SI vuO6 I 15Ivfo97 (2.SI VU 055)
(ISI VP091) h I @1 l; 27,(28)
SHUTDOWN ,,Cha a J C
OM 4,{
COOLING .Tm Ia 151VSO 60 (ISIVS oS4')
2.51 YQlli k i }'
2.
SIT i @; JN
[]
au >c l W FILL & DRAIN ,
]AIN !
b E~ 2 51 YQ(d AMENDMENT NO.1 (10/82)
FIGURE WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6.2 36c Nuclear Projects 3 & 5 FINAL SAFETY ANALYSIS REPORT h h [ ,/ $ -
PENETRATION NO./ INSIDE OUTSIDE SERVICE CONTAINMENT CONTAINMENT NOTES m
- =
STEEL' N
'J' CONCRETE SHIELD 2 CONTAIN.
VESSEL N*. BLDG. '('OC~06EA 1CC-B52,0 30,(31) +- e COOLING WATER M ', 'M Y O ,/l ,
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TO RCP IccyHol2 NO 6, (2Ccyggy3] 7,(.
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l oN CONNECTS TOLINE PENETRATION 36 THIS PENETR CONNECTS TO PENETRATION 209.
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ANALYZER _
+
SUCTION & . :
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% , db q mlm a s, w 3, da IS L vPlo2.~ LSLYPl03 AMENDMENT NO.1 (10/82) k FIGURE WASHINGTON ?UBLIC POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6.2-36d Nuclear Projects 3 & 5 FINAL SAFETY ANALYSIS REPORT
PENETRATION NO./ INSIDE OUTSIDE SERV!OE CONTAINMENT CONTAINMENT NOTES m : : :
STEEL I CONTAIN.
h 41 CONCRETE SHIELD VESSEL -['i. BUILDING
. :s 37 -
SPARE T-I.i n
k w
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.; ic.MV Po l l ac.w veo11 ,)gj AMENOMENT NO.1 (10/82) age FIGURE f.-[q WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6.2-36e Nuclear Projects 3 & S FINAL SAFETY ANALYSIS REPORT $O + / h Sc V77
PENETRATION NO./ INSIDE OUTSIDE SERVICE CONTAINMENT CONTAINMENT NOTES
- O :
i STEEL % CONORETE CONTAIN. ,- SHIELD VESSEL 4- BUILDING 41 ._ f CHARGING +
5 " )
LINE & i r u g a. c. n g o g o AUX SPRAY l ' .3 11.HYQC64 4 2
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th.
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AMENDMENT NO. 1 110/82)
&& WASHINGTON PUBLIC FIGURE 5.c;,? u
"' POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6.2-36f Nuclear Proj,ects 3 & 5 FINAL SAFETY ANALYSIS REPORT h . / h 5(NF77
PENETRATION NO./ INSIDE OUTSIDE SERVICE CONTAINMENT CONTAINMENT _ NOTES
~ ~ ~
..~ )
--' STEEL CONTAIN.
[g$ SHIELD CONCRETE VESSEL BUILDING 2 CHg VS 041 gf 45 17 ; A MAKE UP TO a a "
REACTOR
' P[ 1; e, p
COOLANT DRAIN TANK y
A Ic.HVsfot
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p 29, O N
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SG DOWNCOMER 8G [] L5 SAMPLE
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- CONTAINMENT PRELIMINARY ATMOSPHERE " "
SAMPLING i a
r e G ,
S dHM,- RETURN a ..
AMENDMENT NO.1 (10/82) 40%
FIGURE
$ WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6.2-36g Nuclear Projects 3 & 5 FINAL SAFETY ANALYSIS REPORT . Ih l
PENETRATION NO./ INSIDE OUTSIDE SERVICE CONTAINMENT CONTAINMENT NOTES
=
STEEL hCONCRETE DURING REFUELING CONTAIN. 3d SHIELD THE PENETRATION IS REFUELING 50 VESSEL ]I BUILDING FULL OF WATER. AT ALL OTHER TIMES, IT CAVITY PURIF. 2,PC YWolo N 50 a kvwot t a " m' m e CO AINS AIR.
SUPPLY r &_ m ,,
J r'N bJ rN Jl Vm %
1""* O 51, ggyyy REFUELING 51 i;f '
C M i "
'(.-j 2
CI,' '
CAVtTY PURIF.
RETURN U
'" h "3
DURING REFUELING
), *' THE PENETRATION -
- MAY CONTAIN GAS.
AT ALL OTHER TIMES, i Al AIR.
FAILED FUEL F ', e, DETECTOR N!
( w
-TEST CONNECTION
( $ SEE NOTE FOR PENETRATION 50,51.
"O" RINGS Q
M) 6uritisisiin 53 FUEL TRANSFER " ' ' ' ' ' '
CANAL U /.J kW $
N 2Hrn-VEoSB 2 N rr-VE 057
== A A 54 H w NITROGEN M$
y $- M SUPPLY n. .
M hi M
W AMENOMENT NO.1 (10/82)
{: ; - FIGURE
- g. ;.y WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6.2 36h Nuclear Projects 3 & 5 FINAL SAFETY ANALYSIS REPORT h hEO./
C L i4 S~11 8
e PENETRATION NO./ INSIDE OUTSIDE SERVICE . CO_NTAINMENT, CONTAINMENT _ NOTES
< 3, STEEL CONTAIN.
hCONCRETE SHIELD gg, VESSEL
. BUILDING 55 SPARE -
N
- C
?:.
asL-YP tok 2SL-yPloh k S l [ X i g r! ,
- _1; 9 v
- _;
- : }/RC HOTLEG SAMPLE RC HOT LEG '
SAMPLE, == -
PRESSURIZER STEAM ll
" h A SPACE SAMPLE @[ ;:C ,. 3-C $ I! 2 2.sk Y P io4 a sw-vP o5 w
l s VENTED TO
\ CONTAINMENT $ S
! N ;_; GAS / AIR 57 POST ACCIDENT J' SAMPLING N. @
RETURNS
- g iM e
- : WATER I sh:
vg 4:
l 58 l SPARE k
n-AMENDMENT NO.1 (10/82)
Y'#$$ FIGURE WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTAINMENT ISO'LATION VALVES ARRANGEMENT 6236
Nuclear Projects 3 & 5 g ,, g FINAL SAFETY ANALYSIS REPORT
- PENETRATION NO./ INSIDE OUTSIDE SERVICE CONTAINMENT CONTAINMENT NOTES
, = C =
? , c: .
- {Q I STEEL CONCRETE CONTAIN. SHIELD VES$EL ,
BUILDING se SPARE y-"j 4
4
- Py-(2 py io IPV-B co ! 'cPENETRATION 60 (2.PV-8 g t o) CONNECTS TO g -f g PENETRATION 62.
CC(AND 70)
HYDROGEN psk ---*
PENETRATION 70 CONNECTS TO H M', jp l# H PENETRATION 72.
PURGE "C NC EXHAUST y
- ? 'E sten 61 (AND 71) Mm 'e SHIELD BLDG "SS" u [v SUPPLY F s1 (.9i) ,
- () ,TO SH!!LD BUILDING a ,/r, i e VENTILATION CLEAN-UP SYSTEM 62 (AND 72) _ _ _ . . _ . . . . , . .
SHIELD BLDG i EXHAUST ', ; i
'm 1'11'
,,,. m.O .',! m '".?m--
62.C'th DOES NOT HAVE A
. - SHIELD BUILDING PENETRATION.
63 . .g
- I
-/
SPARE
(-
AMENDMENT NO.1 (10/82)
FIGURE WASHINGTON PUSLIC POWER SUPPLY SYSTEM CONTAINMENT ISCu', TION VALVES ARRANGEMENT 6.2-36 Nuclear Projects 3 & 5 FINAL SAFETY ANALYSIS REPORT
. e-
- P'ENETRATION NO./ INSIDE
, OUTSIDE SERVICE CONTAINMENT CONTAINMENT NOTES QD) 1.g; $y M
M .<
65 (AND 75) , ,gj.l RB .,
' 7' r,
" 'j'i e r-ny )
VACUUM 4 i
's (2.P v - 6 t t i') s EEE N
l '
A A 66 (AND 76)
SB ,
- . e VACUUM " 'I.' '.I.' '
MAINTENANCE NO NO p)
(U TO CONTAINMENT VACUUM e RELIEF SYST. 'A' CONTAINMENT
- TO ANNULUS DIFF. PRES.
f INSTRUMENTS 67 e TO SHIELD BLDG VENT SYST 'A' INSTRUMENT 3 ; ANNULUS TO OUTSIDE RAB DIFF.
LINES PRES. INSTRUMENTS
[ .
] SPARE 2 91f Bol(* STEEL CONCRETE 2 PV-6 Olfl (2PV-B tiQ 68 C8 PV 8 Ill) CONTAIN. SHIELD RB PURGE \ VESSEL .
BLDG. p2py 80'S (zPV 8113) g ',
SUPPLY y es 4----
lll C: l$> $l 73 NC s
NC NC 7
RB PURGE EXHAUST k X3 M
LO AMENDMENT NO.2 (12/82)
FIGURE CJ WASHINGTON PUBLIC f$,%Y POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6.2-36k Nuclear Projects 3 & 5 gg , f g FINAL SAFETY ANALYSIS REPORT g3p
l g
PENETRATION NO./ INSIDE OUTSIDE SERVICE CONTAINMENT CONTAINMENT NOTES
c
,d' STEEL CONTAIN.
g CONCRETE 9 <.g SHIELD VESSEL t,6,. BLDG.
M so ::
CONTAINMENT VENT MAKE UP
]d '
A ,, ::
/ a 6
'/
g 5
2PV Vo21 ==
s . g py,gg N
1 l
2 OM I 2rv Bite 4
== A A <
l 81 a 6 % 4 !
CONTAINMENT -ll / :: / $ l VENT EXHAUST l[ l
= ,
I
($ \
j p ist. to 2 Ec. tooi 82 i CHILLED WATER r, M
- =M
, SUPPLY 1 r s2 3 r '
i d b &d h W ,, ; l/H.
l l Wl; , /NO ,
(83) j g 83 NO g -
CHILLED WATER u n. l
'i:
RETURN -
'93 l fh dW @ '
n k:, x TO CONTAINMENT VACUUM RELIEF 84 SYST. 'B' CONTAINMENT TO ANNU- :
INETRUMENT LUS DIFF. PRES. INSTRUMENTS l
"" "' SPARE LINES s -
.S*t AMENDMENT NO.2 (12/82) )
FIGURE ,
WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6 2~36"1 Nuclear Projects 3 & 5 Csqrfp , ,1/
FINAL SAFETY ANALYSIS REPORT 3cN q77 l . - - . - . __ - . _ _ . .
9 3
PENETRATION NO./ INSIDE OUTSIDE SERVICE CONTAINMENT CONTAINMENT NOTES
'. 2 CONCRETE vt SHIELD bt!. BUILDING TO SB ANNULUS VACUUM MAIN-89 2 ; TAINANCE SYST. 'B' ANNULUS TO OUTSIDE RAB DIFF. PRES.
INSTRUMENT "
LINES : J SPARE N
,, A *j.g A g"' i.'{
VENT 90 REACTOR DRAIN
TANK VENT, g AND REACTOR -- A = 2.CH -VS o APT DRAIN TANK / NITROGEN SUPPLY NITROGEN SUPPLY g _ %
2.c.H ys e.#9 :
91 SEE PENETRATION 48 92 SEE PENETRATION 25 k 1.cM vQcos' 2CH V4oodo
%~
F -
93 4 e SEAL INJECTION HEADER
% 'A y a
C >
e r r7
_k = v ;; 1 ,
% J k Q o ,. s a AMENDMENT NO. 2 (12/82) _
. (b.M '
WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6.2 36o Nuclear Pro.iects 3 & 5 6,4/<sO, I S FINAL SAFETY ANALYSIS REPORT gg g $77
- g.
. PENETRATION NO./ INSIDE OUTSIDE SERVICE CONTAINMENT CONTAINMENT NOTES C --
G
- N CONTAIN.
STEEL N r.*
CONCRETE SHIELD VESSEL 1,'*. BUILDING 94 SPARE __
i.t.
- N CbCRETE STEEL CONTAIN. SHIELD .g yp.yno 06 VESSEL :
BUILDING f 1FP-VHeas ;, ,
r,
" u
.A w
)
95 & 9 y l I J J k F'e e t-ProiecNion 4 ; ,.
J N
w Asp VSif.h ni 2.5 0 VS 12.7 8 ?$;. I A
96 5 $
3 PUMP 'n/ --.-
DISCHARGE
$N l 94 M,
st 5
9'.
9, STATION *--
AIR TO i ,
SHIELD BLDG. . ),F 2d M
AMENDMENT NO. 1 110/82)
FIGURE WASHINGTON PUBLIC POWER SUPPLY SYSTEM CONTAINMENT ISOLATION VALVES ARRANGEMENT 6.2-36p Nuclear Projects 3 & 5 Cu%'O ,,g FINAL SAFETY ANALYSIS REPORT $CW 5'77
(
- 4 Question No.
480.21 Section 6.2.5.2.1 states that manual operator action from the (6.2.5) control room is required to actuate the containment hydrogen analyzers. Discuss and justify the emergency procedures that will alert an , operator of the need to actuate the hydrogen
- analyzers.
Response
The WNP-3 Emergency Procedures have not yet been fully de-veloped. These procedures will, however, be based on the guidance of CEN-152 " Combustion Engineering Emergency Proce-dure Guidelines", which include discussions on containment combustible gas control. The subject CE guidelines have re-ceived an SER, per NRC letter dated July 29, 1983; D. G.
Eisenhut to R. G. Wells of CE Owners Group; Safety Evaluation of " Emergency Procedure Guidelines".
F-
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